ML20205C885

Safety Evaluation Supporting Amend 126 to License DPR-50

ML20205C885
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Site:	Crane
Issue date:	03/20/1987
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UNITED STATES

.- g i NUCLEAR REGULATORY COMMISSION q k' ~. W E

WASHING TON. D. C. 20555 x..~. f /

s a j SAFETY EVAll'ATION PY THE OFFICE OF Fl' CLEAR REACTOP REGUI ATIGF SilPPOPTif:r-IFENDMENT FO.126 TO FACIIITY OPERATING LICENSE im. DPP-50 METROPOLITAN EDISON COPPANY JERSEY CENTRAL POWEP At!D LIGHT COMPANY PENNS)LVANI A EL ECTRIC COMPANY GPl> hl:Cl EAP COPPORATION TFFEE MILE ISLAND NUCLEAP STATION. UNIT NO. 1 DOCKET NO. 50-289

1.0 INTRODUCTION

By letter dated November 3.1986 (Fef.1), the GPU Nuclear Corporatior (GPil or the licensee) made application to amend the Technical Specifi-cations of the Three Mile Island Nuclear Station, Unit No. 1 (TMI-1).

The proposed chances would modify the Technical Specifications to permit operation for a sixth cycle (Cycle 6). The safety analyses performed ard the resulting modifications to the TMI-1 Technical Specifications are described in the Cycle 6 reload report (Ref. ?).

The safety analysis for the previous fifth cycle of operation at TMI-1 is being used by the licensee as the reference cycle for the proposed sixth cycle of operation. Cycle 5 operated with no anomalies that would effect Cycle 6.

Where conditions are identical or limitina in the fifth cycle safety analysis, our previous Safety Evaluation (Ref. 3) continuer to apply.

By letter dated July 16, 1986, the licensee applied to amend the Technical Specifications for TMI-1. The proposed amendment (Technical Specification Chanae Request No. 159) would allow withdrawine of the axial power shaping rods under end of Cycle 5 conditions and would chance the order of preference for instruments used to measure quadrant power tilt. L'ithdrawal of the axial power shaping rods was approved ir Amendment No. 120 issued September 2, 1986.

Our evaluation of the safety analysis for the TM1-1 Cycle 6 reinad and changina the order of preference for instruments used to monitor quadrant tilt follows.

7.0 DESCRIPTION

OF THE CYCLE 6 CORE The TMI-1 core consists of 177 fuel assemblies, each of which is a 15 y 15 array containina 208 fuel rods,16 control rod guide tubes, and one incore instrument guide tube. The fuel manaaement scheme is basically a low-leakaae B703300285 870320 PDR ADOCK 05000289 P

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i design with loading pattern and enrichments chosen to provice a Cycle f length of 475 e 15 effective full power days (EFPDs',. The loadire patterr consists of 49 Batch 6 fuel assemblies shuffled to new locations with t1 fuel assenblies distributed in a checkerboard pattern in the interior of the core and 8 fuel assemblies placed on and rear the core peripher.*

on the core flats, 5? Patch 7 fuel assemblies stuffled to new locatirr?.

I un trd near the core periptery, and 76 fresh fuel assemblies distributec in o checkerboard pattern in the interior of the core.

Sixty-four of tF(

'rrsh fuel assemblies corprise Patch EA, while !? cororise Patch 8R.

The Fatch 6 fuel assemblies are characterized as heirg twice-buri.cd, and the Batch 7 fuel assemblies are once-burned. The iritial fuel enrichment for Patches 6, 7 and 8B is ?.85 weight percent (w/o) uranium-?35, while the initial fuel enrichment of Batch EA is ?.9E w/o uranium-235.

Reactivity control #or Cycle E will be provided by 61 full-lenoth silver-indium-cadmium control rods, 68 burnable poison rod assemblies (EPRAs) containing varyinc amounts of B,C admixed witt fl,0, 8 Incenel exial pcvtr 3

shaping rods (APSRs), and solubTe boron in the coolant. The APSPs have been changed for Cycle f to Inconel (a " gray" absorber in contrast to tFe previous APSR absorber material) and provide for control of the axial power distribution. All cf the core locations, except the peripheral core locations, will contain either a control rod (in the once-and twice-burned fuel) or a BPRA fin the fresh fuel).

