ML19344E242

Safety Evaluation & Eia Supporting Amend 50 to License DPR-40

ML19344E242
Person / Time
Site:	Fort Calhoun
Issue date:	08/15/1980
From:	Office of Nuclear Reactor Regulation
To:
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ML19344E237	List: ML19344E236 ML19344E238 ML19344E242 ML19344E244
References
NUDOCS 8008280223
Download: ML19344E242 (17)

Text

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION y

E WASHINGTON, D. C. 20555 r

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SAFETY EVALUATION AND ENVIRONMENTAL IMPACT APPRAISAL h

BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

SUPPORTING FACILITY OPERATION AT 1500 MW FOR i

FACILITY OPERATING LICENSE N0. DPR-40 OMAHA PUBLIC POWER DISTRICT FORT CALHOUN STATION, UNIT N0. 1

[

DOCKET N0. 50-285 I.

INTRODUCTION By application dated July 17, 1979, as supplemented October 30 and December 4, 1979, January 30, 1980, February 12 and 25,1980, March 12,1980, May 21,1980,

~

and June 2c,1980, Omaha Puelic Pcwer District (0 PPD or the licensee) requested an amendment to the Technical Specifications (TS) appended to Facility Operating License No. DPR-40 to allcw operation of Ft. Calhoun at 1500 MWt. We completed portions of our review cf the OPPC application to increase Ft. Calhcun power tc 1500 F".!t as part of the Cycle 6 reload review which was approved by Amendment 47 issued on April 1, 1980. The following evaluation includes tnese pertinent portions of that review and the results of further reviews cerformed by the staff for 1500 FWt operation.

I!.

DISCUSSION AND SAFETY EVALUATION The Ft. Calhoun core for Cycle 6 operation contains 40 fresh Exxcn Nuclear Company, Inc. (ENC or Exxon) fuel assemclies and 93 burned Combustion Engineering Company (CE) assecolies.

ENC performed all the safety analyses for Ft. Calhoun Cycle 6 with two exceo: ions.

OPPD performed the calculations te cetermine the input to the CECOR procram used to calculate the core pcwer distribution and CE performec the Small Ereak Loss cf Ccclant Accicent (LOCA: calculaticn for Ft. Calhoun with Exxcn fuel in the core.

Tnese are discussed in later sections of the Safety Evaluation.

The licensee proposed to operate Ft. Calhoun at a' power level cf 1500 l*W: (5.65 above the currently licensed pcwer of 1420 MW:).

He have reviewed tne acclicaticn and supclemental information, and have reached the fellowine conclusiens.

FUEL CESIGN ine Ft. Calhoun reactc consists 5f 133 fuel assemblies, each having a 14xl* fuel roc array.

Each fuel assembly contains 176 fuel rods and five guice tuces.

The fuel refs consis cf slightly enriched n

8008280 213

=

+

d 2

1. i a

i U02 pellets inserted into Zircaloy cladding.

The control element

?

assembly (CEA) guide tubes,and instrumentation tubes are also made f.

of Zircaley. Each ENC assembly contains nine Zircaloy spacers

]

with Inconel springs; eight of the spacers are located within the i

active fuel region.

=

5 The Cycle 6 loading pattern ir, shown in Figure 3.1 of Reference 19.

The initial enrichments of the various fuel batches are listed in Table 3.1 of Reference 19. The beginning of cycle (SOC) 6 expcsures

~;

along with the batch ids are: shewn in Figure 3.2 of Reference 19.

The core consists of 40 fresh ENC assemblies at 3.5 w/o U-235 and

='

93 exposed CE assemblies scatter-loaded throughout the interior cf the core.

Pertinent fuel assembly parameters for the Cycle 6

=

4 fuel are given in Table 3.1 of Reference 19.

