Text

Forwards Draft Evaluation of SEP Topics VI-2.D, Mass & Energy Release for Possible Pipe Break Inside Containment, & VI-3, Containment Pressure & Heat Removal Capability. Comments Requested within 30 Days
	ML17194A384
Person / Time
Site:	Dresden
Issue date:	12/28/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	Del George L; COMMONWEALTH EDISON CO.
References
	TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR LSO5-81-12-081, LSO5-81-12-81, NUDOCS 8201040103
	Download: ML17194A384 (40)
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Docket No. 50-237 LSOS12-081 Mr. L. _Del George December 28, 1981

-- Director of _Nuclear Lfcensfng Commonwealth Edison Company Post Offfce Box 767 Chicago, Illinois 60690 Dear Mr.

SUBJECT:

Del George:

SYST~MATIC EVALUATION PROGRAM (SEP) FOR THEGr:t~~e.!!:.,2:

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NUCLEAR POWERfSTATION;-EVALUATION REPORT ON TOPICs-vr:-a:-1r*~-'-----*~~----~

AND VI-3

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Enclosed is a copy of our draft evaluation of *SEP Topics VI-2.D, "Mass and Energy Release for possible Pipe Break Inside Containment," and

.v:1~'3, "Containment Pressure and Heat Removal Capab11 ity." This evaluation compares your facility, as described in Docket No. 50-237, with the criteria currently used by the regulatory staff for licensing new.faci-lities. Appendix,A to 9ur draft evaluation is a draft Technical Evaluation Report from our. c;ontractor, Lawrence Livermore National Laboratory.

Please inform us if your as-built fac1lfty differs from the licensing*

basis assumed in our ~~sessment. Comments are requested within 30 days of the rec~1p.:t of this letter so that they may be considered 1n our final eval-uation.:

This evaluation will be a basic input to the integrated safety assessment for you*r facility unless you identify changes needed to.reflect the as-buflt conditions at your facility. This assessment may be revised in the fu~ure.~f your facfl ity design f s changed or ff NRC c1t]teria relating to this subjeet are modf;fied before the. fntegrated assessment fs completed.

sincerely,l?

Dennis M. Crutchfield, Chief 8201040103 81 i228!,

PDR ADOCK 05000237 P

PD~

Operating Reactors Branch N.o. 5 /} L

Division of Licensing

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Enclosure:

Draft SEP Topics VI-2.D and VI-3 cc w/enclosure:

See next page f 1{r

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NRC FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-33!>-960

UNITED STATES NUCLEAR.REGULATORY COMMISSION WASHINGTON, D. C. 20555-

December 28*, 1981 Dotket No. 50-237 LS05-81-l2-081 Mr. L. Del George Director of Nuclear Lfcens1ng Commonwea 1th Edi son Company Post Offtce Box 767 Chicago, Illinots 60690

Dear Mr. Del George:

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM (SEP) FOR THE DRESDEN 2 NUCLEAR POWER STATION - EVALUATION REPORT ON TOPICS VI-2.D AND VI-3 Enclosed is a copy of our draft evaluation of SEP Topics VI-2.D, "Mass and Energy Release for Possible Pipe Break Inside Containment_,.. and VI-3, "Containment Pressure and Heat Removal Capability." This _evaluation compares your facility, as described in Docket No. 50-237, with the criteria currently used by the regulatory staff for licensing new facil-ities. Appendix A to our draft evaluation is a draft Technical Evaluation Report from our contractor, Lawrence Livermore National Lab9ratory.

Pl ease -inform us if your as-built fac11 ity differs from the 1 icensing basis assumed in our assessment.

Comments are requested within 30 days of the receipt of thfs letter so that they may be considered in our final eval-uation.

This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This assessment may be revised in* the future if your facility design is changed or if NRC criterfa relating to this subject are modified before the integrated assessment is completed.

Enclosure:

Sincerely,

~Dennis M. Crutchfield, Chief D.

Operating Reactors Branch No. 5 Divisio~ of Licensing Draft SEP Topics VI-2.D and VI-3

cc w/enclosure:

See next page

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SAFETY EVALUATION REPORT ON CONTAINMENT PRESSURE AND HEAT REMO.VAL CAPABILITY SEP TOPIC VI-3 AND MASS AND ENERGY RELEASE' FOR POSSIBLE PIPE BREAK

INSIDE* CONTAINMENT

SEP TOPIC VI-2.D FOR THE

e.

