ML17157C056

Safety Evaluation Supporting Amend 91 to License NPF-22

ML17157C056
Person / Time
Site:	Susquehanna
Issue date:	10/28/1992
From:	Office of Nuclear Reactor Regulation
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ML17157C055	List: ML17157C054 ML17157C056
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NUDOCS 9211060157
Download: ML17157C056 (19)

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'0 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT N0.91 TO FACILITY OPERATING LICENSE NO.

NPF 22 PENNSYLVANIA POWER 8I LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.

SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 DOCKET NO. 50-388

1. 0 INTRODUCTION By letter dated June 30,

1992, as supplemented July 21,

1992, and October 19,

1992, the Pennsylvania Power and Light Company and Allegheny Electric Cooperative, Inc. (the licensees) submitted a request for changes to the Susquehanna Steam Electric Station, Unit 2, Technical Specifications (TS).

The requested changes would make changes to the TS in support of the Unit 2, Cycle 6 reload.

Changes have been made to the following TS and Bases:

a.

B 2.1 b.

3/4.2.1 c.

3/4.2.2 d.

3/4.2.3 e.

3/4.2.4 f.

3/4.4.1 g.

B 3/4.1 h.

B 3/4.2 i.

B 3/4.4.1 j.

5:3.1 Safety Limits Average Planar Linear Heat Generation Rate APRM Setpoints Minimum Critical Power Ratio Linear Heat Generation Rate Recirculation System Reactivity Control Systems Power Distribution Limits Recirculation System Fuel Assemblies.

The July 21, 1992, letter provided clarifying information and the October 19, 1992 letter provided an affidavit for the July 21, 1992 letter.

They did not change the initial proposed no significant hazards consideration determination.

The Susquehanna 2 Cycle 6 (S2C6) reload will consist of 236 fresh (unirradiated)

SNP 9x9 (S(B-5) fuel assemblies, 232 once irradiated SNP 9x9 (ANF-4) assemblies, 204 twice irradiated SNP 9x9 (XN-3) assemblies, 88 thrice irradiated SNP 9x9 (XN-2) assemblies, and 4 thrice irradiated SNP 9x9 (XN-1) assemblies.

The new 9x9 S(B-5 fuel has similar operating characteristics (mechanical, thermal-hydraulic, and nuclear) to the previously used SNP 9x9 reload fuel.

In support of the S2C6 reload, the licensee submitted a reload summary report (Reference 2).

9211060157 921028 PDR ADOCK 05000388 p

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Cycle 6 is the third reload cycle for Unit 2 that PP&L designed by using NRC-approved steady state physics methods (Reference 3).

However, this will be the first reload cycle for Unit 2 where the licensing analyses were performed by PP&L using safety analysis methods recently approved by the NRC (References 4,

5, 6, 7, 9 and 24).

The licensing analyses that PP&L performed are:

shutdown margin; standby liquid control system capability; control rod drop accident; loss of feedwater heating; rod withdrawal error; fuel loading error (both rotated and mislocated);

generator load rejection without bypass; feedwater controller failure; recirculation flow controller failure; and, ASHE overpressure compliance.

SNP provided some supporting analyses including the loss of coolant accident (LOCA) and minimum critical power ratio (HCPR) safety limit analyses.

In addition, SNP has previously performed fuel storage criticality, single loop operation, and fuel and equipment handling accident analyses.

2. 0 EVALUATION 2.1 Fuel Mechanical Desi n

The S2C6 core reload will include 236 SNP 9x9 fuel assemblies with the designation'gB-5.

These reload assemblies contain 79 fuel rods and 2 water rods in a 9x9 array.

Of the 236 SgB-5 fuel assemblies, 148 will contain 9

burnable poison rods with 4.0 weight percent (w/o)

Gdz0s at a bundle average enrichment of 3.51 w/o U-235 and 88 will contain 10 burnable poison rods with 5.0 w/o Gdz0~ at a bundle average enrichment of 3.40 w/o U-235.

The fuel design and safety analysis are described in the Susquehanna 2

specific report PL-NF-92-001 (Reference 2) and the generic mechanical design report XN-NF-85-67(P)(A), Revision 1 (Reference

10).

The NRC has approved the latter report and issued a Safety Evaluation Report on July 23, 1986 (Reference 11).