DEM0hSTRATION ASSEMBLY The licensee precosed to retain ene CE fuel assembly (criginally loaded in Cycle 2) during Cycle 6 to aid in obtain.19 data on high burnup fuel perf ormance. Tne licensee stated (Reference 36) tnat cischarge burnup of this assemoly could be as hign as 45,000 Megawat: cays per Metric Ton Uranium (Mwd /MTU).

c The licensee's performance evaluaticn for the demonstration assembly was based en a Cycle 5 length of 10,500 Mwd /MTU and a projected Cycle 6 length of 10,000 Mwd /MTU.

It is applicaole

~

to-any combination of cycle lengths no greater than a two-cycle length cf 20,500 Mwd /MTU. The licensee stated that this safety analysis is applicable tc the operating conditicas and TS of Cycle 3 or :ne procesed conditic s and TS cf Cycle 6 including che ir. crease in ocwer level to '600 MW:.

Several aspects of fuel cenavior beccme mere important to safety at high burnuo.

Among these are fission gas release which increases the internal cressure of the fuel rod, fuel red bewing, cladding collarse, ccerosien and hydriding.

Increased fissien cas release has its most significant effects en clacding collapse and LOCA

erformance. The licensee has demonstrated, as discussed belew,

na: :ne respense of :ne fuel rce to both these considerations is accectable.

The hc: red gas ;ressure was calculated to be below the minimum cperating ccolant pressure allowed by the TS.

The licensee performed an analytical prediction of cladding cree;-

cellarse time fer the cemenstra:icn assemoly. Using the ccmouter 8

9

=

7 i

s:=

a. i

+

code CEPAN (Reference 30), the licensee concludes that no creep-i collapse will be experienced by this assembly during Cycle 6.

i Time to cladding creep-collapse for the demonstration assembly

.g is predicted to be greater than 45,000 effective full power hours 3

(EFPH), while the cumulati.ve exoosure expected at the end of l

Cycle 6 is less than 40,000 EFPH.

Fuel rod bowing effects en departure from nucleate boiling (DNB) margin for the high burnup demonstration assembly during Cycle 6

+

have been evaluated with the guidelines set forth in Reference 31.

Since the demonstration assembly reached a burnup cf less than j

45,000 Mwd /MTU at end of Cycle 5, the fuel rod bowing penalty T

on DNS prescribed by Reference 31 would be less than 7%. Hcwever, the assembly never achieves radial peaks within 30% of the maximum radial peak in the core at any time during Cycle 6.

Therefore,

~

no penalty on core DNB margin is required.

T The licensee evaluated the perfcrmance of this demonstration fuel assembly by performing an analysis according to the Acceptance Criteria fer Light-Water-Cocied Reactors as presented in 10 CFR 50.46 (Reference 17). The calculated peak cladding temperature was 1303 F and the maximum amount of local Zirconium / water reaction was 0.13%.

These values are well below the criteria of Reference 17.

The analysis was perforned for a core power of 1420 MWt.

The licensee stated that an increase to 1500 MWt wculd not significantly affect the results. We concur.

Sufficient data are available to provide reasonable assurance :nat tnis fuel assembly will not experience excessive corrosion er hydriding during the planned irradia: ion period as icng as the usual cperating coolant chemistry is maintained.

fiUCLEAR DESIGN The nuclear design of the Ft. Calhoun ccre for Cycle 6 at 1500 f:Wt cperation was done with cethods used by Exxon in the past (References 5 :nrough 7 and 9).

These have been approved cy :ne NRC staff.

The peakinc f ac:crs FI and FI remain unchanced.

The shutdcwn rargin also remains UNEhanged Ss does the power dependen insertion limits.

Fuel rce bewing has an effect on the nuclear design since lateral revement of a fuel rod changes the local moceraticn.

Exxon has submitted a tecical report to the NRC staff describing the methces crcpcsed fer fuel compatible wi n CE reactors such as Ft. Calhcun.

9 y

,e-+

C 4

5 j

.4 I

she conclusion of the repcrt is that, at the maximum expected i

burnup of an Exxon fuel assembly, no aeditional multiplier i

on Fq need be applied to account for fuel rod bowing.