DRESDEN 2 NUCLEAR POWER PLANT DOCKET NO.

50-237~;

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- t TABLE OF CONTENTS I ntr.oduCt ion*.. ****..*...****.*.**..*.**.**, *.**.*..* ** *****.**...*. *** ***.** *** *.*.*

3 II. Review Criteri*a..........

,................... 3 III. Related Safety Topics ***********************.* ** ********* t! ********,

4 IV.

Review Guidelines............. *....... *.................. **........... 4 v-.

Evaluation....*.*...............*.....*....... ~...*.*.......*....*.. 5

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VI.

Conclusions......................... *.. ~.......................... 6.

Appendix A:

Containment Analys.is and Evaluation for the Dresden 2 Nuclear Power Plant..............................

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I.

INTRODUCTION The Dresden 2 Nuclear Power Plant began commercial operation in 1970 Since then the. staff's review criteria have changed.

As part of the Systematic Evaluation Program (SEP),. the containment pressure and heat removal capability (Topic VI-3) and the*mass* and energy rel ease for possible pi'pe break inside containment (Topic VI-2.D) have been re-evaluated.

The purpose of this evaluation is *to* document the deviations from current safety criteria as they relate to the containment pressure and heat removal capability and the mass/energy for possible pipe break inside containment.

Furthermore, independent analyses in accordance with current criteria were performed to determine the adequacy of the containment design basis (e.g.,

design pressure and temperature). Thesignificance of the identified devi-ations, and recommended corrective measures to improve safety, wi 11 be the subject of a subsequent, integrated assessment* of. the Dresden 2 plant.

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u: REVIEW CRITERIA * * ' * -*-

The review criteria used in the current evaluation of SEP (Topics VI-2.D and VI-3) for the Dresden 2 plant are contained in *the following documents:

(1) 10 CFR Part 50, Appendix A, Seneral Design Criteria (GDC) for Nuclear Power Plants:

(a)

GDC 16 - Containment design; (b)

GD.C 38 - Containment heat removal; and (c)

GDC 50 - Containment design basis.

(2) 10 CFR Section 50.46, "Acceptance Criteria for Emergency C~re Cooling System for Light Water Nuclear Power Reactors."

10 CFR Part 50~ Appendfx K,. "ECCS Ev~1uation Models".

NUREG. 75/087,. Standard Review Plan for the Review of Safety Analysis r Reports. for Nuclear Power Plants (SRP 6.2.1, Containment Functional Design).

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.e III. RELATED SAFETY TOPICS The review areas identified below are not addressed in this report, but.

are related to the SEP topics of.mass and energy reiease for possible pip~

. break inside containment, and/or containment. pressure* and* heat removal.

capability.

(l).. III.-1, Classification of Structures, Components and Systems (Seismic.*

and Quality)

*(2) III-12, Environmental Qualification of Safety Related Equipment' (3)

VI;..7.8, ESF Switchove~ from I~jection to Recircuhtion Mode (Automatic ECCS Realignment)

(4)

IX-3, Station Service and Cooling Water Systems (S)

X, Auxiliary Feedwater System.

(6)

USI-A24, Qualification of Class.IE Safety Related Equipment IV.

REVIEW GUIDELINES General Design Criterion (GDC) 16 of Appendix A to 10 CFR Part 50 requires that a reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment.

In addition, GDC 16 requires that the containment design conditions important to safety are not exceeded for as long as the postulated accident conditions require.

GDC 38 requires.,.a containment heat removal system be pr~vided whose system safety function shall be to reduce the containment pressure and temperature following any loss-of-coolant accident (LOCA) and maintain them at an acceptably low level; furthermore, the system safety function shall be achievable assuming a single failure.

GDC 50 requires that the containment st~ucture and the containment heat *-removal system shall be designed so that the structure can accommodate, with sufficient margin, the calculated pressure and temperature con~~ions resulting from any LOCA.

This margin as obtained from the conservative

calculation of mass/energy release andthe containment model is discussed in the Standard Review Plan (SRP) Section 6.2.1~ Containment Functional Design.

