ML17156A301

Safety Evaluation Supporting Amend 45 to License NPF-14

ML17156A301
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Site:	Susquehanna
Issue date:	05/22/1985
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION AMENDMENT NO. 45 TO NPF-14 SUS(}UEHANNA STEAM ELECTRIC STATION UNIT 1 DOCKET NO. 50-387 1.0 Introduction By letter dated January 15, 1985 as supplemented on February 21, 1985 from the Pennsylvania Power It Light Company, to the Director of Nuclear Reactor Regulation (Reference 1), Technical Specification changes were proposed for the operation of Susquehanna Unit 1 for Cycle 2 (SIC2) with a reload using Exxon manufactured fuel assemblies and Exxon analyses and methodologies.

Enclosed were the requested Technical Specification changes and a

number of reports (References 2-6) discussing the reload and analyses done to support and justify the second cycle operation with General Electric (GE) and Exxon fuel and the Technical Specification changes.

The subsequent letter dated February 21, 1985 (Reference 7), priI-,@rily provided results from several newer analyses covering (1) operation of Cycle 1 for longer exposure, (2)

LOCA at full power but 87 percent flow (Reference 8), and (3) feedwater controller failure analysis at reduced power, resulting in revised MCPR operating limits (with main turbine bypass inoperable).

Also submitted (and reviewed) in connection with this reload was a generic supplement on the fuel mechanical design (Re erence 11).

The submitted documents contained a large number of references, which describe and justify the Exxon methodology used to design the core and analyze the reactor and its components and relevant transient events.

These are given in References 9 through 36.

Cycle 2 for Susquehanna will be the first use of Exxon fuel and analysis in this reactor.

However, similar reloads with Exxon fuel have been done for Dresden 2 and 3, and these reloads and the associated Exxon methodologies were extensively reviewed and approved (see for example Refprence 25).

These methodologies are generally applicable and were used for SIC2 analyses.

Beyond the switch to Exxon-provided reload fuel, there is nothing unusual about SIC2, and,the proposed Technical Specification changes are entirely related to the use of Exxon fuel and accompanying analyses and methodology, terminology or related operational approaches..

The submittals (References 1-8) discuss (a) the reload core description, (b) the fuel mechanical

design, core thermal hydraulics, nuclear design and fuel storage critically, (c) the use of POWERPLEX for core monitoring, (d) transient and accident analyses, (e)

LOCA and ECCS analysis, and (f) the proposed Technical Specification changes.

The submittals have been reviewed by the staff.

The analyses and proposed Technical Specification changes have been found to be acceptable.

The following will cover some 8506060339 850522 PDR ADOCK 05000387 P

PDR

aspects of the review and will discuss all of the Technical Specificatiori

-changes.

For the most part Exxon methodology involved in this reloao has been reviewed in connection with the previous reload reviews and will not be discussed here.

Those methodology reports for which the review has been officially completed are noted in the references by the (A) in the report number.

Those few reports for which this is not yei thus indicated have either been reviewed in the course of this review (e.g.,

Pe.erence il) or the incomplete areas are not relevant to this review (e.g.,

Reference 15 and 19, for which the incomplete review of statistical uncertairty analysis is no. needed since bounding analyses were used for transient analyses).

The evaluation of the two reports on LOCA analyses required evaluation of the EKC topical report "Generic Break Spectrum Analyses for BWR 3 and 4

with hodified Low Pressure Coolant Injection Logic," XH-NF-84-71(P),

December, 1984 (Reference).

Although the staff approval of this generic report for the general class of plants for which this report was intended has not been completed, its application to Susquehanna Unit 1 is addressed and approved in this evaluation.

2.0 Fuel hechanical Desi n

2.1 The Susquehanna Unit 1 Cycle 2 core will consist of 192 fresh Exxon XH-1 Bx8 fuel assemblies, and 572 GE 8x8 fuel assemblies.

The Exxon XN-1 8x8 fuel design is described ir, the approved generic report on the jei-pump (JP)

B'liR fuel design (Reference 10).

However, several conditions of approval on Reference 10 are attached.

These conditions are:

(1)

The licensee must confirm that the design power profile shown in Fig. 5. 10 of Reference 10 bounds the power limits for the application in question.

