ML20011A844

Forwards Addl Info Requested in NRC 810918 Ltr Re Transient & Accident Analyses in Basic Safety Rept

ML20011A844
Person / Time
Site:	Millstone
Issue date:	10/27/1981
From:	Counsil W NORTHEAST NUCLEAR ENERGY CO.
To:	Clark R Office of Nuclear Reactor Regulation
References
A02000, A2000, TAC-54199, NUDOCS 8111030274
Download: ML20011A844 (11)

Text

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gg g General offices e S.Iden Street, Berlin, Connecticut 3 572fJUECEf~ $ fro *nN CONNECTICUT 06101

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<5 Docket No. 50-336 Ao2mo eg h L Il0V021981> 8 Direr: tor of Nuclear Reactor Regulation Attn: Mr. Robert A. Clark, Chief 8'* %$ s[

\ S Operating Reactors Branch #3 ,p ,4%

U. S, Nuclear Regulatory Commission (gg Washington, D.C. 20555 -

References:

(1) R. A. Clark letter to W. G. Counsil, dated September 18, 1981.

Gentlemen:

Millstone Nuclear Power Staton, Unit No. 2 Additional Information on Basic Safety Report In Rer'erence (1) , the NRC Staff requested Northeast Nuclear Energy Company (NNECO) to provide additional information regarding the transient and accident analyses presented in the Millstcnt. Unit No. 2 Basic Safety Report (BSR) .

In responsa to that reqtaest, NNECO hereby provides the attached information.

We trust you find this information responsive to the Reference (1) requests.

Very truly yours,

'iO?'DIEAST NUCLEAR ENERGY COMPANY 8/II W. G. Counsil RE Senior Vice President 1

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Docket No. 50-336 J

9 Attachment >

Millstone Nuclear Power Station, Uldt No. 2 Additional Information Basic Safety Report '

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k October, 1981

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Question 1 Provide an analysis of the .Feedwater Line Break Accident for Millstone Unit No. 2. Include a specific discussion relative to the applicability

of the analysis to the Millstone reload applications.

i

Response

! - The Feedwater Line Break Accident _ Scenario is not included in .the design bases for Millstone Unit No. 2. As such, a roload specific analysis' of this event is not included in Section 5 of the Basic Safety Report (BSR) .

NN2CO has provided the NRC Staff the results of an analysis of.a feed-water line rupture event in the W. G. Counsil letter to R. A. Clark, I dated June 16, 1980. This analysis was provided to the Staff for in-F formational purposes only, and NNECO references it here for identical reasons.

Question 2 Clarify the reference to the THINC code in Section 5.2.6.5 of the BSR 1.e., verify that an NRC-approved version of THINC is used'in reload

j. applications. If not, provide documentation for the staff's review.

Response

The THINC code used in the Millstone Unit 2 application is the same as that currently used on W 14x14 and 15x15 plants. This code is

, briefly described in Section 3.4 of the BSR and had been documented in References 1, 3, and 4 of Section 3.8 of 'tlua BSR. These referenced WCAP's have been provided to the staff.

Question 3 Provide a qualitative discussion of thc broken pump shaft accident and, if more limiting than a pump siezure, perform;a quantitative analysis -

to show that die acceptance criteria of a calculated dose rate which is a small fraction of 10CFR Part 100. guidelines and a maximum pressure' within 110% of design are met.

l-

Response

The broken pump shaf t . accident is not included in the design bases for

(

j Millstone Unit. 2. 10ECO has reviewed the broken shaft scenario and has determined that this event is bounded by the consequences of a locked H reactor coolant pump' rotor. event based on the following.

Both events rely on the' same reactor protection ' system trip function,

! namely the low flov trip set at 89% 'of design reactor coolant flow.

While the locked rotor event assumes an instantaneous flow decrease- to .

i' -

75% with core power at 100%, a broken' shaft scenario would allow the pump impeller to spin *o a stop at some definite rate due to 'the _ momentum associated with the component. Until the reactor trip, the DNB be-

= havior for both events is similar. 'After-the reactor trips, flow will-

{ decrease much more rapidly in the locked rotor. scenario resulting in,a larger reduction in' DNB margin. then for the case of the broken shaft

, scenario.

