ML20065E232

PWR Main Steam Line Break W/Continued Feedwater Addition (B-69) Pse&G,Salem Nuclear Generating Station Units 1 & 2, Technical Evaluation Rept

ML20065E232
Person / Time
Site:	Salem
Issue date:	09/28/1982
From:	Vosbury F FRANKLIN INSTITUTE
To:	Peter Hearn NRC
Shared Package
ML18087A554	List: ML18087A553 ML18087A556 ML20065E232
References
CON-NRC-03-81-130, CON-NRC-3-81-130 TAC-46858, TAC-46859, TER-C5506-139, NUDOCS 8210010059
Download: ML20065E232 (26)

Text

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-. - -. a TECHNICAL EVALUATION REPORT PWR MAIN' STEAM LINE BREAK WITH CONTINUED FEEDWATER ADDITION (B-69)

PUBLIC SERVICE ELECTRIC AND GAS COMPANY SALEM NUCLEAR GENERATING STATION UNITS 1 AND 2 NRC DOCKET NO. 50-272, 50-311 FRC PROJECT C5506 NRC TAC'NO. 4685.8, 46859 FRC ASSIGNMENT 5 NRC CONTRACT NO. Ol5C-03-81-130 FRC TASK 139 Preparedby Franklin Research Center Author:

F. W. Vosbury 20th and Race Street Philadelphia, PA 19103 FRC Group Leader:

R. C. Herrick Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer:

P. Hearn i

September ~28, 1982 s

This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or impiled, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such thire.

party would not inf ringe privately owned rights.

Prepared by:

Reviewed by:

Approved by:

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TER-C5506-139 FORMORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Comunission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.

t Mr. F. W. Vosbury contributed to the technical preparati6n of this report i

through a subcontract with WESTEC Services, Inc.

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l... INTRODUCTION 1.1 PURPOSE OF REVIEN This Technical Evaluation Report (TER) documents an independent review of the Public Service Electric and Gas (PSEEG) compliance with the Nuclear Regulatory Cosuaission's (NRC) IE Bulletin 80-04, " Analysis of a Pressurized Water Reactor Main Steam Line Break with Continued Feedwater Addition" (1], as 4

I it pertains to the Salen Nuclear Generating Station, Units 1 and 2.

This evaluation was performed with the following objectives:

o to assess the conformance of PSE&G's main steam line break (MSLB) analyses with the requirements of IE Bulletin 80-04 o to assess PSE&G's proposed interim and long-range corrective action j

plans and schedules, if needed, as a result of the MSIa analyses.

1.2 GENERIC BACKGROUND In the sur-ar of 1979, a pressurized water reactor (PWR) licensee j

submitted a repo.

e NRC that identified,a deficiency' in the plant's i

' original analysis of tne containment pressurization resulting from a MSLB. A

,u reanalysis of the containment pressure response following a MSIa was performed, -

l and it was determined that, if the auxiliary feedwater (APW) system continued f

to supply feedwater' at runout conditions to the ste'am generator that had experienced the steam line break, containment design pressure would be exceeded in approximately 10 minutes. The long-term blowdown of the water supplied by f

the APW system had not been considered in the earlier analysis.

i on October 1,1979, the foregoing information was provided to all holders of operating licenses and construction permits as IE Information Notice 79-24 l

(2]. Another facility performed an accident analysis review pursuant to l

receipt of the information in the notice and discovered that, with offsite l

electrical power available, the condensate pumps would feed the affected steam

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generator at an excessive rate. This excessive feed was not previously considered in the plant's analysis of a MSLB accident. g 42 Franklin Research Center ac aor m remo.aw u.

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_.. ' E - ' ' n' 2F TER-C5506-139 A third licensee informed the NRC of an error in the MSLB analysis for

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their plant. During a review of the MSLB analysis, for zero or low power at the end of core life, the licensee identified an inco' erect postulation that the startup feedwater control valves would remain positioned "as is" during the transient. In reality, the startup feedwater control valves will camp to 804 full open due to an override signal resulting from the low steam generator pressure reactor trip signal. Reanalysis of the events showed that opening of the startup valve and associated high feedwater addition to the affected steam generator would cause a rapid reactor cooldown and resultant reactor return-I to-power response, a condition which is outside the plant design basis.

