ML18082A441

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Forwards Response to IE Bulletin 80-04 Re Analysis of PWR Main Steam Line Break.Auxiliary Feedwater Flow Becomes Dominant Factor in Determining Duration & Magnitude of Steam Flow Transient in Later Stage
ML18082A441
Person / Time
Site: Salem  PSEG icon.png
Issue date: 04/17/1980
From: Schneider F
Public Service Enterprise Group
To: Grier B
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
IEB-80-04, IEB-80-4, NUDOCS 8005220501
Download: ML18082A441 (13)
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Frederick W. Schneider Vice President Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 201/430-7373 Production April 17, 1980 Mr. Boyce H. Grier, Director U. S. Nuclear Regulatory Commission Off ice of Inspection and Enforcement Region 1 631 Park Avenue King of Prussia,, Pennsylvania 19406

Dear Mr. Grier:

NRC IE BULLETIN NO. 80-04 NO. 1 AND 2 UNITS SALEM GENERATING STATION In response to your letter of February 8, 198"0, transmitting NRC IE Bulletin 80~04, which was received on February 11, 1980, the attached response is hereby submitted for your review.

If you require additional information on the enclosed submittal, we will be pleased to* discuss it with you.

Sincerely, Attach.

CC:

Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D. C.

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The following response corresponds to Item Nos. of the NRC Bulletin 80-04.

1.

Response to this item is already covered in our re-sponse to the Question 5.106 which was transmitted to the Director of Nuclear Reactor Regulations:

Attention Mr. Parr, on January 31, 1980.

(A copy is attached hereto~)

This response to Question 5.106 is valid for both Units 1 and 2 of the Salem Gener-ating Station.

2.

We have reviewed the assumptions made for main and auxiliary feedwater flow as they apply to Salem Units 1 and 2 licensing basis steamline break transients.

Several of the relevant assumptions used in all core transient analyses follow, and are further explained in the Salem Generating Station FSAR.

(ll The reactor is assumed initially to be at hot shutdown conditions, at the minimum allowable shutdown margin.

(2)

For the Condition IV breaks, i.e., double-ended rupture of a main steam pipe, full main eeedwater is assumed from the beginning of the transient at a very conservative cold temperature.

(3}

All auxiliary feedwater pumps are initially as-sumed to be operating, in addition to the main feedwater.

The flow is equivalent to the rated flow of all pumps at the steam generator design pressure.

(4)

Feedwater is assumed to continue at its initial flow rata until feedwater isolation is complete, approximately 10 seconds after the break occurs, while auxiliary feedwater is assumed to continue at its initial flow rate.

(5)

Main feedwater flow is completely terminated fol-lowing feedwater isolation.

Based on the manner in which the analysis is performed for Salem Units 1 and 2,the core transient results are very insensitive to auxiliary feedwater flow.

The first minute of the transient is dominated entirely by the steam flow contribution to primary-secondary heat.transfer, which is the forcing function for both the reactivity and thermal-hydraulic transients in the core.

The effect of auxiliary feedwater runout (or failure of runout pro-tection where applicable) is minimal.

Greater feedwater flows during the large steamline breaks serve~ to reduce secondary pressures, accelerating the automatic safe-guards actions, i.e. steamline isolation, feedwater iso-lation and safety injection.

The assumptions described above are therefore appropriate and conservative for the short-term aspect of the steamline break transient.

The auxiliary feedwater flow becomes a dominant factor in determining the duration and magnitude of the steam flow transient during later stages in the transient.

However, the limiting portion of the transient occurs during the first minute, both due to higher steam fTows inh~rently present early in the transient and due to the introduction of boron to the core via the safety injection system.

In conclusion, PSE&G and Westinghouse have evaluated the effect of runout auxiliary feedwater flows in the core transient for steamline break, and based on ~bis evalu-ation, have determined that the assumptions presently

~ade ~re appropriate for use as Salem licensing basis.

The concerns outlined in the introduction to IE Bulletin 80-04 relative to~ 1) limiting *core conditions occurring _

during portions of the transient where auxiliary feedwater flow is a relevant contributor_tc;> plant cooldown; and 2) incomplete isolation of main feedwater flow, are not representativ~ ot the Salem Generating Station, Units 1 and 2.

3.

Based on our response to items 1 and 2 above, poten-

. tial for the containment overpressure does not exist

and the potential for the reactor to return to p6wer

- -. does not worsen with due considerations to the -NRC Bulletin 80-04.

PSE&G has determined that no cor-rective actions are required at Salem Units 1 and 2 based on the NRC Bulletin 80-04.

MRD:aec 16/17C.

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...._oFILE COPY

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  • LIC. AND ENV.

s'fai~ COOE L-~----------~~~ry 31, 1980 Director of Nuclear Reactor Regulation U. s. ?~uclear R~gulatory Commission Washington, D. c.

