ML20215N891

TMI-1 Cycle 6 Reload Rept

ML20215N891
Person / Time
Site:	Three Mile Island
Issue date:	10/31/1986
From:	BABCOCK & WILCOX CO.
To:
Shared Package
ML20215N870	List: ML20215N869 ML20215N877 ML20215N886 ML20215N891
References
BAW-1977, NUDOCS 8611100036
Download: ML20215N891 (72)

Text

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I I BAW-1977 October 1986 I

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INREE MIIE ISIAND UNIT 1 CYCLE 6 RELOAD REPORE I

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E COCK &WijCOX 8611100036 861103 , a McDermott company ll PDR ADOCK 0500028

,E P PDR

BAW-1977 October 1986 I 'IHREE MILE ISIAND UNIT 1 CYCLE 6 RELOAD 1<.t.a.m-I I

BAECDCK & WIILOX Nuclear Power Division P. O. Box 10935 Lynchburg, Virginia 24506-0935 Babcock & Wilcox a McDermott company

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e Page 1.0 DTIRODUCITQi . ............... . . . . . ..... 1-1 E

L 2.0 OPERATD G HISIORY ..... . ...~..... . . . . ..... 2-1 c 3.0 GENERAL DESCRIPTIQi .... ..... . . . . . . . . ..... 3-1 l

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4.0 FUEL SYSTEM DESIGT . .... ............ . ..... 4-1 F 4.1. Fuel Asserbly Mechanical Design . . . . . . . . ..... 4-1 L 4.2. ruel Rod arxi Gray APSR Designs . . . . . . . . . . . . . . 4-1 4.2.1. Cla Ming Collapse ...... . . . . . ..... 4-1 7

4.2.2. Cla M ing Stress ..... . . . . . . . ..... 4-2 L 4.2.3. Cladding Strain ..... . . . . . . . ..... 4-2 4.3. 'Ihermal Design . . . . . . . . . . . . . . . . . . . . . . 4-3 4.4. Material Design .. ....... . . . . . . . ..... 4-3 Operatirq Experience . . . . . . . . . . . . . . . . . . . 4-4

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5.0 NUCLEAR DESIGT . .. .... . . . . . . . . . . . . . ..... 5-1 5.1. mysics Characteristics ... . . . . . . . . . ..... 5-1 5.2. Ane1ytice1 Ingue . . . . . . . . . . . . . . . . . . . . . 5-2 5.3. Changes in Nuclear Design .... . . . . . . . ..... 5-2 6.0 THERMAIr-HYDRAULIC DESIOGi . ........ . . . . . ..... 6-1 7.0 ACCIDENT AND TRANSIEtTP ANALYSIS 7-1

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7.1. General Safety Analysis .... . . . . . . . . ..... 7-1 7.2. Accident Evaluation . .... . . . . . . . . . ..... 7-2 8.0 PROPOSED FDDIFICATIONS 'IO TE0iNICAL SPECIFICATIONS . . ..... 8-1

[ 9.0 STARIUP PROGRAM - PHYSICS TESTDG . . . . . . . . . . ..... 9-1 9.1. Precritical Tests . . ... . . . . . . . . . . ..... 9-1 9.1.1. Control Rod Trip Test 9-1

[ 9.1.2. RC Flow ........ . . . . . . . . .....

9-1 9.2. Zero Power mysics 'Ibsts . . . . . . . . . . . . . . . . . 9-1 9.2.1. Critical Boron Concentration . . . . . . . . . . . 9-1 9.2.2. 'Ibnperature Reactivity Coefficient . . . ..... 9-2 9.2.3. Control Rod Group / Boron Reactivity Worth . . . . . 9-2 9.3. Ibwer Escalation Tests . . . . . . . . . . . . . ..... 9-3 9.3.1. Core Power Distribution Verification at Intomadiate Power level (IPL) and 100% FP With Noninal Control Rod Position . . . ..... 9-3

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I CONIDTIS (Cont'd)

Page 9.3.2. Incore Vs. Excore Detector Imbalance Correlation Verification at the IPL ....... 9-4 9.3.3. Temperature Reactivity Cbefficient at 100% FP . . . . . . . . . . . . . . . . . . . . 9-5 9.3.4. Power Doppler Reactivity Coefficient at 100% FP . . . ................. 9-5 9.4. Procedure for Use if Acceptance Criteria Not Met . . . . . 9-5

10.0 REFERENCES

. . . . . . . ............... . . . . . 10-1 I

List of Tables 4-1. Fuel Design Parameters and Dimensions . . . . . . . . . . . . . 4-5 5-1. M-1 Cycle 6 Ihysics Parameters .......... . . . . . 5-3 5-2. Shutdown Margin Calculation for M -1 Cycle 6 . . ....... 5-5 6-1. Maximum Design Conditions, Cycles 5 and 6 . . ......... 6-3 7-1. Comparison of Key Parameters for Accident Analysis ...... 7-4 7-2. Bounding Values for Allowable IOCA Peak Linear Heat Rates for M -1 ..................... 7-4 8-1. Reactor PIctection System Trip Setting Limits . . . . . . . . . 8-11 List of Ficures Figure I

3-1. Core Irsading Diagram for 'IMI-1, Cycle 6 . . . . . . . . . . . . 3-3 3-2. Enrichment and Burnup Distribution for 'IMI-1, Cycle 6 . . . . . 3-4 3-3. Control Rod Iccations and Group Designations for

'IMI-1, Cycle 6 ........................ 3-5 g 3-4. IEP Enrichment and Distribution for M-1 Cycle 6 . . . . . . . 3-6 3 4-1. Gray Axial Power Shaping Rod ............ . . . . . 4-6 5-1. BOC (4 EFPD), Cycle 6 Two-Dimensional Relative Power Distribution - Full Power, Equilibrium Xenon, APSPs Inserted ... ........................ 5-6 8-1. 'IMI-1 Core Protection Safety Limits ............. 8-5 8-2. Core Protection Safety Bases ...... ........... 8-6 8-3. PIctection System Maximum Allowable Setpoints . . . . . . . . . 8-12 8-4. Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance ................ 8-13 Babcock & WHcox iv a McDermott company

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r List of Ficures (Cont'd)

L Figure Page

( 8-5. Rod Position T.imits for Four-Pump Operation Fran 0 to 30 +10/-0 EFPD, 'IMI-1 . . . . . . . . . . . . . . ..... 8-19 8-6. Rod Position T.imits for Four-Pump Operation Fran L 30+10/-0 to 25G10 EFPD, 'IMI-1 . . . . . . . . . . . ..... 8-20 8-7. Rod Position T.imits for Four-Pump Operation After 250110 EFPD, 'IMI-1 . . . . . . . . . . . . .. . . . ..... 8-21 8-8. Rod Position T.imits for 'Ihree-Pump Operation From u O to 30+10/-0 EFPD, 'IMI-1 . . . . . . . . . . . . .. ..... 8-22 8-9. Rod Position T.imits for 'Ihree-Pump Operation Fran 3O+10/-0 to 250t10 EFPD, 'IMI-1 . . . . . . . . . . . ..... 8-23 8-10. Rod Position Limits for 'Ihree-Pump Operation After

{ 250i10 EFPD, 'IMI-1 . . . . . . . . . . . . . . . . . ..... 8-24 8-11. Rod Position Limits for Two-Pump Operation Fran p 0 to 30&l0/-0 EFPD, 'IMI-1. . . . . . . . . . . . . . ..... 8-25 L 8-12. Rod Position T.imits for 'Dw>-Pump Operation Fran 30+10/-0 to 250t10 EFPD, 'IMI-1 . . . . . . . . . . . ..... 8-26 8-13. Rod Position Limits for Two-Pump Operation After I . 250i10 EFPD, 'IMI-1 . . . . . . . . . . . . . . . . . ..... 8-27 L

8-14. Power Imbalance Envelope for Operation From 0 to 3O+10/-0 EFPD . . . . . . . . . . . . . . . ..... 8-28 L 8-15. Power Imbalance Envelope for Operation After 30F10/-0 EFPD . . . . . . . . . . . . . . . . . ..... 8-29 8-16. IOCA Limited Maximum Allowable Linear Heat Rate, 'IMI-1 . . . . 8-30 L

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1. INIRODUCITON AND

SUMMARY

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This report justifies the operation of the Three Mile Isihrxi Nuclear Station Unit 1 (cycle 6) at a rated core power of 2535 MWt. Included are the required analyses as outlined in the USNRC document, " Guidance for Proposed License AnnuAents Relating to Refueling," June 1975.

To support cycle 6 operation of the TMI-1 nuclear station, this report employs analytical I techniques and design bases established in reports that have been submitted and received technical approval by the USNRC (see references).

I Cycle 6 reactor parameters that are related to power capability are summarized in this report. All the accidents analyzed in the FSAR1 have been reviewed for cycle 6 operation. The cycle 6 characteristics proved to be conservative with respect to all applicable safety limits and criteria.

The Technical Specifications have been reviewed, and the modifications required for cycle 6 operation are justified in this report. Based on the analyses performed, which take into account the postulated effects of fuel densification and the Final Acceptance Criteria for emergency core cooling systems (ECCS), it has been concluded that TMI-1, cycle 6 can be safely operated at the rated core power level of 2535 MWt.

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2. OPERATING HISIORY

, 'Ihe reference cycle for the nuclear and thermal-hydraulic analyses of the

'Ihree Mile Island, Unit 1 plant is the operating cycle 5. Cycle 5 achieved criticality on October 3, 1985, and after zero power testing began power operation on October 6, 1985. Cycle 5 is scheduled for completion in i November 1986 after 290 i 15 EFPD. No operating anomalies have occurred 1 duriry the fifth cycle. 'Ihe operation of cycle 6 is scheduled to begin in March 1987. 'Ihe design cycle length is 425 i 15 EFPD.

