ML20148F989
| ML20148F989 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear, Vermont Yankee, Crystal River File:NorthStar Vermont Yankee icon.png |
| Issue date: | 09/14/1978 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1532, NUDOCS 7811100068 | |
| Download: ML20148F989 (466) | |
Text
{{#Wiki_filter:_ _ _
*1*
,,h p
ha TABLE OF CONTENTS E [,.J,;J" a[s u D" 93 g n MINUTES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 h y3Qy[jj ]u]s$ 5 g ( 14 r ,fgf 9
* *{D'9" d g;- :'
WASHINGTON, DC BgS- MyA I. D$ Of7f Ch a i rma n ' s Re p o r t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 A. Reviewers ....................................................... 1 B. New Nucl ea r L.i ce ns i ng tii l l s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 C. ACRS Fellows Program ............................................ 2
- g. D. Testimony of Comnittee Before Senate Subcommittee o n Nu c l ea r Re g u l a ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 E. Vermont Yankee Vs. National Resources Defense Council . . . . . . . . . . 2 II. Meeting on Arkansas Nuclear One, Unit 2, Nuclear Power Plant ........ 2 A. Subconaittee Rcport ............................................. 2 B. Appli cant 's Pres e nta ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 1. Introduction ................................................ 4
- 2. Core Protection and Cal culator Sys tem . . . . . . . . . . . . . . . . . . . . . . . 4
- 3. Functi onal Desi gn of the CPCS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 4. CPCS Algori thms and Uncerthi nti es . . . . . . . . . . . . . . . . . . . . . . . . . . . 4
- 5. CPCS Hardware and Software Design ........................... 5
- 6. Ge ne ral Que s ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 !
C. Status of NRC Staff Review ...................................... 5 '
- l. Outs tandi ng i tems Rel ati ng to the CPCS . . . . . . . . . . . . . . . . . . . . . . 5
- 2. Review of CPCS .............................................. 6
- 3. Safety Si gni fi cance of CPCS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
- 4. Data Li nks to Pl a nt Computer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
,w 110 DON ---.- .~~_. ... .
__.._--.A d
. . . o
,j s 1-us J
216TH ACRS 11EETING i TABLE OF CONTENTS '
\
i 5. Sta tus of Proj ect Revi ew . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 ;
- - + . . ,
_ 6. Applicant's Response ....................................... 9 9 D. Caucus ................................'......................... i III. Meeti ng on Liqui d Pathway Generic Study . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 i A. S ub commi ttee Repo rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - 9 I B. NRC Sta ff Pres e ntati o ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
- 1. I n t ro d u c ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 Offshore Power Systens Company Concl usi ons . . . . . . . . . . . . . . . . . . . . . 11 C.
11 D. Tech ni cal P res enta ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 1. Do s e Compa ri s o ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .' . . . . . . . . . 11
- 2. Core Melt-Through Penetration Mode and Steam Explosions .... 11
- 3. Applicant's Response to HRC Staff's Presentation o f S te am Expl o s i o ns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
- 4. Magnitude of Pressure Transient Following a Steam Ex p l o s i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
- 5. Coupli ng of River, Estuarine , and Ocean Doses . . . . . . . . . . . . . . 13
- 6. liRC Staff's Response to Applicant's Presentations . . . . . . . . /. 13 l E. Caucus ......................................................... 14 3 llecti ng on McGuire Nuclear Station, Uni ts 1 and 2 . . . . . . . . . . . . . . . . . . 14 IV.
S ub commi t tee Re po r t . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 A. Sta tus of NRC Staf f Revi ew . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 B. ,
)
I n tro d u c ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 C. Eme rg e n cy P l a n ni n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 j D. ECCS Design .................................................... 19 E.
-ii I
I -
-,,__,._.,,.,._m.,- ~ . _ _ . , . . , , m. . . . .,
. n D
J 21bTH ACRS MEETING TABLE OF CONTENTS
- 1. UHI Analys i s vs . Me a s u rement . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
)
- 2. ' 53 e'ci fi c P l a nt An al ys i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 20
- 3. UHI Analysis ...............................................
20 F. S tu d B o l ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . G. E' 'ct on Fuel of Core Radial Differential Pressure 23 fru Asymne tri c LO C A Lo a ds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 H. Ge ne ra l Que s ti ons . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 I. Caucus ....... ................................................. V. Meeting with the NRC Staff on Recent Operating Experience, Licensing Actions, Generic Matters Relating to LWRs and 24 Future Agenda ...................................................... A. Davis-Besse-1: Implementation of ACRS Recommendations from the ACRS Report of January 14, 1977 ....................... 24 27 B. Oconee: Mi c r os e i smi c i ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . C. Combustion Engineering Plants: Control Element Guide Tube Wear . 27 Failure of Burnable Poison Rod Assembly ...... 28 D. Crystal River 3: E. Implementation of Regulatory Guide 1.97, " Instrumentation 29 to Foll ow the Cours e of an Acci dent" . . . . . . . . . . . . . . . . . . . . . . . . . . .
- Moni tori ng on the Eas tern Seaboard . . . . . . . . . . . . . . . . . . . . . 30 F. Sr 31 G. Moh .oring Neutron Exposure at Nuclear Facilities ..............
31 H. Future Agenda .................................................. 32 VI. Executive Sessi ons (0 pen to Public) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 A. Regul atory Acti vi ties Subcommi ttee Report . . . . . . . . . . . . . . . . . . . . . . 32
- 1. Revi s i on of 10 CFR 50.44 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
32
- 2. Re g ul a to ry Gu i d e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 3. Regulatory Activities Subconmittee Agenda for its 33 May Me e ti n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
iii
, , . )
Q 1 216TH ACRS fiEETfNG ' TABLE OF CONTENTS B. ACRS Quarterly Report to Commi ssione rs . > . . . . . . . . . . . . . . . . . . . . . . . . 33 C . Testimony. tp Senate Subcommittee on Nuclear Regulation . .... . .. . . 33 D. Acti vi ti es o f th e Memb e rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 34
- 1. Mr. Shewmon .................................................
34
- 2. M r . Mo el l e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
34 E. Proposed Independent Nuclear Accident Review Board .............. 34 l F. Propos ed Meeting wi th Groupe Permanent . . . . . . . . . . . . . . . . . . . . . . . . . . 34 ; G. Reorgani zati on of ACRS Subcommi ttees . . . . . . . . . . . . . . . . . . . . . . . . . . . . 34 H. ACRS Reports and Letters ........................................
- 1. Letter to Dr. E. J. Sternglass .............................. 34 35 VII. Executi ve Sessions (Closed to Publi c) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
35 l A. New Members ..................................................... 35 B. ACRS Reports a nd Le tte rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Arkansas Nuclear 1, Uni t 2 Nuclear Power Pl ant . . . . . . . . . . . . . . 35 1. McGui re Nuclear Stati on, Uni ts 1 and 2 . . . . . . . . . . . . . . . . . . . . . . 35 2. 35
- 3. Li qui d Pa thway Ge n e ri c S tudy . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
F iv 1
,n,
216TH ACP.S MEETING l
' TABLE OF CONTENTS 1 Appe ndi x I - Atte nde es . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-1 Appe ndi x II 'e *ACRS Eu ture Age nd a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-8 i I
' Appendix III - Ltr, Rep. T. Bevill to Chmn. Hendrie on ACRS Fellowship Program ....................................A-10 1 Appendix IV - ANO-2: Proj ect Status Report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-11 l Appendix V - AND-2: Core Protection Calculator System (CPCS) . . . . . . . . . . . . A-21 Appendix VI - ANO-2: ACRS Cons ul tants ' Report . . . . . . . . . . . . . . . . . . . . . . . . . . . A-23 i
Appendix VII - ANO-2: Core Protection Cal culator Sys tem . . . . . . . . . . . . . . . . . A-39 i Appendix VIII - ANO-2: Core Protection Calculator System Fun c ti o nal De s i g n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A- 55 Appendix IX - ANO-2: CPCS Al gori thms and Uncertainties . . . . . . . . . . . . . . . . . . A-68 Appendix X - ANO-2: CPCS Hardware and Sof tware Design . . . . . . . . . . . . . . . . . . . A-75 Appendix XI - ANO-2: NRC Staff Review of CPCS ........................... A-85 Appendix XII - ANO-2: Data Links to the Plant Computer . . . . . . . . . . . . . . . . . . A-104 Appendix XIII - ANO-2: Status of Project Review . . . . . . . . . . . . . . . . . . . . . . . . . A-108 Appendix XIV - Liquid Pathways Generic Studies: Project Status Report ... A-ll7 Appendi x XV - Cons ul tants ' Reports . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-120 Appendix XVI - NUREG-0440: Major Conclusions ............................ A-130 Appendix XVII - h0 REG-0440: OPS Conclusions ............................. A-132 Appendix XVIII - NUREG-0440: Dose Comparisons for Land-Based and Fl oa t i ng Pl an ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-14 3 Appendix XIX - NUP,EG-0440: Core Melt-Through Penetration Mode and Steam Explosions ..................................... A.154 Appendix XX "UREG 0440: Analysis of Pressure Transients Following ( a S te am Expl o s i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A- 16 2 Appendix XXI - NUREG-0440: Cu oling of River, Estuarine, and Ocean Doses ... .......................................... A-168 l V I~ 4
.g 1
216TH ACRS MEETING IABLE OF CONTENTS Appendix XXII - McGuire 1 & 2: -Project Status Report .................. A-173 Appendix XXif f '- McGuire 1 & 2: Site, Organization, and Layout ........ A-208 Appendix XXIV - McGuire: UHI Analyses Compared with Measurement ....... A-213
-Appendix.XXV - McGuire 1 & 2: UH I An a l y s i s . . . . . . . . . . . . . . . . . . . . . . . . . . . . A- 2 21 Appendix XXVI - Davis Besse 1: Status Report .......................... A-228
- Appendix XXVII - Oconee: Mi cro s ei s mi c Acti vi ty . . . . . . . . . . . . . . . . . . . . . . . . A-231 Appendix XXVIII - CE Reactors: Control Element Guide Tube Wear ........ A-238 Appendix XXIX - Crystal River 3: Burnable Poison Rod Assembly Failure . A-262 Appendix XXX - Microseismic Networks i n Eastern U.S. . . . . . . . . . . . . . . . . . . . A-273 Appendix XXXI - Neutron Exposure: Request for Office of Nuclear Regul atory Research Program . . . . . . . . . . . . . . . . . . . . A-276 Appendix XXXII - Neutron Exposure: Dosimetry Methodol ogy . . . . . . . . . . . . . . A-289 Appendix XXXIII - Request by E. J. Sternglass for ACRS Review of Changes ir. Cancer Mortality in the Vicinity of Several Nucl ear Pl ants . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-298 Ippendi x XXXIV - Ltr to E. J. Sternglass . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . A-411 Appendix XXXV - Regulatory Guides , ACRS Actions . . . . . . . . . . . . . . . . . . . . . . . . A-412 Appendix XXXVI - Report on ACRS Activities, Dec.1977 - April 1978 ..... A-413 Appendix XXXVII - Report on Arkansas Nuclear One, Unit 2 . . . . . . . . . . . . . . . A-417 Appendi x XXXVIII - Report on McGui re, Units 1 and 2 . . . . . . . . . . . . . . . . . . . . A-420 Appendix XXXIX - Additional Documents Provided for ACRS' Use . .... . .. .. . A-423 4
vi
8 I 11877 ' N0flCES The Commlnsjon has determined yr G;* r.n O Drwht, Alt,mie F.afety an<t 1.1 Commiwinn's Public Document Room. that the 1.uunnee of this amendment , 1717 !! Etreet NW., Wn' hmeton. D.C.,
... .:ns: Ikrard Panet. UK the l w HwHa" and at the Rochester Pubhc Library, will not result in any signifscant envi. l tory Commeon, wuhmgton D C. M55 ronmental impact and that pursuant @ J:, mew C. IAmts !!! 313 Woodr.aven 115 South Ave nue, Rochester. N.Y.
to !O CFR 51.MdH4) an environmental P. cad. Chawl Ell N.C. 2%I4. 14G27. A copy of items (2) and (3) may impact statement or negative declara-be obtained upon request addressed to Lion and environmental impact a p- ! day DsLtd at Bethesda, of March 1978. Md., this 16th the U.S. Nuclear Hr gulatory Commis- praisal need not be prepared in sion. Washington. D C. 20555. Atten. nection with issuance of this amend- j For the Atomic Saf ety and Licensing tion: Director. Division of Operating E ,ard Panel.
~
t mn Reactors. Jiur.s R. YonE, l Chairman. Dated at Bethesda, Md., this 1st day this action, see (1) the spplicatic.n for or March 1978. asnendment dated November 28,1977 j fnt Doc.18-1520 nled 3-21-18; 8 45 am) (2) Amendment No.16 to License No. i For the Nuclear Regulatory Com. DPR-18, and (3) the Commission's re- ' rnission. lated Safety Evaluation. All of these [7590-01] DENNIS h ZIEMANN. items are available for public inspec-gDocket No. 50-244,' Chic /, Operaffnp Reactors tion at the Commission's Public Docu. I Branch No. 2, Dit'irion of Op' ment Room, 1717 H Street. NW., I ROCHt1TER C A1 & tt!CTRIC CORP. craung Reactors. Washington, D.C., and at the Roches-busnee of Amendment to Prov. ,isional IFR Doc.18-iS25 Piled 3-21-78; 8.45 am) ter Public Library,115 South Avenue,
- Rochester, N.Y.14627.
OPeteling bsent. A copy of items (2) and (3) may be The U.S. Nuclear Regulatory Com. (7590-01) obtained upon request addressed to z.!!slon (the Commission) has issued the U.S. Nuclear Regulatory Commis-Amendment No.15 to Provisional Op- (Docket No. 50-244) sion, Washington, D.C. 20555. Atten-trating License No. DPR-18 issued to tion: Director. Division of Operating ROCHESTIR GA5 & itICTRIC CORP. Rochester Gas & Electric Corp. which # revised Technical Specifications for Istvonce of Amendment to Provisional Dated at Bethesda, Md., this 8th day ' eperation of the R. E. Ginna Plant lo- op.,,,;ng tic.n.. of March 1978. ts:cd in Wayne County, N.Y. The The U.S. Nuclear Regulatory Com- For the Nuclear Regtdatory Com-ef:r.endment issuance. is effective as of the date mission (the Commission) has issued mission. Amendment No.16 to Provisional Op. DENNIS b ZIEMANN, The amendment incorporates fire Operating Reactors protection Technical Specifications on erating License No. DPR-18. Issued to Chic /, the existing fire protection equipment Rochester Gas & Electric Corp. (the Branch No. 2, Division of Op-and adds administrative controis relat. licensee), which revised Technical eraHng Redors. ed to fire protection at the f acility. Specifications for operation of the R. IF7t Doc. 78-1527 nled 3-21-78; 8:45 am) This action is being taken pending E. Ginna Plant (facility) located in completion of the Commission's over- Wayne County, N.Y. The amendment c11 fire protection review of the facill- is afThe fective as of its date of issuance. [7590-01] amendment changed the Tech-ty. ADV150RY COMWTTEI ON REACTOR The application for the amendment nical 1. Speelfications to' Delete the requirement for an sugouAgos compiles with the standards and re- Annual Operating Report, while re-quirements of the Atomic Energy Act taining the specifte requirement for an h *Has of 1954, as amended (the Act), and the Annual Report of Occupational Expo- In accordance with the purposes of Commission's rules and reculations. sections 29 and 182b of the Atornic The Commission has mnde appropri- sure, 2. Modify the submittal date for the Energy Act (42 U.S.C. 2039,2232b the ste findmgs as required by the Act and Monthly Operating Report to the 15th Advisory Committee on Reactor Safe-the Commission's rules and ret ula. instead of the 10th of the rnonth fol- guards will hold a meeting on Apr:1*6-tions set forthin 10 in CFR I, which are lowing the calendar month covered by 7,1978, in Room 1046, 1717 H Street Chapteramendment, the beense NW., Washington. D.C. Prior public notice of this amendment the3.report.- Delete the Respiratory Protection The agenda for the subject meeting mas not required since the amendment Program based on your comphance will be as follows: does not involve a significant hazards with 10 CFR 20.103 smee this item is Tuvasory Arnu.6,1978 consideration. The Commission has dciermined now included in 10 CPR Part 20 of the e:so a.w.-e:s s 4.w. r.uctmvs ssssion (orex) that the issuance of this amendment Commission's regulations, and
- 4. Add a shock suppressor (snubber) The Committee will hear and discuss the util not result tonmental in any hupact andsignificant envi- to the safety.rclated listing of suppres.
that pursuant 'h
' N i[n(ous m t te atfn to m tS ac-to 10 CPR 51.5t dx4) an environmental sors in Specification 3.13-1.
The applention for the amendment tivttles. The Committee mill hear and ch3 impact 6tatement, or nerathe declara- cuss the report of the ACTIS Subcommittee tion and environmental impact als complies with the standards and re* pralsal need not be prepared in con. quirements of the Atomic Energy Act and consultants who may be present rc nection with issuance of this amend.Commission's of 1954, as amended (the Act), and the ing the request for an operattne Itce rules and regulations. the Arkansas Nucicar One, unit 2 poser. rnent, plant. Portions of this session will be closed For further detalls with respect to The Commission has made appropri. if necessary to discuss proprietary informa-this action. see (1) the application for ate findings as required by the Act r.nd ti n appheable to this project. , amendment dated July 19, 1977, as the Commhaton's rules and regula. oas ax.-as as A.w. Awo :s s r.u.-s:s s P.u. I supplemented December 13, 1977, (2) tions in 10 CFR Chapter 1, which are ARKANSAS NUcLt.AR ONe, UMT s (oPEN) Amendment No. 15 to License No. set forth in the lleense amendment. The Committee will hear and discuss pre-DPR-GI, and (3) the Comminion's re- Prior public notlee of this amendment was not required since the amendment sentations by representatives of the Nnc lated Safety 1: valuation dated Novem- does not involve a significant hazards. staff and the applicant related to the rd ber 25,1977. All of these items are quest for operation of Arkansas Nudear Cvalthble for public inspection at the consideration. HDIRAL ttGISTER, VOL. 41, NO. SbWIDNt$D AY, MARCH 22, MS l 7- - - ,-
.31878
_N oriegs One, tmit'2. Dortions of this nennlon will be and ihr sositlennt rrgarelme s tic trauest fur Further informntion . regard;r g
- closed il necewary to divuss proprietory in. sn operatens bre.ns e i..e inc u, burr imric. toples to be discussed. whelhcr the firmation appbcable Lu thui rnatter, ar station. isud I aent a r.g ts. en .il slun mrrting has born cancelled or rosche 3:ss r.w.-s:4s P x. rxtevTrvs anston torent """'"I"'d"'""*"F'"""" dated, the Chairmnn's ruling on rw proprwtary Information applu nhlc to this Ths Comriuttee w!!! discuss the report of matter, Qur.sts for the opportunity toallott prP*ht its Subcommittee and consultants w ho may oral statements and the time be present regardinn the laymd pathway g'e. 4 P M *8;88 P.C suretrTeva saaston (OITM/ therefor Can be obtained by a prep 1;g nette study for floats ur and land based nu. et.naron telephone call to the ACRS Executae
. . einr pourrplants (NUltEG-0440). Portions The Committee mill dAcuss 6ta pn. posed Director. Mr. Raymond P. Praley, tek.
af this sezion will be closed if rectured to reports to NRC ri nrdmg Arkanw Nuclear phone 200-634-1371 between 8:15 a.rs protect proprietary infortnation regarding One. Unit 2 n:sd the M.timre smrarar sta- andfp h c.s.t. thls rnatter ' tion, unita 1 and 2. Thn aceton mill be Dated: March 20,1978. s:ss r.u.-s ex tJQUtD PATHWAY CENEstC catory pruccedino.
,7,,,gg,g,3 The Committee util di.scuss reporta of its JOHN C. Ifort.r
.Ad isory Commifice
'Ihe Committee sill hear and discuss re. rnemtsers regarding miwruancous ACllS ne.
- Afanagement O//icer*
ports of representatsbes of the NRC staff taittes such as propowd trurganizattun of and offshore power syhtems regard.nc the ACRS Subcommitts ra anal workmg groups [MI Doc. 78-7712 P11cd 2-21-78: 8.45 aml
!! quid pathway genene study (N UR EG. and the quahhcations of ranslutatim pro-0440) concerr.ing impact.s of accidental ra. posed for appointment to the Omnnuttre.
dioactive releases to the hydrosphere frorn Portions of this nession sill t>c closco in dis-Donting and land-based nuclear power. cuss maternal s hich if renciued sould repre. [7715-01] plants. Portions of this se.uion atil be closed sent an unaarranted amasun of pcrnonal if required to discuss proprietary informa- privacy. * --{QSTAL M. RATE COMMISSION ) Larn regardmg this matter. Procedures for the condtiet o'f and s r.w.-s:so ex exrecT vt sessron (orcx) participation in ACRS meetings were POSTAL ltATE AND f!I INCREASES,1977 The Committee will hear and discuss re. outlined in the hDMA1. RFC!sTut on g,,9 g,cu,,,, ports of Subcommittees, working groups. October 31,1977, page 50972. In accor. and members un a number of generic mat. dance with these procedures, oral or MARett 15.1978. ters related to reactor safety including written statements may be presented Oral argument in this case previous-review of NRC sating policies and practices, by members of the public, recordirms
. proposed changes in NRC rerulatory guides, will be permitted only durmg those ly scheduled for March 24,1978, his and NRC procedures for review of proposed ' portions of the meeting wheti a tran. been re scheduled for Tuesday, Mar ~*
cperatons at increased power levets. This script is being kept, and Questions may 28,1978, at 0:30 a.m. portion of the meeting mill t e wen to the be uked only by members of the Corn. A further notice will be issued ade.s. public. The Committee wW ul discuss its mittee. Its consultants, and staf f. Per. in; counsel of the time allotted to proposed report to the NRC regardmst oper' sons destring to make oral statements each for oral argument. Etion of Arkansas Nuclear One, unit 2. This should notify the ACRS Executive Di* sessicn sitt be closed to ctiscuss matters in. volved in an adjudicatory proceedmg. rector as far in advance as practicable DAVID F'HAan9""gts* , 50 that appropriate arrangements can Farnsv. Aran.7,1978 be made to allow the necessary time (Mt Doc. 78-7510 Pued 3-21-78; 8:45 sud during the meeting for such state-c:so Ax-si Ax urztzNo WITH NRC STAFF ment 3, torrx) I have determined in .ccordance [8025-01] ! The Committu will hear presentations with section 10td) of Pub. L. 92-4G3 from and hold discussions with rnembers of that it is necessary to close portions of SMALL BUSINESS ADMINISTRATION the Nuclear Herulatory Commission staff the meeting as noted above to' protect ILicense No. 06/06-01751 regardwg recent I censmg actions and oper. proprietary information (5 U.S.C. cling experience includ.nz reports of in* 550b(c)(4 )), to permit discussion of SMAtt BUSINESS INVESTMINT CAPITAL. INC. I f e Oc ce n car i nt,lalur of " pr r;i;,, ,, appia,,;,, ,,, A pp,,,,; ,, , c.,,y ,; burnable potson. rod assembly in the Crystal fe^e 5 SC 5b 1) ad * " " " protect infortnation the release of cf C recor en at or s t a. din g h which would represent an unwarrant. Notice is hearby given, pursuant to Davts-Desse nuclear power station, unit 1. ed invasion of personal privacy (5 1107.1004 of the regulations govermng The NHC staff wtti also report to the ACitS U.S.C. 55 b(c)(61). Eeparation of factu. small businesti investment companas en generic matters related to nuelcar pomer. al information from information con- (13 CFR 107.1004 (1977)), by the Sm:C plant safety taciudirig development of crite- Sidered exempt from disclosure during Business Administration (SI)A) of & rts for instrument.ation to todow the course closed portions of the meeting is not conflict of interest transaction b e-ci a serious accioent, and criteria for combi- considered practical. tween the Small Business Investment nation of dyTismic loads in the design of nu. Background information concerning Capite.l. Inc. (Licensee). 10003 New clear powerplants. Th2 future schedule for iterns to be consiuered durme thi.s Benton Highway. Little Rock. Ark. s1d r t y Llie C Lee a be d[ meeting can be found in documtnts un 72203, a Federal Licensee under the cussM. file and available for pubbe mspect:on Smhl! Business Investment Act cf in the Nuclear Regulatory Comm:s. 19t8 as amenced (the Act) (15 U.S.C. si ax-s s:so a.C ExEctrrivr SLsstoN (orrN) Slon's Public Document Ituom, l The Cornmittee will hear and discuAs the Etreet NW., Washington D C. _717 .25 Il 651 et scQ.l. and an Associate. Tlie Licens(e was licensed by SBA report of its buccommittee and consultants and in the followit.g pubhc oc umt nt on March G.1975. It is wholly owned who may be present renrding the request rooms: by Sriur.Valu Stamps, Inc.,10003 New for operation of the McGute nuclear sta. tion, unita 1 ard 2. Portisas of th.s sc.n:on yecoggg ,peggg,3747:op vNgTseaNo U"uton list:hway. Little Rock. Art I sitt be closed af necest,ary to discuss propre. Pubhe LLhrary of Charl.;,tte and M , 1 n. yg;t./. %h.cr.'irl turn is owned approw 1 stary information apphenble to this project. burr Coimty. 310 North Tryon r t, Charlotte. N.C. 28:02. torch. Inc., a cooperative of retall gr> s::33 4x-t s:so P,M. AND :30 F ht.-4 r.C cm, and 55 percent by present an:1 CCC GUlaE STATION UNtra ) AND (orEN) ARKANSAS NUCLEAR oNE. UNIT 2 former rDPtreberS of the cooperative. The Cornmittee will hear and dhcuss re. Arkansas polytechnic collerc. ruurcille, it tx prol.'ed that the L:ct rive lo'Ln ports of reprnentatsves of the NI(C staff ArL 72801. $150,Ouu to hit. Jctry KcIly to build & l FtD!t At PEGISTit. VOL. 43. NO. 56-WION150AY, M ARCH 22, 1973
- * " ' * * ' N news mr- e..<-=-am m as eme. e ---.*w, e.% messes ===** * ,. - -
- ,. .,~ ,
Issue Date: MINETTES OF THE p p p y[ 216TH ACRS MEETING ji 3 y y d ' j APRIL 6-7, 1978 wj dl uh WASHI!Uf0N, DC The 216th Meeting of the Advisory Committee on Reactor Safeguards, held at 1717 H St. N.W., Washington, DC, was convened at 8:30 a.m., Thursday, April 6, 1978. The Chairman noted the existence of the published agenda for this meeting, and listed the items to be discussed. He noted that the meeting was being held in conformance with the Federal Advisory Committee Act (FACA) and the Government in the Sunshine Act (GISA) , Public Laws 92-463 and 94-409, respectively. He noted that no requests had been received from members of the public to present oral statements. He also noted that copies of the transcript of some of the public portions of the meeting would be available in the NRC's Public Document Room at 1717 H St. N.W., Washington, DC, within approximately 24 hours. [ Note: D . Isbin was not present Thursday.] I. Chairman's Report (Open to Public) [ Note: Raymond F. Fraley was the Designated Federal Employee for this portion of the meeting.] A. Reviewers The Chairman named Messrs. Siess and Shewnon as reviewers for the 216th ACRS meeting. B. New Nuclear Licensing Bills The Chairman noted that several proposals are before the Congress for changes to the nuclear licensing procedures. He noted two specific items included in the Administration's version of the proposed bill that are of direct interest to the committee: e The proposal for non-mandatory review of all license applica-tions is more limiting than that proposed by the Committee. i In the current form, the only license applications exempt from mandatory review are those for follow-on standard plants.
- The name of ACRS is proposed to be changed to the Advisory Committee on Reactor Safety.
I i 1 7
. . a MINUI'ES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 C. ACRS Fellows Program
... The Chairman noted that appropriations for the ACRS Fellows
~
Progr'am are not included in the current appropriations bill l for fiscal year 1978. Members suggested that a letter should be l prepared octlining the Committee's need for such a program and l sent to the appropriate Senate and House Oversight Committees, accompanied by a request that these letters be forwarded to the pertinent committees or subcommittees within each House (see Appendix III). l D. Testimony of Committee Before Senate Subcommittee on Nuclear Regulation I l The Chairman noted that the Committee has been invited to testify before the Senate Subcommittee on Nuclear Regulation, the Committee on Environment and Public Works, Senator Hart, Chairman. The Chairman noted that a draft of proposed testimony has been distributed to Members, and requested that they provide comments and suggestions regarding this testimony as soon as possible. He also noted that he would be accompanied to the t hearing by the Executive Secretary, the Vice Chairman, and Messrs. Bender and Siess. He welcomed the presence of any other I Members who could participate. E. Vermont Yankee Vs. National Resources Defense Council The Chairman noted that the U.S. Supreme Court has handed down a decision on the litigation between the Vermont Yankee Nuclear Plant and the National Resources Defense Council. Included in this decision also was the litigation between the NRC regarding the Midland Plant and Aschliman. The decision, copies of the summary of which have been provided, was generally favorable to the position taken by the NRC. II. Meeting on Arkansas Nuclear One, Unit 2, Nuclear Power Plant (OL) (Open to Public) [ Note: Gary R. Quittschreiber was the Designated Federal Employee for this portion of the meeting.) A. Subcommittee Report Mr. Carbon, Subcommittee Chairman, discussed the review of the Arkansas Nuclear One, Unit 2 design. He noted that this plant is a CESSAR-80 plant, but is using the newly designed core 2 ._ -. __ _ --y-- .
< r- .
MINUTES OF THE 216TH ACRS MECTING APRIL 6-7, 1978 protection calculator system (CPCS). This system has been de-signed since the construction permit review of this plant, and therefore was not discussed at that time. He discussed the gen-eYal design of the plant; the fuel rod configuration, noting that this plant will _be the first to use the combustion Engineering 16x16 fuel rod assembly; the functions of the CPCS, the core oper-ating limits supervisory system (COISS), which will be used for
-the first time in this plant; and the outstanding issues. He said that the subcommittee believes that the two major areas of ,
, interest are the CPCS, and the large number of outstanding items l remaining to be resolved at this stage of the review. (For details, see Appendix IV; for a description of the core protection calculator system, see Atp2ndix V; for consultants' reports, see Appendix VI.)
Mr. Carbon noted that an NRC Staff reviewer, J. Calvo, iden-tified 39 techthal issues which needed to be resolved as of last Spring, but the NRC Staff claims that all of these items have since been resolved. Mr. Kerr offered his opinion that he believes that the CPCS has been very thoroughly reviewed by the NRC Staff, its consult-ants, and the ACRS' consultants. He noted that the Applicant
- desires to keep the CPCS connected to the plant computer, which is not part of the plant protection system, and therefore is not safety grade. In this way the Applicant hopes both to mnitor the operation of the CPCS, and to use the information gained for other calculaticns, making wre efficient use of information generated by the CPCS. The NRC Staff has concluded that such a connection can be allowed during the startup phase of the reactor, but should not be permitted during normal long-term operation. Mr. Kerr sug-gested that the Committee may wish to explore this matter.
W. Lipinski, ACRS consultant, suggested that the connection of the CPCS to the plant computer may improve the reliabilit,y of the CPCS, and also provide a means for recordirrj the data genera-ted by the CPCS. If the connection is not permitted, there will be no recording of this data. E. P. Epler, ACRS consultant suggested that the Committee should proceed with caution in permitting a non-safety item to be tied to a safety system. He noted however, that there are advan- ' tages to this connection as well as possible pitfalls. ;q E [ Note: D. Rueter , Arkansas Power and Light Company (APLC) , $ coordinated presentations for the Applicant; D. Martin, for the y NRC Staff.) tj 3 4
MIlUI'ES OF THE 216TH ACPS MEETI1G . APRIL 6-7, 1978 B. Applicant's Presentations
- 1. Introduction
~
~ ' N '. Moore, APLC, discussed the status of construction and testing of Arkansas Nuclear One, Unit 2 (ANO-2) . He noted that the plant is 93.51 complete, based upon labor require-ments, and 99.25% complete, based upon equipment installation.
261 of 287 systems have been accepted from construction for pre-operational testing and startup activity. He noted that pre-core hot functional testing was completed on February 18, 1978. The status of the pre-operational testing program is as follows: o 135 of 185 tests are complete, o with regard to tests required for fuel loading, 102 of 136 tests are complete. The present schedule for fuel loading is May 15, 1978. It is anticipated that criticality will be achieved during the week of July 3,1978, and it is hoped that commercial operation will begin in late October 1978.
- 2. Core Protection and Calculator System A. Spinell, Combustion Engineering (CE) , discussed the design and review of the CPCS and discussed the COIES and how this system aids the plant operator in maintaining some of the limiting conditions for operation (see Appendix VII) .
- 3. Functional Design of the CPCS W. Gill, CE, discussed the functional design of the CPCS, including the relationships of the calculator to the remainder of the reactor protective systems, the design bases events, system inputs and outputs, system functions, algorithms, power distribution methods, mthods to control departure from nucle-ate boiling, and treatment of uncertainties (see Appendix VIII) .
In answer to a question, he noted that the plant will trip for an accident involving a sheared shaft on a main coolant pump, although this is not a design basis accident for the CPCS.
- 4. CPCS Algorithms and Uncertainties R. Humphries, CE, discused the algorithms used in the CPCS functions, and discussed the manner in which uncertainties are treated (see Appendix IX) .
4
MINUI'ES OF THE 216TH ACRS MEETING APRIL 6-7, 1978
- 5. CPCS Hardware and Software Design E. Brown, CE, discussed the CPCS hardware and software
. '" design (see Appendix X) . He described how the functional requirements of the CPCS were implemented in the hardware and the software systems of the core protection calculator, the on-line testing features, and the verification program for the CPCS. He said that he believes that the test procedures pro-vide for a wider range of response than hopefully would be obtained from an on-line plant.
In answer to a question, E. Brown said that the CPCS sys-tem automatically fails safe on loss of power. W. Lipinski, ACRS consultant, questioned whether the test procedures actually provide for dynamic response of the system under truly transient conditions.
- 6. General Questions In answer to a question, N. Moore, APIf, said that the Applicant believes that any modifications made in the control rod element system as a result of resolutions of the current CE control rod problems, will be capable of being backfitted into ANO-2. For example, additional guide tubes can be inser- l ted in the top of the assemblies to protect the current guide tubes from the full vibrations. If resolution is obtained in the very near future, it is possible that this nodification can be made before the reactor goes into operation.
Mr. Shewmon requested that the Applicant provide him with l the specification for hold-down bolts for Class-1 equipment. l C. Status of NRC Staff Review '
- 1. Outstanding Items Relating to the CPCS R. Martin, NRC Staff, said that with review of the CPCS, out of the 27 issues which have been raised by tb NRC Staff, only 3 continue to be outstanding and must be resolved prior to the issuance of the operating license:
o Position No. 26: Optical Isolators. Resolution will be required prior to the issuance of the OL. Resolution is anticipated by June 1. 5' 4 p.-- -~=,.q. 4-m.+= __..g. w,.
.MINIJfES OF THE.216TH ACRS MEETING APRIL 6-7, 1978 e Position No. 14: Seismic Qualification Review. Resolu-tion is anticipated to have progressed
. . . . , ' to a point by June 1 that remaining portions of the issue can be addressed with conditions to the OL, without undue impact on the issuance date of the license.
e Position No. 19: Software Change. Resolution should have progressed to a point by June 1 that remaining portions of the issues can be addressed with conditions to the operat-ing license, without undue impact on the issuance date of the license. l
- 2. Review of CPCS L. Beltracchi, NRC Staff, discussed the Staff's review of the CPCS, including the current status, the safety overview, '
.the criteria for the review, the methodology used, and the test audits (see Appendix XI).
- 3. Safety Significance of CPCS W. Hodges, NRC Staff, said that the design basis for the CPCS' is the loss-of-flow event. The CPCS is also capable of protecting against a number of the anticipated operational.
occurrences, which include uncontrolled control element assem-bly withdrawal, control element assembly misoperation, boron dilution, excess heat renoval, and steam generator tube rupture. In addition, the CPCS also.provides some protection for a steam-line rupture event with loss of off-site power, and for a main coolant pump shaf t seizure when loss of flow can be detected. The NRC Staff has investigated the backup protection i that is available for each of these postulated accidents in the event of the failure of the CPCS concurrently with one of these ; accidents. For several of the postulated accidents, the CFCS l does not provide the_ first trip even when it is operating, J therefore there is no change in the consequences to the plant at such times, whether the CPCS is operating or not. The only
- accident for which there is not automatic backup for the CPCS is a control element assembly misoperation, which requires manual backup by the operator. In such an event, the operator is assisted by alarms from the COLSS, and also by indications on the control panels. If the CPCS does fail to operate in the control element assembly misoperation event, the plant is in the same mode of operation in which it would have been had there not been a CPCS installed. For this particular event, Combustion Engineering plants always required a manual. trip.
~
6
..._ _ _.. _ . m ., ._m.__ _. _
-.m .m
APRIL 6-7, 1978 MINETTES OF THE 216TH ACRS MEETItG W. Lipinski asked whether a program has been developed to
-examine the performance of the CPCS system after it is placed into operation to determine that it does indeed function in
---the manner in which it is intended to function.
W. Hodges said that during startup testing, a Icss of flow test will be run at no power. Power inputs will be simulated to the CPCs at several power levels durire these trips and the function of the CPCS will be monitored. There will also beHe a
-l' zero power trip based upon the input frm the real pumps.
said that the purpose of the vendor's Phase 2 testirsg of the system, af ter the software had gone through a rather extensive testing program to determine that there were no codire errors, was to check the system functionally to see whether the inte-grated system would perform as designed. During this Phase f 2 testing, the simulator that was used was noisy, and may have I been nonrepresentative of the type of noisy signals which might be received from an operating plant. Evaluation was obtained of the behavior of the dynamic components of the algorithms themselves used for the DNBR calculations. During these tests, although the scatter of the signals fejl outside that pre-dicted, the time to trip was an order of magnitude less I than the required time to trip as demonstrated by design analysis codes. Further, the simulator provided a capability , to dynamically vary all input simultaneously. l l W. Hodges, noted for an example, that the most limiting ' transient is the core pump loss of flow. This. case was ana- l lyzed on the simulator, all of the inputs were ramped in a pre-dicted manner based on design calculations, all of the inputs were varied simultaneously, the trip output was obtained, and the time to trip was compared with the required time of trip as predicted by the . design codes, and this was found to be con-servative. Fur the r , to ensure that there were no coding errors, the test cases were reperformed on the single channel test system at Windsor, where the dynamic inputs could be held l steady. The results of these tests were compared with the FOKfRAN coding, and the difference was very small between the DNBR outputs and the local power density outputs. Based on these tests, the NRC Staff was satisfied that the scatter arising from the Phase 2 testing was not due to coding errors, but rather to a noisy simulator. However, the NRC Staff is requiring that process noise be evaluated in the plant during startup testing to insure that excessive trips owing to the dynamic component of the thermal power algorithms themselves 7 cannot result. i 7 . t i m...,.,..~_,, , _ _ . _ _ _ . . . _ _ _ _ -
o MINUI'ES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 W. Gill, CE, discussed noise levels in operating CE plants. He noted that measurements have been made in a num- I ber o plants, including St. Lucie, and that a spectrum of l
'~~ frequencies and amplitudes have been identified which would l be expected to be found in ANO-2, assuming that ANO-2 will l behave as the other plants have. The single channel CPCS I system at Windsor, is being exposed to this spectrum of amplitudes and frequencies, and this program is approximately i 80% complete. !
1 I In a discussion regarding the linking of the CPCS to the reactor computer, F. R. Naventi stated that the NRC Staff made a decision (position 20) not to conduct a safety review of ! these combined systems in the case of ANO-2. Since this area l has not been reviewed as a safety system, and the possible l interactions and consequences not considered, the combined use ( of the two systems will not be authorized during routine , operation. As a continuation of this discussion, E. P. Epler ! I suggested that, before use of the CPCS and the reactor computer as a combined system should be allowed, the interactions between the two systems should be considered, so that failure of the safety system from such events as common node failure or interference by the nonsafety system can be precluded. In answer to a question regarding the issues raised by J. Calvo, NRC Staff, in his memo of June 24, 1977, L. Beltracchi said that many of these concerns were also NRC Staff concerns. Changes have been made in the ANO design to resolve these issues. R. Tedesco, NRC Staff, said that J. Calvo was not involved in the last part of the ANO-2 review, and that his concurrence on the resolutions of these issues has not been obtained.
- 4. Data Links to Plant Computer F. R. Naventi discussed the data links to the ANO-2 i computer (see Appendix XII) . l
- 5. Status of Project Review l l
D. Martin, NRC Staff, discussed the overall status of the project review, noting recent resolutions to formerly open items, the status of open items and the expected resolution, and new items that have been identified recently (see Appendix XIII). 8 1
- .~.mm._, , - _ . .,_ _
MINUTES OF THE 216Til ACRS MEETING APRIL 6-7, 1978
- 6. Applicant's Response D. Rueter said that the Applicant agrees with the inter-pretation of the NRC Staff regarding the status of the pro-j ject in the review. i In answer to a question regarding the number of exper-ienced personnel to be transferred from Unit 1 to Unit 2, l R. Terwilliger , APLC, said that five shift supervisors, five i I
plant operators, five assistant plant operators, and four waste control operators who will operate Unit 2 have also ' had experience on Unit 1. He noted that, in agreement with the NRC Operator Licensing Branch, because of the dissimi-larity between Units 1 and 2, there will be very few opera-tors that are licensed for both units. D. Caucus The Members provided their opinions concerning matters dis-cussed above and identified those matters that they believed should be addressed in a report concerning this review. l The Committee agreed unanimously that it would try to write i a favorable report on the Arkansas Nuclear One, Unit 2, Nuclear l l Ibwcr Plant. Mr. Kerr requested that the NRC Staff compare the ANO-2 docket with other dockets recently reviewed to determine trends in the number of open items in the reviews being considered by the Committee. L. Crocker, NRC Staff, agreed to provide such a ' summary. III. Meeting on Liquid Pathway Generic Study (Open to Public) [ Note: Gary R. Quittschreiber was the Designated Federal Employee for this portion of the meeting.] A. Subcommittee Report Mr. Moeller, Floating Nuclear Plant Subcommittee Chairman, discussed the history of the development of the Liquid Pathway Generic Study (NUREG-0440), noting that this study compares the radioactive material releases from Class-9 accidents from floating nuclear plants with similar accidents at land-based plants (see Appendix XIV) . He noted that reports were received from a number i I of ACRS consultants (see Appendix XV) . 9
, , - c -
MINETIES OF THE 215TH ACRS MELTING APRIL 6-7, 1978 Mr. Moeller stated that the Committee's mission in this meeting is to ascertain whether NUREG-0440 adequately addresses the , questions raised by the Committee at previous meetings. Of particul'ar interest is the methodology used in the evaluation. He suggested that subsequent meetings may be needed to review the acceptability of the risks involved or the need for design changes to make the risks acceptable.
]
R. Foster, ACRS consultant, pointed out that while the re- l sults of both atmospheric and liquid pathways have been examined, i there is a very important difference in the two pathways in that l the liquid pathway may provide a prompt release of sump water, as l contrasted with a second more delayed release from the molten core. It is possible that the prompt release of the sump water may in fact lead to consequences which are substantially greater 1 than those from the core melt. I. Catton, ACRS consultant, offered the following opinions: e The NRC Staff calculations of leach rate of debris as presen- l ted in NUREG-0440 are best estimate calculations, rather than : upper bounds. e Fragmentation of the rtolten core from a steam explosion has I not been adequately considered. [ Note: R. Vollmer coordinated presentation for the NRC Staff.] B. NRC Staff Presentations
- 1. Introduction R. Vollmer summarized the major conclusions drawn from the current revision of NUREG-0440, Liquid Pathway Generic Study
- (see Appendix XVI) . In his conclusion, he stated that the NRC Staff believes that, in the very unlikely event of a core melt, the floating plant presents a greater hazard to the health and safety of the public, in that the sump water, and a large amount of the contained radiation in the core, would be removed through the liquid pathway before effective interdiction could be instituted. Further, the NRC Staff believes that effective liquid pathway interdiction would result in a high social or economic impact. Actual cost figures are currently being developed.
10
l MINCTTES OF THE 216TH ACRS MEETItG APRIL 6-7, 1978 R. Vollmer said that the Liquid Pathway Generic Study represents a last major effort in the completion of a safety
. . evaluation report (SER) and an environmental statement for the
' floating' nuclear plant. It is important to the NRC Staff that the Committee provide c. report containing the Committee's view on its assessment on the adequacy of the study, as input to both the environmental impact statement and the SER.
C. Offshore Power Systems Company conclusions D. Walter, Offshore Power Systems Company (OPS), discussed the Liquid Pathway Generic Study, specifically both the NRC Staff and the OPS conclusions. Included ir: this discussion were the dose pathways considered for the study, the FNP liquid pathways dose consequence calculations made by OPS, NRC Staff conclusions as included in NUREG-0440, the scope of the NRC Staff Liquid Path-way Generic Study, the results of recent Sandia Corporation leach tests, calculated dose consequences from Class-9 accidents, a com-parison between liquid and air pathways, air pathways dose conse-quences, the calculated accident risk for air pathways, and a comparison of the risks from air pathways and liquid pathways (see Appendix XVII) . D. Technical Presentations j
- 1. Dose Comparisons G. Chipman, NRC Staff, compared calculated doses released from Class-9 accidents postulated in both floating and land- j based plants (see Appendix XVIII). i In answer to a question, G. Chipman noted that, if the airborne material is released through the bottom of the contain-ment,'such release caused by a melt-through, the airborne release for a floating plant would be approximately the same as for a land-based plant.
Mr. Bender suggested that it would be uss ful to examine a 4 variety of current land-based sites to consider what the economic and environmental ~ consequences of a Class-9 accident might realistically be.
- 2. Core Melt-Through Penetration Mode and Steam Explosions T. Spies, NRC Staff, discussed several postulated modes for penetration of containment by a core melt and also the consequences of postulated steam explosions (see Appendix XIX).
He addressed the concerns regarding steam explosions that were raised during the March 22 subcommittee meeting. 11
,_. g
4 MItRTTES OF 'n1E 216TH ACRS MEETING APRIL 6-7, 1978 In a discussion among Members, ACRS consultants, and T. Spies, it was established that very limiting conditions are
..... required in order for a steam explosion to take place. Criti-cal to this interaction are the temperature of both the cooling mass and the heating mass, and the further requirement that the heating mass must be a liquid. One of the prime requirements for a steam explosion is a very high rate of heat transfer.
- 3. Applicant's Response to NRC Staff's Presentation of Steam Explosions D. Walker, OPS, said that it is important to note chat the calculated releases from the leaching of the molten fuel and those from the sump produce comparable dose consequen:es. The radionuclides most effective in producing these dose conse-quences are cesium and strontium. Important to the overall consequences are the leach rate, and whether or not the sump material can be interdicted. If significant leaching takes a week, and the sump material finds itself in the sea water almost immediately, the contribution to dose from the leaching becomes insignificant. He noted the Applicant's disagreement with the NRC Staff conclusion that the realistic concern for dose to the public should not be, as the NRC Staff contends, Category-5 releases. He said that the Applicant believes that both the Category-5 and 7 releases are insignificant because of the' low probability of their occurrence.
- 4. Magnitude of Pressure Transient Following a Steam Explosion D. Walker , OPS , summarized the Applicant's analysis to determine the magnitude of pressure transients following a steam explosion, noting the calculated initial peak pressure, the physical configuration of the reactor pressure vessel and the barge, time factors involved, and heat transfer calcula- j tions (see Appendix XX) . He noted that both the Applicant and l the NRC Staff have concluded that steam explosions are unlikely 1 in the event of a core melt accident, and that, even if the '
l
-accident should occur , it is unlikely that the interactions would involve any large fraction of the core debris in a single event.
I. Catton, noting that the model used by the Applicant involved a pressure pulse from a noncondensible gas, questioned whether these calculations were applicable to a steam bubble. 12
~
hw epte a +4e e= = mwho
MINLrfES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 In a discussion regarding possible damage to the second unit at an FNP site from a steam explosion occurring beneath the first . unit, B. Haga said that the neighboring plant could
.. , pe shutdown safely even in the event that its barge was rup-
~
tured, sank, and was in a tilted position. D. Walker said that the study indicated that the time > the material would be held in the sumps was in the range of 10 to 35 days. He said that, following the core melt-through, the plant would still float, even with a hole, and that tidal flushing of the sumps would not occur.
- 5. Coupling of River, Estuarine, and Ocean Doses J. A. Nutant, OPS, discussed the estimated liquid pathway results (man-rem per core-melt), plant site rankings based on liquid pathway man-rem consequences without interdiction, airborne releace estimates for a steam explosion inside the hull, diameters of released particles, and dose consequences from particle transport (see Appendix XXI).
In answer to a question, D. Walker said that the airborne. releases are approximately the same for all plants of equiva-lent size, and that the dose consequences are determined by the population distribution in the vicinity of the plant, and vary from site to site. l D. F. Bunch, NRC Staff, noted that, in these presentations, the Staff made the same assumptions that were used in NASH-1400, 1 so that comparisons could be made in familiar terms. l
- 6. NRC Staff's Response to Applicant's Presentations l G. Chipman noted that the NRC Staff made its comparison on the basis of expected results, what could occur, and what would be expected to occur if the core-melt did occur. The NRC Staff analysis showed that risk from the air pathway for a floating plant is comparable tc that from a land based plant. He said that the NRC Staff disagrees with the Applicant's analysis with regard to sump water, and that more than 10% of the sump water would be released within one week.
P. Haga said that he believes that the Applicant has shown that, when realistic interdiction is considered, the consequen-ces of accidental releases of radioactivity to the liquid path-way are not very important when compared to the consequences of accompanying releases to the air pathway. The conclusion is that, for accidents beyond the design basis, floating nuclear plants have been shown to be comparable to land-based plants. 13 p____._.u._..- __ _ . _._ - - . _ . .g
MI!PJTES OF TILE 216Tl! ACRS MEETING APRIL 6-7, 1978 , Therefore, the applicant sees no need for any extra design features, and does not plan any. He offered the hope that
,...this meeting would be sufficient for the Committee to prepare a' final report on the OPS application.