Table 2. 1 of XN-NF-85-67(P)(A), Revision 1, gives the pertinent design data for SNP 9x9 fuel.

Neutronic values specific to the S2C6 reload are given in PL-NF-92-001 (Reference 2).

The analyses for S2C6 support fuel assembly average discharge exposures of 40,000 HWD/HTU which is based on the approved SNP topical report XN-NF-82-06(P)(A), Supplement 1, Revision 2 (Reference 12).

In addition, Appendix A of PL-NF-92-001 provides information for extending the burnup limit of the four XN-1 SNP 9x9-2 fuel assemblies to 48,000 HWD/HTU.

Based on the poolside inspections performed on these assemblies during cycle 5

and on the mechanical design analyses presented in Appendix A, the staff concurs that these four fuel assemblies have the capability to attain an exposure of 48,000 HWD/HTU without exceeding any design limits.

These four assemblies will be placed in the center of the Cycle 6 core and will not be thermally limiting.

They are expected to attain a maximum assembly exposure somewhat less than the proposed licensing limit of 48,000 HWD/MTU.

Therefore, based on our review of the information presented, we find the proposed extension of the maximum assembly average exposure limit on the four XN-1 fuel assemblies to 48,000 HWD/HTU for Cycle 6 to be acceptable.

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The generic analyses to evaluate the steady state strain, transient strain, cladding fatigue, creep collapse, cladding corrosion, hydrogen absorption, differential fuel rod growth, and grid spacer spring design for the SNP 9x9 fuel design were performed by SNP.

The

RODEX2, RODEX2A, RAHPEX, and COLAPX codes were used in the generic mechanical design analyses.

These codes have been approved and/or previously accepted by the NRC and the results indicate that all parameters meet their respective design limits as described in Reference 10.

Therefore, we find the mechanical design of the SNP 9x9 fuel for the S2C6 reload acceptable.

A figure of linear heat generation rate (LHGR) limit versus average planar exposure (HWD/MTU) is incorporated into the Susquehanna 2 TS.

This figure was previously approved to reflect the design values which have been reviewed and approved for the SNP 9x9 fuel in connection with the staff's review of XN-NF-85-67(P)(A), Revision 1 (Reference 10).

Since an extended burnup limit to 48,000 HWD/HTU has been approved for the four XN-1 assemblies in Cycle 6, this figure has been separated into two figures, Figure 3.2.4-1 applicable to all SNP 9x9 fuel except XN-1, and Figure 3.2.4-2, which is applicable only to XN-1 9x9 fuel.

These figures, as well as those for the average power range monitors (APRH) setpoints (Figures 3.2.2-1 and 3.2.2-2),

are given in terms of average planar exposure, which is the average exposure of all the rods in a given fuel assembly in a given horizontal plane (actually a 6-inch slab).

The planar exposure limits given are those necessary to support the average assembly exposure limits and are acceptable.

The licensee has discussed the mechanical response of the SNP 9x9 fuel assembly design during LOCA-seismic events in Reference 2.

The discussion compares the physical and structural properties of the SNP 9x9 fuel and the previously used GE 8x8 fuel.

The staff has reviewed this information in connection with a previous review and has confirmed that the physical and structural characteristics of the SNP 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, the staff concludes that the original analysis is applicable to S2C6 and the analysis indicating that the design limits are not exceeded is acceptable.

2.2

~N1 D

The nuclear design methodology used for S2C6 is that presented in PPKL topical reports PL-NF-87-001-A, PL-NF-89-005, and PL-NF-90-001 (References 3,

4 and 5),

and corresponding supplements (References 6, 7, 9 and 24).

These reports have been reviewed and approved by the staff for application to Susquehanna core reloads (References 22 and 23).

The minimum value of shutdown margin occurs at a cycle exposure of 8,250 HWD/HTU and is 1.05X ~k/k.

The beginning-of-cycle (BOC) shutdown margin is 1.31X ~k/k.

Thus the minimum shutdown margin at any time during Cycle 6 is well in excess of the minimum 0.38X ~k/k shutdown margin TS requirement.

The Standby Liquid Control System (SLCS) also fully meets shutdown requirements.

The existing new fuel storage calculations are based on the value of k-infinity (k) of the fuel assembly.

