ML17193A038
| ML17193A038 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/12/1980 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Peoples D COMMONWEALTH EDISON CO. |
| References | |
| TASK-06-10.B, TASK-6-10.B, TASK-RR NUDOCS 8009180312 | |
| Download: ML17193A038 (19) | |
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-- Mr. o. *Louis Peoples.
01 rector of Nuclear Liceris*in*g..
. : Commonwealth.Edison Cofii)any_
- Post Office* Box 767_
- SEP--PF -* :
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Chicago, -I 111 no~s : 60690._
Dear Mr~ Peoples*:
RE:
SE'.P TOPIC VI-10.B,--*"SHAAED :ENGINEER.ED. SAFETY FEATURES" (DRESDEN_2)
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En~losed is a *copy' of our evaluation _of Systematic Evaluation Program -
Top;c VI-10.B, "Shared Engineered Safety Features 0 (Dresden Unit No. 2) *.
Th1s assessment corrpares *your ~ac.ilfty ;-_ a~:. des.cr1 bed 1 n Docket No. 50-237,
. with the cr1teri_a cu*rrently used *by the regulatory staff for Hcensing new
_ facil_it'1es.-* Please -.1 nform us if your.* as-built facility differs from the
-licensing _bas1~ assu_i:ned.in our assessinent within 30 days of*rec~1pt of this letter *.. *
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~-
- This evaluation wnl be*a ba-s1C input tp.the *'integrat~d*:~af~~Y-assessment for your fad11ty _unless*you ~dent;fy changes*- needed.to reflect th~ as-built conditions.at your facility~. T.h,.1s top1c assessment may be revised in the ~
_*future if your faci l1ty design_ ts changed or 1f NRC criteria relating to this topic are modified before: th<rint'egrated assessment is cof!!Pleted.
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,. * ~- *Sincerely, *
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' * **~1s M; Crutchfield, *Chief.
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Enclosure:
perat i ng Reactor5. Branch 15.
tv1_s1on of Operatfog Reactors Colli> leted SEP*.
. Topic. -
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NRC FORM 318 (9-76) NRCM 0240
_. -. {r.u.s. GOvERNMENT:PR1N'r1NG o~FICE: 19i9*289*3G_9
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 August 12, 1980 Docket No. 50-237 Mr. D. Louis Peoples Director of Nuclear Licensing Commonwealth Edison Coll'l)any Post Office Box 767 Chicago, Illinois 60690
Dear Mr. Peoples:
RE:
SEP TOPIC VI-10.B, *sHARED ENGINEERED SAFETY FEATURESN (DRESDEN 2)
Enclosed is a copy of our evaluation of Systematic Evaluation Program Topic VI-10.B, "Shared Engineered Safety Features" (Dresden Unit No. 2).
This assessment CO!ll>ares your facility, as described in Docket No. 50-237, with the criteria currently used by the regulatory staff for licensing new facilities. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment within 30 days of receipt of this letter.
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 topic 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 coll'l)leted.
Enclosure:
Completed SEP Topic VI-10.B cc w/enclosure:
See next page.
