Text

Report Entitled Palisades Core I Reanalysis ECCS Performance Results
	ML18347A784
Person / Time
Site:	Palisades
Issue date:	07/09/1975
From:	; Consumers Power Co
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML18347A784 (139)
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Consumers Power Co.

Rpt: Loss-of-Coolant Accident Analysis ~n conformance *with 10 CFR 50, App K.*..

(Rec'd w/ltr 7-9-75.*..........* #7364)

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I.

Palisades Core I Reanalysis ECC_S Performance Results Introd 1Jction and Summary On January4,.-1974, the Atomic Energy Commission* issued New Acceptance Criteria for Emergency Core Cooling Systems for L_ight-Water-Cooled Reactors(l).

The analysis presented-her~in d~monstrates that the Palisades Core I ECCS design satisfies these n*ew criteria.

This analysis has been performed using the approved CE large break evaluation model (3) including proposed modifications.(B)".

For comparison, the worst break calculations wer~ repeated without the probosed modifications.

The results of this analysis, which are presented in Section lI, cover primary system ruptures in the pump discharge leg larger than the pipe flow area, 4.909 ft2*

As-demonstrated in CENPD-137(2)-, smaller breaks are not limiting.

Therefore, a small break spectrum analysis has not been performed.

The total system flowra.te wasreduc~d from 130 x 106 lbs/hr to_l24 x 106 lbs/hr for this Core I. reanalysis. This cdolant flow ~as selected as the minimum which would be reached based on the la test measurements and includes a 11 owances for measurement error.

Using the proposed modifications to the CE large bre~k evaluation model, peak clad temperature calculatio'ns were performed for J:he entire spectrum of break sizes at a peak linear heat g~ner~tion rat~ (LHGR} of 11.0 kw/ft.

Sirice at thi.s LHGR the peak clad temperature for the worst break (l.O x DES/PD ) was

.only 2l35°F, this.case was rerun at a peak LHGR of 11.3 k~/ft.

This latter calculation produced a peak'clad temperature of 219a0i=, whi~h-is still below the criteria limit of 2200°F ;thus demonstrating that operation at a peak LHGR of 11.3 kw/ft is acceptable.

Usi~g the approved CE large break evaluation model.without' proposed modifications, the revised peak clad temperature results for the worst break (1.0 x DES/PD) did not alter the determi nation of the acceptab 1 e peak LHGR of 11. 3 kw/ft.

DES/PD = Double-Ended Slot at the Pump Discharge 1

l

The resuits~of this study supercede those reported. i.n *References 9 a.rid 10 and show that the plant meets the AEC.Acceptance Criteria published in the Federa1 Register on January 4, 1974~. Conformance is.summarized as follows:

C'riterion (l) Peak Clad Temperature.

11The ca1Culated maximum fuel element 0

Criterion. {2) cladding temperature shall not exceed 2200 F 11 Using the approved CE evalUation model with or without proposed.

modifications, the spectrum analysis yielded a peak clad temperature of 2198°F for the 1.0 x DES/PD break at a peak linear heat generation rate of 11.3 kw/ft.

. Maximum Cladding Oxi.dation.

11The calculated total *Oxidatian of the cladding shall nowhere exceed. 17% of the.total cladding thickness,before oxidation 11 With the proposed modification~*, the spectrum analysis yielded a local peak clad oxidation percentage of 2.67%.for the 0.6 x

DEG/PD break.

Without the proposed ~odifications* the 0.6 x DEG/PD break yielded ~ local peak clad oxidation percentage of 5.74%.

Criterion (3)

Maximum Hydrogen Generation.

11i"he calculated total amount of hydrogen generated from the che.mical ~eaction~ of the cladding with water 6r steam shal~ no~ exceed 1% of the hypothetical amount that would be generated if all of the metal in the cladding. cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react 11

With the proposed modifications, the 0~6 x DEG/PD break produced the highest core-wide oxidation which was <0.421%.

Without the proposed modifications, the 0.6 x DEG/PD break produced the highest core-wide oxidation which was <0.529%.

2

CriteriOn (4) Coolable Geometry.

11Calculated changes in core geometry shall be such that the core remains amenable. to coolingli.

Criterion (5)

The ~lad swellin9 and rupture model which is part of the CE Evaluation Mo.del (3) accounts for th.e*effects of changes in core geometry ff such changes are predicted to.occur.

With these core geometry changes, core cooling was enough

to 1 ower temperatures.

No.further rupture* can occur si nee the calculations were carried to the point a.t which the

. temperatures were decreasing'. Thus, a coo lab 1 e geometry has been maintained.

Long Term Cooling.

"After any calculated successful.

. initi~l o~eratiorl of the.ECC~~ the calculated core temperature shall* be maintained at an acceptably.low'. value. anc:I ~ecay heat shall be removed for the extended period of time required by

. the 1 ong-1 i ved ra.di oacti vi ty remaining in. the core 11 The spectrum analysis presented in this report sh~ws that the.

r~~id ~nsertion of borated watet from the ECCS will suitably limit the peak clad temper~ture and cool the cpre within a short period of.time. Subsequently, the safety injection pumps would supply cooling water* from the refueling water tank.to remove decay heat resulti.rig from the _lon~-.lived radioactivity remaining in the core.

When the refueling water tank is nearly empty, the safety i nj ecti on pumps would then be 1 i ned up to

. recirc~tate water. from the containment sum~;* 'rn this ~anner,

. the core would be co'oled for *a~ i.ndefinite periOd of time.

II.

Larg~ Break Analysis A.

Method-of Calculation The calculations reported in this section were performed using Combustion Engineering's large break evaluation mode1(3) with the following prop6sed modifications (B):

3

1.

The cont~inment wall noding technique has been revised in ord~r to provide a converged wall temperature solution.

2.

B~sed *on a recent review of steam~water mixing data, the res1~tahce across the ECCS injection section during the period after the safety injection ta~ks have empti~d has been ~avtsed:

In order to confirm that the model. changes did not affect the peak LHGR, two additional calculations were performed; the break producing the peak clad temperature (l.O x DES/PD) a.nd the. break with the highest local and core-wide oxidation percentages (0.6 x DEG/PD) were repeated using the approved model without the proposed modifications....

In this model the CEFLASH-4A(4) computer program i~ used to determine the primary system flow parameters during the bJowdown phas.e :and.the COMPERC-.II(S) computer program is used to describe the system behavior during the refi i'l and refl cod phases.

The core. fl ow ancl thermodynamic* parameters from these. two codes are used as input to the STRIKIN"'.II* program(G) w.hich calculate*s the hot rod clad* temperature transien_t.

The peak clad te1J1peratureand peak local clad oxidation percentage are therefore obtained,from.. the STRIKI~,.;II calculation.

The core-wide clad oxidation percentage is obta~ned from the results ~f both the STRIKIN-II and the COMZIRC(~, SUpplement l)computer programs.

B.

Emergency Core Cooling System Assumptions The Emergency Core Cooling System consists of three high.pressure pumps,

+

two low pres.sure pumps and four safety injection tanks.* Automatic operation of the pumps is. actuated by either a low-low pressurizer pressure signal or*

a high containment pressure signal.

Flow is initiated from the safety injection tanks when the cold leg pressure drops below 215 psia plus the elevation head.

Parameters pertinent to the calculati.on of the LOCA are presented.in Table II-1.

In performing the LOCA calculations, conservative assumptions are made concerning the availability of safety injection flow.

It is assumed that two high pressure pumps are operable at the time of the accident. Furthermore, it is assumed that off-site power is lost and all pumps must await diesel startup before they can begin to deliver flow.

(It is assumed, however, that off-site power is 4

available for the containment spray system}. Also, it is assumed that all safety injectlon flow delivered to the brok~n cold leg is lost.

An analysis of* the possible si:ngle failures that can occ.ur within the.ECCS has shown that the worstsingle failure for the la~9e break results is the faflure of one of the low pre~Sure pumps to start( '* Thus, only one low pressure pump is used in the c~rrent LOCA analysis for Palisades.

A review was made of the effects of a single failure or operator error that causes any manually-controlled, electrically-operated valve to move to a position that could adversely affect the ECCS.

In conformance with Branch Technical Position EICSB 1801\\plarit modificati.ons'and changes to the*.

technical speciffcations have. been initiated to protect agair:ist any loss of safety function caused by this type of failure.*

The above assumptions. lead to the. conclusion that the following safety ~njection flows are available:

75%of the flow from tw'o high pressure pumps (one of the three H.P. pumps is. not energized on SIAS}

75% of the flow from one low pressure pump Flow from three safety injection tanks

'*'J**

In the analyses repo~ted in this section,, nb credit is take~ for ~ump;flow unti 1 *the tanks are empty.

t. Core, System and.Containment Parameters The significant core and syste~ paramete~s_used in the larg~ break calculations are presented in Table U-1. *The peak linear heat rate was assumed to o6cur in ihe top 6f the core, the conservative location as identified in Sectioh IV.A~4 of Ref. 3. A conservative beginning~of-life moderator

(

-4 0 }

temperature coefficient +.5 x 10 6p/ F was used for all cases.

The gap conductance at the hot spot, as determined by the FATES computer program(?}, represents the minimum value for the remainder of the first cycle.

5.

1'...._.,

The study of peak clad temperature versus burnup presented in Ref. 3 shows that the peak clad temperature is maximized when the gap conductance is minimized.

Containment parameters.as presented in Tabl~ 11~2 ~re chosen to minimize containment. pressure such that a conservative determination of core reflood rate is made.

Pressure suppression equi~ment start-up times are selected

at their minimum values corresponding to off-site power being available.

D.

Break Spectrum In general 1 all possible break locations are considered in a LOCA analysis.

However, it was demonstrated in Reference 3 that ruptures in the cold 1 eg pump discharge location produce the highest clad temperatures. *This' is due to the minimization of core flow and reflood rate for this break location. Since core flow.is a function of the break size, the Palisades large break calculations J

have been performed for the cold leg pump discharge breaks for both guillotine and slot breaks over a range of break sizes. *This. range was selected in order to show that the peak clad temperature is maxi~ized with r~spect to break size.

E.

Results Table II-3 presents a listing of the large break sizes analyzed in this study along; with th~ figure n~mber presenting the pertinent transient data for each break.

As noted in Table II-3 the results for each of the breaks analyzed are displayed graphically in Figures 11.1 through ll.10. For each of the breaks shown in Figures II.l through II.B_the nine variables listed in Table II~4 are plotted as a -function of time~ For the break having the highest peak clad temperature (l~O x Doubl~-Ended Slot) the additional quantities listed in Table II-5 are also presented.

