ML19210B068
| ML19210B068 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/27/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19210B066 | List: |
| References | |
| NUDOCS 7911020490 | |
| Download: ML19210B068 (9) | |
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NUCLE AR REGULATORY CO*.W WON 8 +,, f 'g V. ASHINGTON. O C. 700$$
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.....f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACT 00 REGULATION SUPPORTING AME'iDuENT NO. 39 TO FACILITY OPERATING LICENSE NO. DPR-50 METROPOLITA'; EDISO'i COMP ANY JERSEY CENTPAL PC).'ER klD LTGR' C0f tPANY
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PENNSYL'JNIA ELECTRIC C0"PANY THREE__ MILE ISLAND NUCLEAR STATION, UNIT N0. I DOCKET NC. 5_0_-289 O))
g Introduction By letter dated January 9,1978, l'etropolitan Edison Company (Met Ed) requested amendment of ?poendix A to facility Cperating License No.
DPR-50 for Three Mile Island Nuclear Station, Unit No.1 (Tt:I-1).
The requested chance, as revised by the liet Ed letter of April 3,1978, would arend the it:I-l Technical Specifications to reflect plant oper-ating limits applicable during the first 125 + 5 effective full po.ter days (EFPD) of operation with the fuel loadinii to be used durino Cper-ating Cycle 4.
By letter dated April 7,1978, we requested additional inforr3 tion concerning the proposed arandment. This inforration was furnished by l'et Ed in a letter dated April 10,197E.
i'et Ed has stated that they will make apolication at a later date for anendrent of the Till-l Technical Specifications as necessary to establish operatinc limits in Cycle 4 for the period fron 125 + 5 EFP. to the end of the cycle (approxir.ately ES3 EFPD).
Supolemaniary inforration concernine the reactor high pr essure trip and pressurizer code safety valve relief settings for Cycle 4 was provided by "et Ed letters of April 17 and 20, 1978.
===.
Background===
The Met Ed striaal of January 9,1978, was presented to support operation for a full eperating cycle (Cycle 4) followir.g the refuelir.g perforrad at the end of Cycle 3.
As such, the ar.alysis presented in the sutrittcl was based on the exoected exposure of Cycle 3 (?70 + 10 EFPD), and the irtanded cxpcsure of Cycle 4 (26E + 15 EF;D).
Susecuent to rakin; this s?ittal, ar d in the ab scr.ce of 070aress to. P eds early settlerent of the rational coal strike, !'et Ed rec,ested bv letter dated February 17, 1975, an:ndr.ent of the T"I-l Technical Specifications as necessary to eurmit e> tension o' Cycle 3 opratic" to 315 EFPD.
A;preval of ".is re r st us granted by our letter of "2rch 7,197S,
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1564 240 1911020
Shortly after receiving authorization for the extension of TMI-1 Cycle 3, Met Ed determined that it v.as not in their interest to utilize the full term of the extension.
Accordingly, they terminated Cycle 3 on March 17, 1978, after 2S7.1 EFPD of operation and comenced refueling operations for Cycle 4.
Because the fuel burnup in Cycle 3 was greater than assured in the original Cycle 4 analysis transmitted by F.et Ed's letter of January 9, 1978, Met Ed, by letter dated April 3,1978, submitted an amendnent to their January 9,1978 request which took into account the effect of the authorized extension. This amendment also proposed a revised fuel load-ing arrangenent, whico on the basis of experience with a similar Sabcock
& Wilccx-designed facility, is expected to provide a more uniform neutron flux distributinn.
Met Ed stated that because of the short time interval between the decision to terminate operation in the extended Cycle 3 and the projected conpletion of refueling for Cycle 4, there was insufficient time to perform the revised analyses necessary to support operation over the full tern of Cycle 4 Accordingly, Met Ed in their letter of April 3, 1978, only proposed technical specifications applicable to the first 125 + 5 EFPD of operation in Cycle 4.
They state that a subsequent sub-mittal covering the balance of Cycle 4 (to 265 + 15 EFPD) will be made in May,1978.
Evaluation By references 1, 7, and 8, Pet Ed requested changes to the Technical Specifications appended to the TMI-l Operating License for Cycle 4 operation.
The TMI-l reactor core consists of 177 fuel assenblies.
