ML20041F937

Forwards Responses to Accident Evaluation Branch Series 450 820212 Request for Addl Info Re Steam Generator Tube Rupture Analysis.Response to Question 450.3 Will Be Submitted by 820319

ML20041F937
Person / Time
Site:	Seabrook
Issue date:	03/12/1982
From:	Devincentis J YANKEE ATOMIC ELECTRIC CO.
To:	Miraglia F Office of Nuclear Reactor Regulation
References
SBN-238, NUDOCS 8203170566
Download: ML20041F937 (17)

Text

.

i suanom sum IPUBLIC SERVICE LV._ _ : *, Office:

Companyof NewHampel* e 1671 Worcester Road Framinoham Massachusetts 01701 (617) - 872 - 8100 N

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March 12,1982

[

-]

SBN-238 Co ff'A Q T.F. B 7.1.2 o

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United States Nuclear Regulatory Commission b

, g N' Washington, D. C. 20555 ce li "

Attention:

Mr. Frank J. Miraglia, Chief Licensing Branch #3 Division of Licensing Re f e re nc es :

(a ) Construction Permits CPPR-135 and CPPR-136, Docket No s. 50-443 and 50-444 (b) USNRC Letter, dated February 12, 1982, " Request for Additional Information," F. J. Miraglia to W. C. Tallman

Subject:

Responses to 450 Series RAIs; (Accident Evaluation Branch)

Dear Sir:

We have enclosed responses to the subject RAIs, which you forwarded in Reference (b), except 450.3 which will be provided by March 19, 1982 (response or schedule for response).

Very t ruly yours, YANKEE ATOMIC ELECTRIC COMPANY

/~

John DeVincentis

/

Project Manager S

JDV : ALL: dad f

Enclosure 8203170566 82031'2 PDR ADOCK 05000443 A

PDR

QUESTION 450.4 The applicant's steam generator tube rupture analysis does not contain enough information for us to complete our revies.

The following list of parameters and figures describes information which is ~ presently missing from the FSAR and which the staff uses in estimating the radiological consequences following an SGTR.

Therefore, provide the following:

A.

Figures 1.

Steam generator tube leak rate versus time.

2.

Steam generator pressure versus time in affected steam generator.

3.

Total steam flow per steam generator versus time.

4.

Primary system pressure versus time.

5.

Core coolant temperature versus time.

6.

Primary system mass inventory versus time.

7.

Steam generator mass inventory versus time for both the affected and unaffected steam generators.

8.

Water level in affected steam generator with respect to the top of the tube sheet versus time.

B.

Tables l

1.

Sequence of events as a function of time for the steam generator tube rupture with a loss of offsite power, including.the various

'E setpoints for system actuations.

1668Q:1 Y

RESPONSE TO QUESTION 450.4 The sequence of events for the steam generator tube rupture with a loss of offsjte power is given in Table 450.4-1. The principal setpoints are listed in Table 450.4-2.

The following is a list of figures of pertinent time dependent parameters:

Figure 450.4-1 Core Pressure Reactor Coolant System Tenperature Figure 450.4-2 Figure 450.4-3 Steam Generator Pressure (Faulted Steam Generator)

Steam Generator Temperature (Faulted Steam Figure 450.4-4 Generator)

Pressurizer Water Volume Figure 450.4-5 Steam Generator Flow (Fau'lted Steam Generator)

Figure 450.4-6 Feedwater Flow to Faulted Steam Generator Figure 450.4-7 Figure 450.4-6 Faulted Steam Generator Safety / Relief Valve i

Flow Rate l

Faulted Steam Generator Break Flow Rate l

Figure 450.4-9 Steam Generator Mass Figure 450.4-10 l

Figure 450.4-11 Faulted Steam Generator Liquid Volume l

l 1668Q:1

TABLE 450.4-1 SEQUENCE OF EVENTS Event Time (seconds)

Tube rupture occurs 0.0 Reactor trip signal 280 Rod motion 282 Steam generator safety valves open 287 Main feedwater flow terminated 302 Safety injectica signal 555 Safety injection 580 duxiliaryfeedwaterinjection 616 Faulted steam generator isolated 1800 i

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1668Q:1

TABLE 450.4-2 PRINCIPAL SETPOINTS 1.

Reactor trip Overtemperature AT 61.4170F 2.

Safety injection Low pressurizer pressure 1760 psia f

3.

Steam generator safety valve High secondary side pressure 1236 psia t

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~ SB142 FSAE RAI 4'50.5.(1 S e7.4) There is a diocrepancy as to the closure time of the purge system isolation valves. Table 6.2-83 lists a closure time of 5 seconds which correspons to the Proposed Technical specifications while the accident analysis in Section 15.7.4 uses a closure time of 3.0 seconds. Alao, the exact location of the radiation monitor's signal is not clear. With the present description of the air flow 3'attern and the location of the radiation monitors, it may not be possible fc>r the radiation monitor to detect a release from a FRA inside the reactor vessel in sufficient time to assure isolation of the containment air purge system prior to the release of activity. 9. In order for the staff to complete the review, provide the following informaeion 1. The smact location by drawing, of any radiation monitors which affect the containment air purge system. Also, supply the following information relative to these monitors: safety classification, and technical specification a. requirementsi and b. a description of instrument response following a FRA within the reactor vessel taking into account instrument location and air flow pattern. * '2. Clar.lfy thq isolation valve closure time. .RESPONSIt 1. Radiation Monitors which Affect the Containment Air Purne System There are two monitors: RM 6535A is associated with valves 1 and 43 RN 65355 is associa'ted with valves 2 and 3. Both monitors are IEEE Class 1E orea monitors located on the manipuistor crane. In the event of a fuel handling sesident, these monitors isolate the containment on-line and c ff-11ae purge isolation valves (refer to section 12.3.4, F.12.3-5, F.12.3-6). The description f the instrument response following a fuel handling accident w ithin~the reactor vessel is given in F3AR section 6.2.4. 2. Isolation valve Closure Time The valve closure time. listed in Table 6.2-83 (5 seconds) is the total design ties requi' red for the accomplishment of the safety function (valve closure plus instrument response time). The values shown in Section 15.7.4.3 (3.6 seconds total) were prelimir.ary estimates. This subsection is being revised in Amendment 45 to delete these estimated values. -}}