ML20215N886

Proposed Tech Specs Supporting Cycle 6 Operation

ML20215N886
Person / Time
Site:	Three Mile Island
Issue date:	11/03/1986
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20215N870	List: ML20215N869 ML20215N877 ML20215N886 ML20215N891
References
4194F, NUDOCS 8611100031
Download: ML20215N886 (43)

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I. TECHNICAL SPECIFICATION CHANGE REQUEST (TSCR) NO. 149 GPUN requests that the followino chanced replacement pages be inserted into the existing Technical Specifications:

Revised pages:

, vii, viii, 2-2, 2-3, 2-5, 2-6, 2-7, 2-9, 3-34, 3-34a , 3-35, 3-35a ,

3-36, 3-36a, 5-4 Revised ffpures:

2.1-2, 2.1-3, 2.3-1, 2.3-2, 3.5-2A, 3.5-28, 3.5-2C, 3. 5-20, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, 3.5-21, 3.5-2J, 3.5-2K, 3.5-2L These paces are attached to this chance request.

II. REASON FOR CHANGE This change is requested to provide Technical Specifications for operation of TMI-1 Cycle 6.

III. SAFETY EVALUATION JUSTIFYING CHANGE i

The attached Reload Report (BAW-1977, October 1986) and the revised Technical Specifications provide the desian and administrative bises to support the conclusion that Cycle 6 can be operated at full power without exceedina the established safety criteria.

Cycle 6 has been designed as a typical B&W lumped burnable poison rod core based on an in-out-in fuel management strategy. The low neutron leakage cycle desian is consistent with GPUN reactor vessel fluence reduction efforts for TMI-1 as described in our response on the Pressurized Thermal Shock rulemakina (GPUN letter 5211-86-2007, Janua ry 23,1986).The reference cycle for core analyses is Cycle 5 which is scheduled for completion in early November 1986 af ter 290 + 15 EFPD.

Operation of Cycle 6 is scheduled to start in March 1987 with a desion cycle lenath of 425 + 15 EFPD.

Core Loadina Cycle 6 loading of Batch 8 includes 12 fresh Mark B4 fuel assemblies at 2.95 weiaht percent enrichment (Batch 8A) and 64 fresh Mark B4 fuel assemblies at 2.85 weicht percent enrichment (Batch 88). The remainder of the core is comprised of 52 once-burned Batch 7 and 49 twice-burned Batch 6 assemblies with an initial enrichment of 2.85 weicht percent.

Reactivity Centrol Cycle 6 will be operated in a feed and bleed mode. Reactivity is controlled by 61 full-length Aa-In-Cd control rods, 68 burnable poison rod assemblies and soluble boron control. Control rod aroup locations and desionations have been changed to be consistent with the extended-cycle desian.

4194f 8611100031 861103 PDR ADOCK 05000289 P PDR

1 ? h Cycle 6 will utilize advanced desian "oray" axial power shapino rods which decrease fuel duty due to local power chanaes while providina adequate power peakina and imbalance control.

Fuel System Desian The Batch 8 Mark B4 fuel assembly desian is identical to that of the Batch 6 and 7 Mark B assemblies previously utilized in Cycles 4 and 5.

Positive lockina retainer devices will be installed on all BPRAs and on the two reaenerative neutron source assemblies. The retainers were also used for Cycle 5.

Fuel rod thermal and mechanical analyses were performed usina previously approved codes. Cladding collapse analysis was done with the approved CROV code (BAW-10084P, Rev. 2) using conservative power histories.

Results showed that the collapse time for the limitina fuel assembly (Batch 6) in Cycle 6 is areater than the pro.iected residency time.

Claddina stress and strain evaluations showed that the Batch 8 fuel is bounded by the analysis in the TMI-l fuel Densification Report

( B AW-1389) .

Thermal desian analyses for the Cycle 6 fuel were performed usina the approved TACO-2 code (BAW-10141P-A, Rev.1), which utilizes a time dependent fuel densification model. All fuel batches were determined to have a linear heat rate capability to centerline fuel melt of 20.5 kw/f t which is preserved by the core protection limits. Maximum assembly burnup at E0C is less than 33,700 MWD /mtU and E0C internal pin pressure for the hiahest burnup fuel rod is less than system oressure. Both conditions are within the 88W desian limits.

The results of the TNI-1 Fuel Densification Reoort remain boundina for all Cycle 6 fuel since those analyses were based on a lower initial pellet density and an assumption of densification to 96.5% TD. Results of resinter tests on fuel pellets representina all fabrication lots comprisina Batch 8 have confirmed the conservatism of densification '

characteristics assumed in the TACO-2 analyses. Information reaardina resinter data for insertion Batch 8 fuel has been provided previously via GPUN letters 5211-86-2103 (June 9,1986) and 5211-86-2176 (October 13, 1986).

Thermal-Hydraulic Desian Batch 8 fuel is hydraulically and aeometrically similar to Batches 6 and

7. The Cycle 6 thermal-hydraulic desian utilized the approved LYNX crossflow codes (BAW-10129A, BAW-10130A, BAW-10156) for the THI-l DNB predictions. Insertion of the 68 BPRAs decreased bypass flow from 10.4%

in Cycle 5 to 7.6% for Cycle 6. A conservative value of 8.4% was used in the analysis. Also for Cycle 6 the rod bow penalty on DNBR was eliminated based on an approved evaluation (BAW-101478-A). These chances have resulted in a 5% improvement in steady-state DNBR for Cycle

6. This DNBR improvement results in additional marain to the DNB-based RPS trip setpoints for variable low-pressure, low pressure, hiah temperature and flux / flow (Table 2.3-1).

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,e i Nuclear Design Core design chances for Cycle 6 include the increase in cycle lifetime to 425 EFPD, the use of burnable poison rod assemblies (BPRA) to aid reactivity control and the use of gray axial power shapino rod assemblies (APSR). The gray APSRs contain a lonaer poison rod with a weaker absorber made of Inconel replacing the Ag-In-Cd poison APSRs used previous 1v. Nuclear design calculations were performed usino the approved codes PDQ 07 (BAW-10117P-A) and N0ODLE (BAW-10152A). The latter was applied to TMI-1 for the first time. Core relative power distributions were done usino PDQ 07; physics parameters were produced usino N0ODLE.

The lonaer cycle lenoth and use of BPRAs resulted in laroer changes in certain physics parameters than normally occurred between annual cycle reloads for TMI-1. Cycle 6 BOC critical boron concentrations increased because the additional core reactivity is not completely offset by the BPRAs. Control rod worths, includino ejected rod, differ due to chances in radial flux and burnuo distributions. Cycle 6 physics parameters are shown in Table 5-1 of the attached Reload Report and are considered to be reasonable and consistent with the core desian chances. The B0C and E0C shutdown margins shown in Table 5-2 of the Reload Report have increased from the Cycle 5 values and meet the 1% maroin criteria.

Accidents and Transient Analysis The chances in Cycle 6 parameters have been evaluated to determine effects on the FSAR accident analyses.

For each accident or transient the key core thermal, thermal-hydraulic and kinetic parameters for the cycle were compared to FSAR, TMI-1 Densification Report and/or reference cycle values to determine whether the event remained bounded by the previous analyses. Cycle 6 parameters were also compared to the generic LOCA analyses.

