ML17194A318

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Forwards Evaluation of SEP Topics XV-3,loss of External load,XV-4,loss of Ac Power to Station Auxiliaries & XV-19, LOCA Resulting from Spectrum of Postulated Piping Breaks within RCPB
ML17194A318
Person / Time
Site: Dresden Constellation icon.png
Issue date: 12/12/1981
From: Crutchfield D
Office of Nuclear Reactor Regulation
To: Delgeorge L
COMMONWEALTH EDISON CO.
References
TASK-15-03, TASK-15-04, TASK-15-19, TASK-15-3, TASK-15-4, TASK-RR LSO5-81-12-018, LSO5-81-12-18, LSO5-81012-18, NUDOCS 8112100090
Download: ML17194A318 (21)
v • d • e


Text

Docket No. 50-237 LSOS 1 2-018..

  • Mr. L. Del George Director of Nuclear Licensing Connnonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690 Daar Mr. Del George:

December 12, 1 981

SUBJECT:

DRESDEN'- SEP TOPICS XV-3, XV-4 ~~D XV-19 (systems)

By letter dated October 15, 1981, you sutmitted safety assessment reports for the above topics. The staff has reviewed these assessments and our conclusio~s are presented in the enclosed $afety evaluation reports, which complete these topics for Dresden 2.

These evaluations will be a basic input to the integrated assessment for your facility. The evaluations may be revised in the future 1f your facility design is changed or tf NRC criteria relating to these topics are modif1e~ before the integrated assessment ts completed.

Enclosures:

As stated cc w/enclosures:

See next page 8112100090 011212, PDR ADOCK 0500023i p

PDR J

Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch.No. 5 Division of Licensing NRCFORM318(10-BO)NRCM0240 OFFICIAL RECORD COPY USGPO: 1.981-335-960

Mr. L. DelGeorge cc Isham, Lincoln & Beale Counselors at Law

.One First National Plaza, 42nd Floor Chicago, Illinois 60603 Mr. B. B. Stephenson Plant Superintendent Dresden Nuclear Power Station Rural Route #1 Morris, Illinois 60450 Natural Resources Defense Council 917 15th Street, N. w.

Washington, D. C.

20005 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 Morris Public Library 604 Liberty Street Morris, Illinois 60451 Chairman Board of Supervisors of Grundy County Grundy County Courthouse Morris, Illinois 60450 John F. Wolf, Esquire 3409 Shepherd Street Chevy Chase, Maryland 20015 Dr. Linda W. Little 500 Hermitage Drive Ral~igh, North Carolina 27612

EIS COORDINATOR 230 South Dearborn Street Chicago, Illinois 60604 Dr. Forrest J. Remick 305 East Hamilton Avenue State College, Pennsylvania 16801 The Honorable Tom Corcoran United States House of Representatives Washington, D. C.

20515

  • .
  • e
  • DRESDEN 2 SEP Topic XV-3; Jos~;...rof EXterna 1 Load; Turbine Trip, Loss* of' Condenser Vaccum, Closure of Main* Steam Isolation Valve (~WR), and Steam Press.ure Regulator Failure (closed)

A loss of generator load causes a power/load unbalance which produces.a turbine con-.

trol valve fast closure: The sudden. loss of heat removal causes reactor pressure and neutron flux to increase. A reactor trip results directly from the power/load unbal-ance (load rejection scram) in about 0.01 seconds.

The licensee has performed two analyses on loss of external load.

One analysis assumes the reactor at 100% po~er? fa i 1 ure ~f the steam _bypass and the reactor tripped b.y the turbiF1e stop valve* closure in 1.5 s~sonds (Ref~ 1). A secon.d analysis assumes

~he steam bypass properly functioning and the reactor tripped by the load rejection scram in ~bout 0.01 secbnds (Ref. 2).

4!"'~--

II.

REVIEW CRITERIA

(

Section 50.34 of 10 CFR Part 50 requires that eac.h applicant for a construction permit or operating license provide an analysis ~n~ evaluaii~n at the design and ~erforma~ce of structures, systems, and components of the facility with the objective *Of assessing the risk to public health.and safety resulting fro~ operation of the facility, includ-ing detennination of the margins of safety during normal operation and transi.ent conditions anticipated during the life of the facility *.

