ML17252B572

NUREG-0823, Supplement No. 1, Integrated Plant Safety Assessment, Systematic Evaluation Program

ML17252B572
Person / Time
Site:	Dresden
Issue date:	10/31/1989
From:	Office of Nuclear Reactor Regulation
To:
References
NUDOCS 9009240082, NUREG-0823, Suppl 1
Download: ML17252B572 (62)

Text

DRESDEN 112 CECo INTEGRATED PLANT SAFETY ASSESSMENT Rec'd W/ M~mo Dtd 9/18/90 9009240082

-NOTICE-THE ATI ACHED FILES ARE OFFICIAL RE-CORDS OF THE RECORDS & REPORTS

. MANAGEMENT BRANCH. THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED TIME PERIOD AND MUST BE RETURNED TO THE RECORDS & ARCHIVES SERVICES SECTION .

P1.,..122 WHITE .FLINT. PLEASE DO NOT SEND DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY PAGE (S) FROM DOCUMENT FOR REPRO-

• DUCTION MUST BE REFERRED TO FILE PERSONNEL.

-NOTICE-

NUREG-0823 Supplement No. 1 Integrated Plant Safety Assessment Systematic Evaluation Program Dresden Nuclear Power Station, Unit 2 Commonwealth Edison Company Docket No. 50-237 I Report U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation October 1989

AVAILABILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room, 2120 L Street, NW, Lower Level, Washington, DC 20555

2. The Superintendent of Documents, U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20013-7082

3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publica-tions, it is not intended to be exhaustive.

Referenced documents available for inspection *and copying for a fee from the NRC Public Document Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investi-gation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRG-sponsored conference proqeed-ings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regula-tions in the Code of Federal Regulations, and Nuclear Regulatory Commission Issuances.

Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies and reports prepared by the Atomic Energy Commission, forerunner agency to the Nuclear Regulatory Commfssion.

Documents available from public and special technical libraries include all open literature items, such as books, journal and periodical articles, and transactions. Federal Regi~ter notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.

Documents such as theses, dissertation~. foreign reports and translations, and non-NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draft reports are available free, to the extent of supply, upon written request to the Office of Information Resources Management, Distribution Section, U.S ..

Nuclear Regulatory Commission, Washington, DC* 20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copy..:

righted and may be purchased from the originating organization or, if they are American National Standards, from the American National Standards Institute, 1430 Broadway, New York, NY 10018.

NUREG-0823 Supplement No. 1

.* IntegrMed Plant Safety Assessment Systematic'~Evaluation Program Dresden Nuclear Power Station, Unit 2 Commonwealth Edison Company Docket No. 50:-237

• a l Report U.S. Nuclear*.:-Regulatory Commission Office of Nuclear Reactor Regulation

• October 1989

• ABSTRACT The U.S. Nuclear Regulatory Commission (NRC) has prepared Supplement 1 to the final Integrated Plant Safety Assessment Report (IPSAR) (NUREG-0823), under the scope of the Systematic Evaluation Program (SEP), for the Commonwealth Edison Company (CECo) Dresden Nuclear Power Station, Unit 2, located in Grundy County, Illinois. The NRC initiated the SEP to provide the framework for reviewing the design of older operating nuclear reactor plants to reconfirm and document their safety. This report documents the review completed by means of

.the SEP for those issues that required refined engineering evaluations or the continuation of ongoing evaluations subsequent to issuing the final IPSAR for Dresden Unit 2.

The review was provided for (1) an assessment of the significance of differences between current technical positions on selected issues and those that existed when Dresden Unit 2 was licensed, (2) a basis for deciding on how these differ-ences should be resolved in an integrated plant review, and (3) a documented evaluation of plant safety. The final IPSAR and this supplement forms part of the bases for considering the conversion of the existing provisional operating license to a full-term operating license. *

• Dresden 2 SEP, Supp. 1 ii i

-.-. CONTENTS Page ABSTRACT ...............*........... -. ............................... *. iii ACRONYMS AND INITIALISMS............................................ vii 1 INTRODUCTION .................................................. . 1-1 2 TOPICS THAT REQUIRED REFINED-ENGINEERING ANALYSIS OR CONTINUATION OF ONGOING EVALUATION ............................ . 2-1 2.1 Topic III-1, Classification of Structures, Components, and Systems (Seismic and Quality)......................... 2-1 2.1.1 Radiography Requirements ...... ~.................... 2-1 2.1.2 Fracture Toughness .................. :~************* 2-2 2.2 Topic III-2, Wind and Tornado Loadings and Topic III-4.A, Tornado Missiles.......................................... 2-4 2.2.1 Ventilation Stack.................................. 2-4 2.2.2 Components Not Enclosed in Qualified Structures.... 2-6

2. 2. 3 Roof Decks ...*.. ;.................................. *2-8 2.2.4 Load Combinations.~................................ 2-9 2.3 Topic III-4.A, Tornado Missiles........................... 2-9 2.3.1 Service Water System............................... 2-9 2.3.2 Diesel Generator Ventilation....................... 2-12 2.3.3 Exterior Tanks..................................... 2-12

• 2.3.4 Masonry Walls...................................... 2-12 2.4 Topic III-4.B, Turbine Missiles........................... 2-13 2.5 Topic III-5.A, Effects of Pipe Break on Structures, Systems, and Components Inside Containment................ 2-13 2.5.1 Jet Impingement on Target Pipe..................... 2-14 2.5.2 Broken-Pipe Impact on Target Piping................ 2-14

• 2.5.3 Detectability Requirements......................... 2-15 2.5.4 Criteria Implementation............................ 2-15 2.6 Topic III-6, Seismic Design Considerations................ 2-15 2.6.1 Mechanical Equipment............................... 2-16 2.6.2 Qualification of Cable.Trays....................... 2-16 Dresden 2 SEP, Supp. 1 v

CONTENTS (Continued) 2.7 Topic III-7.B, Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Design Criteria ......... . 2-16 2.8 Topic III-10.A, Thermal-Overload Protection for-Motors of Motor-Operated Valves .................................

• 2-17 .

2. 9 Topic V-5, Reactor Cool ant Pressure Boundary Leakage ***.:

Detection ................... -........ *.................. :r.:-" 2-17 2.9.l System Sensitivity ............................... *.:.. 2-17*

2.9.2 Seismic Qualification.............................. 2-18 2.10 Topic VI-4, Containment Isolation System ......... :.~ ...* ~. 2-I8 2.II Topic VI-7.C.I, Appendix K--Electrical Instrumentation and Control Re-reviews ......................... *... '. .... ;.. 2-I9

2. Il. I Breaker Adequacy ............................... . 2-I9

2. Il. 2 Disconnect Links .... : ............... *.* .......... r"-.: 2-20

2. Il. 3 Operation With Failed Battery~ ..................

• 2-20 2.11.4 Isolation of Class IE Sources From Non-Class IE Loads .............................. ; ........ . 2-20 2.12 Topic VI-IO.B, Shared Engineered Safety Features, Onsite. -

Eme~g~n~y Power, and Service Systems for Multiple-Unit*:**.

Fac1l1t1es............................................... .2-20 2.I3 Topic VII-I.A, Isolation of Reactor Protection System From Non:-safety Systems, Including Qualification of * '- *.

Isolation Devices........................................ 2-2I 2.I3.I Reactor Protection System Control Systems....... 2-2I 2.13.2 Process Computer ....... .-......................... 2-22*

2.I4 Topic VIII-3.B, DC Power *System Bus Voltage Monitoring and Annunciation.*.... *....................................

• 2.,.23 3 IPSAR TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL*

SPECIFICATIONS ....................* -.: ................ *......... . 3-1 3.I Topic VI-7.C.I, Appendix K--Electrical Instrumentation and Control Re-reviews .................................. : 3-I 3.2 Topic XV-I6, Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment*

and Topic XV-I8, Radiological Consequences of a Main Steam Line Failure Outside Containment................... 3-I 4 IPSAR TOPIC RESOLUTION CONFIRMED BY NRC REGION III OFFICE ..... . 4-1 5

SUMMARY

.............................. -... *.......................

• 5-1 6 REFERENCES .................................................... . 6-1 APPENDIX A--NRC STAFF CONTRIBUTORS AND CONSULTANTS Dresden 2 SEP, Supp. I vi

-* ACRONYMS AND INITIALISMS ACI American *Concrete Institute ADS* automatic depressurization system APRM average power range mon-itor -

A~ME: American Society of Mechanical Engineers BWR boiling-water reactor CE Co Commonwealth Edison Company CF.R Code of Federal Regulations DEi dose-equivalent iodine DG diesel generator ECCS . emergency core cooling system El&C electrical instrumentation and control EPRI Electric *Power Research Institute FSAR final safety analysis report FTOL ful~-term operating license**

f yi e l.d stress

• Y ..- .

GDC general design criterion(a) '

GE General Electric Company HPCI high-pressure coolant injection IE Office of Inspection and Enforcement IEEE Institute of Electrical and Electronics Engineers IGSCC intergranular stress corrosion cracking IPSAR Integrated Plant Safety Assessment Report LCO limiting condition for operation LPCI *

• low-pressure coolant injection LST lowest service temperature MCC motor control center MCF maximum credible fault MOV motor-operated valve.

NOT* nil duct.i 1i ty temperature .

NRC U.S. Nuclear Regulatory Commission POL provisional operating license RCPB reactor coolant pressure boundary RG regulatory guide

• RPS reactor protection system Dresden 2 SEP, Supp. 1 vii

SEP SER SRP SSE STS sws Systematic Evaluation Program safety evaluation report standard review plan safe shutdown earthquake Standard T~chnical Specifications service water system TER technical evaluation report TMI Three Mile Island TOL thermal overload TS technical specifications I ... ~.

UFSAR* . * , updated final safety ana lys i's report USl

• unresolved safety issue I*

, :' j

,1.

I .

T '* "'

~ ~ .

l * *~

• Dresden 2 SEP, Supp. 1 viii

INTEGRATED PLANT .SAFETY ASSESSMENT REPORT SUPPLEMENT NO. 1 SYSTEMATIC EVALUATION PROGRAM DRESDEN NUCLEAR POWER STATION, UNIT 2 * *  ;.. :",

1 INTRODUCTION The U.S. Nuclear Regulatory Commission (NRC) initiated the* systematic Evaluation P~ogram (SEP) to provide the framework for reviewin~ the designs of-olde~ operat-ing. nuclear power plants to reconfirm and document their safety.. The review pro-vides (1) an assessment of the significance of differences between current tech-nical positions on safety issues and those that existed when a particular plant was licensed, (2) a basis for deciding how these differences should be resolved .

• in an integrated plant review, and (3) a documented evaluation.of plant safety.

The NRC published initial results of the SEP review of the Dresden- Nuclear Power Station, Unit 2, in the final Integrated Plant Safety Assessment Report (IPSAR)

• in 1983. The review compared the as-built plant design with current review criteria in 137 different areas defined as "topics. 11 During the review, 49 of the topics were deleted from SEP consideration because (1) the topic was being reviewed in other programs (unresolved safety issues or Three Mile Island Action Plan tasks), (2) the topic was not applicable to Dresden Unit 2, or (3) the items to be reviewed under that topic did not exist at the site.

Of the original 137 topics, 88 were, therefore, reviewed for Dresden Unit 2; of these, 54 met current criteria or were acceptable on another defined basis.

From the review of the 34 remaining topics, certain aspects of plant design were found to differ from current criteria. Of these 34 topics, 7 were found to meet current criteria or were acceptable on another defined basis after modi-fications were made during the topic review and were, therefore, not evaluated in the integrated assessment. The remaining 27 topics were considered in the integrated assessment of the plant, which consisted of evaluating the safety significance and other factors of the identified differences from current design to arrive at decisions on whether modification was necessary from an overall viewpoint of plant safety. To arrive at these decisions, engineering Judgment was exercised and the results of a limited probabilistic risk assessment study were employed.

In general, the staff's positions resulting from the evaluations of the integ-rated assessment fe 11 into one or more of the fa 11 owing categories: (1) equip-ment modification or addition, (2) procedure development or technical specifi-cation changes, (3) refined engineering analysis or continuation of ongoing

• evaluations, and (4) no modification necessary. Table 4.1 of the IPSAR summa-rizes the staff's integrated assessment positions and documents the licensee's agreement with those positions as of February 1983.

For those positions classified above in either of the first two categories, the IPSAR lists the scheduled completion dates agreed upon by the staff and the licensee. NRC's Region III office verified the implementation of these positions as noted in Section 4.

Dresden 2 SEP, Supp. 1 1-1

For those positions in the third category, the licensee has provided the results of the ongoing evaluation to the staff for review, and this supplement provides the staff's evaluation of these issues. For Dresden Unit 2, 25 issues under 14 SEP topics required refined engineering analysis or were continued under an ongoing evaluatio~. The evaluation of these issue~ or *their status.at issuance of this supplement is discussed in Section 2. In those c~ses i~ which.an~ly~es are continuing, the staff's evaluation identifie~ the analyses to b~ ~erf6rmed anq-the acceptance criteria that will be used to.determine the. optim~m .Plant modifications.; . * . . * *r. *.*

. , . . . .* . . .  :.c; r: i:: . .

This supplement summariies the status of all actions to be implei:n~n~~d a.s a .. ; .

,. re$ult of the SEP. review and presents this summary in. Table: :2. l.. A..flY pr:eimple-meritation staff revi~ws required for these ong.oing analyses*,.aften~~:J:hi.s ~upple-*

_ment has. been issued, wil,l be summarized in individua.l s~fety evaJ,~~tion, reports as ..t~e,an~lyses are co.mpleted.* * ._ . *, * . . . . , .. , *' ...... ;*; *; .. * .

• Dr!=!sden' .unh 2 ; S: on:~ o( the' four remaining pl ants being, .revi ~w'ed :~*~*de~_.-the ,:SEP, t~~-~ ~as r:iot .re.ceived a full-term .operati_ng Hcense .(FTOL) ... '.A: safety."evaluat:ion report~ .CS.ER) to support the ~onv~rsion of the prov:isional ope.rat.i119 1 license (POL) to an FTOL will be prepared. . ..

.' ~ *. '*\.,: *. .1*,*. \ ,,*

• ,, .\.._ .

• , 1

• .,I '*

' ) !' " . ~ : ,

f* ., *

i',' ,,, .. .. *. ~ . , . -..; ~~

. ; .*:*.1 -

,1 ,'

.I:.*.!

i;- ..... *-I~* J. ** ': ' ' ' * *I

• . \ ', I* * .. '.. :. \ *

.~ .\ ' . ,'

.. 'l I I~: C

., ~I' j ,

_*:1.=*

Dresden 2 SEP, Supp. 1 1-2

2 TOPICS THAT REQUIRED REFINED ENGINEERING ANALYSIS OR CONTINUATION OF ONGOING EVALUATION The litens~e (Commdnwealth Edison Company or CECo) ~ubmitted*an evaluation for each of the issues that required refined engineering analysis or that were part of a continuing ~nalysis. The staff reviewed each evaluation-and classified it ~~ (1) m~eting c~rrent criteria, (2) acceptable on another ~efined basis, (3) ~~quirind*corrective action; or (4) requiring further an~lysis. Factors* con-sidered in" reaching these conclusions include the perceived safety significance of* the di ffererice from current 1i cens i ng c*ri teri a, a qua li tat i ve asse~sment. of the financial and exposure costs to make a modification and, to a lesser e*tent, implementation impact and schedule. In evaluati"ng these issues, the staff also.

cohsidered*any applicable risk perspectives developed for the integrat~d ~sses~

nient and* descr-i bed in the* IP SAR, and re 1ated corrective. actions* proposed by the licensee as part of the *ihtegrated assessment bras a result.of the f6llow-on

• evaluations.

Each of the"outstanding issues is discussed below. Each evaluation references*

the more 'detailed licensee evaluation and staff topic evaluation. References for the pertinent safety evaluation report (SER) and technical evaluation report, (TER), where applicable, appear in Section 6 of this first supplement to the IPSAR report (NUREG-0823). Appendix A contains a listing of the staff contributors and consultants.

The final status of each of these issues is summarized in Table 2.1 along*with the status of all SEP issues for Dresden Unit 2.

• stems Seis-2.1.1 Radiography RequirementsSection III of the American Society of Mechanical Engineers (ASME) Boiler. and Pressure Vessel Code requires that Categories A, B, and C weld joints be radio-graphed. Furthermore~ ASME Code,Section III, 1977 Edition, requires that weld joints for Classes 1 and 2 piping, pumps, and valves be radiographed.

In IPSAR Section 4.2.l, the staff stated it was in general agreement with the current requirements except that the following items were not addressed:

(1) ASME Code Class 2 vessels built to Class C requirements and containing Category C joints, along with the examination technique employed, should be i dent ifi ed.

(2) The actual examination given to the recirculation system pump casing (this is an ASME Code Class 1 component built to Class C requirements) should be described.

The staff identified the following components for which information on radiography requirements was needed:

Dresden 2 SEP, Supp. 1 2-1

• emergency system isolation condenser

• low-pressure coolant injection heat ~xchangers .

