ML17194A347
| ML17194A347 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 12/15/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Delgeorge L COMMONWEALTH EDISON CO. |
| References | |
| TASK-15-05, TASK-15-09, TASK-15-5, TASK-15-9, TASK-RR 1S5-81-12-45, LS5-81-12-45, LSO5-81-12-045, LSO5-81-12-45, NUDOCS 8112210339 | |
| Download: ML17194A347 (13) | |
Text
Docket No. 50-237 6805 12-045 Mr. L. Del George Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690
Dear Mr. Del George:
December 1 5, 1 981
SUBJECT:
DRESBEN @-SEP TOPIC XV-5, LOSS OF NORMAL FEEDWATER FLOW AND XV-9, STARTUP OF AN INACTIVE LOOP OR RECIRCULATION LOOP AT AN INCORRECT TEMPERATURE AND FLOW CONTROLLER MALFUNCTION
- CAUSING AN INCREASE IN BWR CORE FLOW RATE By letter dated October 15, 1981, you submitted safety assessment reports for the above topics. The staff has reviewed these assessments and our,.
conclusions are presented in the enclosed safety evaluation repotts, which.
complete these topics for Dresden 2.
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These evaluations wil 1 be a basic input to the integrated assessment for *.
your fac111ty.
The evaluations may be revfsed in the future ff your facfl it:Y. * ~'...
design is changed or ff NRC criteria relating to these topics ar~ modiff'ed::..\\
before the integrated assessment f s completed *. :
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Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 Division of Licensing 5fp'I \\
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Enclosures:
As stated cc w/enclosures:
~ee next page 8112210339 811215 PDR ADOCK 05000237 P
POR I
, NRG FORM 318 (10-80) NRCM 0240 OFFICIAL RECORD COPY f)S:~s, 6 (, D I
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USGPO: 1981~335-960
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, Maryl and 20015 Dr. Linda w. Little 500 Hermitage Drive
- Raleigh, North Carolina 27612 DRESDEN 2
- Docket No. 50-237
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Illinois Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Springfield,~Illinois 62704 U. S. Environmental Protection Agency Federal Activities Branch Region V Office ATTN:
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.
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DRESDEN 2 SEP TOPIC XV-5:
LOSS.OF NORMAL FEEDWATER FLOW I.
INTRODUCTION Loss of fe.edwater fl ow could occur as a result of the simultaneous tripping of all feedwater pumps, or a feedwater controller failure that closes the feedwater
. control.valves.
When the feedwater flow drops, to 20% of f.ull flw, the re-circulation loop control system is designed* to reduce the speeds of the motor-generators (M-G's) to a minimum to protect the recirculation and jet pumps from cAvitatiCin.
Loss of feedwater causes the water level. in the reactor vessel to drop. -The reactor protecti.on system is designed to trip the reactor when the water level drops _to the Jew-level set point.
Due to the decay heat and the release of steam, the water*level will continue to drop until the low-low water level set poipt is reached at which time the protection system is designed to close the main steam.isolation valves, (MSIV's), trip the recirculation pumps, and start High Pressure CoQllant Injection, (HPCI}.
An isolation condenser is designed to take away the decay heat after the MSIV's close.
Should the HPCI
. system fail, the isolation condenser is designed to independently maintain the
~ater level above the top of the active fuel.
The C~mmonwealth Edison Company (CECO) submitted two analyses of the loss of feedwater event *at Dresden 2, on November, 1968'. (Reference 1) and another in August, 1979 (Reference 2). This event was also assessed in a generic study for a topical report (Reference 3) which was published in May, 1977.
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II.
REvIEW CRITERIA
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Section *so.34 of 10 CFR Part SO requires *that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objectf"ve *of assessing the risk to public health and safety i:-esulting from operation of the facility, including determination of th~ margins of safety during normal operations and transient conditions anticipated during the life of the facility.
Section 50.36 of 10 CFR Part ~O requires the Technical Specifications to include safety limits which* protect the integrity of the physical barriers.
wnich guard against the uncontrolled release of radioactivity.
The General Design Crtteria (Appendix A to 10 CFR Part 50) establish minimum requirements for the principal design criteria for water-cooled reactors.
GDC 10 "Reactor Design". requires that*the core and associated coolant, control and protection systems be designed:wfth appropriate mar.gin to assure that specf ff ed acceptable fuel d~.sign limits* are not exceeded during normal *operation, including the effects of anticipated operational occurrence.
