ML20248D047
| ML20248D047 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/22/1989 |
| From: | William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| GL-88-20, TXX-89720, NUDOCS 8910040109 | |
| Download: ML20248D047 (26) | |
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- E File # 10015 227 !Z~ Ref. # 10CFR51.95 1UELECTRIC September 22, 1989 W.J.CahlH Enecutwe Vice President U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET N05. 50-445 AND 50-446 SEVERE ACCIDENT MITIGATION DESIGN ALTERNATIVES (SAMDAs)
Gent? amen:
In accordance with the request of the NRC staff in telephone conversations on September 7 and 15, 1989, TU Electric hereby provides in the attachments to this letter the following information:
o Identification of potential severe accident mitigation design alternatives for CPSES based on available PRA studies, and an evaluation of existing capabilities at CPSES (Attachment 1); and o Cost evaluation of such alternatives (Attachment 2).
This preliminary information is being provided at this time because of the NRC staff's request. For economic evaluation, the costs in Attachment 2 should be multiplied by 1.35 to cover contingency factors and corporate costs such as interest, administrative an'd general expense, taxes, etc. In accordance with Generic Letter No. 88-20, Supplement No.1. TU Electric will submit to the NRC its method, approach and schedule for performing an Individual Plant Examination which will quantitatively evaluate potential design and procedural improvements.
Sincerely.
WLW &&
William J. Cahill, Jr.
By: -
4 Roger D. Walker Manager, Nuclear Licensing RSB/vid l
Attachments c - Mr. R. D. Martin, Region IV I)O~
Resident Inspectors, CPSES (3) , ,
Mr. J. H. Wilson, OSP-NRC 8910040109 DR 090922 ADOCK O5000445 400 nnh obvc streer La si vattas, Texas 75201 PNV
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.' -' Attach:::nt 1 to TXX-89720 Pag 2 1 of 11 ATTACHMENT 1 COMANCHE PEAK STEAM ELECTRIC STATION PRELIMINARY REVIEW 0F POTENTIAL SEVERE ACCIDENT MITIGATION DESIGN ALTERNATIVES BASED ON AVAILABLE PRA STUDIES
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'- Attachstnt'l to TXX-89720 Page 2 of 11'
. 1.0 OBJECTIVES The objectives of this preliminary review are to:
- 1) Identify potential. Severe Accident Mitigation Design Alternatives (SAMDAs) for Comanche Peak Steam Electric Station.(CPSES) based on the results of a number of available PRAs, and
- 2) Identify'the existing capabilities at CPSES which could be used to perform the same functions as the SAMDAs described'herein.
2.0 ENGINEERING EVALUATION NUREG-1150-(Reference 1), NUREG/CR-5102 (Reference 2), and the Zien, Indian Point, and Seabrook PRAs (References 3 through 5) were utilized to identify the potential SAMDAs which could be applicable to CPSES.[1] The NUREG-1150 study includes severe accident risk assessments for five nuclear power plants (Surry Unit 1. Zion Unit 1 Sequoyah Unit 1, Peach Bottom Unit 2, and Grand Gulf Unit 1). The NUREG/CR-5102 study evaluates the vulnerability of current pressurized water reactor designs to an interf acing systems Loss of Coolant Accident (LOCA). The nuclear power plants included in this study were Inaian Point Unit 3, Oconee Unit 3, and Calvert Cliffs Unit 1.
2.1 IDENTIFICATION OF SAMDAs The SAMDAs included in the aforementioned studies as design improvements or corrective actions were selected for evaluation. Then, the ones that obviously did'not apply to CPSES due to plant differences were excluded from the list. For example, one SAMDA required the addition of a third Auxiliary Feedwater ( AFW) pump which did not apply to CPSES since the AFW system at CPSES currently consists of 3 pumps (two motor-driven pumps and one turbine-driven pump) and is already a highly reliable system. The remaining SAMDAs were retained for further investigation to determine whether CPSES already has capabilities which could provide the same accident mitigation functions as the SAMDAs. Table 1 presents the potential SAMDAs selected for further evaluation for CPSES. A limited-scope cost analysis, which is outlined in Attachment 2, wrs also performed to estimate the costs associated with implementing each of these SAMDAs.
