ML19322A210

Forwards Tech Spec Change Request 86 to Amend App a of License DPR-50 & Certificate of Svc.Change Provides Tech Specs for Cycle 5 Operation of Unit 1 W/Anticipated Startup of 790314

ML19322A210
Person / Time
Site:	Crane
Issue date:	12/28/1978
From:	Herbein J METROPOLITAN EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
FOIA-79-98 GQL-2068, NUDOCS 7901100163
Download: ML19322A210 (72)

Text

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Metropolitan Edison Company j

e Post Oft:ce Box 542 RtuJnq Penn ;yNan:n 19r )3 a

J15 929 3601 December 28, 1978 GQL 2068 Director of Nuclear Reactor Regulation Attn:

R. W.

Reid, Chief Operating Reactors Branch No. 4 U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Sir:

Three Mile Island Nuclear Station, Unit 1 (TMI-1)

Docket No. 50-289 Operating License No. DPR-50 Technical Specification Change Request No. 86 Enclosed are three signed originals (sixty conformed copies sent separately) of Technical Specification Change Request No. 86 requesting amendment to Appendix A of Operating License No. DPR-50, as well as a check for $12,300.00 as required, per 10 CFR 170.22 for a Class IV Amendment.

Also enclosed is one signed copy of Certificate of Service for proposed Technical Specificatien Change Request No. 86 to the chief executives of the township and county in which the facility is located.

Technical Specification Change Request No. 86 provides the specifications for Cycle 5 operation of TMI-1.

The anticipated Cycle 5 startup is March 14, 1979.

Therefore, the Licensee requests that Operating License No. DPR-50 be amended for Cycle 5 operation no later than March 14, 1979.

Sincerely, J. G. Herbein Vice President Generat ion JGH:WSS:p1w Enclosuras:

1)

Technical Specification Change Request No. 86 2)

Certificate of Service for Technical Specification Change Request No. 86 3)

Metropolitan Edison Company Check No. 102245 4)

Three Mile Island Unit 1, Cycle 5 Reload Report,

December 1978 7 9 01 10 0 1(o3p p mn a n conws a veneonne cenaar ecc um saem

r METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY NED PENUSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT I Operating License No. DPR-50 Docket No. 50-289 Technical Specification Chn,nce Recuest No. 86 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR 50 for Three Mile Island Nuclear Station Unit 1.

As a part of this request, proposed replacement pages for Appendix A are also included.

METROPOLITAN EDICON COMPANY By Sworn and subscribel to me this 2 day of '

, 1978.

w >-to&l W Notary Publif i

NOW?Y PUF4UC

C ':,i; k'y COirtm.$4 L " TM IOv 13.I3')

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UNITED STATES OF M4 ERICA NUCLEAR REGULATORY COMMICSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No.86 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Commission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:

Mr. Weldon B. Archart Mr. Harry B. Reese, Jr.

Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 1705:

darrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY By

[

DEC 2 8 M Dated:

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Three ' tile Island Nuclear Station, Unit 1 (TMI-1)

Operating Liernse No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 86 The Licensee requests that the attached changed pages replace pages vi, vii, 2-3, 2-6, 3-19, 3-20, and 3-34 through 3-36 and Figures 2.1-1, 2.1-2, 2.1-3, 2.3-2, 3.5-2A, 3.5-2B, 3.5-20, 3.5-20, 3.5-2E, 3.5-2G, and 3.5-2H.

Figure 3.5-2F is to be deleted.

Reasons for Change Request This change is requested to provide Technical Specifications for operation of TMI-1 for Cycle 5.

This change was developed with the intent of bounding future cycles.

Several of the changes are administrative in nature and are identified as such below.

Safety Evaluation Justifying Change Pages vi and vii Changes to these pages were strictly administrative in nature. The changes reficct the correct figure titles.

Page 2-3 The maximum thermal power for three pump operation changed from 87.1% to 87.2%

as a result of roundof f.

This change is administrative and of no safety significance.

Figure 2.1-1 This figure was changed to delete the " Cycle" reference. This figure is valid for Cycle 5 operation and should be bounding for future cycles. This change is administrative in nature and of no safety significance.

Figure 2.1-2 The core protection safety limits are determined by finding power levels / core imbalances where violation of the minimum DNBR criteria or linear heat rate for fuel melt would occur for various pump operating modes. However, the limits removing instrument presented in Figure 2.1-2 were derived from Figure 2.3-2 ?

error and calculational uncertainty in flux / flow determinc ~ons.

In all cases, the limits of Figure 2.1-2 are the more restrictive of the two determinations above. The Cycle 5 specific limits were compared to the safety limits of Figure 2.1-2 and were bounded by the given limits (see Table 1, attached). In addition, the allowable radial-local peak for Cycle 5 was reduced from 1.78 to 1.71 to ensure that all previous DNBR limiting Tech. Specs. will remain limiting for Cycle 5 and should be limiting for future cycles.

Figure 2.1-3 This figure was changed to delete the " Cycle" reference and to change the Curve 2 power from 87.1% to 87.2%.

This figure is valid for Cycle 5 operation and should be bounding for future cycles. The change from 87.1% to 87.2% is the result of roundoff. These changes are administrative and of no safety significance.

Page 2-6 The reactor coolant ' low rate was changed from h9 1% to 49 2% as a result of roundoff, f

and for consistency with Figure 2.1-3.

This change is administrative and of no safety significance.

Figure 2.3-2 This figure is unchanged for Cycle 5 and is therefore submitted to delete the " Cycle" reference. Furthermore, this figure is the basis for deriving the safety limits of Figure 2.1-2 which was bounding for Cycle 5 - specific limits.

It follows that the Cycle 5 - specific setpoints are bounded therefore by the setpoints of Figure 2.3-2.

Pages 3-19 and 3-20 These pages were changed to increase the volume of concentrated boric acid solution required in the boric acid mix tank and reclaimed boric acid storage tank, and the volume of boron as boric acid solution required in the BWST. Due to the increased volumes, the. injection times increased somewhat.

These boric acid requirements are Cycle 5 - specific, and may not be bounding for future cycles.

Page 3-3h The quadrant tilt as determined using the full incore detector system was reduced from +3.6h% for Cycle h to +3.52% for Cycle 5 This reduction in measured tilt is the result of increased uncertainties due to detector depletion.

l Page 3-3ha i

The quadre tilts as determined using the FIT, MIT, or OCT, have been reduced fron l

+26.75% to.16.80%, +15.21% to +9.50%, and +22 96% to +14.20% respectively. These

]

limits are bounding for Cycle 5 and should be bounding for future c: :les.

I Pages 3-35 and 3-35a i

These pages have been changed to delete reference to Figure 3.5-2F, which has been deleted. This change is administrative and of no safety significance.

Page 3-36 The quadrant power tilt was reduced to +3.52% to be consistent with page 3-3h.

This change is administrative and of no safety significance.

4 i

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, Figures 3.5-2A, 3.5-2B, 3.5-20, 3.5-2D, and 3.5-2E These figures were established through examination of previous feed-bleed operating cycles that have been analyzed by B&W for their 177 - lowered loop plants of the 2568 MWt power classification. These figures are bounding for Cycle 5 and should be bounding for futur' yeles. A comparison of the Cycle 5 - specific data to the key limiting Tech S: -.alues is presented in Table 2 and Table 3, attached.

The limiting values for che CRA position curves are LOCA kv/ft based, and provide acceptable imbale-es with the 1% shutdown margin limit defining the area where operation is "

ellowed". Ejected rod limits fall within the " Restricted" region

~

and hence do not affect normal operation.

The rod position limits for 2 and 3 pu=p operation vere developed by power scaling (i.e., % of maximum power allowable for 2 and 3 pump operation) the limiting boundary for the " Permissible"' region of the h-pump rod position limits. The shutdown margin limit was retained for all pump combinations, adding additional conservatism to the 2 and 3 pump curves.

The Cyc'.e 5 minimum shutdown margin is 2.10% Ak/k well above the 1% Ak/k requirement.

The ma71 mum ejected rod worth for Cycle 5 is 0.71% Ak/k at Hot Zero Power and 0.25% tk/k at Hot Full Power, well below the limits of 1.0% Ak/k and 0.65$ Ak/k, respectively.

The power imbalance limits of Figure 3.5-2E cover the full cycle rather than change in mid-cycle, as done in previous cycles; therefore permitting deletion of Figure 3.5-2F.

The power imbalance limits are based on LOCA kv/ft limits. A comparison of the Cycle 5 - specific imbalance limits to the limiting Tech Spec values is presented in Table 2, attached.

The EOC value for these figures where applicable is 280 EFPD's for Cycle 5 The specific number was not included to provide cycle - independent figures that can be used for different cycle lengths if bounding.

Figure 3.5-2F Deleted as discussed above.

Figure 3.5-2G This figure is valid for Cycle 5 and should be bounding for future cycles. This figure is therefore submitted to delete the " Cycle" reference. This change is administrative and of no safety significance.

Figure 3.5-2H The APSR limits, based on LOCA kv/ft limits / Imbalance limits, are more restrictive than those of previous cycles. The most limiting APSR positions are those positions, that when coupled with the most limiting full length control rod positions, vill not exceed the LOCA kv/ft limits / Imbalance limits. The Cycle 5 CRA position limits are bounded by the Tech Spec values, and therefore, the given APSR position limits are bounding for Cycle 5 operation, and should be bounding for future cycles. 1

Conclusions Based on the above, it can be concluded that the proposed Technical Specifications are applicable to and/or are bounding for Cycle 5 operation, and support a full power Cycle 5 operation for 0 to 280 EFFD's without endangering the health and safety of the public.

Amendment Clasq (10 CFR 170)

The Licensee has determined, based on NRC guidance, that because this Amendment is a Reload submittal, it is a Class IV Amendment, and the appropriate fee therefore is $12,300.00.

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f 5.

TABLE 1. Comparison of Cycle 5 to Limiting Core Protection Safety Limits Tech Spec Value Cycle 5 Value (5 Imbalance, % Pull Power)

($ Imbalance, % Full Power)

(+30, 112)

(+hh.8, 112)

?

