ML20125B299
| ML20125B299 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 10/24/1979 |
| From: | Brown S VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| 792A, NUDOCS 7910290316 | |
| Download: ML20125B299 (4) | |
Text
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October 24, 1979 Mr. James P. O'Reilly. Director Serial No. 792A Office of Insoection t Enforcement PSEC/GLP:mec: wang U. S.
luclear Requistorv Crvr.miesion Tsegion II Jeck1t 'lo. 9-P 9 10' Aarietta Street, Snita 1100 AtIant3, Geornia 30303 Daar Mr. O'Reilly:
Or September ?a, in70, a report was made under the previsions of 10CFR50.M fel concerning interaction between non e$fety grida svstems end safety orade systems.
'da ' ave reviawed thesa conc.?rns and have made the followino evaluations.
1 Wes+1nahouse. as a cart of its ervironmental goalification activities for IEEE M -197d, reviewed original assumotions it mace for safety analvs4s reoorts.
Specifically, could a severe environment cause a f ailure of a non-orotacticn grade component that was oreviously issumed to remain es is" and alter toe results of the design basis analysis? Westinghcuse addressed tha f ailure of a control system due to an adverse envircoment insida or j
outside containment following a nigh enerqy line ructure which could negate a orotective function perforned by a safety grade system. Thev determined that ootential inte*'ictinns ex4sted for the followino systems in conjunction with a feedline ruoture event:
1.
Steam gererator c?dar operated relief va'va centrol system 7
Pressuri::er ocwer operated relief valve control system 1
Main feedw'tter control system TFey furthar det?- + ad that a potential intar9ction axisted for the automatic od m t-ol system in con 4 unction with 29 intermafiate steam line ruotura event.
These fou* consequential f ailures were found to violate internal West 4nghouse saf?ty innlysis crite*ia.
So*cificallv, kot 101 Anil4no could occur following a feedline rupture with a consaouantial f ailure and minimum 0904 -nulr4 Fall $n'ow 1,t0 prinr to 3 "??CICr trio foll1 Wing an dntermedi3te stesm lina mmture Sith a consacuential f ailure.
A revlw of the 'forth Anna F4nal Safety Ansivsis paport was performed to datar91ne if this new information was outside the bounds of the FSAR.
Hot leg boilinq was cer9itt?i kv the analvsis of a feedli"e ruoture event IFSM sectieft 3.1.E.M.
As oreviousiv stated, r? cant destischrusa crit?ria di'i act cermit this result.
Therefore, whila a NSSS vendor crite-ion has not keen met, this rew Tnalysis does not axceed the existing WR snCveis. ForNorth6 w0 t
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Anna, an intermediate steam line rupture event is directly comparable to 3n excessive load increase incident (FSAR Section 151.11). This new information does not avcW the nxistina FSM analysis.
SnecificWy, thras rvetn*
nrotection circuits are provided (the first is assumed to f all in this new analysisi:
Ovar tamnaratura Delts-T 115 Powar Parice (Wah Flux)
Overpower Delta-T rurthar clari'icstion nf *e anvel pe of protecticn provided by overtmperature Delta-T and ovarpo>ver ?elt,-T is s'iown in FSM Figure 15.1.1..
Olote that this new Analysis by tha 1555 ven,10r fi d.iot coas t de* Other protecti:n circuits which.sould m ye to mitigite this ? vent, 1.3., Saf?ty injection Trio f-nm Steam l be Differential N ssura nr Cont *9ent He Pressure.1 info-mitf c is providad belov for North Anna control systems with resoset 3
to t"? soecific conce-ns of I.E. Information Notico 79-?', iiswd Septem%r la, 1970 Sta?n cene ater nowe onorated ralie# valva control svstMi:
These valves 2-a loc:ted in the dain Steam Valve House.
They could be subject to an adverse environment in the event of a main feedwater line rupture in this building.
(The new destinghousa analysis sssumes a break between the containmant oenetration and the first upstream check valve. This run of pipe is very short and is ANSI 33*.7 piping.)
The steam generator PORV's fail to a closed position upon a loss of air or electrical signal. These valves are sir operated with an electrical to oneumatic interface device which converts the alectrical control signal to a sna m tic control signal.
FSAR Suppl eent 510.11 states that the aux 111ary feedwater system meets the guidelines of 3 ranch Technical Position APCS 3 No.
'.0-1.
