ML20024E650

Submits Addl Info Re Steam Generator Tube Rupture Event Analyzed in Support of Cycle 6 Operation,Per 830725 Request. Current Steam Generator Tube Rupture Analysis Adequate to Support Cycle 6 Operation

ML20024E650
Person / Time
Site:	Millstone
Issue date:	08/08/1983
From:	Counsil W NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES
To:	Clark R Office of Nuclear Reactor Regulation
References
A03413, A3413, TAC-49798, NUDOCS 8308160300
Download: ML20024E650 (12)

Text

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August 8,1983 Docket No. 50-336 A03413 Director of Nuclear Reactor Regulation Attn: Mr. Robert A. Clark, Chief Operating Reactors Branch #3 U. S. Nuclear Regulatory Commission Washington, D. C. 20555

References:

(1) R. A. Clark letter to W. G. Counsil, dated July 25,1983.

(2) W. G. Counsil ater to R. A. Clark, dated July 15,1983.

(3) W. G. Counsil letter to R. A. Clark, dated April 13,1983.

(4) R. W. Reid letter to W. G. Counsil, dated May 12,1979.

Millstone Nuclear Power Station, Unit No. 2 Request for Additional Information on Cycle 6 Reload Steam Generator Tube Rupture Analysis In Reference (1), the NRC Staff requested Northeast Nuclear Energy Company (NNECO) to provide additional information concerning the Steam Generator Tube Rupture Event as analyzed in support of Cycle 6 operation. The information reauested is listed below.

Question

1. Provide an analysis of the radiological consequences following a design basis steam generator tube rupture which is consistent with current SRP Section 15.6.3. The analysis should include the effects of having 15% of the steam generator tubes plugged and should also incorporate the loss-of-offsite power coincident with the reactor scram. Also provide the following additional information until such time as releases from the affected steam generator are terminated:

a. A description of the sequence of events which includes an identification of all operator actions and when these actions are expected to occur. Also include descriptions of the automatic initiations and actuations as they occur chronologically,

b. The following parameters as a function of time:

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1) the primary system pressure; 8308160300 830808 PDR ADOCK 05000336 l k P PDR

_ . _ . _ ~ _ . . _ _ ._ _

2) the tube rupture flow rate and integrated tube rupture flow;

3) the secondary liquid water mass and level in both steam generators;

4) the primary system liquid mass;

5) the secondary system pressure in both steam generators;

6) the integrated mass released out of the atmospheric relief valves or safety valves for both the affected and unaffected steam generators;

7) the primary liquid and secondary liquid temperatures;

8) the pressurizer level;

9) the extent of upper head voiding, if predicted; and

10) the steaming rate for all steam generators through the atmospheric dump valves, if used.

Response

Attachment 2 of Reference (3) presented the reanalysis of the Steam Generator Tube Rupture (SGTR) Event including radiological results in support of Cycle'6.

The radiological consequences were analyzed conservatively in accordance with Standard Review Plan (SRP) Section 15.6.3 with additional guidance from NUREG-0409 and NUREG/CR-2683. The SGTR analysis included the effects of 15.3 percent of the steam generator tubes plugged and was performed with offsite power available.

The NRC Staff has documented their acceptance of the SGTR event with offsite-power available as the Millstone Unit No. 2 licensing' basis in Reference (4).

Therefore, the request for reanalysis for this event without offsite power available constitutes an attempt to modify the Plant licensing basis with no documented basis other than the fact that the Staff. position is consistent with a provision contained in the current SRP.

NNECO would like to clarify that the SRP is not a compendlun; of licensing requirements but has been prepared as guidance for NRC Staff reviewers within the office of NRR. . As such, there are many provisions in the current SRP with which Millstone Unit No. 2 does not explicitly comply. Utilizing the SRP to impose backfit requirements is inconsistent with its intended purpose, it is our view that the imposition of new analytical requirements upon licensees should be shown to contribute effectively and significantly to the health and safety of the public such that both Staff and licensee resources are expended in an optimal

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fashion. NNECO has not been provided with information supporting the need for

,any add!tional analytical effect beyond the SGTR analysis already provided to the Staff in Reference (3). There is no basis known to NNECO that uniquely treats this particular issue.

Therefore, NNECO considers the current Steam Generator Tube Rupture Analysis adequate for the purpose of supporting Cycle 6 operation. This is consistent with previously docketed analyses supporting operation of Millstone Unit No. 2 through five fuel cycles. We consider this position to be in concert with the direction to the Staff from the Commission provided by COMSECY 83-3, Backfitting Guidance.

Northeast Nuclear Energy Company is providing the following additional information in response to items 1.a and 1.b.

With respect to item la, a sequence of events for the reference analysis

• was contained in the original submittal of the Steam Generator Tube Rupture event in support of Cycle 6 operation with up to a 15.3 percent Steam Generator tube plugging level (Reference (3)). Table 1 of Attachment 2 of Reference (3) indicates the major automatic plant actions which occur up to 1800 seconds, at which time operator actions are assumed to begin.

At 1800 seconds, the affected Steam Generator (SG) is isolated and its atmospheric steam dump setpoint is raised to 975 psia. These actions are directed in the Millstone Unit No. 2 operating procedures when the hot leg temperature (TH) has been reduced to 5300F (approximately.100F below the saturation temperature for 975 psia). Figure 7 of this submittal shows that TH is approximately 5300F at 1800 seconds. Thus, releases from the affected SG are terminated at this time.

NNECO is providing the following information on micro-fiche and in .the attached figures in order to address item 1.b. One copy of this micro-fiche is .

being provided to the Millstone Unit No. 2 Project Manager for the Staff's review. The parameters which were requested for the reference analysis are identified as follows:

PRES 22 - Pressurizer Pressure (psia)

MIXL 22 - Pressurizer Level (ft) - NOTE: Pressurizer Height = 33.5 ft.

COUT Total Break Flow (Ib/sec) - Through U-Tube and Tube Sheet COUT Integrated Break Flow (Ib)

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l MIXL 23,24 -SG #1, #2 Mixture Level (ft)

LIQM 23,24 -SG #1, #2 Liquid Mass (Ib)-

PRES 23,24 -SG #1, #2 Pressure (psla)

WP*

• 48, 50 - SG #1, #2 Atmospheric Dump Flow (Ib/sec):

COUT Total Integrated Atmospheric Dump Flow (Ib)

WP** 30,34 - SG #1, #2 Safety Valve Flow (Ib/sec)

COUT Total Integrated Safety Valve Flow (Ib)

TEMP 2,5 - Cold Leg, Hot Leg Temperature (oF)--

COUT RCS Average Temperature (oF)

SATT 23 - SG #1 Secondary Liquid Temperature (oF)

Case 3 of Reference (3) is referred to as the reference analysis in this ^

report.' Case 3 represents NNECO's position as the intended licensing basis analysis for Cycle 6 operation.

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Figures 1 through 8 show the parameters plotted versus time which were requested in items 1.b.1 through 1.b.8 and 1.b.10. Concerning item 1.b.4, the initial reactor coolant system (RCS) liquid inventory in the reference analysis was 459,000 lb. The net outflow rate (breakflow - charging flow - HPSI flow)is shown in Figure 4. The results shown in this figure may be integrated to obtain the RCS liquid inventory versus time. Regarding item 1.b.9, details on the subject of upper head voiding were addressed in Reference (2).

We trust y'ou will find this information responsive to your Reference (1) request.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY 11L W. G. Counsil ' V Senior Vice President

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