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UNITED STATES y
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E WASHINGTON, D. C. 20555
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QCICBER 1 A F#
Docket ?!o. 50-313
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Mr. William Cavanaugh, III Vice President, Generation and Construction Arkansas Power & Light Company P. O. Box 551 Little Rock, Arkansas 72203
Dear Mr. Cavanaugh:
We have reviewed your submittals dated April 24, May 16, June 6 and June 8,1979, concerning your requests for amendments and Technical Specification changes to Facility Operating License fio. DPR-51 for Arkansas Nuclear One, Unit No. 1, as a result of IE 79-05A, IE 79-05B and the Order of liay 17, 1979.
In order to complete our review, we need additional information. You are requested to provide the enclosed additional information within 30 days of receipt of this letter.
Sincerely, j
f3-Robert W. Reid, Chief Operating Reactors Branch #4 Division of Operating Reactors
Enclosure:
Request for Additional Infomation cc w/ enclosure:
See next page 8201110485 810403 PDR FOIA l
MADDEN 80-515 PLR i
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' Arkansas Power & Light,Corpany F
cc:
Phillip K. Lyon, Esq.
House, Holms & Jewell 1550 Tower Building Little Rock, Arkansai 72201 Mr. David C. Trimble Manager, Licensing Arkansas Power & Light Company i
P. 0. Box 551 Little Rock, Arkansas 72203 I
Mr. James P. O'Hanlon F
General Manager Arkansas Nuclear One l
P. O. Box 608 Russellville, Arkansas 72801 Mr. William Johnson U. S. Nuclear Regulatory Commission P. O. Box 2090 Russellvill-e, Arkansas 72801 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division r
Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Troy B. Conner, Jr., Esq.
Conner, Moore & Corber 1747 Pennsylvania Avenue, N.W.
Washington, D.C.
20006 Arkansas Polytechnic College Russellville, Arkansas 72801 t
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Enclosure REQUEST FOR ADDITIONAL INFORMATION ARKANSAS NUCLEAR ONE, UNIT NO.1 1.
In your proposed Technical Specifications (TS), Sections 4.8.1, the operability of the emergency feedwater (EFW) trains, and hence the limiting conditions for operation, are determined by surveillance requirements on specific com-ponents within the trains. Our position is that the following additional revisions should be made to these in order to address the concerns of the bulletins and orders issued since Three Mile Island with regards to EFW:
a.
Add a surveillance requirement for the EFW flow instrumentation which states when the detectors will be calibrated and tested.
b.
Revise your requirements to include that the trains shall be tested on a staggered test basis.
2.
We have some concern as to the capability of your turbine driven EFW pump to cool the plant down to the RHR cut in temperature of 280*F since Psat for 280 F is approximately 50 psia. Demonstrate that your turbine driven EFW pump has this capability or provide a TS that requires an additional appro-priate action during a cooldown when only the turbine driven ERi pump is available.
3.
Your revised TS for limiting conditions for operation (LCO) in Section 3.5.1.7 regarding the operability of the reactor trip circuitry upon loss of main feedwater or turbine trip as written allows power operation (10 to 20%) when either of these trips are inoperable. Revise items -1, 2 and 3 as follows such that the trips may be bypassed up to these power levels but not inoperable and such that hot shutdown is the required action within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> rather than 24:
1.
Reactor trip upon loss of main feedwater shall be operable (as determined by Specification 4.1.a. items 2 and 36 of Table 4.1-2) at >5% reactor powe'.
(May be bypassed up to 10% reactor power).
r 2.
Reactor trip upon turbine trip shall be operable (as detemined by Spect-fication 4.1.a, items 2 and 42) at >5% reactor power.
up to 20% reactor power.)
(May be bypassed 3.
If the requirements of Specifications 3.5.1.7.1 or 3.5.1.7.2 cannot be met, restore the inoperable trip within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or bring the plant to a hot shutdown condition.
3a.
Provide the basis for the bypass setpoints of 10 and 20% power for the loss of main feedwater and turbine trips anticipatory reactor trips'.
4.