Core thermal and thermal-hydraulic parameters are common to all ,

accidents. Cycle specific parameters in both of these categories were shown to be conservatively bounded by previous analyses. These comparisons are shown in Sections 4 and 6, respectively, of the Cycle 6 Reload Report. For all reload dependent accidents, the initial conditions defined by the cycle-specific parameters will produce less severe transients than the initial conditions assumed in previous analyses. Cycle 6 transients are thus bounded by previously accepted analyses, and no re-analysis was necessary. The accidents examined included:

Rod Withdrawal Accidents Moderator Dilution Accident Cold Water Accident Loss-of-Coolant Flow Stuck-out, Stuck-in, or Dropped Control Rod Steam Line Failure Feed Line Failure Rod Ejection Accident Uncompensated Operatino Reactivity Chances 4194f

e 9 )

The EOC Doppler coefficient is more neoative than the value used in the FSAR. Althouoh it is not conservative to have a more negative EOC Doppler for dropped rod accident, the transient results are still conservative with respect to the FSAR analysis as a result of the smaller maximum dropped rod worth and the less neoative moderator coefficient for this cycle. Also, the moderator effect is more than an order of maanitude areater than the Doppler.

The consequences of certain transients and accidents are only affected by a reload if core isotopic inventory chances, i.e., they are not affected by thermal or reactivity parameters but rather by radiolooical considerations. These include:

Loss of Electric Power Steam Generator Tube Failure Fuel Handling Accident Maximum Hypothetical Accident LOCA Analysis Waste Gas Tank Rupture Acceptability is based on offsite doses. Since the potential radiolooical releases are based on Technical Specification limits or core power rating, the consequences of these events normally need not be reevaluated. Improved fuel utilization and improved calculational methods have resulted in a hiaher plutonium / uranium fission vield.

Thus, the radiolooical consequences of these accidents were recalculated to evaluate the Cycle 6 core isotopics effects.

Thyroid dose for FSAR Chapter 14 accidents that are affected by isotopic inventory increase by a maximum of 17% for Cycle 6 compared to FSAR values and remain Well within the limits of 10 CFR 100.

Small and larae break loss of coolina accident (LOCA) analyses are primarily dependent upon overall system response rather than reload core characteristics. Cycle 6 key core thermal and reactivity parameters have been compared to the Ifmitino values used in the generic analyses done for these accidents and determined to be bounded. Therefore, the ECCS Final Acceptance Criteria will be met by observance of the Ifnear heat rate (LHR) limits for Cycle 6 shown in Table 7-2 of the Reload Report. The impact of NUREG-0630 cladding swellino and rupture models have been included in the calculation of the boundino LHRs.

Proposed Modifications to Technical Specifications The Technical Specification limit curves were revised for Cycle 6 operation to account for chances in power peakina and control rod worths that can be caused by conversion to a BPRA-loaded extended cycle. The new curves also reflect the implementation of the crossflow thermal-h.ydraulics model and the impact of the NUREG-0630 claddino swell and rupture model.

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j Revised RPS instrumentation errors, as recommended by BW and reviewed and accepted by GPUN, specific to TMI-1 instruments and based on updated

(

BW error combination methods have been used to establish the Cycle 6

, setpoints ifmits. Also, evaluation has confimed that the limits. .

i provide protection for the overpower condition that has been oostulated ^-

durino an overcooling event due to transient-induced nuclear "

i instrumentation errors. ,

l Pace vil and viii Administrative chances to Table of Contents. .,

L Pace 2-2 s i

j The description of the P-T curve on Figure 2.1-1 is chanced to indicate 'C that the curve no longer represents minimum DNBR conditions.

• 1 .

] The page includes the chance to a 1.65 axial desion peaking factor, _ ,.

] Fl,consistentwithimplementationofthecrossflowmodel,alsothe ',

i chance in total peak F to 2.82 consistent with the new 1.65 axial .

peak. '

]

d P

The CFM Linear Heat Rate capability has changed from 20.15 kw/f t in . 4 I Cycle 5 to 20.50 kw/f t in Cycle 6 due to the chance from the TAFY to the '

} TACO-2 code.

I Pace 2-3 di .

The description of the P-T curves of Ff oure 2.1-3 is chanced to indicate

! that the curve no lonaer represents minimum DNBR conditions. j j .

.s i Chance reflects truncated ad,iustments to 89.3 from 89.36 and 80.6 from t

80.67 used in Cycle 5. Also, slicht increases in maximum allowable i therwal powers for 2 and 3 pump operation have been made due to

, application of revised RPS instrument errors. (See discussion below for I Fiqures 2.1-2). -

4 Chances to References are administrative to be consistent with updated -

FSAR.

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i' Ff oure 2.1-2 (Core Protection Safety Limits)  %

i j The figure establishes imbalance windows because power peakino is not directly observable. The limits preserve the center-line fuel melt (CFM) and/or DNBR criteria always applying the more restrictive '

condition to establish the limit. The curves are derived from Fioure 2.3-2 (RPS Setpoints) by removal of instrumentation error and  ; .-

l calculational uncertainties in flux / flow determinations. Thus, since,.

certain RPS instrument errors have increased relative to those used for

~

Cycle 5 Technical Specifications, the maximum power level limits have increased s11ohtly for the 2- and 3- pump curves. These limits remain ,

f conservative since setpoint power levels of Fioure 2.3-2 are ratioed i

from the conservative 1.08 flux / flow setpoint discussed under that -

figure. The imbalance window has been widened to accommodate potential r changes due to the core desian but preserve the DNBR and'CFM criteria.

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(_ -. . , , _ _ . , , _ _ _ . -

7 Fiqure 2.1-3 (Core Protection Safety Bases)

This figure also reflects the sliaht chances in 2- and 3-pump allowable power levels due to the revised RPS errors.

Paae 2-5 The chance in nuclear overpower is shown which is also due to implementation of the revised RPS errors.

Paae 2-6 The chanae reflects a truncated ad.iustment to 80.6 from the 80.7 roundoff value used in C.ycle 5.

Paae 2-7 The chance in high temperature trip setpoint strina accuracy due to application of the revised RPS errors is shown: 1.2*F vs. l'F for Cycle

5. This chances the setpoint to 618.8'F from fl9'F.

Table 2.3-1, Paae 2-9 (RPS Trip Settina Limits)

The table reflects the overpower and hiah temperature trip setpoint chanaes due to revised RPS errors as discussed above. Al so, typographical correction is made to footnote (1).

Fiqure 2.3-1 (RPS Maximum Allowable Setpoints)

The figure shows the revised hiah temperature setootnt.

Fiaure 2.3-2 (RPS Maximum Allowable Setpoints for Power Imbalance)

This fiaure conservatively establishes power and imbalance limits based on Cycle 6 analyses. A flux / flow setpoint of 1.08 has been maintained to provide additional maroin to the Cycle 6 pump coastdown anal.ysis which would allow a flux / flow limit of approximately 1.13. The 3- and 2-pump overpower values are ratioed from the 4-pump 1.08 value based on flows of 74.7% and 49.2%, respectively, of design flow. The imbalance setpoints conservatively envelope the thermal limits.

Pace 3-34 Section 3.5.2.4.a notes the chanae in allowable quadrant tilt usino the full incore system due to adjustment of the incore detector uncertainty factor caused by depletion.

Section 3.5.2.4.d deletes reference to the power level cutoff as discussed below under Page 3-35.

