ML17194A216
| ML17194A216 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 10/19/1981 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Delgeorge L COMMONWEALTH EDISON CO. |
| References | |
| TASK-04-02, TASK-4-2, TASK-RR LSO5-81-10-035, LSO5-81-10-35, NUDOCS 8110210286 | |
| Download: ML17194A216 (5) | |
Text
.r-"'
- I
~._.-..:-
\\
-~*.;:. j
~
.. '\\..rl
- *. J October 19, 1981 Docket No. 50-237 LSOS-81-10-035
. *f1ft*~~:{t.: sI~Ab.e[::,~
Director of' Nutlear Licensing Commonwealth Edison Company Post Office Box 76?
Chicago, Illinois 60690
Dear r.{p. Abel:
SUBJECT:
SEP TOPIC IV-2, REACTIVITY CONTROL SYSTEMS - DRESDEN 2 Enclosed g**a c;:opy of our evaluation of SEP Topic IV-2, Reactivity Control Syste'if:iS;-~for Dresden 2. This assessment compares 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.
Your response within 30 days of the date you receive this letter is requested.
If no response is received within that time, we will assume that you have no comments or corrections. This evaluation will be a basic input to our review of your evaluation of Topic XV-8 and the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This assessment may be revised in the future if your facility des.gjgn is changed or ff NRC criteria relating to this subject are modified.before the integrated assessment is completed.
Enclosure:
As stated cc w/enclosure:
See next page
'N.RC FORM 318 (10-80) NRCM 0240 Sincerely,
. Dennis M. Crutchfield, Ch~1~f Operating Reactors Branch No.
Division of Licensing OFFICIAL RECORD COPY USGPO: 1981-33~960
Mr. L. DelGebrge*
cc
- z IL Isham, L incoln.:*&*:aeale*~ :::
Illinois Department of Nuclear Safety l 035 Outer Park. Ori ve, 5th Fl oar Sp~ingfield, Illinois 62704 Counsel ors... *a*t* -LaW * **-***
One First National Plaza, 42nd Floor Chicago, Illinois 6060l Mr. B. s.. Stepnenson.. _.. -. :.* :
Plant Su~erfritendent ** - * * **
Dresden Nuclear Power Station.
Rural Route #1
- Morris, Illinois 60450 Natural Resources _Def,ense Couricil 917 15th Str~et~ N. w~ '.
Washington,.~* C *... 20005.
U. S. Environmental Protection Agency
- -.Federal Activities.Branch.
- ,-:
Region v office-* * *- *
.ATTN: *Regional Radiation Representative 230 South Dearbo~n *street Chicago, Illinois 60604 Dr. Forrest J. Remick * * * *.. _ -
... 305. East Ha.rn.il ton Avenue
- State.College; Pennsylvania' 16801
- .. *
- The H~n6rat>1e Tam* Corcoran ---.-
U. S. Nuclear Regulatory Conmission
. United.States House of Representatives Resident* Inspectors OffiCe
. : -_. < Washington, D. C
~
20515 '- '***' :
Dresden Station RR #1
.~yS(C~ '*.
I
~... l *;
Mor~i~, Illinois 60450 Mary: Jo Murray *.
- '. t I *.
~ * :
~ (* !-
Assistant Attorney General Environmental Control Division..
188 W. Randolph Street Suite 2315 Chicago, Illinois 60601 Morris Public Library 604 Liberty Street Morris, Illinois 60451 Chairman Board of Supervisors of Grundy County Grundy County* Courthouse..
Morris, Illinois 60450 John F. Wolf, Esquire 3409 Shepherd.Street Chevy Chase*, Maryl and 20015 Dr~ Linda w. Little 500 Hermitage Drive Raleigh, North Carolina 27612
I.
SAFETY EVALUATION REPORT SEP TOPIC IV-2, REACTIVITY CONTROL SYSTEMS INCLUDING FUNCTIONAL DESIGN AND
---*.PROTECTION AGAINST SINGLE FAILURES DRESDEN 2 NUCLEAR POWER PLANT
. ~-..
'... ;. ~.;:.. *. :.
DOCKET.-NO. 50.-237 *.*.*
.... *= ~-
.l. *-*......
~--
~: ::. *.
- I'.
INTRODUCTION
~ '.
~,.,.
.. '~.'..... *.
The purpose of.this *evaluation is. to insure that the desigr:i basis for.
I
, I'..
the Dre.sden 2 reacti_vity control :systems is co~sistentwith analyses performed ~o verify_ -.that the protection.sys tern meets General Design_ Criterion
- 25.
Ge.nera l Design. Criterion. 25* re qui res that.. the. reactor protection. sys tern
- be designed to~~s~~re.. th~t specified acceptable fuel design limits are not exceeded for any. single malfunction of the reactivity control systems.
such a~ acci d~nt~~ ~ith;d~*~w~l.6f'.. ~6gt~~ l : r~<lL :*~ Reac ti vi tY t6riti-61 sys terns*
- . *. ~~ ::.
