ML17194B265

Forwards NRC Responses to Util 820406 Comments on Evaluation of SEP Topics VI-2.D & VI-3.Revised Table 4 to 820610 Evaluation Encl.Evaluation Final & Will Be Used as Basic Input to Integrated Safety Assessment

ML17194B265
Person / Time
Site:	Dresden
Issue date:	08/19/1982
From:	Oconnor P Office of Nuclear Reactor Regulation
To:	Delgeorge L COMMONWEALTH EDISON CO.
References
TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR LSO5-82-08-038, LSO5-82-8-38, NUDOCS 8208240398
Download: ML17194B265 (12)

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August 19, 1982.

~Pocket No. 50-237 LSOS 08- 038 Mr. L. De 1 George Director of Nuclear Licensing Commonwealth Edison Company

• Post Office Box 767 Chicago, Illinois 60690

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM (SEP) FOR DRESDEN NUCLEAR POWER STATION, UNIT 2 - EVALUATION REPORT ON TOPICS VI-2.D AND VI-J (DOCKET NO. 50-237)

References:

1.

Letter from Q. Crutchfield to D. VandeWalle, dated **

June 10, 1982, on the same subject.

2.

Letter *from T. J. Rausch, Commonweal.th Edison Co. to P. O'Connor, NRC, dated April 6, 1982, on *the same subject..

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By letter dated Apri.1 6, 1982, (Reference 2),.you provided comments on ~""

our evaluation of SEP Topics VI-2.D and VI-3 for the Dresden 2 Plant.

(Reference 1).

• Enclosed is our responsi) to these comments which should AoP*

be attached as Appendix B to our Evaluation Report on the subject topics. a,.(vfa./i"'Cf.

In addition, a transcriptional error was found in Reference 1, Table 4

-of Appendix A, Page. 17. Since the correct value was used in the analysis,

- our conclusion is not affected.

Please replace Table.4 with the enclosed revised sheet.

1

~ - With this correction, we consider our June 10, 1982 evaluatfon final and 1 it will be a basic input to the integrated safety assessment for your*

• facility.

These topic assessments may be changed in the future if your a-MIS facility design~s changed or if NRC cr1teria relating to. these topics

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are modified be't_~,re. the integrate~ assessment *is. completed.

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nas Operating Reactors Branch ~o. 5 t//7rr'?.;,

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. 7 /1~/82 Lr:tfl**7 /82 D1vis1on of L fcensing

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DATE*. *. * * * ** *... *. * * * * * * *. *... * * * *. *. * * ** * *. * * * * * *. *..*. ** * *. * *** ** * ** * * * * ** * *.z1.'i.1.a2._......... z1.'!..1.a2...........7L!:1.8.z..........z1. 1\\.1.az........

NRC FORM 318 (10-80) NRCM 0240 0 FFICIAL RECORD COPY USGPO: 1981-335-960

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~1,r. L. DelGeorge cc Robert G. Fjtzgibbons, Jr.

Isham, Lincoln & Beale Three First National Plaza Suite 5200 Chic~go, Illinois

• 60602 Mr. Doug Scott Plant Superintendent Rural Route #1 Morris, Illinois 60450

. u. s. Nuclear Regulatory Commission Resident Inspectors*Office Dresden Station RR #1.

Morris, Illinois 60450 Mary Jo Murray Assistant Attorney General Environmental Control Division 188 w. Randolph Street Suite 2315

. Chicago, Illinois 60601 Chairman Board of Supervisors of Grundy County

.Grundy County Courthouse Morris, Illinois 60450 111 inoi s Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Sprin~field, Illinois 62704 U. S. Environmental* Protection Agency.

Federal Activities Branch Region V Of ff ce AJTN:

Regional Radiation Representative 230 South Dearborn Street Chicago, Illinois 60604 The Honorable Tom Corcoran United States House of Representatives Washington, D. C.

20515 John H. Frye, III, Chairman

• Atomic Safety and Licensing Board

u. s. Nuclear Regulatory Commission*

Washington, o. c..20555 Dr. Martin J. Stefndler Argonne National Laboratory 9700 South Cass Avenue Argonne, Illinois 60439 Daniel Mintz, Esquire.

Counsel for P*etitf oners...