Section 50.36 of 10 CFR Part 50 requires the Technical Specifications to include safety limits which protect.the integrity of the physical *barriers which guard against the. uncontrolled release of radioactivity.*

The General Design Criteria (Appendix A.to 10 CFR Part 50) establish minimum require-ments for the principal design criteria for water-cooled reactors.

e*

e GDC 10 "Reactor Desi9n 11 requires that the core and associated *coolant, control and

~(

protection systems *b_e designed with appropriate margin to assur~ that specified

. design conditions of the reactor cool ant pressure bqundary are not exceeded during normal operation, including the effects'of anticipated operational occurrences.

GDC 26 "Reactivity Control System Redundance and *c*apabil ity 11 requires that* the reac-tivity control system be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated operational

  • occurrences, and with appropriate margin for' malfunctions such as stuck rods,
  • specified acceptable fuel design limits a re not exceeded.

-~-

III.

RELATED SAFETY TOPICS Various other SEP topi~evaluate such items as the reactor protection system. *The

(

effects of single fan ures on safe shutdown capabil i ~)' a re considered under Topic VII-3.

IV.

REVIEW GUIDELINES The review is conducted in accordance With SRP 15.2.1.

The evaluation includes review of the analysis for the event atid identification of the features in the plant that miti-gate the consequences of the event as well as the ability of these systems to function as required.

The ex ten*~ to which operator action is required is also evaluated.* Deviations from the criteria specified in the Standard Review Plan are identified.

I*

.* - 3 ~

  • V.

EVALUATION The results of the licensee's first analysis indicate a maximum p.ressure_ of 1244_.ps~g *

(1375 p~ia allowable), ~nd a MCPR bf 1.07 (1.06 allowable). This analysis is !.

~

extremely conservative in that it assumes a delayed reactor trip.* Therefore, the.

additional over pressure that -m~y result from an i~itial reactor* power ot" 102%,

-required by th*e SRP Section 15.2.1, as opposed to the 100% assum~d in_the analysis would not be significant.

The r~sults of the licenseeis second analjsis.indicate a MCPR of 1.13. The maiimum pressure for this analysis is not stated in the results. However, ~e believe that this pressure is significantly lower than the 1244 psig of 'the first analysis.

. VI.

CONCLUSIONS As part of the SEP review for Dresden 2, we have evaluated the licensee's analyses of loss of external load and have concluded that the foitial *conditions ~acceptable and the peak pressures and MCPR,reached during the.transient, are within the accept-ab 1 e 1 i mits.

(

TURBINE TRIP I.

INTRODUCTION A turbine trip can occur as *a result*of many causes.includin,g loss of load. It produces a reactor isolation and_ increase in system *pressure with reduction in voids.

A reactor trip would be initiated on 10% closure of the.turbine stop valves in about 0.1 seconds.

The licensee has performed two analyses on turbine trip.

One analysis assumes the*

reactor at 100% power, failur~*of the steam bypas~ ~nd the reactor tripped by 10%

turbine stop valve closure in about 0.1 seconds (Ref. *2). A second analysis assumes the steam*bypass properly functioning and the reactor tripped by 10% turbine stop

- 4 valve closure in ab.out 0.1 !conds (Ref.* 2).

-~~!:.r.

II.

REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each*applicant.for a* construction permit or operating license provide an analysis and evaluati~n of the design and performance of structures, systems, and components of the facility with the objective of assess-ing the risk to public health and safety resulting froni operation of the facil_ity, including determination of the margins of safety during normal operations and tran-.

sient conditions anticipated during the. life of the facility.

Section 50.36 of 10 CFR Patt 50 requires the Technical S~ecifications to include safety limits which protect the integrity of the physical barriers which guard against the uncontrolled release. of radioac~ivity.

'The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum require-ments for the principal design criteria for water-cooled reactors.

-GDC 10 "Reactor Design" requi.res that the core and associated coolant, control and

~

~protection systems be designed. with appropriate margins to assure that specified acceptable fuel design limits are not exceeded durin*g normal operation, including the effects of anticipated operational' occurrences.