The licensed core power level is 2535 MWt. The safety analysis provided in the reload report (Ref. 2) demonstrates the safe operation of TVI-1 throuchout Cycle 6 at full power. The follcwing sections describe our evaluation of the safety analysis.

3.0 FVALUATION OF THE FUEL SYSTEM DESIGN 3.1 Fuel Assembly Vechanical Desion and Gray APSR Design The 76 Babcock & Wilccx (B&W) Mark B4 15 x 15 fuel assemblies to be loaded as Batches 8A and 8P for Cycle 6 operation are mech-anically interchanaeable with Batches 6 and 7 fuel assemblies previously loaded at TMI-1. The cladding stress, strain, and collapse analyses are bounded by conditions previously analyzed for TMI-1 or were analyzed specificelly for Cycle 6 using methods and limits previously reviewed ard approved by the NRC. The licensee similarly evaluated the gray APSRs that are to be inserted in Cycle 6 and established that the gray APSR design ret applicable limits on claddinn stress, strain, and co' lapse.

3.7 Fuel Rod Design All batches in the TMI-1 Cycle 6 core utili7e the same B&W Mark BC fuel desian, and the Batches BA and 8B fuel parameters are virto-ally identical to the previously loaded Batches 6 and 7 except that the enrichment of Batch FA fuel has been increased from 2.85 to 2.95 w/o uranium-235.

4 Resinter tests were rer'ormed on all fuel pelle t lots comprisino Fetches 8A anc EB fuel assenblies. The tests were based on a ntdified samplino plar and are described in References 4 anc' E.

The resinter tests confirm the corservatism c.f densification characteristics assur.ed ir the TACO? (Ref 6) analvses. The licensee states that tha results of the TFT 1 Fuel Densificatir.r Report (Ref. 7) remain boundino for all Cycle 6 fuel since those analyses were based on a lower iritial pellet density ar# an as-sumption of censification to 96.5i of theoretical density. Ve have reviewed the information presented in References 4 and E en the resinterino tests based on a modified samplire plan anc find it acceptable.

3.3 Fuel Rod internal Pressure Section A.? of the Standard Review Plan (Pef. 8) addresses a number of acceptance criteria used to establish the design bases and evaluatior of the fuel system. Among those which may affect the operation of the fuel rod is the internal pressure limit. Our current criterion is that fuel rod internal cas pressure should remain below normal sys-tem pressure durino permal operation unless otherwise justified.

GPL:

has stated that fuel rod internal pressure will net exceed nominal system pressure during normal operation for Cycle 6.

This analysis is based on the use of the approved B8W TACO 2 code (Ref 6). The staff concludes that the fuel rod internal pressure limit has been acceptably cor.sidered for Cycle 6 operation.

3.4 Fuel Thermal Desion There are no majer changes between the thermal desien of the new Batches 8A and 8B fuel and previous batches that will be reinserted in the Cycle 6 core. The thermal design analyses were performed with the app' roved TAC 02 code (Pef. 6). The Cycle 6 core protectinr.

limits are based on a linear heat ceneration rate (LHGCi to center-line fuel melt of 70.5 kW/ft, which is applicable to all Cycle 6 fuel batches. The results of the thermal design evaluation show no significant differences between the rew Batches 8A and BB fuel and the previous Patches 6 and 7 fuel. We have reviewed the fuel thermal design parameters for normal operation and find them acceptable.

3.5 Loss of Coolart Accident (LOCA) Initial Conditions In addition to the steady-state conditions, the average fuel tempe rature as a function of LPGR and lifetime fuel pin pressure data used in the LOCA analysis (see Section 7.2 of Reference ?) are.also calculated with the TAC 02 code (Ref 6). The reload report (Prf. ?)

,, states that the fuel temperature and fuel pin pressure data u W in the generic LOCA eralysis (Fef. 91 are ccrservative compared tr these calculatec for TMI-1 Cycle 6.

The bounding values of the allowoble LOCA LHGRs (see Table 7.? of Peference 7) include the effects of NUREG-0630 (Ref. 10) regarding fuel cladding swellinc and rupture behavier durina LOCA.

3.0 Conclusion Or Cycle f Fuel Systen Desian We have reviewed the fuel system design ano analysis for Cycle 6 operation and find it acceptable, as discussed abcVe.