Because of these deficiencies identified in original MSLB accident analyses, the NRC issued IE Bulletin 80-04 on February 8,1980.

This bulletin required all PWRs with operating licenses and certain near-term PWR operating license applicants to perform the following:

4 "1.

Review the containment pressure response analysis to determine if the potential for containment overpressure for a main steam line break incide containment included the impact of runout flow from the aux

• eedwater system and the impact of other energy sources, such

'miation of feedwater or condensate flow. In your review, consider y....bility to detect and isolate the damaged steam generator from these sources and the ability of the pumps to remain operable af ter entended operation at runout flow.

I 2.

Review your analysis of the reactiv'ity increase which results from a main steam line break inside or outside containment. This review should consider the reactor cooldown rate and the potential for the reactor to return to power with the most reactive contsol rod in the fully withdrawn position. If your previous analysis did not consider i

all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indicated the report of this review should includes a.

The boundary conditions for the analysis, e.g., the end of lif e shutdown margin, the moderator temperature coefficient, power level and the net effect of the associated steam generator water i

inventory on the reactor system cooling, etc.,

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b.

The most restrictive single active failure in the safety l,

injection system and the effect of that failure on delaying the delivery of high concentration boric acid solution to the reactor coolant system, i

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c.

The effect of extended water supply to the affected steam j

generator on the core criticality and return to power, j

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The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the Minimum Departure from Inacleate Boiling Ratio (MDNBR) values for the j

analyzed transient.

3.

If the potential for containment overpressure exists or the reactor-return-to-power response worsens, provide a proposed corrective action and a schedule for completion of the corrective action.

If the unit is operating, provide a description of any i

interim actica that will be taken until the proposed corrective action is completed."

'i 1.3 PLANT-SPECIFIC BACKGROUND

't PSE6G responded to IE Bulletin 80-04 in a letter'to the NRC dated May 2, 1980 (3) and provided additional information in a letter to the NRC dated July 26, 1982 (4]. The information in References 3 and 4 has been evaluated along with pertinent information from the Salen Nuclear Generating Station Final Safety Anal.-8* Report (FSAR) (5) to determine the adequacy of the Licensee's compliance wi illetin 80-04.

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2.

ACCEPTANCE CRITERIA i

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The following criteria against which the Licensee's MSLB response was evaluated were provided by the NRC (6):

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PWR licensees' responses to IE Bulletin 80-04 shall include the following information related to their analysis of containment pressure and core reactivity response to a MSLB within or outside containment:

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a.

A discussion of the continuation of flow to the affected steam generator, including the impact of runout flow from the AN system and the impact of other energy sources, such as continuation of feedwater or condensate flow. A N system runout l

flow should be determined from the manufacturer's' pump curves at no backpressure, unless the system contains reliable anti-runout provisions or a more representative backpr. essure has been l

conservatively calculated. If a licensee assumes credit for anti-runout provisions, then justification and/or documentation used to determine that the provisions are reliable should be provided. Examples of devices for which provisions are reliable are anti-runout devices that use active components (e.g.,

automatically throttled valves) which meet the requirements o.f IEEE Std 279-1971 (7) and passive devices (e.g., flow orifices or cavitating venturis).

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b.

A determination of potential containment overpressure as a result of the impact of runout flow from the AN system or the impact of other energy sources such as continuation of feedwater or condensate flow. Where a revised analysis is submitted or where refererice is made to the existing FSAR analysis, the analysis

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must show that runout AN flow was included and that design containment pressure was not exceeded.

c.

A discussion of the ability to detect and isolate the damaged steam generator from continued feedwater addition during the MSLB 4

accident. Operator action to isolate AN flow to the affected i

l I steam generator within the first 30 minutes of the start of the MSLB should be justified. If operator action is to be completed within the first 10 minutes, then the justification should address the indication available to the operator and the actions I

required. Where operator action is required to prevent exceeding j

a design value, i.e.rcontainment design pressure or specified A.

acceptable fuel design limits, then the discussion should include I

the calculated time when the design value would be exceeded 'if no operator action were assumed. Where operator actions are to be performed between 10 and 30 minutes after the start of the MSIB, the justification should address the indications available to the -

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operator and, the operator actions required, noting that for the i

first 30 minutes, all actions should be performed from the control room.

d.