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.e. The containment design basis includes the effects of stored and generated energy in the accident. Calculations of_the energy available for release should be done in accordance with the requirements of 10 CFR Part 50, Section

?0.46 and Appendix K, paragr~ph *I.A, and the cons~rvatism as specified fo SRP 6.2.l.3. The mass and energy release to the containment from a LOCA should be

- .. considered in terms of the mass and energy release during blowdown.

Break..

locations should include recirculation line breaks and steam li.ne breaks.. The review also* includes the ar:ialysis* of a postulated single.active failure.

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By reviewing the licensee's analysis*, deviations from the current.

criteria are identified and independent analyses are.performed, as* required, to evaluate the significance of these deviations*.

In* our analyses, "the best estimate" method _is used; i.e., by using actual plant design data, a best estimate,. but still. reasonably conservative containment analysis, is*.

. obtained.

Th~ evaluation is completed by' comparing the results with the containment design basis.

V.

- EVALUATION In the case of BWRs it is necessary to evaluate the effect of pipe breaks below the core for maximum containment pressure and pipe breaks above the core for maximum containment temperature.

Based on our review of the existing docket of Dresden 2, the only break locations analyzed were breaks below the core.

In the Dresden 2 FSAR a spectrum.of recirculation line breaks were analyzed for calculating the peak post-accident pressure.

All of the resultant peak calculated pressures were-determined to be below the containment design pressure of 62 psig. The maximum calculated peak drywell

pressure was determined to be 48 psig resulting from a OBA LOCA.

The calculated peak drywell temperature was 295°F.

The resultant peak wetwell pressure was 27 psig. The resultant wetwell temperature was.~oot reporter\\.

In

. addition to reviewing the applicant's analysis, a confirmatory analysis was performed which is presentec;I in Appendix A of this report *. Mass and energy release rates utilized in the analysis were calculated using RELAP-4 MOD 7 in accordance with current criteria. *ca lcul ati on of the post-accident containment pressl!re and temperature was done using CONTEMPT-LT/028.

In this case* the analysis which was run was a OBA LOCA.

The calculated transient

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reflects a post-accident peak containment pres~ure of 47 psig and a peak temperature of 295°F.

The resultant peak *wetwel 1 pressure and temperature was 17 psig and 169°F.

Both drywe11 and wetwell design pressures for Dresden 2 are 62 psig.

Both the utility and our analysis *showed the peak.

pressure is wel 1 below the containment design value.

In addition to recircul_ation 1 ine breaks, the current criteria state that steam 1 ine breaks. above the core Must be considered~ The licensee had*..

not performed a steam. line break analy~is:.; **fher~fore*, an: or_igjnal ~team :i;ne break. anal_ysis was-per~fo~ed for 'the *;t~_ff 0

.. by-LLNl~:~ TJi'i-s

1tgive.n iil. Appen-dix A.

This analysis was for two individual steam line breaks whiCh con-sisted of break sizes of 0.01 and 1.0 sq. ft.

The containment response to a steam line break was calculated assuming that the only heat sink available was the pressure suppression pool.

ln addi~ion, the presence of heat slabs was not accounted for.

The results of these analyses showed that of the two breaks examined a l.0 sq. ft. break resulted in the most severe temperature conditions in the containment.. The peak drywell temperature was 335°F. The peak. dr_yWell pressure was 18 psig_.

Therefore, for a steam line break the calculated peak pressu.re is less than design.

The temperature profile is more severe than that.resulting from a LOCA.

Therefore, the temperature profile resulting from a 1.0 sq. ft. MSLB should be used to support equipment qualification efforts.

VI.

CONCLUSIONS We have reviewed the OBA LOCA ana*lyses submitted by the licensee and have found. it to be within the design limits of the Dresden 2 Nuclear Po.wer Plant.

in addition to the docket review, a confirinatory analysis was performed and is reported irr Appendix A.

The confirmatory analysis was found to be com-parabl ~to the licensee~s results and, therefore, w~ conclude that the.licensee has satisfactorily demonstrated the adequacy of the*containment func~ional de-sign.

The 1 icensee has not subnitted a steam line break analysis, therefore, an independent analysis was performed and is reported in Appendix A.

lhe con-tainm~nt. atmosphere co1iditions during a 1.0 sq. ft. steam line break may be

-:utt'l'.tze:d* as input to the equipment qualification of safety related. equipnent*-

effort-:-:

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APPENDIX A ~

SEP Containment Analysis and Evaluation for the Dresden 2 Nuclear Power Plant TABLE OF CONTENTS Page

1.0 INTRODUCTION AND BACKGROUND

- ...................................... 4 2.0 CONTAINMENT FUNCTIONAL DESIGN ******************* ~................