~DC 15 "Reactor Coolant System Design" requires that the reactor coolant and associated protection syste~s be designed with sufficient mar_gin to assure that the design condi-tions of the reactor coolant pressure boundary are not exceeded during normal opera-tion, including the effects of anticipated operational occurrences *.

GDC 25 "Reactivity Control System Redundan.ce ~nd Capabi.l i ty" *requires 'that the reactivity control systems be-capable of_ reliably controlling reactivity changes to assure that under conditions of normal operation, incl ucii ng anticipated operational occurrences, *and.with ~ppropriate margin for malfunctions such.as stuck rods, speci-fied acceptable fuel design limits are not exceeded.

5 e**

III.

RELATED SAFETY TOPICS a:.

Various other SEP topics evaluate such items as the reactor_protection system.

The effects of s~ngl e fai 1 ures on safe shutdown *~apabi l i ty are considered under Tohi c IV *. REVIEW GUIDELINES The review is conducted in accordance with SRP 15.2.2.

The evaluation includes review of the analysis for the event and identification of the features in the_ pl ant that mi ti gat~ the consequences of the ~~e~t as well as the ability" of these systems to function as required.* The extent. to which operator action is retjuired is also evaluated.* Deviations from the criteria specified in

  • the Standard Review Plan are identified.

V.

EVALUATION The results of the 1 i tens~e'-s first aria,ly~i.s i.ndicate *a maxi.mum pressur~ 1209 psig and a MCPR of 1.05 reached when the rea~tor is tripped assuming the steam bypass not available. The MCPR of 1.05.reached by this analysis is below the 1."06 a.n'owable for.moderate frequency events.

However, the inavailability of the steam bypass makes thjs transient an infrequent.event.

The results of the licen~ee's seco~d.analysis indicate a maximum pressure of 1182 psig is reached when the reactor is* tripp_ed assuming the steam bypass ava.ilable.

  • The MCPR is not reported for this analysis. However, it* is* demonstrated that this trim***

sient is bounded by the main steam isolation valve closure event which has been evalu-ated separately.

VI.

CONCLUSIONS As part of the SEP review for Dresden 2, we have evaluated the licensee's analyses of the turbine trip transient and have concluded.that this event is adequately boun~ed

/-:-.

  • by the mafo steam isolation valve closure event.*

~'...

LOSS OF CONDENSER VACUUM *,

. ;~' -

An instantaneous lbss of c~ndens~r vacuum i~ similar to a turbi~~ tri~ wi~h

.*bypass failu.re.

Therefare*,. a' se~arate analysis is not required.. for* this event

  • _:*~ ":_

INTRODUCTION

~"... ~.

Closure or the MSIVs results in loss.o:f steam removal fro~ the reactor and overpressur"i-.

zation of the primary system. -A direct reactor trip wo.uld* be initiated on 10% closure of the va 1 ves in about 0. 3 seconds.

/

The licensee *has performed two analyses on inadvertent closure of the main*steam isolation valves.

One.analysis assumes the reactor at 100% power,.complete failure of the steam relief valves arid the reactor tripped by high neutron flux in 3.0 seconds (Ref. 1).

A* second an~lysis assumes the steam relief valves properly func_tioning and the reactor tripped by 10% steam isolation valve closure in 0.3 seconds (Ref. 2).

II.. REVIEW CRITERIA

.. ----;-*---... -.... ~--. *- *...... ".

~

Section 50.34 of 10 CFR Part 50 requires that each appl~cant for a construction permit

  • or operating 1 i cense pro vi de an analysis and eva 1 uation of the design and performance of structures., systems, and components of _the i;aci 1 ity w1 th the objective of assess-ing the ri~k to pub.lie health and safety -resulting 'from operation of the facility, iricl udi ng determination. of the margin of safety during normal opera ti on and transient conditions anticipated durtng the life of the facillty.

~.:.: ::..: -. ---.:... ~-- -------- *-*-.

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t I

~:

Sectiori 50.36 of 10 £FR Part 50 requires the !echnical Specificat~ons to iriclude safety limits which protect the integrity*of the physical ba.rriers ~hich gt;Jard

  • 1 against the WlControlled release of radioac,tivity.