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3. GENERAL DESCRIPI' ION 2e 'IMI-1 reactor core is described in detail in section 3 of the Final Safety Analysis Report for the Unit. 'Ihe cycle 6 core consists of 177 fuel assemblies (FAs), each of which is a 15 by 15 array containirg 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube. 'Ihe I fuel is comprised of dished-end, cylindrical pellets of' uranium dioxide clad in cold-worked Zircalcy-4. All fuel annamblies in cycle 6 have a constant nominal fuel loading of 463.6 kg of uranium. 'Ihe undensifie:1 nominal active fuel lengths, theoretical densities, fuel and fuel rod dimensions, arri other related fuel parameters are given in Table 4-1 for all fuel assemblies.

Figure 3-1 is the fuel shuffle diagram for 'IMI-1, cycle 6. 'Ihe initial enrichment of batches 6, 7, and 8B is 2.85 wt% U-235; batch 8A has an enrichment of 2.95 wt% U-235. All batch 4 and 5, and three batch 6 assemblies will be discharged at the end of cycle 5. 'Ihe remaining 49 batch 6, and all batch 7 assemblies will be shuffled to new locations. The 76 fresh batch 8 annamblies will be loaded in a symmetric checkerboard pattern throughout the core. Figure 3-2 is an eighth-core map showing the assembly burnup and enrichment distribution at the beginning of cycle 6.

Cycle 6 will be operated in a feed-arrl-bleed mode. Reactivity is con-trolled by 61 full-lergth Ag-In-Od control rods, 68 burnable poison rod assemblies (BPRAs), and soluble boron shim. In addition to the full-length control rods, eight Inconel axial power shaping rods (gray APSRs) are provided for additional control of the axial power distribution. 'Ihe cycle I 6 locations of the 69 control rods and the group designations are indicated in Figure 3-3. 'Ihe cycle 6 locations and enrichments of the BPRAs are shown in Figure 3-4.

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3-1 I Babcock & Wilcox a McDermott company

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'Ihe ncminal system pressure is 2200 psia and the core average ncminal heat rate is 5.65 kW/ft at the rated core power of 2535 NWt.

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I Figure 3-1. Core Loading Diagram for TMI-1, Cycle 6 7 7 6 7 7 "A NO3 LO2 008 Ll4 N13

'M 7 8B 7 88 6 8B 7 88 7

'o' "25 '25 I' ' ko2 ' ct' ' '

7 8B 7 8B 7 8B 7 88 7 EB 7 C H15 F NO2 F M02 F M14 F N14 F R08 7 8B 6 88 6 8B 6 88 6 8B 6 8B 7 D E07 F C09 F E09 F G13 F A06 A10 F C07 F 88 7 8B 6 8A 6 88 6 SA 6 8B 7 8B E B04 F F B12 F H06 F D06 F DIO F F08 F IF 7 C12 7

A09 88 F

6 GOS 8A F

6 H02 8B F

6 G09 8B F B08 6 8A F

6 Gil 8B F

7 A07 7

C04 7 88 7 88 6 8B 6 8A 6 88 6 8B 7 88 7 5G B10 F B11 F F04 F F06 F F10 F F12 F B05 88 F

6 B06

, 6 6 88 6 88 6 8A 6 8A 6 88 6 6 H

H04 E03 F G03 F C07 F 011 F K09 F K13 F M13 Ht2 i

7 8B 7 8B 6 8B 6 8A 6 8B 6 8B 7 88 7 K

P10 F Pil F LO4 F LO6 F L10 F L12 F P05 F P06 l

!W 7 7 8B 6 8A 6 8B 6 88 6 8A 6 8B 7 7 L 012 R09 F K05 F P08 F K07 F H14 F Kil F R07 004 88 7 88 6 8A 6 88 6 8A 6 8B 7 8B M F P12 F LO8 F N06 F N10 F H10 F PO4 F 6 88 6 8B 6 88 6 BB 6 8B 7 lg l N 7 RIO BB F K03 F M07 F 007 F M09 F 009 F R06 l 7 88 7 88 7 8B 7 8B 7 88 7 0 A08 F 002 F E02 F E14 F 014 F H01 P 7 8B 7 88 6 8B 7 8B 7 F01 F G01 F 005 F GIS F FIS l

7 7 6 7 7 R F02 N08 F14 013 D03 I .

10 11 12 13 14 13 4 5 6 7 8 9 1 2 3

- Batch

= Previous Cycle Location; F = Fresh Fuel Assembly i

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I Figure 3-2. Enrichment and 8urnup Dis'tribution for E TMI-1, Cycle 6 E I

8 9 10 11 12 13 14 15 H

2.85 19,779 2.95 0

2.85 18,934 2.85 0

2.85 17,249 2.85 0

2.85 19,745 2.85 20,033 l

2.85 2.85 2.85 2.85 2.85 2.85 2.85 I

19,167 0 15,739 0 9,411

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0 10,550 l

2.85 2.95 2.85 2.85 2.85 2.85 19,155 0 16,987 0 7,596 10,330 2.85 2.85 2.85 2.85 M

17,883 0 6,772 0 2.85- 2.85 2.85 N

17,249 0 6,513 2.85 0

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X.XX Initial enrichment XX,XXX BOC burnup, mwd /mtU E*U"** **

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h Figure 3-3. Control Rod Locations and Group

! Designations for TM1-1, Cycle 6 X

l Fuel Transfer l Canal )

A B 1 6 1 C 3 5 5 3 D 7 8 7 8 7 E 3 5 4 4 5 3 F 1 8 6 2 6 8 1 G 5 4 2 2 4 5 H W- 6 7 2 4 2 7 6

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K 5 4 2 2 4 5 L 1 8 6 2 6 8 1 ft i 3 5 4 4 5 3 N l 7 8 7 8 7 0 3 5 5 3 l

P l l 1 6 1 R I i Z

1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 X Group Number Grouo No. of Rods Function 1 8 Safety 2 8 Safety 3 8 Safety I 4 6

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12 8

Safety Control Control

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Control APSRs i Babcock & Wilcox 3-5 a ucoermore company

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Figure 3-4. LBP Enrichment and Distribution for TMI-1 Cycle 6 8 9 10 11 12 13 14 15 I

H 1.03 1.03 0.80 I

K 1.03 1.03 1.03 0.20 L 1.03 1.03 0.80 M 1.03 1.03 0.80 I

I N 1.03 0.80 0.20 0.80 0.80 0.20 I

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BPRAs wt % B 4C X.XX LBP Concentration (wt % B4 C in A1230) '

20 0.80 16 0.20 Total if6 Babcock & Wilcox 3-6 a Mcoermoir company g

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I I 4. EUEL SYSTEM DESIGT 4.1. Ebel Assembly Mechanical Desian

'Ihe types of fuel nnnomblies and pertinent fuel design parameters for 'IMI-1, cycle 6 are listed in Table 4-1. All fuel assemblies are identical in concept and are mechanically interchangeable. Retainer annamblies will be used on the two fuel assemblies containing the regenerative neutron sources (RNS) and the 68 assemblies containing BPRAs. 'Ihe justification for the I design and use of the retainers described in references 2 aM 3 are applicable to the RNS retainers in cycle 6 of 'IMI-1. All results, references, and identified conservatisms presented in section 4.1 of the cycle 5 reload report 4are applicable to the cycle 6 reload core.

4.2. Fbel Rod and Grav APSR Desians I 'Ihe mechanical evaluations of the fuel rod and gray APSR are MW sned below.

4.2.1. Claddirn Collanse Creep collapse analyses were performed for the 4 different fuel batch assembly power histories. Because of its longer previous incere exposure time, the batch 6 fuel is more limiting than the other batches. 'Ihe batch 6 assembly power histories were analyzed and the most limiting assembly was determined.

'Ihe power history for the most limiting assembly was used to compare with a conservative generic creep collapse analysis. 'Ihe collapse time for the I most limiting asscably was conservatively deternined to be more than 35,000 EFTH (effective full power hours), which is greater than the maximum I projected residence time (Table 4-1) . 'Ihe creep collapse analysis was performed based on the m nditions set forth in reference 5.

'Ihe gray APSRs that are to be used in cycle 6 were designed to improve creep life. Cladding thickness aM rod ovality control, which are the prirary 4-1 Babcock & Wilcox I a McDermott company

I factors controlling the creep life of a stainless steel material, have been improved to extend the creep life of the gray APSR. We minimum design claMing thickness of the Mark B APSR is 18 mils, while that of the gray APSR is 24 mils. Additionally, the gap between the end plug and the Inconel absorber material was rwhM. Finally, the ovality in the gap area will also be controlled to tighter tolerances. 'Ihe gray APSR.is shown in Figure 4-1.

4.2.2. Claddina Stress We 'IMI-1 stress parameters are enveloped by a conservative fuel rod stress analysis. For design evaluation, the primary membrane stress must be less than two-thirds of the minimum specified unirradiated yield strength, and all stresses (primary and secondary) must be less than the minimum specified unirradiated yield sL.w yth. In all. cases, the margin is in excess of 30%.

'Ihe following conservatisms with respect to 'IMI-1 fuel were used in the analysis:

1. Im post-densification intamal pressure.

2. Im initial pellet density.

3. High system pressure.

4. High thermal gradient across the cladding.

'Ihe stresses reported in reference 6 for core 1 fuel represent conservative values with respect to the cycle 6 core.

We gray APSR design was analyzed to demonstrate that it meets specified design requirements. 'Ihe APSR was analyzed for cladding stress due to pressure, torperature, and ovality. It was fourd that the gray APSR has sufficient cladding and weld stress margins.