-~ <. *: :: ~--. *:. ". *:*..
- ~ s C"._-;:-.. *-.-.. -:*; :.-\\~ r:~.
need not be single failure proof.
However, the protection system must be
~.. : :.*
capable of ~ss~ring'~fia~ ~c~ept~b1~ fuel de~ig~ Ti~~ti.are not exceeded in the event of a single failure in the reactivHy control systems.
The review criterion,_coveted in this ev~luation, is addressed in Section II.
Review areas that are not covered, but are related and essential to the completion of this topic, are covered by other SEP topics addressed in Section III.. The scope of the SEP topics is defined in the "Report on the Systematic Evaluation of-Operatin.g Facilities" da.ted November 25, 1977.
This repor~ is limited*t~.:~he i_dentification of inadvertent control rod withdrawals and malpositioning of controls rods which may occur as a result of single failures in the control rod drive system.
-II.
REVIEW CRITERION The review criterion for this topic is based upon Se:tion 7.7, Part II of the NRC Standard Review Plan.
In the speci fie case of the reactivity
4
- ~..
- control systems a single failure shall not cause plant conditions more severe than those. for which the reactor prot~ction system is designed.
- *1:ic-::,** REL-ATED S*A,FETY TOPICS Th~ following listed review areas are not covered in this report, but
- ~.. :.
are related_and.essential: to the completion of this topic. These review
.areas are covered.~y other SEP topics as indicated below~
1.. A~alys~s of the consequences of control rod withdrawals and the
-** malposition'ing 'of contro:l rods which may occur* as* a r*esult of single fai1ures. i.n the e:l~c-~rical _circuits of the reac.t.i,vity __ c,o:nt~ol
. sys terns -a re covered by.SEP. To.pi c XV~8, "Con-tro 1. Rod Mi sop~ra tfon.
i.. ~
}~..
~-.
l :: *.. ".. :.!.. ~
~~ ~.>:. _(Sys tern ~a 1 f~nc ~.~gn. O!"_ o_p*er.~ to_~_ E:ro_r} II..
~o.,. :_.
- , *-.'~\\. -~ *i ~--
- 2. Araly°ses of re~2ti.vi-ty" ins,ertibns occurring as a result of -i°nadvertent.
reactor coolant pressute incre~ses ~~ temperature decreases are covered in SEP Topic XV-1, "Decrease in Feed\\'.Jater Temperature, Increase in Feedwater Flow, Intrease in Steam Flow and Inadvertent Openirig of Steam Generator Relief or Safety Valve;"
SEP Topic XV-3, "Loss of External Load, Turbine Trip, Loss of Condenser Vacuum, Closure of Main Steam Isolation Valve (BWR), and Steam Pressure Regulator Failure (Closed);"
and SEP Topic XV-9,-"Startup of an In~ctive Loop or Recir_culation Loop at an Incorrect Temperature, and Flow Controller Malfunction Causing I
ai Increas~ in BWR Core Flow Rate."
IV.
REVIEW GUIDELINES The purpose of this evalu~tion is to identify inadvertent control rod withdrawals and malpositioning of control rods which may occur as a result of single failure in the control rod drive system for the Dresden 2 Nuclear
, ~
~
f'J
' *e Power Plant.
- v.
EVALUATION
,~~:::;'j*~format\\~n ~a-S- 'p:rovided in Commonwealt~*~Ed~sbn~:'fe'.ti~t.*:_-ci~te.d Jurie'a,::i9af:
~~.:.. *. ; * ' ' * > *:. ' * *. * ":"
- .' ~~ ; ' -".
f *
[-
- ,..O.'
~
~ *: '. ~ * {
',."a~ * *,.
- c! :_describing s.ingle *failures.within the control rod drive system* which can
. :::-:' ~ause *control rod withdrawals and malpositioning of cont,rol. rods *:a*t the*
~-~ _;_~Dresde~ 2. Nuclear Po_~er ~lant. Also incl~ded ~a*s.a descrl-ption. of design features which limit reactivity insertion rat~s and rod malpo:s-itionings res_ulting from single failures. Based.upon. an audit review of the information
, *, :... ~
~: ~:..
~.
_:**: pro vi de_dby the _1_ i_c_ensee we conclude that the fol 1 owfng m~y occur as *a;
- result of single failures:
~..
l) A control rod may move more than one notch when a sin~le
.~*
- 1* **
I J
. no"tch.*movement. is commanded.
~:~.. _ 2) *.A control rod may drift in or out of the core.
This conclusfon is based upon the design of the control rod drive system which precludes operation of more than one control rod at a time (excluding reactor trip).
VI.
CONCLUSION
- The results of this topic wi 11 be used by the *staff in our review of the licensee report on Topic XV-8.
- '