ML20086K110
| ML20086K110 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 01/25/1984 |
| From: | Bauser D GENERAL PUBLIC UTILITIES CORP., SHAW, PITTMAN, POTTS & TROWBRIDGE |
| To: | Clements W NRC OFFICE OF THE SECRETARY (SECY) |
| References | |
| NUDOCS 8401260231 | |
| Download: ML20086K110 (16) | |
Text
{{#Wiki_filter:.- . .
~ . .- - _ - _ - -- . .- .-~.- . . .
SHAw. PITTMAN. PoTTs & TROWBRIDGE
. .r. o,-cu,ouco r.o.
DOCKETED isoo w sin.cr. w. w. UMPC WASHINGTON, D. C. 20036 m v o. .ovt.. ..c. sown ruott. .c WOMS blOOO ,,,,,,,, ,,y,,, , LaTJh,* 2 . c. i 1".?t'ti, ;i"A,'"::!: !!SE,^t^ m ti" M,,,',,a",3,
.ma ,$U.
ag4
"'s? "., '.".2:^t.
o ., nAti. .:," . M"t'"h: .c 4 !W m ;;m./."."r:rs."e ="ff vracoar. u.<,.
- ggy,ygg eas.t c .
- g. a.-a.. .y,,
- ==,.g,
..v m s" ,',- = = .t m er
,, .m .g.,g ,
p'T!"".Eid2 .c . r co N$3!^,.^; ".'!*via! ' L ' tut".i"c".;*!rc^."" '
. .%,.1. *a .yM"M " " '
aun=uw' _ . - , , n = = ". w "" us::.v.w:a <
$lt" b*. , 1T!.!!:ta"Al'!." . ^k!"'4?'s"?J.t"t*****"
k'i"*d n& nam F":'e s==,l'A.V!':ve""~
,,a. v:." uw k=w.v"
, p"'*?J
- .!c':' E .m 71,."!gg*12" e "$s**f,* f'A!"oavo.
*^".".It o
.e '^tt"#t.'
.. .u....s.. .. i e..a ....u.-
",'"J " !t1;^"
*'t"J.A.'t".::"#'
"d"'
n'n.'t^.#;**" nT' "!,'Ja'" i E.TNA. l"'.;.f."c*-
- s. T'.'A'J."a"."e" v. e _
"'ga".".* Jat??!""'" t'E"ott"," ,~
s'"
"^t".M*Nf $1'"~t 8'kk *t.' so- r.oc tv- 12"f."? %.IJ"." "' C i"E "d '. = "'"
"o'AM.". ^a!&ou .. . corr ev.,ca. s.. "'*
,.o:'"!t:c*.^"!.c . s.. "*""*d".T."."'."*.'J" u
i".1?.*".n.",e'J')-{:'
..mes :,
OR.N O. .vL.C.st.
, C. ..
. at'~
Mo.. w w
.. H..M.Frg.M.N .o..... w,ven.t=
HELEN To ELLI
==:= ..c...
.iL.E.T ...M*NGToN It
. m...
..,ve......e,. 6 u ...
January 25, 1984 822-1215
. HAND DELIVERED:
William L. Clements, Acting Chief Docketing and Service Branch Office of the Secretary-U.S. Nuclear Regulatory Commission -
- -Washington,.D.C. 20555 o In the Matter of Metropolitan Edison Company
-(Three Mile. Island Nuclear Plant, Unit No. 1) , Docket No. 50-289 (Restart)
Dear Mr. Clements:
In your Memorandum of January 9, 1984, you notified the parties to the restart proceeding that the Commission had agreed with the Staff's suggestion that the parties be given an oppor-tunity to. comment on the Staff's response to Licensee's manage-1 r ment proposal, as reflected in Mr. Dirck's Memorandum to the i Commission of January 3, 1984. Your Memorandum directed that comments should be provided to your Branch by the close of business on January 25, 1984. Enclosed are Licensee's Comments which: (1) urge the earliest practical completion of OI's evaluation of the leak rate procedure at TMI-1; (2) express the opinion that a temporary limit in the range of 40-45% of full power would be more meaningful in terms-of plant conditions and operator
. experience.than the 25% proposed by the Staff; and (3) provide ,
8401260231 840125 {DRADOCK 05000209 PDR
. SH W. PITTMAN. PoTTs & TROWERIDGE A mentNgaswsP OF pmOFESSsONAL COmpomataONS Mr.-William L. Clements January 25, 1984 Page Two clarification or' additional information on several other portions.
