ML19224C164
| ML19224C164 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/04/1975 |
| From: | Schwencer A Office of Nuclear Reactor Regulation |
| To: | Mallay J BABCOCK & WILCOX CO. |
| Shared Package | |
| ML111230347 | List: |
| References | |
| NUDOCS 7906290233 | |
| Download: ML19224C164 (3) | |
Text
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RSSoyd ei 211. Ja: ass F. Mali n7 VAMoore i Eanaggr Licen31n3 TR ads I Babcock I, Wilcox Cc =peny TR BCs Euclezr Pcvar Conorntien LWR BCs
- 7. O. Lox 1260 IE (3)
- Lynchburg, Virgi-4 a 24503 ACES (16) l l LWR 2-3 LPMs
Dear Mr. Malley:
RL AD:
I These During the past year, several SARn for BW reactors have been revieved.
3 and 4, C cervood 2 and 3 Nor*.h l.nnn includs the 2allefent: 1 c=d 2, Sur r
Sinco a concurrent 3 and 4. UT?SS 1 cn: 4, nnd Febble Spri 2s 1 cnd 2. plants.
review of E-SA".-241 ucs in progress, requests for informatica,en generic itens vero oubnitted on the 3-SAR reviev vith the intent. to apply the information
, obenined to the.chove plants.
Eowever, tha E-SAR-241 application was withdrevn before rc ponses to these generic requests were sub=itted..
!!! We ' cave decided that it is inappropriate to require responses to all these
' Eequests f ron a cingle applic.nt (e.g., PCI for Pebble Springs).
Eovever, this Accordingly, the I deneric infor tion nuet bo previded in a t'2ely canner.
info
- - which.: require hn beca grouped into categories (see enclosure),
dod we prcpose that t2ese be use.' as bases for 3 W topical reports.
These icports should apply to all current LW plants.
The enclosure is not = cant to
' ~-'Aual repo rts.
The nunber of reports to be cubnitted define tho scope c' dad the categorics of infernation to be covered in each report is lef up to SW.
lf that SW subait detailed outlines of the proposed topical reports in,
'e propose ix vecks and that a =ceting then be held between 31W and the staff to discuss o
the outlines.
It is requestod that the reports be aubtitted by the endaof this Year.
l:If you have any questicns concerning our request for these reports or if you Ifeel that you can not nact this schedule for sub=ittal, please contact us.
I Sincerely, Cp w g 5
A. Sch n ee, Chief e
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D h 6S k Division of Reactor Licensing E
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En closure:
Topics Requiring Additional k[ey p,
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Infor=ation fro: BW
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FRO:i B&W 4
1.
Integrated Control System
~
It is stated in all license applications involving a B&'.1 f:SSS that the ICS is a non-safety system.
In some applications, the results i
of an accident analysis performed for the ICS are presented.
Submit additional infornation regarding this analysis to show the following for each case analyzed.
i (1) The assumptions used and their justificat'.cn.
l (2) The worst c se conditions and their bases.
j (3) The specific subsystem of ICS involved in each action, the 1
effects of its failure or malfunction on other sub:iystem
~
. of ICS, and the ultimate effect on safety of the piant.
(4)
Interfaces with sys'tems'not s'opplied by BSN and the specific criteria for these interfaces if related to safety.
.(5) A complete list of functions performed by the ICS in all ll modes of operation.
.t
!N.
For each of the accidents analyzed in Chapter 15.0, provide a detailed i
listing of all systems and equipment for which you take credit in makinc your assumptions in each analvsis.
T f
fN 7
2 N
3.
For the case et a s:eam line break accident inside containment it assumed that the main steam line isolation valves and the feedwate isolation valves to the affected steam cenerator clost.
l Demonstrate that in the event of a steam line break inside contair..ent concurrent with loss of offsite power, ne single failure in the main
}
steam and feecwater isolation valves will defeat these two protective i
actions.
Identify.all instrumentation, controls and elec.rical equipment j
that are not desicned to Class IE requirements for unich credit is takeri l
in this analysis.