(2)

Unless RODEX2 (Xt<-t<F-81-58) is approved without modificaiior., the licensee must confirm or redo the following analyses, which were reviewed on the basis of RODEX2 results:

design straip, external corrosion, rod pressure, overheating of fuel pellets, and pellet cladding interaction.

(3)

Until such time that the Exxon revised cladding swelling and rupture modes (Xh-liF-82-07) are approved and incorporated in the Exxon ECCS evaluation model, a supplemental calculation using the hUREG-0630 cladding models must be provided on a plant specific basis each time a

new ECCS analysis is performed.

(4)

The licensee must make sure that the fuel performance code that is used to initialize Chapter 15 accident analyses has current l PC approval.

We have evaluated these four conditions during the course of our review, and our conclusions are described in the following paragraphs.

The licensee stated in the Susquehanna Unit 1 Cycle 2 reload submittal (Reference

3) that the expected power history is bounded by the design profile-in Fig. 5. 10 of Reference 10.

We thus conclude that the Cycle 2

power history is within the design limit, and Condition 1 is satisfied.

2.1.2 RODEX2 -- Strain, Corrosion, Rod Pressure, Overheating'f Fuel Pe lets, and Pe et Clad Interaction PCI Anal ses The analyses of strain, corrosion, rod pressure, overheating of fuel pellets<

and PCI were described in the approved jet pump BWR fuel design topical report.

We have completed the review of the RODEX2 code used in this analysis and approved it with some modifications for licensing applications (Reference 26).

Using the approved version of the RODEX2

code, Exxon provided supplemental calculations to demonstrate that the design limits on these physical parameters would not be exceeded throughout the entire lifetime (Reference 11).

Since these analyses bound the Cycle 2

applications, we conclude that these analyses are acceptable for Cycle 2.

2.1.3 Claddin Swellin and Ru ture The cladding swelling and rupture models in Reference 18 (Exxon Nuclear Company ECCS Cladding Swelling and Rupture tlodel) have been approved (Reference

27) for use in the Exxon ECCS.evaluation model and have been incorporated in the approved, Exxon EXEH/BWR ECCS model.

Since Exxon used this approved swelling and rupture model for cladding in ECCS analysis, Condition 3 has been satisfied.

2.1.4 LOCA Initial Conditions Exxon used the recently approved steady state

code, RODEX2 (Reference 12) to calculate Cycle 2 LOCA initial conditions including stored energy and rod pressure f'r the Exxon EXEH/BWR evaluation model.

Thus Condition 4 is satisfied by the use of the approved code RODEX2.

2.2 LHGR Limits For GE fuel SIC2 will retain the Cycle 1 Technical Specification LHGR limits (13.4 kw/ft) and APRN setdown for excessive peaking factor at part power operation.

For Exxon fuel, however, a specific Technical Specification LHGR limit and setdown are not required.

Operation for SIC2 will remain within the limits given in Figure 5.10 of Reference 10 and thus fuel design limits will not be exceeded during overpower conditions.

To assure that the limits of Figure 5.10 are met, daily surveillance of power distributions relevant to this limit will be carried out using the POWERPLEX monitoring system.

2.3 Seismic -

LOCA Mechanical Res onse The mechanical response of Exxon fuel assemblies to design Seismic-LOCA events is'ssentially the same as-it is for GE assemblies.

The channel boxes were manufactured for these assemblies to GE design criteria and dimensions.

.The analysis indicating that design limits are not exceeded is acceptable.

3.0 Thermal-H draulic Desi n

Exxon thermal-hydraulic methodology and criteria are presented in the reports of References 9, 13, 14, 19, 20, 21, and 22 which have been reviewed and.approved with the exception of some statistical aspects of Reference 19 which are not needed for SIC2 since bounding transient analyses are used.

These methods have been approved for the Dresden review.

These reviews concluded that hydraulic compatibility between GE and Exxon fuel is satisfactory.

The calculation of core bypass flow and the safety limit MCPR are also acceptable.

The core stability, for which Susquehanna has Technical Specifications implementing surveillance for detecting and suppressing power oscillations (approved in Reference

23) is also satisfactory.

MCPR limits are discussed below and are'also discussed in connection with the Technical Specifications.

3.1 Recirculation Pum Run-u Events

.2 ~kd The minimum critical power ratio -(MCPR) operating limit at full recirculation flow is determined by calculating the plant response to anticipated operational transients which are expected to be the most limiting transients at rated conditions.