Your proposed TS do not include the revised temperature / pressure limits of Appendix G which are to be applied during emergency conditions (do not shutdown HPI flow until system pressure is twice that allowed by the previous brittle fracturelimits). Revise your specifications to include new emergency tempera-1
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' ture/ pressure limits. The revised specifications should clearly indicate 1
when the emergency limit is to be applied, and when the previous Appendix G limit applies.
5.
Verify that the number of acceptable scrams as limited by the TS include the additional scram functions (that is, reactor trip on main turbine trip and reactor trip on main feedwater pump trip).
If not, provide a list for these additional scram functions.
6.
The proposed TS do not address station operation with the electromatic relief isolation valve closed or electromatic relief valve inoperable.
In this regard, include reporting requirements for a special report on the status of a power operated relief valve which has been inoperable or isolated for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when reactor coolant system pressure is above the low reactor coolant system pressure trip setpoint associated with the reactor protection system.
This report is to be prepared and submitted to the staff within 10 days following inoperability or isolation of the power operated relief valve for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
7.
For all the proposed TS changes, provide a revised bases Section in your TS that support the proposed changes, l
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Distributfen Centrai file COT t ! :373 E00 Reading NRR Reading thR #3 File H.Dentoo
. E? CRC /,Eli FOR:
Chairman Hendrie E. Case Comissioner Gilinsky D.
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- y. q Ccmissioner Xennedy f
S..Varg. " p C'omissioner Bradford a
Ccmissioner Ahearne iams.
'(Signed) Las V. Gossick, THRU:
Executive Director for Operations
' A. Bournia ff. Rushbrook FRCM:
Harold-3.;Denton, Director,, NRR
SUBJECT:
INTERIM REPORT-Off SENSIT,IVITY STUDIES OF THE B&W REACTOR.
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DESIGR-The sensitivity of B&W' reactor designs.to transients and accidents hasibeen discussed previousTy with the Comission. Some preliminary findings and con-clusions were presented in an April 25,. 1979 NRR report (Ref.1)'which served in part as.the basis for confirmatory shutdown orders for operating plants with B&W reactors (Ref. 2-7). A more complete exposition-on this subject tsr in the staff report NUREG-0560 (Ref. 8).
. The NRR staff has reviewed the responses to the short-t'erm portions of the confirmatory orders and is in the process of. reviewing compliance with the icng-term order provisions Some of.the design changes al. ready accomplished have reduced ~the sensitivity aspect of the. B&W design, in particular the addition of anticipatory trips from abnormal. conditions in either t,he feede.
water or steam flow / pat.hs:of the steam, generator.
Hcwever, our review of the sensttivity of the B&W design has continued; and we -
are no.t yet fully sapsfied on this question. Recently we and.the Office of Research hav~ e initiated a. detailed study of the risk, on a relative basis,.
of a B&W design;. We, have' selected the Crystal River plant for this purpose.
Further' details'are provided in Ref.-9-(which is also attached here as
.Enel'osure. A).
As the-reliabiTity study may take 3-6. months, and't:enversion to licensing strategy would take further time,. we -are also considering whether it is necessary to halt portions. of the construction of B&W plants, pending the outcome of the reliability assessment. As a preliminary consideration, we have identified those systems and comconents that may_be impacted by possible design changes as a result of this. study. A survey was made of.the status
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ultiple Addressees
- of these systems an.d components for B&W plants under construction. Enclosure B indicates that in general, the desfgn, procurement and fabrication of these items are essentially' complete. The construction (or installation)' status of these items-Varies from zero.to 100 percent complete, with an estimated over-all average of about 50 percent complete,.except for the North Anna units, which are about 10 p'ercent complete..
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I have-today notified holders 'of cps. withr 83W design' of this ongoing study s
and under the authority of.10 CFR' 50.54(f),. requested them to' provide infor--
nation-to allow us to determine whether it is necessary to halt. all or -
portions, of the construction of S&W plants. Enclosure C is a copy of my
( ~ ~ ' ~~ -letter to one. 6f the utflities I intend to direct the utilities ro respond by. December 3,1979 and expect our evaluation to be completed i
.- by December -l 5,1979.
