ML19260A070

Tech Specs Change Request 30,Amend 2,to Amend DPR-50,App a to Conservatively Account for Maximum Effects of Fuel Rod Bowing Possibly Encountered During Cycle 2 Operation. Certificate of Svc Encl

ML19260A070
Person / Time
Site:	Three Mile Island
Issue date:	04/02/1976
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19260A067	List: ML19260A066 ML19260A070
References
NUDOCS 7910290694
Download: ML19260A070 (10)

Text

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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT CC:GANY -

AND PENUSYLVANIA ELECTRIC COMPANY TEREE MILE ISLAND NUCLEAR STATION UIIIT 1 '* LS..+}&

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Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Reauest No. 30 Amendment No. 2 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 ir . 'ided.

METROPOLITAN EDISON COMPANY By ) (

Vice Predhent-Generhion Sworn and subscribed to me this day of , 1976.

...,s.. o.. a ubis tuv.19. H79 Notary Putfic 1479 231 391029 M

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Metropolitan Edison Co. (Met-Ed)

Three Mile Island nuclear Station Unit 1 (TMI-1)

Operating License No. DPR-50 Docket No. 50-289 Technical Specification Cnange Reauest No. 30 Amendment No. 2 The licensee requests that the att' ached changed pages (2-2, figure 2.1-2, and figure 2.3-2) replace page 2-2 figure 2.1-2 and figure 2.3-2 of Change Request No. 30 A=endment #1.

Reason for' Amendment No. 2 This supplement to our previous request is necessary to conservatively account for the maximum effects of fuel rod boving which may be encountered during cycle 2 operation.

Safety Analysis Justifying Amendment The method of analysis and the effect of fuel rod bow are discussed in the attached B&W Fuel Rod Bov Evaluation. This evaluation has conservatively

- accounted for the effects of the maximum predicted fuel rod bow during Cycle 2 operation. Further, this change does not constitute an unreviewed safety question in that incorporation of rod bow penalties into the TMI-1 Technical Specifications imposes additional, more conservative limits than those required by the original TMI-1 FSAR analyses, and the ECCS analyses requirements of 10 CFR 50.46.

The attached changes to technical specifications incorporate fuel rod bow penalties and, when combined with those requested in our request of February 11, 1976, provide safe operating limits for continued operation of TMI-1 at rated power of 2535 MWt during Cycle 2.

1479 232

a conservative margin to DNB for all operating conditions. The difference between the actual core outlet pressure and the indicated reactor coolant syste= pressure has been considered in determining the core protection safety li=its. The difference in these two pressures is noninally 45 psi; however, only a 30 psi drop vas assumed in reducing the pressure trip set pointe to correspond to the elevated location where the pressure is actually measured.

The curve presented in Figure 2.1-1 represents the conditions at which a mini =u= DNBR of 1.3 is predicted for the maximu= possible ghernal power (112 percent) when the reactor coolant flow is 139.8 x 10+ lbs/h, which is less than the actual flow rate for four operating reactor coolant pumps. This curve is based on the following nuclear power peaking factors (2) with poten-tial fuel densification and fuel rod bowing effects; N N N F = 2.6T; F = 1 78; F = 1 50 q M z The 15 axial peaking factor associated with the cosine flux shape provides a lesser =argin to a DNBR of 13 than the 17 axial peaking tactor associated with a lover core flux distribution. For this reason the cosine flux shape

_- and the associated.F_ =-140 is more -limiting _ and thus the more conservative assu=ption.

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The 150 cosine axlar flux shape in conjunction with FAH = 178' define the reference design peaking condition in the core for operation at the maxi =um overpower. Once the reference peaking condition and the associated ther=al-hydraulic situation has been established for the hot channel, then all other combinations of axial flux shapes and their accompanying radials must result in a condition which vill not violate the previously established design criteria on DNBR. The flux shapes examined include a vide range of positive and negative offset for steady state and transient conditions.

These design li=it power peaking factors are the most restrictive calculated at full power for the range frc= all control rods fully withdrawn to maxi =u=

allovable control rod insertion, and form the core DUER design basis.

The curves of Figure 2.1-2 are based on the more restrictive of two ther=al li=its a .d include the effects of potential fuel densification and fuel rod boving;

a. The 13 DNER limit produced by a nuclear power peaking factor of FN = 2.67 of the combination of the radial peak, axial peak, and position of the axial peak that yields no less than a 13 DUER.

b. The co=bination of raciel and axial peak that prevents central fuel telting at the hot spot. The limit is 19.6 kW/ft.