20555

---At tent-ion:-**- -Mr. -Olan D. Parr, Chief Light l*later Reactors Branch 3 Division of Project.f.ianagement Gentlemen 1 RESP0~1SE TO REQU:CSTS FOR AD~ITlO!~li.L INFO?..!*jATION NO. 2 m:IT SJ..l..:S?-~ NUCLEAR GENERATING STATION DOCKET rm. 5 0-311 FILE No. ----

CODES ___ _

Public Sc!rvice Electric and Gas Company hereby transmits si;::ty (60) copies of its response to your September 21, 1979 letter requesting additionai infomation relating to how Auxiliary Fe~dv;ater Flow was accounted for in the original Contain~cnt Overpressurization Analysis (Question 5.106).

The ir.f or.:iation cont~ined herein will be incorporated into the Salen FSAR in an amendment to our application. *

  • Should }*ou have any questions, please do not hesitate to contact us.

Very truly yours, R.1!~(1

  • General Manager

. p.f; ~tctA.

Licensing and Environment Engin~erinq and Construction RWS t::a,*

Enclosure BCC:

Manager -

Salem Projects

.Manager -

Nuclear Opcra7ion nanagcr -

Salem Chief Controls Engince Chief Mcch~nical Engineer.

R. W. Skwarek

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QUESTION 5.106 In a letter dated September 10, 1979, the NRC was informed by Virginia Electric and Power Company that overpressuriza-tion of the containment at North Anna 3 and 4 could occur as a result of a main steam line break inside containment.

This overpressurization resulted when auxiliary feedwater flow was included in the analysis.

NRC is currently assess-ing the generic implications of this letter.

To assist us in determining if a similar circumstance could occur at your facility, you should take the following actions.

1)

Review your original analysis of this event, and provide NRC with the assumptions used during this analysis.

Particular emphasis should be placed on describing how auxiliary feedwater flow (AFF) was accounted for in your original analysis.

(Reference to previously submitted information is acceptable if identified as to page num-ber and date.)

Any changes in your design which would impact the conclusions of your original analysis *should be discussed.

We are particularly concerned with design changes that could lead to an underestimation of the containment pressure following a MSLB inside containment.

2)

Specifically, provide the following information for the analyses performed to determine the maximum containment oressure for a spectrum of postulated rnai:i steam line Sreaks. for various reactor power levels:

a.* Specify the auxiliary feedwater flow rate that was used in your original containment pressurization analyses.

Provide the basis 'for this assumed flow rate.

b.

Provide the auxiliary feedwater rated flow rate, the*

run out flow rate, and the pump head capacity curve of your current design.

c.

Provide schematic drawings to show the auxiliary feedwater system arrangement in your current design.

a.
e.

Provide the time span over which it was assumed in.

your original analysis that AFF was added to the affected steam generator following a MSLB inside containment.

Discuss the design provisions *in the auxiliary feed-water system used to terminate the auxiliary feed-wa ter flow to the affected steam generator *. If QS.106-1

g.
h.

ANSWER o~erator action is required to perform this func-tion, discuss the information that will be available to the operator to alert him of the need to isolate the auxiliary feedwater to the affected steam gener-ator, the time when this information would become available, and the time it would take the operator to complete this action.

If termination of auxiliary.feedwater flow is dependent on automatic action, describe the basic operation 0£ the auto-isolation system.

Describe the failure modes of the system.

Describe any annunciation devices associated with the system.

Provide the single active failure analyses which specifically identifies those safety grade systems and components relied upon to limit the mass and-energy release and the containment pressure response.

The single failure analysis should include, but not necessarily be limited to:

partial loss of contain-ment cooling systems and failure of the auxiliary feedwater isolation valve to close.

For the single active failure case which results in the maximum containment atmosphere pressure, provide a chronology of events.

Graphi~ally, show the con-tainment atmosphere pressure as a function of time for at least 30 minutes following the accident.

For this case, assume the auxiliary feedwater flow to the broken loop steam generator to be at the pump run out flow (if a run out control system is not part of the current design) for the entire transient if no automatic isolation to auxiliary feedwater is part. of the current design..

For the case identified in (g) above, provide the mass and energy release data in tabular form.

Dis-cuss and justify the assumptions *made regarding the time at which active containment heat removal _sys-tems become effective.

l)

The analysis of the MSLB inside containment is contained in --the responses to Questions 5.62, 5.63, 5.82, 5.83, 8.84, a.as covered in the November 28, 1978 submittal and Question 5.64 which is in the October 28, 1978 submittal.

The assumptions used in the analysis are discussed in these respon.ses.

Q5.106-2

---c. The Auxiliary Feedwater System is actuated shortly after the occurrence of a st~am line break.

As discussed in the response to Question 5.82~ the mass addition to the faulted steam generator from the Auxiliary Feedwater System was conservatively determined by using the follow-ing assumptions.

a.

The entire Auxiliary Feedwater System was assumed to be actuated at the time of the break and instantane-

    • ously pumping at its maximum capacity.
b.

The affected steam generator was assumed to be at atmospheric pressure.

c. **

The intact steam generators were assumed to be at the safety valve _set pressure.

d.