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GOC 15 uReactor Coolant System Design" requires_that the reactor coolant arad associated protection *systems be designed with sufficient_ piarg_~~ to _a~sµry..: that the design conditf o_ns of th_e reactor coolant pressure boundary are not exceeded during normal operation, including-the effects of anticipated operational occurrences *.
GDC 26 "Reactivity Contro1*system Redundance and Capability" requires that the reactivity control systems be capable of reliably controlling reacfivfty changes to assure t~at under conditions of normal operation, including anticipated operational occurrences, and with appropriate margin for malfunctions such as stuck rods, specified acceptable fuel design limits are not exceeded.
III.* RELATED SAFE1Y TOPICS*
Various other SE~ topics evaluate such items as the reactor protection :system.
The effects of single.failures on safe shutdown capability.are considered under Topic VII-3.
IV.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.2.6.
- I The evaluation includes review of the analysis for the event and identification o~ the features in the plant that mitigate. the consequences of the ev4!nt as well as the ability of these systems to function as required. The extent to which operator actio" is required is al~o evaluated. Deviations from the
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criteria specified in the Standard Review Plan are_ i.dentified.
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V.
EVALUATION I*
Since it is assumed that the reactor trips on a low-water-level signal and that the power then rapidly decreases, the critica.l power ratio does not decrease be-low its initial, steady-state value.
The main concern in this event is the loss of too much water so that the fuel elements are uncovered and inadequately cooled.
From the analyses presented in Reference 2, CECO concluded that:
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Dresden 2 is adequately equipped to mitigate the consequence of the loss of feedwater event, as it relates to core cooling, with-ou~ operator assistance and with or without a stuck-open relief valve.
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Operator actions are required only when there is a complete loss of high pressure injection (HPCI) and other water inventory sys-tems.
In *such an event, timely manual depressurization followed by injection from low--pressure systems will mitigate the conse-quentes.
In its evaluation (Reference 4) of this report (Reference 2) the NRC stated that the basic requirements for this event, which are given in General Design Criteri~ 10 and 15, appear to have been met, even' in the light of the TMI-2 experience.
In a generic review of incidents of moderate f:requency, which includes the loss of feedwater flow event, Reference 3 concludes th~t a turbine trip is more limit- -
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- in Reference 5 agreed with this conclusion..
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CONCLUSION
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As part of the SEP-review of Dresden 2, the-analysis for loss of feedwater has been evaluated and we have concluded that the basic requirements of GDC 10 and 15 are met.. The-MCPR consequences of this event are bounded by those of a turbine trip. This aspect is reviewed under SEP Topic XV-3.
REFERENCES
- 1.
Dresden Nuclear Power Station, Units 2 & 3, Safety Analysis Report; Commonwealth Edison Company, Chicago, Illinois; Chapter II; Novemqer, 1968.
- 2.
General Electric Company; Additional Information Required for NRC Staff Generic Report on Boiling Water Reactors; Revision l; NED0-24708A; August, 1979.
- 3.
General Electric Company; Licensing Topical Report General Electric Boiling Water Reactor, Generic Reload Fuel Application; NEDE~24011-P-A; May, 1977; pps 5-11, 5-68, 5-69.
4.. U. S. Nuclear Regulatory Commission; Generic Evaluation of Feedwater Transients and Small Break L6ss-o~-too1ant Accidents in GE-Desi~ned Generati~g Plants and Near-Term Operating License Applica.tions; NUREG-0626; January, 1980; page G-3.
- 5.
U. S. Nuclear Regul~tory Commission; Safety Evaluation for the General Electric Topical Report, Generic Reload Fuel Application, (NEDE-24011-P); April, 1978; page C-62.
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DRESDEN UNIT 2
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TOPIC XV-9
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Startup of an Inactive Loop or Recirculation Loop at an In.correct Temperature and Flow Controller Malfunction Causing an Increase in BWR Core Flow Rate-INTRODUCTION
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The objective of this review is to assure that the consequences of core flow increase. transients are acceptable, i.e., that the increase in core flow or the introduction of cooler water into the core does not lead to an unacceptable loss of fuel clad integrity or overpressurization of the primary system.
The imprope*
startup of an idle recirculation loop would cause a power increase due to a com-*
bination of a negative moderator coefficient reactivity addition and increas-ing recirculation flow.
A flow controller malfunction can cause the scoop tube position to move at its maximum speed in the direction of increasing pump flow.