Section 3.5.2.4.e.1 contains an administrative change to be consistent with Sections 3.5.2.4.e.2 & 3.

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Page 3-34a Changes are to fioure ' numbers and are administrative. The chance to existina Technical Specification Section 3.5.2.4.f is beino processed separately as part of TSCR No.159 (GPUN letter 5211-86-2125, July 16, 1986). A marain bar has been inserted since the amendment for this change has not yet been issued. u Paae 3-35 Section 3.5.2.5 b has been chanced to delete references to APSR position limits. APSR position limits evaluations are performed in con.1 unction with the control 2 rod position and power imbalance LCOs to preserve the transient ~ analyses. Cycle 6 has implemented the lower worth advanced design " gray" APSR which will areatly reduce power peakina due to APSR movements. -The' Cycle 6 analyses have confirmed that no APSR position limits are required to , preserve the thermal criteria. ,

(

Cycle 5 Figures 3.5-2H dnd 3.5-2I for APSR Position Limits have also been deleted.

Section3.5.'2.5chadbeendeleted. This section contained restrictions on operation above the power level cutoff ,(PLC) of 92% full power.

Removal of the PLC was based on c.ycle-spec'ific transient xenon analyses y using approved B&W models to simulate the transients and calculate power -

distribution peakina factors. The analyses considered power level changes, control rod positions, imbalance and time in cycle. Results were compared to Cycle 6 LOCA and initial-condition DNBR peakino conditions derived usina standard methods (equilibrium xenon and aeneric xenon factors) upon which the Cycle 6 limits are based. The transient xenon conditions were determined to be bounded by the standard analysis thus precluding the need for a power level cutoff below 100% full power.

Other chances are to figure numbers and a're administrative.

Paae 3-35a Since a figure is no lonaer requf red for APSR Position Limits due to the implementation of oray APSRs a santence has- been added to note that the effects of the arays have been included in constructino the LCOs.

Item f in the bases for Section 3.5.2 has been added to indicate that the control rod and power distribution limits for Cycle 6 include the effects of increased radial peakina due to the LBP extended-cycle desian which causes the loss of coolant flow transient to become limitino rather than a LOCA for portions of the cycle.

Other changes are to fioure numbers and are administrative.

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Pace 3-36 Chanae provides explanation of the bases for the 16.8% tilt value.

Pace 3-36a The chance in nuclear overoower is shown which is due to implementation of the revised RPS errors.

Other chances are editorial only.

Ficures 3.5-2A to 3.5-21 (Rod Position Limits for 4, 3, and 2-Pump Operation)

The LC0 control rod index fiaures were established based on the Cycle 6 analyses to preserve shutdown marain and accident analysis criteria.

Curves are provided for three burnup windows rather than the two windows used for Cycle 5 to reflect the B0C impact of NUREG-0630 claddina damaae model s. Rod limits for 2- and 3-pump operation were developed by power scaling from 4-pump values. Separate curves have been provided for 2-and 3-pump operation showina actual power levels on the scale rather than superimposed curves as in the past with a "% maximum allowable power" scale. Tnis was done for ease of use for the operators.

The "Not Allowed" reaion is defined by the 1% shutdown marain criteria.

This region has been decreased since Cycle 5 because it is based on actual Cycle 6 analyses values, whereas the Cycle 5 limits were set to bound future cycles. Ejected rod limits are bounded by the shutdown marain limits. The " Restricted" region is defined by thermal accident criteria. In the past for annual out-in cycles this curve was established by LOCA analysis results. Because of the conversion to the LBP extended cycle and a consequent increasina radial peak during the cycle as the LBPs deplete, the controlling criteria became initial condition DNBR rather than LOCA. The " Restricted" curve preserves total allowable peaking limits and DNBR criteria during the limitina transient. LOCA limits are also bounded by the curve.

Fiaures 3.5-2J and 3.5-2K (Power Imbalance Envelope)

The power imbalance LCOs are developed in conjunction with the rod.

position limits and provide imbalance limits that preserve the transient analyses. The Cycle 6 analyses served as the basis for the curves.

Figure 3.5-2J includes the NUREG-0630 cladding model impacts.

Also, editorial chanaes made to renove redundant term " axial" from description.

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Fiaure 3.5-2L (LOCA Limited Maximum Allowable LHR)

The Cycle 6 LOCA kw/f t limits reflect three burnuo windows due to revisions in the ECCS evaluation since Cycle 5. The most severe restriction is from 0-1000 MWD /mtU (about 30 EFPD) due to the combined effects of implementation of the NUREG-0630 cladding damaae models and the TACO-2 fuel model into the ECCS analysis. These limits are chances from the Cycle 5 interim LOCA limits and do not include an offset to the NUREG-0630 penalty through the use of the FLECSET heat transfer coefficient improvement model. The limits preserve the LOCA Final Acceptance Criteria.

Pace 5-4 The core design description is chanced to reflect the current core design.

Startup Proaram - Physics Testina The startup physics test program for Cycle 6 provides the testing necessary to determine that the characteristics of the operating core are consistent with the design predictions.

Major parameters to be tested include core excess reactivity, control rod worth, temperature and boron coefficients, radial and total power distribution and Doppler coefficient. These are measured and compared to the same acceptance criteria as used for the reference cycle.

Since Cycle 5, tests included in the startup proaram primarily to benchmark desian calculations have been eliminated. These include:

ejected rod worth, dropped rod worth, multiple coefficient tests, distorted Dower distributions and atypical peakino confiaurations.

Removal of these. f asts is .iustified by the broad database developed for B&W 177 Fuel Assembly plant startups with the standard in-out-in low leakaoe core design similar to Cycle 6.

Conclusion Based on the discussions aiven above, it is concluded that the desian and Technical Specification limits developed for Cycle 6 support full power operation at 2535 MW for a cycle lenoth of 0-425 + 15 EFPD without endanaering the health and safety of the public.

IV. No Significant Hazards Considerations GPUN has determined that this Technical Specification Chance Request poses no.sionificant hazards as defined by NRC in 10 CFR 50.92.

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1. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

The FSAR Chapter 14 accidents that depend on the. core reactivity parameter changes due to Cycle 6 reload have been reviewed. The results of this review confirm that those reload-dependent events remain boundina for Cycle 6 operation. The thyroid doses for Chapter 14 accidents that are only affected by core isotopic inventory increase by a maximum of 17% for Cycle 6 compared to the FSAR values and remain well within the limits established by 10 CFR 100. Therefore, Cycle 6 operation in accordance with the proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from an.y accident previously -evaluated. Review of Cycle 6 core loading design, fuel assembly thermal-hydraulic desian, and analytical methods confirms that there are no safety concerns regarding the Cycle 6 core desian. The controlling Cycle 6 reactivity - dependent parameters of sianificance in transient and accident analyses are conservatively bounded by the corresponding FSAR values used in the docketed transient and accident analyses.

Cycle 6 fresh fuel assemblies are hydraulically and aeometrically similar to previously irradiated fuel assemblies. Cycle 6 is designed as a standard B&W in-out-in low leakaoe core. Therefore, it is concluded that Cycle 6 operation in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed amendment would not involve a sionificant reduction in a marain of safety.

Cycle 6 characteristics are conservative with respect to previous accident analyses, and all safet.y criteria as described in the Technical Specification bases are preserved by the revised limits.