• reactor. shutdown *cooling sys tern . hea:t exc.han.gers

• recirculation system pump casing *

• In the Apri 1 20, 1987 .1 etter, the* 1i cen.see' provi.ded a copy of the. ori gi na 1 Genera 1 Electric speci fi cation for. each comp()rient. Sub sect ion C of ASME Sec-tion III (1965 Edition) descr~bes the'requirements for Class t ve~sels. P~ra-*

graph. N-211l ;sta~es that requ*irements of Section VIII qf the Code shal.l apply.-

The .radiography required on ea:ch of 'the. subject ASME. Section IIT (1965 E~ition)

Class 2 vessels was defined. On the. basis of G.e.neral E:lectric Spe.c.ificatio~ .*

21 A1208, radiography was required for the recirculation system pump- casing in accordance with ASME .Section III, Paragraph N323., 1965 Edition. Tn.e technique for radiography was to be in accordance with Paragraphs N624.2 through N624.7 of.this .code. After-at.least one .solution .he.at.trea.tment, the. pr~~sure-conta i ni ng casings were to be radiog~ap~ed. for. th_e f.i nal ti me. .

• In the December 9, :198.7 lett~r, the Hcensee provided proposed revisipns to the updated Dresden FSAR describing the . radiography that was required.,;, The staff evaluated the additional information provided by the licensee and determined the General Electric specifications; an.d* :t.h.e mar:tufacturer' s *data. rep_o,rt or the .

certific~te :of shop inspection provide :the.-.infqrmatfo~...defin_ed in _,N,UR;q-0823_...

T~er.ef.ore_, .the staff .concludes that.the ,tw.o. letters submitted by the~*licen,see.

resolve the issue on radiography requirements* and that. additionaf_correcti*ve ~-"

measures are not required;

.Fract_~r~ Toughne.ss

2. L2 ,.. .J: ..

J.l). rSect,ion 4.*2. 2 .of._the I_PSAR, thel_staff ~stated th.a~ the it:censee).as not pro-:.

vi_ded the .actual requirements :imposed qn t~e test .r.esults requested in. the . .. .

staff's .draft evaluation.forwardedby.le_tter dated March 9,*1982 to determine if the components: identified m~et ,~9re recent fracture ,toughness reqyi rements >:

The SEP plants, i.ncluding Dresden .Unit 2, were .generally.designed.~nd'cohstructed between the late 1950s and the Jat~ 1960~;. They were .designed. according to_ .

codes and criteria in effect at that time; however, since then, the codes and criteria have been revised to jncorpor~te the.results~{ .addition~l research . .

Thus,-._earl_ier*plants may .have. bee~ design~d .~ocording .to *criteria a11d codes nq.

longer accepted by t.he NRC . .-The purpqse *Of SEP Topic III-;l is to ~ompare the classification ~f. structuresl:syste~s and.compon~nt~ of the as-built pJants to the requirements in later editions *of the,,ASME Code. The .. staff reviewed the earlier plants to.determine whether thefr~mat~rials met the fracture toughriess requirements in .the ASME Code, Section. 1 IIJ~ 1977 Edition, Summer. 1978 Addenda ..

The staff:determined that fracture t()ughnes~,requirements could differ signifi-cantly between the.as-bui]t,Oresd~n.l)nit;2 plant and*~he r:-equirements i.n later editions of the ASME Code. In addition, the licensee was asked to determine whjch.~omponents could need.impac:t*testing,to m~et the fr.actur.e :toughness .. _

requirements of the*later.ASME Code editions.

In a letter* dated.July° 16,.1982,_ th~ 'iicensee' determined that impact testing could be required to determine the fracture toughness of the following components:

'

• 4*.

• .core* spray system P,ump casing ..... * ' *

• low-pressure coolant injection (LPCI) pump casing Dresden 2 SEP, Supp. 1 2-2

• LPCI heat exchangers--shell side

• high-pressure coo 1ant inject ion (HPCI) pump casings * "'

• HPCI piping, fittings, and valves with *nomihal pfpe diameter :greater than 6 inches *

• • condensate/feedwater system piping from reactor vessel to outermost

• • • containment.isolation valve * * .*' .*

• **maih s.team sy:stem pipi.ng, *valves*, and fittings... . .. .'

• f. **

• i .,* . * . * . . . ..  ! .

In Enclosu'rt(.5 to a January 19, 1983 letter, the staff indicated to the licensee that 'CO!llP HJrice with .the fractur(;! toughness requirements of 1ater edit i ohs of ....

the ASME Co'de>could be demonstrated by one* of the following: * * ** * " : .:

.

• O:~. . .. .

(1)

• .t ~j g

prqvi dfri.. test results 'that m~et the *ASME Code *requirements *

(2) determfr{ing that the component's lowest Sf;!rvi ce temperature ( LST) *is: *h~.Qh

• enough to exe~pt the materi~ls'*from testing * *

(3) deterinfnlng' that the tompo.nent' s failure will not :result in 'unacceptable:*,'.

consequences .:: ** * * * * ' *: * * * ** * * . * ., .. *. *: 1

' . . . :'i;.. . . . . . ... '.. ,

In lettersdated 'April 20,-'1987 and Jan,uary 6; 1989,: the* licens.ee** provid~d :. ... i;*

i nform'afi on* to* demonstrate' that' *the compon*erits

• i,dent if fed in: its* Ju.ly 16-/ 1982*

• letter wolild'irieet the*fracture toughness *requirements 'of later editions of the.

ASME Code. *:; . :-. ' . :,_**.... . .r . . ' .. .'

The core spray pump casings, the LPCI pump casings, and the HPCI pump casings .

• were designated as Class B Quality *Group and were constructed' of.carbon steel

• A-216, Grade~CB material. According to the ASME Code, these components wo~ld be exempt from *testing, *'if the LST exceeds '60°F; The core spray and* LPUl *pumJ?s were designed "for'* a* 'temperature range *from 60°F*to.I65°F, and normally oper~te:

at app*roxfmately '95°F; The HPCI pumps *were de'signed for a temperature range ...

from 40°F *to'140°F,:*an*d normally operate around;95°F. *.Although' the lkensee

• has not determined the LST, the design and normal operating temperatures .. for these ~omp9nents indicate that the materials will meet the intent of the materi~l ~~sting ~xemption requirem~nts in t~e. ASME Code.

The shell *~ide of the: LPCI heat exchangers were designated as Class *c Quality *~

Group ~~d ~~te tonstructed*of carbon steel A-212, Grade B material.* 'Accordtna_

to NUREG.:.0577, 11 Potentia.1 for' Low Fracture Toughness and Lamellar Tearing*on .*

PWR Steam. Generator arid React.or' Cool ant Pump Supports, II the 90% *confidence *.' ..

upper'bouhd NOT (nil ductility temperature) for this"material would be 77°F ......

On the b~~is of latef ASME Cod~ requir~ments for this class of components, .th~

mat~ri~Ts ~~*the' shell side of th~ LPCiihe~t exchangers would be exempt ftom*!*

• testing if"the* LST exceeds*77°F*. The licensee indicates that the operating . .-

temperature of. the she 11 side of ~he LPCI he~t. exchanger is* about 40°F. * *: -~ ....

' . .. . . . . , . . *r .; .

The HPCI *piping, fi tt i hgs, and va 1ves. were designated as Cl ass B Qua Hty *Group .1 and were* constructed of carbon steel* A-106, Grade B 'material. Fbr this ~lass

• component, the materials are exempt from testing if the LST exceeds 150°F.

The license~ *iridicates ' .

th~t the lST for these .

components exceeds 150°F.* * ~

'\

The feedwater system piping from the reactor vessel to outermost containment isolation valve and main steam system piping were designated as Class A Quality Dresden 2 SEP, Supp. 1 2-3

Group and were constructed of carbon steel A-106, Grade B material. According

• to later editions of the ASME Code, Class A components are not exempted from testing.

• Howevef,* according to Section 6.2.1 of the Dresden Updat.ed_'Final Sa,fety Analysis Report (UFSAR), the emergency core cooling system (ECCS) is desi~ned to prevent the melting of fuel cladding for any mechanical failure of the primary system up to and including a break equivalent to the largest primary system. pip-ing. In addition, the licensee indicates that except for hydrotesting, the LST for these components is 150°F or greater. At 150°F, these materials will behave in a ductfle .fashion and will not. be.subject to brittle fracture. Hence,. based

' on the LST for these components and 'the makeup capability of ECCS, _the. f eedwater system piping f~om the reactor vessel to outermost containment isolation valve and main steam piping have adequate fracture toughness.

The staff has concluded that based on the information provided by the licensee, all components, except for the shell side of the LPCI heat exchangers, have adequate fracture toughness~ .The materials on the shell side of the* LPCI heat exchangers will have adequate fracture toughness if their LST exceeds 77°F: In a letter dated May 1, 1989, the staff requested.that the.licensee determine whether the LST exceeds 77°F for the shell side of the heat exchanger. The li-censee was also asked to identify (within six months of the date of the letter)

(1) the operating conditions when it does not, (2) the. LST*during these operat-ing conditions, and (3) the* design changes necessary so. the LST exceeds 77°.F. *.

Tornado. Mi ss*i 1es 10 CFR Part 50 (GDC 2.) as implemented by.Standard Review Plan (SRP} Sections 3.3.1 and 3.3.2.and Regulatory Guides (RGs) 1.76 and 1.117, requires that the plant be d~signed to withstand the effects of such nat~ral phenomena as wind and tornadoes, including tornado missiles .. **.

Issues relatiog to wind and iornado lo~dings and tornadD missiles were identified for further evaluation in the IPSAR. Each of these issues is discussed below.

The issue of *load. combinations including.wind loads is discussed in Section 2.6 of this supplement, "Design C.odes, Loads, and Load Combinations."

2.2.1 Ventilation Sta~k The vent iJ at ion stack damaged by tornado winds co.ul d, in turn, damage safety-re lated structures. As a resylt of such damage, its capability under tornado loadings, and thus the possible risk, had to be assessed. In its letter dated November 22, 1982, .the..licensee concluded that the stack will 11 fail 11 if subjected to tornado windspeeds greater than 210 mph. This analysis is based on American Concrete Institute (ACI) ultimate strength methods using a reinforcing steel*

stre,ngth c.orrespondi ng to a .yield stres~. ( f y) of 40 ks i.

In the same letter, the licensee also concluded that the stack will adequately survive tornado windspeeds up to 255 mph, based on an analysis that uses ACI ultimate strength methods and-a reinforcing steel strength equal to its ultimate strength. From SEP Topic II-2.A, the licensee found that the probability of ..

tornado windspeeds exceeding 255 mph is approximately 2.35x10- 6 per year. Con-sidering that the impact area of a falling vent stack damaging buil~ings or Dresden 2 SEP, Supp. 1 2-4

eq~ipment is ~pproximately 36° out of 360°, the licensee concluded that the ... 1 probability ofa failed stack damaging buildings or.equipment is appr,oxi.mately:

2.35x10- 7 per: year'. On the basis of this.probability, the lic.ensee ~onC:luded

• *~hat m'odifi ~affo'ns are. not warrar:ited.

• j

• ' * * * ... * . *. ** ~

The staff' concludes that the proba~ility of. exceeding the windspeeds of

. i nte'rest. ar¢'; as fo 1i ows: ' .

Probabi .1 i ty o1\ exc*e~di n'g NRC ~*i Q.dspeed/yr. .'

,... : 210 ~* . .*

5xi0- 5

• '-1 255 8xi0- 6 .

'!.J

_***~ :~.' * . *, . ' 'I, * !

These probabilities -are based on the staff's estimates for the-Dresden site~

The probabi:.lities presented by the licensee correspond to expected values (not* upper:-95th* percenti.le) which are. contained in. the* November 22,

• 1982 .. * *

,letter. ~*\ **~: _,* * ..

' . ' .. ; ~

.~ ..

The staff1°with assistance'fro~:the Franklin=Research Center~ has independ~ntly analyzed .".the.:-stack* using .. the ;ACL u.ltimate* strength method *.for-. steel 'yield :. . **

• stresses of 40 ksi and 50 ksi and concludes that the limiting windspeeds and assoc.iated:,pr.obabi-1 ities based* om NRC staff estimates are' those .presented below.

The -TER *refe.renced in- this *evaluation** provides* details* of* the stac'k analysis .

.. . l

' * * **! *,, P.robabi l ity of

. '.*. :'* _, *.**1 *)

exceeding* NRG*

Analysis Limiting windspeed (mph)~ wi n'.dspeed/yr* *.* * **

• Ultimate strength des*ign* :

6x10-~

• 1.

(fy=40 ks,;) ... :* ' 189 Ultimate strength design (f =50 ksi) 199 5x10- 5 The current. code for us.e in stack design (ACI 307-78) specifies the use of working* stress, design l:>~cause there are little experimental data on hollow concrete-..cy'1inders1. However, discussi.on with members of the ACI-307 Code

.committee~indicates. a growing acceptance of ultimate st~ength design methods~'

for stacks *based. on testing which has been. performed* at the University of '

• Michigan. The issue of concern related to* the Dresden stack is actual stack collapse upon adjacent structures a.nd not such items as deflection: and cracking 1

normally associated with service lo~d requirements. For these reasons, the staff h.~s.used ultimate strength .. methods*in the* analyses ... ,

• The staff: concludes that the probability of the stack collapsing is extremely*

1ow based on the ca 1cul ated strength and* the associated wi ndspeed probabi,l'i.ty.

• Given that: structures and components wi 11 be .damaged only *if the stack fai l*s .; n specific directions, since it is not .completely surrounded.. by safety-related*

I Dresden 2 SEP, Supp. 1

structures, the probability of damage to riearby structures is even lower.

Should some p 1ant structures be damaged, the p 1ant might st i 11 be shut. down by using other systems. Therefore; the staff concludes that modifications are not warranted.

2.2.2 Components Not Enclosed in Qua~ifi~d Structures In the lPSAR,the sta-ff concluded that.the licensee should evaluate the effects of tornado winds on comp~nents not enclosed in qualified structures} thi~ item is closely related to ~taff concerns on tornado missile protection -b'ecause such coniporients are also vulnerable to being damaged by tornado missiles? -In* order to adequately address this i tern, damage from both wind and missiles '.:mus~ be **c:on-5 i dered. Tornado wfod effects are discussed here and missile *effec'ts :are diScussed in Section 2.3.of this evaluation. *' r*::

In the November .22, i982 letter; the 1i tense*e discussed this item in the context of effects of failure* of vulnerable components on safe-shutdown capabi'l ity.

• The condensate and deminera'lized water tanks are located outside ana thu's may be damaged by a tornado wind. Should these tanks fail, the HPCI syste.m c_an take suet i cin from the suppression poo 1 and the i so 1ati on condenser can b1if supp 1i ed makeup water from the firewater system neader by either the servi-ce wate*r* pumps or the emergency di es~ l fi'rewater pumps.

• The supports for diesel generator exterior intake, exhaust, and exnaust silencer are also exposed . . The Tkensee's intake and exhaust support ca*paci ti es for the di ese 1 generator and-' probabfl it ies of exceeding expected wtndspeed ~lorig with the NRC estim~t~ are.as follows:

Tornad-o Probability of Probability _ci*f winds peed exceeding expected exceeding NRC Support .(mph) windspeed/yr windspeed/yr Unit 2 exhaust silencers 105 9xl0- 5 5x10- 4 Unit 2/3 swing diesel exterior intake 170 2xl0- 5 9x10- 5 Unit 2/3 swing diesel exhaust silencer 200 9xl0- 6 5xio- 5 Should the intake and/or exhaust of diesel generator (DG) 2 and DG 2/3 become damaged so that both diesels become inoperable, hot shutdown could be accom-plished using the isolation condenser or th~ HPCI system, depending on the supporting equipment that is available. Assuming both diesels are inoperable because of intake and exhaust damage, the plant can be shut down safely in the following ~anner:

(1) The 'isolation condenser can i.nitially be used for cool down. The isolation condenser provides reactor core cooling if the reactor becomes isolated Dresden 2 SEP, Supp. 1 2-6

from the main condenser. This system is initiated from the 250-V de reac-tor building motor control center No~ 2 (bus 2A) and bperates by natural circulation; the system does not require ac power for operation. Cooldown can continue in this manner for approximately 20 minutes, at which time shellside water on the i~olation condenser.will boil 6ff and makeup water to the shell side of the condenser will be required to remove more h~at.

Table 2.2, extracted from th~ licensee's J~ne 12, 1984 submittal, .describes sources of .makeup water for the isolation condenser. Although a number of sources exi~t, the licensee has not demonstrated that structures housing the wat~r ar.e resistant td tornado winds; also, no source of makeup water

• is completely protected from tornado missiles since the tanks and diesel-driven fire pumps are. vulnerable (th~ pumps required to transfer makeup water to the condenser require ac power which is assumed to be lost as a result of DG 2 and DG 2/3 intake and exhaust stack failure). The licensee has calculated a probability of a tornado missile strike to the Unit 2/3 cribhouse to be 2.3xl0- 6 per year.* Actually striking and damaging the diesel-driven firewater pump would result in a lower probability. Thus, from these.,probabilities, it is unlikely that a diesel-driven firewater pump from Uriit 1 or Unit 2/3 (both of which feed into one common header to provide water to Unit 2/3) will be damaged by tornado missiles. No struc-tural capacities were calculated for this portion of the cribhouse.

• This portion of the cribhouse is not expected to supply a high resistance to tornado wi~ds. Failure of the surrounding structures may damage these pumps or leave them vulnerable to tornado-entrained items such as dust and debris which may affect pump operability. Should any of these water sources .and associated.transfer pumps not requiring ac power remain operable, cooldown via the isolation condenser can continue until hot shutdown is achieved.