While we do net yet agree that this cenclusion is valid at the i

maximum expected burnup, we do agree that it is valid for the expected length of Cycle 6 at Ft. Calhoun.

Befere Cycle 7 the validity of the Exxen red bewing medel fcr Ft.

=

Calhoun must be reassessed.

THERMAL HYORAULIC DESIGN Cycle 6 centains both fresh Exxcr fuel and CE fuel which has been ourced for one or more cycles.

7 The Exxon fuel is designed to be ccccatible with the already burned CE fuel assemblies.

The licensee has conducted hydraulic compatibility tests which serve twc purposes.

These tests de:cn-strate that the flow cistributien in either fuel type is net certurbed excessively by the presence of the other.

The tests also provide :ne input data recuired to cerfcrm the safety analyses.

These tests were cencucted similarly to those cescriced in Reference 29.

The results of these flow tests incicate that the Exxon fuel has a higner flow resistance nan :ne CE fuel.

This difference is acccunted for in the safety analyses.

Table i belcw gives a brief thermal-hydraulic ccmaarison cf the Exxon and CE fuel casigns fcr F:. Calheun Cycle 6.

. n.u. : 1 Cc bustion Oesign Fac:ce E:C Fuel Engineering Clac ID (in.)

0.27E C.38c Ciac CD (in. )

C.;;2 0.440 Cen:rcl Rod CD (in.)

1.llE 1.115 Eare rec ficw area (in )

26.25 35.lE

%ettec perime:er (in. )

261. 9 260.3 Meatec cerimeter (in. )

220.0 243.3 O

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?i hence results in an initial MDNSR lower than actually anticipated.

t Thus, the initial MDNER is set at a low value providing additional i

conservatism in the initial plant conditions to evaluate thermal j

margins.

i

- Reference 2 reported the results of the postulated Uncontrolled CEA Withdrawal event for Ft. Calhoun Cycle 6.

The results of the r

analysis of this event show that the Thermal Margin /Lew Pressure (TM/LP) trip was the first Reactor Protection System respense over the _whole range of reactivity insertion rates (Figure 3.12 of Reference 2).

Normally, protection from an uncontrolled CEA g

withdrawal is provided by several different trips (e.g., high

~

pressure, high flux and TM/LP) rather than cnly onc.

i he e

licensee explained that this was because the T!!/LP equaticn which was used was generated througn the use of a single limiting axial power profile for values of pcwer >100% which 4

is calculated to be more limiting than would ac ually be allcwed i

by the Axial Power Distribution (APD) Limiting Safety System Setting (LSSS); hence, a higher than required sensitivity of Fvar with respect to power results.

This increased sensitivity results in a large enange in Pvar with rescect to a small enange in power to the extent of initiating a reactor trip via TM/LP oefere the overpower er hign pressure trip setpcints are reacned.

ihe maximum reactivity insertion rate assumed by the licensee in the safety analyses report (Reference 2) was 1.0 x 10-4 ac/sec.

The maximum reactivity insgrtion rate calcula:ec by the licensee for Cycle 6 is 1.725 x 10-'

/sec.

An additicnal CEA witncrawal analysis was perferred with a withcrawal rate of 1.725 x 10-"

/sec.

1.0 x 10 jhe results are not significantly different from the ac/see witnerawal r::e and still above the safety limit.

In analycing the Red Drep eve.nt, which is the limiting A00 for Ft. Calhcun Cycle 6, the licensee's calculations show :nat at 70 seconds the reacter pcuer tends :c return to its initial value, the cere inlet temperature cecreases acproximately 7'F and the system cressure decreases accu: 105 ;sia. The MDHER value cccurs at this tire.

The licensee states (Reference 36) that sensitivity studies cerferred with XC0 ERA-IIIC indicate that a conservative evaluation cf :ne MONBR for the Rec Dro; event can be cc:ainec by assu=ing initial core conditions plus the ceaking augmentation anticipated fcr :nis event.

