ML19270F495

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Minutes of 780525 Meeting in Washington,Dc to Review Request of Util to Operate Plant Up to Power of 2,630 Mwt
ML19270F495
Person / Time
Site: Maine Yankee
Issue date: 05/25/1978
From:
Advisory Committee on Reactor Safeguards
To:
Advisory Committee on Reactor Safeguards
References
ACRS-1540, NUDOCS 7902140327
Download: ML19270F495 (60)
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Text

{{#Wiki_filter:* a %, r, .3 iJ ] b MINUTES OF THE ACRS l Ml6 ).O e0 a.J b. aL, SUBCO@ilTIEE FIETIIG a ca MAINE YANKEE NUCLEAR PIMr N $YO Na 1GS C. /D /d a/>/7c The ACRS Maine Yankee Subco=nittee held a meting on May 25, 1978 in Room 1046, 1717 H Street, N. W., Washington, D. C. The purpose of the meeting was to review the request of the Maine Yankee Atomic Power Corp. to operate this plant up to a power of 2630 MN(t). Notices of this meeting were published in the Federal Register en April 17, 1978 and May 11, 1978; copies are included as Attachment A. The schedule for this meting is presented in Attachment B, and the list of attendees is Attachm nt C. Mr. E. G. Igne was the Designated Federal Employee for the meting. The entire meting was open to the public. EXECUTIVE SESSICN Dr. Kerr, the Subconnittee Chair:ran, stated the purpose and scope of the meting. He noted that the items to be discussed would include safety-related topics associated with an increase in power. The topics would include mteorology, fuel performance, and transient be-havior of the plant. SITE DESCRIPTION (P. Bergeron, Maine Yankee) f Maine Yankee is located in Wiscasset, Maine, in Lincoln County. The site contains approximately 740 acres. The minhaum exclusion radius is 2000 ft. The low population ::one is 6 miles. The major pcpulation centers near the site are the twn of Wiscasset (4 miles north, 50-309 Tloa19o32 r [

r population approximately 2000), the town of Bath (7 miles southwest, population 12,000) and the town of Lewiston (26 miles north, northwest; population 45,000). The low population =cne is expected to have 2 pep-ulation density of 63 people per square tr 2 by the year 2000. Slides 2-6 in Appendix D contain additional details on the site location and demog-raphy. PIANT DESCRIPTION (P. Bergeron) Maine Yankee uses a Corrbustion Engineering NSSS, not a standard design. Each of the three loops of the NSSS contains a reactor coolant pump and a steam generator. The reactor core is composed of 217 fuel asse:rblies, and 85 control element asse:rblies (including 8 part-length rods). The fuel used at Maine Yankee is the standard Co:rbustion Engineering fuel,14 by 14 fuel rods per asse:rbly. Slides 7 through 10 of Appendix D illustrate additional details re-lated to the NSSS, core instrumentation, and fuel design. LICENSING AND OPFSATING HIS'IORY ( P. Bergeron) The chronology of the licensing and operating history is presented in slide 11. The original core used unpressurized, low density fuel. Mr.ine Yankee experienced scun fuel leakage during its first cycle. The leaking fuel assarblies were replaced for cycle 1A (the second cycle) oporation with pre-pressurized high density fuel which iJrproved fuel performance. Before the third cycle began, the total core was replaced with pre-pressurized fuel.

s, .r . Currently, the plant is operating in the fourth cycle at 2560 FA(t), the FSAR stretch power level. The NRC approved a license amendment for operation at this level on May 15,1978 (slide 12). RELATIONSHIP OF VARIOUS CORE POWE:. LEVEIS (P. Bergeron) Mr. Bergeron discussed the relationship of the various core power levels presented in slide 13. Primary system pressure, flow, nuclear peaking and core inlet temperature are listed with their corresponding power level in slide 13. It was noted that in applying for the higher power level, credit was taken for the neasured primary mass flow rate which was greater than the design value. Credit was also taken for the fact that the power distribution is limited. INCIDENTG CONSIDERED (P. Bergeron) It was noted that the entire set of accidents and transients considered in the FSAR was :e-analyzed for the new increased power safety analysis. There were three categories of incidents considered (slide 14). The first category of accidents and transients considered affect the core, but do not release radioactivity to the environment. The recond cate-gory considers the potential for the release of radioactiv ty through possible ruptures of the primary or secondary system. The final category of low probability of events considered has the potential for releasing radioactivity, but is not associated with the core. RADIOIDGICAL ANALYSIS (J. Di Stefano, Maine Yankee) The analyses performed for the 2630 MnT(t) increased pwer level were cor:pleted assuming a core power of 2683 MN(t) taking into account

1 . an uncertainty factor of two percent. The power level of 2630 !G(t) was chosen because it corresponds to the m ximum capacity of the turbine generator. The re-evaluation of design-basis-accidents for radiological dose assessment also reflects current criteria for eval-uating DBA's presented in the Standard Review Plans for the applicable accidents and uses the most current meteorological data available at the time of the application for power increase (data for the 1975-1976 operating period). Four DBA's were re-evaluated for the 2630 !G(t) operating power condi-tion. It was felt that these analyses were the ones most affected by the power increase. They are: Steam generator tube rupture, loss of coolant accident-dose assessment, main steam line failure outside containment, and fuel handling incident (slide 16). A number of major parameter changes were identified in connection with these accidents. They included the following items listed belcw: 1. Increase in primary and secondary activity levels based on higher power operating level. 2. Evaluation of the radiological consequences based on pre-existing and coincident primry coolant iodine spiking. 3. Radiological dose consequences based on latest atmos-pheric dispersion factors at the time of submittal. 4. Decontamination factor of 10 applied to iodine re-leases from affected steam generator (in connection with a main steam line failure outside containment). 5. Increased fuel assembly activity inventory based on higher operating level. e 1 9

~ . 6. Increased core halogen and noble gas inventories based on higher power operating level. 7. Factor of two reduction in the elemental iodine removal rate used for the sodium hydroxide spray system. 8. Credit for particulate iodine removal rate constant for the stretch power submittal. No credit for particulate iodine removal by the spray system was taken in the FSAR. 9. Post-ICCA hydrogen purge dose contribution was calculated based on the increased power operating level.

10. A post-IOCA engineered safeguard feature leakage dose con-tribution.

Slides 17 through 24 list the various postulated accidents considered and the associated parameter changes, together with the resulting doses. The results cale. lated by the Maine Yankee Staff were below the 10CFR 100 dose limits. Slide 25 shows the X/Q values used in the recent analyses and 1977 data are presented to indicate the consistency of the meteorolog-ical data. The X/Q dilution factors presented are not expected to be ex-ceeded more than 5% of the time. Slides 26 and 27 show the parameters and results from a revised IOCA analysis. This analysis was done after the stretch power submittal. Major parameter changes included the following: 1. Revised primary containment spray model. 2. Assumed primary containment leak rate = 0.10% per day ( previous assumed leak rate = 0.15% per day ). 3. Radiological dose calculations based on 1977 meteorological data. 4. Revised hydrogen purr dose contribution based on additional zinc corrosion rates. The doses calculated by the Mair.e Yankee Staff were again ' elow 10CFR Part 100 limits. c

. SAFETY ANALYSIS ASSCMPTIONS (P. Bergeron) The plant is restricted to three loop operation; it is not allowad to operate with a reactor coolant pump out of service. In some tran-sients the nuclear peaking factor was reduced from that used in the FSAR analysis, to achieve acceptable performnce. The plant is limited by technical specifications to a nonpositive moderator temperature coefficient. Control assembly shutdown characteristics were reviewed. (Slide 28) Reactor protective system trips were reviewed. The trip setpoints, setpoint uncertainty, and delay times are presented in slide 29. Additional trips at the Maine Yankee plant include a thermal margin low pressure trip, a symmetric offset trip, a rate trip, and a variable overpower trip. STEADY STATE THERMAL HYDRAULiq (P. Berceron) Slide 30 shows a conparison of cycle 3 2 eady state thermal hydraulics at 2440 mi(t), and a power of 2630 31(t). It was noted that the same safety margin to DNB is calculated for the cycle tiree increased power case as was present in the FSAR analysis. This wat made possible in part by taking credit for the actual primary system flow rate which is greater than that assumed in the FSAR. Mr. Bergeron presented a summary of accident analyses in slides 31 and

32. The summry compares results from the 2630 Mi(t) analysis and the FSAR analysis.

It was noted that the methods used in the 1968 FSAR analysis are somwhat different than those performd for the analysis at 2630 MW(t). The limiting anticipated operational occurrences for Maine

. Yankee are the rod drop for linear heat rate, melt concerns, and DNB, and the loss of load for over-pressuri-M on. In the original FSAR analysis of a rod drop credit was taken for an automtic turbine runback feature (which throttled the turbine to 75% of initial flov). The system was not a Class I-E system and credit was not given for this feature. In the present analysis, the limiting set of power distributions is applied along with the associated uncertainties. To predict the limiting power distri-bution, a set of possible xenon oscillations is calculated and the power to DNB is predicted as a function of symmtric offset. The peak pressure predicted in the analysis of the loss of load inci-dent is now higher than the FSAR predicted peak. This is due to the higher power level, and the way safety valves are nodeled (they are assumed not to activate; pressure is assumed to be reduced by a plant trip caused by high pressurizer pressure). The peak pressure in the FSAR was calculated to be 2490 psla; it is now calculated to reach 2540 psia. IDCA results were presented in slide 33. Results of both large and small breaks are summarized. The Appendix K criteria for peak clad temperature, clad oxidation, and maximum hydrogen generation are m t. MAINE YANKEE REAC'IOR PICTECTION SYSTEM (P. Bergeron) Mr. Bergeron noted that in addition to a number of other plant trips (slide 34), Maine Yankee has a thermal mrgin low pressure trip and a synmetric offset trip. The symmetric offset trip has the principal