Analysis of recirculation pump run-up events is needed to determine the need for an increase in the above MCPR when operating from initial conditions at less than rated recirculation pump capacity.

These analyses are necessary since increase in pump flow can cause significant increases in reactor power.

3.1.3 Evaluation For recirculation pump run-up events during manual flow control operations which could occur, for example, as the result of faulty signals, the required increase in the MCPR at the initial low flow is that required to prevent the MCPR from dropping below the minimum critical power ratio safety limit during the transient.'or EHC fuel in Susquehanna Unit 1, this safety limit is 1.06.

For recirculation pump run-up events during automatic flow control operation, the minimum critical power ratio during the transient should not decrease below the full power, full flow, operating limit.

Hence, the increase in the minimum critical power ratio operating limit,at low flows for automatic flow control operation should be larger than that for manual flow control operation.

The calculations of Reference 28 deal only with manual flow control.

Hence, automatic flow control should be prohibited in the Technical Specifications for

~ Susquehanna Unit 1.

The staff understands that automatic flow control operation is not part o-the Susquehanna Plant design.

The calculations of Refelence 28 were made for both single and two pump excursions ouring manual low control operation.

The methodology used is consistent with that used and approved for licensing the Dresden units (reference 29).

The calculations for single pump excursions, ranging from gradual and intermediate to rapid pump run-ups, indicated thai ihe safety limit of 1.06 was not vic.ated during the events.

Hence s',ngle pump run-up events cic nor require an increase in the minimum critical power ratio operating limit at low flow.

Simultaneous increase in the speed of boih recirculation pumps could result from a faulty signal in the master flow controller.

6ecause Gf 'ihe design of the control

sysiem, ihe expected pump responses to aulty signals originating at the master flow controller are gradual increases in pump speed.

In this case, the relatively slow pcwer increases are accompanied by approximatelv eqUivalent increases in the fuel su)face heat flux.

Calculations simulating a gradual increase in controller demand

( less than 1 percent rated speed per second) were made using a voic reactivity feedback 25 percent more reactive thar. expected and a Doppler feedback 10 percent less reac-.ive than expecied.

The results

',nd;cated that the critical power ratio could Crop below the safety limit during the ransieni.

Hence, an increase in the minimum critical powel l 2 io operating limit was needed a

reduced flow.

Thc calculated power/flow ratio durino this ever,.

was conservatively extrapolated past the predicated scram point to the maximum allowea flow of 105 percent flov'.

This power-flow relation was ther; used to calculate the minimum critical power ratio operating limit a various reouced flows.

The procedure involves calcuia-.ion of the change in the t~iCPR alone this path while main aining the NCPR at the 105 percen-. flow poiri ai the safeiy limit.

The results of this calculation are presented in Figure 1. 1 of Reference 28.

The cvcle specific HCPR operating limit for Susquehanna Unit 1 was stated

'.n Re.erence 28 to be the maximum of the reduced flow operating limit of this figure or the full flow HCPR operating limit.

He conclude that this calcula"ion of the reduced flow hCPR operating limits for Susquehanna Unit

's acceptable, provided operation in the automatic flow control mode is prohibited.

The staff has confirmed with the licensee that automatic flow control operation is noi part, v-the Susquehanna plant design.

As a result the licensee has stated thai this mode woula not be and could not be used.

4.0 Nuclear Design Exxon nuclear desigr methodologies have been approved (Reference 14).

The SIC2 reload replaces about one quarter of Cycle 1 fuel with new Exxon fuel.

The loading pattern is a normal type of scattered corfiguration.

The axial maximum planar average enr icllmeni of the new assembly is 2.8'. percent U235.

The beginning of cycle shutdown margin is calculated to be 3.63 percent

h. k, the R fac7or is 1.45 percents k, and thus the cycle minimum shutdown margin is 2.18 percent b K, well in excess of the required G.38 percent b k.

The Standby Liquid Conirol System also fully meets shutdown requiremer.s.

The existing new fuel storage calcu',ations are based on k

~f of the asser'b'.y.

if the maximum enrichment zone is such that k

. is less ihan 1.3C at 1 imitlllg state condi tions then the required cr itical ity 1 imits are r e' For the Exxon fuel k f under these conditions is 1.13 and the criterion is met.

The existing spent fuel pooi criticality calculations have met ef criteria using a U235 asserbly avelace enrichment of 3.25 percen ard nc burnable poison.