For Figure II.9 (1.0 x DES/PD) only those quantities which were affected by the removal of the proposed modifications are presented.

(Refer to Figure*II.l for typical behavior of the other variables).

Only the local clad oxidation percentage is presented for the 0.6 x DEG/PD break using the approved model without the proposed modifications (Figure II.10-0). (Refer to Figure 11.7 for typical behavior of the other variables).

6

P~ak clad ~emperature calculati6ns for the entire break spectrum are reported at a peak 1 inear *heat generation rate (LHGR) of 11.0 kw/ft~* The calculation for the 1.0 x DES/PD was repeated at a peak LHGR of 11.3 kw/ft.

The resulting peak clad temperature and local clad oxidation for this latter calculation are included in Figures 11.1-H and. 11.1-0, res~ective1y. This case was repeat~d usi~g the approved model without the proposed mcidifications at a peak LHGR of 11.3 kw/ft., and th~ results are shown in Figure 11.9.

Times of interest for the various breaks are shown *in Table Il-6 while Table II-7 summarizes peak clad temperatures and clad;oxidation percentages.

The STRIKIN-11 calculations ~ere run to 280 seconds with the proposed modifitati6ns and to 420 seconds without th~ modi fi cations, in *order to fully terminate the clad oxidation reaction~

The peak local clad oxidation results for a LHGR of 11 ~O kw/ft, given in.

Table II-7, show that the 0.6 x DEG/PD break had the highest local percentage.

To determine the maximum local clad oxidation percentage at the limiting LHGR c

of 11. 3 kw/ft, t.he ca lcul a ti on for the O. 6 x DEG/PD break was repeated. A special figure (II.7-0) was inserted into the r~sults for th~ 0.6 x DE~/PD break, showing the transient local clad oxidation percent~ge a~ the limitin~

LHGR.

This case.*was also repeated w*ithout the proposed modifications at_ a peak LHGR of 11.3 kw/ft, and the results shown in Figure II.10-0.

It should be noted. that the hot assembly regfon.:power in CEFLASH-4A was based on a peak LHGR of 11.91 kw/ft, thus conservatively allowing flexibility during the determination of the allowable peak LHGR based on the STRIKlN-II prediction of the hot rod thermal behavior. This procedure, however, leads to a

~

conservative core~wide clad oxidation calculation since the CEFLASH~4A hot.

assembly fuel and clad temperatures are used to initialize COMZIRC at the.

beginning of the reflood. Thus, the actual values for core-wide clad oxidation would be less than those reported in Table 11-7.

Figure 11.11 shows peak clad temperature plotted versus break size and type.

It is noticed that the worst break is the 1.0 x Double-Ende~ ~lot break, which has a peak clad temperature of 2198°F at 11.3 kw/ft.

7

Mass ~nd energy release to the containment during blowdown* is presented in Tables II-8 and II-9, for the cases with and wlthout proposed modifications.

respectively. Also shown in.these tables is the steam expulsion data during reflood.

The ECC water spillage and containment spray flow rates are presented graphicall~ in Figure II.12.

III. Computer Code Version Identification The following versions of the Combustion Engineering ECCS Evaluation Model computer codes were used for this analysis, including the 'proposed modifications:

'CEFLASH-4A:

Version No. 74329 STRIKIN-II: Version No. 7506,6 COMPERC-II:

Version No. 75097' COMZIRC Version No. 75055 For the calculations performed without the propqSed modifications:

CEFLASH~4A: Version No. 74329.

STRIKIN-II:

Version No. 75066 COMPERC-II:

Version No. 75055 COMZIRC Version No. 75055 8

IV..

References

1. Acceptance.Criteria for Emergency Core: Cooling Systems for Light-Water-Cooled Nuclear Power Rf!actc;>rs, Federal Register, Vcl. 39, No. 3 - Friday, January 4, 1974.

2.

CENPP-137, "Calculati9nMethods for the.CE Small Break LOCA Eva}..uation Model", Combustion Engineering Pr,oprietary R~port '*

August, 1974 (Proprietary).

3.

CENPD-132, "Calculative Methods for* the C-E.Large Break*LOCA Evaluation Model'i, August 1974 *. (Proprietary) *

4.

s.

CENPD-132, Supplement 1, "Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model", December i974.* (Ptoprietary).

CENPD-133; "CEFLASH-4A, A FORTRAN IV Digital Computer Program.for Reactor Blowdown Analysis", April 1974 (Prop):'ietary) ~.

CJ.!:NPD-133,. Supplement. 2,. "CEFLASII-4A*, A FORTRAN. IV Digital Computer Program for Reactor Blowdown Analysis (Modific~tion) ", December 197l*

(Proprietary).

CENPD-134, "COMPERC-II, A Program for Emergency Refill-Reflood, of the Core", April 1974 (Proprietary).

CENPD-134, Supplement 1, "COMPERC-II, *A Program for Emergency Refill-Reflood of the CoJ:'e (Modification)", December 1974 (Proprietary).

6 *. CENPD-135, "STRIKIN:;_iI, A Cylindrica~ Geometry Fuel Rod Heat Transfer Program, April 1974

(!"ropd.etary).

CENPD_::i3~, Supplement 2, "STRIKIN..;.Ir, A Cyl:l.nd~ical Geometry Fuel Rod Heat *Transfer Program (Modification)", December 1974 (Proprietary).

7.

CENPD-139; ncE Fuel Evaluation Model",* July 1974 *(Proprietary) *

8.

DP-606, Letter from F. M. Stern (CE) to Victor Stello, Jr.* (NRC) dated April 14, 1975, re:

Request for Modification of Approved C-E ECCS Model of November 27, 1974.

9

9.

Letter from Consumers Power Company (R. B. Sewell} to the AEC (L. M.

Muntzing} dated.October 21, 1974, 11 PaliSades Plants - Final Acceptance Cr'iteria".

10.

Letter from Consumers Power Company (R. B. Sewell) to the AEC (L. M.

11.

Muntzing} dated Dec. 16, 1974,

Subject:

Docket #50-255, License DPR-20, 11 Pali sades Pl ant - ECCS Reanalysis 11 Branch Technical Position EICSB 18, 11Application of the Single Failure Criterion to Manually-Controlled Electrically-Operated Valves 11 10

........ ~--..-.... -,. _ _..

Table II-1 Palisades cm*e I Genera1 System.Parameters*

Quantity.

Reactor Power Level (102% of NominalJ...

Average Linear Heat Rate (102% of Nom.inal)

Peak Linear Heat Rate Gap Conductance at Peak Linear Heat Rate Fuel Centerline Temperature at Peak Linear Heat Rate Fuel Average Temperature at Peak Linear Heat Rate Hot Rod Gas Pressure Mod era tor Tempera tu re Coefficient at

'*".litial Density

~stem Flow Rate (Total)

Core Fl ow Rate *.

Initial System Pressure *.

Core Inlet Temperature Core Outlet Temperature Active Core Height Fuel Rod OD Number of Cold Legs Number of Hot Legs Col 9 Leg 'oiameter Hot Leg Diameter

Safety.Injection Tank Press_ure

._,. fety Injection Tank Gas/Water Volume

Value

. 2244 4,73

.1L3 403.3

.* 3647 2617.*..

190

+o.5 x lo-4 s*

'124.0 x 10

. 120.3 x 106 1800 530 576

11. 0 0.4135 4 *'

2 30 42 215 908/1103 MWt

kw/ft kw/ft BTU/hr-ft2-°F.

psi a

/Sp/OF lbs/hr lbs/hr psi a OF 0 F Ft.

In..

In~

Iri.

psi a Ft3

Table II-2 Palisades Core I Containm~nt J>11ysical.Paraindters Net Free Volume Initiation Time for:

Spray Flow

. *Fan Cooler~; 3 Fans 4th Fan Containment Initia.l Conditions:..

Temperature Pressure Relative Humidity Containment* Spray Water':

Temperature Flow Rate (Tota~, 2 pumps)

Fan Air Cooler Capacity (per Cooler)

Vapor Temperature (°F) lleat Sinks:

40 104 184 244 283 1.64 x 106 ft 3 30.0 sec.

0.0 sec

27.5 sec.

90° F.

. 14.7 psia 50%

40° F1 3250 gptn Capacity (BTU/sec.)

0 700 9530 22900 34600

'!'ABLE II-2 (Cont'd) :

A.

Heat Sink I

Totai Mass (Lbm)

Total Surface Area (Ft2)

I. Tanks and Piping

2. Heating* and Ventilation Ducts Reactor'Crane

3. ~nternal Concrete Structures

4. Gratings Roof Trusses 351,000 80,000 315,000 11, 500, 000.

. 54,450 91, 600 19,332 20,072 13,946 55,845 8,880 12, 116

5. Corit~inmeri~ Dorne 7,270.

Lirter*:Plate **

73,000 Concrete 338,000

6. Containme1'1t Dome Base 11, 000 Liner Plate

. 110,750 Concrete 12,750,000

7. Contaj.rufient Wall Liner Plate Concrete J

8. Storage:Pool Floor Liner;Plate Concrete*

. 546, OOQ

.29 1 71~,ooo 6,200

._,470,000 -

.9. Shielded Internal Walls Liner Plate 26,650 Concrete

.2,420,000 54,400 821 3.635

10. Containment Base Slab

. 8, 229 Liner Plate 83,000 Concrete

.15, 350, 000 11 *. Biological Shield 2, 340 (ID)

Steel Lining 38, 300.

Insulation Concrete.*

8,580 3, 720,000...

Net Surf ace Area (Ft2) 20~072 13,946 9,401 8,880.

12, 116 7,270 11,000 50,600 821 3,635 8,229 2 1 340 (ID)

12. Strµctural, Suppoit Steel 57'*,000

'26,320

- 26,320 Thicb~ess (Inch~s)

.453

.100

, 2.35 3.3.0

.154

.1875

. 0.25 36.0 0.25 92.7.5

0.25 42.0

,003 0.187 5 45.8 0.18 75 53.4 0.25 149.0 0.406 4.00 90.0 S.40 *.

l

. TABLE II ~2 (Cont t.d)

B.

Minimum* for protective coat.ings with reference to the t,able on the previo.us paga:

I Itc:m 1.

Four safe :f.niect, aper floor in dome inorganic zinc Item 2.

Item 3.

3 M-DFT (mi.ls, Dry Film Thickness) with one coat inorganic_

white (Titanium Di.oxide) 1!2 M-DF1'.

