ML17194A379

Forwards Safety Evaluation of Util 811015 SAR Re SEP Topic XV-1, Decrease in Feedwater Temp,Increase in Feedwater Flow & Increase in Steam Flow. Tech Spec Changes Required for Turbine Bypass Sys

ML17194A379
Person / Time
Site:	Dresden
Issue date:	12/29/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Delgeorge L COMMONWEALTH EDISON CO.
References
TASK-15-01, TASK-15-1, TASK-RR LSO5-81-12-095, LSO5-81-12-95, NUDOCS 8112310316
Download: ML17194A379 (15)

Text

{{#Wiki_filter:* December 29, 1981 DOcket No. 50-237 LS05 12-095 RECEIVED OECS_O 198t ~ Mr. L. Del George Director of Nuclear Licensing COnnnonwealth Edison Company

• Post Office Box 767 Chicago, Illinois 60690

SUBJECT:

DRESDEN 2 - SEP TOPIC XV-1, DECREASE IN FEEDWATER TEMPERATURE, INCREASE IN FEEDWATER FLOW AND INCREASE IN STEAM FLOW In your letter dated October 15, 1981, you submitti'd a safety assessment report on the above topic. The staff has reviewed your assessment and our conclusions are presented in the enclosed safety evaluation report. Our report completes this topic evaluation for Dresden 2. As noted in the evaluation of the feedwater controller malfunction event, the staff will requi~ Technical Specifications changes to conform with current licensing practice if credit is to be given for operation of the turbine bypass system in the analyses. The enclosed safety evaluation will be a basic input to the integrated safety.assessment for your facility. The assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this topic are modified before the integrated. assessment is completed_. Sincerely, Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing

Enclosure:

As stated cc w/enclosure: --:-- ____ Seel nex ~- p_a_g!_ _ ~-----{:al*l-2310316 811229 D:, I . PDR*' ADOCK 050002371 I P.DRI ... _P_*,'-**--.-*... L ( NRC FORM 318 (10-BO) NRCM 0240 OFFICIAL RECORD COPY USGPO: 1981-33$-960

UNITED STATES

• NUCLE.AR REGULATORY COMMISSION Docket No.

50~237. . LS05 12~095.

• Mr~ L Del George

. Director of Nuclear' Licensing Conunonwealth.Edison Company

• .Post Offi.ce Box 767 Chic~go, Illinois 60690

Dear Mr. Del George:

WASHINGTON, D. C. 20555

• December 29,.1981.

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SUBJECT:

DRESD_EN 2 -SEP TOPIC XV-1, DECREASE IN FEEDWATER TEMPERATURE,*.* INCREASE IN FEEDWATER FLOW. AND INCREASE IN STEAM FLOW In your letter dated October 15, 1981, you submitted a safety assessment report on the above topic. The staff has reviewed your assessment and our conclusions are presented in the enclosed safety evaluation report. Our report completes this topic evaluat~on for Dresden 2. As noted in the eva 1 uation of the feedwater contro 11 er malfunction event; the staff will -require Techni-cal Specifications* changes to conform with . current 1 i cens i ng practice if credit *is to be given for operation of the - turbine.bypass system in the analyses. The enclosed safety evaluation will be a basic input* to the integrated* safety assessment for your faci 1 i ty. The assessment may be revised in the future if your facility design is changed or if NRC criteria relating to this* topic are modified before the integrated assessment is completed.

Enclosure:

• As stated cc w/enclosure:

-See.next page*,,-;\\~;"~\\-,". Sincerely,

• .. tJ~li-~

k Dennis.M. Crutchfield, Chief d.* ~ Operating Reactors Branch No. 5. Division of Licensing ~*~:-.; :.. ':. ..... *?.

Mr~ L. De1George . cc Isham, Lincoln & Beale -., Counsel ors at Law \\_. One Fir~t National P1aza~*42nd Floor. Chicago, Illinois 60603 -*. _ ~M~. B. s. Stephenson Plant* superintendent Dresden Nuclear Power Station Rural Route #1 Morris~ Illinois 60450 Natural.Resources Defense Council 917 15th Street, N. w. Washingto~. O. 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 C_ounty Courthouse Morris, Illinois 60450 JohnF. Wolf, Es.quire 3409 Shepherd Street

. Chevy Chase,_ Maryl and 20015 Dr. Linda-W. Little 500 Hermitage Drive Ral~igh, North Carolina 27612 ~ -~ DRESDEN 2 ' .*Docket No~.50-237 .. *':i.-... ; Illinois *Depa~tment of. Nuclear.:Safety* ~

• 1035 Outer Park Drive, 5th Floor Springfield, Illinois 62704 *. *.. *'.