of the Staff's response. Respectfully submitted,
,9a 4. 4-" - ~
Deborah B. Bauser Counsel for Licensee DBB:jah Enclosures , cc: Service List
00CKETED USilRC UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION +84 JAN 25 P3:27 .. 7 fy y C[E. '
')
Before the Commission In the Matter of )
)
METROPOLITAN EDISON COMPANY ) Docket No. 50-289
)
(Three Mile Island Nuclear ) . Station, Unit No. 1) ) SERVICE LIST Nunzio J. Palladino, Chairman Administrative Judge U.S. Nuclear Regulatory Commission John H. Buck Washington, D.C. 20555 Atomic Safety & Licensing Appeal Board Victor Gilinsky, Commissioner U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Washington, D.C. 20555 Administrative Judge Thomas M. Roberts, Commissioner Christine N. Kohl U.S. Nuclear Regulatory Commission Atomic Safety & Licensing Appeal Washington, D.C. 20555 Board U.S. Nuclear Regulatory Commission. James K. Asselstine, Commissioner Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Administrative Judge Ivan W. Smith, Chairman Frederick Bernthal, Commissioner Atomic Safety & Licensing Board U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission-Washington, D.C. 20555 Washington, D.C. 20555 Administrative Judge Administrative Judge Gary J. Edles, Chairman Sheldon J. Wolfe Atomic Safety & Licensing Appeal Atomic Safety & Licensing Board Board / U.S. Nuclear Regulatory Commission' U.S. Nuclear Regulatory Commission Washington, D.C. 20555 - Washington, D.C. 20555
. _ _ = _ , . - . _ _ - - - _ . . _ . - . _ . . . -
- 2-Administrative Judge Mr. Henry D. Hukill Gustave A. Linenberger, Jr. Vice President Atomic Safety & Licensing Board GPU Nuclear Corporation U.S. Nuclear Regulatory Commission P.O. Box 480 Washington, D.C. 20555 Middletown, PA 17057 -
Docketing and Service.Section .(3) Mr. and Mrs. Norman Aamodt Office of the Secretary R.D. 5 U.S. Nuclear Regulatory Commission Coatesville, PA 19320 Washington, D.C. 20555 Ms. Louise Bradford Atomic Safety & Licensing Bokrd TMI ALERT Panel 1011 Green Street U.S. Nuclear Regulatory Commission Harrisburg, PA 17102 4 Washington, D.C. 20555 Joanne Doroshow, Esquire Atomic Safety & Licensing Appeal The Christic Institute l Board Panel 1324 North Capitol Street U.S. Nuclear Regulatory Commission Washington, D.C. 20002 Washington, D.C. 20555 Ms. Gail Phelps Jack R. Goldberg, Esq. (4) ANGRY /TMI PIRC Office of the Executive Legal 1037 Maclay Street Director Harrisburg, PA 17103 U.S. Nuclear Regulatory Conunission Washington, D.C. 20555 Ellyn R. Weiss, Esq. Harmon & Weiss Douglas R. Blazey, Esq. 1725 Eye Street, N.W., Suite 506 Chief Counsel Washington, D.C. 20006 Department of Environmental Resources Michael F. McBride, Esq. , 514 Executive House LeBoeuf, Lamb, Leiby & MacRae l P.O. Box 2357 1333 New Hampshire Avenue, N.W. Harrisburg, PA 17120 Suite 1100 Washington, D.C. 20036 John A. Levin, Esq. Assistant Counsel Michael W. Maupin, Esq. Pennsylvania Public Utility Hunton & Williams Commission 707 East Main Street P.O. Box 3265 P.O. Box 1535 Harrisburg, PA 17120 Richmond, VA 23212 David E. Cole, Esq. Smith & Smith, P.C. 2931 Front Street Harrisburg, PA 17110
-,-vvy- - - + , - , - - - , , ,yw ---g w-n-- -.----,mi,-1-ey,v.----
Joseph Gray, Esq. Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 - Steven C. Sholly Union of Concerned Scientists 1346 Connecticut Avenue, N.W.