All of the Batch 3 assemblies v;ill be discharged at the end of Cycle 3.
burned Satch I assenblies, with an initial enrichrsnt of Thirteen eg"L'pand eight Catch 2 assenblies, with an initial enrichrent 2.06 wt ;
of 2.75 wt f ' "U, wili be reloaded into the interior portion ofMe core.
Batches 4 and 5 with initial enrichments of 2.64, and 2.85 wt t "~U, res-pectively, will be shuffled to new locations within their present quadrant.
Batch 6, whic consists of 52 fresh asse.cblies with an initial enrichment of 2.85 wt '; p3 U, will occupy the core periphery.
Fuel Asserbly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters and dimensions for TMI-l Cycle 4 are listed in Table 4-1 of the attachrent to reference 1.
The Park Ea fresh fuel assecblies (Batch 6) are identi-cal in concept and are mechanically interchangeable with those added in TMI-l Cycle 3.
1564 241
. D " * ~Dl W lii M h o wn NdhhiM The Batch 6,15 x 15 (Mark B-4), the B3tch Ic,15 x 15 (Mark B-2), and the Batch 2b,15 x 15 (Mark B-3), fuel assembly designs have been pre-viously reviewed and accepted by the NRC staff for use in TMI.-l.
- Also, these types of assemblies have been operated in TMI-1.
The reload assemblies, therefore, do not represent any unreviewed change,in mech-anical design fron the reference cycle.
Met Ed has taken each fuel assembly design into account in the various mechanical analyses. The Batch 2b fuel is generally limiting because of its relatively low initial fuel pellet density, lower prepressuriza-tion, and previous incore exposure.
The results of these analyses have shown that the mechanical design differences between fuels for Cycle 3 and Cycle 4 are negligible and are acceptable.
Creep collapse analyses were performed by Met Ed for three-cycle assembly power histories.
The Batch 2b fuel is more limiting for cladding collapse i
due to its previous incore exposure time.
The creep collapse analyses were performed based on the conditions (g forth in reference 9 which have been previously found acceptable.
The collapse time for the rest limiting assembly was conservatively determined to be more than 30,000 EFPH (effective full power hours), which is longer than the maximum design exposure for the total of three cycles.
Met Ed stated that the TMI-l stress parameters were enveloped by a con-servative fuel rodstress analysis.
The following conservatists with respect to TMI-l fuel were used in the analysis:
lower post-densification internal pressure, lower initial pellet density, higher system pressure, and higher thermal gradient across the cladding.
The licensee hu referenced the report BA'.!-lM9 which presents calcula-tions of cladding stress at various power levels and fuel burnups for TMI-l fuel.
These calculctions show that in no case does the stress exceed the yield stress.
Inis is acceptable to the staff.
The fuel design criteria specify a limit to the cladding plastic circur-ferential strain of 1.0..
lhe pellet design is cstablished for plastic cladJing strain of less than 17 at values of maximum design local pellet burnup and heat generation rate, which are considerably higher than the values for TMI-l fuel.
This will result in an even greater margin than the analysis demonstrated.
The strain analysis is also based cn the maximum manufacturing specifications value fc$r the fuel pellet diameter and density and the icwest permitted manufacturing specifications toler-ance for the cladding ID.
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The linear heat generation rate (LHGR) capabilities are based on certe{-
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line fuel melt and were established by Met Ed using the TAFY-3 code with fuel densification to 96.5% of theoretical density.
All the fuel assemblies in the Cycle 4 core are thermally simil'ar. The fresh Batch 6 fuel inserted for Cycle 4 operation introduces no signifi-cant differences in fuel thermal perfornance relative to the other fuel remaining in the core, and its LHGR limit has been established as 20.15 KW/ft.
Met Ed's thermal analysis of the fuel rods assumed in-reactor densifica-tion to 96.5% theoretical densj+y The analytical methods utilized are the sane as those for Cycle 3.tM These analyses were based on the lower tolerance limit of the fuel density specification and assumed isotropic diametral shrinkage and anisotropic axial shrinkage resulting from fuel densification.
The Batch 6 fuel assc-blies are not new in concept, nor do they utilize different component mterials.
