L-83-429, Forwards Corrected Safety Assessment as Result of Thermal Shield Removal,Submitted w/830623 Ltr

Forwards Corrected Safety Assessment as Result of Thermal Shield Removal,Submitted w/830623 Ltr

ML17214A342
Person / Time
Site:	Saint Lucie
Issue date:	07/27/1983
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
L-83-429, NUDOCS 8308030197
Download: ML17214A342 (37)

Text

I'EGULATORY INFORMATION DISTRIBVTI SYSTEM (RIDS)

ACCESSION NBR 8308030197 DOC ~ DATE 83/07/27 NOTARIZED ~

NO FACIL:.50-335 St ~ Lucie Plant~

Uni~t 1~ Flor ida Power 8 Light" Co ~,

AUTH ~ NAME AUTHOR AFFILIATION UHRIGg R ~ E, Florida Power

& Light Co, REC IP NAME RECIPIENT AFFILIATION CLARKgR,AD Operating Reactors Branch 3

SUBJECT:

For wards corrected safety assessment as result of thermal shield removal~submitted w/830623 ltd DISTRIBUTION CODE:

A001S COPIES RECEIIbtED:LTR, ENCL

SIZE, TITLE:

OR Submittal:

General Distr ibution NOTES'OCKET

¹ 05000335 RECIPIENT ID CODE/NAME NRR ORB3 BC 01 COPIES LTTR ENCL 7

7 RECIPIENT ID CODE/NAME COPIES LTTR ENCL INTERNAL~

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O PLORIOA POWER & LIGHT COMPANY July 27, l983 L-83-429 Office of Nuclear Reactor Regulation Attention: Mr. Robert A. Clark, Chief Operating Reactors Branch f73 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Clark:

Re:

St. Lucie Unit I

Docket No. 50-335 Reactor Vessel Internals and Thermal Shield; P lant Recover Proaram In our letter of April l9, 1983 (L-83-230), Florida Power 8

Light Company committed to provide a safety assessment as a result of thermal shield removal.

This assessment was subsequently transmitted to you on June 23, l983 (L-83-369).

It has been determined that our June 23rd letter contoined three minor errors in the text, that do not change the conclusions made therein.

These changes are.

indicated in Attachment I of this letter.

This attachment should replace the attachment of ovr June 23rd letter in its entirety.

No other changes hove been made to our original submittal of June 23, l 983.

Very truly yours, Robert

. Uhrig Vice President Advanced Systems and Technology REU/DAC/cab cc: Harold F. Reis, Esquire Enclosure 8308030197'30727~

'Q~'DR

"*DOCK-05000335

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PDR, iI-"u PEOPLE... SERVING PEOPLE

ASSESSMENT OF SAFETY IMPACTS OF REMOVAL OF THERMAL SHIELD FROM ST-LUCIE UNIT 1

7he Gllying a

Pchmen s pr'ovi lpac 5 o~ removing he hermal Speci ica.lly, P.

ac>aient 1 aoc'.r ex-core Qe ector Qecal'bration, co e Qhysxcs; r ttacfDient 2 pres

o. "he Loss.o= Coolan-Accic;ent s.'elc per=ormeQ by exxon.

Qe an assessmen shielc

~rom Si.

esses he e zec h>Q, aul ic 1 i-.i 0 en" s the resl'1-s (LOCA ) wx hou i o"

he sa ety Lucie Uni l.

on rpnsien s,

orces ana o

a reanalysis h e th Brmal aQci ion -'o he ma. erial con a'neQ here'n, oo n =-xon anQ Combis ion =ncineering have per~ormeQ a Ie ie o: cur

~ Bn plan Technical Speciiica ions.

Both have concluceG "ha no fuel:

rela eQ techn'a.l specif ications neeQ be changeQ support o

- herm-1 shi e 1Q remova.l.

"asec on -'he material presenteQ, no ma)or lmpac to sP.'

v cue o

herma.l shield removP1 is seen.

