ML17207A771

Forwards Best Estimate Main Steam Line Break Analysis to Assess NSSS & Containment Response W/Automatic Auxiliary Feedwater Actuation,In Response to NRC

ML17207A771
Person / Time
Site:	Saint Lucie
Issue date:	01/24/1980
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
L-79-36, NUDOCS 8001310294
Download: ML17207A771 (42)

Text

REGULA I DRY IN'MATION DISTRIBUTION SYST (RIDS)

ACCESSION NBR!8001310294

'OC,DATE: 80/01/24 NOfARIZEO:

NO FACIE:50-335 St.

Luc.)e Planti Uriit 1r F'lor ide Power 4 Light Co ~

AO'fHsNAHE 'Ul'HOA AFFILTA'l'IOA UHH/t IN'D Florida Power 8 Light Co,.

HECIP",NAME AECIPIEAT." A'FF ELIATION RE Ibid s W ~

Operat'ing Reactors Branch 4

SUBJECT!

Forwards best estimate main steam line break analysis to.

assess NSS8 8 containment'esponse'w/automatic auxiliary feedwater actuationiin 'response to NRL'91221 ltr.

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P BOX 529100, MIAMI,FL 33162 FLORIDA POWER 4 LIGHTCOMPANY January 24, 1980 L-79-36 Office of Nuclear Reactor Regulation Attention:

Mr.

R.

W. Reid, Chief Operating Reactors Branch 84 Division of Operating Reactors U.

S. Nuclear Regulatory Commission Washington, D.

C.

20555

Dear Mr. Reid:

Re:

St.

Lucie Unit 1

Docket No. 50-335 Auxiliar Feedwater S stems The attached information is submitted in response to your letter of December 21, 1979 on the subject of auxiliary feedwater systems at St.

Lucie Unit l.

Please call if you have further questions on this subject.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems 8 Technology REU/MAS/cph Attachments (2) cc:

Mr, J.

P. O'Reilly, Region II Harold Reis, Esquire 8001 810 ~~/

PEOPLE... SERVING PEOPLE

ATTACHYNNT Re:

St. Lucie Unit l Docket No. 50-335 Auxiliar Peedwater Systems BEST ESTItSTE t>SLB l'HALYSIS TO ASSESS HSSS MD COHTAIHHEttT RESPOi'ISE !IITH AUTOS'IATIC AUXILIMYFEED!STER ACTUATIOih

1.0 IWTRODUCTIOW A set of calculations has been performed on a generic basis with plant characteristics representative of CE operating plants to model containment building pressure and temperature response and overall WSSS behavior.

including core reactivity, following a Hain Steam Line Break (tiSLB) inside containment.

The intent of these calculations is to determine if the containment building response (pressure) and the core reactivity response (return to power) are acceptable following a HSLB when auxiliary feed-water is added without regard to the identification of the affected steam generator.

The auxiliary feedwater flow is assumed to be activated at the initiation'f the transient to maximize its effects.

Hain feedwater flow

- including post trip rampdown is simulated.

Wo isolation of main or, auxiliary feedwater is considered unless a high water level condition is reached.

2.0 ASSUHPTIOWS AWD CASES Assunptions for the analyses are given in Table 1.

The four cases analyzed are listed in Table 2.

3.0 DISCUSS IOW OF RESULTS Haximum containment pressure and least negative core reactivity for the four cases are listed in Table 3.

Both the containment pressure and the reactivity (return to power) values are within 'acceptable limits.

Hain feedwater flow, auxiliary feedwater flow, core reactivity change, core power, containment'pressure, primary loop temperatures, and steam generator secondary temperatures for the four cases are detailed in Figures A-1 through A-7, B-l through B-7, C-1 through C-7, and D-1 through D-7, respectively.

The results of the analyses using best estimate models for steam generator moisture carryover and containment passive heat sink heat transfer demonstrate that the additional auxiliarv feedwater has a negligible

'impact on containment peak pressure.

The containment peak pressure is determined primarily by the initial inventory in the ruptured unit-This

inventory is released within the first few minutes, depending

.upon the break size, so that the contribution of auxiliary feedwater flow to the ruptured unit over this time frame is small.

Over the lonqer tim

frame, the secondarv inventorv is boiled off at essentiallv the decay heat rate which the containment active heat removal'vstems can accommodate while reducing containment pressure.

Th'e excess,feedwater which is not boiled off remains in the steam generator, causing the secondary level to rise.

The containment peak pressure is essentially an initial inventory limited phenomenon.

The results of the analyses also show that the additional auxiliary feed-water has a negligible impact on core reactivitv.

Cases h and C assume no stuck rods and a best estimate moderator cooldown curve.

For comparison, Cases B,and D assume that the most reactive rod is stuck and that the

-moderator cooldown curve is a licensing curve.

All case took credit for boron injection via three charging pumps; however, safety injection boron credit was not taken.

These cases do not have a return to power for the following reason.

The initial primary loop temperature decreases are linited 2

by the two-phase blowdown process associated with large break P'2 ft),

since much of the break flow is saturated liquid which has not absorbed significant amounts of energy from the primary loop.

For smaller break areas

((2 ft ), the blowdown is pure steam which does require large amounts of energy per unit mass to boil via primary to secondary heat transfer; however, the rate of primary-to-secondary heat transfer is controlled by the blowdown flowrate which in turn is limited by the small break area.