Where all water sources were r.ot considered in the previous analysis, an indication should be provided of the core reactivity change which results from the inclusion of additional water sources. A submittal which does not determine the magnitude of.

reactivity change from an original analysis is not responsive to the requirements of IE Bulletin 80-04.

2.

If containment overpressure or a worsening of the reactor return-to-l power with a violation of the specified acceptable fuel de, sign limits described in Section 4.2 of the Standard Review Plan (8)

(i.e.,

increase the core reactivity) can occur by the Licensee's analysis, the Licensee shall provide the following additional information:

I a.

The proposed corrective actions to prevent containment overpressure or the violation of fuel design limits and the-schedule for their completion.

l b.

The i'nteria actions that will be taken until the proposed corrective action is completed, if the unit is operating.

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The acceptable input assumptions used in the licensee's analysis of the core reactivity changes during a MSLB are giv,en in Section 15.1.5 of the Standard Review Plan (9]. The following specific assumptions should be used unless the analysis shows that a different assumption is more limiting:

Assumption II.3.b.:

Analysis should be performed to determine the most conservative assumption with respect to a l

loss of electrical ~ power. A reactivity

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i analysis should be conducted for a normal l

power situation as well as a loss of offsite power scenario, unless the licensee has previously conducted a sensitivity analysis a

which demonstrates that a particular

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assumption is more conservative.

l Assumption II.3.d.:

The most restrictive single active failure in the safety injection system which has the effect of delaying the delivery of high concentration boric acid solution to the reactor coolant system, or any other single

-ac'tive failure affecting the plant response, i -

should be considered.

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Assumption II.3.g.:

The initial core flow should be chosen such j

that the post-MSLB shutdown margin is minimized (i.e., maximum initial core flow).

The acceptable computer codes for the licensee's analysis of core reactivity changes are, by nuclear steam supply system (NSSS) vendor, the following: CESEC (Combustion Engineering), IAFTRAN (Westing-house), and TRAP (Babcock & Wilcox). Other computer codes may be used, provided that these codes have previously,been~ reviewed and found to be acceptable by the NRC staff. If a computer code is used which has not been reviewed, the licenseo must describe the method employed to verify the code results in sufficient detail to permit the code to be reviewed for acceptability.

1 4.

If the APW pumps can be damaged by extended operation at runout flow, l

the licensee's action to preclude damage should be reviewed for t

technical merit. Any active features should satisfy the requirements of IEEE Std 279-1971. Where no corrective action has been proposed, this should be indicated to the NRC for further action and resolution.

i 5.

Modifications to electrical instrumentation and controls needed to detect and initiate isolation of the affected steam generator and feedwater sources in order to prevent containment overpressure and/or unacceptable core reactivity increases must satisfy safety-grade requirements. Instrumentation that the operator relies upon to follow the acciderit and to determine isolation of the affected steam generator and feedwater sources should conform to the criteria contained in ANS/ ANSI-4.5-1980, " Criteria for Accident Monitoring Functions in Light-Water-cooled Reactors" (10], and the regulatory positions in Regulatory Guide 1.97, Rev. 2, " Instrumentation for

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Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" (11].

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6.

AFW isystem status should be reviewed to ensure that system heat l

removal caparf.ty does not decrease below the minimum required level l

as a result of isolation of the affected steam generator and also l'

that recent changes have not been made in the system which adversely affect vital assumptions of the containment pressure and core reactivity response analyses.

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The safety-grade requirements (redundancy, seismic and environmental j

qualifications, etc.) of the equipment that isolates the main feedwater (MFW) and AFW systems from the affected steam generator I

should be specified. The modifications of equipment that is relied f

upon to isolate the MFW ind APW systems from the affected steam generator should satisfy the following criteria to be considered safety-grade:

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Redundancy and power source requirements: The isolation valves.

should be designed to accommodate a single failure. A failure-modes-and-effects analysis should demons,trate that the system is capable of withstanding a single failure without loss of function. The single failure analysis should be conducted in accordance with the appropriate rulee of application of ANS-51.7/N658-1976, " Single Failure Criteria for PWR Fluid Systems" [12].

o Seismic requirements: The isolation valves should be designed to Category I as recommended in Regulatory Guide 1.26 [13].