4.0 FV/ll'ATION OF TPE Nt!CLEAP DESIGN l

To support Cycle f eperation of TMI-1, the licensee has provided analyses l

using analytical methoos and design beses established in licensing topical reportr. that have been approved by the NRC. The licensee has provided a comparison of the core physics parameters for Cycles 5 and 6 as celculated with these approved trethods. The parameters for Cycle 5 were generated using PDQ07 (Pef. 11), while the parameters for Cycle 6 were generated usino the N000LE code (Ref.12). The two codes give comparable results when compared to measured data. There are differences in the neutronic par-ameters compared between Cycles 5 and 6.

These differences can be at-tributed to differences in cycle lenoths, BPRA loadinc, and fuel loadina pattern. All of the transients and accidents analy7ed in the Final Safety

~

Analysis Report (FSAR) were reviewed for Cycle 6 operation. The Cycle 6 pararreters were conservative when compared to analyses accepted for previous cycles, and no new transient and accident analyses are included in the reload report (Ref. ?).

The control rod worths and shutdown margin requirements at beginnine-of-cycle (BOC) and at the most limiting time at end-of-cycle (EOC) for the Cycle 6 nuclear design are presented in Table 5-?

of Reference ?.

At EOC 6, the reactivity worth with all control rods inserted, assuming that the highest worth control rod is stuck out of the core, is 6.48%. This reactivity worth also assumed a reduction in worth of 10% for uncertainty and a reduction in worth due to control rod absorber burnup. The reactivity worth required for shutdown, includinc the contribution reovired to accommodate the reactivity effects of the steamline break evert at E0C 6, is 4.85%. Therefore.

sufficient control rod worth is available to accommodate the reactivity effects of the steamline break event at the worst time in core life allowing for the most reactive control rod stuck in the full withdrawr position and allowing for calculational uncertainties and control absorber burnup. Ve have reviewed the calculated control rod wortFs and uncertainties in these worths based upon corrparisons of calculations with experiments in other B&W repcrts. On the basis of this review, we corelude that GPU's assessment of reactivity cortrol is suitably conservative and that adequate negative reactivity worth has been

. crovided by the centrol syster to assure shutdown capability assurinc the most reactive centroi rod is stuck in the full withdrawn position.

We conclude that the licensee's predicted neutronic paraneters are acceptob1r because they were obteined usino approved metFods, the vtlidity of which has beer, cemonstrated throuch nany cycles of predictichs, includiro startup tests, for this and other reactors. /e a result of this review of the neutronic parametert compared to Dreviou?

cycles, we concur with the litersee's conclusions reoardina the Cycle f trunsient and accident analyses.

The licensee has made a number o' chances in the nuclear desion of Cycle 6.

These changes are: (1) the increase in cycle lifetine to 425 EFPDs with the incorporation of BFRAs to aid in reactivity control, (?) the use of aray APSRs, (3) the use of the NOODLE code to calculate the physics parameters for Cycle 6, and (4) the removal of the power level hold requirerents for xenon in Technical Specifications 3.5.2.4.d and 3.5.?.5.c.

The effects of the chance in the cycle lifetime. o' the use of BPPAs, and of the use of gray APSRs have all been taken into account in the nuclear desion.

In particular.

the licensee verified that the gray APSPs provide adequate axial power distribution control. The ND0DLE code has been reviewed and approved by the NRC staff (Ref.13). An extensive analysis has been perforced by B&W for the licensee (Ref.14) to justify removal of the pc6tr level cut-off requirements. This power level cut-off had been utilized to accommodate. transient xenon effects on power peakinc factors before ascendirq to 100% power. The analysis showed that the 5% total xenon factor applied in the computatior of LOCA margin provides conservative operating limits. The 2.5% radial xenon factor applied in the evaluation of initial condition departure from nucleate boiling (DNF) margin was al.so shown to be ccnservative.

We conclude that these changes in the Cycle 6 nuclear design are acceptable since.the nuclear desion and resulting Tectrical Specifications for Cycle 6 include the effects of the changes calculated with approved methods.

5.0 EVALUAT10F 0F THE THEFVAL-HYDRAl! llc DESIGN The thermal-hydraulic design of Cycle 6 is nearly identical to that of Cycle 5 as shown in the comparison of raximum design conditions in i

Table 6-1 of Reference 2.