(Citizens for a Better Environment)

Suite 1600, 59 E. Van Buren Street Chicago, Illinois 60605 Dr. Robert L. Holton School of Oceanography Oregon State University Corvallis, Oregon 97331 James G. Keppler, Regional Administrator Nuclear Regulatory Commission, Region III 799 Roosevelt Road Glen Ellyn, Illinois *60137

APPENDIX B:

STAFF RESPONSES TO LICENSEE-COMMENTS This appendix contains the NRC staff's responses to comments received from Commonwealth Edison Company (Reference Bl) regarding our draft Safety Evaluation Report, Appendix 'A.

THe licensee's corranents are also:presented as an *attadunent~

to Appendix B.

The staff's responses were prepared by the Containment Systems* *;

Branch.

The following is a brief summary of t°he*containment analyses.. perfonned,. the.

• response to Commonwealth Edison Company corranents,.and*our conclusion.

Summary of the Dresden 2 Containment Analysis The containment analyses were perfonned (Referen*ce 82) *for-postulated recir-*

culation line and main ~team line breaks (MSLBs).

The recirculati6n line.*

break analysis, a 5.6 2 ft-. 2 break, produced the maximum drywell pressur*e *re-

_sponse of 48 psig and wetwell response of 27 p~ig. :Both values are :below the.

the containment design pr~ssure of 62 psig.: The MSLB*analyses were perfonned

• to.determine the maximum dry~ell temperature.response.

Two break sizes were*

• examined. the 0.01 ft. 2 and the 1.0 ft. 2 breaks. *Steam flo~ was maxi~ized by assuming an infinite phase separation velocity in the upper plenum. *Since heat*

sink data were not available to the NRC, the only.containment heat removal system

. modeled was the residua~ *heat removal system.

The containment response was cal-culated (as with the recirculation line break). using CONTEMPT*LT/28.

The maxi-mum drywel l_ temperature obtained was 335°F which resulted from the 1 :*o ft,:2 break.

The maximum emergency environment for Dresden 2 specified in the FSAR

• is 320°F for two.hours.

The temperature limit is, therefore, exceeded und*er these assumptions.

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• Response* to Comments A.

Design Basis LOCA A.l

. The litensee has indicated that a break size~of-4.77 ft. 2 is used in the current Exxon LOCA analysis. The LOCA analysis performed in Appendix A was based on a break size of 5.62.ft. 2, the ~~lue given in FSAR Section 5.2.3.2. The LOCA analysis perfonned in Appen-dix A yielded a containment pressure response which was below the *

. design value of 62 psig. Since the Appendix A analysis does not exceed the design value the staff does not believe that a reanalysis *

. is necessary even though the resulting containment responses would be lower with the smaller break size.

B *. Large Steam Break B. l The SER blowdown assumptions included the use of an infinite p_~_as~

separation in the vessel during the transierit-wnich the -stafr-acknow:..:

ledges to be very.conservative. This is used to maximize the steam released to the drywell.

In the absence of a licensee pool swell.

• analysis to determine the specific break size which separates the

• single.steam phase flow from the two phase flow pattern the more

• conservative.alternative, namely, the inffnite phase separation must be* chosen.

B.2* The licensee coriimer.t~d t.hit the reactor would scram at so:ne time earlier ~han the two second point used in the analysis.

The staff believes that this may be the case; the two second point was

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.i e* an input assumption to the RELAP computer runs. However, the ef-:

feet of the change in time would not be significant. The RELAP analysis performed in.Appendix A far* the large MSLB did not factor in the effect of neutronics, i.e., the reactor consisted of aves-sel containing the mass and energy corresponding to 102% power.

It did not include, however, the energy ~eveloped from continuing *

• power operation so that in effect the reactor was scrammed*at time zero. This effect was counter-balanced by the isolation of the steam lines. at time zero also which, in reality, would not occur until the system trip signal of low steam line pressure would occur.

The st~ff believes that these two effects yield a ~onservative:re~

sult with the additional effect that the only real impact. of the two second scram time is to isolate...the fe.edwater.line-at -that time.

The difference between feedwater isolation at two seconds or some.

time less than that ~ill not substantially affect the ~eactor blow-down.

The sraff do~s not believe that a reanalysis is warranted.

C.

Small Steam Line Break C.l Although the assumption of infinite phase separati6n is unrealistic for the DBA*analysis, the staff disag.rees with CECo that it is un-realistic for a small MSLB because vessel lev~l swell is reduced in a small break accident and the water level in the vessel will not.

rise to the level of the break elevation.

C2.

Refer to the response-to comment 82.

C3.

The staff agrees that operator action taken at 10 minutes may have significant* effect on* both the duration and peak temperature_s *reached.