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o Environmental qualification: The isolation valves should satisfy the requirements of NUREG-0588, Rev.1, " Interim Staff Position l

on Environmental Qualification of Safety-Related Electrical j

Equipment" [14].

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o Quality standards: The isolation valves should satisfy Group B l

quality standards as recommended in Regulatory Guide 1.26 or similar quality standards from the plant's licensing bases.

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TER-C5506-139 3.

TECHNICAL EVALUATION The scope of work included the following:

1.

Review the Licensee's response to IE Bulletin 80-04 against the acceptance criteria.

2.

a.

Evaluate the Licensee's MSLB analyses for the potential of overpressurizing the containment and with respect to the core reactivity increase due to the effect of continued feedwater flow.

b.

Evaluate the Licensee's proposed correc'tive actions and schedule for implementation if the findings of Task 2a indicate that a potential exists for overpressurizing the containment or worsening the reactor return-to-power in the event of a MSLB accident.

l 3.

Prepare a TER for each plant based on the evaluation of the l'

information presented for Tasks 1 and 2 above.

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This report constitutes a TER in satisfaction of item 3.

Sections 3.1

-through 3.3 of this report state the requirements of IE Bulletin 80-04 by subsection, summarize the Licensee's statements and conclusions regarding these requirements, and present a discussion of the Licensee's evaluation t

followed by conclusions and recommendations.

3.1 i

. REVIEW OF CONTAINMENT PRESSURE RESPONSE ANALYSIS j

i The requirement from IE Bulletin 80-04, Item 1, is as follows:

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" Review the containment pressure response analysis to determine if the I

j potential for containment overpressure for a main steam line break inside l

containment included the impact of runout flow from the a,txiliary j

feedwater system and the impact of other energy sources, such as i

continuation of feedwater or condensate flow. In your review, consider 1

your ability to detect and isolate the damaged steam generator from these l

sources and the ability of the pumps to remain operable after extended operation at runout flow."

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3.1.1 Summary of Licensee Stateme'nts and conclusions In regard to the review of the containment pressure response analysis for Salem Units 1 and 2, the Licensee stated [3]:

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" Response to this item is already covered in our response to the Question 5.106...

This response to Question 5.106 is valid for,both Units 1 and 2 l

of the Salen Generating Station."

The following is a summary of the Licensee response in Question 5.106:

"The Auxiliary Feedwater System is actuated shortly after the occurrence of a steam line break...the mass addition to the faulted steam generator

• from the Auxiliary Feedwater System was conservatively determined by using the following assumptions.

a.

The entire Auxiliary Feedwater System was assumed to be actuated at the time of the break and instantaneously pumping at its maximum capacity.

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b.

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

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

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Flow to the affected steam generator was calculated from the Auxiliary.Feedwater System head curves assump,tions b and c above and the system line resistences. 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.

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The failure of auxiliary feedwater runout control was considered f

separately as a single failure.

The auxiliary feedwater system has not been changed in any way that would affect conclusions of the original analysis.

i a.

The [MSLB) analysis...used the following auxiliary feedwater flow l

rates i

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

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

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

In the analysis the auxiliary feedwater flow to i

the faulted steam generator was assumed to exist from the time of the i

rupture until realignment of the system was completed. The Auxiliary j

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

i Therefore, the analysis assumes maximum auxiliary feedwater flow to a

,j depressurized steam generator for full 10 minutes.' The actions taken to terminate auxiliary feedwater to the faulted steam generator are discussed...below.

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

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, indication provided to monitor plant conditions.

i The information available to alert the operator of the need to isolate auxiliary feedwater to the affected steam generator is mounted on the j

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 generators this allows the control room operator to readily view and compare the steam flow of one steam generator to the others.

The suction and discharge pressures of each auxiliary feedwater pump are indicated on the control console. The auxiliary feedwater flow indicators for each steam generator are mounted on the control console next to each other, allowing the opera *.or to easily view and compare flows.

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In addition to the above mentioned indications, high steam flow, low steam pressure, and steam-feed flow deviation conditions for each steam a

generator are alarmed on the main control console in the control room.

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

l Several failures can be postulated which would impair the performance of ljl var'ious steam break protection systems and therefore would change the not energy release from a ruptured line. Four different single failures were analyzed for each. break condition. These were: 1) failure of a e

safeguards train 2) failure of a main feed isolation valve; 3) failure of a main steam isolation valver and 4) failure of auxiliary feedwater runout protection equipment...