The thermal-hydraulic design evaluation utilizes, for the first time, the LYNX series of codes (Refs. 15, 16 and 17) for crossflow modeling for DNB predictions. These LYNX codt.s have been reviewed and approved by the NRC staff. The application of cross-flow modeling for reload cores is describec in Reference 10. The fresh Ettches 8A and 8B fuel assenblies are hydraulically and geometrically similar to irradiated Batches 6 and 7 fuel assemblies. No departurr fror rucleate boiling ratio (DNBP) penalty is required since the approvec rod bow topical report (Ref. 19) shows that the reduction in power

~

l

-6.

s procuttien capaHlity more than offsets ery rod bow ef fects as Eurrup increrses.

The bypass flow for Cycle 6 decreased 10 7.6% from 1C.4*.

for Cycle 5 because of the insertion of the CE BPDAs. The thermal-hydraulic analysis used, however, a conservative value of 8.R for the bypess flow. Based on the thermal-hydraulic similarities of Cycle 6 with Cycle 5 and the use of approved models ard methods, we conclude thbt the thernal-hydraulic desior c' Cycle 6 is acceptable.

6.0 EVALUATTON OF TPAt!SIENT AND ACCIDENT ANALYSES The licensee has examired each FSAR transient and accident ar.alysis with respect to changes in the Cycle 6 parameters to ersure that the calcul;ted consecuences still meet applicable criteria. The key parameters havire the greatest effect on the outcome of a trarsient or accident are the core thernal parameters, the thermal-hydraulic parameters, and the physics static and kinetic parareters.

Fuel thermal analysis values are listed in Table 4-1 of Reference 2 for all fuel batches in Cycle 6.

Table 6-1 of Reference ? compares the thermal-hydraulic parameters for Cycles 5 and E.

These parameters are either the same for both cycles or exhibit di#ferences due to, for example, modelino changes. The physics parameters are pro-vided in Table 5-1 of Reference ?.

A comparison of key kinetics parereters from the FSAR and the densi#icetion repurt with predicted Cycle 6 values is provided in Table.7-1 of Reference P.

These data indicate that the FSAR data are bounding for most of the parameters.

For those paraneters not bounded by the FSAR data, the licensee states that their ef#ect on affected transients or accidents will produce less severe consequences than in previous bounding analyses. The effects of fuel densification i

on the FSAR accident analyses have also been evaluated.

A generic LOC # analysis for the B&W 177-fuel assembly, lowered loop plant design has been performed using the Final Acceptance Criteria (F/C) emergency core cooling system (ECCS) evaluation model (Ref. S as updated by Refs. 20-and 21). That analysis used the limiting values of the key parameters for all plants in the 177-fuel assembly, lowered loop class and is, therefore, bounding for TMI-1 Cycle 6 plant operation. Table 7-7 of Reference 2 presents the limiting values of the allowable LOCA LHGRs for TMI-1 Cycle 6 fuel.

The radiological dose consequences of the accidents presented in the FSAR have been reevaluated for Cycle 6.

The reason for the reevaluation is the increaseo amount of energy produced by fissioning plutonium caused by the extended cycle fuel nanagement strategy. The bases used in the radio-looical dose evaluation are the same as in the FSAP except for two factors-(1) the fission yields and half-lives used in the Cycle 6 evaluation are based on more current data, and (2) the steam generator tube rupture (SGTF) accident considers the increased amount of steam released to the environ-ment because of a post-TPI modification. The radiological doses are still a small fraction (10'4) of the 10 CFR Part 100 limits and are consistert with those of the reference cycle, f

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s

. L4 conclude fror the review of Cycle 6 core therral and kinetic parameters, with respect to previous cycle values and with respect to the FSM selues, that this reload core will rct adversely affect TVT

• ple.r t's ability to operate safely during Cycle C.

7.0 TECHr? CAL SPECIFICATIONS I.s indicated ir our evaluation of the ruclear design, provided in Secticr. 4, the operating characteristics of Cycle 6 were calculated vith well-established, approved methods. The proposed Technic 61 Erecificetiers are the result of the cycle-specific analyses for power peakiro, control rod worths, and quadrant tilt allowance. The analyses performed include the implementatior of a low leakage fuel shuffle patterr:, the impleren-tation of a crossflow thermal-hydraulic analysis, and the removal of the power level hold reouirements for transient xenon. The removel of thr power level cut-off to accommodate transient xeren effects is discussed in Section 4 We conclude that the Technical Specification chances prcrosed by the licensee in Reference 1 and repeated in Section 8 of the Cycle 6 reload report (Ref. 2) are acceptatle. The proposed Techricel Specification changes are as follows:

1.