.. However, in perfoming the analyses in Appendix A,.the staff felt. that

• neither ADS actuation nor drywell containment spray actuation would be:

appropriate.

ADS operation causes a rapid cool down* of the vessel, in excess of the l00°f/hour GE recommended maximum rate.

For* this: reason*

the operator may choose not to actuate ADS for *a small steam l foe :break.

The drywel 1 containment sprays are not safety grade* *and, although: they*

may be actuated in this case, the staff does not believe it conservative to rely on them..

~on-safety grade systems may not be-relied upon as a first.l inc defense.

Consequently, the staff does not feel that a reanalysis is necessary~

General Comments D.l The staff agrees that the two points analyzed for the steam line

_ breaks do not constitute a spectrum.

The-two pojnts were c;_hosen_*

as representative of the high and low end of the MSLB sp~ctrum.' Al-though they do not constitute a spectrum of break sizes it was felt that t~ey are the most: significant bre~k sizes which mu~t*be examihed.

[1.2 Passive heat slabs were not considered due to lack of infomation.

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The peak* drywell temperature was obtained from the 1.0 ft.2.MSLB at 10 seconds into the'transient. Passive heat sinks would not have

~*significant effect on reducing the temperature rise tb the peak value during the initial few:seconds of the transient.

0.3 The licensee has noted a difference in feedwater temperature of appro-ximately 32°F whereby the Appendix A analysis uses 309.75°F and the*

OPL-3 value is 341°F.

1Since the feedwater flow is stopped at the time of reactor scram the effect of this temperature difference wou*l d be

small and would impact predominantly the drywell pres.sure response, which already is well below the design.value of 62 psig.*:The-staff, 1

therefore, does not believe that this effect would.merit reanalysis.

,0.4

• The staff agrees that the initial wetwel l temperature of l-25°F is 30°f greater than the Technical Specifications allow.

• The effect of this would be to slightly raise the drywell and wetwell pressures *. Since, as _previously mentioned, these are well within allowable values the staff*

does not believe this to be a si~nificant factor in affectirig the con-cl us ions reached: concerning the containment pressure response~ *

'"j D. 5.

The staff agrees. that the drywel l pump ba*ck system will increase the drywell pressure 1 psig *. The analysis was based on FSAR Table 5.2.l

.. 0. 6 which lists the nonnal internal pressure as oein9-atmospher-ic.

The.. -,

. staff believes that this, too, does not alte*r the conclusions regard-ing the acceptability of the containment pressure respens~..

FSAR Table 5.,?."2 g'ives the range of relative humidity in the primary containment as 20% t6 100%.

As a. realistic approach we must use the

• 100% value, since it is. likely at least some time.

• o.7 The drywell,pool surface area of 700 ft. 2 was selected as an estimate based on our best engineering judgement.

Our experience with the CONTEMPT.LT/28 code is that it is not very sensitive to the value specified for the:;drywell pool when a pressure*

suppressfon type containment is modeled.

Consequently, :we do not 'feel that the results would change significantly with~ different fnput value fot the drywell pool.

~

Conclusions The licensee has stated through their comments (

Reference:

l) that "the NRC.

staff cannot make a safety determination for this topic unless it performs the analysis again using realistic assumptions and heat sinks." The staff has attempted to show in this Appendix that the comments made:by.the licensee will.

not have a significant effect on the overall conclusions reached in ~he SER.

concerning SEP Topics VI-2.D and VI-3~ i.e., the containment pressure and tern-*

perature response to LOCA and MSLBs._ The staff feels that although ma~y of the

• comments made were reasonable, in total they do not point to.the need* for a reanalysis~

Consequently, the SER (Reference 2):predicts adequatelyw and without undo conservatism, the containment temperature and pressure responses and no reanalysis is felt to be necessary.

References Bl. Letter from T. J. Rauche, Cormtonwealth Edison Company, to P. O'Connor, NRC, dated April 6, 1982.

82. Letter from O. M. Crutchfield, NRC, to L. DelGeorge, Commonwealth Edi son Company, dated December 28, 1981.

'l'tT y Commonwealth Edison One First National ~

Chicago, Illinois Address Reply to:

Office Box 767 Chicago. Illinois 60 90 Mr. Paul O'Connor Operating Reactor Branch D5 Division of Licensing U.S. Nuclear Regulatory Commission Washington O.C. 20555 April 6, 1982

Subject:

Dresden 2 SEP Topic VI-2.D, Mass and Energy Release for Possible Pipe Break Inside Containment and VI-3, Containment

Reference:

Pressure and Heat Removal Capability

• NRC Docket No. 50-237 D.M. Crutchfield letter of December 28, 1981 to L. DelGeorge

Dear Mr. O'Cqnnor:

Commonwealth Edison has reviewed the above referenced letter.