[1] Although the PRAs for Zion, Indian Point, and Seabrook have been used in identifying potential SAMDAs, it would be expected that the cost effectiveness of such SAMDAs would be substantially lower at CPSES in light of the lower population within 10 miles of the site (see Table 2).
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- Page 3 of.11-i 2.2 EVALUATION OF EXISTING' CAPABILITIES at CPSES~
k ' The CPSES Final Safety ' Analysis Report' (FSAR) (Reference 6), Emergency-Response Guidelines (Reference 7), various-plant and system drawings and other pertinent procedures were carefully reviewed during this study..
' Appropriate Operations personnel were also interviewed,to identify the t, capabilities that already exist at CPSES for mitigating a severe-
' accident.. .The capabilities associated with.each SAMDA are summarized below.
'o- 'CPSES Capabilities Associated with SAMDA ho. I
- 1. Provide an additional Station Service Water (SSW) Pump for each-unit to improve the reliability of'SSW system.
Steps have already been taken to improve the reliability of the Station. Service. Water; System at CPSES beyond what was originally proposed in the Unit I design. Both of the Unit I service water. pumps (each with 100% capacity) operate continuously. In addition,'a commitment'has'been made to ensure that at least one of the Unit 2 service water pumps is always available. The Unit 1 and Unit 2 service water systems can be' cross-tied with three manually operated valves. These manual actions are described in ABN-501A, Revision 1 " STATION SERVICE WATER SYSTEM MALFUNCTION" and SOP-501A. ABN-501A also
.provides instructions intended to provide ~ alternate cooling to t vital plant equipment should the service water system be out of
' service.
-Because the aforementioned feature essentially duplicates the function of.SAMDA No. 1; a cost evaluatinn has not been included in Attachment 2.
o' CPSES Capabilities Associated with SAMDA No. 2
- 2. Provide an additional diesel generator for each unit to increase the reliability of the AC power system.
TV Electric has implemented the following procedures for station blackout events in response to Generic Letter 81-04, dated February 25, 1981:
ECA-0.0, Rev. 3 " Loss of All AC Power" ECA-0.1, Rev. 3 " Loss of All AC Power Recovery Without SI Required" ECA-0.2, Rev. 3 " Loss of All AC Power Recovery ,
With SI Required" 1
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-Attach; Intl 1 to TXX-89720
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.These procedures provide guidance.for the operators to safely shutdown the. plant following a total loss'of offsite and onsite (diesel generators) power. These procedures will.be enhanced, as necessary, pursuant to the requirements of.the Station
, Blackout Rule (10CFR50.63).
- o. CPSES Capabilities Associated with SAMDA No. 3
- 3. Improve the DC batteries capability to provide 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of.I&C-power as opposed to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in the event of a station blackout.
Currently, the only method of extending the' operating life of the batteries during a station blackout event is by sequentially removing selected equipment from the DC bus.after its' function has been completed. Via the Emergency Operating Procedure ECA-0.0, Revision 3, " LOSS OF.ALL AC POWER", the loads on the safeguards DC buses are minimized. In addition, the maintenance of a suitable environment for the batteries is addressed in ECA-0.0 and ABN-202A, Revision 0, " LOSS OF NON-SAFEGUARDS VENTILATION SYSTEMS". The amount of additional time which may be gained by following the above procedure has not
- been determined, o CPSES Capabilities Associated with SAMDA No. 4
- 4. Provide a system to automatically switch the suction of the safety injection and centritagal charging pumps to the Residual Heat Removal (RHR) pump discharge to initiate the high pressure recirculation phase of the Emergency Core Cooling System (ECCS).
CPSES does not have automatic systems for switching the suction of the safety injection and centrifugal charging pumps to the RHR pump discharge: however, the reactor operators are provided with explicit instructions for manually performing the switchover to cold leg recirculation. In addition, there is an automatic switchover capability which is used to switchover the suction of the low pressure safety injection (RHR) pumps from the RWST to the containment sumps following the receipt of a Low-Low Refueling Water Storage Tank (RWST) level alarm (at 40%
RWST level).