- This increas1ng pump flow will sweep. voids at a* faster rate and therefore cause an increase* in neutron flux. The review consi~ts of evaluating the licensee's analysis of the ~equence of events, the analytical model, the values of para-meters used in the analytical model and the predicted consequences of the transier II.
REVIEW OF CRITERIA Section 50.34 of 10 CFR Part 50 requires that ea~h applicant for a construction permit or operating license provide an analysis and evaluation of the design and performan~e of structures, systems, and com~nents of the facility with the objective of assessing the risk to public health and safety resulting from opera-tion* of the facility, incl tiding determination of the margins of. safety during
- normal operations and transient conditions anticipated during the life of the facility. Section 50.36 of 10 CFR Part 50 requires that Technical Specifications include safety limits which protect the integrityof the physical barriers which guard against the uncontrolled release of radioactivity.
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The General Design Criteria {Appendix A to *ro CFR Part 50) set forth c"riteria
- l for.* the design of water-coo 1 ed reactors. *
- GDC 10 "Re~ctor Design" requires that_ the core and associated cooling, control and protection system be designed with appropriate margin to assure that speci-fied acceptable fuel design limits are not exceeded during normal operation, including the effects of anticipated operational occurrences.
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 ~ormal operation, including th~ effects of anticipated operational occur-rences.
GDC 20 "Protection System Functions" requires that the protection system be de-signed to initiate* Jutomatically the operation of reactivity control ~ystems to assure that specified acceptable fuel design *1 imits are not exceeded as a result of anticipated o~~rational occurrences.
GDC 26 "Reactivity Control System Redundance and Capability" requires that the reactivity control system be capable of reliably controlling reactivity changes to assure that und~r conditions of normal operation, including anticipated opera-tibnal occurrences, and with appropriate margin for malfunctions s~ch as stuck rods, specified acceptable fuel design limits are not'exceeded.
GDC 28 "Reactivity Limits" requires that the reactivity control systems be de-signed with appropriate limits on the potential amount and tate of reactivity increase to ensure that the effects of postulated reactivity accidents can neither (1) result in damage to the'reactor coolant pressure boundary greater than limited local yielding nor (2) sufficiently disturb the core, its support structures or other reactor pressure*~essel internals to impair significantly the capability to cool the core.
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.III.
RELATED SAFETY TOPICS Various other SEP topics evaluate such items as the reactor protection system.
The effects of single failures_on safe shutdown capab.iJity are considered under Topic VII-3.
VI.
REVIEW GUIDELINES The review is conducted in accordance with SRP 15.4.4 and 15.4.5.
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 which operator action is required is also evaluated.
Deviations from the criteria specified in the* Standard Review Plan 'are i'.dentified.
V.
EVALUATION Both o.f the subject transients were assessed in the General Electric generic reload ropical (Reference 1) and it was determined that the inadvertent flow increase transient was the most limiting of the transients at reduced flow.
For the FSAR analysis of the inactive loop startup, *the inactive loop is assumed to be filled with l10°F water.
The initial power level was 30% of full power.
. The analyses were performed at the end of equilibrium fuel cycle exposure condi-tion in which the scram and void character'istics would be worst.
The critical power ratio (MCPR) remains above the safety limit of 1~06 for both events.. The general analytical methods have been.approved by the staff (Reference 2).
For the flow controller malfunction transient, the assumed initial conditions are 65% power and 50% flow.
The failed motor ~enerator set speed,controller causes the scoop tube position to move at a rate of 10%/second for nine seconds.
The use of generic Kf factors to.adjust the MCPR operating limits will ensure
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that the operating limit MCPR will not be violated for this event. Appropriate Kf factors have been reviewed and approved in Referenc.e 2. Additional dis-cussion of these events was provided in Reference 3 in which appropriate opera-tor attions were identified~
VI.
CONCLUSIONS As part of the SEP review of Dresden 2, the startup of an inactive loop and flow controller malfunction transient were reviewed against the acceptance criteria of SRP Section 15.4.4 and 15.4.5. The initial conditions and analy-tical methods have been reviewed and found to conform to the requirements of the SRP.
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REFERENCES
- 1.
NEDE-24011-P "Generic Reload Fuel Application, with Revisions," May, 1977-March, 1978.
- 2.
Staff Evaluation GE Generic Reload Application, May 12, 1978.
- 3.
NED0-24078-A Additional Information Required for NRC Staf~ Generic Report on Boiling Water Reactors," Volume K, August, 1979 Revision 1-December, 1980.