L

Question 4 For the loss of feedwater event, provide an analysis assuming the pressurizer power-operated relief valves fail to open.

Response

The loss of normal fee ter accident was reanalyzed assuming that the pressurizer.

power-operated relia' .alves are inoperable. All other analysis assumptions remain unchanged ' .< chose described in the BSR.

The protect 0 against a loss of normal feedwater is provided by a reactor trip on low sta jenerator water level in any steam generator. Following the reactor J turbine trip from full load, the water level in the steam generators will f_. due to the reduction of steam generator void fraction and because steam flow through the safety valves continues to dissipate the stored and generated heat. Ten minutes following the initiatinn of the low level trip, the two motor-driven auxiliary feedwater pumps supplying a total of 600 gpm are manually started, reducing the rate of water level decrease.

The capacity of. the auxiliary feedwater pumps is such that the water level in the steam generator being fed does not recede belcw the lowest level at which sufficient heat transfer area is available to dissipate core residual heat wi';hout water relief from the RCS safety valves.

Figures 1 through 4 show the significant plant parameter transients following a loss of normal feedwater, with the pressurizer PORV's assumed inoperable.

The sequence of events for this accident are listed in Table 1.

Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system since the auxiliary feedwater capacity is such that only steam, not reactor coolant water, is discharged through the pressurizer safety vahas, and the water level in the steam generator receiving feedwater is maintained above the tubesheet.

TABLE 1

, TIME SEQUENCE FOR THE LOSS OF NORMAL FEEDWATER -

(PRESSURIZER PORV'S FAIL TO OPERATE)

EVENT TIME (SEC) h Loss of normal feedwater 0.0 High pressurizer pressure trip setpoint (2422 psia) reached - this trip is ignored 15.6 Steam generator safety velves begin to open (1000 psia) 16.0 i Pressurizer safety valves open (2500 psia) 18.0 High oower level trip setpoint reached (12 r.arcent) -

this trip is ignored 26.0 Maximum pressurizer pressure (2570 psia) 26.0 Low SG water level trip setpoint (36 percent of span) reached - reactor trip process begins 26.3 CEA motion begins 27.2 Minimum DNBR (1.75) 28.0 Maximum pressurizer water volume (1130 cu. ft.) 30.0 Maximum steam pressure (1070 psia) 30.0 Pressurizer safety valves close 30.0 Auxiliary feedwater delivery begins 600.0 Minimum steam generator mass (tubesheet is not uncovered) 654.0

1.2000  :  :  :  :  :

l.0000 - -

i o

z b .80000 -- --

u m

6

~

m

. 60000 - - --

LaJ 2

o a.

m .40000 -- --

u o

z

.20000 - - --

0.0 -

o l

o o

o o o o o o o o o e o o e o o o

. o o o o o o o - ~ m w e w TIME (SEC) .

1 FIGURE 1 MILLSTONE 2 - SAFETY ANALYSIS LOSS OF NORMAL FEEDWATER NUCLEAR POWER VS TII.1E 1

6

- y ., -.m+.w_p..,p_.~%,. .., ..m,,,y-,n w,,.- , , ,-.p._.,y. ,, . , ..q.,_.,_._y, -

,y,.w,_,,e,_. .,,e,,,,.., ,e-,y.m.,.. .,

,#7., ,.mw., .-.c ..--.

~ _

2600.0 i  ;  ; .-

q 2500.0 - ..

- 2250.0 - ..

w Q.

[ 2000.0 - - ..

E g 1750.0 -- __

a.

ce w

5 1500.0 - - --

g 1250.0 -- --

c.