ATTACHMENT 1 IMPACT OF REMOVAL OF THERMAL SHIELD ON TRANSIENTS The removal of the thermal shield in St. Lucie Unit 1 will result in an increased flow (about 1/2%) and an increase in downcomer liciuid volume (about 20%).

The increased flow will result in slightly more rapid transients in the primary loop.

In particular the excess load,'he loss-of-load and the steam line break will tend to respond more rapidly (by less than 1/2 second) at design flows-However, the increased volume delays cold leg transients by about 1/'3 second.

The net impact on transient timing will not be greater than a few tenths of a second at design flows, nor a few seconds at natural circulation flows-Hence no significant impact could be expected for temper-ature transients.

Flow transients are conservatively estimated in transient analysis.

Any impact on flow coastdown is undetectable.

Even the most rapid flow transient, the seized rotor, should not be noticeably affected by the increased flow.

The increased volume serves to increase the stored energy in the fluid and offsets, to some extent, the tendency of the pump to coast down more rapidly from higher flows.

The major effect of increased flow, for which there are no offsetting effects, is the increase in MDNBR caused by the increased flow.

Thus, the net effect on transient analyses of removal of the thermal shield should be negligible and probably beneficial.

EX-CORE DECALIBRATION DUE TO REMOVAL OF THE THERMAL SHIELD The ex-core neutron flux detectors are distal to the thermal shield and the downcomer.

Absolute calibration of the detectors is based on the in-core monitoring system and depends on the'mount of attenuation present between the core 'and the detectors.

Decalibration of the ex-core detectors during large system cooldowns is an expected phenomenon due to the density increase associated with the cooldown.

The current. safety analysis addressed this effect with the thermal shield present.

Without the thermal shield,'he larger volume of water between the core and the ex-core detectors will show a slightly larger decalibration effect.

This impact of the removal of the thermal shield could produce a minor penalty on the analyses in that it would reduce the pressure margin in the TM/LP trip by 10 PSI or so and the margin in the LPD LSSS by 0.6% for the worst cooldown events.

Neither of these values are significant when compared to the available margin.

a

~

HYDRAULIC LIFTOFF Increased flow through the reactor core will result in increased levitational forces on the bundles.

Based on a

quadratic flow dependence for pressure losses, it is anticipated that the lift-offforces will be about 1%

greater with 1/2S more flow.

For a dynamic pressure drop of approximately 40 PSID, which conservatively bounds the actual pressure ~i~ the increased levitation forces on a

9"X9" assembly would be 32 pounds.

Since the margin with the thermal shield exceeds 400 lbs, this increase repre-sents a loss of about 8% of the margin and thus lift-offis not a concern.

CORE PHYSICS The impact of the thermal shield removal on the core physics is negligible'o quantitatively assess the impact of the thermal shield removal upon the core reactivity and upon the core power distribution a two dimensional pin by pin PDQ'alculation was made with and without the thermal shield.

The core reactivity and core power distribution were identical between the two calculations.

~

0 performed with bounding ENC 14 X

14 fuel parameters and power history corresponding to:

beginning-of-life stored

energy, end-of-life fission gas

release, and end-of-life actinide decay power.

The fuel design parameters are shown in Table 1.

Calculated event times for this break are

.'given in Table 2, and final temperature and metal-water reaction results are shown in Table 3.

Results from the previous analysis are compared in Table 3 with those calculations in this study.

The system nodalization is identical to that used in the previous analysis and is shown in Figure 1.

The result of the analysis show a decrease of 59oF in the Peak Cladding Temperature (PCT) as compared to previous case.

The reduction in PCT is due primarly to improved heat transfer calculated during the blowdown phase of the analysis.

The improved heat transfer results from reduced coolant temperatures in the new analysis from about 10 seconds to the enC of blowdown.

The new calculated containment pressure is 0.35 PSIA higher than the previous case for the first llO seconds of the transients.

Th increase in the containment pressure is caused Jy an increase in the m ss flow to the containment building.

The increased containment pressure results in reduced steam binding which improves reflood rates.