Therefore, it is concluded that Cycle 6 operation does not involve a significant reduction in' a margin of safet.y.

The Commission has provided guidelines pertainino to the application of the three standards by listing specific examples in 48 FR 14870. The proposed amendment is considered to be in the same cateaory as example (iii) of amenoments that are considered not likely to involve significant hazards consideration in that the proposed change results .

from a nuclear reactor core reloadino, no fuel assemblies sionificantly l different from those found previously acceptable .to the NRC for the previous core at TMI-l are involved, and it has been adequately demonstrated that the acceptance criteria for the Technical Specifications have not been significantly chanaed, that the analytical i

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methods utilized to demonstrate conformance with the Technical Specifications and reaulations are not significantly chanced, and that the NRC has previously found such methods acceptable. Thus, operation of the facility in accordance with the proposed amendment involves no sionificant hazards considerations.

V. IMPLEMENTATION It is requested that the amendment authorizing this chance become effective upon startup following the Cycle 6 refuelina outaae.

VI. . AMENDMENT FEE (10 CFR 170.21)

Pursuant to the provisions of 10 CFR 170.21, attached is a check for

$150.00.

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ATTACHMENT Technical Specification Changes l

LIST OF FIGURES Figure Title 2.1 -1 TMI-1 Core Protection Safety Limit 2.1-2 TMI-l Core Protection Safety Limits 2.1-3 TMI-1 Core Protection Safety Bases 2.3-1 TMI-l Protection System Maximum Allowable Set Points 2.3-2 Protection System Maximum Allowable Set Points for Reactor Power Imbalance, TMI-l 3.1 -1 Reactor Coolant System Heatup/Cooldown Limitations (Applicable to 5 EFPY) 3.1 -2 Reactor Coolant System, Inservice Leak and Hydrostatic Test' Limitations (Applicable to 5 EFPY) 3.1 -3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter HO 2

3.5-2A Rod Position Limits for 4 Pump Operation from 0 to 30+10/-0 EFPD, TMI-1 3.5-28 Rod Position Limits for 4 Pump Operation from 30+10/-0 to 25010 EFPD, TMI-l 3.5-2C Rod Position Limits for 4 Pump Operation af ter 25010 EFPD, TMI-l 3.5-2D Rod Position Limits for 3 Pump Operation from 0 to 30+10/-0 EFPD, TMI-1 3.5-2E Rod Position Limits for 3 Pump Operation ~ from 30+10/-0 to 25010 EFPD, TMI-l 3.5-2F Rod Position Limits for 3 Pump Operation af ter 25010 EFPD, TMI-l 3.5-2G Rod Position Limits for 2 Pump Operation from 0 to 30+10/-0 EFPD, TMI-l 3.5-2H Rod Position Limits for 2 Pump Operation from 30'+10/-0 to 25010 EFPD, TMI-l 3.5-2I Rod Position Limits for 2 Pump Operation af ter 25010 EFPD, TMI-1 3.5-2J Power Imbalance Envelope for Operation from 0 to 30+10/-0 EFPD, TMI-l vii Amendment Nos. J A , A/, 29, ~ 39, Ah, 50, 59, 72, AM, J A9, 120

LIST OF FIGURES Figure Title 3.5-2K Power Imbalance Envelope for Operation af ter 30+10/-0 EFPD, TMI-1 3.5-2L LOCA Limited Maximum Allowable Linear Heat Rate

3. 5-1 .Incore Instrumentation Specification Axial Inbalance Indication,5 TMI-l 3.5-2 Incore Instrumentation Specification Radial Flux Tilt Indication, TMI-l 3.5-3 Incore Instrumentation Specification 3.11-1 Transfer Path to and from Cask Loading Pit 4.17-1 Snubber Functional Test - Sample Plan 2 -

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5-1 Extended Plot Plan TMI 5-2 Site _ Topography 5 Mile Radius 5-3 Site Bnundary for Gaseous Effluents 5-4 Site Boundary for Liquid Effluents 6-1 GPU Nuclear Corporation Organization Chart 6-2 TMI-l Onsite Organization viii Amendment Nos. 77, 77

a conservative margin'to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been considered in determining the core protection safety limits. The difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was assumed in reducing the pressure trip set points to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1 represents the conditions at which a DNBR of 1.3 or greater is predicted for the maxs num possible thermal power (112 l percent) when the reactor coolant flow is 139.8 x 10+6 lbs/h, which is less than the actual flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors (2) with potential fuel densification and fuel rod bowing effects; N N N F = 2.82, F = 1. 71 ; F = 1. 65 q AH z The 1.65 axial peaking factor associated with .the cosine flux shape provides a l lesser margin to a DNBR of 1.3 than the 1.7 axial peaking factor associated with a lower core flux distribution. For this reason the cosine flux shape and the associated Fy = 1.65 is more limiting and thus the more l consawaMy assumption.

. The 1.65 cosine axial flux shape in conjunction with F AH = 1.71 define the l reference . design peaking condition in the core for operation at the maximum overpower. Once the reference peaking condition and the associated thermal-hydraulic. situation has been established for the hot channel, then all other combinations of axial flux shap0s and their accompanying radials must result in a condition which will not violate the previously established design criteria on DNBR. The flux shapes examined include a wide range of positive and negative offset for steady state and transient conditions.

These design limit power peaking f actors are the most restrictive calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion, and form the core DHBR design basis.

The curves of Figure 2.1-2 are based on the more restrictive of two thermal limits and include the effects of potential fuel densification and fuel rod bowi ng:

a. Thg 1.3 DNBR limit produced by a nuclear power peaking factor of F 2.82 of the combination of the radial peak, axial peak, and l poS=ition of the axial peak that yields no less than 1.3 DNBR.

b. The combination of radial and axial peak that prevents central fuel melting at the hot spot. The limit is 20.50 kW/ft. l Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power inbalance produced by the power peaking.

2 and 3 The specified flow rates for curves 1the expected three minimum flow pumps, rates and one with pumpfou,r in pumps,of Fi each loop, respectively.

2-2

The curve of Figure 2.1-1 is the most restrictive of all possible reactor coolant pump-maximum thermal power combinations shown in Figure 2.1-2. The curves of Figure 2.1-3 represent the conditions at which a DNBR of 1.3 or greater is predicted at the maximum possible thermal power for the number of reactor. coolant pumps in operation or the local quality at the point of minimum DNBR is equal to 22 percent, (3) whichever condition is more restrictive.

The maximum thermal power for three pump operation is 89.3 percent due to a l power level trip produced by the flux-flow ratio (74.7 percent flow x 1.08 =

80.6 percent power) plus the maximum calibration and instrumentation error. l The maximum thermal power for other reactor coolant pump conditions is produced in a similar manner.

Using a local quality limit of 22 percent at the point of minimum DNBR as a basis for curve 3 of Figure 2.1-3 is a conservative criterion even though the quality at the exit is higher than the quality at the point of minimum DNBR.

The DNBR as calculated by the B&W-2 correlation continually increases from the point of minimum DNBR, so that the exit DNBR is always high'er and is a function of. the pressure.

For each curve of Figure 2.1-3, a pressure-temperature point above and to the lef t of the curve would result in a DNBR greater than 1.3 or a local quality at the point of minimum DNBR less than 22 percent for the particular. reactor coolant pump situation. Curve 1 is more restrictive than any other reactor coolant pump situation because any pressure / temperature point above and to the lef t of this curve will be above and to the lef t of the other curves.