In ctner wercs, nc creci: is taken for the cecrease in core inlet temperature and no penalty is taken fer SO 9

8

--n

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---

n

i

+

i

=

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the pressure decrease. The licensee stated that "Since the credit E

associated with the decrease in inlet temperature exceeds the E

penalty associated with the pressure decrease, the analysis of the E

MDNBR for the Rod Drop event is found to be more conservatively 5

calculated using the initial conditions (7* F higher core inlet 5

temperature, and 108 psia higher pressure) than the asyrptotic conditions which exist in the transient." We evaluated this assertion with sensitivity factors based on the W-3 correlation and agree with the licensee that his assumpticn for this

=

combination of pressure drcp and inlet temperature drop is

=

14 conservative.

N The steam line b 6 analysis for Ft. Calhoun was done using several conservat.ve assumptiv.,s.

In the analysis, the location of the steam line break was assumed to be at the outlet nczzle of the steam dome. The fastest cooldown of the primary system is thus achieved.

The discharge coefficient was assumed to be one so that maximum pessible discharce rate could be

~

realized.

Break flew was calculated each time step based g

on a chcke ficw model precortional to the steam generator pressure.

This gives the maximum flow rate whicn in turn gives i

the maximum cooldown.

The steam was assumed saturated; ccmpu-

ations duri1g the transient indicated the quality was essentially unity. With the cuality equal-to unity the steam leaves the steam generator with :ne hignes energy content.

Ereak flow was computed based en the ideal gas flew =cdel

=

and results in a greater flew than calculated using Mccdy's results.

Inus, the abcve model is judged :c result in a more rapid cooldown anc nence an increased likelihood of l

return to power.

Tnese assumptiens, althcugn censervative, are consistent with the usual acsumptions made in analyzing

he steam line break.

We have reviewed the results of the safety analyses cf costula:ec non-LOC

• ACOs anc acticants fcr Cycle 5 a:

Ft. Calheun as presented in P.efere~ces 2 and 19.

In acdition, :ne licensee has providet a ccmputer listing of :ne input to tne FTSFWR comcute code usec in cerfcrming the Uncontrc11ed ',CA

ndrawal event.

We have reviewed both the rei'ef tces and the computer listing anc conclude that the itcensee has adecuately calculated :ne consequences of :nese events.

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5 The licensee analy:ed the Cesign Basis Large Break LCCA using jg NRC approved methces.

The results are repcrted in References M

3 and 35. The results cf the licensee's calculaticns are 5

su=marized in Table 2 of this Safety Evaluation. The most limiting peak cladding temperatures were calculated to occur M

~

at the end of life of both the CE and ENC fuel.

This is due 2

E to the increased fissicn gas release, cc:bined with the use E

hy the licensee of an NRC mecel for flow bicckace due to p

swelling and rupture cf the fuel rces during a LOCA.

g In the past, condi:icns cicse te beginning of life have been

.g worse because cf fuel densificaticn effects.

s

~

The results cf the calculations given in Table 2 show tha:

the ENC fuel reets the emergency c:re eccling system {ECCS)

=

i criteria (Reference 17) with :ne desigr. peaking f ac:cr f5 Of 2.53.

Fcr the CE fuel, the criteria are met with 4

F6 ecual to 2.53 up to a peak ;ellet burnup cf 32,600 lb Mwd /MTU.

In crder fcr the CE fuel to meet the ECCS F

criteria at greater burnurs, an F(, rec;ced by 3% was used.

This redue fcn in F6 is less than would c cur a: a burnu; cf 32,500 Med/MT" anc is :herefere c:nservative.

1 By letter dated May.21,1950 (Reference 40) and su pienen:

cf June 26,1950 (Reference 41), Omana Public Fower Distric:

provided a small break ECCS perfcrmance evaluaticn ir. su;;cr:

cf c:eration at an increased ;cwer level of 1500 M :.