Pages vii and viii are changed to accommodate new Cvcle C figures. Since these chances are administrative, they are acceptable.

2.

Basis page 2-2

.a The description of the curve on Figure 2.1-1 is changed tn indicate that the curve no longer represents mininue DNPP conditions. This change is acceptable since it more accurately describes the curve presented on the fioure.

The axial peaking facter is changed to 1.65 (ard concomitantly the total peaking factor is changed to 2.F2). These changes are acceptable sirce they are reflected in the Cycle 6 analysis and, in particular, the implementation of the crossflow medel in the DNBF. analysis.

+

The centeriire fuel melt LHGR is chanced to 20.5 W/ft. This chance is acceptable since it is reflected in the Cycle 6 analysis and, in particular, the chance to the TAC 02 fuel analysis methcdolocy.

3.

Basis race ?-3 The descriptier of the curves on Figure 2.1-3 is changed to indicate that the curves no longer represent minimum PhER 7

conditions. This change is acceptable sirce it more accurate y describes the curves presenteo on the #icure.

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6

-F-The thermal pcwers for three pumo creration are truncated f rom two dccimal to ore decimal fictre. This chanae is acceptable. since it reflects more rehlistically the accuracy in measured therral povers.

Refererres to FSAP. sections arr changed. These changet tre administrative since they reflect the updated FSAR.

4 Figure 2.1-P, Core Protection Safety limitt-The curves on the figure present the power-imbalar.ce envelopei, for various pump operation, that meet the cer'erline fuel melt and DNBP criteria. These curves are derivec from Figure F.3-2 by removal of instrumentation error and calculational uncer-tainties in flux / flow determinations. The curves reflect widening of the power-imbalance envelopes as well as increased Reactor Protection System instrument errors.

Figure 2.1-7 is acceptable since it reflects Cycle 6 analysis by beino derivec from Ficure P.3-2.

5.

Figure 2.1-3, Core Protection Safety Bases The curves on this figure have been revised to reflect sracil chances in the 2 and 3 pump allowable power levels. This chance is acceptable since it reflects the revised Reactor Protection Systen instrument errors.

6.

Paoe ?-5, Basis For Technical Specification P.3.1 The nuclear everpower at which a reactor trip occurs has beer revised slightly. This change is acceptable since it reflects the revised Reactor Protec' tion System instrument errors.

7.

Page 2-6, Basis For Technical Specification 2.3.1 The power level at which trip occurs for three pump operatior has been revised. This change is acceptable since it merely reflects the truncation of a number to 80.6 instead of 80.7 8.

Page 2-7, Basis For Technical Specification 2.3.1 The high reactor coolant outlet temperature trip setting limit has been changed to 618.8"F from 619 F.

This chance is acceptable since it reflects the revised Feactor Protection System instrument error, which ir this case is 1.?*F instaad of the previous l'F.

9.

Page 2-9, Table 7.3-1, Peactor Protectier System Trip Settirg Limits

~

This table is chanced to reflect the overocwer and high terperature trip setpnint changes discussed in items 6 and E above.

9 o.

These changes are acceptable for the reasons stated in Itens f and 8 above.

10. Figure 2.3-1, Reactor protectinn System Maximum Allowable Setpoints This figure has been revised to reflect the change in the high reactor coolant outlet temperature trip setting discusseo in Item P above and is, therefore, acceptable.

11.

Figure P.3-2, Reactor Protection System Maximum Allowable Setpoints for Power Imbalance The figure is based on the Cycle 6 analyses discussed in previous sections of this Safety Evaluation. A flux / flow setpoint has been maintained to provide additional margin fer the Cycle 6 pump coastdown analysis.

Since the power / imbalance limits on the figure conservatively bound the thermal limits, they are acceptable.

12. Technical Specification 3.5.2.4, Quadrant Tilt Technical Specification 3.5.2.4.a has been changed to reflect an adjustment in the incore detector uncertainty factor caused by depletion. Therefore, the change in the allowable quadrant tilt to +4.12% from +3.52% is acceptable.

Technical Specification 3.5.2.4.d has been changed to delete reference to the power level cut-off for transient xenon.

This is acceptable as discussed in Section 4 above.

Technical Specification 3.5.2.4.e.1 has been changed for consistency with Technical Specifications 3.5.2.4.e.2 and 3.5.2.4.e.3.