R. F. Foster, ACRS consultant, summrized his conclusions as follows: e The air pathway is going to be at least as important as the liquid pathway, and perhaps as much as an order of magnitude more important. e The sump release is at least as important as the leach rate from the core, and perhaps five times more important. 1 l e Interdiction is practical in terms of the liquid pathway, l although probably not at the source. I e The problems are economic as well as risk to health and safety. I. Catton summarized his conclusions; e He questioned the size of the calculated steam explosion. e ile questioned whether the steam explosion was properly addressed with regard to the particular materials involved. l l e The overall consequences from a Class-9 accident may be only a factor of two different between floating and land-based plants. E. Caucus The Committee agreed unanimously that it would attempt to write a report on NUREG-0440, Liquid Pathway Generic Study, at this meeting. [ Note: Time did not permit the completion of this report at the 216th ACRS neeting. It is anticipated that this report will be completed at the 217th ACRS meeting, in May,1978.] IV. Meeting on McGuire Nuclear Station, Units 1 and 2 (OL) (Open to Public) [ Note: Richard P. Savio was the Designated Federal Employee for this portion of the meeting.) 14
., c - .
MINUTES OF TdE 216TH ACRS MEETING APRIL 6-7,1978 A. Subcommittee Report
- Mr . Plesset, discussed the history of the Subcommittee's review of the application for an operating license for the McGuire
. Nuclear Station, Units 1 and 2, discussed the overall plant param-
~
eters end site location, the outstanding issues, and briefly compared this station with other stations of similar design (see Appendix XXII) . He noted certain changes made in the ice con-denser portion of the plant as a result of the operating experi-ence obtained at the D. C. Cook Nuclear Plant. He noted that McGuire is the first plant to be reviewed that utilizes an Upper-Head Injection (UHI) system as part of the ECCS. [ Note: K. S. Canady coordinated presentations for the Applicant; R. Birkel, for the NRC Staff.) B. Status of NRC Staff Review R. Birkel discussed the chronology of the review of the application for an operating license for the McGuire Nuclear Sta-tion, and he noted that there were no differing technical views expressed by members of the NRC Staff relating to the review of the McGuire safety analysis report, as summarized in the Safety Evaluation Report (SER) , NUREG-0422. He noted that the environ-mental portion of the public hearing was completed on April 22, 1977, and that it is anticipated that the radiological safety portion of the AS&LB hearing will be resumed in late June,1978. This hearing is being contested. He noted that, with the excep-tion of a few outstanding issues, the NRC Staff has completed the review of the McGuire Station. He noted the status of the 21 issues listed as outstanding in the SER as of March 1,1978: 10 issues have been satisfactorily resolved, 3 are partially resolved and generic in nature, and the remaining items are expected to be resolved in the near future. Referring to section 1.6, begin-ning on page 108 of the SER, he listed the status of outstanding items as follows: l Items Requiring Information from the Applicant: Item 1: Duke Power Company has orally committed to providing the required information and justification no later than July 1, 1978. NRC Staff finds this acceptable. Item 2: The Applicant is committed to submit his stress anal-ysis report by October 1, 1978. NRC Staff finds this acceptable. Item 3: Resolved. Item 4: Resolved. 15
.c -
MINUTES OF THE 216'ill ACRS MEETING APRIL 6-7, 1978 Item 5: Resolved. Item 6: Resolved. htem 7: Generic, partially resolved.
~
Item 8: Resolved. Item 9: The Applicant has stated u.at a response will be I provided by May 8, 1978. The NRC Staff finds this satisfactory. 1 l Item 10: Generic, partially resolved. The Applicant has agreed l to submit additional information to identify the model ;
. number and the requalification document reference for !
information for class-1E equipment. The NRC Staff will review this information and compare it with the Westinghouse Topical Report. The Applicant has also agreed to submit additional documentation on balance-of-plant electrical connections and terminals inside l containment. The NRC Staff believes that the review of the requalification program and the correction of ! any equipment deficiencies can be completed prior to I the issuance of an OL, currently scheduled for Decem-ber, 1978. Item 11: Resolved. Item 12: Resolved. Item 13: Resolved. Item 14: The NRC Staff has not yet completed its evaluation of the steam line break accident which was recently filed as Revision 28 to the FSAR. Issues Requiring NRC Staff Evaluation Item 1: This issue is open. The NRC Staff review is in pro- - gress, but major problems are not expected. Item 2: Partially resolved. The NRC Staff has approved the l LAll ECCS Generic Evaluation Model. The Applicant will . sutxnit the Appendix K analysis using this model by May 8, 1978. Major problems are not expected. f 7 Item 3: This issue is resolved. 9 16
- ~ , . .
, o MINUTES OF THE 216TH ACRS MIETING APRIL 6-7, 1978 Item 4: This issue is resolved.
,_,, Item 5: The NRC Staff review of the McGuire fire protection design is in progress. A revised fire protection analysis was sutxnitted by the Applicant on March 22, 1978. A site fire protection review is scheduled for the week of April 17-21, 1978, with staff positions to
' be issued by May 15, 1978. The Applicant's response is due on June 29, 1978. The review is projected to be completed by mid-August.
Item 6: A request for additional industrial security informa-tion was issued April 7,1978, with response from the Applicant due in mid-May. The review is scheduled to be completed by mid-July. j Item 7: UpSated financial information from the Applicant was l filed April 7, 1978. With regard to ACRS generic matters, a discussion of NRC Staff efforts leading to satisfactory resolution is contained in a letter to Chairman Bender dated October 25, 1977. Appendix B of the McGuire SER provides the current status of these matters as they relate to this plant. R. Birkel concluded that the NRC Staff does not consider any I of the remaining eight open items or the three generic items to ( pose serious or major problems. Each item will be resolved to the NRC Staff's satisfaction well in advance of a decision to isFue an operating license. Mr. Plesset identified two additional items for which the NRC Staff was requested by the subcommittee to be prepared to answer questions: o The effect of inadvertent injection by the upperhead injection system when the plant is being shutdown and pressure in the primary system falls. e At the subcommittee meeting, D. Riley, California Environmen-tal Study Group, had noted concern that the headbolts that secure the reactor vessel head to the vessel could fail in an
unzippering" node of all the bolts, and that the head might be blown through the containment.
In answer to a question, R. Birkel identified the two conten-tions raised in this case by intervenors: 17 i - - ,. ,
I l l APRIL 6-7, 1978 MINETTES OF THE 216TH ACRS MEETING 1 e financial qualification of the Applicant, and e the effects on the McGuire Plant of seismic events occuring
... , outside the tectonic province in which the plant is located.
Both issues have been addressed in the SER. C. Introduction 1 K. S. Canady, Duke Power Company (Duke) discussed the loca-tion of the site, the Duke service area and transmission grid, the Duke generating capacity, the corporate organization of the
-company, and the general plant layout (see Appendix XXIII) .
In answer to a question, K. S. Canady noted that security i I protection for McGuire will be contracted to an outside organiza-tion. He also noted that quality assurance and control organiza-tions report to a corporate quality assurance director, whose i I organization is outside that of the steam production department. He also noted that there is a decommissioning task force developed within Duke. ; 1 In answer to a question regarding the use of emergency diesel generators at McGuire, instead of an emergency line from a nearby hydroelectric station as at Oconee, K. S. Canady said that Duke was unable to reach agreement with the NRC Staff on the use of hydropower for emergency power at this station. In answer to a. question regarding the NRC Staff's reluctance to permit McGuire to rely upon the hydro station at the Collins - Ford Dam for emergency power, K. Kniel, said that the NRC Staff requirements are that emergency power must meet Seismic Category-1 qualifications and tornado missile protection: the Collins - Ford hydro station does not meet these requirements. In answer to a question regarding hydroelectric generators, C. Wylie, Duke, said that at Keeowee Dam, the experience with the a l hydro station since the two units went in service in April 1971, was that out of approximately 500 manually initiated starts, there , have been 7 failures to start. These failures were not in the ! hydroelectric unit itself, but in the manual initiation circuits. Once the initiation signal was received, the turbines started. With respect to diesel reliability at McGuire, with 4 diesel gen-erators being tested, and with a criterion of reaching speed in 11 seconds, in 545 starts, 544 were successful, and the other reached i speed in 12 seconds. 18 i 1 k w . . - . . - _ . . - _. _ m , se
., c
- MINUTES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 D. Emergency Planning (Questions Only)
In answer to a question regarding relationships with the
, ' North Carolina Department of Human Resources, Radiation Protection l Branch, L. Lewis, Duke, said that Duke is working with this cogni- l zant agency in North Carolina to prepare the draft of the State of North Carolina's emergency plan. This plan will comply with the l NRC requirements, the NRC Regulatory Guide and Checklist, and sup-plements to the Checklist. Appendixes will be prepared for each l of the nuclear stations within the State of North Carolina. He l noted also that Duke has well-established communications with the South Carolina Department of Health and Environmental Control, Radiation Protection Branch, which were developed with the licens-ing and operation of the Duke's Oconee Nuclear Station in South :
Carolina. In addition, the Southern Emergency Response Council, ( made up of representation of all of the Southern States, effec-tively interacts with the states and provides help to each state in the event of an emergency. I E. ECCS Design ;
- 1. UHI Analysis vs. Measurement ;
N. Lauben, NRC Staff, discussed the Staff's review of , the UHI evaluation model from January,1975 through Deceraber, 1977; identified the issues considered since December, 1977, discussed the confirmatory comparisons for the split downcomer model and downcomer model sensitivity; and compared accumulator ; flow rates, upperhead temperature, guide tube flow, and support column flow, as calculated by the Westinghouse UHI models and the Sandia UHI model (see Appendix XXIV).
- 2. Specific Plant Analysis S. Israel, NRC Staff, said that when the Applicant subnit-ted a IDCA analysis for the large break, using the August,1977 model, the analysis indiccted that the double-enSed cold-leg guillotine break with a Moody discharge factor of 0.6 was the most limiting break, and calculated a peak clad temperature of 2164 F. More recently, the Applicant has submitted small break analyses which indicate that a six-inch break is the nost limiting break with peak clad temperature of approximately l 1500 'F. The large break calculations took credit for a finite containment backpressure based on the IOCA calculations. These .
calculations are still under review by the NRC Staff. In ! reviewing the IDCA calculations, the Staff is interested in sensitivity studies regarding the application of UHI. Of interest also is the break spectrum, to assure that the most l 19 L . .m _ , . . . _ _ , ._ _ _ . _ _ - , _y
1 MINIJfES OF-THE 216TH ACRS MEETING APRIL 6-7, 1978 limiting breaks have been identified. Following this discussion, the Applicant will present analyses of these matters, which the NRC Staff .has not seen yet. However, based on the experience
--obtairled during the long review of the UHI evaluation model, the NRC expects that the results will show peak clad tempera-tures below 2100 F for peaking factors of around 2.3.
- 3. UHI Analysis W. Johnson, Westinghouse, discussed the UHI analysis, including predicted ECCS performance, the analysis conditions, compliance with Appendix K, modifications to the ECCS model, loss-of-coolant / temperature analysis code, zircaloy-water problem description, and the calculated peak clad temperatures using the approved UHI evaluation nodel for a double-ended cold-leg guillotine break (see Appendix XXV) . .He discussed the recently discovered metal-water problem in the Westinghouse evaluation model. This error was originally discovered by Fromateme, Westinghouse's French licensee, when they found that the total heat generation rate calculated from the metal-water reaction was low by a factor of 2. Further investigation i showed that the evaluation model had a logic error. He men- i tioned the various reasons why Westinghouse had not found this error previously. Westinghouse believes that this error, in terms of licensing requirements, is significant, but that in terms of margins to safety, it is not.
D. Ross, NRC Staff, disagrees with Westinghouse that the margin of safety was unchanged. In answer to a question, J. Cermak, Westinghouse, said that for non-UHI plants, the effect of this error is in the neighborhood of from 0.15 to 0.25 in peaking factor. With respect to UHI plants, the effect is approximately 100 in the Appendix K calculated temperature. In answer to a question regarding the advantage of higher peaking factors, K. S. Canady said that, with a low peaking factor, the plant must be operated to closer limits than with a higher peaking factor. If there is a relative high peaking factor, the plant can be operated in either a base loaded or load-following mode; the higher peaking factor gives more I operational flexibility. F. Stud Bolts R. E. Tome, Westinghouse, said that his presentation was pre-pared as a result of the question of stud bolt failure in the pressure vessel head as raised at the subcommittee meeting. Each 20
- e. -
l
. l
. MINtTIES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 reactor vessel has 54 7-in. diameter closure studs threaded at each end. There are 8 threads per in. , and the length of the thr.ead engagement between the stud and the vessel flange is 9 in.
The length of the engagement between the stud and the nut is 7 in. Studs, nuts, and washers are made from SA-450 Class-3 material with an ASME code specified minimum room temperature yield strength of. I 130,000 psi, and an ultimate strength of 145,000 psi. The closure stud assemblies for McGuire- Units 1 and 2 were designed, analyzed, fabricated, ' and inspected to the requirements of the 1971 edition of the ' ASME code, Section III, and December,1971 addenda. The design conditions for the stud assemblies are to resist an inter-nal pressure of 2500 psi in the reactor vessel at 6500F, while the I normal operating conditions are to resist 550 psi pressure at 550
*F. There are several stress limits that must be met by the studs during plant operation. The studs are sized so that the average membrane stress in the studs is less than 1/3 of the code speci-fled minimum yield strength for the design conditions. The shear ,
stress in the threads is less than 0.6 of this allowable membrane l stress limit. The maximum membrane stress during normal plant operation occurs at the end of heatup, and is 47.6 ksi, vs. a code allowable of 73.6 ksi, which is 2/3 of the specified minimum yield strength at operating temperature. The maximum membrane plus
- bending stress during normal plant operations also occurs at the end of plant heatup, and is 98.6 ksi vs. a code allowable limit of 110 ksi, which is the specified minimum yield strength at 550 'F.
During ' steady state operation the stresses in the studs are at a membrane stress of 24.4 ksi, and a combined membrane plus bending stress of 57.5 ksi. The hypothetical case for one stud failing instantaneously at the end of the heatup was considered. The dynamic effect of suddenly applying all the loads from the stored energy in the flange below the failed stud and the pressure load increase from the failed stud on just 2 adjacent studs next to the failed stud was calculated to be an average stress across the two studs of 65.7 ksi, which is still below the code allowable limit , of 73.6 kai for normal operation. The maximum calmlated membrane
. bending stress in those studs was found to increase to 116.4 ksi, which will result in some localized yielding. Therefore, there would not be a progressive stud failure mode, or zipper effect from this assumed stud failure. The most critical stress loca-tions on the studs, the threads, were evaluated using the ASME- I specified fatigue reduction factor of 4. The significant stress cycles on the studs were 57 specified startup cycles including a tensioning and untensioning operation, and 143 additional startup cycles without the studs being tensioned or untensioned. The total usage factor for the studs was calculated to be 0.6 vs. code
-allowable value of 1.0, using the ASME design peak curves. The ASME Code provides design fatigue curves for temperatures between 21
s 4 MINUTES OF THE 216TH ACitS MEETING APRIL 6-7, 1978
- f66m temperature and 700 F, which covers the operating temperature of McGuire. The temperature effect is compensated for by code re-quirements. Bolts will be tensioned to a maximum membrane stress
.f 55 ksi, the nut will be tightened, and the tensioner relaxed, which would reduce the membrane stress in the stud. The elonga-tion applied to each stud by the tensioner is controlled by a hydraulic pressure gauge on the tensioner pumping unit. Elonga-tions of all studs are held to 0.0051 inches, with a tolerance of plus or minus 0.0002 inches for normal operation. In addition to other requirements for the material, it also must meet a Brinell hardness between 302 and 388. In addition, Section 3 of the ASME Code requires Charpy impact testing be performed on the material.
Charpy and tensile tests are performed on samples removed from one bar of each heat of material represented or for each charge of 10,000 lb. of material, whichever is less. A review of the record for the McGuire stud material processed to these requirements indicates that the axial strength and ultimate strength in all cases exceed the code-specified minimum allowables. Periodic non-destructive testing, both volumetric and surface, are required I for these studs during fabrication and plant life. The stud material receives a 100% volumetric examination by ultrasonic testing (Ur) in both the radial and axial direction after heat treatment and prior to threading and must meet the ASME Section III acceptance standards. After final machining, the studs receive a 100% magnetic particle examination to detect nonaxial surface indications. The studs then receive a 100% Ur axial scan af ter hydrostatic testing in the shop, prior to their ship-ment to the plant site. The studs receive a 100% visual examina-tion and a 100% Ur examination from one end. During the plant life, the studs must be 100% reinspected on both surfaces, either a Ur or MT and Ur exam, in 10 years with 5 to 33% of the studs being inspected every 40 nonths. R. Tome concluded that the Applicant believes that the reac- ' l tor vessel studs, designed and fabricated to the requirements of Section III of the ASME Code, and inspected to both Section III and XI requirements of the code as required by the federal regula-tions, will not fail and will operate in a reliable manner for the life of the plant. This conclusion has been borne out by experi-ence on operating Westingho.se plants. l Mr. Siess noted that this presentation did not answer the questions raised by Mr. Riley, regarding .possible failure at a pressure of approximately 7000 psi under AIWS conditions, and his claim that 18% of the stud bolts may be below standard strength. 22 i N 0'l9'% N . AA8'WM ' AM__ h as , , ,
MINUTES OF TIIE-216TH ACRS MEETING APRIL 6-7, 1978 R. Tome claimed that none of the studs are below strength, that any, defective studs are replaced before operation, and that there is no concern that the system could reach 7000 psi pres-sure and rupture the studs, since other components would fail first. 1 1 Mr. Shewmon requested informath regarding the specifica- 1 tions to which the holddown bolts ' 'e auxiliary feedwater tur-bine are manufactured. K.S. Canad 'd to provide the informa-tion. G Effect on Fuel of Core Radial Diffe' cressure from Asymmetric LOCA Ioads i l J. Cermak, Westinghouse, said that Westinghouse has performed 1 a scoping calculation of the forces from a pressure wave which moves the core barrel and then the rarefaction wave moving back i across the inside of the core causing a differential pressure ! across the fuel assembly. This load would be the equivalent of 10% of the hydraulic load on the fuel. Westinghouse believes that the conservatism in the design can easily absorb this 10% , pressure, which was not considered in the initial design. Further, i calculations of the forces on the fuel assembly of McGuire, con- 1 sidering both the seismic and hydraulic loading on the fuel assem- ! bly, totaled less than 2000 lbs. Adding the 10% calculated in the bounding calculation, the total loading would be less than 2200 lbs. Both of these loads are less than the lower limits of the experimental data, which is 3200 lbs. Westinghouse concludes that there would be no significant deformation of the fuel assem- i bly, and therefore believes there is no problem. 1 l H. General Questions l In answer to a question regarding whether an operator action at a remote shutdown panel could cause an unsafe condition, T.C. ; McMeekin, Duke, said that there is protection against such a i situation. A rem te shutdown panel alarm will register in the 1 control room if the panel is opened. Further, a set of transfer I switches must physically be opened to transfer control from the j control room to the remote control panel. , i I. Caucus 4 1 The Committee agreed unanimusly that it believed it could l write a letter favorable to the application for an operating j license for the McGuire Nuclear Station, Units 1 and 2. j G 23
* * * " " * = * *
- 4e- _%. w.p.-... .,,,,,,,,._m ,,
MINT 7fES OF THE 216TH ACRS MEETING APRIL 6-7, 1978
- V. Meeting with the NRC Staff on Recent Operating Experience, Licensino Actions, Generic Matters Relating to IMRs and Future Agenda (Open to Public)
[ Note: Thomas G. McCreless was the Designated Federal Employee for this portion of the meeting.] A. Davis-Besse-1: Implementation of ACRS Recomendations from the ACRS Report of January 14, 1977 L. B. Engle, NRC Staff, discussed the implementation of the
. ACRS recommendations included in the Ccamittee's report of January 14, 1977. (For background, see Appendix XXVI.) He said that the purpose of this report is to update the Committee to the current status. He noted that Supplement No.1 to the SER was issued on j April 22, 1977, the same date that an OL was issued for Davis- l Besse, Unit 1. Although the OL was issued authorizing full l power, 960 MWe, the operation of the facility was restricted to a sequence of operational modes until preoperational test, startup test, and other items were completed to the satisfaction of the NRC Staff. This OL stipulated 19 conditions that imposed limita-tions on plant operations, and required special reports and/or modifications to be completed at specific timas following the date of the issuance of the license. Since the issuance of the license, 6 conditions to the license have been removed by amendments,
. supported by SERs, and 2 conditions have been revised.
The reactor attained criticality on August 12, 1977, initial electricity was produced on August 23, and 75% of full power was attained on January 23, 1978. A reactor power of 90% was obtained for 2 days on February 15 and 16, 1977. However, condenser prob-lems required the Licensee to reduce operating power to 75%, until April 3, when power was increased. 100% of rated power was reached on April 4. Based on recommendations by Babcock and Wilcox (B&W), the NSSS vendor, on April 5 the Licensee reduced plant operation to 3 pumps to reduce flow and is currently operating at about 75% of full rated power. Davis-Besse has burnable poison rod assemb-lies, and is therefore affected by the failure of one of these assemblies in the Crystal River Plant. The cumulative service factor has been about 75%, and the unit forced-outage rate has been about 25%. The shutdowns were required for repairing and servicing equipment primarily in the secondary system. The first scheduled refueling outage is planned for late 1979. In answer to a question, L. B. Engle said that the condenser problems to which he referred involved some tube leaks. It has not yet been determined whether these leaks are from flow-induced vibration as a function of power. Tubes will be inspected during a planned outage. 24 = __-_ = _ - - - - ---
,_q
-MINITIES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 L. B. Engle reviewed the items raised in the Committee's January 14, 1977 report:
~
e' Increase of seismic design basis from 0.15g to 0.29 The NRC Staff stipulated in the operating license that the Licen-see shall subnit a seismic reanalysis and evaluation to the NRC for its review in sufficient time tc obtain Commission approval'of the adequacy of the plant systems needed to accomplish safe shutdown of the plant and continued shutdown heat renoval prior to startup following the first regularly scheduled refueling outage. In performing the reanalysis, a safe shutdown earthquake of 0.2g shall be applied at the foundation level of the plant, and the response spectra shall be as specified in Reg. Guide 1.60. The NRC Staff is in the process of developing guidelines for this seismic reanalysis. e ECCS. The NRC Staff has reviewed revised nucleate boiling logic proposed by B&W, which does not allow return to nucle-ate boiling after critical heat flux conditions are reached The NRC Staff has determined that the revised logic was en appropriate change to be incorporated in the B&W evaluation
- nodel, that the overall effect of the change on peak clad temperature was insignificant, and that it met the Acceptance Criteria. The Applicant has subnitted additional analyses correcting for fuel pin pressure errors and erroneous flow resistance values for the reactor vessel inlet nozzle. The NRC Staff has determined that the ECCS analysis for Davis-Besse, Unit 1, is in accordance with Appendix K.
o Large Break Analysis. The operating license stipulated that, within six months from its issuance, the Licensee shall provide additional supporting analysis for the large break spectrum to document the exact margins, and should provide to the NRC Staff reactor coolant system flow data. The Licensee submitted the large-break spectrum on October 21, but, be-cause of delays in plant operation, the coolant system flow data were rot available. The license condition was revised to require that within 30 days following 2 weeks of sustained , reactor power operation at a power level of 90% or greater of l rated thermal power, the Licensee provide operating reactor ! system coolant flow data. However, during the 2 days that 1 the plant was at 90% rated thermal power, the Licensee was 1 able to obtain the system flow data, and is getting ready to I subnit the information to the NRC Staff by mid-April. 25
. , a_ . erw e 8P *= r* ****=%ndh,we scopwr. , m ,. ,e,ws,,-pe.w.,, ,w,ee--.%e
APRIL 6-7, 1978 f MINtfrES OF TIIE' 216TH ACRS MEETING i e Improved Radiatic' Surveillance in Ohio. The State of Ohio has indicated to the NRC Staff that it is initiating a pro-
...,, gram which will eventually qualify for an NRC state contract
~
for' technical aid. e long Term Capability of Hermetic Geals. Environmental quali-fication of equipment is being pursued by the NRC Staff as part of a Category A generic activity, Task Action Plan A-24, Qualification of Class IE Safety-Related Equipnent, to assure proper performance of seal materials during plant operation. The e Instrumentation to Follow the Course of an Accident. and a NRC Staff has issued Regulatory Guide 1.97 (Rev.1), technical activities steering committee was established on August 22, 1977. The NRC Staff has concluded that the instrumentation to monitor post-accident conditions ::0! the NRC Staff criteria and was acceptable. However, Davis-Lesse 1 does not meet Position C.3 of Regulatory Guide 1.97, Revision 1, and guidance is being developed in this area. e A'IWS. This is a generic matter. The NRC Staff has prepared a draft generic technical report on AIWS which incorporates the comments and concerns of the industry, including the Babcock and Wilcox Company. This draft report is currently being reviewed by NRC management and the NRC Staff. An open meeting will be held on April 19, 1977, to discuss the stud-ies concerning AIWS. e Fire Protection. The Licensee has submitted its fire hazard analysis report, and the NRC Staff has determined that the report was not adequate for determining the fire protec-tion program in accordance with Appendix A to the Branch : Technical Position. Therefore, a condition was placed on the license, stipulating that within three years from the date of l issuance of the license, the Licensee shall increase the level of fire protection in the facility to the levels recom-mended in Appendix A, or with alternatives acceptable to
- the NRC Staf f. Prior to startup following a first regularly scheduled refueling outage, the Licensee shall implement Section B of Appendix A, Administrative Procedures, Controls, and Fire Brigade, and Section C of Appendix A, Quality Assurance Program. Since the issuance of the OL, the reevalu-the fire protection program has continued. The ation of Licensee has presented additional information on the installa- The tion of fire retardant seals for electric penetrations.
~
26
l MItCI'ES OF THE 216TH ACES MEETING APRIL 6-7, 1978 ! NRC Staff has concluded that these seals through fire barri-ers as installed at Davis-Besse are conditionally acceptable. However, the NRC Staff h'as required that the Licensee perform
- ~ .certain full-scale testing to verify the adequacy of the configuration and installation of the seals that represented the worst-case departure from sections originally tested under the ANSI E119 standard. The NRC Staff is scheduled to complete the Davis Besse Unit 1 fire protection review before the end of of calendar year 1978.
e Industrial Security. The NRC Staff has reviewed the Licen-see's amended security plan as required by NCFR 5073.55, and
- completed their Phase 1 review in September of 1977. The Licensee has sutmitted a nodified security plan. This plan is being evaluated.
4 B. Oconee: Microseismicity J. Kelleher, NRC Staff, discusscd a microcarthquake swarm occurring at the Oconee site in January,1978. He noted that the cause of this swarm is not clear. The maximum magnitude of these microearthquakes was in the range of 2 to 2.5 on the Richter scale. He noted that there was a network of microseismometers located at this site. In answer to a question, J. Kelleher said that there are very few areas in Eastern United States that are currently being noni-tored for microseismicity, and that it is possible that similar swarms of microcarthquakes could be taking place elsewhere without their being detected. Regarding the reported . swarm, only one earthquake was re-ported as being " felt". (For the data presented, see Appendix XXVII.) !
- C. Combustion Engineering Plants: Control Element Guide Tube Wear H. Levin, NRC Staff, stated that a problem of cracking of control element guide tubes was first identified in the Millstone 2 Plant on December 14, 1977. He reviewed the chronology of the actions taken since this problem was first identified, and dis-cussed the design of the components, the safety considerations involved, the observations made on the worn components, the interim fixes accepted by the NRC Staff, the bases for continued operation, and the susceptibility of other NSSS designs to guide tube wear (see Appendix XXVIII).
27 J4e 4% f-"-- s egg _ _ gg gg -ve*Lt-u W*+4- DN@ *-et Q eem M.4ema 4 me _.
~ - - l MINtJfES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 1 Mr. Bender suggested that _.plicants establish whether there are suitable out-of-pile vibration test arrangements for all l f6 actor internals configurations to evaluate the effect of repre-sentative working conditions. D. Crystal River 3: Failure of Burnable Poison Rod Assembly K. Seyfrit, NRC Staff e first discussed events leading up to the identification of a burnable poison rod assembly failure at Crystal River, Unit 3. On September 12, 1977, following recovery from a scram, a quadrant tilt of 7% was noted, which disappeared after several hours of operation. On January 1 and again on January 3, 1978, there was an alarm on the loose parts monitor associated with the B steam generator; this disappeared after a short period of time. On February 17, there was another alarm on the loose parts monitor associated with the B seam generator, which persisted. The Licensee performed a number of investiga-tions, including examining the chemistry of the primary coolant, and looking at other loose parts monitors. No other abnormalities were noted at that time. Because of the persistence of the, noise in B steam generator, one of the reactor coolant pumps was shut down, and power was reduced to about 78%. This reduction in power and flow eliminated the noise at that point. A B&W investi-gation team was called in, evaluated the data that was available, and confirmed that there were some 1cose parts in the top of the B steam generator. Operation continued from February 18 to March 3, at which time the Licensee was able to determine that there had been a small amount of steam generator tube leakage, on the order of a gallon a day. The unit was taken off-line on March 3, and a cool-down begun. Observation through a manhole on B steam genera-tor identified some loose parts, including the coupling and spider for the burnable poison rod assembly B-47. In addition, there j were other parts identified, including pieces of cladding, aw some evidence of damage of the tube ends. On March 13,,the reactor vessel head was removed, and a second burnable poison rod, B-52, was observed sticking up. Several poison pins were broken off. Examination of the latching mechanism indicated that one of the balls was missing, and grooves were found in the hold-down latch assembly that corresponded to the location of the missing ball. She remainder of the fuel assembly appeared to be in good condition. The Licensee has postulated three possible causes for j the damage, and is investigating to determine the actual cause. ; The first two assumptions involve a manufacturing deficiency and j the possibility that the assembly was not initially latched. The ] third assumption is that the wear was caused by vibration. 28 4 I 8
MINUI'ES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 K. Seyfrit stated that at this time there are only two other plants in operation using these burnable poison assemblies, Davis-
'Besse-1, and Three Mile Island-2. It is likely that the burnable poison assemblies will be removed from Davis-Besse within the .
next ten days to two weeks, when the plant is shut d,own for other planned maintenance. He noted minor damage to the steam j generator, and inferred that part of the problem is the extension ( of the tubes approximately 3/4 in, above the tube sheets. The l I probable fix for this problem may be the removal of the extension I of the tubes above the tube sheet. (For details, see Appendix ) XXIX.) , E. Implementation of Regulatory Guide 1.97, " Instrumentation to Follow the Course of an Accident" F. Hebdon, NRC Staff, stated that when the work had been com-pleted on Regulatory Guide 1.97 (Rev.1), the NRC Staff began work on Task Action Plan A-34 to develop detailed acceptance criteria and guidance to be used by Applicants, Licensees, and NRC Staff reviewers to support implementation of this guide. During the development of this Task Action Plan, it was recognized that cer-tain instruments were described in the Guide with such clarity that implementation on that part of the Guide could proceed more quickly than could implementation of the entire Guide. Therefore, the NRC Staff decided to divide implementation of Reg. Guide 1.97 into two phases: o Phase 1 incorporates the recommendations of position C.3 of the Guide. Position C.3 describes the specific instrumenta-tion to be used if accident conditions degrade beyond those assumed in the FSAR. It was believed that position C.3, by itself, constitutes an interim solution that could be implemented on all operating plants in a timely manner, while i the more time consuming case-by-case review described'in the ! remainder of the Regulatoy Guide is completed. e Phase 2 incorporates the remainder of the regulatory posi- i tions in Regulatory Guide 1.97. The principle position is position C.1 which states that for postulated accidents listed in Chapter 15 of Regulatory Guide 1.70, the Applicant shall perform a detailed safety analysis to determine the parameters that should be measured to provide the operator with essential information concerning the nature of an acci-dont and the response of available safety systems. 29 t w h e en 4.s -o wm=.es. >- =ww
--N6=4-- *"NhM*-' '
--Q
o 3 3 . d MINlTTES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 LaSalle County and Watts Bar were selected for Phase 1, position C.3, and Allens Creek and Sundesert were selected for Phase 2, or for full implementation of Regulatory Guide 1.97. Subsequently, in response to the. ACRS report on Diablo Canyon, Diablo Canyon was added as a lead plant for implementation of position C.3. In Aua"st 1977, Regulatory Guide 1.97 (Rev. 1) was issued. The NRC Staff has characterized this revision as Category 3, backfit require 6 for all applications in review, and further NRC Staff consideration of individual cases required in order to determine the need for backfitting for all operating plants. Two problems have developed in dealing with Applicants:
- 1. Technical questions, such as the definition of identifiable release points described in position C.3 have been raised.
- 2. Philosophical problems concerning the apparent commitment in ;
position C.3 to include instrumentation to monitor accidents l that go beyond Class 8 have been raised. The NRC Staff has made some progress in clarifying its position and in resolving the Applicants' concerns in these areas. How-ever, work remains to be done in both areas. In March, the NRC Staff sent additional guidance to the Applicants and requested the proposal for implementation of Reg. Guide 1.97 be submitted no later than May 1,1978. F. Seismic Monitoring on the Eastern Seaboard C. Stepp, NRC Staff, described the seismic monitoring net-works in Eastern United States, noting that the NRC Staff is fund-ing entirely or partially the operation of these networks. He said that there are two principle networks along the eastern sea-board, the Northeastern Network, and a network around Charleston, SC. The Northeastern Network includes about 30 microseismic monitoring stations that will detect local earthquakes of smaller than magnitude Richter 3. The Charleston network includes about 12 microseismic stations. Another network is currently being installed in the central Virginia region, and will consist of 5 microseismic stations. In addition, another 5-station micro-seismic network is being established in southwestern Virginia, in the Giles ' County earthquake zone. It is proposed that these microseismic stations in Virginia be operated by Virginia Poly-technic Institute. j i 30 Y f lr 4
t es s MINUTES OF THE 216TH ACRS MEETING 3.PRIL 6-7, 1978 C. Stepp said that four of the microseismic instruments cur-rently located in the temporary network at North Anna will be
-incorporated into the central Virginia network. He noted that there ar'e currently 17 stations in the Lake Anna area. The effect of the reduction of the number of microseismic stations will be loss of accuracy in pinpointing epicenters of seismic disturbances.
The accuracy of the proposed network would be only approximately plus or minus 5 km. Because of the concentration of the four sta-tions at North Anna, the accuracy in this particular area would be on the order of from 1 to 1.5 km. (For details see Appendix XXX.) l The Committee raised no objections regarding the proposal to reduce the number of instruments at the North Anna seismic network. G. Monitoring Neutron Exposure at Nuclear Facilities S. Block, NRC Staff, said that, as a result of questions arising regarding the exposure of nuclear plant personnel to unknown amounts of neutrons, and the possible inadequacy of neutron detection and recording instruments, this problem is being considered by the NRC Staff. He noted that the Office of Nuclear Regulatory Research has been requested to initiate a program for the purpose of collecting data on the effectiveness of personnel l neutron dosimetry programs at the operating nuclear power plants (see Appendix XXXI). He pointed out that, for certain energy ranges, adequate dosimeters have not been developed, and he ~ recommended that such instruments be developed. He reviewed the neutron dosimetry records that are available from operating plants, and was unable to identify a significant problem. He suggested that the likelihood of overexposure, in the range where instrumentation has not been available, is highly unlikely. However he recommended that further investigations be nude. (For ) details on personnel neutron dosimetry methodology, see Appendix XXXII. ) H. Future Aaenda The Committee approved a tentative future schedule (see Appendix II). L. Crocker, NRC Staff, in discussing the scheduling of a re-view of Indian Point Nuclear Power Plant, Unit 3, for an increase to full rated power, said that the SER has been published and is
- being delivered to the Committee. However, at this time, the impact of the programing error discovered in the Westinghouse ECCS evaluation model, noted earlier in the meeting, is not known.
1 31 p .. - _ _ _ . . . . _ _ . _ _ _ _ _ _ __ __._ .
, ,,s
+e, e,
.o 4 4 MINUTES OF THE 216TH ACRS MEETING APRIL 6-7, 1978 Mr. Bender requested that the Staff provide the Committee with a written statement regarding the details of the error, and
'how it was found.
The Committee again requested that the NRC Staff inform the Committee of any significant design changes regarding the currently-proposed Allens Creek Nuclear Plant and the design pre-viously reviewed by the Committee. VI. Executive Sessions (Open to Public) [ Note: James M. Jacobs was the Designated Federal Employee for this portion of the meeting.) A. Regulatory Activities Subcommittee Report
- 1. Revision of 10 CFR 50.44 Mr. Siess, Subcommittee Chairman, recalled that at the 1 215th ACRS meeting, the Committee declined to approve the !
proposed changes in wording of 10 CPR 50.44, relating to : combustible gas control following a IDCA. He said that the NRC ! Staff has decided that their position to not permit repressuri-zation, along with purging, is the conservative position, and it.is their intention to propose the changes to the rule to the Commission without the ACRS' blessing. They do recognize, however, that there are questions about the desirability of repressurization, and therefore have requested the Office of Nuclear Regulatory Research to include a study of repressuriza-0 Lion in connection with its proposed research project on advanced containment concepts, involving vented containments. It is hoped that risk information can be developed eventually to resolve the question. The subcommittee does not believe l that this matter requires additional action at this time by the ] Committee. ,
)
- 2. Regulatory Guides 1
The Committee approved the following Regulatory Guides: l e Regulatory Guide 1.29 (Rev. 3), Seismic Design Classifica-tion, and .
- Regulatory Guide 1.68 (Rev. 2), Initial Test Program for Water-Cooled Nuclear Power Plants. j s
32 g i I
2 , g ek J MINUTES OF THE 216TH ACRS MEETING , APRIL 6-7, 1978 During the discussion of Regulatory Guide 1.68 (Rev. 2)
-Mr. Bender recommended that the NRC Staff allot some technical assistance funds to establish how this Regulatory Guide is
,. 'being applied throughout the nuclear industry with the antici-pation that the Guide will ultimately be modified to define j effective preoperational test practices for nuclear power plants. The current Guide does not, in his opinion, provide adequate information for the purposes of regulation.
- 3. Regulatory Activities Subcommittee Agenda for its May Meeting Mr. Siess noted that the following items are scheduled to l
be considered at the May 3, 1978 meeting of the Regulatory ' Activities Subcommittee: e Regulatory Guide on Lightning Protection, l e Regulatory Guide 1.9 (Rev. 1), Selection, Design, and ; Qualification of Diesel Generation Units Used as On-Site J Electric Power Systems at Nuclear Power Plants, i e Regulatory Guide 1.63, Electrical Penetration Assemblies,
, Regulatory Guide 1.130, Service Limits and Ioading Combin-ations for Class I Plate and Shell Type Component Supoorts
- nd l 1
e Proposed Amendment to 10 CFR 55 Appendix A, Codes and l Standards. B. ACRS Ouarterly Report to Commissioners The Chairman noted that Members have received a draft copy of I proposed quarterly report to the Commissioners for the period,
, December, 1977 through March, 1978. He requested that Members pro-vide the ACRS Office with their comments on this report within the next week.
C. Testimony to Senate Subcommittee on Nuclear Regulation The Chairman and the Executive Director stated that they would prepare a final statement to be used as testimony before the Senato Committee on Environment and Public Works, Subcommittee on Nuclear Regulation, Senator llart, Chairman. He noted that a draft
-copy of the testimony had been provided the Members, and requested ,
that their connents be subnitted to the ACRS Office as soon as possible. The Chairman noted that he would be accompanied to the 33 l
-~ < ~,~ - - .. _ .. .__ ,_,_ _ __ ,, ,
' MINUTES OF THE 216TH ACTS MEETING APRIL 6-7, 1978 hearing by the Executive Secretary, the Vice Chairman, and Messrs.
Bender and Siess. He welcomed the presence of any other Members
-who could participate.
D. Activities of'the Members
- 1. Mr. Shewmon It was the consensus of the Committee that Mr. Shewmon should not act as a consultant-without-pay to the Westinghouse Material Research Laboratory.
- 2. Mr. Moeller The Committee offered no objection to Mr. Moeller's pre-paring a paper for the British Nuclear Energy Society's November 1978 meeting. The paper will be on radiation pro-tection in the ' fuel cycle, and will consist primarily of an abridgment of chapter 7 of the Committee's Annual Report to Congress (1977) (NUREG-0392) and new material developed for the 1978 report.
The Committee offered no objection to Mr. Moeller 's lecturing at an MIT safety course. j E. Proposed Independent Nuclear Accident Review Board The Chairman designated a working group to develop a position on a proposed independent nuclear accident review board. Messrs. Bender and Shewann volunteered to serve on this working group. F. Proposed Meeting with Groupe Permanent
.The Committee agreed to defer until the 217th ACRS meetino l the setting of dates for a meeting with the French Groupe Perma-nont, to give more Members a chance to check their personal appointment calendars.
1 G. Reorganization of ACRS Subcommittees ] i The Committee agreed to defer until the 217th ACRS meeting a discussion of the reorganization of,ACRS Subcommittees. H. ACRS Reports and Ictters
- 1. Letter to Dr. E. J. Sternglass The Committee prepared a letter to Dr. E. J. Sternglass, '
1 in response to his request for ACRS review of changes in cancer mortality in the vicinity of several nuclear plants. i 34 _m ~ . _ _ .
APRIL 6-7, 1978 g MINCTfES OF THE 216TH ACRS MEETING ll (For background material on the request, see Appendix XXXIII;
....for. Committee reply, see Appendix XXXIV.)
s VII. Executive Sessions (Closed to Public) s [ Note: James M. Jacobs was the Designated Federal Employee for thic portion of the meeting.] A. New Members The Committee agreed ' to nominate [ J J
} .:' jfor appointment to the Committee.
B. ACRS Reports and Ietters l
- 1. Arkansas Nuclear 1, Unit 2 Nuclear Power Plant A report was prepared advising the Commissioners that the Committee believes that, subject to certain conditions, there l is reasonable assurance that the Arkansas Nuclear One, Unit 2 i l
Nuclear Power Plant can be operated at core power levels up to 2815 MWt without undue risk to the health and safety of the public (see Appendix XXXVI). r
- 2. McGuire Nuclear Station, Units 1 and 2 A report was prepared advising the Commissioners that the .
Committee believes that, subject to certain conditions, there is reasonable assurance that the McGuire Nuclear Station, Units 1 and 2, can be operated at power levels up to 3411 MWt without undue risk to the health and cafety of the public (see Appendix l XXXVI).
- 3. Liquid Pathway Generic Study Although the Committee completed its review of the NRC Staff's draf t report, NUREG-0440, Liquid Pathway Generic Study, time did not permit completion of a report. The Committee agreed to table completion of this report until the 217th ACRS meeting. ,
The 216th ACRS meeting was adjourned at 8:00 p.m., Friday, April 7,1978. 1 35 i L
OU ACRS Meeting Meeting Dates: April 6-7,1978 APPENDIX I ATTENDEES ADVISORY COMMITTEE ON REACIOR SAFEGUARDS Stephen Lawroski, Chairman Max W. Carbon, Vice-Chairte.an Myer Bender Harold Etherington Herbert S. Isbin - William Kerr J. Carson Mark Dade W. Moeller Milton S. Plesset Paul G. Shewnnn Chester P. Siess ACRS STAFF , Raymond F. Fraley, Executive Director , Marvin C. Gaske, Assistant Executive Director \ Herman Alderman Andrew L. Bates , Paul A. Boehnert Sam Duraiswamy Elpidio G. Igne James M. Jacobs Morton W. Libarkin Richard K. Major Thomas G. McCreless John C. McKinley Robert E. McKinney Ragnwald Muller Gary R. Quittschreiber Jean A. Robinette Richard P. Savio Hugh E. Voress Robert L. Wright 03NSULTANTS Ivan Catton Ellert P. Epler Richard F. Foster p Walter C. Lipinski L R-I
1 NRC ATTENDEES O 21cT" ^cas "To-April 6, 1978 Div. of Systems Safety _ Div. of Systems Evaluation J. P. Joyce L. Belfracchi R. Codell G. Chipman M. W. Hodges R.Schemel , R. L. Tedes:o L. Phillips H. Berkson l G. Lainas T. A. Ippolito D. R. Muller ! R. A. Vollmer j D. Pickett S. Israel l B. Turovlin D. F. Bunch A. J. Szukiewicz L. G. Hulman P. W. Baranowsky H. E. Polk ' H. Conrad l I RESEARCH Div. of Project Management R. DiSalvo F. Manning R. E. Martin M. A. Taylor D. B. Vassallo J. A. Murphy F. R. Haventi L. P. Crocker J. F. Stolz O~ p s NRC Consultants I f.P.
, ;, g A. R. Markese J. B. Bullock * ,
K. Kniel l ACRS Consultants Attending ll. C. Lioin ni E. P. Epier R. Foster I Catton O Wa ,. s ,
l ATTENDEES - APPLICANT
)
l 216TH ACRS MEETING April 6,1978 l ARKANSAS POWER & LIGHT _ I C0!CUSTI0:1 ENGIflEERIrlG' I J. R. Perdue . R. R. Mills G. H. Miller W. J. Gill M. Cajanaugh T. A. Jones O. R. Sikes A. B. Spinell, Jr. G. G. Young H. E. Heuschaefer D. G. Wardis E. H. Kennedy fleal A. Moore
. R. Hamphries J. R. Marshall E. M. Brown D. Rueter T. M. Starr D. H. Williams J. F. Church B. A. Terwilliger F. C. Sernatinger T. G. Shultz OFFSHORE POWER SYSTEMS BECHTEL ~J.A. Nutant 1 e E. H. Smith H. J. Stumpf b J. C. Bradford M. S. Iyer K. C. Perry V. W. Campbell i
i P. B. Haga ! B. Z. Cowan
' C. A. Pelletier SAI D. C. Aabye i
R. L. Ritzman H. A. Capo i 1 A. S.Caerdrin D. Howaeesee G. R. Collin D. llalker J. E. Tabugen, Consultant to OPS T. Pudlin, consultant to OPS E. M. Buchak, consultant to OPS H- 3 - . - n.,
PUBLIC ATTEl!0EES 216TH ACRS MEETING Thursday, April 6,1978 P. E. Grossman, Jr., Ebasco Services, In., NY R. W. Prados, Louisiana Power & Light Co. 2425 Ramsey Drive, New Orleans, LA R. Borsum, B&W, Derwood, MD
~
P. S.Damerell - MPR Associates W. W. Little, Westinghouse Hanford Co. , S. J. Weems, MPR Associates , ^ (',.)i 1 O e-Y
-_ x
1 l (_)s URC ATTENDEES 216TH ACRS MTG. i April 7, 1978 Div. of Project Management Div. of Operating Reactors C. Garland
- L. Engle R. E. Johnson L. P. Crocker G. F. Lauih C. W. Moon E. L. Conner D. B.Vassallo G. S. Vissing A. W. Drumerick H. A. Levin J. F. Stolz F. D. Coffman I K. Kniel E. Moler '
C. Van Niel 1 Div. of Systems Safety J. Stepp Inspection & Enforcement C. F. Miller K. Seyfrit H. C. LI H. A. Wilber M. Hartzman D. Suy G. N. Lauben W. Milsted S. Israel () RSLB S. Berggren R. G. Fitzpatrick N. H. Wagner R. J. Bosnak i MIPC C. H. Hofmayer ) R. Denning D. F. Ross J. P. Knight ! Div. of Systems Eval. M. D. Houston F. Hebdon M. R. Hum G. Chysman F. C. Cherny G. B. Staley G. Mazetes W. P. Gammill l J. Grieves Stds. Development A. Huite RSR J. Kelbher Nuclear Reactor Regulation DA B. Grimer R7 K. Grahm R. B. McMullen EEB ORB 4 S. Block R. M. Reid
%j)
[ s
/1- s r 1 3
l'h -) Applicant Attendees i 216TH ACRS Mocting April 7, 1978 e i l Westinghouse l Duke Power N. J. Kiparalo D. B. Blackmon E. V. Somers G. A. Copp C. R. Srerrett l R. F. Wardell W. J. Johnson L. Lewis J. O. Cermak i' . P . Harrall R. S. Howard R. A. Pace R. S. Borgraber T. C. McMeckin F. Landerman K. S. Seidle A. Ball, Jr. l K. S. Canady R. E. Tome W. J. McCabe W. F. Beaver R. B. Priory () D. L. Camup W. Parker, Jr. Toledo Edison R. E. Sund W* LOWGf D. G. Owen C. J. Wylie R. A. Pearce L. Dail Law Engr. P. M. Abraham C. E. Sams G. Cage D. C. Holt B. Rice J. Foley,J r.
~
N' ' . r
PUBLIC ATTEllDEES
,-3
(_ ) 216TH ACRS MEETIllG Friday, April 7,1978 - A.M. June Allen - NAEC - Charlottesville, VA R. Borsum - B&W - Derwood, MD R. S. Bhatnagar - Duke Power Co. - Charlotte, NC Paul Grossman - Ebasco Services, Inc. - flew York, NY Richard M. Kac.ich - Northeast Utilities - Hartford, CT James B. McIlvaine - Bechtel Power Corp. - Frederick, MD R. R. Mills - Combustion Engineering - Windsor, CT R. M. fleil - VEPC0 - Richmond, VA R. C. L. Olson - Baltimore Gas & Electric Co. - Lutherville, MD Thomas R. Robbins - Pickard Lowe & Garrick (Toledo Edison Co.) - Crofton, M: l Scott Sunde - Greenville flews - Greenville, SC R. E. Schaffstall - GE - Reston, VA David Sokolsky - Self - San Francisco, CA l P. B. Haga, OPS, Jacksonville, FL . B. A. McIntyre, Westinghouse Electric Corp. , Irwin, PA S. E. Jacobs, Westinghouse, Pittsburgh, PA M. Young, Westinghouse, Pittsburgh, PA
,7 U
1 l l 1
^%
V
- -7 r
1 APPENDIX II - O- ACRS FUTURE AGENDA 4/3/78 ACRS MEETING TYPE OF REACTOR SER ISSUE PROJECT REVIEW VENDOR DATE YE MAINE YANKEE. POWER INCREASE CE 4/3/78 INDIAN POINT 3 -FULL POWER W 4/3/78 9 ME NEW ENGLAND 1&2 CP W 5/1/78 DIABLO CANYON 1&2 OL W 5/1/78 DAVIS BESSE 283 CP B&W 5/1/78 n1 RESAR-414 STD NSSS W 6/1/78 ALLENS CREEK 1 CP GE 6/1/78 S8G SP - 6/1/78 AUGUST ERIE 182 CP B&W 7/3/78 FFTF SP - 7/3/78 SUNDESERT 182 CP 4 7/3/78 NORTH C0AST ESR- - 7/3/78 O g- r
. - ~
'l O ACRS FUTURE AGENDA qf3f73 l ACRS MEETING TYPE OF REACTOR SER ISSUE PROJECT REVIEW VENDOR DATE SEPTEMBER SEQUOYAH 182 OL W 8/1/78 O 1 1 M
.O ig- 7
E' 1
.. APPEllDIX III Letter, Rep. T. Bavill to Chr.:n. IMndrie _-
.... . ...,~ - on ACRS Fellowship Procram ue n e,.u., o... n -- -
,_ i
~ . , . . . . - _ . . . . _ '
.= .:=._::
- , - . . ~ . . .