The larg r downcomer volume produced by the removal of the thermal shield, slight'.y retards the reflood rate during the early portion of the reflood transient due to reduced downcomer liquid head.

Sufficient water however exists to fill the downcomer, within lO seconds after the start of reflood.

The improved blowdown heat transfer and higher containment pressure more than offset the effect on PCT that reduced downcomer liquid heads have on reflood rates during the early portion of the reflood transient.

System blowdown results for the 0.4 DECLG break are given'n Figures 2-10, hot channel results in Figures 11-14, containment pressure and

~

~

normalized power results in Figures 15-16, reflood results in Figures 17-19, and TOODEE2 heatup result in Figure 20.

The calculated peak cladding temperature for the 0.4 DECLG break is 2000oF with less than 3.% maximum local metal-water reaction,.and less

'thaf lf core-wide metal-water reaction.

These results show that for St.

Lucie Unit 1 operating at 2700 Hath and maximum LHGR of 15.30 KM/ft, including K power uncertainty, the emergency core cooling system will meet the acceptance criteria as presented in 10 CFR 50.46 with the thermal V

shield removed.

Table 1

CE 2x4 PN Data Primat.y Heat Output, NMt Primary Coolant Flow, ibm/hr Primary Coolant Volume, ft3 Operating Pressure, psia inlet Coolant Temperature, oF Reactor Vessel Volume, ft3 Pressuriier Volume, Total, ft3 Pressurizer Volume, Liquid, ft3 Accumulator Volume, Total, ft3 {one of four)

Accumulator Volume, Liquid, ft3 Accumulator Pressure, psia Steam Generator Heat Transfer Area, ft2 {one of-two)

Steam Generator'econdary Flow, ibm/hr Steam Generator Secondary

Pressure, psia Reactor Coolant Pump Head, ft Reactor Coolant Pump Speed, rpm Nceent:of Inertia, ibm-ft2/rad Cold Leg Pipe, I.D., in Hot Leg Pipe, I.D., in Pump Suction Pipe, I.D., in 2700>>

1.394 x 108 19,214' 2250 549 4402 1500 800 2020 1090 230

?4,722 5.899 x 106 885 280 88e 101,900 30 30

~ Primary Heat Output used in RELAP4-EN Nodel - 1.02 x 2700 2754 N~t.

~Includes total accumulator and pressurizer volume.

Table 1

(Continued)

Fuel Assembly Rod Diameter, in*

F~el Assembly Rod Pitch,'n" Fuel Assembly Pitch, in*

Fueled

{Core) Height, in*

Fuel Heat Transfer Area, ft~

Fuel Total Flow Area, ft~

Steam Generator Tube Plugging (Assumed Uniform)

~

ENC Fuel Parameters

Table 2

Large Break Results Time Sequence of Events Start Event Time of Event (seconds DECLG QC

- 0.4) 0.0 Initiate Break Safety Injection Signal Pressurizer Empties Accumulator Injection, Broken Loop.

Accumulator Injection, Single Intact Loop Accumulator Injection, Double Intact Loop End-of-Bypass Safety Injection Flow, SIS Start of Ref lood Accumulators Empty, Single Intact Loop Accumulators Empty, Double Intact Loop Peak Clad Temperature is Reached 0.05 0.92 8.8

20. 98

22. 95 22.95

27. 81

30. 96

46. 05

73. 81 74.11 165.0

Table 3

Analysis Results for 0.4 DECI.G Break w/Shield(1) w/o Shield Anal sis Results Peak Clad Temperature, oF Peak Clad Temperature Location, ft from bottom I ocal Zr/H20.Reaction (max),

X (at 395 sec)

Local Zr/H20 Location, ft from bottom Total H2 Generation, X of Total Zr Reacted

. Hot Rod Burst Time, sec Hot Rod Burst Location, ft from bottom Peak 'near Reit Generation

Rate, BOCREC, kW/ft DECLG 2059 9.22 4.0 8.7 c 1.0 39.8 8.0

.729 DECLG Vii~4~

2000 9.47 3.3 9.22 41.41

7. 47.