Flow to the affected steam generator was calculated from the Auiiliarv Feedwater Svstem h~ad curves assumptions b anc'c above and the system line resis-*

tances.

The effects of flow limiting devices were considered.

e.

The flow to the faulted steam generator from the Auxiliary Feedwater System was assumed to exist from the time of rupture until realignment of the system was completed.

f.

The failure of auxiliary feedwater runout control* was considered separately as a single failure.

The auxiliary fee~water system has not been changed in any way that would aff.ect the conclusions of the orig1nal analysis.

2)
a.

The analysis presented in responses to Questions 5.82 and 5.85 used the following auxiliary feedwater flow

- - * - rates:

l.* With runout protection operational, a constant auxiliary feed flow of 1840 gpm to the faulted steam generator.

2.

Failure of runout control was simulated by assuming a constant auxiliary feedwater flow of 2040 gpm to the faulted steam generator.

The above flow rates were held constant from time of break until realignment, which was assumed at ten minutes*.

QS.106-3

~

,. The assumptions used to maximize flow to a faulted steam generator *are provided in the response to Ql, above.

b.

R~fer to Table 4 of the Question 5.85 response and figures QS.106-1 and QS.106-2.

c.

Refer to figure QS.106-3.

d.

The auxiliary f eedwater system is actuated shortly after the occurrence of a steam line break.

In the analysis the auxiliary feedwater flow to the faulted steam generator was assumed to exist from the time of the rupture until realignment of the system was com-p"leted.

The Auxiliary Feedwater System is manually realigned by the operator after 10 minutes.

There-fore, the analysis assumes maximum auxiliary feed-water flow to a depressurized steam generator for full 10 minutes.

The actions taken to terminate auxiliary feedwater to the faulted steam generator are discussed iri response toe., below.

e.

In the event a postulated main steam line break occurs, auxiliary feedwater to the affected -steam generator*must be terminated manually.

Present design criteria allows ten minutes for the operator to recognize the postulated event and perform the necessary actions.

However, the operator is expected to terminate auxiliary feedwater flow to the affected steam generator in much less time due to the amount

  • of Class lE indication provided to monitor plant.

conditions.

The information available to alert the ooerator of the need ~o isolate auxiliary feedwater to the affected steam generator is mounted on the control console in the control room.

The pressure in each steam generator is monitored and displayed by two independent channels of* instrumentation.

Also, a bank of pen recorders indicates steam and feedwater flows for each steam generator; thi*s allows the _con-trol room operator to readily view and compare the

The suction and discharge pressures of each auxiliary

. ~ feedwater pump are indicated.on the control console *

.. The auxiliary feedwate"r flow indications for each steam generator are mounted on the control console next to each other, allowing the operator to easily view and compare flows.

QS.106-4

-s-In addition to the above-mentioned indications, high steam flow, low steam pressure, and steam-feed flow deviation conditions for each steam generator are alarmed on the main control console in the control room.

Alarms for these conditions are also provided on the overhead annunciator.

Since a sufficient number of trains of instrumenta-tl.on must be available for normal plant operation, steam generator instrumentation will be in operation at the time of the postulated event.

Therefore, changes in steam generator pressure and steam flow will be detected as they occur.

The only delay expected in transmitting the information to the con-trol room is the time reauired for the instrumenta-tion to* react to the changing conditions.

  • This delav is expected to be no more than a few seconds.
f.

Several failures can be postulated which would impair the performance of various steam break protection systems and therefore would change the net energy releases from a ruptured line.

Four different single failures were analyzed for each break condition.

  • These were:
1) failur~ of a safeguards train; 2) failure of a main feed isolation valve: 3) failure of a main steam isolation valve; and 4} failure of auxiliary feedwater runout protection eq*uipment.

These are discussed in the response to Question 5.82.

The effect of these failures is to orovide additional

.fluid which may be released to the containment via the break or reduce the heat removal capability of the containment safeguard systems.

Failure of the auxiliary feedwater isolation valve to close has not been considered.

The maximum auxiliary

, _* feedwater flow that can be delivered to a faulted

-steam generator has been assumed in the analysis for

  • ten minutes with two cases being considered:
1) run-out.protection operational; 2) failure of runout pro-
. tection.

Only after ten minutes the operator takes

At that time if the remote control-

. led auxiliary feedwater isolation valve fails to close, the operator can.trip the two auxiliary feed-water pumps feeding* the broken steam generator until this valve or another in the line is manually closed.

QS.106-5

I g.

An analysis of a spectrum of steam line breaks at various power levels, assuming several different single failures, has been provided in the response to Question 5.82.

These analyses include cases assuming failure of auxiliary feedwater runout protection.

The containment temperatures and pressures resulting from all cases considered are presented in the response to Question 5.82 along with pertinent trips and trip times.

Also graphical results showing con-tainment atmospheric temperature and containment pressure are provided.

h.

The mass and energy release data for the worst cases analyzed are provided in tabular form in the response to Question 5.82.

The time at which active containment heat removal systems become effective is discussed in the*response to Question 5.62.

M P80 12 04 1/6 QS.106-6

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