The General Design Criteria* (Appendix.. A to 10.CFR Part 50) establish* minimum require-ments for th~ pri nci pa 1 design criteria for water-cooled reactors. *

-~. ~~~~ ~-==~:~~.-_:._ __ ~~-==-~-~=~-*** *_:~~ ~-~_:_~ -.*.~~-(*;.[~,;~:.;*;*~::::~~~~=~~-;:~~;~zjJ,'.:~j~:.:>~::Z~*:s:~r-,~~~-~~-*... _::~J-:_*. ----------*------1 GDC 10. "Reactor De.si gn.11 requires that the. cbr~ and associated cool ant, control and protection systems be designed with appropriate margin to assu.re that specified.

acceptable fuel design *limits are not.exceeded dur.ing no~l operation, including the effects of anticipated operational occurrence.

. GDC 15 "Reactor Cool ant System Desi gn 11 requires that the reactor cool ant and associated protection systems be designed with sufficient.margin to.assure that the design condi-

_t~ °.~~.. ~.!. the. re.ac::!~t:.. ~ool ~nt_p_!:~~~~!~~~oundary a!:~ *.not exceeded duri n~~o~~~--~o~~.~~-ion,

including the effects of anticipated operational occurrences.
l.

'GDC 26 "Reactivity Control System Redundance and Capability" require*s that the reactivity control systems be capable of*. reliably controlling reactivity changes to assure that under conditi ens. of normal opera ti on,.including anticipated opera ti anal occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptabl~ fuel design li~its are not exceeded.

II I.

RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection system~ The effects of single failures on safe shutdown capability are considered under Topic vu.:.:a..

  • - *--- -----*--*----~*---------;-*- ___,,_.

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~.-*:.:~::... _.,..:_: ___ ~ __ -~,:. '.*:;. ~.,~... : -~.*: _ *-*-":~~-.. d'...*,_.*,_,..... --*":........ *.:.;;.:.:;~.*-.. * :.~. ~ ::.:.--..:..~~..: --":""'-.::iS;...'"-'*"~c,.. ___..... -**----,-~--., ------*------**-. _

IV.

REVIEW GUIDELINES The review is conducted in accordance*with SRP 15.2.4~(BWR only) and 15.2.5.

The evaluation includes review of the analjsis for the event and id~ntification of the features in the plant that mitigate the.consequences of the ~verit as well as the ability of the_se systems to function as required.

The exten~ to which operator

  • action is required is aJso evaluated~ Deviations from the criteria specified in the Standard Review. Pl an are identified.

C'*

(

-V.

EVALUATION Tne results of the licensee's first analysis indicate. a maximum pressure of 1375 psig (110% of vesse*1 design} is reached when the reactor is tripped by the i-mfi rect high neutron flux trip, assumi~g the steam relief valves not available. This analysis is extremely conservative in that it assumes a delayed reactor trip. Therefore, the additional overpressure that ~ay result from an initial ie~ctor powef of* 102%~ required by the SRP Section 15.2.3, as opposed to the 100% ass*umed in the analysis wciuld not be significant.

The results of the licensee's second analysis indicate a maximum pressure of 1210 psig is reached when the reactor is* tripped by the direct is.elation valve closure trip.

~-- ----*-**-

--~:...---~---~ ____.... ____,,_._,. _______.. _

..... :*:--- ***-**-**"***.*-**-....--**:*:*-*--:-----_-*_:-r--*-**.-~--------_...,, __.. ~- --- --~----------------------.:._: :..:..: *-..

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.It is further demonstrated that the main steam isolation.valve closure event is bounded by the loss of feedwater-transif;!'1t which. is* sepa~ately evaluated i_n.SEP*

Topic XV.:.5 *.....

-:~;,

The MCPR was not calculated for this transie~*L". However-, the results-of th~---l~ss of_

. external load event with failure>of the steam.bypass.*and a delayed reac.tortrip

... *~

?

. :* ~.,.. *."

~....

vr.

CONCLUSIONS As part of _the SEP review for Dresden 2, we have evaluated the licensee's analyses of the closure of the main ~team isolation v~lves ~nd.hav~ concluded_tbat~he jniii~J~

conditions are acceptable and the peak pressures reached during th~ transient are

  • within the acceptable limits.

The MCPR i'S-~ addressed_ in the evaluation* of the bound'i_ng analysis of loss-of load le!'-;,_-;-*

event reviewed under the same topic herein.