Additionally, FSAR Suco! ment resconsa 10.135.2, in effect, already addressed the new Westinghouse analysis in that it fiscusses the loss of tha turbine-driven Dumo and one noter-91ven cup.
Additionally, FSAR Suoplement $15.13 indicates that one steam garerator is 1
sufficient to provide cooling and that FSAR 15.4.2.2. is the bounding an11ysis for the ere-on-one auxiliary feedwitar system.
Pressurizer power coereted relief valve control system: These valves are incated inside the containment. They could be subject to an adverse mvi*onment in the event of a main feedwater line ruoture in this
- u11 ding.
The pressurizer PORV's f ail to a closed position upon a loss of air or elactrical signal. These valves are air operated with solenoid valves in the air lire.
An additional design feature is provided that couli mitigate the affects of a PCS inressurintion if a PORV were opan. Air is ramoved from the control air syst-em for these valvas by a signal derived from protection grade equipment upor 1CS eressure Isensed in the pressurizer) fallinq 5elow a preset value.
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Viin feedwater centrol svstam: Thase valv?s tra located in the Service SUIlding and are upstream of the containment onnetrations and check velves. They could be subject to an adverse anvirorment in'the avent of a mein fee:9 ster line rupture in this building.
(Note that a orsek in this area would Se unst-eam of tha chack valve fer aach line.
FSAR 1:4.?.2 states that this break location would be treated as a loss of normal fe*dwater which is cove ed "" F#^R 1".9.1.
This srctinn assumes that the steam generators are at a nominal 0% indicated level. This is, then, already censistart with the naw Westinghouse ansivsis for this censequentici failure.)
l T'm '?adutt?r *egulatirg valves f ail to a closed oosition uoan a loss of He or electrical signal. These valves are air oper3ted dt" solanoid valves in the air line and an electrical to oneumatic inte*fsce device ubich convr-ts t'e elect-ical control signal to 3 pnaumatic contr?1 cina2l.
a tomatic rod control system: The excore oower ranga detectors are u
a.ssomad to f ail in the naw 1SSS vendor analysis.
These detectors are located within the neutron shield tank. The d?tectors are liftad into dry.
soaces within the tank.
These spaces are *de:1-ended" and, thus, the detectors vould not be axposed to a severe environment. Additionally. the largest ooening into the trea under the vessel ir at the RfiR mer?aninc.
-@ich is 35cvc the bottom of the snield tank.
This further reduces the octe9tial for an advarse envirenment to affect the detactors and c3ble connectors.
Incore detector guide tubes (and thimbles) pass through much of this o! ping. N!S cable is run in corduit from th9 detector to the containment electrical oenetration. The detectors themselves do not have an environmental qualification. However, they can operate at 1750 for eight (8) hours and prohaSI.v could operste at en einvated tencerature for shcrter dur3tions. The cabla has been cualified for 3000F for 15 minutes followed by ?S20F for the ceriod of 15 minutes to 13 days following the event.
(The cable is radittien qua11f'ed to 2 x 103 rads and is qualified for a spray solutten of boric acid and sodium hydroxide.)
Additionally, FSAR 5.2.? states that the containment dasign tenperatura is exceeded for a major steam line break for only one (11 minute.
The Sceak size assumed by Westinghouse in this new analysis is 0.1 to 0.25 square feet at a reacter oower level of 70 to 100%. This break is small comoared to the large v in Steam line break. Thus, the peak containment a
temocratura would be rasched later in the event since the Slowdown rate is lower. Consequentiv, tS9 NIS onwer range detectors should remain operable to provide the orotection for axcessive core power.
Lne, Tar, a tion c
Licensed casetor coerators and licensed seninr reactor coerators will
-aview this *esconse so that they will be informed c' ilew informatinn cererated hv Mestinghouse.
This review is intended t? orovide an understariisq cf *ecent NSSS vendor work and to develoo an awaraness of system inter =:tions.
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Based on this review, we have cercluded that thi, dots not constitute a 311nificant safety concerr.
'42 consider this c'tr final report on these four system interaction concerns; if 'itether
'toreation is required, oleise advist.
...,19ey truly snurs.
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senior Vica Dreeident Dowar Stttice Engineering and Construction cc: 1. ' lictor Stollo. Director Jffice af Insocction & Enforcement Mr. Hircld 9. Denton, Director Office of Nuclear.teactor Reculation l
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