4*

2.1 Review of Dresden 2 Containment Design ****.** ******* ** ** *.*

6 2.2 Pipe Breaks Inside the Rea~tor'Pressure Coolant Boundary ** ~.

7 2.3 Reanalysis of Dresden 2 Containment Design

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7 3.0.RECIRCULATION LINE BREAK ******************** ~ ********* ~: ** ~......

1 3.1 Contairiment Response to a Recirculation Line Break *********

9 3~2 Containment Response Results *******************.*****.*****

10

4. 0 MAIN STEAM LINE PIPE BREAK * * * * * * * * * * * * * * *.* * * * * * * *. * * * *. * * * * *. * * *

10 4~1 Containment Response to a Main Steam Line Break *****.******

11 4.2 Containment Response Results..****.****.********...*..*....

12

5.0 CONCLUSION

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6. 0 REFERENCES... *.......... **.. *.*.. *....................................

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LIST OF TABLES l

.Recirculation LineBlowdown Mass and Energy Release Rate Data * * * * *.* * * * * * * * * * * * * * *.* * * * * * * * * * * * *. * * * * * * * * * * * * * * * * * * *.

14 2

Cont*ainment Model* Input Data......................... *'........ 15 3

. Main. Steam Line Break Mass and Energy Release Rate Data

(.01 ft2 break) ********************************************

1.6 4

Main Steam Line Break Mass and Energy Release Rate Data (1.0 ft2 break) ********************************************

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1.0 INTRODUCTION

ANO BACKGROUND As part of the Systematic Evaluation Program (SEP), the containment functional design capability of the Dresden 2 Nuclear Power Plant has beeti.

re-evaluated *. The purpose of thjs report is to document the resolution of SEP Safety Topic VI-2.D.

Mass and Energy Release for Possible Pipe sr*eak Inside*

Containment, and Safety Topic VI-3:, Containment Pressure and Heat Removal

. Capability, and deviations _from current safety criteria as they relate to the containment functional des.ign. The significance of th~ identified deviations

.and recommended corrective measures will be the subject of a subsequent integrated assessment of the Dresden 2 plant.

The containment structure encloses the reactor and is the final barrier against the release of radioactive fission products in the event of an*

accident. The containment structure must~ therefore, be capable of withstanding, without loss*of function, the pressure and temperature conditions resulting from postulated LOCA and steam line break accidents.

Furthermore, equipment having a post-accident safety function.*must be environmentally qualified for the resulting adverse pressure and temperature conditions.

2.0 CONTAINMENT FUNCTIONAL DESIGN Dresden 2 is a 2527 MWt General Electric Mark l BWR which has a primary containment consisting of a drywell, a pressure suppression chamber, and interconnecting vent pipes. The pressure suppression chambe~ is a steel pressure vessel in the shape of a torus located below and encircling the drywell. The chamber is approximately half filled with water. The vent system from the drywell terminates below the water level in the pressure suppre*ssion chamber, so that in the event of a pipe failure in the drywell, the released steam passes directly to the water where it is condensed.

This transfer of energy to the water pool rapidly reduces the pressure in the drywell and substantial~y reduces the potential for subsequent leakage from the primary containment.

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':*."':'.. ** In addition to t_he pressure absorption chamber, independent auxiliary coo 1 i ng systems are provided for the re a-ct or and containment coo 1 i ng under various normal and abnormal conditions. These are:

(1) A LPCI/containment cooling system is provided which serves three.

functions:

a.

To'.inject water into the reactor vessel subsequent to a postulated LOCA rapidly enough to reflood the core and prevent fuel clad melting *.

b.

To remove heat from the water in the sup*pression chamber.

c.

To spray water into the drywel 1 as an *augmented means of removing energy from the drywell as required.

(2) A shutdown cool~ng system is provided to remove reactor decay heat during shutdown.

(3)

(4)

An isolation condenser is provided for removal of decay heat from the care when the reactor ii isolated.