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The effect of these failures is to provide additional fluids which may be released to the containment via the break or reduce the heat removal capability of the containment safeguards systems.

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l 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 anslysis for ten minutes with two cases being considered: 1) runout protection i

operationair 2) failure of runout protection. Only after ten minutes the operatoe' takes action to isolate auxiliary feedwater to the broken steam generator. At that time if the remote controlled auxiliary feedwater isolation valve fails to close, the operator can trip the two auxiliary feedwater pumps feeding the broken steam generator until this valve or another in the line is manually closed."

In response to a' request to provide information regarding operator response time, the Licensee stat (4):

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" Existing Westinghouse containment analyses use 10 minutes for isolation i

of a faulted steam generator and do not project containment peak pressure beyond such assumption. This ten minute isolation time criteria is the I

design basis for Westinghouse units, inclusive of Sales.

In the event that the postulated main steam line break in the containment occurs, auxiliary feedwater to the affected steam generator is manually' terminated by pushbutton operation in the control room. The control room operator, upon evaluating the alarms and indications symptomatic of this very recognizable accident, must depress the ' shut' pushbuttons for either one or two electrically operated control valves supplying the faulted steam generator. The number of valves to be closed depends upon the number of pumps feeding the affected. steam generator at that time.

The instrumentation available to alert the operator of the need to isolate' auxiliary feedwater to the affected steam generator is mountad on -

i the control console in the control room. The pressure in each steam i

generator is monitored and displayed by several independent channels of instrumentation. Also, pen recorders indicate steam and feedwater flows for each steam generatoer this allown the control room operator to readily view and compare the flows of one steam generator with the others.

I The suction and discharge pressures of each auxiliary feedwater pump are indicated on the control console. The auxiliary feedwater flow indica-tions for each steam generator are mounted on the control console next to each other, allowing the operator to easily view and compare flows.

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In addition to the above mentioned indications, high steam flow, low l

steam pressure, and steam / feed flow deviation conditions for each steam generator are alarmed on the main control console in the control room.

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

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Based on the number of class 1E indications provided to monitor plant conditions, the number of alarms provided to annunciate this accident, and the minimal operator actions required to terminate auxiliary feedwater to the faulted steam gerierator, operator action within 10 minutes is easily achievable and is justifiable as a design base.".e MJ Frankhn Research Center rc m.en.v snm u.

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i TER-C5506-139 3.1.2 Evaluation

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The Licensee's submittals (3, 4) concerning the' containment pressure response following a MSIA and applicable sections of the Salem FSAR [5] were reviewed in order to evaluate whether the following portions of the acceptance criteria were a tt o Criterion 1.a - Continuation of flow to the affected steam generator I

o Criterion 1.b - Potential for containment overpressure o Criterion 1.c - Ability to detect and isolate the damaged steam generator 1

o criterion 4 - Potential for AN pump damage l

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!i o Criterion 5 - Design of steam and feedwater iso.lation system o criterion 6 - Decay heat removal capacity I

o criterion 7 - Safety-grade requirements for Ml'W and AN isolation valves.

The Salem Units 1 and 2 are vir'tually identical Westinghouse-designed four-loop plants.

In the event of a MSLB, the following functions provide the necessary protection:

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o Safety injection system actuation from any of the following:

a.

Two-out-of-three low pressurizer pressure signals b.

High differential pressure signals between steam linos l

High steam line flow in two main steam lines (one-out-of-two per c.

line) in coincidence with either low-low reactor coolant system average temperaturo or low steam line pressure in any two lines j

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Two-out-of-three high containment pressure.

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o 'Ihe overpower reactor trips (neutron flux and differen.ial 1.

temperature) and the reactor trip occurring in conjunction with receipt of the safety injection signal.

o Redundant isolation of the main feedwater lines: normal control l

action closes the stain feedwater regulating valves. A safety '

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t TER-C5506-13 9 injection signal kill rapidly close all feedwater. control valves, trip the main feedwater pumps, and close'the feedwater inlet stop valves (safety-grade).