This ECCS

erfcrmance evaluatien has

een cone by C :00sti:n Engineering's

~=

small break =cde' an: re:hedciegy as cascri:ec in Reference 40.

This =ccel and meth ::Ic;y has been previcusly found acceptable and in c cliance with 10 CFR 50 A;;en:ix X as cc:ucented in

eference 41.

This evaluaticn assumed the mest limi:ing single failure, i.e., the failure of cne emergency diesel generator with no Offsite ;cwer, which minimizes the safety inje: icn ficw to the reketer vessel and maxici:es delay tire defere safety I

injecticn.

Additionally, the evaluati n assumec that cnly three Of the f gr safety injet:icn tar.ks fur::icnec.

The reac:cr initial :cwer level was assurec to be IO2 cf the ;r:pesed 1500 En: stretch roter level. The ::ber key syste parare:ers were als: assured to be a: c nservative values fer this analysis.

Thus, the ECCS perfcreance evaluaticn analy:ed a s:e: ru= of c lc leg breaks in the rea:::r 00 lan: ;um: cischarge piping, e.

the : s: limiting small tr6 k Icca:icn.

The treak spectru:

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TABLE 2 TORT CAL l100fl Exposure lleatup Analyses Ress its for ENC and CE fuel ENC FUEL Cf FUCL l

Exposure,PPBU(MWD /MIM DOL 48,000 (E0L)

BOL 32,600 42,400 (EOL) j Total Peaking, f 2.53 2.53 2.53 2.53 2.46 g

Peak Clad Temperature (PCT), Of 1980 2195 2012 2188 2190 Hax. Local Zr/Il 0 - Reaction, percent 4.6 9.1 6.2 9.5 10.1 2

I ilot Rod Durst Time, sec 31.6 29.3 28.5 26.6 26.1 7

i, llot Rod Durst location, ft 7.47 7.47 7.47 7.47 7.47 Time of PCT, sec 208 252 229 235 254 PCT Location, ft 8.22 8.22 8.22 8.22 8.22 Max, Zr/Il 0 Reaction Location, f t 8.22 8.22 8.22 8.22.

8.22 2

Linear lleat Generation Rate, kw/fL at 80CREC 0.8218 0.8682 0.8206 0.8596 0.8338 k

Total 11 Generation, % f total Zr reacted <1%

< 1%

< 1%

<1%

< 1%

2 t

1 e

^k l

• - g 5

analysis covered the range of small break sizes and showed that M

the most limiting was the 0.075 ft break.

This spectrum analysis i

varied from the previous analyses in that the 0.5 ft break was E

no longer limiting. This is due to an improved low pressure SI Ei pump flow which reduced PCT for the larger, small break ansalyus.

E The improved low pressure 'SI pump flow has been verified exper -

imentally and for the purpose of this analysis has been conserv-atively reduced to account for measurement uncertainties (Refcrence 2j 41).

[

In our review of CE small break LOCA analysis as related to the 1:.

TMI-2 event (NUREG-0653), we have required several mcdifications 3

to reactor operation, operator training, etc.

The licensee has i

verified that the recommendations of NUREG-0635 are being implemented (Reference 41). The staff, therefore, finds the

-E small break LOCA analysis for the increased power level to

E be acceptable.

5 RADIOLOGICAL CONSEOULNCES OF POSTULATED ACCIDENTS

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1.

F el Handlinc Accident, Control Rod Ejection Accident, Less-J of-Coolant Acc1 cent (LOCA) d 4

We have reviewed the evaluation of the potential radiological consequences of these accidents in the Safety Evaluation Report (SER) dated August 9, 1973. Since the calculations for the potential radiological consequences included a power level of 1500 Mut, the conclusions reached in the SER are still valid, i.e., the radiological consequences are within the guidelines of 10 CFR 100 and are acceptable.

2.