Since the maximum corresponding enrichi;en.

o.

-he new fuel is 2.81 percenT. the previous calculations are still acceptable.

Si'scuehanna will use the Exxori PCL'ERPLEX col"e lioni i.oring sysTem o nonitor reactor parameters.

Ve have not specifiicallg reviev'ed details o

this system (nor have we in the past reviewed details of the GE process computer monitorino sysi.em),

but we have reviewhc the principal methodologies

. involved in the sy'tem and consioer them to be appropriate and acceptable.

The system has been in use during Cycle 1 ard has provided suitable mionitoring and predictive results.

5.0 Transient and Accident Analyses The Exxon transient methocology is described in Re erence 5.

This methodology is generally approved and was ir. Dresden analyses.

Aspects oi the methodoiogy review riot et completed involve statistica anaiyses which were not used in the SIC2 analyses since boundinc parareters were used in the calculations (because a non-transient local event, Roa 'k'ithcrawal

Error, v as flCPR limiting).

1 Exxon examined the desigr, events discussed in Refereiice

'.5 arc the SIC2 submittal s (Peference 2, 3, and 4) presented results for the more limiting events.

These included Gerierator Load Rejection without Bypass (LRWOB),

Feedwater Cnriti'olier Failure (FMCF) and Rod Withdrawal Error (RWE).

Results for these events were also presented in Reference 7 for the extended burnup for Cycle 1, and the FWCF evert v~as also reanalyzed at 80 percent power.

The transients were also analyzed with End of Cycle Recirculation Pump Trip (EOC-RPT) or with tiain Turbine Bypass (HTB) inoperable.

The RWE was analyzed for a range of Rod Block tlonitor

settings, including values of 1.06 and 1.08 used in the Technica:

Specifications.

These various analyses v.ere used io determine zhe Techi'lica, Specification MCPR limits.

in general the RWE is the limiting even:

(by a larce margin) but for the non iiormal operation with HTB inoperablE and exierded Cycle 1

burnup, the FWCF, at 80 percent

power, became limitirg.

Since initially (and for normal operatiori and for EOC-RPT inoperable)

EWE v as liiriting, the statistical convolution o

parameters was not used for the zransient analyses.

Also, the Stanoard Technical Speciification scram times were used for all analyses so no scram speed adjustmenzs to the HCPR lim'.ts is necessary in the llew specifications.

Reduced flov operation (presented in Reference 5) v.'as aralyzed for manual flow conTrol mode only.

tutor.;aiic flow control is therefore riot allowed.

These results became part of the Techr!ical Specification HCPR 1 iiiiis anG became a factor belcw about 55 percent of rated core flow.

The rod drop accioent was analyzed with approved methodology

{Re. erence 14).

The resultinc maximum fuel enthalpy o

153 cal/gm is well below the limit of 280 cal/gm.

The analysis assumed a control rod reactivity worth whiich requires the use of GE's Banked Position Withdrawal Sequence.

Our review of the transient and accident analyses done for SIC2 reload indicates

.hat appropriate methodology and input have been used and the results provide a suitable basis for the Technical Specification changes made ir, support of cycle 2 operation.

5.

~OCA 5.1.1 Break S ectrum Calculations 5..!

On December 31, 1984, the Exxon Nuclear Company, inc.

(ENC) submitted the topical report XN-NF-84-117(P) "Generic LOCA Break Spectrum Analysis BllR 3 and 4 with licdifiea Low Pressure Coolant Injectior, Lcoic," for staff review and evaluation (Reference 24).

As noted in the submittal, the Pennsp'lvania Power and Light Company was planning tc reference this report in their application for a license amendment or the Susquehanna Steam Electric Station, bnit 1.

The subject report is based on ENC methods for ECCS

-calculations for BWR jet pump plants that are describeo ir the ENC licensing topical reports of References 30 throuch 33.

Reference 30 includes an overall description of EXEY the'ECCS Evaluation Hodel for 6oi iing Water Reactors and a discussion of the conf'crmance of the model to the requirements of 10 CFR 50, Appendix K.

References 31 ano 32 ceal with model changes to the non-jet pump model to account for iet pump features.

Reference 33 describes the veri ication and qualification stud',es of'he model.

he staff reviewed and evaluated

.he above r.",ethods and approved them in Reference 34.