. function of protecting against exceeding a linear heat rate of 21 kilowatts per foot. The symetric offset is a measure of the axial power distribution (slide 35). In order to reach higher power, the thernal margin low pressure trip system was modified. The modi-fled system contains an additional function that relates Dra to sym-metric offset. Plant operation will be somewhat limited in that the operating band of the symmetric offset trip will be narrower than it was for 2440 MW(t) operation. The nodified trips permit the assumption of a less severe axial power shape in transient analyses. FUEL PERMR'4ANCE ( P. Bergeron) Fuel performance characteristics are summarized in slide 39. It was noted that the effect of higher heat flux on fuel temperature was acceptable. The iodine activity in the primary coolant was displayed graphically for core 2 and core 3 in slides 40 and 41 respectively. Peaks in iodine concentration were attributed to plant load changes. Core 3 iodine levels (slide 41 right hand side) were shown to be lower than core 2. EFFECTS ON MA70R EQUIPMENT ( P. Bergeron) Slide 42 shows that maximum expected operating conditions ( temperature and pressure at 2630 MN(t)) of the plant fall within the design parameters of the equi;: ment in both primary and secondary sides of the plant. Analyses were perforned to show that safety-related equignent would

HIGHLIGilTS OF '1HE MAINE YNJKEE SUDCOMMITTEE MEETI!G MAY 25, 1978 HASHING'IGN, D. C. 1. The purpose of this meeting was to review the request of the Maine Yankee Atomic Power Corp. to operate this plant beyond the FSAR power of 2560 FM(t) to 2630 IM(t). 2. The Maine Yankee NSSS was built by Combustion Engineering. The plant is a nonstandard three loop design. 3. The plant is operating in its fourth cycle at 2560 M;1(t), the full FSAR rated power. The NRC granted permission to operate at this level on May 15, 1978. 4. In applying for the increased power level, credit was taken for actual primary side mass flow rate, which represents an increase over the FSAR assumed value. 5. Credit was also taken in noving to the increased power level for the ability to control the pcwer distribution within a symmetric offset level, which will be core narrow than the FSAR level. 6. The stretch plus power level of 2630 FM(t) represents the maximum capa-city of the turbine generator. 7. The current safety analysis was performed using current NBC criteria. 8. Maxinum expected operating conditions of temperature and pressure at 2630 FM(t) fall within the design paraneters of coth primary and secondary side equip: rent. 9. Since initial operation, Maine Yankee has operated with all volatile chemistry. There has been some denting in the steam gener tor .mes, but it was considered minimum at tnis stage.

10. It was noted that the meteorological data collection, control room habitability analysis and ESF component leakage outside containment analysis were found adequate by the NRC Staff to support a power increase, but warranted further review.

O O

ATrAcapen7 A NO11CES 1r. portance and Interest to the rencral with representativcs of the NRC Staff, FOR FURTIIER INFORMAT-public to be announced in the FEnrHAt. the Maine Yankee Atomte l'ower CONTACT: REctsvra as a mec attendance and obs, ting open for public Corp., and their consultants, pertinent ervationc to Ihis review. Mr. John E. Finlay, 202-:45-414; Decause of space !!m:tations those The Subcommittee may then enueus SUPPLEMENTARY INFORMAT-who wish to attet d should make prtor to determine whether tiie matters The Postal Service propous to r-- arrangements with the Chairman. Dr. ident! fled in the inttlal sen:fon have Richard A. Lesh. 202 232-7745. or by been adequately covered and w hether a routine use of USPS 10 0'O. - letter to him.t the Division of Science the project is ready for review by the Clainu--Tort Claims Recoru e.: E Education D ihement and Research, full Committee. add the same use to USPS CD 0; spection Requirements-Invest:. Science Educat ra Directorat e. Na. In vidttlon, it may be necessary for u

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tional Science F,. Nation. Washing. the Subcommittee to hold one or more y;le System. One purpeee ci closed sessions for the purpose of ex. cliange is to eliminate the Trrd ton, D.C. 20550. r. M. Rr:TcCA WIMtE*t. ploring rustters involving proprietary t!stical" from the text of the r-Comm:.we Management Informatlon. I haVe detcfmtned. in MC. . The.%N. Coordinafor. cordance with subsection llhd) of Pub. vice.W trad!ttonally prondM ce-I. gg.4G3 that, should such sessions dera t information to the.L.e.- ArntL 12,1973 be required, it is necec.c.ry to close IU# f ' ca Assoetation Indes ts : '.z-fFR Doc. 78-1CT ; n'ed 4-14-78: 8:45aml. thee sessions to protn

  • r roprictary a e.

ative effort to deteL:o - 1. rmation (5 U.S.C. 5L ' a 4 )). frec e:es and patterns of sc-2 .urther informatio*: vgarding anu .:ics on a nationa;.va'e. [7590-01] tc:cs to be discussed. ..etner the the ; never intended. the vert - NUCLEM REGULATORY meeting has been cancelled or resche. tist:- was inadvertently inc.. duled, the ;hairman s ruling on re. wit h" USPS 180.010 as ort.:11; 7 CO MMISSION quests for the opportunity to present scre a n the FCERA!, RrG!$ D. 7 oral statements and the time allotted addu en of the use into US?S " ADVISORY COMMfE ON REACTOR FE* therefor can be obtained by a prepaid GUAJtD$ $UswwfTIE ON THE NE telephone call to the Designated Fed-refle* R the fact that infc: a YANKEE NUCtIM PLANT eral Employee for this meeting. Mr. disc:c.ed from that system as $- Tinse proposals do not re f'xt Elpidio G. Igne. telepnone 202-G34-1920. between 8:15 a.m. and 5.00 p.nt, chance in disclosure p 1:* ires. ' Die ACRS S- 'mmtto + on the e.s.t. rather rnore accurately w :.te. 'laine Yankee b u P! ant will hold Background information concerning long standing practice m a meeting on * - 2.19"3 in Room items to be considered at this meeting fomation to the Amer ' .-v "- - 1046,1717 H St: NW., Washington, can be found in documents on file e.nd s>sa"clati n Index Syste a res ::c. d'3 D.C. 20555, to > w the request of available for public inspection at the s of mcom the Maine Y-ee Atomic Power NRC Public Document Room.1717 H A amplete statemer: et the er Corp. to operat - ns plant beyond the Street NW., Washington. D.C. 20555, ter.ce uut character of

  • m s7 stem :

FSAR power t,.3C0 MW(t) to 2630 and at the Wiscasset Public Library, pearr a m the F:mrA.:. arc:s ta MWt t). HIch Street, Wiseasset, Maine 04578. S* V. - t-30.1977 O. R MOT). In accordane-Tith the procedures Dated: April 13, 1978* outlined in th0 fDERAL REG!srER on este i Nrsons are' m.- 'd to fu::- October 31, IST: page 56972, oral or SAM EI. T. Cmtx, wr;. data, views c-st;;=ents written statements may be presented Secretary of the CJmmissfort the posed routtr> use cht: p by members of the pubile, recordmgs fFR Doc. 78-10470 nled 4 14-73: 9.33 amj FM c Ntice will be pn dished af er r will be permitted only during those portions of the meeting when a tran. time for public comir act has ela::ct. script is being kept, and qu%tions may W 10-171 The specific change roposed fcr : be asked only by memuers # the Sub. records system descr stion fcr CS.1 committee, its consultante. ed staff. 2057AL SERVICE 180.010 follows: Persons desiring to make u state-ments should notify the. runated Tyrcy or tNFORMAT10N Roudne tisu of records vintained in t system, including catte s of pers t. Federal Employee as far in uvance as

  • * ""* E""d8 Mad 3 &*'i*n the purpose of such use :

practicable n that appropriate nr. A rangemente. - be made to allow the s;CY: U.S. Postal Service. Change second rou-P to rw. necessary t. furing the meetinet for

h....O N: Advance notice of a Pro-

"2. To provide mer N

r.: A= e-st*:h statem:.

pmd Routine Use Modification. can Insurance Ar g { qqgy'-{ ~ The agen.:a for subject meeting gy g shall be as follows:

SUMMARY

This document proposes ing to accidents and inr2 ries."