Steam piping insulation with* aluminum cladding, no coating.*

H~V ductwork. no coating.

Reactor crane, one coat in-.

organic zinc* 3 M-DFT.

Concrete floor~. organic epoxy de-con tamable coating, 15 M-DFr.

Concrete walls six feet up. from floor., coating* 10 M-Dr""T.

Item 4.

Grating, galvanized, no coating.

Handrails, organic9 alkyd coat:l.ng, 3. M-DFT.

Item 5.

Item 6.

Roof trusses, contain dome and liner plate, inorganic zinc 3 M.:...DFI' with l~ M-DFr in-organic white.

The. liner plate* at.the *floor and ~ix feet above was finish

coated over the inorganic zinc with six mils DFT of an

. e.poxy organic coating.

All concrete,* except floor and six foot wainscot, no -coating.

Same as Item 5.

Item 7.

Same-as Item 6i Item 8.. Storage* pool stainless steel~ no coating.

Item 9.

Same as Item 6.

Item. 10.

Same as Item 5~

Item 11.

No coating.

Item 12.

Platforr11 columns I 'supports,. etc~ alkyd organic primer and finish coat 3 mils DFT.

In Addition:

Steam Generator nasis:

organic epoxy si'!' mils DFl'.

Four air.cooiers*, organic alkyd.:...epoxy

Four mils DFl' Four Hold-Up Tank~: no coating.

(

. TABLE II-2 (Cont'd)

. I.

. 1 '

I.

I i.

c.

Thermal C~nduct_ivi.ty (Maximum)

And -*

Volumetric Heat Capadty (Max:l.mU111)

Materials Organic Protective Coli tings Inorganic Protectiye Coatings*

Stainless Steel Liner Plate Carbon Steel Liner Plate Structural Concrete Thermal Conductivity

* (:Btu/}lr-ft~OF)'

0.3 2*

11 28 0.9 Specific Heat (Btu/lbffi-OF) 0.12 0.12 0.23 D. Heat Transfer Coefficients

a. : Conta_inment atmosphere to. sump:

500 BTU/hr-ft2-°F

b.. Sump to base s.1 ab:

20 BTU/hr~ft 2 ~

0 F-Volumet:.ic Heat C& :->£. ::~ :*-*

.(Btu/{t~~?)

62 62 59 59 33

c. Containment structure to ~nclosure building atmosphere~ 10.0 ~TU/hr-ft 2 -°F

Table II-3 Palisades C6ie I Large Break Spectrum*

Approved CE Evaluation Model With Proposed Modifications Break Size, Type and Location 1.0 x Double-Ended Slot Break in Pump Discharge Leg 0.8 x Double-Ended Slot Break in Pump Discharge Leg 0.6 x Double-Ended Slot Break in Pump Discharge Leg 0.5 x Double-Ended Slot Break in Pump Discharge Leg 1.0 x Double-Ended Guillotine Break i.n Pump Discharge* Leg 0.8 x Double-Ended Guillotine Break

' Pump Discharge Leg 0, 6 x. Doubl.e-Ended Guillotine Break in Pump Discharge Leg 0.5 x Double-Ended Guillotine Break in Pump Discharge Leg Abbrevia d.on 1.0 x DES/PD O. 8 x DES/PD.

0.6 x DES/PD 0.5 x DES/PD

.LO x DEG/PD 0.8 x DEG/PD 0.6 x.DEG/PD 0.5 x DEG/PD Approved CE Evaluation Model Without Proposed Modifications 1.0 x Double-Ended Slot Dreak* in Pump Discharge Leg 0.6 x Double-Ended Guillotine.Break in* Pump Discharge Leg

1.0 x DES/PD 0.6 x DEG/PD Figure I1.l II.2 II.3 II.4 II.5 II.6 II. 7 II.8 II.9 II.10

Table II-4 Variabl~s Plotted as a ~un~tion of time for Each ~a rge Break* in the Spectrum Variable Core.Power Pressure:in Center Hot Assembly Node Leak Flow Hot Assembly Flow (below hot spot)-*

Hot Assembly Flow (above htit spbt).

Hot Assembly Quality Containment Pressure Mass Added to Core During Reflood

__ "'eak Clad Temperature

~-

Figure Desi gna ti c_:>_ll A

B c

D. 1 D.2 E

F G

H

Variables.

Mid Annulus Flow.

Table n-s Additional Variables Plotted as a Function of Ti me fo~ the Large Break* Ha vi ~g.

the Highest Clad Tempera.ture Qualities Above and Below the.Core Core Pressure Drop Safety Injection Tank Flow into Intact.Discharge Legs Water Level in Downcomer During Refiood Gap Conductance Local Clad Oxidation Clad Temperature, Centerline Fuel Temperature, Average*

.._,fuel Temperature and Coolant Temperature for Hottest Node Hot Spot Heat Transfer Coefficient Hot Spot Heat Transfer Coeffic1ent.During Reflo*od

Containment Temperature Sump Temperature

Figure Designation I

J K

L M

N' 0

p Q

R s

T

Break 1.0 x DES/PD

'0,8 x DES/PD 0.6 x DES/PD 0.5 x DES/PD 1.0 x DEG/PD

.'-._.,,,J x DEG/PD 0.6 x DEG/PD 0.5 x DEG/PD 1.0 x DES/PD 0.6 x DEG/PD Tabie lI-6 Palisades Core I Times of Interest for Each Large Break

{Seconds)

Approved CE Model With Proposed Modifications Hot Rod Rupture

SI Tanks On 16.2 16.5 17.6 18.7 16.3 16.7 17.9 19.5 End of B~pass 16.2 16.5.

17.6 18.7 16.3 16.7 i7.9 19.5 End of Blowdown 18.9 19.2 20.3 21.5 19.0 19.5 20.6 22.3 Start of Re flood 30.65 30.95 30.05 33.28 30.76 31.28 32.33 34.03 Approved CE Model Without Proposed Modifications 16.2 17.9 16.2

'17. 9 18.9 30.43 20.6 32.12

'ot Rod Rupture is not Predicted to Occur SI Tanks

. EnEY._

56.0 56 * '*

57.4 58.5 56.1 56.5 57.S 58.2

54. 72 56.39

Table II-7 Palisades Core I Peak Clad Temperatures a*nd o:kidation Percentages

~or the Break Spectrum AEEi"oved CE.Evaluation Model With ProEosed Modifications Peak Clad Clad oxidation %

Break Temperature (OF)

Local

Core-Wide 11.0 kw/ft 1.0 x DES/PD 2135 2.14

<0.3761 0.8 x DES/PD 2124

1. 76

. <O. 3099 0.6 x DES/PD 2111 1.69

<0.2906 0.5 x DES/PD 2044 1.69

<0.2501 1.0 x DEG/PD 2131

. 2.05

.<0.3469

~*

0~8 x DEG/PD 2118

. 2.01

<0.3426

'-' x DEG/PD

. 2083

. 2.42

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FIGURE II.1-A PALISADES CORE I REANALYSIS 1

1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG CORE POWER I

I

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FIGURE II. 1-C PALISADES CORE I REANALYSIS I.

1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG 1

LEAK FLOW

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FIGURE II.1-D. l PALISADES CORE I REANALYSIS LO x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG FLOW IN HOT ASSEMBLY - PATH 16, BELOW HOT SPOT u

w.J

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FIGURE 11.1 ~D.2-.

PALISADES CORE I REANALYSIS

---i 1.0.x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG I

FLOW IN HOT ASSEMBLY - PATH 17~ ABOVE HOT SPOT 3 0 Ii 0 0 0 ~----...--------.-----------.------------....-------.

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. FIGURE 'It l~E

PALISADES CORE 1 REANALYSIS.

1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG

.*HOT ASSEMBLY QUALITY

--NODE 13, BELOW HOTTEST REGION

- ~~-NODE 14, AT HOTTEST REGION

-. * -.

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FIGURE II. 1-F

i PALISADES CORE I REANALYSIS

.. I.

1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE I

. r 50.000

~

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... FI.GU RE 11.1 ~G PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED SLOT BREA!< IN PUMP DISCHARGE LEG.

MASS ADDED TO CORE DURING REFLOOD 12.0 0 0 0 * ;._* ______

. 100000.

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FIGURE Il.1-H

~2.00 PALISADES CORE I REANALYSIS

1. 0 x DOUBLE ENDED SLOT BREAI< IN PUMP DISCHARGE LEG PEAi< CLAD TEMPERATURE 11 I I

I I

I 2.000 I

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TIME' SECONDS

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, FIGURE II.1-1 PALISADES CORE I REANALYSIS.

LO x DOUBLE ENDED. SLOT BREAK IN PUMP DISCHARGE LEG fylID ANNULUS :FLOW 15000.i....-.--

10 0 0 0 * '-_..:_--t-_

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'G FIGURE II.1-J PALISADES CORE I REANALYSIS

1. 0 x DOUBLE ENDED SLOT BREAI< IN PUMP DISCHARGE LEG QUALITIES ABOVE AND BELOW THE CORE

.*1.0000 I

n8000

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.I FIGURE II.1-K PALISADES CORE I RtANALYSIS t-4 V> a..:

L.LI"'

~

I.Ox DOUBLE ENDED SLOT BREAK. IN PUMP DISCHARGE LEG CORE PRESSURE DROP

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PALISADES CORE I REANALYSIS

1. 0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG SAFETY INJECTION TANK FLOW INTO INTACT DISCHARGE LEGS 10 20 30 40 50 TI.ME, SECONDS.

60

~,.

FIGURE Il.1-M PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG WATER LEVEL IN DOWNCOMER DURING REFLOOD

'I

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T

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00

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0

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'~00*.----

FIGURE II.1-N PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG GAP CONDUCTANCE

. 6 0 0 1------1-----+-----1-----+---1------+--------l u..

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'16 14

12.

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z 0

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0 FIGURE 11.1-0

PALISADES CORE I REANALYSIS I.Ox DOUBLE ENDED SLOT BREAI< IN PUMP DISCHARGE LEG LOCAL CLAD' OXIDATION

~.

11.3 KW/FT

/-:-----

{;. - 7_;;:.--

J

11. 0 l<W/FT I

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. 3500 3000

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~

a:::

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~ 1500 l.J.J I-1000 500 0

PRL

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FIGURE Il.1-P PALISADES CORE I REANALYSIS

1. 0 x DOUBLE ENDED SLOT BREAI< IN PUMP DISCHARGE LEG CLAD TEMPERATURE, CENTERLINE FUEL TEMP.,,. AVG. FUEL TEMP AND COOLANT TEMP FOR HOTTES 1 NODE

11. 0 KW/FT fif UEL CENTERLINE

~

---==~

~ER:GEFUEL l?'