(2) As noted above, the sources of makeup water to the isolation condenser.may be affected by the tprnado .. In this case, hot shutdown could be accom-plished using the HPCI system. After the existing shellside water boils off in the isolation condenser, th~ ~PCI system could inject water into the reactor from the torus. Water from the primary system would discharge to the tor.us through relief valves which have remote manual operating capa-*

bility using 125-V de power. After. several hours, it will be necessary to cool the torus water via the LPCI heat exchangers. This would require (a) containment cooling service water pumps, which are protected from missiles ~s discussed in Section 2,3~1, and (b) ac power for the pumps.

The ac power can be supplied from offsite (if restored) or by the diesel generators (if repaired). Since the postulated missile damage is to the intake, exhaust, and exhaust silencers, and not to the diesel generators themselves, it should be possible to repair these components. The licensee estimates that the damaged components could be repaired or removed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> so that the diesel generators would be operable. The necessary equipment, for example, cutting torches, and cranes, are available at the site. Once the diesels are repaired, cold shutdown could then be accom-plished using the automatic depressurization system (ADS), the LPCI system, and the LPCI .containment cooling system. The staff will confirm that the l i c.ensee has the necessary equipment on site to repair or remove the damaged componen~s. of. the diesel generators.

• The initial siages of this scenario up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> are similar to those rif a station. blackout. The licensee has stated that the assumptions in this Dresden 2 SEP, Supp. 1 2-7

shutdciwri scenario, with regard td operable* equipment,* are*tonsistent with those in its station blackout submittal dated April 17, *1989. On this basis,' the *Staff concl~des *that there is r~a~onable assura~ce that Drss-den 2 will be able to attain safe shutdown in the event a tornad6 caused damage to components not enclosed in qualified structures. If the staff determines that port~ons of the station blackout submittal that affects equipment taken credit for in this scenario are unacceptable, the licensee wi 11 be required to modify its shutdown scenario for tornado loadings and nii ss i 1es tci be consistent with. the po~ it ion accepted* by the* staff for

• • station blackout. *

• 2.2.3 Roof Decks In IPSAR Section 4.3.4, the staff noted that th~ capacities of the built-up roof decks had not been evaluated. 'The licensee, iri the November 22,.1982 letter, gave the following windspeed capacities for the concrete roof slabs in the teactor and turbine buildings:* *

.

• Allowabie uplift Ultimate.uplift Probability Probability of exceeding of.exceeding expected * . expected ..'. '

Pr.~ssure. windspeeci wi. ndspeed/yr  : Pressure. Windspeed *winqspeed/yr T 118 psf 250 mph 1. 7xl0- 6 130 psf 270 .mph. * . 1. Ox10~ 6 .

. NW. 73 p*sf 200 mph . 0.9xl0- 5 * .. * ..

T=-~ornado; NW~ normal wind.

These capacities were determined assuming the siding is not blown ~ff* by the wind. The licensee further concludes that if the building is considered to be open, that.is, the internal pressure has been. relieved due to losscof siding, the ultimate capac it¥ would corr.espo_nd to a 320-mph ~ornado wi ndspeed, with .a probability of 2x10- . per yea~. On the basis of the capacities pre.s~n.ted above by the_ licensee, the.staff.calc~lated the following windspeeds*and the associated probabilities using the staff estimate curve: These windspeeds_were calculated.

on the basis of capacity for pressure drop as compared to the licensee 1 s assessment for dynamic pressure.

Allowable ~plift Ult i'mate uplift Probability of Probability of exceeding NRC *'exceeding NRC PressLJre.'

• Wi ndspe.ed wi ndspeed/yr

• Pressure * . Wi ndspee*d * ** .*'wi. n*dspeed/yr TPD 118 ps f

• 151' mph 2x10- 4 130 psf i59 mph * .* 1x10- 4

~WP* 73 psf

• 202 mph 5x10- 5 130 psf 269 mph * *' SxI0-6 TPD = tornado pressure drop; TWP = tornado windspeed pressure.

Dresden 2 SEP, Supp. 1 2-8

Safe-shutdown equipment is not located immediately below either of these roofs; it is located a floor below. Thus, even if the roof decks .fail, capability to attain safe shutdown is not likely to be affect~d. *This fact, in conjunction with the probabilities of exceeding the tornado windspeeds, leads the staff to conclude that no modifications are ~arranted.

2.2.4 Load Combinations The lic;ensee states that pip-e re~iction -loads *and thermal loads-.were considered in combination with wind loads under SEP Topic III-7.B. These are the operating loads specified to be combined with wind load. This aspect is addressed in the staff's evaluation of the licensee's response to Topic III-7.B in Section 2.7 of this report.

2.3 Topic iII-4.A, Tornado Missiles (NUREG-0823, Section 4.5)

In IPSAR Section 4.5.l, the staff noted that the cribhouse was not fully pro-tected from tornado missiles. In particular, the staff was concerned about the service water system that provides cooling for the control room ventilation system and for the auxiliary electrical equipment room ventilation system.

2.3.l Service Water System In the June 28, l982 SER and in Section 4.5.l of the IPSAR for Dresden Unit 2 (NUREG-0823)* the .staff concluded that the station cribhouse (intake structure) did not ad~quat~ly protect from tornado missiles that part of the service watef system that pro vi des coo 1i ng water for the contro 1 room ventilation* system and the auxiliary' electrical equipment room ventilation system. The staff further concluded in NUREG-0823 that backfitting was not required because of the licensee's commitment to address tornado missile protection as part of the TMI Action Plan, Item III.D.3.4, "Control Room Habitability." The* staff's fin.al evaluation of TMI Item III.D.3.4 concluded loss of the seismic water system and subsequent loss of control room ventilation would not prevent safe plant shutdown.

The loss of the auxiliary electrical equipment room ventilation system*, which houses safety~related equipment, was not addressed by TMI Item III.D.3 . '4. Since this equipment room houses safety-related electrical equipment, the licensee performed a probab*ilistic analysis* to show that the effect of loss of. service water as a result'of tornado missiles need not be considered because of its **

1ow probabi 1ity :*

By letters dated February 3, 1983 and June 12, 1984, the licensee provided the results of a probabilistic analysis of tornado missile strikes on the cribhouse at Dresden Unit 2. The methodology used in the analysis was developed by the Electric Power Research Institute (EPRI). Specifically, the analysis references EPRI reports NP-2005 and NP-768/769. Two independent staff consultants reviewed these reports.and each consultant prepared a TER. The consultants reported favorably, recommending a few areas in which additional conservatism might be used. The staff prepared an SER dated October 27, 1983, regarding the EPRI reports. On the basis of the 1icensee 1 s February 3, 1983 submittal , the con-.

sultants 1 *eval~ation~ of the EPRI reports, the completion of TM~ Item III.0~3.4, Dresden 2 SEP, Supp. 1 2-9

and the completion of the Appendix R alternate shutdown review, the staff has completed its evaluation of protecting the cribhouse at Dresden Unit 2 from tornado mi s.s il es.

The licensee's analysis, using the EPRI methodology, resulted in an estimated probability of a tornado missile striking the cribhouse of approximately 2.3x

~0~ 6 per year compared with staff's estimate of 1.1x10- 5 per ye~r .. The staff's criterion is. that tornado missile .protection need not be provided if the proba-biHty of exceeding the radiation dose ~uideli.nes of .10 CFR Pa.rt .100 due to tornado missile damage is less than 10- per year. The probability. estimated by the licensee is only for a cribhouse stri.ke. The probability that the service water system would be damaged to the extent of inoperability, plus the probabil-ity that loss of cooling water to the control room and auxiliary electrical equipment room ventilation systems resulting in a 10 CFR Part 100 release, is le.ss than the missile strike probability.

. . . . *. ~

A further reduct ion in probability o.f, exceeding* the 10 CFR Part 100 guide 1-i nes is immediately realized if credit is given for the east turbine room Ventilation system, which can be used to cool the auxiliary electrical equipment room, as discussed below.

Control Room. Ventilation System *

• For loss of control room ventilation due to loss of service water, the staff

• conclµdes that 10 CFR Part. 100 radiation .. dose guidelines would not be. exceeded;

.this conclusion is ~ased, in part, on the staff's previous conclusions in the*

final evaluations of TMI Item IIL0.3.4 and.Items !ILG, and !ILL; of Appendix. R to.10 :CFR Part 50. The licensee, as part .of modifications. to meet Appendix R to:10 CFR Part 50, has provided alternate shutdown capability in the event of a fire in the ~ontrol room . . The .alternate* shutdown method assumes service ~ater is ava*ilable. However, the staff's safety evaluation for, SEP Topic VII-3, 11 Sys-tems Required- for Safe Shutdown, 11 desc:ri bes ..a method for shut.down .when no sta..,.

• tion service water is available . . Thi*s method relies on the automatic:: depressur-faat ion system; low-pressure cool ant *injection, LPCI containment cooling heat

• exchangers, and the containment cooling service water system. All of these systems have been protected from tornado missiles and would be avail.able for al.ternate shutdown outside the control .room, assuming (1) no control .room fire and (2) availability of the di es el generators, which is further discussed in Section 2.3.2 of this IPSAR supplement . . E~en though this method -0f shutdown can be accomplished from outside the control room, this method assumes no con-trol room fires since electrical circuitry is such that it may be damaged by a control room fire and render this method inoperable. Plant fires and tornadoes are.not postulated to occur at the same time. The staff, in .a memorandum from L G.. Hu.lman to W. T. Russell, dated May 25, 1983, concluded that the plant can be safely shut down and maintained should control room ventilation be lost as a result of loss of service water due to tornado* missile strike.

Auxiliary Electric Equipment .Room Ventilation System In the event that the auxiliary electric equipment room ventilation system were lost as a result.of a missile strike,* the auxiliary electrical equipment room can*be cooled by the east turbine building ventilation system, which does not Dresden 2 SEP, Supp. 1 2:-10

use a cooling water medium. Circulated outside air Js sufficient to maintain ..

ambient temperature in the* room within design limits because of the high.capacity of the east turbine building ventilation system; however, the east turbine room ventilation system is powered by a nonessential motor control center (MCC 26-4) and, therefore, would be unavailable if offsite power were lost~ a likely occur-rence during a tornado. In the event that both ventilation systems were lost following a tornado, the heat load in the room could be reduced by deenergizing such unnecessary loads as the reactor protection system motor generator sets.

The equipment in the ro.om would remain operable for some time after ventilation.

was lost, and, ~ith manual actions to open doors and provide s6me means of ai~

Circulation; the. time could be extended.

Loss of station ,service water will have no immediate' effect on plant shutdown .

since the diesel generators, high-pressure coolant injection system, and con~*

tainment coolers, which are used to remove reactor heat when using low-pressure

.core injection cooling in the torus cooling mode, are cooled by water systems

'that have been protected from tornadoes.* (In IPSAR Section 4. 5. 3., the staff found the intake and exhaust st~cks *for DG 2 and DG 2/3 are vulnerable to tornado misiiles; see Section 2.3.2 of this supplement.} As a result of Appen-dix R (10 CFR Part 50) modifications, the licensee also provided alternate shutdown capability in the event of a fire in the electrical equipment room.

The alternate shutdown capability relies on the station service water system.

• However, if the auxiliary electrical equipment room were lost as a result of fire,. the service water system alone could provide co*ld shutdown. The iso*lation condens~r or high-pressure coolant.injection *~ystem could maintain the plant *'

in safe hot shutdown without the \1se of service water .. The .staff concludes that the isolation c'ondenser (if the condensate star.age tank and condensate transfer pum~s are a~ailable} or*the high~press~re coolant injection system could main-tain hot shutdown following a tornado. In addition, safe cold shutdowri could eventually be achieved using the automatic depressurization system, low-pres~ure coolant injection system, containment cooli'ng system, and the instrumentation and controls provided as alterriate shutdown equipment in the *event of a.fire in the auxiliary e*lectrical equipment room. *This method of safe shutdown is described in more detail in Section .2.2.2 of this IPSAR supplement. Hence, the only concern relative to the ability to shut the plant down if service water were *lost is *the effect of loss on ventilation on the electrical controls and.

instrumentation in the auxiliary electrical equipment room associated with the*

safe cold shutdown systems. It is assumed here that the station power is af- **

fected.by a tornado-initiated transient that can interr~pt offsite power arid.

can ~ause other tornado-rel~ted damage but is not concurrent with any loss-of~

coolant accident.

• On the basis of the licensee's estimate of probability of a missile strike on the cribhouse of .2.Jxl0- 6 (NRC estimate is 1.lxl0- 5 ), the probability that other non-tornado-~issile-protected equipment would be available, the alternate shut~

down capability outside the equipment :room, the possibilities of obtaining some ventilation, and the overly conservative assumption that all equipment in the room would be unavailable following.loss'of service water, the staff con-cludes that the probability of exceeding 10 CFR Part 100 guidelines should be very low. It shou1d be emphasized that the 10 CFR .Part 100 fission product*

• release criteri~ are deterministic; nevertheless the likelihood of radioactive release due lo tornados at Dresden Unit 2 is very low and therefore acceptable; Dresden 2 SEP, Supp. 1 2-ll

-On the basis of its review,-the staff concludes that backfitting tornado missile protection for the unprotected portion of the Dresden 2 cribhouse that houses the service water system that provides cooling water for the control room ven-tilation system and the auxiliary electrical equipment room ventilation system, is not required since the probability of such a strike leading to a radioactive release exceeding the guidelines of 10 CFR Part 100 is very *1ow. The staff based its conclusions*, in part, on the April 24, 1981 Dresden 2 SER for SEP Topic VII-3, "Systems Required for Safe Shutdown"; the May 11, 1983 SER for TMI Action Plan Item III.D.3.4, "Control Room Habitability"; the.January 19, 1983 SER for Appendix R, Items III.G and III.L, "Alternate Shutdown Capability";

and the SER for SEP Top.i c IX-3, u Station Service and Cooling Water Systems, 11

• dated Jane:30, 1981.

2.3.2 Diesel Generator Ventilation In lPSAR.Section 4.5.3 (NUREG-0823), related to DG 2 and 2/3~ the staff concluded that the air intake and exhaust systems for DG 2 and the swing diesel (DG 2/3) are not protected against tornado missiles and that damage. tb the intake and

• exhaust structures could result in unavailability of the diesel generator.

By letter dated June 12, 1984, the licensee concluded that*the probability of a torh~do missile striking the intake and exhaust pipes and silencers is 2.4x10~ 7 per year fot DG 2 and 3.9x10- 7 per year for DG 2/3 (NRC estimates are 1.2x10- 6 and 1:9xl0- 6 , respectively). These probabilities were obtained using the same methodology described in Section 2.3.1. Considering that these values are the probability of strike, the probability that these systems would be damaged to the extent of diesel inoperability would be lower.

Should the intake and exhaust structures .of both diesels become so damaged that both diesels become inoperable, .hot shutdown could be accomplished using the isolation condenser or the HPCI system, depending on the supporting equipment that is available, as previously described in Section 2.2.2.

On the basis of the licensee 1 s estimates of missile strike probability of the DG 2 and DG 2/3 intake*, exhaust, and s i 1encers, the probability. of damage to*

• the 'extent of inoperability of the diesel generators and the alternate shutdown capabilities, and the probability that other equipment not protected from tornado missiles would be available, the staff concludes that the probability of exceeding 10 CFR Part 100 guidelines is very low and ther,efore acceptable.

2.* 3. 3 . Exterior Tanks Safe shutdown with the loss of the ~ondensate storage tank or of other exterior tanks would be accomplished using the isolation condenser, with fire water pumps, or using HPCI, ~s discussed in Sections 2~2.2, 2.3.1, and 2.3.2 above.

-On the basis of .these discussions, this concern is considered resolved.

2.3.4 Masonry Walls In a June 12, 1984 submittal on this topic, the licensee identified two masonry walls--No. 105F on column line 31* of the Unit 2. turbine building ~nd one wall on column row C between. the radwaste building and the turbine. bujlding--that Dresden 2 SEP, Supp. 1 2-12

may be vulnerable to tornado missiles. The licensee *reported and the staff con-curred .that the probabilities of striking either of these walls *is much Tower than 10- 7 -per year. Therefore, this issue is considered resolved.

2.4 Topic 111~4.B, Turbine Missiles (NUREG-0823, Section 4;6)

• 10 CFR Part 50 (GDC 4), *as implemented by RG 1.115 and SRP Section J.5.L3, requires that structures, systems, and components important to safety" be *'

appropriately protected against dynamic effects, inc 1udi.ng potent i a1 missiles:

To provide assurance that turbine missiles will not be generated at either design speed or at destructive overspeed, t~rbine *inspections and overspeed

• protection system testing are required. *

• Th~ staff co~cluded that Dresden Uhit 2 does not comply with the staff recom-mendation that an inservice irispection.program be developed and i~pl~mented: *

• that requires turbine disassembly at approximately 3-year intervals and that all normally inaccessible parts be inspected in accordance with procedures suggested by*the.turbine manufacturer. -

By letter. dated October 8, 1982~ the licensee provided the staff with the pro::.*

posed inspect ion schedule for the low-pressure port ions. of the turbine and *the basis* for the proposed schedule. . (~

The staff,: irl:. a Jetter dated May 31, 1983 to the licensee, found the Dresden** . .