Based on SNP calculations of 9x9 fuel, an average lattice enrichment of less than 3.95 w/o U-235 and a k for the cold (68'F), moderated, uncontrolled fuel assembly lattice in reactor geometry at BOC of less than or equal to 1.388 will meet the acceptance criterion of k-effective (k,<<) no greater than 0.95 under dry or flooded conditions.

More recent evaluations of new fuel vault criticality for temperatures as low as 32'F have caused PP&L to reduce their k criterion to 1.385.

Since the zone average enrichment of both SgB-5 fuel assembly types are 3.63 w/o U-235 and 3.51 w/o U-235 and the maximum cold, uncontrolled, beginning-of-life (BOL) k of the two SNP fuel assembly enriched zones are

1. 1206 and 1.0745, the staff's acceptance criterion is met for the new fuel storage vault under dry and flooded conditions.

To preclude criticality at optimum moderation conditions, watertight covers, criticality monitors, and appropriate procedures are used.

These are acceptable.

SNP also performed analyses for 9x9 fuel stored in the spent fuel pool.

A maximum enriched zone of less than 3.95 w/o U-235 meets the staff acceptance criterion of k no greater than 0.95.

Since the S(B-5 fuel has zone average enrichments of Y.63 and 3.51 w/o U-235, the staff's acceptance criterion for spent fuel storage is met for the SgB-5 9x9 fuel.

Susquehanna will continue to use the SNP POWERPLEX core monitoring system to monitor core parameters.

The system has been in use for a number of cycles for both Susquehanna Units 1 and 2 and has provided acceptable monitoring and predictive results.

However, the POWERPLEX input for Cycle 6 will be based on the CPM2/PPL methodology (Reference

3) reviewed and approved by the NRC.

The application of CPH2/PPL generated input in POWERPLEX is described in Reference 5 as supplemented by References 7 and 9, and has also been approved by the NRC.

Although the current SNP POWERPLEX power distribution uncertainties were shown to be conservative relative to those obtained using CPM2 generated input, the NRC has concluded that the safety limit HCPR POWfRPLfX uncertainties should remain at their presently approved values when monitoring the core with CPM2/PPL generated input (Reference 22).

The licensee is conforming to this requirement for Cycle 6.

2.3 Thermal-H draulic Desi The minimum critical power ratio (HCPR) for the S2C6 reload was determined by the licensee to be 1.06 for all fuel types.

The methodology for S2C6 is based on the SNP methodology in XN-NF-80-19(P)(A), Volume 4, Revision 1 (Reference 13), which has been approved by the NRC.

The XN-3 critical power correlation used to develop the HCPR safety limit has been approved for the SNP 9x9 fuel.

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SNP has determined that this correlation provides sufficient conservatism such that there is no need for any penalty due to channel bow for S2C6.

Susquehanna is a

C lattice core and uses channels for only one fuel assembly lifetime.

For such cores, SNP has determined that the conservatism is greater than the maximum expected 6 CPR.

The staff has reviewed the SNP channel bow analysis methodology and it is acceptable for this analysis for S2C6.

The core bypass flow fraction has been calculated to be 8.4 percent of total core flow using the approved methodology described in PL-NF-87-001-A (Reference 3).

This is used in the HCPR safety limit calculations and as input to the S2C6 transient analyses and is acceptable.

In response to Bulletin 88-07, Supplement I (Reference 14) on BWR thermal-hydraulic stability, PPKL developed restricted operating regions on the power/flow operating map which were in compliance with the NRC recommendations.

Calculations similar to those setting up the restrictive boundaries were done for S2C6.

TS implementing these regions have previously been approved by the NRC for Susquehanna.

Stability tests have been conducted in Susquehanna 2 with various amounts of SNP 9x9 fuel from succeeding

reloads, including all 9x9 fuel.

These have indicated no significant deterioration of decay ratio as compared to BWRs loaded with standard Bx8 fuel.

A report from Oak Ridge National Laboratory (Reference 8),

an NRC consultant, provided the same conclusion based on their analysis of the Susquehanna stability test data.

TS Figure 3.4.l.l.l-l implementing the changes has been submitted.

This review concludes that the analyses are suitable and the changes to the TS are acceptable.

2.4 Transient and Accident Anal ses Various operational transients could reduce HCPR 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 Critical Power Ratio (CPR), that is, 6 CPR.