#1101 Washington, D.C. 20036 James M. Cutchin, IV, Esq.
Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Chauncey Kepford Judith H. Johnsrud Environmental Coalition on Nuclear Power 433 Orlando Avenue . State College, PA 16801 h
Comments on W. J. Dircks Memo of January 3, 1984
- 1. Page 3, 1st paragraph The sentence that starts, "In addition, assignment to func- .:
tions involving overview assessment . . ." should be deleted and replaced by the following, which is a direct quote from the GPUN June 10, 1983~ letter. GPUN "will reassign personnel such that those functions which provide an overview assessment, analysis, or audit of plant activities specifically; General Office Review Board Independent On-Site Safety Group Shift Technical Advisors Q/A Audit Q/A and Q/C Site Staff Licensing Radiation Control Emergency Preparedness will contain only personnel with no pre-accident involvement as exempt employees at TMI-l or-2." ,
- 2. Page 3, Section III, Item 2 The Staff recommended temporarily limiting power level to approximately 25%. GPUN has reviewed this situation and believes that a lLait in the 40-48% range would be more meaningful in terms of plant conditions and operator experience. A copy of the GPUN evaluation was given to the Staff at a December 16, 1983 meeting. A copy of the evaluation is enclosed.
- 3. Page 3, Section III, Item 3 ,
In the June 10, 1983 letter, GPUN stated that:
"We will, prior to restart, add full time on shift operational quality assurance coverage by degreed engineers until the open issues are resolved. We would defer to the NRC should it wish to provide full time on shift resident inspector coverage of TMI-l operations."
In view of the Staff position, we believe care will be needed to avoid excessive or conflicting oversight which might detract from safety. We will continue with our preparations to implement the OQA program as described , and will work with the Staff on the details of implementa- / tion and schedule.
- 4. Page 5, 5th paragraph This paragraph identifies six safety oversight groups for TMI-1. It also indicates that item (1) a plant safety review group (our TMI-l PRG) and (2) a corporate level -
safety review group (our TMI-l GORB) are required of all plants. It should be noted that for TMI-1 these two groups are in addition to the " Safety Review" organization required by technical specifications. Therefore, they are activities undertaken by GPUN that exceed regulatory requirements.
- 5. Page 6, 1st paragraph This paragraph states that "further changes to the control room design may be necessary as a result of a detailed control' room review which is still required to be per- l formed." Actually, the detailed review and all the identified modifications have been completed. A supple-mental review, associated with implementation of the ATOG procedures, is all that remains outstanding.
- 6. Page 8, 3rd paragraph -
This paragraph implies that the letter from the Department i of Justice will have some impact on the OI evaluation of the leak rate procedure at TMI-1. GPUN believes that the Justice request should have no impact on the OI evaluation of the leak rate procedure at TMI-l and strongly urges the earliest practical completion of the OI effort. I l l
. nS r - - , - - , - - - - - . - - ., ,----------r,.p,-,,-, . , - - - - - , , , -g -------a m - - n- r- -
m - - - - w
~
. ----....__ . ~
GPU NCCLEAR CORPORATION December 12, 1983
. CONSIDERAT[0XS IX LOW POWER OPERATION OF TML-1 Inicial scarcup of THI-1 involves operating a planc which has been shut down for approximately five years. To miniziae risk to che public and ensure che readiness of che planc and scaff for full power operacion, GPCN prcposed a.