Moreover, for those events where safety injection is required, e foldout page and appropriate cautions are provided in the Emergency Response Guidelines to instruct the reactor operators to initiate the switchover to cold leg recirculation whenever the RWST level drops to 40% of span. Then, in accordance with EOS-1.3, Revision 4. " TRANSFER TO COLD LEG RECIRCULATION". the reactor operators are instructed to manually realign the suction of the safety injection and centrifugal charging pumps to the discharge of the RHR pumps. l 4
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-Attach cnt'l to TXX-89720
.Page 5 of 11 o CPSES Capabilities Associated with SAMDA No. 5
- 5. Provide a means to rapidly depressurize the Reactor Coolant System (RCS) to enable low pressure safety injection / low.
pressure recirculation.
.The failure of high- and intermediate-head safety injection.
and/or-cold leg recirculation is significant for those accident scenarios involving a small LOCA. Following a small LOCA, the pressure in the reactor coolant system would fall to the saturation pressure at the hottest temperature in the system.
To reduce the pressure to below the-shutoff head of the low pressure injection pumps, the coolant temperature must be lowered by approximately 2000F, This temperature reduction
! must be achieved through the use of steam generator cooling (feed and bleed cooling at these pressures requires either a centrifugal ' charging pump or a safety injection pump).
- a. Qualify the pressurizer power operated relief valves (PORVs) for harsh containment environments.
The pressurizer Power Operated Relief Valves (PORVs) are currently qualified as active valves for normal containment environments. Based c7 comparisons to similar valves, there is reasonable assurance that the CPSES.PORVs would also perform their intended function in harsh environments: hence the cost of further qualification of the PORVs is not included in Attachment 2.
- b. Enlarge the pressurizer vent valve to be of an equivalent size to the pressurizer PORV for use as a backup for the feed & bleed cooling function.
- In accordance with the background documerts supplied'with the generic Westinghouse Emergency Response Guidelines, both pressurizer PORVs are required to be opened to support successfp1 feed and bleed cooling. The Emergency Response @!jdel snes instruct the reactor operators to open the pressurizer and reactor vessel head vents if one of the pressurizer PORVs cannot be opened; however, the size of these vents is very small relative to the size of the PORY, thus limiting their effectiveness as alternate pressure relief pathways, o CPSES Capabilities Associated with SAMDA No. 6
- 6. Mitigation of Interfacing Systems LOCA
- a. Install pressure or leak monitoring instruments (permanent pressure sensors) between the first two pressure isolation valves on low pressure injection lines. RHR suction lines, and High Pressure Inspection (HPI) lines.
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[t ' .Attachrnt~1 to TXX-89720 Page.6 of 11-The CPSES design has no permanent pressure sensors between
'the. first two pressure isolation valves 'on the . low pressure injection,1ines, RHR suction lines, and high L pressure-injection' lines which can be used to detect and
[ identify the location of an ~1nterfacing system LOCA.from the. control room. No. instrumentation-is available to identify the source of leakage from the RCS through any of the redundant pressure boundary isolation valves.
P b. Develop a leak testing program for the pressure boundary isolation valves. The leak test should be performed L during refueling outages and after specific valve Amintenance.
.CPSES has implemented a program for testing the reactor coolent system pressure isolation valves. This program, described in EGT-712A . Revision 4 " REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVE LEAKAGE TESTING", was designed to verify that the leak rate through these valves is in accordance with the plant Technical Specifications (T/S 4.4.5.2.2) (Reference 8). This procedure requires that-the pressure boundary valves be tested for leakage at-least once per 18 months, prior to entering Mode 2 after the plant-has been in cold shutdown, and after specific valve maintenance. The procedure also requires that, in accordance with ASME Code,Section XI, paragraph IWV-3427(b) for valves 6" and larger:
a) the. test frequency he doubled if the leakage rate has increased by a certain amount, and, b) if the test shows a trend toward an increasing leakaae rate, and the next leakage rate will exceed the allowable limits by the next scheduled test, the valve will be replaced.
This leak testing program frequency of interfacing systems LOCA.
Because the aforementioned program satisfies SAMDA Sb, a cost evaluation has not been included in Attachment 2.
o CPSES Capabilities Associated with SAMDA No. 7
- 7. Install a system to vent the containment through a vent path routed to an external filter.
The CPSES containment design includes three ventilation systems:
- 1) Lu..tainment Purge Exhaust System;
- 2) Containment Pressure Relief System;
- 3) Containment Hydrogen Purge System.