Station mainte~ance procedures and schedules, as described in the final SER for Topic 111-4.B, and in Section 4.6-of the lPSAR, *constitute adequate interim assurance of the low probability for turbine missile generation, until a gen-erically based~maintenante*and inspection program can be eJtablished. Should the . results o'f. the generic General Electric (GE) models and methods *i ndiC:ate

• the need to ch~nge the existing ~aintenance and inspection ~rogram ~t Dresden .. '

Unit 2, such changes~will be implemented generically for*.all facilities.with GE turbires following compl7tion'of the staff's review. * *

• Cons~quently,-~he staff*concludes that the action described in Section 4~6 of the IPSAR for Dresden Unit 2 is compl~te.' * **

2.5* To it 111~5.A .Effects of Pi e Break on Str~ctures S stems *and Com onents

• ns1 e.* onta1nment ect1on 10 CFR Part 50 (GDC 4), as implemented by SRP Section 3.6.2,* requires, ~n ~art, that structures, systems, and components important to safety be appropriately protected against dynamic effects, such* as* pipe whip *and 'discharging fluids-~

that may result from equipment failures. If a pipe should break inside *the *

• containment, this review ensures that the*plant could be safety shut down *with~

out a loss of containment integrity and that the break *would n'Ot be more severe than those analyzed as design-basis accidents.

The staff compared the licensee's proposed evaluation method, submitted by let-ter dated*Augu~t 23, 1982, with the criteria and methods presently used for licensing new f~cili.ties; In general, the licensee 1 s program, methods.of approach, :and tri teri a *u*sed for the* evaluation were adequate, *but there were Dresden 2 SEP, Supp. 1 2-13

four issues in which the licensee's methodology differs* from current criteria.

They were: jet impingement on target pipe, broken-pipe impact on target piping, detectability requirements, and criteria implementation. The licensee. provided a 'fi na 1 report on November 11, 1982 on the effects *of hi.gh-energy 1i ne breaks on systems, structures, and components inside.the containment. The ~taff re-viewed this report fo a safety evaluation issued on October 27, 1983; the reso-lution of each of these issues is given below.

2.5.1 Jet impingemen~ on Target Pipe In IPSAR Section 4.7.1, the staff found the licensee used the assumption that a jet or whipping pipe would inflict no damage on other pipes of equal or greater size and equal or greater thickness. However, the staff held that the energy-absorption mechanism for a pipe-to-pipe impact is different from that for jet impingement on a pipe and, therefore, required the licensee.to evaluate and address the effects of jet impingement regardless of the ratio of impinged and postulated broken pipe sizes.

In re~ponse, the licensee submitted the assumptinns and criteria used in its final e'valuatian of jet impingement effects. On the basis of this information, the .staff concluded- that the licensee had. reassessed its jet impingement eval-uation in accordance.with the staff's position, and therefore, that the licen-see's evaluation is acceptable.

• 2~5.2

• Broken-Pipe Impact on Ta~get Piping I *' '

In IPSAR Section 4.7.2, the staff determined that the licensee:u~ed the criterion that the limiting factor for an applied equivalent static load is that the result-ing strain in the target piping material should not exceed 45 percent of the minimum ultimate uniform strain of the material at the appropriate temperature.

This criterion is acceptable for avoiding cascading pipe breaks. However, some piping systems are required to deliver certain* rated flow and should be designed to retain dimensional .stability when stressed to the allowable limits associated with.the emergency*and faulted conditions;. that is, the functional capability of the piping is required to be.demonstrated. The licensee wa~ asked to show that the target piping will remain functional as a result of jet impingement and pipe whip impacts.

The. 1i censee performed a parametric study covering a range of geom*etri c and 1oad parameters. In reviewing the example in the licensee's parametric study, sub-mitted on August 23, 1982, the staff found that the 45 percent of the minimum uniform ultimate *strain reached at the point of load application was a global strain, because a beam model was *used for analysis. The licensee was asked to show that the resulting localized deformation (the.flow area reduction) would not affect the System IS Capabi 1ity tO de 1i Ver the reqUi red fl Ui d fl OW.

In responses dated November 17, 1982 and January 10, 1983, the licensee indi-cated that a more detailed shell model analysis showed that the:maximum strain was 25 percent of the minimum uniform ultim~te strain of the material and that the maximum flow area reduction was 20 percent. The licensee further stated that the functional capacity of the target pipe is not significantly affected as a result of*~ postulated 20-percent reduction in flow area. After reviewing this information, the staff determined in its October 27; 1983 evaluation that the licensee's target pipe evaluation is acceptable.

Dresden 2 SEP, Supp. 1 2-14

2.5.3 Detectability Requirements The staff reviewed the licensee's proposed fracture mechariics evaluation for

• resolving the pipe break issue and found it unacceptable, as noted in IPSAR Section 4.7.3. The.licensee's analysis was based on pipe crack caused by fa-tigue failure of the pipe. The *staff's position is that piping failures are -

more likely caused by other mechanisms (e.g., stress-corrosion cracking) and therefore, the licensee should follow the staff's guidance for resolving the pipe break interactions. .. - -

During the licensee'~ detailed. review*of break interactions, as discussed in the licensee's November 17, 1982 submittal, all interactions were resolved without recourse to the* fracture mechanics/leak detection approach. Therefore, the .

staff cQhcluded, in its October 27, 1983 safety evaluation, that this issue* is*

no l on~er .applicable. * ~* *:

• 2.5.4 Criteria Implementation In IPSAR Section 4. 7.4, two areas were identified during the topic review in -

which the licensee's approach was found to be generally acceptable pending comple~ion of staff review of the results.

The staff asked the licensee to provide information on pipe whip load for~ula~

ti on, including a discussion of how the kinetic energy of the whipping segment was determined. After reviewing*the information provided in the .licensee's sub-mittal of October 3, 1983, the staff concluded that the licensee's methodology for pipe whip load formulation is- acceptable*.

The second item* identified was the integrity -of the drywe 11 .1 i ner. The 1 i censee was asked if there were any break locations that could result in sharp edges_

perforating ~he liner as the- result of large piping (14 inches or* greater) im-pacting the. liner. - Pre_vi ous tests have showed that when the 1 i ner

• is 1oaded over a large:enough area, deformation of more than 3 inches can occur.without failure of the 1i ner. However, .the staff was concerned that the 1i ner could be punctured if a* jagged edge whipped into the liner over a loc~l area. *-

The licensee presented an analysis of the containment liner in a November 11; 1982 submittal which shows that the* liner would not be perforated. Missile impact test*results were used along with the above statit l~ad test to support the analysis. Also, the licensee provided additional information by letter*

dated September 12, 1983 on the one case in which a feedwater line reducer (18-inch to 12-inch pipe reduction) contacts the liner. The whipping feedwater pipe forms a plastic hinge and then the reducer contacts the liner along the -

pipe and thus flattens against the liner. Therefore~ the staff concluded that the consequences of this case are bounded .by the above tests and that the inte-grity of the drywell liner is assured.

2.6 - Topic III-6, Seismic Design Considerations (NUREG-0823, Section *4.9) * --. '

10 CFR Part 50 (GDC 2), as implemented by SRP Sections 2.5, 3.7, 3.8, 3.9, and 3.10 and SEP review criteria such as NUREG/CR-0098, requires that structure~,

systems, and components important to safety be designed to withstand* th_e effects of such natural phenbmena as,earthquakes.

Dresden 2 SEP, Supp. 1 2-15

2.6.1 Mechanical Equipment In IPSAR Section 4.9.2(1), the staff identified a lack of* information- with re-gard to pipe stress resulting from motor-operated valves. The staff will re-view this concern under the resolution .of Unresolved Safety Issue (USI) A-46.

In IPSAR Section 4.9.2(2), the staff stated it lacked sufficient *information to evaluate the structural integrity of the reactor vessel and internal supports.

The staff also stated it will review the analyses of the Oyster Creek reactor vess*e l and internal supports to determine* its applicability to***tlle Dresden Unit 2 design. The Oyster Creek analysis is currently under review and, when completed, the staff will evaluate its applicability* to Dresden ljnit 2*.

• In IPSAR Section 4.9.2(3), the staff stated it lacked sufficient information to eva.luate -the structural integrity of the recirculation pump and supports.* By letter dated March 1,.1989,' the staff questioned the long-ter~.use*of Cod~ Case N-411 damping values for limited applications, such as.recirGulation piping and supports, without the benefit of RG 1.84. The staff further stated that the licensee should re-evaluate these applications for long-term.operation* utilizing RG 1. 84~ . By 1etter dated April 11, 1989; the 1i censee stated *that this* method-

.. ology was only used for* interim operabi*lity justification:for Dresden Unit 2.

During the February 1987 refueling outage, the recirculation pump supports were upgraded to full FSAR compliance with work completed as of June 1987. The 1i-

.,. censee .further *stated that it. does not believe there are any,*applicatiOns*_'.of.

• the ~se:of Code Case.N~411 *damping* values.on other,sy~tems* at.Dresden*Unit 2~ .

.T~is .was confirmeq in a. May 31, 1989 *letter. On the basis .of th.is information, the staff has concluded that the iSsue is .resolved; .. * "' .. ~-* *

• 2*! 6. 2 Qua 1ifi cation ..of Cable Trays" * . ,. *,  : . ; ~' '*~ :

In IPSAR Section* 4.9.3, the staff stated. it lacked sufficient information.to con-clude that. the ~eismic qualification.of electrical cab.le* trays *is-acceptable*.

Since that time, the staff and the Seismic Qualification Utility Group have de-cided tha,t this issue will -be resolved on a.generic*ba*ses within the*framework of the staff's resolution of USI A-46. This issue, with. re~ard.to thi~ program, is therefore considered closed.

~ *: - *

• I ' * , * .*' '

• I 'f

• j" ,

2.7.*.Topic III-:-7.B,*Design Codes; Design Criteria, Load Combfoations., and

• Reactor Cavity Design Criteria (NUREG-0823, Section,4.10} * : 1 ":' * *.

I * *

• 10 CFR Part 50 (GDC 1, 2, and 4), as implemented by SRP Section 3.8, requires that structures be* designed for* the loadings that may be im*posed* on them and that the.structures conform to applicable codes and standards .

Code, load, and load combination changes affecting specific structural elements have been i dent ifi ed for which safety margins may be reduced *from* those requi:red by current standards. Therefore, the staff position in the IPSAR was that the licensee.,should.provide information regarding .the applicability of the code changes 'and the safety margins. **

The licensee provided this* information in letters dated August 2, .1982 ~nd July.11~ 1984., The staff's contractor,. Franklin Research Center\ reviewed the Dresden 2 SEP, Supp. i 2-16

design code changes and load combination issues and issued a TER (C5506-425) dated June 3, 1986. Section 6 of the TER identifies items not addressed, items addressed by the licensee using a sampling philosophy, and items requiring further clarification. By letter dated.July 26, 1989, the staff requested ad-ditional information to close these open items identified in the TER. The li-censee, by letter dated August 30, 1989, provided the requested information which is under staff review.

Thermal-Overload Protection for Motors of Motor-a erated ect1on 10 CFR 50.55a(h), as implemented by Institute of Electrical and Electronics*

Engineers (IEEE) Std 279-197 and 10 CFR Part 50 (GDC 13, 21, 22, 23, and 29),

requires that protective actions be reliable and precise and that they satisfy the single-failure criterion using quality components.

• RG 1.106 pre~ents. th~:

staff position on how thermal-overload protection devices can meet these requirements.

This topic was reviewed. to provide assurance that the application of thermal-overload (TOL) protection devices to motors associated with safety-related

• motor-operated valv~s does not needlessly hinder the performance of valve safety

.functions.

In IPSAR Section 4.12, the staff concluded that the licensee should either bypass thermal overloads*on an emergency core cooling system initiation signal or demon-strate that the TOL device trip set points are conservatively established. By letters dated November 4, 1983 and.January 2, 1985, the licensee described the procedure used to size thermal-overload heaters and included a sample calcula-tion. On the basis of this information, the staff concludes that the licensee's procedure will result in trip set points with all uncertainties resolved.in fa-vor of compJeting the safety-related action: Therefore, by letter dated July 9, 1985, the staff concluded that the licensee 1 s actions have resolved this issue.

2.9 Topic V-5, Reactor Coolant Pressure Boundary Leakage Detection

• (NUREG-0823*, Section 4.13) .

10 CFR Part 50 (GDC 30), as implemented by RG 1.45 and SRP Section 5.2.5, pre-scribes types and sensitivity of systems and their design criteria necessary.for detecting leakage*of primary reactor coolant to the containment or to other interconnected systems.

In IPSAR Sections 4.13.1 and 4.13.2, two issues related to leakage-detection capabilities for the reactor coolant pressure boundary required further evalua-tion. These* issues are addressed below.

2.9.l System Sensitivity In TPSAR Section 4.13.1, the staff noted that the reactor coolant pressure*

boundary (RCPB) leakage detection systems did not satisfy the sensitivity re-quirements of RG 1.45. Also in IPSAR Section 4.7.3, the staff found that the necessary sensitivity of these systems should be associated with the resolution of high energy line break effects (refer to Section 2.5.3 in this supplement .

• for resolution of this issue).

Dresden 2 SEP, Supp. 1 2-17

As a result of* the staff's review of intergranular stress corrosion cracking (IGSCC), independent of .the SEP, additional monitoring requirements and tighter limits on unidentified leakage were added to the Technical Specffications for

.. Dresdeh Unit 2 by the issuance of Amendment 75 to Provisional Operating License NQ .. DPR-19i on* April 7, 1983. .

The staff's review of high energy line breaks, issued October 27, 1983, *did not identify a need for improved leak detection*sensitivity. The existing systems are monitored once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> rather than continuously as recommended by current licensing criteria. This difference in sensitivity is small, particu-1arly in the absence of any compelling high energy 1i ne break concern. There-fore; the staff concluded in a letter dated February 13, 1984:that this issue is' resolved. The staff notes, however, that the continuing review of IGSCC in

• .RCPB piping may identify a need for further improvements in leakage-detection capability.

2.9.2 Seismic Qualification In IPSAR Section 4.13.2, the staff noted that the RCPB leakage-detection systems are not qualified for a safe shutdown earthquake (SSE). The licensee committed to implement procedures that would specify actions to be taken following seismic events that caused leakage detection equipment failures. The* implementation schedule was to be.established following resolution of the open issues in IPSAR

• .Section 4.7. These issues were resolved by letter dated October 27, 1983 and the licensee provided in a *letter* dated*April 14~ 1984, a*copy'of the Emergency Plan Implementing Procedure that satisfied the required implementation schedule.

2.10 Topic VI-4, Containment Isolation System (NUREG-0823~ Secfion 4.~8) 10 CFR Part 50 (GDC* 54, 55, 56, and 57}, as implemented by SRPSection 6.2.4 and

• RG 1.11 and 1.114, requires isolation provisions for lines penetrating:primary containments to maintain an essentially leak-tight barrier against the uncon-trolled release of radioactivity to the environment.

• In IPSAR Section 4.18.2, an issue related to containment isolation.required further evaluation. The licensee responded to the identified differences by letter dated November 18, 1982, and the staff reviewed that information. The staff concluded. in an evaluation dated February 13, 1984 that the presence of

• sump level monitors satisfies the requ*irement for a system to '.detect* leaks in containment penetrations with remote manual isolation valves. However, the procedures associated with that system do not satisfy the concern regarding the point at which the remote-manual valves should be isolated once a leak is detected.

The licensee has stated that the switches for remotely operated valves 1501-22A and Band 1402-25A and Bare located in the main control room. The staff.re-

. quires that the leakage conditions under which these valves should be isolated should be incorporated in the emergency procedures; this action wi 11 **re so 1ve *the

• is~ue. In a letter dated April 14~ 1984f the licen~ee committed t6 incorporate these leakage conditions into the appropriate procedures. The staff will confirm this.

Dre~den 2 SEP, Supp. 1 2-18

2.11 Topic. VI"."7.C.l, Appendix K--.Electrical Instrumentation and Control Re-reviews (NUREG-0823, Secti-0n 4.21) 10 CFR Part 50 (GDC 2, 4,-17, and i9), as implemented by SRP Sections 8.3.l and 8.3.2, requires that the onsite electric power distribution* system be designed to provide (1) redundancy of safety-related components and systems, (2) electri-ca.l i ndependente between redundant portions, and .( 3) phys i ca 1 separation. between redundant components of the system. **

In IPSAR Section 4.21, the staff found that action should be taken with respect to issues of breaker adequacy, disconnect links, use of bus tie breakers during power operations", operation with a failed battery, and isolation of Class*.lE sources from non-Class lE loads. Commitments to modify plant operating proce-dures for the.* use of disconnect 1inks and bus tie breakers during power. opera-tions were made in the IPSAR (see Section 4 of this supplement for confirmation).

The resolutions of the three remaining issues--that ensure electrical power for the plant electrical instrumentation and control (EI&C) features.will be a\ia.il-able--are described in the sections that follow.

2.ll;l Break~~ Adequacy In. IPSAR Sec~ion 4.21.li the.staff .identified concerns on' the electric~l inde-

.pendence between *.redundant portions of -the electric power distribution,* system.

The staff 1 s. position was that the licensee should-verify the adequacy of protec-

tive . rlaying for battery charger breakers, for.the 125-V dc*automatic trans;fer I~~

and for -the- 250*:-Y battery chargers.' " .