Therefore, the chemical compatibility of all possible fuel-cladding-coolant assembly interactions for the Satch 6 fuel assemblies are identical to those of the present fuel.
This fuel as prooosed for reload in TMI-l has had considerable operating experience.
The Ratch 4, 5, and 6 fuel assemblies are not new in ccn-cept and do not use different co ponent materials.
The fuel assemblies for Cycle 4 operation will not exceed any design life limits. We con-clude, therefore, that the fuel rechanical design for Cycle 4 operation is acceptable.
fluclear Analysis Table 5-1 of the attachnent of reference 1 corpares the core physics parameters of Cycles 3 and 4 The values for both cycles were cenerated by Met Ed usinc PD007.
Since the core has not jet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles.
The extended Cycle 3 produced a larger cycle differential burnup than is expected for Cycle 4.
The accumulated average core burnup will be higher in Cycle 4 than in Cycle 3 because of the presence of the once-burned Batch 1c and 2b fuel and the extension of Cycle 3.
The critical boron concentrations fcr Cycle 4 are aporoxirately the same as for Cycle 3.
The csr. trol rod worths are sufficient to -aintain the required shutdo..n r.argin.
The maxinum stud. rod worths foi Cycle 4 are less than those in Cycle 3.
The adecuacy of the shutdym rargin with Cycle a red worths his been de:enstrated analytically by Met Ed.
I':t Ed's shutdosn calculatiM : conserwitively used a poisen raterial depletion allc.
ce and 10 ainty on net r,:4. wru r
1564 243
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.za The same calculational rethods and design infornation were used by Met Ed to obtain the nucler.r design parareters for Cycles 3 and 4.
The node of reactor operaticn has been changed from a rodded to, an unrodded feed-bleed mode.
I;o changes to the rakeup and purification system were necessary for this node o# operation.
For operation in the unrodded mc6 t% -equired feed-bleed capabilities are the same as for operation in the rodded nade with the addition of adjusting the Rf 3ctor Coolant System (RCS) boron concentration to maintain the regulatii,
'ds wiLhin specified control bands.
The plant raneuverability is limited by the ability of the waste processing system to handle the waste generated.
Met Ed had intended (I) to cross-core shuffle the fuel
- for the Cycle 4 reload.
However, due to quadrant flux tilt problers encountered at another Babcock & Wilc.z-) designed facility using cross-core shuffle, this plan was changed.U All fuel shuf fling for Cycle 4 will now be limited by the Met Ed to the quadrant in which the fuel resided in Cycle 3.
This methad of fuel shuffling tends to reduce possible carry-over effects of any burnup asynnetry that right be present in the pre-vious cycle. The lowest indicated tilt (1.25), which occurred at the end of Cycle 3, should be further reduced by this method of fuel shuffle.
Met Ed requested a chance in the technical specification limit on quadrant tilt to increase the allowable maxinur tilt from 3.41F to 4.92' The submittals (8,16) on quaarant tilt indicate that the change will restore the allcwable tilt level to that permitted in Cycles I and 2 for TM1-1. Tne lower tilt limit used for Cycle 3 operation was to offset a required peaking penalty due to fuel rod bow.
For Cycle 4, TMI-1 has used a statistical corbination of peakina factors, renoved the densifica-tion pcwer spike fron ECCS-% pendent technical specification limits, and reduced the core peaking factor.
By use of data in the Babcock & Wilcox (C&',l) report BA,;-10073 and data obtained fren Oconee Unit 1, Cycle 4, Met Ed has demonstrated that an increase in allowable tilt to 4.92:. for TMI-1, Cycle 4 is acceptable.
The infora,ation presented has been reviewed and found acceptable.
In vien of the above and the fact that startup tests (to be condacted prior to power operation) will verify that the significant aspects of the core perfomance are within the assumtions of the safety analsyis, we find Met Ed's nuclear analysis for Cycle 4 to be acceptable.
Theral-Fydraulic Analysis The Injor acceptan:e crit-ia dich a r ' u=ed for the thernal-hydraulic desip are specified in Standard Rnir./ Plan (57) 4.0 These criteria establish acce; wie li lt' or. departuru f ror nuci t.a t o boili ng (D"E ).
