ML19256D345
| ML19256D345 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/10/1970 |
| From: | Case E US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7910170849 | |
| Download: ML19256D345 (9) | |
Text
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T '.,' 3 10 '6970 P. A. ?!crris, Director Divicien of Ecactor Licencing QU2STIO:iS n"IATEG TO DSTRI'"I?trATIC:I A '.D DC20CCY PGIER; THTIE lHL" ISIAND L"JIT fl; DCCIET #50-239 Please include the attached questions a=eng those in preparation for transmittal to the applicant.
Original +4ned b3 E. G. Case ESB-67 Edson G. Case, Director DRS:ES3:DFS Divisien of Reactor Standards
Enclosure:
Questions - Instrumentation
& Emergency Power cc w/cncl:
Distribution:
" *A R. Boyd h uppl.
R. DcYoun DR Reading D. Skovholt DRS Reading C. Lon3 ESB Reading D. Ross bec:
E. G. Case V. Moore D. Sullivan 1454
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Form AEC-4&e (Rev. 9-316 AECM 02 0 e s. aovamanant eeintros or,'c i ee e-see-en 0
791017
21REE !!ILE ISLAND, UNI';' #1 A.
In::trierntatien 1.
In regard to the protection syste:u which actuate reactor trip and engineered safety feature action, the following informatica should be providod:
a.
A list of those systems designed and built by Babecck &
Wilcox that are identical to those of the Oconce Nucica:
Station (as documented in tha Oconce FSAR) and a discussion of any design differences; b.
A list of those systems and their suppliera that arc designed and/or built by suppliers other than Babcock & Wilcox; and c.
Identification of those features of the design which differ from tha criteria of IEEE 279 and the Cc:xsission's proposed Cencral Design Criteria and an explanation of the reasons for any differences.
2.
In regard to the Babcock & Wilcox designed centrol systems, the following information should be provided:
a.
Identification of the major plant control systems (e.g.,
primary tocperature control, primary water level centrol, secam generator water level control) which are identical to those in the Oconce Nuclear Station; and 1454 327
Three IEle Island 91 2
b.
A list and a discussion of the deci;n diflarences b1 those systems not identical to those used in the Cccas:
Nucicar Station. This discussion ahculd include en evalua-tion of the safety significance of cach design chan;2.
3.
State the seismic design criteria for the reactor protectica system, engineered safety featuro circuits, and the energ2ncy power syste.1 including the station batecries. Tiu criteria should cover (1) the capability to initiato a protective action durit.3 peak acceleration, and (2) the capability of the enginecred nm-4 safety feature circuits to withstand scianic disturbances during poot-eccident operation. Dcocribe tha qualification testing requirements which will be used to assure that the critoria are satisfied and the means by which these requircaents will be i= posed on equipment suppliers.
4.
Describa the quality assurance procedures which apply to the equipment in the reactor protacticn system, engineered naf2ty featura circuits, and the emergency electric power synten. This description should include : (a) quality assurance precedurca used during equipment fabrication, shipoent, fic1d storaga, field installatien, and system ec=ponent chacheut; and (b) records pertaining to (a) above.
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Three Pdic Island G1 3
5.
To supplement the infor:stica prescated on p:ges 3-6.7 of the rc.in, submit the criteria and their bases which establish the mininua requirements for preserving the independence of redundant reacter protection systana, engineered safety facture sys te=a and Cla;a ;;"
Electrical Systems through physical arrangecent and separation and assure mini:nas availability during any *desiy basis event.
~_t c submittal should include a discussion of tha adainistrative rec-pensibility cnd centrol to be provided to assure conpliancu sfiu these critoria during the design and insta11atian of these sycteca.
The criteria and bases for the installation of cicctrical cable for these systoca should, as a minimn, address:
a.
Cabic darating.
b.
Cabic routing in cor.cainment, penotration areas, cabic spreading rooms, centrol rooms and other congested or hostile areas.
c.
Sharing of cabic trays with non-safety related cabica er with cables f the same system or other systene, d.
Fire detcetion and protectien in the areas where these cables are installed.
c.
Cabic and cabic tray markings.
f.
Spacing of viring and components in centrol boards, pancis,
and relay racks.
3 Circuit overload protecticn.
- Cicas II cicctrical systems and design basis events are dcfined in th:
Proposed IZZZ Criteria for Class IE E1cetrical Systens for Nuclear Pa.ac Generating Stations (IZZ1-303).