A HPCI system is provided for removal of decay heat and to provide coolant inventory control and heat dissipation from -the core to the I

suppression chamber under postulated slow depres~urization accidents..If the HPCI system *should fail to operate an automatic depressurization by blowdown will be employed through automatic opening of relief valves which vent steam to the suppression pool.

This blowdown will depressurize the vessel in sufficient time to allow the core spray or the LPCI function of the ECCS to adequately c6ol ~he core and prevent any clad melting *

(5)

Two core spray systems are provided which are designed to pump water under accident conditi ans from the pressure suppress ion chamber pool directly to the reactor.core by a separate spray header or sparger mounted in the reactor vessel above the core.

(.6) *A standby coolant supply system is provided by a cross-tie between the service water system and the feedwater system which maKes available an inexhaustable supply of cooling water from the river to-the reactor core and containment independent of all other cooling water sources.

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::-::-:--**- In the event of loss *of offsite power and failure of one diesel generator, minimum conta1nment cooling is provided by two low pressure core injection (LPCI) pumps.

These pumps are manually switched from core injection mode into the.residual heat removal (RHR) mode.

In the RHR mode; water from* the wetwell is passed through two RHR heat exchangers and returned to the wetwell. A containment spray system in the drywell is provided, but is not Safety Class* 1 and, therefore,. is not given credit in the containment analysis.

2.1 -Review of Dresden 2 FSAR Containment Design Analys'is There are two separate calculations which make up the containment design analysis. The first calculation is the mass and energy release analysis for postulated LOCA' s. This pro vi des the time d~pendent mass and energy input into the containment structure. The second calculation is the containment.

response to the mass and energy input.to the structure. This results in the time-dependent containment temperature and pressure profile. The severity of the containment response depends on the magnitude and nature of the break location. If the break is below the core the break flow will be initially single_ phase liquid. This results in a fast blowdown of the mass and energy release to the containment at a relatively low enthalpy.

If the break is above the core the break flow will be mostly single phase steam. This results in a much longer blowdown of the mass and energy release to the containment at a much high~r enthalpy.

Because of these effects, breaks below the core are found to produce the most severe pres~ure response in the containment and steam line breaks above the core produce the most severe temperature response.

The acceptance criteria used to evaluate the Dresden 2 Containment Design Analysis was based on the *Standard Review Plan (SRP).

For the containment design analysis to be found acceptable both the mass and energy release and the containment response calculation must meet the acceptance criteria specified in the SRP.

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' 2.2 Review of Dresden 2 Pipe Breaks Within: the ~Reactor. -Contai_riinent--.System.

The SRP specifies several acceptanc.e criteria applied to the mass and.

energy r~lease analysis for primary system pipe breaks.

Among these are the break location.. In the Dresden 2 FSAR the most severe mass and energy release

rate calculated for containment design was* done assuming a double-ended recirculatfon line. break *. The calculated peak post-accident containment pressure resulting "'from a double-ended recirculation line break is 48 psig *.

The peak wetwell p~essure ~as 27 psig. The peak drywel l temperature was

295°F.

The wetwell temperture profile was not reported.

In addition to recirculation. line breaks, the current criteria states*

that steam line breaks above the core. must be considered.

The licensee* did not per.form a steam line break analysis.

2.3 Reanalysis of Dresden 2 Containment Design As mentioned earlier in Sectiorr 2.1, Review of Dresden 2 Containment

. Analysis, there are two separate *calculations which make up the containment design analysis, the mass and energy release rate and the containment response.

The mass and energy release rate calculation can be the result of

either a recirculation line break or a steam line break. The recirculation line break results in the limiting condition for calculating the peak pressure inside the containment.

The steam li11e pipe break analysis is the most limiting case for temperature conditions inside the containment~ Both of these analyses were performed and are discussed below.

3.0 RECIRCULATION LINE PIPE BREAKS For a recirculation line br~ak a OBA LOCA generates the highest containment temperatures and pressures for breaks which occur below the core mixture level. The LOCA analysis was performed using the RELAP4-MOD7 computer code.

The RELAP4 input deck was obtained from Commonwealth Edison Company at*

the request of NRC.

The deck was used for analyzing operational transient using the computer code RETRAN.