[_CLAD

\\)

~

~COOLANT I

I 0

40 80 12.0 160 2.00 2.40 TIME, SECONDS

2.8C

eso FIGURE II. 1-Q.

PALISADES CORE I REANALYSIS LO x DOUBLE ENDED SLOT BREAK.IN PUMP DISCHARGE LEG -----

HOT SPOT HEAT TRANSFER COEFFICIENT

,Y-120 11--w.-.~-1-~---~~-:...~4-'--~~~~~----+-"--~-----r-~---;

I

')!....

u..

~

c -

. ~

~ 100 1-l-W---~-1-~--~~*~~+--_;._~"-+-~-'---+~~--"r-~---;

CXl **

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z

<C

~

I-

~

LIJ

c G 0 *IJU-J-'-oll---i----~---"--+-----...;....__j..------~------'"----~

'11. 0 KW/FT o*

40 ao 120 180 2.00 2.4 0

'.. \\*

1'

.f' TIME-' SECONDS

~:~

c *

c

FIGURE II.1-R PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG HOT SPOT HEAT TRANSFER COEFFICIENT DURING REFLOOD 30.000 25.iOOO LJ..

0 I

~

I 0::::

c: -

20.000 I-*

al

. i..:

2 L&J t-4 u

. t-4 u..

l:b 0 u

'BJ LJ..

Vl I

(

~ v

/

~

15.000 10.000 2 <

i=

I-

~

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0 0

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0 0

0

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0 0

a*

0 0

C\\J (Y)

.q-LO TIME RFTER CONTRCT ( SEC ).

FIG.URE II.1-S

PALISADES CORE I REANALYSIS

1. 0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG CONTAINMENT TEMPERATURE.

260---------------------------------.

220 200

.* LL.

0 LL.J°' 180.

~

I-

<(

0::::

L.l.J

~ 160.

L.l.J I-120 100 o""---1--00 ___

2__,_oo------'30_0 ___

4'--oo-. ----'soo TIME, SECONDS

FIGURE II.1-T PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG SUMP TEMPERATURE 240----------------....----

L.t..i"

~

J 220 200 180

<( 160 a:::

w.

o_

~

l..W I-140 120

100 0~~~~1~0~0~~~20~0~~-3*0~0~~-4~00~~---Jsoo TIME, SECONDS

FIGURE Il. 2-A PALISADES CORE I REANALYSIS

0. 8 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGELEG CORE POWER

1. 0 0 0 0 1-4"---t-------+-------+-----+-----*

aooo a::

w 3:

0 (L

_J

.6000 a:

I-0 I-

.4000

2.0 0 0.--~\\ ---+------+-----+----1-------1

\\_

~r---:----~~-----J-~__:___J_~~-

0.00000 0

0 0

0 c

0 0

0

(\\j (V)

"Cf-10 TIME IN SEC

. ~--.....

FIGURE II. 2-B PALISADES CORE I REANALYSIS 0.8x DOUBLE ENDED SLOT BREA!< IN PUMP DISCHARGE LEG PRESSURE IN CENTER HOT ASSEMBLY NODE 2400.0 2000.0 1600.o

<C 1-1

(./)

a.....

~ 1200.o '

(./)

~

a..

soo.o 400.0 o.o 0

0 0

0

~

0 C) 0 0

0 0

0 LO

.-1 TIME

~

0 0

a 0

C)

O 0

0 0

LO C)

LO

.-1 C\\J

'C\\J IN SEC

FIGURE II. 2~C PALISADES CORE I REANALYSIS 0.8 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG LEAK FLOW.

u L.LJ V> -

V>

ca 12.0000.

100000.

80000.

,_..J 60000 * ~

~

s: g 40000.

u..

2.0000.

o.

C>

0

C>

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/

I~

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0 C>

C>

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C>

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ru ru TIME IN SEC I

l I

Ii I

I I

i I

I I

' i I j. I I

I l

I I

I

1 I I

-*.. -------------------*~ --*-*

FIGURE II.2-D.l PALISADES CORE I REANALYSIS o.s*x DOUBLE ENDED SLOT BREAI< IN PUMP DISCHARGE LEG FLOW IN HOT. ASSEMBLY - PATH 16, BELOW HOT SPOT 30.000 2.0iiOOO

~

\\

ar w1

\\ /

v

_:...2.0.000

-30.000 a

0 C>

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0 b

0 C) t:>

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6 C>

C)

C>

0 0

ii C>.

LO C>

LO 0

LO C\\J C\\J TIME* IN SEC

' I I

~*

.. FIGURE II.2-D. 2

..... PALISADES. CORE I REAN'ALYSIS 0.8 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG FLOW IN HOT ASSEMBLY -PATH 17, ABOVE HOT SPOT 30.000*-*~----__.;..;.---.--.-----~~----.-~----,

'~

(..)

Lr.J

. 10

0 0 0 ll-4----+---------l,,__ ______ --f-_____ "-1 _____ _

V>. -

V>

a:::i Lr.J....

i--

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~

LJ..

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0 0

C>

0 0

0 lO 0

LO 0

. LO ru c\\J TIME IN SEC

>-l-

.H

_J' a:

'.:)

CJ

FIGURE IL 2-E PALISADES CORE I REANALYSIS 0.8 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG

. HOT ASSEMBLY QUALITY

---*NODE 13, BELOW HOTIEST REGION

- - NODE 14,. AT HOTTEST REGION..