• The staff had identified several mechanisms by which.breakers might fail:

• (l) failure of *a load b~eake~ to cleat a fault before the bus feeder r~ach~s its. trip set point (may be caused* by* a: lack of adequate protection curves resulting. from failure to revise ~ettings as new loads and/or sources are added to a system) ,,

(2) a ground fault tripping an ac feeder.breaker instead of a load break~r

. (caused by inadequate ground-fault,protection)

• j **

(3) .protective re-lay set point drift -outs.ide the error band assumed *in the coordination of load arid feeder breakers (caused by infrequent testing.of the_ relays)  :. **

I.

By letter dated December 3o, 1983, the licensee addressed the staff 1 s concer~s with regard to protective relay coordination. The licensee provided a study of the relay coordination for the breakers mentioned above. The licensee noted that

. ground fault protectjon is not provided for these circuits. After reviewing that submittal, the staff concluded in a January 24, 1984 safety evaluation that<set point drift* does.not-pose a significant problem based on the nature of the over-current devices. Further, the staff concluded that there is sufficient margin in the protection curves to. accommodate larger faults .than those calculated in the analysis: This issue is resolved~

• Dresden 2 SEP, Supp. 1 .2-19

2.~1.2 Disconnect Links In IPSAR Section 4.21.2, the staff identified a concern with regard to the inter-conhection of the Division 1 main de bus with the Division II reserve de bus through molded case circuit breakers that cannot be open. The licensee stated that these breakers are only used during maintenance operations and has committed to institute a pre-startup procedure to verify that 125-V de breakers and dis-connect links are in proper position before startup. The licensee has since provided the staff a copy of 'this procedure for review. On the basis of this review, the staff finds that the proposed administrative control is in accord-ant~ with *staff requirements and is, therefore, acceptable.

2.11.3 Operation With Failed Battery In IPSAR Section 4.21.4, the staff ldentified a concern with regard to the Dres-den Unit 2 technical specification time limit for continued operation with a battery out of service. The licensee has addressed this concern and it is dis-cussed in Section 3.1 of this supple~ent..

• 2.11.4 Isolation of Class lE Sources From Non-Class lE Loads In IPSAR Section 4.21.5, the staff identified a concern with regard to the iso-lation of Class lE sources from Non-Class lE loads. The licensee*was asked*tri demonstrate by suitable short circuit analyses and coordination curves, that a11 Non-Class lE loads* are adequately isolated from Class lE sources by at least two coordinated circuit breakers. The licensee in a letter dated March 6, 1985 ~ro

'vided the fault current for**a bolted fault at the bus end and has demonstrated that* non-Cl ass IE loads wi 11 be isolated from Cl ass lE sources* by two properly coordinated breakers~ Based bn the information provided the staff has determined that the two breakers feeding the non-Class lE lc:iads are properly coordinated with the bus supply breakers and that the connection of the non-Class lE loads to the Clas~ lE system does not significantly degrade the Class *lE system ~nd is, therefore, acceptable.

3)

• " 10 CFR Part 50 (GDC 5) requires that structures, systems, and components impor-tant to safety shall not be *shared among nuclear power units unless it can be shown that such*sharing *~ill not significantly impair the ability of these structures, systems~ and components to perform their safety functions.

In IPSAR Section 4.23.1, the staff raised a concern regarding paralleling of the 125-V de and 250-V de systems. There are no physical features nor administrative procedures* to prevent paralleling during power operation. Furthermore, plant

• procedures require that the 125-~ de batteries be paralleled as part of the ground-fault detection process.

• Therefore, the staff position was that the licensee should provide assurance that the shared de syste~ will not be paralled during power operation .

Dresden 2 SEP, Supp. 1 2-20

In a l~tter dated June 9, 1987, the licensee provided a procedure in place at;:

Dresden that allowed paralleling of the de systems as a trouble-seeking tech-nique for ground detection. By.letter~ated December 22, 1987, .the.staff com~

mented on this Dresden procedure and. cone 1uded ,that the procedure was not in accordance with GDC 17 of Appendix A to 10 CFR Part 50 and IEEE.Std 308:-1980.*

• In a.letter dated~September 28~ 1988, the licensee stated it agrees wit~ the.*

staff position and paralleling of the 125-V *de and 250-V. de systems shall no.**

1anger be all owed during power, op~rat ion of either Dresden .uni.t; the ground . . ,

detection proc;ec;lures for both u_ni t~. Cir.e.. b.ej ng revj sed accordingly.. The .staff

• has reviewed the ,licensee's submittal and determined that it resolves the. staff concern and is therefore acceptab 1e. NRC '*s Region I.II personnel wi 1.1 verify *.

that the ground detection procedures have been appropriately modified.

Isolation of Reactor ua 1f1cat1on o 10 CFR 50. 55a(h), through IEEE Std 279-1971 and 10 CFR 50 (GDC 24),. requ*i res that a protection system be isolated from non-safety systems and that no credi-ble failure at the. output of ~n isolation dev*ice shall pr.event .the associated.

protection .

system channel from meeting* minimum

, . ~

performance requirements.

'*. l ,. 1 *

• .' . ,. . ,. . . i**. . .

2.:13.1: *Re.actor:Protectjon System Control Systems. . * . -:.:*

' ,: *'.  : ( ~, .  :. . '  :. I i ~ * .. , . * . . ' . * ~ . , .' \ . . ' l : ;

ln.lPSAR :Section *4 .. 24.1, the staff ,identif.ied a co_ncer11 with regard to the* ele.c-

.trical .. isolation between the neutron flu~ monitoring-system .and .the co*ntrol-.r.oom process;recorders.and jndicating meters not.satisfying the requirements of:lEEE Std.-,279 *. I.n a ,letter dated December 6; 1982-, the licensee committed to verify

_. that the neutron flux i:nonitoring system is sufficiently isolated _from the. con-tr.ql

• roomjndicators -and, if not, to install Class 1E .isolators .

. *in .. letter..'dat~d Janua~y 9, 1987., th~ licensee proposed_ to i.nstall. Class.iE signal i sol at ion devices at the inputs to each recorder. The l icens.ee al so stated i.t is evaluating the possibility of installing Class lE recorders that would elimi-nate*the need.for*electrica] isolation devices .

. (' . *.. .

The staff concluded that either of the proposed methods (Class lE devices or Class 1[ recorders) meet the .electrical separati6n requirements and1are therefore acceptable .. "However; -in.a letter dated Noveml?er ~O; J987, 1 the staff. reques:ted that: the lfoen.see submit the test results or evaJuatiqn that demonstrate the,.

ability of.-.the-*specific.components. selected.for' installation to meet the sepa:-

ration requirements. The NRC staff criteria for reviewing isolation devices are listed below ..

is~lation, de sc;ribe .

1 (1), For :the :type of device used to accomp.iish electrical.

. the* specific testing performed to demonstrate 1;.hat. th.e device.is acc,eptab le for its application(s). This description should ,includeelem,entary d.ia-._

grams necessary for indicating the test configuration and how the maximum

• .~redibl~ifaults were applied to the devices.

(2) Oat~ to verify that the maximum credible faults applied during the.tes't were the maximum voltage/current to which the device could be exposed, and

• to define how the maximum voltage/current was determined.

Dresden 2 SEP, Supp. 1 2-2i

(3) Data to verify that the maximum credible faults applied to the output of the.device in the transverse mode (between signal and return) and other faults were considered (i.e., open and short circuits).

(4) Define the pass/fail acceptance criteria fo-r each type of device.

(5). Provide a commitment that the isolation devices* comply with the environment qualifications (IO CFR 50.49} and with the Seismic qualifications that were the basis for plant licensing.

(6) Describe the measures taken to protect the safety systems from electrical interference (i.e., electrostatic coupling, e*lectromagnetic interference, common mode, and crosstalk) that may be generated by the safety parameter display system.

(7) Provide information to verify that the Class lE isolator is powered from a Ciass lE source.

By letter dated February 2, 1989, the licensee submitted test reports of the i so 1at ion devices selected for the modi fi cations.* . The 1j c':ensee a-1 so stated that the modifications for Dresden Unit 2 were completed in December during the refueling outage that took place in the fall of 1988. The $taff will cbnfirm the installation of these isolatlon devices.

The staff h.as reviewed the test reports submitted by the 11cens.ee. The isolation device: used by the licensee is a Moore Industries signal isol.atQf, Mpdel.No.

SCT l-V/O-l-V/24-V de. For Dresden, the licensee determine<:! that .the maximum credible f.ault (MCF) is 120-V ac at 50 amps. As described .in the test report, the Moore Industries isolator wa~ subjected to 12 separator MCF tests. The fault tests produced an increase in the noise on the source (lE) terminals but did not affect the IE.signals or equipment and was found to be acceptable.

The tests performed included functional tests, hi-pot tests,' the various MCF

• _tests included above, and seismic qualification testing. The isolators are in a mild environment which does not require environmental qualification. The licensee has taken appropriate measures to protect the equipment from electrical interference.

On the basis of the staff's review of the licensee's submittal, including testing results, the staff concludes that these devices are acceptable for.use at Dresden 2 to provide isolation'fr6m the recorders to the RPS and, therefore, this issue is resolved.

2.13.2 *Process Computer Section 4.24.2 concerns the electrical isolation between the average power range monitor {APRM) and the process computer. The APRM scram function is derived from.relay actuation resulting from amplified analog signals sensed by those relays. The amplified analog signals are input directly. to the process computer.

The licensee, in a letter dated January 9, 1987, stated it has modified the inputs to include "flying capacitors" which are switched by relay contacts from the input signals to the proce.ss computer. Since there is always .an open contact between either the capacitor and .the computer or. the capacitqr. and the input, .

the staff finds this acceptable.

Dresden 2 SEP, Supp. 1 2-22.

2.14 and Annunciation In a letter dated October 5, 1982~ the licensee proposed the use of .reduridant process computers to monitor the following indications and alarms in the eontrol room: (1) battery current, (2) battery charger current, (3) battery breaker open alarms, and (4) battery charger (ac and de) breaker open alarms. The staff, as part of its review of this issue in Section 4.28 of the IP'SAR found this proposal acceptable.

Subsequently, by 1etter dated August . 11, 1987, the licensee indicated that.

it had performed additional analysis and* concluded that a different scheme, .

which does not include the indications and alarms listed above, would provide more effective monitoring for de systems. the licensee's scheme includes the following *indications and alarms in the main control room at Dresden Unit 2.

• battery voltage monitor (voltmeter)

• battery low-voltage alarm

• battery discharge current high alarm

• de.bus loss of voltage (or undervoltage) alarm

• de ground alarm (for undergrdunded systems)

• battery cbarger trouble (ac/dc failure) alarm The specific requirements for de *power* system monitoring derive. from the gen-.erfr requiremehts embodied in Sections 5.3.2(4), 5.3.3(5), and 5.3.4(5) of IEEE Std 308-1974; In .summary, thes.e general requirements simply state that th~ de system (batteries, distribution systems, *nd chargers) shall be ~onitored to th~ extent that it is shown to be ready to perform its intended function. Accorc;lingly, the guidelines* used in the licensing* *review of the de power system designs follow.

The following indic~tions ~nd alarms *of the Class lE de* power system status.

shall be pr.ovi ded in the.. contro 1 room:

(1)

• battery current (ammeter-charge/discharge)

(2) battery charger output current (ammeter)

(3) de bus voltage (voltmeter)

(4) battery high-discharge-rate alarm (5} de bu~ undervoltage alarm (6) de;: bus ground 'alarm (for ungrounded system)

(7) -battery breaker(s) or-fuse(s) open alarm (8) battery charger output breaker(s) or fuse(s) open alarm (9) battery charger trouble alarm (one alarm for a number of abnormal conditions which are usually indicated locally)

The staff concludes that the monitoring cited above, augmented by the periodic test and surveillance requirements included in the Technital Specifications, provide reasonable assurance that the Class lE de power system is ready to perform its .intended safety function.

• Except for battery and charger current indication (items 1 and 2) and battery and charger breaker (or fuse) open alarm (items 7 and 8), the Dresden Unit 2 monitoring stheme for de systems meets the guideline& cited above and is, there~

fore, acceptable. The staff evaluation of these exceptions is described bel~~.

Dresden 2 SEP, Supp. 1 2-23

Battery and Charger Current Indication By 1etter ~ated Augµs*t 11, 1987, the license!=<: in _just i fyfog its ':-scbeme that '

.. does no~ Jnc~ude moni~oting of *de_ system currents:fl'.'om the_control room, argued that capacity does not indicate the condition of a de system. -The mgin c9ncern for determining the condition of a de system is that the battery is charged, the c~argers are supplying powery antj hjgh r~sistance _connections do-not exist~ .. In 1i eu. 9f moni,tori ng current, _the 1i,censee. i npi cate,d that -C~paci ty :~es ts,. battery

-high. discharge current a1arms;* .an<;i undervql.:tage_ ,a 1arms Vli.ll. be U!)ed to ensure the._ de* syste'!l is'. operational .. _.Performing ..c;apacity tests.,demo~str~tes the battery has its capacity and high resistance connections do *not exist o_n the *battery _-

cell. interconnect:ipns{, Battel'.'.Y high dfscharge and unde~voltage_ al_a_rms alert: an

' ,opera;to_r th~~. ~he,. batt.e.ries are lo_sing,capacity. . . -- * .. ,. -:- .> _. _-.

8/.1e'tter dated March 25*, 1988, the licensee"provi'ded the foHowing addit1o~al information in support of its scheme for monitoring de systems:

(1) The monitoring scheme includes local de system voltage and current meters as well as control room indicators and alarms described above. These moni-*

tors comply with the requirement, contained in GDC 13 of Appendix A to 10 CFR Part 50, that instrumentation be provided to monitor systems over their anticipated ranges for accident conditions*as appropriate to ensure adequate safety. -

(2) The existing voltmeter, which monitors de bus*voltage in the main control

-room, meets the requirement contained in Section 5.3.3(5) of IEEE Standard 308-1974 that indicators be provided to moni.tor the status of the battery supply.

(3) .In regard to Compliance with the requirement contained in Table 1 of Regu-latory Guide 1.97, Revision 2, (i.e., status indication of de bus currents shall be monitored in the control room in order to help the operator make appropriate decisions i~ using the de system in mitigating the consequences of an accident), the licensee argued (a) that de bus current does not pro-vide any useful information not already provided by the main control room voltage meter and undervoltage alarms and (b) that de bus current will be monitored locally and not in the main control room.

The staff concludes that the proposed scheme that includes de bus current being monitored locally (i.e., float current to the batter9 is metered with appropriate accuracy, monitored, and recorded at the same frequency as would be done for an instrument located in the main control room) provides an essentially equivalent level of monitoring and meets GDC 13 of Appendix A to 10 CFR Part 50, meets IEEE Std 308-1974, meets RG 1.97, and is, therefore, acceptable for the range of design-basis events for which the Dresden Unit 2 station was licensed.

Battery and Charger Breaker (or Fuse) Open Alarm By letters dated August 11, 1987 and March 25, 1988, the licensee indicated that the functions of a battery and charger breaker (or fuse) open alarm are provided by alternate means; specifically, operating procedures and alarms activated by low voltage and high battery current. The staff concludes that this alternate Dresden 2 SEP, Supp. 1 2-24

means, together with periodic monitoring of battery current discussed above,..

provides reasonable assurance that the battery is connected to the de distribu-tion system, is being charged, and is, thus, ready to perform its*inter:ided function. On this basis, the staff considers the licensee's propos~d ~onitoting

• scheme acceptable.

• On the basis of the considerations discussed above, the staff has concluded that the monitoring scheme for de 5-Ystems proposed for use at Dresden Unit 2 wi 11 provide reasonable assurance that the Class lE de power system is ready to per-form its intended safety function, meets IEEE Std 308-1974, RG 1.97; *Goe 13 of Appendix A to 10 CFR Part 50, and is, therefore, acceptable. The.staff, in Inspection Report 50-237/85-30 confirmed that battery voltage indication is*

installed in the control room. The staff will further confirm that a monitor-ing scheme for determining local de current and voltage for the de systems has been installed: * * **

• Dresden 2 SEP, Supp. 1 2-25

Table 2.1 Summary of IPSAR and supplement evaluations

'~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~---'-~~~~~~~~~~~~~~~~~~-

~ SEP IPSAR . Licensee's Licensee's

g. Topic Section requirements Supplement requirements

i No. No. Title from lPSAR Section N.o. from supplement N ~~~~~~~~~~~~~~~~~~~~~---""~~~'--~~---~~~~~~~~~~~~~~~~~~~~-

~ 11-3.B, 4.1.1 Design-Basis w-o 11-3.B.1 Groundwater Level V'I

.§ 11-3~ C 4.1.2

• Probab°Te Maximum Floo*d

-0 4.1. 3 Roof Loadings Mod Hy parapets* to. Efrisur~ . 4 Complete ponded water* is within *...

structur~l capaC:ity Of r:oof.

4.1.4 Flood Emergency Plan Modify existing procedures to 4 Complete address the ability to cope with.probable maximum flood, 111-1*' 4.2.1 Rad*i ography (1) Identify Class 2 vessels 2. r.1 Co'mp l ete N

I Requirements bui.lt to .Class C requirements N

CTI containing Class C jqints and their exam*; nat i cin techniques (2) Describe examination given 2.1.1 Complete to recirculation pump casirig.