3 21I
-.~ .LOF Cong:esof tijcEInitchsbtated
~.
05.5....'.iC- _. C.'a.'J.00.?
- . . . . K)oudeof Ecprcdctitatibed AC ~. _.~'.. ,. .TA 35F,63,,..;;- Cottttttitter att Tipproptintiottd Man!)fngfon, D.C. 20515 E3Q L" *. "'O. *~L E
~
y M.., .u..
,,, ,,y; , r,r i 1
'll"77.,"n.7. ";;""'- March 16, 1978
~
""" '.lll',.*,.'me {
. ~.
- u. -.u. , % -
. . -m ,
lllt7al'"'.00 - '"'-== l lllll";'. "."#r,'! ** *""'-
. ".";;;T'.
.O ;
1
"'?llCu".*:k7.*- .
tll."' t:..?.'.O.IS u. '
~:.,:' ,:';"; .": -'. - %
Honorable Joseph H. Hendria
- Chairman U.S. Nuclear Regulatory Cou=iission U.tshington, DC 20555 -
~ '
O - -
~
The Committee has received and considered your , February 28, 1978, request to reprogran $300,000 in FY 1978 appropriations for a fcilowship program which would assist the Advisory Co=nittee un Reactor Safeguards. During hearings on the FY 1979 budget request, the Co- 4ssion identified several important areas where the regulatory progran needs improvenen t, in order to reduca nucica: licensing times and to increase confidenca in reactor safety. These include the inprovc=ent of internal processing schedules, additional on-site inspections, additional ad=inis-trative support and others. In view of these higher priority prograti needs, the coi=nittee does not approva che reprogramming of funds ce f"#*te a fellowship program. Sincerely, I, TOM BEVILL Chairman -
. Public Work.s Subcoc=tittee .
O . - - -
/ -/0 .
r-
APPEi! DIX IV AfC-2: Project Status Report ' O PICIECT STATUS REPORP ARKANSAS NUCLEAR CNE, UNIT 2 BACKGROUND: , he NRC Staff issued the original Safety Evaluation Report for Arkansas Nuclear One, Unit 2 (ANO-2) Operating License review on November 11, 1977 and a Supplement No. 1 in early March 1978. me ANO-2 project was reviewed at ACRS 93==ittee meetings in Russellville, Arkansas on June 24, 1977 and in Washington, D.C. on February 2, 1978 and March 20, 1978. %e Cm-bustion Engineering Core Protection Calculator System (CPCS), which is being reviewed as part of the ANO-2 docket, since ANO-2 will be the first 1 operating plant to use this new system, was reviewed by the CESSAR System 80 Subemmittee on February 28, 1975 in Windsor, Connecticut and by the Electrical Systems,' Control and Instrumentation Subcommittee in Windsor , Connecticut on May 20, 1977 and in Washington, D.C. on June 30, 1977 and l March 20, 1978 (Highlights attached). %e CPCS was not reviewed by the NRC or ACRS at the ANO-2 construction permit stage since this is a newly developed system subsequent to the construction permit review. The ACRS, during its 214th meeting, February 9-11, 1978 partially reviewed O the splication of Arkansas Power and Light Cmpany for a license to oper-ate ANO-2. All areas pertinent to the oprating license review were covered except for the CPCS/COLSS. At the conclusion of that meeting, it was pointed out that it did not appear that any significant new items were opened up by the Committee beyond those identified by the NRC Staff. %e ACRS decided not to write an interim letter on ANO-2 at that meeting due to the large number of outstanding issues and the inccuplete review of the CPCS/COLSS. PLANT DESCRIPTION: Many features of the design of ANO-2 are similar to Calvert Cliffs 1 and 2 except that the ANO-2 plant will use fuel assemblies with a 16 x 16 fuel rod array, while the Calvert Cliffs 1 and 2 use 14 x 14 fuel rod assemblies. %e initial power of ANO-2 core is 2815 MWt (approximately 912 net MWe), ccupared to the initial power level of 2560 MWt (approximately 810 net MWe) for Calvert Cliffs 1 and 2. A reactor design caparison between ANO-2 and Calvert Cliffs 1 and 2 is included as Attachment 2. The ANO-2 reactor will be the first to use Ccubustion Engineering 16 x 16 fuel rod assembly design. tis fuel will be longer than the previous Ccabustion Engineering 14 x 14 design. Because of this and the fact that there will be more fuel rods per fuel assembly, the fuel rods will operate at lower linear heat generation rates. me cladding also has a larger thickness-to-diameter ratio than the 14 x 14 design.
#- d --
f3 y The ANO-2 will be the first in the United States to use a digital com-puterized Core Protection Calculator System (CPCS) as part of the reactor protection system. Se remainder of the reactor protection system is conventional analog hard-wired equipnent. n e CPCS, in conjunction-with the overall reactor protection system, is designed to provide at least the same level of protection to the core as a conventional, hard-wired system. n e CPCS is designed to initiate a reactor trip for the following events: (1) Uncontrolled control element assembly (CEA) withdrawal from a critical condition. (2) CEA misoperation. (3) Uncontrolled boron dilution. (4) Total and partial loss of reactor coolant forced flow. (5) Excess heat removal due to secondary system malfunction. s pi . Q (6) Steam generator tube rupture with and without a concurrent loss of offsite alternating current (AC) power. Backup trips are available to limit the consequences of each of the above events, even with failure of the CPCS, except for the CEA misoperation. It is not clear how limited the consequences will be in the event of CPCS system failure. n e NRC Staff has not done an independent evaluation of these consequences. The ANO-2 will also use a new reactor monitoring system, designated as the Core Operating Limit Supervisory System (COISS), to continuously monitor important reactor characteristics and establish margins to cper-ating limits. Bis system will use the output of the incore detector system to synthesize the core average axial power distribution. This is not considered to be a safety system and as a result has not been reviewed in detail by the NRC Staff. OUTSTANDING ISSUES: Attachment 3 provides a list of outstanding issues discussed at the last Submmmittee meeting held on March 20, 1978. O) L
,Q- / &
HIGHLIGHTS ]v COMBINED ELECTRICAL SYSTEMS, CONTROL AND INSTRG4ENTATION AND ' HIE l ARKANSAS NUCLEAR 04E UNIT 2 NUCLEAR PLANT , SUBCOMMITTEE MEETING WASHING' ION, D.C. MARQi 20, 1978 9
- 1. Ten outstanding NRC Staff Positions on the Core Protection Calculator System were outstanding. These include the'following:
Position 1
- Uncertainty values in CPC data base nust be e.xperimentally qualified - requires measure- ,
ments at'startup l Position 4 Separation Criteria between the optical isolator cards in the CEAC - Contingent on Position 26 l Position 5
- Cable Separation - Susceptibility to noise - '
requires check during startup l Position 12
- Noise susceptability - Resolved pending confir-O maeion mee uremeate durias eertue i
Position 14 Adequacy of seismic loads - report of tests is i under NRC Staff review Position 15 Range limits - Resolved with the exception of limit values on shape annealing matrix constants Position 18 Software burn-in test - awaiting final NRC Staff review of test data Position 19 Qualification of Software Change Procedure - l Requires fully qualified software change l procedure Position 20 Data links to plant computer - Applicant has just recently agreed to disconnect data links - Re-solvel Position 26 Qualification of optical isolator as an isolation device
- Positions 1, 5 and 12 require startup of the plant to obtain data; therefore, O ener ce##oe de re 1vea "#t11 erter ie eemce or the oeeretics 11ceese-
/9- i 3 ALL J l ..
ESCI/ANO-2 Highlights March 20, 1978 ;
- 2. The NRC Staffs Supplemental Safety Evaluation Report provided for this i meeting reported 24 non CPCS related outstanding issues. Two of these l items were resolved at the' time of this meeting and four additional items l were identified leaving a total of 26 outstanding issues. Both, the NRC Staff and Applicant agreed that all of these items could be re-solved prior to the end of May 1978. Scheduled fuel load date is May 15, 1978. ,
- 3. Topics discussed during this meeting which appeared to trouble the Subcomittee included:
.a. Determination of periodic CPCS test interval determination without a reliability analysis of the CPCG.
O b. NRC Staff requirement to disconnect the CPCS plant computer data links
- c. Large number of Phase II test cases were out-side the initial acceptance criteria
- d. Lack of qualified software change procedure
- e. Noise tests, similar to actual plant noises, should have been used to determine the affect on CPCS operation.
- f. Total number of outst,anding issues may be excessive to get a favorable ACRS report in April 1978.
- 4. The Submittees recomended that the CPCS and ANO-2 be brought to the ACRS for consideration at the April 6-8 ACRS meeting. The Applicant was informed that due to the large number of outstanding issues (10 CPCS and 26 non CPCS) the ACRS might not be able to write a favorable report at the O April ACRS meeting.
h-/ Y
i O esc 1/^ao-2 aeetime sarch 20, 1,2e Documents Provided to the Subcomittee for this Meeting
- 1. Presentation Schedule (Attachment B) ,
- 2. Copies of viewgrap.:s (Attachments 1-33) . A couplete set of handouts is availab1e in the ACRS official copy of these minutes.
- 3. Supplement No. I to the Safety Evaluation Report for the Arkansas Nuclear One Unit 2 plant, dated March 1973.
9 0 O
/7- /[~
i l O V REACTOR DE5!CN COMPARISON l l THERMAL AND HYORAULIC OESIGN PARAMETERS (NCMINAL) ANC-2 Calvert Cliffs 1 & 2 Performance Characteristics Reactor Core Heat Outout, thermal megawatts 2815 2560 Reactor Core Heat Outout, millions of British 9608 8737 thermal units per hour . Systee Pressure, pounds per square inch ansclute 2250 2250 : Minimum DN8A at Nominal Conditions 2.26 2.18 l (full power) ) 1 Coolant Flow , Total Flow Rats, millions of pounds per hour 120.4 128.8 Average Velocity Along Fuel Rods, feet per second 16.4 14.2 Avertge Mass Velocity, millions of pounds per hour 2.60 2.33 per square foot Coolant Temperature, degrees Fahrenheit Nominal Inlet 553.5 543.4 Vessel Outlet 612.0 595.0
. Average in Vessel 582.75 569.2 Nominal Outlet of Hot Channel -
652 642.9 (full power) - Heat Transfet at 100 percent Power Active Heat Transfer, Surface Area, scuare feet 51,000 48,400 Average Heat Flux, British thermal units per hour 182,200 178,000 per square foot O Maximue Heat Flux, British thermal units per hour 425,800 per square foot 527,900 Average linear heat rate of fuel rod 5.34 5.94 only, kilowatts per foot Maximum Clad Temperature degrees Fahrenheit - Clad Surface at Nominal Pressure '57 657 Fuel Tempereture, degrees Fahrenheit *
,. Maxiouur at 100 percens 'cwor 3420 4170 CORE MECHANICAL DESIGN 5ARAMETERS l Fuel Rod Array 16x16 14x14 Number of Fuel Assemblies 177 217 Fuel Rods per Assemoly 224-236 164-176 Fuel Assemblies Overall 7.980 x 7.980 7.980 n.980 Ofaensions, inches Number of Spacer Grics per Asseenly 12 8 Fuel R.ods Number 40,716 36,896 Outsic.e Of aseter, inches 0.382 0.440 Clad Thickness, inches . 0.025 0.026 ,
Clad Material 11rcaloy 4 Zircaloy 4 l Fuel Pellets , Material Sintered Pellets $1ntered Pellets ' Length, inches 0.390 0.650
. /7-- / 6 .
- .N .. ._ a=11
~W
I STATUS OF PRnJECT REVIEW (2). SAFETY EVALUATION REPORT ' j ISSUED - NOVEMBER 11, 1977 , , f!RST ACRS MEETING - FEBRUARY 9, 1978 - SUPPLEMENT No. 1 TO SAFETY EVALUATION . REPORT ISSUED - MARCH 6, 1978
~
SCHEDULED PROSPECTIVE DECISION DATE FOR ISSUANCE OF THE OPERATING LICENSE - JUNE 1, 1978 II ITEMS Resolven SINCE PREPARATION OF SSER No. 1 FUEL ASSEMBLY BURNABLE POISON DESIGN VERIFICATION (4,0)
'CEA SURVEILLANCE' PLAN FOR 2A103 - BqC (4.0) ,
O 111 FACH MFW ITCM SINCF PogpARATION OF SSER 40, 1 CONTAINMENT PURGE VALVE CLOSURE (SECTION 6.0) REGULATORY GUIDE 1.44 (SECTION 5.0) [NVIRONMENTALQUALIFICATIONS. .
-OR ?OLYETHEYLENE CABLES (SECTION 3.11) ,
ECCS PUMP ROOM LEAKAGE (SECTION 3 5.4 4) . , , s . l l O e O AHaL. -l 3
,9-i7
~
O IV- EACH OUTSTANDING ISSUE IDENTIFIFD IN SSFR %.1 (21 iTcMS) SEISMIC QUALIFICATION (SECTION 3,10) ENVIRONMENTAL QUALIFICATIONS (SECTION 3.11) CEA Gu!DE TUBE WEAR (SECTION 4.0) CONTAINMEllT PRESSURE DUE TO MAIN STEAM LINE BREAK MASS AND ENERGY RELEASES (SECTION 6.2) CONTAINMENT LEAKAGE TESTING PROGRAM (SECTION 6.2.6)
' ENVIRONMENTAL QUALIFICATIONS OF SAFETY RELATED INSTRU-MENT TION FOR MAIN STEAM LINE BREAK INSIDE CONTAINMENT (SECTION 6.2.1)
EVALUATION OF EMERGENCY CORE COOLING SYSTEM (SECTIO CONTAINMENT SUMP TESTS (SECTION 6.'3.4) VERIFICATION OF IMPLEMENTATION OF INSTRUMENTATION & CONTROL S',' STEMS DESIGN (SECTION 7.1) ~ ' INPUT FAULT AND SURGE TESTING OF POWER SUPPLIES
~
(SECTION 7.2.2) e EVALUATION OF ADEQUACY OF PARAMETERS ESSENTIAL FOR ACCIDENT AND POST-ACCIDENT MONITORING (SECTION 7.5.1) REDUNDANT VALVE POSITION IllDICATION (SECTION 7.6.3) SEPARATION CRITERIA FOR C0tjDUITS (SECTION 7.9.4) FIRE PROTECTION (SECTION 9.7) O. eEEDWATER N^MMER iN STEAM GENERATORS (SECTION 10.6) PREOPERATIOf1AL TESTS (SECTION 14.0) EMERGENCY' PLAN (SECTION 13.3) 8-[
O
' RCP SEIZURE AtlALYSIS US NG CESEC CODE (SECTION 15.4.2)
- REVIEW OF MAIN STEAM LINE BREAK ANALYSIS (SECTION 15.4.L FINANCIAL QUALIFICATIONS (SECTION 20.0)
GENERIC ISSUES - SPECIFIC N.'O-2 ACTIONS . O. o> REACTOR vESSec Suge0RTS (SECT 10N S:e.S) (2) OVERPRESSURE PROTECTION [ LONG TEV1 (SECTION 5.7)
~
(4) 0FFSITE GRID STABILITY (SECTICN 8.2)
. .9 # g
, ./
e O. [7- t f
9 Y CONCLUSIONS FOR OPERATIffG LICENSF ISSUANCE , (A) TOTAL NUMBER OF ITEMS - NON-CPCS (28) (B) .TWO ITEMS RESOLVED AND ARE CURRENTLY AWAITING PUBLICATION IN AN SSER OR APPLICANT DOCUMENTATION MAJOR ISSUES (C), THROUGH LATE(0) PROJECTED f1AY, 19/8 TO BE UNDER REVIEW CEA GUIDE TUBE WEAR
- CONTAINMENT LEAKAGE TESTING PROGRAM ENVIRONMENTAL QUALIFICATIONS OF SAFETY RELATED INSTRUMENTATION FOR MSLB INSIDE CONTAINMENT
, CONTAINMENT SUMP TESTS .
INPUT FAULT AND SURGE TESTING OF POWER SUPPLIES () FIRE PROTECTION RCP SEIZURE ANALYSIS USING CESEC CODE OFFSITE GRID STABILITY , j (D) f CORE PROTECTIO y ggQULATOR SYSTEM REVIEW STATUE
.. IN THE MARCB_p, IM/o SSER 22 POSITIONS 12 POS'ITIONS RESOLVED 3 POSITIONS RESOLVED FOR FUEL LOAD (REQUIRE STARTUP DATA) 7REMAINOUTSTANDING(THESEINCLUDEShISMIC ,
QUALIFICATIONS AND POSITION 20)
/
o o .
... APPEi? DIX y THE CORE PROTECTION CALCULAT01 hore Protection
-ystem (CFCS) Calculator !
t
/3 1. What --is it? A system for on-line calculation of core power g
distribution and DNDR, and for providing a reactor trip signal when either linear power density or DNER reach selected levels. The system also calculates CEA position, and primary coolant flow rate, and provides a trip signal when flow declines to a selected value.
- 2. How does the protection system M a,"CPC reactor" differ from that in,the reactor without CPC? The CPC system produces 2 trips out of a total of 14 trip functions which can be distinguished in the protection system. The other 12 trip functions are un-changed. The non CPC reactor uses a measurement of VP in the steam generator to indicate primary coolant flow rate. The DNBR trip replaces (in some sense) the thermal margin trip in O i the non CPC systems. It should be noted that each of the trips V~ provided by the CPC has an identified backup so that if the ex-pected trip does not provide the necessary shutdown, a backup trip is available.
3 How does I_t, work? The CPC system makes use of six minicomputers, one in each of four separate channels, to calculate DNBR; and two to calculate CEA position. Using as input CEA position, the readings from 12 (4 sets of threc) ex-core neutron detectors, primary flow rate as calculated from primary pump speed, pressurizer pressure core VT as determined by the difference between measured values of hot leg and cold leg temperatures. The CPC makes virtually real time calculations of core powcr distribution and of DNBR.
- 4. What are the problems? Since this is the first U.S. reactor in which on-line digital computers are to be used as part of the
/f' M /
i a : l i 1 reactor protection system the NRC staff has been concerned about. (a) hardware re1iablIity (b) system independence (both independence'of individual channels one-from-the-other, and , independence of the protection systems and components, from non protection systems and components, (c) and about software validity and reliability. Acceptable methods for testing and maintaining both hardware and software have had to be developed by the vendor, and checked and accepted by the NRC staff. 5 h the new system safer than the old? It may be a standoff. With the new system one should know more about what the core power distribution is than one knew with the old. However, l O the new system is more complicated than the old one. In my view, the additional information available makes the change worthwhile. 4 0 O ,. j7 ~ "
, .i,.,-
' APPENDIX VI InDr=wmdEwF AHC-2: ACRS Consultants' Report g9GONNE NATIONAL LABORATOR'i Tdtplcrdl2/972 4639
.nhCAss Avu%Andc iis' mr lilhois 6009 (4636)
AoyisonY to:#.trTEr. 0:1 FTS 972-4639 (4636) s ItcAtTOP. SAFEtVAnus U.S. N.R.6 MAR 311978 March 29, 1978
, ) 'o '
. . . P.tl ,
l
'ggdgOt t h9ill1 2 01 4:5:0 7
- Advisory Committee on Reactor Safeguards
-, U. S. Nuclear Regulatory Commission Washington, DC 20555 Attention': Mr. G. R. Quittschreiber
Subject:
Core Protection Calculator System (CPCS) for Arkansas j Nuclear One - Unit 2 .
Reference:
Letter, W. C. Lipinski to G. R. Quittschreiber, subject: Revico of CESSAR System BO, dated May 5,1975. ; The above referenced letter discussed CPCS issues which were of concern at the time of writing and was based on the review of non-proprietary
'nformation. Subsequently proprietary information was reviewed and several rcas of concern were resolved. In addition, the NRC staff and its consul-tants have conducted an in-depth review of the CPCS. The documentation resulting from the NRC review has been transmitted to me and reviewed by me.
It is to be noted that the reactor trip system is based on thirteen (13) trip functions. Each trip function is comprised of four redundant and independent protection channels. Only two (2) of the thirteen (13) trip fune- l tions arc derived from the CPCS. These are: (1) High local power density and (2) Low departure from nucleate boiling ratio (DNBR) . The remaining cleven (11) trip functions are derived from hard wired analog systems. If the CPCS were to fail, backup trips will function. In order to assess the degree of protection provided by the backup trips, the NRC staff conducted a review beyond that normally performed for reactor protection systems. The results of this review are documented on pages D-1 through D-3 in Supplement No. 1 to NUREG-0308. The NRC staff concluded that the backup to the CPCS is acceptable. I concur with this finding. During the course of the CPCS review, the NRC staff developed twenty- . seven (27) positions. Of these, seventeen (17) positions are resolved and J closed, and ten (10) positions are still outstanding. This is not unexpected because for several issues operating data is required. The outstanding positions are:
- 1. Uncertainty Associated with the Algorithms. Resolution:
^b- Experimentally qualify adequacy of uncertaintics by
, T1.. t L - g .s.. . . ( T4 .t . . .
/V-23 A.... ,IL& __
l
ricory Committea on Reactor Safeguards
- h 29, 1978 y gs Two .
performing confirmatory measurements during startup to demonstrate the adequacy of the axial power systhesis.
- 4. CEAC Separation Criteria at the Output of the Optical Isolator Cards. Resolution: Contingent on position No. 26,
- 5. Cable Separation. Resolution: Applicant will reevaluate design where safety-related control rod drive position sensor cables are run together with nonsafety cable and will advise NRC staff as to its resolution. Concern is that nonsafety cables will induce noise in safety cables.
- 12. Electrical Noise and Iso.lation Qualification. Resolution:
Noise and DHI readings to be made in plant to verify that the noise spectrum is within the susceptibility envelope used during system test.
- 14. Seismic Qualification. Resolution: NRC provided applicant with current criteria for multi-frequency input and sine beat tests. Submittal date for a satisfactory seismic quali-g-
fication plan and a review completion date to be determined.
\-
- 15. Addressable Constants. Resolution: Software has been rede-signed to reject entry of unreasonable constants by operator and was tested by NRC staff during Phase II test audit.
Resolved for all addressable constants with the exception of limit values on the nine (9) shape annealing matrix constants. l l l
- 10. Burn-In Test. Resolution: Software burn-in test on fully l configured system completed at ANO-2 during February 1978.
A preliminary review indicates no major problems. Final review required for resolution
- 19. Qualification of Softwarc Change Procedures. Resolution: l Qualify software change by either:
A. Final test on plant system (1) Defino a test cor. figuration acceptable to NRC staff. (2) Define an acceptabic test program for cach change or for a pre-defined category-of software changes. or .
,() B. Final test on a single'channci system (1) Qualify the single channel system (2) Defino an acceptable test program for each change or for a pre-defined category of software changes.
$~ LY
# k V isory Committee on Reactor Safeguards l cch 29, 1978 Pege Three Changcc to ANO-2 software will be prohibited until a change procedure has been fully qualified in accordance with position No. 19.
- 20. Data Link to Plant Computer. Resolution: NRC staff will only allow links to plant computer to be connected during initial startup and refueling startups. The applicant must submit procedures and test criteria and methods for NRC review. If this is not acceptable, the NRC will require the applicant to remove the data links.
- 26. Optical Isolator. Resolution: Optical isolator must be qualified as an isolation device by applying 125 volts alternating current or 125 volts direct current at the input and output of the device. These optical isolators are installed between the two (2) CEAC computers and the four (4) CPC computers.
Of the above outstanding positions, three (Nos. 1, 5, and 12) require y reup data / analysis and seven (Nos. 4, 14, 15, 18, 19, 20, and 26) require
;olution prior to plant startup. In addition, the NRC staff requires that No. 26 he resolved prior to the issuance of an operating license.
I disagree with the NRC staff resolution of position on No. 20 Data Link to Plant Computer. The NRC staff bases its position on GDC 24 and IEEE 279-1971, Section 4.7. CDC 24 - " Separation of protection and control systems. The protection system j shall be separated from control systems to the extent that failure of any single l control system component or channel, or failure or removal from service of any ' single protection system component or channel which is common to the control i and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Inter-connection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired." IEEE 279-1971 - "Section 4.7 Control and Protection System Interaction. 4.7.1 Classification of Equipment. Any equipment that is used for both protective and control functions shall be classified as part of the pro-tection system and shall meet all the requirements of this document. 4.7.2 Isolation Devices. The transmission of signals from protection system equipment for control system use shall be through isolation devices which shall be classified as part of the protection system and shall meet O all the requirements of this document. No credibic failure at the output (O of an isolation device shall prevent the associated protection system l channel from meeting the minimum performance requirements specified in I the design oases.
~k
! s . . o Q 'ivisory Committee on Reactor Safeguards reh 29, 1978
. age Four l l
Examples of credible failures include short circuits, open circuits, grounds, and the application of the maximum credible ac or de potential. A failure in an isolation device is evaluated in the same manner as a failure of other equipment in the protective system. 4.7.3 Singic Random Failure. Where a single random failure can cause a control system action that results in a generating station condition requiring protective action and can also prevent proper action of a protec-tion system channel designed to protect against the condition, the remaining redundant protection channels shall be capable of providing the protective action even when degraded by a second random failure. Provisions shall be included so that this requirement can still be met if a enannel is bypassed or removed from service for test or maintenance purposes. Acceptable provisions include reuucing the required coincidence, defeating the control signals taken from the redundant channels, or initi-ating a protective action from the bypassed channel. 4.7.4 Multiple Failures Resulting from a Credible Single Event. Where a credible single event can cause a control system action that results in a condition requiring protective action and can concurrently prevent the pro-tective action from those protection system channels designated to provide principal protection against the condition, one of the following must be met. 4.7.4.1 Alternate channels, not subject to failure resulting from the same single event, shall be provided to limit the consequences of this event to a value specified by the design bases. In the selection of alter-nate channels, consideration should be given to (1) channels that sense a set of variables different from the principal channels, (2) channels that use equipment different from that of the principal channels to sense the same variabic, and (3) channels that sense a set of variables different from those of the principal protection channels using equipment different from that of the principal protection channels. Both the principal and alternate protection channels shall meet all the requirements of this document. 4.7.4.2 Equipment, not subject to failure caused by the same credible single event, shall be provided to detect the event and limit the conse-qucnces to a value specified by the design bases. Such equipment shall meet all the requirements of this document." CDC 24 and IEEE 279 do not forbid the connection of protection and control equipment. (In this case, the term control is used in the broad sense where the operator is used to close the loop between the information he receives from the plant computer and the actions he may take in operating the plant.)
/7 A description of the Core Operating Limit Supervisory System (COLSS),
c Core Monitoring Computer, and the Plant Computer appears on pages 9, 10, and 11 in the above reterenced letter.
/Y-RG
*e I'/h
(_ visory Committee on Reactor Safeguards rch 29, 1978 . l Page Five ! 1 1 4 During the NRC staff presentation on March 23, 1978, the following verc offered as reasons to support the NRC staff isolatlon position:
- 1. The plant computer is not safety grade, therefore it should not perfo:m safety functions.
- 2. The CPCS has to have additional programs to generate the output data to be transmitted to the plant computer.
- 3. The plant computer will send interrupt signals to the CPCS to start data transmission from the CPCS to the plant computer.
I would like to comment on each of the above points:
- 1. If the CPCS were not digitally implemented, the plant computer would still be used to perform the same calculations. Plant computers are used in all other nuclear power plants to supply the operator with 1 information on plant status. This information, coupled with operator judgement, is used to operate the plant. The plant computers in all
(~)
\/
cases supplement and do not replace plant protection equipment. If 4 the NRC staf f has developed a new position on the use of plant com- l puters in general, this position should be better stated and added to ! the General Design Criteria if the position has merit.
- 2. It is true that the CPCS digital program has been expanded to provide i for the data transmission to the plant computer. As to whether this l added feature has compromised the CPCS can only be determined by exam-ining overall CPCS reliability. More discussion of reliability is presented later under discussion of Position No. 8.
- 3. The feature by which the plant computer sends an interrupt signal to the CPCS to start data transmission does place the CPCS as a slave to the plant computer rather than the reverse. There is a solution to this problem in which the CPCS would seed an interrupt signal to the plant computer to tell it that data is to be transmitted. The appli-cant should be given the opportunity to discuss whether slaving the plant computer to the CPCS is acceptable. l The NRC staff, in taking its final position in allowing the CPCS to be connected to the plant computer during initial startup and refueling starteps, is inconsistent for the following reasons:
- 1. The CPCS softwarc will not be changed. The same softwarc that is used during startups will be used during power operation. Any concern the g) . staff may have with respect to a reduction in system reliability
\_/ because of sof tware complexity reemins unchanged.
847
- . .\ ._ . Lu . . .
OAdvisoryCommitteeonReactorSafeguards reh 29, 1978 ge Six , j
- 2. The plant computer will continue to execute algorithms and display information to the operator. The operator will use this information to run the plant. The plant computer will not receive data from the CPCS and the algorithms based on this data will not be operational.
It is my recommendation that the applicant be allowed to connect the CPCS to the plant computer with properly qualified isolation devices and that the plant computer be sloved to the CPCS for data transmission. The NRC staff in Position No. 8. Time Interval of Periodic Testing, has required that the test interval should be significantly more frequent chara the. proposed 30 days during the first six months of operation and that the applicant develop an acceptable analysis of the CPCS reliability in accordance with applicabic IEEE standards. Based on the test data acquired during the first six months and the reliability analysis, the test interval can then be s modified. The NRC staff has not provided the applicant with guidance on a relia-bility goal. If the test interval is to be modified, it must be done on the basis of meeting a specified requirement. It is rec.ommended that the NRC staff dev
>elop a reliability requ1rement for the CPCS and provide this information to O PP11c =t- .
The Phase II Test and Test Report are covered under staff position No. 24. Of the 36 static test cases, 16 were outside of expected DNBR range and 6 were outside of expected LPD range, but based on addi,tional analysis and testing were found to be acceptable by the NRC staff. Ten of the 26 dynamic test cases did not meet acceptance criteria for " time to trip" but explanations are acceptable to NRC staff. Satisfactory final performance of the CPCS is determined by testing. l It is imperative that the system pass all static and dynamic tests without explanations as to why a particular test was not passed. A proper simulation ! of the reactor should be used such that the test results are not dependent on ; a poor simulation, and explanations therefore have to be used to qualify a cost ) as acceptabic. Furthermore, the dynamic tests have been based on variation of a single input parameter witn time and with all other inputs held constant. From previous static and dynamic test cases, coding errors of a fixed point multiplication overflow and a floating point multiplication underflow were detected. It must be clearly demonstrated that similar coding crrors still do not exist for the case of variation of multipic input parameters with time. Alternately, it must.be shown that if the system is still not properly scaled, that trips will occur sooner because of improper coding or scaling. Sincerely, O L%k W Walter C. Lipincki WCL/at Senior Electrical Engineer cc: ~Dr William Kerr.
/fM I . . .
]
. ~ _ . _.. ._ ._ . _ _. -- . . . _ . . . . _ _ _ _ _ _ _ _ .
o
% W 1:
f . s' - . s?
. - 7 ELBERT P. EPLER g ,,, 8(?[,3Y . NUCLf.AR StiTEMS CONSULTANT g g g,, ,
712 FLORIDA AYENUE '"'* Calc luDGE, T]ESSEE 37830 ' .
.2. 1978 3-30-/'I78 @s, n, , . .,. ,. ,.A #, 3 s, .,, ,.., , ,1 '
wm ked; ' Cholen1an , Elec /rie a l Sy s lem s <>nd D C L l>com"n IIee BJ lhe t'lareh 2 o '3v b ccmnwflec iv,eeliii<;, held {c rewbu lhe c ec s k/ nWo un; / .2, o q ue.3l ion' avose of pe(iodiiconceminc;[cr}le//k les/il2<j o- -lhe lo<c Pro /ec-llbn lc/ /heColc [ve7venctj v /n Stjs*fel?2. /i?OJmv c l} O3 lhi) ecj ut,ongc'nl /s {il'sl c l c% l< mc), iI hns been ma<>le a ve,nuiremend /ha /, c)ven*:
.lh e 50'3 h S ] Y lo 10 n ~l]23 0hOjc]GYOlldi1,//,,cj2VVICr.>]il h eS Y b e 3 lcf r)i A O r)-ll n1 dye fr'C'q' u en l ll) un /he cu3 om a f L' 0 6 r,.lo l J ,, Wf do ho t uere/, have l/1IOfH)Gblbn re/nY YV bo 'beS [fe7Jen(y Whdh 8063 '
11ol Of;)eo/ /f) hh e l'he/n lUfe . ln U)etu a-l kh e concerns <>xpressed, & cuoubt he tuor/hwAtle La j)r('321)l l1e fele VO n h ( O f.t a.
/> sa,,ilov s;/vaJan exisle<d a J orwt w hei., a netu 1sjJYe/n lo/ prolec llo11 05 lh e N f IR. w n .5 clweloped, making use oi solm .s lade ronyone>>]s -
' in } <n IculodianoI -/cchniques, for cohs h no P
i
. z O
rehabiliI asda o I once pe/ eighJ hove shif-l exisled. p lesJ(<eyvene7hed wa., ihilibil e. slo bli , hase a very pessanislic esliinale of componen/ lativ<e tales. This les) fre<juency posed no probJem inosmuc h. as an on-hhe sijsdem enabled each channel k he lesled, from .sen.sov lo aclua-/w, b y pas hine a .s/nc; /e hvHon nl Jhe opevalo/s' tonsole, 77w Jesl,,by c a vsl a loca) perlu/halion el.lhe process ravnbble, vevilks opevol>;hly allhe channel b y m/errupliny /se cun eni in one of Ihe 4h<ee macynelcoils Jo/ ecich rod, ne enlire 3;.slem,. con, pas /A 7 21 chav>n cis , con he Jeslen M o boud Jen minodes, The once.p e>< shill lve9 uence
- 0. sca s /n lev' c hang ed to onc e g ev cda .
Du>ny. -lh e n ' y ea rs o ( op era libn , J his ho s res o iden in /co, coo <ho m lesIs being ferlormed. - 1 Three fa; lures hnve occurved in n c,eous as Jolla tu s. ;
- a. Rs a resul) of a. desicjn error /he nev-Ivc, flux nmplefiers cuere /cund J-o I, e incapnb lo o f p roleclih 7 ocjonu-l a -lvansionI ha vnic ce very sha// pericci. when .sunila / eguipmc cva s appUer) to eu Sas/ />vrs} reneJo/, ?J wn.s O .
S-30 .
O dis c o vered lla> / J h e p o w e / bv <s) o ve r lo v oled and: des lroyed Ihe lieId eflec/ Jransis/ovs aJ Jhe ihyvl l-o /he amp liIii>rs, fo/ op,wo n;nn lek, dwo <7 ears Jhe reo c1c/ had been vnp,oJecled \ nejonu) a desiqi, hosis evenv ,l 77as condi//on was l nol disc c>reveol h., pen brlic les4s.
- b. The Jhree jornz o-libri c hnin h ers a l /he Fnv/lt
\
Fuel Eleinen
-leslihq of a relic /f voJ,ysdem 'were wel-lt-d lve, Th e e<j u ip men wos c)u < i dvd d and relurned lo servic e , when Jhe ren c lo/ vu n.s broug hJ l-o juowe ond lhe periodN 4es) jwvlo/mec), ll wa s dbec> reved -lho) one csJ Jhe lhvec cha mloers re mai>1 ed in ep er'n b le.
C
/)n ene,in ee/, in Jeo u b le s hoodi kq , inoc)/edo ,r]I
/ef.) n c lip lead in plac e, The lnop ern ble chonnel wos di> c overed b y p en'adi< 4es].
This a buns e of componen) Enilure conic as a su < prise, as il wo uld b e e xp e< lcd Jh.> / duri,,c; lh e.s e l'z tj enn , sereeni da2.en componen } to:/u es tuovId have been cdiscovevec). }} ?5 aIvibuIe Ju Jl> c e xtelleirl <,volil of sve veilla nr e a nd prevend{re noitu-lenance }hn), a llho vejh conf an en/s hn we been 53/
~
replaced .n.s n resul) of degraded pc>rlorniance, Jiiere
'unve heen bul Jhese Jwo ins la n c.es of fo livi c Jo respa n,/. - -O The cons eq ueor e of fniIv,c of a prole <Ja e leo hee depend s upon Jhc fre9vene al u hic h Ihe sysicin >s '
chnIleng ed, and on lhe nun >be/ ol d>>erse den Ju r es ' wilhin t h e s q.s d em y whnh n, c a votIn ble Jo meel each c h o llenij ei Duvlnc; Ihe / 2 y e o <.s of op evalion J h e, e hn re b e en bd Jhvee occasions on w hie. h
- a. .s c va rn be c a n i e ir e c e.ss a , :j . l These weve ecch brouqhI n bou) bu, on in e e e a.s c , ;., :
dempwo-lvre of Jhe coolcind a ] Jhe renckr'in/ed. ' h, o ' c c h c a se n mo r>1e>> la re dij, in voldo g e co v.s c><>) o n e J Jhe lh<ee mn]n ci<culallmq 7, umps k J,ip , ivilh Q pun,p remahuhq Ji,e (/o w w a s redu c ed /c go yL, The con 4rol sy.slem duh {ull reduced Jhe veock/ powe/
+o 5070 la ma rvinin lhe j) cop e/ //ux - lo - flo <u rolh, hotvere/ c< h en one pump -/vipp ed ils heol er cha ng e/
becaine una vn1la l> le , A.s a resu ll. w~,11, s oys //ow and t, '? fo h e o ) ex t h a nc, e / cap a c il , -/h e r c a /a n) -lemp vo /. rose /c lhe sc va an j> oind. n/Jhouqh Jhere /> ore been Jhree cho//enges lo l -lh e z oleI remperolure p<oIeck<e lealu> c, i ha i, n<ovides ll,e s ole pro.lec bbn . i. e., ./here is no d&er.s e : O
,/7'31
Loch up / colure, Jhe<c An re been n a to mpo n en J ilures. The fec9 vent y of periodic. les) hns /herefore Onoi heen n e v1lic a l mo Hev, lla d lhe c ha llenges b e en, no) o n in c ven se s 'n leinpeen Jure, buJ a s lo tu in c <co se in renclo/ potvev, lhe even) vs auld ha ve been even less . serious bu & wouId h a ye been seen /> 3 Jhree diver.se leo Juves, 4he Wev Iro n Flu x, lh e Th erm o / Po uv ev (now n or) n nd
/ he In lol mr>perrrlvre, There h a s however, b eeni no Inc reaso in rec e los po tv e v la c ho llen<7 e / h e sujslein not bo re }herc been a ns; roiryon end follvres.
i A failur e al Jhe I?ca c /c / ,L o cu fressu/c len /v> e QuouId he mac h eno ve .s erio u s . Foifure lo scrain on loss o f j>< css u <e due la n leo k, wo uId 'resu /J in Jissibn p< oduc l.s b eing dis chn eged ihlo lhe reo c lo/ huildliq, D ere is n o dl<ers e f ro-le rl/bn Sor' lhi s even), ho tucVe/ -lhere ho,s been ne;lher a ch n }}eng e nor ct
< c>in()ov e irl ln ? lv r f-In es In blishin7 n de sd frer, vc>irre; 10< Jhis sysleni rl enn b e s een /h ol ,, e> l h er co mp o n en d /n ilv ec n w -fec7ven n; c-( clin /len7e has been ex si gnific a n ) -fo c lov, The inilioI 4 es) -l*re7ve ncy a/ o nee pe< shiil wa 5 ' lev' cha ng ed lo o n e c p e/ do c\ . Th e sp lem co n lin v es t
if- 33 .
- ~. , . - .
1 l to bc lesled fregueiclh Leenu.se lhe a bel;L k O .s ee J h e s ujsdern respon;d k n c ha ng e in /he vn nn ble is n. <n lun ble nid k suvvetJin n c e. nic r._oquidd les) fre7uency , however; i) once p e< n,<c lif e, i. e, 23 do y s, I-( lor nny reas on il should beconic snpossiisle c>< vndesien l>le k )es], -lhe ren tlov' c uo u/d noI haye k be > ), v./ do wn unn er es> n y;ly a .s cu i resol J od <m u n een l, >lh rey vir en> en J 4 < p enod,*e l
-lesJi,9 - _
The IIFIR ha3 &enion>l<n-led Jho] tomf anen] In)Ivre ha.s noJ been a p, o blen1, a nd Jhn ) l SurveiIfnnee, nol p eviodic (es/, hn's />een 1he ingai Jonl l on-l<ihulcr' k relin b> lllc; , 7~he freq venc e of Je3l l 0 co n J herclo<e he based c'n/Nel y on co>,venienc e. This tan also be den >ons/roJed k De deve k ,' l comm ercin I veo clo vs, Ie 0 LV0 f e *} c'))r l'S , s ere n
/b)f.) f V Y' L}Fu/,
a s ever e Jra n.si en J on a r c o <<n n e e o J n. ,Lu, I>in o -), : , p .
<u 1 Th e reo < lor' fo wer' co n in e een s e a s mvch o.s Ja c Jo, o f Jen in less lhon a second, and to;lv -e lo .sc va in wo uld .r uvel resv14 in m e /Ji sq . T/> eis c 3 c'r en c h n lle og c.s en > > be us el c'n lhe / /h e JVe Ji oix
)>., d F(v v o r Ih e diter sc Pr cssu re p<o ec li< e lenlv<cs. l This is in Sl>ory con lrosi la J/se Hr/Il whal, ha 5 ;
O n,e,a c,,c ca .ti,,c e nn is 4,~,c,n.i <e e w,o cu n .Itool J c O ,n veni eid -les ) i;, Je <n i o f s o . an p i s o d c,7 un de .
The ku o - a 1 - fou/ Jogie ceinglo y ed i,, e iva s , a nd oI cvw end in lete.s/ lo Jhis dis cus.ubn, pro uides er dep)h of de len.s e anoins/ rondon, componen J fo r Ivre . Three failures wo uld be i e,r ulved to pr ev e).J prolc e dWe a c //on . 7tu s inn h e 1/lu s /<n ded a s ,L A ws, I eJ > -I b e a s.s um ed -/h a / eveli c honnel foil., one e p er u;e n /, In a J-tv o - o l- fov/ sejs /em a n d n 3 a clo JesJ in}erval, Ihis wov /d y ie ld n n un n < a iln bil;l e,J
/, ( 6 >' t o ' lor a <jiten prodeclive fro Ivre . The lu. bin e
' vip -lvo n sien-l I s , ho w e <e r; .s ens ed h, ho / h Alev d> un Fivs and Pr es.su r e 7i n'n q a com bin ed un rela b;):) ol 1 3 Y t o '. The 7 fran s i e n /s pe/ y e o / 'ws v /d yield et to,Ivec ro le oJ io r whah vuo vid be o u ep do b ie -
The a ssumer) fo rivre en de of once ,> w ye o / Jar ea c h c h n noe I cu o ul,l, ),0 cv ere/, he en /Veh g unn a ej>/a h/c , e<ea kw hv a s e ):old Igle lu e, o nyp Jin n e e r, l= <e n .s o , t v , ] h ~ lw a - of fou/ sysJem ond n <on venien/ /c.sJ inder voi o 1 30 da s;.s , Jion hiq hl, un ,ehn'hle hn <duvore evo vid be ndequale /c mee] even .seven c lio lle n 7 e.1 p e< y ear, 7 tic n hor e rep <esen)s e spe<ien c e wd h <o n <en li:ns,,I
' o rd suoec, av hn h doe.s nol- n eces.s a </13 o,y19 tu Jhe cecs, lh 3Y
we en,, .see .sereenI diflavances Eh e h m,yb / o flecJ hoc
,iIv<e en Je oJ J h e new .sy.sl<w ,
Il has been dencon5/<nded Jhn / w;Ih h/gh <,unlil, >
someo iln n t e , ano /o <j hnrdwn /c fativres en,> b e - dele < /cd, of everu- a ndic o'pnl col- // is lo b'e <>
cs mil r)elecl )he vos) nro) ort These hvo fo c lo rS In l< en J-cy elh e/' s h o vtol lea ve re/nlir,cI few to,np on en / fo r /v <e.s k, he sedec/ed h 3 lhe j'criodic -le.s$ Q Who ) Is c>rert more nApo fl0 nl1 is Ihe ede9 re e ol clire/sil te laici, p 1 0 c e.s a }l>ni l o n Jhe ro n.s e 9 v en c c ol [n l lv te o f n. 9 i/en pro.Iecl ire len/vtr. The Ic //o rv ,;>. 15 q u o f erl / r e cn lhe .sen,o-2 " Li,v :le d Ev e ) d o ,ms c, e f.5 17 0 Y (Oll$ / / e rc'N i,8 ) s e l* O SIcfni(ilOn Sn e' L / 0 ) ) f Pr <1 * **c* The s lo ff is eonsiote<ing (n,'Ivce of Jnc c)<g ; /o / -frty s., slen, lo p e</onn ils eles/ n fun c J &n . wl, 'le o u / <evsv 7 is incoinglede o r>J a wn]lr'19 .subinifln ) o f seine rc7 ve.s /cd anni;se.s irom Jhe opp licon i N nypenrs Jhol ol Jhn slo<je o/ out review ] ha] b o e l~p o nn log } oip.; avilj'o/ i,, n e , e ,, / 'shuldown m ee hn n's s in s . vuo uIr} h*>vll Jh e tons eg u ene e.s al Jhis Is,pe 01 << lo t Iv e r lo pi er en d undu e ris h. ' /s J he publi [e l ". } oil h a nd sn /-l~ 3 L IJ uovid appeuJhs) w,4h Jhe <e d en J io n oj han) 1 e 9 vip m eni iha I eren. nuo, e I h n er Jhe 0 'd n usua l d&e<>i no loc;l exis h w;JI, /pe resu /J Jho ) con,pon e,,/ ledlvre in :thn sysJem, a s in lh e pa s), wovid b e o f j rninor con ceviv The s /n(/ posilibn se7v;n ;,q n ,, i,u , e n s e in le, J frec,v ency w Jh e i;nlin t s >x mon /Ja , o pp en <s k be i sou n d <=<en on in c ren s e o f a lo cle o / Lw a , in ~ Ju.o - o f - Covi se s)ein , wo vid ' reduc e /n1/u, c palsob:)il 5 s
- a. [a c k< o f e,igi, J .-
'7hese Jo c kr s , e sp e c'in ii Ih e malle/ a 1 d,; ers e Q JecJiun, o <e peculicu lo h e A No s ys dem . we en rt expec J in Jhe Culure to s ee di gila i sys dems sinnd>;g n io n e , i. < , w J hov) b e;> >gt w c)< ed u,, b. An,;I w :,,cd s.plen a.
ha lhalense lh e tons e <f uer> < e a l <smp; on en! Coliv< e be ver"; nwc I t g ren le V', we en n , ho w e re,', eyp e c l lc leo nt muc I [< c> n g Jhis d ila l A Wo t
in >Jo llnli6n s o n .s lo be bel l0y j), e,)o v ed,i,t lhe l-v]v a e, lo ern /v a le )h e < ole o f 1,erad,t. ie s l-We sho uld b e mim]fu ( J hn -l in Jhe pnsl, e amp onenl
-fo tiv ,'c h a s never been a. s/ 7nila <> n d con /> ils vlo>' lo p<olevlan syslen, snilu/c , ond tho.1 p erio sa Jes-Is 9~>< h, mea lo, Ihe ,,v,,,ose o( He)erJay ccinpo,,en J
/7~ 3 7 .
an-aa .-s -nay.eu.u aao m. = a+-,-s,s n -wa ea, m -- r. --- a.xm.,> p - -s.s ~
,.e O fo, . n e mn.(Jt,erclor e he perlwnwd al toiwen> en-l inde<vols . tue wc,vid he />of h so,p As ed a nd dius,, pain Jer) il -lhe diqilo I ego ip rn end we< e L> p r ove
,Jo b e a o pron e la compo n eir ) in o'Iv /c J ha J j> cdo d i les) n'>q uiovM n e ed L lie ye,'lo, nved nron />-c7 ve>< // \
$ ? fj h b r 4
O ,
1 l
4 0
e R- 3 r
nFFEiiDIX VII s
Al .0- 2 : Core Protection Calculater
( ) Sy:; ten CORE PROTECTION CALCULATOR SYSTEM 4
- 1. DESIGN AND OVERVIEW
- 2. FUNCTIONAL DESIGN
- 3. POWER DISTRIBUTIOWDNB METHODOLOGY / UNCERTAINTIES
- 4. HARDWARE / SOFTWARE DESIGN e
4 l
g- 3 7 9
i O .
b O j THE ANO-2 PLANT PROTECTION SYSTEM (PPS) IS COMPOSED OF TWO
! SUB-SYSTEMS:
- 1. AN ENGINEERED SAFETY FEATURES ACTUATION SYSTEM (ESFAS), AND
- 2. A REACTOR PROTECTION SYSTEM (RPS) -
- ~
THE CORE PROTECTION CALCULATOR INITIATES TWO 0F THE TEN TRIPS IN THE REACTOR PROTECTION SYSTEM, THE LOW DNBR TRIP AND THE HIGH t LOCAL POWER DENSITY TRIP.
~
PLANT PROTECTION SYSTEM h RPS CPC ANALOG TRIPS ESFAS I i i 2 DIGITAL 8 ANALOG TRIP 7 ANALOG SAFETY TRIP FUNCTIONS FUNCTIONS .
FUNCTIONS .
O .
i 1
DESIGN CRITERIA AND OVERVIEW
- l. DESIGN CRITERIA
- 2. DESIGN APPROACH
- 3. CONVERSION TO DIGITAL SYSTEM l
O 4. DESIGN FEATURES x
V
O ,
CRITERION NUMBER 10 (10CFR:50, APPENDIX A) l "THE REACTOR CORE.AND ASSOCIATED COOLANT, CONTROL, AND l
PROTECTION SYSTEMS SHALL BE DESIGNED WITH APPROPRIATE O MARGIN TO ASSURE THAT SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS ARE NOT EXCEEDED DURING ANY CONDITION OF NORMAL OPERATION, INCLUDING THE EFFECTS OF ANTICIPATED OPERA-T10NAL OCCURRENCES."