Main Steam Line Failure. Steam Generator Tube Failure and Waste 6as Decay Tank Failure The August 1973 SER states "...we have concluded that the con-sequences of these accidents can be controllad by limiting the primary and secondary coolant system racioactivity concen-s trations and raximum gas decay tank activity so tnat pctential offsite-doses are small. We will incluca limits in the Technical Specifications en primary and secondary cc slant radioactivity ccncentrations such that the potential two a

nour deses at the exclusion radius for these accidents

+

will be well within the guidelines of Part 100."

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i This conclusion was based en calculations perfereed during

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the operating license review fer the streten power level cf E

1500 MW: and, therefore, is n:: altered by the preposed I

amendment.

j 3.

cel Handlinc *cciden Inside Containren:

l The SER regarding the radiological c:nsecuences of tne Fuel

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Handling A::icent Inside Centainrent was issuec in a letter

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to the licensee dated A;ril 27, 1979. Since the assured 2

power level in the calculaticns re; rted in the SER was i

1500 MWt, the c:nclusien cf acce; ability witnin the cuidelines

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cf 10 CFR 100 is un:nanged.

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The subject of ::ntrol rect na:itability fcilewing an a::iden:

I was n:: addressed in the Augus 1973 SER. However, as par:

E cf the Three Mile Islanc Action 'lan, the licensee establisne:

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treatren systers and the selte waste canageren: syster.

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9 The station vents through a 2" diameter line to control 5

containment pressure about half of the time; automatic is closure signals are provided to the valves in this 2

line for Safety Injection Actuation Signal, Containment Actuation Signal, Containment Spray Actuation Signal,

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Containment High Radiation Level and Stack High Radiation

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. Level. According to Station personnel, Safety Injection

j Actuation will take place in 160 seconds for break sizes 3

as small as 0.007 square foot; this is the si:e of the Power Operated Relief Yalve.

For 1/4" diameter instrument line

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failure (0.0005 square foot), the leakage rate would be near the statien Technical Specification limit of 10 gallons / minute. The setpoints of the radiation monitors

,!u are chosen based on 10 CFR 20'-limits, considering such factors as meteorology, background in the area, and s

efficiency of the particular instrumentation.

We find the redundancy available to provide automatic closure signals for the vent line (as well as for cther actions)

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acceptable for control of possible doses froc small

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LOCAs as well as from other potential accidents.

TECHilICAL SPECIFICATIONS j

1.

Aporoved Chances for Cvele 6 As discussed previously, the licensee ;roposed to operate Cycle 6 at a power of 1500 Wt rather than the currently licensed 1420 MWt.

The TS changes to allcw 1500 MW:

operation were submitted to the HRC staff for review and approval.

The licensee chose to use power distri-bution limits generated for 1500 MW:, although they are more restrictive, fer Cycle 6 operation.

These limits

.x are the APD (Figure 1-2 of the TS), TM/LP Safety Limits (Figure 1-1), TM/LP LSS (Figure 1-3), Allowable Peak Linear Heat Rate vs. Burnup (Figure 2-5), LCO for Excore Monitoring of Linear Heat Rates (Figure 2-6), LCO for DNB Monitoring (Figure 2-7) and Flux Peakjng Augmentation Factor (Figure 2-8).

ThevaluesofF{,Fj and axial tilt y

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are unchanged from the previous cycle. These changes were i

found to be acceptable fer Cycle 6 operation as discussed i

in the SER issued with Amendment 47 on April 1,1980.

i We have reviewed these changes and have determined that j

they are acceptable for 1500 MWt operation.

+:

Pi 2.

Changes Necessary for 1500 MWt Ooeration i

The only change to the TS required for 1500 MWt operation is definition of Rated Power. The stated value has been changed Fi from 1420 MWt to 1500 MWt. The License Condition for Maxinum Power Level (3.A.) has likewise been changed from 1420 to j

1500 megawatts thermal.

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SAFETY CONCLUSION We have concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and

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safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this a indment will not be inimical to the coar.on defense and security or to the health and safety of the public.

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9 h III.