. - *...:._ * -. NODE 15,. ABOVE HOTTEST REGION

.A,

~\\vi/

I I

\\,

ii

i 1

/ I 1.0000 1

I

/ /

.sooo~

1 1

~~~~

1

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I I*

I

2.0 o*o 1

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o.oooo~/ ___

--4-___________

..__ _________ ~

0.

0 0

tn

\\ *..

FIGURE II. 2-F PALISADES CORE.I REANALYSIS..

0.8 x DOUBLE ENDED SLOT BREAK.IN PUMP DISCHJ,RGE LEG CONTAINMENT PRESSURE

!j 40.ooo

\\

~

LU..

3 0 * *o Q Q ~---4-___.:.:~:::::::::f==m-"--------+--~+-.-....;..,._----j 0::::

J V')

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a.. ' ' 2.0

0 0 0 11-----'--------1-----+--,----t-----------j o.ooo 0

0 0

o 0

0 0

!I 0

0 a

0 b

0 C\\J 00 0

0

<D

(\\j

('f')

TIME AFTER RUPTURE~ SEC

FIGURE II. 2-G PALISADES CORE I REANALYSIS

0.8 x DOUBLE ENDED SLOT B.REAI< IN PUMP DISCHARGE'. LEG MASS ADDED TO CORE DURING REFLOOD

r 10 0 o.o 0.1-----4-_.;._..;__---:-:l-----t-----1----j Vl

~. 8 0 0 0 o.* 1-. --~---~-----------+_:.._--+7"-----j LL.I

~

0 u g

6 00 0 0.1-----1-----r--?'-+-'"*----j---~

Q.

UJ Q

,Q I

Vl

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40000.~--.-----~__,.c__-4..;.._---~+----------+-------1

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o.

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0 C).*

0 0

C) 0 0

0' ci' 0

C) 0 C)

(J)

~

ru C)

C) 00 ru

('\\')

~

TIME RFTER CONTRCT-' SEC

FIGURE II. 2-H PALISADES CORE I REANALYSIS

0. 8 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG

. ;> 20 0..-----

PEAK CLAD TEMPERATURE

(

~

r--.....

__ -:-t-__

I '-.../

. 16QQ U-l--1-,-J----~-~1--~-l----=:::::~-=~r----_~~-t----;

ILO KW/FT I

.~.

14 0 0 l-J--l~--i-----+----l----+-----f----1----i LLI.....

~12.QQW--~.~--+~~~+-~_;_;_-4--~~-+-~_.__--i~~~-+-"""--'~--t I-

<(

0:::

LLI 0..

~

~lOQQH-~~-+-~~-b~~.,-1-~_;_;_~-+-~~-+-~~-+~~--1 c

5 u

0 40 8 0 120 160 2.00 TIME-' SECONDS

-,_ 1

I I

-~

w

3::

0 0...

_J a:

l-o l-FIGURE II. 3-A PALISADES CORE I REANALYSIS

0.6 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG CORE.POWER..

1.2001.--~-----.-~~~--~~~-----------"T~~~--.

I

.1.ooooVl I

I I

I.

I

~

.eooorr~~-t-~~---r~~~r--;.__~-f-~~_J

6 0 0 0 ~~-----~1--..;__;..--1-----+---------r----------i I

.400Ql--"--~-4~~---+-------r.-----t------1

.2000i---+-~--t-~--+-----t-~----t-----1 0.00000 0

0 0

0

o 0

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C>

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0 C>

I

0

- -I C\\J

('t)

-.it-in

.TIME IN SEC

~

V>

a....

LL.I

~

FIGURE II. 3-B PALISADES CORE !*REANALYSIS 0.6 x*DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG PRESSURE IN CENTER HOT ASSEMBLY NODE 2.400.0 2.000.0 1600.0 12.00.o V>

~

- ~

V>

UJ

~

.c...

800.o.

400.0 o.o 0

0 0

0

~

. ~

~

0 0

0

LO 0

0 0

0...;

0 0

0

to TIME IN SEC

~

0 0

0

0 ru 0

0 0

to ru

FIGURE II. 3-C.

PALISADES CORE. I REANALYSIS u

l.J.J V) -

V)

CQ

....J 0.6 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG

. LEAK FLOW 12.0 0 0 0

r---~----..,.,---..,--..------------..

100000.

I 8 0 0 0 0

J---~---f-----+-------------l------1

~"', 60000 II

<(I e::::

~

'....J LL.

40000 ** ~~--"r-l~~~---t-~~~---t---~~-+-~~~-1 lOOOO.t--~~---t~~~---t----~~---t-~~~-+-~--~-1.

o

0 0 c Q

0 0 c

LO.

0 0

0

0.;

TIME IN I

0 0

0 LO 0

LO

.-i.

C\\J C\\J SEC

u LU Vl -

Vl a:l

....J..

LU e:::::

s:

g LL.

FIGURE II. 3-D. l

. PALISADES CORE l REANALYSIS 0.6 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG

FLOW I~ HOTASSEMBLY - PATH 16, BELOW HOT SPOT 3 0

0 0 0.---------.;,-----------.....---------i 10~000 o.ooo

-10.000

\\*

-2.0.000H------+----+-----+------+----i

-30.000 0

.0 0

0 0

o.

0 0

o 0

0 0

0 L()

0 LO 0

L()

.-f

~

C\\J C\\J TIME IN SEC

r-"

FIGURE II. 3-D*. 2.

PALISADES CORE I REANALYSIS 0.6 x DOUBLE ENDED SLOT BREAK *IN PUMP DISCHARGE LEG FLOW IN HOT ASSEMBLY - PATH 17, ABOVE HOT SPOT 30.000

' *2.0.000 u'

LL.I 10.000 V') -

V')

co

....J o.ooo

-2.0.000

-30.000 0

0 0

0

~.

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~

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0 LO

.TIME IN I

0 0

0

LO _.

SEC 0

d 0

0 C\\J 0

0 0

LO C\\J

H

_J er

J CJ FIGURE II. 3-E PALISADES CORE I REANALYSIS

0. 6 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY QUALITY 1.0000 n

I!

i; '

'.-sooo I

I I

1

'.6000 I v I

I.

I. I I

I I

112.000 I

0.0000 ~

o*

0 0

0

/

0 0

0

tn

-- NODE 13, *BELOW HOTTEST REGION

- - - --*NODE 14, AT HOTTEST REGION NODE 15, ABOVE HOTTEST REGION 0

o 0

0.-;

! ' v I

I I \\

I \\

j

}

I

\\

I l I \\

\\ I.

\\

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1.

I

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v 0

0

.0

LO TIME IN SEC r.

I I

I

o.

0 0

0

(\\j 0

0 0

LO

(\\j I

I

r----~~--------*

L&.J..

c::::

l

(/')

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. c::::

c..

FIGURE II. 3-F PALISADES CORE I REANALYSIS 0.6 x DOUBLE END.ED SLOT BREAK IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE 6 0. 0.0 0,----.---~r---~--r---,_;...._-,~---

I 20.0001r-----t----:--l-----+-~-~1--~-~

o.ooo 0

0 0

0

o

0.

0

0.

0

.0 0

0

.o 0

0 0

0 C'U 00

-.:.t-0

o

<..O

.-i

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TIME AFTER RUPTURE, SEC.

V'>

cc

-I L.&J 0:::

0 u 0 I-0

. L.&J 0

Q <

. V'>

V'>

~

I FIGURE II.3-G * *

PALISADES CORE*I REANALYSIS.

0.6 x DOUBLE ENDED SLOT "BREAK IN PUMP DISCHARGE LEG.

. MASS ADDED TO* CORE DURING-REFLOOD 10 0 00 0. i----~--~1--...-.....;;..-~---1~---l 80000.

60000.

40000~

2.0 0 0 0 * ~---+--f------+-----li------1---~

.0.

0 0

o*

0 0

a Ii 0

c..o

""4-C\\J a

.o 00

~

C\\.J (Y)

.q.

TIME RFTER CONTRCT~ SEC I

.- I

-'I

~.2.0 0 2.000 1800

. 1600

.u..

0 1400

~

.. 12.0 0

=>

r-

<C 0:::

LLI a..

~ 1000 r-e s

(.,)

800 600

~ 400 t..

l.J 0

FIGURE II. 3-H PALISADES CO-RE I REANALYSIS

0~6x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG PEAK CLAD TEMPERATURE r ~

v I

11.0 KW/FT

~

I 40 80

- ... 12.0 :, 160 2.0 0 2.4 0 2.8 0 TIME-' SECONDS.

. FIGUREIL4-A

. PALISADES *CORE I REANALYSIS

0. 5 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHJ.\\.RGE LEG CORE POWER

1. 2. 001*-----.----....,_--------,r-----r-------i r

1.oooo---------------~--1----~---.~----------------~

o.ooooo~.----0------------------------------'

0 0

a*

0 0

a 0

0 0

0

~

0

'.-I C\\.I (Y)

LO

.TIME IN SEC

FIGURE II. 4-8 PALISADES.CORE,I REANALYSIS o~s x DOUBLE ENDED SLOT BREAI< IN PUMP DISCHARGE LEG '

PRES-SURE. IN *CENTER HOT ASSEMBLY NODE 2.000.0 16*00.o t-4 1*

t.n a....

12.0o.o i

LU I.

~

J t.n r

t.n LU l

I

~

I 0..

soo.o 4 00.o ---------.----.......-r-------1

FIGURE II. 4-C PALISADES CORE I REANAL VS IS,

0.5 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG LEAK FLOW 12.0000,.

.

100000 II u

80000.

I.LI V>

V>

cc.

...J..

I.LI I-60000.

<C 0:::

\\

~

3:.

0

...J LL.

40000.

.~

r-...

~

2.0000.

o.

0 0

0 a

0 0

0 ill 0

LO 0

0 LO.

C\\.l TIME IN SE.C 0

0 0

II LO C\\.l

u LU Vl -

Vl en LU t-

<(

0::

~

FIGURE II. 4-D.1 PALISADES CORE I REANALYSIS 0.5 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG FLOW IN HOT ASStMBLY - PATH 16, BELOW HOT SPOT 30.-o 0 0 2.0.. ooo 10-.000.\\t

~.

.~ ~

~

o.ooo LL, -,10.ooo

-2.o.ooo

-30.000 C>

C>

0 0

C>

0 0

0 C>

0 o

o 0

LO C>

0 lO

.-t

(\\j TIME IN SEC

.0 0

0

LO

(\\j

FIGURE II. 4-D.2

_ PALISADES CORE I REANALYSIS 0.5 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG FLOW IN HOT ASSEMBLY - PATH 17, ABOVE HOT SPOT 30.000 2.0.000

(._)

~ 10.000 Vl co

...J u..i'.

1-

<C.

0::

o.ooo o'

...J u_.

i-10.000

-2.0.000

--30. 0 00 0

0 0

0

~

~r

~ vv

~

0 0

a*

0 0

0' 0

Ln 0

.LO N

TIME IN SEC 0

0 0

Ln N

-----~

1-.

H

_J CI

J G

FIGURE It 4-E PALISADES CORE I REANAL VS IS 0.5 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY QUALITY 1.0000 i :-

.aooo

6000 j I )

I' i: A I'

!1 11.,

~

.2.000 l I

I I

0.. 0 0.0 0 r 0

0 0..

0 I v*

J.

0 0

II 0

t.n

.-i TIME NODE 13, BELOW HOTTEST REGION NODE 14, AT HOTTEST REGION NODE 15, ABOVE HOTTEST REGION

\\ \\ / I

' \\

.\\

\\_ I u

I

\\

l

\\

I.

/.

\\.

I

\\...

0 0

0 t.n 0

t.n

.-i N

ru IN SEC

V) c..

*r FIGURE II. 4-F.

.. PALISADES CORE I REANALYSIS.

0.5 X DOUBLE ENDED SLOT BREAK TN PUMP DISCHARGE LEG CONTAINMENT PRES SU RE *.

5 Q ii Q Q 0 i-----t-----------11-----1--_..;_-'-'.;,__---1 20

00011-----t----i-----'----4-----l~---l o.ooo 0

0 0

o' 0

0 0

C>

C) 0 0

C\\.l co

-.:t-0 0

U)

C\\.l

(\\")

TIME *AFTER* RUPTURE,. SEC

-,, ___._.,,;;___.:.;;,;;;;;===!

r FIGURE II. 4-G PALISADES CORE I REANALYSIS

0. 5 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG MASS ADDED TO CORE DURING REFLOOD V')

CQ

...J 8 0 0 0 0

l---'-----4----4---__;_+.---~:..___---1 Q

UJ Q

Q c::c V')

V')

c::c 40000.f.-----+--~~-+--~--+-~--+------I

E

o.

0 0,

0 0

0

{D

'"'<t" ru

o.

0 00 ru

('t)

"'<t" TIME RFTER CONTRCl-' SEC

e

2.2.0 0 2000 1800 1600 400 u_

0 uJ' 12.0 0 0::::

=>

I-

<(

E5

~ 1000 LU l-o s u

{" :t_..

.i, '

~

800

r~; ** *4oo

-~

1*:1,....

... :~: *... *<'

li((~ '

..,5'....

v 0

FIGURE II. 4-H PALISADES CORE I REANALYSIS..

0.5 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG PEAK CLAD TEMPERATURE

(

' ~

~

r---__

I I

-~

11. 0 KW/FT

,;. \\

f',.* '

~: **.

.~

\\., --

..f.

. ' '*'. ;;tl

-~

~

40 80 12.0.

160 2.00 2.4 0 2.8 0 TIME-' SECONDS I

§ I

n::

LU 0

0...

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r-0 I-FIGURE II. 5-A

PALISADES CORE I REANALYSIS..

1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CORE POWER

1. 0 0 0 0 F-+--------1----------+----4

8 0 0 0 1--+-----+----+------1-----+-------1

.* 6 0 0 0 i--r---t-----t----+-----+----~

.4000*~t----T----+-----+---~-+-------I

2. 0 0 0 l--\\-+---t---+-------+---....-_.,_-'-----1

. "----i---f---~-~~-

0.00000 0

0 0

d

.-4'

(\\j (Y)

LO TIME IN SEC

FIGURE Il.5-B PALISADES CORE I REANALYSIS

1. 0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PRESSURE IN CENTER HOT ASSEMBLY NODE 2.400.0 2.000.0 1600.0 LI.I..

'-~

12.00.o ' ~

V'l.

V'l LI.I

~

0...

aoo.o

~

400.0 o.o

~

........ -***. *- *-* *****--** --. *-**-*-----~------* --

/

FIGURE II. 5.:.c PALISADE.S CORE t' REANALYS.IS.

1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG.

LEAK FLOW

---PUMP SIDE

~ -- -REACTOR VESSEL SIDE 12.0 0 0 0 * --:----.--..:.-----.,---~-___;_-..---___;_-

.. l00000.1----:---~-+-------+----~__;..~-~~---1

40000.1~~-~,~---t-----+--~---+----+----J I

".\\

ll.. '

\\

2.0000.

o.

0 0

0

0

\\

0 0

0

lO

0.

0 0

0 -i 0

0 0.

lO

-i TIME* IN SEC 0

0 0

0 C\\J 0

0 0

lO C\\J

. FIGURE II. 5-D.1 PALISADES CORE I REANALYSIS

1. 0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG*

FLOW IN HOT ASSEMBLY - PATH 16, BELOW HOT SPOT 10.*000-

.(..)

I LU V> -

V>

a::l

-I o.ooo LU I-

<(

~

...J -10.000*

u...

-20.000. -**--*

-30.000

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0 0

0 C>

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0 o*

0 0

0 Ln C>

Ln 0

Ln C\\.J

(\\J TIME IN SEC

~.

_./

(...)

LJ.J V'> -

V'>

FIGURE II. 5-D.2

. PALISADES CORE I REANALYSIS