The long-term pro' gram for B&W operating reactors already addresses several of the concerns of the Crystal River study. As the results from the Crystal Rfver. study becomes available, they wilt be applfed to operating reactors.
as apprcprfate.
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r liarald R Denton, Ofrector
'. Offfee.of~. Nuclear Reactor' Regulation
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Enclosures:
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As Stated.-
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cc: SECY 7
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CPM: LWR #3*
Concurrence see previous yellcw..
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l p nogk UNITED STATES NUCLEAR REGULATORY COMMISSION f $.y;, g[(, j WASHINGTON, D. C. 20555
%,,'~-~.9h/jl-OCT : 5 mg MEMORDANUM FOR:
Chairman Hendrie Commissioner Gilinsky Commissioner Kennedy Commissioner Bradford Commissioner Ahearne THRU:
Executive Director for Operations '
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FROM:
Harold R. Denton, Direc+or, NRR
SUBJECT:
INTERIM REPORT ON SENSITIVITY STUDIES OF THE B&W REACTOR DESIGN The sensitivity of B&W reactor designs to transients and accidents has,been discussed previously with the Commission. Some preliminary findings and con-clusions were presented in an April 25, 1979 NRR report (Ref.1) which served in part as the basis for confirmatory shutdown orders for operating plants
'with B&W reactors (Ref. 2-7).
A more complete exposition on this subject is in the staff report duREG-0560 (Ref. 8).
The NRR staff has reviewed the responses to the short-term portions of the confirmatory orders and is in the process of reviewing compliance with the long-term order provisions. Some of the design changes already accomplished have reduced the sensitivity aspect of the B&W design, in particular the addition of anticipatory trips from abnormal conditions in either the feed-water or steam flow paths of the steam generator.
However, our review of the sensitivity of the B&W design has continued, and we are not yet fully satisfied on this question. Recently we and the Office of lesearch have initiated a detailed study of the risk, on a relative basis, l
of a B&W design. We have selected the Crystal River plant for this purpose.
Further details are provided in Ref. 9 (which is also attached here as Enclosure A).
j As the reliability study may take 3-6 months, and conversion to licensing strategy would take further time, we are also considering whether it is necessary to halt pcrtions of the construction of B&W plants, pending the outcome of the reliability assessment. As a preliminary consideration, we have identified those systems and components that may be impacted by possible design changes as 3 result of this study. A survey was made of the status I
e
Multiple Addressees.
of these systems and components for B&W plants under construction. Enclosure B indicates that in general, the design, procurement and fabrication of these items are essentially complete. The construction (or installation) status of these items varies from zero to 100 percent complete, with an estimated over-all average of about 50 percent complete, except for the North Anna units, wnich are about 10 percent complete.
I have today notified holders of cps with B&W designs of this ongoing study and under the authority of 10 CFR 50.54(f), requested them to provide infor-mation to allow us to determine whether it is necessary to halt all or
-l portions of the construction of B&W plants. Enclosure C is a copy of my letter to one of the utilities. I intend to direct the utilities to respond by December 3,1979 and expect our evaluation to be completed by December 15, 1979.
The long-term program for B&W operating reactors already addresses several of the concerns of the Crystal River study. As the results from the Crystal River study becomes available, they will be applied to operating reactors as appropriate.
AYk e
Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
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ENCLOSURE A l
OCT 4 1979 t
MDGANDUM FUR: Frank H. Rowsone. Acting Director Probabilistic Analysis Staff. RES l
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FRCH:
J. A. Harphy i
Probabilistic Analysis Staff. RES H. A. Taylor Probabilistic Analysis Staff. RES l
SUBJECT:
IREP - INITIAL PLAltT STUDY As requested, we have attesoted to develop a general framework for the conduct l
of a limited risk assessment of a B&W reactor aimed at identifying any unique risk-impacting sequences relative to the Reactor. Safety Study. An absolute i
detemination 'of risk is not intended. Tentatively. we have selected Crystal l
River 3 a plant owned and operated by Florida Power Corporation. Vor analysis.