Sincerely, td {)_~A.fav-t~--
~
Dennis M. Crutchfield, Chief
{'- Operating Reactors Branch 15 Division of Operating Reactors
Mr. D. Louis Peoples cc w/enclosure:
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 Susan N. Sekuler 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 August 12, 1980 Department of Public Health ATTN:
Chief, Division of Nuclear Safety 535 West Jefferson Springfield, Illinois 62761 Director, Technical Assessment Divis ion Office of Radiation Programs (AW-459)
- u. s. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460
- u. s. Environmental Protection Agency Federal Activities. Branch Region V Office ATTN:
EIS COORDINATOR 230 South Dearborn Street Chicago, Illinois 60604
- Resident Inspector Dresden Nuclear Power Station c/o U. S. NRC Rural Route #1 Morris, Illinois 60450
SEP TECHNICAL EVALUATION REPORT SEP TOPIC VI~lO.B SHARED ENGINEERING SAFETY FEATURES DRESDEN NUCLEAR STATION UNIT 2
CONTENTS
1.0 INTRODUCTION
2.0 CRITERIA
- 3.0 DISCUSSION AND EVALUATION 3.1 3.2 3.3 3.4 Discussion of Shared Systems Key Systems for Mitigating Accidents and/or Maintenance of Safe Shutdown Capability to Supply Required Loads 3.3.1 3.3.2 3.3.3 3.3.4 Large Break LOCA * *
- Sma 11 Break LOCA * *
- Safe Shutdown and Cooldown
- Evaluation *
- Sharing of DC Systems 3.4.1 Evaluation **
3.5 Stored Energy at the Site 3.5.l Evaluation **
3.6 Capacity of Onsite AC Power Supplies 3.6.1 Evaluation **
3.7 Capacity of Onsite DC Power Supplies
- 3.7.1 Evaiuation **
3.8 Coordination Between Unit Operators 3.8.l Evaluation * * * *
. 3.9 Independence of Onsite DC Supplies 4.0
SUMMARY
5.0 REFERENCES
ii for 1
2 3
3 4
4 5
6 7
9 10 10 10 11 11 11 11 12 12 13 13 13 14
SEP TECHNICAL EVALUATION REPORT ELECTRIC.U., INSTRUMENTATION, Ai~D CONTROL PORTIONS OF SHARED SYSTEMS DRESDEN NUCLEAR STATION UNIT 2
1.0 INTRODUCTION
The purpose of this review is to determine if the Electrical, Instrumentation, and Control (EI&C) portions of shared systems for Dresden Nuclear Station Unit 2 are in compliance with current licensing criteria as outlined by SEP Topic VI-10.B~ Specifically, this review is to determine whetaer or not the interconnection of Engineered Safety Features (ESF), onsite emergency power, and service systems between units are not such that failure, maintenance, or testing operation in one unit will effect the accomplishment of the protective function of the systems in other units.
Additionally, this review will evaluate whether or not the required coordination between unit operators can cope with an incident in one unit and safe.shutdown of the remaining units, and whether or not system overload conditions will arise as a consequence of an accident. in one unit coincident with a spurious acci-dent signal or a:iy other single failure.
I Specifically excluded from this review is evaluation of EI&C por-tions of the fire protection system covered by SEP Topic IV-6.
Although not specifically mentioned in the SEP Topic VI-10.B outline, a review of some of the EI&C portions of the "Systems Required for Safe Shut-down" (SEP Topic VII-3) is required~ Since no such review has yet been completed with respect to EI&C system performance, this review must, of necessity, evaluate some of the EI&C functions of those systems identi-fied by the safe shutdown review.
This review only deals with the affect of EI&C shared systems on Unit 2, although Unit 3 systems are, for the ruost part, identical to those for Unit 2.
However, detailed evaluation of the capabilities of the Unit 3 safety systems is not included in this evaluation as Unit 3 is not an SEP plant.
1
2;0 CRITERIA The criteria for shared EI&C systems between nuclear units of a station are contained in Regulatory Guide (RG) 1.81, "Shared Emergency and Shutdown Electric Systems for Multi-Unit Nuclear Power Plants;"
Branch Technical Position (BTP) EICSB.-7, "Shared Emergency Electric Power Systems for Multi-Unit Generating Stations;" and IEEE Stan-dard 308-1974, "Criteria for Class lE Power Systems for Nuclear Power Generating Stations."
Sections of these criteria used primarily in this review are listed below:
(1)
Assure that a single failure, including false or spurious accident signal, does not reduce the capa-bility to supply, automatically, mini.mum ESF loads in any unit and safely shut down the remaining units assuming a loss of offsite power (BTP EICSB~7 and RG 1. 81 ) *
(2)
Provide onsite power capacity sufficient to ener-gize siesmic category I equipment to attain a safe and orderly shutdown of all units assuming a single failure and a loss of offsite power (BTP EICSB-7).
(3)
Limit the ii;iteraction between unit ESF electrical circuits such that any allowable coobination of maintenance and test operations in the units will not affect the capability to supply power auto-matically to minimum ESF loads in any unit (BTP EICSB-7).
(4)
Minimize the coordination required between unit operators in order to accomplish (1), (2), and (3) above.
Although each design will be evaluated on an individual basis in this regard, all shared onsite power systems should meet the following (BTP EICSB-7):
(a)
Coordination between unit operators should not be necessary in order to provide for (1) and (2) above.