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The therral-hydraulic design in support of Cycle 4 operativn utilized the methocs and models described in referenccs 12 and 13.
f!et Ed stated that Cycle 4 analyses and resulting setpoints have been based on 106.5' of the design reactor coolant (RC) system flow rate.
The core configuration for Cycle 4 differs slightly from that of Cycle 3 in the proportion of Mark E2, Park B3 and Park B4 fuel assemblies con-tained in the core.
Specifically, 52 P. art B assemblies will replace an equal nur.ber of Mark E3 assemblies used in Cycle 3.
Park C4 assemblies differ from the Park S2 anc E3 primarily in the design of the end fittinc, which results in a slight reduction in flow resistance for the E4 desiga.
No credit was taken by."et Ed in the analyses for the increased flow to the Mark B4 assemblies, located in the hottest core locations, as a result of the presence of the Park B2 and B3 assemblies.
Met Ed used the CAW-2 CFF correlation
) for thermal-hydraulic analysis of Cycle 4.
This correlation has been reviewed and approved for use with the Park B fuel asser.bly design.(15)
The effect of fuel densification on the minim r' departure from nucleate boiling ratio (CNER) is prirarily a result of the reduction in active fuel length, which increases the average heat flux.
Met Ed's Cycle 4 DNSR analysis was based on a cold densified active length of 140.2 inches, a value selected to apply generically to a number of B&W plants.
This is a conservative rethod of applying the densification effect since all the fuel assemblies in Cycle 4 have longer densified lengths ano because no credit is taken for axial thermal expansion of the fuel column.
The pote 1 effect of fuel rod bow on D"5P was previously evaluated for Cycle 3.
The effect cf fuel rod bow on CNEP would be unchanced durinc
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Cycle 4 operaticn.
Therefore, the previous evaivation of this potentia, ef fect remains unchanged.
Accident and Transient Analvsis The accident and transient analyses as provided by Pet Ed denonstrate that the TMI-l FSAR analyses conservatively bound the predicted conci-tions of the T'C-1, Cycle a core and are, therefore acc'estable.
Fet Ed has stated that each FSAR accident analysis has beer. exa.ined, with respect to chances in Cycle 4 paraneters, to determine the effects of the reload and to ensure that perfor ance is not degraded during hypothetical transients.
The core thc-rral parareters used in tne FSAR accident analysis..ere desi n operatinc values based on calculated values plus uncertainties.
FSAP values of core thermal parareters were cercared with those calculatea in " e Cycle a mn'ysis.
The e'fects of fuel den-sification on the FSTP accident resul' c 'un been evaluated and are 1564 245 j@
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reported in the T!!I-l fuel dens i'ication rei art.(12) Since Cycle 4 reload fuel assenblics contain fuel rods with a density higher than those considered there, ti e conclusians dorived in that report are valid for T!4-1 Cycle a.
Calculatieral techniaues and rethods for Cycle 4 analyses remain consistent with those used for the FSAR.
With respect to radiation doses, !'et Ed reported that because of improved fuel utilization and improved calallational methods, they noet estimate Cecause they are achievina a higher plutoniun-to-uraniu, fissien ratio.
plutonium has a hicher iodine fission yield then uranium, rare iodine will be produced. ?'et Ed estimates that the increased iodine produc-tion will increase the 2-hour thyroid doses given in the FE/R by 8 to 15 . We have reviewed the effects of this increase and conclude that the consequences of all accidents remain well within acceptable limits.
Met Ed has clarified (8) the manner in which errors in reactor power measurement are incorporated in their analyses. The setpoints used in the accident analyses assumotions contain the required calorimetric power measurement uncertainty of 2t' plus a 4~' uncertainty to account for errors in neutroa power measurement.
The 25 calorimetric uncer-The 4" tainty accounts for steady-state power measurement error.
neutron power reasurement uncertainty allows for steady-state and transient errors following maneuvecina transients. We conclude that this manner of accounting for these uncertainties is acceDtable and therefore the resultant analyses of postulated accidents and transiants are conservative.
Reactor High Pressure Trip and Pressurizer Code Safety Valve Settings By letter dated April 6,1977, we authorized Met Ed to increase the T"I-l reactor high pressure trip setting from 2355 psig to 2405 psig, and to increase the relief setting of the pressurizer code safety valves fro-2435 psig to 25C0 osig.