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

The specified flow rates for curves 1, 2, and 3 of Figure 2.1-2 correspond to the expected =inimum flow rates with four pumps, three purps, and one pump in each loop, respectively.

1479 233 2-2

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Thermal Power Level, f, UNACCEPTABLE OPERATION .-120 (112) (+35.112)

(-33,I12) *

- 110 OI ACCEPTA8LE 4 PUHP

- 800 OPERATION Kw/ft Limit . Kw/f t Limi t

(-56,90) *- 90 (86.7) (+58.90)

(.33,86.7) (+35,86.7) h __ 80 ACCEPTABLE 3 & 4 PUMP OPERATION

._ 70 ,

( .56.64.7) (+58,64.7

(-33,59.1) __ 60 (59,;) (+35.59.I) -' '-

, . . . .* -: ;-. c -' ' ," T ~. :-!

ACCEPTA8LE _.

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- 50 2,3 3 4 PUNP - - -

(-46.47) OPERATION ( +47,48)

.- 40

.- 30

.. 20

.- 10 l l I I I t

-60 -40 -20 0 20 40 60 Reactor Power lobalance, 5 i.

CURVE REACTOR COOLANT FLOW (Ib/hr) .

6 I 139.8 x 10 6

2 104.5 x 10 3 65.8 x 10 UMIT l, CYCLE 2 CORE PROTECTION SAFETY LIMITS F i gu re 2.1 -2

Power Level, 7 UN ACCEPTABLE -- 120 OPERATION

(-20,108) .- 110 (108) (+21,108)

..100 +

g l +

N' e i

'/

ACCEPTABLE 90 iPp

// 4 PUNP .-

+N OPERATION l (80,7)

(-46,80) 80 l .. 70 ACCEPTABLE .

. . . . . _. 2.& 4-P. UMP ~

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-OPERATION,_,_.-- _

E ,0

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h- )

0'

(-46,52.7) -- 50 l l

.. 40 I I ACCEPTABLE

-- 30 2,3 & 4 PUMP ( +4 8',2 5. I )-

(-4s,2s.1) OPERATION 20 g 7 O N to

' E Y -- 10 +

11 11 11 l ~

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i i i i i t

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-50 -40 -30 -20 -10 0 10 20 30 40 50

Power imbalance, 7, 1479 235 UNIT I, CYCLE 2 PROTECTION SYSTEM HAXIHUM ALLOWABLE SET PolNTS Figure 2.3-2

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F  ; "D PDN EVA!J!ATION

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An evahation of the effects of fuel rod bowing onTMI-l cycle 2 operation h'as been performed. It is concluded from the results of this evaluation that modifications to the proposed Cycic 2 Technical Specification limits are required to maintain acceptabic fuci design limits with the increase in rcwer peaking associated with fuel rod bow. This report described the analyscs used to evaluate the effects of

, fuel rod bowing and the resulting modifications to the Cycle 2. Technical Specifica-tions.

IDCAL p0WER PEAKING ANALYSIS FOR FUEL PDD TN EFFECTS The tecnhiques used to analy e the local. power peaking effects of fuel rod bowing are similar to those currently used for fuel'densification analyses. Briefly .

these techniques employed the two-dim,nsional transport theory code Dar (discrete

! ordinates technique) to detenaine the clianges in local power in the center fuel rod as a function of the bowing (in both magnitude and amathal orientation) of the

[ surrcunding fuel rods. Analysis was perfomed to establish neutron group structure, quadrature order, mesh line spacing for rectangular rod representation, boundary i

i conditions and buffer zone characteristics surrounding the discrete rod array, and

! the validity of superposition techniques for subsequent Mante Carlo analysis. Results frca these studies support the use of 6 neutron grcups and S4 quadrature for the DUT calculations. The 5x5 fuel red array used 15 mesh lines per fuel cell and was surrcunded by a buffer region of homogeneous fuel. The cnalyses were perfomed for a range of peak magnitudes of deflection based on the predictions in table 1, i

! and for a=uthal orientations of the deflection in 45 increments.

This model was used to detenaine the % power change per rod as a function of bow rngnitude and angle for each type of rod location in the 5x5 array. This data fonned the input to a B5W Monte Carl'o code. .