.728 Anal sis In ut License Core Power NWt Power Use for Analysis, IQt Peak Core Linear Power kW/ft Total Allowable Peaking Factor 2700 2754 15.00*

2.42

• Does not include 2C power uncertainty.

5KONDARY

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REf.Af'4-EN Bfoftdofttf'ystem Nodaffzatfon For St. Lucfe Unft 1

ST LUCIK lNXT le O.i OECLC SON I

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I 1O EO TXNE'SEC)

Figure 3

Olowdown TOtal Break,liinct)ntl Flow Bate, 0.4 DECLG Oreak

Attachment 2

ST. LUCIE UNIT 1 LOCA-ECCS ANALYSIS ARITH THERMAL SHIELD REMOVED Prepared by T. Tahvili This Attachment presents results of a LOCA-ECCS analysis performed for the St. Lucie Unit 1 Nuclear Reactor utilizing Exxon Nuclear Company (ENC) fuel.

This analysis differs from previous ENC analyses(1) in that the thermal shield was removed from. the Reactor Vessel.

The analysis was p rformed using the current NRC approved ENC evaluation models(2~3 4~5.6).

The results of this analysis show that the St. Lucie Unit 1 reactor can operate in conformance with 10 CFR 50.46 Appendix K criteria(7) with an allowed linear heat generation rate, including 2X for power uncertainty, of 15.30 KM/ft.

The total allowable peaking (FqT) remains at the valuo justified in the previous analysis of 2.42.

The methods used to perform the LOCA-ECCS analysis with the therm=-',

shield removed are the same as those used in the previous analysis(1).

The analysis assumed no increase in reactor vessel flow from the valu established in previous ENC analysis.

Revisions were made to the input to reflect changes in the lower downcomer flow area,

volume, and heat transfer

.area resulting from the thermal shield removal.

Pressure distributions vere revised to re-establish initial steady-state conditions.

Calculations were made for the limiting break identifi'ed in the previous analysis(1),

i.e., the double-ended cold leg guillotine break with a discharge coefficient of 0.4 (0.4 DECLG).

The analysis was

1C TIVE f SE'C1 Figure 0

01owdown Pressurizer Surge Line F1ow Rate, 0.0 OECLG Break

ST LUCXE'HXT 1 0 l OE'CLC'80N

~ I vxve <sec)

I-lqiire 5 Sing1r fntict. 1.nop hccimnt1at.or F1ow Rate, 0.4 OECLG Break

S'T LVCXE'NXT le 0.4 OE'CLC 80H TIVE (SEC)

F/gure 6

Ooub1e intact Loop Accumu1ator Flow Rate, 0.4 OECLG Break

Figure 7

Blowdown Average Channel Inlet Flow Rate, 0.4 DECLG Break

~Pb 8

ST LVCXC UHXY l. O.l DECLC SON I

~ g T.THE (SEC)

Figvre 0 Olowdown Average Channel Outlast Flow Rate, 0.4 OECLG Break

12.

Figure 9

Dlowdown Hot Chinnel Inlet Flow Rate, 0.0 DECLG Dreak

I

~..

r r

Figure 10 01owdown Hot Channel Pht1et F,low Rate, 0.4 DECLG Break

ST LUCID UHXT lo 0

5 OEGLC HC Q5 6I It I

tD tl vzvc onER ORE'AX

( Sec<

, s Figure 11 Blowdown Hot Rod Cladding Surface Temperature, 0.4 DECLG Break

ST LUCIE lNXT 1 0 l DEGLC HC R

N ZO U

TIH'E AFTER BRE'RK (SEC)

Ffqttrr. )7 lllnwrlown Ifnt. R>>il Ynl>>mrl.rir: Rvr,rage Tr.mprritt~rn, 0.4 OECLG Or<ok

ST LUCIC VHIT le 0. I OEQ C HC' I

R Q

t0 tl TIVE AFTER BRCAK t SEC)