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rne architect-engineer for this Babcock and Wilcox reactor was Gilbert Associates.
It began concercial operation in March 1977.
i The project, as presented in Figure 1. will require the following tasks:
1.
A survey of the LER files as now established in ort 1 and A0 reports, as t
well as the Sandia and Fluor-Zion systems interactions studies to identify
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interactions and conson mode failures which have occurred in staflar l
P ants. This survey should precede construction of systee logie models land event trees since it will ensure that actual experience is incorporated l
into the assessments perfomed.
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1.
Event trees for loss-of-coolant accidents and transient conditions.
Specific attention will be given to more frequent LOCAs and these will include a feedwater transient tree which incorporates experience at E1W plants and will explore the post-TMI modifications. Emphasis will be I
given toward understanding the human coupling interaction between systems at the event tree sequence level.
Fault trees for the key systems identified in the event trees. Tney will
-t 3.
be constructed to tne component level and will include contrui, actuation, i
and electric power considerations. human errors will be included as well.
as the ability of the operator to cope in the time span available. Our prelizicary' opinion is that simplified fault trees will be required for
-- the following systems: auxiliary feedwater and secondary steam relief, e
high pressure emergency core cooling in the injection and recirculation i
modes, low pressure energency core cooling in both injection and recirculation andes. containment spray and containment heat removal systems and a l
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2 GCT 4 1973 Frank h. Rowsom::
li=itad study of. lass of AC power probably done in detail for the 4a0 and 4160 busses and tne emergency diesel generators, with limited analysis of high voltage switchyard faults. Separate fault trees will probably be
, quired for ECCS and AFWS initiation logic and the systas trees must include the contribution from auxiliary systests such as instrument air.
ventilation, component cooling, etc., and control-induced failures.
l Tmacatan of the fault trees will be pemitted pmvided a written basis is provided. This basis will prese.nt the rationale why no coupling of cutsets is expected frem further development of the tree.
4.
As investigation of the adequacy of high pressure-low pressure interfaces.
5.
Acquisition of an appropriate data base for component failure rates and human errors. At the present time. it seems advisable to request that i
till Yesely update the RSS data base as required to reflect the data analyzed in his programs. This would include a tabular presentation of Nman error failure rates. It should be recognized that utilization of an updated data base will significantly cocplicate any meaningful caparison of the results with WAsti-1400.
C.
Quantification of the fault trees and event trees using error prvpagation to tne extent necessary. It may be desirable to separately pertom sensi-tivity analyses to demonstrate the effect of various distributions on systen error bounds.
7.
Analysis of the physical phenrrwna associated with drainant seque5ces to obtain esti=ates of the magnitude of releases fran the containment. This will aid in categorizing releases into appropriata release catepries.
i To conduct a program of this magnitude in a short time period delays associated with acquiring and transferring infomation must be minimized. Optimally, the event tree and fault tree analysts should share a conman location during the initial portion of the project.
(A. Garcia has indicated space can be made avalleble at the Air Rights Building for approatmately 12 persons.) As the fault trees pigrss below the top logic, however, the analysts should be i
located at or near the site with imediate access to as-built drawings and pivcedures as well as a representative of the plant operations staff. This l
will paruit verification of engineering and procedural details and will minimize I
information transfer and print reproduction. Access should also be arranged between the fault tree analysts at the site, the remaining teen in Scithesda.
tne architect-engineer, and the vendor. Such an arrangement will be costly i
and will require considerable efforts on the part of upper management to solicit the support of all parties involved.