O g- V2-
O DEFINITTON OF ANTICIPATED OPERATIONAL OCCURRENCES
" ANTICIPATED OPERATIONAL OCCURRENCES MEAN THOSE CON-DITIONS OF NORMAL OPERATION WHICH ARE EXPECTED TO OCCUR Q ONE OR MORE TIMES DURING THE LIFE OF THE NUCLEAR POWER UNIT..."
\
QUOTE FROM 10CFR:50, APPENDIX A 4
I l
O:
O O O l
SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS
- 1. LHR CORRESPONDING TO CENTERLINE MELT
~2' , DNBR .
~
EQUAL TO 1.3 (W-3 CORRELATION) x X
! O O o_ i CRITERION NUMBER 20 T10CFR50, APPENDI'X.A): ,
"THE PROTECTION SYSTEM SH ALL' BE DESIGNED '1) TO INITI ATE AUTOM ATIC ALLY THE OPER ATION OF APPROPRI ATE l
SYSTEMS INCLUDING THE REACTIVITY CONTROL SYSTEMS,
% TO ASSURE THAT SPECIFIED ACCEPTABLE FUEL DESIGN .
LIMITS ARE NOT EXCEEDED AS A RESULT OF ANTICIP ATED
~
i l
OPER ATION AL OCCURRENCES AND 2) TO SENSE ACCIDENT CONDITIONS AND TO INITIATE THE OPERATION OF SYSTEMS AND COMPONENTS IMPORTANT TO S AFETY."
O O O.
~
~ ~ ~ l CRITERION NUMBER 25 ,
Il0CFR50,. AP PEN DIX A) '
~
"THE PROTECTION SYSTEM SH ALL BE DESIGNED T0.: -
~
' kSSURE THAT SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS ARE NOT EXCEEDED FOR SINGLE MALFUNCTION
\.
D ..
OF THE REACTIVITY CONTROL SYSTEMS,, SUCH . AS ACCIDENTI AL WITHDR AWAL (NOT EJECTION OR DROP-OUT) 0F CONTROL R'0 DS ."
l O O O l
OVERALL REQUIREMENT I
l THE NSSS DESIGN AND TECHNICAL SPECIFICATIONS WHICH GOVERN ITS OPERATION ARE SUCH THAT:
- 1. THE SPECIFIED ACCEPTABLE FUEL DESIGN LIMITS (e.e., DNBR = 1.3)
AND OTHER SAFETY LIMITS ARE NOT VIDIATED AS A CONSEQUENCE OF ,
ANY ANTICIPATED OPERATIONAL OCCURRENCE (e.c., A R0D DROP), AND
- 2. THE CONSEQUENCES OF ANY OTHER POSTULATED ACCIDENT (E.G., STEAM
~
GENERATOR TUBE RUPTURE) WILL BE ACCEPTABLE, k
k PROVIDED THAT N 1. ACTUAL PLANT CONDITIONS ARE WITHIN THE LIMITING CONDITIONS FOR OPERATION, AND
- 2. ACTUAL SAFETY SYSTEM SETPOIflTS ARE EQUAL TO OR CONSERVATIVE RELATIVE TO LIMITIrlG SAFETY SYSTEM SETTINGS, AND
- 3. EQUIPMENT OTHER THAN THAT CAUSING OR DEGRADED BY THE OCCURRENCE OR
~
ACCIDENT OPERATES AS DESIGNED, INCLUDING ALLOWANCE FOR DESIGN MALFUNCTIONS SUCH AS A STUCK R0D OR OTHER SINGLE FAILURE.
.O'
)
i THE MARGIN REQUIRED BY CRITERION 10 IS DESIGNED INTO THE NSSS; HOWEVER THE' REACTOR OPERATOR MUST OPERATE THE PLANT SUCH THAT THIS MARGIN IS MAINTAINED.
1 ALLOWED OPERATION IS DEFINED BY TECHNICAL SPECIFICATION
(]) LIMITING CONDITIONS FOR OPERATIONS (LCO),
COLSS,dDIGITALMONITORINGSYSTEM,AIDESTHEOPERATOR I IN MAINTAINING SOME OF THESE LCO's.
\
1
. l j7- YY ,
O -
l LOFA (ANTICIPATED OPERATIONAL OCCURRENCE)
. l LSSS: LOW DNBR TRIP (CPC) l LCO : CORE OPERATING LIMIT ON THERMAL MARGIN WITHDRAWN ROD WORTH SCRAM DELAY TIMES l O R0D DROP TIME
- MAINTAINED WITH HELP OF COLSS
\
S f
O .
S~ f
'O . O O CORE PROTECTION CALCULATORS
- (CPC)
THE CORE PROTECTION CALCULATORS ARE DESIGNED TO PROVIDE THE FOLLOWING PROTECTIVE FUNCTIONS:
A. INITIATE AUTOMATIC PROTECTIVE ACTION SUCH
. (f THAT THE SPECIFIED FUEL DESIGN LIMITS ON c DNBR AND LOCAL POWER DENSITY ARE NOT-EXCEEDED DURING SELECTED ANTICIPATED OPERATIONAL OCCURRENCES, AND B. INITIATE AUTOMATIC PROTECTIVE ACTIOF DURING CERTAIN ACCIDENT CONDITIONS TO AID THE
~
ENGINEERED SAFETY FEATURES SYSTEM IN LIMITING THE CONSEQUENCES OF SELECTED ACCIDENTS. 1
O EVOLUTION OF LICENSING CRITERIA l
NRC INTERPRETATION OF CRITERIA At1D INDUSTRY KNOWLEDGE AND VIEWS IN REACTOR PROTECTION DEVELOPED AND CHANGED IN THE EARLY 1970's.
- 1. IN GENERAL SINGLE FAILURES OF AN ACTIVE COMPONENT SHOULD BE C0tlSIDERED AS A POSSIBLE INITIATING O MECHAtllSM FOR AN A00.
R0D MISOPERATION EVENTS
,- SINGLE R0D WITHDRAWAL
- OUT OF SEQUENCE INSERTION AND WITHDRAWAL
- 2. IN MOST CASES, OPERATOR ACTION SHOULD NOT BE RELIED UPON TO PREVENT THE SPECIFIED ACCEPTABLE FUEL DE-SIGN LIMITS FROM BEING EXCEEDED.
- AXIAL FLUX PERTURBATIONS O ,
jf- s/
BASED ON THESE CONSIDERATIONS AND THE POTENTIAL IMPACT OF THE RESTRICTIONS IN TERMS OF OPERATION, IT WAS CONCLUDED THAT THE PROTECTIVE SYSTEM MUSTi !
i
- 1. SENSE THE POWER DISTRIBUTION WITH INCREASED l ACCURACY.
)
O 2. INCLUDE MEASURED CONTROL R0D POSITION AS INPUT.
- 3. PROVIDE INCREASED ACCURACY IN DNBR THERMAL MARGIN BY ,0N-LINE INTERPRETATION OF RELEVANT COOLANT SYSTEM PARAMETERS.
TO EFFECTIVELY IMPLEMENT THE ABOVE REQUIREMENTS, DIGITAL PROCESSING TECHNOLOGY WAS INCORPORATED INTO THE PPS.
4 4
O g-sa
O O O CPC ADVANTAGES RELATIVE TO ANALOG COUNTERPART
- 1. IMPROVES PLANT SAFETY MORE DIRECT MEASURE DE FUEL DESIGN LIMITS REDUCES RELIANCE ON OPERATOR ACTION x 2. IMPROVES PLANT PERFORMANCE M PROVIDES MORE ACCURATE MEASURE OF FUEL DESIGN LIMITS PERMITS PLANT PARAMETERS TO BE TRADED OFF AGAINST ONE ANOTHER U(
A SUCH THAT MARGIN TO FUEL DESIGN LIMITS IS UNCHANGED
- 3. IMPROVES PLANT FLEXIBILITY SIMPLIFIES TASK OF ACCOMMODATING CHANGING CONDITIONS DURING PLANT LIFE
O
[PC DESIGN FEATURES AN.0N-LINE PROTECTT00 SYSTEM USING 3 LEVELS OF EX-CORE DETECTOR INFORMATION A COMPLETE SYSTEM 0F DEDICATED DIGITAL CALCULATORS TO PROVIDE FOUR CHANNEL REDUNDANCY USES AN AXIAL / RADIAL SYNTHESIS TO CONSTRUCT POWER DIS-TRIBUTI0tlS USES MEASURED CEA POSITION INPUT FLOW DETERMINATION BASED ON RCP SPEED MEASUREMENTS LINEAR HEAT RATE AND DNBR CALCULATED ON-LINE OPERATOR'S CONSOLE PROVIDES COMPREHENSIVE DATA DISPLAY O
l
/h 3 .
l l
W 6 /// Cf 9
AFPEiiDIX VIII
, F.. .v- 2 : Core Frctection Calculater Systan Functional Desicn OBJECTIVE TO PROVIDE A FUNCTIONAL DESCRIPTION OF THE CPC/CEAC .
SYSTEM O
l l
l I
O
- g. sr
(2) .
DUTLINE PART 2 FUNCTIONAL DESCRIPTION
- 1. RELATIONSHIP OF CPC.IO REMAINDER OF RPS-
- 2. CPC DESIGN BASES EVENTS
- 3. SYSTEM INPUTS AND OUTPUTS
- 4. FUNCTIONAL BLOCK DIAGRAM
() 5. ALGORITHMS PART 3 METHODS AND UNCERTIANTIES
- 1. POWER DISTRIBUTION METHODS
- 2. DNB METHODS
- 3. TREATMENT OF UNCERTAINTIES l
O g-sc
O i
REACTOR PROTECTION SYSTEM THE ANO-2 REACTOR PROTECTION SYSTEM IS COMPOSED OF TEN TRIP FUNCTIONS. THESE INCLUDE:
A. EIGHT ANALOG TRIP FUNCTIONS CONSITING OF SINGLE VARIABLES (E.G. PRESSURE) WHICH ARE COMPARED TO O TRIP SETPOINTS, AND B. TWO DIGITAL TRIP FUNCTIONS CONSISTING OF MULTI VARIABLES WHICH ARE PROCESSED BY DIGITAL COMPUTERS AND COMPARED TO TRIP SETPOINTS.
1 i
REACTOR PROTECTION SYSTEM DIGITAL ANALOG TRIP TRIP ,
FUNCTIONS FUNCTIONS 2 DIGITAL J8 At[dLOG TRIP TRIP FUNCTIONS FUNCTIONS O
l R-s7
ANO-2 REACTOR PROTECTION SYSTEM TRIPS INPUTS: ' TRIPS: i 2
FROM SENSOR AND SIGNAL PROCESSING .
< BISTABLE NUCLEAR FLUX- '
COMPARATOR > HIGH LINEAR POWER '
POWER FROM LEVEL TRIP EX-CORE DETECTORS
, BISTABl.E # HIGH LOGARITHMIC COMPARATOR POWER LEVEL TRIP
+
~
CEA POSITIONS > Dh I TRI COLD LEG '
DIGITAL TEMPERATURE
> GALCULATIONS HOT LEG ~~ ~ ~ ~ ~
TEMPERATURE > LOW DNBR TRIP
~
REACT'OR C00MNT )
PUMP SPEED + l BISTABLE ,"HIGH PRESSURI7.ER
+ COMPARATOR ' PRESSURE TRIP REACTOR COOLANT PRESSURE FROM . BISTABLE ,_ LOW PRESSURIZER
> ' PRESSURE TRIP PRESSURIZER COMPARATOR hf!hhffR0p ^ m BISTABLE > PRESSURE LOW STEAM TRIP GENERATOR STEAM GENERATORS COMPARATOR BISTABLE ,,, LOW STEAM GENERATOR I FEEDWATER LEVEL IN STEAM GENERATORS _, COMPARATOR WATER LEVEL TRIP i
61 STABLE ,_ HIGH STEAM GENERATOR
+ COMPARATOR ' WATER LEVEL TRIP !
O REH W COM M MENr c0RienfTia
>MSui1TtP fya 5'P -
O-1 CEC THE CORE PROTECTION CALCULATORS'ARE DESIGNED TO PROVIDE THE FOLLOWING PROTECTIVE FUNCTIONS:
A. ' INITIATE AUTOMATIC PROTECTIVE ACTION SUCH THAT THE SPECIFIED FUEL DESIGN LIMITS ON DNBR AND LOCAL POWER DENSITY ARE NOT EXCEEDED .DURING ANTICIPATED OPERATIONAL OCCURRENCES, AND B. INITIATE AUTOMATIC PROTECTIVE ACTION DURING CERTAIN ACCIDENT C0flDITIONS TO AID THE ENGIEERED SAFETY FEATURES SYSTEM Iii LIMITING THE CONSEQUEiiCES OF THE ACCIDENTS.
O
. R-cr
O CPC DESIGN BASES EVENTS MAJOR ANTICIPATED OPERATIONAL OCCURRENCES
- 1. UNCONTROLLED AXIAL XENON OSCILLATIONS
- 2. CEA RELATED EVENTS, INCLUDING SINGLE R0D WITHDRAWAL, SINGLE DROPPED ROD, SUB-GROUP DEVIATION AND OUT-0F-SEQUENQEWITHDRAWALANDINSERTION 3.. EXCESS LOAD LOSS OF LOAD 4
Lt .
- 5. L SS,0F FORCED REACTOR COOLANT FLOW O 6. UNCONTROLLED BORON DILUTION POSTULATED ACCIDENTS
- 1. STEAM GENERATOR TUBE RUPTURE
- 2. REACTOR COOLANT PUMP SHAFT SEIZURE l
i l
1 O
R-4o -
O '
BACK-UP TRIP FUNCTIONS FOR CPC ANTICIPATED OPERATIONAL OCCURRENCES FAILURE OF CPC WITH CONCURRENT A00 IS NOT A DESIGN BASES EVENT FOR THE ANO-2 PLANT.
AN EVALUATION BASED ON THE CENPD-158 ATWS REPORT, WAS PER-FORMED TO DETERMINE BACK-UP TRIP FUNCTIONS.
1 RESULT:
EVENT BACK-UP TRIP l
UNCONTROLLED CEA WITHDRAWAL FROM A CRITICAL CONDITION llIGH PRESSURIZER PRCSSURC UNCONTROLLED BORON DILUTION HIGH PRESSURIZER PRESSURE TOTAL AND PARTIAL LOSS OF REACTOR COOLANT FORCED FLOW HIGH PRESSURIZER PRESSURE EXCESS HEAT REMOVAL DUE TO SECONDARY SYSTEM MALFUNCTION LOW STEAM GENERATOR WATER LEVEL STEAM GENERATOR TUBE RUPTURE LOW PRESSURIZER PRESSURE CEA MISOPERATION MANUAL TRIP O
l l G /
([)
CPC MONITORED PLANT VARIABLES MONITORED VARIABLE NUMBER OF SENSORS.PER CHANNEL RCP ROTATIONAL SPEED A (1,PER PUMP)
COLD LEG TEMPERATURE 2 HOT LEG TEMPERATURE 2 PRIMARY PRESSURE 1 EX-CORE DETECTOR FLUX 3 DETECTORS, IN AXIAL STACK CEA POSITION 20 (1 PER CEASUBGROUP:)
O CEAC MONITORED PLANT VARIABLES MONITORED VARIABLE NUMBER OF SENSORS PER CHANNEL.
CEA POSITION 81 (1 PER CEA)
O g- c x
O O O 1 CORE PROTECTION CALCULATOR SYSTEM J
~
CEA CEA CALCUl.ATOR CALCULATOR NO.) , NO.2
. I I
~
DEV1ATION ~ DEV1 ATION 4 ISOLATED 4 ISOLATED N . DATA LINKS- DATA LINKS M -
u.
p -
I bu .
Vk Ik kV kY CORE CORE CORE
' CORE PROTECTION PROTECTION PROTECTION
~
PROTECTION CALCULATOR CALCULATOR CALCULATOR A B CALCULATOR C D a -
- f. --
a -
a .
V V 97 'P OPERATORS OPERATORS MODULE OPERATORS OPERATORS MODULE MODULE -
,', MODULE
~
~
O.
O O .
~ ~
CPC FUNCTIONAL BLOCK DfABRAM CORE POWER ,
CALCULATION DNBR -
^ #
N > CALCULATION - .
, LOW DNBR .,
o ~
TRIP SIGN'AL FLOW K CALCULATION
~
HIGH LOCAL t -
~'
POWER DENSITY g-CALCULATION HIGH LPD POWER ?,
TRIP SIGNAL
' DISTRIBUTION -
CALCULATION .
~
O O O.
~
CPC FLOW C.ALCULATION ,
PUMP SPEED s TO TRIP' . .
RPM 1 > ' LOGIC .
TRIP AND ~
RPM 2 > -
PUMP -
RPM 3 > -
DEPENDENT s PART LOOP ,
RPM 4 > ' CONSTANTS -
LOGIC ,
FLbWCALIBR.ATION 00NSTANT .
T C
s PRIMARY COOLANT s ORMLIED ES .
( TO DNB ,
SPECIFIC VOLUMES I i FLOW RATE > CALCULATION D H T '
AND LOOP s
' V RESISTANCE is n v ix ,
RPH1 RPM 2 RPM 3 RPM 4 m
3 O -
I
O .
O O CPC POWER: CAL,CULATION FLUX POWER -
. FLUX POWER CALCULATION ,
- . . :ALIBRATION CONSTANT
' SHAPE ANNEALING. ,
01 - TEMPEPATURE .s AND R00 SHADOWING OT SHA00 MING O
TC s/
(
s CORRECTION
) y D2 CORRECTIONS ' .
03 ) i, fs .
'Q I
CEA POSITIONS
- D TO DNB Y Y CALCULATI0t O . HIGH
-> SELECT -TO LOCAL PC
-> DENSITY CALCULATIO!
l0RE THERMAL P S THERMAL POWER 20WER CALCULATION "
~
sf CALIBRATION CONSTANT ,
CORE ENTHALPY
.s THEPML T - RISE C -AH s POWER- s DYNAMIC T > CALCULATION
' CALCULATION V ADJUSTMENT-,.
H T
H
O "
i CORE PROTECTION CALCULATOR SYSTEMS ANO-2 DESIGN 5 C-E 3410 MWT CLASS PLANTS 18 C-E SYSTEM 80 CLASS PLANTS ,
THE CPC DESIGNS FOR THESE PLINTS ARE IDENTICAL TO THE ANO-2 1 DESIGN EXCEPT FOR DIFFERENCES DUE TO !
- 1. NUMBER OF CONTROL RODS 1
- 2. PLANT SPECIFIC DATA BASE CONSTAilTS-
. PLANT SPECIFIC HARDWARE QUALIFICATION CRITERIA i O 4. ADVANCEMENTS IN METHODOLOGY TO IMPROVE PLANT PERFORMANCE s
O R- (- 7 . - .
O O O -l CPC ALGORITHMS AND UNCERTAINTIES -
_ m. s TOPICS POWER DISTRIBUTION
, LOCAL POWER DENSITY k' -
DNBR >
F D
UNCERTAINTY ASSESSMENT
[ .
t 0
DETAILS 3%
3.3 CPC UNCERTAINTY TOPICAL REPORT $5 -
gR (CENPD - 170 AND SUPPLEMENT l~-P) l 3-d 3-
?
CPC CALCULATION OF CORE POWER DISTRIBUTION O CEA POSITION NORMALIZED EX-CORE INFORMATION DETECTOR RESPONSE
'I SHAPE ANNEALING CORRECTION -
PERIPHERAL POWER INTEGRALS if R0D SHADOWING y CORRECTION IN-CORE POWER INTEGRALS if 1f RADIAL PEAKING FACTOR AXIAL SHAPE TABLE LOOKUP SYNTHESIS F(Z)
R FZ (Z) 1r ,r HOT PIN POWER DISTRIBUTION SYNTHESIS HP(Z) = FR (Z)
- FZ(Z)
V if HOT PIN POWER 3-D DISTRIBUTION PEAK R'll 1 _ .
O O O LOCAL POWER DENSITY CALCULATION VARIABLE INPUT FIXED INPUT CORE POWER ALGORITHM CONSTANTS 3-D PEAK , TILT MAGNITUDE
' UNCERTAINTY FACTOR u o D LOCAL POWER DENSITY CALCULATION O
MAXIMUM LOCAL POWER DENSITY
o .,
O ,
0 DNBR CALCULATIO.1 l VARIABLE INPUT FIXED INPUT
. SYSTEM PRESSURE CHANNEL GEOMETRY l CORE INLET TEMPERATURE CORE POWER ENGINEERING FACTORS POWER DISTRIBUTION l[; COOLANT FLOW RATE
+
ALGORITHM CONSTANTS
, t x CPCTH
- 1. EQUIVALENT POWER & MASS VELOCITY N
- 2. SUBCHANNEL MIXING FACTORS i
- 3. ENTHALPY AND QUALITY 4.DNBR
~
!. < r MINIMUM DNBR -
O O O CPC UNCERTAINTY ASSESSMENT REACTOR SIMULATION CORE CONDITIONS EX-CORE SIGNALS
,7 CEA POSITION
& 4 DESIGN . CPC Q \
CALCULATIONS ALGORITHMS
" ACTUAL" "SYNTHESIZEli N Fg AND DNBR F g AND DNBR 4 4 COMPARIS0N TO DETERMINE ACCURACY
\/
CPC UNCERTAINTY FACTOR ON F g ANo DNBR
-----n--
9 J~
A l U _
e
\ s
. \. ,
2
, N.. '
4 W
o
\s"wL
'+ a O
La
~
ce m Of.
O 4M E
=
CQ
- z-V 9: = Y. j A 25 O v- - .a O
e CC Lu a M
- C~ U
.4 S.
h b
M. E l s = 1 m
em I
L s
. ~
i i
\
i t i I l l l l I \5o k k. b.
m m N '8 O aunaaaztsaa19As.,
\
/-7 'l 3
O O O
~
~
~~~~iCDPC FQ COMPARISON
~
;.1 ,
(MOC) .VM
- ' 4 . ,; .
'l 4.000 l
~
+
i+
+
D 3.000 -. Y o .
+
++.
z %
m . .
- 2. 000 '
.f-*
9 I
- N
\ ~
I I I 1.000 -
1.000 2. 000 3.000 4.000 t n4j ACTUAL' F0
a- ,
APPEl;CIX X AI.U- 2 : CPCS Hardware and Software Design l
.StrLIFIED PIACTOR PROTECTION CHANNEL BLOCK DIAGRAM CHANNEL A PROCESS INPUT SIGNALS CEA CALCULATOR 1 PENALTY FACTOR
- HOT LEG TEMP's CEA CALCULATOR 2
- COLD LEG TEMP,s PENALTY FACTOR'
- EX-CORE POWERS
- PRESSURIZER PRESSURE
- REACTOR CGOLANT PUMP SPEEDS
- CEA POSITION y y y CORE PROTECTION TYPICAL FOR FOUR CALCOLATOR CHANNELS
'A' J LOW DN3R HIGH LPD o
G a:A = L A SINGLE VARIABLE REACTOR TRI,P SIGNALS
<! il 1 I
CHANNELS CRUNEL A B,C & D REACTOR PROTECTION IDENTICAL TO SYSTEM A
! U t j O TRiPSiamLS TO REACTOR TRIP CIRCUIT BREAXERS
, / ~~)
- - . . _. . u. .
.n. me. . s .,e _. w. ~
O
$1MPLIFIED CEA CALCULATOR BLOCK DIAGRAM CEA POSITION l CEA POSITION INPUT SIGNALS INPUT SIGNALS A A r 3 r 3
< >1 r 1, q, ir y ir <r tr CEA -
~ l -
~
CEA CALCULATOR CALCULATOR g
1
.. __ _ 1 -. __ .
_ _ _.] DISPLAY GENERATOR L _" _ _ _.
I O
o BAR CHART
. CRT DISPLAY CEAC 1 .. .. CEAC 2 PENALTY .
- PENALTY FACTOR FACTOR T T CPC (TYPICAL)
NOTE: * = OPTICAL ISOLATORS TYPICAL 6 PLACES TRIP SIGNALS TO REACTOR ,
PROTECTION '
. LOGIC g&G.
_ ___....___..____:_..._..._==
O CORE PROTECTI0tl CALCULATOR SYSTEM DISPLAY AIG INDICATION OPERATORS MODULE ,
SYSTEM STATUS INDICATION OPERATOR DISPLAY OF SETPOINT AND CALCULATED VARIABLES OPEPATING BYPASS CONTROL AND lt0ICATION KEYLOCK ADMINSTRATIVE CONTROL.FOR CALCULATOR SECURITY ADDRESSABLE CONSTANT ENTRY FOR CALIBRATION
, CPC ANALOG INDICATORS DEDICATED ANALOG METERS FOR O DNBR MARGIN TO TRIP SETPOINT LPD MARGIH TO TRIP SETP0ldT CAllBRATED HEUTRON FLUX POWER ALARM ANNUNCLATORS STATION ANNUNCIATORS ARE PROVIDED T0 INDICATE TRIP STATUS AND OPERABILITY OF THE CPC SYSTEM ANAtOG PROCESS INDICATION DEDICATED ANALOG METERS DISPLAY EACH CPC SENSOR INPUT VALU CEA POSITION AND REACTOR COOLANT PUMP SPEED CEA30SITION DISPLAY A CRTDISPLAYS THE POSITION OF ALL 81 CONTROL ELEMENT ASSEMBLIE THE DISPLAY IS SWITCH SELECTABLE TO EITHER OF TWO REDUNDANT SIG CHANNELS, R- 9 7
O -
O_ .
ol t
U
~
11 0 l .
l ~1 ~l l l .
XIO'
- I I '
U U I I L .I 3 0 I
~
I J - ' * ~
VALUE D .
POINT ID ,
CEAC CPC OFF ON ON OFF OFT ' ON -
OO.
OO OO O O.
b 7 8- 9 -
p g g .
j
@p -
3 3.g. -
- CALCULATOR TRIP ktEMORY FUNCTCN SELECT BYPASS PROTECT KEYS
- NJ I 2 3 xto".
7 +
Et4TER '
* ~~ '
O . -
DISPLAY 4 i CHANGE CAi4CEL EXECUTE i VALUE
( ^ ( ( CEAC j CEAC CPC DISP CEACs CEA CPC ,
SENS SENS TEST [
FAIL FAIL FAIL Fall, TEST INOP DEV TEST e
?
v Operator's Hodule .
-g '
,_ _ , _ - . _ _ ~~
qc 42 m gagig3g3g3ll3 ll ag 33 :3gl eagiegi ie l 333:g3g 3ggg3 gg333 ',3ll:a3ig giagiIll I I I W 150 ............................ ................... .....'..........**.*.... 150 140 ,
140 ,
130 130 120 120 110 110 i
100 joo ,
90 go :
(
80 80 70 t 70 i x
N 80 eo 8
V 50 i 50 M 40 i
g . 40 h
30 4 30
, J 3 to ,
10 l 0
0 2
t a l i l l i l a l l i l l i t i l l i l l l'I l l l l l i l l 1888888111118150188348888818111181888888118191113 1-+ - =-II-Ile- Il~1H il- IlW1.-II- CW .
1 2 3 4 5 6 PL-1 PL-2 SHUTDOWN CEA POSITION DISP L AY Floure 7 1
' .~
., ,1
~
~
O. t CORE PROTECTION CALCULATOR SYSTEM TESTING AUTOMTIC ON LINE TESTING EACH CALCULATOR IN THE CPC SYSTEM PROVIDES A RAPID SELF DIAGNOSTIC CAPABILITY TO ASSURE A FAIL SAFE RESPONSE TO DETECTED HARDWARE.
FAILURES. .
POWER Fall AUTOMATIC CHANNEL TRIP 011 LOSS OF INPUT POWER MACHINE AUTOMATIC CHANNEL TRIP ON INTERNAL CALCULATOR MAL-MALFUNCTION FUNCTION PUT/0UTPUT AUTOMATIC CHANllEL TRIP ON DEFECTED MALFUNCTION I
. BSYSTF11 LOSS OF POWER -
NO RESPONSE CAllBRATION VOLTAGE CHECK MEMORY AUTOMATIC CHANNEL TRIP ON PARITY INDICATES HARDilARE FAILURE l CHECKSUM INDICATES FAILURE OR CHA' IGE OF MEMORY CONTENTS SENSOR RANGE AUTOMATIC AtlNUNCIATION ON SENSOR FAILING HIGH OR LOW WATCHD0G TIMER AUTOMATIC CHANNEL TRIP AND LATCH IF THE COMPUTER FAILS REQUIRES MANUAL RESET CALCULATION AUTOMATIC CHANNEL TRIP IF THE RANGE OF A CALCULATION REAS0!! ABILITY IS EXCEEDED O
/9- po .
O -
PERIODIC TESTING SURVEILLAf1CE -
ALARMS, INDICATORS AND OPERATORS MODULES PROVIDE TIMELY INDICATION OF THE STATUS AND OPERABILITY OF THE CPC SYSTEM OFF LINE TEST .
A COMPREHENSIVE TEST CAPABILITY IS PROVIDEDTO ALLOW THE OPERATOR TO CHECK THE HARDWARE AND SOFTWARE OPERABILITY OF THE CPC SYSTEM, THE TEST IS MANUALLY INITIATED WITH AUTOMATIC TEST ROUTINES AHD A HARDCOPY PRINTOUT OF TEST RESULTS O SIGNAL INJECTION TEST THE CAPABILITY FOR I!1 JECT 10N OF " LIVE" PROCESS SIGNALS IS PROVIDED TO ALLOW PERIODIC VERIFICATION OF THE COMPLETE SIGNAL PATH WITHIN THE CPC SYSTEM ISOLATION TEST PERIODIC VERIFICATION OF THE OPTICAL ISOLATORS AND CEA POSITION ANALOG ISOLATORS CAPABILITY FOR ISOLATION IS PROVIDED l
Q QUALIFICATION PROGRNi THE QUALIFICATION PROGRAM IS DESIGNED TO DEMONSTRATE THAT THE CPC SYSTEM WILL PERFORM ITS REQUIRED FUNCTION CONSISTENT WITH THE DESIGN BASES OF THE NUCLEAR POWER GENERATING STATION.
HARDWARE A COMPREHENSIVE PROGRAM 0F TEST AND ANALYSES H.AS BEEN PERFORMED TO DEMONSTRATE THAT THE HARDWARE IS CAPABL.E OF PERFORMING ITS REQUIRED FUNCTloriS.
O siniRoanEnT SEISMIC TESTING AND ANALYSIS TEMPERATURE / HUMIDITY TEST ELECTRO-l%GNETIC NOISE TESTS DESIGN FEATURES IS01.AT10N VERIFICATION TESTING ACCURACY / DRIFT TESTING l DESIGN SPECIFICATION TESTS RELIABILITY 5 MONTH FACTORY BURN IN TEST / ANALYSIS SITE BURN IN TEST ,
O R- Px
. . . . - ~ . . .. . . . -
. __ . . . . __ = 2. .
/ ;
~
O .
4 SOFTWARE '
SIMILAR TO THE HARDWARE, THE SOFTWARE FOR THE CPC SYSTEM HAS UNDERGONE EXTENSIVE TEST AND ANALYSIS TO DEMONSTRATE ITS ADEQUACY.
PHASE I Q EACH MODULAR ELEMENT.OF THE SOFTWARE WAS TESTED TO ASSURE THAT IT CORRECTLY REFLECTED THE DESIGN REQUIREMENTS INPUT SWEEP THE INTEGRATED SYSTEM WAS THOUROUGHLY TESTED OVER THE ENTIRE PARA RANGE OF REQUIRED SYSTEM OPEPATION TO ASSURE CORRECTNESS OF IMP TATION AND TO DETERMINE THE UNCERTAINTY COMPONENT DUE TO THE DIGIT l
COMPUTER CALCULATIONS.
O
O .
4 ANO-2 INTEGRATED SYSTEM QUALIFICATI0tUFIELD TEST PHASE 11 TESTS THE ANO-2 CPC SYSTEM WAS THOROUGHLY EXERCISED UTILIZING " LIVE" SIGNALS DRIVEN FROM A SPECIAL PURPOSE SIMULATOR. RESULTS WERE COMPARED TO 0FF LINE PREDICTIONS OF CPC PERFORMANCE.
O PRE-0PERATIONAL TESTING THE CPC SYSTEM WAS INSTALLED IN THE FIELD AND THE OPERABILITY OF THE SYSTEM WAS DEMONSTRATED.
I SITE BURN-IN TEST THE ANO-2 CPC SYSTEM WAS TESTED IN THE FIELD WITH A STATIC SIMULATOR TO VERIFY THE SYSTEMS PERFORMANCE AND OPERABILITY UNDER FIELD CONDITIONS.
O
. 8- 7,V
O O '"" " ' '
' ' O' '
ACRS MEETING .
AGENDA CORE PROTECTION CALCULATOR SYSTEM ARKANSAS NUCLEAR ONE UNIT 2 '
g DOCKET 50-368
~
APRIL 6,1978 g
- o. &
SAFETY OVERVIEW n
REVIEW CRITERIA [g REVIEW METHODOLOGY IS en REVIEW STATUS ~
i 5
u a. mm .n -,a- - , . . -
m a _z,,.. a a - - pes,s e--- - . a--
O 4
9 0
2 SART/ OET/IS/ ,
O .
l J
i f
i I
l O
-R- 6 "'M'- - - ---.3 ,,_
O O 6 '
PROTECTION CHANNEL FUNCTIONAL COMPONENTS SENSORS ,
INITIATION DEVICES CPCS BISTABLES l LOGIC MATRICES r
ACTUATION DEVICES v
ACTUATED DEVICE
O POSillON O FOSITION O'
TRANSMITTER TRANSMITTER O OPitCAL ISO 4 AIOft QOlllEftSENSORS ( T
[>lSOL AllON AMPilflEll h CATilODEll AY IUBE
'r 'r 20 CE AS 20 CE AS 20CEAS PO T ON 20 CE AS POSIT ON INDICATOR INDICATOR
(" POSI (" POSI 41CEAS 41CEAS
' ' CRT 'r CEA P CEA (p CALCULATOR CALCULATOR q N
q g 1tBl =q QC P .
2tC) yy" yyy" A B C D
, ,A B C D 1r PC CPC CPC CPC C.. A~ *B" *"C" ~~ D "
>- + <- <-
O .. .. . ..
O .. .. ..
O .. .. ..
O
~~ ~~ ~~ ~
TTT TTT TTT CftWP ALARMS .
DN!!R TillP LPD Title CPS HARDWARE CONFIGURATION BLOCK DIAGRAM
O O O CHARACTERISTICS TRIPS 2 Digital Lo DNBR Hi LPD 12 Analog Hi Flux Hi Press Etc. .
s DIGITAL N
Sensor Signals ,
- Q A/D Conversion of Continuous Signals Discrete Logic Execution Protection Algorithms Real Time Computer Output - Logic Matrix -
Communication - Information Readout Periodic Test & Surveillance
~
O O O IS IT FUNCTIONALLY ADEQUATE?
) WILL IT OPERATE WHEN NEEDED?
CAN IT BE RELI ABLY MAINTAINED? ,
5
_ _L . __ - =
-_______.__=______-____________-_-_a_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
O -
O e
e PE/IB4 CRITERIA O
O
- /9-7/
~
O REVIEW CRITERIA 1
-A9JWA9E G JC Reg. Guices nc ustrv Stanc arcs SO =~~WA 9 E G JC O Reg. Guices ncustry Stancarcs Surveys
\ uc ear
-a cen 3roject
<WU 3 1oenix S93 S-Wa re Structu re Qua ity Contro 0_ - .
xx aerience
/-7~ 9%
4 , , _ a ,a ,_, w _~---- ---- -- ---,-,-,-,,-, --,--,-------- _ ,----- - ----- _ _ - - , . , - ---- - - - - - - - - - - - - - - - - -
P I
O l
6
*a mme O ,
l O I
O O O REVIEW METHODOLOGY STAFF l CPB Physics AB T-H <
Design Basis
% \
ICSB EE N CONSULTANTS R
S-Ware Eng EXECUTION Review Plan ~
Task Force Meetings Audits Working Meetings
- , . . i '
O N
O I
T CE p E R i r _
T U T p OL l e
v i
T r
R I e
P FA i p
i p L r
e r
u _
r r e s E T T t s e
D P C e r e r WP a r _
I u u r r O VC e s
s s s
e t o
a t
o O F P r
P r r e
a r
e R O e r r e n n P z z e e G G T
i i r r H N u s u mma s s s a C E I
e r er t e t e
P H V E h g
P w
S w
S w
o o o W N i
H L L L S I e e e e P
I R
T _
O t
q4k -
!llllN
O O . O t
w FIRST BACK-UP TRIP FOR EACH EVENT e CEA Withdrawal - High Pressurizer Pressure (CPC Not 1st Trip) e Boron Dilution - High Pressurizer Pressure -
^
g o LOF - High Pressurizer Pressure
!No e Excess Heat Removal - Low Steam Generator Water Level gs (CPC Not 1st Trip) e Steam Generator Tube Rupture - Low Pressurizer Pressure (1600 PSI A) e CEA Misoperation - Manual (COLSS Alarms and Control Board Indication) .
O O O MULTIPURPOSE INPUT /
CEA ACQUISITION CENTRAL PilOCESS
> OUTPUT > PROCESS CONTHOL CONTHOL j k SYSTEM UN'T SOFTWARE INTERFACE PLANT
' ~
COMPUTER CEAC 3
CPC ll OPERATORS '
MODULE l
SOFTWARE
\ INTERFACE T EMPEll ATUHE CENTRAL PROCESS )f PROCESS PLANT MULTIPURPOSE 2r PRO-INPUT /
ACQUISITION TRON
" OUTPUT bb CONTROL CONTROL HEACTOft '
COOLANT = ?-
PUMP SPEED CHANNEL B CEAC/CPC
O O O FUNCTIONS AND DOCUMENTS ASSOCIATED WITH REDESIGN AND REQUALIFICATION OF STORED COMPUTER PROGRAMS Functional Test
* ' a opment =
Hettuitements I' CEN 44( Al-P CEN S3(Al-P CEN 67( A).P CEN 6S( Al-P CPC Functional Description CPC/CEAC Data Base CPC/CEAC Program Assembly Phase i Test Audit N and Supplement 1(P) and Supplement 1(P) Listing CEN68(Al-P Supplement 2(P) Supplement 2(P) Phase il Test Audit I Supplement 3(P) CEN S7( Al-P CEN72(Al-P
% CEN-4b( A) P CEAC Functional Desciiption CPC Sof tware Specification and Supplement 1(P) ihase i Test Heport CEN73( A) P CEN SU( Al-P Pliase il Test Heport CE AC Soltware Specification CEN SS A Core Protection Calculator 1%ase il Test Procedure Integrated System Burn-in and Supplement 1(P) Test Proceduse CEN 69(A).P CPC/CE AC Executive System Sof tware Specification
l i
l TYPICAL HIGHLIGHTS QUALITY ASSURANCE PLAN (16)
QUALIFICATION OF SOFTWARE CHANGE PROCEDURE (19)
PHASE ll TEST AND TEST REPORT (24)
M L
BURN IN TEST OF SYSTEM (18)
DATA LINK TO PLANT COMPUTER (20)
OPTICAL ISOLATOR QUALIFICATION (26)
O O O TEST AUDITS HARDWARE BURN-IN TEST PHASEITEST PH ASE ll TEST -
.s PROCESS PROTECTIVE CABINET 2 THER VIAL TEST OPTICAL ISOLATOR QUALIFICATION TEST EMI NOISE IMMUNITY TEST .
lO l
l l
l FE/IEd STATUS
~
O l
l l
l O
l
/Y- / b I
o o o REVIEW STATUS Positions Defined 27 l%
Positions Reviewed 21 y and Closed i Positions Outstancing 6
O O ~O
SUMMARY
- Positions Outstanding 6 Start-Up Data / Analysis 3 (1,5,12)
Resolution Required Prior to 1 (26) License . f P License Condition 2 (14,19? W Plus Detailed Start-Up Procedures Start-Up Test Audit Start-Up Test Report Technical Specifications
L Q O O w DATA LINKS TO PLANT COMPUTER' ,
) BASES G DC - 24 o ADDED DESIGN COMPLEXITY % 0 ADVERSE FUNCTIONAL FEEDBACK D o DATA COLLECTION FOR DESIGN BASES g ,h ANALYSES EVALUATION h ee RESOLUTION in E E5 o FOUR CHANNELS CONNECTED DURING [5 INITI AL STARTUP AND REFUELING STARTUPS E -
o DISCONNECTED DURING OPERATION E 1
O O O GDC 24 FACTORS IMPACTING SAFETY
- ADDED DESIGN COMPLEXITY o INCREASE PROBABILITY OF DESIGN ERROR e COMMON MODE FAILURE ,
4 o ELECTRICAL FAILURE PROPAGATION FROM NON-lE INTO IE
$ o ADVERSE FUNCTIONAL FEEDBACK o ANALOG SYSTEMS.
- SIMPLE PARAMETERS SUBJECT TO OPERATOR JUDGEMENT ,
oCPCS
- COMPLEX MULTIVARI ABLE PARAMETERS NOT
, F ASil.Y FVAl llATFD BY OPFR ATOR
O O O
~
m
. GDC 24 R EQUI R EM ENTS
- ELECTRICAL ISOLATION O INDEPENDENCE
~
x
- REDUNDANCY 7 .
; o SINGLE FAILURE D
- SAFETY IS NOT SIGNIFICANTLY IMPAIRED o RELIABILITY AND COMMON MODE FAILURE
, o ADVERSE FUNCTIONAL FEEDBACK e ADVANTAGES
~
o o o I PROTECTION AMD 2 CONTRO L INTERACTION
- 1. Hardwirec Between Protection and Automatic Control System .
$ 11. Set Point / Calibration o" Protection System Using Operating System Ill. Incorporating Additional \ on-Safety Design Features into t7e Protection System. .__-.__.__m . _ _ _ . _ . . _ _ _ . _ _ _ _. . _ . , _ . _ . _ _ _ __ _ . _ _ . _ . _ _ _ _ _ _ _ _ _ _
APPEl! DIX XIII , Af;0-2: Status of Project Review l i n l V . I
- 1. STATUS OF PROJECT REVIEW )
-FSARDOCKETEDlilAPRIL1974
- SAFETY EVALUATION REPCRT (SER) ISSUED ON NOVEfGER 11,1977
-SUPPLEIGitTONETOSERISSUED0;!f%RCH6,1973 ,
- ACRS ELECT:tICAL SYSTEl'.3, CO:4 TROL AND INSTRUf',ENTATIO:1 St&-
CO:Til EE "EETli:GS ON T!? CPCS WERE HELD CN l'AY 20,1977, JUNE ]3v,15/7 NlD 1%RCH 20, l378
- ACRS ANO-2 SUBC0l441TTEE fiEETIIF3S WERE HELD ON JUNE 24,197/ ;
A!!D FEBRUARY 2,1970 ) l
- ACRS f2ETING II;CLUDI:!G THE CORE PROTECTION CALCULATOR SYSTEI4 l WAS HELD ON FEBRUARY 9, 137C
- ESTifMTED DATE OF CCMPLETION OF ALL f TATTERS It! PARTS II THROUGH VI Ill SU? PORT OF ISLUANCE OF Rl GPERATIN3 LICENSE -
JUNE 1973 O l l l
/9-/d ?
l l 's O l 1I. 1TEl"3 RESOLVED SIflCE FEBRUARY 9,1978 ACRS MEETIf!G l A Sb?PLElEllT TO THE SER HAS fl0T YET BEEi! PREPARED FOR THE FOLLO'.!!NG ITEt'3. HChiEVER THE CO:?iUNICATIONS BETr:EEN THE STAFF l AND THE APPLICNiT IN RECENT WEEKS INDICATE THAT THESE ISSUES 1 l HAVE BEEll RESOLVED N!D A SUPPLEi2NT TO THE SER REPORTING OlJ3 EVALUATION WILL BE PREPARED If! THE I: EAR FUTU.iE. l l FUEL ASSEl'BLY EUP.RADLE POISON DESIGil VERIFICATION (4.0) 1 CEA SLOVEILL/d4CE PLAN 23FCR A1C 0 - c4 (4'0) CONTAll: MENT PRES $URE DUE TO f%IN STEAMLINE BREAK IMSS AND , ENERGY RELEASES (0.2) . EVALUATION OF EMERGENCY CORE CCOLING SYSTEM PERFORfMNCE (6.3) EVALUATICN OF ADECU/CY OF PARN2TERS ESSEiff!AL FOR ACCIDENT NIDPOSTACCIDENTIONITORiiiG(7.5.1) FINANCIAL CUALIFICATidNS (20.0) O c0NTain:EuT Sur's res>S (s.s.4) i I R- /o f
liI. [1Eh' ITES3_.S.IILCE FEBRUARY UZO ACRS 11EETIfJG c0:rrAlfRErf PURGE VALVE CLOSURE (6.0) l l REGULATORY GUIDE 1.!!!! (5.0) ECCS PUPP ROOM LEAKAGE (15.11.6)
\
1 l O 4 O R-uo
o V IV, M2LETE. LISTING OF NiO-?JsEVIEW ISSjlES (28 ISSUES 1 l SEVEN Of_T1DE ARE RE3QLYER
** THREE OF WESE NAYLfEEN IDEi!TIEJED SINCE FEBRUARY 1.197[3 SEISMIC CUALIFICATION (3.10)
ENVIRONi'EitTAL QuiLIFICATIONS (3.11) I 1
- FUEL ASSEMBLY BURt!ABLE POISON DESIGtt '5RICATION (4,0) j CEA SURVEILUiNCE PLAN FOR A1 0 -ngC (4,C) l 23 CEA GUIDE TU3E WEAR (4,0) l
" REGULATORY GUIDE 1.44 (5,0)
** CONTAINi'ENT PURGE VALVE CLOSURE (6,0) !
CONTAINMENT PURESSURE DUC TO M'ilil STEN 1 LINE EREAK t%SS g ANDENERGYRELEASES!.$27 V CONTAINMENT LEAL %CE TEST!!!G PROGRN1 (6.2.6) ) ENVIRON:Et!TALQUALIFICATIONSOFSAFETYRELATED,1(ISI6U:'.ENTATION FOR l%IN STEN 4 LINE EREAK lilSIDE CCNTAll' F ENT (o.m,1) EVALUATION OF EtERGENCY CORE COOLING SYSTE'M FERFORiuNCE (6.3)
- CONTAlllMENT SUMP TESTS (6.3,4)
VERIFICAT10:10F IMPLETENTAT10:10F II 5TRUMENTAT10N 8 1 CONTROL SYSTEMS DESIGN (7.1) INPlK FAULT NID SURGE TESTING OF POWER SUPPLIES (7.2.2)
. EVALUATIO 10F ADEQU.\CY OF PARNETERS ESSENTIAL FOR ACCIDENT AND POST-ACCIDE .T IDNITORING (7,5,1)
REDUNDNIT VALVE POSITICM INDICATION (7.3,3) SEPARAT10:1 CRITERI A FOR CONDUlTS ('i .S.4) FIRE PROTECTION (9.7) FEED;.'ATER IV.'IR lil STER. GENERATORS (19.C) PRE 0?EPATici!AL TESTS (14,0)
\ EiiERSENCY PLN! (13,5)
R-H/
O RCP SEIZURE At!ALYSIS USli1G CESEC CODE (15.4.2) REVIEW OF l%Ifl STEAM Lil!E BREAK ATMLYSIS (15.4.2) ECCS PUI'P ROOM LEAKAGE (15.4.6) Fli1NICIAL QUALIFICATIONS (20.0) 0FFSITE GRID STABILITY (3.2) EEfERIC IS.SWS - SPECIFIC NO-2 ACTICM 4 REACTOR VESSEL SUPPORTS (3.9.3) l OVERPRESSUREPROTECTION, LONGTERM (5.7) j i O l i l l l i O hz
l l O V. SHEDULE FOR RESO' VJ.LO:.10F IIFE13 A, THE FOLLOW!i.'G TEll ITEMS ARE EXPECTED TO BE RESOLVED FOR-THE ISSUANCE OF Tile OL N!D REPORTED IN A SUPPLEE!!T TO THE SER BY JL(!E 1,1978 EINIRON!'EllTAL CUALIFICATIcNS (3.11) REGULATCRY GUIDE 1.44 (5.0) j CONTAINf ENT sui? TESTS (G.3.4) doc /vec/ VERIFICATION OF If ?LEMENTATION OF INSTRIJ '.Et1 TAT 10N AND CONTROL SYSTEMS DESIGN (7,1) SEPARATION CRITERIA FOR CONDUlTS (7.9,4) l FEEDWATER HArF,ER IN STEAM GENERATORS (10.3) l EMERGE: ICY PLN1 (13,3) REVIE'd 0F thlN STEN'.LINE BREAK ANALYSIS (13.4,2) I i REACTUR \'ESSEL GUFPCRTS (3,9.3) OVERPRESSURE PROTECTICL LON3 TEidi B. THE SCHEDJLE FCR RESOLUTION OF THE FOLLO'c.'IfiG TWELVE ITEF.S IS D2'ENCENT PRIt'/,RILY ON THE STAFF'S FINDINGS RESULTING I FROM THE REVIEW CF PRE 3Ei!TLY SL21ITTED INFOR.'ATIC:! OR ON THE DATE OF SU2MITTAL CF CURHENTLY OUfSTAIDING INICRiMTICN, SEISMIC CUALIFICATICN (3,10) CEA GUIDE TU3E WEnt (4,0) CONTAll!M?l!T PURGE VALVE CLOSURE (6.0) CONTAINMENT LEAKAGE TESTING PR03RN1 (G.2.6) E!! VIP.CFENTAL CUALIFICATICN3 CF SPETV RELATED INSTRU"ElfiATION FCR THC l'all INSIDE CCNTAIN:ENT W,2.1) INPUT FAULT n!D SURCE TESTING OF POWER SUPPLIES (7.2,2) REDE1 DANT VALVE POSITION lid! CATION (7.6.3) FIRE PROTECTICN (2,7) l-J'-i / 3
2 O eaEcariOm TESTS uun (LOSS OF OFFSITE POWER TESTS) l RCP SEIZURE NIALYSIS USIt!G CESEC CODE (15 A,2) (CESEC VERIFICATI0;l TESTING PROGP&O ECCS PUB'P RCCM LEAKAGE (15 A.6) 0FFSITE GRID STABILIT/ (0,2) (OFFSITE POWER SYSTEM DEGPADATIO:0 l O . O R-/ W
VI. ITEMS h*.iOSE RESOLUTION FOR THE OL f%Y INCLUDE COT DITIONS TO O THE OL ENVIR0!CT!!TAL QUTLIFICATIO:!S (3.11) CONTAINMEHT PURGE VALVE CLOSURE (5.0) CONTA!NMEllT PRESSURE DUE TO ilSLB IMSS A!S ENERGY RELEASES (6.2) REDU'O.ilT VALVE POSITICN INDICATI0tl (7.6.3) . FIRE PROTECTION (9.7) REOPERATIOi!AL TESTS (l!4.C) RCP SEIZURE ANALYSIS USING CESEC CODE (15.II.2) REACTOR VESSEL SUPPORTS (3 9.3) OVERPRESSURE PROTECTION, LONG TERM (5.7) CPCS POSITIONS:
- 1. U! CERTAINTY ASSOCIATED WITli ALGORITHMS
- 3. C/C,LE SEPAraT!C:!