Edison believes that the staff's analysis is overly conservative and *as a result is a conservative upper bound. for:~

containment temperatures.

Edison does not-beli-eve-the analysis provided constitutes a valid basis for equipment qualification~

Edison's specific comments concerning the Staffs analysis:

are as follows:

A.

Design Basis LOCA -

l. Break size of 5.62 Ft2 is too large (~ssumes cross~connection of recirculation system not allowed by operating license)._ A break size of 4.77 Ft.2 is utilized in current Exxon LOCA analysis.

B.

Large Steam Break (l.O Ft.2).

1. The assumption of -infinite phase separation utilized in the RELAP Slowdown model is unrea-listic.

Level swell and resultant moisture introduction to the vessel dome significantly alter the mass energy '

release and yield lower peak containment temperatures.

2. The Rx scram at 2 psi drywell pressure should occur earlier than the 2 second time point used in the Slowdown Analysis.

This time is at odds with the NRC Drywell pressure trace. p~ovided with this report.

8204190223 820406 PDR ADOCK 05000237 P

PDR

c.

Small Steam Break (0.01 Ft.2)

IL

l. The assumption of infinite phase separation utilized in the RELAP Slowdown model is-unrealistic.

Level swell and resultant moisture introduction to the vessel-dome significantly alter the mass energi release and-yield lower peak containment temperatures.

2~ The Rx* scram at 2 psi drywell pressure should ~ccur

" earlier than the 65 second point used in the *analysis.

Note NRC Drywell Pressure trace ~rovided in this report.

3. Operator action taken'at 10 minutes to terminate this transient may have a significant effect on both the:

duiation a~d peak temper~tures reached, i.e.

. containment spray, or rapid depressurization.

General Comments

• J1. The two points analyzed for steam breaks do riot constitute a spectrum.

2. Passive heat slabs were not considered, due to lack of information, but can have a ~igni ficant effect.

3. Feedwater temperature of 309~-750F is' low when**

compared to OPL-3 value of 34loF.

4. The initial wetwell temperature of 125oF is 300 greater than the Technical Specifications allow.

5. Drywell initial pressure is 15.7 psia due to pumpback system.

6. :A drywell relative humidity of 100% seems high.

80% is more representative.

7. Drywell pool surface area of 700 Ft.2 is low.

1100

• Ft.2 is ~ore typical *

. It should be noted that one of the original concepts of SEP was to determine the real margins in older plants.* When the.staff.

follows curtent guidance and makes conserva~ive, or as was done several times in this analysis, grossly conservative assumptions, the goal of determining how much margin the plants really have is

• lost.

The original intent of SEP was to review the plant against :

current licensing procedures and where the plant was at variance with current procedures determine whether or not the variances have

'significant safety implications.

The staff cannot make a safety*

determination for this topic unless it performs the analysis again

• using realistic assumptions and heat sinks.

The plant conditions.

should be based on the as-built plant conditions.

~*

e* Please address any questions you may have concerning this matter to *this office.

One (1) signed original and thirty-nine (39) copies of this transmittal have been provided for your*use~*

Very truly yours,

. r.~J!¥.

~J.

Rausch '

Nuclear Licensing Administrator Boiling Water Reactors NPS:mnh/lm cc:. RIII R~sident Inspector, Dre~den 3837

h.

• e --

TABLE 4 MSLB Mass and Energy Release Rate Data (1.0 ft2 Break-)

Time

. (seconds) o.o 1.0 10.0 20.0 40.0.

60.0 80.0 120.0.

160.0 200~0 400.0 Decay Heat Energy Addition Rate Time.

{seconds}

400.0 1000.0 4000.0 10000.0

• AOOOO.O Flow

( lbm/sec) 2218.0 2218.0 2232.0 I 1090.0 930.0 970.0 861.0

. 610.0 530.0 412.0 412.o Energy Rate (Btu/sec},

6.893E4 5.426E4

3. 75_4E4 2.83E4 9.943E3 Energy (Btu/lbm) 1198.

1198.

1202.

. 1201.

1198.

1196.

1196.

1197

• 1198.

1198.

1198.

ENCLOSURE 1

...