The core wide transients which resulted in the largest &PR were the generator load rejection without bypass (GLRWOB) and the feedwater controller failure (FWCF).

These were analyzed based on an average scram speed of 4.4 feet/second and the minimum allowed TS scram speed.

Therefore, the power dependent MCPR operating limits for S2C6 are given in the TS as a function of scram speed.

The results of the required scram speed time testing (TS 4.1.3.3) will be used to adjust the HCPR operating limits to assure the validity of the Cycle 6 transient analyses.

The recirculation flow controller failure (RFCF) event, conservatively analyzed at the TS scram

speed, was the limiting event in determining the TS flow-dependent HCPR operating limits for Cycle 6.

In response to the staff's question, the licensee has confirmed the recently discovered error in the calculation of the TS flow dependent HCPR operating limit curve (Reference 25) has been corrected in TS Figure 3.2.3-1 for Cycle 6.

The loss of feedwater heating (LOFWH) event was found to be bounded by these other three core wide transients.

The calculations of the thermal margin were performed with approved methodology and the resulting required HCPR operating limits as functions of core power and core flow proposed in TS Figures 3.2.3-1 through 3.2.3-4 are acceptable.

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s It was assumed for the above analyses that the turbine bypass system and the end-of-cycle (EOC) recirculation pump trip (RPT) were operable.

Analyses were also performed to determine HCPR operating limits with either of these systems inoperable.

This resulted in increased HCPR limits which are also proposed for Cycle 6.

These calculations follow standard procedures and operation within the proposed HCPR operating limits with either the main turbine bypass system inoperable or the EOC-RPT inoperable is acceptable for S2C6.

The two limiting local transients were analyzed using the approved methodology described in References 3, 5, 7 and 22.

The control rod withdrawal error (RWE), was analyzed to support a rod block monitor (RBH) setpoint of 108 percent and resulted in a MPR of 0.20.

The fuel loading error, which included analysis of both rotated and mislocated fuel assemblies, was also analyzed and the rotated assembly analysis resulted in the larger hCPR of 0.22.

Both of these events are bounded by the GLRWOB and are, therefore, non-limiting for Cycle 6.

Compliance with the ASHE Code overpressurization criterion of 110 percent of vessel design pressure (1375 psig) was demonstrated by analysis of the main steam isolation valve (HSIV) closure event assuming HSIV position switch scram

failure, an HSIY closure time less than the current TS minimum closure time, and six safety relief valves out of service.

Haximum vessel pressure was 1334.9 psig, within the limit of 1375 psig.

The calculation was done with approved methodology and the'results are acceptable.

The LOCA analyses for the Susquehanna plants (Reference 15) was performed by SNP for a full core of SNP 9x9 fuel.

These analyses have covered an acceptable range of conditions and have been performed with approved methodology.

SNP confirmed that the Haximum Average Planar Linear Heat Generation Rate (HAPLHGR) limits in Reference 15 also ensure that the peak cladding temperature for the SgB-5 fuel remains below 2200'F, local.Zr-H 0 reaction remains below 17X, and core-wide hydrogen production remains below 1X for the limiting LOCA event as required by 10 CFR 50.46.

Therefore, the resulting TS HAPLHGR values for the SNP fuel remain acceptable for the S2C6 residual and reload SNP fuel.

Because of the large amount of margin between calculated and limiting peak clad temperature (i.e., 532'F) for the S(B-5 fuel at an assembly average exposure of 48,000 HWD/HTU, the licensee initially extended the HAPLHGR curve to 48,000 HWD/HTU for all SNP 9x9 fuel loaded in S2C6.

However, since this burnup limit has only been approved for the four XN-1 assemblies, the TS figure was

amended, based on the staff's request, by a note specifying an assembly average exposure limit of 40,000 HWD/HTU for all SNP 9x9 fuel except the XN-1 fuel (Reference 26).

The control rod drop accident (CRDA) was analyzed with approved PP&L methodology (Reference 16).

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 (Reference 17).

The maximum fuel rod enthalpy was calculated to be '164 cal/gm, which is below the clad failure threshold of 170 cal/gm, and well below the prompt fuel rupture threshold of 280 cal/gm.

The staff concludes that the analysis and results for the SZC6 CRDA are acceptable.