- l 1
In addition to basic plant data, detaministic calculations cuty be required to understand the behavior of the plant under off-nomal conditions. This may also involve real-time simulation at an appropriate simulator to the extent i
possjble. The arrangements with the vendor should cover this possibility and i
it may be desirable to have confirmatory calculations made by one of the NRC contractors (e.g.. I::EL) on a selected basis.
i
OCT 41979 3
Frank H. Rowsone l
As previously noted, accurate sequence categorization To expedite the preparation of a final report, technical writing, graphics, Advan and pmsfreading support would be most helpful.and direction of the rl presented.
i Assuming the analysis is perfomed over a four-conth period, the following resources are considered necessary:
Event trees and contractor direction - M. Taylor, J. Harpby mentially full-time with review by either J. Curry, P. Baranowsky, or M. Cunningham 1.
on a 25 percent part-tirte basis.
l Fault trees - At least eight analysts chosen from the following:
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SAI E
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f L. Conradi S. Asselin J. Kelly M. Fedele M. Stewart J. Young G. reib A. Garcia A. Giacobbt S. Atkinson l
G. Loyd D. Leaver W. Cramond P. Wood F. Leverenz Fault tree review - J. Pittnan, F. Hanning essentially half-time.
3.
Fault and event tree quantification - At least two of the following:
r These E. Lofgren, R. Liner, J. Powers (all SAI), W. Ycsely, F. Manning.
4.
will be required full-time over a two-conth period.
I Accident analysis - R. Wooten. P. Cybulskis on an on-call consultant 5.
basis.
Peer review by NRR - Periodic review and consultation with a T
6.
of the following:
In addition. it would be desirable to H. Rabin. V. Lcfave,W. Minners.
inclu:la specific IE personnel such as the cognizant site inspector and IE training personnel on an as-needed basis.
f Quantificattos review - F. Goldberg and W. Vesely approximately half-time 7.
ever a two-month period.
- l Considering tho' cost of rented space at the site and Bethesda and the travel required, we esti= ate that an expenditure of about 5700K is required to comp
'[
PAS effort will be approximately 1.4 person-years this snort-ter-a program.
over this four-conth period.
We envision a enapilatios The nature of the final report needs to be detemined.
and description of the analyses performed', and a narrativs ao the importaat I
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i OCT 4 37g 4
Frank h. K wso=c l
engineering insights applicable to various plant systas and control featum r
with deterministic recoramendations suitable for licensing to supple =ent tne I
Additional inforcation f
Sundard Review Plan and Standard Technical Specifications.
i will t4 generated which could benefit PAS studies and operator, staff, and l
inspector training.
i i i
J. A. Nrp.hy Probabilistic Analysis Staff Office of huclear Reculatory Research
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H. A. T'aylor
. Prebabilistic Analysis Staff Office of naclear Acgulatory Research i
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1 FIGUME 1 1
1 i
4 ANALYSIS OF EVENT TREE d
ACCIDENT CONSTRUCTION PROCESSES a
AS NEEDED LER SURVEY PULL TOGETilER EXISTING INFO ALREADY C0tlPILED V
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^TI N A
E
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QUANTIFICATION
-k LICENSING FE T
CONSTRUCTIDH RECOHHENDATIONS 1
TREE SEQUENCES DETERMINISTIC, CALCULATI0 tis BY VENDOR AS NEEDED, ANAlfSIS OF 111Cl! PRESSURE -
SUMMARY
REPORT LOW PRESSURE INT.
OF RESULTS L
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MIDLAND UNIT 1 System / Component Design Component Component Construction Completed Procurement Fabrication Completed
(% ) (l &2 )
(%) (1)
(") (1)
(%)
HPI System 100 95-100 90-95 75 EFW System 100 95-100 90-95 85 DHR System 100 95-100 90-95 55 CFT System 100 95-100 90-95 20 RCS Pressure Control System 90 95-100 90-95 60 Makeup / Letdown System 100 95-100 90-95 70 SG Pressure Control System 100 95-100 90-95 60 Steam Generator 100 95-100 90- 9 5 50 Pressurizer 100 95-100 90-95 40 Quench Tank 100 95-100 90- 9 5 50 l
Control Room Layout 100 90 80 70(3) l l
RCS Piping 100 95-100 90 -9 5 50 l
NOTES:
(1) Units 1 & 2 show same percents due to parallel design and procurement for both
. units.
(2 ) All of these values are based upon the present design and does not include
" lessons learned" changes resulting from TMI-2.