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-u~**s. Environmental. Protection Agency~*, -**.* Federal Activities Branch Region V Office ATTN: EIS CO!JRDINATOR. 230 South

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• Chicago, _Illinois 60604_ **

-~... -..... ;. ';. ~* Dr. Forrest _J. Remick ** _ :~, 305 East Hanri.lton Avenue ****.. . State College. Pennsylvania*. 16801.

• The Honorable Tom Corcoran:

United States House _of Repre:sentatives.* Washington, D. C. _ 20515 . '. ~*

DRESDEN 2 . ~

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~~ '* SEP. TOPIC XV-1:.. DECREASE IN FEEDWATER, TEMPERATURE~ INCREASE IN.FEEDWATER.. :*,;.__, FLOW, I NC REASE IN STEAM FLOW ~:.~- "... :.: -~*,--... '. ::. }:: :~;-:<--: \\,::{ :. ~*. ;.

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I *. 'INTRODUCTION .\\ : .~~~-----~-.- ..... -.. I .1.. Feedwater heating can be lost-by closure* of the._steam extraction lines to the;. ryeaters or the.bypassing of feedwater around ttie heaters. In either case the*

• reactor vessel receives cooler feedwater and there is an increase in core inlet subcooling.

The decrease in coolant void fraction and the negative* void, react-* ivity coefficient. result in a gradual initial increase in reactor power. The event could occur with the reactor in either manual or au.tomati-c control mode *. .. In the. automatic control mode*, there is some compensation* for reactor power increase by modulation of core flow and the event general'ly is less severe... In the manual control mode, and as*sumi ng no *corrective operator actions, the reactor power could either reach a higher equilibrium value below the scram setpoint or increase sufficiently to cause automatic scram on high neutron flux. The power

• history depends on both the assumed maximum decrease in feedwater temperature and the feedwater temperature time constant.

The lass of feedwater heating event results in a mild transient in which the fuel surface heat*flux increases to a maximum value below that corresponding to steady state operation at the scram setpoint. This increase in power to the coolant is partially offset by the benefiCial effect of the increased core inlet subcooling on the critical power ratio. However, this event can be one of the limiting.events with* respect to minimum critic;al power ratio, and is considered i*n reload ana*lyse.s. II. REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide_ a*n analysis and evaluation of the design and* performance of structures, systems, and components of the facility with the ob-

2 - je~tive of assessing the risk to public health and safety resulting from operation of the facility, including detennination of the margi.ns of safety during nonnal

• operations and transient 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 ba.rriers wtiich guard a~ainst the uncontrolled release of ra'dioactivity. The General Design Criteria (Appendix A to 10 CFR Part 50) establish minimum re-quirements for the. principal design criteria for water~cooled reactors. .GDC 10 "Reactor Design 11 requires that the core and associated cool ant, con.trol and protection systems be designed with appropriate margin to assure that specified. acceptable fuel design 1imits are not exceeded during normal operation, inGluding the effects of anticipated operati.onal occurence. GDC 15 "Reactor Coolant System Design" requires that* the reactor coolant and* associated protection systems be designed *with sufficient margin to assure that the design conditions of the reactor coolant pressure boundary are not* exceeded during normal operation, including the effects of anticipated operational occurrences. GDC 26 "Reactivity Control System Redundance and Capability" requires that the reactivity control systems be capable of reliably controlling reactivity changes to assure that under conditions of normal operation, including anticipated* op~rational occurrences, and with appropriate margin for:malfunctions:such a~ s.tuck rods, specified acceptable fuel design limits are _not exceeded.*

- 3 *- "I I I.* RELATED SAFETY TOPICS \\. Var~ous other SEP topics evaluate such items as 1the reactor protection system~*< The~ effects of single fai 1 ures on safe shutdown capabil itY ~~ co'ns'idered under . Topic VII-3. .*~ ' -~*:: -.-

• • Iv.