In the Monte Carlo analysis, a normal distribution was assumed for the amount of bow and a unifona distribution for the angle of bow. For cach axial intenal L

j47g 2}6

(distance between spacer' 9-), the Monte Carlo code comp-e- th r. total spike on

', ' the central rod 100,000 times to develop a good statistical sample. This data is used to deter =ine power spikes at each axial interval in a minner similar to densification power spike analysis. The power spikes presented in table 1 were calculated such that less than one rod in the core will exceed the given power spike at a 95% confidence level.

EFFECT OF ROD BOW ON DNBR Analysis was performed with the COBRA III-C code to determine the effect of a fuel rod bowing into the hot channel and reducing the flow area of that channel.

The results demonstrate that rod bow of the magnitude predicted is adequately co=pensated for by the flow area reduction factor. Rod bow away from the hot channel was also analyzed. In this analysis, the effect s of a power spike was added'

' to the hot rod in the area of the minimum DNBR. This analysis also demonstrated that the current TliI-1, Cycle 2 DNBR results conservatively accc"nt for the effects of fuel rod bowing.

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TECHNICAL SPECIFICATION CIIANGES The imbalance limits for the reactor protection system are determined to preclude violation of the central fuel melt criteria (DNBR is less limiting in TMI-1, Cycle 2. To deter =ine these li=its, a correlation is developed which includes the following items:

A.

Power distribution effects (all in combination)

1. Fuel depletion

2. Power transients.

3. Full length control rod position (normal and abnormal)

4. Axial power shaping rod position (normal and abnormal)

5. Fuel dencification spikes B. Uncertainties (all in combination) }k7h 237

1. Hot channel faccors

2. Nuclear uncertainty

3. Spacer grid effect

4. Instrumentation (Observabil[tv, cicetronics and calibration).

. Including the power spike impact of rod bow in t he analysis results in the necessity to reduce the RPS inhalance limits (technical specifications 2.1 and

2. 3. )

The normal operations limits (technical specirication- 3.5.2) are determined to preclude violation of the ECCS linear heat rate criteria (See lapical Report BAW-10078 Rev.1, " Operational Parameters for B&W Rodded Plants", Proprietary, and BAW-10079, non proprietary), shutdown margin criteria, and ejected rod worth criteria. Evaluation of the impact of the rod bow power spikes on these limits indicates that the rod bow penalty (1.6%) is offset by the conservatism of the current limits. A densified average linear heat rate of 5.80 kw/f t was assumed in the analysis and a slightly more conservative power spike *han required *

(1.022 instead of 1.018 for instance, at the 30" elevation). The actual limiting average linear heat rate is 5.73 kw/ft at a power level of 2535 Mat. These conserv.tisms provide an excess margin slightly greater than the required 1.6%.

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, TABLE 1 FUEL ROD E0W POWER SPIKE PREDICTED FUEL ROD BOW

'AT OPERATI:G CONDITIONS Mid Span Elevation EOC 1 EOC 2 EOC 3 Top 1 5.55 6.45 7.53 2 6.66 '9.16 12.33 .

3 7.71 11.26 15.55 '

4 8.81 13.76 19.21 5 9.46 e 15.00 20.54 - ~

~6 9.23 14.01 19.'86 Bottem 7

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9.26 13.66 18.45 - 4

. POWER SPIRE RESULTING FROM PREDICTED FCEL ROD EOT ~

Ed Sp PERCENT POWER SPIRI Elevation EOC 1 EOC 2 EOC 3 Top 1 0.6 0.7 0.8

• 2 0.7 1.0 1.3 3 0.8 1.2 . 1.6 .

4~ 0.9 1.5 2.05 5 1.0 1.6 2.15 I 6 1.0 1.5 2.1 Bottes 7 1.0 1.5 1.7  :(

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U'IITED STATES OF AMERICA ,

k g .jg g y N' m NUCLEAR REGULATCRY COMMISSION M y IN THE MATTER OF 4

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DCCKET NO. 50-289 LICE:ISE NO. DPR-50 FMCPOLITAN EDISON CCIGANY ggp; Qg {

This is to certify that a copy of Technical Specification Change Request No. 30, Amendment No. 2 to Appendix A of the Operating License for Three Mile Island Nuclear Staticn Unit 1, dated April 2,1976 and filed with the U. S. Nuclear Regulatory Cor:sission on April 2,1976, has this 2nd day of April 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 fo11cvs:

Mr. Weldon B. Arehart, Vice-Chairman Mr. Harry B. Reese, Jr.

Scard of Supervisors of Board of County Conmissioners Londenderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 k m 0POLITAN EDISCN CC?GANY By Vice President-Generation 3SSI see 1479 240