Figure 13 Hot Rod 81owdown Heat Transf'er Coefficient, 0.4 OECLG Break

ST LUCIE UHlT le 0.$

OEGLC HC w5 MQ XD jE N

5)

LE 1l tl Kl TAHE AFTER SREAK

( SEC )

Figure 14 Hnt Rod 01nwdown Oepth of 7.irconium-Water Reaction, 0.4 OECLG Oreak

St. Lucie Unit 1, 0.4 OECt.G Containment Pressure 50

. 8l0-'50 300 TINE e SEC 350 450 Figure 15 Containment, Back Pressure versus Time, 0.4 DECLG Break

ST lUC?E'HIT j.

HP 0.$

OECLG Lf0 f00 t50

%0 TIVE AFTER BRE'Alt.'SECOHDS 550 I'igure 16 Normal'izr.d Power versus,T/me, 0.4 DECLG Break

Si'VCXE'HXT 1 O. I DECLG lD 1'

QO, f90 t40.

tl0 TI<<E, nf'TFR STnRT Or REFLOOD, SCCOIIDS I Iqirrr 17 I:nr~ F1nnrfinq Rntr.. 0.4 OFI;I.G Brrak

~

~

SY l.UCXK UNIT 1 O.i OECLG IO 15)

~ 1N '0 K40 tN TINE, AFTER START Ol-REFLOOD, SECONDS Fight>re 18 Downcome Mixture Leve1, 0.4 DECLG Break

LZO

'CO:

tN KiO tl0 llME, AFTER START OF REFLOOD, SECONDS Figure 19 Core Mixt.use Level, 0.4 DECLG Break

St. Lucie Unit. I, 0.4 OECN Hnt Rod Heatup a.

FCT @RE t~ ta AT S.ir ff. l t.

NPTWK9 NSC fNSK' AT t.ff FTe 1

40 0 td-d u0.0 aeo.e ee.e ee.e TIVE' SE'CORDS t!0.0 a0.4 "ig<<e 20 TOODEE7. Calculated Cladding Surface Temperature (PCT), 0.4 OECLG Break REFERENCES

,.a.

Exxon Nuclear

Company, St.

Lucie Unit 1

LOCA Anal sis Unpin the EXEM/PWR ECCS Model, XN-ecember upp ement anuary upp ement

- Revision 1 January 1983.

2.

I Exxon Nuclear Company, ENC ECCS Evaluation of a CE 2 X 4 PWR usin the EXEM/PWR ECCS Model Lar e reak xam e

ro em, XN-N - -,

up-p ement 3, u y 982.

3.

Exxon Nuclear

Company, Exxon Nuclear Com an WREM-Based Generic PWR ECCS Evaluation Model, N-a.

Volume I, July 1975

- b.

Volume II, August 1975 c.

Volume III, Revision 2, August 1975 d.

Supplement 1, August 1975 e.

Supplement 2, August 1975 f.

Supplement 3, August 1975 g.

Supplement 4, August 1975 h.

Supplement 5, Revision 5, October 1975 i.

Supplement 6, October 1975 j.

Supplement 7, November 1975 4.

Exxon Nuclear

Company, Exxon Nuclear Com an WREM-Based Generic PWR.

ECCS Evalution Model U

ate N

W M-N-

a.

July 1976 b.

Supplement 1, September 1976 c.

Supplement 2, November 1976 5.

Exxon Nuclear

Company, Exxon Nuclear Com an WREM-Based Generic PWR ECCS Evaluation Model U

ate N

M-A, XN-N 0.

6.

Exxon Nuclear Company, Exxon Nuclear Com an Evaluation Model EXEM/-

PWR ECCS Model U dates, N-ebruary a-Supplement 1, Narch 1982 b.

supplement 2, March 1982 I,

7.

"Acceptance Criteria for Eeergency Core Cooling Systems for Light Mater Cooled Nuclear Power Reactors,"

10 CFR 50.46 and Appendix K of 10 CFR 50; Federal Re ister, Volume 39, Number 3, January 4, 1974.

I