(b)
Complete information regarding the status of shared systems should be provided for each operator.
2
(5)
DC sys~e::i.s in multi-unit nuclear stations should not be sha=ed (RG 1.81).
(6)
Stored energy at the site shall have the capability to ope=ate the standby power supply while supplying post-a=cident power requirements to a unit for the longer of the following (IEEE Standard 308):
(a)
Seven days.
(b)
Time required to replenish the energy from sources away from the generating unit's site following the limiting design basis event.
(6)
Each battery supply will be capable of starting and operating all required loads, and each battery supply shall be independent of the other battery supplies (IEEE Standard 308).
(7)
Each battery charger shall have sufficient capacity to restore the battery from the design minimum charge to its fully~charged.state while* supplying normal and post-accident steady state loads.
Each battery_charger shall be independent of other bat-tery chargers (IEEE Standard 308).
3.0 DISCUSSIO~i.:\\i'iD EVALUATION 3.1 Discussion of Shaied Systems Both the AC and DC power systems and their distribution systems are shared between Dresden Units 2 and 3.
The AC systems share the services of diesel generator 2/3 (DG2/3) which can supply either unit's 4160 V AC system.
The AC systems are also connectable via tie breakers ihat link the switchgear supplied by ~iesel generator 2 (DG2) with that supplied by diesel generator 3 (DG3).
The 125 V and 250 V DC systems are shared such that each of the two batteries in each system provides power to loads in both units.
Means exist in both systems to connect the load groups :for either unit to either or both battery and charger trains.
The main computer can be powered from the essential bus of either unit by ::::a:is of installing a copper link.
With the ex=eption of control power for DG2/3, there are no ESF electrical circuits shared by the two units.
Tne DG2/3 control circuit 3
receives input fro:i the ESF circuits of both plants s.uch that a loss of coolant accident (LOCA) in one plant concurrent with a loss of offsite power would cause DG2/3 to provide power to the plant experiencing the LOCA.
Otherwise, DG2/3 serves whichever plant is either without off*
site power or has an accident condition.
This generator will only supply power to one unit at a time.
A normal/bypass switch for each unit allows/prevents DG2/3 from automatically supplying power to that unit.
The switch is a keylock switch and is normally placed in "nor-mal" when the unit is operating.
However, there are no LCO require-ments so specifying.
Several other systems related to accident mitigation and safe shutdown are shared between Units 2 and 3, including the SWS, the Reactor Building Closed Cooling Water System (RBCCWS), the Standby Gas Treatment system, and the Radioactive Waste Control System (RWCS).
3.2 Key Systems for Mitigating Accidents and/or for Maintenance of Safe Shutdown In determining the effect of failures in the EI&C systems shared between units, the effect on those systems necessary for mitigating accidents and those necessary for attainment and maintenance of safe shutdown must be considered.
The systems of particular importance are the Core Spray system, High Pressure Coolant Injection (HPCI) system, Low Pressure Coolant Injection (LPCI) system, Auto-Depressurization System. (ADS), Shutdown Cooling System (SDCS), RBCCWS, SWS, and Isola-tion Condenser System (ICS).
3.3 Capability to Supply Required Loads BTP EICSB-7 and RG 1.81 require that the capability to supply, automatically, minimum required ESF loads in one unit and safe shutdoWn of the other units should not be reduced assuming a single failure coincident with a loss.of offsite power.
The cases to be evaluated, therefore, involve large break LOCA, small break LOCA (i.e., one in which Core Spray/LPCI systems alo~e are not sufficient to mitigate the 4
e e
a"cc ident), and safe shutdown.
Each case will be evaluated based on the required perfor::ance indicated in the Final Safety Analysis Report (FSAR), Section 6.
The availability of vital indications of reactor parameters (level, temperature, pressure) and key systems.status is being reviewed under SEP Topic VII-3, "Systems Required for Safe Shutdown."
The instruments which provide control room indication are totally independent of those which initiate the Reactor Protection System (RPS) and Engineered Safety Feature (ESF) system, and are not covered by this report.
The ability of the RPS and ESF system to initiate has been considered in this review.
3.3.1 Large Break LOCA.
Section 6 of the FSAR identifi'es the Core Spray and LPCI systems as the required ESF systems necessary to prevent core damage from a large line break LOCA.