a change to a routine use for a pub-TUESDAY, MAT 2.1978 lished system of TcCords and the addi-The addition of the routine use a tion of the same use to another pub-U3pS 000.010 follows 8:30 a.m. unt!! the conclusion of lished system. The routine u:e as pres. business. ently reported refers to providing cer-1toutine use, of records maintainad in :L. The Subcommittee may meet in Ex. tain information to the American In-system, includine catotories of acts a.r. i the purpose of such unes: surance Associatmn Index Syst err, eeutive Session, with any of its consul-tants who may be present. to explore The purpose of the proposed change is Add a new routine use to rest !! and exchani:e their preitminary opm. to further clarify the nature of the m. To er vide memtiers of the A er :a; lons res:ardmg matters which should formation involved. Insuraatce Association Index Sys:c-- be considered durme the meeting and DATES: Comments regarding the pro-with certain mfonnation relatir.g : to formulate a report and recommen. posed routme ur.cs m'ist be received on accidents and injuries." dations to the full Committee. or beicre May 17,1978. Rocr.R p. CzAIc. At the conclusion of the Executive ADDRESSES: Records Officer, U.S. Deputy General Cvease. Session, the Subcomrmttre will hear Postal Service, Washington. D.C. Doc.7amaa m 41448 m as presentations by and hold discussions 20060. FICERAL RIGl1 Tit, VOL. 43, NO. 74--MONDAY, APtit 17, 1973

P NOTICES 202C Dated:'May 3.1978. which considers the ifndings made in For the Nuclear Regulatory Csa-JOHN C. IIOm. the FF.S. Parts I and II. (NUREG-75/ mission. Adr.f sorv Comm ettre 091 nnd NUREG-00M), the Draf t Ad. blassagemen t Officer. dendum to the Final Enstronmental Geonce W. Kn cirrow Chief. Environ mental Projects (FR Doc.7812544 Filed 510 78; 8:45 mm! Statement (FES). Part II (NUREG* Bra nch L Dit tsson o/ Sau 0056; Addendum) and the Part III sa/ctv and Environmente.J 8 [7590-01] D*" 3'****'u"nutd Pathway Genertc A "*8I*' t The Final IFR Doc. 78- :933 tiled 5-10 78; 8.43 arsj l ADYl50tf COMMirff t ON ttAC704 SArt. Study (Nt;R1;G-0440) was used as a i cuAmos sucCoMwntt ON THE MA6NE primary re!Prence by the NRC staff in g YANKit NUCt1A2 PLANT the prepa.Ntton of this Part III Enyt. [4910-53] ) ronmenta: :tatement. The Notice of } D " " #**""' ~ availabihe, of NUREG-0440 was pub. NATIONAL TRANSPORTATION The postem t May 2.1978 meeting ILhed in tne Fratztat. RrctsTra on SAFETY SOARD cf the ACIt., f aucomm.ttee on the v uen to 073 (43 F1t 0391). Maine Yr wuetear Plant has been . F.evned DES is available for fn. IDocket No. SA-48t1 reschedul- . 9 held on May 25. The .- sy the public in the Commis. meetinc w

  • a t at 10 a.m. on Thurs.

, ]iblic Document Room at 1717 r il t s pertaininit to this '"t NW., Washington D;C.: the lavestigeflea Hearing ville Public Library,1... orth 2 <e ttru - vt the same as an. u.a.'ci o.n FrozRAI. REGISTER On . arect. Jacksonville. Fla. 32::01 Notice is hereby given thst the N-y. . mton State College Library. tions! Transportation Safety B e ar-- .. e, t . :mor. N.J. 08:40: and the New Or. will convene an accid:nt investiga:t

  • .W May 9.. h.3.

~,. u "-slic IJbary. Business and Sci. hearing at 9 a.m. (local time) on Mi-JOHN C. HovLt. - - * *. aon. 29 Ieyola Avenue. New 23, 1978 in Conference Room H. Adetsory Committee m;.. La. ';0140. The Revised DES Center House. Seattle Center Ccnfer. Jf.Inagement O//icer. 13 a;.o 'tng made available at the ence Center. Seattle Wash. (FR Doc. 78-1:33: PJed 5-10-78: 8.45 ami Dureau :( Intergovernmental Rela. The public hearinst will be held ' : e 6 tions. Dtasion of State Planning. De. connection with the Safety Ucard s ta. [7590-01] partme f Admhitstrau n. 6@ Apa. vesucation of an accident invom s lachee nrkway. Ta:hhassee. Fla. Columbia Pacif!c Airhnes. Inc. De*c. t CDOCITr NO S'nt 50-4373 3:304 and at the Jacxsanv111e Ares 01 N190EA. which occurred Februa_7 OFF5HOtt POWit 3 MTMS. FtOAftNG Planninc 30ard. 330 K. Hay Street. 10.1978. at Richland. Wash. NUCLEAR POW 4 MNf5 ac nr a Fla. 3:202. 1 Pursuan;

  • 3 10 CFR Nrt 51, inter.

MARTIN SrEisER. Av.:landlity of navned c.. - f.vironmentes ested pers.. may sucmit comments j Steisment. 7 m.tl on the Re ed Draft for the Commis. .w 3.1978.' Pursuant to the Nr. Mnal Environ. sion's cor.meration. Federal. State. . 3 Doc. 78-1:s f: Filed 5-10 78;8:45 amJ n: ental Policy Act o* MG3 and the Nnd specumd local agencies are being United States Nuc'rar Regulatory provided a tth copics of the Revised Commission's (Comtn uion) regula. Draft. Cemments by those officials or [4910-58) tions in Appendix M of 10 CFR Part other per ons received by the Conunis. 50 and 10 CFR Par. 51 notice is sion will ce available for public inspec. MED PARM. NUNICA110N5 hereby given that th: Commission's tion at the Commission's Public Docu. Sen8 F*r Office of Nuc! car Resext Ret:uiation has prepared a Revisca : mf t Environ. ment Room in Washington. D.C., and the othe' ' cations mdicated above Nctice is here e afven that on Ma.,. m mental Etatment. Part i Revision 4.1973, the NTC3 adopted the fo1107

1. NUR EG-0107 L Upon nsideration of comments submitte : ith respect to this Rettsed ing policy statement by unanimous The ord Draft F ' Mnmental Draft, t.t ammisc. ion's staff will pre, vote:

Statemete , consiocr i the com. pare a - 31 Environmantal Sta te. It is the pol!cy of the Nation-1 E*',"[ '..[nue ment. P:.r t III. the availabt!!ty.of Transportation Safety Doard that the r nts an

t. clear plants :L.:.ociated

%hich ws;; be published in the FGEaAL Members thercOf Will not PCit Cr calental release of radioac. hrstra. subject themseWes to thW .' ce mal to the noucous environ. Comments on the Revised Draft municartoos regarding ar /.atter a wtice of availauthty of the Irom interested members of the public pending before the Board at er the no-c: Draft Environmental State, should be sent to the U.S. Nuc! car tation with a notation nut %cr has to. Part III. was puuhshed in the Regulatory Commission. War.hington, been received by the Doard. h.

4. Rtatsrta on Getober 18.197G D.C 20055. Attention: Director. Dist-MAacAarT L. Fisitrn.

1411 R 4$699). Comment.s received on ston of S6te Safety and Environmental federal Register the eriginal DES have been included Analysis and are s'te by July 3.1978. Lietson O//icer. Es /.;pendlic A. end discussed m sec. Copics of the immmitcaon's Revised MAY 5.1978. Lacu {of the Revised DFS. Dra.it. Part III. y be obtained by re. '2..~ Revised DES dif fers from the Int Doc. 78-1:sa: n:ed 5-10-78; 8:45 aml quest addres'.e : to the U.S. Nuc! car crir.h11 DES in that it comdders the ovum! risk rnd conaequences between n t'ulatory Ccmmission. Washincton D.C. 20555. Attention: Director. Davi. flon: X and land-used nticicar planta resul..nc from the accidental relcurs sion of Technical Information and ' Enforcement opinions and orders are oc4 of ro. ' activity to the environment. It Document Control. ABC' Ud'd '*** L h'Y " "d )""*""# thercey subicet to the 11ourts rules cord Go U vides a total cost benefit bat. Dated at Bethesda, Md. this 5th day cerning Ex l'arte co.mntaucauun tis cra Lace IJr the proposed hcensing action. of May lidtl. [ 821.60 and 825.40). ftDttAt RIOliitt. VOL. 43. NO. 92-THUt3 DAY, MAY ll, IUa j

A ' / Ac MWC"Y & D TENTATIVE SCIEDULE ACRS SUBCC".'41'1'IEE "ECrIlC CN FBINE YNJKEE 11ay 25, 1978 Washington, D. C., , s . The purpose of this subcommittee meeting is to review Yx '.ee Atomic Electric Cc.Tparr/'s request to increase power at the Mair.: Yankee Atomic Power Station frcm the RSAR power 1cvel at 2560 IG(t) to 2630 fG(t). The approved power level {is 2440 IG(t). F Aporoxifnate Time .10:00 a.m. I. EXECUTIVE SESSICN (OPEN) 10:10 a.m. II. LICENSEE PRESufrATICN A. Brief plant and site description. ~ ~ ~ - O'.. B. Licensing History C. Corrents on Power Increase ~ approved power - 2440 tW(t) ~ ESAR - 2560 trd(t) ~ Stretch-plus-2630 IG(t) D. '$echnical presentation ~ Radiological consequences Accident analysis Transient analysis u Fuel performance 6