4.2.3. Claddina Strain

'Ihe fuel design criteria specify a limit of 1.0% on cladding circumferential plastic strain. We pellet design is established for plastic cladding strain of less than 1% at values of maximum design local pellet burnup and heat generation rate, which are considerably higher than the values the 'IMI-1 fuel is expected to see. We strain analysis is also based on the maximum specification value for the fuel pellet diameter and density and the lowest pemitted tolerance for the cladding ID.

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I I he gray APSR was analyzed for claMig strain due to thnmal and irradiation swelling. We results of this analysis showed that no claMirq strain is induced due to themal expansion or irradiation swellirg of the Inconel abrher.

4.3. Wemal Desian All fuel in the cycle 6 core is themally similar. All themal design I analyses for cycle 6 fuel used the TACD2 code, as described in reference 7 for fuel taperaturn and fuel rod internal pressure preciictions.

We results of the themal design evaluation of the cycle 6 core are summarized in Table 4-1. Cycle 6 core protection limits were based on a linear heat rate (IHR) to centerline fuel melt of 20.5 kW/ft as detnmined by the TACD2 code. %e maximum fuel assembly burnup at BX: 6 is predicted to be less than 33,700 mwd /mtU. We fuel red intennl pressures have been evaluated with TACD2 for the highest burnup fuel rods and are predicted to be less than the ncminal reactor coolant pressure of 2200 psia.

In the cycle 5 reload report,4 the batch 6 and batch 7 pellet diameter, stack height, and ncminal linear heat rate were provided in Table 4-2 based on the assumption of instantaneous fuel densification.- We TACD2 code utilizes a time deperxlent fuel densification model. With the implementation of the TACD2 code for cycle 6 evaluations, the Table 4-1 parameters are provided based on nominal design dimensions.

4.4. Material Desian he chemical compatibility of all possible fuelmlaMig-coolant-assembly interactions for batch 8 fuel asserblies is identical to that of the i previous fuel batches.

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I 4.5. Operatirn Experience NWk & Wilcox operating experience with the Mark B 15x15 fuel ambly has verified the adequacy of its design. As of April 30, 1986, the following experience has been accumulated for eight B&W 177 fuel assembly g plants using the Mark B fuel a-bly: E Cumulative Current Max FA burnup, mwd /mtU(a) electric Reactor Cycle Incore Discharned output.MWh(b)

Oconee 1 10 33,710 50,598 62,028,968 Oconee 2 8 38,100 37,326 55,785,115 Oconee 3 9 37,714 39,229 55,385,714 Three Mile Island 5 28,440 32,400 25,105,483 Arkansas Nuclear 7 41,960 36,820 48,299,124 One, Unit Rancho Seco 7 26,100 38,268 39,078,111 Crystal River 3 6 24,970 31,420 35,863,252 Davis-Desse 5 31,020 32,790 25,233,177 I

(a)As of April 30, 1986.

(D)As of January 31, 1986.

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M M M M M M M M M M M' M M M M Y M M M Table 4-1. Ebel Desian Parameters and Dimensions Twice- Once Fresh Fresh burned assys burned assys fuel assys fuel assys Batch 6 Batdi 7 Batch 8A Batch 8B I

i Ebel assembly Mark B4 Mark M Mark M Mark M type No. of acelies 49 52 12 64 l Ebel rod OD, in. 0.430 0.430 .430 .430 Ebel rod ID, in. 0.377 0.377 0.377 0.377 Undensified active 142.25 142.25 141.8 142.25 fuel length, in.

Ebel pellet OD 0.3695 0.3695 0.3686 0.3695 i (mean specified),

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Ebel pellet initial 94.0 94.0 95.0 94.0 density, % 'ID Initial fuel 2.85 2.85 2.95 2.85 enrichment st%

235U Burnup (DOC) mwd /mtU 17,900 8,500 0 0 Claddity collapse >35,000 >35,000 >35,000 >35,000 time, EFIU

,l M imated residence 23,700 17,000 10,100 10,100 time, EFTU (max) tg S$ Ncs;inal timar Heat 5.7', . 5.73 5.74 5.73 5 L' Rate at 2568 FWt, 3O kW/ft 85 y= Average Ebel Temperature ag at nmtinal UIR, F (BOL) 1399 1399 1100 1399

" Minimum UIR to melt, 20.5 20.5 20.5 20.5 I

W/ft

m m

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I NIX 2 EAR DESIGN 5.1. Physics Garacteristics Table 5-1 lists the core physics parameters of design cycles 5 and 6. The values for cycle 5 were generated using PDQ07 8 and the values for cycle 6 were mloHated with the NOODId code. Figure 5-1 illustrates a representative relative power distribution calculated by PDQ07 for the beginniry of cycle 6 at full power with equilibrium xenon and nominal rod positions.

We initial BPRA loadirg, longer design life, and revised shuffle pattern create differences in the physics parameters of cycles 5 and 6. The BOC I critical boran u.u.v_ailcations for cycle 6 are higher hamum the additional E

""^"'i"i'Y '" ~ "7 ' " th* "**" "Y * * * " ' '"" "'" Y *** "Y ***

5 burnable poison. ^

The c.wiuul red worths differ between cycles due to charges in radial flux and burnup distributions. This also accounts for the smaller ejected red sorths in cycle 6 cmpared to cycle 5 values, calculated ejected red vorths and their adherence to criteria are considered at all times in life and at all power levels in the development of the red position limits presented in section 8. All safety criteria associated with the control red worths are met. The adequacy of the shutdown margin with cycle 6 stuck rod worths is deum'uated in Table 5-2. The following -

conservatisms were applied for the shutdown calculations:

1. Poison material depletion allorance.

2. 10% ur. h inty on net rod worth.

3. Flux redistribution penalty.

Flux redistribution was accounted for since the shutdown analysis was calculated using a two-dimensional model. The reference fuel cycle shutdcun margin is presented in the TMI-1 cycle 5 reload report.4 I The differences in the paraneters between cycles 5 and 6 are mainly due to the larger cycle 6, and the presence of burnable poison.

5-1 I Babcock & Wilcox-a McDermott company

I 5.2. Analvtical Irnit h cycle 6 incore measurement calculation constants to be used for cmputirq core power distributions were prepared using the same approved methods as for the reference cycle.

5.3. Chames in Nuclear Desian Core design charges for cycle 6 include the increase in cycle lifetime to 425 EFFD with the ira.m.@ ration of BPRAs to aid in reactivity contwl, and the use of gray APSRs. Gray APSRs, which are longer and use a weaker absorber (Inconel) , replace the silver-indium-cadmium APSRs used in all previous cycles. Calculations with the standard threhional model verified that these APSRs provide adequate axial power distribution control.

As stated in section 5.1, the IOODE code was used to calculate the physics parameters for cycle 6. h ICODE mvialNg of the tm hcnogenized 3 fuel assembly is the same as that used in PDQ07. However, the analytical expression ICODE uses for the spatial flux solution provides more accurate results than the finite difference expression used in PDQ07 when there are few flux solution points per sWly. Reference 9 illustrates the calculational accuracy attainable with NOODLE in cccparison to measured results for various physics parameters. PDQ07 results are compared to measured data in referen 10. 'Ihese cmparisons show NOODE to be as accurate as PDQ07.

h calculational methods used to obtain the irportant nuclear design parameters for this cycle were the same as those used for the reference cycle. h operating limits (Technical Specifications changes) for the reload cycle are given in section 8.

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Table 5-1. 'IMI-1 Cycle 6 Ihysics Parameters (a) l Cvele 5(b) Cycle 6(c)

Cycle length, EFPD 280 425 Cycle burnup, MNd/mtU 8,650 13,129 Average core burnup - EOC, mwd /mtU 19,162 20,588 Initial core loading, mtU 82.1 82.1 i critical boron - BOC, ppm (no Xe)

HZP(d), group 8 inserted 1255 1418

'g HFP, group 8 inserted 1064 1253 5 Critical boron - EOC, ppm (eq Xe)

HZP, group 8 inserted 311 237 HFP, group 8 inserted 29 0 Control rod worths - HFP, BOC, % ak/k Group 7 1.44 1.07 Group 8 (maximum) 0.45 0.21 l Control rod worths - HFP, EOC, % ak/k Group 7 1.56 1.10 lg Group 8 (maximum) 0.50 0.22 EB Max ejected rod worth - HZP, % ak/k(e)

IOC 0.65 0.47 EOC 0.71 0.32 Max stuck rod worth - HZP, % Ak/k BOC 2.24 2.31 EOC 2.03 2.17 Power deficit, HZP to HFP, % Ak/k BOC -1.34 -1.39 lI EOC '

-2.06 -2.23 Doppler coeff - BOC,10-5 (3 pop) 100% power (no Xe) -1.47 -1.49 I Doppler coeff - BOC,10-5 (gg) 100% power (eq Xe) -1.58 -1.84 Moderator coeff - HFP,10-4 (AWOF)

BOC (no Xe, 1253 ppm, group 8 ins) -0.77 -0.40 EOC (eq Xe, O ppm, group 8 ins) -2.63 -2.64 I 5-3 Babcock & Wilcox a McDermott company

I Table 5-1. (Cont'd) cvele 5(b) Cvele 6(c)

Baron worth - HFP, pp:V% Ak/k l

BOC 105 116 l

EOC 93 102 Xenon worth - HFP, % Ak/k BOC (4 EFPD) 2.64 2.60 EOC (equil.) 2.73 2.73 Effective delayed neutron fraction - HFP BOC 0.00583 0.00644 EOC 0.00517 0.00524 I

(a) Cycle 6 data are for the conditions stated in this report; the cycle 5 core corditions are identified in reference 4.

(b) Based on 277 EFPD at 2535 Nt, cycle 4.