In Reference 35, ENC applied the approved EXEM model to a generic break spectrum analysis.

This report was to be referenced as a lead plant analysis for BWR 3 and 4 plants with Low Pressure Coolant Injection (LPCI) loop selection logic in support of the break spectrur.. analysis required by 10 CFR 50, Appendix l',.

This report was also approved hy the sia f in Reference 34.

The subject report is a similar application of the approvec EXEh model to BWR 3 and 4 plants which have the modif'ied loop seleciior 1 ogle.

5.1.3 Evaluation The subject report, "Gene> ic LOCA Break'Spectrum Analysis GWR

"- and 4 with Hodified Low Pressure Coolant Injection Logic," uses EflC methods for ECCS calculations that were apprcved by the staff in Refer<rice 34.

These methods are described in several EHC licensing topical repor.s (Pef'erence 30 through 33).

The ENC gercric model, EXEfl, incorporates the RELAX code for system blcwdowr~ caicuiations, the FLEX code f'oI refill/reflood calculations and HLXY/BULGEX for the hoi assembly heatup calculations.

As noted in the subiect report, some minor updates have been r;,ade to RELAX and FLEX since the previously approved calculational results.

It was stated by EHC that these changes would have little or no e feet on the calculated results.

Hence, we conclude that the methodology used to obtain

.he results described in the subject report is acceptable.

The lead plant chosen by EHC is Susquehanna Unit 1, a

BWR 4 with a 251 inch diar eter pressure vessel and modified loop selection logic.

For a

BWR 4 with modified loop selection logic, the HSSS vendor selected the James A.

=Fitzpatrick Hu'clear Power Plant, a

BQR 4 with a 218 inch diameter pressure

vessel, as the lead plant (Reference 36).

The lead plant calculations by the NSSS vendor indicated that the limiting break was located in'he recirculation loop piping.

The largest diameter recirculation loop pipes are a) the suction line between the reactor vessel and the recirculation pump and b) -the discharge pipe between the recirculation pump and the toroidal header.

The most limiting single failure was the failure of'he Low Pressure Coolant Injection (LPCI) injection valve in the intact loop to open.

For this failure, the NSSS vendor calculations indicated that the most limiting break location was in the pump discharge line.

On the basis of the previous NSSS vendor calculations, EHC selected the injection valve failure to open as the most limiting single failure.

For this failure, the safety systems assumed to be operational were the high pressure coolant injection system, the LPCI on the broken loop (assumed during blowdown only), two low pressure core spray systems and the automatic depressurization system.

The break location analysis bv EHC involved double-ended guillotine (DEG) breaks on either side of the recirculation pump, with discharge coefficients of 1.0.

These calculations confirmed that the limiting break location was on the discharge side of the recirculation pump.

The break spectrum calculations at the limiting break location by ENC included double-ended guillotine breaks in the discharge pipe with discharge coefficients of 1.0, 0.8, 0.6 and 0.4.

Split break calculations included breaks ranging in size from the smallest equivalent guillotine break (2.8 ft~) down to areas of 0.3, 0.2, 0.1 and 0.05 times the double-ended break cross-sectional area of 7.0 square feet.

All calculations were for a core composed of EHC fuel at nominal beginning of life.,conditions.

The calculations for the guillotine break show an increase in the peak clad temperature (PCT) as the discharge coefficient was reduced,,with

=he maximum peak clad temperature occurring at the lowest discharge coefficient of 0.4 considered in the analysis.

Values of the discharge coefficient below 0.4 are considered unrealistic.

The calculations for the split breaks indicated lower peak clad temperature values than those gu-;llotine breaks at the same area (2.8 ft~) and exhibited an abrupt decrease in peak clad temperature at the smaller break size of 0.7 ft'.

EHC stated that the results of the report were intended to be generic for a%1 BWR 3 and 4 plants with modified loop selection logic.

The s ~aff has not completed the evaluation.of this generic application.

However, we conclude that the results are acceptable for the Susquehanna Unit 1

submittal in determining the limiting break and meeting the requi rements of 10 CFR, Appendix V. with respect to the break spectrum calculations.

5. 1.4 l~APLHGR Limits

- As discussed

above, the generic break spectrum analyses of Re erence 24 indicated that the limiting break for Suscuehanna Unit 1 is a double-ended guillotine (DEG) break in the recirculation system discharge piping with a discharge coefficient of 0.4.