RETRAN is an EPRI funded code based on

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The input deck was carefully reviewed for code options, initial and boundary conditions. The plant physical description was assumed to be as built. Additional information required was taken from the Dresden 2 FSAR.:

The initial and boundary conditions used in this analysis* were defined to satisfy the requirements of the. Standard Review Plan as-discussed below. -. The following is a listing of the initial conditfons and a summary of the assumptions used in thi~ analysis~-

(!) The reactor is operating atl02% of design power at *the time the recirculation pipe breaks. -- This maximizes the core heat generation rate.

(2) A complete_ loss of normal a-c power occurs simultaneously with the pipe break.

In addition, the single failure-assumption causes a*

1 ass of one di ese1 generator. _

(3)

The recirculation loop pipelin~-~s considered to be instantly severed. This results in the most rapid coolant loss and depressurization with coolant being dischargedfrom-both_ ends of the break.

The_ break area is ass urned to be 5. 62. sq. ft: and represents a double-ended break of one of the 28 inch diameter recirculation lines with the equalizer -line open.

The break area assumes the flow will choke in.the equalizer val~e arid in the nozzles of the ten jet pumps on the jet pump header of the broken l--i ne.

(4)

The. reactor is assumed to go subcritical at the time of accident in_itiation due to void formation in the core region.

Scram would also occur in less than one ~econd due to a high drywell pressure signal. The difference betwee*n shutdown at time zero and one second is negligible.

(5)

The sensible heat released in cooling the fuel and the core decay heat are included in the reactor vessel depressu~ization calculation. The rate of energy release is calculated using a conservatively high heat transfer coefficient throughout the depressurization. This maxi~izes the heat removal rate into the containment. Calculations of heat transfer from surfaces *exposed to liquid were based on nucleate boiling heat tr an sf er.

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0. 5 seconds after the accident, and are assumed to be fully clo~ed within 3 seconds.

By assuming rapid closure of these valves, the.

reactor ves~el is maintained at a high pressure, which maximizes.the discharge of high energy steam and '!'later into the primary containment.

{7)

The feedwater flow is assumed to stop instantaneously at time zero.*

This conservatism is used because the relatively cold feedwater

fl ow, if considered to.continue, tends to depressurize the reactor.

vessel, thereby reducing the discharge of steam an9 water into the primary containment*.

(8) The vessel depressurization flow rates are calculated using a discharge coefficient of 1.0., with _the Henry Fauske correlation of subcooled and Moody corr~lation*for saturated fluid.* A 14~7 psia back pressure was assumed to maximize mass and energy release throughout the blowdown.

The blowdown calculation was run until the primary system pressure dropped below the dr.,YWell design pressure of.. *.

62 psig. At* this time the 1.2 ANS decay heat curve was used.

The ll'esults of this analysis are the time dependent*IT!ass and energy release rates presented in Table 1.

3.1 Containment Response Calculation to a Recirculation Line Break The input data for the containment response calculation consists of the mass and energy re 1 ease to the cont a i ~ment, a* descriptive of the.containment heat remova*l systems and containment heat sink data.

The mass and energy release rate data used were taken from the blowdown of the recirculation line presented in the previous section~

T_he containment heat removal system consists of a pressure suppression pool, a residual heat removal system, and containment sprays. However, since the containment sprays are not Class 1 Safety.Grade, they were not taken into account. The containment heat sin.k data was also omitted since this analysis

is primarily for determining the maximum post-accident pressure and the effect.

. of heat sinks would be negligible in the short term *

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The residual heat removal (RHR) system consists of 4 low pressure pumps which take water from the suppression pool and pass it through two heat exchanges before it returns to the suppression pool. With loss of offsite.

power this system is reduced to 2 low pressure pumps.

The RHR system is manually operated and assumed to _start at 600 sec. after the break.

Prior to activation of the RHR system, the low pressure pumps add liquid*

to the reactor vessel. At the end of b.lowdown this flow discharges through the rec i rcu lat ion line break into the drywe 11 poo 1.

The* containment respo~se calculation was done using the CONTEMPT-LT/28 computer code. The program model represents the containment as three regions:

the reactor vessel, the drywell and the wetwell.

The physical model was obtained from the Dresden 2 FSAR.

A summary of the containment input model is given in Table 2.

3.2 Containment Response Results

.The containment pressure and temperature responses to a recirculation line break are shown in Figures 1 through 5.

The calculated transient reflects a peak post-accident containment drywell pressure of 47 psig and a temperature of 295°F.