- - Attachmnt 1 to TXX-89720-p Page 7 of 11 Both the Containment Purge Exhaust System and the Containment Pressure Relief System vent through the Primary Plant Ventilation System. The Hydrogen Purge System contains its own filters and vents through the plant stack. The Containment Pressure Relief System does not have sufficient capacity to provide pressure relief under a'ccident conditions. The l Containment Purge Exhaust System is normally used only for operations in Modes 5 and 6. Manual, non-proceduralized actions could be performed to provide containment pressure relief. The Hydrogen Purge System is normally used for post-accident cleanup of the containment. However, the " pressure ratings" of these systems are significantly less than the pressures expected if the containment integrity were to be challenged (~50 psig). Thus, it is expected that the usefulness of these systems as filtration systems is small. In addition, the isolation valves may not be operable with the large containment-to-atmosphere pressure differential expected during a severe accident. However, if the redundant isolation valves can be operated with the pressure differential, these systems may be used to vent the containment directly to the atmosphere in lieu of allowing the containment to fail in t.n unpredictable and possibly uncontrollable manner. No procedural guidance is available at this time for the use of any of these filtration systems during accident conditions.
Other systems which are available at CPSES to improve containment performance are discussed under SAHDA No. 8 and 9.
o CPSES Capabilities Associated with SAMDA No. 8
- 8. Install a Hydrogen Ignition System which includes _ hydrogen concentration monitors and measures to control the ignitors.
CPSES does not have a Hydrogen Ignition System. The CPSES design does include (i) Containment Hydrogen Monitoring System which provides continuous indication in the control room and (ii) hydrogen recombiners to limit the hydrogen concentration in containment. The recombiners use natural circulation to draw the containment air through a heating chamber where the free hydrogen is oxidized to form water. There are two, independent, safety-grade recombiners in each containment.
Each recombiner is sized to accommodate the hydrogen expected to be released following a design-basis large break LOCA. In accordance with several of the Function Restoration Guidelines, the hydrogen recombiners are manually started in accordance with S0P-206A, " ELECTRIC HYDR 0 GEN RECOMBINER SYSTEH", if the hydrogen concentration exceeds 0.5%.
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Attachzent I to-TXX-89720 Page 8 of 11 o CPSES Capabilities Associated with SAMDA No. 9
- 9. Install a system to flood the reactor cavity to promote debris coolability, minimize core-concrete interaction and provide potential fission product scrubbing (e.g., Zion reactor cavity flooding system).
CPSES does not have a system whose sole purpose is to flood the reactor cavity: however, there are several methods available by which the reactor cavity can be flooded if necessary.
Due to the relative elevations of the containment sumps and the reactor cavity, the cavity will be partially filled if the sumps overflow. Following a LOCA and/or containment spray actuation, the containment will be flooded with most of the contents of the Reactor Coolant System (~90,000 gallons) plus the contents of the RWST (~500,000 gallons).
If this water is insufficient to fill the sumps (sump level monitoring is provided), a direct floe pda from the RWST to the containment sumps can be m3de availabit to force the overflow of the sumps. In addition, the P.WST can be replenished in accordance with Attachment 2 of ECA-1.1 " LOSS OF EMERGENCY COOLANT RECIRCULATION". The RWST can be replenished from the Reactor Makeup Water System, the Unit I spent fuel pool or the Unit 2 RWST. In addition, a non-proceduralized option is to replenish the RWST with firewater pumped from two 500,000 gallon tanks or directly from Squaw Crcek Reservoir.
Thus, the overflow of the containment sumps into the reactor cavity can be assured.
2.3 OTHER KEY SAFETY FEATURES at CPSES There are other_ safety features / systems available at Comanche Peak which are very important in the mitigation of severe accidents, including the following:
o The CPSES auxiliary feedwater system is a very reliable system. The auxiliary feedwater system consists of two motor-driven auxiliary feedwater pumps and one steam turbine-driven auxiliary feedwater pump. The capacity of the turbine-driven auxiliary feedwater pump is equivalent to the capacity of both of the motor-driven auxiliary feedwater pumps. Either the turbine-driven auxiliary feedwater pump or either one of the motor-driven auxiliary feedwater pumps is sufficient to provide adequate cooling following any of the design basis accidents.
o If the auxiliary feedwater water system should fail entirely, CPSES has Functional Restoration Guidelines which provide instructions to the reactor operators. In accordance with FRH-0.1, Revision 4,
" RESPONSE TO LOSS OF SECONDARY HEAT SINK", the main feedwater pumps could be used to provide water to the steam generators. If the main feedwater pumps were unavailable, the condensate pumps could be used.