deliberace scep by step program of poser escalacion in the planc Scarcup Specificacion (reference 5? 1101-06-008, Rev. 1). Thac Tesc Specification idencified a period of approximately 30 days at abouc LS: of power operacion followed by about 30 days ac 75% power. More recenc suggestions by the NRC ,' Scaf! center around 25_: power operacion for greacer chan 30 days. The reduccion in risk associaced wich reduced power operacion cencers principally abouc reduced decay heac at reactor shutdown and correspondingly grescer cime co take operacor accion in che event of planc accidenti. BACKGROUND: In developing che planc Scarcup Test Scecificacion over a :.ea.- ago, che follawing basic criteria were used:
- 1. Ec wduld be. desirable to limit inicial extended power operacion co less chan 2005 in order to minimize risk. .
- 2. The first extended low power placeau should be at a power level ac which all essencial planc systems are being exercised and the planc scable.
- 3. It was highly desirable chac che-planc equipmenc be overating in 2 ccde similar, if not identical, to che canner in which it operates ac full power (i.e., similar valve lineup, similar automacic concrol icops, ecc.).
- s. Planc solcware, i.e., procedures, ecc., should also be used in a manner similar to full power operacion sichout excessive number of cnan;v notices
[ or other orocedural modif1cacions. In development of the inicial Tesc Specificacion, che above criceria aere j . sacisfied wich some margin with che seleccion of a 45: pouer operating placeau. The determining item which sec chis level was a desire to have che plant in a mode where it was least suscepcible co fsed system induced cransients. To l l achieve chis, che feed water crain would have two feed pumps, cwo condansoce. I and cuo condensate booster pumps in operacion. This feature puts the feud 1 system in its normal full pouer lineup and makes the planc least susceptible i co feed svscem induced transients. - I l RE-ASSESS. VENT OF LOW pCWER OPERAT:ON: l l The quescion of extended low power spEYacion has been further revieued given che'recent incerest in potencia11y operacing ac an even icwer extended power
. becu bar 12, 1983 pg3 Two plateau and ac a perceived lower risk. If one examines the various planc condicions at various power levels chey. find the following:
Low Power End: Ac che low power end, certain condicions sec technical limics on operacion. The curbine is normally brought on line in che 12-15: power range. In the range of 15-20 power, the once chrough sceam generaccrs are not fully operacing in che once chroug6 mode but cend to operace more like recircula-cion steam generacars, and over an extended period of cine, would probably concentrace feed impuricles. (No on line blowdown capabilicy exiscs for che OT5G's.) In the 15-13: range, che curbine cannoc be loaded for extended periods because the last stage blading will be operacing far frem its aero-dynamic design point wich increased probability of blade buffecir.g or fluccer. Ac 22-23 power, boch trains of condensace and condensace booster pumps are operacing but only one of the sceam driven feed water pumps is in operacion. At chis power level, feed water flow is just being cransferred from che scarcup feed water regulating valves co che main feed wacer regulacing valves and feed system concrol requires manual accencion. Ac abouc 25: planc power, che excraccion steam system has marginal capability for normal feed water heacing and other syscens are well off their design poinc. For example, at abouc 25:, only two powdex vessels are in operation in che condensate syste=, and chey would be operacing ac abouc 2000 gpm each vs. a design value of abouc 3300 gpm each. In che power range of 23-30%, che "cenderness' of the planc is largely eliminated, che feed system is controlled on the full flow regulating valve, all of the planc automatic control loops can be utilized for operacion includine chac for feed pump control, and feed wacer heaters and drains would all be operacing in a normal mode. Ac che 28-30 power level, che steam generacors should be operacing in a once chrough mode, secondary side chemistry should be maintainable and other auxiliary systems and planc
' behavior would be expected to be normal. Ac che 25-301 power level, however, cne planc would still be on one feed wacer pump, and chus, che lineup of the feed train would be different chan for normal extended operation.