The post-accident containment wetwell pressure and temperature are 31 psig and 169°F.

The containment design pressure for both the drywell and wetwell is 62 psig. There is, therefore, a substantial margin between t~e peak calculated pressure and the containment design pressure.

4.0 MAIN STEAM LINE PIPE BREAKS Analyses of the containment response to a steam line break were also made. This analysis is performed to d~termine the most severe long term press*ure and temperature condition in the containment following a pipe break.

The containment long term response is calculated for a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period~ To determine the most severe long term conditions in the containment the most limiting steam line break size must be found. For BWRs the most limiting break.size.*is-found by running a spectrum of break sizes.and taking into consideration, vent clearing in the suppression pool and the rate of

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The smallest break size of interest is the one which just clears.*

the vents. The largest break size of interest is the one which can permit reactor depressurization in no less than six hours, i.e., no ECCS required_.

The single active failure assumption was loss of offsite power with the failure of one diesel generator. The initial_ power was specified as 102% of the* design rating. The input and boundary conditions for this analysis were defined to satisfy the requ.irements of Section 6.2.1 of the Standard Rev1ew Pl an and are. discussed below.

The break was. assumed to be in.one of the main steam lines. The break discharge coefficient was set equal to unity. The*

critical flow models used were extended Henry-Fau.ske'for subcooled flow and Moody for saturated flow at the break junction. The RELAP 4 input deck used w~s based on tlfe same one used in the recirculation break analysis. The break location was moved to the. main steam Jine. up~tream of the MSIV in that line.

The break sizes that were run cons:isted of a O~-Of arirl l:.o *sq~.ft. break.

In order to be conservative with the respect to the containment response, it i.s necessary to maximize the steam that would exit from the break. This was done by assuming an infinite phase separation velodty' in. the upper plenum which prevented any entrainment *. The feedwater flow was assumed constant at 552.5 lbm/sec at 309.7S°F until the reactor scrammed on high drywell pressure of 2 psi g.

The steam fl ow was al so assumed constant at 552. 5 1 bm/sec at 559.2°F.

For the 0.01 sq. ft. break, the reactor scrammed at 65 seconds. For the 1.0 sq. ft. break this occured at 2 seconds.

Following blowdown, the mass and energy release rate reduces to that of a steam decay heat curve.

The mass and energy release rates for.the 0.01 ft2 and 1.0 ft2 steam line break are pres~nted in Tables 3 and 4, respectively.

4.1 Containment Response to a Main Steam Line Break T_he input data for the containment response calculation consist of the mass and energy release to the containment, a description of the containment heat removal systems and the available containment heat sink data.

The mass and energy release rates were taken from the blowdown rates presented in the previous section. Following the blowdown, the mass and energy release rates are reduced to decay heat steam.

As before, the only containment heat remova 1:

system modeled was the residual heat remova*l system mentioned in Section 3.1, ;

Containment Response to a Recirculation Line Break *. The containment heat sin~i data was also omitted.since this data was unavailable.

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4.2 Containment Response to Main Steam*Line Break Results The containment response result$ over a six hour period were calculated using CONTEMPT~LT/28. A two volume model representing the drywell and wetwell was used over the entire period.

The containment pressure and temperature

respo~se to the.01 ft2 steam line break is shown in Figures 6 through 10*.

The calculated: transient. reflects a peak post-accident drywell pressure of 15 psig and a temperature of 328°F. The resulting wetwell pressure and

temperature is 14 psig and 125°F. The response to. a.1.0 ft2 steam line.

break is shown in Figures ll through 15. The calculated transient in this

case reflects a peak post accident drywell pressure of 18 psig and a temperature of 335°F.

The resulting wetwel1 pressure and temperature is 17 p~ig and 161°F.

Both the peak drywell and wetwell pressures are substantially below design for both cases. The post acc.i dent temperature for the 1.0 ft2 steam line break results in the most severe temperature conditions and shoul~ be used for equipment qualification of safety related equipment.*

S:O CONCLUSIONS Based on* our review of the Dresden 2 docke.t and our subsequent analysis, we conclude that the Dresden 2 containment design pressure meets current NRC criteria. The containment atmosphere conditions as a result of a 1.0 ft2 steam line break provides the most se~ere temperature conditions for equipment qualification of safety related equipment.

6.0 REFERENCES

1.