Supplement your discussion with a sketch shouing the feedwater/ steam system and identify all ccmoonents that need to be actuated in all cases of a feedwater/ steam line break inside end cutside i
i containment.
Submit the results of a failure mode and effect analysis that j
demonstrates that the requirements for feecwater/ steam line isolation
/
are met while allowing erercency core cooling
- 4.
Supplement the analysis presented in Section 7.4.2 to shou that the RPS and CRDCS designs meet' the requirements of GDC 25.
Considering recent PWR operating experience and the fart that the coerator has provisions to manually control sincle rod motion, the single rod withdrawal accident should be analyzed as an anticipated I
event.
Specifically, provide information justifying the selection of worst-case conditions of time-in-cycle, power level, power distribution, peaking factors, control rod worth, control positions, etc.
Jus ti fy any differences between these selected values and those used in th7 rod ejection accident.
257 032 3
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4 Provide a plot of maximum fuel centerline temperature as function of time for the control rod misoperation accident.
Demonstrate that the worst case initial conditions have been analyzed.
events do not Justify that the following control rod misoperation have to be analyzed.
(1)
Inadvertent withdrawal of two or more control rods at the same time.
(2) Leaving one or more rods behind (i.e., stuck rods) during rod bank withcratal.
(3) insertion of a rod bank with one or more bank rods stuck.
Show the changes in ninimum CS5F. and minimum kw/ft for the
- control rod withdrawal and misoperaticp events. -
s
, Describe the startup and full pcuer control rod withdrawal and mis-operation events.
5.
It is not cle.c fech the information presented in the PSAP. what your desiqn provisions are with regard to disconnecting of the Reactor Coolant pumps frcm their electric pc.:er busts because of excessive grid frequency decay rates, to assure that tht-cumps kinetic energy is available for ficw coas-down.
Prcvide additionti information to shcw whtt ycur reouirementes are in acceiyant analyses presented in Chaptar 15.0.
Your respons'e snould state the
,llaiting frequency decay rate H /sec.
6.
There is not sufficient informatica in SAP.s to determine that your design, utilizing the high and low pressurizer level signal, in lieu of the high containment pressure signal, uill provice an ecuivalent degree of assurance that the reactor uiil trip prior to or coincident with ECCS actuation.
The staff concludes that if the analysis for the effectiveness of the ECCS performance takes credit for a reactor trip, we require that both diverse signals actuating the ECCS be used to trip the reactor, thero-fore, either (a) modify your cesign to include a high building pressure trip to trip the reactor; or (b) demonstrate that a high and low pressurizer level signal will perform satisfactorily for all.ccident conditions, and will trip the reactor prior to or coincident with ECCS actuation, thereby assuring effective emergency core cooling.
Include in your response what I
assurance is provided that this level measurement will maintain its accuracy during blowdcun.
t 7.
Provide a list of all safety related setcoint settings used for reactor For each setting tabulate (1) the setpoints used, (2) the operation.
Technical specification limits (or range), (3) the maximum and/or minimum value alleued (as derived from the safety analysis) prior to onset of 257 033
conditions that are unacceptable from the standpoint of plant safety.
(i.e., safety limi t)
The listing should clearly define the relationship bet'. teen the Technical specificatica limits, the setpoint settings, and the mcximum safety limit assumed in the safety analysis, and demonstrate that adeouate margins are provided that insure actuation in a timely manner.
~
~
The listing should include the following, in additicn to the reactor protection and ESFAS setpoints:
(1)
Core Flooding Tank Isolation Valve interlock setpoints.
(2)
Decay Heat Isolation Valve interlock setpoints. '[,
(3) Auxiliary Feedwater Actuation parcceters (i.e., level and pressure setpoints).
(4)
F.ain Steam Line Isolaticn actuation setpoints (i.e., pressure).
(5)
Setpoints used in change ever frem injection to recirculation m de of opera tion.
(6)
Other safety related interlock setpoints.
e 257 034 e
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