4.2.2 Fracture Toughness Demonstrate fracture toughness 2.1. 2 Determine if LST ex-for components or demonstrate (Open) ceeds 77°F, identify failure is accepta~le. the operating condi-tions when it does

.not, and the design changes necessary so the LST exceeds 77°F.*

• 111-.2 4.3.1. Reac.tor Building None Structure Above the Operating Floor.

0

""'5 (I)

(,/'I

• Table 2.1

• (Continued) 0.

(I)

SEP IP SAR Licensee's Licensee's

J Topic Section

• requirements Supplement requirements N No. No. Title from IPSAR Section No. from supplement

(/)

m

" III-2 4.3.2 Ventilation Stack Ensure stack failure will not 2.2.l None

(/)

affect ~afe shutdown.

c

-c

-c 4.3.3 Components Not Enclosed Identify and ensure components 2.2.2 None in Qualified Structures . can with~tand 1 tornado loading or their loss will.not affect safe operatio.n .

4.3.4 Roof Decks . Demonstrate failure of roof 2.2.3 None

• deck~ will not ~ffect plant safety. * *

~ . . .

4.3.5 Load Combinations Will be addressed in Topic N

I III-7.B ...

N

-....J IIl-3. C 4.4.1 Flow-Regulation Station None' * ..

4.4.2 Intake and Discharge None Structures 4.4.3 Inspection Program Modify procedures to ensure 4 Complete (1) supervision *by qualified

.Persorinel and (2) ihspections fo l'l owing extreme events.

III-4.A 4. 5.. 1 Service Water System *Demonstrate* auxiliary electri- 2.3.l None

. (SWS) cal equipment room has adequate ventilation as part of TMI con-trol room habitability (Part

. of TMI *Action Pl an 'Item

• }II . D. 3. 4) .

4.'5.2 Station Battery Systems None Ii

Table 2.1 (Continued) 0

-s (I) "

CJ') SEP IP SAR <-.

Licensee's Licensee's

~

(I) Topic Section JO **

requirements- Supplement requirements

~

N9. No. Title' frorn IPSAR Secti.on N9. _fr9m s~pp l ement N

(/)

m III-4.A 4.5.3 Diesel Generator Ensure that DG 2 and DG 2/3 2.3.2 Procedures/adminis-

~

-0 Ver1tiJation. wi.fl remain operable if venti- tr..ative controls to

(/)

.. fa ti on

• i s

• 1o*s t; *

~ *:~

repair diesels if c: ~ '* ** H > '

damaged "O

"O

....... 4.5.4 Exterior

• ~** : . -

Tanks Enstire safe shutdown can be 2.2.2 Plant procedures for a~tompl{shed'using ~i~~ii~- alternate shutdown p~otected s~stems. ** means III-4.B 4.6 Turbine Missiles P~o~id~ schedule and basis for 2.4 None inspection of lriw-preisure turbines. * ..

III-5.A 4.7.l Jet Impingement on EValuate*and*~d8~e~s effects of 2.5.l None N ,. Target Piping jet impingement regardless of N ratio of pipe 'sizes.

CX> '. '.. *

• 1..

• 4.7.2 Broken-Pipe Impact on Demonstrate deformation asso- 2.5.2 None Target Piping ciated with global strain would not affect function-ability of target pi~e;

• 4.7.3 Detectability Ensure detectability for 2.5.3 Non~*

Requirements through-wall cracks of high-energy piping systems ..

4.7.4 Criteri~ I~plementation (1) Provide criteria and 2.5.4 None results for pipe whip load

' r *.,:.

formulation .

(2) En~ure pi~e ~h1p and 2.5.4 **None jet impingement will not affect containment liner .

Table 2.1 (Continued)

CJ

-s Cl)

VI SEP IPSAR Licensee's Lfcensee 1 s

• 0.

Cl) Topic Section requirements Supplement requirements

l No. No. Title from IPSAR* Section No. from supplement N

(/")

rn 111-5.B 4.8 Pipe Break Outside None

-0 Containment

(/")

c:

-0 111-6 4.9.1 Piping Systems Addressed.~s part 0¥ IE

-0 Bulletin 79-14 effort.

4.9.2 Mechanical :Equipment (1). Supply information 2.6.l To be addressed regardirig motor~operat~d under USI A-46 valve (MOV) lever arms and ljm{ting m6ments.

(2) Staff wi 11 use. Oyste.r 2.6.1 (Open) Same as IPSAR Creek.~nalysfs to d~termine

• seismic capability of ..

~eactor i~ternal

~

s~~~orts.

" ~ ~

(3)'Provide :fu~th~r 2.6.1 Complete information regarding sei sm1c *capability of .

rec i rcu lat ion .puinp *and .

• supports. ***

• 4.9.3 Qualification of Cable To be determined fo 11 owing 2.6.2 Trays comp.1 etion, of owners. gr9up program ..

4.9.4 Ability of Safety- *SEP owners group program will Related Equipment to develop USI A-46 criteria.

Functioi;-i ****

-- '.~ . - ..

111-7. B 4.10 Desi~n C6des, Desigri tfooe 2.7 (Ope~J Complete_

Cfiteria, Load Com- _.. #, ..... . . . *>\""'.***

... *~

binations, and Reactor

~avity D~sign Criteria

• ,,.. _ _ * ._ # -~*

Table 2.1 (Continued) c

-s t'D

(/I SEP IP SAR Licensee 1 s Licensee's

c. Topic Section :requirements t'D _Supplement ,req~irements

J No. .No.* .;

Title ~ *from IPSAR Sec.tion No. from supplement N

Vl

!Tl 111-8.A 4.11 Loose-Parts Monitoring *None '\

~ and Cor.e Barre 1 <

Vl Vibr~tioQ Monitoriri~

c:

-0

-0 111-10. A 4.12.1 Thermal Overloads Bypass thermal overload pro- 2.8-- None

....... tect ion of MOV 0,r ~nsure:

~et poi rit. _adequacy ..

4.12.2 Torque Switches N,one V-5 4.13.1 System Sensitivity Evaluate leakage detection in 2.9.1 None.

conjunction with pipe break inside containment.

N I 4.13.2 Seismic Qualification Demonstrate,reliability and 2.9.2 None w procedu"res corresponding to 0

seismic events ..

V-5 4.13.3 System Testability None V-6 4.14 Reactor Vessel None Integrity V-10.B 4.15 Residual Heat Removal Addressed in SEP Topic Vll-3 System Reliability V-11. A 4.16 Requirements for None *'

Isolation of High- *.. . .  ;.

and Low-Pressure ~ ;

Systems V-11. B 4.17 Residual Heat Removal Addressed in SEP Topic Vll-3 --

System Interlock Requirements

c

-s l'D

(/) SEP IPSAR Table 2.1 (Continued)

Licensee's Licensee's 11 f* 0.

Topic Section requirements Supplement requirements l'D

'  ::J No. No; Title . from* IPSAR Section No. from supplement N

~

m VI-4 4.18.1 L:ocked-Closed Valves Provide mechanical locking 4 Complete

-0 devices and administrative

~ ' . procedures to ensure valve c closure. Verify by Region "C

"C IIL I-'

4.18.2 Leakage Detection Modify procedures for post- 2.10 Modify emergency accident leakage. procedures to address conditions under which remote manual valves should be isolated.

4.18.3 Manual Isolation Valves Provide locking devices for 4 Complete N

I valves.

w I-'

4.18.4 Check Valves as None

~solation Valves 4.18.5 Valve Location* None 4.18.6 Branch Lines With Provide second locked-closed 4 Complete Single Isolation valve.

Valves VI-6 4.19 Containment Leak None Testing VI-7.A.4 4.20 Core Spray Nozzle None

• .i Effectiveness VI-7.C.l 4. 21.1 Breaker Adequacy *verify adequacy of protective - z.*1i..1 No*ne

~elayirig .

Table 2.1 (Continued)

Cl

-s l't)

(J') SEP IP SAR Licensee 1 s Licensee's 0...

l't) Topic Section re_qui r~~ents Supplement requirements

I No. No. Title trorii

. .IPSAR **.

Section No. . from supplement N

VI-7.C.l 4~21~2 'Discdnn~ct Links Provide procedures to verffy . Complete

(/)

rn 4

-0 disconnect links are open.

(/)

c: Complete

-0 4.21.3 Use of Breakers During Provide administrative con.trol -2.11.2

-0 Power Operations to ensure breakers.are.not used during power op_erat ions*,

4.21.4 Op.eration.With Fai.led Limit time for operation with 2.11. 3 Complete Bat_tery ,.... ** * *

• fail.ed battery.

4.21.5 Isolation of Class lE P~ovide short-circuit analysis 2.11. 4 Complete

• *sources from No.n-_Cl ass t.o demonstrate adequate

.*. lE *loads

• i so fat ion.

N I VI-10.A 4.22 Testing of Reactor Trip None* .. ** ~ ~ .:J w

N

• .* System and Engineered

• ~ Safety 'Feat4res ,. lncJ ud- ..,

. ing Reponse-Time Testing VI-10.B 4.23.1 Sharing of DC Systems Prohibit pqralleling of shared 2.12 Complete de systems during power operations.

4.23.2 Diesel Generator Bypass Prohibit pla~ing.DG 2/3 switch in 11 bypass 11 during normal operations.

4. 23. 3 Battery Stp,tus. s*ee. Topi*c: \;f1{.:.*3.B.

Indi cat fon *

4. 23:4 Battery Room See Top_i c IX-5.

Ventilation

0

-s (1)

(J')

o.. SEP

• IPSAR Table 2.1 Licensee's (Continued)

Licensee's

~ Topic Section requirements Supplement requirements N No. No. . Title from IPSAR Section No. from supplement V'>~~~~~~~---'-"'~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~-

m

~-o VII-1.A 4.24.1 Reactor Protection Ensure common-mode electrical 2.13.1 Complete System (RPS} Control faults will not disable neutron Systenis * -. flu~ *nionit'orfrig systems.

4.24.2 Process Cqmputer Ensure RPS ii prritected from 2.13.2 Complete common-mode electrical faults.

4.24.3 RPS Channel Power Instali *tlass lE protection at 4 Complete Supplies interface of_ *.RP*s and RPS power supply.*

• VII-3 4.25.l Proc~dure~ for Shutdown Provide procedures for 4 Complete N

I From', Outside* Control athieving cold shutdown from w

w Room

• outside control room. Part 6f fire protection review.

4.25.2 Use* of *safE!ty-grade Nohe

• System~

• 4.25.3 Residual.Heat Removal None Single-'failure Criter.ia 4.25.4 Inservice Testability Provide procedure's for. testing 4 Complete "o'f shutdown' c.oo.l i ng' sys,.tem .

'. i tempe,tat1Jr_E!

• i *nt~r.l O!=KS. ** ..

VIII-2 4.26.1 Annunciators None 1.. ****

4.26;2 ., Prote~t.iv~*rr'ips* Bypass p~ro'ted:ive trip's **on .. 4..

• ~i e.sels _under acci de.nt tondi tion*s

0

-s Table 2.1 (Continued)