REFERENCES (1) FSAR, Section 3.2.3.1.1 (2) FSAR, Section 3.2.3.1.1.3 l (3) FSAR, Section 3.2.3.1.1.11 i

1 l

1 2-3 l

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Thermal Power Level, %

- 120

_ (-39.4,112) (38.1,112) 1

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ACCEPTABLE ~ 110 4 PUMP OPERATION

- - 100

_(-39.4,89.3) 90 (38.1.89.3) 1 ACCEPTABLE

(-61.3,80.4) '

3 & 4 PUMP -

- 80 L (44.2,80.4)

OPERATION

- 70

_(-39.4.62.6) (38.1.62.6) 3

(-61.3,57.8)

60 '

ACCEPTABLE (44.2,57.8) 2,3, & 4 PUMP OPERATION

- 50

- 40

(-61.3,30.4) -

- 30 (44.2,30.4)

-- 20

-- 10 i i i i i , e i e i e i I ' ' '

70 50 30 10 0 10 20 30 40 50 60 70 80 Reactor Power Imbalance, %

Curve Reactor Coolant Flow (Ib/hr) 1 139.8 x 10 6 2 104.5 x 10 6 3 68.8 x 10 6 TMI-I CORE PROTECTION SAFETY LIMITS Figure 2.1-2

t i

2400 T

~

h

? 2200 E [

• 2 it E

E 2000 )

/ - T

=

5

(,

1800 - ,

1600 800 820 640 660 560 580 Reactor Outlet Temperature. *F REACTOR COOLANT FLOW CURVE (L8S/HR) POWER PUMPS.0PERATING (TYPE OF LIMIT) 1 139.8 106 (1005)* 112; Four Pume.s (DNBR Limit) 2 104.5 x 106 (74.75) 89.3% Tnree Pumps (DNBR Limit) 3 88.8 x 106 (49.25) 62.6% One Pump in Eacn Loop (Quality Limit)

• 106.55 of Cycle 1 0: sign Flow TMl-1  ;

CORE PROTECTION SAFETY BASES Figure 2.!-3

2.3 LIMITING SAFETY SYSTEM SETTINGS, PROTECTION INSTRUMENTATION Applicability Applies to instruments monitoring reactor power, reactor power imbalance, reactor coolant system pressure, reactor coolant outlet temperature, flow, number of pumps in operation, and high reactor building pressure.

Objective To provide automatic protection action to prevent any combination of process variables from exceeding a safety limit.

Specification 2.3.1 The reactor protection system trip setting limits and the permissible bypasses for the instrument channels shall be as stated in Table 2.3-1 and Figure 2.3-2.

Bases The reactor protection system consists of four instrument channels to monitor each of several selected plant conditions which will cause a reactor trip if any one of these conditions deviates from a pre-selected operating range to the degree that a safety limit may be reached.

The trip setting limits for protection system instrumentation are listed in Table 2.3-1. These trip setpoints are setting limits on the setpoint side of the protection system bistable comparators. The safety analysis has been based upon these protection system instrumentation trip set points plus calibration and instrumentation errors.

Nuclear Overpower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions too rapid to be detected by pressure and temperature measurements.

During normal plant operations with all reactor coolant pumps operating, reactor trip is initiated when the reactor power level. reaches 105.1% of rated l power. Adding to this the possible variation in trip set points due to calibration and instrument errors, the maximum actual power at which a trip would be actuated could be 112%, which is the value used in the safety

. analysis (1).

a. Overpower trip based on flow and imbalance The power level trip set point produced by the reactor coolant system flow is based on a power-to-flow ratio which has been established to accommodate the most severe thermal transient considered in the design, the loss-of-coolant flow accident from high power. Analysis has demonstrated that the specified power to flow ratio is adequate to prevent a DNBR of less than 1.3 should a low flow condition exist 'due to any malfunction.

2-5

The power level trip set point produced by the power-to-flow ratio provides

-both high power level and low flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases. The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all modes of pump operation. For every flow rate there is a maximum permissible power level, and for every power level there is a mininun permissible low flow rate. Typical power level and low flow rate combinations for the pump situations of Table 2.3-1 are as follows:

1. Trip would occur when four reactor coolant pumps are operating if power is 108 percent and reactor flow rate is 100 percent, or flow rate is 92.5 percent and power level is 100 percent.

2. Trip would occur when three reactor coolant pumps are operating if power is 80.6 percent and reactor flow rate is 74.7 percent or flow l rate is 69.4 percent and power level is 75 percent.

3. Trip would occur when one reactor coolant pump is operating in each loop (total of two pumps operating) if the power is 53.1 percent and reactor flow rate is 49.2 percent or flow rate is 45.3 percent and the power level is 49 percent.

The flux / flow ratios account for the maximum calibration and instrumentation errors and the maximum variation from the average value of the RC flow signal in such a manner that the reactor protective system receives a conservative indication of the RC flow.

No penalty in reactor coolant flow through the core was taken for an open core vent valve because of the core vent valve surveillance program during each refueling outage.

For safety analysis calculations the maximum calibration and instrumentation errors for the power level were used.

The power-imbalance boundaries are established in order to prevent reactor thermal limits from being exceeded. These thermal limits are either power peaking Kw/f t limits or DNBR limits. The reactor power imbalance (power in the top half of the core minus power in the bottom half of core) reduces the power level trip produced by the power-to-flow ratio so that the boundaries of Figure 2.3-2 are produced. The power-to-flow ratio reduces the power level trip and associated reactor power / reactor power-imbalance boundaries by 1.08 percent for a one percent flow reduction.

b. Pump Monitors The redundant pump monitors prevent the minimum core DNBR from decreasing below 1.3 by tripping the reactor due to the loss of reactor coolant pump (s). The pump monitors also restrict the power level for the number of pumps in operation.

2-6

l l

, .

• j i

c. Reactor coolant system pressure During a startup accident from low power or 'a slow rod withdrawal from high power, the system high pressure trip set point is reached before the nuclear overpower trip setpoint. The trip setting limit shown in Figure 2.3-1 for high reactor coolant system pressure has been established to maintain the system pressure below the safety limit (2750 psig) for any design transient (6). Due to calibration and instrument errors, the safety analysis assumed a 45 psi pressure error in the high reactor coolant system pressure trip setting.

J The high pressure trip setpoint was subsequently lowered from 2390 psig to 2300 psig. The lowering of the high pressure trip setpoint an raising of the setpoint for the Power Operated Relief Valve (PORV), from 2255 psig to 2450 psig, has the effect of reducing the challenge rate to the PORY while maintaining ASME Code Safety Valve capabil ity.

The low pressure (1800 psig) and variable low pressure (11.75 Tom 5103) trip setpoint were initially established to maintain the DNB ratio greater than or equal to 1.3 for those design accidents that result in a pressure reduction (3,4). The B&W generic ECCS analysis, however, assumed a low pressure trip of 1900 psig and, to establish conformity with this analysis, the low pressure trip setpoint has been raised to the more conservative 1900 psig.

Figure 2.3-1 shows the high pressure, low pressure, and variable low pressure trips.

d. Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (618.8F) shown in Figure 2.3-1 has been established to prevent l excessive core coolant temperature in the operating range.