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(3 ) This is for panel installation only. It does not include field cable terminations,'
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' HVAC and lighting. With all these included, a value of about 30-35% applies.
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T MIDLAND UNIT 2 System /Companent Design Canponent Component Construction Completed Procurement Fabrication Completed
(%)(1 & 2)
(%) (1)
(%) (1)
(%)
HPI System 100 95-100 90-95 80 EFW System 100 95-100 90-95 85 DHR System 100 95-100 90-95 70 CFT System 100 95-100 90-95 50 RCS Pressure Control System 90 95-100 90-95 60 Nakeup/ Letdown System 100 95-100 90-95 75 SG Pressure Control System 100 95-100 90-95 60 Steam Generator 100 95-100 90-95 65 Pressurizer 100 95-100 90-95 50 Quench Tank 100 95-100 90-9 5 75 Control Room Layout 100 90 80 70(3)
RCS Piping 100 95-100 90 - 9 5 85 NOTES:
(1)
Unit 1 and 2 show same percents due to parallel design and procurenent for both units.
(2 )
All these are based upon the present design and does not include " lessons learned" changes resulting fran TMI-2.
(3 )
This is for panel installation only. It does not include field cable tenninals, HVAC and lighting. With all these then included, a value of about
, -35% applies.
30
NORTH ANNA UNITS 3 AND 4 System / Canponent Design Canponent Component Construction Completed Procurement Fabrication Completed
(%)
(%)
(%)
(%)
HPI System 60 90 90 10 EFW System 60 90 90 10 OHR System 60 90 90 10 CFT System 60 90 90 10 RCS Pressure Control System 60 90 90 10 Makeup / Letdown System 60 90 90 10 SG Pressure Control System 60 90 90 10 Steam Generator 60 90 90 10 Pressurizer 60 90 90 10 Quench Tank 60 90 90 10 Control Room Layout 60 90 90 10 RCS Piping 60 90 90 10 l
5
BELLEFONTE 1 System / Component Design Component Component Construction Completed Procurement Fabrication Completed
(%)
(%)
(%)
(%)
HPI System 85 90 80 45 EFW System 98 90 85 65 CHR System 95 90 80 45 CFT System 98 80 70 30 RCS Pressure Control System 99 100 100 50 Makeup / Letdown System 85 90 80 45 I
SG Pressure Control System 99 100 1 00 50 Steam Generator 100 100 100 95 f
Pressurizer 100 100 100 50 i
Qu'ench Tank 100 100 100 100 l
Control Room Layout 99 100 100 60 RCS Piping 100 100 100 90 t
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BEU i.3GNTE 2 System / Component Design Component Component Construct 1on Completed Procurement Fabrication Completed
(%)
(%)
(%)
(%)
HP! System 85 90 60 5
EFW System 98 90 60 0
DHR System 95 90 70 5
CFT System 98 80 45 5
RCS Pressure Control System 90 100 100 10 t
Makeup / Letdown System 85 90 60 5
SG Pressure Control System 90 100 100 10 Steam Generator 100 100 100 80 Pressurizer 100 100 100 50 l
Quanch Tank 100 100 100 100 Control Room Layout 95 100 100 40 RCS Piping 100 100 100 45 L
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r WFPSS 1 & 4 System / Component Design Component Component Constructior Completed Procurement Fabrication Completed (1)
(%) (2)
(%) (4)
(%)
HPI System 100 100 95 10 EFW System 100 100 95 5
CHR System 100 100 95 5
CFT System 100 100 95 0
RCS Pressure Control System 100 100 100 0
Makeup / Letdown System 100 100 95 10 SG Pressure Control System (1) 100 100 95 0
Steam Generator 100 100 100 0
Pressurizer 100 100 100 0
Quench Tank 100 100 100 100 Control Room Layout (3) 100 100 0
0 RCS Piping 100 100 100 0
NOTES:
l (1)
Only level control system.
(2 )
All equipment on order but may not have been delivered.
(3 )
Panel Layout complete and instrumentation for panel.
(4)
Major component fabricated, but some special order item not yet delivered.
l