This change is acceptable since it is admir-istrative in nature.

13.

Page 3-34a Technical Specifications 3.5.?.4.e.2 and 3.5.2.4.e.3 are being changed to incorporate new control rod group linits and power imbalance limits derived from the Cycle 6 safety analysis.

These changes are, therefore, acceptable.

Techriical Specification 3.5.2.4.f is changed to revise the order of preference for selecting the system that shall be used to determine quadrant power tilt. The purpose of the change is to insure that the most accurate system available is used to determine quadrant tilt.

It does not change any setpoint, required system accuracy, or surveillance interval.

Therefore, we conclude that the change is acceptable.

p ses

-S

. 14 Page 3-35 Technical Specification 3.5.P.5.6 has been changed to delete reference to the APSPs.

The Cycle 5 figures for APSP position limits are aise deleted. These changes are acceptable since the Cycle 6 safety analysis concludes, and we concur, that oc position limits are required for the Cycle 6 cray APSPs.

Technical Specification 3.5.2.5.c has been deleted since, as discussed in Section 4 above, no power level cut-off for transient xenon is required for Cycle 6.

Technical Specification 3.5.P.S.d specifies the new power imbalance limits for Cycle 6 with renumbered figures.

This is acceptable since renumberino the figures is administrative and the figures are based on the Cycle 6 safety analysis.

Technical Specification 3.5.P.7 reflects a new figure rumber.

This change is acceptable since it is administrative.

15. Page 3-35a, Bases 3.5.2 A sentence has been added to the bases to indicate that the effect of the oray APSRs has been included in the derivation of the power imbalance limits. This change is acceptable since APSR position limits are no longer specified for Cycle 6.

Item f has been included in Bases 3.5.? to indicate that the loss of coolant flow transiert is limiting for portions of Cycle 6 rather than LOCA. This is acceptable _ since it reflects the Cycle 6 safety analysis.

Bases 3.5.2 also includes changes to various figure numbers.

These changes are acceptable since they are administrative.

16. Page 3-36, Bases 3.5.2 A sentence has been added to explain the 16.8% auadrant tilt value used in Technical Specification 3.5.P.4 This change is acceptable since it provides a clarification of the quadrant tilt Technical Specificatior.

17.

Page 3-36a, Bases 3.5.F The trip setpoint for Test Power greater than 757 power has been changed to 105.17. This is acceptable sirce it is based on revised Reactor Protection System errors.

Other chances on the page are administrative and, therefore, acceptable.

F ars

. 18. Techrical Srtcificaticr. 3.5.2.5. Cor. trol Roo F0sitions New control rod insertion lir.its (Ficure 3.5-?A throuch figure 3.5-?I} are provided for 4, 3 ar d ? pump operatinn, as well as a function of burnup interval. These Figures are acceptable since they are based on the Cycle 6 safety analysir discussed previousl).

19. Technical Specification 3.5.?, Peactor Power Imbaurce New reactor power imbalance limits (Figure 3.5-?J and Figure 3.5-?K) are provided for two different burnup intervels.

These Figures are acceptable since they are based on the Cycle 6 safety analysis discussed previously.

20. Fiqure 3.5-PL, LOCA Lirited Maximum Allowable linear Heat Generation Rate For Cycle 6, the LOCA kk'/ft limits have beer. changed der to revisions in the LOCA analysis since Cycle 5.

LOCA limits are provided for three burnup intervals. These changes to the (OCA limits are acceptable since they are based on approved charcer to the LOCA analysis.

Pl. Technical Specification 5.3.1, Reactor Core Chances are'made to this Technical Specification to more accurately reflect the nominal design features of the core including the active fuel assembly length, the average enrichment of the current core, and the use of pray APSRs. These changes are acceptable since they provide for an accurate description of the reactor core.

8.0 STARTUP TESTING We have reviewed the startup physics testing program for TFI-l Cycle 6 presented in Reference P.

We conclude that this program is acceptable since it will provide confirmation that the as-loaded core conforms to the Cycle 6 nuclear design and since the data required by the Technical Specifications will be satisfied.

9.0 SUMPARY We have reviewed the fuel syster desian, nuclear desicn. thermal-hydraulic design, and the transient and accident analysis information presented ir the TMI-I Cycle 6 reload submittal. We have concluded that the propor.ed reload and associated modified Technical Specifications are acceptable.