(+43.1,100)

(+53, 100)

(+34.0, 80)

(+64, 80)

(+36.9, 50)

(+60, 50)

(-29, 112)

(-53.8, 112)

(-39 2, 100)

(-80, 100)

(-31.8, 80)

(-105.6, 80) 1 TABLE 2. Comparison of Cycle 5 to Limiting Power Imbalance Limits Tech Spec Value Cycle 5 Value

(% Imbalance, % Full Power)

(% Imbalance, % Full Pover)

(+10.0,102)

(No Limit, 102)

(+10.0, 92)

(No Limit, 92)

(+10.0, 80)

(No Limit, 80)

(-15 0, 102)

(-23.8, 102)

(-15 0, 92)

(-30.3, 92)

(-20.0, 80)

(-37.6, 80) 4 TABLE 3. Comparison of cycle 5 to Limiting

" Permissible" and " Restricted" Values Permissible Region (LOCA Based)

Tech Spec Value Cycle 5 Value (Rod Index, % Full Power)

(Rod Index, % Full Power)

O to 125 i 5 125 i 5 to EOC 0 to 125 5

125 i 5 to EOC (270 9, 102)

(270 9, 102)

(267.1,102)

(250,102)

(270.9, 92)

(270.9, 92) more margin than at 102% Full Power (2h8.2, 80)

(248.2, 80) more margin than at 102% Full Power Restricted Region (Shutdown Margin Based)

180, 102)

(250, 102)

(70, 102)

(170,102)

(130,

(

50)

(190, 50)

-( 5, 50)

(100, 50)

( 70, 15)

(1h0, 15)

( 0, 15)

( 30, 15)

(_ 0, 0)

(

0, 0)

( 0, 0)

(

0, 0)

LIST OF FIGURES FIGURES TITLE 2.1-1 TMI-l Core Protection Safety Limit 2.1-2 TMI-l Core Protection Safety Limits 2.1-3 TMI-l Core Protection Safety Bases 2.3-1 TMI-l Protection System Maximum Allowable Set Points 3

2.3-2 Protection System Maximum Allowable Set Points for Reac.cor Power Imbalance, TMI-l 3.1-1 Reactor Coolant System Heat-up/Cooldown Limitations (Applicable to 5 EFPY) 3.1-2 Reactor Coolant System, Inservice Leak and Hydrostatic Test Llmitations (Applicable to 5 EFPY) 3.1-3 Limiting Pressure vs. Temperature Curve for 100 STD cc/ Liter H O 2

3.5-1 Incore Instrumentation Specification Axial Imbalance Indication, TMI-l i

3.5-2 Incore Instrumentation Specification Radial Flux T2.lt Indication, TMI-l 3.5-2A Rod Position Limits for 4 Pump Operation From 0 to 125 1 5 EFPD, TMI-l 3.5-2B Rod Position Limits for 4 Pump Operation from 125 1 5 EFPD to EOC, TMI-l 3.5-2C Rod Position Limits for 2 and 3 Pump Operation from 0 to 125 i EFPD, TMI-l 3.5-2D Rod Position Limits for 2 and 3 Pump Operation from 125 i 5 EFFD to EOC, TMI-l 3.5-2E Power Imbalance Envelope for Operation from 0 EFPD to EOC vi

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FIGURES TITLE 3.5-2F Deleted 3.5-2G LOCA Limited Maximum Allowable Linear Heat Rate - TMI-l 3.5-2H APSR Position Limits for Operation from 0 EFPD to EOC 3.5-2I Deleted 3.5-2J Deleted 3.5-2K Deleted 3.5-2L Deleted I

3.5-2M Deleted 3.5-2N Deleted 3.5-3 Incore Instrumentation Specification, T?tI-l i

1 4.2-1 Equipment and Piping Requiring Inservice Inspection in Accordance with Section XI of the ASME Code 4.4-1 Ring Girder Surveillance, TMI-l 4.4-2 Ring Girder Surveillance Crack Pattern Chart, TMI-l 4.4-3 Ring Girder Surveillance Crack Pattern Chart, TMI-1 4.4-4 Ring ';irder Surveillance Crack Pattern Chart, TMI-l

. 4.4-5 Ring Girder Surveillance Crack Pattern Chart, TMI-l 6-1 Met-Ed Corporate Technical Support Staff and Station Organization Chart vil w-

Tae curve of Figure 2.1-1 is the most restrictive of all possible reactor v

coolant pu=p-caximus thermal pover combinations shov in Figure 2.1-2.

Tne curves of Figure 2.1-3 represent the conditions at which a minimum DIiBR of 1.3 is predicted at the maximum possible ther-al power for the nu=ber of reactor coolant pu=ps in operation or the local quality at the point of mini =um D!iBR is equal to 22 percent, (3) hichever condition is nore restrictive.

The maximum thermal power for three pump operation is 87.2 percent due to a l

power level trip produced by the flux-flow ratio (Th.T percent flov x 1.08 =

80 7 percent power) plus the maximum calibration and instrumentation error.

The maxi =us ther=al power for other reactor coolant pump conditions is produced

,d in a similar =anner.

Using a local quality li=it of 22 percent at the point of minimun 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 cale ilated by the 3&'l-2 correlation continuey increases from the point of =ini=us D:iBR, so that the exit DN3R is always higher and is a function of the pressure.

For each curve of Figure 2.1-3, a pressure-te=perature point above and to the left of the curve vould result in a DIGR greater than 1.3 or a local quality at the point of =ini=u= DIGR less than 22 percent for that particular reactor (l

coolant pu=p situation.

Curve 1 is more restrictive than any other reactor coolant pu=p situation because any pressure /ce:perature point above and to the left of this curve vill be above and to the left of the other curves.

RETERETCES (1)

FSAR, Section 3.2.3.1.1 (2)

FSAR, Section 3.2.3.1.1.c (3)

FSAR, Section 3.2.3.1.1.k v

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v 2-3

2600 2400

.=a ACCEPTA8LE E.

OPERATION f

2200

/

2 n.

5 2000 e*

8 UNACCEPTABLE OPERATION 1800 7

1600 560 580 600 620 640 660 Reactor Outlet Temperature, F TMI-1 CORE PROTECTION SAFETY LIMIT Figure 2.1-1

CR Thermal Power Level, %

-- 120 (ll2)

_l (30,112)

(-29,112)

ACCEPTAF.LE 4

4 PUNP

-- 300 (47.I,96.5)

OPERATION

( -46. 4,91. 5) 30 (87.2)2

(-25.4,87.2 (25.8.87.2)

ACCEPTABLE

~~

3&4 PUMP OPERATION 70 (46.4,69.2)

__ 60 (59. 6_ j 3

ACCEPTABLE 2,354 PUMP ~~ 50 OPERATION (47.5.41.6)

( -48. l. 36. 6 )

30 20 10 t

f f

f f

I I

I f

f f

f

-60

-50

-40

-30 10 0

10 20 30 40 50 60 Reactor Power imbalance, f>

Curve Reactor Coolant Flow (lo/hr) 6 I

139.8 x 10 6

2 104.5 x 10 6

3 68.8 x 10 CORE PROTECTION SAFETY LIMITS Figure 2.1-2

2400 1

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~

5 2200 E

{

d 2

5 U

E 2000

/

3 3

E o

1800 y'

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

REACTOR COOLANT FLOW CURVE (LBS/HR)

POWER PUMPS OPERATING (TYPE OF LIMIT) 1 139.8 x 106 (100%)*

1125 Four Pumas (DNBR Limit) 2 104.5 x 106 (74,75)

M.25 Three Pumps (DNBR Limit) 3 68.8 x 106 (49.25) 59.65 One Pump in Eacn loop (Quality Limit)

• 106.5% of Cycle 1 Design Flaw

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TMI-l CORE PROTECTION SAFETY BASES Figure 2.1 3 1

c: "

The povsr level trip sst point produced by the povar-to-flow ratio provides both high power level and lov flow protection in the event the reactor power level increases or the reactor coolant flow rate decreases, s,f The power level trip set point produced by the power to flow ratio provides overpower DNB protection for all =cdes of pa=p operation.

For every flov rate there is a =aximu= per=issible power level, and for overy power level there is a mini =u= per=issible lov flow rate.

Typical power level and lov flow rate co=binations for the pc=p situations of Table 2 3-1 are as follows:

N 1.

Trip would occur when four reactor coolant pu=ps 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 pu=ps are operating if power is 80.7 percent and reactor flow rate is 74 7 percent or flow rate is 69.h percent and power level is 75 percent.

3.

Trip vould occur when one reactor coolant pump is operating in each loop (total cf two pu=ps operatin6) if the power is 53.1 percent and reactor flow rate is L9 2 percent or flow rate is l

45.3 percent and the power level is h9 percent.

Tne flux / flow ratios account for the maxi =u: calibration and instru=entation errors and the maxi =u: variation fro = the average value of the RC flow signal in such a =anner that the reactor protective syste= receives a

nservative indication of the RC flov.

Ne penalty in reactor coolant flov through the core vas taken for an 0;en core vent valve because of the core vent velve surveillance progra=

during each refueling outage, p:

safet; analysis calculations the =axi=ur calibration and instrumentation errors for the power level vere used.

The pover-i= balance boundaries are established in order to prevent reactor ther=al limits frc= being exceeded.

These ther=21 limits are either power peaking Kv/ft li=its or DU33 li=its.

The reactor power i=':alance (pc.er in the top half of the core minus power in the botto=

half of core) reduces the power level trip produced by the pover-to-flov ratio so that the boundaries of Figure 2.3-2 are produced.

The power-g'

-flow ratio reduces the power level trip and associated reactor pover/ reactor

ver-i= balance boundaries by 1.08 percent for a one percent flow reduction.

b.

Pu=p nonitors 2ne reoundant pu=p =enitors prevent the =ini=u= core DUBR fro =

decreasing below 1 3 by tripping the reactor due to the loss of reactor coolan pt=p( s ).

The pur.p =enitors also restrict

• porre level for the number of pu=ps in operation.

2-6

a Thermal Power Level, 5 l

_ 120 J

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j

(-17,108) 110 (108) (17,108)

I 100 l

M2 = -1.0 M1 = 1.28 ACCEPTABLE

\\

4 PUMP

- 90 (35,90)

GPERATION

(-35,85)

I (80.7) i j

f

-- 80

/ I 70 ACCEPTABLE i

3 & 4 PUMP (35,62.7)

OPERATION

(-35,57. 7 )

l

-- 60 t

(53.1)l l

50 l

l ACCEPTABLE 40 2,3 & 4 PUMP

( 5,35.1)

OPERATICN

( -35,3 0.1 )

l

__ 30 I

I

-- 20 o

I o

s E

?

lC

~i

-- 10 i

6-l E

I e

1 t

t I t

i t

I t

i 1

-70

-60

-50

-40

-30

-20

-10 0

10 20 30 40 50 60 70 Reactor Power imoalance, 5 l

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

l

3.2 MAKEUP AND PURIFICATION AND CHEMICAL ADDITION SYSTEMS Apulicability Applies to the operational status of the makeup and purification and the chemical addition systems.

Objective To provide for adequate boration under all operating conditions to assure ability to bring the reactor to a cold shutdown condition.

Specification The reactor shall not be critical unless the following conditions are met:

3.2.1 Two makeup and purification pumps are operable except as specified in 3.3.2.

3.2.2 A source of concentrated boric acid solution, in addition to the borated water storage tank, is available and operable. This can be either:

a.

The boric acid mix tank containing at least the equivalent of 906 ft3 of 8700 ppm boron as borie acid solution with a temperature l

of at least 100F above the crystallization temperature. System piping and valves necessary to establish a flow path from the tank to the makeup and purification system shall also be operable and shall have at least the same temperature requirement as the boric acid mix tank. One associated boric acid pump shall be operable.

b.

A reclaimed boric acid storage tank containing at least the equivalent of 906 ft3 of e700 ppm boron as boric acid solution 0

with a temperature of at least 10 F above the crystallization t emperature.