1. 0 x DOUBLE ENDED GUILLOTINE BREAK IN.PUMP DISCHARGE LEG FLOW IN HOT ASSEMBLY - PATH 17, ABOVE HOT SPOT 3 a*. o o o 1-rn-------r-...___ __

~_,___---:----...-----

' 2.0

0 0 0 ll-Hll------1-------1----1-------+----

5 0.ooo l

.~

i.J I-<

0::: s g -10.oooi...-----r-----.-~---1-~--+----1 LL..

-30.000 0

0 0

0 C>

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LO C>

I.(')'

-i

.-i

(\\J

(\\j TIME IN SEC

/ **

>-t-H

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.::J 13

. FIGURE II. 5-E PALISADES CORE (REANALYSIS.

1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT ASSEMBLY QUALITY

1.0000 fl

. I

~I,,

1

.

800 0 I

t I

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I

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NODE 13, BELOW HOTTEST REGION

--.- -. NODE 14, AT HOTTEST REGION

-* - **-. NODE 15, ABOVE HOTTEST REGION

~ \\

I

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FIGURE II. 5-F PALISADES CORE I REANAL VS IS L 0 x DOUBLE ENDED GUILLOTINE.BREAK IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE

'I 5 0

0 0 0 1------1--------1------+------f----~

s 40.000

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I FIGURE IL 5-G PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED GUILLOTINE BREA!< IN PU.MP DISCHARGE LEG MASS ADDED TO CORE DUR ING REFLOOD 12.0000 ll~

. I

lOOOOOa---~~--~~__,~~~---~~--~~~-1 V>

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FI GU RE II. 5-H PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG 2.2.0 O ~

PEAK CLAD TEMPERATURE 2.000t-t-t------f---,---t-------+-----+---1---,--~-+----J f\\

1800H-t---f-tt,_-

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~

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40 80 12.0 160 2.00 2.40 2.80 TIME~ SECONDS

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.[

3 0

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a.

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0 I-FIGURE lL*6~A PALISADES CORE I REAl\\JALYSIS 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CORE POWER la2.001

/'

1.0000

8000 116000

. 4 00 0

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~

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FIGURE II. 6-B PALISADES CORE I REANALYSIS o~s x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PRESSURE IN CENTER HOT ASSEMBLY NODE 2.400*0 2.000.0

\\

1600.0

<C

' ~

~

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tn 0

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. i.

I

i--~----~~~~~~~~~~~~~~~~~~~~~~~

FIGURE II.6-C PALISADES CORE l REANALYSIS

0. 8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG LEAi< FLOW*.

-'----PUMP SIDE

-REACTOR VESSEL SIDE.

. a o*o o o

~ ----+--_..,...,--'1------1----'--+----~

(.)

L.LJ V> -

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~

.~

s: '.

0 LA:!

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\\

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2.000011

~

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FIGURE II. 6-D. l PALISADES CORE I REANALYSIS 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG FLOW IN HOT ASSEMBLY - PATH 16, BELOW HOT SPOT

! u LL.I V> -

30.000 2.0.000

~

10.000

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FIGURE II.6-D.2 PALISADES CORE l REANALYSIS*

. 0.8 x DOUBLE ENDED GUILLOTINE BREAl(IN PUMP DISCHARGE LEG FLOW IN HOT ASSEMBLY - PATHJ7, ABOVE HOT SPOT 10 *.0 0 0 o.ooo

g. -10.000 LL.

-2.0

0 0 0 r----r--~_,. _

_;_;_.;..._..+---~...___;.--

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TIME IN SEC

>-r-1-t

_J

([

J '

G

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FIGURE II. 6-E

. PALISADES CORE I REANALYSIS

0. 8 x DOUBLE ENDED GUILLOTINE BREA!< IN PUMP DISCHARGE LEG HOT ASSEMBLY QUALITY

~..

r 0

NODE 13, BELOW HOTTEST REGION NODE 14, AT HOTTEST REGION I

d NODE 15, ABOVE HOTTEST REGION 1.0000

~

I \\

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l

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fr I

I

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0 0

o lO

s V>

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LL.I 0:::

a..

PALISADES CORE I REANALYSIS FIGURE II. 6-F '

I 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE1EG CONTAINMENT PRESSURE 50.000

r

- - - - ~

.. ~. :

40.000 f"

\\

~. '

30.000 2.0.ooo 10

0 0 0 i-----.......+----i--------~--------11----1 o.ooo 0

0 0

C>

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('t)

TIME.AFTER RUPTURE, SEC

V):

al !

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c:::

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~

FIGURE IL 6-G *

. *

PAlJSADES CORE I REANALYSIS 0.8 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE.LEG

. MASS ADDED TO CORE DURING REFLOOD

r aoooo.~-------1-----'--+---~-+-----+-r--~

c : b 0 0 0 0 ~ 1---------+----+---__,..~----+-~-~1 LLJ Q,

o.*

<C V)

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---~ -~-* - '.

2.000 1800

. 1600 1400 u..

0

~

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~

Ll.J 0..

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I;!: 1000 0

5 U.

800

~*

'\\-....

600 i

~

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FIGURE II. 6-H PALISADES CORE I REANALYSIS

0. 8 x DOUBLE ENDED GUILLOTINE BREAI< IN PUMP DISCHARGE LEG PEAK CLAD TEMPERATURE I

I I

/\\

\\

~

11. 0 KW/FT

~

I l

40 80 120 160 2.0 0 2.4 0 2.8 0

'TIME-' SECONDS

~

w

==

D 0...

_J

([

r-D r.-

FIGURE It7-A PALISADES CORE I REANALYSIS

~

0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG CORE POWER *.

1.2.001

('

1.0000

8000

.sooo

.4000

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0 0

0

\\ c 0

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ru

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(V) lri IN SEC

*: ::s I'

V)

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0::

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FIGURE II. 7-8 PALISADES CORE I REANALYSIS

0. 6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG

PRESSURE IN CENTER HOT ASSEMBLY NODE 2.400.0 2.000 mO 1600.0

~

12.00.0

~

r.....

400.0 o.o 0

0 C>

a a

C>

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d C>

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lO C\\J C\\J TIME IN SEC

FIGURE II.7-C,

. PALISADES CORE I REANALYSIS

0. 6 x DOUBLE ENDED GUILLOTINE BRtAI< IN PUMP DISCHARGE LEG LEAK FLOW 12.0000.

... 100000.

u 80000.

LLJ V> -

V>

CQ

...J

~

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. <C a::::

~ce 40000.

2.0000.

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l I

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TIME IN SEC I

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FIGURE II. 7-D. l PALISADES CORE I REANALYSIS 0, 6 x DOUBLE ENDED. GUILLOTINE BREAK IN PUMP DISCHA.RGE LEG

FLOW IN HOT ASSEMBLY - PATH 16, BELOW HOT SPOT

.. 30.. ooo 2.011000 u

ro.ooo I.LI V> -

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0 0

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...J FIGURE II.7-D.2 PALISADES CORE I REANALYSIS

-e

0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHJl.RGE LEG FLOW IN HOT ASSEMBLY - PATH 17, ABOVE HOT SPOT o.ooo

-30.000 0

0 a*

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FIGURE II. 7-E..

PALISADES CORE I REANALYSIS 0.6x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG HOT ASSEfy\\BLY QUALITY 1.0000 II

I l t I~

.sooo

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~ -

~ NODE.14, AT HOTTEST REGION

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NODE 15, A BOVE HOITEST REGION

\\

I

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Li.J' FIGURE II. 7-F PALISADES CORE*I REANALYSIS 0.6 x DOUBLE ENDED GUILLOTINE BREAI< IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE

SO.000 50.000 40.000 (\\

~

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c:::

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0

~

0 C\\.I

('I')

FIGURE II. 7-G.

. PALISADES CORE I REANALYSIS

0. 6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG MASS ADDED TO CORE DURING REFLOOD..

12.0 0 0 0

r---.....,_ _ ___,__..;..;---~.....---------.~-----

, I 100000.

I.

V) co

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( ') 0 2.000 1800

1600 1400 LLJ..

§3 l2.0 0 1-

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LLJ

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! 1-1000 c

s

<..)

800 600 400

\\

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0 FIGURE II. 7-H PALISADES CORE I REANALYSIS 0.6 x DOUBLE ENDED GUILLOTIN[ BREAK IN PUMP DISCHARGE LEG

. PEAK CLAD TEMPERATURE t

- ~

I I

~ -i---_.

II. 0 KW/rr------r---__

i. I

.* 1 l.

--~

i 40 80 12.0 2.80.

160 2.0 0 2.40 TIME-' SECOND$*

Figure II. 7-0 PALISADES CORE I REANALYSIS 0.6 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG

- i8 LOCAL CLAD OXIDATION 16 14

12.

I I

8

~ -

J z

0

I=!

Q b

t-4 f3 I

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. 2.4 0 2.8 TIM[~. s~:coNns.

0:: w

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l-0 l-FIGURE II. 8-A PALISADES CORE I REANALYSIS 0.5 x DOUBLE ENDED GUILLOTINE BREAI< IN PUMP DISCHA.RGE LEG CORE POWER

1. 2.001-*---------------...------.,..---.

I l*O 0 0 0 ~---f-----+---+----+------1

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6000

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\\

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lO TIME IN SEC

r **

1-f Vl a..

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e:::::

=>

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a..

I FIGURE II. 8-8 PALISADES CORE I REANALYSIS 0.5 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG PRESSURE IN CENTER HOT ASSEMBLY NODE 2.400.0 2.00Q.O 1600.0

~

~-

~

12.00.0 800.0 400.0 o.o 0

0 0

0 II 0

Ln 0

. Ln o*

lO

(\\.)

C'U TIME IN SEC

~

FIGURE II. 8-C

\\..---'

I I

u.

. LU V') -

V')

al

...J

PALISADES CORE I REANAL VS IS 0.5 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG.

LEAK FLOW *..

PUMP SIDE

REACTOR VESSEL SI DE

... --... ' '~

~oooo~~~.~\\--~,---------~_+_------.~. --~-+---------1

~ -,,,

0 *.

0 0

0 p

0 0

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TIME IN SEC*

\\*.

0 0

0

0 C\\J 0

0 0

lO C\\J

FIGURE 11..8-0.1

. PALISADES CORE! REANALYSIS

.Q.5 X DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG

. FLOW JN HOT AS~EMBLY - PAT,H 16,. BELOW HOT SPOT 3 0

0 0 0.-------------.------------.,---.----.

-20.000------------------------------*

-30.000 0

0 0

0

o.

0 0

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LO C\\J C\\J TIME IN SEC

-*~*---~..

u LLJ Vl -

Vl co LLJ I-

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FIGURE II. 8-D.2 PALISADES CORE I REANALYSIS

0. 5 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG FLOW IN HOT ASSEMBLY - PATH 17, ABOVE HOT SPOT 20.00011-+-~~-+---------+~~~4-~~-4-~~~

10.000 o.ooo

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-10.000

-30.000 0

0

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0 0

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C)

C) b CJ CJ C)