(

STEAM PRESSURE REGULATOR FAILURE*

A turbine pressure regulator failure *can occur in either zero output or maximum out~

put.

A failure to zero output is terminated by operation of the backup regulators.

When the fa;Jed regulator attempts to close the valv.es,_ pressure rises and the backup regulator.takes over.

This is a relatfvely minor transient and i.s similar to a pressure setpoirit increase which is.normal plant operation.

This transient can result in a.main steam valve isolation. Therefore, this transient is bounded by the main steam isolation valve closure event

  • i

. JI>*

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REFERENCES

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~

  • _._..; *'~... ~:. -. *.. :
  • f.....

L Letter from Thomas J.Rausch, Co~onwealth* Edison~***to*o*. M. *c~utchfield,,'.

NRC,

Subject:

Dr~sden 2 SEP Topic:*~ Accident and Transient Topics XV-1, 3, 4, 5, 7, 8, ~, 11, 13, 14, 15, 18~ 19 and 20.

Dated October 15, *198L *

2.

General Electric, "Additional lnfonnation Required for NRC Staff Generic Report on Boiling Water React~rs Volume l,'

11

NED0"'.'24708A, ~ciass I,. August 1979, Revision 1, December 1980.

(

DRESDEN.2 SEP TOPIC XV-4..

.; ~...

I.

INTRODUCTION Loss of AC power to station auxiliaries can occur as a resu.lt*of plant electrical. faults.

or. faults in the external power system.*.Following loss of power the.reactor will trip from loss of power t.o scr~m soienoids and later, within a few seconds,.the steam line isolation valves close.

The highest pressure transient resulting from loss of power to stat~ion auxiliaries is less severe than the main steam isolation valve closure.event or loss of all recircula-tion pumps.

The licensee has performed an analysis on loss of power to station auxiliaries (Ref. 1 & 2) *. The analysis assumes the.reactor at 100% power.

Heat remQVal *is 4!-'~--

accomplished by action of the relief valves and isolation condenser.

II.

REVIEW CRITERIA Section 50.34 of.10 CFR Part 50 r~quires that each applicant for a construction per~i~*

or bpeFating license provide an analysis and eval~atio~ of the 'design and performance of structures' systems: and components of t.he facility_ with the objective. of assess-ing the risk to public.health and safety resulting from o_peration of the facility.

t including detenninati on of the margins of safety durfng *normal. operations and tran-sient conditions anticipated during.the life of the facility *.

. *---*--~---**--*--*-**-__,.--** *---.----.----=.--------*******-*...,,, ------------.

  • The General Design Criteria (Appendix A to 10 CF.R Part 50} ~s.tabl ish minimum require-ments.for the principal design criteria for water~co6led reactors.

~ 2 -

GDC 10 "Reactor De~rgn" requires that the core and as.sociated coolant, control and protection systems be designed with ap~ropriate margins to assur~ that specified acceptable fuel design limits* are not exceeded during normal operation, including the effects of anticipated operational occurrentes.

GDC 15 "Reactor Coolant System Design" requires that the reactor coolant and associated protection sys_tems be designed with sufficient* margin to assure that the *design condi-tions of the reactor coolant pressure boundary are.not exceeded during normal opera-tion, including the effects of anticipated operational occurrences.

GDC 25... Reactivity Control System Redundance and Capability" requires'that the reactivity control systems be capable of r'eliably controlling reactivity changes to assure that under conditions bf normal oper~tion, including ant1cipated operational occurrences, and with appropriate margin for.~al functions such as stuck rods, speci-fied acceptable fuel design limits are not exceeded.

UL RELA.TED SAFETY TOPICS

'iarious other SEP topics evalua.te. such. i.tems as the. reactor protecti.on system.

The effects of single fai 1 ure on safe shutdown ca pa bi l ity are '.cons i.dered under Topk VII-3.

IV.

REVIEW GUIDELINES The review is conducted in accordance with SRP 15.2.6~

The evaluation includes review of the analysts for the event and identification of the features ih the pl ant that mi ti gate the corJsequences ~f the event as well as the: ability*of these systems to function as* required.

The extent to which operator action is required is also.evaluated. Deviations from the criteria specified in fhe Standard Review Plan are identified.

e*

V.