REVIEW.*GUIDELINES Th~ review ls ctinducted in accordance with SRP 15.1-.~, 15.1.2, 15.1~3and15~1.~~* The evaluation includes review of the analysis for the:event and identification* of the features in the plant that mitigate the consequences of the event as* well as the ability of these sYstems to function as required. The extent to whi.ch operatc>"r action is required is also evaluated. Deviations.firm the* .criteria specified in 'the Standard Review Plan are identified. V.. EVALUATION In reference l, *the 1 i censee reported on cal~ul ati ons of reactor response to a loss of feedwater heating event in which feedwater temperature decreased by 145°F. Reactor power increased. to th~ scram setpoint in about 90 seconds with a *corres-ponding decrease in critical power ratio of 0.17. This event was analyzed for an initial power of 100 percent instead of 102 percent as required by SRP Section 15.1.1. Use of the _higher initial power could result in a slightly larger . decrease in critical *power ratio.. However, for Dresden 2 the generator load. rejection without bypass and rod wi thdrawa 1 at power events result in a decrease in critical power ratio of about.0.18. Hence reevaluation of the loss of feed-water heating event with 102 percent initial power should not change the operating mfnimum critical. power ratio 1 imit. \\':: VI. CONCLUSIONS

• As part of. the SEP review for Dresden 2, we have evaluated the licensee's analysis

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of the loss of feedwater heating event. ** We conclude that the loss of feedwater heating event is bounded by generator load rejection without bypass. We,.there-.*. ~,. > ~-.' fore,,find results accep_table even though an initial powe~- of lOtn was assumed._;:*.. *.. instead of 102% as required *by the SRP acceptance criteria. * ~ - : *: -~ References 1: Yl003JOlA06, "Supplemental Reload Licensing Submittal.for Dresden Nuclear Power Station Unit 2 Reload 5", October, 1980.

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. 6RAN 2, SEP TOPIC XV-.1 EVALUATlc9*. INCREASE.IN FEEDWATER FLOW

• • ~..

I. INTRODUCTION Failure of the feedwater controller to maximum demand res:ults in an* increase* iri re-. ac.tor power and vessel* inventory. There is a gradual initial in~reas*e in power be-cause of the increased core inlet subcool in~ and the negative void coefficient of _ reacti vi.ty. The steam/feedwater flow mismatch.leads to a ~igh vessel water l.evel *.* trip of the main turbine. The turbine trip results in a pressurization transient, with attendant power transient, which is mitigated by reactor scram due to turbine stop/control valve closure and initiation of the turbine b.ypass system. The limi-ting conditions occur during the pressurization portion of* the overall event. C A feedwater controller failure at partial power gives a larger steam/feedwater.flow mismatch. However, failure at rated power can be more severe in terms of maximum reactor pressure and minimum critical power ratio. A feedwater control failure event at rated power is similar.to the turbine trip event at -rated power with tur-bine bypass op.erable. However, for the feedwater controller event:o the turbine trip signal occurs when the reactor is at above rated power. ~ence this event can be limitin.g with respect to minimum critical* power and is evaluated in reload analys~s. To meet. current criteria, surveillance of the turbine bypass system is required.. Si nee.the bypass system was ass.urned to operate*. f n the analysis of.this event:o 1 imitations to either reactor power or minimum critical powe.r ratio - would. be required in the Technical Specifications* to cover the case where the bypass system

• is found inoperable.

II. -REVIEW CRITERIA. Section 50. 34 of 10 CFR Part 50 requires that each apf)l icant for a. construction ~ermit or operating license provide an analysis and evaluation of the design and ~ ; I

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• per.fonnance of structures, systems, and components of the facility with the object-.:.

ive \\,, .-~: of assessing the risk to public heai'th 'and safety resulting from _operation.of.:. faci 1 i ty;.including. determinati o~*- of. the rria'rgins of *safety,,du~i*~*~:;.;ri~~;1*;*~;~~:>:.: *~**:*_ ~ -~ ~' h the .~. ations and transient conditions anticipated during' the" life of the. faci 1 ity *.. :~"... ~.. *:,~-:-. '*,. :. : *:.... . ~. ;.. .*. _: -::* :.7'. <.. ~(: Section 50~ 36 of 1 O CFR Pa rt *so safety: 1 imits *which *protect the requires the Technical Specifications to include*... integ.;ity::of th'e physical *b~r~i~~ ~hi~h9~~d-_;.:;.* against the uncontrolled release of radioactivity *. The General Design Criteria (Appendix A* to 10 CFR Part 50) establish.minimum re-. quirements for the principal design criteria for water"'.cooJed reactors::-* GDC 10 "Reactor Design" requires that the core and associated *coolant, control

• and protection systems.be designed with appropriate margin to assure t_hat speci-fied acceptable fuel design limits are not exceeded during nonnal operation inclu-ding the effects of anticipated operational occutences *.