Trip of the fast acting steam line stop valves (designed to close in o

less than 5 seconds) ons i-i High steam flow in two mair.' steab lines in coincidence with either, a.

low-low reactor coolant system average temperature or low steam

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line pressure in any two lines i

b.

High-high containment pressure.

j Fast-acting isolation valves are provided in each steam line that will fully close within 10 seconds of a large break in the steam line. For breaks f

downstream of the isolation valves, closure of all valves would completely i

terminate the blowdown. Pbr any break, in any location, no more than one steam generator would blow down even if one of the isolation valves fails to close.

s The two motor riven AFW pumps automatically start on:

i o Ioss of offsite power o Ioss of MI'W flow k

o Iow-low level in one steam generator i

o Safety injection system actuation.

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h The turbine-driven AEW pump will start on o Ioss of offsite power

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o Iow-low level in two steam generators o 4-kV bus undervoltage.

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1 Finally, the AEW system is equipped with a runout protection system' which f

regulates the discharge flow *of the motor-driven pumps and controls the i

turbine-driven pump governor to 1_imit flow so that the pumps will not ' sustain damage from operation during a MSIS.

All of the equipment required to mitigate che MSI2 accident is safety-grade and complies with the requirements IEEE Std 279-1971.

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The environmental qualification of safety,-related electrical and mechanical components is being reviewed separately by the NRC and,is not

^l within the scope of this review.

The review did not determine if the instrumentation upon which the operator relies to follow the accident and isolate the affected steam generator conforms to the criteria in ANS/ ANSI-4.5-1980 (10] and Regulatory Guide 1.97 (11].

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i A spectrum of blowdowns covering four power levels and three different l

break sizes were evaluated by the Licensee. The three break sizes considered f

, at each. power level (0, 30, '70, and 102% of nominal) were: a full double-ended

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g rupture upstream of the steam line flow restrictor, a full dou'ble-ended d

/ rupture downstream of the steam line flow restrictor, and the largest split rupture that will not result in generation of a steam line isolation signal from the primary plant protection equipment. In the analysis of the third (split) break, reactor trip, feed line isolation, and steam line isolation are generated by high containment pressure signals. Additionally, all blowdowns used in the analyses were assumed to concist of dry steam.

For each break condition, four different single failures were evaluated.

These were (1) failure of a containment safeguards train, (2) failure of a

.j main feed isolation valve, (3) failure of a main steam isolation valve,'and l

failure of the ADi runout protection equipment.

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(4)

For both the limiting large break and small break analyses, failure of i

one of the containment spray trains proved to be the most limiting single

,j failure. Both analyses used an AMi flow rate, assuming runout protection f

operational, of 1840 gpm to the affected steam' generator until the flow was manually isolated 10 minutes into the accident. In the limiting large break

'2 case, a peak pressure of 39.1 psig occurred at.657 seconds after a 1.4-f t break at 70% power.

In the limiting small break case, a peak pressure of 42.8 psig occurred at 810 seconds after a 0.86-f t split at 1024 power. The peak

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pressure for both cases remained below the containment design pressure of-47 psig.

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TER-C5506-13 9 Sufficient indications and alarms are available to the operator to i

determine that a MSLB has occurreds once this determination'has been made, the operator has to perform minimal actions to isolate AFW flow to the affected steam generators. It is conservative to assume that the operator will complete the required actions within the 10-minute time frame. Therefore, it, is concluded that there is no potential for containment overpressurization.

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3.1.3 Conclusion i

The Licensee's responses [3, 4] and the Salem FSAR [5] adequately address j

the concerns of Item 1 of IE Bulletin 80-04.

The containment pressure L

response analysis and the the design of the mitigating systems meet the NRC's acceptance criteria. Regarding Item 1, it is concluded that there is no potential for containment overpressurization resulting from a MSLB with continued feedwater addition. The AFW pumps are protected from the effects of runout flow and therefore can be expected to carry out their intended function during the MSLB event.

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, 3.2 REVIEW OF REACTIVITY INCREASE ANALYSIS The requirement from IE Bulletin 80-04, Item 2, is as follows:

" Review your analyt ts of the reactivity increase which results from a main steam line break inside or outside containment. This review should consider the reactor cooldown rate and the potential for the reactor to l'

return to power with the most reactive control rod in the fully withdrawn position. If your previous analysis did not consider all potential water sources (such as those listed in 1 above) and if the reactivity increase is greater than previous analysis indiqated the report of this rev.iew should include:

,i The boundary conditions for the analysis, e.g.,

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a.