. lO CJ lO CJ LO C\\J C\\J TIME IN SEC

1-H

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C?J FIGURE II. 8-E.

PALISADES CORE I REANALYSIS 0.5 x DOUBLE ENDED GUILLOTINE BREAI< IN PUMP DISCHARGE LEG HOT ASSEMBLY QUALITY 1.0000

! I I

. I

. -I '

.

I I

.~

11

aooo ii

1

. I'

.6000 i! '

.4000,,

r I

I

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0.0000,Y 0

0 0

LO I

NODE 13, BELOW HOTTEST REGION.

NODE 14, AT HOTTEST REGION NODE 15, ABOVE HOTTEST REGION

\\

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0 0

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s V>

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FIGURE II. 8-F PALISADES CORE' I REANALYSIS 0.5 x DOUBLE ENDED GUILLOTINE BREAI< IN PUMP DISCHARGE LEG CONTAINMENT PRESSURE so.coo 50.000 40.000 (\\

\\

30.000 I'---_

20.000 10.000 o.ooo 0

0 0

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TIME AFTER RUPTURE, SEC J'

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FIGURE 11. s~G. **

PALISADES CORE I REANALYSIS 0.5 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG MASS ADDED TO CURE DURING REFLOOD.

'I 100000.

80000.

60000 Ii 40000.

2.0000.

0.

a a

0 a

a a

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- 1

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FI GU RE II. 8-H 1

, PALISADES CORE I REANALYSIS 0.5 x DOUBLE ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG

,,... *') 2.0 u PEAK CLAD TEMPERATURE 2..oo 0 1800

. 1600 u...

0 Lu°"

1400 gs 12.00

<C 0:::

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~

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5

u.

' 800

(

. t. :

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600 400

!~

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~

40 8 0 12.0 160

. 2.0 0 2.4 0 TIME.-', SECONDS

figure II. 9-F

PALISADES CORE REANALYSIS 11

1. 0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG

'I, CONTAINMENT PRESSURE I:

i :

APPROVED MODEL 4 0

0 0 0 ~ -*-* -----*----1*-- ---*--------------1

'S V>

c..

u.. r 30 aOOO 0:::

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. - -.. v--* * -*,,, *.., *** *... ""'r.. C7-..---* -*-'1"......,.... -........ -:___-"'*._.... ~- ** ***

.. ' '* *..,,.*..,-. **.......... '*'..,., ~--...,,,_,.,._. --**.* --

Figure II. 9-:-G PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED SLOT BREAI< IN PUMP DISCHARGE LEG

.

MASS ADDED TO CORE DURING REFLOOD i--__

V) al LL.I 0:::

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0

<C V)

V)

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80000a 60000D 40000A

o.

0 0 *

,o

. APPROVED MODEL TIME {SECS}

REFLOOD RATE {INS/SEC}

0.0-10.0 1.762 10.0-48.0 1.107 48.0-400.0 0.587 I

0 C) 0 a

a 0

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RF.TEI~ CONTr-1CT.9 SEC

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Figure IL9-M PALISADES CORE I REANALYSIS 1.0 x DOUBLE ENDED SLOT BREAK IN PUMP DISCHARGE LEG WATER lEVEL IN DOWN COMER DURING REFLOOD APPROVED MODEL r

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CONSUMERS POWER COMPANY DOCKET 50-255 REQUEST FOR CHANGE TO THE TECHNICAL SPECIFICATIONS LICENSE DPR-20 For the reasons hereinafter set forth, it is requested that the Technical Specifications contained in Provisional Operating License DPR-20, Docket 50-255, issued to Consumers Power Company on October 16, 1972 be changed as described below:

I.

Changes A.

On Page ii, Table of Contents, add the following:

Section 3.18 Description Linear Heat Generation Rate Limits Associated With LOCA Considerations Page 3-82 B.

Add a new Item d to Section 3.1.5 as follows:

"d.

The primary to secondary leakage in a steam generator shall not exceed 0.3 gpm for any period greater than 24 consecutive hours."

c.

Add a new paragraph at the end of the basis of Section 3.1.5 as follows:

"The 0.3 gpm primary to secondary, leakage limit was originally provided to ensure that if clad.collapse existed which in essence locked the fuel pellet stack that transients such as a loss of load transient accompanied by a secondary relief valve sticking open would not cause 10 CFR Part 20 limits to be exceeded at the site boundary.

Fuel inspections performed during the fall of 1973 after a core exposure of approximately 7,000 MWd/MT have shown that the collapse of unpressurized fuel rod cladding is not likely to occur because significant pellet gaps do not exist.

Therefore, from the standpoint of fuel densi-fication the limit is no longer required.

The limit is retained to provide a stringent limit on primary to secondary leakage during the transition from phosphate to volatile secondary water chemistry.

This transition is due to the wastage attack experienced during 1973 due to the phosphate secondary water chemistry control."

2 D.

Change Section 3.11 to read as follows:

"3.11 IN-CORE INSTRUMENTATION Applicability Applies to the operability of the in-core instrumentation system.

Objective To specif'y the functional and operability requirements of the in-core instrumentation system.

Specification

a.

Sufficient in-core instrumentation shall be operable whenever the reactor is operating at or above 75%

rated power in order to:

(1) Assist in the calibration of the out-of-core detectors, and (2) check gross core "power distribution.

As a minimum, 10 individual detectors per quadrant, which shall include 2 detectors at each of the four axial levels, shall be operable.

b.

For power operation above a power level of 85% of the level permitted by Section 3.18, in-core detector alarms generated by the data logger shall be set, based on the latest power distribution obtained, such that the.

peak linear power does not exceed the limit specified in Section 3.18. If four or more coincident alarms are re-ceived, the validity of the alarms shall be immediately determined and, if valid, power shall be immediately de-creased below alarm set point and a power distribution map obtained. If a power distribution is not obtained within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the alarm conditions, power shall be reduced to 85% of the value defined in Section 3.18.

c.

The in-core detector alarm set points shall be estab-lished based on the latest power distribution maps, normalized to the kW/ft limit defined in Section 3.18.

d.

Power distributions shall be evaluated every week or more often as required by plant operations.

e.

The data logger can be inoperable for two hours.

If at the end of two hours, it is not available, the power level shall not exceed 85% of the kW/ft limit defined in Sec-tion 3.18.

3

f. If the data logger for the in-cores is not operable for more than two hours, readings shall be taken and logged on a minimum of 10 individual detectors per quadrant at least each.two hours thereafter or the reactor power level shall be reduced to less than 75% rated.

Bases A system of 45 in-core flux detector and thermocouple assemblies

. and a data display, alarm and record functions has been pro-vided. (l)

The out-of-core nuclear instrumentation calibra-tion includes:

"a.

Calibration (axial and azimuthal) of the split detectors at initial reactor start-up and during the power escala-tion program.

b.

A comparison check with the in-core instrumentation in the event abnormal readings are observed on the out-of-core detectors during operation.

c.

Calibration check during subsequent reactor start-ups.

d.

Confirm that readings from the out-of-core split detec-tors are as expected and that the ratio of the top and bottom detector readings is acceptable.

Core power distribution verification includes:

a.

Measurement at initial reactor start-up to check that power distribution is consistent with calculations.

b.

Subsequent checks during operation to insure that power distribution is consistent with calculations.

c.

Indication of power distribution in the event that abnormal situations occur during reactor operation.

If the data logger for the in-core readout is inoperable, for more than two hours, power will be reduced to 85% of the limit specified in Section 3.18 to provide margin between the ac-tual peak linear heat generation rates and the limit and the in-core readings will be manually collected at the terminal blocks in the control room utilizing a suitable signal detec-tor. If this is not feasible with the manpower available, the reactor power will be reduced to less than 75% rated to

4 minimize the probability of exceeding the peaking factors.

The time interval of two hours and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the core power distribution to detect significant changes until the data logger is returned to service.

Reference.:.. (1)

FSAR, Section 7.4.2.4."

E.

Add a new Section 3.18 as follows:

3.18 Linear Heat Generation Rate Limit Associated With LOCA Consider-ations Applicability Applies to fuel linear heat generation rates.

Objective To delineate the requirements regarding fuel linear heat genera-tion rates associated with a postulated Loss of Coolant Accident.

Specification 3.18.1 The linear heat generation rate with appropriate consider-ation of normal flux peaking, measurement-calculational uncertainity (10%), engineering factor (3%), increase in linear heat rate due to axial fuel densification (1.75%),

power measurement uncertainity (2%), and flux peaking augmentation factors which vary from near 0% at the bot-tom of the core to approximately 3.8% at the top of the core, as shown in Figure 3-6, shall not exceed that limit which causes calculated ECCS performance, as predicted by an evaluation model approved by the NRC as satisfying the requirements of 10 CFR 50, Appendix K, to exceed the "Acceptance Criteria for Emergency Co:r:e Cooling Systems for Light Water Cooled Nuclear Power Reactors" as given in 10 CFR 50.46(b)."

Bases "To maintain the integrity of the fuel cladding under the condi-tions of a postulated Loss of Coolant Accident (LOCA), the Emergency Core Cooling Systems (ECCS) must satisfy certain criteria set forth by the US Nuclear Regulatory Commission in

1.06 1.05 1.04 H

0

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l 1.02 c:i:

1.01 1.00 0

20 Bottom Augmentation Factor Versus Height Monte Carlo Computation i'or Palis.ades 40 60 80 100 Active Core Height in Inches Figure 3 - 6 r---

120 140 Top

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l20 l40 Top

5 10 CFR 50.46(b).

These criteria assure that under LOCA condi-tions the temperature and oxidation of the cladding will be controlled such that the uranium dioxide pellets will be main-tained in a coolable geometry.

These criteria are summarized below:

l)

The calculated maximum fuel element cladding temperature shall not exceed 2200°F.

2)

The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxida-tion, including effects of cladding thinning and rupture.

3)

The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hYJ?othetical amount that would be generated if all of the metal in the cladding cylinders sur-rounding the fuel, excluding the cladding surrounding the plenum volume, were to react *

4)

Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5)