-GDC 15 "Reactor. Cool ant System Design 11 re qui res that the reactor cool ant and associated protection systems be assigned with sufficient margin to assure that the design conditions of the *reactor cool ant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurrences.

..... * ---* - :-**:*.. ;~-~.;.-*".'. __....... ;. GOC 26 11 Re~ctivity.Control System Redundance and Capability" requir.es that the III.. RELATED SAFETY TOPICS. ~ --~1 Various other SEP :topics *evaluate such items.as the react.or protection *systein. --c on safe.~hutdown capability are considered

• J~e effects ofi'Single failures

~- ~-:- under Topfc VII-3. 0' -c

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REVIEW GUIDELINES The -review is conducted in accordance with SRP 15.1.1, 15,. :t. 2, 15.1. 3 and *

15. L 4.

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V.. EV"ALUATION \\. ' In Refer.ence l, the licensee evaluated the consequences of a*feedwater* contro*ller*failure.leading to an increase in feedwate~.flow to"1il0%~*,.The:*.. initial power was 100% instead of 102% as required.in SRP Section 15.1.2. However,* the r*eduction in critical power ratio was only 0.09 compared* to.a* reduction of ':',, 0.18 for load rejection without bypass. Hence, reevaluation of 'this *event for an initial*power of 102% would not result in:a reduction in the* operating minimum critical power ratio limit. VI. CONCLUSIONS As part of the SEP review of Dresden. 2, we have evaluated. the li"censee's analysis of a feedwater controller failure. We conclude that tnis event is =. bounded by load rejection without bypass. We, therefore. find' the results acceptable even though an initial.Power of 100% was* assumed*.instead of 102% as required by the SRP acceptance criteria * ...... -. *-*-*=-- _,, ___ -- --.----------* ***--~-----*- -*- *' *-* - --- ------. ---. -- -*. -- .=- __, -*-*.---.:****---:;'-:::**~*-.:*-=~------::-~-~-=---=::-;* --~---... -..-. *-*---.-* -------. To meet current.criteria, surveill a_nce of the turbine bypass system is 'required* Since the bypass system was assumed to o_perate in t_~e an~ly~i.s of tM~ _event,. 1 imitations to either reactor power or minimum critical power ratlo would: be required in the Technical Specifications to cover the c.ase where the bypass system is found inoperable. REFERENCES*

1.

• Yl003J01A06, "Supplemental Reload Licensing Submittal for Dresden Nuclear Power Statfon Unit 2 Reload 5, 11 October 1980.

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. ".~ '.":' ~. -:.-.. ~:. ,.,**.... -. -... --c-*~.{i:'~::./~-<*' 1,.. :.*... - '*'.i. DRESDEN 2, SEP TOPIC XV-1 EVALUATION.

• INCREASE IN STEAM FLOW I.

INTRODUCTION Fai 1 ure of the pressure regulator in an open position results in. full opening of the turbine admission valves and partial opening of the l°urbine bypass valves. The total steam flow rate resulting froin the *regulator failure is limited to approximately 110 p*ercent of rat~d flow by a maximum fl ow limiter. The increased

• steam fl ow rate results in a drop in reactor pressure and inventory.