The net result is that over. approximately the first 100 seconds of the event, the amount of core and loop cooldown is about the same regardless of break size.

This time frame is most important since.the presence of delayed

=neutrons minimizes the amount of cooldown needed to produce a core criticality problem.

without a return to power (via primary loop cooldown and delayed neutrons),

the remainder of the.transient is a gradual increase in reactivity <<9 <<

. loop cooldown which is coupled to the containment pressure, plus a decrease in reactivitv due to boron injection.

In time (approximatelv 300 second ),

the reactivitv decrease due to boration overtakes the reactlv'Itv increases due to loop cooldown; thereafter, the total reactivitv steadil<<<<<<>>es The ruptured steam generator is at the containment backpress<<e and ""th

0 RCps operating the sensible heat from the non-ruptured unit is quickly removed resulting in RCS and SG secondary temperatures essentially in equilibrium with the containment conditions in about.10 minutes.

With licensing assumptions, the peak in the reactivity transient is calculated to be within the first two minutes of the event.

A two minute time delay, if added to the automatic actuation circuit, would justify a statement that automatic auxiliarv feedwater actuation will not impact existing SAR core cooldown HSLB analyses.'.0 COMPARISON 1lITH LICEHSItlh CALCULATIOllS The following items are important in comparing. the results contained herein with those'obtained with traditional licensing models and assumptions.

1.

The moisture carryover model used is a best estimate model which gives a two-phase blowdown for large break areas.

The two-phase blowdown-results in a lower containment pressure and less initial primary loop cooldown than a pure steam blowdown.

Chapter 15 analyses assume a pure steam blowdown regardless of break size.

. 2.

Chapter 15 analyses assume that the most reactive rod is stuck.

tloreover, the remaining rod worth is assigned a conservative value in conjunction with a conservative moderator cooldown curve.

3.

A best estimate containment heat transfer model provides containment pressurization results significantly lower than those provided in Chapter 6 analyses.

ASSUNPTIOllS t<SSS Initial Conditions Power Core Inl.et Temperature Primary Pressure Secondary Pressure Secondary Temperature 2700 H';lt 548'f 2250 PSIA 875 PSIA 529'F Containment Data Free Volume Design Pressure Heat Sinks Heat Transfer Hodel Number of Fan Coolers Fan Cooler Capacity, each Fan Cooler Actuation Setpoint tlumber of Sprays Spray rate, each Spray Actuation Setpoint Other Data 2.5 x 10 ft 44 psig CESAR values Best estimate model 4 (no single failure) 68 x 10 B/hr at 280 F

6 containment temperature 100'F CCH Temperature Fans are operational 9 t = 0 2 (no single failure) 2700 GPf1 10 PSIG + 60 seconds Steam Generator Isolation Signal (t<SIS) setpoint Decay Heat Curve Hain Feedwater Flow Ruptured Unit:

500 psia AHS-5 Pamped to 10'.l over 60 seconds following Reactor !rip: (10,.

represents twice the bypass nominal value or 5':, this accounts for purp run-out wi" reduced backpressure),

temperature is r=duc=d to lo" to account for turbine off-Flow terminated if the elevation of upper 'vel tao.

is reached.

See Figures A-l, B-l, C-l, and D-l.

'BLE 1 continued Hain Feedwater Flow continued Unaffected Unit:

Auxiliary Feedwater Flow Ruptured Unit:

Unaffected Unit:

Reactor Coolant Pumps CEA Insertion Worth All rods in (API) ttost reactive rod stuck t'1oderator Worth SAR Yalue Best Estimate Yalue

. Doppler 'Worth Noisture Carryover On Steam Generator Secondary Side Boron Injection Parameters Safety Injection Charging Pumps f(umber of Pumps Flow Pate Actuation Time Boric Acid Concentration Boron llorth Boric Acid Conversion Factor Hixing Model Used Loop Transi" Time Same as ruptured unit except that flow is ramped to 5~.

See Figures A-l, B-l, C-1 and O-l.'

Initiated at t = 0..

Flow rate is a function of unit pressure.

All control valves assumed to be fully opened.

Wo flow; all flow i,s totally diverted to the ruptured unit.

Operating during the transient.

-8.9Ã (no stuck rod)

-7.12~

(best estimate)

See Figure 1

See Figure 2

See Figure 3

Best Estimate Hodel Credit !1ot Taken 3

44 GP!1 per pump SIAS 8Ã bv weight 00 PP!1/"

1709 PP!1 boron/;l by weight boric acid Slug Flow l1odel 10.5 seconds

TABLE 2 CASES Case CEA Scram worth ttoderator Curve Break Area Ft2 D

-8.9

-7.12

-0. 9

-7.12 Figure 2 Figure 1

Figure 2 Figure 1

6.63 6.63 1.99(2)

(1)

Double-ended severance of main steam line (trio-phase blovIdovin).

(2)

Largest break area corresponding to pure steam bio@down.

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'TABLE 3 RESULTS Case 1east'f/egative

'"'ontainnent Peak Pressure (PSIh)

Core Reactivitv" 29.7/83.0 (sec.)

29.7/83.0 (sec.)

35.0/231.9 (sec.)

'5.0/231.9 (sec.)

-0.31

'2.34

-3.54

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( SEC.)

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200 400 600 T I f'IE

( SEC

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200 400 600 T I YiE

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