ENVIRONMENTAL IMPACT APPRAISAL DESCRIPTION OF PROPOSED ACTION i

E By letter dated July 17, 1979, Omaha Public Power District (the licensee) requested an amendment to Operating License DPR-40 F

for Fort Calhoun Station Unit 1 to allow operation at 1500 MW thermal power level. The licensee's letter included an Environmental Assessment as Attachment A.

ENVIRONMENTAL IMPACT OF PROPOSED ACTION The NRC has evaluated the potential environmental impact associated with this proposed license amendment as required by the National Environmental Policy Act (NEPA) and Part 51 of 10 CFR.

We have reviewed the Final Environmental Statement (FES) of August 1972, related to the operation of Fort Calhcun Station Unit 1.

Although the operation of the station was initially i

to be at only 1420 MWt, a " stretch" power of 150u MWt was

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anticipated at that time. Therefore, all radiological 2

assessments in the FES were considered on the basis of the higher power level and the present request for operation at 1500 MWt represents no increase in radiological impact.

The radiological consequences of accidents in the licensee's Environmental Assessment are lower than those discussed in 2

the FES based on an increase in the exclusion area size authorized by Amendment No. 36 to the station operating license.

However, the conclusions reached in the FES remain valid.

Implementation of the proposed amendment will therefore not increase normal radiological effluents from the plant compared to tne estimates in the FES.

Implementation will also not allow the licensee to discharge concentrations greater than the maximum allowed nor to discharge more radioactivity in a year than the maximum allowed.

Compliance with the Technical Specifications as modified by the amendment will adequately control releases such that there

.~

will be no appreciable effect on the environment due to operation under these proposed changes and the conclusions reached in the FES remain valid.

CONCLUSION AND BASIS FOR NEGATIVE DECLARATION On the basis of the NRC evaluation and information supplied by the licensee, it is concluded that the implementation of the proposed amendment to Operating License DPR 40 will have no environmental impact other than that which has alreac!y been predicted and described in the Comission's Final Environmental Statement for the Facility dated August 1972.

Having reached these conclusions, the Commission has determined that an environmental impact statement need not be prepared for the proposed license amendment and that a Negative Declaration to that effect should be issued.

Dated: August 15, 1980

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REFERENCES t

9 1.

XN-NF-79-70, Generic Mechanical Design Report for Exxon Nuclear Fort Calhoun 14x14 Reload Fuel Assembly, September 1979.

6 i

2.

XN-NF-79-79, " Plant Transient Analysis for the Fort Calhoun E

~

Reactor at 1500 MW:," September 1979.

3.

XN-NF-79-89, Fort Calhoun LOCA Analysis at 1500 MWt Using ENC WREM-IIA PWR Evaluation Model", September 1979.

4.

XN-NF-78-44, " ENC Generic Rod Ejection Accident Analysis",

January 1979.

5.

XN-75-27(A), " Exxon Nuclear Neutronic Design Methods for Pres-r surized Water Reactors", April 1977.

6.

XN-75-27( A), Supplement 1 to Reference 5, April 1977.

XN-75-27, Supplerent 2 to Reference 5, December 1977.

7.

" Cycle 5 Core Reload Applicatien, "0 PPD, August 7,1978.

8.

XN-CC-28, Rev 4, "XTG: A Two Group Three-Dimensional Reactor 9.

Simulator Utilizing Coarse Mesh Spacing and Users Manual (PWR Version)", July 1976.

10. XN-NF-507, " ENC Setpoint Methodology for CE Reactors", June 1979.

11. " Definition and Justification of Exxon Nuclear Comcany DNS Correlation for PWRS," XN-75-45, October,1975.

12. " Plant Transient Analysis of tne Palisades Reactor for Operation at 2530 MW*h, XN-NF-77-lS, July 18, 1977,

13. "Cescription of the Exxon Nuclear Plant Transient Simulation Model for Pressurir.ed Water Reactors (PTSPWR), XN-7a-5, Revisien 1,May 1975.