'i shutdown margin, the moderator temperature coefficient, power le' vel i

and the net effect of the associated steam generator water inventory on the reactor system cooling, etc.,

b.

The most restrictive single active failure in the safety injection Ij system and the effect of that failure on delaying the delivery of i

I high concentration boric acid volution to the reactor coolant system, The effect of extended water supp"ly to the affected stcan generator c.

on the core criticality and return to power, ndn Research Center m n.n.~,m.

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The hot channel factors corresponding to the most reactive rod in the fully withdrawn position at the end of life, and the Minimum Departure from Nucleate Boiling Ratio (MDNBR) values for the analyzed transient."

3.2.1 Summary of Licensee Statements and conclusions In regard to the reactivity increase resulting from a MSLB with continued feedwater addition, the Licensee stated (3):

"We have reviewed the assumptions made for main and auxiliary feedwater flow as they apply to Salem Units 1 and 2 licensing basis steam line break transients. Several of the relevant assumptions used in all core transient analyses follow, and are further explained in the Sales Generating Station FSAR.

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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 feedwater is assumed from the beginning of the transient at a very conservative cold temperature.

3.

All auxiliary feedwater pumps are initially assumed 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.

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4.

Feedwater is assumed to continue at its initial flow rate until l

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 following feedwater isolation.

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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 l.

by the steam flow contribution to primary-secondary heat transfer, which is the forcing function for both the reactivity and thermal-hydraulic tiansients in the core. The effect of auxiliary feedwater runout (or failure of runout protection where' applicable) is minimal. Greater feedwater flows during the large steamline breaks serve to reduce secondary pressures, accelerating the automatic safeguards actions, i.e.,

steamline isolation, feedwater isolation and safety injection.' The,

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

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e TER-C5506-139 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 lLaiting portion of, the transient occurs during the first minute, both due to higher steam flows inherently l

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 this evaluation, have datermined that the assumptions presently made are 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 to plant cooldown, and

2) incomplete isolation of main feedwater flow, are not representative of the Salem Generating Station, Units 1 and 2."

1 3.2.2 Evaluation The Licensee's analysis of the core reactivity increase resulting from a MSLB with continued feedwater addition was reviewed in order to evaluate whether the following acceptance criteria were met:

o Criterion 1.c - Ability to detect and isolate the damaged steam l

generator o Criterion 1.d - Changes in core reactivity increase f

o Criterion 3 - Analysis assumptions.

The Elcensee's response and the FSAR analysis of the reactikity increase l

resulting from a MSLB were reviewed. From that review, it was determined that the analysis is conservative in its assumptions and that the assumptions are in 'accordance with those in Acceptance criterion 3, with the exception of considering runout AFW flow.

I The worst-case analysis assumed complete severance of a pipe inside.the containment at the outlet of the steam generator with the plant at no load conditions and offsite power available. This analysis determined that, although a return-to-power is preTicted, there is no violation of the.specified acceptable fuel design limits.

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The Licensee's conclusion that the cote transient for the MSLB is insensitive to runout AFW flow is valid for the following reasons:

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TER-C5506-139 the primary to secondary heat transfer rate Early in the transient, is several l

(from the blowdown of the iritial steam generator mass) o orders of magnitude greater than that contributed by the additional f

f AN flow due to runout.

Later in the transient (when the majority of the initial mass has blown down), AN flow becomes a dominant factor in determining the o

magnitude and duration of the transient,

'Ihe limiting core conditions will occur within the first minute due to the initial high cooldown rate contributing to the reactivity o

addition which is terminated by the introduction of boron into the core region.

Since the limiting core conditions occur before the AN flow becomes a major contributing factor, it can be concluded that the core transient is insensitive to the contribution of AN flow, and therefore the assumptions of I

the FSAR analysis remain valid.