. ?,, O TDTTATIVE SCIEEULE - !%INE YN1KEC D. 'Itchnical presentation (continued) effects of power increase on safety-re-lated systems, equipment & components. Technical specification changes. 12:10 p.m. WNOI 1:10 p.m. III. NRC PRESENTATION Response to licensee presentation. Other open issues if any. Operating history /I&E, including excessive wear in control element assembly (CL'A). Generic issues. .] 2:00 p.m. IV. CAUCUS BY THE SUBCCI'24ITTEE (OPEN) 2:10 p.m. V. MEETING WITH THE LICENSEE AND STAFF 2:30 p.m. VI. AnJOUEFEh'r 4 l t Y

... :.. a. ~ m 0FFICIAL USE ONLY ATTENDEE LIST ATIACHPENT C MAINE YANKEE SUBCOK4ITTEE MEETIfG MAY 25, 1978 ACRS MAINE YMEEE W. Kerr, Chairman P. Bergeron H. Etherington D. Canton M. Ficst, Consultant S. Schulty P. MacCready, Consultant G. Solan J. Lee, Consultant J. Turnage E. Igne, Designated Federal Employee J. Garrity R. Major P. Littlefield NRC YANKEE A'ICMIC ELECT. CO. C. Nelson J. Distefano J. Fairobent P. Guimonal J. Giannelli D. Pounder S. Weiss R. Shone L. Olshan J. Laynor R. Reid G. Cwalina E. Adensam B. Grime 0FFICIAL USE ONLY

I ppen cli 4 MI!!E YA.1XEE PRESE:iTATI0il e SITE CESCRIPTI0il O PLAtlT DESCRIPTIO;1 0 LICEf1Sli!G AND OPERATING HISTORY RELATI0iiSHIP OF VARIOUS CORE PO'.lE.k' LEVhLS 0 - o TECHNICAL PRESENTATION e INCIDEilTS C0t4SIDERED o RADIOLOGICAL.AtlALYSIS SAFETY ANALYSIS ASSUf1PTIONS e 0 STEADY-STATE THERMAL HYDRAULICS o SUf4 MARY OF RESULTS-ACCIDENTS AtlD IRAtlSIEllTS O DISCUSSION OF LIMITING EVENTS o RPS SETPOINTS. 0 FUEL PERFORMAtlCE ? o-EFFECTS ON MAJOR EQUIPMEtlT 0 IECHill C AL SP EC I F I CAT I 0tl. CH AN GES N e m. e

n, ~ IMIfE YA!'KEE SITE DESCRIPTIO1 1 LOCATI0ti 0 WISCASSET, LInc0L:1 COUNTY, HINE 740 ACRES 3AILEY POINT. RIDGE OF BEDROCK o 4 MINIMUM EXCLUSION RADIUS 2'000 FT. o LPZ 6 MILES o POPULATION TOWN OF WISCASSET 4 MILES ~2000 0 ii LE5IiSTON 26 MtLES~I45,000 o BATM 7 MILES-12,000

o. LPZ 63 PEOPLE /MI2 YR. 2000

~ =L o O 9 O . ~,, ~ \\. G e

7 + i I CIT!ES: POPUL AT I 0il Over 2500 '0-50."les Ba::cd on lHO U.S. Cen us { ,6 1 l l = Fairficid o Watorvillo .:. }

  • "L::1ou y

Livor:cy noltect { Fa a gugusta o Enllevall - n crdiner Cciden f R6c$.nd Lai en n \\t xn \\,' W F (I // Lic en Fal' $290 \\ 'o i "g, Bath j,, n g, 30 go 3-Brun.ick o 0 </,/ f' ) g niles fA vy Mb yn 3 p \\ /' bW ' Fair ,1 L Lincoln Cbunty g C.3 I Westbro.k o ?3r land i S. crtland Caeg(:)f::both j 1

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\\... ClTIES: POPUL ATI O ! 07cr 25,000 [] 0-100 l' ales Based on 1960 U.S. Consud = f J / i e'. f 1 i = j j + ~' ~+I i l r t y$t ) .t m t / rf/,nJ +*i \\\\ f s' / / / Auburn l 21s,h13 / q / Is h on 4 / 6, 0014 Pledh Sitc 'h Dan ar \\/, y Po-t.Inntlo 'D 3"'b$.

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POPULATIONDISTRIBUTION i / ; CIRCLE PoPULATIota 1960 POPULATI0tt 2000 RAD!us PoPutAT!ok ~ ilirii Sunnea Nor ItactuDirin EllLE1_ __ E60 _h_'21DEllT.S SUldt1EILRESIDt'.tLIS. t 0-1 St1 61 78 1-2 303 112 7 128 1 2-3 112 5 556 189 5 3 11 1,09/1 1,151 1,587 11 - 5 1,177 1,273 1,719 0-5 3,053 3,fl68 ~ li,307 .c l W e o G e

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I l 1 l I '/, i i -l - ; {a/[ I I i ! i I I f ; 8 ! I 'p I q . n, p --- ,i !., i, . /,, ' u\\ l,m t <. ' THE0 A1. i, ' i.i>u i i i i i i i i Ti : i4 <!}};) SHIELD y'l' M i t. :- i- ;. i i, m. --.i /. e ! i ) l ik' i[/ U. *. I I i i 1 I e.,7 s.. , l A.- I e i = s g. e ' Xs? l % &' Q.Q. , I bhy/l L-4' O f f'f V ' i CORE SUPPORT N N.. - / ' B ARREL g-. C,0B 0 "i'/,',/p'%.. g.. S n.~ C U D - 4',/gi y,,yyj,/,j,'y/' y j g i GUIDE TUBE ~ FUEL ROD / 0.440" OD s' i.w - m-y.~;.u =x-mn..: w-.,::e.u,4 , e. 4... -.. a v.. m_ w% . u .n ...{ m..a. -. T ~ _. : - t\\./ i -- --< % )- j UH V.--,. . J'3 l I p s. h j r, .txi,,.... . w c.2 cms .... a i>&1 ti 7, m - : L.!!!!+i T.... !,--- .a T 0. 5 50" 4.640 " ,-._). - _,. m' . n =%,,"-H ...,.., E,j, ~: ? -N -H \\ ,D -l-,1.. i -, -,. H. i. Ti 9 3!Y-h, - 3 + r r--- ' il l. l i! ~""! ?.d ik'-"h i H 1h. s i4 rt f,,-- j',qH-i 4 2 k: - ' 4-Ji'A a.a.sze-y-m':g e.4+,w-u-+e-- --t m :u: m O 7 020" 0.580" =, - -,r0.140" OU'T'(10 F .ph0D -FUEL. ICD'S F i 0.200" W ATER ' GAP i ~. rigure umn v.wia:n ,un.ge i num e.,. Reactor Core Cross-Section .ame u..,g. 3.6-1

j o. ~ f = m m__c j FUEL m.._c (N HOLDDOWN sid.=8s n=== 2s m. =. ~ m T.1 5 %s..\\il -GUIDE TUBE W 4 '.e u pigs sun .o linllM DN/d_1 .bdjjj$ n ~ 45 qw. ;'. <n.TP aOe i e J:-o 0 9+, i[.c]1; f0/.e \\O__+i ml! :;in.., y,, t- -i i. 4.. .i w ? th y 157 " iph,n TOP V1EW y/,,,:j(.l :l ;,.'g. a FUEL RODqW N.1;'i c l i l. ,d1 !,g! ' t hMg !$i !iO!! NN'l!m136.7", iN ! ! L !! Nii & ij!k,lii! l U Hl!iji!! ACTIVEl i!H' Ej [f,$ f, LE?.'GTH: S PACER GR I D Q'Ij ;j.-llill j-l:4 tly m v m i li!j l 'i FLOW HOLES Wijn! hi !lf ..f../......_ k6-[-P: i !!!!l <Nn ... 2.. m_ ::::: r; i.p,,n p; u e

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~' Il ~ s, ,gg n.= y v,ec LICEi!SI?lG A",3 OPERATING IIISTORY t g, 0CToBER 1960 c' CONSTRUCTICM PERMIT ,.fSAR-AUGUST 1970 o ACRS LETTER JANUARY 1972 E'EBRUARY 1972 S MFF SER. OPERATING LICENSE (75% 'oF o SEPTEMBER 1972 2440 (@!T) DECEMBER 1972 c Cc:'.:<ERCI AL OPERATION DECEMBER 197,) o.0PERATING LICENSE (2440 i'.'JT) =. CYCLE 1 (10,367 is!D/!iT)

  • DEC.1972 - JutY 1974 o

5 CYCLE 1A (4500 IO!D/ lit) OCT.1974 - lhY 1975 CORE 2 (17,100 islD/i4T) . JUNE 1975 - thy 1977 .o CYCLE 3 (9700 islD/iiT) JUNE 1977 - PREsENT .o O g e ~. / l e ~~~.

