L-83-369, Forwards Assessment of Safety Impacts of Removal of Thermal Shield from Facility,Per 830419 Commitment.Assessment Addresses Cycle 6 Reload Safety Analysis of Plant Recovery Program

Forwards Assessment of Safety Impacts of Removal of Thermal Shield from Facility,Per 830419 Commitment.Assessment Addresses Cycle 6 Reload Safety Analysis of Plant Recovery Program

ML17214A255
Person / Time
Site:	Saint Lucie
Issue date:	06/23/1983
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
L-83-369, NUDOCS 8306280250
Download: ML17214A255 (37)

Text

REGULA Y INFORMATION DISTRIBUTI SYSTEM (RIDS)

AGCE'SSIOA~ NBR~8306280250-,

DOC ~ DATE: 83/06/23 NOTARIZED; NQ FACIL:50 385 St'< Lucie Planti Uni't li Florida Power L Light Co>

AUTH,NAME AUTHOR AFFILIATION UHRIGr R ~ Ee Florida Power L Light Co, RECIP ~ NAME RECIPIENT AFFILIATION CLARKpR.Ai Operating Reactors Branch 3

SUBJECT:

Forwards assessment of safety impacts of removal of thermal shield from facilityiper 830019 commitments Assessment addresses Cycle 6 reload safety analysis of plant recovery DISTRIBUTION CODEX'05AS COPIES RECEIVED:LTR ENCL SIZE:.

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TITLE:

OR Submittal:

Steam Generator Sleeving Review, for PNR's NOTES:

DOCKET 05000335 RECIPIENT ID CODE/NAME NRA ORB3 BC 01 INTERNAL: AEOD 16 IE/DEPER 0IR NRR/DE/MEB 06 NRR/DL/ORAB 10 R ~/M

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i' Ct FLORIDA POWER & LIGHT COMPANY.

June 23, 1983 L-83-369 Office of Nuclear Reactor Regulation Attention:

Mr. Robert A. Clark, Chief Operating Reactors Branch 83 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; Plant Recovery Program (Cycle 6 Assessment of Safety Impacts of Removal of Thermal Shield from St. Lucie Unit I)

In our letter of April l9, l983 (L-83-230), Florida Power 8

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

As an enclosure to this letter, please find such a safety assessment addressing the impact to the current (Cycle 6) Reload Safety Analysis submitted in our letters of January 20, l983 (L-83-27) and February 8, l983 (L-83-57). The attached evaluation supports the conclusion that the removal of the thermal shield causes no significant impact to these analyses.

As a result, the conclusions of the Cycle 6 Reload Safety Analysis remain valid and support Cycle 6 operation without the thermal shield.

Very truly yours, Robert E.

rig Vice President Advanced Systems &'Technology REU/CGO/cab cc: Harold F. Reis, Esquire Enclosure 8gQQQEIQg?50 03~5 pDR,ADOCK

pDR, PEOPLE... SERVING PEOPLE

ASSESSMENT OF SAFETY IMPACTS OF REMOVAL OF THERMAL SHIELD FROM ST.

LUCIE UNIT 1

SUMMARY

The following attachments provide an assessment of the safety impacts of removing the thermal shield from St. Lucie Unit l.

Specifically, Attachment 1 addresses the effect on transients, ex-core detector decalibration, hydraulic liftoffforces and core physics; Attachment 2 presents the results of a reanalysis of the Loss of Coolant Accident (LOCA) without the thermal shield performed by Exxon.

Xn addition to the material contained herein, both Exxon and Combustion Engineering have performed a review of current plant Technical Specifications.

Both have concluded that no fuel related technical specifications need be changed in support of thermal shield removal.

Based on the material presented, no major impact to safety due to thermal shield removal 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 volume (about 20%).

The increased flow will result

.in slightly more rapid transients in the primary loop.

In particular the excess load, the 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 temperature 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 fromhigher 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 amount 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 decrease associated with the cooldown.

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

Without the thermal shield, the 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.

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-off forces will be about 1% greater with 1/2% more flow.

For a dynamic pressure trip of approximately 40 PSID, which conserva-tively bounds the actual pressure drop, the increased levitation forces on an 9"X9" assembly would be 32 pounds.

Since the margin with the thermal shield exceeds 400 lbs, this increase represents 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.

To 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 calculation was made with and without the thermal shield.

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

I I

Attachment 2

ST.

LUCIE UNIT 1 LOCA-ECCS ANALYSIS WITH THERMAL SHIELD REMOVED Prepar'ed 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 performed 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(>).with an allowed l.inear heat generation rate, including 2X, for power uncertainty, of 15.30 KW/ft.

The total allowable peaking (FqT) remains at the value 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 value 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 were revised to re-establish initial steady-state conditions.

Calculations were made for the limiting break identifi'ed in the

(

previous analysis(>), i.e.,

the double-ended cold leg guillotine break with a di scharge coefficient of 0.4 (0.4 DECLG).

The analysis was

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 end of blowdown.

The new calculated containment pressure is 0.35 PSIA higher than the previous case for the first 110 seconds of the transients.