1000.00 -- --

800.00 o o

o o

o o

o o o 'o o o o b O b o S 2 8 8 8 TIME (SEC)

FIGURE 2 MILLSTONE 2 - SAFETY ANALYSIS LOSS OF NORMAL FEEDWATER PRESSURIZER PRESSURE VERSUS TIME f

% ,vn,-a,-ge,. , - , , - , , - - , . -,w yw m,-,-~,e-, -- e ,--n. .-y, rey, ,.e y ,4 e,---v- --m - -- - -- = g r rw w

1550.0  :  :  :  :  :

1400.0 -- --

2 3

C -

1200.0 -- --

d

> A e []

W 1000. 00 -- --

4 3 _

3 800.00 -- --

5 0

e 600.00 -- --

400.00  :  :  :  :  :

8 8 8 8 8 8 o o e o o e o o R R S S l TIME (SEC)

FIGURE 3 MILLSTONE 2 - SAFETY ANALYSIS LOSS OF NORMAL FEEDWATER PRESSURIZER WATER VOLUME VERSUS TIME

650.00 - l  :  ;  ;  :.- t 625.00 - -

600.00 - -

C o

La e 575.00 - --

' o r

550.00 - -

525.00 - -

500.00  :

:  : =

o o o o o o o o o o o o e o o o o o- o-6 2 8 2 8 8 TIME (SEC)

FIGURE 4 MILLSTONE 2 - SAFETY ANALYSIS LOSS OF NORMAL FEEDWATER T VERSUS TIME AVG W

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.b . ,

m

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Question 5a

'For the steamline rupture event (Section 5.3.15) , provide a qualitative '-

-discussion of the event assuming loss of offsite_ power.

Response

In the case of a steam line rupture with a loss of offsite power, an additional delay would be assumed following the generation of a safety '

injection signal to start the diesel generators and to commence loading the -safety injection _ equipment onto them. Since the reactor coolant -

~

pumps are coasting down with a' loss of offsite power, the ability of the emptying steam generator to extract' heat from the reactor ' coolant" system is reduced. The minimum approach to criticality would occur later -in the transient and the core, power increase would be slower taan in the similar case with offsite power available. The peak power-for this case would remain well below the nominal full power. value and the DNB design basis would not be violated.

Questior'lb

' For the steamline rupture event 1(Section 5.3.15) , provide the results of an analysis under full power conditions assuming the main feedwater valve fails to close (with offsite power available).

Response

The main feedwater isolation system at Millstone Unit No. 2 is part of the engineered safeguards system and, as such, ~ failure'of this system to isolate main feedwater on a main steam isolation _ signal is an overly conservative assumption.

In additinn, NNECO has demonstrated .in previous analyses of a steam line rupture that initiation of the event at no-load conditions is more limiting with respect to primary side cooldown and return to power considerations than the case at~ full power conditions.-

. Question 5c For the steamline rupture event (Section 5.3.15) , state the assumption made in all analyses rel tive to auxiliary feedwater isolation to the ruptured unit.

Response

In the steamline rupture analysis, all auxiliary feedwater was con-servatively assumed to be delivered to ' the ruptured steam generator.

No isolation of auxiliary feedwater is assumed during the first ten minutes of the event.

b , . .,- .

Question G For the reactor coolant depressurization event (Section 5.3.8), describe the small-break LOCA model-used and provide the results of the determination of pertinent plant variables and peak clad temperature.

s

Response

The reactor coolant depressurization event described in Section 5.3.8 of the Millstone 2 Basic Safety Report is a Condition II event resulting from the inadvertent opening of both pressurizer relief valves. Condition II criteria specify that no fuel rod failure or RCS or secondary system overpressurization should occur during such a transient and that the plant will be capable of returning to operation following a reactor trip. The results of the analysis of the reactor coolant depressurization event are described in Section 5.3.8 and show that Condition II criteria are met. No RCS or secondary system over-pressurization occurs during this transient, and the margin to DNB is calculated to show that no fuel failure occurs. No small-break LOCA model was used in this analysis since this is not a LOCA transient. The assumptions used in the small break LOCA analysis can be found in WCAP-9528 and Addendum I to WCAP-9528.

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