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  • Operaring Irictory Pouer

?r Prensuro (ilut) Durnup (UP) (p:: i.a) Daten . l' u< r l. - 2009 12/72-4/73, unprennurized i Cycle 1 1830,1,2 10,367 530

  • 1000 4/73-7/74 lou d.nnity o

530 ,4,500 535 1000 10/74-5/75 - -# 72 preunurized Cycle.1A 244D2 ~ . remainin.: Cycic o o 'Coro 2 2440 17,100. 537 2100 6/75-5/77 ' pre-prennurized ~ liigh dennity ' 24403 542 2250 9/76 g o Cycle 3 24404 _9,700 542 2100 6/77 procent ' pre-prennurized .high dennity 5 550 2250 5/70 65 Cycle 1A 2560 1 OL 7'it of 2440 MWh 9/72, OL 2440 MNt 10/73' t. 2 tiever.achiev'c 2440 MWt due td 1) bay tcmp., 2) LIIGR limit, 3) EdS Act. / 3 ~Special test 4 During Cyclo.3 refueling RPQ modn to allow 2630 tst. ... f 5/15/70 operated at 2560.MWt,' 2250 psi'a and 550'F 0 5 1 ~ g e e

r MAINE YANKEE RELATIONSHIP OF VARrIOUS CORE POWER LEVELS / / 0 CORE POWER. Ic PRESSURE LOW NUCLEAR 2 IMWT1 (OF1, (_ESJAL ( (cerd EEAgInti FSAR 25601 Sll6 2250 3 211 DES Ic'rt MAIf1E YANP,EE OPERATION 211'10 535-546 1300-2100 360 REDUCED 2 STRETCli PtWER 2.250 55fl 2250-360 REDUCED 1. IllERMAL HYDRAULICS 2fil10 MWT, ACCIDENTS AND TRANSIENTS 2560 MHT '2 IMPOSING SPECIAL RESTRICTIONS ON PLANT OPERATIONS;PDIL AND MONITORING OF CORE POWER DISTRIBUTION ~ 3. 5 511 F AND 2250 EaUlVALENT TO Sil6 AND 2100 ~ ,i e e O e

81 a a e s incidents C:nsidered Category 1: c,,i. w a. -.., i .. C m. Boron Dilution .CSA.Droo 'A f-Halpositioning of ?::: Len;th CIA's 's . Loss et Cco., cat :.3 ca Excess Leal Loss of Lead Loss of Feeds.2ter Stcas Linc ?.upture (SLR) Category 2: ~ Steen Gencrnte r Tub s ?.unture CSA Ejc.ct:.0: Loss cf Ccelant ,v - a ;. e.. e. _...-. e -..,..

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~ Regii0;c3,1 :( v nsc:* 2nces ef Fecc.:ccc: 2.;ne 5:ec..s , vu t s zue v u.::. i.iae u. Category 3:- e... Contcin=ent ?rcssure Analysis Fuel.%r.dli-, + L'aste Ces Systcc r a:.3 ure .s Spent Fuel Ccsi Orcp Rcdiocctive Liquid.::ste Syste: Leak or Failure e S N 9 ~. m. g e e e / G / O

BASIS FOR RE-EVALUATION OF THE DESIGN BASIS ACCIDENT FOR RADIOLOGICAL DOSE ASSESSMENT I - INCREASE IN STRETCH POWER OPERATING CAPACITY FR 2560 MWT TO 2630 ISIT MAINE YANKEE FSAR ACCIDENTS EVALUATED AT 2611 M STRETCH POWER SUBMITTAL ACCIDENTS EVALUATED (NET - 3% POWER INCREASE) II - CURRENT CRITERIA FOR EVALUATING DBA'S AS PRE THE STANDARD REVIEW PLANS FOR THE APPLICABLE III - MOST CURRENT METEOROLOGICAL DATA AVAILABLE A TIME OF THE STRETCH POWER SUBMITTAL (DATA ACCUMULATED FOR THE 1975-1976 OPERATING PERIOD) s e t we 4 9 I

-'Ib. DBA'S THAT WERE RE-EVALUATED FOR THE ~ 2630 MWT STRETCH POWER SUBMITTAL 1. STEAM GENERATOR TUBE RUPTURE 2. LOSS OF C00LAllT ACCIDEllT - DOSE ASSESSMENT A. LEAKAGE FROM ENGIllEERED SAFETY FEATURES OUTSIDE CONTAINMENT B. POST LOCA HYDROGEN PURGE c. CONTAINMENT LEAKAGE CONTRIBUTION 3. MAIN STEAM.LItlE FAILURE OUTSIDE CONTAINMEllT 11. FUEL HANDLING INCIDENT e =e e e e* e f G e 2

STEAM GENERATOR TUBE RUPTURE FSAR VS. STRETCH POWER SUBMITTAL - MAJOR PARAMETER CHANGES: 1. INCREASE Ill PRIMARY AND SECONDARY ACTIVITY LEVELS BASED ON HIGHER POWER OPERATING LEVEL. 2. EVALUATION OF THE RADIOLOGICAL CONSEQUENCES BASED ON PRE-EXISTIflG AtlD COINCIDENT PRIt%RY C00LAtlT IODINE SPIKING. 3. RADIOLOGICAL DOSE C0flSEQUENCES BASED @ LATEST ATMOSPHERIC DISPERSION FACTORS AT THE TIME OF SUBMITTAL. O 4 o S 3

TABLE 4.11-2 0FFSITE DOSES FROM STEAM GENERATOR TUBE RUPTURE (0-2 HR) (0-30 DAY) SITE BOUNDARY DOSE (REM) LPZ DOGE (REM) THYROID WHOLE BODY _ THYROID WHOLE BODY CONSERVATIVE CASE '6.4 + 0 6.0 - 1 3.0 - 1 3.0 -2 '1.2 - 4 3.1 - 3 6.1 - 6 1.5 - 4 REAL:allt CASE CONSERVATIVE CASE WITH COINCIDENT IODINE SPIKE 1.1 + 2 1.1 + 0 5.4 + 0 5.0 -2 CONSERVATIVE CASE WITH PRE-EXISTING IODINE SPIKE 1.1 + 1 7.0 - 1 9.0 - 1 3.0 .:2 '6.4 + 0 = 6.4 x 10'

MAIN STEAM LINE FAILURE OUTSIDE CONTAINMENT ~ FSAR VS STRETCH POWER SUBMITTAL MAJOR PARAMETER CHANGES: 1. INCREASED PRIMARY AND SECONDARY ACTIVITY LEVELS BASED ON HIGHER OPERATING LEVEL. 2. EVALUATION OF THE RADIOLOGICAL CONSEQUENCES BASED ON PRE-EXISTING AND COINCIDENT PRIFARY COOLANT IODINE SPIKING. 3. RADIOLOGICAL DOSE CONSEQUENCES BASED ON LATEST ATMOSPHERIC DISPERSION FACTORS AT THE TIME OF SUBMITTAL. i ~ 11. DECONTAMINATION FACTOR Oi 10 APPLIED TO IODINE RELEASES FROM AFFECTED STEAM GENElATOR. m e. me w M O = 9 S

IABLE 4.14-3 0FFSITE DOSES FROM STEAM LINE BREAK - Th k h0fkODf" Tl 0 h blEBODY CONSERVATIVE CASE 4.8 - 1* 2.1 - 3 2.4 - 2 1.0 - 4 REALISTIC CASE 4.4 - S 5.4 - 7 9.7 - 8 1.2 - 8 b0lbb D!" 0Dhh! b E ~ ~ ~ ~ m b k b IOD N "IKE ~ ~ ~ ~ "4.8 - 1 = 4,8 x 10-1 o O

FUEL HANDLING INCIDENT FSAR Vs. STRETCH POWER SUBMITTAL ~ MAJOR PARAMETER CHAilGES: 1. INCREASED FUEL ASSEMBLY INVENTORY BASED OU HIGHER OPERATING LEVEL. 2. RADIOLOGICAL DOSE CONSEQUENCES BASED ON LATEST ATMOSPHERIC DISPERSION FACTORS AT THE TIME OF SUBMITTAL. ~ e ~ ~ ~. ~ ~~ e eg g e ~ m - a w

TABLE 4.17-2 ~ DOSES FROM FUEL HANDLING INCIDENT REALISTIC CONSERVATIVE CASE, REM CASE, REM DOSE POINT THYROID WHOLE BODY-THYROID WHOLE B0DY 1.6 - 04 1.4 - 03 . 2.6 + 1 3.4 + 0 {fgjgRY .0F '0PULATION ZONE lot il) RY ACClu NT()10F 3.6 - 06 3.1 - 05 ~1.3 + 0 .1.7 - 1 (DURA 10

  • 1.60 - 04 = 1.60 x 10-4 4

e

LOSS OF COOLANT ACCIDENT FSAR VS. STRETCH POWER SUBMITTAL ~ MAJOR PARAMETER CHANGES: 1. INCREASED CORE HALOGEN AND NOBLE GAS INVENTORIES BA ~ HIGHER POWER OPERATING LEVEL. 2. FACTOR OF TWO REDUCTION IN THE ELEMENTAL IODINE RE CONSTANT (AC HR-1)USED FOR THE SODIUM HYDROXIDE SP ~ -1 Ae USED IN FSAR ANALYSIS = 28.5 HR ~1 ~ Ae USED IN STRETCH POWER SUBMITTAL = 14.0 HR 3. CREDIT FOR A PARTICULAR IODINE REMOVAL RATE CONSTANT OF A, = 0.708 HR-1 FOR THE STRETCH POWER SUBMITTAL, NO CREDIT FOR PARTICULATE 10 DINE REMOVAL BY THE SPRAY SYSTE WAS TAKEN IN THE FSAR. 4. RADIOLOGICAL DOSE CONSEQUENCES BASED ON LATEST ATMOSPHE DISPERSION FACTORS AT THE TIME OF SUBMITTAL. POSTLOCAHYDROGENPURGEDOSECONTRIBUTIONWASCALCU 5. BASED ON THE INCREASED POWER OPERATING LEVEL. 6. POST LOCA ENGIN O ED SAFEGUARD FEATURES LEAKAGE DOSE CONTRIBUTION. 5 ~ h g e 9