System piping and valves necessary to establish a flow path from the tank to the =akeup and purification system shall also be operable and shall have at least the same temperature requirement as the reclaimed boric acid tank. One associated reclaimed boric acid pump shall be operable.

Bases The makeup and purification system and chemical addition systems provide control of the reactor coolant boron concentration.

(1) This is normally accomplished by using any of the three cakeup and purification pumps in series with a boric acid pump associated with the boric acid mix tank or a reclaimed boric acid pump associated with a reclaimed boric acid storage tank. The alternate method of boration vill be the use of the makeup and purification pumps taking suction directly from the borated water storage tank. (2) 3-19

The quantity of boric acid in storage from either of the three above mentioned sources is sufficient to borate the reactor coolant system to a one percent.suberitical margin in the cold condition at the vorst time in core life with a stuck control rod assembly. Minimum volumes (including a 10 percent safety factor) of 906 ft3 of 8700 ppm boron as concentrated l

boric acid solution in the boric acid mix tank or in a reclaimed boric acid the borated water storage tank (o( 2270 ppm boren as borie acid solution in l

storage tank or 32,112 gallons 31 vill each satisfy this requirement. The specification assures that at least two of these supplies are available i

whenever the reactor is critical so that a single failure vill not prevent boration to a cold condition. The nini=um volunes of boric acid solution given include the boron necessary to account for xenon decay.

The primary method of adding boron to the reactor coolant system is to pump the concentrated boric acid solution (8700 ppm boron, minimus) into the makeup tank using either the 10 gpm boric acid pumps or the 30 gpm reclaimed boric acid pumps. Using only one of the two 10 gpm boric acid pumps, the required volume can be injected in less than 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />. The alternate method l

of addition is to inject boric acid from the borated water storage tank using the makeup and purification pumps. The required 32,112 gallons of boric acid can be injected in less than four hours using only one of the makeup and purification pumps.

Concentration of baron in the boric acid mix tank or a reclaimed boric acid storage tank may be higher than the concentration which would crystallice at ambient conditionc. For this reason, the boric acid mix tank is provided with an immersion electric heating element and the reclaimed boric acid tanks are provided with low pressure steam heating jackets to maintain the tempera-ture of their contents well above (10 F or more) the crystallisation tempera-0 ture of the boric acid solution contained in them. Both types of heaters are controlled by temperature sensors i=mersed in the solution contained in the tanks.

Further, all piping, pumps and valves associated with the boric acid mix tank and the reclaimed boric acid storage tanks to transport boric acid solution from them to the makeup and purification system are provided with redundant electrical heat tracing co ensure that the boric acid solution vill be maintained 10 F or more above '.ts crystallisation temperature. The electri-cal heat tracing is controlled by the temperature of the external surfaces of the piping syste=s. Once in the makeup and purification system, the boric acid solution is sufficiently well mixed and diluted so that normal system tempera-tures assure boric acid solubility.

References (1) FSAR, Sections 9 1 and 9.2 (2) FSAR, Figure 6.2 (3) Technical Specification 3.3 3-20

f.

If c control rod in the regulating or axial power sheping groups is declared inoperable per Specification h.7 1.2.,

operation may continue provided the rods in the group are positioned such that the rod that was declared inoperable is maintained vithin allowable group average position limits of Specification b.7 1.2.

g.

If the inoperable red in Paragraph "e" above is in groups 5, o, 7, or 8, the other rods in the group =ay be tri==ed to the sa=e bosition.

Normal operation of 100 percent of the ther=al power allowable for the reactor coolant pump ccmbination may then continue provided that the rod that was declared inoperable is maintained within allevable group average position limits in 3525 3 5 2.3 The verth of single inserted control rods during criticality are limited by the restrictions of Specification 3.13 5 and the control Rod Position Limits defined in Specification 3 5 2.5 3 5 2.h Quadrant tilt:

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

+3 525 as determined using the full incore detector system, b.

When the full incore detector systen is not available and except for physics tests quadrant tilt shall not exceed +190%

as dete ined using the mini =us incore detector systes.

When neither incere detector syste= above is available and c.

except for physics tests quadrant tilt shall not exceed +1.99 as deter =ined using the power range channels displayed on the console for each quadrant (out of core detector system),

d.

Except for physics tests if quadrant tilt exceeds the tilt limit pover shall be reduced ic=ediately to below the power level cutoff (see Figures 3 5-2A, and 3.5-23.

Moreover, the power level cutoff value shall be reduced 2 percent for each 1 percent tilt in excess of the tilt limit.

For less than four pu=p operation, thermal power shall be reduced 2 percent of the ther=al power allovable for the reactor coolant pump combination for each 1 percent tilt in excess of the tilt limit.

Within a period of k ht uts, the quadrant power tilt shall be e.

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

1.

The protection cystem t eactor power / imbalance envelope trip setpoint.s shall be reduced 2 p-ercent in power for each 1 percent tilt.

r 3-3h 0

5

2.

The control rod group withdrawal limits (Figures 3 5-2A, 3.5-23, 3.5-2c, 3.5-2D, and 3.5-2H, shall be reduced 2 percent in power for each 1 percent tilt in excess of the tilt limit.

3.

The operational irbalance limits (Figures 3 5-2E, and 3 5-2F) 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 +16.807, 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 +14.2d determined using the out of core detector system (OCT) when neither the FIT nor MIT are available, the reactor vill be placed in the hot shutdown condition. Diagnostic testing dur-ing power operation with a quadrant tilt is permitted provided that the thermal power allovable is restricted as stated in 3 5 2.h.d above.

g.

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

(

3-3ka i

4 3.4.2.5 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 and axial power shaping control rods.

Except for physics tests or exercising control rods, the regulating control rod insertion /vithdrawal limits are specified on Figures 3 5-2A, and 3 5-23 for four pump operation and Figures 3.5-2c and 3.5-2D three or two pu=p operation. Also excepting physics tests or exercising control rods, the axial power shaping control rod insertion /vithdrawal limits are specified on Figure 3.5-2H.

If any of these control rod position limits are exceeded, corrective measures shall be taken Lanediately to achieve an acceptable control rod position.

Acceptable control rod positions shall be attained within four hours.

c.

Except for physics tests, power shall not be increased above the power level cutoff of 92 percent of rated thermal power unless one of the following conditions is satisfied:

1.

Xenon reactivity never deviated more than 10 percent from the equilibrium value for operation at 100 percent of rated thermal power.

2.

Xenon reactivity deviated more than 10 percent and is nov vithin 10 percent of the equilibrium value for operation at 100 percent of rated thermal power and asymptotically approaching stability.

3 Except for Xenon free startup (when 3.5.2.5.c.2 applies) the reactor has operated within a range of 87 to 92 percent of rated thermal power for a period exceeding 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in the soluble poison control = ode.

d.

Core imbalance shall be monitored on a minimum frequency of once every two hours during power operation above h0 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 Figure 3.5-22.

If the imbalance is not within the envelope defined by Figure 3 5-2E, corrective measures shall be taken to achieve unacceptable imbalance.

If an acceptable inbalance is not achieved within four hours, reactor power shall be reduced until imbalance limits are met, e.

Safety rod limits are given in 3.'. 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.

3-35

3 5 2.7 A power map shall be taken at intervals not to exceed 30 effective full power days using the incere instrumentation detection system to terify the pcVer distribution is within the limits shown in Figure 3 5-20.

Bases The pover-imbalance envelope defined in Figure 3.5-2E is based on LOCA analyses which have defined the maximum linear heat rate (see Figure 3.5-2G) such that the maximum clad temperature vill not exceed the Final Acceptsnee C;1teria (2200F). Operation outside of the power imbalance envelope alone does not constitute a situation that vould.cause the Final Acceptance Criteria to be exceeded should a LOCA occur. The power tsbalance envelope l

represents the boundary of operation linited by the Final Acceptance Criteria only if the control rods are at the withdrawal / insertion it=its as defined by Figures 3 5-2A, 3-5-23, 3.5-20, 3.5-2D, 3.5-2H, and if quadrant tilt is at tuu limit. 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 1

i e.

Postulated fuel rod bow effects 4

The Rod index versus Allowable Power curves of Figures 3.5-2A, 3.5-23, 3.5-20, 3.5-2D, and 3 5-2H 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 4,

four hours is not considered a violation of a limiting condition for operation. The limiting criteria within the Restricted Region are potential ejected rod worth and ECCS power peaking and since the probability of these accidents is very lov especially in a h hour time frame, inadvertant 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.

i l

l 3-35a L

I-

- - - - - mnetcJerlGGMcy;racp betvxn successivo control rod groups ic allowed cinco tha v;rth cf a rod is lovr et th2 upp2r and 1cvar part cf tha ctroka.

Control rods ara arranged in groups or banks de ' 'ned as fellcvat Groun Function 1

1 Safety 2

Safety 3

Safety b

Safety 5

Regulating

,1 6

Regulating 7

8 Regulating (Xenon transient override)

APSR (axia3. power shaping bank)

Control rod groups are withdrawn in sequence beginning with group 1.

j 5, 6 and 7 are overlapped 25 percent.

Groups group 7 to be partially inserted.

The nor=al position at power is.for De red position limits are based on the =ost li=iting of the following three 1

criteria: ICCS power peaking, shutdovn nargin, and potential ejected rod vorth.

is ensured by the rod position linits.As discussed above, compliance with the The =inimum available rod verth, by reactor trip at any ti=e, assuning the highest vorth control vithdrawn re=Ains in the full out position (1).

The red position W its also ensure that inserted red groups vill not contain single rod vortha greater them: 0.65% Ak/k at rated power.

the safety scalysis (2) of the hypothetical rod ejection accident.These values have single inserted control rod vorth of 1.0% Ak/k is allowed by the rod positii:

A maximum 7d-its at hot cero power.

4 A single inserted control rod worth 1.0$Ak/k at beginning of life, hot, zero pover vould result in a love:r transient peak themal power and, therefore, less severe envrion= ental consequences than 0.65% Ak/k ejected red vorth at rated power.

Le plant co=puter vill scan for tilt and imbalance end vill satisf/ the technical specificat; ion require =ents.

If the ec=puter is out of service, than

=t.nual calculation for tilt above 15 percent power and imbalance above h0 percent power t:uat be perfomed at least every two hours until the ec=puter is retur=ed to service.