Hi2h Power End: On cne upper end of a band of low power extended operacions, there are some regions which incuid be avoided. At approxi=aceiy 007 pouer, che planc exhibics ics masc pronounced concrol instabilicies wich che maximum pressure oscillations michin the sceum generacor secondary side. These pressure oscillacions are speculaccd by some to be concributing to secondary side cube failure problems experienccd ac some plancs. Poact levels boca above and below 604 show reduccions in these pressure oscillacions. At decreasing power levela from 6C, che moderator temperature coefficient becomes less negative wnich is undesirable for many plant transients. Ac 4>)-42: power, che second feed water pump can be brought on and the feed system placed in the nor=al operating mode. At about '0 power, it is likely chac the main steam safeties would be challenged in che event of a turbine ~ crip. Ac 28-305 poker, che saiecies would probably not be challenged by such a crip. Ac 452 poser and above, che first and/or second banx of safeties (out of five) would be 12cced for short_duracions. (See Accacnment 1-Parcial Load Fluid Syste: Operating Characteristics)
. kgo Three
.Safetv & Transient Considerations for Oceration with Reduced Power:
An evaluacion was made of the time to RCS saturacion and to core uncovery for a loss of henc sink event wich no HPI and with one HPI pump, with iniciacion at 1200 seconds for 100%, 40% and 25 power operacion. The results shown in Accachmenc 2 indicace chac there is an increase in time to core uncovery for "
, a reduced power condicion when no heac sink and no HPI are available. With one HPI pump available ac 30 minutes, core uncov'ery may occur for full power operacion but noc for operacion at either 40% or 25 pcwer.
SB LOCA evencs of 0.01 and 0.005 ft ' were also evaluated with no EFW and no '
. HPI and again an improvement *in cime available prior co core uncovery i's found ac 40% and 25 power. The time available to regain HPI co prevent core uncovery is correspondingly longer for lower power operacion as can be seen froe che accachmenc.
For' a curbine crip, che benefits of the reactor scram on curbine crip can be seen for all ooser levels in terms of che reduced need for steamline relief following che crip of the curbine. For a crip from 100% power, three out of five banks of stnaaline rellei valves will lifc. Only one bank will lifc from 40 power. The curbine bypass and ADV capacicy is sufficient to prevenc any relief valve liicing for 25: power. . Based upon sensitivity scudies documenced in BAW - 1610, ic is concluded chat chore will be a reduccion in peak RCS pressure following a loss of feed water wich failure of che RPS to crip the reaccor from lower inicial power levels.
. The scudy indicaces a 500 psi reduction in peak pressure for a 15: reduccion in inicial pwoer. Ac 40% or lower power, we would expect pressure to stay less chan the recommended pressure limic of 2750 psi. Any benefic of 25:
power over 405 power would probably be seen in the longer term response co an ATWS racher chan in che inicial pressure spike. - Since only one feed water pump is in operacion at 25 power and since che ICS reaccor runback feature prevencs a scram from 100* power, a parcial loss of feed water is more severc from 25: chan 100% because ic results in a loss of all cain feed wacer, and che need for EFW iniciacion. At 40 power, cwo feed water pumps are running, but one pump is more chan sufficient and thus, chere , is no transient on parcial loss of feed water from 40$ power. See Accachmenc 2 for more quantified informacion regarding plant cransients iniciated ac differenc power levels. Qualicacive Risk:
- Da'ca is noc available to make a realistic quancicacive assessment of public risk as a function of power operacion in the 25-42 range. Qualicacively our judgemenc is chat risk, defined as the frequency of occurence cimes consequence, is clear 1v subscancially reduced as one moves down from 100% power to che 40
- range. Ac 40 power, the frequency of a cransienc or accidenc is not c perceived to change (compared to full power) but che planc and accidenc conse-quence should be less given the lower decay henc and greacer cime for opera-clonal response. Most of the FSAR accidents would be judged qualicacively to
P:ge Fdur be of lesser consequence. Over che range of 40% down to abouc 26: power. the continuing reduccion of decay heac is judged to furcher reduce risk as well as furcher increase the cime for operacional response to an accident condicion. Qualicacively most accident scenarios would seem to be less severe but these have not been examined in any significanc depch. However, we have reviewed the dominant risk sequence from other B&W NSS planc PRA's and conclude chac ' scme of the sequences would be less frequenc (due to addicional time for operator accion) in the 25-42 range compared to 100% power. In the 28-40 range, chere is an offsetting factor in chat the plant is running on the one feed water pump, and it is judged chat the probabilicy of feed system iniciaced crips is increased. Increased frequency of feed system upsecs will lead to
. grescer frequency of challenge to reaccor proteccion systems and hence provide some increase in overall risk. The relative crend of cocal risk due to chese competing factors is not able co be quantified at this cime. At approxi acely 28 power and belcw, che greater degree of manual control and "cenderness' of the planc is judged to offsec any risk reduccion which might occur from lesser decay heac and, cnerefore, leads to no risk advantage below 28-30 power.