The increase in core void fraction produces an initial decrease in core. power and increase in v~~sel level. The vessel level increase can cause trip of the main turbine. Re~' actor scram then results from turbine stop/control valves closure. If the turbin*e trip signal or high water level is not reached, an *MSIV closure* on low steam pressure occurs. Reactor scram then results from position switches on the MSIV's. Since the turbine trip or MSIV closure occurs at relatively low reactor pressure* and power, the pressure.regulator failure event is*not of consequence with resp~ct to peak system pressure or minimum critical power ratio *

........... -.~""--* .. -~- M_ * - - - - * - - - - - - * - 2 --- - ---~..... *---*- -*-*-*---.;_.. - II. REVIEW CRITERIA . *'....... *... --..,....... - *"'.' ~---. Section 50.34 of 10 CFR Part 50 requires that each appl 1cant for a construction_ r

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. p'erin1t or operathlg license pro~1de an analysis and e.valuat1on of 'ih~ des~gn and.. >.*.. < '. ~~-onnanc~ of -:~r~:~=~~s, ~Y~ t~ms :* ::;-~~;:~:~:~" ;:~~ ;:~-;~,t::. w;;~;~h: ~~j~c~~ ******** - \\ .............. I. ~ ... _._. / i ve of assessing the risk* to.public heal th and safety resulting from operation of.. :*; *:

• I I

r the facility, including determination Of the margins Qf safety "during* nonnal oper- ;:';'*_-*:..

• I ations and transient conditions anticipated during the life of the facility.

Section 50.36 of 10 CFR Part 50 requires the Technical. Specif~cations 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 re-quirements for the principal design criteria for water-cooled reactors

• GDC 10 "Reactor Design" requires that the core and associa.ted coolant, control and protection systems be designed with appropriate margin to assure that speci-fied acceptable fuel design 1 imfts are not exceeded during nonnal operation inclu-ding the effects of anticipated operational occurences.

GDC 15. "Reactor Coolant System Design" requires that the reactor coolant and associated protection systems be assigned with sufficient margin to assure that the design cond'itions of the reactor coolant pressure boundary are not exceeded during normal operation, including the effects of anticipated operational occurrences.

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.. :;.. ~... GDC! 26. 1tReact i vi ty Contro 1 System Redundance and Capabil i ty11 re qui res that the reactivity. co'ntrol sys~ems be capable of ~elia.bly. controlling reactivity ~hanges.:. .... ~-.... -~;... *... -....... ~.... ~:*. -~.. -: ~~*:~~- *- -*~

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~*.. *.. *.... *_-*: :-) - *; ~ :: >. *... :* :. -~ -:: ~- -.. -~. to assure tnat under conditions of;.r>orinal operation,. i'ncludi*ng anticipa~,d~*~~:~*~.~-~:~~*.*

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~*_-.*::'? --~~*-** *.;.;.*~:;..:.... :c::-:~~_;.::.-:-:~.:..;.--*+~.. -:.**~::-- ~-"-.:..*;~-:--.. :;J operatio"nal occurrences,.and wi~h appropriate-'margi.n *f'_or, mal~unctio~s *~uch.*._as...,.\\ ':

.~... ~ stuck rods, specified acceptable fuel design limits are not exceeded~ ~-* - III. ~- ... -.. :...,.. :*.,.\\ RELATED.SAFETY TOPICS*.-_.-~. '.. * --~* ..:. -*-:f Various other SEP topics *evaluate such items as the react.or pn>t~c'f!ion system. 'a- "The effects of~single failures on safe.~hutdown capability are consi°dered under Topi~ VII-3. ~; -*.~. . IV. REVIEW GUIDELINES The review is conducted in accordance w.i th SRP 15.1. 1, 15.1. 2, 15.1. 3 and 15.1.4. The evaluation includes review of the a*nalysis for the event and identification { of the features in the plant that mitigate the consequences of the event as well as the abili'ty of these syste.rns to function as required. The extent to which operat.or action is required i~ also ev~luated. Deviations from the .criteri~ specifi~d in the Standafd Review Plan are identified.

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EVALUATION \\ _*.;lf ~,;~) :

The licensee considered this event in the FSAR* b*ut did not reanalyze 'the event.for::*.

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• later fuel cycl.es.

The event is not limiting with respect t~ pe.ilk;'sys'f~~- pr~ss~~--*.~*-*.

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. or minimum critical power ratio.* . :)*;* :(:.. ;~;>\\:;*/,{' *:* ' ',.... '~. ~ VI. CONCLUSIONS As part of the SEP revi enr for Dresden 2, we have evaluated the li-censee 's treat-ment of the fa i1 ure OT a pressure regulator to the. open po5ition. We conclude,.. that the event is bounded by load rejection without bypass and: is in confonnance. with SRP Section 15.l;J.. . i . ~ :,

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