.12. ELECTRICAL !!31SE AIID ISOLATION CUALIFICATION I
1 l l l 1
/7-us- .
,Q V VII. GECIMLC31 . FOR OPERATitlG LICEf!SE ISSUAt!CE
- 1. I!ON-CPCS TOTAL flu:CER OF ITEi!S (20)
A. SIX OF THESE 20 ITS13 ARE I:~7.1 RESOLVED B. TEll VCP.E OF THESE 28 ITEMS AT.E EXPECTED TO EE RESOLVED ' AND 'REFORTED IN AN SSER BY JU:'E 1,19/C C. THE DATE OF RESCLUTION OF THE REI'AINIflG L'ELVE IS DEPEITCEi;T PRIFARILY ON THE RESULTS OF ONG0 LNG STAFF REVID.'S (4) OR THE.DATE OF SLZHITTAL OF CURRENTLY ClifSTANDING INFOR:MTIC?! (U)
- 2. CORE PROTECTIOil CALCULATOR SYSTEM STATUS CN SEVEN OUTSTANDING POSITIC::S li! SSER i:0,1
- 4. CEAC SEPARATION CRITERIA - SEE POSIT C:t 26
- 14. SEIS: llc CUALIFICATIC:!S - OW^TANCU:G O 15. ADDRESSABLE CONSTANTS - H!iS'JLV! D
- 10. BUR: If1 TEST - RESOLVED
- 19. QUALIFICATION OF SOFTWARE CR^s :SE PROCEEURE - OLTISTANDli:3
- 20. DATA LillKS TO PLANT CO PUTER - RESOLVED
- 25. OPTICAL ISOLATCRS - OUTSTANDING
/7-// 6
t -- - - ggygggg ggp, m,, -
- Liquid Pathways Generic Stuides: _
,.* ,_ Project Status Report HIGILIGHTS PIDATING NUCLCAR PIANT SUBCOttiITIT.E MEEIT.NG LIQUID PATHWAY GCIERIC STUDY Washington D.C.
March 22, 1978
- l. The FNP radioactivity release to the hydrosphere consists of the pronpt release and a long term leach release. The NRC Staff has .
assumed that the prompt release would come from the stap water l discharging into the ocean upon hull mit-through and would censist of from 10% to 80% of the iodine and cesium inventory which is about 100 million curies. The leach release for the FNP results frcm the core debris sitting on the ocean floor and leaching of the cesium and strontium. The NRC Staff has asstrred that ahout 50% of the total l cesium and strentium in the debris would leach the first week which is about 10 million curies. l
- 2. Sandia laboratory tests using 7 gram molten corium samples being .
dropped into seawater had cesium leach rates of 0.075% minimium to 0.80% maxi:nrn and strentitra leach rates frem nen-detectable to 2.5%
'O for the first 3 days and about the same amount for an additienal 29 days. This indicates that the NRC Staff assumption of 50% leaching within the first week may be conservative. ., -
a 3 The NRC Staff warns against j6st looking at the numlers to make a decision regarding the LSP and FNP conparision. They suggest 1 coking at the qualitative conclusions of NUREG-0440. '
- 4. The assumption used in the LPGS is that the prcbability of a core melt '
accident is the same.for a LEP and FNP and that in case of a core mel" the airborne release is the same. No additional censideration was I given to the FNP airborne release. S. The ACES Consultants present at the end 6f the.Subec=:ittee neeting l all indicated that they had no major disagreements with the existing study. Several suggestions were made for additional censideration but it was felt these would have a minor effect on the overall conclusions.
- 6. A suggestien by Dr. Fester which appeared to deserve further considera-
\
tion was that since the prompt sump water release is assumed to release 100 millien curies while the leach release is considerably smaller (10 million curies the first week) that consideration should be given to ensure that the sump water which contains the in-rushing seawater after molt-through( remains in the hull.
/ // 7 e -
k - i 2 1 i
- 7. It was noted that no NRC Staff mmber has expressed disagreement with the results and conclusions of the LPGS report;_however, it was ;
also noted that som NEC Staff views indicate that changes in the l FNP design or siting configuration my be needed to conclude that ! the FNP design does not pose an undue risk as a result of a Class 9 ) event. -
- 8. The Subcomittee recommended that the NRC Staff and OPS come to the ACRS at the April 6-8, 1978 Meeting and diseass LPGS. It was noted that a letter from the ACES commnting on the LPGS Report would be ,
appropriate. , i 1
. i 1
l a 9 1 i l l l e 4 4 4 4 19-u ? . A d A
~ . . , - .
L NRC STAFF CONCLUSIONS IN NUREG-0440 Based on the results and analyses of this study the following conclusions were reached:
- The risks associated with uninturdicted releases to the liquid pathway at an f FNP are generally less than for an LBP for the spectrum of design basis events. -
- The liquid pathway risks do not involve acute loss of life, although as dis-cussed above, some long-term effects could be manifested and economic impacts could be large. The significance of the differences in the liquid pathway related risks between FNPs and LBPs depends, in part, on the risks of the 1tguld pathway as opposed to the air pathway. Based on the information ~
reviewed and the staf f's independe'nt analyses, for most sites the risks to the public of any of the various categories of accidents (Class 1-9) are likely to be dominated by the air pathway. However, in the case of the FNP, the release of large quantities of fission products to the water resulting from a ceremelt accident is expected to result in economic and other impacts greater than for an L8P (although the impacts may be different in kind) and approaching those associated with the air pathway. O The expected liquid pathway impacts resulting from a core-melt accident at an FNP are different ,from and greater than the expected impacts from an LBP. This results primarily from the fact that measures to isolate releases to the immediate vicinity of the site are not feasible for an FNP for the first few days following a core-melt accident. During this time, significant quantities of radioactivity would be released to the open water body with resulting impacts that are greater than those associated with an LEP where isolation (interdiction) at the source would essentially eliminate off-site impacts. This study has as its objective an examination of the comparaollity of the risks associated with accidental releases via liquid pathway at an FNP to those at a similarly designed L8P. Based on the present design of the FNP and its site structure design envelopes, the overall conclusion is that, while the liquid pathway risks are small for both types of plants, the core-melt impacts are not comparable with the FNP Impacts being greater. The staff results indicate that the consequences associated with core-melt releases to the liquid pathway at an FNP are higher than those associated with an LBP and that prompt interdiction measures to keep the initial releases (within about I week) from entering the open waterbody (liquid pathway) are not feasible for an FNP. The staff considers this comeination of differences in release magnitude and interdiction potential to be significant. The impacts from releases to the liquid pathway from FNPs could be reduced to the b
%./
level of impacts from LBPs if the ability te prevent the rapid release of large quantities of activity to the open water body is provided. An evaluation of the environmental, economic, and social significance of the above g findings will be performed as part of the overall assessment of the FNP concept.
I
,+
SCHOOL OF NUCLEAR ENGINEERING a, Georgia 30332 . (404) 894-3720 RECEIVED ADYlSORY COMMITTEE Ota apcTOR SAFEGUARDS U.S. H1c March 27, 1978 MAR 31.1978 Appenerx xy
) .. -( p w pia n 's eports ggg0ittl{l345tg g li12 l 11 Mr. G. R. Quittschreiber Senior Staff Engineer 1
Advisory Committee on Reactor Safeguards United States Regulatory Commission Washington, D.C. 20555 4
Dear Mr. Quittschreiber:
Since I found it necessary to leave the meeting of the ACRS i Subcommittee on the Floating Nuclear Plant at 5:00 p.m. (the scheduled time for conclusion of the meeting), I am not sure the consultants were requested to submit a written statement. However, since this is our usual practice, I am enclosing a few brief comments. , l
, O Sincere fy, a /
Kar Z..I5fgahl Nec y Professor - K2M:rs i i Enclosure l l l l pa a c
)
(A .
. - \
b i
\
Report of Karl Z. Morgan on the Meeting of the 1 Subcommittee on the Floating Nuclear Plant (FNP) Held in Washington, D.C. March 22, 1978 i i I think it is important to emphasize that for conditions up t:o Class 9 core melt accidents the FNP has a better radiation safety score than the LBP.
' I interpret NUREG-0440 to indicate that the '
risk for airborn radioactive contamination is greater than that for 1 the contamination released to the water and that the risks from air- } born contamination are about equal for the FNP and the LBP. I consider that NUREG - 0440 is over conservative fur the FNP and that on the average the ocean cited FNP would be more than twice as safe as the LBP from risks associated with airborn contamination and much safe from the standpoint of radioactive water pollution. I do not agree with NUREG-0440 that in case of a Class 9 core 4 melt the LBP would be safer than the FNP. The difference in the two cases in general is that the radioactive contamination of the water would take place almost immediately in the case of the FNP while with the LEP it might take years, decades, or centuries for the peak of the water radioactive pollution to reach the human environment. NUREG - 0440 considers this time factor a plus for the LBP while I am convinced it vould be a negative safety factor. From my own experience with accidents, I have observed that la the population dose (man. rem) is less when the risk veil defined and comes early and disappears soon rather than in th
- m s.here it e comes at some indefinite time in the future and lingers over a long period of time.
With the FNP Class 9 accident immediate, effective, and heroic measures (such as I summarizedo inow-Ann our September my report f ll 29, 1977 meeting) would be taken to minimi ze the man. rem dose. After this, full advantage would be taken of dilution and dispersion in the large body of water and a large fraction e of th Cs and body Sr would of water. settle and be buried in the thebottom mud ofatthe With the LBP,'however, the risk probably would show up as radioactive contamination in the water supply of a future un suspecting
/ / 1 /
('
,..s S
Report of Karl Z. Morgan Meeting March 22, 1978 Page 2 public. If there is a serious radiation risk, it is better to face it and dispose of it as soon as possible rather than wait to face it in the indefinite future. Ground water contamination from a Class 9 LBP for 100 years at an average dose of only 10 mrem por year to 5x10 6 persons is 5x10 6man. rem or about 5x106 x3xio-4 = 1500 radiation induced malignancies while 5 rem average to 1000 persons during the year follow-ing a Class 9 FNP accident is only 5,000 man. rem or only 1 to 2 mali-gnancies. I think it would be difficult to find an offshore FNP site that would present a cancer risk from radiological pollution of the water that would be as great as that from some of our presently sited land based plants. Since radioisotopes of iodine present one of the major risks in a Class 9 Core melt accident of a LBP or a FUP, it is easy to show that dilution with stable lodine at the source could almost climinate the risk of radiation induced thyroid carcinoma. One gram of stable iodine in the proper chemical form would reduce the radiation risk of 100,000 Ci of radiciodine during a major reactor accident, by more than a factor of 2. Isotopic dilution would not be as simple in the case of 89,90Sr and 134,137Cs, but it still could be effective. The stable isotopes could be introduced into the sump water at the time of the accident and the basement floor of the barge could be covered with several feet of silicon sand impregnated with KI. The Si would tend to reduce tho-solubility of the reactor core mix while the KI would reduce the radiation hazard from radioiodine radionuclides in proportion to the reduction in spec 1fic activity (Ci/g). l O -
.l
-p %
'a
.- m O -.
ADV1s0RY COA 1.iHTTEE ON , l mm smco- u.s.""
. @Battelle "a""'"""""~'"'"'""""
APR 3 1978 AM yf ;) y Fil v o. n.,em u hi..e w.nsine., . m - i March 28, 1978 71891 1 1 hilil2i l:2 3:4:5:6 1 r,w .,,,.rm 942-5011
$ r,+., mm ,
. 1 Mr. G. R. Quittschreiber Senior Staff Engineer Advisory Committee on Reactor Safeguards s Nuclear Regulatory Commission Washington, D.C. 20555 ,
1 Dear Mr. Quittschreiber. This letter documents my impressions of the potential consequences of the liquid pathway following a major accident at a floating nuclear plant on the basis of the ACRS Subcommittee meeting in Washington, D.C. on March 22, 1978, the material presented in NUREG-0440, and other previous reports and presentations by the staff and the app.licant. It is my understanding that a major purpose for undertaking the Liquid Pathway Generic Study (LPGS) was to detennine whether the consequences of a major accident at a floating. nuclear plant (FMP) would be substantially l greater than for a land based plant (LBP) because of the liquid pathway. Implicit in the urpose would seem to be the objective of reaching a decision as to whether the consequences of the liquid pathway for a FNP are sufficiently adverse that some design change is necessary in order to ' make such plants acceptable from a health and safety aspect. Since the LPGs was begun several years ago, a great deal of attention by the ACRS Subcommittee, the staff and the applicant has been given to the parameters, choice of assumptions, modeling methodology and compara-
, bility of treatment of the FNPs vs. LBPs. In my view, the major flaws that were identified in earlier reports have been eliminated and, although there are still many uncertainties involved, the radiation doses as calculated and presented in NUREG-0440 represent a reasonable basis for comparing the consequences of accidents at FNPs and LBPs via the liouid pathway. I would have preferred to see the summary tables and figures of NUREG-0440 focus on the dose to individuals rather than " man-rem."
It is the dose to individuals that would detennine the nature, extent and duration of interdiction and thus the socioeconomic costs. Although l man-rem provides a simplistic common unit for comparison, it is so muddled with a range of dose rates (promptly lethal to fractions of natural bac co"kground), somers or population groupsrisn) -(users eadof orthe . contaminated beach to - O the coete imetee modes or expos #re (exteroe'
Q . dr. G. R. Quittschreiber Page 2 OBattelle March 28, 1978
' ignoring beta, to long retained internal emitters) that it tends to obscure the underlying factors needed for rational decisions. However, if man-rem is to be the basis for judgment then, as a minimum, the .
decision makers should be provided with a clear picture of the portions of the total dose that are associated with: 0 the sump water vs. the molten core o Cs, Sr and possibly a few other nuclides - e fish consumption, beach exposure, and swimming. Most (if not all) of this infonnation is contained in NUREG 5440 or in the Applicant's Report T.R.22A60. However, it. is not presented in NUREG-0440 in a way that make the relationships stand out. Such relationships are fundamental to considerations of what needs to be contained at the source and which pathways may require the most effective interdiction. Another feature of NUREG-0440 that clouds the basis for decisions about the liquid pathway is the absence of perspective in relation to the atmospheric pathway. Apparently the rationale is that the population dose consequences O sf the atmospheric pathways for LBPs and FNPs are about the same; therefore,
. hey can be eliminated from further comparison and attention can be focused just on the liquid pathway. Such a rationale would be alright if the atmospheric pathway consequences were about the same or substantially less than the liquid pathway consequences. On the other hand, .if the atmospheric pathway consequences far outweigh those of the liquid pathway, then the worthiness of directing attention just to the liquid pathway is questionable.
The applicant has provided population dose estimates for the air pathway that are on the order of ten fold higher than liquid pathway. The NRC staff also seemed to believe that consequences from the air pathway would be more severe than for the liquid pathway, but apparently have used the values in WASH-1400, which they point out were not derived in the same manner as 'the dose for the liquid pathway. At this point it is not at all clear to me whether the worst case air pathway doses and worst case liquid pathway doses each assume virtually all of the available volatile fission products to be released via that
.one pathway. Obviously this can not be the case. My perception is that there can be releases to the atmosphere without a core melt-through which would initiate the liquid pathway; but, that there will not be a core melt-through to the basin without prior rupture of the containment (an initiation of. the air pathway). We need a better perspective of the relative contributions of the liquid vs. the air pathway for the same
, h " worst case"_ scenario (s). In order to accomplish the most good (in
.V 2duced dose to people in the neighborhood) is it better to:
, A. minimize the release to the atmosphere?
/h _
~~
i . ,_
^ ~
. _ . . . . . . . . _ . . _ _ . . , _ ~ ~l _ . ... _.. . . . _. O - CBattelle Mr. G. R. Quittschreiber Page 3 March 28,1978 B. minimize the loss of sump water? or C. minimize the loss of radionuclides from the molten core into the sea water? Another consideration that warrants some attention under liquid pathways is the potential for small. " hot particles" being transported away from the accident site by ocean currents and deposited on the beaches. The nature of this potential source makes it difficult to incorporate into generic dose models, but.it would be of interest to have some order of magnitude estimates of the dose rate from a particle that is small enough to be transported by the water and deposited on the shore. Sincerely yours, R. F. Foster 3enior Staff Advisor 4
.O .
m : -- . 50 C1trd 'M; CW.7.Js U.S. Ith' . n4 m @ Bane 00e
" ~ --m . i - -
14\ ,\, , .z .
. n.os sen.. no enlm .n si
, . - h lif.nni. W.r.liisq: tun 'rWel Telepluum 0fru March 31,1978 '"'"' '2 "
l Mr. G. R. Quittschreiber ; U.S. 'lluclear Regulatory Commission
- Advisory Committee on Reactor Safeguards ,
Washington, OC 20555
Dear Hr. Quittschreiber:
'I want to thank you for the opportunity of participating in the ACRS subcommittee meeting on the Floating fluclear Plant Liquid Pathway Generic 1
Study in t!ashington, DC on March 22, 1978. The presentations made by the NRC staff were exec 11ent and the report, NUREG-0440, clearly documents the findings of the study in a comprehensive manner. I have little doubt that the dose to the public via the liquid pathway .would be greater for a floating nuclear plant in the event of a core meltdown than for the same event in a land based nuclear plant. However, I believe the flRC staff has taken an overly conservative stance on the issue of leaching of radioactivity from the core melt debris. Additio'nal tests, such as the
/-* aching tests conducted by Sandia, would be helpful in resolving the Que but I doubt that complete resolution would be possible without costly large scale tests and further confirmation of the core meltdown scenarios. In view of the very low probability of a core melt event, the expense of a large scale testing program does. not appear warranted at the present time.
I find it difficult to accept the staffs' view that the leaching characteristics of the core melt debris would more closely resemble cal-cines or poorly formed concretes than crystalline or glassy material. Calcines and concretes are typically very porous which accounts in a large measure for the relatively high leachability of these materials. Calcines are formed with little or no molting of the final product and i concretes involve no melting at all. Mclting is important with respect to leach rate since it tends to produce a dense material with a low porosity. Although concrete may be involved in the meltdown, it is lighly unlikely that the core molt debris would resemble concrete upon contact with water. The formation of concrete requires very finely divided particles (coment) to form the crystalline hydrates which " glue" the particles, sand. and aggregate together. Meltdown debris would not be expected to exhibit the degree of fincncss needed to form concrete. A crystalline or glassy material, such as formed in the small scale pro-liminary Sandia tests, would, in my opinion, he the more likely result of core melt entering or contacting water. The core melt debris may bc G
/F / y
. v.
* . .,/ * . * *
/
tir. G. R. Quittschreiber , pI liarch 31, 1978 . V Page 2 porous and fracturci 6)ut will have a porosity much less than either calcines or concrete. The 5% leach fraction in one week used by the applicant in conputing doses appear more realistic than the staffs' 50% leach fraction in one week. I believe the Sandia leach test results, although preliminary, are the best data available for estimating leach rates from core melt debris. Tests where the melt is poured into water may produce different results than quenching in a crucible as in the initial Sandia tests. However, I believe a substantial portion, perhaps most, of the core melt will not be dropping into water but will be covered by water rushing into the breach made by the initial melt-through. Quenching the melt in a crucible would more closely simulate this case than pouring the melt in water. Very truly yours, M 'kW/
#Basil W. Mercer, Manager .
Water and Waste Management Blet:mae G V O
/0 a 7 . . - .
_ _ _ - . . _ ~ ~. - - . .
~ *
$a nrog
~ f+$ % UNITED STATES g ;g ' . ,g ,
NUCL5AR REGULATORY COMMISSION
%' /? 8 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
$ .Y, WASHINGTON, D. C. 20555
***** April 6, 1978 1
'IO: 'G. R. Quittschreiber l FIOM: Ivan Catton .
SUBJECI: FNP SUBC0t011TIEE MEETING ON THE LIQUID PATHWAY GENERIC STUDY, MARCH 22, 1978 Quest ions raised'at the September 22, 1977 FNP Subcommittee meeting on steam explosions have not yet been fully answered. The Staff position on debris leach rates and that of the Applicant are as far apart as ever. In the following paragraphs, I will reiterate the remaining concerns about steam explosions and my view of the leaching rate. The steam explosion was assumed to take place under the barge. A very conservative e M.imate of the energy release was made and calculations of the shock impact on the adjacent barge was made. Two questions were raised. First, the pressure wave will reflect off the water surface as a rarefaction wave and, as pointed out by Dr. Plesset, may cause note damage than the pres-sure wave. Second, the large steam bubble will collapse and cause a water O hammer that may result in damage. The collapsing steam bubble will drive a great deal of water into the barge. This may increase the prompt release. A mechanistic view of the neltdown process leads one to conclude that the l steam explosion could take place in the lower bulkhead. In my opinion, the process is as follows: (1) The molten core melts through the four foot thick concrete pad below the vessel. (2) After the concrete pad is penetrated, the concrete-fuel-steel mixture falls twelve feet to the lower bulkhead. (3) The impact of the fuel debris on the lower bulkhead will cause the heat transfer to be high and the penetration rapid. (4) The water pressure outside the hull will be equivalent to 30 feet of water (depth below the surface) above the pressure inside. Geysering will follow penetration. (5) . . Intimate mixing of the fuel debris and water will result because of the water driving pressure. O j9'- /2 Y l
FNP Mtg 3/22/78 - Catton April 6, 1978 (6) The mixing process - water up and fuel debris dwn - will te highly susceptible to a steam explosion in the lower bulkhead. (7) A steam explosion in the lower bulkhead could result in enhanced radioactivity being driven out the hull ' annulus or up through the hole in the pad. It is my understanding, from conversations with Dr. Speis and Mr. Marchese, that all of the above aspects of the steam explosion are being resolved. I have not seen the resolutions. The question of what leach rate is proper is still open. The Staff and the Applicant differ by about a factor'of twenty. The Applicants arguments are based on their view of hw the Canadian leach data should be used and their interpretation of the recent SANDIA leach data. Details of the SANDIA test are not available as yet. The Applicant visited SANDIA to obtain information for his position. In many respects the SANDIA leach data are non-prototypic. A small (7gm) sample of cor rete-fuel in a platinum crucible was rapidly quenched by n immersion in water. The s'ug of material was then put in water and leach V rates were determined. The sample was glass-like with only a moderate enount of cracking. My experience has been that significant cracking and fragmentation occurs when nolten glass is poured into water. Further, if the pour results in a steam explosion, very fine fragmentation occurs. The SANDIA experiments found a sample surface area of 100 cm'/gm and 2.5% loss of Strontium in three days. A nederate interaction of the fuel-concrete mixture will probably increase the surface area by a factor of ten. Tempera-ture effects and buoyancy driven circulation could easily lead to another factor of two so that one obtains the Staff result from suitably interpreting the SANDIA data. My conclusion is that the Staff estimate of the leach rate is a best estimate. Measurements of leach rates are very difficult and the results are quite variable. At the " meeting of experts on leaching", one of the experts in-dicated that they usually throw away the first few weeks worth of data because it is too unreliable. He also indicated that the same test repeated in the sano laboratory can yield results that differ by a factor of 10. A factor of 10 difference exists between the two sets of data from SANDIA and the time period is that usually considered to be too unreliable to be of interest. As a final note, the second barge must nuintain its cooling and as a result circulation in the basin should be a consideration in any methods of miti-gation under study. O y df -
APPENDIX XVI NUREG-0440: Major Conclusions NRC STAFF ASSESSMENT ON I LIQUID PATHWAY GENERIC STUDY l l l l l l O l PRESENTED AT ACRS MEETING
- 1 APRIL 6,1978 l
l l O
~
1 N-/3d
~ ~ ~ ~
4 SlM4ARY OF MAJOR RESULTS t t
- IMPACTS VIA LIQUID PATHWAY FROM DBA'S SMALL AND SIMILAR FOR FNP AND LBP
- i
-- CONSEQUENCES OF CORE-MELT RELEASES, FROM CORE DEBRIS AND SUMP WATER, HIGHER FOR FLOATING PLANTS
- PROMPT SOURCE INTERDICTION AT FNP NOT LIKELY d - PATHWAY INTERDICTION EFFECTIVENESS PROP 0RTIONXL TO EFFORT APPLIED W
- x ,
o O ___________________________________..__._._,m -_- -
- - - _ m _ . - . - . - ->
l p c.,)icL.2 s O O D;'y GENERIC LIQUID PATHWAYS STUDY
- STUDY INITIATED TO ADDRESS ACRS CONCERNS
- STUDY INTENDED TO PROVIDE BASIS FOR COMPARING FNP AND LBP DOSE CONSEQUENCES VIA LIQUID PATilWAYS
- OPS AND NRC STUDIES CONSIDER RANGE OF RATIONAL SOURCE TERMS g BASED ON ENGINEERING JUDGHENT AND USED TO CALCULATE DOSE N
CONSEQUENCES (j - DOSE CONSEQUENCE CALCULATIONS GENERIC RATilER THAN FOR SPECIFIC SITES
- OPS PERFORMED CALCULATIONS FOR FNP'S, NRC FOR LBP'S
- LIQUID PATilHAYS COMPARISON APPEARS IN NRC REPORT, E NUREG 0440
- COMPARISON Of LIQUID At;D AIR PATilWAYS RISK IN OPS REPORT ED BUT NOT IN HRC REPORT S mx ,
O o-2- 5 t [ en
O O O V l , DOSE PATI! WAYS CONSIDERED,f0R LIQUID PATilWAYS STUDY
- SEAFOOD CONSUMPTION DOSE PATHWAY -
- BEACil EXPOSURE PATiiWAY
- SWIN11NG EXPOSURE PATI!WAY N - DRINKING WATER INGESTION DOSE PATilWAY -
N\ t t t
O O O JPS CONCLUSIONS - FNP LIOUID PATilWAY DOSE CONSEQUENCE CALCULATIONS
- FISil liiGESTION IS DOMINANT DOSE PATilHAY TO MAN FOR OCEAN, BEACH, AND ESTUARINE SITES
- FOR ESTUARINE SITES, DOSE CONSEQUENCES ARE ABOUT FACTOR OF 3 GREATER TilAN OCEAN SITES
- FOR RIVERINE SITES, DRINKING WATER IS PRINCIPAL DOSE PATilWAY; DOSE CONSEQUENCES LESS TilAN OCEAN SITES t
- POPULATION DOSES AND MAXIMUM INDIVIDUAL DOSES DO NOT INCREASE SIGNIFICANTLY AFTER 2 YEARS
. - INTERDICTION TO REDUCE LIQUID PATilWAYS DOSE CONSEQUENCE IS FEASIBLE, COULD BE N
N 1 APPLIED IN SUFFICIENT TIME AND WOULD SUBSTANTIALLY REDUCE DOSE CONSEQUENCES
- RISK VIA LIQUID PATilWAY SIGNIFICANTLY LESS THAN VIA AIR PATilWAY CONSIDERING REALISTIC INTERDICTION PRINCIPAL CllANGE IN OPS PERSPECTIVE SINCE OPS REPORT
- INCREASED POSSIBILITY FOR CONTINUED PUl1 PING OF SUMP LIQUID AFTER POSTULATED VESSEL MELT TilROUGil l
! 0; - o o 1 .. NRC CONCLUSIONS - NUREG 0440
- RISKS FROM LIQUID PATilWAYS RELEASE 25 to 300 TIMES GREATER FOR LBP FOR THE SPECTRUM OF DESIGN BASIS EVENTS.
- RISKS FROM LIQUID PATilWAYS RELEASE 6 TO 30 TIMES GREATER FOR FNP FOR EVENTS BEYOND DESIGN BASIS.
- UNLIKE RELEASE TO AIR PATilWAYS FOR SEVERE ACCIDENTS, LIQUID PATHWAYS RELEASES DO NOT POSE-AN IMEDIATE RISK (ACUTE
-FATALITIES). ,
- UNDER REALISTIC (CONTINUED) USE CONDITIONS, SIGNIFICANT DOSES TO INDIVIDUALS ARE NOT EXPECTED VIA LIQUID PATHWAYS.
- LARGE AND RELATIVELY PROMPT RELEASE OF RADI0 ACTIVITY IS 4 LIKELY FOR FNP BUT NOT FOR LBP.
bJ - FOR FNP, TIMELY AND EFFECTIVE INTERDICTION OF LIQUID
@ PATHWAYS DOSE ROUTES IS FEASIBLE AND LIKELY (WITH ATTEN-DANT COSTS).
- FOR FNP, ECOSYSTEM IMPACTS VIA LIQUID PATHWAYS ARE IMMEDIATE, TRANSIENT & REVERSIBLE.
- FOR LBP,. ECOSYSTEM IMPACTS VIA LIQUID PATHWAYS ARE DELAYED AND MAY BE PREVENTED. IF THEY OCCUR THEY ARE LIKELY TO BE LONG LASTING. ~
~
O O O 4 NRC LPGS STUDY AREAS OF OPS DISAGREEMENT WITil NUREG 0440 PERSPECTIVE
- 1. HIGH LEACil RATES
- 2. DOSE PATilWAYS INTERDICTION CUT 0FF AT S REM i
- 3. LAKE AND SMALL RIVER LAND BASED SITES NOT INCLUDED IN SU!HARY TABLES
- 4. NO COMPARISON TO AIR PATHWAY .
x AREAS OF OPS AGREEMENT, NUREG 0440 N l. DESIGN BASIS ACCIDENT EVALUATION Q{N. 2. 3. MODELS FOR LIQUID PATilWAYS DISPERSION & DOSE STEAM EXPLOSION EVALUATION P
l O RECENT SANDIA LEACH TESTS
~
, TEST SET #1 CORIUM: CONCRETE CATION AMOUNT LEACHED IN 3 DAYS (%)
(WT RATIO) 250C 900C 9:1 Cs 0.075 0.17 Sr 1.2 2.5 7:3 Cs 0.099 0.25 0.68 Q Sr 1.0 5:5 Cs 0.020 0.80 ! Sr -- 1.1 TEST SET #2
.1) THREE DAY LEACH VALUES SIMILAR TO THOSE FOR TEST SET #1
- 2) INCREMENTAL LEACH FOR ADDITIONAL 29 DAYS OF LEACH SIMILAR TO THOSE FOR INITIAL 3 DAY LEACH PERIOD l 1 2
- 3) MEASURED SURFACE TO MASS RATIO (BET), 100 cm /gm i
O.
-____ _ _ - . B-es7
h
.O O 0; i
CALCULATED DOSE CONSEQUENCES, CLASS 9 ACCIDENTS , ESTIMATED LIQUID EST3 MATED LIQUID PATiiWAYS C0llSEQ PATn AYS CONSEQUENCES, NOINTERDICTIONgCES, INTERDICTION SITE PLANT TYPE (MAN-REM) (MAN-REM) OCEAN FNP 106 103 LBP 105 10 2 7 ESTUARY FNP 10 104 LBP 106-107 103-104 6 RIVER FNP 10 103 N LBP 105 jg2 N LAKE LBP 106-107 (j)
'SMALL 6 LBP 10 -107 (1)
RIVER (DOSE CONSEQUENCE BASED ON CASES DESIGNATED AS "EXPECTE0" ON EARLIER TABLES) (1) NO CALCULATED VALUE IN NUREG 0440 (2) BASED ON TABLES IN NUREG 0440 _ _ _ _ _ _ _ _ _ _ _ _ ___ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ . _ - _ _ _ - --___.______-__ - -______ _____ _ _______ _ - ______m--
O O ' O COMPARISON
SUMMARY
INCLUDING AIR PATilWAY g EFFECTIVE EllVIR0ff- TIME BEFORE PATilWAY w SOURCE EXPECTED BIOTA MENTAL ACTIVITY INTER- CALCULATED (V MAN REM INTERDICTION SOCIAL KILL CONTAM- REAOlES DICTION ACUTE g PER CORE MELT ABILITY COSTS ZONES INATION PATilWAY ABILITY FATALITIES ()) STAFF CASE (2} LIQUID PATilWAY Flip N0 llIGli YES MAJOR Sil0RT NO YES LBP YES LOW N0 MINOR LONG YES NO APPLICANT CASE (3) LIQUID PATilWAY fitP PARTIAL MODERATE NO MODERATE S110RT YES NO LBP YES LOW NO MIfl0R LONG VES NO AIR PATilWAY I) FNP N0 ilIGli PROBABLE MAJOR Sil0RT NO I4) YES LBP N0 llIGil PROBABLE MAJOR SHORT NO I4I YES (1) ASSUMES NO SOURCE INTERDICTION (2) ASSUMES llIGil LEACil RATE FOR FNP (3) ASSUMES 15% TOTAL LEACil, 15% PROMPT RELEASE * (4) NO INTERDICTION FOR DIRECT EXPOSURE OR DIRECT INGESTI0ft; FOOD PATilWAYS INTERDICTION FEASIBLE
AIR PATHWAY DOSE CONSEQUENCES, SEVERE ACCIDENTS (ATLANTIC C0AST SITE-0NSHORE) 10-4 N N
+444e ++++.5,
---- -~ ~~
,+;?,+ 9~4 N' 'qh k
10-5
- N M \
+ \
k \
+ \
r \ w 1:-6 &
% g\
\
p.e
- k
\)\
Oh k e . d k U k \ b ' m 1. 10 7 - ; 5 Y g
$ WASH-14C0 , 1 ACCIDENT DISTRIB JTION, +
LAND BASED PWR +g 4 NRC RESEARCH + 'g ACCIDENT DISTRIB JTION Y - 10-8 LAND BASED ICE CONDEN$ER YI ,
++++ NRC RESEARCH SMO )THED Y, $
ACCIDENT DISTRIB JTION, 1 OPS RESULTS FOR NP' Y
,1 tl
+1 I
10-9 +I O O 101 102 103 104 105 106 107 108 109 TOTAL MAN-REM (X) l-7 i n l
i-O. O O i CALCULATED ACCIDENT RISK FOR AIR PATilWAY SITE CASE MAN REM / YEAR ATLANTIC COAST SITE, DNSil0RE OPS-FNP- 70 t NRC RESEARCll, LBICP 180 N -
- WASil-1400 PWR 270
\ EASTERN RIVER SITE WASit-1400 PWR b4b NRC RESEARCll, LBICP 610 ,
t ATLANTIC C0AST SITE, OFFSHORE- OPS-FNP 70 NRC RESEARCil, LBICP 180
- r - - + - - . , ~ . , ~ ..a w r- - - ---- . .- -- - ,_.
O. O O t' LIQUID PATilWAYS - AIR PATilWAYS RISK COMPARISON MAN-PEM/ YEAR INTERDICTION NO INTERDICTION FNP 75 - 150 100 --200 N LBP 75 - 600 100 - 800 N FNP 10-I'to 10-2 50 - 100 LIQUID PATHWAY -2 LDP 10 to 10 -3 5- 70
O O , O DOSE COMPARISON FOR MOST LIKELY SENARIO FNP LBP INTERDICTION 106-7 M-R 3-4 10 M-R SOURCE PLUS LOW PATHWAY EFFORT 5 3 10 M-R 10 M-R SOURCE PLUS HIGH PATHWAY EFFORT 1 N 9 - e
- 2 "8
3 .. 2 aw
- c. 8 4 3*T a==
O +-e 2, EL
t ?
,.4 E' .
[N
\-l l.'
M e =
- W L Q W U H m
^ J d D * -H M C3 D C of5
- C Z =
- O c -
. Z W M W < Q M
C" w W J ,
. G O c W e w ,
8 A = d M
- E.
Q O
- W D
= A M< % Q D W .
E W
~
,> = W W W
- a -
H = M 6 C d C C W . - H
- w <
W M > >H H C d . A
= = M W W J w 3 Q
- 3 4 c y U w W W - - -
d l
> M W
M M W J o' C C m O J C
=
=
O W O
=
H M W W % < < - D K W
< H
% M W
w O o H M J J
% M J J J C H @ W -
- C G W W < M A y Z ~
M L H 3 W
- O
= Q < W J L O .
W - M' - J J = 3 < c -
. e.R W < = o W = =
m o - o e c W o = H -
- g W W H - - C = - C 3 "
M.H . a f M H
=
=
H
< =
m o Q H H O a
. a*W .
f J - J H = C U < > =M o M L O O M - = d J g ( O = m W w o c
- o d m N - C - 6 W Q c U W =
7 6 M W W M 3 C
$ 3 W o =
- W 4 - < H m H
W - l 4 A c U *
$ W < W m
* =
< 3 H e > I
= > W > - J J M W c C
- C O W O J W U q w
- 1 N G C O c w W - - W W V Q H - M d W = E W d W D - Q H w
> d C H c <
M H E M c W M U - H M W W > W C D H = W x u H w = < - w - = - m H U W U < W H< 3 M < W C M C q H y - = L < W M 3 = U ~
~
W W = -
^ % 4 < H J W =
J o w a e . W C H = W L H = H 4
- = m w c W -
W G *. E = * = W O J A
= H C C 6 > A W U = > Q = 3
- < O = Q M O < H J - J N O O H O C W = W M J U
=
- J M C M M N = = U Z - 4 3 ~ J 1
4 U - o
- J s e m U W d -
= x H w
= U = W H 6 Q
1
= W
" p W I - W
= O M A R m W = 6 W W W c H W
3 e e < m m W m J < 6 - U % W H W < = 0 W G W W O & J W W < E W - H 4 J W J - < O W
- = O H
= c W M J W H E C C C d = =
d C J V = 6 - H Q J L W W c W * - H k H 1 = M G C 2 d -
= - =
m V r U = x g m H c s - c - a e a W e e e C C = C C E W = > - 2 -
= H J 6 c W M G J m W U W
< Q J O O W c e 1 w J = w H W < = G < 6 O M M J - x J W
= W W C M ~
C < R ! W W W
~ J
% A 1
v
/ /
7s n ~. A N COMPARIS0N
SUMMARY
- EXPECTED CASE LIQUID AIR LBP FNP & LBICP*** PWR FNP TIMELY SOURCE INTERDICTION NO NO YES NO ABILITY YES YES YES PATilWAY INTERDICTION ABILITY YES SMALL LARGE SMALL RELEASES TO OPEN ENVIRONMENT LARGE TIME BEFORE ACTIVITY REACliES StiORT S110RT LONG StiORT PATilWAY 10 6-7* 10 3-4* 2 x 10 6
4 x 10 3 MAN REM PER CORE MELT ** ** N 10 10 N' EXPECTED SOCIAL COSTS LARGE SMALL LARGE SMALL N YES NO BIOTA KILL ZONES __ NO NO NO N0 PATilWAY FATALITIES t l
- SOURCE PLUS MIN. PATilWAY INTERDICTION
** SOURCE PLUS MAX. PATilWAY INTERDICTION
*** SIMILAR RESULTS WILL BE EXPECTED FOR BWRs l
I
O i l l l PATHWAY INTERDICTION I
. POPULATION DOSE = f(INTERDICTION LEVEL)
. ESTUARY SITE
' l LEVEL POPULATION DOSE
~ l (REM) JEDUCTION FACTOR
,5 20 1 100 0.5 200
. OCEAN SITE LEVEL POPULATION DOSE i
_,( R EM )_, R_ EDUCTION FACTOR 5 10 1 20 0.5 30 l l 1 i O
# /r4
O O , O AIR PATHWAY COMPARISON COMPARISONS OF MOST LIKELY SCENARIO WASH-1400 ICE CONDENSER
- PWR RELEASE CATEGORY MOST LIKELY TO RESULT PWR 7 PWR 5 CORE INVENTORY RELEASED 0.6% Xe-kr 90% Xe-kr
-3 2 x 10 %I 0.2% I 6
MAN-REM 4 x 10 2 x 10 ACUTE FATALITIES <1 <1 i COMPARISON MEAN CONSEQUENCES -N (ALL CATEGORIES) 6 N MAN-REM 2 x 10 8 x 10 6 ACUTE FATALITIES <1 <1
- SIMILAR RESULTS WILL BE EXPECTED FOR BWRs
O O O IMPORTANCE OF DEBRIS LEACH RATES SCENARIO RESULTANT SCENARIO AIR PATHWAYO) LIQUID PATHWAY (2) MAN-REM !' RELEASES PROBABILITY EXPECTED CASE SLOW LEACHING RAPID LEACHING
- 1. 5 0-y 2
AIR + 36% 2 x 10 6 2 x 10 6 3 x 10 6 DEBRIS + TML-y SUMP
- 2. 5H-y 2
AIR + 22% 1 x 10 7 10 5 10 6 . DEBRIS 3]g_Y
- 3. 5 HF-y ,6 AIR +
2 DEBRIS 17% 3 x 10 7 - 10 5 10 6 i N N og
- 4. STEAM EXPLOSION AIR + 10%I4) 10 - 10 8 7 (3) 10 6
2 x 10 6 DEBRIS + IN OR BELOW SUMP BULKHEAD CONCLUSION: DIFFERENCE BETWEEN APPLICANT AND STAFF LEACH ASSUMPTIONS HAVE LITTLE EffECT ON RISK.
- 1. Average of several east coast sites, variation less than 50%.
- 2. Assumes effective source interdiction after one week for FNP and pathway interdiction to 5 rem integrated max individual dose.
- 3. Estimate. -
- 4. Conservative estimate of likelihood of SE given core melt. ,
~a _ . _ _ - - . . - _ - _ - _ - _ _ _ . _ _ _ _ _ - - - _ - _ - _ - _ _ . _ - - . - . - - - - _ _ _ _ . - - _ - _ _ _ _ - _ w , c -w,,- a ~
O O O PRINCIPAL FNP CORE MELT SCENARIOS AND RELEASES TO THE ENVIRONMENT .; SCENARIO RESULTANT SCENARIO AIR PATHWAY LIQUID PATHWAY COMPONENT RELEASES PROBABILITY COMPONENT DEBRIS SUMP
- 1. 5 0-y AIR + 36% % 100% Xe-kr s 0% Xe-kr s 0% Xe-kr :
2 DEBRIS + 1% I 3% I 96% I ,
'TML-y SUMP 1% Cs 5% Cs- 94% Cs s 0% Sr 89% Sr 11% Sr s 0% Ru 92% Ru 8% Ru
- 2. S H-y AIR +
2
;. Sj H-y DEBRIS 22% s 100% Xe-kr s 0% Xe-kr 1% I 3% I 4% Cs 5% Cs 0.3% Sr 89% Sr 0.9% Ru 92% Ru .
- 3. S HF-y j6 AIR + 17% s 100% Xe-kr -
S 0% Xe-kr 2 DEBRIS 6% I 3% I N, 14% Cs 5% Cs N'l 2% Sr 89% Sr 1% Ru 92% Ru 4. S.E. IN AIR + 10% s 100% Xe-kr s 0% Xe-kr s 0% Xe-kr '
\ ' S(
43 OR BELOW DEBRIS + 6% I 1% I I 93% I BULKHEAD
- SUMP 8% Cs 20% Cs 72% Cs '
1% Sr 89% Sr 10% Sr 11% Ru 9% Ru 80% Ru
*- ASSUMES S.E. P = 10%, WATER DF = 10, RELEASES AVERAGE OF SCENARIOS 1-3 INCLUDES 3% VAPORIZATION OF_ SUMP a
6 0 m m - , -~r
O O O COMPARISOff 0F STEAM EXPLOSION RELEASES TO MORE PROBABLE RELEASES Xe-Kr I Cs Te Sr Ru Average Release from 90% of Core Melts 1.0 .02 .05 .07 .005 .006 Release from SE below RC 1.0 .06 .08 .15 , .01 .11 s NI N V I O Release from SE in RV 1.0 .27 .67 .40 .08 .43
) .O ! i 2 00SE CONSEQUENCE FOR PARTICLE RELEASE I
.PARTICLESIZE-idMICRONS
. CORE MASS. INVOLVED - 1%
~
. BIOTA UPTAKE MODEL -
, - GI-TRACT TRANSFER (ICRP-2)
-RETENTION'PER NUREG-0440 l
. CONSEQUENCE TO MAN
' ESTUARY Cs & Sr - 1. x 10 7 Particies 5. x.10 6 TOTALev/2.x17 OCEAN
' Cs'& Sr - 3. x 10 6
- Particles 1. x 10f TOTAL 4. x 10"
. CONSEQUENCE TO BIOTA
- GI-TRACT EXPOSURE LIMITING
- INCREASED ADULT MORTALITY 4
OL g-M
ama, ,.4 ..g _.p- p 3A , .hi - e.w._ d ld.,m , _.3.W m,JJ_4k a ,- ; sh e
~
O RIVER-ESTUARY-0CEAN C00PLING ESTUARY-0CEAN EFFECTSNOTADDITIVE lARGE RIVER-0CEAN EFFECTSADDITIVE
' ~
LARGE RIVER-ESRIARY EFFECTSADDITIVE
' CGPARABILI1Y OF FNP-LBP NOT CMr3ED i \
l l O. f- /S~ *
- p. _7 Q
.i1 : .j . 1. .j.. mi -
- h : . u in.:
p . _b _
, , i
. . . . "i :... .e .1. .
..,.,,,. a.._ . . . , , _ ,'- _ . . _ _ . . , _ . _._ _
.;e 4.. y. q ;, ,
r g jy : .[ - .
. 7 #.
-..- : . . .=r
...J-. . . - , . .'.. .y w
_ g ._ i
. ..- a n ;; . . , , ._ .
ri_L E. .fi 0 :- E ' :" " . r
!. ' 1
.L.
f' k; !! 'il- 9'ir
\ . .[ q__\ g
- I
_ p. . _ p. _ i ;
~. %v ,'-~ 4k hl .
. . d.
.p. .-J - . . . . .
-i. _.a _. .
F. _ g . 3 J ~.;; , .p:
' i i t
.g ;
,i-- Q Q M C m
[ . - r- --
--{ 3 _. _s:_ . -.4 k
u .; .
. _L _.
- : f."
3p c_. ._; . _ .p . ,_ h i% , A./ G id. -T.. _.
.v_. -
1 .
;.. -_ ; . _ . tu - - %.~ . .: -. 7.7 x.-=
..;_.nu .- ..;__-_,.._;.
m.__.; L L __ _. y .-. 7 .- % ;:_ .
- t ., %s=_- . _ . . _ _ . u.. .:: 7
(-
"b .W
~
= ::. t i l
. i l =:r
~ ~~-- : W
. , i i
" ~~ #
G .~ . . I , c- - ,- .j.r , % ...,.2 F:_ c: . . . :4 -
; du. - . i. . *
... n t x. . . . . -"- =
--- . 9
.o :--- - .-- ; ..
. ,r...
.... . _ . _ . . \, ~ ~~..
..~- -
.. t : . ,
. p. . ut. :L ....t.- . b--- .
. . . --..r=-- -
=r p,: _; e 3 _ ;- ; ;
p c!" E.. mtm . - . . .
.!.- q .q.
=..c.; __:p a b. . . . .
. .--r- : r- r "-
nr:
. . .r.=.=.p:.
. . '.. r.. _=. . -
.. y .
.i . . _ .. t :
. .r . -
' ~: '~=
c..
.;= := -r;=. =: y.
Ftir: ME' in!H __:. . __ l" . " I.. . = * . {i. =' i: . h . L '! (.
- g. . ^h r.
t.. -
. :. .f f5 Fi f,,.."_ [C'{I{.~ _
5 Y h - Nif NSC . T:b i- E 1 l.h 5 :: :. 3 1- Nb.
~ .- :
, _r.--: .: ,. . . . . ... .-- _ . - . . _
.. ."] ._-{*'"-: . ^,.f.h * . ]{.'~E;'-.~~
^ ~ ~
. . *.* { ".:' ' , . *Lj
-~ "
.,{'**
.~~ ,
- I.c. . p ..
+(
+
.p. .. p. L ._: . t. ._. . . l'i. n. 2_iG-_
- _: ' i~. . . !. . !.:. . . .. . h'
. ..L _,
s:= zi n mur .$, -i -1' i 4. T h * " F "
- 4. : . M9 5E
,M @ . _ . . .
E!N _ IICg )T l k g
!iE I *
.!M55
...c.
ti=4 m: ".h? ,N ! , y !' :b i- -- r d
- .:::u=.:._.b. =.1'_ :
- !,_gh i .g. 4.'. ..
~
,. . F.
.;: ._=. .r. . . !.:.n.5.
. . Y. . . cE"E
-G fi Sii : 4r. !Mi- I _ b: "
? !.: i I:i ~ .bi 4
-- I. : F=R '
"EFi~- : I :~ C
( .. . j~.....g.g:
. . . . , .. _ ??.y .rl' . .*. " E:~y
=._;_--._.=.. l,..
*:{ C. '
. .'7=.. .
. ; L. t. . :._. : .*' I..
.. .=.(~....
a
~.=.?p . . ' .Q )? ~- .:..._A. .%. . [2-~~
O - 55 58~ _b -d=" Mri: =:r };:i'- . r5 ~'h h " Hr.P 1 '23 C p pi EK.( Z
..iD. t H_.. f._
_mk. .:Mi.j d ; 't K.'#-
.=_: lE@W :HiI:- d 4: _ p.; =@.=J:., .h. :.F_. g.M_-~
3Ets .iii GI d -i:Is Cin Mlh : Ef ubE F .' 2:G - *:. # ' d= ~ N.hr"i 2._=j=E. .=._E:? t Wb=..EMN ,s ct g .i JyI_ . 3_= C t r. ru; E. . . .
- r. . H. =. ::hr. - j ". ' h.. ..-[~ . .'? %
~
. : Bi " h.?. E.. .
* . yj : e ' -- B M-
$ .'O. -ii. ' '.t.... . :E-
- i~ i'-
f=.M.i*i-E1.Ubi
. . . . _ JL.L yA - ,.
=... a_.. j.-
. a... l,- .g - k. . . . ... . .
_ . _ . . . =. . =. _ . g .: ,. E-. . .- _. . . . . . .1. . .. _. . . N ' __Y . b
.g ==:U.: . ;. ' V .
1 l ..
. . 5.... . i.
..[ ' ..,
. t. % l
;. ; (
.... . . N ..
~..'T-*..~
=.. 4.. . . . Au ;.: % ,L . ~
. . . . .. . . N--- -
..._p......=_.
=.; _: q, .y r.- - -6 .. .. .
... g. . O :=.. .
.: b:
g . _.
==n -
.w - -
m r i r: .e
= ;um
- m: ~ -
- - L. . -gg
-- .t- ... : ---. . - -3yt-. . N Q a . ..