SUMMARY
In the 20-30 power range, a nu=ber of planc syscams and control loops are operacing ac the end of their performance range or require manual intervencion. At the 40-42: power range, che ability to place che second feed water pump inco operacion places che planc in its normal operacional lineup with all controls in aucomatic. In addicion, operacion in the 40-42: range and above allows viccually all planc operacing procedures co be used without modificacion whereas extended operacion below the 30 level will require a number of cemporary' procedural changes. Operacion ac 40-42: is judged qualicatively to clesrly be a .reduccion in risk as compared to 100% power and would fully meet all of the cesc criteria inicia11y identified. Operacion in the range between 300 and 40-42 involves running wich a single feed water pump but the resc of the plant is in its normal mode, and other chan increased frequency of feed wacer upsecs, there is liccle basis on which to judge variabilicy of risk. Excended operacion (greater chan one or two weeks) at below 30 power is not recoc= ended. We recoamend furcher chac che original cesc objective be maincained. Adding a 10 day or so operacing period ac 25-30 pcwer would be accepcable. O e
GPU NUCL.AR CORPCPATION W har 12, 1983 ATTACHMENT 1 THREE MILE IS.AND NUCLEAR GENEPATING STATION, UNIT 1 Partial Load Fluid System Ooeratine Characteristics .
. PARAFETER OR 9 25% @ 4C%
ISSUE OF CONCERN LOAD CAPACITY LOAD CAPACITY
- 1. OPERATION OF CONDEN5 ATE SYSTEM
- a. Pumo(s) operating Yes Yes at minimum or -
are ater flow.
- b. Nunber of Condensate Two. Second pump Two Pumps operating. brought on line just prior to going above
. 25% load.
- c. Number of Condensate ditto above Two Booster Pumos oo era ting . ,
- d. Condensate Polisher Fewer polisher vessels More polisher vesseis Ooeration, can be in service; in service provioing more effective polisner performanc e.
'~ 2, OPERATION OF MAIN F EE DWA TER SY ST EM .
- a. Pu nts coera ti ng a t Yes Yes (But a transient causing ( No rec i re. cycli ng minimum recuired flow or cre ater, reduction in requireo problems foreseen) flow could cause recire.
l valve to cycle open & l closed.) e hp
ATTACHMENT 1 P ag e 2. PARAPETER OR 0 25% @ 40% ISSUE OF CONCERN LOAD CAPACITY LOAD CAPACITY
- b. Feedpumo tu.rbine-drive operating at most desirable overating point.
- one (1) oumo Yes Yes operation.
- two (2) pumo Some potential for operation increased blace Yes er ro sion. Less flexibility in speed contro l ra ng e. ,
- c. Tra.n sf er from Transf er occurs at Transf er te main feeo-start-uo feedwater 22-23% power. Feeo water regulati ng val ve-reculatino valve to system control completec well :etore main feedwater recuires manual 40% loac.
regulating valve. attention.