LOSS OF COOLANT ACCIDENT FSAR VS. STRETCH POWER SUBMITTAL ~' - MAJOR PARAMETER CHANGES: 1. INCREASED CORE HALOGEN AND NOBLE GAS INVENTORIES BAS HIGHER POWER OPERATiiiG LEVEL. 2. FACTOR OF TWO REDUCTION IN THE ELEMENTAL IODINE REM CONSTANT DC HR-1)USED FOR THE SODIUM HYDROXIDE SPRA le USED IN FSAR ANALYSIS = 28.5 sa-1 -1 le USED IN STRETCH POWER SUBMITTAL = 14.0 Ha 3. CREDIT FOR A PARTICULAR IODINE REMOVAL RATE CONSTANT OF 1, = 0.708 HR-1 FOR THE STRETCH POWER SUBMITTAL. NO CREDIT FOR PARTICULATE IODINE REMOVAL BY THE SPRAY SYSTEM WAS TAKEN IN THE FSAR. 4. RADIOLOGICAL DOSE CONSEQUENCES BASED ON LATEST ATMOSPHERI DISPERSION FACTORS AT THE TIME OF SUBMITTAL. 5. POST LOCA HYDROGEN PURGE ::SE CONTRIBUTION WAS CALCULATED BASED ON THE INCREASED P0'.,E~. OPERATING LEVEL. 6. POST LOCA ENGINEERED SAFEGUARD FEATURES LEAKAGE DOSE -CONTRIBUTION. e e 9

DOSES FROM LOSS-OF-C00LAtlT ACCIDENT LOW POPULATION ZONE ~ DOSE (REM) SITE (BOUNDARY DOSE (asM) (0-30 DAYS) 0-2 HOURS) TYPE OF ANALYSES THYROID WHOLE BODY-THYR 0lll WHOLE BODY ' CONSERVATIVE CASE 164 9.4 12.4 0.60 REALISTIC CASE 0.86 0.03 0.05 0.001 YN0hkL DOSES FROM ESF COMPONENT LEAKAGE THYROID DOSE (REM 1 WHOLE BODY DOSE (REM gUgNAREABOUNDARY 1.5 - 4 1.9 2 2.4 - 5 kOWPPULATbONZONE 0-0 DAY ) e DOSE CONTRIBUTI0il FROM POST LOCA HYDROGEN PURGE LOCATIOM THYROID DOSE (REM) WHOLE CODY DOSE (peu) LPZ (0-30 DAYS) 1.4 0.4 10

t ACCIDENT ATMOSPHERIC DILUTION FACTORS (X/0) EXCLUSION RADIUS (610 METERS) ~ ~ 3 3 DATA PERIOD DILUTION MODEL s s VALUES USED FOR FSAR PASQUILL "F" STABILITY-6.08 x 10 'I 6.48 x 10 'i SUDMITTAL CLASS 1 M/SEc INVARIANT WIND APRIL 1975 - SECTOR INDEPENDENT (5%) 8.07 x 10-4 5.63 x 10 'I MARCll 1976 O JAllUARY 1977-SECTOR DEPENDENT (SE - 2.2%) 6.24 x 10-4 JULY 1977 (SSE-3.2%)- 5.12 x'10-4 JAllUARY 1977-SECTOR DEPENDENT (SE - 2.11%) 6.22 x 10 'I SEPTEMBER 1977 (SSE-3.'1%) 5.05 x 10 'I JAllUARY 1977-SECTOR DEPENDENT (N - 2.1%) 5.93 x 10-4 DECEMBER 1977 (SSE-3.4%) 5.05 x 10 'I O e f

REVISED LOCA ANALYSIS ~ PARAMETER CHANGES: 1. REVISED PRIMARY CONTAINMENT SPRAY MODEL. ~ A. SPRAYED VOLUME - 47.34% OF TOTAL FREE VOLUME ELEMENTAL IODINE SPRAY REMOVAL CONSTANT ( l E) = 10 HR ELEMENTAL IODINE DF = 100 B. UNSPRAYED VOLUME WITH GOOD COMMUNICATION VOLUME = 32.27% OF TOTAL FREE VOLUME MIXING RATE BETWEEN SPRAYED AND UNSPRAYED = 10 HR' C. UNSPRAYED VOLUME WITH POOR COMMUNICATION VOLUME = 20.39% OF TOTAL FREE VOLUME MIXING RATE BETWEEN SPRAYED AND UNSPRAYED = 2 HR ~ 2. PRIMARY CONTAINMENT LEAK RATE = 0.10% DAY ~ 3. RADI0 LOGICAL DOSE CONSEQUENCES BASED ON 1977 METEOROLOGICAL DATA (12 MONTHS) \\ 4. REVISED HYDROGEN PURGE DOSE C0ilTRIBUTION BASED ON ADDITIONAL ZINC 'AND flRC ZIJC CORROSION RATES. O P

  • eo e

e ..e. em. e g 9 9 e g a 12

ca DOSES FROM LOSS OF COOLANT ACCIDENT - REVISED TWO HOUR SITE EOU.1DARY DOSE 30 DAY LPZ (REiD THYROID NHOLE BODY IHRYOID WHOLE BODY )flTAINMENT LEAKAGE 176

COMPONENT

.EAKAGE 1.7 iT LOCA HYDROGEN ' URGE N/A 2.7 1.4 e

  • e I

O m O 9 e G 9 a 13

~ SAFETY A.ALYSIS ASSU::PTIO:iS l o 3 LOOP OPERATION DESIGil POWER DISTRIBUTIO:iS.'- c s, Fz=1.68,Fy=1.49 j3 o REACTIVITY COEFFICIENTS .fiTC - 0.0 -2.74 x 10-4 cf / F 6 o CEA SHUTD0;'il CHARACTERISTICS ~ A~ssusen Cycts 3 BOC HFP 4.0% Af 6.47% o? HZP 2.0% osP 4.13% 6 f E0C HFP 5.7%z?P 6.80% ttf + HZP 2.9% Af 4.33%LLf

  • EOC HFP.SLR REcutaEs 6.5%

'o RPS TRIPS ' ~ ~. e x \\ ~ / e t

~ REACTOR PROTECTIVE SYSTEM TRIPS i. SFlP_QIRI UlLCERTAINTY ~ DELA.Y_I_INrdSECd / ~ o HIGH'flEUTRON FLUX 10,6.57. i5.5% 0.4 0 LOW REACTOR C00LA'NT Flow 93% 2% 0.65 0 ll!GH PRESSURIZER PRES 5URE 2I100 PSIA 22 PSI 0.9 t2' PSI 0.9 o LOW STEAM GENERATOR PRESSURE 500 PSIA t c LOW STEAM GEtiERATOR WATER LEVEL 60 IN 10 IN 0.9 0 LOW PRESSURIZER PRESSURE 1750 PS.IA 22 PSI 0.9 1600 PSIA 22 PSI' 0 SAFETY IIlJECTION SIGt1AL + l j 0 IHERMALMARGIN/LOWPRESSURE ~ 0 SYV. METRIC OFFSET O RATE TRIP O var.I APLE 0'/EP.?O'.fER d

~ STEADY-STATE TilERMAL HYDRAULICS ~ cycle _1_ Cx.cte 3 Cy_cLFJ_S' RET _C11 o TOTAL llEAT OUTPUT M!lT 2IIII0 24Il0 2630 106 BTuhlR 8328 8328 8976-o Pressure ~ il0MiNAL PsiG 2235 2085 2235 Mininun esIG 2185 2035 2185 MAXIMUM esiG 2285 2135 2285 o DEsiGil INLET TEMPERATURE OF SliG SIIS SSIl o TOTAL REACTOR COOLANT 6 10,og/gR 122.0 -136.0 134.6 FLOW o C00LAllT FLOW TilROUGH CORE 106 ts/HR 117.5 132.1 130.7 i o CORE AVERAGE IlEAT FLUX BTU /llR-FT2 170,200 165,830 . 178,740 ~ ,2 117,700~ 48978 118978 o TOTAL llEAT TRANSFER AREA FT i o IIAXIMUM CLAD TEMPERATURE - 0F 657 GI16 656 o FILM COEFFICIENT BTU /HR-FT2-oF 5500 5660 , 5640 o AVERAGL'.LillFAR llEAT RATE KW/FT 5.711 5.60 'e 6.03 0 Cour I'llIll/,l.I'Y RISE IITll/LB 68.5 63.1 63.7 o lu i r l AL S'l EAlf/ STATE 2.05 " 2.32* 2.0l* 2.52** 2.~9** MininuM D;.im Drsial e0'..cR nicTaintiTION

1. inn inn i o n:n nis rRinurion AT S0 PRE-TRIP ALARM i

,,,f i l-

/ Summary of Results ~ Incident Section Appropriate Criteria Result at 2630 MWt FSAR .C CEA Withdrawal 4.2 MDNBR>1.30' MDNBR>1.51 MDNBR>1.4 - f RCS pressure <2750 psip RCS Pressure < 2500 psia RCS pressure < 2500 psis l Boron Dilution