Additionally, we conclude that changing the order of preference of instruments used te monitor quadrant power tilt is acceptable.

<f v

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10.0 Et'vlRONFFf TAL CCNSIDERATION This amerdment involves chances in the installation or use of a facility component located within the restricted area as defined in 10 CFF rart

20. He heve deternined that the arendment involves no sienificar.t increase in the amounts, and no sier.ificant change in the types, of any effluents thet may be released offsite. and that there is no sionificert ircrease ir individual or cumulative cccupational radiatier exposure.

The Commission has previously issued a proposed findin<; that thi:.

amendmert involves no significant hazards consideraticr. and there. har been.no public comment on such findirt. Accordingly, this anendment meets the eligibility criteria for categorical exclusier, set forth in 10 CFR 51.??(c)(c).

Pursuar.t to 10 CFR 51.??(b), no environme.ntal impact statement or environmental assessment need be prepared in connection with the issuance of this amendment, ll.C CONCLUSION We have cor.cluded, based on the considerations discussed above, that:

(1) there is reasonable assurance that the health and safety of the public will not be endancered by operation in the proposed manner, and (?) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendmert will not be inimical tc the common defense and security or to the health and safety of the public.

Dated:

March 20,1987 Principal Contributor:

D. Fieno

e

. :T.0 REFEPENrES 1.

P. D. Fukill (GPUI let ter (5711-Et-?l8, to J. F. Stolz (NPC) on Technical Specification Change Reauert No. 149, dated November 3, 1986.

T.

"Three Mile Island Unit 1 - Cvcle 6 Peload Report." BAV-1977, Babccc) T Uilcox Cctrpany Report. October 19P6 (transmi*ted with Deference I abovel.

3.

Letter from R. K. Peid (NRC) to J. G. Herbein (Metropclitan Edison Company), dated March 16, 1979.

a.

H. D. Hukill (GPU) letter (5?ll-86-?)03) to J. F. Stolz lhPC) on Fuel Pesinterirg, dated June 9, 1986.

5.

H. D. Hukill (GPU) letter (F?ll-P6-?)76) to J. F. Stolz (NPC) on Fuel Pesintering, dated October 13, 1986.

6.

Y. Hsii, et al., " TACO? - Fuel Pin Performance Analysis,"

Babcock & Wilcox Company Report BAV-10141P-A, Rev. 1, June 1983.

7.

"TMI-1 Fuel Densification Report," Babcock & Wilcox Cerpany Report BAV-1389, June 1973.

8.

Standard Feview Plan, Section 4.2 Rev. 1,

" Fuel System Pesion.

U. S. Nuclear Regulatory Corurission Report NUREG-0000, July 19E1.

I 9.

"ECCS Analysis of B&W's 177-FA Lowered-loop NSS." FfW-10103, Pcv. 3, Babcock & Vilcox, Lynchburg, Vircinia, July 1977.

10.

D. A. Powers and R. O. Meyer, "Claddino Swelling Podels for LOCA Analysis," U. S. Nuclear Pegulatory Corsnission Feport NUREG-0630, April 1980.

11. " Babcock & Wilcox Version of PD0 - User's Fanual," Hassan, H.

P.,

et al., BAF-10117P-A, Babcock & Wilcox. January 1977.

10.

"N0ODLE - A Multi-Dimensional Two-Group Reactor Simulator,"

Mays, C. W., et al., BAW-1015?A, Babcock & Wilcox, June 1985.

13. NPC remorandum from L. S. Rubenstein to D. M. Crutchfield on revier of the N0ODLE code, April 10, 1985.

14.

"TMI-Power Level Cutoff Removal Analysis," T. N. Ake, Babcock & Wilcox Corpany, September 1986.

15. " LYNX 1:

Reactor Fuel Assenbly Thermal Hydraulic Analysis Code,"

BAV-101?9A, Babcock & Wilcox, July 1985.

-r

.: 16.

"LYNXF:

Subchannel Thermel Hydraulic Analysis Code," BAV-10130A, Babcock & Wilcor., July 1985.

17.

"LYNXT: Core Transier.t Thermal Hydraulic Frecram," BAV-1G156, Babccck & Wilcox, February ifs 4 18.

" Thermal Hydraulic Cros: flow Applict.tions," BAV-1829, Fobcock & Wilcox, May 1964 10

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