, . .g Attachxent 1^to-TXX-89720 Page'9 of 11 o: If none'of the feedwater systems could be useduto provide feedwater cooling-to'the steam generators prior to the time the steam
. generator water level falls below a specified setpoint, the reactor operators are instructed tol initiate feed and bleed cooling function. Any one of the centrifugal charging pumps or safety.
injection. pumps will provide. sufficient makeup to the Reactor
.Icolant System (RCS). Both of the pressurizer PORVs.would then be mened to provide the bleed pathway.
3.0 REFERENCES
- 1. 0;S. Nuclear Regulator: Commission, " Severe Accident Risks:
An Assessment for Five U.S. Nuclear Power Plants,"
-NUREG-1150,. June 1989.
.2. G. Bozoli,- P. Kohut and R. Ritzpatrick, " Interfacing Systems LOCA: Pressurized Water Reactors," Brookhaven National . Laboratory, NUREG/CR-5102, BNL-NUREG-51235, February 1989.
- 3. " Zion Probabilistic Safety Study," Commonwealth Edison Company, September, 1981.
4' . "Indien Point Probabilistic Safety Study," Power Authority of the State of New York, Consr?idated Edison Company of New. York, INC., 1982.
- 5. "Seabrook Station Probabilistic Assessment," PLG-0300, December 1983.
- 6. Comanche Peak Steam Electric Station, " Final Safety Analysis Report".
- 7. Comanche Peak Steam Electric Station, " Emergency Response Guidelines Manual".
- 8. Comanche Peak Steam Electric Station Unit 1, " Technical Specifications".
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- Attach $ent'1.toTX'-89720 X Page 101of 11 Table l' . Potential SAMDAs Considered for Evaluation for CPSES
- 1. Provide an additional Station Service Water-(SSW) Pump for each unit to improve the reliability of SSW system.
- 2. Provide an additional diesel generator for each unit.to increase the reliability of the AC' power system.
- 3. Improve the DC batteries capability to provide 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of I&C power as opposed to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in the event of a station blackout.
- 4. Provide a system to automatically switch the suction of the safety injection and centrifugal charging pumps to the Residual Heat Removal (RHR) pump discharge to initiate the high pressure recirculation phase of the Emergency Core Cooling System (ECCS).
- 5. Provide a means to rapidly depressurize the Reactor Coolant System (RCS) to enable low pressure safety injection / low pressure recirculation
- a. Qualify the pressurizer power operated relief valves (PORVs) for harsh containment environments,
- b. Enlarge the pressurizer vent valve to be of an equivalent size to the pressurizer PORV for use as a backup for the feed & bleed cooling function.
- 6. Mitigation of Interfacing Systems LOCA
- a. Install pressure or leak monitoring instruments (permanent pressure sensors) between the first two pressure isolation valves on low pressure injection lines,RHR suction lines, and High Pressure Injection (HPI) lines.
l b. Develop a leak testing program for the pressure boundary isolation valves. The leak test should be performed during refueling outages and after specific valve maintenance.
- 7. Install e system to vent the containment through a vent path routed to an external filter.
- 8. Install a Hydrogen Ignition System which includes hydrogen concentration monitors'and measures to control the ignitors.
- 9. Install a system to flood the reactor cavity to promote debris coolability, minimize core-concrete interaction and provide potential fission product scrubbing (e.g.. Zion reactor cavity flooding system).
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Page,l11Eoff 113
-A Table ~ 2.- Population - Datia. Within 10. Miles ' off Plant' Site
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Total. Pooulation (1980 estimates)
PLANT-0 2' miles- .0'- 10 miles Comanche Peak 81 :.15,488 Zion- 15,506' 23'4,180
~285,000'
- Indian Point ~25,000 Seabrook
- * 9,464. 104,884
- - Tota'l population for Seabrook'is based on 1986 estimates.