The Core Spray system consists of two independent loops, either one of which is cap-able of providing 100% of the needed flow to the core to protect it in this scenario.
The LPCI system is also capable of providing the needed cooling water capacity to the core provided three of four pumps are operating.
Cooling water to the LPCI system is provided by the Con-tainment Cooling Serv'ice Water System (CCSW).
There are no postulated single failures of the EI&C features or these systems such that the min.imum required emergency core cooling would be prevented, provided the DG2/3 normal/bypass switch in Unit 2 is in "normal".
Specifically, there. are no single failures which could result in failure of both Core Spray subsystems (only one is required).
There are sever a 1 single failures which result in the inability of the LPCI system to provide the required flow to mitigate this accident.
Loss of either DG2 or DG2/3 (or any loss of power to buses 23-1 or 24-1) results in only two of four LPCI system pumps available (three of four are needed).
Furthermore, loss of Unit 2 125 V DC Reserve or Main Bus power (control power for the pump circuits) results also in only two of four pumps being available.
Any failure on the LPCI system swing. bus (buses 28-7 or 29-7) wo~ld result in inoperability of at 5
least half, 9 possibly all, of the LPCI sys. cooling flow due to loss of power to the LPCI system discharge valves, which are normally closed.
However, since there are no single failures resulting in loss of more than one LPCI system pump and both Core Spray system loops, the capability to provide power to minimum required ESF loads, assuming a large break LOCA coincident with a loss of offsite power, is in compli-ance with BTP EICSB-7 and RG 1.81.
However, if the Unit 2 DG2/3 normal/
bypass switch is in "bypass," then single failures resulting in the loss of DG2 cause a loss of all 4160 V and 480 V AC power in Unit 2 upon loss of offsite power.
This results in the inability of either Core Spray or LPCI systems to mitigate a LOCA, and is not in compliance with current licensing' criteria.
3.3.2 Small Break LOCA.
The small line break LOCA is defined as one in which sufficient reactor pressure reduction to enable the Core Spray or LPCI systems to perform the required cooling functions does not occur pr_ior to automatic initiation of these systems.
Two systems exist to_ provide the necessary pressure reduction to allow either Core Spray or LPCI systems to provide the required cooling.
These systems are the HPCI system and ADS.
Operation of either system is sufficient to provide the required depressurization.
The HPCI system 1uses a turbine driven by steam from the reactor to operate two pumps which provide water from either the suppression pool or the condensate storage tank.
This system allows for rapid depres-surization as well in injection 6f cooling water prior to the avail-ability of Core Spray or LPCI system cooling.
The ADS consists of five relief valves (four electromatic and one electro-pnuematic) which rapidly reduce system pressure by relieving steam directly to the suppression pool.
No make-up cooling water is provided by this system.
The HPCI system is capable of being rendered inoperabale by single failures of EI&C-related equipment.
Failure of the Unit 2 250 V DC Reactor Bus (or the 125 V DC Reserve Bus) will result in a lack of 6
motive power (or control power) for the motor-operated valves (MOVs) required to auto=atically initiate the steam flow to the HPCI system turbine.
The ADS is not susceptible to single failures of Er&c components which would result in system failure.
However, the power supply to both control and motive power to the five relief valves involves an automatic relay-operated switching between Unit 2 125 V DC Reserve and Mai~ Buses.
This method of supplying power is not in compliance with requirements for supplying power to load groups from redundant power supplies, and is being reviewed under SEP Topic VI-7.C.l, "Independence of Redundant Ons i te Pm.*er Systems."
Any changes made to the ADS as a result of that topic would necessitate reevaluation of the system per-formance under this topic.
There are no,single failures of EI&C components which could result in failure of both the HPCI system and ADS *. Combined with the single failure effects on the Core Spray and LPCI systems discussed previously, there are no single failures of EI&C features which would result in an inabiity to automatically provide the minimum required ESF necessary to mitigate this accident as required by BTP EICSB-7 and RG 1.81.
As before, placing the u'nit 2 DG2/3 normal/bypass switch in "bypass" ren-ders the Unit 2 ESF capability ineffectual following a single failure involving DG2.