The calibrated range of the temperature channels of the RPS is 520*

to 620*F. The trip setpoint of the channel is 618.8F. Under the l worst case environment, power supply perturbations, and drif t, the accuracy of the trip string is 1.2*F. This accuracy was arrived at l by summing the worst case accuracies of each module. This is a conservative method of error analysis since the normal procedure is to use the root mean square method.

Therefore, it is assured that a trip will occur at a value no higher than 620*F even under worst case conditions. The safety analysis used a high temperature trip set point of 620*F.

The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drif t, linearity, repeatability, etc. This does not imply that the equipment is restricted to operation within the calibrated range.

Additional testing has demonstrated that in fact, the temperature channel is fully operational approximately 10% above the calibrated ra nge.

2-7

l Tabla 2.3-1 (

I

, REACitR PRDIECTIN SYSIDE 'IRIP SEITDG LDT15(6)

) Pbur Reactor Coolant 'Ihree Reactor Ooolant One Raar+w Omlart .

j Ptaps Operating Pumps Operating 1%mp Operating in l (maninal operating (N ainal Operating Endt Icop (Mainel Shutdown t Power - 100%) Power - 754) Operating Power - 49%) Bypass j 1. NLaclear power, Max. 105.1 105.1 105.1 j  % of rated power 5.0(3) i a 2. Ntaclear power haaari on 1.08 ti m f1 w 1.08 ti m flw 1.08 times flw mismas Bypommed flw (2) and inhalance minus reduction dias mirmas redLaction dine redtaction dias to max. of rated power to inhalance to inhalance inhalance

3. Ntaclear power hn==1 NA NA 55% Bypommed '

(5) on ptmp monitors, j Max. % of rated i power i

4. High reactor coolant 2300~ 2300 2300 1720(4) i n systaa pressure, psig 1 4 *ax-

5. Iow reactor coolant 1900 1900 1900 Bypommed systan pr==ne, psig '

i min.

I

6. Variable low reactor (11.75 Tout-5103) (1) (11.75 'Ibut-5103) (1) (11.75 'Ibut-5103)(1) Bypommed coolant system pressure psig, min.

2

7. Reactor coolant teep. 618.8 618.8 618.8 618.8 4

F., Max. 6 I

8. High Reactor Biilding 4 4 4 4 pressure, psig, max.

(1) Tout is in degues Fahrenheit (F) i (2) Reactor coolant system flow, % li' (3) Adiministratively controlled reduction set only during reactor shutdown.

(4) Automatically set when other segments of the RPS (as specified) are bynaW.

(5) 'Ihe ptmp monitors also produce a trip on: (a) loss of two reactor coolant ptmps in reactor coolant loop, and (b) loss of one or two reactor coolant ptmps during two-punp operation.

(6) Trip setting limits are setting limits on the setpoint side of the protection system

l

.- l l

l l

l 2500 2300 P = 2300 psig as KCEPT.ABLE

}, NERADON T = 618.8 F l 3

2100 ,

/

5/

2 &

. y

$ P = 1900 puis gj e 1900 J

w UNACCEPTABLE I DPERATION W

2 1700 1500 540 560 580 600 620 640 Essetor Outlet Temperature, F DC-1 FROTECHON SYSTZM MAIDC;M ALLOWABLE SET POI:CS Figure 2.3-1

I i

)

Themal Power Level, %

1

-- 120

(-30,108) - 110 (25,108)

ACCEPTABLE ,

g4 PUMP -- 100 g mg = +2.405 gOPERATION m l 2 = -5.277 l

90 l

!(-30,80.6) (25,80.6)I i ACCEPTABLE 80

(-45.8,70) l3 & 4 PUMP -- 70 1 (32.2,70) l OPERATION l 1 I 60

!(-30.53.1) (25.53.1)!

IACCEPTABLE 50 l2,3, & 4 PUMP ' (32.2,42.6)

(-45.8,42.6)  ;

iOPERATION -- 40  ;

I

.i l -- 30 l l I e I -

- 20 l

" (32.2,15.1)

(-45.8,15.1) d oI .I -

i! il 1 -- 10 7I 7 E fl El J i i i i i i i i i ili i i -i i i 70 50 30 10 0 10 20 30 40 50 60 70 80 Reactor Po'wer Imbalance, %

PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS FOR REACTOR POWER IMBALANCE TMI-1 Figure 2.3-2

f. If a control rod in the regulating or axial power shaping groups is declared inoperable per Specification 4.7.1.2., operation may continue provided the rods in the group are positioned such that the rod that was' declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2.

g. If the inoperable rod in Paragraph "e" above is in groups 5, 6, 7, or 8, the other rods in the group may be trimmed to the same -

position. Normal operation of 100 percent of the thermal power allowable for the reactor coolant pump combination may then continue provided that the rod that was declared inoperable is maintained within allowable group average position limits in 3.5.2.5.

3.5.2.3 The worth of single inserted control rods during criticality is limited by the restriction of Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.

3.5.2.4 Quadrant Tilt:

a. Except for physics tests the quadrant. tilt shall not exceed

+4.12% as determined using the full incore detector system. l

b. When the full incore detector system is not available and except for physics tests quadrant tilt shall not exceed +1.96%

as determined using the power range channels displayed on the console for each quadrant (out of core detection system),

c. When neither detector system above is available and, except for physics tests, quadrant tilt shall not exceed +1.907, as determined using the minimum incore detector system.

d. Except for physics tests if quadrant tilt exceeds the tilt limit power shall be reduced 2 percent for each 1 percent tilt in excess of the tilt limit. For less than four pump operation, thermal power shall be reduced 2 percent of the thermal power allowable for the reactor coolant pump combination for each 1 percent tilt ~in excess of the tilt li mi t.

e. Within a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the quadrant power tilt shall be.

reduced to less than the tilt limit except for physics tests, or the following adjustments in setpoints and limits shall be made:

1. The protection system reactor power / imbalance envelope trip setpoints shall be reduced 2 percent in power for

~

each 1 percent tilt, in excess of the tilt limit. l 3-34

2. The control rod group withdrawal limits (Figures 3.5-2A to 3.5-21) shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

3. The operational imbalance limits (Figures 3.5-2J and 3.5-2K) shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

f. Except for physics or diagnostic testing, if quadrant tilt is in excess of +16.80% determined using the full incore detector system (FIT), or +14.2% determined using the out of core detector system (OCT) if the FIT is not available, or +9.5% using the minimum incore detector system (MIT) when neither the FIT nor DCT are available, the reactor will be placed in the hot shutdown condition. Diagnostic testing during power operation with a quadrant tilt is permitted provided that the thermal power allowable is restricted as stated in 3.5.2.4.d above.

9 Quadrant tilt shall be monitored on a. minimum frequency of once every two hours during power operation above 15 percent of rated power.

3-34a 1

l l

3.5.2.5 Control Rod Positions

a. Operating rod group overlap shall not exceed 25 percent +5 percent, between two sequential groups except for physics tests, _

b. Position limits are specified for regulating control rods. Except l for physics tests or exercising control rods, the regulating control rod insertion / withdrawal limits are specified on Figures 3.5-2A, 3.5-2B, and-3.5-2C for four pump operation and Figures 3.5-2D, 3.5-2E, and 3.5-2F for three pump operation. Two pump operation is specified on Figures 3.5-2G, 3.5-2P, and 3.5-2I. If any of these control rod position limits are exceeded, corrective measures shall be taken immediately to achieve an acceptable control rod position. Acceptable control rod positions shall be attained within four hours.