(

ne 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 s

of +h.92% Which is equivalent to a +3 52% tilt measured with the full incere instru=entation with measure =ent uncertainties included.

l During the physics testing program, the high flux trip setpoints are admini-s ratively set as follovs to assure an additional safety margin is provided:

Test Power Trip Setoofnt

)

0

<5%

15 50%

7h l.

h0 50%

50 50%

75 855 9

>75 105.5%

3-36 t

l i

270.9.102 100

'I

l 90 NOT ALLOWED RES1RICTED 248.2.80 80 g

70

=

./

orn 60 130,50 140,50 50 Ey 40 PERMISSISLE 30 20 70,15 80,15 10 0

5,4 0

25 56 75 100 125 150 175 200 225 250 275 300 Rod index, 5 Withdrawn D

M M

M m

I i

t i

1 10b #

0 25 50 75 a

e I

a 1

0 25 50 75 100 Group 6 R00 POSITION LIMITS FOR 4 PUMP OPERATION FROM 0 TO 125 1 5 EFPD TMI-I Group 5 Figure 3.5-2A e

t 210.9.102 250.102 100 210.0.02

~

248.2.80 00

~

NOT ALLOWED 10 E

g@

b 60 0

==

50 190,5 200.50 d

j 40 PERMIS$1BLE 30 20 140.1 150.15 10 0.0 0,

i i

I t

t t

0 e

i 25 50 15 100 125 150 115 M

M

\\

Rod index. 5 Withdrawn 0

25 p0

,M I,"

Grcup i q0

,0 25 50 15 R00 POSITION LlulTS FOR 4 PullP OPERAllR O

25 50 15 100 "E

FROM 125 1 5 EFPD 10 EOC 1,

t I

I UI-I Group 5 F6gure 3.5-2

t 180,102 190,102 248.2.102 RESTRICTEO 100

- FOR 3 PUMP 90 80 NOT ALLONED 3

70 160,70 f

60 d

130,50 140,50 50

~o 40 PERMISSIBLE h

30 h

20 70,15 80,15 10 0

O 25 50 75 100 125 150 175 200 225 25G 275 300 i

Pod index, 5 Iltnaraan 0

25 50 75 100

,a e

i i

e I

Group 7 0

25 50 75 100 R00 POSITION LIMITS FOR 2 1 3 PUMP O

25 50 75 100 Group 6 OPERATl0N FROM 0 TO 125 1 5 EFPO i

a a

e i

TMI-l Group 5 Figure 3.5-2C e,

f 250,102 260,102 100 90 80 NOT ALLOWE0 70

?

A gg d

4 190,50 200,50 50 se t.

40 s

E a.

30 20 PERMISSIBLE 140,15 150,15 10 0

i 5,0,

e i

i j

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod indes, 5 Witndrawn

)

0 25 50 75 100 i

s i

a e

i I

Group 7 O

25 50 75 100 R0D POSITION LIMITS FOR 2 & 3 PUMP Group 6 OPERATION FROM 125 + 5 EFPD TO EOC 0

25 50 75 100

~

l ing.;

i i

i I

1 g

Tigure 3.5-20

\\

Power, % of 2535 NWt RESTRICTED REGION

- l 10

-15,102 10.0,102

-15,92 10.0,92 o

90 80 <> 10.0,80

-20,80<

70 PERMISSIBLE OPERATING 60 REGION 50 40

-- 30 20

__ 10 l

I I

1 I

I I

I

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Power imbalance, %

l i

1 l

POWER IMSALANCE ENVELOPE FOR OPERATION FRCM 0 EFFD TO E0C Figure 3.5-2E

Figure 3 5-2F Deleted

d N

Q 21 20 1g 5

18

[h

\\

E 17

/

/

h e

/

K is 5

/

n g

15 a

1' I

s_

j 13 12 0

2 4

6 8

10 12 i

Axial location of Peak Power Front Bottom of Care. f t v

LOCA LIMITED MAXINU.*4 ALLD7ABLE LINEAR HEAT RATE - TMI-!

v Figura 3.5-2G 0{

J

t

,,6.I,102 45.0,102 RESTRICTED 100 REGION 90 6.I,92 45.0,92 57.9,80 0,80 80 70 100.70 g

1 W

60 E

o 50 x

PERMISSIBLE d

OPERATING

[

40 REGION 30 20 10 I

i 1

1 I

f

?

I O

10 20 30 40 50 60 70 80 90 100 APSR, % Withdrawn l

l APSR POSITION LlHITS FOR OPERATION F3GM 0 EFFD TO ECC I

Figure 3.5-2E

i, e

December 1978 THREE MILE ISLAND UNIT 1 CYCLE 5 RELOAD REPORT I

7

i CONTENTS Page 1.

INTRODUCTION AND

SUMMARY

1-1

+

2.

OPERATING HISTORY 2-1 3.

GENERAL DESCRIPTION 3-1 4.

FUEL SYSTEM DESIGN 4-1 4.1.

Fuel Assembly Mechanical Design 4-1 4.2.

Fuel Rod Design.

4-1 4.2.1.

Cladding Collapse 4-1 4.2.2.

Cladding Stress 4-1 4.2.3.

Cladding Strain 4-2 4.3.

Thermal Design 4-2 4.4.

Material Design 4-2 4.5.

Operating Experience 4-2 5.

NUCLEAR DESIGN 5-1 5.1.

Physics Characteristics 5-1 5.2.

Analytical Input 5-2 5.3.

Changes in Nuclear Design 5-2 6.

THERMAL-HYDRAULIC DESIGN 6-1 7.

ACCIDENT AND TRANSIENT ANALYSIS 7-1 7.1.

General Safety Analysis 7-1 i

7.2.

Accident Evaluation 7-1 8.

PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS.

8-1 9.

STARTUP PROGRAM - PHYSICS TESTING 9-1 9.1.

Precritical Tests 9-1 9.1.1.

Control Rod Trip Test 9-1 l

9.1.2.

RC Flow 9-1 9.1.3.

RC Flow Coastdown 9-1 9.2.

Zero Power Physics Tests 9-2 9.2.1.

Critical Boron Concentration 9-2 9.2.2.

Temperature Reactivity Coef ficient 9-2 l

9.2.3.

Control Rod Group Reactivity Worth. 2 l

9.2.4.

Ejected Control Rod Reactivity Worth 9-3 l

9.3.

Power Escalation Tests 9-3 t

t l

l I

- ii -

CONTENTS (Cont'd)

Page 9.3.1.

Core Power Distribution Verification at approxi-mately 40, 75, and 100% FP With Nominal Control Rod Position 9-3 9.3.2.

Incore Vs Excore Detector Imbalance Correlation Verification at% 40% FP 9-5 9.3.3.

Temperature Reactivity Coef ficient at% 100% FP.

9-5 9.3.4.

Power Doppler Reactivity Coef ficient at% 100% FP 9-5 9.4 Procedure for Use When Acceptance Criteria Are Not Met 9-6 10.

REFERENCES 10-1 List of Tables Table 4-1.

Fue? Design Parameters and Dimensions 4-3 4-2.

Fuel Thermal Analysis Parameters 4-4 5-1.

TMI-1, Cycle 5 Physics Parameters 5-3 5-2.

Shutdown Margin Calculation for TMI-1, Cycle 5 5-5 6-1.

Thermal-Hydraulic Design Conditions 6-2 7-1.

Comparison of Key Parameters for Accident Analysis 7-3

'7-2.

Bounding Values for Allowable LOCA Peak Linear Heat Rates 7-3 List of Figures Figure 3-1.

Core Loading Diagram for TMI-1, Cycle 5 3-2 3-2.

Enrichment and Burnup Distribution for TMI-1, Cycle 5.

3-3 3-3.

Control Rod Locations and Group Designations for TMI-1, cycle 5 3-4 5.1.

BOC (4 EFPD), Cycle 5 Two-Dimensional Relative Power Distribu-tion - Full Power, Equilibrium Xenon, APSRs Inserted 5-6 8.1.

TMI-1 Core Protection Safety Limits 8-2 8-2.

Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance 8-3 8-3.

Rod Position Limits for Four-Pump Operation From 0 to 125 5

EFPD, TMI-l 8-4 8-4.

Rod Position Limits for Four-Pump Operation from 125 5 EFPD to EOC, TMI-l 8-5 8-5.

Rod Position Limits for Two-and Three-Pump Operation From 0 to 125 5 EFPD, TMI-l.

8-6

- iii -

m

..w w--

-1, Figu'res (Cont'd) 1 Page Figure-Rod Position Limits' for Two-and Three-Pump Operation From 8-6.

8-7 125 5 EFPD to EOC,-TMI-l 8-8 Power Imbalance Envelope for Operation From 0 EFPD to EOC.

j 8-7.

8-9 APSR Position Limits for Operation From 0 EFFD to EOC 8-8.

9 9

4 i

J l

I T

- iv -

9 1

1.

INTRODUCTION AND

SUMMARY

This report justifies the operation of the Three Mile Island Nuclear Station Unit 1 (Cycle 5) at a rated core power of 2535 MWt.

Included are the required analyses as outlined in the USNRC document, " Guidance for Proposed License Amendments Relating to Refueling," June 1975.

To support Cycle 5 operation of the TMI-l nuclear station, this report employs analytical techniques and de-sign bases established in reports that have been submitted and received tech-nical approval by the USNRC (see References).

Cycle 5 reactor parameters that are related to power capability are summarized in this report and referenced to Cycle 4.

All the accidents analyzed in the FSAR have been reviewed for Cycle 5 operation.

In all cases Cycle 5 charac-teristics are bounded by those analyzed for previous cycles, therefore no new accident analyses were performed.

The Technical Specifications have been reviewed, and the modifications required for Cycle 5 operation are justified in this report.

Based on the analyses per-formed, 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 5 can be safely operateo at the rated core power level of 2535 MWt.

Because of performance anomalies observed at other B&W plants, arifice rod as-semblies will not be used in TMI-1, Cycle 5.

The removal of orifice rod as-semblies has been accounted for in the analyses performed for Cycle 5.

In ad-dition, retainer assemblies will be installed over the two regenerative neutron sources to provide positive retention during reactor operation.

1 1-1

1 2.

OPERATING HISTORY The reference cycle for the nuclear and theenal-hydraulic analyses of the Three Mile Island, Unit 1 plant is the operating Cycle 4.

Cycle 4 achieved criticality on April 28, 1978, and af ter zero power testing began power opera-tion on May 2, 1978.

Cycle 4 is scheduled for completion in February 1979 after 265' 15 EFPD.

No operating anomalies occurred during any previous cycle that would adversely af fect fuel performance during Cycle 5.

The operation of Cycle 5 is scheduled to begin in March 1979.

The design cycle length is 265 15 EFPD.

2-1 m-n n

n.,

3.

' GENERAL DESCRIPTION The TMI-l reactor core is described in detail in Section 3 of the Final Safety Analysis Report for the unit.

The Cycle 5 core consists of 177 fuel assem-blies (FAs), each of which is a 15-by-15 array containing 208 fuel rods, 16 control rod guide tubes, and one incore instrument guide tube.

The undensi-fied nominal active lengths of the fuel rods are 142.6 inches for Batch 4B and 142.25 inches for Batches 5, 6, and 7.

All fuel assemblies in Cycle 5 maintain a constant nominal initial fuel loading of 463.6 kg of uranium.

The cladding is cold-worked Zircaloy-4 with an OD of 0.430 inch and a wall thickness of 0.0265 inch.

The fuel consists of dished-end, cylindrical pellets of uranium dioxide (see Table 4-1 for data).

Figure 3-1 is the core loading diagram for TMI-1, Cycle 5.

The initial en-richment of Batch 4B was 2.64 wt % uranium-235.

Batches 5,6, and 7 have an enrichment of 2.85 wt % uranium-235.

Thirty-one Batch 4 assemblies will be discharged at the end of Cycle 4, and the remaining 25 Batch 4, and all Batch 5 and 6 and assemblies will be shuffled to new locations.