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2.5 Sin le Loo 0 eration SLO Current TS for Unit 2 permit plant operation with a single recirculation loop out-of-service for an extended period of time.

However, because of the increased measurement uncertainties, the HCPR safety limit must be increased by 0.01 (Reference 18).

SNP has previously performed SLO analyses for the Susquehanna units (Reference 19).

From these analyses, it was determined that the transients considered for two-loop operation bound those for SLO conditions.

In addition, it was also determined that postulated accidents under SLO conditions, with the exception of the single-loop pump seizure

accident, were non-limiting when compared to the postulated accidents under two-loop operating conditions.

For SLO, it was shown that operation of the Susquehanna units with the single-loop HCPR operating limits conservatively meets the dose requirements of a small fraction of 10 CFR Part 100 for the single-loop pump seizure accident.

These analyses have been previously accepted by the NRC (Reference 20) and are applicable to S2C6.

SLO for Cycle 6 continues to maintain the 80 percent recirculation pump speed restriction because.of the previous General Electric (GE) vessel internal vibration

analysis, as discussed in Reference 21.

2.6 Technical S ecification Chan es The following TS changes have been proposed for operation of S2C6.

(1)

TS 3/4.2.1 The changes to this specification extend the HAPLHGR limits in Figure 3.2. 1-1 to an assembly average exposure of 48,000 HWD/HTU.

A note is added which indicates that all 9x9 fuel except the XN-1 assemblies are limited to a burnup limit of 40,000 HWD/HTU.

Also, references to "ANF" are updated to "SNP."

The changes are acceptable.

(2)

TS 3/4.2.2 - The changes to this specification add the LHGR for APRH setpoints (Figure 3.2.2-2) which are to be used for the XN-1 fuel assemblies.

In addition, references to "ANF" are updated to "SNP" and words that were inadvertently reversed are corrected.

The changes are acceptable.

(3)

TS 3/4.2.3 Figures 3.2.3-1 and -2 are changed to reflect the new calculations of flow and power dependent HCPR operating limits using the parameters of S2C6.

In addition, Figures 3.2.3-3 and -4 are added.

The limits calculated for Cycle 6 will also be a function of scram speed as discussed in Section 2.5 of this SER.

As previously discussed, these analyses have been approved and the changes are acceptable.

(4)

TS 3/4.2.4 - The changes to this specification adds the LHGR (Figure 3.2.4-2) which is to be used for the extended burnup XN-1 fuel.

In

addition, references to "ANF" are updated to "SNP."

The changes are acceptable.

(5)

TS 3/4.4. 1. 1 - Figure 3.4. 1. 1. 1-1 is changed to reflect the calculated changes in the regional stability boundaries.

The changes are acceptable.

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(6)

TS 3/4.4. 1.2 Reference to new Figures 3.2.3-3 and -4 for SLO MCPR operating limits is added.

These changes are the result of the MCPR SLO analyses discussed in Section 2.6 of this SE and are acceptable.

(7)

TS 5.3. 1 The proposed changes reflect that the Cycle 6 core will contain only 9x9 fuel, and the reference to Zircaloy-2 cladding has been editorially relocated for consistency with the wording in the Unit 1 TS.

The changes are acceptable.

In addition, there are several administrative and descriptive changes to other TS and to the Bases reflecting removal of errors or the reasons for the TS changes discussed above.

There are also editorial changes to the TS for clarification and consistency.

These changes are acceptable.

2. 8

~Summar The staff has reviewed the report submitted for the Cycle 6 operation of Susquehanna Unit 2 and concludes that appropriate material was submitted and that the fuel design, nuclear design, thermal-hydraulic design and transient and accident analyses are acceptable.

The TS changes submitted for this reload suitably reflect the necessary modifications for operation in this cycle.

3.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Pennsylvania State official was notified of the proposed issuance of the amendment.

The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 and changes surveillance requirements.

The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant 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 the amendment involves no significant hazards consideration,'nd there has been no public comment on such finding (57 FR 42776).

Accordingly, the 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 or environmental assessment need be prepared in connection with the issuance of the amendment.

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The Commission 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, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors:

L. Kopp J.

Raleigh pate:

October 28'992

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I REFERENCES 2.

3.

4, 5.

6.

7.

8.

9.

10.

12.

13.

14.