~ ~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~---'-~~~~-

g- SEP lPSAR Licensee's Licensee's

J Topic Section *requirement~ Supplement requirements N No. No. Title . .from IP SAR . Section No . from supplement

~~~~~~~~~~~~~~~~~~~~_;;;_~~~~~~~~~~~~~~~~"--~~~-'--~~~~~~~~-

w" VIII-3.A 4.27 Station Batte_ry Capa- None

~ city Test Requireme~ts c:

"'C

~ VIII-3.B 4.28 DC Power System Bus* Mod1fy exiSting de power 2.14 Complete Voltage Monitoring and

• system monitoring for .

Annunciation

• breaker or fuse position

• and battery availability.

IX-5 4.29.1 Battery Room Ventilation None 4.29.2 Low-Pressure Cooling None Injection (LPCI)/Core Spray and Diesel N

1- Generator Rooms w

~ XV-1 4.30 Increase in.Feedwater None Flow XV-16 4.31 Radiological Conse- -Modify plant technical specifi-* *3-.2 Complete quences of Failure of *cation limits on primary cool-Small Lines Carrying _ant iodine activity.

Primary Coolant Outside Containment XV-18 4.32 Radiological Conse- Modify plant technical specifi- 3.2

• Comple_te quences of a Main Steam cation limi~s'.on primary cool-Line Failure Outside ant iodine ac't i vi ty;

• Containment

Table 2.2 Sources of makeup water

• System Resource Transfer equipment; Storage tanks Clean demineralized 200,000-gallon storage Two clean demi nera.l i zed water tank water pumps Located in turbine building Condensate water Two 250,000-gallon Two condensate makeup pumps, storage tanks (90,000 one condensate transfer gallons of one tank is: jockey pump, two conde~sate exclusive for HPCI) transfer pumps.

Total available for All in turbine bui~ding I.e .. is 410,000 gallons Contaminate de- 200,000-gallon storage Uses same five-pump ~ystem mineral ized water tank* as above River or cooling lake Service water sys- Un~t 2/3 cribhouse Four' service water. pumps, tem piped to fir~ one ~tandby (swing) water main pump--ac_ power Fire water Un~t 1 cribhouse Unit 1 ;_ two screen wash __,

• pumps-,..~*c power Fire water* Unit 2/3 cribhouse Diesel-~n~ine-driven pump,. ..

local fuel and remote reserve fuel Fire water Unit 1 cribhouse Unit 1, two sc'r~en.wash pumps--ac .power .

Fire water Fire hydrants through- Portable, engine~driven pump out the* facility* ... *

• • Dresden 2 SEP, Supp. 1.

3 IPSAR TOPICS RESOLVED BY CHANGES TO PLANT TECHNICAL SPECIFICATIONS During the integrated assessment for Dresden Unit 2, a number of issues were resolved by commitments from the licensee to revise plant technical specifica-tions. The changes and the license amendment associated with the changes are summarized in Table 3.1.

3.1 Topic Vl-7.C.l, Appendix K--Electrical Instrumentation and Controi Re-reviews In IPSAR Section 4.21.4, "Operation With Failed Battery, 11 the staff ~{quired the licensee to provide a suitable technical specification for a limiting condition for operation when a battery system is out of service. The proposed change sub-mitted on February 10, 1984, limits operation to 2 h.ours with a failed battery system. A cumulative time-out-of-service limit (on a refueling cycle bas.is) is made for testing and maintenance.* A second cumulative time-out-of-service limit is made for cell replacement.

• The 2-hour limit satisfies the requirement specified in Section 4.21.4 -0f the IPSAR and is therefore acceptable. The staff reviewed the time-out-of~service exceptions and considers them necessary to permit operation of Units 2 and 3 while routine testing and maintenance are being conducted. The technical spec-ifications were approved.in Amendment 87 to the license on May 31, 1985.

3.2 10 CFR Part 100, as implemented by SRP Sections 15.6.2 and 15.6.4, requires that the radiological consequences of pipe breaks outside the containment be limited to small fractions of the exposure guidelines of 10 CFR Part 100.

In Sections 4.31 and 4.32 of the IPSAR, the staff concluded that the licensee should adopt the BWR Standard Technical Specifications (STS) limits for dose-equivalent iodine-131 (DEI-131) in the primary coolant to minimize the radio-logical consequences of events involving a release of primary coolant outside the containment without significant core damage. Further, the staff concluded that the licensee should develop appropriate plant-specific actions to be taken in the event that these limits a*re exceeded. The limits and actions should be incorporated into the limiting conditions for operation (LCOs) in the plant technical specifications (TS) and, as documented in the IPSAR, the licensee has agreed.

By letter dated February 10, 1984, the licensee proposed revised TS for Dresden Units 2 and 3 for primary coolant activity. The licensee reformatted and revised these proposed TS and resubmitted them on August 2, 1984. The proposed revision of the Dresden TS would incorporate an LCO for the primary coolant activity that limits both the equilibrium and maximum DEI-131 concentration to those contained

. in the STS, in accordance with the staff's recommendation in the IPSAR. The Dresden 2 SEP, Supp. 1 3-1

action statements proposed by the licensee ensure that appropriate corrective

• measures will be taken when primary coolant activity is so high that the poten-

• tial for an accidental release of primary coolant with ah activity higher than '

the equilibrium limit is acceptably small. On this basis, the staff concluded in Amendment 87 to the license, dated May 30, 1985, that the proposed revision to the Dresden TS is acceptable and, therefore, the issues identified in Sections 4.31 and 4.32 of the IPSAR are resolved.

Table 3.1. Modifications to Dresden*Unit 2 Technical Specifications 1

I PSAR Sect ion . Subject

• Amendment No.

4. 21. 4 Limiting condition for operation 87 with a battery system out of service 4.31 Radiological consequences of failure 87 of small lines carrying coolant outside* containment (limits on primary coolant activity) 4.32 Radiological consequences of main 87 steam line failure outside contain-ment (limits on primary coolant activity)

Dresden 2 SEP, Supp. 1 3-2

4- IPSAR TOPIC RESOLUTION CONFIRMED. BY NRC REG.ION Ill OFFICE During the integrated ass*esslilent. for .Dresden Un*it 2, a number of issues were reso*l ved by commitments made by the licensee for specific pl ant modifications o*r procedural changes. After the staff issued the IPSAR for Dresden Uni't- 2, the Region IIL sJ;aff _was _asked ~hrough Task Interface Agreement 83-'45, to verify that .plant modifications had bee*n implemented and to review *change~- -that the 1i cerisee had _made tO' pl ant 6pe*r~t i ng procedures. Region I II has al so been asked to confirm 111~difications and procedures resulting from the te*s.olut-ion C)f jssues contained* in t~is supplement to the IPSAR. Table 4.1 li~ts IPSAR and IPSAR supplement act ions that Region I II ha.s been asked to confirm.

Region lit:. staff conducted onslte inspections for each 1tetir identified iii Table 4.-1 as. complete .. The inspections consisted of examin~tions of installed equipment* as. well as a review of supporting procedures and other documentatiOn.

Regibn III s:t_aff cohcli.Jded that the. licensee had met the commitments documented in the 1-PSAR for the items in Table _4.1 where the.status is designated*as 11 com-plete.11 InspectiOn findings with the results df the_se r'eviews are: documented in the inspection reports identified in Table 4.1. The remain*ing items. identified as open will b~ closed. in futur~ i osp~l ion reports.

I .

. *i

i .

• ** * *HJ~-~-f~*,..

Dresden 2 SEP, Supp. "*:lfT'?':' ,' *": *:

- -- -- - . :- *~-:.. -

Table 4.1 Items for con.firm~tion .bY NRC Region III Office

• Item

• NC:L : -:D'esc.ri (1) ptfon of confirmation Install scuppers to pr~vent ponded w~ter on roofs:

SEP/IPSAR ref ere nee Topic l_l-3.B, 4.1.3

• I Status (inspection,_

Report No.).

Complete .{85~39)

(2) Revise emergency plan to Topic IF3'.B.1 *. Pa'rtiai. (8,S~30) cope with design-basis 4.1.4 . . . ' , . . ,, ' ',t:- .* .. '

flooding.

(3) Modify procedures for Topic III-3. C, * 'Complete '(83-32) inspection -0f water control 4.4.3.

structures*.* * * .' * ' .I (4) Review containme~t penetra- Topic VI-4,' Open* .. ., :.,

tions and provide*locking* 4.18.1" . 'Level gaug¢ installed in devices and administrative Unit 2/3 cribhouse not control procedures as . *~ .

• confirmed

... t'. ;*:*

necessary~ *

.

"-*~ *.: '* : ~ - *:.{.

(5) lock valves closed.and modify Topic VI.:.4, P.art i a1. (a*5_..:3Q) procedures for manual isota- 4.:18.,3 .. , .Conf i rmatt6n:. not comp 1ete tion valves identified.

' **, I )*

..... - ' *.' * . * . ** (~, I *' *

(6) Provide second locked closed Topic VI-4, Partial (89~30)

• isolation valve on identified 4.18.6 ... Redundant' valve i n*sta 11 a lines.

• tion not .confirmed * * : t I<

(7) Provide procedures to en$ure. Topic v1:7;c.;1_", *:,

Open* '

.~ '

disconnect links are*pro~e~l~ 4. 21. 2 '

-. I .

~*:  ; ..

• positioned following main- .  :: .. _ ;*

tenance.

(8) Provide procedures to addres~ Topic;:* VI-7.C~l, Complete (83;.:32) use of bre~kers 252-2~29 4. 21. 3 .' . .  !.. ' ' ' ' .

during power operation:

(9) Install Class lE protection Topic'vfl-~*.A,_ *. Complet~. (~3~32) between RPS power supply and 4.24. 3 ..

RPS.

(10) Implement*~rocedures,for. .*. Topic VII-3, testing shutdown cooling sys* 4.25:4 .'

tem temperature interlocks. ~\ *.. , '. . . '* .

. . . . { '  :

(11) Bypass DG underfrequency Topic VIII-2, Partial _(85-30) protective trips during 4. 26. 2* . . " DG 2/3. not confirmed emergency operations.  : , ~ ,r (12) Provide monitoring of de Topic VIII-3.B, Partial (85-30) system in control room. 4.28 Local de monitoring scheme installed not confirmed Dresden 2 SEP, Supp. 1 4-2

• • • Item No.

Table 4.1 (Continued)

Description of confirmation SEPIIP*SAR ref ere nee

. Status (Inspect i cin .

Report ~o... ). ..

(13) Confirm that the plant pro- Topic.III-2,.

• Qpen cedures adequately address 2.2.2 (Suppl. 1) al.ternate means of shutdown if components not enclosed in .! .  :

qualified structures are loit as a result of wind and tornado l oadi rg.s.

(14) Confirm that the license~ Topic III.:.4.A, . Op~n .

has the equipment necessar:y 4.5.3 and on site to repair or remove 2.2.2 (Suppl. 1).'

the damaged components of the diesel generators. . ...

(15) Confirm that the licensee has Topic III-6, Ope11 . ,

implemented a plant-specific 4.9.3 analysis of the structura~

integrity of cable trays to

• ensure their ability to main.-

tain and that the cable tray support systems have been .>

modified where necessary. '*'

(16) Confirm that the licensee has Topic VI-4 ,,t Open installed proper leak-rate 4.18.2; ' ' "

test taps on the .reactor Topic 'VI-6,. '

building closed cooling 4.19 . ~ .

i

• '1, water lines.

(17) Confirm that the leakage con- Topic VI-4 Open ditions under which the remote 4.18.2 and manual isolation valves on the 2.10 (Suppl. 1)

LPCI and core spray systems should be isolated are incor-porated into the emergency procedures.

(18) Confirm that the licensee no Topic IV-10.B, Open longer permits paralleling of 2.12(Suppl .. 1)  ;

the 125-V de and 250-V de sys-terns during power operation of either Dresden unit and that ground detection procedures fQr both unit~ have been* revised

• accordingly.

Dresden 2 SEP, Supp. 1 4-3

Table 4.1 * (Conttnued)

Item SEP/IPSAR Status (Inspection No. Description of confirmation reference Report No.)

(19) .Confirm that operating *proce- Topic VI-:10.B, Open .*

... (" :; dures have; been:" changed' to' re- .. 4. 23. 2

.:** quir*e a nor*mal;'*n'orm~l .* a*Ugn~.-.

11 11

. . . .: "

• ment: of tne generator 2/3 * . ., *

,{ .:_*i . .;: normal /bypass switches. . .. ; -~. " *.

! ...... ~ ,~: . .' ..* . : . ' . . . . .* . ' . ' . ..

(2or :.:~*Conti r~':;that th~ licensee* has * . Topi"c VII-1. A, Open u.. * .. **

.,, . . :'. fo'stall ed. Cl ass* IE.signal"** *4. 24.1 and . '~' -, :: .

• ,, ..,_ : j sol afi on dev*i ces at the.. '

• 2.13.l*

inputs of each control room (Suppl. 1) retarder that monitors the RPS as committed to in the January .9, 1987 *and February 2, 1989 lett~rs. * *

(21) Confi~m that the licensee has Topic VII-1. A, Open provided isolation between the 4.24.2 and APRMs and the process computer 2.*13.2.

by the installation of "flying (Suppl. 1) capacitors. 11 *

(22) Confirm that the bypass diesel Topic VIII-2, Open generator underfrequency trip 4.26.2 modifications are completed for diesel generator 2/3.

Dresden 2 SEP~ Supp. 1 4-4

5* *

SUMMARY

Within the framework of the SEP, the staff revi~~ed 27. top.ics where ce.rtai n ~s,;..

pects of the Dresden Unit 2 plant design were. found to* difter from the current criteria.**** The staff's review of these topics is* contained iii the Integrated Plant Safety Assessment Report (IPSAR). Of the 27 topies ad.d.ressed in the IPSAR, 14 topics (25 issues) required refined engineering analysis or ~ere continued under an ongoing E!Valuation;. This supplemen.t closes all but two of the.open;.

topics. These remaining topics will ..be addressed in the. staff's SER fo.r~ the. . .

conversion of the Dresden Un.it.2 provisional operating license.to.a full.,term .*

operating license. * * * '* .,  :  :* .

.... *( * .."'

. r \

_i: .

,:'l * *,,r,* I*,  ::1',

.. ~ .. ; ' ~* \ ~ i"'.<";

• Dresden 2 SEP, Supp. 1 5-1 .* ,*

6 REFERENCES Lette~s December 15, 1980, from D. M. Crutchfield (NRC) to J.S. Abel (CECo),

Subject:

SEP Topic II-2.A April 24, 1981, from D. M. Crutchfield (NRC) to J. S. Abel (CECo),

Subject:

SEP Topics V-10.B~ V-11.B, and VII-3 .(Safe Shutdown Systems Report)

June 30, 1981, from D. M. Crutchfield (NRC) to J. S. Abel (CECo),

Subject:

SEP Topic IX-3 March 23, 1982, from T. J. Rausch (CECo) to P. O'Connor (NRC),

Subject:

SEP Topic II~-4.A, Tornado Missiles June 11, 1982, from T. J. Rausch (CECo) to P. O'Connor (NRC),

Subject:

SEP Topic~III-2, Wind and Tornado Loadings August 2,* i982, from T. J. Rausch (CECo) to P~ O'Connor (NRC);*

Subject:

• .Dresden 2 SEP Topic III-7.B, Design Codes, Loads, and Load Combinations August 23, 1982, from T. J. Rausth (CECo) to P. O'Connor (NRC)~

Subject:

Dresden 2 SEP Topic III-5.A, High Energy Pipe Break Inside Containment September l5, i982, from P: O'Connor (NRC) to L. DelGeorge (CECo),

Subject:

Dresden 2 IPSAR Topic III-2, Wind and Tornado Loadings v

September 16, 1982, from P. O'Connor"(NRC) to L. DelGeorge (CECo),

Subject:

Dresden 2 IPSAR Topic III-4.B, Turbine Missiles September 21, *1982, from P. O'Connor (NRC) to L. DelGeorge (CECo),

Subject:

Dresden 2 IPSAR. Topic III-5.A, Effects of Pipe Break on Structures, Systems,.and Components Inside Containment I .

October 5, 1982, from T. J. Rausch (CECo) to P. O'Connor (NRC),

Subject:

Dresden 2 IPSAR Topic VIII-3.B, DC Power System Bus Voltage Monitoring and Annunciation *

• October 8, 1982, from T. J. Rausch (CECo) to P. O'Connor (NRC),

Subject:

Dresden 2 SEP Topic III-4.8, Turbine Missiles November 17, 1982, from T. J. Rausch (CECo) to P. O'Connor (NRC),

Subject:

Dresden 2 SEP Topic III-5.A, High Energy Pipe Break Inside Containment November 18, 1982, from T. J. Rausch (CECo) to P. O'Connor (NRC),

Subject:

Dresden 2 SEP Topic VI-4, Containment Isolation Systems Dresden 2 SEP, Supp. 1 6-1

-::- - - - - _--=:....;;; - - ---- :, ---=..- --;:::;---=-...::-=- = --- :=.;; -:--_-

November 22, 1982, from T. J. Rausch (CECo) to P. 0 Connor (NRC); '

Subject:

1 Dresden 2 SEP*Topic III-2, Wind and Tornado Loadings Novembe~ 24, 1982, from Z. Zudans (Franklin Research Center) to C. Grimes (NRC),

Subject:

Dresden 2 Review of SEP Integrated Plant Safety Assessment Report December 6, 1982, from T. J. Rausch (CECo) to P. 0 1 Connor (NRC),

Subject:

Dresden 2 Integrated Assessment Jar:i.uary.:10, 1983, .from T. J. Rausch (CECo) to P. 0 1 Connor (NRC), SUbject: *

• Dresden 2 SEP Topic III-5.A, High Energy Pipe Break Inside Containment January 19,.1983, from D. M. Crutchfield (NRC) to L.* DelGeorge (CEC'o),'

Subject:

SEP Integrated Assessment Status for Dresden Unit 2 ~

February 3, 1983, from. T.. J. Rausch (CECo) to R. Gilbert (NRC), Subj,ect:

Dresden 2 SEP Topic III-4.A, Tornado Missiles

• February 9, 1983, from B. Rybak (CECo) to R. Gilbert (NRC), *Subjectt:* D.resden 2, IPSAR Topit* III-2, Wind and Tornadq Loadings March 16, 1983, from D. M. Crutchfield (NRC) to D. L~ Farrar {CECo):~ Subjett~