The Batch 7 assemblies will occupy the periphery of the core.

Note that the designation 4B is used to identify the Batch 4 assemblies being reused for Cycle 5; Batch 4A is the remainder of the Batch 4 assemblies that are being discharged at the end of Cycle 4.

Figure 3-2 is an eighth-core map showing the burnup of each assembly at the beginning of Cycle 5 and its initial enrichment.

Like Cycle 4, Cycle 5 will be operated in a rods-out, feed-and-b leed mode.

The l

l core reactivity control will be supplied mainly by soluble boron and supple-l l

mented by 61 full length, Ag-In-Cd control rod assemblies (CRAs).

In addition to the full-length control rods, eight axial power shaping rods (APSRs) are provided for additional control of axial power distribution. The Cycle 5 lo-cations of the 69 control rods and the group designations are indicated in Fig-ure 3-3.

The core locations of the 69 control rods for Cycle 5 are identical to those of the reference Cycle 4.

The nominal system pressure is 2200 psia, and the core average densified nominal linear heat rate is 5.72 kW/ft at the rated core power of 2535 MWt, as in Cycle 4 3-1

Figure 3-1.

Core Loading Diagram for TMI-1, Cycle 5 1

2 3

4 5

6

'7' 8

9 10 11 12 13 14 15 A

y y

y y

y D7 P10 D9 B

y y

y F

F F

4B 6

4B C

F6 N3 E6 K1 D8 K15 E10 N13 F10 p~.

p 5

6 4B 6

5 6

4B 6

5 013 C7 L1 DS P6 D11 L15 C9 03 D

F F

F F

4B 5

6 5

6 5

6 5

4B C12 G3 G7 D6 N2 B8 N14 D10 G9 G13 C4 E

F F

6 5

5 5

6 4B 6

5 5

5 6

FS A10 F4 Pil E7 R8 E9 P5 F12 A6 Fil F

F F

y p

4B 6

5 6

5 6

5 6

5 6

4B G4 A9 E4 B12 G5 M14 F8 M2 Gil B4 E12 A7 G12 G

y p

4B 6

5 6

5 6

5 6

5 6

5 6

4B F14 H4 L14 H2 H15 H6 N12 H10 H1 R14 F2 H12 L2 H

F 7

6 5

6 4B 6

5 4B 5

6 4B 6

5 6

K4 R9 M4 P12 KS.

E14 L8 E2 K11 P4 M12 R7 K12 K

y 7

4B 6

5 6

5 6

5 6

5 6

5 6

4B L5 R10 L4 Bil M7 A8 M9 B5 L12 R6 Lil L

y y

F F

4B 6

5 6

5 6

5 6

5 6

4B 6

8 4

0 K9 U3 04 p

M F

6 5

5 5

6 4B 6

5 5

5 6

Cl3 07 F1 N5 B10 N11 FIS 09 C3 F

F 4B 5

6 5

6 5

6 5

4B L6 D3 M6 G1 N8 G15 M10 D13 L10 0

F F

5 6

4B 6

5 6

4B 6

5 N7 B6 N9 P

y y

y y

y y

4B 6

4B l

R F

F F

F F

l l

l Note: All fresh fuel (F) is Batch 7.

xx Cycle 4 location x

Batch number 3-2 l

Figure 3-2.

Enrichmsnt and Burnup Distribution for TMI-1, Cycle 5 8

9 10 11 12 13 14 15 2.64 2.85 2.85 2.64 2.85 2.85 2.85 2.85 H

24,996 18,351 7,245 24,136 10,142 17,789 10,142 0

2.85 2.85 2.85 2.85 2.85 2.64 2.85 K

8,638 21,056 6,204 18,597 7,223 24,312 0

2.85 2.85 2.85 2.64 2.85 2.85 8,641 16,045 6,233 24,172 0

0 L

2.85 2.85 2.85 2.85 18,357 15,603 9,723 0

l l

2.64 2.85 2.85 N

21,157 0

0 2.85 0

16,002

'P i

l R

i x.xx Initial enrichment xx,xxx BOC burnup, Mwd /mtU 3-3

Figure 3-3.

Control Rod Locations and Group Dasignations for TMI-1, Cycle 5 X

l A

8 3

7 3

C 1

6 6

1 0

7 8

5 8

7 E

1 5

2 2

5 1

F 3

8 7

5 7

8 3

G 6

2 4

4 2

6 H

7 5

5 3

5 5

7

~

K 6

2 4

4 2

6 L

3 8

7 5

7 8

3 M

1 5

2 2

5 1

N 7

8 5

8 7

0 1

6 6

1 P

3 7

3 l

R 1

Z 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 Group No. of rods Function 1

8 Safety l

2 8

Safety X

Group Numoer 3

9 Safety l

4 4

Safety 5

12 Control 6

8 Control 7

12 Control 8

8 APSRs total # 69 3-4

4 4

E j

4.

FUEL SYSTEM DESIGN 4.1.

Fuel Assembly Mechanical Design The types of fuel assemblies and pertinent fuel design parameters for TMI-1, Cycle 5 are listed in Table 4-1.

All fuel assemblies are identical in con-cept and are mechanically interchangeable.

Retainer assemblies will be used on the two fuel assemblies containing the regenerative neutron sources (RNS).

The justification for the design and use of the retainers described in Refer-ence 8 is applicable to the RNS retainer s in Cycle 5 o f TMI-1.

All results, references, and identified conservatisms presented in Section 4.1 of the Cycle 4 Reload Report are applicable to the Cycle 5 reload core.

Twenty-five Batch 4B Mark-B4 assemblies remain in the core for their fourth cycle of irradiation. However, total end of cycle burnup and residence time for these assemblies are within the conservative design limits for Mark-B4 fuel assemblies.

4.2 Fuel Rod Design The mechanical evaluation of the fuel rod is discussed below.

4.2.1.

Cladding Collapse Creep collapse analyses were performed for four-cycle assembly power histories.

Because of its longer previous incore exposure time, the Batch 4B fuel is more limiting than the other batches. The Batch 4B assembly power histories were analyzed and the most limiting assembly was determined.

The power history for the most limiting assembly was used to calculate the fast neutron flux level for the energy range above 1 MeV.

The collapse time for the LLniting assembly was conservatively determined to be more than 30,000 most EFPH (ef fective full power hours), which is greater than the maximum projected residence time (Table 4-1).

The creep collapse analysis was performed based on the conditions set forth in References 3 and 4.

t 4-1 1

4.2.2.

Cladding Stress The Cycle 5 stress parameters are enveloped by a generic 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 total stresses (primary plus secondary) must be less than the minimum specified unirradiated yield strength.

In all cases, the margin is in excess of 30%.

Th e following conservatisms with respect to TMI-1 fuel were used in the analysis:

1.

Lowest post-densification internal pressure.

2.

Lowest initial pellet density.

3.

Highest system pressure.

4.

Highest thermal gradient across the cladding.

The stresses reported in Reference 4 for Core 1 fuel, which alto provide sufficient margins, were calculated using even greater conservatisms than those used for the generic stress analysis.

This furcher substantiates that eladding stress in Cycle 5 is.not a concern.

4.2.3.

Cladding Strain The fuel design criteria specify a limit of 1.0% on cladding circuoferential plastic strain.

The pellet design is established for plastic cladding strain of less than 1% at values of maximum design local pellet burnup and heat gen-eration race', which are considerably higher than the values the Cycle 5 fuel is expected to see.

The strain analysis is also based on the maximum Specifica-tion value for the fuel pellet diameter and density and the lowest perm.tted tolerance for the cladding ID.

4.3.

Thermal Design The incoming Batch 7 fuel is thermally and geometrically similar to the Batch 6 fuel of Cycle 4.

There fo re, the description in Re ference 2 is appli-cable.

Significant thermal parameters are shown on Table 4-2 based on assumed densification to 96.5% TD.

4-2 J

The TAFY fuel pin analysis performed for Batch 6 is applicable to Batch 7.

A thermal analysis was also done for Batches 6 and 7 using the fuel performance code TACO. Where dif ferences occurred between corresponding calculated values of each code the more conservative values were chosed for Batch 7 design.

The average fuel temperature shown in Table 4-2 for Batches 6 and 7 were taken from an average pin analysis using the TACO code.

The value shown represents the BOL (100 mwd /mtU) average fuel temperature at 5.80 kW/ft.

The average temperature decreases with burnup to a value of 1120F at 38,000 mwd /mtU. The average fuel temperatures shown for Batches -4 and 5 were taken from a BOL TAFY code analysis. The linear heat rate capabilities based on centerline melt are from TAFY results for all batches. The TACO analysis for Batches 6 & 7 gave a higher LHR capability.

4.4.

Material Design The chemical compatibility of all possible fuel-cladding-coolant-assembly in-teractions for the Batch 7 fuel assembies are identical to those of the pres-ent fuel.

4.5.

Operating Experience Babcock & Wilcox operating experience with the Mark B, 15x15 fuel assembly has verified the adequacy of its design. As of July 31, 1978, the following experience has been accumulated for the eight operating B&W 177-fuel assembly plants using the Mark B fuel assembly:

Maximum assembly Cummulative net Current burnup, mwd /mtU elec t rical Reactor cycle Incore Discharged output, HWh Oconee 1 4

30,400 25,300 22,624,262 Oconee 2 3

31,000 26,800 17,792,292 Oconee 3 4

17,300 29,400 18,370,115 TMI-1 4

26,610 32,200 20,133,775 ANO 1 3

27,100 28,300 16,399,947 Rancho Seco 2

25,789 17,170 12,067,322 Crystal River 3 1

10,700 4 O'3,690 Davis-Besse 1 1

4,000 1,734,732 4-3

Table 4-1.

Fuel Design Parameters and Dimensions Thrice-Twice-Once-Fresh burned assys burned assys burned assys fuel assys Batch 4 Batch 5 Batch 6 Batch 7 Fuel assembly type Mark B4 Mark B4 Mark B4 Mark B4 No. of assemblies 75 48 52 52 Fuel rod OD, in.

i.430 0.430 0.430 0.430 Fuel rod ID, in.

C.377 0.377 0.377 0.377 Flexible spacers, type Spt?ng Spring Spring Spring Rigid spacers, type Zr-4 Zr-4 Zr-4 Zr-4 Undensified active fuel length, in.

142.6 142.25 142.25 142.25 Fuel pellet OD (mean specified), in.

0.3700 0.3695 0.

5 0.3695 Fuel pellet initial density, % TD 93.5 94.0 94.

94.0 Initi fuel enrichment we%g}5U 2.64 2.85 2.85 2.85 Burnup (BOC) mwd /mtU 23,762 17,758 7,967 0

Cladding collapse time, EFPH

>30,000

>30,000

>30,000

>30,000 Estimated residence t ime, EFPH (max) 26,402 26,978 26,808 26,880 Table 4-2.

Fuel Thermal Analysis Parameters

(#

Densified fuel parameters Batch 4 Batch 5 Batch 6 Batch 7 Pellet diameter, in.

0.3643 0.3646 0.3646 0.3646 Fuel stack height, in.