Dresden Nuclear Power Station Unit 2 and 3, Safety Analysis Report.

-- --*-------**----'---.-----*------*-*---*----------*-*:"'""I'-*-,.*-*--':-.....:. __._.....

'* TABLE 1 Double-Ended Recirculation Line Break Release Rate Data (5.62 ft2 Break)

Time Flow Energy

{,;;..se=c~o.;.;..nd=s~)--------_....;.;:;.;.;;.:...;;..;;="'--------~..;;;.:...;~

( 1 bm/sec)

(Btu/lbm}

0.0 27211.0

' 552.0 27211.0 552.0 l

27211.0 553.0 I.

LO 2~0 3.0 4.0 '

5.0 27211.0 553.0 '

10.0 15'.0 20.0 25.0 30.0 35.0 45.Q 55.0 100.0 Decay Heat Energy Addition Rate

Time (seconds) 100.0 400.0 1000*.o 4000.0 10000.0 40000.0

*-***-----.-*---,.~

27211.0 27211..0 27211.0 27218.0.

19412.0 '

3012.0 1659.0 1037.0 516.0*

291.0 '

291.o Energy Rate (Btu/sec) 9.7096E4 6.8938E4 S.426E4 3.754E4 2.830E4.

9.943E3 554.0 556.0 562.0 572.0 591.0 570.0

711.0 1198~0 1189.0 1180.0 1180.o

)*_*_.

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TABLE 2*

Containment Model Input Data (Taken from Dresden 2 FSAR)

Drywell/Wetwell Data Free Air Volume (Cu.ft.)

Initial Pool Water Volume (cu.ft.)

Initial Temperature of Atmosphere (°F)

Initial Temperature of Pool (°F)

Initial Pressure (psia)

Relative Humidity (percent)

Pool Surface Area (sq.ft.)

HTC Multiplier Mass Transfer Multiplier Vent System Vent Pipes Number Internal Diameter Vent Tubes Fl ow Area, Total

Oowncomer Pipes Number Submergence Below Absorption Pool Water Level

Vacuum Breakers Number Vent Area Actuation set-point RHR Heat ~Exchangers _

Number Heat.Load

Primary Side Flow Secondary Si de Fl ow.

Drywell 158,236.0 '

o.o 150.0 lS-0.0 14.7 100.0

... 700 ~ 0 1.0 1.0 8

6 ft. 9 in.

285 sq. ft.

96 3 ft. 5 in.

12

. 2 2, 715 in Wetwell

'117,245 *. 0 112, 203. 0 125.0 125.0 14.7 100.0.

9*.900~ 6 '

1.E6.

1.0 0.5 psi for full open

2

l.02E8 Btu/hr each (at design temp.)

10,700 gpm 7,000 gpm

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Time*

(seconds) 0.0 1.0 10.0 1000.0.

4000.0 Decay Heat Energy Addition Rate Time (seconds) 4000.0.

10000.0 40000.0 Flow

( lbm/sec) 22*.1a 22.18 22.18 22.18 22.18 Energy Rate (Btu/sec) 3.754E4 2.833E4 9.943E3 Energy (Btu/lbm)'

.. 1197.

1197.

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Flow Energy (seconds)

{lbm/sec)

{Btu/lbm}

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1.0 2218.0 1198.

10.0 1218.0 1202.

20.0 1090.0 i201.

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80.0 861.0 1196.

120.0 610.0 1197.

160.0 530.0 1198.

200.0 412.0 1198.

400.0 412. 0 1198.

Decay Heat Energy Addition Rate Time Energy Rate (seconds)

(Btu/sec}

400.0 6.893E4 1000.0 5.426E4 4000.0

3. 75.4E4 10000.0 2.83E4 40000.0 9.943E3

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ML17194A384

Contents

Text

SUBJECT:

Enclosure:

Dear Mr. Del George:

SUBJECT:

Enclosure:

1.0 INTRODUCTION AND BACKGROUND

5.0 CONCLUSION

1.0 INTRODUCTION

6.0 REFERENCES

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ML17194A384

Text

SUBJECT:

Enclosure:

Dear Mr. Del George:

SUBJECT:

Enclosure:

1.0 INTRODUCTION AND BACKGROUND

5.0 CONCLUSION

1.0 INTRODUCTION

6.0 REFERENCES

Navigation menu

Search