EVALUATION The re~ults of th~ licensee's analysis indicate that the in1tial portion of the.

I transient is,..similar to the loss of recirculation pumps.

The reactor trips in* about 2 seconds with no significant ihcrease in fuel temperatures or decrease in MCPR.

The MCPR was not calculated for this transient. However, the resul_ts of the main

.. **1 steam isolation valve closure event adequately bound. this transient *

... I VI.

CONCLUSION As part of the SEP review for Dresden,2, we have e'valuated the lfcensee's analysis of the loss of power to station auxiliaries and have concluded that this transient is adequately bounded by analyses performed on main steam isolation valve closure

  • .and loss of all recirculation pumps, discussed in SEP Topic XV-3 and XV-7 respectively.

REFERENCES

1. Letter from Thomas J. Rausch,,Commonwealth Edison, to D*. 'M. Crutchfield, NRC,

Subject:

Dresden 2 SEP Topic: Accident ~nd Transient Topics XV-1, 3, 4, 5, 7, 8, 9, 11, 13, 14, 15, 18, 19 and 20.

Dated October 15, 1981.

2.

General Electric, "Additional Information Required for NRC Staff Generic Report on Boil_ing Water Reactors Volume l, 11

  • NED0-24708A,,Cl-ass I,.August 1979, Revision 1, December-1980~

I

I.

INTRODUCTION TOPIC XV-19 (SYSTEMS)

I I

_I LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHI~ THE REACTOR COOLANT PRESSURE BOUNDARY DRESDEN UNIT 2 The objective of this review is to assure that the consequences of a Loss of Coolant Accident (LOCA) are acceptable, i.e., that the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 are met.

Loss-of-coolant accidents are postulated accidents th~t would r~sult fro~ the loss of reactor cooJant, at a rate in excess of the capability of the reactor coolant make:'t~~ system, from piping breaks in the reactor coolant pressure boundary.

The review consists of evaluating the licensee's analysis of the spectrum of loss-of-coolant accidents including break location, break size, and initial *co.nditions assumed, the evaluation model used, failure modes and* the acceptability of auxiliary systems used.

II.

REVIEW CRITERIA Section 50.34 -of 10 CFR P~rt 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance.of systems provided for the prevention of accidents and the mitigation of.the consequences of acci~e~ts.

e** Section* 50.46 of 10 CFR Part 50 requires that all light water reactors with zircaloy cladding shall be provided with an emergency core cooling system designed so that i ~s performance fa 11 owing a LOCA satisfies the criteria set forth in that section.

Pe.rformance is calculated with an:* eyaluation model satisfying the requi"rements of Appendix K to 10 *cFR 50.

The General Design Criteria {Appendix A to 10 CFR Part 50) set forth the criteria for the design of water-cooled reactors.

GDC 35 "Emergency.Core CoolingQ requires that a system be provided to provide abundant emergency cor~ cool-ing whose function is to tran~fer heat from the core following a loss of coolant such that (1) fuel and clad ~amage that could inter-fere with continued effective core cooling is *prevented and (2) clad metal water reaction is limited to negligible amounts.

The system shoul.4--mave suitable redundancy and i n*terconnect*i ans such that sys tern fun ti on can be maintained assuming* a single failure and assuming availability of only on.site or only off-site power supplies.

. (

III. RELATED SAFETY TOPICS.

Topic Ii'I-5.A, "Effects of Pipe Breaks ori Structures, Systems and Com-ponents ~nside Containment". ensures that the ability to achieve safe shutdown and to mitigate the con~equences df a pipe break accident is not impaired by the dynamic effects of the break.

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.. - 3,....*

Other SEP.topics consider the emergency power supplies, eff~cts of* flooding of safety-rel~ted equipment (vr-7.D) as well as failure modes of the ECCS (VI-7.C).

~

In addition, such areas as containment integritj and isolation, post accident chemistry and Engineered Safety Feature systems a~e considered as part of SEP topics. Topics VI-2.D and VI-3 address the capability of the containment heat

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

                  • -*****--*. *7..* *- -----:*-**,- *-;-*

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

removal systems* __ ~o-* ~~~ev_i~te_:~e pressure/t~mpera~ure transient so_ that the cony

~*.** - -*-

--~..*...

tainment is not overpressurized.