(c) Based on 290 EFPD at 2535 Nt, cycle 5. 3 (d)HZP denotes hot zero power (532F Taq); HFP denotes hot full 5 (e) Ejected power (579F rod T b) . for groups 5 through 8 inserted.

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I I Table 5-2. Shutdown Mamin Calculation for 'It4I-1 Cvele 6 BOC. %Ak/k EOC. %Ak/k(a)

Available Rod Worth Total rod worth, HZP(D) 9.47 9.79 Worth reduction due to burnup of poison material -0.42 -0.42 I Maxhnum stuck rod, HZP -2.31 -2.17 Net worth 6.74 7.20 Isss 10% uncertainty -0.67 -0.72 Total available worth 6.07 6.48 Reauired Rod Worth Power deficit, HFP to HZP 1.39 2.23 I Max allowable inserted rod worth Flux redistribution M 0.33 0.44 1.18 Total required worth 2.24 3.85 Shutdown Mamin Total available minus total required 3.83 2.63 LKZIE: Required shutdown margin is 1.00% ak/k.

(a)425 EFPD.

(b)HZP denoted hot zero power (532F Tavg); HFP denoted hot full power (579F Tavg)*

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I Figure 5-1. B0C (4 EFPD), Cycle 6 Two-Dimensional Relativ,e '

Power Distribution -- Ful1 Power, Equilibrium Xenon, APSRs Inserted

• 8 9 10 11 12 13 14 15 I

H 0,94 1.08 0.99 1.14 1.11 1.21 0.84 0.41 I

0.96 1.10 1.10 1.22 1.26 1.05 0.50 I

X L 1.02 1.20 1.11 X 1.25 1.01 0.43 M 1.10 1.24 1.26' O.94 I

I N 1.06 1.03 0.56 l

0 0.60 I

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a I I

• Calculated results from two-dimensional pin.by-pin PDQ07.

Inserted rod group no.

I X.XX Relative power density Babcock & Wilcox I

5-6 a ucoermoir company  ;

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I I 6. 'IHERMAIcHYEERAULIC DESIGN

'Ihe fresh batch 8A ard batch 8B fuel is hydraulically ard gecnetrically similar to the previously irradiated batches 6 ard 7 fuel. 'Ihe thamal-hydraulic design evaluation supporting cycle 6 operation marks the first implementation for 'IMI-1 of crossflow m,daling with the LYNX <,nax (referen s 11, 12, and 13) for DiB predictions. 'Ihermal-Hydraulic I crossflow applications for reload cores are described in BAW-1829.14

'Ihe thermal-hydraulic design bases for cycles 5 and 6 are sumarized in I Table 6-1. Differences between the two cycles reflect the incorporation of the LYNX mam, a reduction in the core bypass flw fraction, a revised I design axial power shape, and the ese of the nominal undensified fuel length. 'Ihe LYNX code introduction and use of the urdensified lergth are both di m W in BAW-1829. 'Ihe design axial flux shape was increased frm 1.50 to 1.65 which increases the available operating margin for most power distributions. 'Ihe reduction in core bypass flow fraction is a result of the introduction of burnable poison rod asenblies into the core for cycle I 6, which rarhv'm the core bypass flow fraction frm 0.104 in cycle 5 to 0.076 in cycle 6. 'Ihe cycle 6 thermal-hydraulic model conservatively a m mad a bypass flow fraction of 0.084.

I 'Ihe reactor protection system (RPS) setpoints for the DiB-based variable low pressure and flux / flow trip setpoints will remain the same for cycle 6. I21B inprovement gained with crossflow modeling has provided increased margin to these setpoints.

Previous fuel cycle evaluations included the calculation of a red bcw penalty for each batch based on the highest fuel burnup in that batdl. A rod bow topical report,15 which addresses the mechanisms ard resulting conditions of rod bow, has been submitted to ard approved by the NRC. 'Ihe topical report concludes that rod bow penalty is insignificant ard is offset l

by the reduction in oower prcduction capability of the fuel assemblies with lI 6-1 Babcock & WHcox i

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I I l irracliation. 'Iherefore, no departure frm nucleate boiling ratio ([NBR) reduction due to rod bow is required for cycle 6.

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I Table 6-1. Maximum Desian Conditions. Cveles 5 and 6 I Cvele 5 Cycle 6 Design power level, MWt 2568 2568 System pressure, psia 2200 2200 Reactor coolant fl w , % design 106.5 106.5 Core bypass flw fraction 0.104 0.084 Vessel inlet / outlet coolant temp at 100% power, OF 555.6/602.4 555.6/602.4 l n m modeling Closed-channel Crossflw Reference design radial-local power peaking factor 1.71 1.71 -

Reference design axial flux shape 1.5 cosine 1.65 cosine Hot channel factors Enthalpy rise 1.011 1.011 Heat flux 1.014 1.014 Flw area 0.98 0.98 Active fuel length, in. 140.5(a) 141.8 I. Avg heat flux at 100% power,-

103 Btu /h-ft2 175(a) 174 Max heat flux at 100% power, 103 Btu /h-ft2 450(a) 492 CHF correlation B&W-2 B&W-2 Minimum D E at 112% power 1.98 (a) 2.08 i

(a) Based on densified length.

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I I 7. ACrTmE AND 'IPANSIDE ANAIXSIS 7.1. General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in cycle 6 parameters to determine the effect of the cycle 6 reload and to ensure that thermal performance during hypothetical transients is not degraded.

The effects of fuel densification on the FSAR accident results have been evaluated and are reported in reference 6. Since batch 8 ' reload fuel assemblies contain fuel rods whose theoretical density is higher than those considered in the reference 6 report, the conclusions in that reference are still valid.

'Ihe radiological dose consequences of the accidents presented in Qiapter 14 of the FSAR were re-evaluated for this reload report. 'Ihe reason for the I re-evaluation is that, even though the FSAR dose analyses used a conservative basis, charges in fuel management techniques have increased the amount of energy prrr W by fissioning plutonium. Since plutonium-239 has different fission yields than uranium-235, the nixture of fission product nuclides in the core changes slightly as the plutonium-239 to uranium-235 fission ratio changes, i.e., plutonium fissions produce more of same nuclides and less of other nuclides. To evaluate the effect of this change-in the cycle 6 fission product O mm ition, the radiological consequences of the Chapter _14 accidents were recalculated. 'Ihe bases used in the dose calculation were identical to those presented in the FSAR except for the following two differences:

1. 'Ihe fission yields and half-lives used in the cycle 6 calculations I are based on more current data.

2. The steam generator tube rupture accident evaluation considers the increased amount of steam released to the enviro: rent via the main steam relief and atmospheric dunp valves because of the slower depressurization due to the reduced heat transfer rate caused by 7-1 I Babcock & Wilcox a McDermott company

I tripping of the reactor coolant pumps upon actuation of the high pressure injection (a post-DII-2 modification) . 'Ihis assumption is g conservative for 'IMI-1 which trips on loss of subcooling margin. 5 A camparison study has been performed to determine the change in the thyroid doses for the major accidents in Chapter 14 of the ESAR that would result fram the cycle 6 iodine inventory. 'Ihe results show that, although the thyroid doses for cycle 6 increase by a maxinn of 17% over the FSAR, the cycle 6 doses are still a small fraction of 10 CFR 100 limits and are consistent with those for the reference cycle.

7.2. Accider.t Evaluation

'Ihe key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in tPree major areas: core thermal parameters, thermal-hydraulic parameters, and kinetics parameters, including the reactivity f

• ck coefficients and control rod worths.

Core thermal properties used in the FSAR accident analysis were design operating values based on calculational values plus uncertainties. 'Ihermal parameters for fuel batches 6, 7, and 8 are given in Table 4-1. 'Ihe cycle 6 thermal-hydraulic maximum design conditions are campared with the previous cycle 5 values in Table 6-1. 'Ihese parameters are common to all the accidents considered in this report. 'Ihe key kinetics parameters fram the FSAR and cycle 6 are campared in Table 7-1.

A generic IOCA analysis has been performed for the B&W 177-FA lowered loop NSS using the Final Acceptance Criteria ECCS evaluation model; this study is reported in reference 16 arxi updated per references 17 and 18. This analysis is generic in nature since the limiting values of the key parameters for all plants in this category were used. Furthermore, the combination of the average fuel temperature as a function of linear heat l

rate and the lifetime pin pressure data used in the IOCA limits analysis is conservative campared to those calculated for this reload. 'Ihus, the analysis and the IDCA limits reported in references 16,17, and 18 provide conservative results for the operation of HII-1 cycle 6 fuel. Table 7-2 g

shows the bounding values for allowable IOCA peak linear heat rates for 'IMI- a 1 cycle 6 fuel.

I 7-2 Babcock & Wilcox l

l a McDermott company

I I It is concluded frtsn the examination of cycle 6 core thermal and kinetics properties, with respect to acceptable previous cycle values, that this core I reload will not adversely affect the 'IMI-1 plant's ability to operate safely durirg cycle 6. Considering the previously accepted design basis used in I the FSAR and 91W cycles, the transient evaluation of cycle 6 is considered to be bounded by previously accepted analyses. 'Ihe initial corditions for the transients in cycle 6 are bounded by the ESAR, the fuel densification report, and/or si W cycle analyses.