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- ATTACHMENT 2
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[. COMANCHE' PEAK STEAM ELECTRIC STATION
-PRELIMINARY COST-EXTIMATES ASSOCIATED WITH IMPLEMENTATION OF POTENTIAL SEVERE ACCIDENT ,
' MITIGATION DESIGN ALTERNATIVES
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SUMMARY
OF COSTS' ASSOCIATED WITH SAMDAs g
The followingl assumptions were applied by Bechtel Corporation in' estimating costs related to each of the options:
o All 'estimat'es assume that the work will! be performed 'in a nuclear operating plant environment which considers labor productivity effects due to work inside the' security fence and radiological control areas of the plant.
o These estimates are order-of-magnitude. :They are based on design concepts as identified by Texas Utilities.
o- Pricing is based on 1989 costs, o' No overtime premiums for manual labor have been included.
o- Health Physics support was based on direct labor. costs requiring RWPs.
o TUECO Support costs were based on 40 percent of Bechtel Engineering and Other Home Office costs.
o Unitized costs are' based on percentages of work as they relate to Unit 1 and common space, o Where direct craft labor manhours were developed, they were priced at
$20,00/ hour. based on current Texas construction experience.
o Costs represent the total scope for one unit.
o Order-of-magnitude quantity development was based on input from Bechtel engineering using preliminary design concepts. A11cwances were included based on engineering / estimating judgment for quantities for which -j preliminary design was not available. 1 o Material pricing is based en current Bechtel pricing information, o Labor rates reflect recent nuclear construction experience.
o Distributable manual and non-manual labor costs are included as a percentage of direct labor costs.
o Engineering and other home office costs are based on individual estimates for each of the options studied.
o The preliminary cost estimates have used the two following items as basic assumptions:
- Environmental qualification of components added to the modification is not required unless failure of the equipment could have an adverse impact on other qualified equipment.
- Structures, systems and components added by the modification will not be safety-related or seismic Category I unless they adversely impact
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- 3. [the functjon of s'afety-related equipment.. structures or materials. -
_ Cases requiring Seismic Category 1: treatment are so,noted.
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o' No. escalation. costs are included..
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.o. No~ costs.. associated-with unit unavailability during~ installation of the
-SAMDAs have been included.
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.Page!4 off.14
- g. a i-STATION BLACK 0UT DIESEL (SAMDA- NO. 2) -
S , . Base Option: Non-1E: Diesel, Non-Seismic. Building' H! _
, : Description ' Cost (Millions)
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Equipment,. Materials- '$c3.3
'and. Subcontracts' f i : Inst a11 ati on' (Labor. .. .
0.'8 Overhead'and Supervision)'
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- Enginee' ring and QA' 3.2 i.
Dwner's' Support Cost
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-TOTAL $ 8.5 L
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- 'Does not include Owner's. cost.such as operations and maintenance, radwaste disposal, or anti-contamination clothing.
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i Attachment 2 to TXX-89720
'Page 5 of 14 STATION BLACK 0UT DIESEL (SAMDA NO. 2) t Option 1: Non-1E Diesel, Seismic Category 1 Tornado Protected Building Description Cost (Millions)
Equipment, Materials $ 6.5 and Subcontracts Installation (Labor. 0.8 Overhead and Supervision) l Engineering and QA 4.5 Owner's Support Cost
- 128 TOTAL $13.6
- Does not include Owner's cost such as operations and maintenance, radwaste disposal, or anti-contamination clothing.
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' , STATION BLACKOUT DIESEL (SAMDA NO. 2)-
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Option Zi _ Class 1E Diesel; Seismic Category 1 Tornado Protected Building ~.
, Description; ' Cost (Millions)
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Equipment, Materials $12.0 and Subcontracts-E P Installation (Labor,- 1.0 Overhe.ad and Supervision);
-Engineering and.0A 7. 0 -
Owner *s Support Cost *
. _Z.d TOTAL $22.8-o,
- Does no't include Owner's cost such as operations and maintenance, radwaste disposal. or anti-contamination clothing.
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' IMPROVE DC. BATTERIES.(SAMDA NO. 3) t'4 j'
- Des crioti on' - Cost (Millions)
.4 : Equipment,iMaterials? $ 0.6 rJ Land Subcontracts b- ,
- Insta?lation (Labor , . . . . ..
0.9 M Overhead::and Supervision)- -
p Engineering,and QA 0.5-n M
K 0wner's Support Cost *-
TOTAL '$ 2.2 i
- ' Does' not include 0wner's' cost such as ' operations and maintenance, radwaste
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U disposal', or' anti-contamination clothing.