Eecause the assumpticns used by !a t Ed in justi-e fying these changes were applicable to Cycle 3, our approval of the increesed settings was limited to that cycle.
By letter dated April 17, 1978 as amended by letter dated T.pril 20, 1978, Met Ed submitted their evaluation in support of the continued accepta-bility of these settings. Their evaluation indicates that under Cycle 4 operating conditions and assuming a more conservative instrument error, the conservatively calculated peak reactor coolant systen pressure result-ing from the feedwater line break (the limiting accident) is increased from 2734 psig (the Cycle 3 value) to 2749.3 psig.
Since this value is less than the safety linit for reactor coolant system prescure of 27El psig as stated in Technical Speci fication 2.2.1, we conclude that reten-tion of the present row or hich pressure trip and pressurizer cede ra'%.
valve settings for Cycle - is acceotable.
Althouch there is a reducticn in maroin, with sore portion due to use of additional ccoscrvatism, t :
vessel stress is still within the csde requir m ts for relief valve
- capacity, For this reasen the reduction in m rain is not siqnificant.
1564 246
, Ird Startup Tests The physics startup test progran for Cycle ' as stated in Section 9 of Additional informe-Met Ed's January 9, 1978 submittal has been reviewed.
tion vias requested and supplied in the April 10, 1978 Met Ed sobr.ittal.
The physics startup test prograr includes ze-o pcwer reas Power distribution rod worth and control rod group reaccivity worth.
measurenents will be made at higher powers.
the NRC staff and found to be acceptable.
This prograr, has been reviewed byBecause there are areas in Met Ed's safety tion by the physics startup test program, we have requested Met Ed to Met Ed has acreed to submit a report of the results of these tests.
submit such a report within 90 days of thE completion of the' tests.
We find this acceptable.
ECCS Analysis B&W has recently discovered a deficiency in the method used to calculate This matter is being Emergency Core Cooling System (ECCS) performance.
reviewed by the NRC staff.
Concl usien We have determined that the am:ndment does not authorize a change in effluent types or tctal arounts nor an increase in poner level and will not result in any significar,t environrental irpact.
Having made this deternination, we have furthe*' concluded that the amendrent involves an action which is insignificant from the standpoint of envirormental i ;'3ct and, pursuant to 10 CFR 151.5(d)(a), that an envircr, rental irpact state-rent, or negative declaration and envircr.r. ental irpact agoraisal need not be prepa ed in connection with the issuance of tnis arendrent.
With thc eiception of the ratter of ECCS perfccmance, we have concluded, based on the considerations discussed above, that: (1) because the in:rease in the probability or amend-er.t does not involve a siani ficar.:
involve a consecucnces of eccidents previcusly cccc. 2 red and does not significant decrease in a sa e:y marcin, the arrfar.t 'dces not involve r
a sianificant hence consideraticr; (2) there is rcasonable assurarce that the baalth ard u fety of U.e outlic.;ill'not t:c e Hancered by csed rar.ner; and (3) such activities will be operation in the p-conductea in comiicrce with the Cor :ission's repulniors and tne isst.ance of this e y 3.ent will not be ini-ical to the comon defense Our conclusicns and secu-ity or to the nealth and safety of the cublic.
in regard to ECCS mrferrance are addressed in the acccccanying E xcr7 tio n.
Cated:
April 27, 1973 1564 247
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References:
(1)
Letter J.r. Herbein (Met Ed) to R.W. Reid (r:RC) dated January 9, 1978.
(2) Letter J.G. Herbein (Met Ed) to R.W. Reid (NHC) dated February 17, 1978.
(3) Letter J.G. Herbein (Met Ed) to R.W. Reid (NRC) dated March 1, 1978.
(4) Letter J.G. Herbein (Met Ed) to R.W. Reid (NRC) dated March 13, 1978.
(5) Letter J.G. Herbein (tiet Ed) to R.W. Reid (l:RC) dated March 13, 1978.
(6) Letter J.G. Herbein (Met Ed) to R.W. Reid (t:RC) dated liarch 14, 1978.
(7) Letter J.G. Herbein (Met Ed) to R.W. Reid (t:RC) dated Arpil 3, 1978.