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I Table 7-1. Ocuparison of Fey Parameters for Accident Analysis

- ene densification Predicted I

Parameter levud value value Doppler coeff (BOC), Ak/k/ FC -1.17 x 10-5 -1.49 x 10-5 Doppler coeff (EDC), Ak/k/ F0 -1.33 x 10-5 -1.84 x 10-5 Moderator coeff (BDC), Ak/k/ FC +0.5'x 10-4 -0.40 x 10-4 Moderator coeff (EOC), Ak/k/C F -3.0 x 10-4 -2.64 x 10-4 All-rod group worth, % Ak/k 10.0 9.47 Initial boron concentration, ppn 1200 1273 Boron reactivity worth (HFP), 75 116 pprVl% Ak/k Max.' ejected rod worth, % Ak/k 0.65 0.26 Dig rod worth (HFP), % Ak/k O.46 $0.20 I

Table 7-2. Bounding Values for Allowable IOCA Peak Linear Heat Rates for 'IMI-1 Allowable Allowable Allowable Core Peak IHR, Peak IHR, Peak IHR Elevation, 0-1000 mwd /mtU, 1000-2600 mwd /mtU, after 2600 NNd/mtU, ft kW/ft kW/ft kW/ft 2 13.5 15.0 15.5 4 16.1 16.6 16.6 6 16.5 18.0 18.0 8 17.0 17.0 17.0 10 16.0 16.0 16.0 l I l

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8. PROIOSED IODIFICATIONS 'ID TECHNICAL SPFxmCATIO?E 1

'Ihe Technical Specifications have been revised for cycle 6 cperation to account for changes in power peakirq and control rod worths. For cycle 6, a l low leakage fuel cycle and crossflow analysis were implemented in the fuel I cycle design. 'Ihe IDCA linear heat rate limits used to develop the Technical Specification Limiting Conditions for Operation include the impact of NUREG-0630 claMing swell and rupture model. In addition, an analysis was coMucted to verify removal of' the power level cutoff hold requirements of Technical Specification sections 3.5.2.4 and 3.5.2.5. -

A cycle 6 specific analysis was conducted to generate Technical Specification Limiting conditions for Operation (rod index, power irbalance, quadrant tilt) . APSR position limits have been deleted based on the implementation of lower worth gray APSRs which reduce power peaking due to rod movement. Figure 3.5-2L is also provided, which illustrates the burnup-dependent allowable IDCA linear heat rate limits used in the analysis. 'Ihe I analysis also verified the 4.12% quadrant tilt setpoint referenced in Technical Specification 3.5.2.4. 'Ihese revisions also provide protection for the overpower condition that could occur during an overcooling transient because of nuclear instrumentation errors.

Based on the Technical Specifications derived from the analyses presented in this report, the Final Acceptance Criteria ECCS limits will not be exceeded, nor will the thermal design criteria be violated. 'Ihe following pages contain the revisions to previous Technical Specifications.

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I a conservative margin to DIB for all operating conditions. 'Ihe difference between the actual core outlet pressure and the indicated reactor coolant system pressure has been censidered in determining the core protection safety limits. 'Ihe difference in these two pressures is nominally 45 psi; however, only a 30 psi drop was ammul in reducing the pressure trip setpoints to correspond to the elevated location where the pressure is achmily measured.

'Ihe curve presented in Figure 2.1-1 represents the conditions at which a DM of 1.3 or greater is predicted for the maximum possible thermal power l (112 percent) when the reactor coolant flow is 139.8 x 10 6 lbsA, which is less than the actual flow rate for four operating reactor coolant pumps.

'Ihis curve is based on the following nuclear power peaking factors (2) with potential fuel densification and ' fuel red bowing effects; a g~

/q# = 2.82, [AH = 1.71; [Z = 1.65

'Ihe 1.65 axial peaking factor associated with the cosine flux shape provides l a lescer margin to DM of 1.3 than the 1.7 axial peakire factor associated with a lower core flux distribution. For this reason the cosine flux shape andtheassociated(=1.65ismorelimitingandthusthemoreconservative assumption.

The 1.65 cosine axial flux shape in conjunction with M = 1.71 defines the lg reference design peaking condition in the core for operation at the maximum E 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 shapes and their ammpanying radials must result in a corrlition which will not violate the previously established design I criteria on D M . 'Ihe flux shapes examined include a wide range of positive and negative offset for steady state and transient conditions.

'Ihese design limit power peaking factors 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 DM design basis.

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a McDermott company

I I %e 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 bwing:

a. 'Ihe 1.3 Da limit pr*mi by a nuclear power peaking factor of F"'= 2.82 of the ccabination of the radial peak, axial peak, ard l 9

positier.c of the axial peak that yields no less than 1.3 DM.

b. 'Ihe cxnbination of radial and axial peak that prevents central fuel

~

melting at the hot spot. 'Ihe limit is 20.5 kW/ft.

l Power peaking is not a directly observable quantity and therefore limits have been established on the basis of the reactor power imbalance previncM by the power peaking.

'Ihe specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspord to the expected minimum flow rates with four pumps, three pumps, and one pump in each loop, respectively.

'Ihe 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 DR 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 DM is equal to 22 percent, (3) whichever condition is mere restrictive.

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

80.6 percent power) plus the maximum calibration and instrumentation error. l t 'Ihe maximum thermal power for other reactor coolant pump conditions is prrrincM in a similar manner.

I Using a local quality limit of 22 percent at the point of minianum na 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 maximum I DM.

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'Ihe DM as calculated by the B&W-2 correlation continually increases from the point of minimum DE, so that the exit DM is always higher and is a function of the pressure.

2-3 Babcock & Wilcox 8-3 ,y,g,,,,,,,,,,,ny

I For each curve of Figure 2.1-3, a pressure-temperature point above and to the left of the curve would result in a na greater than 1.3 or a local E

quality at the point of minimum DE less than 22 percent for that E 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 left of this curve will be above and to the left of the other curves.

i<tzt2<rBCES

1. FSAR, Section 3.2.3.1.1

2. FSAR, Section 3.2.3.1.1.3

3. FSAR, Section 3.2.3.1.1.11 I

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I Figure 8-1. TMI-1 Core Protection Safety Limits O

R Thermal Power Level, %

I _ (-39.4,112)

_7 120 (38.1,112) l I

ACCEPTABLE. 110 4 PUMP OPERATION -- 100

_(-39.4,89.3) 90 (38.1,89.3) ~

ACCEPTABLE 3 & 4 PUMP -

- 80 (44.2,80.4)

I(.-61.3,80.4)' OPERATION

-- 70

_(-39.4.62.6) (38.1,62.6) 3 I(-61.3,57.8)

ACCEPTABLE

~

~ 60 l L (44.2,57.8) 2,3, & 4 PUMP l OPERATION

-- 50

-- 40

(- 61.3,30.4) -- 30 (44.2,30.4)

- 20

-- 10  ;

i I I t i I t I i l l l i t I I I 70 50 30 10 0 10 20 30 40 50 60 70 80 Reactor Power Imbalance, %

I Curve Reactor Coolant Flow (lb/hr) 6 1 139.8 x 10 6

2 104.5 x 10 3 68.8 x 10 6 TMI-1 CORE PROTECTION SAFETY LIMITS Figure 2,1-2 Babcock & Wilcox I 8-5 a McDermott company l

l I Figure 8-2. Core Protection Safety Bases

' I 2400-T 3

l l 2 2200 h

E {

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- a

= 3

! = /

E 2000 l

/ -

1 E

/ E l

l 5 1800 , ,

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1600 560 580 600 620 640 660 g Reactor Outlet Temperature, 'F I

REACTOR COOLANT FLOW (LBS/HR) POWER PUNPS.0PERATING (TYPE OF LIMIT)

CURVE 139.8 x 106 (1005)* 112: Four Pumc.s (DNBR Limit) 1 89.3% inree Pumps (ONER Limit) l

! 2 104.5 x 106 (74.75) 62.6% One Pump in Eacn Loop (Quality Limit) 3 68.8 x 106 (49.25)

• 106.55 of Cycle i Design Flow TMI -I CORE PROTECTION SAFETY BASES F i gui e 2.1 -3 I 8-6 Babcock &Wilcox a McDermott company 9-- w ww-,-

2.3 LIMITING SAFETY SYSTEM SETTINGS. PROTECTION INSTRUMENTATION Anolicability 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.

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

Soecification I 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 I 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 I listed in Table 2.3-1.

on the comparators.

setpoint side These trip setpoints are setting limits of the protection system bistable The safety analysis has been based upon these protection system instrumentation trip set points plus I calibration and instrumentation errors.

Nuclear Overnower A reactor trip at high power level (neutron flux) is provided to prevent damage to the fuel cladding from reactivity excursions I too rapid measurements.

to be detected by pressure and temperature During normal plant operation with all reactor coolant pumps I

operating, reactor trip is initiated when the reactor power level j reaches 105.1% of rated power. Adding to this the possible l variation in trip set points due to calibration and instrument I 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).

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2-5 8-7 Babcock & Wilcox a ucoermor company

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a. Overpower tripped 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.

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 minimum permissible low flow rate. Typical power level and low flow l 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 l

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 l i

rate is 74.7 percent or flow 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.

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l 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 I prevent reactor thermal limits from being exceeded.

thermal limits are either power peaking Kw/ft limits or DNBR These limits. The reactor power imbalance (power in the top half I

i 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 I power-to-flow ratio reduces the power level trip associated reactor power / reactor power-imbalance boundaries by 1.08 percent for a one percent flow reduction.

and

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

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 I 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 I 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.

The high pressure trip setpoint was subsequently lowered from 2390 psig to 2300 psig. The lowering of the high I pressure trip setpoint and 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 PORV while maintaining ASME Code Safety Valve capability.

The low pressure (1800 psig) and variable low pressure (11.75 TOUT - 5103) trip setpoint were initially established I 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 I 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 I pressure trips.

I 2-6 8-9 Babcock & Wilcox a uconmort company

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d. Coolant outlet temperature The high reactor coolant outlet temperature trip setting limit (618.80 F) shown in Figure 2.3-1 has been established l to prevent excessive core coolant temperature in the operating range.