3.3.3 Safe Shutdown and Cooldown.
In the loss of offsite power scenario, there are three systems available for initial cooling and depressurization.
These are the res, ADS, and HPCI system.
The single failure effects on ADS and HPCI system were previously discussed.
The res serves to cool and depressurize the reactor by allowing steam to flow through tubes immersed in water where it gives up heat.
The cooled steam/water then returns to the reactor by natural circula-tion.
Thus, the systen serves to both depressurize and provide cooling water to the reactor without loss of inventory.
7
\\
The ICS is susceptable to single EI&C failures which would render the system inope=a~le. Failure of 250 V DC Reactor Bus power or fail-ure of the motor ope=ator for valve 2-1303-3 results in inability of the system to autouatically initiate.
Table 6.2.1 of Section 6 of the FSAR states that the design pro-visions for coping with a loss of normal auxiliary power are both the ADS and ICS or the HPCI system.
There is a single failure (loss of 250 V DC Reactor Bus power) which results in disabling both the ICS and*
HPCI system.
Therefore, a single failure potential exists which would.
result in the inability to automatically supply the required loads necessary for a*safe shutdown as required by BTP EICSB-7 and RG 1.81.
Long-term cooling for achieving and maintaining cold shutdown following a loss of auxiliary power (with or without an accident) is provided by *either the LPCI system or SOCS.
The EI&C failures associ-ated with the LPCI system have been previously discussed.
The SOCS consists of three pumps (all three needed) and heat exchangers which. take a suction on the recirculation. system and return water to the core via the recircuiation system.
Cooling for the SOCS heat exchangers is prpvided by the RBCCWS which consists of three pumps.
(two needed) and heat exchangers.
One of the three pumps and heat exchanger loops is shared between both units.
Cooling for the RBCCWS is provided by the SWS, which consists of five pumps (two for each unit and one shared) supplying a common header.
Two pumps per unit are required for cooling.
There are multiple EI&C single failures which render the SOCS inoperable or greatly reduced in capability.
Loss of OG2 reduces cooling capability to one-twelfth normal due to loss of one of three needed SOC pumps, two of three (two needed) RBCCWS pumps, and two of
- three (two needed) SWS pumps.
Loss of OG2/3 results in a loss* of the 120 V AC instrument bus which provides control power to all of the SOCS's normally-closed isolation valves, rendering the system inoper-able.
Furthermore, even if the valves are manually opened, cooling 8
~apacity 1s luced to one-third normal due t~he loss of two of three needed SDC pumps.
Additionally, loss of either the 250 V DC Reactor Bus or bus 28-1 results in system inoperability due to lack of motive po~er for various system isolation valves.
Another single failure of EI&C equipment exists which has signifi-cant effects on the ability to cope with the loss of auxiliary power accident.
Assuming, as required, that the loss of offsite power occurs coincident with a LOCA in Unit 3, DG2/3 will start and supply power to Unit 3 at bus 33-1.
A single failure of the 125 V DC Reserve Bus power or any failure in the DG2 control power (or the generator itself) results in a total loss of 4160 V and 480 V AC power in Unit 2.
While the ICS, ADS, and HPCI system should initiate (operated by AC relays which deenergize to initiate or DC relays not effected), the longer-term cooling of the Core Spray system, SOCS, and LPCI system would be unavai-lable.
If the loss of DG2 was caused by a loss of 125 V DC Reserve Bus power, then the ability to supply power to bus 24-1 from DG3 would also be lost since there would be no control power for closing the 34-1 to 24-1 bus tie breaker (2434).
Since DG2/3 would only be supplying Unit 3, the result would be an extended loss of AC power at Unit 2.
A similar scenario can be postulated for Unit 3 with an accident at Unit 2 coincident with a loss of offsite power.
3.3.4 Evaluation.
Dresden Nuclear Station Unit 2 is not in compliance with current licensing requirements with respect to the capability to supply, ~utomatically, minumum required ESF loads in one unit and safely shut down the other units coincident with a loss of offsite power as stipulated by BTP EICSB-7 and RG 1.81. Specifically, there are postulated single failures in the shared DC systems and/or AC systems resulting in the inability to provide the required safe shut-down loads.
Additionally, the Unit 2 DG2/3 normal/bypass switch has no interlocks or LCO requirements preventing positioning in "bypass" during Unit 2 operation.