In Reference 6 the limiting break boundary conditions were used to calculate the exposure dependen.

IQPHGR limit for

'ENC uel from beginning of life to an assembly exposure of 35 GWD/tlTH.

The calculations were made using staff approved methods arid were for an initial full power, full recirculation flow condition.

The YAPLHGR results for Susquehanna Unit 1 with ENC 8x8 Reload Fuel are presenteo in Figure 2.1 of Re erence 6.

Reference 8 gives the results of ENC calculations to determine if the lJPLHGR limits of Reference 6 that were established at 100 percent power/'00 percent flow conditions would be applicable to the range of f low concitions allowed by the power-low operating region at Susquehanna Unit 1.

In Reference 8, it is stated that the 87 percent flow operating poirt is the lowest flow at which full power operat'.on is peni'iizzed at Susquehanra Unit '.

Hence, an ECCS analysis of the limiting break was made fo1 iniitial conditions of full power and 87 percent flow.

Other iriitial conditions were the same as those for the full power, full flow case.

The maxiruv.

pcwer in the limiting assembly v'as based on ar assumed minimum critical power ratio of 1.24 for both calculations.

To maintain the same hCPR at the lower flow, the hot channel ccrc-wide radial peaking -"actor was reduced.

The lower assembly power gave'improved fluid conditions during blowdown, and hence, lower peak clad zemperature.

A calculatiori at an assembly exposure of 19 G'llC/l'Tti (the most limiting exposure in the tVPLHGP. calculations for the full power, full riow case) indicated a 28'F decrease in the peak clad temperature.

This type of decrease in peak clad temperature would be expected over

.he expected range of fuel exposures; lie conclude that the ENC analyses of VAPLHGR limits at Susquehanna Unit 1 are applicable to the full range of power/flow conditions permitted at the plart and are acceptable.

6.G Technical S ecificaiion

.Changes The following Technical Specification changes have been requested to accommodate the change to Exxon uel and methodology.

(1), Definition of Average 8undle Exposure:

This is a necessary addition to match the parameter used in Exxon analysis methodology for MAPLHGR and is acceptable.

(2) 3/4

~ 1. 2 and 8 3/4. 1. 2:

The change to the definition of reacti viiy anomaly from one of a control rod density anomaly (measured-predicted deviation) to a monitored core k ff anomaly reflects the use of a more direct parameter for the anomaly.

Rod density has been used as a less direct indicator.

PORERPLEK, which maintains a consistent methodology between active oetermination and prediction, can monitor k ff directly.

This change is acceptable.

~

~

,l )

(s fs

- 10 (3).

(4) 3/4.2.1 and 8 3/4.2. 1:

This is a change to the use of the Exxon definition of Average Bundle Exposure for Exxon fuel and the removal of l1APLHGR curves for the GE fuel no longer used, and the addition of HAPLHGR curves for the GE and Exxon fuel as calculated with Exxon methodology.

The GE methodology discussion and Table are removed.

These changes are acceptable.

3/4.2.2 and 8 3/4.2.2'.

This change notes that the requirement to lower the APRt) setpoint for excessive peaking at part power (NFLPD exceeds FRTP) for GE fuel does not apply to Exxon fuel.

This was previously discussed in Section 2.2.

The change is acceptable.

3/4.2.3 and 8 3/4.2.3:

This change removes the elements of the GE methodology for determining HCPR limits, including the variation with scram insertion time and the K

function, and replaces them with the results of the Exxon methodology and analyses for SIC2.

The new SICPR limits are principally single value functions of (1)

GE or Exxon fuel, (2)

RBN setpoint, (3)

EOC-RPT operability and (4) tlTB operability.

51CPR is also limited, however, by reduced flow operation.

As previously discussed these values are the results of Exxon's calculations of transients and are primarily controlled by the RWE.

The values to be used for Table 3.2.3-1 are not those of the original submittal, but those of Reference 7 from analyses using the revised burnup parameters and the FWCF analysis at reduced power.

These changes are acceptable.

The Bases changes primarily reflect the change to reference of Exxon methods and are acceptable.

3/4.2.4:

This=change indicai'es, as discussed previously, that the 13.4 kw/ft LHGR limit applies only to GE fuel. It is acceptable.