After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an ac-ceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioac-tivity remaining in the core.

The computer codes which predict cladding temperature and oxida-tion under LOCA conditions must be approved by the US Atomic Energy Commission in accordance with Appendix K of 10 CFR 50.

The re-sults of these computer code calculations depend on ECCS performance characteristics and fuel design.

Analyses performed with approved codes and techniques for each fuel design tYJ?e taking into consid-eration anticipated operating conditions will be kept on file at the plant and at the General Office.

These analyses provide safety limits given in terms of allowable linear heat generation rates in kW/ft for each fuel design type.

Appropriate factors for measurement-calculation uncertainty, engi-neering factor and shortening of the fuel pellet stack are specified to insure that linear heat generation rate limits are not exceeded during steady state operation."

6 F.

Change Table 4.2.1 to read as follows:

"TABLE 4.2.1 Minimum Frequencies for Sampling Tests

1.

Reactor Coolant Samples

2.

Reactor Coolant

3.

SIRW Tank Water Boron Sample

4.

Concentrated Boric Acid Tanks 5,

SI Tanks

6.

Spent Fuel Pool 7,

Secondary Coolant

8.

Liquid Radwaste

9.

Radioactive Gas Decay Tanks

10.

Stack-Gas Monitor Particulate Samples Test Freg,uenc:,y Gross Gamma by Fission Continuous( 5)

Product Monitor Quantitative gamma 3 Times/Week(l) spectral analysis or gross beta ga.mma radio-activity analysis by internal proportional counter and qualitative gamma spectral analysis.

Tritium Radioactivity Weekly Chemistry (Cl and o2) 3 Times/Week Radiochemical Analysis Semiannual( 2) for E Determination Boron Concentration Twice/Week Boron Concentration Monthly Boron Concentration Monthly Boron Concentration Monthly Boron Concentration Monthly Gas Radioactivity by Continuous( 6 )

Air Ejector Gas Monitor Coolant Gross Radio-Twice/Week(6) activity Iodine Concentration Weekly( 3)

Radioactivity Analysis Prior to Release of Each Batch Radioactivity Analysis Prior to Release of Each Batch Iodine 131 and Partic-Weekly( 4) ulate Radioactivity FSAR Section Reference None None None None None None None 6.1.2 9.4 None None None 11.1 11.1 11.1 (1)

When radioactivity level exceeds 10 percent *of limits in Specification 3.1.4, or 3.1. 5, the 'sampling frequency shall be increased to a minimum of once each day.

7 "TABLE 4.2.1 (Contd)

(2)

Redetermine if:

(a)

The primary coolant radioactivity increases by more than 10 µCi/cc from the previous determination, and (b) upon each start-up only after a two-week equilibrium adjustment period shows a 10 µCi/cc in-crease from the previous determination in accordance with Specification 3.1.4.

(3)

When radioactivity level exceeds 10 percent of limits in Specification 3.1.5, the sampling frequency shall be increased to a minimum of once each day.

(4)

When iodine or particulate radioactivity levels exceed 10 percent of limit in Specifications 3.9.6 and 3.9.9, the sampling frequency shall be in-creased to a minimum of once each day.

(5)

A daily sample shall be obtained and analyzed if fission product monitor is out of service.

(6) If the air ejector gas monitor is out of service, the secondary coolant gross radioactivity shall be measured once per day to evaluate steam gener-ator leak tightness."

H.

Delete Appendix B "Interim Special Technical Specifications" in its entirety.

II. Discussion This proposed change to the Palisades Technical Specifications provides for safety limits governing reactor operation based upon the evaluation of the fuel and ECCS in accordance with the Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors as stated in 10 CFR 50.46(b) and Appendix K to 10 CFR 50.

The evaluation was performed by Combustion Engineering (CE) and transmitted to the Director of Regulation by letter dated July 9, 1975.

The evaluation is for first core fuel operating at a peak linear heat generation rate of 11.3 kW/ft prior to any cladding collapse.

The evaluation results and corresponding ECCS criteria for the worst primary pipe break (l.Ox double-ended slot break) are as follows:

Parameter Peak Clad Temperature Maximum Cladding Oxidation CE Calculation 2198°F*

0.061*

Criterion 2200°F 0.17

These numbers were derived using a containment pressure evaluation based on Branch Technical Position CSB 6-1.

8 The 11.3 kW/ft linear heat generation rate has been shown to f'ulfill the requirements of 10 CFR 50.46(b) and Appendix K to 10 CFR 50; it will be used as an operational limit as specified in the proposed Section 2.1.2 of the Technical Specifications until such time as other appropr.iate calcu-lations are performed.

During the Palisades Plant outage which lasted from August 1973 to September 1974, an interim fuel examination was conducted by Combustion Engineering

.at the request of Consumers Power Company.

One facet of this fuel examina-tion was a gamma scanning program to ascertain the location and size of any fuel column gaps that may have occurred as a result of densification and to determine fuel column lengths in order to clarify the extent of densifica-tion.

The gamma scanning equipment was positioned on the spent fuel racks in front of the fuel elevator at an angle of 60° to the south pool well.

A sodium iodide crystal was used with a photomultiplier tube and a single channel pulse analyzer.

The analyzer was calibrated to detect peak gamma activity for the zr95~ Nb95.

decay at approximately 0.75 Mev.

Calculations had been done to show that the activation of.zr95 in the claddfng was negligible by comparison to the activity of the zr95 fission product.

This decay energy peak, therefore, provided a reliable indication of fuel pellet occupancy in the fuel rod.

As a result, interpellet gaps were observed as sharp decreases in the recorded gamma signals.

A strip chart recorder was used to record the gamma ray intensities and the movement of the fuel rod as it passed the collimator aperture.

'.l.1he aperture width was set at 0. 060".

The four corner rods in each of twenty assemblies were scanned at a rate of 4" per minute.

Ten rods were found to have gaps greater than 0.060" wide.

A summary of those gaps and gap locations is presented in Table 1.

With the exception of the large gap at the bottom of rod A-38 NW, these

9 gaps are quite small and infrequent.

Because of the position of the gap in A-38 NW, it is not likely that this gap is due to densification.

Rod A-25 NE A-12 NE A-22 NE B-50 SE B-50 SW B-01 NW B-01 SW B-67 SW A-35 SE A-38 NW*

Possible TABLE 1 Summary of Fuel Stack Gaps Greater Than 0.060" Gap Width Gap Location (Inches)

{Inches From Bottom) 0.680 114 0.125 28.25 0.125 36 0.545 67.5 0.370 129.5 o.490 65.5 0.130 55.5 0.330 127.0 0.120 99.0 0.140 130.0 2.160 1.25 Loading Anomaly The infrequent occurrence of gaps, c*ombined with the small size of the gap~ observed, leads Consumers Power Company to the conclusion that a change in the*densification and collapse m0.del that has been applied to Palisades is appropriate.

A new model for calculation of flux peaking augmentation factors, based in part upon the data observed during this fuel examination is included in CENPD-139 "Fuel Evaluation Model." The proposed flux peaking augmenta-tion factors are based upon this model.

We further believe that the size of gaps observed in Palisades fuel pre-cludes the type of fuel rod collapse observed in selected other pressur-ized water reactors.

Not a single gap which can be attributed to densification has been observed that is as long as 3/4 of an inch.

10 Clearly, the stability of small gaps to support collapse must be far supe-rior to the 2 11 and greater gaps which have resulted in collapse at other pressurized water reactor facilities.

Because of the results of the Palisades fuel examination, we do not believe fuel collapse will occur.

Therefore, it is not necessary to maintain separate "post-collapse" rules for core average exposure above 10,265 MWd/MTU.

We believe that the Appendix K criteria provide more than adequate protection for required fuel integrity in the unlikely event of a Loss of Coolant Accident. It follows from this reasoning that the analysis supplied by Combustion Engineering applies to Palisades fuel at least through the remainder of the first cycle of operation.

10 CFR 50, Appendix K, requires that the evaluations take into account the effects of possible fuel pellet shrinkage.

Prior to the issuance of 10 CFR 50, Appendix K, fuel shrinkage was accounted for in calculations that were separate to the Interim Acceptance Criteria.

As these calculations were performed after the initial evaluation of the Palisades Plant under the Interim Acceptance Criteria, and at the time the calculations were performed, it was assumed that at some point in the future the need for performing fuel shrinkage calculations would no longer exist or they would be combined with some other criteria; the operating limits associated with fuel shrinkage were kept separate from the original Palisades Technical Specifications.

These operating limits were included in Appendix B to the Technical Specifications titled, "Interim Special Technical Specifications."

As the evaluation performed in accordance with 10 CFR 50, Appendix K, in-cludes the effects of fuel shrinkage, we have included in this Technical Specifications revision, changes which incorporate the still applicable "Interim Special Technical Specifications" limits into the original Technical Specifications.

Requirements which were included in the Interim Special Technical Specifications which we deemed to be no longer applicable, have been deleted.

These requirements are associated with postulated clad collapse; namely, limits,of rate of power increase, limits on primary sys-tem operating pressure, flux peaking augmentation factor and the linear heat rate limits associated with the then postulated post-collapse opera-tion.

11 The primary to secondary leakage limit, which was associated with collapsed clad, has been incorporated in the original Technical Specifications because of the tube wastage that has. been previously experienced at the Palisades Plant.

Even though the specific leakage limit is not justifiable solely on a technical basis, it is deemed prudent to continue a limit of this nature in effect Un.til tube wastage has shown to have been halted.

III.

Conclusion This change has not been reviewed by our Palisades Plant Review Committee or the Safety and Audit Review Board; however, based on its similarity to the proposed change submitted by letter dated November 4, 1974, we believe that their review of this proposed change will result in the conclusion that this change does not involve a significant hazards consideration.

Consumers Power Company R. A. Lamley, Vice PresjJ{'ent Sworn and subscribed to before me this 9th day of July 1975*

Sylvia B. Bftll Notary Public, Jackson County, Michigan My commission expires May 18, 1976 I

~

ML18347A784

Text

Subject:

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