. . : =.:. . t_ : . .-
- p. ,
- T
- m- -
- t M .g
- m. f~ .:=. .
.. .-.% :. (Y k . m..... K~' .
.(
- K :-
d.. &
=:= = =_. ._.. 1___..v
,}_
5 -dig .J' - Q .A
- iir . .i. =iilE. !E +
r _~ [ .= -
- .;EE
~^ ~ **
bd '.
.' h ... . _ .' ' C 'i.: ~. .= -.dh.;bbi-
.b._k.i!5 5' !. ! _ _*. L**"**"*b.~*-b .... :$ ." .. . . - . " }T*'
_. .. . { . '. ~.~k..-- ' '. ' b .
= w. " . . .+' '
% g- s. . g i.. . .f.3 # . 4 f.r=3 -. . . e.E-
- 5Esi2 T:!P= kd#i ci C v:= tii. i@ : l- =
=:FE -i- r
.@' F 7AM r .'!:::
r r ' IP : 'N .k,
~
= _ . . *E.y.
. . =:..E-i'_~h.t.:
_ __. ..::U. .. n_ . F :: _ _ u..E. . . iE_ d.i.ni .
;". i.:. i. ,. ' .r' g
. = d =- -" ::-i@:ii".E .
=;;; . .. t c. m =. . . .a.:-. ... ....;--- :
* ; . .c c... L:.. . . u=n .
. . . .:*-~y, :' *~~~'_
=_. 1 * '.C"_. T.t**~l'
=t.~J.:-
i.: rtr'*; ::.- !r---
- " r. . U*':12 . . . . . . . . . . . ..
{- l * .
,' . . U _.'
'. : p..'P".'.
. . . .. ... " _ - ' . ~ . ~ .
g - . . . - . g . ., . , , , , , , , , , _ , , , , ,, , ,, . , , , , , ,, , , 2: . . . ;. ... n . . . . . . . u .1 r g: ...;=; m rn[,2 .; ., .
- .n; c . ;-
__p=...
=. :2*:tt:. =;n= :t --
+
- fr. .c r. u; =. u- .- . * :; ;.a
_. n .
- n:f-- . . .- - - - - .I..-
_-[---*-~
---t..-
d- - _ - _ h:-
- - - ..- .=.-
_-. _ .. t _. 3--- - . . . 5$N ...x.. M* .N =";'. 4 7-55fb
. i' ~: M3 h:ifC. . . . ~ *A.[~ 'TS D 2-. . .!! E t) .lI e ~.E I h! ~ ~
' N 3 ~~' _ ~[Z O 5I5' ~5Si: si.N 5Sh !55 %!!E Mi 5!5i. .=:= iii" ' :! ::Di m:b 1. . . . . . . . .
.u.:j rM E- 7%=
=EiiE EiPi~ El5.55~ E E!!Mi @i 5E 4% 5 nb 9 ::4.i %$. M +=h=H=-M5i!
"Y % m w.4 % ,,, m , u., , - g ,$
APPEi1 DIX XIX (^' L hUREG-0440: Core fielt-Through Penetra-tion liode and Steam Explosions STAFF ASSESSMENT ON CORE MELT-THROUGH PENETRATION MODE AND STEAM EXPLOSIONS O l PRESENTED AT ACRS MEETING APRIL 6, 1978 l O 1 l
C A.5E - 1 : R3UR-M?QU G H PMWR&TM MODE , ., .. . . . ,... lt s "/ g 6=* 4 s- 4 \ % I. t
,. *: an[~
n..: c...:. i..,,,g . ( ey oJ ; e
*I . f --
s '; () !
<l 1> s.
._. cos. -
- 9. $ .
. g, *
);' ,6?s .
e.g - Q '
' ll.
I'o c -. -f s
~ ~ , \, 9 s .. /
- /; = , > v, :,0
/ , o r. ;
'. w. /~ a.
.' e n.. ' uo\ r ,a y#
s.. s.
. . e. i
() ,- D. is.. j 0h ,. l' . . \ . D l ',. h
- s [ .
l'.'sq.ck,)i_g.go':n?! t
'Cbi'.'CniTE^ y 2~, - ,c ;.o 6 % ,i: ? u'e.*. I,
, ?, .' . ',,, (WJ....'~.n s ,- ='
A m.w, . . o, c ,. ' , s' ;,
~
C. <W E. 3. : UW/,=O:= .*:1 , p.7 2.E 7;'.=. T / o.n/ m o O. F..
; '. M
> < __.RtAcrog
..'. :: ' VEs3 St.
(, H I ';~. *s^~,...- wo, ,; 1,
-c . - '. C6RE
',. 6 3
- ! n ce, i B
, (
e, ,
, 0. ' 'ev-..- y.
'\. .. . . ,, ; ;
.c ., ,.
e j te , t. i b1ELY '.,,; E A.
~
.. s.,,_ ~ . - s a.-. ~ .. - ~ - , ;- -l , ;
b gl _ __I ' i > j !
- c. I I __.. _ . _ _
,l l. : -
.I 1:
',-/ l
. I g
- . . ~ .. -. . . ...._..t
.. .i AgowTH i
t' .
' ~ , c. ; , -
,, ' . t
,,' ,,. C Q W,0 Wf5 ' , , ;-[ lNi t .
fjik.TTE* RNl (APPRDX. s
== fb i _ -
l O
~
CORE fiELT-THROUGH PEilETRATION MODE THROUGH REACTOR VESSEL WASH-1400 CORE MELTDOWN SCENARIO ESTIMATES: CASE 1: MOLTEN CORE MATERIAL IS TRANSFERRED (POURED)- FROM THE CORE REGION TO THE LOWER HEAD IN A SLOW CONTINUOUS FASHION - THIS TRANSFER (OR POUR) WILL MORE LIKELY LEAD TO THE DEVELOPMEIIT OF A LOCAL HOLE IN THE REACTOR VESSEL. O CASE 2: ON THE OfHER HAND (LESS LIKELY SCEilAR10) IF THE MATERIAL IS RAPIDLY COLLECTED IN A LOUER REACTOR VESSEL REGION IN A POOL AND A SLOW MELTING OF THE VESSEL OCCURS, WIDESPREAD FAILURE OF THE LOWER REACTOR VESSEL HEAD IS POSSIBLE. I i l STAFF UTILIZED THESE TWO SCENARIOS TO ASSESS THE SUBSEQUENT EVOLUTION OF THE MELTDOWN SCENARIO AtlD ITS OUTCOME, l O I
,9-/s~C 1
O CORE MELT-THROUGH PENETRATION MODE THROUGH CONCRETE CONTAlflMENT BASEMAT DEPENDS ON INITIAL CONDITIONS: CASE 1: INITIAL HEMISPHERICAL MELT GE0 METRY (CAUSED
^~
BY MELT BURROWING A HOLE IN CONCRETE FLODR) WILL CREATE A NEAR HEMISPHERICAL GROWTH PATTERN IN CC" CRETE RESULTING IN A GRADUAL O POUR-TilROUGH PENETRATION MODE. CASE 2: INITIAL CYLINDRICAL MELT GE0 METRY (CAUSED BY MELT SPREADIllG OVER CONCRETE FLOOR) WILL CREATE A NEAR CYLli!DRICAL GROWTH PATTERN l IN COIlCRETE RESULTING IN A UNIFORM MELT-THROUGH PENETRATION MODE. l O
$- /S~ ?
0 C0RE. MELT-THROUGH PENETRATION MDDE THROUGH THE FNP STEEL HULL PLATE CASE 1 (POUR-THROUGH PENETRATION MODE OF CONCRETE BASEMAT): _ WILL MOST LIKELY RESULT IN A POUR-THROUGH PENETRATION 1
~
MODE OF STEEL HULL PLATE O - CASE 2 (UNIFORM PEliETRATI0i! MODE OF C0i' CRETE BASEMAT): WILL MOST LIKELY RESULT IN A UNIFORM PENETRATION MODE j OF STEEL HULL PLATE I i C)
,7 S V
O STEAM EXPLOSION (SE) BUBBLE DYNAMICS ASSUMPTIONS USED IN THE ANALYSIS - NO HEAT LOSSES DURING BUBBLE EXPANSION PHASE l NO BUBBLE LEAKAGE INTO BARGE AB0VE SE BUBBLE
- BARGE BEHAVES AS A RIGID BODY; NO PRESSURE RELIEF BUBBLE EXPANSION IS NON-SPHERICAL (BY VIRTUE OF IMPOSED GE0 METRY)
O _ OBSERVATIONS RATIO 0F lSMALLASBUBBLERADIUSINCREASES LARGE ENERGY CONTENT IN BUBBLE (BY VIRTUE OF THE CHOSEN INITIAL CONDITIONS) BUBBLE / WATER BOUNDARY VERY HOT O l-7 -/S'?
O ,0 O O
^
b 102 25% Core Mass {y[ [ 18% Efficient (Overall) w g 25% Efficient (Theoretical) y @ 20% Core Mass 3 - G.7% Efficient (Overall) 10% Efficient (Theoreticall 20% Core Mass 1 3.1% Ef ficient (Overall)
\ 5% Efficient (Theoretical)
D 10' D - N i _
~
\ l 10' t i' ' 'i1 ' ' ' ' i'i 10: 1G* 105 108 1/Q 3 'I'.7
. . , . -s a 9.. (i- g 3) e
O O O STEAM EXPLOSION BUBBLE EVOLUTION TIME = 0.0 SEC TIME = 0.2 SEC TIME = 2.0 SEC PRESSURE = 120. ATMOS PRESSURE = 13. ATMOS PRESSURE = 0.5 ATMOS RADIUS = 9 FT RADIUS = 3t, 1 T' RADIUS = 140 FT VELOCITY = 0.0 VELOCITY = 120 iT/SEC VELOCITY = 0.0 l I FNP #1 FNP #2 y 137' > < 120' > WATER LINE 32' Y Y 15'
////////////////////
O O * * '" 6 4 ^
SUMMARY
OF ANALYSIS TO DETERMINE MAGNITUDE OF FOLLOWING PRESSURE PULSES AFTER INITIAL STEAM EXPLOSION PRESSURE PULSE (CALCULATIONS ASSUME 10% OF CORE DEBRIS PARTICIPATES IN INTERACTION) PRESSURE INITIAL RISE RADIAL EXTENT LOCATION OF' INTERACTION PRESSURE SECOND PRESSURE TIME OF BUBBLE GROWTil BELOW PLATFORM PULSE PULSE (MSEC) (FT) (FT) (PSIG) (PSIG) k d 4.18 25 0 1019 135 v 4.18 200 0 1019 40 h 3.20 200 44 3370 95 "G e2 28. i ' Es ? EGm 5*-0 a l . m ,, t h ag
?"
Ea 8
=
4 _ - - - . . _ . _ . - . _ - _ _ . _ _ _ - - - - _ . - _ _ _ - _ - - . _ - - _ - - - - - + , -- -
+
O O O INITIAL PEAK PRESSURE = 1019 PSIG 5
= n-i /
~
Pn a. g s
~
10% OF CORE DEBRIS FRESSURE RISE IIr1E = 4.18 ftSEC g SUBi1ERGENCE = 0 N ~ DAMPING = 0 RADIAL EXTENT OF BUBBLE GROWTH = 200 FT. SECOND PEAK PRESSURE = 40 PSIG 0
)( ' '
O.f5 A ' A ' A O.'50 A ' A ' A O.'75 Tli1E (SECONDS)
'. O FLOODED COMPARTMENTS AFTER I CORE MELT-THROUGH 1
f EL l27'l h.
'. - l l ,. i
.. w-1:6.,c TF v ....a;;.
2
.[.
- A(
E.
.- 's,- 3G.n* , *. .** t..._*
* % *
- W . h **'
,i . '.
,. ~w .~ ;F'
-EL.l!7 IO" #'2 i
",i:7":r:i
- --y
_ >a .. 1
,- .;.yf.. : ir. 1; ~
f.Yh. REACTOR 'j - ,l-t
-1006 DECK .', Sc 4- .28i,' . ..
~
~
N - EL_ 'lOI-O' '.
!? '.?hk!S .
.r; b[5 e '
.G'*T' '. l.M.
. .. ' .*f.' 'p'y j
.TT*-
o! . Pi ,' a
- l. .;a p:.]e-....ne -7;y. o
.. .iIM ;A. a'sc E.T . 4..: W,. Q.
K d. -} C
, h. . g -. ., 55.y;r;. P. . . ,
~
i .. : .e ;y [ t t ,t t l t t t .- r:
.s r- Q \'
f, j.) , .
.y [EFE c,v..
51.0 AREA SQ FT }W)[ F i,n: 4 / ' E L 9_ 0,- O,, ___ r._i. . av
- :- g ...,
(T %) @r. s Q , ,
-a n.
w .., r a.:,,. c. y , ,
.e k; -
-,C .
EL.76'-9 . .e.i ,
. . .y.
h t - 7? .4%va.yM:;: g M yt.9 m::<.:n...:.,;'t' ': u EL.72 9__- .i.i>.* ::h;.gl+' 1 -
%cyg:?,rt.,X_Wg.Mw 7.g _g q- p , M,;"i m g g l M : OCEAN LBP 5.2 X 10 6 I FNP - 6 X 10 6
\ LARGE RIVER :0CEAN 5
h LBP 3 X 10 i SMALL RIVER 6 __
> ESTUARY : OCEAN LBP 7 X 10 g E
R 8 C O m? 24M 4 28 TE e C E of 2 .' E ,
- o
- l 8
,m -t
O PLANT-SITE RANKINGS BASED ON , LIQUID PATHWAY MAN-REM CONSEQUENCES ' WITHOUT INTERDICTION
- 1. LBP - OCEAN
- 2. LBP - LARGE RIVER
- 3. FNP - OCEAN INCREASING MAN-REM 4. LBP - LAKE O CONSEQUENCES
- 5. LBP - ESTUARY
- 6. FNP - LARGE RIVER
- 7. LBP - SMALL RIVER
- 8. FNP - ESTUARY O
W-/ C 2
O O O AIRBORNE RELEASE ESTIMATES FOR STEAM EXPLOSION INSIDE HULL l ASSUMPTIONS
- 3% AND 10% OF DEBRIS MASS PARTICIPATES IN INTERACTION AND IS FINELY FRAGMENTED
- ALL OF STEAM GENERATED IS ASSLHED TO REPRESSURIZE CONTAINMENT
- ACTIVITY IN DEBRIS UNDERGOING INTERACTION ASSUMED TO BE ENTRAINED
- ENTRAINED ACTIVITY REDUCED BY FACTOR OF 10 TO ACCOUNT FOR PLATE 0UT (AVAILABLE ACTIVITY)
\
- FRACTION OF AVAILABLE ACTIVITY RELEASED TO ENVIRONMENT SSUMED PROPOR-
.NI TIONAL TO FRACTIONAL. OVERPRESSURE N
NJ DOSE CONSEQUENCE O DEBRIS AVAILABLE AC- ESTIMATED DOSE MASS PARTICIPATING REPRESSURIZATION TIVITY VENTED CONSEQUENCES A P(PSI) (MAN-REM) - 4 3% 0.8 6% 10 5 10% 2.6 18% 10 ________o________ _ _ _ _ _ _ _ _ _ _ _ _ _ _
l
) 10,000 _
i
~
l
- 1 4,000 -
-/ I 2
/pl
/
2,000 /N # m g1,000 , 7 g 3a a 3b 5
/ /
e g 400 7 jj-/ r 2 s / ,
/
d 200 y j j ('m N <
/
? / /
2 100 . , <
/
/ /
1-ARAKERI (UCLA) 50 2-GIBBY 3-SANDIA 30 PARTICLE SIZE DISTRIBUTIONS OF GLASS /H2 0, Uo2/H 02 AND CORIUM/H O 2 EXPERIMENTS
,, .., ,,,, .....,,,, ,.,, .... .,, ,,,, ,, s.,, ,, . , , , , , ,
10 , , , ,, .. , . . . , , , , , , , ,
.01 .1 1 10 30 50 80 95 99 99.9 05 WEIGHT PER CENT FINER o
b
/7-/7/
l o DOSE CONSEQUENCES FROM PARTICLE TRANSPORT l 1 METHODOLOGY
- 1) ASSUMED 10% OF CORE DEBRIS IS FRAGMENTED WITH A TOTAL OF 108 CURIES.
- 2) ANALYZED PARTICLE SIZES OF 10, 40 AND 100 p WITH A DENSITY OF 4 G/CM3.AND UNIFORM ACTIVITY.
RESULTS
- 1) FNP - 0FFSHORE SITE: BEACH POPULATION AND MAXIMUM INDIVIDUAL DOSES ARE EQUAL TO OR LESS THAN WATER SOLUBLE CASE. SEAF0OD INGESTION POPULATION DOSE MAY BE THE SAME ORDER OF MAGNITUDE :
AND MAXIMUM INDIVIDUAL DOSE MAY BE AN ORDER OF MAGNITUDE HIGHER l THAN WATER SOLUBLE CASE. l
- 2) FNP - ONSHORE SITE: MAXIMUM INDIVIDUAL BEACH DOES MAY BE AN ORDER OF MAGNITUDE HIGHER THAN WATER SOLUBLE CASE. OTHER O doses AS roa Tae orrSs0aE StTE.
1 l O l M- n a
. _ _ ,_ /. . . . . . . . .. .
q .. . . _. . ..- . . _ . 3/29-30/78 g-APPEllDIX XXII l!cGuire 1 and 2: Project Status Report STATUS REIORT FOR THE MCGUIRE NUCLEAR STATION Introduction . l The McGuire Nuclear Station is located on the shores of $ake Norman, I in Mecklenburg County, North Carolina. The site is approximately 11 miles northwest of the city limits of Charlotte, North Carolina, and is located in a rural area, 'having no unusual characteristics. The plant is designed for an SSE of .15g and a OBE of .08g.The NSSS l and the initial fuel loading will be supplied by the Westinghouse Electric Corporation. The containments will be of ice condensor O type and the fuel will be the 17x17 R grid design. McGuire will be the lead plant with UHI. The ice condensor contain ment is similar t,o that used on Cook Units 1&,2 (Cook Unit 1 utilizes the 15x15 grid < fuel design. Cook Unit 2 utilizes the 17x17 R grid fuel design). The core design is very similar to that used in the Cook Unit 2 and Trojan. Westinghouse will provide the steam turbines. Tables com-paring the design features of McGuire to similar plants and some figures illustrating sone features of the design are attached as Attactment 1. The Applicant will act as the architect engineer and construction contractor as per the Applicant's usual practice. Construction was initiated on June 23, 1971. Unit 1 is now approxi-mately 93% conplete, and Unit 2 is approximately 58% complete. Fuel loading for Unit 1 is scheduled for' December 1978, with the fuel loading for Unit 2 scheduled for October 1980.
/b /
\ } .. l . McGuire Status Report - The Staff issued their Safety Evaluation Report ont he McGuire Nuclear
. Station on March 3,1978. -The outstanding issues which the staff has identified will be discussed in this report. A number of the outstand-ing issues identified in the Staff's safety Nvaluation Report have since been resolved. The Staff has in addition issued a Safety Evaluation
( Report on the UHI analytical models and appears to have reached an agree-ment with the Applicant on the ECCS analyses for McGuire. Outstanding Issues in the Staff's Safety Evaluation Report The March Safety Evaluation Report contained a total of 21 outstanding O issues, 7 oc which ere oueseendine beceuse the Seeff eve 1ueeton ce these f areas has not yet been completed. Six of these outstanding issues have l j since been reso1ved and it is likely that others may be resolved in the l near future. The current status of the outstanding issues are as follows: , (1) A Staff requiremant that the Applicant to submit justifiction for the use of augmented inservice inspection in lieu of complete pipe break protection - The Applicant is seeking exceptions to the Staff criteria (Reg Guide 1.49) for protection against pipe whip and wishes to substitute augmented inservice inspection for the installation of certain pipe restrains. The Applicant has not l (4 ,
/f- /W ~
i
. s .
O McGuire Status Report formally either identified the pipe locations for which excep-tions will be sought or submitted a justification. This informa-tion will be submitted by July 1, 1978. (2) A Staff requirement for the submittal of a summary of the dynamic
~
analysis applicable to Seismic Category I piping which would in-clude the location of all postulated breaks and stress comparison under design load combinations - This is a documentation require-ment which the Staff is requiring prior to the completion of their review. The Applicant will submit this information by October 1, O 1978-(3) A Staff requirement for additional information regarding low pressure over pressurization protection - This issue is now resolved. (4) A Staff requirement for the installation of leak detection capability for leakage from the reactor coolant system into the research heat ' j' removal and safety injection pump systems - The Staff is requiring that the Applicant commit to the installation of suitable leak detection equipmant. The Applicant will submit a response by April 7, 1978. g-/74' -
McGuire Status Report - l (5) The Staff is requiring that the Applicant commit to a submit pro- ) l cedures for the leak testing of eight dual bellows penetrations, l 1 l located outside of the secondary containment, which could cdn-stitute potential bypass leakage path - This item is now resolved. (6), A Staff requirement for a residual heat removal system interlock , 1 to prevent pump runout duri,ng switch over to the residual spray - In the event that only one containment spray train is operable, ECCS water will be pumped to the auxiliary spray headers. This
, will'be a.ccomplished by isolating the direct injection of the !
residual heat removal pumps into the cold legs, and diverting O this discharge to the safety injection train and the auxiliary spray headers. To preclude potential pump runout at least one of the valves in the cross between the discharges of the residual l heat renoval pumps must be closed before the residual heat removal I sprayline valve can be open. The Staff is requiring that these valves to interlocked. The Applicant will submit a response by April 7, 1978. . (7) A Staff requirement for the evaluation of water hammer potential - The Staff is concerned over potential water hammer in the ECCS (due to void collaspe in the lines) and the steam generator (due to void collaspe in the feedwater preheaters). The Staff is requiring that the Applicant provide additional infornation which will denonstrate ( g that damaging water hamner would not occur. g-/7c G
.--.-s
O McGuire Status Report . 1 (8) .A Staff requirement for a means of detecting of post-loss of coolant accident leakage - The Staff wi11 require that the Applicant provide information identifying all possible sou'ces r r of significant . leakage and the m2ans of monitoring this leakage. The Applicant wil1 provide a response by Apri1 7, 1978. 1 (9) A Staff requirement for the evaluation of ECCS during off design conditions - The Applicant's procedures ca11 for-the blocking ; of the safety injection signal during p1 ant cooldown and c1osing
-and locking out power to the cold leg accumulator valves during shutdown operations. The Staff is requiring an analysis of the O the zCCS verformance for ehese coaditione, end wi11 review the effect of implemanting these procedures. The Applicant will provide a response by April 7, 1978 .
1 i (10) A Staff requirement for additional information on the qualification i of Nuclear Steam Supply System Equipment and the balance of planned Class IE equipmant - The Applicant has stated that the seismic and environmental qualification of equipment will be in accordance with the WCAP-7744 - 1971 and WCAP-7817 - 1971. The Staff has reviewed these topical reports and has found that some procedures which are specified are not acceptable. The Staff will require that all I clements of the Applicant's qualification program be acceptable, f Resolution of this issue is expected in the near future. i 4 (11) A Staff requirmant for additional information on the offsite poser i
. system design - This issue is now resolved.
77 l m
O McGuire Status Report I a (12) The Staff is requiring that an alarm be installed on the non-isolation of the unborated water supply during startup or
- shutdown - The Applicant now uses information from the neutron l
- l detectors to indicate the unintentional injecticn of unborated e
water into the primary coolant during startup or shutdown. The ! Staff is requiring a more direct indication of the satisfactory I isolation of the unborated water supply. The Applicant will submit a resonse by April 7, 1978. (13) 'A Staff'requirment for the evaluation of accident trips for new trip items - The Applicant has changed som of the technical O specifications on the trip delay times. This issue is now resolved. (14) The Staff is requiring additional information on the steamline break accident - The Staff is requiring additional documentation which is intended to clarify the Applicant's previous responses on questions on this item. Revision 28 to the FSAR has recently been submitted and should resolve this issue. l The following issues are currently outstanding pending Staff evaluation: (1). The evaluation of Unit 2 reactor-vessel fracture toughness data - The Staff's review of this item is not yet completed. There appear to be no unique problems in this review. f (2) The evaluation of conformance with Appendix K to 10 CPR Part 50 - 4 These are ECCS/UllI items. The Staff has issued an Safety Evaluation jy-/ 7 Y
O McGuire Status Report . Report on the ECCS analysis for plants with UHI. It appears that this ECCS analysis for McGuire will show conpliance with the Appendix K. (3) The evaluation of the documentation of the test program results of the electrical penetration qualification - This issue is now resolved., (4) The evaluation of the Applicants justification for shared elec-trical paher supply system - This item is now resolved. O (s) rne eve 1ueeton or ene rire erotectioa ^aetrete - zue stert rire protection review has not yet been completed. l (6) The evaluation of the Industrial Security Plan - The Staff's review of the Industrial Security Plan is not yet completed. It is expected that this issue will be rsolved by mid-July. (7) The eveluation of financial requirements of the Applicant - It is l l the Staff's practice to not complete this evaluation until time near the end of the review to assure that the nest current information has been used in the evaluation. It appears that the Staff's con-l clusions will be favorable. Completion of this' review is expected by mid-May, 1978. l 1 l p-Q , jf-/if
.__________._________.m______.___ __._.____.________________________m.___ _.__.]
McGuire Status Raport u.; - Action on ACPS Generic Matters The status of the Applicant and the NRC Staff action on the ACRS generic items is as follows: (The items which are marked with a asterisk are recommended for inclusions in the generic paragraph. Grcup II: (1) Turbine Missiles - This items is resolved in that the facility has a peninsular turbine arrangement. (2) Effective Operation of Containment Sorays in a Ioss of Coolant Accident - The McGuire Plant utilizes a ice condensor containment. Sodium Tetraborate has been I added to the ice makeup solution to enhance the iodine ' O absolution characteristic of the ice. The technical l specifications will require a minimum ice ph of 9.0 l whenever the reactor is critical. The containment sprays will use borated water.
* (3) Possible Fracture of the Pressure Vessel Post Loss of Coolant Accident by Thermal S.. - This item is not resolved for McGuire and is under generic review by the Staff.
* (4) Instruments to Detect Severe Fuel Failures - The McGuire !
Station utilizes gamma monitors on a hot leg sampling line. The adequacy of this type of instrumentation
, to detect failures associated with very rapid events has I
not yet been established.
//-/ P0 l l
l
e Mci ~ Status Re, port * (Sa) Monitoring for Icose Parts Inside the Reactor Vessel - The Applicant has conmitted to a installation of a loose parts monitoring system. ,
* (5b) Monitoring for Excessive Vibration Inside the Reactor Pressure Vessel - The Applicant has not comitted to either the installation of equipment for neutron noise analysis or the installation of vibration detection equipment on other reactor pressure vessel components, in the event that the usefulness of such devices is established.
O * (6) Non Random Multiole Failures - This item is unresolved for this facility. ; 1 1
* (7) Behavior of Reactor.Puel Under Abnormal Conditions -
This item is unresolved for this facility. (8) Boiling Water Reactor Recirculation Pumo Over Soeed During a Loss of Coolant Accident - This item is not applicable to the McGuire Nuclear Station.
* (9) The Advisability of Seismic Scram - The Applicant has not proposed the use of seismic scram for the McGuire Nuclear Station. The Staff has indica':ed that they do not intend to require such a scram on McGui're.
p#
' McGuire Status Report .
'(10), . Emergency Core Cooling' System capability for Feature Plants - The Staff has indicated that this item is .
under generic review and is considered to be unresolved for the McGuire Plant. The McGuire design does however utilize ti e 17x17 R grid fuel and ' upper head injection. Group IIA:
* (l) Ice Condensor-Containments - The McGuire Plant is.the second ice conden'ser containment station to come before the Comittee for a operating license review. (D.C.
Cook, Units 1&2 was the first). The Staff has developed some capability for performing a independent analysis , h of the ice condensor performance but have not used these analysis tools on McGuire. Results to date indicate a favorable comparison with the Westinghouse calculations for D.C. Cook.
* (2) Pressurized Water Pumo Oversoeed During a Loss of Coolant Accident - The Staff has indicated that this matter is unresolved and is under generic review.
* (3) Steam Generator Tube Leakage - The Staff has indicated that this items is considered to be resolved in part by the requirements for inservice inspection. The steam g_ / P 2,.
. l
= - h
...M m
c] McGuire Status Report generators used in McGuire were manufactured prior to the implementation of the latest Westinghouse steam generator design fixes.
*(4) ACRS NRC Periodic Ten Year Review of All Power Reactors -
This item is unresolved and is under generic review. Group IIB: (1) Computer Reactor Protection System - This item is not applicable to the McGuire Station. Devices of this type
- . are not used at the McGurie Station.
* (2) Qualification of New Fuel Geometries - The Westinghouse 17x17 fuel assembly is to be used in the McGuire reactors. A number of tests and surveillance programs are ongoing at this time. The Trojan Reactor a full 17x17 core loading fuel. Corrmerical operation of ,
Trojan was begun on 5/20/76. j (3) Behavior of Boiling Water Remtor Mark III Containments - This item is not applicable to the McGuire Nuclear Station. l (4) Stress Corrosion Cracking in Boilino Reactor Piping - This l 1 item is not applicable to the McGuire Nuclear Power Station. Group IIC:
* (l) Incking Out of Emergency Core Cooling System Power Ooerated Valves - The NRC Staff has accepted valve lockout and the l l
/Y-/ 0 1
l
l
, McGuire Status Report \ '
, 3 l
1 administrative controls 9stablished by the Applicant for the McGuire Station and considers this item to be' resolved on this basis. This does not seem to be con- ; sistant with the Comnittee's position on a acceptable i resolution for this item. i
* (2) Design Features to Control Sabotage - The NRC Staff has not yet completed,their evaluation of the applicant's Security Plan. The methods planned for the McGuire
' Station appear to be consistant with the current state-
~
of the art.
}
l Q * (3a) Decontamination of Reactors - This item is unresolved and i is under generic review by the Staff.
~ 4
* (3b) . Decommissioning of Reactors - This items is unresolved and 1 i
under generic review by the Staff. l (4) Vessel Succorts Structures - The load analysis has been per-formed for this facility using the approved Westinghouse nodels and the structures have been found to be adequate. The Staff has concurred in this analysis. This items is , 1 considered to be resolved for the McGuire Nuclear Station.
* (5) Waterhammer - This item is unresolved and is addressed in item 7 of the Staff's outstanding issues.
~
/fW
P f McGuire Status Report .
*(6) Maintenance and Insoection of Plants - The NRC Staff con-siders this item to be resolved for the McGuire Nuclear Station and that the Applicant is in compliance with ;
current NRC requirements. This does not appear to con-sistant with the Coranittee position on ' acceptable reso-lution for this item. (7) Behavior of Boiling Water Reactor Mark I Containment - This item is not applicable to the McGuire Nuclear j l Station. Group IID: O (1) seretv ne1atea 1nterreces setweea the neector t=1ead and the Balance of Plant - The Staff has indicated that thisitemisnotapplicabletotheMcduireNuclear Station since the McGuire Station is a custom design. ' Duke Power Conpany acts its own architect engineer and construction contractor and has this advantage in handl-ing interfaces between the NSSS, vendor supplied systems and the balance of plant.
* (2) Assuance of Continuous Lono Term Capability of Hermetic Seals on Instrumentation and Electric Eauipment - The NRC Staff has indicated that they have not address this item in the Safety Evaluation Report, except as general requiremnt for environmntal qualification of equipment.
l~f
.g
McGuire Status Report . The Subcomittee may wish to discuss this further with the NPC Staff. Group IIE:
* (l) Soil Structural Interactions - The NRC Staff has in-dicated that this item is currently being evaluated as part of the Task Action Plan A40 " Seismic Design Criteria". This evaluation is scheduled to be com-pleted by September 1978 and could lead to the nodification of current criteria for seismic input
'and soil structural interaction. The foundations for nost of the mjor structures at McGuire are on O sound rock.
1 Intervenors/Sigriificant Differences of Ooinion Among the NRC Staff The Carolina Environmental Study Group has raised a concern regarding the probability of a major earthquake in the eastern part of North Carolina, and-has cited anomlous changes in land elevation and ground-water behavior as possible predictors of such a carthquake. David Stewart, David Dunn, and S. Duncan Heron has raised this issue on the Dunswick Docket and has reported that such anomlous conditions exist in the vicinity of Southport, North Carolina The Carolina Power and Light Conpany has been operating a micto seismic network in the Southport area for approximately 1-1/2 years and have identified no local carthquakes. The historical record g-/7L t ._ J
McGuire Status Report for this area indicates a very low level of seismicity. The McGuire site is approximately 200 miles from the area of the postulated anomalities and is in a different tectonic province. ; The Carolina and Environmental Study Group had submitted a written statem nt to the ACRS on March 6,1977, (J. Riley to ACRS, see Attachment 2) during the Comittee's review of the Perkins/ Cherokee application. This material as distribued to the Comittee during its review of the Cherokee /Perkins application. The Comittee appeared to be satisfied that these issues had been properly addressed by the Applicant and the NRC Staff. Mr. Riley raised these same issues at the McGuire Subcomittee meeting. O
~
The NRC Project Manager has informd me that there are no significant dissent ing technical views remaining within the NRC Staff on McGuire. The questions as to the testing of the large circuit breakers are believed to have been re solved. Robert Pollard was formally the NRC Project Engineer on the McGuire Plant. The NRC Project Manager does not know of any specific dissenting technical views which were expressed by Mr. Pollard on the McGuire Plant.
/V-/9 7
-f4
l n . PAGE 1 U . DESIGil PARAMETER COMPARIS0fl 1 W. B. McGu1Re .
"1 Af1D"2- D. C', C001c 2 TROJAff ^
DESIGil PARAllETERS l 1 l!SSS POWE'R LEVEL, E,WT . 3fi25 3403 3423 l REACTOR COOLANT PRESSURE, PSIA . 2250 2250 2250 REACTOR COOLANT Flow RATE, 140','3 1346 132 E 106 ggfgg REACTOR COOLANT IEMPERATURE, Op O 5581 541','3 552'.i VESSEL }llLET 588.2 573.8 584.7 VESSEL AVERAGE 618.2 606.4 616.: VESSEL OUTLET STEAM GENERATOR STEAM IEMPERATURE, 0F 544','6 5211 533.: l 1000 820 910 l STEAM PRESSURE t PSIA STEAM FLOW, 10 6 La/HR 15.14 14.74 15.C7 4 [ U
~
j /? Y
i PAGE 2 DESIGli PARNIETER COMPARISON l l l i W. B, MCOUI RE DESIGil PARAtiETERS "1"AND"2' D.'C.' COOK 2 TROJAN ~ l l MINIMun Di!DR AT Il0MIf4AL CONDITIONS
. 1
~~
TYPICAL FLOW CHANilEL . 2,08 3,03 2','04
~
l THIMBLE (COLD WALL) FLori CHANNEL 1,74 2',70 l',71 ! I MituMuM DIlBR FOR DESIGN IRANSIEtlTS 1 1,30 2. l',77 > l','30 l Dl!B CORRELATION "R" (W-3 WITH WRB-1 " R" (W-O MODIFIED SPACE WITH M: FACTOR) FIED S FACTOPS AVERAGE THERMAL OUTPUT, KW/FT 5','44 5,41 5.44 MAXIMUM LICENSED THERMAL OUTPUT 12,6 11,8 12,6 FOR I!ORMAL OPERAT10tl, KW/FT l
~
LIMITING F a VAtuE 2','32 ~ 2 ','18 2,32 1 f~/ .
}lJ
- v. ;
DESIGN COMPARISON' ':e l O W,' B. McGUIRE l Basic COMPol!ENT ' ' l AN D 2 ' D,. C, C00K'2 TROJAN 1 l l REACTOR VESSEL, ID, IN 173 173 173 l l C6RE NUMBER OF ASSEMBLIES 193 193 193 i Ron ARRAY 17 X 17 17 X 17 17 X 17 ROD OD, IN O',374 0','374 0','374 NUMBER OF grins 8 R TYPE 8 R TYPE 8 R TYPE ACTIVE Fust LEl4GTH,, IN 144 144 144 i NUMBER OF CONTROL RODS ! l FULL LENGTH 53 53 53 i PART LEl4GTH 8 8 8 - OTEAM GENERATOR l TYPE PREHEAT FE EDRil4G FEE D'RI NG : l 1 SHELL DESIGN PRESSURE, 1200 1100 1100 PSIA REACTOR COOLANT PUMP TYPE 93A 93A 93A MOTOR ll0RSE POWER 7000 6000 6000 6 O- 1 I
/Y~~ /f 0 ,,
SYSTEM AND COMPO IT DESIGN COMPARISON . OGE1 SYSTEM / COMPONENT SIMILAR DESIGNS PRINCIPLE DIFF$RENCES TROJAN, COOK 2 AND SEQUDYAH 'NONE Fu2L SEQUOYAH AND IROJAN MCGUIRE AND SEQUOYAH WILL HA'VE REACTOR YESSEL INTERNALS UPPER HEAD INJECTION SYSTEMSJ MCGUIRE AND TROJAN WILL HAVE NEUTR0f; PADS; MC6UIRE.ATID SEQUOYAH HILL' BE MCDIFIED FOR INCREASED INLET FLOW BYPASS N Nl- - TO THE UPPER HEAD. , COOK 2, TROJAN AND SEQUOYAH NONE w REACTIVITY CONTROL N ~ COOK 2, TROJAN AND. SEQUOYAH N0flE-NUCLEAR DESIGN , COOK 2, TROJ AN AND SEQUOYAH COOK 2 HAS USED THE WESTINGHOUSE! THERMAL-HYDRAULIC EESIGN IMPROVED THERMAL DESIGN PROCEDURE 7 COOK 2, TROJAN AND SEQUOYAH NONE _ ',' REACTOR COOLANT SYSTEM , COOK 2, TROJAN AND SEQUOYAH MCGUIRE AND SEQUOYAH VESSEL REACTOR VESSEL HEADS INCORPORATE UPPER HEAD 9 PENETRATIONS;
/4
Q SYSTEM AND COMPQNT DESIGN COMPARISON PAC $ i SYSTEM / COMPONENT SIMILAR' DESIGNS PRINCIPLE' DIFFERENCES REACTOR COOLANT PUMP COOK, TROJ AN AND SEQUOYAH MCGUIRE HAS HIGHER COOLANT
~
FLOW DUE TO CHANGES TO IMPELLER, DIFFUSER AND INCREASED' MOTOR HORSEPOMER STEAM GENERATORS COOK, TROJAN AND SEQUOYAH MCGuIRE WILL HAVE PREHEAT , STEAM GENERATORS RESIDUAL HEAT REMOVAL COOK, IROJAN AND SEQUOYAH NONE N SYSTEM N y ( PRESSURIZER COOK, IROJAN AND SEQUOYAH .UONE CONTAINMENT SYSTEMS COOK AND SEQUOYAH ECGu1RE, COOK AND SEQUOYAH f HAVE ICE CONDENSER SYSTEMS EMERGENCY CORE COOLING COOK, TR0JAN AND SEQUOYAH MCGUIRE AND SEQUOYAH HAVE SYSTEM UPPER HEAD INJECTION SYSTEMS
.f REACTOR PROTECTION SYSTEM FUNCTIONS ARE SIMILAR TO . NONE
~
COOKg', TROJAN AND SEQUOYAH Il 9
_ _ _ _....___=.. _ .__.__ _ ._ . _ . _ . . _ _ _ _ _ _ _ _ _ . . . . _ . _ - . _ _ . _- SYSTEM AND COMP 0 T DESIGN COMPARISON PAGE SysTEv/ Cone 0NENT SIMILAR' DESIGNS PRINCIPLE ~ DIFFERENCES
- - . ENsINEEaED SAFETY FEATURE FUNCTIONS ARE SIMILAR TO RONE ACTUATION SYSTEMS COOK 2,-TROJAN AND SEQUOYAH ELECTRICAL SYSTEMS REQUIRED FUNCTIONS ARE SIMILAR TO N0bE FOR SAFE SHuTD0ict COOK 2, TROJAN AND SEQUOYAH SAFETY RELATED DISPLAY FUNCTIONS SIMILAR TO COOK 2, ACTUAL CONFIGURATION MAY
}NSTRUMENTATION TROJAN AND SEQUOYAH DIFFER i
N CHEMICAL AND VOLUME CONTROL COOK 2, T"JJAN AND SEQUOYAH MCGUIRE WILL HAVE BORON y SYSTEM THERMAL REGENERATION N. N
*e I
b e O
. . . = - - . _ - - - -_. - - _ _ _ _ _ _ . - . _ _ _ _ _ _ _ _ _ - r - - m
MAJOR PESICll CllANGES SINCE PStR . . o CHANGt IN TESIGN REASON t i FUEL ASSEMBLY DESIGN CHANGED FROM IMPROVED CORE THERMAL MARGift 15 x 15 TO 17 x 17 ARRAY UPPER INTERNALS MODIFICATION ACCOMMODATE 17 X 37 Allb UPPER liEAD INJECTION THERMAL SHIELD HAS BEEN REPLACED SIMPLIFIED CORE SUPPORT DESIGN BY NEUTRON PADS AND REDUCED PRESSURE DROP t.MD VELOCITY ! ICE BASKETS, SUPPORTS AND LATTICE ACC0l'.M0!'AT E DES IGN LOADING FRAMES FOR THE ICE CONDENSER CONDITIONS BEYOND THOSE HAVE BEEN REDESIGNED COMMITTED IN THE PSAR g]UPPERllEADINJECTIONSYSTEM IMPROVED PERFORF.ANCE OF THE liAS BEEli ADDED EMERGEllCY CORE COOLING SYSTEM STEAM GENERATOR CHEMISTRY CHANGED PROBLEMS ASSOCIATED WITH PHOSPHATE FROM PHOSPHATE TO VOLATILE AMiffE CHEMISTRY TRE ATMENT -
..'n..,r.
.n..
. G . . ,
SEMI-AUTOMATI C SWITCil0VER TO MEET !!RC REQUIREMEllT TO REDUCE ECCS RECIRCULATION HAS BEEN RELI ANCE ON OPERATOR ACTION PROVIDED AS BACKUP TO OPERATOR ACTION O R- / 9Y y
. w
1 fSGE1 Ov i 17 x 17 FUEL SURVEILLANCE PROGRAM i l SURRY" DEMONSTRATION PROGRAM l 0 (2) 7-cRID 17 x 17 ASSEIGLIES 1RRADI ATED IN EACH OF THE l TWO SURRY UNITS ! l 0 FIRST AND SECOMD CYCLE EXAMINATIONS COMPLETED FOR BOTH j UNITS lilCLUDES TELEVISION, ROD BOW MEASUREMENTS, P ROF I LO'1ET RY, GAMMA SCAll, EDDY CURREllT, ETC. o SURRY UNIT 1 DEMO ASSElGLIES DISCHARGED SURRY U;i1T 2 DeM0 ASSEMDLIES n0u uriDEao01iie TslRD CYCtE OF O' 1RRADI AT 100 (EXPECTED DISCHARGE DATE: MARCH, 1979) 17 x 17 FULL CORE EXAMINATIONS . I O VISUAL EXAMINATION OF FIRST TWO 17 X 17 FULL CORE REACTORS TO RELOAD FUEL) 1.E., IROJ AN AllD BEAVER VALLEY UNIT 1 0 EXAMINATIONS TO INCLUDE DESERVATIONS FOR CLADDING DEFECTS, ROD DOUING, CORROSIOM, CRUD DEPOSITION, AND GEOMETRIC DISTORTION 1 l n U l R- /p s y:. \
.- .. . .. .. . .- . - . .-. .~ . . . . .
.
- I tL 9tA' .$
f =.=t- 1 , l a y e% 3*,,, ,
/ \t l' .
I , v, r 1 1< 'id.. _ .
.f .
l QI y s, . y E
.~ - ~qf . ,
. j ii .
i
) N.
i M. ' . . . . Sf 6.' ... . . r R,l)j. t.w -
.t i
'\ . .
-1
-<. t ca w.-to r
-w 3
. g!-
.. i '
4 .
. > - sic.u
' C(N! A 41Qft
.u.y,u
.d- n,., pw
',, .,.\/ p/. _.ip_s E I-I -- .t:1,;n -;
. . , ; . r-, t
.. . .;L,, ;
)
i
.' 1' JLN 1-LL . ., .
., as ,can
== p,p + 1 mP 3
t.4 8-( O _j,
- l Y. ,' .
=r , r4 l .,: ~ - -
-j,
<773
~. g't -
.LIJ.L%L.
( ,.n , j
/ .' . at,c w vtut6
'. 1
- 1L'u;L
": i 'r =
. ; , J_ L =
i
, ' * = ,L J .
g.
. *.'. j vW~nW.-v 1
' ~ ~
... I
* * .stsu.(
l cectn. . .touw tc. uA r.e,rnff rtrfn,. k.l.'.S.2OL!U_^U?.$.!'/ .
. . . . 9 MCCUlitC lluCLt.All STAT 10tl l
/7- if L ' ' * ~ _m, . _,__ ,
\
e f'~"% ( ) R.; VIRGINI A TE NN E SS E E e"4 ~4 [ g- v ..
,o
- G O.& SITE s -
\
1 e *
, . FIORTH CAROLIN A ;
__ f f m q g p , ll , l 1 Q L $, . ' .. .,.< ::; i 1 4 W .. .'
. ie , r #p.. s s..
.' 1
%f*W ,. . . '
- . ;;o p p'; ' g- .
v j' -._; _. --) , - e yQ
,s ,: < r. ; r%'T .;7. G* * ,
*r. ,Q
,m, GEORG1A
( ".:'1, $[ Ye2[.':- ' f.,? - 0 V, l (,1 v,, ,q7' SOUTH CAROt.lNA -f ' , i
< ,.-~..:- ,. . u u x, .u.<
, ; ,n, ys.
s
- >g$z .
I Q, "#hl'~.. {l ' &' '
$ "~ h, ~,.'. :
- 9
-i.j:? * - .
N
.l 5
50 0 50 100 bd.sm# - .= a,.7m APPROXIMAT E SCALE IN MILES DUKE P0',,'ER COMPANY ERVICE AREA
' .:n M.
mi rreo
?. McGUlRE NUCLEAR STATION
$*r-
. ,, .es-~ .
/: /- . - - . . , , .
... . . c.
. 3.. . . , ., i<
- y. s. .
~.
s
.. . p .. ... .,. \ 1 n t e -
e,9<. *
\ei U /e.r .-
m-
,. L ,sa, ,
' /e
}
r "vg aNJ /
\ * ?.-c '**
p, ..,';.
.l 9 ~
' a ,#
"f[
GAltrifyiLLE - j : ! fg~.8[ @f.7 s .' m f... ~ o t
.'p;,s gl%. ' '
Ik.-hp.d.,uu.oduf[..:.df@fU( s
.1 p:LM se.s nr 5 ww w c .m .
i
.... S h i p .'.W.S [Wb' %.(h
- n. (~ Wy?..fl}.f.k"ay'$k r
)kl . . . .Q~ .l f.
- r. :,,
.n,
;e. .s . . .- ..
> $' Wp
*; j .". .
,' . :y ;' j . Qs, t ,-- d V: N 3.Q,., -),'g, .,,.g.
.. s . f
- Q". N ,$ '/. . .
3
. ,s m .,.,!#. f. .
.~
\ *::..?' s.ffj. %c --y:c.n 'M, ji w -
4a
.. .....: w' % ,
,i h f' N Uk{ #., '?.i d ".'i
- h %) h k :~',.'l W L s ' ' y.
irt we, 3 - . m#,3:3S. -/ ta!n:: : :. ~,
~.6 c..m,,. a. m.. i .
w .
. v
.?.. . ,
.e .
! / . t , ~, t . u2,ec/w' r '. . 0, * ;f. .' .,
s,.
- . 4
,, - ~
;.- .y .s : -(. =< ... -
'.u< -
- r. ,. i
<> ,.4 . .s v .;
. u . it.,w..y. e: .. c. ,, - ..- , , -
cu. .: .'
..1, m .a:~.6:. ,cJ. , /
b m z .
-.m n
.I Qusf..,i '% l{ ., ' ' . . ' . .
.m. , i ..q. . ; w:...,s <. s V. ,a.
. m .
.?
ul .m - . .
*. M e.,6sM ,; CJV/ .2;.\. " 1 5
; y .
. ,.; .r .s_ ....a-
- . w =* , p - /,
p .e . . s . s. :
' h..chft s
44o n ,. %r M
- s,
, ~,e M..
g a.t.tlu d 4.L u tt u k
- du u u.L
) a 'ef ..
.x n,s ;- s : q ; .
n
- v. e ..s f .,h NEd , ,),' MW 5. %, p -
.- . mTATe a & .. .a .
< .r v1H
. . , . e.-o , . * ** ( . - ,
w. Jauf. m .~g - - G'h i ) C.% * .l". wvw ; -
/ /
l
/ . -
1 w= i . m
, ,\. ,
s-a -
/
.c ,, ..
t # . s / V; 7 i N AN00
- s'. ._. ,.
s<, .o re z e **
- hMe '
y bysm 1 p e[
'f,n ',, V 1/ ..,/
/. .-
5 D, .,.a ., ..in,.i.j,',;,y,7 .. < v },M.- t
.; f; O ' '>
g*y %s'."' t,' p.' d
. , , .. n
,f,lt< s. . :
'A.w
~ ;'. ~ _ --
. . .J f.,t
- /
a.'(% N2 r.gf.*g.:~,a
.. M -:-
. Q4 / g;\t.*,. w a ,
3s . ouce ,,
\ /
- f. . . .
'M a
- f. . .
_ f i C 4". m
\
c,Mytono - su. w on 3 l
- 1. '
9 el-' N. v., J.p.x wie o . d kuntgj [ b iYg
~
d ,[q ,
' Cad Dj,J,,[m-Q[
5 hjudt.Qfb
- c au.[ t> h p .,F ,cuww ;
v4 ia v .
$ s) A %; . \'
I e ,. . . cwnwu s g\; Am t Mgf,j /e'^ ,m())% - - ~
*\
t
,- l i,on ' p ,
t s '-t- , s .- 37 ,, p or Td
. !'. (.. ' . ,.
,. *S/{ } ,.7 l (
'l S
7 <-
- 3- .
,O-wg,on; ;
n i
' N gs *] \, ,. .~ e ~~
t g i
]
~ '
I
.i '/
/L Rf*3s%
1 E ~~. f ^* : - so , / . . .f- k i ;~'p;;.-/ % -- a
--- m . ,
c ,,, ,,. mvua3 A
~'5-.
l 1 1~ 3 n 5,
,. - l ;.-
t j e
-~ m--q
'- ... a . - ,
, i i 9e . ,.# , of,.nf ; e sf> M /'...'./ , .