140.46 140.47 140.47 140.47 Nominal LHR at 2568 MWt, kW/ft 5.80 5.80 5.80 5.80 Avg fuel temp at nominal LHR, F (BOL) 1320 1315 1400 1400 LHR to C fuel mett, kW/ft.

20.15 20.15 20.15 20.15 (a) Densification to 96.5% TD assumed.

4-4

5.

NUCLEAR DESIGN 5.1.

Physics Characteristics Table 5-1 compares the core physics parameters of Cycles 4'and 5; these values were generated using PDQO7 for both cycles.

Since the core has not yet reached an equilibrium cycle, differences in core physics parameters are to be expected between the cycles.

In Table 5-1 the average cycle burnup is slightly hignec i.: Cycle 5 than in Cy-cle 4 because of the difference in analyzed cycle lengths.

The higher accumu-lated average core burnup at the end of Cycle 5 is mainly due te the use of Batch 4B fuel for a fourth cycle.

Batch 4B has an average bdenst of 31,380 mwd /

mtU and a peak assembly burnup of 32,387 mwd /mtU at the end of ():le 5.

Figure 5-1 illustrates a representative relative power distribution fo r the beginning of Cycle 5 at full power with equilibrian xenon and Group 8 inserted.

The critical boron concentrations are approximately the same as those of refer-ence Cycle 4.

The hot, full power control rod worths are similar in both cycles. Control rod worths are sufficient to maintain the required shutdown margin as indicated in Table 5-2.

The differences in the parameters between Cycles 4 and 5 are due to changes in radial flux distribution, isotopics, and the difference in analyzed cycle lengths.

The ejected rod worths in Table 5-1 are the maximum calculated values.

It is dif ficult to compare values between cycles since the isotopic distributions are not identical.

Calculated ejected rod worths and their adherence to criteria are considered at all times in life and at all power levels in the development of the rod position limits presented in Section 8.

The maximum stuck rod worths for Cycle 5 are similar to those for the reference Cycle 4.

The following conservatisms were applied for the shutdown calculations:

1.

Poison material depletion allowance.

2.

10% uncertainty on net rod worth.

3.

Flux redistribution penalty.

5-1

Flux redistribution was accounted for since the shutdown analysis was calcu-Laced using a two-dimensional model.

The shutdown calculation was analyzed at 280 EFPD.

The Cycle 5 power deficits from hot zero power to hot full power are higher than those for Cycle 4 due to the more negative moderator coefficients in Cy-cle 5.

The differential boron and xenon worths are similar for Cycles 4 and 5.

The effective delayed neutron fractions for Cycle 5 show a decrease with burnup as they did in Cycle 4.

5.2 Analytical Input The Cycle 5 incore measurement calculation constants used for computing core power distributions were prepared in the same manner as for Cycle 4.

5.3 Changes in Nuclear Design There were no relevant changes in nuclear design between the reference and reload cycles. The same calculational methods and design information were used to obtain the important nuclear design parameters.

Additional calcula-tions were performed for soluble boron control, shutdown, reactivity control, and radiation analyses reflecting the mode of operation.

As in Cycle 4, both APSRA and CRA position limits, as well as power imbalance limits, are specified based on LOCA analyses.

These operational limits and the RPS limits for Cycle 5 are presented in Section 8 (Technical Specification changes).

i 5-2 l

6 g

Table 5-1.

TMI-1, Cycle 5 Physics Parameters (*

Cycle 4(b)

Cycle 5(

Cycle length, EFPD 277 280 Cycle burnup, mwd /mtU 8557 8650 Average core burnup - EOC, mwd /mtU 18,165 19,162 Initial core loading, mtU 82.1 82.1 Criticy) boron-BOC, ppm (noXe)

H2P

, group 8 (37.5% wd) 1226 1255 HFP, group 8 inserted 1045 1064 Critical boron - EOC, ppm (eq Xe)

HZP, group 8 (37.5% wd, eq Xe) 285 311 HFP, group 8 (37.5% wd, eq Xe) 16 29 Control rod worths - HFP, BOC, % A k/k Group 6 1.12 0.98 Group 7 1.48 1.44 Group 8 (37.5% wd) 0.46 0.45 Control rod worths - HFP, EOC,

% Ak/k Group 7 1.57 1.56 Group 8 (37.5% wd) 0.50 0.50 Max ejected rod worth -HZP, % Ak/k

• BOC 0.81 0.65 EOC 0.81 0.71 Max stuck rod worth - HZP, % Ak/k BOC 2.01 2.24 EOC 2.06 2.03 Power deficit, HZP to HFP, % Ak/k BOC

-1.29

-1.34 EOC

-2.05

-2.06 Doppler coeff - BOC, 10-5 ( Ak/k/ F)

N-100% power (0 Xe)

-1.48

-1.47 Doppler coeff - EOC, 10-5 ( Ak/k/ F) 100% power (eq Xe)

-1.60

-1.58 Moderator coeff - HFP, 10 ' (A k/k/ F)

~

BOC (0 Xe, 1064 ppm, group 8 ins)

-0.71

-0.77 EOC (eq Xe, 17 ppm, group 8 ins)

-2.53

-2.63 Boron worth - HFP, ppm /% A k/k BOC (1150 ppm) 104 105 EOC (17 ppm) 94 93 Xenon worth - HFP, % Ak/k BOC (4 EFPD) 2.65 2.64 EOC (equil.)

2.74 2.73 5-3

Table 5-1.

(Cont'd)

Cycle 4(b)

Cycle 5(

Effective delayed neutron fraction - HFP BOC 0.00583 0.00583 EOC 0.00520 0.00517 (a) Cycle 5 data are for the conditions stated in this report.

The Cycle 4 core conditions are identified in Reference 2.

(b) Based on 287 EFPD at 2535 MWt, Cycle 3.

(c) HZP denotes hot zero power (532F T

)

• • E

HFP denotes hot full power (579F T

).

(d) 277 EFPD in Cycle 4; 280 EFPD in Cycle 5.

(e) Ejected rod worth Groups 5 through 8 inserted.

(f) Based on 277 EFPD at 2535 MWt, Cycle 4.

1 5-4 1

J

l Shutdown Margin Calculation for TML-1, Cycle 5 Table 5-2.

80c, %A k/k EOC, " % A k/k_

Ave.ilable Rod Worth 8.91 8.72 Total rod worth, HZP( )

-0.51 Worth reduction due to burnup of

-0.50

-2.03 poison material, HZP

~2.24 Msximum stuck rod, HZP 5.98 6.37

-0.64 Nst worth, HZP

-0.60 Lass 10% uncertainty 5.38 5.37 Total available worth, HZP Raquired Rod Worth _

2,06 1.34 Power de ficit, HFP to HZP 0.42 0.34 Ms:x allowable inserted rod worth, HZP 1.15_

_p.59 Flux redistribution, HFP to HZP 3.63 2.27 Total required worth, HZP Shutdown Margin 2.10 3.11 Total available minus total required, HZP Required shutdown margin is 1.00% A k/k.

Note:

full (a) 280 EFPD.

(b) HZP denotes hot zero power (532F T "8 ; HFP denotes hot

)

).

power (579F T,y 5-5

Figure 5-1.

BOC (4 EFPD), Cycle 5 Two-Dimansional Ralstive Power Distribution - Full Power, Equilibrium Xenon, APSRs Inserted

• 8 9

10 11 12 13 14 15 H

0.86 1.06 1.26 0.94 1.13 0.99 0.99 0.82 K

1.22 1.07 1.25 1.01 1.12 0.81 0.80 8

L 1.23 1.08 1.05 0.87 1.17 0.69 l

M 0.97 1.00 1.14 1.04 l

N 0.89 1.15 0.73 0

0.57.

P R

l l

l l

x Inserted rod group No.

J x.xx Relative power density

• Calculated results from two-dimensional pin-by pin PDQO7.

5-6 i

6.

THERMAL-HYDRAULIC DESIGN The incoming Batch 7 fuel is hydraulically and geometrically similar to the fuel remaining in the core from previous cycles. The thermal-hydraulic de-sign evaluation supporting Cycle 5 operation used the methods and models described in References 1, 2, and 4 with two exceptions.

The exceptions are flow and reference design radial x local peaking used in the the core bypass analysis.

Fuel assemblies not containing control rods or neutron sources usually have orifice rod assemblies (ORAs) installed in the guide tubes to minimize core anomalous mechanical behavior of the BPRA and CRA latching bypass flow. Recent mechanisms in operating plants made if prudent to remove the ORAs.

For Cycle 5 operation, all ORAs will be emoved. Therefore, with 69 control rods and a total of 106 locations will be vacant.

two regenerative neutron sources, The maximum core bypass flow used for Cycle 4 analysis was 8.34% based on 44 ORAs removed. For Cycle 5 with 106 vacant fuel assemblies, a maximum core bypass flow of 10.4% was used. The two RNS assemblies will have retainers The installa-installed above them to ensure retention of the source clusters.

tion of the RNS with retainers has insignificant impact on thermal-hydraulic analysis.

To offset the ef fect of the increased core bypass flow in Cycle 5 on the ther-mal-hydraulic design, the reference design radial x local peaking f actor (Edh) has been reduced from 1.78 to 1.71.

This reduction in Fah is fully supported by the Cycle 5 nuclear design, for which the maximum predicted radial x local peaking f actor is 1.403.

Reactor core safety limits have been re-evaluated The removal of based on the reduced FAh and the increased core bypass flow.

all ORAs in conjunction with a reduction of Fah to 1.71 has been previously approved in Reference 10.

The Cycle 4 and 5 maximum design conditions and significant parareters are shown in Table 6-1.

The potential ef fect of fuel rod bow on DNBR was considered by incorporating suitable margins into DNB limited core safety limits and RPS setpoints.

The from the of rod bow was calculated according to the NRC guilelines amount following equation:

AC 7- = 0.065 + 0.001449 vtu o

where AC = rod bow magnitude, mils, C, = initial gap (138 mils),

BU = maximum assembly burnup, mwd /mtU.

6-1 we

An 11.2% rod bow penalty based on an assumed burnup of 33,000 mwd /mtU is ap-plied to all analyses that define plant operating limits and to design tran-sients.

No fuel assembly will achieve a burnup as high as 33,000 mwd /mtU d" ring Cycle 5 operation.

A thermal margin credit equivalent to 1% DNBR to of fset the rod bow penalty has been used as a result of the flow area (pitch) reduction factor included in all thermal-hydraulic analyses.

The 1% DNBR is the only credit applied to offset the the rod bow penalty.

Table 6-1.

Thermal-Hydraulic Design Conditions i

Cycles 2, FSAR 3 and 4 Cycle 5 Power level, NWt 2568 2568 2568 System pressure, psia 2200 2200 2200 Reactor coolant flow,

% design flow 100.0 106.5 106.5 Vessel inlet coolant temp (100% power), F 554.0 555.6 555.6 Vessel outlet coolant temp (100% power), F 603.8 602.4 602.4 Ref design radial-local power peaking factor 1.78 1.78 1.71 Ref design axial flux shape 1.5 cosine 1.5 cosine 1.5 cosine Active fuel length, in.