- d. Nur.b er of mai n One (i) (Loss of Two (2) (Loss of cne feedwater cumps puco causes loss pu.To is ccmoensateo sy operati ng per . of feedwater) otner pump picking up nonnal ooerating loac.)
orecedu re s.
- 3. OtCE T908.GH STEAv GENERA TO RS ( OT SG 's )
PERFOR MA NCE . Onc e-th ro uch OTSG in transition OT SG i s i n onc e-th ro ugh a. vs. recirculation from recirculation mode; no impurity mode. to o nce th ro uch i n conc entra tion. ra ng e of 0 - 20% power; imourities will tend to concentrate in OTSG.
- b. OTSG outlet steam In superheat region Slightly gre ater quality. degree of steam soperneat than at 25%.
ATTACHMENT 1 Page 3. PARAMETER OR 9 25% @ 40% ISSUE OF CONCERN LOAD CAPACITY LOAD CAPnCITY 4 MLIN TURBINE-G ENERA TO R PERFORMA NCE a.. full auto control Yes Yes (Transfer to Auto 915% power) .
- b. Light load operation Increased probability No concerns expressec ey affects on last stage of blade buffeting or turbine manufacturer. <
blade loadino , flutter. r .. 5., EXTRACTION-STE AM F EE0WATEM HEATIhG SYST EM PERF09)Rr'CE
- a. All extraction Yes Yes steam block valves oben l
l b. Need for eicht5 Transition from No aux, steam necessary; stace feedwater aux, steam to extraction steam fice heating from auxil- extraction steam is sufficient to ensure iary steam vs. is estimated to procer feeowater neating occur at approx. in low press re f eed-turbine extraction steam. 20% power, water heaters. l l h
- 6. EEDWATER HEATER ORA I A SY 5T ES Transition to Eichth stace Transition to feedwater heater normal drains normal Stn stace drainina nonvally at about 201 load. heater crains sell before 40% power.
(i.e. aux, steam l isolated & aux. steam drains no l anc er beina returned to aux. - boiler) .
AllAustNi 2 GrlHe liecember 12,198.l.- EIF[CI 0F Pinf(R lev [I 011 Pl Alti RLMAV10It
- Page 1 IMattslG 5LLLI.Ille IRAllS1[CIS Affl ACCibfC15 Analysis Method Dr
[ vent Reference Parameter 1005 Power. 401 Power 251 Power Comment s Gass af Ileat Sink C5MP & l. SG dryout time 110 sec = 130 sec 130 sec 205 saturatten results in rapid insurge Cith e3 HPI R[ IRAN 2. PORV actuation Line 130 sec 160 sec lidi sec into pressurlaer with subsequent opening j 3. Ilme to saturation 850 sec 2l00 sec 4000 sec of both pressortrer safety valves. RC pumps
- 4. Core uncuvery 2000 sec 7000 sec 15000 sec off at time sero to simulate loss of offsite
- power.
58LOCA (0.01 f t2 ) CSMP 1. Minimum pressure prior 1550 psia 1430 psia 1400 psia RC pump heat not included. Pump trip on ulth no EFW & no to repressurisation loss of SC margin would eccesr from 500 to HPI 2. Core encovery time 2400 sec 8700 fee 12000 sec 100 seconds for all three power levels. HPI
- 3. Ilme reg'd for HPI Core uncovery 2 hamrs 3 hours initiatiews implies one HPI pump only.
initiat te.:n to prevent dependent upon Boller-condenser needed from full power to core uncovery IfW Initlation depressurlie system and get higher HPI fIow. (estimated) [fW pot needed from 4tM or 251 power because-
- 4. Ilme req'd for [IM 30 mlantes [FW not req'd [fW not reg'd B-C not required.