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AUTOMATE CHANGE 0VER FROM SAFETY INJECTION TO RECIRCULATION (SAMDA NO. 4).
Description. Cost (Millions)
Equipment, Materials $ 0.5 and Subcontracts Installation.(Labor, 0.2 Overhead'and Supervision)
Engineering and QA . 0. 6
~ Owner's Support Cost
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. TOTAL. $ 1.5.
- Does not include Owner's cost such as operations and maintenance, radwaste disposal, or' anti-contamination clothing.
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,i RAPIDLRCS'DEPRESSURIZATIO11 (SAMDA NO. SB)
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- Option1: . Enlarge Pressurizer: Vent;Line.
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Description:
Cost (Millions) 5:.
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. Equipment, Materials . -$ 0.2
- and Subcontracts
. Installation-(Labor,- :0.1 Overhead ~and Supervision).
Engineering and:0A' O.8
<*- :0wner's Support Cost
- M TOTAL' $'1.4-
- Does not;inclu'de 0wner's. cost-such as operations and' maintenance, radwaste
. :- idisposal, or anti-contamination clothing.
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- t" EAttachnnt'2 to TXX-89720'.
PageJ10 of 141 Y , RAPID RCS DEPRESSURIZATIO11'-(SAMDA NO. . 5B)
L0ption 2: ' Add a New Pressurizer Vent Line
.r+ ,
Description Cost (Millions) -i.
a Equipment.. Materials - $ 0.2'
.and Subcontracts Y Installation.-(Labor. .0.1 i Overhead and Supervision).
, Engineering and'0A 0.8 i a
M Owner's:. Support' Cost
- TOTAL'. $ 1.4 i
)
- Does not include 0wner's. cost such as operations and maintenance, radwaste disposal, or anti-contamination clothing.
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,S' Attache:nt 2 to'TXX-897202 t
- b. :; c-JPage,11'ofl14>
iMITIGATION'0F INTERFACING-SYSTEMS LOCA (SAMDA NO.'L6A)'
t-i - . Description- Dst (Millions)
Equipment, Materials -$.0.1
, ~ and. Subcontracts lInsta11ation'(Labor. 1.3 Overhead and Supervision)
Engineering.and QA 0.3-
~ Owner *s: Support Cost * ' M TOTAL' $ 2.0
.[
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'
- Does not' include'0wner's cost such as operations and maintenance, radwaste disposal, or anti-contamination clothing.
ji $l[ :.4 ? . ,
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y' A - JAtOchtent 2'to-TXX-89720
[" P' age 12lof.14, 7
l CONTAINMENT VENTING,(SAMDA' NO.' 37I-c t-l , ,
> : Description Cost (Millionsi n ,
^ ^
. Equi pment',-- Materi al s ; i$ 4.4 F and-Subcontracts'-
Installation (Labor = 9.9' I; . Overhead and Supervision) p l . Engineering and.0A 1.6
'0wner's Support Cost *.- M iTOTAt. $16.5-
- Does.not include Owner's cost such as operations and maintenance, radwaste
' disposal, or anti-contamination clothing.
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r Attach::nt 2 to TXX-89720
, 'Page 13 of 14
'y HYDROGEN IGNITION SYSTEM (SAHDA NO. 8)
Description Cost (Hillions)
Equipment Materials $ 0.3 and Subcontracts Installation (Labor, 0.7 Overhead and Supervision)
Engineering and QA 2.3 E.
Owner's Support Cost
- La TOTAL $ 4.3
- Does not include Owner's cost such as operations and maintenance, radwaste disposal, or anti-contamination clothing.
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, .: Attachn;nt 2 to'TXX-89720.
. <- . Page 14 of 14o
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REACTOR CAVITY' FLOODING SYSTEM'(SAMDA'N0'. 9)
I' - Description Cost (Millions)-
[:
U Equipment Materials $ 0.1 and' Subcontracts Installation..(Labor,- 0.2
.0verhead and= Supervision)
L .
- Engineering:and QA; '1.0 Owner's' Support Cost *. 9.d TOTAL $ 1.7.
~* Does' not include Owner's cost such.as operations and maintenance, radwaste -
. disposal, or.= anti-contamination clothing.
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