141.12 Table 4-2 Table 4-2 Avg heat flug (100% power),

175(,)

175(,)

10 Btu /h-ft 171.5 Ma3 eat flug (100% power),

h 10 Btu /h-ft 458 467 450 CHF correlation W-3 BAW-2 BAW-2 Hot channel factors Enthalpy rise 1.011 1.011 1.011 Heat flux 1.014 1.014 1.014 Flow area 0.98 0.98 0.98 Minimum DNBR (densified fuel, 112% power) 1.46(b) 1.88 1.98 i

(a) Based on densified length.

(b).114% power i

6-2

7.

ACCIDENT AND TRANSIENT ANALYSIS 7.1.

General Safety Analysis Each FSAR accident analysis has been examined with respect to changes in Cycle 5 parameters to determine the effect of the Cycle 5 reload and to ensure thermal performance during hypothetical transients is not degraded.

th at The effects of fuel densification on the FSAR accident results have been evalu-ated and are reported in Reference 4.

Since the new Batch 7 reload fuel as-t emblies contain fuel rods with a theoretical density higher than those con-sidered in Reference 4, the conclusions in that reference are still valid.

The dose evaluations in the FSAR were based on conservative values for fuel burnup and power peaking; however, improved fuel utilization and improved calculational methods have led to a higher plutonium-to-uranium fiasion ratio.

Since plutonium has a higher iodine fission yield than uranium, more iodine activity will be produced and thuse the thyroid doses will be slightly higher than reported in the FSAR.

Generally, the plutonium fission yield for noble gases is lower than for uranium.

Thus, less noble gases will be produced, causing the whole body doses to be lower than reported in the FSAR.

A comparison study has been performed to determine the change in the thyroid doses for the major accidents in Chapter 14 of the FSAR that would result from the Cycle 5 iodine inventory.

The results show that, although the thy-roid doses for Cycle 5 increase by 6 to 19% over the FSAR, the Cycle 5 doses are still only a very small fraction of 10 CFR 100 limits.

7.2.

Accident Evaluation The key parameters that have the greatest effect on determining the outcome of a transient can typically be classified in three major areas:

core thermal parameters, thermal-hydraulic parameters, and kinetics parameters, including the reactivity feedback coefficients and control rod worths.

7-1

analysis were design oper-Care thermal properties used in the FSAR accident Core thermal cting values based on calculational values plus uncertainties.

in Table 4-2.

A comparison p rameters for the batches used in Cycle 5 are given i

FSAR and reference of the cycle 5 thermal-hydraulic maximum design condit ons toThese par Cycle 4 values is presented in Table 6-1. A comparison of the key kinetics ths accidents considered in this report.

7-1.

pcrameters fran the FSAR and Cycle 5 is provided in Table A lowererd loop A generic LOCA analysis has been performed for the B&W 177-F l

this study NSS using the Final Acceptance Criteria ECCS evaluation mode ;

This ana*ysis is generic in nature since the is reported in Reference 7.

i ory were used.

limiting values of the key parameters for all plants in th s categ function Furthermore, the combination of the average fuel temperature as a CA limits rate and the lifetime pin pressure data used in the LO d

is conservative compared to those calculated for this reloa.

of linear heat 7

7 and substantiated analysis Thus, the analysis and the LOCA limits reported in Referenceion of TMI-1, Batch by Reference 9 provide conservative esults for the operatk linear Table 7-2 shows the bounding values for allowable LOCA pea 7 fuel.

rates for TMI-1, Batch 7.

thermal properties heat is concluded by examination of the proposed Cycle 5 coreious cycle values tha It l

and kinetics properties with respect to acceptab e prev to safely operate the the abili'_)

this core reload will not adversely af fectConsidering the previously acc TMI-1 plant during Cycle 5.

luation of Cycle i

used in the FSAR and subsequent cycles, the trans ent eva The initial 5 is considered to be bounded by previously accepted analyses.d by the conditions of the transients in Cycle 5 are bounde and/or subsequent cycle analyses.

Fuel Densification Report 7-2 f

)

Table 7-1.

Comparison of Key Parameters for Accident Analysis FSAR and densif'n Cycle 5 Parameter report value Predicted value

-5 Doppler coef f (BOC), ik/k/ F

-1.17 x 10

-1.47 x 10 5

-5

~

Doppler coeff (EOC), Ak/k/ F

-1.33 x 10

-1.58 x 10 5 Moderator coeff (BOC), Ak/k/ F

+0.5 x 10 '

-0.77 x 10 4 Moderator coeff (EOC), Ak/k/ F

-3.0 x 10 '

-2.63 x 10 4 All rod group worth, (HZP) % Ak/k 10.0 8.72 Initial boron conc, (HFP) ppm 1200 1064 Boron reactivity worth (70 F),

75 74 ppm /1% Ak/k Max ejected rod worth, (HFP) % Ak/k 0.65 0.25 Dropped rod worth (HFP), % Ak/k 0.46 0.20 Table 7-2.

Bounding Values for Allowable LOCA Peak Linear Heat Rates Core Allowable peak linear elevation, ft heat rate, kW/ft 2

15.5 4

16.6 6

18.0 8

17.0 10 16.0 7-3 1

I

.l l

8.

PROPOSED MODIFICATIONS TO TECHNICAL SPECIFICATIONS i

The following Technical Specification figures have been developed with the intent of bounding future cycles, and are bounding for Cycle 5 operation.

1 i

i Y

4 I

1 1

i.

4 1

1 i

r t

8-1 4

er

.'t

,-w.---

---~r-n-

--.v

-e-

--e

,w,w

a Figure 8-1.

TMI-l Core Protection Safety Limits

$R Thermal Power Level, %

- - 120

(-29,112)

(112)

_l (30,112)

ACCEPTABLE

. ;,=,,,.

i=

(47.i.,e.s>

( 4 6.4,91.5) 90 (87.2)2 (25.8,87.2)

(-25.4,87.2 ACCEPTABLE

~~

384 PUNP OPERATION 70 (46.4,69.2)

(-46,64. 2)

( -24. 4,59. 6 ) -

60 (59.6) 3 (24.7,59.6)

ACCEPTABLE 50

~~

2,3&4 PUMP OPERAT104 (47.5,46. G}

(-48. I,36.6 )

30 20 10 l

i e

i i

i i

i

! 50 40 20 -10 0

10 20 30 40 50 60 Reactor Power imbalance, %

Curve Reactor Coolant Flow (1b/hr) 6 l

139.8 x 10 s

2 104.5 x lo 6

3 68.8 x 10 l

8-2 l

l Figure 8-2.

Trip Setpoint for Nuclear Overpower Based on RCS Flow and Axial Power Imbalance Theresi Power Level, 5 120 110 (108) (17,108)

(-17,108) l l

100 l

El

• I 28 ACCEPTABLE E2 = -1.0 4 PUuP

-8 OPER4 TION (35,90)

(-35,85) l (80.7) su

/i ACCEPTA8LE 70 3 & 4 PUNP

/ OPERATION (35,62.7)

(-35,57.7)

/

l

-- 60 l

l (53.1) l

- 50 I

I ACCEPTABLE 40 2,3 & 4 PUMP (35,35.1)

OPERATION

. 30

(-35,3 0.1 )

l I

I o

-- 20 s

la

=,

7 10 W

W

- 10

-l l'

u u

I E

I iI i

i i

i i

i 60 50

-40 20

-10 0

10 20 30 40 50 60 70 l

Reactor Power imoalance, 5 j

8-3 i

l

Rod Position Limits for Four-Pump Operation Figure 8-3.

From 0 to 125 +5 EFPD, TMI-1 1

270.9.102 180.102 100 270.9,92 l

90 RESTRICTED 248.2,80 NOT AI.LOWE0 80 70 g=

60 130,50 140,50

~a 50 g

40 PERNISSIBLE 3g 20 70.15 80,15 10 i-i e

i 5.4 0

25 50 75 100 125 150 175 200 225 250 275 300 0

Rod index, 5 Witnarawn 0 25 50 75 100 I

I i

j t

10 I "'

0 25 50 75 1

f a

e e

Group 6 0

25 50 75 100 i

i I

f f

Group 5 8-4

Figure 8-4.

Rod Position Limits for Four-Pump Operation From 125 +5 EFPD to EOC, TMI-l 250,l'02 100

, 270.9,92 90 248.2,80 80 NOT ALLOWE0 70 ga ws k

80 190,5 200.50 50 IZ 40 PERulSSIBLE 30 20 140.1 150.15 10 f

0.0 0,

0 I

0 25 50 75 100 125 150 175 200 225 250 275 300 i

Rod index, 5 litndrawn 0

2,5 50

,75 1,00 t

e 0

25 50 75 100 8

I I

i e

0 25 50 75 100 Group 6 Group 5 8-5 l

l r

.o e

Figure 8-5.

Rod Position Limits for Two-and Three-Pump Operation From 0 to 125 +5 EFPD, TMI-l 180,102 190.102 248.2.102 l00 RESTRICTED FOR 3 90 pump 80 NOT ALLOWED 2

70 160,70 E

60

=

h

'50 119,50 140,50 40 PERulSSIBLE i

1

,4 20 20 70,15 80,15 10 10 g

0 25 50 75 100 125 150 175 200 225 250 275 300 Rod inder, 5 Withdrawn 0

25 50 75 100 I

t I

t i

Group 7 0

25 50 75 100 i

f f

I e

0 25 50 75 100 Group 6 E

t t

t l

Group 5 8-6

Figure 8-6.

Rod Position Limits for Two-and Three-Pump Opera-tion From 125 +5 EFPD to EOC, TMI-l 250,102 260,102 90 80 3

NOT ALLOWE0 E

7,

• 2 3

60 5

4 190,50 200,50 50 e

ic 40 g

.a 30 PERillSSIBLE 20 140,15 150.15 10 5,0,

0 0

25 50 75 100 125 150 175 200 225 250,

275' 300 Roa Inder, 5 Witnaraon 0

25 50 75 100 l

I I

I I

Group 7 0

25 50 75 100 i

f I

I i

Group 6 0

25 50 75 100 m

I t

l I

Group 5 8-7

i Figure 8-7.

Power Imbalance Envelope for Operation From 0 EFPD to EOC Power, % of 2535 MWt RESTRICTED REGION

- l 10

-15,102 10.0,102

-15,92

<,10.0,92 90

-20,80 <

80 < >

10.0,80 70 PERNISSIBLE OPERATING

- 60 REGION 50 40 30 20 10 i

i i

e i

i i

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Power imbalance, %

8-8 l

e Figure 8-8.

APSR Position Limits for Operation From 0 EFPD to EOC

'6.I,102 45.0,102 o

100 RESTRICTED REGION

,o

,s.i,92 45.0,92 0,80 7.9,80 80

~

100,70 m

W 60 N

o 50 y

PERMISSIBLE t

OPERATING

)

j 40 REGION 1

30 20 10 I

I I

I I

f f

I O

10 20

.0 40 50 60 70 80 90 100 APSR,7. Withdrawn 8-9

me

=

e 9.

STARTUP PROGRAM - PHYSICS TESTING The planned startup test program associated with core performance is outlined below.