In this case, single failure involving DG2 during a loss of offsite power and a LOCA in Unit 2 would result in the inability to supply the required ESF loads to prevent core damage.
9
3.4 Shari~g of DC Systems BTP EICSB-7 states that DC systems in multi-unit nuclear stations should not be shared.
Dresden Units 2 and 3 snare both a 250 V and 125 V DC distribution system.
Each system consists of two batteries, their chargers, and their distribution centers.
The independence of the systems is being reviewed under SEP Topic VI-7.C.l.
The system is designed such that operations or failures in one unit's distribution system can seriously affect the other unit's ability to respond to the postulated conditions expressed in Section 3.3 of this review.
Further-more, there are multiple means of interconnecting redundant trains in each system which are not addressed by the plant's Technical Specifica-tions.
There are no LCO restrictions prohibiting compromising the independence of these systems.
The capability exists to do so with no obvious indication available to the operator.
The simultaneous loss of both 125 V DC Main and Reserve Buses for Unit 2 unaer the conditions described in Section 3.3.1 (Large Break LOCA) would result in neither DG2 nor DG2/3 being able to supply power to Unit 2.
Therefore, none of the required ESF systems would be available.
3.4.1 Evaluation.
The sharing of 125 V and 250 V DC systems between Dresden Units 2 and 3 is not in compliance with current licen-sing criteria~
3.5 Stored Energy at the Site IEEE Standard 308-1974 required by BTP EICSB-7 calls for suffici-ent stored energy at the site to operate standby power supplies while supplying post-accident power for seven days or until fuel supplies can be replenished from offsite sources, whichever is longer.
Dresden 2 Technical Specifications, Section 3.9.C requires 10,-000 gallons of fuel for each DG.
This is adequate for: only two days at full load (four days at half load).
The fuel replenishment tioe is stated to be eight hours.
10
3.5.l Evaluation.
Dresden Unit 2 is not in compliance with current licensing requirements with regard to onsite storage of fuel for the standby power sources.
3.6 Capacity of Onsite AC Power Supplies Assuming a single failure and a loss of offsite power, BTP EICSB-7 requires that sufficient capacity be available to energize seismic category I equipment to attain a safe and orderly shutdown of all units.
The worst case for DG loading occurs during a LOCA when one DG fails to start.
Each DG is rated at 3125 kVA (2500 kW at 0.8 pf.).
The maximum loading would occur approximately two hours after accident initiation and would be approximately 2770 kVA.
3.6.1 Evaluation.
The capacity of onsite AC power systems is in compliance with the 'requirement of BTP EICSB-7 and RG 1.81.
3.7 Capacity of Onsite DC Power Supplies IEEE Standard 308-1974 requires that each battery supply be cap-able of starting and operating all required loads. It further requires that the battery charger supply have sufficient capacity to restore the battery from the design minim~m charge to its fully-charged state while supplying normal and post-accident steady state loads.
Each 250 V battery has a design capacity of 913 amp-hours.
Using the loads listed in Table 8.2.3 of the FSAR, the battery is required to provide power for loads of 162 hp (120 kW) continuously and 58 hp (42 kW) for one minute.
Thus, the total kW requirement is 121 kW (486 amps at 250 V).
The battery has capacity to carry this load for one hour and 52 minutes.
The battery chargers for the 250 V DC system have a maximum output of 90 amps.
If both chargers are used to charge a battery under the design load of Table 8.2~3, they would not be cap-able of recharging the battery, since the loads would be drawing cur-rent in excess of their output.
11
The 125 V battery has a design capacity of 498 amp-hours.
From*
Table 8.2.3 of the FSAR, the battery is required to provide power for long-term loads of 99 amps and short-term loads of various magnitude and duration constituting a drain of 18 amp-hours from the battery.
The battery could support this load for four hours and 50 minutes.
The battery chargers for the 125 V DC system have a maximum output of 34 amps.
If both chargers are used to charge a battery under the design load of Table 8.2.4, they would not be capable of recharging the battery, since the loads would be drawing current in excess of their output.
3.7.1. Evaluation.
The capacity of the batteries for the Dresden 2 125 V and 250 V DC systems meets the requirements of IEEE Standard 308-1974 stipulated by. BTP EICSB-7.