I 3.2.3 Conclusion i

'l The Licensee's response and FSAR adequately address the concerns of Item All potential sources o' water were identified and, f

2 of IE Bulletin 80-04.

although a return-to-power is predicted, there is no violation of the Therefore, the FSAR analysis of the specified acceptable fuel design limits.

reactivity increase resulting from a MSLB remains valid.

i

.3. 3 REVIEW OF CDRRECTIVE ACTIONS The requirement from IE Bulletin 80-04, Item 3, is as follows:

l "If the potential for containment overpressure exists or the reactor-l-

return-to-power response worsens, provide a proposed corrective action If the unit is and a schedule for completion of the corrective action.

operating, provide a description of any interim action that will be taken until the proposed corrective action is completed."

Summary of Licensee Statements and Conclusions 3.3.1

.I The Licensee stated:

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" Based on our response to items 1 and 2 above, potential for the j

i containment overpressure does not exist and the potential for the reactor l

l to return to power does not worsen with.due considerations to the NRC 1 1 i

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4 PSEEG.has determined that no corrective actions are Bulletin 80-04.

required at Salem Units 1 and 2 based on the NRC Bulletin 80-04."

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3.3.2 Evaluation and conclusion The Licensee's analyses determined that neither a containment over-pressurization not a reactor return-to-power with a violation of the spec,ified Therefore, it is acceptable fuel design limits would result from a MSLB.

concluded that no further action regarding IE Bulletin 80-04 is required of PSE&G for Salem Nuclear Generating Station Units 1 and 2.

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CONCLUSIONS i

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Conclusions regarding Public Service Electric and Gas Company's response

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to IE Bulletin 80-04 with respect to Salem Itaclear Generating Station Units 1 and 2 are as follows:

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h ere is no potential for containment overpressurization resulting from a main steam line break (MSIB) with continued feedwater addition.

o The auxiliary feedwater pumps are protected from the effects of runout flow and therefore can be expected to carry out their intended function during the MSLB event.

i o

All potential water sources were identified and, although a reactor return-to-power is predicted, there is no violation df the specified I

acceptable fuel design limits. Therefore, the Final Safety Analyis I

Report reactivity increase analysis remains vglid.

o No further action is required by the Licensee regarding IE Bulletin 8'0-04.

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REFERENCES 1

1.

" Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition" NRC Office of Inspection and Enforcement, February 8,1980 IE Bulletin 80-04 2.

"Overpressurization of the Containment of a PWR Plant af ter a Main Line Steam Break" NRC Office of Incpection and Enforc3 ment, October 1,1979 IE Information Notice 79-24 3.

PSE&G 5

Letter to B. Grier (NRC, Region I)

Subject:

Reponse to IE Bulletin 80-04 i

April 17, 1980 t'

4.

E. A. Lander (PSE&G)

Letter to S. A. Varga (NRR)

Sybjects Additional Information Related to NRC Bulletin 80-04 July 26, 1982 5.

Salem Nuclear Generating Station Units 1 and 2 Final Safety Analysis Report, through Amendment 37 Public Service Gas and Electric Company 6.

"PWR Main Steam Line Break with Continued Feedwater Addition - Review of J.cceptance Criteria" Franklin Research Center, Nc,vember 17, 1981 TER-C5506-119 7.

" Criteria 'for Protection Systems for Nuclear Power Generating Stations" Institute of Electrical and Electronics Engineers, New York, NY,1971 IEEE Std 279-1971 8.

Standard Review Plan, Section 4.2

" Fuel System Design" NRC, July 1981 NUREG-0800 9.

Standard Review Plan, Section 15.1.5

" Steam System Piping Failures Inside and Outside of Containment (PWR) "

NRC, July 1981 NUREG-0800 l t

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" Criteria for Accident Monitoring Functions in Light-Water-Cooled Il1 10.

j Reactors" American Nuclear Society, Hinsdale, IL, December 1980 t

ANS/ ANSI-4.5-1980 l

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e TER-C5506-139 11.

Regulatory Guide 1.97 l

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Ebliowing an Accident" Rev. 2 NBC, December 1980 12.

" Single Failure Criteria for PWR Fluid Systems" American Nuclear Scciety, Hinsdale, IL, June 1976 ANS-51.7/N658-1976 13.

Regulatory Guide 1.26

" Quality Group Classifications and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants" Rev. 3 NRC, February 1976 14.

" Interim Staff Position on Environmental Qualificatio'n of Safety-Related Electrical Equipment" Rev. 1.

NRC, July 1981 N,UREG-0588 j.

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