These tests verify that core performance is within the assumptions of the safety analysis and provide confirmation for continued safe operation of the unit.

9.1.

Precritical Tests 9.1.1.

Control Rod frip Test Precritical control rod drop times are recorded for all control rods at hot fu11 flow conditions before zero power physics testing begins.

Acceptable criteria state that the rod drop time from fully withdrawn to 75% inserted shall be less than 1.66 seconds at the conditions above.

l It should be noted that safety analysis calculations are based on a rod drop time of 1.40 seconds from full withdrawn to two-thirds inserted.

Since the accurate position indication is obtained from the zone reference switch most the 75%-inserted position, this position is used instead of the two-thirds at inserted position for data gathering.

The acceptance criterion of 1.40 seconds corrected to a 75%-inserted position (by rod insertion versus time correlation) is 1.66 seconds.

9.1.2.

RC Flow Reactor coolant ficw with four RC pumps running will be measured at hot shut-down conditions. Acceptance criteria require that the measured flow be within allowable limits.

This test is planned for Cycle 5 only to verify flow perfor-mance without ORAs.

9.1.3.

RC Flow Coastdown The coast down of RC flow from the tripping of the most limiting RC pump combi-

. nation from four pumps running will be measured at hot zero power conditions.

The coastdown of RC flow versus time s ill then be compared to the required RC l

f l

9-1 l

i

a flow varsus time to datormina whethar ccesptsnca criteria are met.

This test is planned for Cycle 5 only to verify flow performance without ORAs.

9.2.

Zero' Power Physics Tests 9.2.1.

Critical Boron Concentration Criticality is obained by deboration at a constant dilution rate.

Once crit-icality is achieved, equilibrium baron is obtained and. the critical boron concentration determined.

The critical boron concentration is calculated by correcting for any rod withdrawal required in achieving equilibrium boron.

The acceptance criterion placed on critical boron concentration is that the actual boron concentration must be within t100 ppm boron of the predicted value.

9.2.2.

Temperature Reactivity Coefficient The isothermal temperature coefficient is measured at approximately the all-rods-out configuration and at the hot zero power rod insertion limit.

The average coolant temperature is varied by first decreasing then increasing temperature by 5 F.

During the change in temperature, reactivity feedback is compensated by discrete change in rod motion, the char.ge in reactivity is then calculated by the summe. tion of reactivity (abtained from reactivity calcula-tion on a strip chart-recorder) associated with the temperature change. Acceptance criteria state that the measured value shall not differ from the predicted value by more than t0.4 x 10~

(Ak/k)/ F (predicted value obtained from Physics Test Manual curves).

The moderator coefficient of reactivity is calculated in conjunction with the temperature coef ficient measurement.

Af ter the temperature coef ficient has been measured, a predicted value of fuel Doppler coef ficient of reactivity is subtracted to obtain moderator coef ficient.

This value must not be in excess of the acceptance criteria limit of +0.5 x 10-4 (Ak/k)/ F.

9.2.3.

Control Rod Group Reactivity Worth Control bank group reactivity worths (Groups 5, 6, and 7) are measured at hot zero power conditions using the boron / rod swap method.

The boron / rod swap method consists of establishing a deboration rate in the reactor coolant sys-tem and compensating for the reactivity changes of this deboration by inserting control rod Groups 7, 6, and 5 in incremental steps.

The reactivity changes that occur during these measurements are calculated based on Reactimeter data, and I

differential rod worths are obtained from the measured reactivity worth versus i

9-2 l

The following tecepttnce critoria era pitecd on the 40% FP test:

1.

The worst-case maximum linear heat rate must be less than the LOCA limit.

2.

The minimum DNBR must be greater than 1.30.

3.

The value obtained from the extrapolation of the minimum DNBR 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 from the extrapolation of the worst-case maximum linear heat rate to the next power plateau overpower trip setpoint must be less than the fuel melt limit or the extrapolated value of imbalance must fall outside the RPS power / imbalance / flow trip envelope.

5.

The quadrant power tilt shall not exceed the limits specified in the Technical Specifications.

6.

The highest measured and predicted radial peaks shall be within the follow-ing limits:

predicted value - measured value measured value

-<8 x 100 7.

The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value - measured value measured value

-< 12 x 100 Items 1, 2, 5, 6, and 7 above are established to verify core nuclear and ther-mal calculational models, thereby verifying the acceptability of data from these models for input to safety evaluations.

Items 3 and 4 establish the criteria whereby escalation to the next power plateau may be accomplished without exceeding the safety limits specified by the safety analysis with regard to DNBR and linear heat rate.

The power distribution tests performed at 75 and 100% FP are identical to the 40% FP test except that core equilibrium xenon is established prior to the 75 and 100% FP tests.

Accordingly, the 75 and 100% FP measured peak acceptance criteria are as follows:

9-4

a the chtngs in rod group position. The differential rod worths of esch of the controlling groups are then summed to obtain integral rod group worths.

Th e acceptance criteria for the control bank group worths are as follows:

I 1.

Individual bank 5, 6, 7 worth:

predicted value - measured value measured value

--< 15 x 100 2.

Sum of Groups 5, 6, and 7:

predicted value - measured value measured value

--< 10 x 100 9.2.4.

Ejected Control Rod Reactivity Worth Af ter the CRA groups have been positioned near the minimum rod insertion limit, j

the ejected rod is borated to 100% withdrawn and the worth obtained by adding the incremental changes in reactivity by boration.

Af ter the ejected rod has been borated to 100% withdrawn and equilibrium boron established, the ejected rod is then swapped in versus the controlling rod group and the worth determined by the change in the previously calibrated controlling rod group position.

The boron swap and rod swap values are averaged and error-adjusted to determine ejected rod worth.

Acceptance criteria for the ejected rod worth test are as follows:

1.

predicted value - measured value (error-adj) x 100

--< 20 measured value (error-adj) 2.

Measured value (error-adjusted) < 1.0% A /k k

The predicted ejected rod worth is given in the Physics Test Manual.

9.3.

Power Escalation Tests 9.3.1.

Core Power Distribution Verification at approximately 40, 75, and 100% FP With Nominal Control Rod Position l

Core power distribution tests are performed at approximately 40, 75, and 100%

full powe r (FP ). The tes t at 40% FP is essentially a check on power distribution l

in the core to identify any abnormalities before escalating to the 75% FP plateau.

Rod index is established at a nominal full power rod configuration at which the core power distribution was calculated.

APSR position is established to provide a core power imbalance correspoc?i 3 to the imbalance at which the l

core power distribution calculations were performed.

9-3

1.

Tha highsst meccured tnd predi+.ted rcdial peaks shall be within the follow-e ing limits:

predicted value - measured value measured value

--<5 x 100 2.

The highest measured and predicted total peaks shall be within the follow-ing limits:

predicted value - measured value measured value

--< 7.5 x 100 9.3.2.

Incore vs Excore Detector Imbalance Correlation Verification at N 40% FP Imbalances are set up in the core by control rod positioning.

Imbalances are read simultaneously on the incore detectors and excore power range detectors for various imbalances.

The excore detector of fset vs. incore detector offset slope must be at least 1.25.

However, this slope criterion may be reduced to 1.15 if the test is repeated at 75% FP after a suitable period of operation at 100% FP.

If the excore detector offset vs. incore detector of fset slope criterion is not met, gain amplifiers on the excore detector signal processing equipment are adjusted to provide the required gain.

9.3.3.

Temperature Reactivity Coefficient at N 100% FP The average reactor coolant temperature is decreased and then increased by about 5 F at constant reactor powe r.

The reactivity associated with each temperature change is obtained from the change in the controlling rod group position.

Controlling rod group worth is measured by the fast insert / withdraw method.

The temperature reactivity coefficient is calculated from the measured changes in reactivity and temperature.

Acceptance criteria state that the moderator temperature coef ficient shell be negative.

9.3.4.

Power Doppler Reactivity Coefficient at % 100% FP Reactor power is decreased and then increased by about 5% FP.

The reactivity change is obtained from the change in controlling rod group position. Control rod group worth is measured using the fast insert / withdraw method.

Reactivity corrections are made for changes in xenon and reactor coolant temperature that occur during the measurement.

The power Doppler reactivity coefficient is i

1.

9-5

e r

e=

e calculated from the measured reactivity change, adjusted as stated above, and the measured power change.

The predicted value of the power Doppler reactivity coefficient is given in the Physics Test Manual. Acceptance criteria state that the measured value shall be more negative than -0.55 x 10 (ak/k)/% FP.

9.4.

Procedure for Use When Acceptance Criteri* Are Not Met If acceptance criteria for any test are not met, an evaluation is performed j.

before the test program is continued.

This evaluation is performed by site test personnel with participation by Babcock & Wilcox technical personnel as required.

Further specific actions depend on 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 per-forming detailed analyses of potential safety problems because or parameter deviation.

Power is not escalated until evaluation shows that plant safety will not be comprised by such escalation.

I 9-6 l

..m

es ; ~v 10.

REFERENCES g,,

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

2 Three Mile Island Unit 1, Cycle 4 Reload Report, BAW-1473, Rev 3, Babcock &

Wilcox, Lynchburg, Virginia, May 1978.

3 A.F.J. Eckert, et al., Program to Determine In-Reactor Perfo nance of B&W Fuels - Cladding Creep Collapse, BAW-10084, Rev 1, Babcock & Tilcox, Lynch-burg, Virginia, November 1976.

4 TMI-1 Fuel Densification Report, BAW-1389, Babcock & Wilcox, Lynchburg, Virginia, June 1973.

5 C. D. Morgan and H. S. Kao, TAFY - Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044, Babcock & Wilcox, Lynchburg, Virginia, May 1972.

6 R. H. Stoud t, et al., TACO - Fuel Performance Analysis, BAW-10087 Babcock & Wilcox, Lynchburg, Virginia, June 1976.

7 R. C. Jones, et al., ECCS Analysis of B&W's 177-FA Lowered Loop NSS, BAW-10103A, Rev 1, Babcock & Wilcox, Lynchburg, Virginia, July 1977.

8 BPRA Retainer Design Report, BAW-1496, Babcock & Wilcox, Lynchburg, Virginia, May 1978.

9 J. H. Taylor (B&W) to R. L. Baer (USNRC), Letter, "ECCS Analysis of B&W 177-FA Lowered-Loop NSS, July 8,1977.

10 SAFETY EVALUATION AND ENVIRONMENTAL IMPACT APPRAISAL BY THE OFFICE OF NUCLEAR REACTOR REGULATION, SUPPORTING AMENDMENT NO. 65 TO FACILITY OPERATING LICENSE NO. DPR-38, AMENDMENT NO. 65 TO FACILITY OPERATING LICENSE NO. DPR-47, AMENDMENT NO. 62 TO FACILITY OPERATING LICENSE NO. DPR-55 DUKE POWER COMPANY, l

OCONEE NUCLEAR STATION, UNIT NOS. 1, 2 AND 3, DOCKET NOS. 50-269, 50-270 AND l

50-287, OCTOBER 23, 1978.

i l

l l

I 10-1

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