3.5.2.5.c Deleted

d. Core imbalance shall be monitored on a minimum frequency of once every two hours during power operation above 40 percent of rated power. Except for physics tests, corrective measures (reduction of imbalance by APSR movements and/or reduction in reactor power) shall be taken to maintain operation within the envelope defined by Figures 3.5-2J and 3.5-2K. If the imbalance is not within the envelopes defined by Figures 3.5-2J or 3.5-2K at the appropriate time in cycle, corrective measures shall be taken to achieve an acceptable imbalance. If an. acceptable imbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met.

e. Safety rod limits are given in 3.1.3.5.

3.5.2.6 -The control rod drive patch panels shall be locked at all times with limited access to be authorized by the superintendent.

3.5.2.7 A power map shall be taken at intervals not to exceed 30 effective full power days using the incore instrumentation . detection system to verify the power distribution is within the limits shown in Figure 3.5-2L.

s 3-35

Bases The power-imbalance envelope defined in Figures 3.5-2J and 3.5-2K is based on l LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5-2L) such that the maximum clad temperature will. not exceed the Final l Acceptance Criteria (2200*F). Operation outside of the power imbalance envelope alone does not constitute a situation that would cause the Final Acceptance Criteria to be exceeded should a LOCA occur. The power imbalance envelope represents the boundary of operation limited by the Final Acceptance Criteria only if the control rods are at the withdrawal / insertion limits as defined by Figures 3.5-2A, 3.5-28, 3.5-2C, 3.5-20, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, 3.5-21, and if quadrant tilt is at the limit. The effects of the gray APSRs are also included. Additional conservatism is introduced by application of:

a. Nuclear uncertainty factors

b. Thermal calibration uncertainty

c. Fuel densification effects

d. Hot rod manufacturing tolerance factors

e. Postulated fuel rod bow effects

f. Peaking limits based on initial condition for Loss of Coolant Flow transients.

The Rod index versus Allowable Power curves of Figures 3.5-2A, 3.5-2B, 3.5-2C, 3.5-20, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, and 3.5-21 describe three regions.

These three regions are:

1. Permissible operating Region

2. Restricted Regions

3. Prohibited Region (Operation in this region is not allowed)

NOTE: Inadvertent operation within the Restricted Region for a period of four hours is not considered a violation of a 3-35a

The 25+5 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke.

Control rods are arranged in groups or banks defined as follows:

Group Function 1 Safety 2 Safety 3 Safety 4 Safety 5 Regulati ng 6 Regulating 7 Regulating (Xenon transient override) 8 APSR (axial power shaping bank)

Control rod groups are withdrawn in sequence beginning with group 1. Groups 5,6 and 7 are overlapped 25 percent. The normal position at power is for group 7 to be partially inserted.

The rod position limits are based on the most limiting of the following three criteria: ECCS power peaking, shutdown margin, and potential ejected rod worth. As discussed above, compliance with the ECCS power peaking criterion is ensured by the rod position limits. The minimum available rod worth, consistent with the rod position limits, provides for achieving hot shutdown by reactor trip at any time, assuming the highest worth control rod that is withdrawn remains in the full out position (1). The rod position limits also ensure that inserted rod groups will not contain single rod worths greater than: 0.65% Ak/k at rated power. These values have been shown to be safe by the safety analysis (2) of the hypothetical rod ejection accident. A maximum single inserted control rod worth of 1.0% Ak/k is allowed by the rod position limits at hot zero power. A single inserted control rod worth 1.0% Ak/k at beginning of life, hot, zero power would result in a lower transient peak thermal power and, therefore, less severe environmental consequences than 0.65% Ak/k ejected rod worth at rated power.

The plant computer will scan for tilt and imbalance and will satisfy the technical specification requirements. If the computer is out of service, than manual calculation for tilt above 15 percent power and imbalance above 40 percent power must be performed at least every two hours until the computer is returned to service.

The quadrant power tilt limits set forth in Specification 3.5.2.4 have been established within the thermal analysis design base using an actual core tilt of +4.92% which is equivalent to a +4.12% tilt measured with the full incore instrumentation with statistically combined measurement uncertainties included. The maximum allowable quadrant power tilt setpoint of +16.8% tilt measured with the full incore detector system represents a +20% actual core tilt and includes bounding measurement uncertainty allowances.

3-36

During the physics testing program, the high flux trip setpoints are administratively set as follows to assure an additional safety margin is provided:

Test Power Test Setpoint 0 <5%

15 50%

40 50%

50 60%

75 85%

>75 105.1%

l REFERENCES (1 ) FSAR, Section 3.2.2.1.2 (2) FSAR, Section 14.2.2.2 l l

l b

, 3-36a '

7---7 . . - - c , - - -,--~.,-.-,-w-----r'r w - vv*-N' -

N -'*e *** -"* - F' = + - -- ----- -

(300,102)

(76,102) (268,102) -

100 - NOT ALLOWED (264,92) 90 -

RESTRICTED a 80 - (200,80)

E y 70 -

N t 60 va g 50 - (32,50) (125,50)

I g 40 -

PERMISSIBLE 30 -

< (0,23) 20 -

10 -

< (0,5) 0 ' ' ' ' ' ' ' ' ' ' '

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 75 100 a t 1 1 1 Group 7 0 2,5 5,0 7,5 1,00 Group 6 0 25 50 75 100 1 I I t I Group 5 R0D POSITIAN LIMITS FOR 4 PUMP OPExATION FROM 0 to 30+10/-0 EFPD TMI-1 Figure 3.5-2A

,o I

I i

(300,102)

(96,102) .  ;

100 - NOT (272,102) 1 ALLOWED l (264,92) l 90 -

- .(248,80) a 80 ,

i E

RESTRICTED y 70 N

~

3' 60 w.

. 50 - (46,50) (232,50) '

I 2 40 -

~

30 20 PERMISSIBLE (0,15.5) 10 0

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25I 50I 75i 100 i

I Group 7 0 25 50 75 100 e i i i I Group 6 0 25 50 75 100 t i I I I Group 5 R00 POSITION LIMITS FOR 4 PUMP OPERATION FROM 30+10/-0 TO 250+10 EFPD TMI-1 Figure 3.5-2B

n

. . t. -

c .

(300,102)

(171,102) 100 - (272,102) - n NOT

.. 1 ALLOWED (264,92) . s  !

90 -

- .j 80 ,0) -

a E 70 -

$ RESTRICTED  !

$ 60 -

, 50 - (94,50) r (232,50) 40 -

E 30 20 -

PERMISSIBLE.