(8) Letter J.G. Herbein (Met Ed) to R.W. Reid (T RC) dated April 10, 1973.
(9) Prograr to Dcterr.ine In-Reactor Perforrance of F.?.'.! Fuels - Cladding Creep Collapse, EAW-lCCEO, Rev.1, Bal ock & Wilcox, Nove.n.ber 1976.
(10) Letter fror, A. Schwencer (NRC) to J.F. Malla'ry (B&'.!) dated January 29, 1975.
(11)
C.D. Morgan and H.S. Kao, TAFY - Fuel Pin Tenperature and Gas
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Pressure Analysi s, BA'..'-10C20, Babcock & Wilcox, l'.3y 1972.
(12) TMI-l Fuel Densification Report, SAW-1399, Sabccck & Wilcox, June 1973.
(13) Three liile Island fiuclear Station, Unit 1, Final Safety Analysis Report, Docket No. 50-289.
(14) Correlation of Critical Heat Flux in a Eundle Cooled by Fressurized Water, BA'.!-lGC00A, Fab ock & Wilcox, June 1976.
(15) Letter fror J. Stolz (NRC) to K.E. Surke (BM!), de ted April 15, lEC.
(16) Letter W.R. Gib:en (CL?) to t !.ar. dry (*;RC) d:ted april 19,197E.
(17) Lette. R. <.. i.;i d
('..w ) to R. C. Ar nai c
'at Ej) Jated I'. arch 7, 13; 1564 248
w UNITED STATES NUCLE 2: P.EG'JLATORY CC'/"ISSIO*
DOCKET NO. 50-2R9 METF.0DOLITA'; EDIS1N CC"PA"Y R
JERSEY CENTRAL PrMER Af;D LIGHT CO?iPANY PEN';SYLVA.';! A ELECTRIC C0"?A'iY NOTICE OF ISSUANCE 07 A"EC'E'!T TO FACILITY
_ OPERATING LICENSE The U. S. Nuclear Reculatory Commission (the Commission) has issued Amendnent No. 39 to Facility Operatina License No. DPR-50, issued to Metropolitan Edison Company, Jersey Ceritral Power and Light Company and Pennsylvania Electric Company (the licensees), which revised the Technical Spe:ifications for operation of the Three fiile Island I;uclear Station, Unit f;o.1 (the facility) located in Dsuchin Ccunty, Pennsylvania. The amendment is effective as of its date of issuance.
This amendment revises the Technical Specifications to reflect plant operating linitations for the first. i30 effective full power days of cperation in Cycle 4.
The application for the amendrent complies with the standards and requirements of the Atoni: Energy Act of 195a, as The amended (the Act), and the Cornission's rules and regulations.
Commission has made appropriate findings as recaired by the Act and.the Conmission's rules and regulations in 10 CFR Chapter I, which are set forth in the license arendment.
Prior public notice of this anendment was not requir:d since the an 2nc~e-t r'aes not involve a sigt.ificent hazards consideraticn 1564 249
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.. The Connission has deternined that the issuance of this ameridment will not result in any significant er.vircnnental impact and. that pursuant to 10 CFR 551.5(d)(4) an environmental impact statement, or negative declaratior, and environmental impact appraisal need not be prepared in conr.ectior. with issuance of this amendment.
For further details with respect to this action, see (1) the application for amendment dated January 9,1978, as supplemented April 3,10,17, and 20,1973, (2) Amendrent No. 39 to License No.
DPR-50, and (3) the Comission's related Safety Evaluation.
All of these items are available for public inspection at the Ccmmission's Public Document Room,1717 H Street, N.W., Washington, D.C. and at the Government Publications Section, State Library of Pennsylvania, Box 1601 (Education Duilding), Harrisburg, Pennsylvania.
A copy of items (2) and (3) may be obtained upon request addressed to the U. S. Nuclear Regulatory Comission, Washington, D.C.
20555, Attention:
Director, Division of Operating Reactcrs.
Dated at Bethosda, Maryland, this 27th day of April 1978.
FOR THE NUCLEAR REGULATORY C0"."ISS103 f}
(f-f g /. y /~; {} I. C4 A /
e s-Robert W. Reid, Chief Operating P.cactors Branch N Division of Cperating Reactors 156A 250 a