IV.

REVIEW GUIDELINES The review of ECCS performance during a LOCA is conducted in accordance with Standard Review Plan Sections 15.6.5 and 6.3.

The plant is considered to be I.

adequately designed against a LOCA if_ the criteria of 10 CFR 50~46 are met.

The radiological consequences are addressed in a separate evaluation._.._7 -

V.

EVALUATION.

Assuming the most pessimistic combination of circumstances which could lead to core uncovery and excessive heatup following a *loss-.of;..coolant accident, fuel*

The ECCS in Dresden Unit 2 provides the necessary protection to mitigate the consequences of a loss-of-coolant accident.

The ECCS is*automatically actuated on low-low reactor vessel water level or on high drywell pressure signals.

The licensee has analyzed the performance of the emergency core cooling system

. (ECCS) in accordance with 10 CFR 50.4~ a~d 10 CFR 50 Appendix K.

The limiting failure for large brea~s ~as ideritiffed as th~ failure of ~ LPCI injection-~alve.

The break spectrum a*nalysi.s. performed w1.th the General Electri.c evaluation model

identified the worst break as the co.mplete rupture of a recirculati_onsuction line (Reference i). The highe~.t peak~.. cla_d_:temperature (2_2_0Q°F) __ is.. reCiched:.for this break; therefore, sma 11 breaks* are bounded by the 1 arge break aha lys is

    • -*~-

~---*-

(Reference 2).

Analyses provided *for::the drilled lower tie plate designs are also bounded by the 1979 analysis (Reference.3) and the r:esults are incorparated in Reference 1. *The large break analyses were accepted* by the

  • staff '-in connection with. its review of. t'he. Cycle 8 re.l cad for Dresden Unit 2 (Reference 4).

Dresden *2 is a BWR 3 which does not currently include the LPCI modification.

Based on a review by General Electric (Reference 5), the systems remaining operable for a large suction line break assuming**a LPCI injection valve failure a~e"the high pressure coolant injecti6i1 system (HPCI) syttem, the ~u~omatic depressuri zation system (ADS) and two core' spray components.

For ~-LPCI mod plants, ccinsideration should be made of a DC p~wer s~urci failure on the availability of ECCS systems.

The GE review showed that for a large sucti:on break, assuming a DC power source failure, one core spray component, *two LPCI pumps and the ADS are available.

The.Dresden 2 desi'gn

~ati.sfi_e~ thi.$ mi.ni.m.um requirement, as stated in_ Reference 6.

In addJtion, the licensee has committed te> modifications to provide automatic switchover from the primary to an alternate 125 volt DC source.

This modification will ensure the qvailability of the HPCI system i~ the event of a DC power source failure.

VI.

CONCLUSIONS As part of the SEP review of Dresden Unit 2, the* loss-of-coolant analysi~ was reviewed against the acceptance criteria of SRP Section 15.6.5 and 6.3.

The initial conditions relative to single failure, break size and location, power level and operating conditi-ons have been reviewed and found to conform to the requirements of the SRP.

The ana.lys is was performed_ with an approved evaluation model and the resµlts were found to be acceptable~

~*i;. :-

REFERENCES

1.

NED0-24146A - April 1979 - LOCA Analysis Report for Dresden Vnit 2 and 3 and Quad Cities 1. and 2.

2.

NED0-24708A - August 1979, Rev. 1 - December 1980, Additional Information Required for NRC Staff Generic ~eport on Boiling_ Water Reactors, Volumes 1 and 2.

3.

NEDE-24094, January 1978, Loss-of-Coo)ant Accident Analysis Methods for BWR 2 and 3 with Drilled Lower Tie Plate~.

4. Safety Evaluation by the Office of Nuclear Reactor Regulation ~iJf)porting Amendment No~ 58 to Provisional Operating License ~o. DPR-19, Common-weal th Edison Company; March 31, 1981.

/ r 5.. Letter, R. E. Engle (GE) to P. S. Check (NRC)*, *"DC. Power Source Failure for BWR/3 and 4, 11 dated November.1, 1978 (with Attachment).

6.

Letter, R. F. Janecek (CEC) to T. A.

Ippoli~o (NRC), June 12, 1980.

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