(42,15) 10 0 i i i ' ' ' ' ' ' ' '

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 5p 7p 1,00 Group 7 0, 25 j0 {5 190 Group 6 0 25 50 75 100 t t i i I Group 5 R00 POSITION LIMITS FOR 4 PUMP OPERATION AFTER 250+10 EFPD TMI-1 Figure 3.5-2C

100 -

90

+> 80 -

NOT 300,77)

E (76,77)

ALLOWED (268,77)(f N 70 - (264,69)

RESTRICTED

% 60 -

(200,60) w c 50 -

I E 40 -

(32,38)

(125,37.5) 30 PERMISSIBLE 20< (0,17.7) 10 -

0 * ' ' ' ' ' ' ' ' ' '

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 75 100 t i I I I Group 7 0 25 50 75 100 a I f f f Group 6 0 25 50 75 100 m I 1 1 B Group 5 R0D POSITION LIMITS FOR 3 PUMP OPERATION FROM 0 TO 30+10/-0 EFPD TMI-1 Figure 3.5-20 ,

1 I

L i

! s

, , r. <

3 s

3 s

100 -

90 -

, 80 -

(96,77) (300,77) g NOT (272,77)

ALLOWED M 70 _

(264,69)

% 60 - 's (248,60)

[ RESTRICTED

. 50 -

E

$ 40 -

(46,38) (232,37.5) 30

\

20 - J 10 ' '

• 0 ' ' ' ' ' ' ' ' '

0 25 50 75 ( 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 7p 1p0 Group 7 0 25, 5,0 75 100

,e i i

( Group 6 0 25 50 75 100 i t  ? I i Group 5- l s

R00 POSITION LIMITS FOR 3 PUMP OPERATION FROM 30+10/-0 TO 250+10 EFPD TMI-1 l

l Figure 3.5-2E s

, , r. -

100 -

90 -

(300,77)  ;

80 -

(171,77) (272,77)-

g NOT m 70 ALLOWED (264,69)

G 60 - (248,60) o 50 -

h -

j 40 (94,38) (232,37.5) 30 -

20 -

PERMISSIBLE 10 - (42,11.7)

< (0,4.2) 0 ' ' ' ' ' ' ' ' ' ' '

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 75 100 e e e t ' I Group 7 0 25 50 75 100 t I i I e Group 6 0 25 50 75 100 t i I f i Group 5 R0D POSITION LIMITS FOR 3 PUMP OPERATION AFTER 250+10 EFPD TMI-1 Figure 3.5-2F

100- -

90 u 80 -

E W 70 -

10 60 (300,52)

• 50 -

NOT (76,52) _

mew ^

y ALLOWED (264,46)

E 40 - RESTRICTED

.,l 200,40) 3 h-(125,25) 20 -

10

( 'I '6) PERMISSIBLE O

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn ,

0 25 50 75 100 Group 7 0 25 50 75 100 t i i 1 e Group 6 0 25 50 75 100 t t I I a Group 5 ROD POSITION LIMITS FOR 2 PUMP OPERATION FROM 0 TO 30+10/-0 EFPD TMI-1 Figure 3.5-2G

100 -

90 -

y 80 y 70 -

N (96,52) (272,52) 50 NOT f

(264,46)

ALLOWED 2 40 -

(248,40)

RESTRICTED 30 -

(46,26) (232,25) 20 -

10 < (0,8.7) PERMISSIBLE i , 0,2,. 5 ) , , , , , , , , ,

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50- 7p 1,00 Group 7

,0 25 50 75 100 Group 6 0 25 50 75 100 t f 1 i i Group 5 l l

R0D POSITION LIMITS FOR 2 PUMP OPERATION FROM 30+10/-0 TO 250+10 EFPD_

1 TMI-1 l

Figure 3.5-2H i

l

' 100 90

~

" 80 E

70

. - 60 (300,52)

• • (171,52) (272,52)-

. 50- -

5. NOT (264,46)

I ALLOWED 2 40 -

(248,40)

RESTRICTED 30 (94,26) (232,25)'

20 -

10 - PERMISSIBLE 42,8.5) 0' (0,3.5 , , , i , , , , , ,

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 27 50 7p 1,00 Group 7 0 25 50 75 100 1 I I I I Group 6 0 25 50 75 100  ;

I I I I I Group 5 l

l l

I R00 POSITION LIMITS FOR 2 PUMP OPERATION AFTER 250+10 EFPD TMI-1 Figure 3.5-21 1

w, - - , , ,, - - - - - , , , ,

?

Eower, % of 2535 MWt RESTRICTED REGION - 110

(-20,102)^ '

-100

(-20.7,92) ' '*

90

(-24,80)< -

- 80 >(32,80)

- 70 PERMISSIBLE

- 60 OPERATING REGION

- - 50

- 40

- 30

- 20 10 I I e 1 I I I l t I 40 20 -10 0 10 20 30 40 50 .

Power Imbalance, %

POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 TO 30+10/-0 EFPD TMI-1 Figure 3.5-2J

Power, % of 2535 MWt RESTRICTED REGION --110

(-20,102) -

(28.5,102) 100

(-22.1,92)<f < (28.5,92)

- 90

(-22.4,80)< > -

- 80 ,(32.8,80)

- 70 PERMISSIBLE 60 OPERATING REGION

-- 50

-- 40 30

- 20

-- 10 g i i I I I I I I I 40 20 -10 0 10 20 30 40 50 Power Imbalance, %

POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 30+10/-0 l EFPD l TMI-1  !

Figure 3.5-2K l

l

, .,, c, .

l l

20 i i i i i i i i i i 18 d

k a -

3 E

/f~,",# _

16

/

E - / s' ~

• /

2 l _

5 14 -

/

E -

l g ,_ _ 0-1000 Wd/mtU ~ '

12 -

-. 1000-2600 Wd/mtU After 2600 Wd/mtU -

i ,

1 1 i i i e i 8

10 12 4 6 8 10 0 2 Axial Location of Peak Power From Bottom of Core, ft LOCA LIMITED MAXIMUM ALLOWABLE LINEAR HEAT RATE TMI-1 Figure 3.5-2L

,v e. s 5.3 REACTOR Applicability Applies to the design features of the reactor core and reactor coolant system.

Objective To define the significant design features of the reactor core and reactor coolant system.

Specification 5.3.1 REACTOR CORE 5.3.1.1 The reactor core contains approximately'93.1 metric tons of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods. The reactor core is made up of 177 rods.t ft/g(aqsemblies.

2i Each fuel assembly contains 208 fuel 5.3.1.2 The reactor core shall approximate a right circular cylinder with an equivalgnt diameter of 128.9 inches and an active height of 142.25 l i nches. L 2 3 5.3.1.3 The average initial enrichment of the current core for Unit 1 is a nominal 2.86 weight percent of U235 The highest enrichment is less than 3.5 weight percent U235, 5.3.1.4 There are 61 full-length control rod assemblies (CRA) and 8 axial power shaping rod assemblies (APSRA) distributed in the reactor core as shown in FSAR Figure 3.2-1. The full-length CRA contain a 134 inch igngth of silver-indium-cadmium alloy clad with stainless steel.(3) The gray APSRA contain a 63 inch length of Inconel.

5.3.1.5 The core will have 68 burnable poison spider assemblies with similar l dimensions as the full-length control rods. The cladding will be zircaloy-4 filled with alumina-boron. l 5.3.1.6 Reload fuel assemblies and rods shall conform to design and evaluation describpd in FSAR and shall not exceed an enrichment of 3.5 percent of U23o, 5.3.2 REACTOR COOLANT SYSTEM 5.3.2.1 The reactor coolant system shall be designed and constructed in accordance with code requirements. (4) 5.3.2.2 The reactor coolant system and any connected auxiliary systems exposed to the reactor coolant conditions of temperature and pressure, shall be designed for a pressure of 2,500 psig and a temperature of 650 F. The pressurizer and prqsqurizer surge line shall be designed for a temperature of 670 F.DJ 5-4

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