Dresden 2*, Topic III-4.A, .Integrated Assessment Follow-up Item (Tornado Missile and Pipe* Break Inside Containment) ,_  :..

April 7, 19~3, from D. M. Crutchfield (NRC) to D. L. Farrar (CECo}/

Subject:

Cycle 9*Re1oad--:Dresden .statiOn, Unit 2 . . * ,._,, ..

Apri1'18,.l983, fro.m D. G. Eisenhut (NRC) to C. E. Norelius (NRC),

Subject:

Dresden 2 SEP Action Items * * *

... . . . *~.. .-

May ,11, 1983, from D. M.. Crutch.field (NRC) to D. L. Farrar .(CECo),

Subject:

TMI Item III.D.3 ..4, Control Room Habitability May 31, 1983, from D. M. Crutchfield (NRC) to D. L. Farrar (CECo),

Subject:

Integrated Assessment Followup Item--Turbine Missiles (IPSAR Topic ILI-4.& *

(Section*4~6~-NUREG-0823)i Dresden Nuclear Power Station, Unit 2* *.

July 22, l983, from B. Rybak (CECo) to R. Gilbert (NRC),

Subject:

Dresden 2, IPSAR Topic III-6 (Section 4.9--NUREG-0823)

Augu'sf 15, 1.983, from R. Rybak (CECo) to R. Gilbert (NRC),

Subject:

• Dresden 2 IPSAR Topfc VII-1.A, Isolation of the Reactor Protection System From Non-safety Systems, Including Qualification of Isolation Devices Sep~ember 12, 1983, from B. Rybak (CECo), to. R. Gilbert (NRC),

Subject:

Dresden Station.Unit 2 Integrated Assessment Followup Item, Pipe Break Inside~contain ment, SEP Topic III-5.A, Section 4.7.4, NRC Docket No. 50-237 October.3, t983, from B. Rybak (CECo) to R. Gilbert (NRC),

Subject:

Dresden Station U~it 2, SEP Topic III-5.1, Section 4.7.4, NRC Docket No. 50-237 October 18, 1983, from B. Rybak (CECo) to R. Gi.lbert (NRC),

Subject:

.Dresden i IPSAR Topic III-4.A, Tornado Missiles Dresden 2 SEP, Supp. 1 6-.2

October 27, 1983., from D.. M. Crutehfield (NRC) to 0. L. Farrar (CECo),

Subject:

IPSAR Topic 111-5.A (Section,4.7, NUREG-0823), Effects of Pipe Break on Struc-tures, Systems, and Components Inside Containment for Dresden Nuclear Power Station, Unit 2 * **

November 3, 1983, from B. Rybak (CECo) to R. Gilbert (NR.C),

Subject:

Dresden 2 IPSAR Topic 111-10.A, Thermal Overload Protection for Motors of Motor-Operated Valves . ** . *. * * .;

November 4,J.983, from B. Rybak (CECo) to R. Gilbert (NRC),

Subject:

• Earthqua~.e Analysis of Vent Stack November 4.,. 1983, from B. Rybak (CECo) to R. Gilbert (NRC),

Subject:

IPSAR Sect ion 4. 3. 2 December 30;_ 1983; from B. Rybak (CECo) to R. Gilbert (NRC),

Subject:

Dresden*

Station Unit 2 SEP Topic VI-7.C.l, Docket No. 50-237 January 9, 1984, from B. Rybak (CECo) to R. Gilbert (NRC),

Subject:

Dresden 2.

IPSAR Topic 111-6, Qualification of Cable Trays January 24, 1984, from D. M. *Crutchfield (NRC) to D. L. Farrar. *ccECo),

Subject:

i; NUREG-0823, *Section 4.21[.1]--SEP Topic VI-7.C.l, Appendix K, Electrical In~tru- 6 mentation and Control (EI&C) Re-reviews

• February 1, 1984, from W. D. Shafer (NRC) to c.* Reed (CECo), Routine safety *

• inspection conducted by Messrs. J. M. Tongue and S. Stasek .and Miss C-; D.

Anderson during period of 11/18/83 through 1/19/84. Docket No.s. 50..;10, ~0-2.37.

and 50-24.9 ',

February 10, 1984, from B. Rybak (CECo) to H. ,R. Denton (NRC),

Subject:

Dresden Station Units 2 and 3 Proposed Amendment to Technical Specifications for Facility Operating License Nos. DPR-19 and DPR-25 125/250 Volt Batteries and Reactor Coolant Iodine Limits, NRC Docket Nos. 50-237 and 50-249 February 13, 1984, from D. M. Crutchfield (NRC) to D. L. Farrar (CECo),

Subject:

IPSAR Topic V-5 (Section 4.13, NUREG-0823), Reactor Coolant Pressure Boundary (RCPB) Leakage Detection, and VI-4 (Section 18.2, NUREG-0823), Containment Isolation-,,.Leakage Detection for Dresden Nuclear Power Station, Unit 2

• March 8, 1984, from D. M. Crutchfield (NRC) to D. L. Farrar (CECo),

Subject:

Dresden 2 lPSAR Topic III-7.B, Design Codes, D~sign Criteria, and Load * *

• Combin-ations for Dresden 2

• April 14, 1984, from B. Rybak (CECo) to R. Gilbert (NRC),

Subject:

. Dresden Sta-tion Unit 2~ Response to IPSAR Open Items Section 4.13 RCPB Leakage Detection -*

Sectfon 4~18.2* Containment Isolation Leakage Detection, NRC Docket No. 50-237 June 12, 1984, from B. Rybak (CECo) to R. Gilbert (NRC),

Subject:

Dresden 2 IPSAR Topic 111~4.A, Tornado Missiles * *

• June 26, 1984, from I. M. Johnson (CECo) to T. Murley (NRC),

Subject:

Dresden 2 IPSAR Topic III-4;A, Tornado Missiles Dresden 2 SEP, Supp. 1 6-3

~-~--=----=- -:;=- -- . - - -* ~--;-;-,;:--- - ----::;-;::~**----=:::-:.-*--:;- __

July 11; 1984~ f.rom .B. *Rybak (CECo).: to R*.:"Gi lb~rt (NRCJ,

Subject:

': '.::*o_resden' ".*:

Statiorf;Unit 2{ SEP' Topic *III-7:8" * * * ,.. * * * . . . . *:

• Augu.st 2, 1984_, fr.om B. Rybak (CECo) to .. H. R. Dent.on (NRC),

Subject:

. Dresden S-tatioil' Urdts* 2 and *-J*, *,Reformatted T~ctlhiea 1* Sped f.iC:at ions Amendme.nt.s ~ NRC

• Docket *Nos*~*>s0;.*231 a:nd '50-249 * * * ~-" * ** * * * * " '*

November 21, 1984_, from B. Rybak (CECo) to R. Gilbert (NRC), Subjec_t: . Dresden 2 IPSAR Topic IIJ-4*;8, Turbine Missiles * :* * * * * :. " * *

• :_:;\'~-  : * * ,:/

• 0 * '.t  :; .,. '* J .\?' H '\~ 0 December 28, 1984, from D. Skolnik (CECo) to R. Scholl (NRC), Subj'ect:

• Dres"den 2 IPSAR Topic III-6

• ~ .* ,; ' ** .j. '..: _:... ~.:~ .~ " . f1,"  :. ~---  ;*:  : * * * . * *_.  :-  :};-~ .~ *~ .. .... * ,' .' -,.~;- * *'.

J~~~uary 2, 1985, from" D~ Skolnik (CECo)'to R. *Scholl '(NRC), Su'bjec.t":'*** SEP Topic VII-1. A

,. ~ *. .  :- . ' ~  : ' .

,,. *.' : ~ l January 2, 1985, from B. Rybak (CECo)* to* R:* 'Gflbert (NRC), *

Subject:

• Dresden 2 IPSAR Topic III-10.~,. Thermal Oyef'.load Prote.ction for Motors of Mo_:t9r-Operated.

Valves:--*':~ .. *.* *-*::.". *. * ** .- *.** * '* .** * <<~! .. . * ** .... 1

".:*.* '.:!';',': '\'(::.J ,~.:". . ......... i':- ';,; *. ~-':*:  :-- *. : t-::*,*~ ..

March 6, 1985, from B. Rybak (CECo) to R. Gil~ert (NRC),

Subject:

JP~AR.

Topic VI-7:C.1(Sectfon4.21.5-NUREG-082.3)' * ,:;, *..

";.*,' -~*;:: ":.";(.'.'"!, ,*: ..*:-. *  :::* f:"..';*~* -; " :i*- '!.**: * "'.* ,; . *I *

• _r!,.~: *,... ......  ;.*

May 30, 198_5, from.B. Ryb~-~ (CECo) to R... Gilb.~rt (NRC),

Subject:

. ~P.~AR TopicVI-7~C.l {Section '4.21.5~NUREG-0823Y:* \ ' .. *, **. ' . *' . ~-~ "

. . -* '; *' ** * ;. :  :

• i.:. ~:,  ;. ~ . . l *  ; " * . * . ' * ' .! .. .-

• July 9, 1985,- from J. A. Zwolinski (NRC) to o. L. Farrar"(CECo},

Subject:

• lPSAR Topic I;I-10.A,_(Sectio~ 4~12.1, NUREG-0823) . . * ' .

• : ,,:*. z~ i:*. -.- ._ ;  : ** *; ; :. . * *,. . * . , :* * * * .:* : * :*: ._-*i * . *; *, ~-'1 *. . .. ,

July 9*t198S-~* froin J'~ A: Zwolinski* (NRC) *to *D'.' L. Farrar' (CECo}~ s*ubject:

• IPSAR Topic VI-7_.~*h .(Section 4 ..?1.5, NUREG-08?.3) .. .

! * *  :~. c:~:,:~. *~.. ' I * '.:.,:\ j ,  :. ,.:~. *._ ,"  :* '\ * .(  ; * * *~** {- , ,.

January 9, 1987, from J. R. Wojna:*rows'ki.. -'(CECo) tcr H. R. Oentorf *'(NRC-}i/

Subject:

Dresden 2. IPSAR Topic VII-1.A, Isolation. of *the Reactor Protection System from N'.o:n::,~~~.~~~~ -~y~t.~IJ)s*" . * * : * . * * :. * * ~: . j .-'

1

-':; *. * ' . . * * , ,, .* -~ <' .... *, *.*

January 15, 1987', from J. R. Wojnarowski (C~Co) to H. *R.' Denton (NRC}, "

Subject:

Dresele.n_.. 2 SEP Schedule f~r SubmJttal of Remai.ning Open Topics *

':~:1_.~*-11 , , *;: :. _r * .. , ~ *. ::*_~'=,: .,. *'  ::* . . *:.1; ~,.****- .' .' . . ..

Apdl'20,**1987, from l. M.**:John.son '(CEC.o) to T'.' L Murle{ (NRC), s*ubject:

• Dresden 2 IPSAR Topic*,:uI:.1~ Classif'ication*;of *Structures*, 'Systenis,' and

• Components.. .

.; *.. ~ ~ . : . . ) ;.":...: ~.

May 12, 1987, from I. M. Johnson (CECo) to T. E. Murley (NRC),' Subjett:

Dresden 2 IPSA.R Topic VI-7.C.1, Appendix K, Electric.al Instrumentatio.n and Control Re;;;reviews*' <* <.-* :::** * * *'* \":. " * : .* * * * * * * ****.. *

• j,**

June 9, ,1987, from I. M..J.ohnson _(CECo), to T.. E. Murley (N.RC),

Subject:

Oresden 1'*2 IP-SAR Topic'* VI-10. B,:'Shared Eng}heeri rig Safety Features, pris.ite Emergency Power, and *Ser*ilite*"System 'for 'Multiple Unit *Faci'liti e*s . ._ '

Junk "'26~. 1987, from 'l. M-; ":J'bhnson ('CEC.o.) 'to' T.: E: ~uri ey (NRC) / SubJe.ct: TPSAR

• Topic III-4.A, Tornado Missiles * ,-, * '* * *. ,. * * * ** " **,_

~  :: .

Dresden 2 SEP, Supp. 1

~****7

August 11, 1987, from I. M. Johnson (CECo) to T. E. Murley (NRC}, *subject: .

Dresden 2 IPSAR Topic VIII-3.B, DC Power System Bus Voltage Monitoring :and*

Annunciation * * *

• November 20, .1987', .from D. R. Muller .(NRC)to \ ..

• o. Bu.tterfi_elci (CECo),

Subject:

Dresden 2 IPSAR Topic VII-1.A, Isolation of the Reactor_;Protection Sys.t.ern '.(RPS).

from Non-safety Systems *

.i.

December 9, 1987, from I. M. Johnson (CECo). to T. E. Murley (NRC},* Sµbject: . : : ,

Dresden 2 IPSAR Topic 111-1, Classification of Structures,*systems, and Components .. ,, , ~ *. , .

' I ' ~ * * "

December 22, 1987, from M. Grotenhuis (NRC) to L. D. Butterfi~l~ (CEt~)~* ~ubj~ct:

Dresden .. 2 I PSAR Topic VI-10. B, Shared .Engineered Sa*fety Features .

January 12, 1988, from M. Grotenhuis (NRC) tol::D. Butterfi~ld (CECo),

Subject:

Dresden 2JPSAR Topic VIII-3.B, DC Power.Systems * * *' *

• March 25, 1988, from I. M. Johnson (CEC~) to T. E. Murley (NRC), Subject~ ...

Dresden 2 IPSAR Topic VIII-3.B, Response: to Request for Additional Information. ._,

.-*. l June 27, 1988, from L. Norrholni (NRC) to H. Bl i:ss (CECo),.

Subject:

. Dresden 2 .. * *,.

IPSAR Topic VIII-3.B, DC Power Systems Bus Voltage Monitoring .~nd A~nunciating

.l*

June 28, 1988, from B. Siegel (NRC) 'to H. E. Bli'ss (CECo)~.

Subject:

Dresden 2 IPSAR Topic 111-1, Classification of Structures, Systems, and Components-- *

• Radiography Requirements ** ,

• '. :.*1 ,,

September 28, 1988, from I. M. Johnson (CECo) to T. E. Mur.ley (NRC), Sub}ect:

Dres~en 2 IPSAR Topic VI-10.B, Shared.Engineered Safety Evaluation F,eat~res January 6, 1989, from I. M. Johnson (CECo) to T. E. Murley (N*R.C),

Subject:

Dresden 2: IPSAR Topic 111-1, Fracture Toughnes~ * ~.

February 2, 1989, 'from I. M. Johnson (CECo) to T.

• E. Murley (NRC)., Sl!bje~t: * . .

Dresden 2 IPSAR Topic VII-1.A, Isolation of the Reactor Protection System {RPS)

From Non-Class.IE Equipment *

• March 1, 1989~ from D. R. Muller (NRC)' to- H. E. s1'iss (CECo) ,* Sub'jec*t: . A$ME Code Case. N-4il, Alternative Damping .Values for Seismic Analyses of'C.l asses 1, ..

2, and 3 Pi.ping and Alternative Seismic Eva}uation. Criteria * *:* . , ,

April 11, 1989, from J. A. Silady (CECo) to T. E. Murley (NRC),

Subject:

Dresden 2 Status -0f the SEP * *

• April 17, i989, from. M. H. Richter (CECo) to T.. E.* Murley' (NRC), Subje~t:

Response to Station Blackout Rule

• May 1, 19g9",' from D.. R. Muller (NRC)_ to .T~ J. Kovach (CECo}, S_ubjec~: *Dresden -2 IPSAR Topic 111-1, Fracture Toughn~ss Evaluation for Dresden 2 May 31, 198~ *. from R, Stolz (CECo) to T. E. Murley (NRC),

Subject:

. ._Al~ernativ.e Damping Value~ for Piping Seismic Analysis Dresden 2 SEP, Supp. 1 6-5

July 26, 1989, from 8. L. Sfegel (~RC) to T. J. Kovach (CECo), Subjec~: Request for Additional Information Related to the Review of CEC0 1 s Responses to SEP Topic III-7.B, Design Codes, Design Criteria, and Loading Combinations for Dresden 2 August 30, 1989,-from J. A. Silady (CECo} to T. E. Murley (NRC),

Subject:

Dresden 2 Updated Status .Report and Su_pplemental Response on. SEP topic III-7.B

\'*

Memorandum October ir, *1983, *from 0. Parr *(OSI) *to C.c: Grimes (DL) tr~~~mitti'ng: SER on

• To*rnado: Mi'ssi~e Risk Analysis " > ,

~ . *~ . .

Other., Reports

• Franklin Research Center',' Technical Eval~atf~n Report, TER-C5506-427, "Review of Wind and Torn-ado Loading Resp~>nses; Tornado* Wind Load and Vi.brat ion Analysis. of the Ventilation_ :stack_;..".'Dresden .2, Februar,y 1'3, 1984 Franklin,.Re*search Ce.nter, Technical Evaluation 'Rep_ort, TER..,C5506-425, 11 Fjnal Supplementary Report Review of Licensee Responses to SEP Topic III-7.B, Design

.Codes, Design*Criteria; and-loading Combinations for Dresden 2, 11 Ju_*ne'.3, 1986 U.S. Nuclear' Regulatory Ccimmiss ion, 1983, . '~Integrated Pl ant* Safety Ass~ssment, Systematic Eval~ation Program, D~esden Nuclear Power Station, Unit 2, Common-wealth Edison Company,.*1Docket No. 50-237-, .. Ffoal Report, 11 NUREG-0823, February*

1983 . . .. . . ' ' ,, . . ..... . .

. ' . .:~'

. i *" ~

$:* ... * ',*.{'

.';I

.~ ' . .; -. \ . ~

. ,. :: \ .  : ~

• . 1.

. . -.  !...... ~ , , ::.

. **,. I:**: *:' .

• Dresden 2 SEP, supp. 1 6-6

• ~

• I *

~ ., .- .

• • i .. **

APPENDIX A . * :.  :.. *' .. :

. . ' ~ ( . .. ' : .q NRC STAFF CONTRIBUTORS AND CONSULTANTS

.' ,* . ~- . ~ .. '~ . :** *: :.:

The safety ~valuation report supplement is '.a product of the NRC staff:*and* its;:.*

consultants. The NRC staff members listed below are.prin<:ipalc.ontri.butor:s to:

this report. A list of consultants follows the list of staff members .

.* . : ' ;'!-* ~

*.  :*:_t Section NRC Staff Title

-~

2.. i.1' M. Hum *: . ,M~ter.i ais<Engi*~~~:r, ~* ,,, \: '. :,.,'.

2.1. 2 B. Elljot .... , * .. S.enio.r Materials<Engineer .,,,.:

2.2, 2.3, 2.4, 2.7 S. Chan Senior Civil Engineer *

2. 3; 2* .. 4 .I .,. P. Chen .'. * :Mechanical* Engineer  : '.';-:"  :*',

2.4 ' ' ' .G. Cwalina. * .. , '*Section- Chief. ..,. * "-:.'~"

2. 5. 1, 2. 5. *2 : ' R. Lee -'  :

• i. ,

• Project Mana*ger. *;, * "* *,

2.5.3, 2.5.4 R. Hermann Section. Chief 2.6. *': .* A.. tee .. : . . *. Mechanical E11gi neer,, .* . **:. j 2.8*.**' :, R.. Scholl :  ::SIMS: Coor.di nator . ! ~*.

2 .. 9. . D~ Chery~. Section .leader... *  : *. ! .

2.10 R. Gilbert Project Engineer 2.11.2, 2.11.4, 2.12 N. Trehan Electrical Engineer 2.13.1 B. Marcus Electrical Engineer 2.13.1, 2.13.2 J. Stewart Electrical *Engineer

2. 14' 2 .11. 1, 2. 11. 3 J. Knox Senior Electrical Engineer Additional NRC Contributors NRC Staff Title C. Grimes Director M. Boyle Integrated Project Manager S. Brown Reactor Systems Engineer T. Cheng Senior Structural Engineer S. DuPont Senior Resident Inspector E. McKenna Senior Integrated Project Manager T. Michael$ Senior Integrated Project Manager D. Persinko Senior Technical Assistant Consultants IPSAR supplement Name _Company Section D. Barrett Franklin Research Center 2.1 T. Stilwell Franklin Research Center 2.7 Dresden 2 SEP, Supp. 1 1 .

• . Appendix*A.

NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER (2-891 IAal;ned by NRC. Add Vol.. SUPP .. Rev.,

NRCM 1102, - Addendum Numbers. II *nv) 3201, 3202 BIBLIOGRAPHIC DATA SHEET (See instructions on the reverse)

IRE AND SUBTITLE NUREG-0823 Supplement No.

Integrated Plant Safety Assessment

3. DATE REPORT PUBLISHED Systematic Evaluation Program MONTH VEAR Dresden Nuclear Power Station, Unit 2 October 1989

4. FIN OR GRANT NUMBER

5. AUTHOR(S) 6. TYPE OF REPORT

7. PERIOD COVERED llnclu*ive Dare*/

B. PERFORMING ORGANIZATION - NAME AND ADDRESS f/f NRC. provide Division. Office or Region. U.S. Nucl**r Regularory Commission. *nd m*iling Mldress: ii conrracror. provide name 1nd mailing Mldreu.J Division of Reactor Projects - III, IV, V and Special Projects Off ice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

9. SPONSORING ORGANIZATION - NAME AND ADDRESS !If NRC. rype "S.me *ubo..,"';if conrrocror. provide NRC Divi*ion. Office or Region. U.S. NuclHr R°egul*rory Comminion.

1nd mailing address.)

Same as 8 above

10. SUPPLEMENTARY NOTES cket No. 50-237 STRACT (200 word* or leul The Integrated Plant Safety Assessment, Systematic Evaluation Program for Dresden Nuclear Power Station, Unit 2 was prepared by the U.S. Nuclear Regulatory Commission to provide the framework for reviewing the design of older operating nuclear reactor plants to reconfirm and document their safety~

12. KEY WORDS/DESCR:PTORS ILiir ""'rrls orph...-1 rh*r wm-1ir ,.,..*rch*n in loc*rlng th* repon.I 13. AVAILABILITY STATEMENT Unlimited Integrated Plant Safety Assessment 14. SECURITY CLASSIFICATION Systematic Evaluation Program . (Th/1P*I Dresden Nuclear Power Station, Unit 2 Unclassified (Thi1 R*ponl Unclassified

15. NUMBER OF PAGES

16. PRICE

UNITED STATES SPECIAL FOURTH*CLASS RA TE POSTAGE & FEES PAIO

. NUCLEAR REGULATORY COMMISSION USN RC WASHINGTON, D.C. 20555 PERMIT No. G*67 OFFICIAL BUSINESS PENALTY FOR PRIVATE USE. $300