Th increase in the containment pressure is caused by an increase in the ma=s flow to the containment building.

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

The larger downcomer volume produced by the removal of the thermal shield, slighty 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 10 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'in Figures 2-10, hot channel results in Figures 11-14, containment pressure and

3 w

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 3.3X maximum local metal-water reaction, and less

'that'X core-wide metal-water reaction.

These results'how that for St.

'-:Lucie Unit 1 operating at 2700 MWth and maximum LHGR of 15.30 KW/ft, including 2X power uncertainty, the emergency core cooling system will meet the acceptance criteria as presented in 10 CFR 50.46 with the thermal shield removed.

Table 1

CE 2x4 PMR Data Rrimary Heat Output, Mlt Primary Coolant Flow, ibm/hr Primary Coolant Volume, ft3 Operating Pressure, psia Inlet Coolant Temperature, oF Reactor Vessel Volume, ft3 Pressurizer Volume, Total, ft3 Pressurizer Volume, Liquid, ft3 Accumulator Volume, Total, ft3 (one of four)

Accumulator Volume, Liquid, ft3 Accumulator Pressure, psia Steam Genorator Heat Transfer Area, ft2 (one of two)

I Steam Generator Secondary Flow, ibm/hr Steam Generator Secondary

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

1.394 x 108 19,214**

2250 549 4402 1500 800 2020 1090 230 74,722 5.899 x 106 885 280 886 101,900 30 42 30

• Primary Heat Output used in RELAP4-EM Model - 1.02 x 2700 2754 NHt.

~~Includes total accumulator and pressurizer volume.

Table 1

(Continued)

Fuel Assembly Rod Diameter, in*

Fuel Assembly Rod Pitch, in~

Fuel Assembly Pitch, in*

Fueled (Core) Height, in*

Fuel Heat Transfer Area, ft~

Fuel Total Flow Area, ft~

Steam Generator Tube Plugging (Assumed Uniform)

.440

.580 8.180 136.7 50,117 53.19 5%

" ENC Fuel Parameters

Table 2

Large Break Results Time Sequence of Events Start Event Time of Event (seconds)

DECLG

~(C

= 0.4) 0.0 Initiate Break Safety Injection Signal

'ressurizer 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 1ood 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 DECLG Break w/Shield(1) w/o Shield Anal sis Results Peak Cl ad Temperature, oF Peak Clad Temperature Location, ft from bottom Local 2r/H20 Reaction (max),

X (at 395 sec)

Local Zr/H20 Location, ft from bottom Total H2 Generation, X of Total 2r Reacted Hot Rod Burst Time, sec Hot Rod Burst Location, ft from bottom Peak 'near Heat Generation

Rate, BOCREC, kW/ft DECLG

~~C~~

2059 9.22 4.0

, 8.7

<1.0

39. 8 8.0

.729 DECLG

~~=-K4) 2000 9.47 3.3 9.22

<1.0

41. 41

~ 7.47

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

2.42

• Does not include 2X power uncertainty.

1

%CO@DARY STEAM GEMRATO%

MCONDARY 4$

,. ST%~ Gft4lATOC 50 TUBS LOOP I LOOP 2 NfSSUOZE R SG TLSES 15 2

32 OVTLE1

~Q NUM INLET PQ NUM HOT LEG 21 U~'Ao Ib ONTK)

ROD SHkOLDS

'IO

~R PLENUM 33 Id HOTLEG 12 14 INLET PLENUM 24 OUTQT PQ' 25 51 42 50 4t 41

~P SUCTION 4'I 42 34 35 P SUCTIQ AM N

44 COLD LEG 37 47 LPSI HPSI 4b

>44 LPf% R CORE b

10 2t COLD uC 22 31 LPSI Ht51 30 21 27 Tt PU QJCT ON ODLD LED 3t ACCUMUlhTOR L

LEG 45 E

5t 5

I 4

I 8

~U

~e 0 4 50vt 12 24 ACCUMULATO'R LPSI HPSI....

42 bl SU~RT ASSEMlLY 5t 54 LOUR ruNUM 51 ACCUMULATOR CONTAINMTNT Fiaure 1

RELAP't-EM Blowdown System Noda1ization For St. Lucie Unit

ST t UCX'E LNXT 1 0-k OECLQ BOB 1C t0 TIME t SEC )

1 I jqtire 2 Blowrlown Sys'.r.'m Pr~"".,ijr~, O.~i DECLG Orr..ak

ST LUCIE'NIT 1e 0 ~ 5 OECLG BON 1C to TIVE' SE'C)

Figure 3

Blowdown Total Break iluncl.)on Flow Rate, 0.1 OECLG Oreak

ST LUCIE UNIT 1 0 ~ 4 OE'CLG BOB LC t6 TIvE (SE'C)

Figure 0

Blowdown Pressurizer Surge Line F1ow Rate, 0.0 OECLG Break

ST LUCIE lNXT 1.

0 5

DE'CLC 8DN TXHE ( SE;C )

Fiqiire 5 Singlr. inLact. I.nop Rrconnilator Flow Rate, 0.0 DLCLG Break

ST LUCXC UNIT le 0 I D'ECI G BDH I.