The calibrated range of the temperature channels of the RPS is 5200 to .6200F. The trip setpoint of the channel is 618.8 0F. Under the worst case environment, power supply l perturbations, and drift, the accuracy of the trip string is 1.2 0F. This accuracy was arrived at by summing the worst l 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 0F even under worst case conditions. The safety analysis used a high temperature trip set point of 620 0F.

The calibrated range of the channel is that portion of the span of indication which has been qualified with regard to drift, linearity, repeatability, etc. This does not imply I that the eqpipment is restricted to operation within the l calibrated range. Additional testing has demonstrated that i in fact, the temperature channel is fully operational l approximately 10% above the calibrated range.

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M M M M M m ab M -1 M Tablo 2.3-1 RFACIOR PROIEC1' ION SYSTEM 'IRIP SEITING LIMITS (6)

Four Reactor Coolant 'Ihree Reactor Coolant One Reactor Coolant Pumps Operatiry Pumps Operating Punp Operating in (Nminal Operatirg (Nminal Operating Eadt Icop (Neinal Shutdown Ruer - 100%) Power - 75%) Operating Power - 49%) Bypass

1. Nuclear power, Max. 105.1 105.1 105.1 5.0(3) l

% of rated power

2. Nuclear power based on 1.08 times flow 1.08 times flow 1.08 times flow minus BypwW flow (2) and imbalance minus reduction due minus reduction due reduction due to max. of rated power to imbalance to imbalance imbalance

3. Nuclear power based NA NA 55% BymnW (5) on pump monitors, Max. % of rated Pwer

4. Ifigh reactor coolant 2300 2300 2300 1720(4) e system pressure, psig

,L4 max.

s

5. Iow reactor coolant 1900 1900. 1900 BymnW system pressure, psig min.

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

7. R2 actor coolant temp. 618.8 618.8 618.8 618.8 l

,[ F., Max.

it F y 8. liigh Reactor Building 4 4 4 4 jy Pressure, psig, max.

8 n" (1) Tout is in degrees Fabrenhelt (F)

J (2) Reactor coolant system flow, %

j lE (3) Administratively (4) Automatically controlled set when reduction other segments setRPS of the only(as during reactor specified) shutdown.

are bypassed.

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

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

Figure 8-3. Protection System Maximum.

Allowable Setpoints 2500 IIl P = 2300 psig

. KCEPTABLE 1a. NERATION T = 618.8 F l y 2100 ,

s)

! So t +

u 4 3 F = 1900 psig gj g 1900 u UNACCEPTABLE 8

OPERATION e

3 1700

I 1500 540 560 580 600 620 640 Reactor Outlet Temperature, F TMI-1 FROTECTION SYSTEM MALM. H ALLCWABLE SET F01NTS Figure 2.3-1 I

8-12 Babcock & Wilcox a McDermott company

I 4

Figure 8-4 Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance Thennal Power Level, %

-- 120 '

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

I my = +2.405 i ACCEPTABLE g 4 PUMP

- 100 i

g

m2 = -5.277 l OPERATION l

- 90 I(-30,80.6)

, (25,80.6).

I (-45.8,70) i ACCEPTABLE l 3 & 4 PUMP l OPERATION

- - 70 80 1

l (32.2,70)

- 60 I l -

• (-30.53.11 (25.53.1).

IACCEPTABLE -- 50

(-45.8,42.6) l2,3, & 4 PUMP ' (32.2,42.6) l OPERATION -

- 40 g

-- 30 I I I

.g co I -

- 20 l (32.2,15.1)

(-45.8,15.1) *  !

m RI e 4 il il 1 -

- 10 T/ I 7 2 El El 2 E i , i i i i , i i iI i i i i i i 70 50 30 10 0 10 20 30 40 50 60 70 80 Reactor Power Imbalance, %

I PROTECTION SYSTEM MAXIMUM ALLOWABLE SETPOINTS FOR REACTOR POWER IMBALANCE TMI-1 Figure 2.3-2 Babcock & Wilcox 8-13 a ucoermott company

I 3.5.2.4 Quadrant tilt:

a. Except for physics tests the quadrant tilt shall not exceed +4.12% as determined using the full incere l detector system.

b. When the full incore detector system is not available and except for physics tests quadrant tilt shal1~ - 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 g exceed +1.90% as determined using the minimum incore 3 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 lim.it.

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. .

2. The control rod group withdrawal limit s (Figures 3.5-2A to 3.5-2I) shall be reduced 2 percent in power for each 1 percent tilt in e::criss of the tilt limit. -

3. The power imbalance limits (Figures 3.5-2J and l 3.5-2K) shall be reduced 2 percent in power for G 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 +9.50% determined using the minimum incore detector system (MIT) if the FIT is not available, or +14.20% determined using the out of core detector system (OCT) when neither the FIT nor MIT are available, the reactor will ce placed in g the hot shutdown condition. Diagnostic testing during g 3-34 8-14 Eabcock & Wilcox a McDerrnott company

I I 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.

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

3.5.2.5 Control Rod Positions I

1

a. Operating rod group overlap shall not exceed 25 percent l 15 percent, between two sequential groups exc pt for  !

physics tests.

b. Position limits are specified for regulating control rods. Except for physics tests or exercising control rods, the regulating centrol rod insertion / withdrawal I 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 g operation is specified on Figures 3.5-2G, 3.5-2H, and g

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 I position. Acceptable control rod positions shall be attained within four hours.

3.5.2.5.c Deleted l

d. Core imbalance shall be monitored on a minimum

g frequency of once every two hours during power

g 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 l imbalance is not within the envelopes defined by g Figures 3.5-2J or 3.5-2K at the appropriate time in l 3 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.

I 3-34a 8-15 Babcock & Wilcox a ucoermocr company

3 3.5.2.7 A power map shall be taken at intervals not to exceed 30 effective full power days using the incore l

instrumentation detection system to verify the power distribution is within the limits shown in Figure 3.5-2L.

Bases The power-imbalance envelope defined in Figures 3.5-2J and 3.5-2K l is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3 . 5-2 L) such that the maximum clad l temperature will not exceed the Final Acceptance Criteria (2200F). Operation outside of the power imbalance envelope alone does not constitute a situation that would cause the Final E=

Acceptance Criteria to be exceeded should a LOCA occur. The 3 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, 3.5-2B, 3.5-2C, 3.5-2D, 3.5-2E, 3.5-2F, 3.5-2G, 3.5-2H, 3.5-2I, 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 unce.i-tainty f actors

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 g Coolant Flow transients. 3 The Rod index versus Allowable Power curves of Figures 3.5-2A, 3.5-2B, 3.5-2C, 3.5-2D, 3.5-2E, 3.5-2F, 3.5-2G, 3 . 5 - 2 H ,- and 3.5-2I describe three regions. These three regions are:

1. Permissible operating Region

2. Restricted Regions

3. Prohibited Region (Operation in t ie -(gion is not allowed) .

NOTE: Inadvertent operation within the Restricted Region for a period of four hours is not considered a violation of a limit.ng condition for operation. The limiting criteria within the Restricted Region are potential E ejected rod worth and ECCS power peaking and since the E 3-35 8-16 Sabcock &Wilcox a McDermott company

I probability of these accidents is very low especially in a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> time frame, inadvertent operation within the Restricted Region for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is allowed.

The 2515 percent overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower I part of the stroke. Control rods are arranged in groups or banks defined as follows:

GrouD Function 1 Safety 2

lI l

3 4

Safety Safety Safety 5 Regulating 6 Regulating 7 Regulating (Xenon transient override) 8 APSR (axial poV9r shaping bank) -

Control rod groups are withdrawn in seqc.ence beginning with group I 1. Groups 5, 6 and 7 are overlapped 25 percent.

position at power is for group 7 to be partially inserted.

The normal The rod position limits are based on the most limiting of the I 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 I 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 I rod posit. ion 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 I analysis (2) of the hypothetical rod ejection accident.

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 A

control rod worth 1.0% Ak/k at beginning of life, hot, zero power I- 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 I computer is out of service, then 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 I returned to service.

I 3-36 8-17 Babcock & Wilcox I a McDermott company

1 I

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 E statistically combined measurement uncertainties included. The 5 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.

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 Trio Setooint 0 <5%

15 50%

40 50%

50 60%

75 85% -

>75 105.1% l References (1) FSAR, Section 3.2.2.1.2 m (2) FSAR, Section 14.2.2.2 g I

I I

I I

e I

3-36a Sabcock & Wilcox a ucoermott company

I Figure 8- 5. Rod Position Limits for Four-Pump Operation From 0 to 30 +10/-0 EFPD, TMI-1 (300,102)

(76,102) (268,102) -

100 - NOT ALLOWED (264,92)

RESTRICTED (200,80) u 80 -

E W 70 -

IN% 60 Ie b

- 50 -

(32,50) (125,50) d 40 -

PERMISSIBLE 30 -

I 20 (0,23)

I 10 0

(0,5) 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 75 100 a f ' l I 0 25 50 75 100 e t I t '

I 0 t

25 t

50 I

75 t

100 Group 6 Group 5 I

R0D POSITION LIMITS FOR 4 PUMP OPERATION FROM 0 to 30+10/-0 EFPD TMI-1 Figure 3.5-2A Babcock & Wilcox 8-19 a ucoumore company

I Figure 8-6. Rod Position.. Limits for Four-Pump Operation From 30+10/e0 to 250+10 EFPD, TMI-1 100 -

NOT (96,102)

(272,102)

(300,102) l 90 -

.(248,80) a 80 -

E E 5 RESTRICTED W 70 t 60

==

- (46,50) (232,50) g 50 l E

2 40 I

30 20 PERMISSIBLE 10 0

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

,0 25 50 75i 100 i i i t

Group 7 g 0 25 50 75 100 5

. . I f i Group 6 0 25 50 75 100

. , i i .