1". " Technical Report en Densification of Exxon Nuclear PWR Fuels,"

USNRC Report dated February 27, 1975.

15. " Computation Procedures for Esalvating Fuel Red Bowing,"

XN-75-32(NP), Supplement 2, July 1979.

I

Fort Calhoun Unit 1 Cycle 6 Reload and Stretch Power 16.

Application, Docket No. 50-285, dated July 17, 1979.

3

17. Acceptance Criteria for Emergency Core Cooling Systems' for Light Water Cooled Nuclear Power Reactors, Federal Register, Vol 39, No. 3 - Friday, January 4,1974 18.

XN-NF-79-89, Revision 1, " Fort Calhoun LOCA Analysis at 1500 Mwt Using ENC WREM-IIA PWR ECCS Evaluation Model,"

September 1979.

19.

XN-NF-79-77, "Fert Calhoun Nuclear Plant Cycle 6 Safety Analysis Report," October 1979.

20.

CENPD-145-P, " INCA Method of Analyzing Incore Detector Data in Pressurized Water Reactors," April 1,1975.

21.

XN-75-21, "XCOBRA-IIIC: A Computer Code to Determine the Distribution of Coolant During Steady-State and Transient Core Operation," April 1975.

22.

XN-207 " Power Spike Model for Pressurized Water Reactor Fuel,"

March 1979, 23.

F. B. Skogen, "XPOSE - The Exxon Nuclear Revised LEOPARD,"

XN-CC-21, Revision 3, Exxon Nuclear Company, April 1975 24 W. R. Caldwell, "PDQ7 Reference Manual," WAPD-TM-678, Westinghouse Electric Corporation, January 1955 25.

R. J. Breen, O. C. Marlowe and C. J. Pfeifer, " Harmony: System for Nuclear Reactor Depletion Computation," WAPD-TM-478, January 1955, 25.

R. B. Stout, "XTG - A Two-Group Three-Dimensional Reactor Simulator Utili:ing Coarse Mesh Spacing," XN-CC-28, Revision 3, Exxon Nuclear Company, April 1975.

27 Cycle 5 Core Reload Application, Omaha Public Power District, October 30, 1979.

23.

Letter from W. C. Jones, OPPD, to R. Reid, USNRC, January 2C,1920.

J 29.

Krysinski, T., et al., " Palisades Thermal Hydraulic Design Report 4

for Cycle 2 Core (Batches A, B, C1, D and E Fuels), Exxon Nuclear g

Company, Inc., XN-76-3, February 13, 1976.

~

30. CEPAN Topical Report, CENPD-187, May 29,1975.

k 31.

" Interim Safety Evaluation Report on Effects of Fuel Rod Bowing

~

on Thermal Margin Calculations for Light Water Reactors,"

NRC Report.

32.

Supplement 3-P (Proprietary) to CENPD-225P, " Fuel and Poison Rod Bowing," June 1979.

I 33.

Letter from U. B. VassalTo, NRC, to A. E. Scherer, CE, June 12, 1978.

3'.

Letter from Walter J. Apley, Battelle Pacific Northwest Laboratories, to Richard Lobel, USNRC, "TM/LP Set Points for the Ft. Calhoun Reactor," February 26, 1980 35.

XN-NF-79-89 " Fort Calhoun LOCA Analysis for the Ft. Calhoun Reactor at 1500 Mwt," Supp.1, January 1930.

36.

Letter from W. C. Jones, OPPD, to R. Reid USNRC, December 4,1979.

37.

Letter from W. C. Jones, OPPD, to R. Reid, USNRC, February 12, 1980.

35. Letter from W. C. Jones, OPPD, to R. Reid, USNRC, February 25,19S0.

3.:.

Letter from W. C. Jones, OPPD, to R. Reid, USNRC, March 12, 1980.

4;.

Letter from W. C. Jones, OPPD, to R. Clark. USNRC. May 21, 1930.

L '..

Letter from W. C. Jones, OPPD, to R. Clark, USNRC, Jone 26, IE80.

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