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Tarce lille Island 61 4
6.*w State the design criteria for reactor protection system and engineered cafety feature r21sted cicctrical and uchanical equi.u nt located in the primary centainment or clu ur.cre :n t
the plant which take into account the potential eff: cts c' radiation en theoc components 6ac to cither nor=al cperatica or accident conditions (superirposed cn len3-tara naral operation). Deceribe the analysis and testin~,perfar:cd to verify compliance with these design criteria.
7.** Identify all safety related equipment and components (c.c.,
motors, cables, filters, pump seals) located in the pri=ary containment which orc required to be operabic during and subsc-qucnt to a loss of coolant or a steamline break accident.
De s -
cribe tha qualificatione tests which have been or vill be perfo rad on cach of these items to inaure their availability in a ec=binud high temperature, pressure, and humidity environ =cnt.
3.
State the criteria which have been established to assure that loss of the air conditioning and/or ventilation cysta= uill not adversely affcet operability of safety reisted central an_
ciectrical equipmnt located in the control reen and oth:r
- Tacse qucaticna relate to the engineered safety feature chapter cc... :
FSAR and should be forusrded to the applicant with other questi:-m :: -
carning that chapter.
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Three Mile Island 91 5
equipacnt rocas.
Describo tho analysis perforrud to ide.ntify the worst casa envirenacnt (e.g., tc=perature, humidity).
State the limiting condition with regard to te=poratur_ that uculd require reactor shutdown, and how this was determined. Describ any testing (factory and/or onsite) which has been or will be performed to confirm satisfactory operability of central and electrical equiptunt under extreen environ =cntal cenditiena.
9.
Describe how reactor protaction system and engineered safety equipacnt will be physically identified as safety related equipcaent in the plant.
10.
Describe the method for periodic testing of engineered safety feature actuation to show it to be consistent with IEC 279.
k'o interpret IEEE 279 to requiro for engineered safety feature actuation the rarn high degree of on-line testability required for the reactor trip system.
11.
Provido a description of the instrumentation systc=s included in your design for remote monitoring of post-accident ecnditien:
within the primary containment. Provido an analysis to shcu that these systems are adequate over the full spectrun of postulated accidents.
12.
Describe what information is available to the operatar to identify 1454 331
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all RPS and EOF channels that are in tect or caintenance. Ecacribe the indication availabic, down to the channel lavel, taidentify which instrutunts initiate a protective action. These descrip-tions should be in sufficient detail to permit a datomination of the system's cocpliance with Sections 4-13 and 4.19 of ECE 279.
13.
Describe any rod spcod limiting features which provent withdrawal rates in exccas of 30 inches / eda.
14.
Do the circuits which prevent iz:: proper sequcacing of the rods conform to the provisions of IEEE 2797 15.
Do the circuits which autocatically terminatu dilution of the primary coolant conform to the provisions of IEEE 279?
16.
Identify the electrical and pneumatic components (valves, pumps, etc.) of the auxiliary cooling systems which should be censidered as portions of tha ongineered safety features.
Do your criteria with respect to the design of the associated instrumentatica and power systems for operation of thcoc co=ponents conflict in any way with IEEE 279 or the Cencral Ecsign Criteric?
3.
E.ernoncy Power 1.
Provide the design criteria and inforcation concerning the ensita c1ccerical pcwer systems as followc:
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Three Mile Island 61 7
a.
The percentage of the continuous rating of each diccci generator that the engineered safety featurcs cicctrical loads will require. The centinuous rating is defined as that centinucus load which will permit cupplier guarant2cd cperation at a 95". availability with an annual naintent.u c period.
b.
The 2000-hour and the 30-cinute dicael generator overload ratings.
2.
Provide your basis for sizing the station batteries to operate for two hours (without boucfit of any station power).
3.
Discuss the analyses to be performed to show that neither loss of a unit of this station nor the loss of the largest generating unit en tha grid will negate the ability to provitc offsite pcuer to this staticn.
4.
Provide an analysis to show that no singic failure within any d-c system (e.3., station battery) adversely affects the shedding of loads and/or the cpeni:q of supply breakers such that adequat:
diesel generator operation is provented.
5.
Are the battery rocca separately ventilated ?
6.
With respect to both tna a-c and J-c cmergency power systems,
describe the elcetrical interlocks which prevent i=preper 1454 333
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operation (e.g., ento a fault) of the manual cross-connections betucen redundant banca.
7.
Identify any heat tracing circuits vital to th operation of the engineered safety features (e.g., any circuits to ensure that bcron rc=ains in solution). If so, describe and justify the design of the circuits and their power scurecs.
8.
Subnic a one-line diagram, similar in format to Figurc G.3, showing the assign:acnt of enginecred safety feature equip::ent to the erergency busos 1D and 12.
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