Group 5 I

I l

ROD POSITION LIMITS FOR 4 PUMP OPERATION FROM 30+10/-0 TO 250+10 EFPD TMI-1 Figure 3.5-2B 8-20 Babcock & Wilcox a ucoermot company

I Figure 8.7. Rod Position Limits for Four-Pump Operation After 250110 EFPD, TMI-l (300,102) 100 -

NOT ALLOWED (264,92) 90 -

80 (248,80)

E -

70 l

W RESTRICTED E 60 -

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

e$

30 -

20 -

PERMISSIBLE (42,15) l 10 (0,5) 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 7p 1,00 0 25 50 75 100 I 0 i

25 t

50 t

75 I

100 a

Group 6 Group 5 I R0D POSITION LIMITS FOR I 4 PUMP OPERATION AFTER 250110 EFPD TMI-1 Figure 3.5-2C 8-21 Babcock & Wilcox I a McDermott company

I Figure 8- 8. Rod Position Limits for Three-Pump Operation From 0 to 30+10/-0 EFPD, TMI-1 l 100 -

I 90 -

u 80 -

NOT (300,77) E E (76,77) (268,77) -

ALLOWED 3

$ 70 - (264,69)

RESTRICTED g 60 -

(200,60) e c 50 -

l (32,38) (125,37.5) 30 PERMISSIBLE E

20< (0,17.7) 5 10 -

0 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 75 100 t t I t t Group 7 E

O 25- 50 75 100 5 e I t t t

Group 6 l

l 0

25 t

50 t

75 I

100 e

l l Group 5 I

R0D POSITION LIMITS FOR 3 PUMP OPERATION FROM 0 TO 30+10/-0 EFPD TMI-1 l

Figure 3.5-2D

' Babcock & Wilcox 8-22 a ucoermore company

I Figure 8-9. Rod Position Limits for Three-Pump Operation From 30+10/-0 to 250110 EFPD, TMI-1 I 100 -

90 -

80 -

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

ALLOWED M 70 (264,69)

E

% 60 - (248,60) j I E

. 50 -

RESTRICTED

$ 40 -

(46,38) (232,37.5) 30 20 -

(0,12.1) PERMISSIBLE l 10 '

0' 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 75 1,00 0 25 50 75 100 I 0 25 50 i

75 i

100 i

Group 6 Group 5 R0D POSITION LIMITS FOR 3 PUMP OPERATION FROM 30+10/-0 TO 250110 EFPD TMI-1 I

Figure 3.5-2E l

Babcock 11Wilcox 8-23 a uconmoir company

I Figure 8-10. Rod Position Limits for Three-Pump Operation After 250+10 EFPD, TMI-1 100 -

I 90 -

(300,77) 80 -

(171,77) (272,77) -

3 NOT ALLOWED (264,69) in 70 0

'" (248,60) 60 _

't 50 -

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

20 -

PERMISSIBLE 10 - (42,11.7)

E i (0,4.2) ' ' '

E O

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 25 50 75 100 e i t i I Group 7 E

0 25 50 75 100 5 g i I I i 0 25 50 75 100 I I t 1 I Group 5 R00 POSITION LIMITS FOR I

l 3 PUMP OPERATION AFTER l 250+10 EFPD l

TMI-1 l l I

Figure 3.5-2F 8-24 Babcock & Wilcox a ucoermoci company

1

'I Figure 8-11. Rod Position Limits for Two-Pump Operation From 0 to 30+10/-0 EFPD, TMI-1 I 100 -

90 -

8o

.I g

M 70 -

o 60 (300,52) 50 -

NOT (76,52) MsZ ^

i ALLOWED (264,46) r E 40 - RESTRICTED (200,40) 30 -

I (125,25) 20 -

(, .5) 10 - PERMISSIBLE

< ,(0,2.5) , , , , , , , , ,

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

25 f

50 f

75 I

100 I

Group 7 Group 6 I 0 t

25 I

50 i

75 I

100 t

Group 5

,I I

R0D POSITION LIMITS FOR I 2 PUMP OPERATION FROM 0 TO 30+10/-0 EFPD I TMI-1 i Figure 3.5-2G Babcock & Wilcox 8-25 a ucoermore company

I Figure 8-12. Rod Position Limits for Two-Pump Operation From 30+10/-0 to 250110 EFPD, TMI-1 100 -

I 90 -

y 80 y 70 -

~

~ I

" (96,52) (272,52) (300,52) 50 g

f NOT (264,46) E x ALLOWED 2 40 - (248,40)

RESTRICTED 30 -

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

5 10 < (0,8.7) PERMISSIBLE

, 0,2,. 5 ) , , , , , , , , ,

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

0 25 50 75 100 Group 6 g

0 25 50 75 100 3 a f I I I Group 5 I

! R0D POSITION LIMITS FOR 2 PUMP OPERATION FROM 30+10/-0 TO 250110 EFPD TMI-1 Figure 3.5-2H Babcock & Wilcox j

l 8-26 a ucoermore company l

r - . -__ _. __ . . _ - - - . . . - . _ _ _ .-.

I Figure 8-13. Rod Position Limits for Two-Pump Operation After 250110 EFPD, TMI-1 I -

100

.I 90 -

y 80 r

M 70 I =

• 60 (300,52)

• • (171,52) (272,52) -

. 50 -

g NOT (264,46) 2 ALLOWED 2 40 -

(248,40) 30 -

(94,26)

I 20 -

(232,25)

I 10 0'

(

( 0,,3. 5)42,8.

5)

PERMISSIBLE

, i i i 0 25 50 75 100 125 150 175 200 225 250 275 300 Rod Index, % Withdrawn 0 2p p0 7p 1,00 I t 0 25 I

50 f

75 I

100 e

Group 7 Group 6 I t 0 25 t

50 I

75 I

100 I

Group 5 I R0D POSITION LIMITS FOR I 2 PUMP OPERATION AFTER 250110 EFPD TMI-1 Figure 3.5-21 8-27 Babcock & Wilcox a ucoermott company 1

I Figure 8-14 Power Imbalance Envelope for Operation From 0 to 30+10/-0 EFPD E

Power, % of 2535 MWt I

RESTRICTED REGION - 110

(-20,102) '

--100 l

(-20.7,92) ' ' *

- 90

(-24,80)o -- 80 >(32,80) .I

- 70 PERMISSIBLE

-- 60 OPERATING REGION

-- 50

- 40

-- 30

-- 20

-- 10 1 I t I l I I  ! I '

=

40 20 -10 0 10 20 30 40 50 Power Imbalance, %

I POWER IMBALANCE ENVELOPE FOR OPERATION FROM 0 TO 30+10/-0 EFPD TMI-1 Figure 3.5-2J 8-28 Babcock & Wilcox a McDermott company

I I Figure 8-15. Power Imbalance Envelope for Operation After 30+10/-0 EFFD I

Power, % of 2535 MWt RESTRICTED REGION --110

(-20,102) -

(28.5,102) '

100

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

{'

I (-22.4,80)< ' -

80 ,(32.8,80)

-- 70 PERMISSIBLE 60 OPERATING I

REGION

.. - 50

-- 40 30

-- 20 I -

10

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

I POWER IMBALANCE ENVELOPE FOR OPERATION AFTER 30+10/-0 EFPD TMI-1 Figure 3.5-2K 8-29 Babcock & Wilcox

. ucoermore company

I Figure 8-16. LOCA Limited Maximum Allowable Linear Heat Rate, TMI-1 l I

i i i i I

20 , , i i i i i I

18 sa ~

. I s ,

"",,s

$ 16 -

a lf ~

/ i E _ / j e /

'E /

o 14 -

f

/

I l 3

• ~ ~~ ~ 0-1000 mwd /mtU ~ ~

12 -

__ _ 1000-2600 mwd /mtU After 2600 mwd /mtU _

I i 10 '

8 10 12 2 4 6 O

Axial Location of Peak Power From Bottom of Core, ft l I I

LOCA LIMITED MAXIMUM ALLOWABLE LINEAR HEAT RATE TMI-1 E

Figure 3.5-2L 8-30 Bat > cock & Wilcox a McDermott company l

l . - - . . - - _ . - _ _ _ .

I I

I

9. SIARIUP PROGRAM - RlYSICS TESTING

'Ihe planned startup test pr @ am a m v'4ated with core performance is outlined below. 'Ihese tests verify that core performance is within the n=nptions of the safety analysis and provide informtion for continued safe operation of the unit.

9.1. Precritical Tests 9.1.1. Control Rod Trio 'Ibst Precritical wiLwl roi drop times are recorded for all control rods at hot full-flow conditions before zero power physics testing begins. Acceptance criteria state that the red drop time frm fully withdrawn to 75% inserted shall be less than 1.66 seconds at the conditions above.

It should be noted that safety analysis calculations are based on a rod drop frm fully withdrawn to two-thirds inserted. Since the most accurate position indication is obtained frm the zone reference switch at the 75%-inserted position, this position is used instead of the two-thirds inserted position for data gathering.

9.1.2. RC Flow Reactor coolant flow with four RC pumps running will be mammxi at hot shutdown conditions. Acceptance criteria require that the mamrud flow be within allowable limits. 'Ihis test is planned for cycle 6 to verify flow performance with IEP wwents.

9.2. Zero Power M1vsics Tests 9.2.1. Critical Boron Concentration I

I once initial criticality is achieved, equilibrium boren is obtained and the l critical boren concentration determined. 'Ihe critical baron concentration is calculated by correcting for any rod withdrawal required to achieve equilibrium boren. 'Ihe acx:cptance criterion placed on critical boron l

1 9-1 l

l I Babcock &Wilcox a McDermott company

I w a.wikation is that the actual baron concentration nust be within 100 lIl ppn boron of the predicted value.