The capacity of the bat-tery chargers does not meet the requirements of IEEE Standard 308-1974
'in that they are µot capable of restoring the batteries to the fully-charge~ state while supplying normal and post-accident steady state loads.
3.8 Coordination Between Unit Operators BTP EICSB-7 requires that, in providing the capability to supply required loads and the capacity to supply those loads discussed in Sections 3.3, 3.4, 3.6 1 and 3.7 of this review, coordination between unit operators be minimized.
Specifically, no coordination between operators should be required to perform the functions discussed in Section 3.3 of this review.
Additionally, complete information of the status of the shared systems should be provided to each operator.
The control panel for DG2/3 is located in the Unit *2 section of the common control room which houses the control stations for Units 1, 2, and 3.
The control panel contains control as well as monitoring of vital parameters such as load, current, etc.
An operator licensed on both Units 2 and 3 is assigned to this panel.
He is responsible to the 12
shift foreman of whichever plant the DG2/3 is supporting.
Though loca-ted with the Unit 2 control station, the indication of status is avail-able to the Unit 3 foreman, if needed.
Additionally, each plant has a normal/bypass *function switch and a set of indicator lights showing whether or not the DG2/3 is supplying power to that unit's switchgear.
Each unit monitors the battery, chargers, and buses used by that unit.
The status of the battery, chargers, and buses of the other unit is not indicated.
Although the status of the Unit 3 battery/chargers which supply the Unit 2 125 V DC Reserve Bus is not displayed in Unit 2, voltage on that bus is monitored so that Un~t 2 operators can monitor the availability of power to the bus.
3.8.1 Evaluation.
No coordination between unit operators is required to automatically provide. sufficient power to the required
Complete information is available to operators in both units of the status of DG2/3.
Com-plete information is not available to operators in each unit of all of the shared DC syst~m components.
However, complete indication of those items supplied by the shared batteries/chargers is available to the effected operator.
3.9 Independence of Onsite DC Supplies IEEE Standard 308-1974 requires that battery supplies be indepen-dent of other bat~ery supplies and that battery chargers be independent of other battery chargers.
This evaluation is being conducted under SE~ Topic VI-7.C.l.
4.0
SUMMARY
Dresden Nuclear Station Unit 2 is not in compliance with current licensing requirements with regard to some of the EI&C features of shared systems.
They include:
(1)
Single failures (as des~ribed in Section 3.3) could result in the inability to provide power to the 13
required safe shutdown loads upon a loss of offsite power coincident with an accident in Unit 3.
(2)
There are no physical or electrical interlocks or LCO preventing parallel operation of the shared 125 V anj 250 V DC battery systems.
Such opera-
- tion, co~bined with a single failure, would result in a loss of capability to supply accident or safe shutdown loads following a loss of offsite power.
(3)
There are no LCO requirements or interlocks preven-ting the normal/bypass switches for the DG2/3 from being in "bypass" during operation of either unit.
Such operation, combined with a single failure, could render.the required accident and safe shut-down loads inoperable following a loss of offsite power.1 (4)
Complete information of the status of the shared DC batteries, chargers, and buses is not available to operators of each unit.
(5)
The 125 V and 250 V DC systems are shared, which is not in compliance with current licensing requireu:.ents.
(6)
Stored energy for DG operation does not meet the seven-day minimum (or time to replenish, whichever is longer) required by current licensing criteria.
(7)
The 125 V and 250 V DC battery chargers are not capable of.restoring the battery to its fully-charged condition from minimum charge conditions during normal arid post-accident steady state loads.
5.O REFERENCES
- 1.
Branch Technical Position EICSB-7 from Standard Review Plant 7 of NUREG 75/087, "Shared E~ergency Electric Power Systems for Multi-Unit Generating Stations."
- 2.
U. S. Nuclear Regulatory Commission Regulatory Guide 1.81, "Shared Emergency and Shutdown Electric Systems for Multi-Unit Nuclear Power Plants."
- 3.
IEEE Standard 308-1974, "Criteria for Class IE Power Systems for Nuclear ?ower Generating Stations."
- 4.
Final Safety Analysis Repo~t, Dresden Nuclear Station Units 2 and 3.
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