(7) 3.3.4.2:

This change reflects the fact that (in 3/4.2.3) l1CPR limits

, are available from calculations with EOC-RPT not in operation.

Thus operation can continue if these t'ICPR limits are met.

This is acceptabl e.

(8) 3.7.8:

This is a minor word change and am<plification.

It is acceptable.

(9) 8 2.0 and 8 2. 1. 1:

These changes refer to Exxon fuel and correlation.

They-are acceptable.

( 10) 8.2.1.2:

These changes remove the discussion of GE GETAB methodology, reports and data and refer instead to Exxon methodology.

They are acceptable.

(11) Bases page 8 2-7:

This addition refers to the previously discussed conclusion that APRH trip setdown is not required for Exxon fuel and is acceptable.

( 12) 8 3/4.1.3 and 8 3/4 1.4:

This is a change to refer to Exxon methodology and reports and is acceptable.

P,

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Conclusions We have reviewed the reports submitted for the Cycle 2 reload of Susquehanna Unit 1.with Exxon fuel and with Exxon methodology and analysis.

Based on this review we conclude 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 changes,and reload parameters.

The NRC staff finds the Susquehanna Unit 1 Cycle 2 Technical Specification changes submitted to support this reload acceptable.

REFERENCES (1)

Letter from B.D. Kenyon of Pennsylvania Power 8 Light Company to Director NRR, "Susquehanna Steam Electric Station Proposed Amendment 59 to License

.No. HPF-14," January 15, 1985.

(2)

NPE-84-015, "Susquehanna SES Unit 1 Cycle 2 Reload Summary Report",

December 1984.

(3)

XH-NF-116, "Susquehanna Unit 1 Cycle 2 Reload Analysis," Exxon Nuclear Co.,

December 1984.

(4)

XN-NF-84-118, "Susquehanna Unit 1 Cycle 2 Plant Transient Analysis,"

Exxon Nuclear Co.,

December 1984.

(5)

XH-NF-84-118, Supplement 1,

"Susquehanna Unit 1 Cycle 2 Plant Transient Analysis:

Recirculation Pump Run-up Results',"

Exxon Nuclear Co.,

December 19G4.

(6)

XN-NF-84-119, "Susquehanna Unit 1 LOCA-ECCS Analysis MAPLHGR Results,"

Exxon Nuclear Co., December 1984.

(7)

Letter from N. Curtis of Pennsylvania Power E Light Company to Director NRR, "Susquehanna Steam Electric Station Supplement to Proposed Amendment 59 to License Number HPF-14," February 21, 1985.

(8)

XN-NF-85-14, "ECCS Analysis for Susquehanna Unit 1 and.at Full Power and 87 percent Flow", February 1985.

(9)

XN-NF-80-19(A), Vol. 4, "Exxon Nuclear Methodology for Boiling Water Reactors:

Application of the ENC Methodology to BWR Reloads,"

Exxon Nuclear Co., September 1983.

(10) XN-HF-81-21(A), Rev.

1, "Generic Mechanical Design for Exxor. Nuclear Jet Pump BWR'eload Fuel",

Exxon Nuclear Co., September 1982.

(ll) XN-NF-81-21(P),

Rev.

1 Supplement 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR 8xS Reload Fuel",

Exxon Nuclear Co.,

March 1985.

(12)

(13)

(14)

(15)

(i6)

(17)

(iS)

(ig)

(20)

(21)

\\

(22)

(23)

(24)

(25)

(26)

(27)

(2S)

XN-NF-81-58(A), Supplements 1&2, Rev. 2, "RODEX2 Fuel Rod Thermal-tlechanical

Response

Evaluation Yiodel", Exxon Nuclear Co., triarch 1984.

XH-NF-525(A), Rev. 1, "Exxon Nuclear Critical Power methodology for Boiling Water Reactors",

Exxon Nuclear Co.,

November 1983.

XN-NF-80-19(A), Vol. 1, and Vol..l Supplements 1&2, "Exxon Nuclear Hethodology for Boiling Water Reactors:

Neutronic methods for Design and Analysis,"

Exxon Nuclear Co., Harch 1983.

XN-NF-79-71(P), Rev. 2, "Exxon Nuclear Plant Transient t1ethodology for Boiling Water Reactors,"

Exxon Nuclear Co.,

November 1981.