N. ! s n u, ' ,.
.~ \
y -. \ _l /,', - f j , . *. .' /... s ,
>J.,1, N i l m ; sh ..'s j ,. q-Ny,u {:
,[NC. , " , '
~'
i.b . p . ,f f/ s uu.L y ,, I ~ L
.[,. &. C.
. . . y,v ')
a,1 i vk, . ./ h_ f l -~ l N,
.I Y\ T'n. s:d-4 I
g 3,. , t% i uso ., .# c
% g_ .,+. 4
\
w fe m u.cassamen5., Ae ,in Cass W Goedsac Seney j- ,[fH[ TAW
- . t d"I % SCALE 1:1,000,000 g.w i 'e 4 f.ONToua inittvAt .c-g ( ,/ ,
% ~
fb b;ee, HM ' HtCHESI TiktAIN ele.oben .4 ' 'C[.'
;. 6634 feef ' 4,i,,J Wia.d .* 15*4d'H .42 te'w lh
'] *
. . . . stSt t
'..~.-..........-(,Y / y Cosmaele.enap,,,, % ,w,,,,,
. , , , , , .sm I**
n..,e,n d t.y w on
"5 f p- ,,' f
. . . -'. AettWON, ., ' .Dowtet6
, % p.,,,, w,, pwoe ,, .. ...; j- saw M a
'- .assA qquc o a s A DC A M AD
*y' M 7 pc t .,g. ,. e.,e ry , e' . s. ' 'u , . u , y 3
,8 % ', Y-;-!?'hp 'y t ,. ("y '4* g , ,
'S '
S q' Gjs
,.. . .i' ." *,~':.9, .),, rp r;.y, ~ ,i:..>'L ., . .;. ,.l,',* ls f y .'. 1 l ,., <.<y ,.afl,'s~/:"~..
- .. - l
~l i ,
4 Y E., sf un x& 3 n& s ,en. y. - l-q - . x.. \ .,9 . [q , ;p t
- Jt:%'dr.%A' *,,lm RYn ,,, ' ~ $ ' . W .,', l. X t . ) v'* Q. ~,'q {\ *'Q%(*'Vp. Y 'v ;.. .
f'r!Y p'I; T wg9, .lh, 0 'Q .j
;)& .. fh5 ?{.fkV','h f0,,*, } uv.es f { . .> ::% ,,
.. . : . & [ ,,
,i t < / ', ..Y -l, Q'
.a y -
m{c$$b . c . s.! &pkf ()/A. p m ~. - 5~
,,as
{O h,,. x : W W;:. ,.W. / pa - u si<.3@.d.Q,.ih,);.g,A"",s,pn,= . , us 1
'I""' "f"%, "it ""'"
%..3 , Q i 7(h, I' ' h*$w.,f tt eC -
$- f*
5' , o .. . ~__,l llf, l _ ,*h . ,'.*)3g, t 7 )4 , ya \ 1 -(
\ o f3 ,!
, , . ko s,,7gy t, t
6.:;,f, .,mn,i.,, , x.r ,,a .v . . s' 4'm< + ,
. M'. '" ' ,[* '. ,#
Q , N - g . 7 - sa .,/ . << % , _ ). . .\ k\ /,--
~ no .'
i .,. - s
, s ..s
}
. ,f'f *- [ ' j ,/ E
[
-a _..
e
,* j ,.' ,,- ,
e <fA608f ( h b*.(Ot y -. .as f
.,'. Y ,h.
" _ . .' .. ./.
.ssf.
, m .-
-1 .. s. F :
% ' .,,,,o /
/
gg;
'.-,. ye>-
5^' .,,d 1-
/, j ,,
;vy' d 3 /y .-4
.c,. s ~ ,
,o p
. . .x . , .
3- -
, k%.., ,[ [ U-D l . :;. l %. ow' '
w[-\!? '
.r 5 b,Z' ~e* ~ "*
. , ,3'
. 80/;r
~~~ . -- f,'!.T A (f4 f- - .
p .'. t ' '!
-\:T
?l f -
', O ' ' k u , i tse j 't 17s *
** [,,,f *(g ,.. (7 t. I N O E{U1 %',,./,/~
.+w
.x <
y/ s.p... m n.n .ca w.35 o .+pt4w+p . ;
.. n 4 ,. 4.a W. m T--.- q -L O \
4 - (.
, ,/
/j *P34*f'L", x -
< \
u! .J '*' si I
,- g
! , [CETH AtCUN
,' y %. O < ; wuu. caw 62 , s to3- : -
.m
, N d '*h *M a ,. Qf 8 - - -
=
4',. c.- , :J ,g wi.. ... s 'w ( ) ,/ .,, /
,/ ~
p< _
<..~
tm
- f s Y- M, <
1 Is 7 %. w.wt o o g *.
-i t,....j ....-
- l. .
L
- 4 IM h . _____ dh m .
< .N POPULATION CENTER DISTANCES WITHIN A 100 MILE RADIUS A
V 100 MILES y ' VA JOHNSON CITY
- g, -
7 5 M TI E 3 ~ ' - ~ ~ ~ (35,770) / ( r= WINSTON-SALE!.1 nt GREENS 00RO 50 MIL ES (1323 ) h 044, ors ) g ,/ d.RLINGTOtJ (35,338) HIGH PO(NT 2 5 MILES (83 4 ' ASHEVILLE E KANN POLIS SdE 4'* f %, CONC o_ GASTOfilA
~
p ~ CHARLD TE SPARTA NBURG (44,540) O* g Q p ROCK HILL' (33#4c) 1_N.C.. _ __ GREENVIL LE
. (6l,4 36) \
s N COLUMDtA (113,542)
. D SOURCE: 1970 CENSUS OF DOPUL ATION POPL'LATIO't CE!!TC,7 Ul STAT:US Willlitt A 100 f tlLE RADill5 i(
O troisih McGUlHE NUCLEAR STATION OGY'
'~
r i ue re z . .i . .i- i
/4 - / 9 9 -
/
i '~ i e> , ,
,9,, , . ,4
,a,l, , , + ' k'?
,.ge , 4 '-)- .,
jj ,4 , I' -,
- g
. 5 '. 6 e J. * 'ht , ,, ,
,lg
, . ; aa 5 ! 'VE * '
< T h ,', ', ";, ~~** ' *#
's O j ,
t[-
. i. .
'[, .' +( *
- . e , <
,, . r ,
jj.! - ; y e 9 j 4 "i=,' ,I i,
. , ,#E{'*- #
-.E
~ ..
.* y n p . +
4 g
* *r I 4 e p %
y, , , + ,
,eL # E. -p$ # #
3 4, s ,, ,(
+ %
& y 4 s' .
( , , . %' # a ..r . g e 3
+ a -
. p- -8 9
y . s , Y. .
. ; f r . .
~
gE
/ J %
*t s _ _i - ....
,/ .
t - a
, esc >.uk rwm h "A Si 6mm que an 3
%h- Aj'f /
. y e-l ;
Q' - 4 .
$d '
xs/N i .n
\.
i ss , i i cm en e g iramws n. w .! 4 l I
\
f . ..
\, ..A s
m..
\.
N lV.c
, aamu
.v
- y. ,
y
/
m
) _s i =c,.
y o , i e D 4 I $ e
- A 9
*f ,
y
- O 4 '" 3 1.- 0 WW . __
M P%r% w
~acaJ e n11
* * ^
, , 'e ,,
. g- a o o i g, x
~-
. = .
- ,,.~.. n....,. ,
o _n. ,..
. a . u.
m p<.s .-u . c w...-w,
. . .. ... . .~
- . mw. ;c , ~,. .
rp- _ q. - w ,.
+ v ,i , v, .e.- r r
. .-n--7 c;n .m y n,. ge,m.M, . -
g:,,,w.+...-e-
-- u. .-
>.s.. , <
..,- > > e , , , , ~ < ' . . . . , . . .
;-g.o.n
,t
.,:... ,y. . , . . -
^ - . . .- f ., ., i. ' *. , j , . ,*5.,: .. 'W ' . , ,
g . . T.
.- .i, '
- d. ' r *
..M r v ..
3,,. ~
. ,- / w .
,i .
.. y- ^ ,. . g ..:. ;
, 4, , .c . .
r;
.;i,. ' - , +c .,< ... .
. . g. ,,,.+v.,.- ... ; , ., r , ..
- m. .:,
, . .,.o ~.,.... ,.
. . . s .,. . . . ,u -
m s
. , , . , m .;%'
. : , v .. *
;,v....a. . , <; . ; :..w/ .. -,s- .. ... , . .. ,:., .; .a.-,o. >. . ; .,,: ;
s ...
, , n., .2 p . . 3 .m- ;. g.
.,,-y.- ,.q,
< ; . , . ;, . y , ,7 ,, ,. , . . . e. , ....
, . ~
', , E.,', ,
ip
'& , O-1..,.',
.,.m 1.
., ,_ t... .
, . " ,. 3. . .: $ . , n ., .
i
,. * ' m
, t. , ls .,;. r
~
., ; e, r. ,g e .
,s y
a, ' x ,s . . n. s . , ,.e..
.. . .. a:. /
. c ,1 )/
>. 'r*. . . 'M f * ' -a t e
.. e f
. , .. ., 3.. . . .- . ,
1
.o. . s, . .
m ,. .
- - to .' ,.* ,
t .,
.. a ~ ' , . ._ , s -
n
._ y
_ e<.- . <'. c* . t, a Y ' , e,
- q-
.m
.,.,_, . ., ,t,,.
y c p
. s c,
g P' e i .
}g ?
I r k
! m.,..
mu. m , r> v 1 .
- s.
, t, ,
,. - r
,I fI b
i 1 N
- L
/ T
- i
~
p_ 4 4 4 a s e { >
- o. , :
i t I
., J .
99. DISTANCES FROM RELEASE POINTS TO THE EXCLUSION ARCA [h;h
' ~
r ..'. .: ,' t < i 1!51.W4 'o:GUlRE NUCt. EAR STATION ! W
'i-I l
- Figure 2.1.2-2 ' -
1I r v . ' i 4 1 y 1 % I
% i 4
.'_U*t$ e . y e ev w g - -' ' _ ' 4 y, .
LAKE fl0RMAtt DISCIIARGE N CAtlAL 7 l I h
'! i'!'
~
IflTAXE ' ' ' ! h CHAttilEL
//- s s
N N - l g
//
// //[ N\
\g -
1
~ s \
CONDEtlSER /
/ NN N
\s CIRCULATIllG '/ 7 I J sN s
N WATER f '/ AUXILIARY sN
- s\ g g\ LillES f
/
BUILDIt!G 's N
' 'e \ \ / # s\I l
/'
Ld i g\ \\ l i e i l
\-- LOW s\ \ \ \ B t RB i I
l
\ \ .\ \
.i,;
LEVEL UllIT UtilT I II INTAKE \ \ \ l1 1 2 i I i l LitlES \ l
\ \\ I-l I
\
\ \\ l_ _
l c\ s' \ \
\ '- -
TURBINE SERV. TURBitlE
~~
\ \ BUILDIrlG BLOG. BUILDIfiG '
\ \ \__
.\
UNIT 1 Uti1T 2
\s g '
s-. i___
. I l
9 100 290 . SCALE Ill FEET '
. , l ADMitl. ;
. BLDG.
11CGUIRE ltVCLEAR STATI0tl O
,' l
/q. g<u g
m - - - - l l- -l l ll 1 - flew FUEL POOL i_ _ l li _i c . l .
- l. --
3
. II - REFUElli;G WATER i- : : '
I 'l -SPEllT FUEL POOL
. 5TORAGE TAtlK 3 i
I il I lI.I 7 I HSSS I I *
- iI --
I - l , _ . - - - - -
/ k,fds w.- _l ,
O 70 l EdiPENT
, l ..______- . ;
I ICTOR ltAKEUP
.q g \ %, / I I _ _ ,,, _,
l l __
' ~ ~ ,I '
i CONTROLI l.-' /
' - - ' DIESEL ..; AGE '
j
I l I RO0li GENERATORS N h ~_-l_\f- ---
I E3 t_ IDL _, _ E1Il l_- _6. MS I /- , t- - - - r- --7 _\. ./, O 1 7 i FEED-l WATER l i FEED- i l WATER l r-ti~'l d 'k i HTRS. l . I IITRS. I rJ ' L-
,' 1
,i
'l g I
1----_ L____J llI ll si g l i I I TURBillE . TURBillE s' t, ' i - GENERATOR GENERATOR 1 I il t i ;g e, g, i l,, a l i. : u_, L_a 0, 1,00 200 . SCALE Ill FEET MCGblREIlVCI.EARSTATION O f-f.z o 2. ,
( ), ;.1 s 2 p- < :
. ,-+ ,
,g g h h '6) g.
~
, .,l ,
! (LT " s
./
e.n
, , ,s. f r *.n r ,
)s. <. r% M < ,
I ['
!smesq j73Eqf.._,]73,
- u. -
i 'g , j.
. .wa.u a p il ,
j p sossu E9 W; _~3u.= t
~
i
.. r. - a..u .ee
.__ L. ,+= .s., i. m# %. . p, i
- W a _ d. .1 % } di- :Mm.
- ..,uv. " y$..% a ;;"c; m.
mm 3 ' -
,_s. L, qA , -i
.. l.- a 3 <i-a*om p.; - - i n,. :-. - ,..<
,[, -- =
a(,_W,
. . . l ,-,
i r g- s i i. a
, 7..u
-3 ,
,u ggf. ;; ,_
i
.__. m
.-- n. , , i 1 ,,,,
i
. n_~_ _+
u . > L
*" i m . v ~ ' a i , j <
usa- .. -
-p i i .saan l 'i' f,. m ,
l
, j
^
r M .
,a..
A. .j . l') * -
,,, I '
% -+---'**' 4 ,;' ti I
. r . . .. .,, '
w i j i d; (,ov I 5 ..u.I
.L.
p- -* FT l u. .- ".a '9.! r, s . i 3
", 2 ._,.-
j J e p . , a - "' $ a endg* g- ,, ,* e g
-e. =.=g=i . , iv n - , . . - . c +
w m uv/e s
~
l'T 8 ,e m.2 3 t1 -
,wf q e t ~_T.an" .o .d!iin ,.s cff ; u,o-- , , t . -
,1_. ;: -
y.a
- im _! ..-
l
- v.;__ _ ,
; ,e ,
e s ._ . . r ;t , ,. , .ict wee . i po y,n y>.
- s r, : m,yy-g . .. .y __ g y
.-t ;
- q- -
te.m, .3 - d =c 7 u .
~,
s J ' 'l raL ~-;-- ,- t;1""* ty. .
~
TOP RDCK 4 y
' / \ ,' gi V \ / \ / E ~ EM f / , %
/ _ \ / \ < \ . \ /\ / \ ' nw, , , ; ;i -
\/ og , bi,; ,. I M . c;L_
, . i h ,', '\'/ $ . /\/\.
4 _. s u v u u "' TOP EXCAVATE<
- q. .
5 W D m)5VE@5_E_q . o o ra m m h
, ., j
, i m%,,A _
SCALE 0
,1 s
Q'Q- Q gs Q'
.s l .' '.Y ,'
- c. - .
^ ly}
..v .* . g
, . ' ~
- e. . . . ,9
,,wa m .
J.6
. J c -;
..', 88W 4
.gv.E g
4, y
,. . gr ww .-. .,
~
1 - QZr--
. .~ = . -t = s.. s+ A m *.-
L m .==:,:u .: m . f
= : m _ =m a.s.: q aa f l , _ a M W'(;'" m W aadr ,;
I
~
i, t ., um
*"a- 1, a cityjg e
X - -
.1.
,MM4, ~_a
,e ',t..,,. c .mm ,
1 i t .
""* cf3.
'*MG 1 pf i.i F MfHw(rtM s
w
.- r,t. ..-" m
. u.
1 ,
%:- w:
._ p: c_% t :.! "
mTu 1J
.J.M,JA.,{. g %,'3 vmo d,, , ,
L
- m. - - _ _ .
m'"'i .--. Q
~~^~i
~, f
~~
s, wusw mn
.. ; l a. <
a L<.,g< 9rW %e, m 1 4 - s em -', w i
.- ....., 1 % :
!p .-
.#a ,
_L;,j --- y-
-- -- 1 ) ; -
-, iig. .
. 6 l y , ,. .
i ,, . w J t- . .m L .k,ii.[ ; L m' ._Q,N,,yd ) , ,.! h n w _ m - t m "(., ,
;4 M*f-f
(" f, *l ; } [(W'p' *^ ' rs lfr u t*
- d-
' O m -- -
's. 1 W
m ,- < oP,. k%','j.j
- -.'g_j 1__Y_... ,.,.cL_w.
f
*"; ?
a;rr*1
- 4
,'.Q.
.v s.
. ,l -._4 C. g_ s.:(". _. f_,
- - - - ~ , 1 _-
~IN .
,. ' y - -
;_ /
~.. -
,s-
.. . . , - ~ . .c . . . . , . . ..
' %< RhCKF!LL "n.AM ,=-w-.IL_{5. ~ %. t R F.7 7 W w wdi. - *t?.-lT e %i s c
- ,n , _
., a - _
f% 7I t. c-1*h .J __L, d ) W IVN/\/' G D ,7 g .,m ' _ & .m't.rt .d = TA -idrw F- % /T/17. _i , _ L
,gg
.__g' _.
s j . 3 ,,
<. o EXCAVATION LINE ,
- 2. -
c r zu
.___ 0.- .
y
'; Lw I/\/ ,\/\/\/\/\/ '
]
i.. b.- -
) .
i eeu_4.CD4TOPBLDC,1 i
.f 2 '[
I
.T
<i en u.< un m
,'(FIG. 2.5.l-1)
N/ t -
.wa J
o etaar nw - w, I I- . y T'.I'7tJE M.4 "5 - STRUCTUREFOUNDAT(ONPROFILE
.. . y d m. p gg. McGL"nE NUCLEA .
STATION l Figure 2.5.1-2 e
l!i j ' i O - M r- 5
< 0 0
,4 ts 2
7 7 0
' 0
,0 2
a s
\
\
\
t k e 0 c e 6 o T 7 - . p .
) 0
. -
- d 2 0 e l
-
- r - - ,6 c a B 1 n E l a I A2 t
- 0 d
r a ht-A e7 r9 i D s Y n u1 og ii. t Fb 0 C c e 0 eeF i7 S c H / ? ' ( S - 1
- ^ J_
O t e r u Y
~
c u r , t e I. i 00 S ' - 8 _' f t n l P a t a 7
. V _
,II 7!l ! I [
0 . , 0 8 0 7 4 l
,t;t r n J
t
$ r, enr I
E k r e e e
/
k 1 e. H ' i ~ / - ~ D /, - .
~ ~ - ~ _ - - . - 0 0 o -
0 0 _ 0 0 0 5 c 5 s 6 6 1 7 b .g _ ' 7 2 w 4$ R y omOEoC>Mz MmG4 R _. O . hp. t
;w,p ,
pp.u> rQ"
. esc 5"2 Corm >=.,.d>H_OZ
, y E,o $ e L 3 O y
- m. n2 K o" e
\
l L~.'QQfi '!D INN 'lsli)3ff:-
- - r.-.ll aQC l
- s. .bi t.. L.tl it 1 ._i1 N (. I?
.d h*)P 1 .4u If s.:
C Ns Tli y!-
. ,_ c'l lr \/ O)lUd \Jni t == . -
--- _ z _
i.:Grch 0, 1977 . O Advicory Comittee on Eco ctor Sufc cuards [ U.S. I!uclear iteculu tory Comiu;1on flachingto n, ii . C . 20050 j Gentlemen: The encloned twenty copica of questions vifth respect to the nufcty of the 1:cGuire, On ta ci bu , Perhina, Chcrohte and Oconce nuclear stations ..cre picpure d fo r cu bmit.cion at the scheduled nectinc of the ACiG in Cl.crlotte on JLa. 10. The ncetint, tiac canceled. Ac I have not letraed of a subccquent cchedulinc of the noctinc I an tchinc the opportu;.ity to f ory:ard thcuo questions to you. I note with interect that none, of th en vicre raised i n 2.~JJEC--0136 or h"diCG-0153. Ishallf56me: hat lena concern if the procedurec for p2D J ccting life perfornnnce of .hc nost houvily fcticued components in the prinnry coolcat systen are ncdc at a hicher level of cophic tication th;n in Ir ovided by the AS:!E codes now in effect. Yours very pa-v, Q jfft '9 /cL/
, . m Jesse L. Rile.
Prssident, C25 ' p n e e. p m 1C.L,0U:
'~'?9 uCpy - it,OtR1Il for tIl0 [_:d3 C2 ti.3 CommittGO .
,.4.- - -em - n .wn-r,- mr.re -u ATTACllMENT 2 1
O / - aes
~
u 5LCONi IKritCY NhCC Cil4L'lOTTC liORTil CAROllith 98007 - (704) 375 43d. ._Jf '
', r-
CAltoLINA ENVlitONMC'NTAL GTUDY GROUP January l'/,19T/
.. n 4 ..r n , , ,, m -
Cfl Atel asf 8 0 #de seel ps s' A sse nt ap4 A an/sef P' il sta 6a letanet
'Fil4 4 # $ 4.14 J Wip4m i spH R A 0 P M Ne itt l et s' A ttitt IN A 3, 9 49 'til11 Lin fin Advinot $' 00mittoo on Ronctor Safoguarda
!!cotir., !a Charlotto, N. C.
InCatauba, to the teatter of Porkins licGuiro and Chorchco mr.1 Oconco nuclear stations and with referenco nuclear ctations - WAS'i-1270 van concerned uith reducing the rick of a common modo failurc involving control rod incortion. At both the Mcnford and Kahl reactorc , thoro had bocn incidenta in uhich the pritt.ary control rod incertion system had fdled to perform. 'dAO!-1270 rcported octimatos of as high 7000 pai in the princry cooling :g.cton during an.~rbicipatt;d trancient without ceram. The Atoaic Encrr,7 Co :miccien in that report required that, cftor a specified date, all reactorn in order to noot licensing requirementu,- vould have to be omtipped uith tuo coparato cordrol red incertion ny:fcems uhich vould be entirely differord with rocpcet to modos of failt.ro co that the rick of complete insertion failuro during a 'trancient uculd be cignif, icant37 reduced. p UASH-1270 did not en11 for retrofit of units licensed before the et teff date. Vas the primry roccon for not requiring rebrofit economic or concern for d concrating capacity? The procent rocerve is :;uch as to per:dt shutdown and retrofit. - The WASb1270 requirement ap.c are not to be in offect hor Catauba, Perkins or Chorokco. initial]y promulgated Phat are the r'ounds for removing from force the reqc'rements S WAm-1270? The ASTM codo for roactor . acccis proceribes the requirenants for bolting materials used in clocure studo. Sinclo tencilo cpocincno cut frca cach cud of a bolting st,och heat euct exceed 130 kips at ambient break. For the 11cCuire and Catauba reactors tho varianco in accepted individual tencile specinenc van cuch that, using nor..a3 ctatictical inforence, the population from 130 kipo. which the cpeciv.cna uore drem contained 10'!, of individuals locc than Scroral questions should bo asked and ansverod. Tho prediction of fniluro, a corious concorn in the NASA spaco progrann, ronulted in tho developmort of probabliutic analysis. Why in thin novo nophinticated, stato-of-tho-art technique not unnd in tho ovaluation of phycical tout, data mvl the catting of cpecifications for reactor utudu? o The ASN! codo annrmon that emn31 tona11e specimons, area loco than 0.9 in", vill provado crJtically Jc. port.::nt ini'ornation an to the olovated temperaturo tennQc h0 in", propert. ion of bolt.n ujth a ntnimum crosa noctioin1 aren of about ( ilan thn relationship betwor u tencilo upocimen proporti on obtnfnod necording te tho codo cinel the olovnted tamporaturn tenni]o characturintien of boltu from tho camo nLock boon dotormined? If so, uhat woro tho riiviingn?
. /f706 ~
r-
'tha autnuspblon una mado in tho.citco of tho McGuiroonctors and Catnuba r ,_ that, of L'00inF.oporetion, no port. ion of tho studholt vould oxened a tenporatura
]s I lartt acm:=ptionn nro mndo for the Porkina and Chorokon r 11au en actual dotorninntion boon mado of the mxiraa oporntin otte t ors? g tottpornturo of utud boltu telth particulttr reforonce to thoco portions ponotrati trrrpor flanco t.nd threaded into tho louer flango? ng the Doca not the departuro fron the initia) rod incertionu prcposed, cetually pronu)cated, requirements for dual and different rick accociated trith an un:hrtcetod, wonk ctud bolt?in WA5i!-1270 incroaco the I Hao rm indoporrlant, uoll qualified orport or groun ofecleultted oxperto the lond contribution frco comproccion to uteda of tho flentlo region vicinal to a failed co=prosced byrecethe ctud ovory nut of of th the j stud for a varioty of cituations including normal operating failed proscuro a the ranCo of preneures Anticipated can develop in tho aboonco of scrars? for the full rango of trcnci ents thich If so, uhat uoro the findings?
,n
( Josco L. Ri of 1 for CEEG l O L) (.) l-hl y- -
AFFEi4IX XXIII l;c3uire 1 & 2: Site, Orcanization, and ( Layout VIRGINIA TENNESSEE
, J~ ,y
[ /
,,,8 fM
. s 0 s TE
. .- -- NORTH CAROLIN A .
- 4.c . . .; . .; ec;.. ,
q f ~pp -_, b -- ; s j _ Tc ,y ~ o { ;g; y; . "-- ,? ~.
.,A. _W - 5.j.N
.- S I.. ~'..s.d..:.~
~
.gt@~--3.'p',.}~,?$o_,K_~"j3*
1.} ,
..E P[;$ jj, [(hI*h:- e pD l h ** ;%h?%Y5:0iY, Y GEORGIA ' #y T,.,$,M..,c
-: as esw!";c M r ' 0 C
O' y, ;tySOUTH CAROttNA- 4
%Shu jx,e:&>:
% U mQ. (:- <..-
\
m.ede; ~ 4$ y' g V
- q. '*$ N-?.
. . ~ , . . .
.. gn .
- i'l '
, g I -
l 5 50 0 50 100 z-urnc--> n APPROXIMATE SCALE IN MILES DUKE POWER COMPANY .TRVICE AREA h% \ IS"A
- p. eu McGUIRE NUCLEAR STATION Y
PRE 510ENT' O EXECUT!YE VICE PRESIDENT SENIOR VICE-PRESIDENT
, PRODUCTION AND TRANSM!5510N 1
l Y!CE-PRESIDENT ~ VICE PRESIDENT Y!CE-PRES! CENT DESIGN ENGINEERING _ CONSTRUCTION STEAM PRODUCTION CONSTRUCTION GENERAL OFFICE STAFF SERVICES STAFF
~
STAFF CHI NEER , y PROJECT MANAGER ENVIRONMENTAL MANAGER NUCLEAR O!Y1510N PRODUCTION l CHIEF ENGINEER ELECTRICAL STATION DIV1510N gg3gggg CHIEF ENGINEER MECHANICAL & NUCLEAR O!V!510N i l l N 0 i McGUIRE NUCLEAR STATION jf ^2d) 9 {mmwW Figure 13.1.1-1
\ E . C - I _ R V
,] -
E G R P E U 7N A S O N R A G G M N I N I A R I T N N N E I O D I I tNN C S l0E I I
$lT N V _
SlN D l _ OAI - TIR O O SE R P P E U V $ l l S D t P - RA E N n U ER C P A i. O - R GI I AS V O U E L E G N R G A N NI R R A C N S AN E G N I O E WI S A O N M lN R C i I I A t V V E N R I E E 3 I N I D N 1 0 5 M l A i
,1.- E t ~ . I U
( (S l 0 0 C R I P, V i V S H Ig I E E0 L M Fi R U E AA E S G - 1M G E A P l A N R D N U - ND N A O ER A W A W R T0 M [ G i l. I0 $ VC E -
$ V U -
S E - l N N S N O [ I RNR G S A0I A I llc _ l il U V D I ~ llA CAN UIA M . H D ~ NSW _ _ P P _ N U 0 O _ l _ R _ l C _ A _ R R E _ l [ C G P D N A O N A N A N A N x N I E I rU ^
'- E - ]; OcQ gz I N b','7 ,4-5 S 0= 'dd o" I I -
T A S M --
?.
' ~ k8 m zco{>M d,.4- o.
g'C i l { l I l l l 1
l l
~
r:,
- i. r.., : r: ,.: l
.?. ----
l i J W : 3 . i.. - ., - - i l l %: 1 g * :. 2 22
. i
- --:: 3. 1 l
=
. a, -
l
.- t =- .. ...- . E, - <
t .. I i i -' - t- - l._ ._ l_ t a l i 8 1
,.1.. .
I I
.i -
I
. m.
.-. C a s
l, ;* : 2: : - l, - . . (
-[
3 ( I
- , i
~,- t
~
i
,, . I
- , .m
. \
.. .n a..
\
- 3. -.
l -=
= . . .
.-~.. ~
=
- . . ,a,
-e .. -. .
.- e ---
e a.
~ - i
. . ~ . . .
e. 1
.... ~ w I - . .- ,
.. . -ag l
- . \
-. g, * - .
- ,b.s .-
. e
. . ==
.--a.
. t ..,... . .
.. .~.e-.
~- .
w
.. - I. . -
. .> . ..~ -. . .
! .. . . s e .. .
m o 2
.: :. ~. _
m .. . 2 e a. C :n
~
8 C n r., . . E g - a 1:I m .m .. . . . - e -2 2s N e .ee
.t 2 9
>~n
. .N . .
g y * .6
.a. o. E .?,
V ."'e . . . . . 98
.e.. Q
.=t .
...e
.s
=
L .. 3
> ..e. . .. .
, 4,. .
o 5 . 2 -
g e Mgp w.y.w .p:. .'? 5.:. s.,y .a.p t. q). . e m. .. c
> im
. . m ,- . n. :. c. . .
}
fs%gg.... . ... thpffixW.':* .n. m ) ~
.....__.'M.,.-
;...G,... .iy,
.v.,,.%.
.. . , . .i. $
.+*+y ' '. - .. j d ,....
44- Jn- ...,p'. ..
.&...z,..gl'[.
l? 4 .'};f g;.; i, . ;. o 'l [sh 'r .W.f. ., ,. .
,gi'r1( u% :g
,s + . - . . e 'r o ;;f' y
- s. g t y , r.,
>k;p:r-,,: %q, g.U'95.5.. r eI ,. . D g::. p >g p'$ .;.
'jN, .'.r 3';": .
V .
'kg .c.. . ;.y .g[.7. 4 1
t .N Y2.. .
.. .J. ,.y.c t ',.,( l c " .e : /c./ ;g
- g' D % ( h.'.k k j.h.,?*Y . , . .
- . .lE-N 5 y'$f4.,kh,.h!m .,..' h'~ .~Id Q-. - f,f *fs, .b' 4tr. ';
- a. ,
+.
'k . .. .@W.
i.
. . ,1,
.f e
. n'. .s gl*h.
./
-7 ' ;' El j.d < ,. . lg + ,: W..,. v.
F3P,f,h f,. W%y'i N.\p,,; *
- g 'I y! -
g ../ k I 4:( - 4 E ,-l. p' ', w, '- 'M
% Qi ,'/
L J.y 4.s
. . . . s..
b.; 5 j ~. =- i1 *
.i..
- c. .. >
;....r..,:p4...c;; % ,_
. . 3 _, .. .
_ _.c.S.. ,, .- . . . _ .. _. . . _ . _ py-9-:Tffy;gn ) i y\
' V' a nn.?;M *
) !Pl;@Ml
,. f , __ . _J -l N.. .,
( iIcM / ..,_. L
. 'I~i.f }e . i['.1y',. %. .
' aJ E g. n;;. M.'
h.
)k. ,E- l g) !g .
.% V . $ m Oh. -
i .
! &, J? , . fkeYW n id . .. .
f..S7
. ,- -. v .
I. ys9 1 ' o G i@ ./..g
' n' s
4,
. M .i/ . 4.;i ' _'
r s.v'dp(W.. . ! ;g ' i c y ,. yu. g .s
)N. . .1, v,.s
=.
5 =E.i ,~*
-- . {. , 3.....a u .
s Q> sva,.p ;y\ 8, v p -
.,- ,:s , . s.
/ f. /h ,.
5.. .. i
...e,... . g: s .
~
*9 l
.]r n ig y' q cf
.h,.g :u n..,.uk..:.. <e f .Q y
' .t .J,a ..;n . ..4.
. .% .., . W, o. . .
4 ..d.,.:3.,k. ,' e+ .[i '*U7 4.i y . . ..-
;. ' A m .1. ' .W.o e .
; r4., p. ':,.x, y.c . , , . .. . . .
.r.w: ; .. . m . ./
.,1 ..
e .,e9.~n).;1:' :
.. w% .p.> y [ot .;i . .i,. $. . pwQM - .
n: . f. e
. D ,0 yi Q <
* ?m! ,.. . :.r yh ? (,..c. .
pTl8;p,.h;g.. I'1r-*T8li.g.Ng..i.jC.,, D, v . 6. ,P.4m hi h4.. h
- s. . . a_ .a 4.T.T$.e ac g .,,'s#.R
..~,.M.
w. N,
.. 0w< . 3. . ,, , - . . . .,-
.i.-_I
+ q..,
.. . .u ,
fn
,I m, =. 4,. .s.y.
+ . a w ,
po a..' QS ..g?,.p85 k. g w.' t . pe.A .
..; n;.n.mmas, m ' .' m:-e
^
. q.
e, in ; ,. a+. e m. i s:. x , a. U,
. .s ..: .r
.. ,,a
- i ..,: . i .
3 i
't , .e: 7 hj..Ar W.W, h. ' M., ',[.,
g{a ..~/,, . ' . f. N ,$' . , . , . .'dA. . 'j4. i M , 9. %y:,i s..,A 'm.}.y% g.. _. 49 3 (%
- y. .
.i
' n' 1.?. 4.-
.;; O .1
'.8 .a..
'.)- '( A- l y[' '
d.6,kr'I ' , g,;:.t b .T .e .1(* '.s,I ./ t,b f 8 4 3< - ,, /.f . F, \> *A -
,.2. . .w. .,~,.--
i ...' ,,li ,[. ' .
-t '
kv f'
* (( .** . f Nf *k' . r 5
i , g , ,, _ -- ef %v .
, -1.
m
, .i..
i ,. - s , q. . a. i . ie Y r $$ c 'I \ ., ',' % ' , f , '. , ~ +- . . r,8 l p r. . ..n .f.
- g ...,v a n ,. .- .~ ..
,0Ly '- .t
's t \
, s
,, y i
t ?,,, N.v, -
.f. v -
g L ;r-Q .
' ,i, . ! ,[ i,.. ,. ,, 'e. , , q .j j * - ,
l ' ;., .. [. 3
' r:
r'. , j: -
.. 1 i.-
' ,- L l Q1 :'\; p, , _ ; . ' i d.,',<k' l 2. y , . ' -' 'j!. f; ,'.;_f.f" '
t$. ?'. ,'% ll \ v n p ;
- j. bM; ;'
4 {' .! . .f i / ,;, f -7 / 2-
l APPEliDIX XXIV F.cGui re: UHI Analyses Compared with Measurement
SUMMARY
OF UHI EVENTS INITIATED REvlEW 0F UHI EVALUATION MODEL JAN. '75 NRC STATUS REPORTS SEPT. '75 AuG. '76 DRAFT SERS JULY '77 DEc '77 ACRS MEETINGS JUNE '75 SEPT. '75
'75 O Dec MARCH '76 JUNE '76 l SEPT. '76 )
JULY '77 l DEc. '77 O !
9 O ISSUES CONSIDERED SINCE 12-77 .
. VERIFICATION OF SPLIT DOWNCOMER
, ACCUMULATOR DELIVERY .
. ROSA-Il ANALYSIS
, SEMISCALE MOD 3 UHI TESTS
, ERROR CORRECTIONS
- UHI L0w FL0w QUENCH LIMIT
-GENERICZlRC-WkTERREACTION
. HEAD COOLING JET MODIFICATION IO REDUCE UPPER HEAD IEMPERATURE 5
O g- a / /
CONFIRMATORY COMPARISONS FOR SPLIT DOWNCOMER liODEL )
. L1 LOWER PLENUM DELIVERY, LOWER PLENUM AND DOWNCOMER STORAGE, END-OF-BYPASS. ,
l
. CREARE SWEEPOUT - LOWER PLENUM STORAGE CREARE TRANSIENT - END-0F-3YPASS AND LOWER PLENUM DELIVERY
. PLANT SENSITIVITIES - PCT'AND SYSTEM EFFECTS l
O l l l O !a
/f~ Of V
O DOWNCOMER MODEL pct SENSITIVITY PERFECT IMPERFECT CASE MIXING MIXING pct aPcT pct APcT ("F) ( F) (*F) ( F) BASE 2020 - 1800 - HEAD FIx 2020 0 1997 197 LOWER PLENUM SEPARATION 2130 110 2150 153 0
. BASED ON THIS STUDY AND CONSERVATIVE COMPARISONS TO DATA, REVISED MODEL IS ACCEPTABLE.
9 0 77- p-/ L
n ACOLTJ1ATOR FLcr4 RATES c LOOO. L ..
\
I [ (' 1 Westingho se UHI i
-f .
l s l 1 \
- e
/ -t-- UHK1 - S2ndia UHI '
I l 3000 - o . I : I } O O I \
/ 1 .
- u o ,
..! l / - UMK1 Sandia - Cold legs I < <
6 } !j s / 0 0 4 4 W 4 t
*l U r L/ \.i<
2000. 4 <<s< l
'[
N, [ \ g r i c < N o . 1 l
- 's M I Westinghouse 'N i
\
t I Cold legs 8
' \
a v-w-%, o k j'. % C . . l ! 1000. .' ' l l 1 L- . : i I t . I l e
- a. .
, r t i 0 i
! !'l l
4 -o1l L.. _ __._.L L_ ._._J J_ O. 20 40 60 80 100 100 Tir.c - secoMs r\ I
)
l l
/ D / }
1 l
8 l 1 m 1 l 1 1
- l L
l I
!o 1 i \ -1 go l
}
f i 8
\,
ni
. r -
l i l ' i () ' i t 1 -s (/ ; g
; j
$ f -
1 8
.' 8= a i
_ "o : i 6 i e 5a. - h 4 w o \
- I s
.\i)
-o e ~
l
/
/ i
\
/
/ .
/V u-t@ . l l
.. ---.1 .g 8e
- 8e $~ $ I s ~
1 3.- .n n w e n : O V 4
- W-2/V
. o
/
/ \
U
~
O
. O
.H
_,e c3 O
,o y o
a ) o g ,. Og to m // , i & n a- _- .yr :.- e
.5 W s g / ,-> . _ ..~ .. . _ _ L ~.W
,l o e __._***%. ~
l a 61 .__, o
-2 I
e D 6
'\ . \. y=
x . o
.L" ~%
4J n .- s4
\
., f~ -
~
,N.
s Xs N s .
% ~ u ..
l t l j
%3"-~% y %.. ,
._-.Q,~,
T2
~,
= - - ~ ~ .. _. ~*"w- O O
- o.' a 8
~ * ; 3l:.M .~.or; $I 1
l l l [\ l l' '. rs ) l
/
1/ l I
a O
,- 0
, I 0 _
1 s _ a{}f; .
- _ l 0
8
^
W e D I s F u a . o d' I
, t 0 N h g
an 6 M n O U L O C i t s e k
~ } s a
T R W N n O P n u F i 'A, l S j( .f i
;a f' i[, d s
n 2 I 0 o
,I $ 4 c e
s . W e i m T _
~ _
~
, 7g .
f 1 0 2 g/ _ 4 _
/I f,
,,i _
ll! jA f o o a o o . c c o 2 i gC O _ N
. x'I Nyq . _
l w O Aeeeriotx xxv ticGuire 1 & 2: UHI Ai!ALYSIS PREDICTED ECCS PERFORMANCE 1 l OVERVIEW l CURRENTLY SUBMITTED RESULTS l l MODIFICATION TO METAL-WATER REACTION RATE CALCULATION O APPROVED UHI EVALUATION MODEL CALCULATED ECCS PERFORMANCE WITH APPROVED MODEL l 1 O g, z i
O - O - O INSIDE OUTSIDE CONTAlfoENT gCOMTAINENT SURGE
, y LANK Jc-55 FT 3
[ 12*
' - NEMBRAME I800 FT3 l [D [ '
I I LIQUID GAS I I2" p\ V
*A UNI , ACCU!4ULATOR YESSELS 5" D' 1
~ ,r -
/ 12' ISOLATION VALVES
\ 8' YALVES CLOSE ON --
h CHECX YALVES 0 LOW LEVEL IN ACCUMULA TO R p S' MP j% W -r-8
/ % 9 Z_3,
- FOUR INJECTION 5 / ORTS (5" 03) t 2 -
~
REACTOR YESSEL HEAD -
-- w figure 2.1 Upper Head Injection System Schematic -
O ANALYSIS CONDITIONS NSSS POWER 3564 (HYDRAULIC ANALYSIS, MWT,102% OF) i CORE POWER (ROD HEATUP ANALYSIS, MWT, 102% OF) 3411 PEAK LINEAR POWER (KW/FT) 12.63 O PEAKING FACTOR (AT LICENSE RATING) 2.32 ACCUMULATOR WATER VOLtDIE (COLD LEG, NOMINAL, FT3 ) 950 ACCIDIULATOR WATER VOLUME PERFECT MIXING (UHI, NOMINAL DELIVERED, FT ) 1000 E50 FT F} IMPERFECT MIXING b u 1
-a23
li O COMPLIANCE WITH APPENDIX K 10CFP.50.46 l
~ ~
RESULT D = 0.6 DECLG D MI G PERFECT MIXING PEAK CLAD TDIP. ( F) 2164. 2163. PEAK CLAD TDIP. 9.0 9.0 I LOCATION (FT) ! LOCAL ZR/H2O 7.02 6.05 REACTION (MAX. %) O LOCATION OF MAX. 9.0 9.0 LOCAL ZR/H2O (FT) ) TOTAL ZR/H2O <0.3 <0.3 REACTION (%) HOT ROD BURST 64.0 60.2 TDIE (SEC) 1 i HOT ROD BURST 6.0 6.75 l LOCATION (FT) ' l l l l [ A
,cf - 2 9- Y
'f5 O l l
l l l I 'l MODIFICATIONS TO ECCS MODELS l l l I l MODIFIED METAL-WATER REACTION RATE l I COLD LEO ACCUMULATOR GAS EXPANSION ( DOWNC0FER CROSSFLOW MODIFICATION LOWER PLENUM SEPARATION MODEL l l l ACCOUNT IS MADE OF INCREASED SPRAY N0ZZLE FLOW AREA. i i i r) ff - 2 ;L S~
l$ . O LOCTA ZIRC-WATER PROBLEM DESCRIPTIOl1
- ZIRC-WATER REACTION, HEAT SOURCE IS ASSOCIATED WITH SURFACE fl0DE
- DEFIllITI0il 0F VOLUMETRIC HE,AT GEf1ERATI0il BASED Off RADIAL f4ESH SIZE RESULTS Ill TOTAL HEAT GEtlERATI0il DUE TO ZIRC-WATER REACT 10il DEIflG REDUCED BY FACTOR OF TWO
- PROBLEll HAS BEEH VERIFIED BY EXAMIflATIO;l 0F HEAT BALAtlCE AT TIME O or ect w"er' "c^T Ifl EauALS HEAT OuT i
- CORRECTI0il HAS BEEi! IMPLEMEllTED FOR ALL FUTURE A:lALYSES .
a O
# aa L
O O O i
. t PEAK CLAD TEMPERATURE RESULTS USING APPROVED UHI EVALUATION MODEL DOUBLE ENDED COLD LEG GUILLOTINE BREAK, Cd=0.6
& E I
s PERFECT MIXING IMPERFECT MIXING A OR IN UPPER HEAD IN UPPER HEAD - SETPOINTS PRESSURE 1200 PSI 1300 PSI , h VOLUME ~ 1080 FT3 ~ 960 FT3
~
x1 :
- PEAK CLAD TEMPERATURE 2190 DEGREES F 1990 DEGREES F R
e Y e I t
, :: %D' fjl February 24, 1978 i
(,(' ' l APPEilDIX XXVI I Davis Besse 1: . Status Report l l STATUS REFORT DAVIS-BESSE 1
]
l The Committee reviewed the OL application for DB-1 on January 6, 1977 and issued its report on January 14, 1977. The plant is now in power ascension, operating up to 75% full power. STATUS OF ITEMS IN ACPS OL LETTER, January 14, 1977 (item nu:rmers are snown in margin of attacned copy of letter)
- items are suggested for discussion.
- 1. Not resolved. OL license conditioned to require analysis prior to startup af ter 1st refueling. NRC Staff has not is' sued guidelines to licensee for the analysis.
- 2. Resolved. Mditional data on reactor coolant flow and pressure drops is to be provided after the reactor reaches
. 90% full power.
- O
- 3. Not resolved. Little has been done on this item. %e NRC Staff wrote to the State of Ohio (in 1973) suggesting they participate in the program and explaining what asrist-ance is available. No. response has been received and n. ;
follow-up action has been taken. *. i
- 4. Not resolved. %is is generic item IID-2. We NRC Staff will elaborate.
- 5. Not resolved. Little has been done on this item for 2-1 or other operating plants. The NRC Staff has not required backfitting of RG 1.97.
- 6. Not resolved. AEG is generic to all plants..
- 7. Not resolved. OL license conditioned to require modifi-cation prior to startup following 1st refueling. NRC Staff has suggested a modification to resolve this item.
- 8. Not resolved. Much progress has been made on fire protec-tion. OL conditioned to require licensee to meet Appendix A to BIP 9.5.1 within three years of issuance of OL. NRC Staff will elaborate.
O g_ 2 2 8
- D 2
Jf,'""' , ,.?!&*
..g . . .
( STATUS REPCIC DAVIS-BESSE 1 2- ,
- 9. Pesolved. NRC Staff wil elaborate.
STATUS OF ACRS GENERIC ITEMS. (Report No. 4, Acril 16,1976) II-l Turbine missiles No change. II-2 Containment sprays following a IfCA
. No change; LCCA doses re-evaluated. t II-3 RPV Failure Post-ICCA by 'Ihermal Shock No change.
II-4. Instruments to detect severe fuel failures No change; resolved for limited failures. II-6 dommon mode failures
. W change. ,, II-7 Behavior of fuel under abnormal conditions I tb change.
i O II-9 Seismic scram No change.
. II-11 Instrumentation to follow course of an accident Resolved generically but CB-1 does not meet BG 1.97 See Above Discussion.
IIA-1 Pressure in containment followine a LOCA Resolved.- Peactor cavity pressure for DB-1 resolved. IIA-4 Rupture of high pressure lines outside containment Resolved. IIA-5 PWR pumo overspeed during a IfCA No change. IIA-7 Steam generator tube leakage ! No change; partially resolved by RG 1.83. t- ! IIA-8 10-year review No change for.DB-1 aO 77-a y i I
- .= . . .
, '. ;;.:.dh .
(g STA'IUS PEPORT DAVIS-BESSE 1 , IIC-1 Locking out of oower ocerated ECCS Valves Isolation of RHR trom primary system resolved, single failure criterion for ECCS satisfied. tb other changes. IIC-2 Fire Protection Progress being made by CB 1. IIC-3 Features to control sabotage . NRC Staff maintains this is resolved fo DB-1. IIC-4 Decontamination and Decommissioning No change. IIC-5 Reactor vessel suctorts tb change for DB-1. IIC-6 Water Hammer Steam generator water hammer not a problem in B&W plants. o f CONDITIONS PIMAINING CN OL LICENSE
/
A
- Seismic re-analysis of certain plant systems for a 0.2g safe V shutdown earthquake and use of Regulatory Guide 1.60 design re-sponse spectra.
. Evaluation of fuel rod bowing effects,.
. Inadvertent closure of decay heat removal system isolation valves during decay heat removal operation.
. Plant operating restrictions with less than t'.ree reactor coolant pumps in operation.
. Evaluation of facility fire pectection capability in accordance with Appendix A to Branch Technical Position APCSB 9.5.1.
CONDITICNS RESOLVED 'IO SATISPACTICN OF NRC STAFF, BUT ABCDEES PI-MOVING THE LICENSE CONDITICNS NOT ISSUED.
. Install flow measuring devices to monitor adequacy of boron dilu-tion modes of plant operation.
. Analysis of the reactor coolant system response to pressure tran-sient that can potentially occur during startup and shutdown.
ALL CUTSTANDI1X3 ITEMS IN SER HAVE EITHER BEEN RESOLVED OR A CONDITION O IMPOSED CN 'IHE OL. O jfp30
^
I Fig. 1
. . STAT!0f( LOCAT10ft5 0F KE0 WEE flET ,
APPEllDIX XXVII l 57'30, c Ac W G 3452'30 i . W l (P [h LAKE KEOWEE
' Ih l .
- a s
4 STATlOh! LOC ATlON . . W .
- SCC 50' i
'- 50t
- TBS
*TB2 K5 0 .
,to ?
d DAM K ,/ ad 4 7'30- + % FS. CONEE 1 % 4; NUCLEAR STN V eLRS !
.- 6 l
*ONH <
1 l
, < l 3445' j 83'00' 57'30" 55' ,
82'5, .Jj l 0 nm t
,jj m
~
N-aG/ l - . . .
.. .. j
e-
-- Fis.3 _y . 6 --. . l
; 57'30" ,
55' .__ g
.h
[]j 35!SMICITY NEAR LAKE MOWEE - y r3
$?
OZ 61km l l c14I54km r, cgc \
.$ }i l i- oyBS 56
- tj i
~
. b
*TB2 *K KS V'.
] '
A e o c a
%kh @s . DAM f.i n
U o e
.g m,
c / d[
- 3. t
% b ( b
( OCONg(N. V M. i - ~ d- + 2 jf} 7y ; o, . u s stRS *
\
c . t
*ON H l
j i Q i t - i I 12'c,o 57'30" 55' $2*$2'? Ikm j j oc4 tion of Seiwie activity near Lake Keowee ( December 30, 1977 - January 16,197?\ O _ ___. h
.." _= / I -2 3 A
.~. :
i " '
- g =. ' .', .
'f.. /,h _,* / fe, , *, .,'*
.. ., ,; h' -.*r*.,
. , # s' ' t a: .. * * *
-.'* .. .; e
r- , s ..,; 3 s.