PLA-3787, "Proposed Amendment 105 to License No. NPF-22:

Unit 2 Cycle 6

Reload,"

June 30, 1992.

PL-NF-92-001, "Susquehanna SES Unit 2 Cycle 6 Reload Summary Report," June 1992.

PL-NF-87-001-A, "gualification of Steady State Core Physics Methods for BWR Design and Analysis," April 28, 1988.

PL-NF-89-005, "gualification of Transient Analysis Methods for BWR Design and Analysis," December 21, 1990.

PL-NF-90-001, "Application of Reactor Analysis Methods for BWR Design and Analysis," August 1, 1990.

PLA-3542, "Response to RAI on PL-NF-89-005,"

Letter from H.

W. Keiser (PPSL) to W.

R. Butler (NRC), Harch 16, 1991.

PLA-3578, "Final Response to RAI on PL-NF-90-001,"

Letter from H.

W.

Keiser (PPSL) to W.

R. Butler (NRC), June 4, 1991.

Letter from James J. Raleigh (NRC) to H.

W. Keiser (PPKL), "Evaluation of 9x9 Fuel Impact on the Stability of Susquehanna Unit 2 at the End of Cycle 4, Susquehanna Steam Electric Station, Units 1 and 2," September 12, 1991.

PLA-3663, "Licensing Methods:

Supporting Information," Letter from H.

W.

Keiser (PPKL) to W.

R. Butler (NRC), October 17, 1991.

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,'" 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 Fuel," Hay 1988.

XN-NF-80-19(P)(A), Volume 4, Revision 1,

"Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads,"

Exxon Nuclear Company, Inc., June 1986.

NRCB 88-07, Supplement 1,

"Power Oscillations in Boiling Water Reactors (BWRs),"

USNRC Bulletin, December 30, 1988.

~ k

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- 11 15.

16.

17.

18.

19.

20.

21.

22.

23.

24.

25.

26.

XN-NF-86-65, "Susquehanna LOCA-ECCS Analysis MAPLHGR Results for 9x9 Fuel," Exxon Nuclear Company, Inc.,

May 1986.

XN-NF-80-19(A), Volume 1, and Volume 1 Supplement 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors:

Neutronic Methods for Design and Analysis,"

Exxon Nuclear

Company, Inc., March 1983.

NED0-21231, "Banked Position Withdrawal Sequence,"

General Electric

Company, January 1977.

ANF-90-050, "Susquehanna Unit 1 Cycle 6 Reload Analysis:

Design and Safety Analyses,"

May 1990.

PLA-3407, "Proposed Amendment 132 to License No. NPF-14:

Unit 1 Cycle 6 Reload," Letter from H.

W. Keiser (PP&L) to W.

R. Butler (NRC), July 2, 1990.

Letter from James J. Raleigh (NRC) to H.

W. Keiser (PP&L), "Susquehanna Steam Electric Station, Unit 2, Cycle 5 Reload (TAC No. 79140)," April 22, 1991.

PLA-2885, "Proposed Amendment 52 to License No. NPF-22," Letter from H.

W.

Keiser (PP&L) to W.

R. Butler (NRC), June 30, 1987.

Letter from James Raleigh (NRC) to H.

W. Keiser (PP&L), Topical Report PL-NF-90-001, "Application of Reactor Analysis Methods for BWR Design and Analysis," Susquehanna Steam Electric Station, Units 1 and 2, November 21, 1991.

Letter from James Raleigh (NRC) to H.

W. Keiser (PP&L), Topical Report PL-NF-89-005, "gualification of Transient Analysis Methods for BWR Design and Analysis," Susquehanna Steam Electric Station, Units 1 and 2, March 1992.

PLA-3729, "Response to RAI on Transient Analysis Methods," Letter from H.,

W. Keiser (PP&L) to C. L. Miller (NRC), February 12, 1992.

PLA-3778, "Administrative Control of Flow Dependent MCPR Operating Limits for UIC7 Operation," Letter from H.

W. Keiser (PP&L) to C. L. Miller (NRC),

May 15, 1992.

PLA-3810, "Proposed Amendment 105 to License No. NPF-22:

Unit 2 Cycle 6

Reload," Letter from H.

W. Keiser (PP&L) to C. L. Miller (NRC), July 21, 1992.