=O MN cD TIME (SE'C)

Figure 6

Double Intact Loop Accumulator Flow Rate, 0.4 DECLG Break

ST LUCXF I+IT X, O.i OE'CLG BDH 1E lS CO TXVE' SEC)

Figure 7

Blowdown Average Channel Inlet Flow Rate, 0.4 DfCLG Break

ST LUCI'E UHXT 1a 0.4 DEC G BDN 8

I-4J 8

Km TX~E rsEc)

Figure 8

Blowdown Average Channel Outlet Flow Rate, 0.4 DECLG Break

ST LUCIE UHIT l.

0 5

DE'CLG BON 12.

10 KO TXHE l SEC )

Figure 9

Blowdown Hot Channel Inlet Flow Rate, 0.0 OECLG Break

ST LUCIE'HXT 1 D.4 DCCLC BOB CD g 4

Og le t1 TIVE' SE'C)

Figure 10 Blowdown Hot Channel Outlet Flow Rate, 0.4 DECLG Break

ST LUCXE UHXT le 0 4'OEGLC HC' ZD t.l TILDE 4FTER BREAK ( SEC )

Figure 11 Blowdown Hot Rod Cladding Surface Temperature, 0.4 OECLG Break

4.

R$

MS K~

o ST LUCIE UNIT lo 0 I DEGLC HC 1t 1l ZO TIHE AFTER BRCAK ( SEC )

3t l-igi)rn 17 lflnwrlown Idol; Rorl VnliimrI.ric Rvrrnge Temperate>rr.,

0.4 OECLG Break

b.

I I-~

LL g 9

CJ ST LUCIE UNIT le 0

4 DEGLC HC'E 1C t0 U

TINE AFTER BREAK (SEC)

Figure 13 Hot Rod Blowdown Heat Transfer Coefficient, 0.4 DECLG Break

ST LUCIE UNIT le 0-4 OEGLC HC I4

~S Oe lt N

t0 zL TIHE'FTE'R SREAK (SEC)

Figure 14 Hot Rod Blowdown Depth of 7.irconium-Water Reaction, 0.4 DECLG Break

St. Lucie Unit 1, 0.4 OECLG Containment Pressure 50 100 F50 300

TXWE, SEC 350 400 450 500 Figure 15 Containment Back Pressure versus Time, 0.4 DECLG Break

ST LUCIE UNIT 1 hlP 0 ~ 5 DECLG I.

I 50 100 150 t00 t50

$00 TINE AFTER SREAKi SE'CONDS 550 400 Figure 16 Normalized Power versus Time, 0.4 DECLG Break

ST LUCXE UHXT 1 0.$

DECLG.

40 1ZO 1CO 5N UO tl0 TINE, RFTER STRRT OF REFI OOD, SECONDS I iqitrr 17 Cnr<

Flnnrlinq Ratr,,

O.A,OFCI.C Brnnk

ST I UCXE UNIT 1 O.h OE'CLC 10 tN ti0 t$ 0 TIME, AFTER START OF

REFLOOD, SECONDS Figure 18 Downcome Mixture Level, 0.4 DECLG Break

Sf 'LUCXE UNIT 1 0.$

OE'CLt:

40 IO u,O m0

.ZOO HO tSO ilME, AFTER START OF

REFLOOD, SECONDS Figure 19 Core Mixt.ure Level, 0.4 DECLG Break

St. Lucie Unit 1, 0.4 DECLG Hnt Rod Heatup lo PCT HO%

tHOOE tX At S.i1 FY. )

t.

R~REO INK (NXK Q AT 7.if FT.

1

~.I..

.0 io.o 00.0 u0.0 eo.o ee.o tio.o TIHE SECONDS t40-0 25.0

%0.0

~ Figure 20 TOODEE7 Calculated Cladding Surface Temperature (PCT), 0.4 DECLG Break

~

3 0 REFERENCES Exxon Nuclear

Company, St.

Lucie Unit 1

LOCA Anal sis Usin the EXEM/PWR ECCS Model, XN-NF-82-98, December

982, 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 Exam e

ro em, XN-NF 0, up-p ement 3, Ju y 982.

3.

Exxon Nuclear

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

- b.

C.

d.

e.f.

g' h.l.

J ~

Volume I, July 1975 Volume II,. August 1975 Volume III, Revision 2, August 1975 Supplement 1, August 1975, Supplement 2, August 1975 Supplement 3, August 1975 Supplement 4, August 1975 Supplement 5, Revision 5, October 1975 Supplement 6, October 1975 Supplement 7, November 1975 4.

Exxon Nuclear

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

ate N

WR M-, XN-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 date ENC WREM-IIA, XN-NF 0.

6.

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

PWR ECCS Model U dates, N-N -

ebruary 9

a.

Supplement 1, March 1982 b.

supplement 2, March 1982

. 7-

"Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors,"

10 CFR 50.46 and Appendix K of IDDFR RD; ~Fd I

R I 9, 31 39.

II 3

3.

1 94,'1914.

0

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