;<.y .
l TABt.E 1 , e s, t N'JitER OF EVENTS IN KECWEE AREA - j ': p < g DATE NtM ER
- y Decemoer 28, 1977, . O J,.
29 2 < 30 1 ' 31 ' 28 January 1,1978 16 . i 2 . 38 3 162 l.. 4 109 L ' l 5 94 [ 6 209 I-7 179 I 8' 8 40 - 9 9 V ( , 10 ,f 55
!- 1 11 l 37 . -
- 12. 12 j 13 13 ,
l i 14 4 ! 1 i~
- Note: ':.e number of events for the period (December 28 - January 2). l are those recorced at the Jocassee net, wnile from January '
3,1978, events ce:orded in the v.eowe net wre used. e i I. t Ic E I l 8 ; ! i l i t l l d c., ., . /f ~ S 0 0 .' o(' . .-.- ,: _ . . . V - : 4. w. y,. v
,r,. ., ~ . .
,., ,,9 = ux.
= -
.s,.~.'..<.
- 9. M , .
. ., K '
.; ....p .s
./
w ... . +.. r . , , .f , , , i x ---_- --_. - - - --_ _--_-. - -l
y i ,. h ~~ l ?
- f "y - 4 ' t+ .i p_jJU., y i v _
6 i o ,. , . , 7..
. 6
.a c ~ ~
h
. r k
4._* , j I t. I ii', m . . ._._ s. >
.~-i ! -
- i er f i !
-_d t*
t i - i.s- ' ' - ' . r *.?
.ar
}
!,. i ', I..>j6
'8* 6 * " .1 e i ! I _. i e 3.... -l 4- . i. {
'. .f.
! r* !.t.4-l -s .
l.. i . i; t
.~. ...,t...
..- l . s I - d .s s_
t I.p 8 . r d .. ..s .), ! A i . . g .. .i - 9s ,
- y. 2. a -gwin: .,, p. 4 , . ' s 5 ! .t a
~
f- *
. . j . a e i .)2 t; - O. . ).. .f I<** 'J-l- t * - i ', ~l #,.**
- r. - -r e6-(.g i t p. s q . .. , .
- t- i -j . -r i i ,- . .- ew . - . - -4
(*/ j 1 .- . 1 i- I. .
. lI1(77b m.9 -6(J +. , ... .+r e. .i 1
g -I"'T N]-.;6 i%- - c , ' ; 1. - i
. s..-
h .h.,w , :.., r L . .C
;- -L,.-
+
N, N r&s.i. ,...!.i , i 1. . . -"-t .. i
- .6-- -
jy 1 J . .j+j.s#.-
. .j, [ ~ 2 .- , == f" i - 3 -
,e . -
e ._e,.
. ,3 .
'.. r m.. r.s a i . 1 ._
(. issh +' - . , * - .- e _;r i* r . ,a.r4 7.e a s , r r' ,- , . ! t r- r -- I-5: m.t . -i. , 9.-r
<h r sJ -(~,.4 t t a- ; - l l' ; . r l r, ' - 4 l*! 8 >*i * -
D . - 7 i'-1 r. l -I , k l's. -.g
# .. [ lei. i ."L 4 - [ J. .
.rp ' j. . p n -a...s.' [ 2; eq 24. .t-I.-. -l -
6
,b N1'-t , .q* : :,
' Q .r-f1+:
g ' ' fa o V. *. ' l-u S i.p...e-r .4 e pAa . . - gr. 4 s. _ -h-t. a + ;. .e;' . L.*3,
.. p-m; LS LtH. , ,
+].t ua -
,l.- . ,
e
- s. r . t.a. .
,! .i I-.. ~.e . . , r ; .
em .e e , e--m. * 't # l 5.
- s*i }1W / . ' . - -
- 3- h ,* f!- i . g _II .'W h ; , I _A . ] 8 d[ J . f !# % .e "i* .
f* *
*g
..t l r%" e b,d.'.e - ,-C s.
l'>
?- % }+.~f D i.j.j .! , j e. -i .k [#"[* '
J 9.M1 "n-
@t'..~ '
-- '- 4
,--+..w'
- 6-~
.r' *- I -
- el. ~l* sv'.Si. =
.A.p
+. s - - l- j. y A, . .g i .: l . '. s .- . ! y e've=}1 g --
t
, ,1 ; .
.A'5 -t
.:-* .: ' *s
- f,.s. ..
b.,4, . 4xv j ~ s.- i . .-o. -$<r-.
$1 J 4:- - . . r. e- T ..ww - _+
n= .n s?=***'-====..' -- [c se';e. . f s - ( i' & f
.. - - l . r ;
. .y 1-l .. 4
$ t- -
- 4 .~ - -
.1 .4t , - a4,.+.,',.r,.sa ^y, u =.n . . . s.e . I ws , .. s ?. - ..
L..r.r w. ' ' w! i g' 4..- -l : :1 1 y. e.i;,.:..--O
.e.
g 8:rt.a.ts 3 . w- *.' . .
. ~av ose " * *
-', .r1 N.
'O . .
- 1. . R- ' ' tu4 C * . t. -r. .**
. '- I / ' I :. F -l i :I4 .i-.. .
- w. .
C** -*t'L * ~- -- . , y rL=.a.i '.e1.,". -r ,f{ p%_. - ~ . .. ~
** Lau. i (.i -ra .i e .pr.m ., . i e n ~ ' ,: % w. s . -_s a i .
~
wr }- ; ..! ." W ?: 0,. l .. %(.i N .,,.' , ->w.. % 4n d !* ' .W-1**L e;ts. e n.. '...: . d W , p. .- h A.. 7. r H. "
. 'ry . p i
- - .s. :,_ r _ . y ,_- - . . rib r-s.4 . '-M rf i t-.*~. " :'.:.-. ,"ir.- Ar
~- - -
r m .*. .". ./ . . . p h. . . . ,r--s p.q... .- - N . ,, -- ,m . 2 g .7. >{ sp- :# ; .'- 4:a .ca e s s m- # m a.r
---, _n.m sus p reer, r+ *(~) H i cl. :* 1 t p j . .. .a -
'i; y ' j . 4 .%=r-
*r* --
et. : /..:ti4.r is :' i 4 s.
~ b r - ta i
- r ..]' yL. l. '.'.:..: ., t .n... [.;.*z * +*.H
- * - ~^
u-. %- .
- g. *' .='.~ .A.-
p p :s,e.E ' l '.-s r %'s b. a s +a, a s.\ k1-c t. ~ .e .' -. t.'s.. - l :'i >.- . H- i s < ' r . . W -t *,' . ,.3 5.,-! tr.r'. u.-l;' tl 9.-.
%3g~ ~
b .,W. v n.:l.-t r . .- l.. ; f *' d , f -i d - L.
+-1 ; i it a. i ..I . b. '. r; - ..
.1 -J E
- r q = .c. . s%r , : s vP.-t - .hl m.i-.4 9 ? r.d. D 3 -
h , @ l .fi r! ':- /.s --
- r .m t *W--*. .'a %e : 1 .- i
.. t d . ! . ! '
i - .-:: ' . ,
;.t .:.i -
; 4. ,.: w .:~..L.v M -~na.g.,
t.r-m : s. .r-r -
. . . . t ;+rt s. : q%u .J 3 %.-H - W.,. * -j ; h. *A,r***"*- -*-
-J-.---m,..;
9** 1-
- s. sm s st. =re is :
& . . . ;.rr.7.:,c
%. .~t, . * - -i . r _- m '. ) __ .
v i w! 4,, r - . -urr-e .-H F = e a s. =.ss
* .N . '. .
at N"'"O #t !M T d F- %C 8d #F ~9 ' *""88
' "P d*2N'-/5.*i" E ' - I -
h l h+ ~C.L*" .~7*, *.* j" p*-sji.A' .2 -U p, .;;" > ; +--, h NTy;).CECT ; M M c.- -3 ~dij s -4;Wlv pWi r./'T .vst a
~. G.'.4 Es i . .d :. f .+s.7'..
- b O'%W.::0-]=, m*.
n IW. '3 ;.i ? CH: . L = U-% '
- . : . a . .- #_sws.- . r, A.-v.
U W f. 2.1 + l s .r-C= ~1 br2 - < Oe 2-8%U. ~ i . ...' y . 4 . .
- m H.{w c--l'.;9:.?,i.m ' . .n.
. :. 9
* ..r .a:.t ,.. ; _ r u j w -,".- m .,. .q p.. .
.: p A sw.1.c _ t . . i ,r-r, .
h,gg,, (H. ;~.- g .w pe
- t. -,t.s.r =g .. 1 4:. .i .,
, 3 ,. ,v .
s-t YN , 6 k+.h.- sbf :irNg.. . f,... i /..? in.r tiiIj d__ea ,' b.he . -[ h : ' . e3}-all$ - F [Y . r 9 E
$" a -
t.;;r .*"+._Lp " %a' -l . -. i r .e l .1 - < 6- Mr4 ; +. L. gf_,{ r%.3 e if:5
,b 3 cr. .
b.it s .):rr.~.h r-m M. /.. ..r-TT~.l,.f.v : ~ \ ;s ' r 3r+ s 2- s :r-h =~ :-- k n 6 y * .ss .: -J - . y: - p r.-.:.6 =,. l i M i. r H in 1, y' 1.' - L.?- ::L.T s-
.'. ~.'
;^ '.
s u a, s.r ." y,. ,y - ,. '( .s e m m:~.. - -4 :. .i :e s- s "L . i s_-r riv.- u . :- f..a v v - r:m-. .:.f t P. .r~ Ge-*p-y r
- ..--i.<.=-- .y $p: t s x-a-- ~
"O
' .r-7dM
- - EU'h.u ,,;_$;_g%,Lp..sr nM-1.z-[.-
r
.~ ::-
NDw w. 4 . .c. YY. .s.."". . A i r.'..P
-;. -i t4 'riis--
. r Q J.r. . .-
'v. z i fa ~d { sh.-!= 4 s p_Y. ah:.dN.,
.
- M =l . ' q 9 m .r m. .. s e lf .' . e.: ,.
- e .- .Nr.' sa ..
(_ ,,JS .:-- .
.w. - '-*-t- .7.N _6-FJ ' -
z.V %. s=
- li :T r s;. .i tl~r e *J.
,2. I s * ~ %l "Je j l.?%.r.;Cli - r' .
1,- ' :b. ~ *-
- g Nz-
- ^
La,. .
. L
- O w. - b U r 5 srH r i-r: .m r .q'~ f p-..%
;.ame-tnrp.r_ g v '. a..** :. . 'r - } . ' ,, n y 6 i .- i - r . -; . l - -.:- . - t i tr. .Lr ,. . - i .r. .
*-li . .
r r y - .~~ry ~ +%pn-r .
'r .%.'- . sa"n ..
- 1. t s5.* t z^- l r te ? *ss i i s ' r . = ; . c
. sh h.a' * . t h n. i r- r i . .- V. .u zti. :. s v,.ri i g.
- 5 Li. . -+"J.: \ 2r :1, .s t- %
b-=-kamm=. I .r
--: f - * .~
*- ' * *J I
- j. vo == = 6 t "- :
ye '" s ~..c4 g... . I r; ; . p u ' . . s ve . ! -? -.-ti ' '" i -p.t#..
.~-'*f8'*- 4'\ **;3.f8 .
. 7 4 ' i -r.. . . r . ', s.f . ,. 0...r
*Q . .y . v." . 4 .,s h i c.i . J, a iJ J.. + z.r; o ,_ .. .eu.n -: c... ( : %.
c- i
, +.g s. . - - -
k7 a g :
- .; i.:-
$. i
- iat i -- f 2, .
I @ ensa.es:f- - y pre . Lm . -s ; s t _q t-- . no l,-t.- _c s- 1 1.. -.i.- ..a a p r . F., ia%. m . a .r i a t . l-r .i . q U.S.r.E-.
~
g T< . T*.l b . .! s . n..i . t..i .'Yi .r- i - i- .. ... a e 5 -
! i . a .m
=:t it : s g o ._.
u szw if.~sr.1 r.
- a\ o i .-* * - 1.;r..; r s .
i i . . . . . .- 5.- - .- - w; .j g s
. g', ,. . A- -.. y e -i (7 irt> t- ,
.i- .i -i . 1
-i i* .i. >
* . - ; , 37 a ll . ; -
. ., 4
.s i -
, . . , . - ,-= f .. y l
~""g.. y. : i .
I , . a . = i -6 1
.+i.-*^
e
.se g . . . .
- i
}
l i l ( ._jl a
! r
-[- { 9 e.- 6...} , , . . i gg - u l e i .- -- . .-e .., t .s s , !
i y l. !
- d. w 4. a N.l l e {. .
3 r l e a s I a/i I I
-g I- i
- t. 6'" 6 t
. .,.I.% - ... ..
i p
,,, i- l. 1 I l g( l W .?*.*
;e i e i i f .. ~ ( 4 ,
l
; i . i . i. i..
l w .e . i i 1 - , i i ;
. s. 1 i i i. . I i
4
. i * ! ! l ! (
! 4J l .i. , l 1- ,- l l l . ,
t ' I l ! l i * .i e 6 I e . . i g l ! - l % e l
.. t.. . . LM
, p. . / i t . g l 1 i- , e 1
p
. ' 4i aJ L { f rv j l I t e
# 4
}. .a
,.,, l .' g 'j 3 . .. .
,* i . i i ki e # . , , i . i
) ?.. t e g .. - ..
- i i --
r- t a (
- j. ----- al . ;
4.; u) i
-L 4- e i 8
i 4 I .o . - I i r i
.4
-4 =.. . ..
4-
,d e .
N.. . 1 % .
.. j . e l . j i l 4 & g e
- ,- { . . . .
a is 5,a. t . .
. i , i ; , ,
i . . i , - Lu .s. -- - -- : wn 6 - i
- i . ;
(
; y i N.y. o i , i e - . i e e
~
, . . I t i i ,
=9-*-a]- - - -- r*~
, {- *1 . ;"" i -}"~~"*
. .. I
**. i- 5 t, , , ,
,i { ,...'i g g e e.. h.
8 1 t 4 - g i I i , _i. t
. { - .1 S ._. ... _ __ e __ . . _ - - - - -
-- g. -
- f*
y, .. . , ,
/ /* e m, % - C l N M A 9 .:f C 'C W l
t gg
%g k: ' 2 *
- j j
1
- 1. . i- 9 -v { %. . ;., l ., - '..;~2)eaq l .. . ; :.,. , ! ,{ 4 ,; ,, i, ,, . . j
.,,/(; *,.a$,. - . O...} ,i .4
-{. >. g. *,- ; .j
- . .. , ;. g
( a f, . . q-- .g ; ; . .;
; t_
~',,-*, . . , . ,,.-t.. s -- .- I> , - - . ,;
- r. , 7.
e . . . , r< g.s .
.! e .
l .. i ..t. . e i ; i g- ,., i 6 .,, . . , ,+3....'- i t
._4. -
8 . l..) p . . t . , t- ; ! . ( ) ,,,, *g _4
. .+- .
i
.;, . .[ **.,
- 7, .. g .} l i j i
- r. . .,, f ..e--,i e
'. t.-
_ _ _i.
.,.. ...3:',
' g. ! I c,.
s 1 5 it .t
, t .
i
.l.,
j '..'
.; :. i.
, a
'f I ' J. - i i
.- l L
r -- a'
..; e i.
.'l r,-
li- I.,
,. r , .
.f f n ) .
- 3. - l ' 5 m,.., ...g* .
Q / .'._.'.*../,* . 8* 1 l* f s.'-tl
'I
, . , e c f8..fb g f _'
a .*s ef
._-i ..p O , m; rM,"+
- g .
. je )
,.CL ji j, i l { _
j- .- .g
. ,. f 4. . .t*
,j { p, 9, g .[ 3 g-.l lj .
t
.,h, , ., ,.Z.. >* l4,
- c. e i . h e . '. t j i . .,. p. .g. .
.'.g, l
3 y, L( ts ,-
.l ' 9 j l l l~ J ,.[.
'.I j .,=j 4), .~~p.. *
[.d.4l A * * * 't *.. { & ". l
,.j. .
- n. -- l { e
**".'"#i ., .
......o.
al0P.i.g, ljl. g k ' ** b _' -" l NJQ q i
- i e . . . ;. ; ) 4 ..
f.' I ".')y 'd,a j:* ' ,y*. .*. < .c .. * ,' b.: , . i.c c he . i. ; c.~ir~c 4 . ,. . , i. ,
- g. , g t .Wx
-3 .--};- ,: ..e 2. ,c ,%. 4. Q
- e. de g '
..** s. ., -- <
. . . f,, ,-.* - %w
?. c.., y*
.%-.. -gp.iy..p... a.
- Q.m*.. .
q ,. , ,,,r .I . g
, b,fg%. ,.*-,,
< , , yg g .;;; . . ., 9 , , , . . g . , ., a.
- m. .. .
_ t...,,,, g
,,.y -
, :, **3 .... . (. . .
- i; ; . O- -i.
-l 3 4 , , c 4 1 . j r, .,
' 'yy *,, g r.'
- g .", . ; .; 4 , . g . ;r., j , ,, g
.i. ! [ ,. 4. . ., ,. g.. ; ,.9 . . ;M e, , ,
M-t%,- p.g
, ^
O , . . , ; c- . ., .., . .t . .e W . %.: w o: %> ,.... -e .% . ., .N. . },. .1., {j.-.l .-". ..i., s { r.* .' .ut,1; kd i f g,; 4 J( l , . ;
,.. . ~ . , .., . . . . pJ. ,
/ -
- . _ --c
,, ( ;l aj ;:. J.r48 l
,""0 . 14 .. 1-./,..
. . _ _l 4:*
e
. - i.- .- . t i
@ :i..t *
- v. p . j . ; n .-
s
., y ,.wv g3. %, .y' ', ". i 3 ' ; * - -i - P---- -.7. , . . .
.,,.e.
,, . f g. L.;.v g . , _z ; . s. ,N,, q. . .. . . .ei.. .. 5. . . . } ,. . . ..c .,,...J...,...
- i .. .- l .. . 1 . , , ,
.e e.d. -gm. t. , .y r '_ ? . i-.,g- ..i. ', t t.:- .. p4
. .; , I,.' , .
..;, .~..
i
'i.. .r. ;
.,r. A ...,,,4 - ., t ,y . 4- ,. H . ! i <. . . . it
(- 5 . 4 - I . l ' '.
. . u .
. u. 1' . .
.- } . j -i.. .
J
'. 4.,,.,b..t
.;;4. w .:s ? i- -.,I j t .o ! . . ! . p i., .. ., . g . . ..,..
. f ." *30l, '. r i @. .b.
- M" P i .' 4 * '[ (J. i ;t:fl . :." l J* j ' .J. l 2 , .
. . l ,. { a.p.4.: j .. . e.. .
e l
,v ,,,,. R l.~~/. ..'
s ', ",: r% ' i;. T t..".s g 9 ; : ,i , ,,. i. , ,i . . ; , 3 g l 1
",..y.', yf.1 , l b f ~ E 4: i .'* j -- ., i . :4'. ,- .-
) ;. .- s-l.. .c t . , i- ,
. ye, ,. . ' .
. ' T,.g .
. g.g. .-i . . si t.s v-. *-
- p. .. q. . . } * - * .
rg
;. )
.n.'
. .A .. >- . &
/ .. ; . . , ;. 4.
1 y . - ) .A . ' ra .. t,p,.4 .
. . . q ,, i -
.p.
1
.j. ,, ., 9 -
p 4. se t u, q s .; . t &+c. ' .y 11 ' . . , { . .) 4. "' % r 4 ' a ,. . ; - 1-
.,.i,. -. - .l ;- '
12
' l., .
s .
- - -.. ' ; 8 'a
. . . .. J. rg . li ':_s! r. ' 'l r . . , ' , . ***.*
.. ' t,,;
I?* *.l..qeg, . , - ( a_ g, { . , . i g j "*
.. , n ajj' g/ .' .#
j[.,.-a,.' .,'.3-
. 6-
. . '_* A yM,P.< - H'. J6j- I' 4 j' . ). (a ; p q ,,, ', , ;,,, .; .,.. . 'ni . efr
. . r+. L . e - i.
ao . I s. . u- . ...i.- .. I. 4-.- .< i- . t . . :. . . . -
. . i. ...;.. t ,..a s ?. e
...: u, . wf . -
.~__:.--
. .. .._f, . .a>. . .
- u.aw m,.
..s,8-c p ,
\
. :.., p. ~..
, 1 .
- ..... ;.-r... w ,,.y.r i-._. u.c. :.,,.,.- .,,.a...
_ ,. - 1
. ,.r...,.=N.....
.: . ;,.,..m . ~,
y... .. 7 4. ..r.. . . , . . .. g2.: n y a.:.g. w _fi.-r.g,.ot.i.u....e
.__ a . s.v.,e m .:-q .e 1,4.,
F..i.,.....,p ;x ..,ye..L c;...j r.;..;1 .O
.- %q..,. q . e.p..; t3 w .r.- g ... n .; . g,e -
-J .
)
c r.w--. y.. =.. . i ....e 4 >..io. ..
. . . , - i..-
.. . . .s w - . . u. i . .
o.i qao
..y.y. -.. ..m , .m ..
,. .w ,i - ; A. . .-. i.. . ,. , . , J ,
w.. ; , y , . . w_ .p.. : .. -y t
- t. . .- p . q ; , . * . a , , c . a . . n. . r . .
.,-.I 4C 9'.*. ? y.$.. ' Q. 3 v! i . * ~ i.. " i- :l; i j-
- f. ! , q . } ,., l.;., .,; .,ac 7 g('[.",
iq,;r,J. t, y .. .'s c.. .. t. ; H t T. - l. .q.i . .. . }
. ..~. . ,. m l g y ,,.;;-.s ..., .
b.; j,.;il.K (.... F . t. { .,c , q . ;, . j i. e. , r, j , L g , ., } .c. .; y.. , ,, .4 . , ,q ,, l, g. cp . J.J.J. .t.p:
- .p g.e.m [..". ,
4.. ..r ...; . .. . ,. 3 - ..i; ,y ; z . . ,.. , .;. .., ,, t. ., g. g h1,*e*r g. 1). ' r * ?' * ;, . * - .* * - j.. v, f, ... t . .rj ., .j. s
< t ,.
P"I.F
!,...;..._ ; a. ;, - a. . c., ; ; p . c , , . . . i. f. . ;.,
P y.; . 4 , ; p ,, j *, H rj.9..'.r.;. . . '- i.: ., 1- G. T ' .ni. -
, y ;.j .- { .. . 4. . . . w-4 .r.-t
' p . i. rc } .y. 4 - ,f..gT ;;. ;.
i , ,,. 4 q y j ,, , , ; . , , f .p.;,'. ,
.?
t r ; Pc.--M i ; I _ : . .- 4 --+ l > .< A < j i - G.m. c., . .a e s 3 , . ,,. ! m.. . .e 5 4..@.
- p. lF y ya i-4-
i;.4 + , p . 4 > c. d-
- s q e.. i -#. p. . *' g* d {i- l I
... , ..r.
- 4. ,.;,..,-.. - N.d.ir .t .. l ..; 7 ..u. n. .:.. ;. ,n.. . , - - , .
- t=(,.,q
. t . . e. .3 s .ss ..+ .z.,,. .
p>') A.,%i-y". y . 3 . .,. . . . .
. t.) . t.. .S - p . 4-. j . .<q. . p .. v, 4 p , I ., . > .p,,. . ; t i
. ..;;. . . g. ;: .
... r , yn , i 3 A e .. i . , (,...;;.., 4 , ,i w 4 . .. g ,, . ,_f g ; , .. j ; ,, , ,, . , , , . . (, ,, ,((
%,;.1. ,
[.%-..
. w.r. . } - - . - -
t .r , n
.~ i .. s .4 . _. T. , ,
-. .a,3. <.
c .i- , N -:....- .e. i
..F ._.t.e l
-ac::. . .{t ' n.. .
g.- .,'
.e ! . :,. i . n ..g .C . p.
, d;. Q,.,g..i gp L r. W.-~ .1_. :. % ; a .c# . .t i...
.~ . w . . t . . ; . 3. . , ! . -
s
. ,. . , . , . i .;._... i u .
..L.,...
.n y
m 9:, m.<._- - w. .a w - . . . .,,.. . . :..,. . . .a. . ,, - o : - u . . . , . y . . : o . .. -... . . . . . , . .e- _ r-s .s 2.,a.., - r . c, %w . t.o.
.e. .-. c>
. . , < ~ - - r. .. . w..
.e,y m-~a,1 %.r -- . ..U. -w rso. . .:.:.a.m. .g.. .c. . ..m. . .. q. . .-.#..,.. ;;.3.,. 2 V4 .
- u,+. mw. ..:. ~ n.,.ce,, , . .:s.- m
.;t..,g. c.- c - .-. .w; ..:...'. .:w.m:-wm v.m. .... .Ew .n3
--L,,.....==?:f.. n tG.y,,,2 .C,g,Ly) _ . . , ..;;pm .-c
%. . v.
- - - - .. . .:. ,,.- .- - . m . .- .. ~ . ~ - . - - - .
- s. -
e
l l 4
. ;n. v. ; .
s.* . . ..*e.
. . .w .w .. ..%n c .. pv., s t.:.y..n .% *
- a. .. ., ., .~. t. g*en :.v ., A ,.,.. . ?; . t .. .sg. .
- v. . * . . . . . ... .
* .e,u.. . ..p...
- y. .* .n j.
=q.
,,=
\
[N * $ *
. . . . .. , . ,. . . , .] l
.u.. ,,...,,u
........i.....i.....-._.....__ .. . . .
... .. . . .. .l- . _. .... .
.. ..r-.- . ~ ..: . -
. ...,y
-f '
.. ....,t.....,.,-........
.r
* .. . . . . . , ,.g . 44 A..
. ...... .j.. ..
..l. ..
.. .. ., .. ,. ,g. . . . .. .-=-=.. - .
i ....4p. . a. y
., D .y,te . . . . . . . .. . , . . . . ... .... .e. "
l.
....,y g., ..>e .. &
r..... . . . . . , .
.. 6i .
.. +
f N P . , , , ,
.. . ... . . . . -. . p ..- . .. .. . . .. u. ...=..... .
l
- e ,
. . . . ~
.. . . . 1A, , . __ ... -
I
. .. . i .. .
..m... 1, .._.
.. . .....l..
.. . . . . . . . . . . . . . il..~ ...
. . . . . . . ....l .5 .
. . . .. .. .. .. .. .. .. .,.... _. .. .. .-...6.. -
- o. ..._.6 r . ..s. . .; . . . . . . . . . .,. . . . .
.4 .
.. .g.b 1
.. ...f.....,- ... . ...
..;i..~.. . p .. . .
..., . . . ... .p ......,...l....
. ..ta n . .. . , ,
i
.i...... . . ... .. .. - . ... '......g.......g-.a
. 6 f. (
......j......... . ._
I ,. - ..i .... ...~. . .
.....g..............
.. s . . .. .
... y8 3...
s .- -. .
...a....
.y.. .
.. I
. . . . .. . ...a.. .- ,
V/ . . , ,
..,.4 I
.L p
6 ....
- t. ...... . s..~......e. .........l.
e......... . . cy. g, 3 i......... p .
,# . ..- e.
...{
...6.**.**.'.**
. .* ..* .a= . * . . .- j
..-**..=.-
- t. s. .. .. . .... ' . s'
.A....
. . . . . . . . . . . . . . .,.....a t
.. i. l
. ... ... .. .. . . . . . . . . . . . . . . . . .... . . . . . . .. 4 O
w
. .. . . .. .... . 3 , . .. ~ . .; .
. . . . .... . ..i_....... 4 ...
...4.
>. . m.. .
. . i.
Q - i 4, .. . ...
- i. . .i...
.g.......1.
... # . , . .!. g ..
.. ... i. .. . . .
...~..g .
.;a.s u 6e
, . . . . . . . . . i
..; . . . . ~ . .. . . . . . w
. . . . . . ; . . .. .L .... w l . , . .
.....I.... .
.. . g ,
. .. . . .. . . a
. . . . . . . ..s... ... , . i.
. . . ail I .. ,.
. U g .
ii.
. .g . . .. . . . . , .. .. . . . . . . . 3
. . . . . , . . . . . . . . . . ll
. . . . . . . . . ..l. . ... -
,. .. . . i . ...j k . . .
..I..
.ip 4 .
( ... ....p... ..e. . . .. . . . ... .. D.= . s . , , . . . .. .a = . g.... a . g i . .,... ..
.i..... . . ...
.. . . . . Inj .. . .. .
- g. .
.j.
y
. ..... ~...
, .. . . . . . . . . . l
.. .. A...
... . ... . ... . . . . .. . g . . . _ ..
j l . . t e , . . . l ......,. ,.. 6.I, e .
- i. .. _.. .. .,,.
- i. :
. .. .~1.a O , .
. .6
.l
..l...g m
- a. ..
. . . . m .
* . .. . 4. g r
'4 . . . .. . '
l g . p., .. g.. l
' i.
.. . .= ..-. .
3
...I.
g . .
.a. ...
6' . ?. .- . . . . 4. .
, .. m.
, r .
.+ .
,4 . l... . .. . . t l , . . . ~
g. l. l .s
.g. j l ....
. . .e ...!
.a.'
9 . . . ... . . . .
.. .. . . . . . . s ...... . . . .. . . . .ll
. 1
,.. L . . . . . . . . ... .I,o .sy) . .
. .= =.,*. . .*
. . .. .s
, . . e. . e . . s .... .4.
... , g . . .
. ....e .. *
- j. a r
, . ,4
. ,4
. .' g
. . . . . ' .* I. ._ ' .
*I 3
.) . . ' . . . . a. ., .. . . la. .
.t j... .. . .l. . . . . . . 1- ..a,.
a.! ... .l... . . . . . 4.. 4 . . .i p.. . . . .e
. l... .e . . g, .
'. . . . 6 1
e
. .. . . ....s. .. . ~ . . .. . ...
i , 6 ., 8 o . -, . 4 . . ... .. .
- s. , . .a
.e .a
.e. 9
... ...e
. ..a
...........-...g _ . . .s..
., J 4
i.. .... _. . ..
.g i .... . .. , . .. .t.
. y ..
"j N ,
t > .$ I #"'
~9g
[] _. . . [ ' ' ~ .. .
~
M, i
, 9j i-
/ o
. g. . ;
/ Q -
. N44 f .
8- . t-o ' k e3;
/ 8 - \
/ g.O .'
/ e i.
t
/
f . 60 U i 0 .
.J .. . , '
i.
', a, ,
O c 1.:- e
' 3 O- @
,/ O ' <
\
N O O oj ,
\
O - m UJi
/ 0 0, cf
'O Ob
.O O , , ', '
O 1 . i e <
\
' O l
}A r p3.y,, ?
1 4QnQ _ l ; l
\ ,
I g (og M55!0NI conosite fault pane soluti C/' O h. Aggito45 Only event: IYi"9 I"9
- an[ in n I" '"i '
p+ _ _'q . 3 7 r o.. . -
.*~- , , .:*.f *=
,,, ( . .. ,. _, u.<. :
,4 Y" ,* E .
~
.. . , . a . . , . .e. . . . . . . , -
.y ..
9
- *s *
~ .
* ^ _
+
i I APPENDIX XXVIII CE Reactors: Control Element Guide Tube CEA GULDE TUB Wear G CHRO.NOLOGY.OF ACTIONS 1 i 12/lt/77 i Cracks observed in Control Element Assembling (CEA) l guide tube during fuel inspection program at t'. ills tone 2. f i i 12/15/77 !!ortheast fluclear Energy Company (liflECO) and Combustion I Engineering (CE) met with IIRC engineers presenting pre- l lininary finoings. We request general meting to . evaluate j the guide tube problem. , 12/19/77 F.2presentatives fron Caltinore Gas and Electric (CG&E), l Flcrida Power and Licht (FP&L), l'. sine Yankee Atomic Power ! Company (ilYAPCO), flNECO, Omaha Public Pcwer District (0PPP), I and CE met with f;RC to present liillstone 2 data on CEA guide tube wear. All effected CE licensees with operating facsilit'iec were 1 12/20/77 j notified to: (1) insert 1/7 of all CEA's, not fully inserad, ; at least 10 steps each day, and (2) provid2 rcquest for i cm:ndr2nt to allow CEA insertico ie/ond the prescat ' full out" position. ] 12/21/77 Engineering Cranch (EB) contracts with Id..ho Lat N .;l Engineering Laboratory to provide car,fi er,t v i ur s'; s- ! p or ihe wif" tube problem. ' d 11/23/77 CE provides proprietary version of slides used at 12/1E'7' nzc ting wi ch in; staf f. 12/27/77 Received recuest for Technical Specification (TS) charmes from all operating CE licensees. 01/04/70 BGLE provides inforration on necessary inhibits to alinw CEA insertion of three inches. Also excepts prcpered T'2: telecopied by sta f f. 0!/00/70 Arendments for Calvert Cliffs 1 and 2, Ft. L:alhoun, Imine Yankee, and St. Lucie were issued authoriziny CEA inser,ien thrce inches. 01/12/73 Second r2eting with facilities and CE was held. CE pro-sented iore ECT inspecticn data and Ascribed the pcc:ible tearrary fix; sleeving of selected CEA quin tel s. 01/lC/7C DDR reques t ISE to ensure the the intent of 1/0/73 am0*cn letter is irpier' anted at each f:cility. 01/17/7? CE pro'.id- non-m eprieta ry versic n c f si:dr . sed ct 12/19/77 cre c in ' '.;i tn the s ta f f. k] d;L3
CEA Guide Tube Pear I Chronology of Actions j m , b l 01/18/78 Twenty-day-lttters sent to BG?.E, FPAL, MYAPC0 and OPPD requiring, pursuant to 10CFR53.54(f), that " justification that excessive guide tube wear does not exist in your facility, or, if unable to assure that such wear does not exist, justification that continued operation of the facility would not create undue risk. .." f!"EC0 was sent a similar letter requiring " additional justification for return to operation. . . . " - 01/ 27/78 Encincering 3rcn:h (EB) contracts with Battelle !!ortinlost Engineering Lcbaratory to : t uvide confirmatory analyses of the guide tube problem. 02/03/70 DDR provides operating experience Ibmarandum I;o.11 on guide tube wear to DSS. 02/06/78 The staff prenoticed the " resolution of the operational problems related to CEA guido tube wear prior to return to peuer operation: for Calvert Cliffs 1 and Millstone 2. 02/14/78 All responses to 20-day letters received. (fioto: Scre respenses late due to adverre Leether condition.) BGSE and FP!.L subr:ited CEi;-79-P for the operating reactors, Calvert Cliffs 2 and St. Lucie. fit;ECO references CET:- 0 02-P for return to operation of Millstone 2. refen acos Cal f ar: Cli f f r.1, but it w:s n:t suh ,.t;d by (CEi;-32-P E.G!m . ) C73 rt ~' ', "rbt there ic no significant ouide tuM 'v ar in v' Fw' C,;heun Staticn L' nit i fuel M u miies.' MPC0 s: *..itt;l crates ;nat "the waun; o? r u i r; t e '. '2 war on I:ai n: i:.nkes fuel that is similnr to th.: fuel curry .ly resia..< in tlm reector is significantly less than the wcar experienced by 1 fuel in earlier cycles. l 02/15/78 CSLE ret with the staff to inforn us that they plan to sleeva the CEA guide tubes in 110 fuel assembiies. 02/17/78 BG".E submitts CE -83(C)-P cn "Calvert Cliffs Unit f:0. 1 Reacter Operation with Ibdified CEA Guide Tubes." 02/17/78 CE recorrends that EST.E place a hold on the handline of 14 selected fuel assemblics in the Calvert Cliffs l' reactor. I 02/11/7D BGII At and CE ret ccaciusion, thesstaith fstaff to anTier f in' arm BGBEour Lhetconcerns onoration onui:ltnin7
.h sleeved CEA guice tuba involves an unrc. ie..ed sa fety a;enica.
1 02/24/70 imECO and CE et uith tha sie ff to pn wot their pu n to I sl eeve appr <.i 'tely 3E guide tt 'as. Thu st; f f noti'is tht" ';I C: ? r.i t i 7'1 U i ' h a F I C 9 VQ d C Els [Uide tuiX , i n V o I '.' ? s p nn u .revien d 5a n ty qt .:I;i . . V l l 1 l
-95 l l
1
/
I CEA Guide Tube Hear Chronology of Actions 02/27/78 FPSL and BGLE, respectively, notified the staff that 02/28/78 "prelininary results of the CE enalysis of the nore - severely worn assenblies identified to date indicate that the stress criteria established in so e guide tube during the limiting seismic excitation (SSE)." They-also state that test program "results continue to support the conclusion that guide tube wear will not prevent CEA's from inserting following an SEE. 03/08/78 Nf1ECO provides additional infornation on CEA guide tube wear in:1uding Amendr.ent 1 to Cell-79-P, CEN-80(fi)-P, CEN-82-P, and CEh-83(B)-P. 03/03/78 fa:ECO requires an anendment to authorize operation of liisstone 2 with sleeves installed in the CEA guide tubes. The letter also transmits CEN-30(N)-P, "l'illstone 2 Reactor Cperation with Ibdified CEA Guide Tubes." 03/15/70 Ic;ECO submits additional information en CEA guide tube wear in:1uding Anandrent 2-P to CEN-79-P, C f -00(N)-P,
,O and CEf; 'D(C)-P and sleeving procedures used at Calvert V Cliffs 1 cad .'illst;ne 2.
03/lE/78 IN.ECO responds to staff questions and submits Anendment 1 to CEN-79-P, CEN-80(N)-P, CEN-82-P, and CEi:-83(C)-P. 03/16/78 BCSE response to staf f questions, including Amendment 1 to CEN-79-P, CEN-C0(N)-P, CE"-82-P, ar.d CEN-83(B)-P. 03/17/70 E" E recinest fer c ondrant to o;'erate with sleend CEl, guida tubos and to rem,ve all p: -t Itngth CE', plus res ;nse to s ta f f qu:stions , in:ludes Ar .:ac. . n' 2-P to CE"-79-P, CEN-83(N)-P, CEN-32-P, t.nd CEN-83(S)-P, " Additional Intora tion on Guide Tube Wear." 03/20/7el DG5E subnits revised reload anslyses for Cyle 3 operation, j 03/20/70 BUT,E reir,onds to staf f questions including resul tr of visual e:acinations of sleeved cuiu tubes and worse case Wear ohfer'~.d Ct CEI Nr* Cliffs I. 03/29/78 U.C p ro v i r.' :
- W ' 'h "t : sions on lo" ' cn , ..o '::
6.ti d r O ! UW #
- 100 tuDl ., itaClure EeCh3niC5, C,'.r? dn . 40, and r27;:Cd HitniC a 'n l y '? 5 .
03/21.'" ' , ' C i s '* " ' ' l, r ' 4 "'~%t '
' " !. w ' ~ r -l ? i n" "' n l . r N rJ; t i ? ',
U fc r Cal ' ci Clifts 1 ci
'E -
h * ' ;3 . l
#A P
, , - - - - , . . . _ . , _w
.CEEMIODF .
s r I , _ , v . 2 .m v . .. a - ! i
- l. 11 . ' . . . . L t . _ , . ._ . . i a L I .t .i I iy rt.r,, ia." a..a r) . u,p r
9 C, 'i . C nL' !, L': i e' i'la
- ;.n r, og.mt 1
]
C , Lml._ i . is r.L i,cFc o 1 6.".4 D -- E,'", tcJ,'c v ..i w. _c v 'uE- E"vi m"ra nvic "m.v. ,c ,r,. . . . .n. o. , i
._ ) . , _ .I . . , _ _ - . . . . .t_
i c~I.
- [.t.!r-pm:. . _ e
- v. p i ot iT,
; n', , i' ;iucnnDD nT't.;m_. .P.
^
'T T'. q
}. ' i1 D C_,'_'T. T f_.11.
. t. ! D,,E __
-- ^ - ----
~~ n bi 5 l . ,i L . . '.] r- ;
Ce"' b ' 'l. a '. kL.:c ' . 'I ~ r. " .' [ . i s t L)
~
m' [ t, j.. } d' ii far r j r- u..:, p I th ]L . . L ): .t. .L. l l l i l i l l l I i l 1 , I I 1 i I l 1 l /93 /
O
'. g
.d L ,.,...W.! O M d ! I' ' .I
%L h _
f f#* $* P 9 fO I*f8 I I (" f a* 4 [ $ ft '[ l I g i*Off' I C e . u '~. Cb . un fa i' k k l l a li d eindbati Ub / h i V. i -) ' I
.. . . , i .r i s '
) Iidii l li d . l ADn. t l T I C",,e,l
, . D.6.T A F".~.'o".
F..U..N,r. _ v,,t0...u.
"rc 0" . Wu ","_u
. F Ui_ti_ l i
,*O "
Of f T}* [9l f O *, * *
, c a Ei ,s L . f #Sc d VOa. RIOt2 , 'O,!"t ti $ l C t i a e . iv vu,108 . . .
j 'f r J e J ' ' t <*'t l'.'.s AT v' lTiTJi Vi
' I r' F r' t .' r*i k L f L e' T 7 : :eI . ,] 6 t' --
ciul{pum nu !".1 '1. ' ', u' un v Q . 's . v:6 J Vi, 0 r s1
. . t. ,
O ,7. , y i7is. n lid /'tb lid i 1 1 4 I I 1 l 1
)
1 0 1
- - 9 ps 1
_m___ _ _ _ _ _ _ __mm____
O . l FUEL ASSEMBLY GUIDE TUBE INTEGRITY I. PROBLEM II. DESCRIPTION OF GUIDE TUBES III. SAFETY AND DESIGN CONSIDERATIONS A. LOADINGS B. COOLABILITY C. SCRAMABILITY IV. WEAR OBSERVATIONS A. NON-DESTRUCTIVE EXAMINATIONS B. DESTRUCTIVE EXAMINATIONS V. INTERIM FIXES A. SLEEVING
- 1. DESCRIPTION OF SLEEVE O 2.
3. STRUCTURAL ANALYSES SLEEVING PROCEDURE
- 4. SAFETY CONSIDERATIONS S. TESTING B. 3" INSERTION VI. BASES FOR CONTINUED OPERATION A. PLANTS WITH SLEEVED AND UNSLEEVED GUIDE TUBES B. PLANTS WITH UNSLEEVED GUIDE TUBES VII, SUSCEPTIBILITY OF OTHER NSSS DESIGNS TO GUIDE TUBE WEAR l
I l i ,
#W
4 FUEL ASSEMBLY GUIDE TUBE INTEGRITY L. PROBLEM . A. ON DECEMBER 13, 1977, CRACKS WERE FOUND IN THE CONTROL ELEMENT ASSEMBLY GUIDE TUBES OF THREE FUEL ASSEMBLIES AT MILLSTONE POINT UNIT NO. 2. Q II. DESCRIPTION OF GUIDE TUBES l 1 1 O g-avv
,\
Nm,- 0 . 9 t. 5 i n~'ie s . CEA 00 0.950 I ' Guide Tube 2irca./oy.4
- 10 1.035-00 1.115 i i Wall 0.040 l .l
- ~ '
I c 0 nr ,, . r . l ~k
// 'a 4.94 i i
le I J I' . \l i li 1s.7s : 1 , TriCrnfin,
; - - , c. j 8. ,, ,,
! ,JUeljd,uaj c I e ik i l
n, 7 ! ij ' i,' ; ' l
)
l 1, ' = i a
, l 'v/
e i : h .i " v ! f , i i 707 ACT:'t! FUIL
- 'i i ;
' : !' i F, ,
s l' l lt o c . i 157.23.1 , 1 ij ', j ' . , 4 i <
', ' 't', i 1h j , '/
\1 V M.v, ,i.m , , .
!, 'i
. i.i(
I li l
, i i 3 $XIAL SWTl0tilfiG In' i
'%.. /
/ -2A
i l \ l &f * . .Y$.f. g Q ~'; 'Ihurface > *'
.,., 9' g7qd
. . , e ..
gf*: ..
^ 7c ~-
_..:. .r,.~
~ . , - 1
. .;y -
s a . -. ~
-.; c ., ; .
..M
- , . m .m -w ~.
? : -g k'
- . -.. ,M. . r.t g -- -- = -- X654 0.3 in.
sc . - - 2..... 1
. h. _ ' . {d.i.k. k *~--
;= = l N~ .m* -
w z : - .3:3 -- J w,+,ms i
, 2. .. :s-
~-
-. P.!Q*-
s .. q y,n)
- v. ~. -.., . .a ..q.. , . u U, i.
.. . . - M.:.i "' $
-. w . ... ,
g:.. - - - ,. ~. ,c:,a s .s . n. ..;
&$. . ' ' ] ~.f.".*E y+?.; . .- . ~ @j ga .. w r ,.. %;m+ . m-ti'C.. '*
- ::- v.
M,. ...i T. : .' .. ,J.. %. . 1
. g n
gx - - - * -
....r
~
.'g T,4 . ~
O
.h._f x. .:' .. - u = z ..
1st Surface > -
;3;.w..n.. A w;4.:.3 ~
z , g : .- e :- -- --
, ~;;. ;. q.
. . .-- Rn. ,
f5 M 2.?YM .q
- - -m P - ' r Z~ , . X653 t#?5$
~
l t I _. _ _ .E. t @ --
.;,4 .:ce* ."
7- _ _ - __
...,.:.-_, R',Q p:
- ~ ~ - * .~ .- -a . . ,:csw ys .
yz 4 ,. . r f. -
> .,g. . .;: ..g M-i. ,
.~,, p. ~ - ,
g.i
- c. ~.
- m. . :
1 a> ', dj
- .;, l
. , 'S &
.t r
. m. '.
y-{ y,
=-
(
%:+:: .
. ... %g- v. - I i
tp. . , ~_. r...
~
. - .' e*,~ . a .
hs.;m __ . .,.. .. . w .c -th C-E 3-5 " O f l OD Accearance of SE Guide Tube Wear V h - 3 YS -
.g w.. ,. p..
s- - - m ;- - O. u2.. r,;+... .
. .- ~
- *4s;'4.~h: +
~
r i,.. )
~. J ' n*
.r$ [-P - .
g U2
;ies,5, .z.
- 5 ..y... c,g~lYyq, y .n . ..~' 't ' -+- -
.,*.**7 . t % , . ." -
~ '
[', -a ',* Y, - h*5 '. e .. . m.e M...o+..-c . . :. .
=
- i. n , ~: .
&Q&+- '.g:~ . .+.,i -
4
. .y ( ;.
~. s n,-
p-
-4 7;qas * = $.. - -
': ;9p I., t_6, ___. . _.? c, eru '-EI. m m. :4.. :r cgq
,# .g . ,.n.:. . e. . . . .
, 4 .; r? . E,. ^<l '. - @ . s' ?ls
- :;.t. +Q<5,Q .:
$a., b , ~ .. '. '."'( L :' :(e. 1
~
?.Q Q . -
, . ;.,
- 5 ; _ g .. 7.,y.5
;q ~,,,,a
.',,,7 :, ,. ,
mntf. v _ g 1irl L E R 9,u. t i(? ; as,. .c::
> .t .
O -.;# ., nit <-.y-% ..+. . . tv:
~
z 'r
..s ~
e
. ; 3-
' x \_ ,,.: _;: ,
~
, a. .. - ..
;.gg,.R t
.. .m 4 ,
,,_w
.e,. ~. -t . , _ . f. .. .;+. ....
%%k2 .8 Q:W g% gl. 2 : f % .i ll W P 2m.N. , .: .; 1 f -
h
--l ,
C-E 21-22 m 0,3 m p- q Montage of SW Guide Tube ID Showing Wear " Opening After Clamshelling R- W7 _ __ . __ _
4 y *Aa. 0 E e e O ; I I I D I I "O 9 l l I I SECTt0ft 30 , g i
'l 1 -j i 1 g
4 C . l O l 1 1 I I I 1 B (_ _s - - J ,.,
. i - --
7 l j 4 l SECT!0.'t 6 3 l 1 O m' i L!W, ! _ I SECTtC'I -a
?It#Esti ATION OF '.!Att 3. C*J::E IL3E -\
g.2yr
e O III, SAFETY AND DESIGN CONSIDERATIONS , A. ScaAMAart17y B. C00LABLE GEOMETRY O l l O R- av7 i
O l IV. WEAR OBSERVATIONS - A. NON-DESTRUCTIVE EXAMINATIONS
- 1. EDDY CURRENT TESTING
- 2. VISUAL INSPECTION WITH BORESCOPE AND/OR PERISCOPE 1
B. DESTRUCTIVE EXAMINATIONS
- 1. METALOGRAPHIC EXAMINATION IN HOT CELL FACILITY O
l l l l l I [. 2 3' -
O EDDY CURRENT TESTING AT OPERATING COMBUSTION ENGINEERING REACTORS NUMBER OF GUIDE TUBES TESTED AVERAGE AZIMUTHAL I PLANT WEAR PROBE WEAR PROBE MILLSTONE POINT #2 474 120 MAINE YANKEE 659 93 CALVERT CLIFFS #1 623 291 FORT CALHOUN 140 0 l O
/7 W ^
i l
O V. INTERIM SOLUTIONS A. REPAIR OF WORN GUIDE IUBES BY SLEEVING .
- 1. DESCRIPTION OF SLEEVES
- 2. STRUCTURAL ANALYSES
- 3. SLEEVING PROCEDURE
- 4. SAFETY CONSIDERATIONS
- 5. TESTING OF SLEEVED ASSEMBLIES B. 3 INCH INSERTION OF CONTROL R0oS O
i l 1 l g- 2s x ' l
CONICAL SECTION I{/g
, l ,' \ \-
t i l
'v ' .l, i
I h 55 E.'1D POST SS .^L__'/:_ -l l ~ l l l { SLO ~5 I e- Ai!NULUS s i I (NOT TO SCALE) , I I i 1 ; i l, ** iia; l} .. Li l Ow=ED
, si' 1
I fl I lij l fill ii] p' '. n in til .. y l , i ~q ir' i
! ! t0 C"': Ng I
- . .,__J
,x_,)
I . I
> l I
i i, Y $5 #9j @
,gp
- as 9 9 9 Cr L. Ii ud.%Ost o. . :.
'l
}l \
l * .to n at m e q .==,***sy j f*l '\ L Ar ML { LL w DC'.,iVii % ,. 3 j/ 4 h l I h /
== j b
g
/g.e*
s aut i
- q fa e
,, s :, , y i
jl , I l J i i _ mI , I\ t
~ ~_
A
? ! %..J f q' I I( f #* 4 9 4s eea ,p.m ma-w r. . . ,. ,; .* ; ,;; ,;,}}