Initiatiosi to prevent core uncovery (estimated) ( SBLOCA (0.0)5 f t? ) C5HP 1. Minimwn pressure prior 1600 psia 1550 psia 1500 psia RC pump heat not included. Pump trip on with no [fW & no to repressurtration loss of SC margin would occur from 600 to . HPI 2. Core uncovery 1ime 3000 sec 8000 sec 15000 sec 1200 seconds for all three power levels. HPl !
- 3. Ilse reg'd for itPI Core un<.overy 2 hours 4 hours initiation implies one HPl pump ogly 401 initiation to prevent dependent upon case repressurlies to PORV sgtpoint and core ;
core uncovery [fW initiation uncovers faster than 0.01 ft break which (estimated) does not repressurire as much. Thus, core
- 4. Ilme reg'd for [IW 30 minutes [FW not reg'd [IW not req'd uncover initiation to prevent 0.01 f tg case.
f asterB-C for only this needed case than from forfull core uncovery power with one HPI pump. Ilot needed from (estimated) 40% or 251 power. 0 p mili
- - _ _ w_ __ _ , ," ,
-c-Method Or .
Page 2 EvenL Neice con e Par.amet er Milli Power 40E Power 251 l'oen.r runecit s ' tbss cf licat Sink CSMP & I. SG dryout time H'.les see IP 1r4 sec eg Belf sec itPI initlation assumed to occur at 1200 with On2 HPI Pep RLIRAN 2. PORV actuallan time 13'#10 sec stofc6 sec g .10tr sec
- Aeconds. ACS salvration ressalts in large (Fe d & 8 teed) 3. Core uncovery Norse None Nune ' insterge into pressurlier and subscequent lift
- 4. Ilme to saturation H50 sec len saturation No saturation on both pressurizer safety values. AC peamps
- 5. Ilme to pressurlier 850 sec seo 5/V lift leo 5/V lif t off at time sers'to simulate e loss of off-5/V- lif t with PORV open , , .
, site _ power. - For 1001 p.,wer, suhtcollag is
- 6. line to regain sub- 3 hours Never lost Never lost u e not required. , <
ccoling ' ~-
~
Lass'of feedwater B N -1610 & l. Feak RCS pressure 3464 psig - 7150 psig 2750 psig Base <l upon Ruf -1610 and sensitivity study AIWS SM -10099 2. Moderator temp to- -1.05 x 10-4 -0.014 x 10-4 -0.737 1 10-4 ishich shows that:
& GPU Calcs efficient 101 MIC = 120 psi (P2-21 of BAlf -1610) 151 power a fA10 psi (Figure 5-1 of BAnf-10099) i7urbine Trip RETRAN 1. Steamline relief Banks 1-3 Banat I only lione full power turbine trip analyred with a (Assume reactor valves lifting recommended set of 105 settings. With scram on turbine 2. PORV actuation time 10 0 lift 90 0 lift No lift proper secnndary side response. RCS pressure 3 rip) 3. Peak RC5 pressure 2230 psia 2230 psla* 2230 psle* reaches pressurifer spray setpoint and then turns around. Only 1001 power case was analysed with PCTRAN. Other power level results estimated from other analyses. 401 secondary safety valve behavior based upon a loss of feedwater analysis. Two small safeties will also lif t for 1001 and 401 but T not for 255 power.
Trip of a $1ngle RITRAN & l. Peak RC5 pressure 2235 psla** slo change 2235 psia ** Two feetheater pumps operational at 401 Main Fecdwater Pump M5 for 2. Steamline safety None None Hone power. One feedwater pump capable of energy
]Parttal L0fW) 1001 power valve lift removal up to 605 pnwer. Thus, trip of one case 3. Integrated control 601, Nm runback Plant scrams pump at 40E power does not result in system euriback power required transient behavior. Power is below 605 run-level back point also. Since single feed pump
- 4. Scram No scran No scram Scram on loss running at 251 power then trip of one pump of feed water is a loss of all MFW and is a more severe transient editch results in strame, init ia t ion of fIW and ACS pressurliat lon.
- Not Analysed M Pressuriser Spray Setpoint
- I i . .}}