9.2.2. Tencerature Reactivity Ccefficient

'Ihe isothermal HZP tenperature coefficient is measured at approximately the all-rods-cut cxmfiguration. During changes in temperature, reactivity f*ck may be -pated by c.uikvl rod movement. 'Ihe change in I

reactivity is then calculated by the summation of reactivity (obtained from 3 a reactivity calculator strip chart recorder) associated with the tmperature change. Acceptance criteria state that the mear.,ured value shall not differ frcan the predicted value by more than 0.4x10-4 AM.

The moderator coefficient of reactivity is calculated in conjunction with the temperature coefficient measurement. After the t eperature coefficient has been wacure:1, a predicted value of fuel Doppler coefficient of reactivity is ^Mui to obtain the unbrator coefficient. 'Ihis value must not be in eyrann of the acceptance criteria limit of +0.5x10-4 AM.

9.2.3. Control Rod Grouo/ Boron Reactivity Worth Control rod group reactivity worths (groups 5, 6, and 7) are mammx1 at hot zero power coniitions usire the borary' red swap method. 'Ihis technique consists of establishing a deboration rate in the reactor coolant system and compensating for the reactivity changes frun this deboration by inserting control red groups 7, 6, and 5 in irceumital steps. 'Ihe reactivity changes that occur during these measurements are calculated based on reactimeter data, and differential red worths are obtained frun the measured reactivity worth versus the change in rod group position. 'Ihe differential rod worths -

of each of the conkulling groups are then smrni to obtain integral red group worths. 'Ihe acceptance criteria for the c.wikul bank group worths are as follows:

I

1. Individual bank 5, 6, 7 worth:

credicted value - measured value x 100 < 15 measured value -

I I

9-2 Babcock & Wilcox a McDermott company

I I 2. Sums of groups 5, 6, and 7:

I Dredicted value - measured value man =1 red value x 100 < 10

%e baron reactivity worth (differential boron worth) is .maamired by dividing the total insertA rod worth by the boron change made for the red worth test. S e acceptance criterion for maa=Lred differential baron worth I is as follows:

1. oredicted value - measured value I maamtred value x 100 -< 15 The predicted differential baron worth is taken fr m the PIM.

9.3. Po w r Escalation Tests I 9.3.1. Ccre Power Distribution Verification at Irtamar14 ate Power Level (IPL) and 100% FP With Nominal Control Rod Position i

Core power distribution tests are performed at the IPL and 100% full power I (FP). %e test at the IPL is essentially a check on power distribution in core to identify any abnormalities before escalating to the 100% FP the plateau. Peaking factor criteria are applied to the IPL core power distribution results to determine if additional tests are required prior to 100% FP operation.

The following acceptance criteria are placed on the IPL tests:

1. The worst-casa maxinn IER must be less than the IOCA limit.

2. The minimum DE must be greater than 1.30.

3. The value obtained frm extrapolation of the minimum DE to the next power plateau overpower trip setpoint must be greater than 1.30, or the extrapolated value of imbalance must fall outside the RPS power / imbalance / flow trip envelope.

4. The value obtained frun extrapolation of the worst-case maximum IHR to the next power plateau overpower trip setpoint must be less than the I fuel melt limit, or the extrapolated value of irbalance must fall outside the RPS power / imbalance /ficu trip envelope.

9-3 I Babcock & Wilcox a McDermott company

E

5. h quadrant pwer tilt shall not exceed the limits specified in the Technical Specifications.

6. h highest mamired and prwiir+=1 radial peaks shall be within the follwing limits:

credicted value - measured value x 100 more positive than -8 mammx1 value

7. h highest mammri and predicted total peaks shall be within the following limits:

credicted value - measured value x 100 more positive than -12 m amired value Items 1, 2, 5, 6, and 7 are established to verify core nuclear and thermal calculational melala, thereby verifyiry the acceptability of data frcm these wrlois for input to safety evaluations.

Items 3 and 4 establish the criteria whereby m 1ation to full power may be accmplished without eyrmdirg the safety limits specified by the safety analysis with regard to DER and linear heat rate.

N power distribution test performed at 100% FP is identical to the IPL test except that core equilibrium xenon is established prior to the 100% FP test. Accordingly, the 100% FP measured peak acceptance criteria are as follows:

1. h highest measured and predicted radial peaks shall be within the followiry limits:

predicted value - measured value x 100 more positive than -5 wamired value l

l

2. h highest mamired and predicted total peaks shall be within the following limits:

predicted value - measured value x 100 more positive than -7.5 measured value 9.3.2. Incore Vs. Excore Detector Imbalance Correlation Verification at the IPL Imbalances, set up in the core by control rod positioning, are read simultaneously on the incore detectors and excore power range detectors.

l l 9-4 l

Babcock & Wilcox a McDermott company

l I 2e excore detector offset versus incore detector offset slope must be l I

I greater than 0.96. If this criterion is not met, gain amplifiers on the excore detector signal processing equipnent are adjusted to provide the required gain.

I 9.3.3. Temperature Reactivity Coefficient at 100% FP 1

The average reactor coolant temperature is decreased and then increased by about 5 0 F at constant reactor power. The reactivity associated with each taperature change is obtained frm the change in the controllire rod group I position. Controlling red group worth is =amtred by the fast insert / withdraw method. %e temperature reactivity coefficient is

.I calculated from the measured changes in reactivity and temperature.

Acceptance criteria state that the moderator taperature coefficient shall be negative.

9.3.4. Power Dopoler Reactivity Coefficient at 100% FP Reactor power is deczna M and then increased by about 5% FP. %e reactivity charge is obtained frm the change in wuLwlling red group position. Control red group worth is measured using the fast insert / withdraw method. Reactivity corrections are made for changes in xenon and reactor coolant temperaturn that emn durirg the measurement.

The power Doppler reactivity coefficient is calculated frm the measured reactivity change, adjusted as stated above, and the maatred power change.

l The fuel Doppler reactivity coefficient is calculated in conjunction with the power Doppler coefficient measurement. The power Doppler coefficient as measured above is multiplied by a predicted conversion factor to cPain the This measured fuel Doppler coefficient must be I fuel Doppler coefficient.

more negative than the acceptance criteria limit of -0.90 x 10-5 ggop,

( 9.4. Procedure for Use if Accentance Criteria Not Met If acceptance criteria for any test are not met, an evaluation is performed before the test prugum is continued. R1rther specific actions depend on l evaluation results. These actions can include repeating the tests with more detailed attention to test prerequisites, added tests to search for anomalies, or design personnel performing detailed analyses of potential

( safety problems Wmm of parameter deviation. Power is not escalated 9-5 I Babcock & Wilcox a McOctmott company

I until evaluation shows that plant safety will not be meu aid by such escalation.

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10. REFERDKES

1. 'Ihree Mile Island Nuclear Station, Unit 1, Final Safety Analysis Report, USNRC Docket No. 50-289.

2. BPRA Retainer Design Report, BAW-1496, R,hmk and Wilcox, Lynchburg, VA, May 1978.

I 3. J. H. Taylor (B&W) to S.

Reinsertion," January 14, 1980.

A. Varga (NRC), letter, "BPRA Retainer

'Ihree Mile Island Unit 1 Cycle 5 Reload Report, BAW-1509, I

4.

R, W k & WilCox, Lynchburg, Virginia, November 1978.

5. Frugscuu to Determine In-Reactor Performance of B&W Fuels - Cladding I creep Collapse, BAW-10084P, Rev. 2, R,Wk and Wilcox, Lynchburg, VA, October 1978.

I 6. 'IMI-1 R2el Densification Report, Lynchburg, VA, June 1973.

BAW-1389, R,Wk and WilCox,

7. TACO 2 : Riel Performance Analysis, BAW-10141P-A Rev.1, R,Wk I and Wilcox, Lynchburg, Virginia, June 1983.

8. R,hk Wilcox Version of PDQ User's Manual, BAW-10117P-A, R, W k &

Wilcox, Lynchburg, Virginia, January 1977.

9. NCODIE - A Multi-Dimensional 'IWo-Group Reactor Simulator, BAW-10152A, R,Wk & Wilcox, Lynchburg, Virginia, June 1985.

10. Comparison of Core Ihysics Calculations with Measurements, BAW-10120, Babcock & Wilcox, Lynchburg, Virginia, June 1978.

11. LYNX 1 : Reactor Fuel Assembly 'Ihermal Hydraulic Analysis Code, BAW-10129-A, R, W k and WilCox, LynChburg, Virginia, July 1985.

12. LYNX 2 : Subchannel 'Ihermal Hydraulic Analyc,is Code, EAW-10130-A, R,Wk arxi Wilcox, Lynchburg, Virginia, July 1985.

13. LYNXT : Core Transient 'Ihermal Hydraulic Frugscua, BAW-10156, R,hk and Wilcox, Lynchburg, Virginia, February 1984.

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14. 'Ibermal-Hydraulic Clussflow Applications, BAW-1829, mWk and Wilcx:oc, Lynchburg, Virginia May 1984.

15. Fuel Rod Bowing in mWk & Wilcox Riel Design, BAW-10147P-A. Rev.1, m W k and Wilcox, Lynchburg, Virginia, May 1983.

16. ECCS Analysis of B&W's 177-FA lowered Icop NSS, BAW-10103, Rev.3, M W k & Wilcox, Lynchburg, Virginia, July 1977.

17. TACD2 Ioss-of-Coolant Accident Limit Analysis for 177-FA IcWered Icop Plants, BAW-1775, Rev. O, NWk & Wilcox, Lynchburg, Virginia, February 1983.

18. Bounding Analytical Aw=nt of NUREG-0630 Models on IOCA kW/ft Limits with Use of FIECSET, BAW-1915P, mWk & Wilcox, Lynchburg, Virginia, May 1986.

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