4.3 Subcritical

Sufficient Subcritical: Suberttical: Time for operation action Refueling - 80 minutes Refueling-65 minutes Critical: MDNBR>1.3 Startup - 3.2 hours .Startup -5 hours Critical: Bounded Critical: Bounded by CEA Withdrawal by CEA Withdrawal CEA Drop 4.4 MDNBR>1.30 hDN11R>l.36 MDNBit>l.38 Local LilGR121 kW/ft Local 1,ilGR = 21,kw/f t Ma1 positioning of 4.5 MD!iBit>1. 3 Bounded by, NA Part length CEA Local LilGR<21 kw/ft CEA Drop MDN11R>1.36 Loss of Coolant 4.6 MDNBR>1.3

  • MDN11R>1.56 MDNBR>1.46 Flow

-l Seized Pump Rotor 4.6 A dufficiently low 2.01% of rods with 7.5% of rods with fraction of rods MDtlBR$1.3 MDNBR51.3 with MDNBRSl.3 Exccus Load 4.7 IIDulu:>1. 3 MDN11R>l.8 MDNBit> 2.0 Loss of 1.oad 4.8 1:CS l'ressure42750 psia Peak HCS pressure Peak lu:S pressure llouh.21.3 <2'240 pula <2410 pula MDN11R>1. 9 11Dubl>1.9 ~ e

continued / Incident-Section Appropriate Criteria Result at 2630 MWt FSAR Loss of 4.9

1. Sufficient time
1. Auxiliary feedwater
1. Auxiliary feedwater for initiation of system required 15 system required 18 auxiliary feedwater minutes following Feedwater i

minutes following event event system.

2. RCS pressure
2. Peak RCS pressure
2. Peak RGS press *re

<2750 psia <2530 psia <2325 psia r ~ j Stea:n line 4.10 Maintain fuci Fuel rod integrity is No return to s Rupture , rod integrity maintained since reactor critical is predicted does not return to for 3 loop operation critical Steac Generator 4.11 10CFR100 Radiological dose well <10CFR100 within,10CFR100 Tube Rupture s CEA Ejection 4.12 10CFH100 No clad damage thus Small f raction of clac no radiological release damage predicted ,g Releane<10CFR100. LOCA 4.13 10CFR100 <10CFR100 - <10CFit100 NA 5.0 10CFil50.46 PCTa 2181 F clad oxidation =15.59% llydrogen generation =.74% / e e Steam line rupture 4.14 10CFR100 Outside Containment NA Feedwater Line 4.15 10CFR100 Bounded ty Steam line rupture ontst.!c rupture ( ou t. liuaen t Co n t.i t u:aen t 4.16 Pe.& pressure <55 psig Inercase of 1.77 pst Peak pressure = l'r u:.u u r e (Containment design over current 50 pata prennure) operating conditions

n LOCA RESULTS LARGE BREAK o RF (12.8 Kw/FT) ~' I PCT 21810F IhX. CLAD OXIDATION 7.54% lhX. HYDROGEN GENERATION < l% e E,F,G,H (16.5 xw/FT) PCT 2168 F ' thXs CLAD OXIDATION 15.59% lhX. HYDROGEN GENERATION < 1% SMAlL BREAK (16.5 Kw/FT) PCT 13480F \\ lhX. CLAD OXIDATION <.08% '00 TROD'CLADOXIDATION <.009% e e f e ~..,

MAINE YAr<EE RPS~ / / o VARI ABLE 11UCLEAR OR AT P0wER / e HIGH RATE OF CHANGE OF P0wER / o HIGH ilEUTR0ff Flux e HIGH PREssuRIzsR PRESSURE e low REACTOR COOLANT FL0w e low STEAM GEllERATOR PRESSURE e low STEAM GENERATOR WATER LEVEL THERi%L IMRGIN/LO'd PRESSURE e 'o SY;95TRIC.0FFSET e 9 N e e, 9. b 0 w e 9 ees.

SYPPETRIC OFFSET TRIP THE SYMMETRIC OFFSET TRIP SYSTEM PROVIDES A REACTOR TRIP BEFORE LHGR ExCEEas 21 fot/FT (FUEL CENTERLI ;5 MELT). A LCO - 100 ----j'- I V' 80 I i 1 4 TRIF BAND l l-I I 0 0 + SYMMETRIC 0FFSET SYMMETRIC 0FFSET = A (SE) + B (PERIPHERAL) A = SHAPE ANNEALING FACTOR 8E = EXTERNAL SY 'l'ETRIC OFFSET FROM EXCORE (LL-L) DETECTORS U+L ~. d e e

THEREAL MARGlil LO'I PRESSURE TRIP TR Pyga!P A Q lB + BT + C Di C = ODilB = A1

  • QR1 1.0 Al QR I

1 / I 1.0 l SYMMETRIC OFFSET 0 1.0 ~~ ~ ~ E.'_~. '. METHODOLOGY DESCRIBES THE DERIVATI0t1' 0F A, B, C, Ai Af1D QR1 N e h 4

as .not4 St4 t P CF MoDtFIE.D TiA/L.P u. TmP Pgg : k kk IE Ogd + B$ + C. (MOOtFtsp ) g TPtP ggg = A Pq s 4 + BTc +C P iG Q lS EQOl\\l ALE 9T To QR i q t.o I n ..o 9W i . I I I l I I i i 1.o

t. o Q

Q I4 DtFtED SVSTEM C o NT Ain s A9 Aoo cnow At O 70N CTt o M THAL' R ELAT E S DMB To S O. e em

LIMITlilG CC:lDITIOi"S FOR OPERATION IHERE A?.E TRA.'!SIE:lTS THAT CO fl0T RECEIVE EITHER A SYMMETRIC OFFSET OR A iM/LP TRIP O LOSS OF FLOW o CEA DROP .PROTECTITI AGAltlST THE VIOLATION OF SAFDL FOR THESE EVErlTS IS PROVIDED BY MAINTAltlING SUFFICIEllT OVERPOWER MARGIll. OVERPOWER MARGIll \\ REQUIRED N / L / \\ % I I w LCO l. I I l-1 I TRIP l I SO 8 ? 3 ~N %O N. e G O -.m

a FUELPERFORfWICE PERFORMAtlCE CllARACTERISIICS LilGR cg FUEL VOLUME {l. A D Cf.6D IKW/FTl S_UBFACE AYEMGE_ _ID_ Ud JIGAE Rf' j / 5.5 1797 967 1356 650 609 597 6.0 1917 992 1424 657 612 619 12 3386 1113 2148 732 643 1084 13 3608 1113 2253 745 649 1217 E,F, G,11 . 5.5 1694 936 1292 654 609 3 675 6.0 1804 959 1355 661 612 700 12 3150 1068 2012 739 643 1266 13 3360 1067 2107 752 649 144b CLAD COLLAPSE o RF 20,000 tiouRs; DistilARGE 12,000 iiouns EFGil - 73II,000 liouRs; DISCilARGE 26A00 Houns o REACTOR C00LAilT SYSTEM ACTIVIlY CORE 2 - ?,Y 2.86x10-14 C /ML I'" 2.19x10-2jtc/Mt Ac/M,L-2.211x10-{JIC/st. CYCLE 3 -l3,Y 2.87x10-1A C/ML-3.ll3X10-t u 1.15x10-3 36tch;L e 4

- - a..r cirv a 2r= en. ll l l lllk I I ,.J i Q ~9 .e l =_. 1 Ml-% l i s -- l F-l 'e i I l ll I d.s l l lll h> ~ / _l 1 _I l I I lll!N lll \\ l l l3 AR I lIlh l ll lll lM I N b $a 4 \\ i I g, v 1 7>. lv Lu l l l_1 h i ,A >ti lpg p' i j I i mii: ja d-z.u i K! I Ill D 2' i l l 4 i l I i kl l 8 l "U .I II i t i I N,i iN liiij k k O b is g 9J s X r 9 -y ' e' l l l l l l I i I l lh llllll l 'U_l lll l l i _e Wj l l i I l T,, b l ? / i i .( lllll l 1 / !N f kN 7 I M __> i;R I i j _m_ _____e m N 03 4 V N ' CD 4 V N y CO 4 v N C4

MAINE Y. ANKEE .. ~ - = _:._.:u:.-a...-=n. - - - - -,.=~:.~r-r 2 k B t' 10 l 'd '.i e!! G.r,. 4j _( li (i e s 2 :ei-I.l 3{ .ee E 10 i.i _ 'o o.-- a. G,, g u 4 i. l 3 d m 2a I 4 l f-to 2 g --) \\ O ', r - - - - - - --- I G1 l: 4 4 y.._. {jj t i 1 2j .l .i [Q-0 ;i 2.._ .....u ..,,.. m.,. ...,m.. ..,,._7.. j i u. O/7G 10/73 II/7G 12/7G 1/77 2/77 3/77 4/77 5/77 G/77 'u 77 C/, 7

., n EFFECTS Oft IAAJOR Ecut preE* T ECS t1AXIMUM ExoEcTED DESIGN VALUE TEMPERATURE 6060F 6500F (PRESSURIZER 7000F) PRESSURE 2300 PSIA 2500 PSIA '$EC0i!DARY ~ TEMPERATURE 5300F 5500F PRESSURE 870 PSIA 1000 PSIA 3AFETY RELATEn Ecuto"ErT PERFORMANCE CHARACTERISTICS USED TO DEMO 1 STRATE THAT APPROF.IIATE CRITERIA IS MET FOR THE VARIOUS POSTULATED ACCIDEt1TS AllD TRAflSIE?iTS. 6 e ,w k