XN-HF-80-19(A), Vols. 2, 2A, 2B,

& 2C, "Exxon Nuclear methodology for Boiling Water Reactors:

EXEM BWR ECCS Evaluation tlodel," Exxon Nuclear Co., September 1982.

.h XN-HF-CC-33(A), Rev.

1, "HUXY:

A Generalized l1ultirod Heatup Code with 10 CFR 50 Appendix K Heatup Option," Exxon Nuclear Co.,

Novmeber 1975.

XN-HF-82-07(P), Pev.

1, "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Yiodel," Exxon Nuclear Co.,

November 1982.

XN-NF-80-19(P), Vol. 3, Rev. 1, "Exxon Nuclear Methodology for Boiling Mater Reactors THERt<EX:

Thermal Limits t1ethodology Summary Description,"

Exxon Nuclear Co., April 1981.

XN-NF-81-11(A), "Generic Statistical Uncertainty Analysis t1ethodology,"

Exxon Nuclear Co., November 1983.

XN-NF-512(A), Rev. 1, and Supplement Rev.

1, "XN-3 Critical Power Correlation,"

Exxon Nuclear Co., October 1982.

XN-NF-79-59(A), "methodology for Calculation of Pressure Drop in BWR fuel Assemblies,"

Exxon Nuclear Co.,

November 1983.

PLA-2339, "Facility Operating License NPF-14 Condition 2.C.(4)(b),"

Pennsylvania Power

& Light Co.,

November 1983.

XN-HF-84-117(P), "Generic LOCA Break Spectrum Analysis:

BWR 3 and 4

with Modified Low Pressure Coolant Injection Logic," Exxon Nuclear Co.,

December 1984.

Letter from D. Crutchfield, HRR, to D. Farrar, Comonwealth Edison,"

"Cycle 9 Reload-Dresden Station, Unit 2" Amendment 75, April 7, 1983.

Letter from C. Thomas, NRC, to J.

C. Chandler,

Exxon, November 16, 1983.

Memorandum from L. S. Rubenstein, NRC, to T. Novak, HRC, October 6, 1982.

"Susquehanna Unit 1 Cycle 2 Plant Transient Analysis Recirculation Pump Run-up Results,"

XN-NF-84-117(P), December, 1984.

~

~

-'3-(29) "Dresden Unit 3 Analysis for Reduced Flow Operation," XN-NF-81-84(P),

January, 1982.

(30) "Exxon Nuclear Hethodology for Boiling l'ater Reactors-EXEH:

ECCS Evaluation Hodel Surrmary Description" Xti-t/F-80-19(P), Volurie 2, tiay 1980.

(31) "Exxon nucleal Hethodology for Boiling Water Reactor s-RELAX-A RELAPSE Based Conputer Code for Calculatina Blowdown Phenomena,"

Xti-NF-80-19(P),

Volume 2A, Hay 1980.

(32) "Exxon tIuclear Hethodology for Boiling ltater Reactors-FLEX; A Conputer Code for Jet Pump BWR Refill and Reflood Analysis," XN-NF-80-19(P),

Volune 2B, tray 1980.

(33) "Exxon Nuclear Hethodology for Boiling Water Reactors

- Verification and Qualification of EXEtl, XN-NF-80-19(P), Volume 2C, June 1981.

(34) Letter from James Hiller, NRR to G.F. Owsley, ENC, "Acceptance for Referencing of Topical Report XN-NF-80-19(P), Volune 2, 2A, 28 and 2C and Topical Report Xtl-ttF-81-71(P)", January 27, 1982.

(35) "Generic Jet-Pump BWR3 LOCA Analysis Using the ENC EXEH Evaluatior, t'iodel" XN-NF-81-81(P)(A) Supplement 1 (P)(A), September 1982.

(36)

"LOCA Analysis Report for James A. Fitzpatrick Nuclear Power Plant (Lead Plant)," NED0-21662-2, July 1977.

Environmental Consideration This amenament involves a change in the installation or use of a facility component located within the restricte'd area as defined in 10 CFR'ar.

20.

The staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any efiluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure.

The Corrmission has previously issued a proposed finding that this amendnent involves no significant hazards consideration and there has been no public corrzient on such finding.

Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.72(c)(9).

Pursuant zo 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

'I Conclusion We have 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 Comission's regulations and the issuance of this amendment will be inimical to the corrmon defense and security or to the health and safety of the public.

Dated:

NY 3 3 585

4 iy