      • e e

g e % *N es

raurwar.u rivuAricativna Itemf Identification Description of Change Basis for Change 1. Operating license Modify Paragraph 2B5(a) of the To allow operation at operating license to read a steady state core power of 2630 MWt. "(a) Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2630 megawatts thermal." 2. Technical Specification Modify the definition of To allow operation at Definitions llated Power to read a rated core power of 2630 MWt. "A steady state reactor core output of 2630 MWt" 3. 2.1 Modify Thermal Margin / Low Reflects operation at Pressure Trip setpoints 2630 MWt and 2250 psia as follows: s b) Thermal Margin /Lc Pressure >A Qggg + BTc + C, or 1835 psig (whichever is larger) where / T = cold leg temperature,*F c A = 2016.2 S.- B = 17.9. C = -10102 i trip 4. Figure 1.1-la. and b. trodify P gg equation on Reflects operation at 2630 MWt these Figures to read:. t ri Pp3;p= 2016.2 Qous + 17 9I -10102 f c

I ATTACHMENT A (CONTINUED) ItemI Identification _ _ Description of Change, Basis for Change 5. Figure 2.1-2 Modify Symmetric Offset Reflects operation at 2630 MWt Trip Function as provided in Attachment B 6. 3.10 Modify LilGR limits in Consistent with LOCA analysis 3.10.B.1 as follows and delete performed at 2630 MWt conditions. shim rod associated peaking Shim red associated peaking uncertainties uncertainties. diminish with Cycle 3 burnup and will be evaluated when approval for 2630 MWT 12.8 kw/ft Peplacement Fuel operation is imminent. 16.5 kw/f t Batches E,F,G, and 11 - ~ - 7. 3.10 Hodify coolant condition Consistent with safety specification 0.F as follows: analysis performed for 2630 MWt. "1. The reactor com at pressure shall be maintained ni 2250 psia i 25 psi (indicated) during ster *y-state i power' operation, /

2. The reactor coolant 'am,.;rature at the inlet to the reactor vessel shall
  • ~

be maintained at or less t!)an a nominal value of 550 F (indicated) during three loop steady state operation." 8. Figure 3.10-3 Replace Figure 3.10-0 Consistent with LOCA analynia with Figure-3.10-3 2, #.ttachment B LIICR limits and operatio t at 2630 MWe. /

PAINE YNEE ETEOPOLOGICAL DATA JULY 367 - JUNE E68 OJSED FOR SER) WIND SPEED, WIND DIRECTICN AT 45.44 3-BuwED AEROVANE (STARTING SPEED ~.94/S) CAU4S DEFINED AS lESS THAN.45 M/s 4CLASSESOFATP.oSPHERIcSTABILITY PROBLEG: P00a CHARACTERIZATION OF LCW WIND SPEED CONDITICNS WIND SPEED REDUCTICN TO PEPRESENT 1CM CONDITIONS lhCERTAINTY IN DEFINITICt! CF CAU4 LIMITED RESCWTICN OF ATMOSPHERIC STABILITY thCERTAINTIES IN DATA REDUCTICn PROCEDURES e 9 s_ O O e o e ~, -- - ~ u i

7, (All'E YNE fBECRCLCGICAL DATA APRIL 1975 - PARcs 1976 (PROVIDED FOR APPENDIX I) WIND SPEED, WIND DIRECTION AT @.7M 3-BLADED AEROVANE (STARTING SPERD ".9 dS) CAU4S DEFINED AS LESS IHAN.9YS 7 CLASSES OF ATMOSPHERIC STABILITY (PER R.G. 1.23) PROBlE G POOR C11ARACTERIZATION CF LOW WIf!D SPEED CCNDITICNS WIND SPEED REDUCTION TO REPRESENT 1CM CCNDITICNS SELECTION OF WIND SPEED r0 REPRESENT CAUi CONDITICNS AT10M O h. m_. o V

  • ~~~

ri l%INE YtME ETEOROLCGICAL DATA JANUARY - DECEtEER 19~ 7 / UPGRADED ['YTEORCLOGICAL PROCRAM (PER R.G.1.23) (Ud IHRESH0LD (.22WS) WIND SPEED AND DIRECTION SENSCRS E 'D SPEED AND DIRECTION IA.ASURED AT 10M E 7 dASSES OF [.fMOSPHERIc STABILITY "BEST AVAILABE" IMPROVED CHARACTERIZATICN CF LOW WIND SPEED CONDITIONS EXTRAPOLATIONSOFWINDSPEEDTOAREFERENCEHEIGHT ARE NOT NECESSARY IMPROVED CUALITY ASSURANCE PROCEDURES PROBLB'S: LICENSEE PEDUCED DATA WITH A " VARIABLE" CLASS (NO WIND DIRECTION ASSIGNED) BASED ON WIND SPEED CRITERION AND NOT RANGE OF FLUCUATICN IN WIND DIP 2CTION l= h WINDS LtSS THAN.75 tvS CONSIDERED "VARIAPLE"

2. WINDS LESS IHAN.75 W S OCCURRED 3.5% OF THE IIME, WIm As0ur 1% IN EACH G & F STABILITY CLASSES s y.

t ATmSPHERICDISPERSIONFCEELS

1. DIRECTION-INCEPENCENT MOCEL DESCRIBED IN R.G.1.4 n4D SPP 2.3.4
2. DIRECTION-DEPDIEDIT POCEL DESCRISED IN DRAFTR.G.1.XXX,"ATtOSPHERICDISPERSION f0DELS FOR POTENTIAL ACCIrENT CCNSEQUENCE ASSESStENTS AT NUCLEAR PCWER PtarrS" 9/23/77*

CONSIDERS:

1. I.ATERAL PLtEE PERIDER
2. ATtOSPHERIC DISPERSION CcNDIT7CNS AS A FuaCTICa OFDIRECTION
3. FREQUENCY OF wit!D IN EACH DIRECTICN 4.

EXCLUSIGN DISTN CE AS A Fil!CTICN OF DIRECTICN (AT Iil, CIRCULAR EAB = 610t0

  • REVIEWED BY ACRS AT NOVEtGER 2 1977 PEETING/ APPROVED FCR INTERIM USE BY RRRC AT PAY 2,1978 PEETING-B n.

-e--e + = +=

I I PAINE YANKEE X/0 VAllES DATA-MODEL 0-2HOURX/0 LOCATION 3 (SEC/M) 1977 DIRECTION-DEPENENT 7.0E-4* SSW AT 610M (R.G. 1.XXX) I 1977 DIPECTION-DEPENTNT 1.0E -3 610M (R.G. 1.41-D7 DIRECTIW-INTPENENT 6.7E-4 610M

  • USEDINAPRIL 11,1978SER o

A

go SLM%RY

1. 1977 DATA ARE CONSIDERED "BEST AVAILABLE" BECAUSE OF IMPROVET' SITS IN DATA C0f I CCTION AND REDUCTION SYSTEM.
2. DIRECTION-DEPENDENT MODEL BEST REPRESENTS EE SITE BECAUSE OF INCORPOPATION OF RECENT EXPERIMENTAL RESLLTS AND CONSIDERATION OF VARIATION OF DISPERSICN CCIDITICMS BY DIRECTION.
3. REQUESTED Atl ADDITIONAL YEAR CF DATA FOR CONFIRi% TION OF X/0 VALUES A. CONCERN THAT 1977 DATA INDICAfE SLIGifiLY Bti LtR DISPERSION (LOWER X/0) WAN AVERAGE ONCREASED FREQUENCY OF If,'l WINDS AT EXPENSE OF SW WINDS)

B.' CONCERN WIE D/.TA REDUCTION PROCEDURES CONCERNIt'G " VARIABLE" CLASSIFICATION.

4. LICENSEE HAS ComTFTED bN FAY 5,1978 LETTER) TO:

A. PRrVIDE mE ADDITIONAL. YEAR OF DATA BY FARCH 1,1979 B. MODIFY EE DATA REDUCTION PROCEDURE FOR "VARIAELE" CONDITIONS y

.\\, BASIS FOR CONTINUED OPERATION 1. SCRAM CAPABILITY IS MAINTAINED CEA INSERTED 3" TO PREVENT HANGUP CEA CONFERS ADDED STIFFNESS TO TUBE CE SCRAM TESTS SHOW POSITIVE RESULTS CEA EXERCISING PROVIDES ASSURANCE 2. CORE STRUCTURAL INTEGRITY IS MAINTAINED NORMAL OPERATING LOADS ARE LOW OVERALL INTEGRITY SATISFACTORY UNDER SEISMIC EVENT

3. I I t.
) e CURRENT STAFF ACTIONS EVALUATE ST. LUCIE DATA TO DETERMINE RATE OF WEAR AT 3" LOWER LOCATION DETERMINE IF MAINE YANKEE AND CALVERT CLIFFS #2 NEED TO INSERT ADDITIONAL 3" WILL REQUIRE INSPECTION OF ASSEMBLIES BEFORE MOVING THEM AT REFUELING

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