ML19209B548
| ML19209B548 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/05/1979 |
| From: | Cavanaugh W ARKANSAS POWER & LIGHT CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 2-109-5, NUDOCS 7910100099 | |
| Download: ML19209B548 (22) | |
Text
.
ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK. ARKANSAS 72203 (501) 371-4422 WILLIAM CAVANAUGH lli October 5,1979 Vice President Generadon & Construction 2-109-5 Mr. H.
R.
Denton, Director Office of Nuclear Reactor Regulation U.
S.
Nu c l e a r Reg ul a t o.- -- Commi s s i on Washington, D.
C.
20555
Subject:
Arkans as Nu cle ar One-Uni t 2 Docket No. 50-368 License No. NPF-6 Interac t ion Be tween Non-Sa fe ty Grade Systems and Safety Grade Systems (File:
2-1510)
Gentlemen:
This letter responds to your September 17, 1979, letter on the subject of a " potential unreviewed safety quest'.on on interaction between non-safety grade systems and safety grade systems".
This potential problem was further ad-dressed in IE Information Notice 79-7.2, issued Sept sber 14, 1979.
In conj unc t ion wi th Combus t ion Engineering we have reviewed the specific non-safety grade systems listed in IE Informa-tion Notice 79-22, as well as others, for potential inter-actions that could constitute a substantial safety hazard.
We have not been able to identify such an interaction.
While, in some cases, we have identified potential vari-ations from the FSAR Licensing bases, the basic conclusion of the FSAR, that these events do no constitute an undue risk to the health and safety of the public, remains uri-changed.
In our preliminary screening for potential adverse ron-mentally-induced failures of non-safety grade equipment, it appears that po t ent i al problems a ri s e onl y veh en su ch f a i l -
ures are combined with other failures or operator errors.
The probability of the type of high energy pipe breaks we are considering is small.
Such breaks are also more likely to be small cracks rather than abrupt failures so that the resulting adverse environment builds up over a period of time providing the potential for detection prior to com-ponent failure. Additionally, our review recognized the 7910100 099 MEMBEA MlOOLE SOUTH UTIUTIEs SYSTEM
2-109-5 Mr. H.
R.
Dent;n October 5, 1979 difference between a demonstrated deficiency (e.g.,
deter-mination that a control component would operate in a fash-ion not within the limits presented in the safety anal-ysis) and a potential unreviewed question.
As previously stated, we have not identified any events that would change the conclusions of the FSAR: nanely that these events do not constitute an undue risk to the safety and healtL of the public.
investigation within the limited As you mus t recognize, our time frame required by your September 17 letter must be considered preliminary and could not include detailed analyses.
Based on our preliminary investigation we are convinced that continued operation is warranted.
Justification for continued operation of ANO-2 is provided in Attachments 1 through 4. Attachment I summa r i z e s Can-bustion Engineering's efforts to date. is the event / int erac t ion ma trix genera t ed by Combus t ion Engineer-ing. Attachment 3 details generic event /interac tion scenar-ios ident i fied by Combus t ion Engineering.
supplements Attachment 3 and provides further plant speci-fic i n f o nma t i on.
In addi t ion Nucle ar Sa f ety Anal ysis Cent er (NSAC) has deter-mined that the probability of severe consequences resulting from one of these high energy pipe breaks is very low for a typical nuclear power plant.
The probabilistic analysis of IE In f o rma t i on No t i c e scenarios was prepared by NSAC as a result of AIF's promotion of an industry wide generic response to this concern.
We have part icipa ted wi th AIF to the extent possible.
The above referenced probability anal 3 sis will be su'omi t t ed to you by NSAC la ter t h i s we ek.
As a result of the Three Mile Island accident, there are a significant number of industry, governmental and regulatory investigations under way examining that licensing bases and the operating procedures of nuclear generating facilities.
These investigations are already identifying areas where studies may result in the consideration of new or revised events as part of the bases for assuring the continued safety of nuclear plants.
NUREG-0578 outlines several such events and suggests remedies.
NUREG-0 5 7 8 requ i remen t s for analyses of po tential safety problems envision the kinds of scenarios identified by Westinghouse and made the subject o f IE In f onna t ion No t i ce 79-22.
Section 3.2 of NUREG-0578, Page 17 states in part,
) ) b 4 ' 9,=3
2-109-5 Mr. H.
R.
Denton
-3 October 5, 1979
"...The NRC requirements for non-safety systems are gene ally limited to assuring that they do not adverse-ly aifect the operation of safety systems.."
Further, on Page A-45 of NUREG-0578,
" Consequential failures shall also be considered..."
We, therefore, believe that the scope of the action re-quired by IE Information Notice 79-22 is :onsistent with the requi rements of NUREG-0 578 and should theref ore be in-tegrated with the planned response sequence for compliance wi th the NUREG.
We are aware of the need to establish a priority of consi-deration of new issues based upon their potent.al impact upon the health and safety of the public.
Such a priority is required so that the resources of skilled engineers and anal ys t s can be applied to the review of the most important concerns first.
Very truly yours,
[
t n.w W. Cavanaugh, MIJ WC / s!OW /ew Attachments 1134
~24
STATE OF ARKANSAS
)
)
SS COUNTY OF PULASKI
)
William Cavar:augh III, being duly sworn, states that he is Vice President, Generation & Construction, for Arkansas Power & Light Company; that he is authorized on the part of said Company to sign anJ file with the Nuclear Regulatory Commission this 5coplementary Infonnation; that he has reviewed or caused to have reviewed all of the statements contained in such infonnation, and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, informa-tion and belief.
]
f si c tm m William Cavanaup( III SUBSCRIBED AND SWORN T0 before me, a Notary Public in and for the County and State above named, this M day of
/h)8L/ 1979.
7 Notary Public/
~f
~ ~
/2)}hr{
02 R
. 1st
/
My Conrnission Expires:
Mt Co. miss o., n;;_.u gy,,Sl m
113.
25 DESCRIPTION OF C-E'S EFFORTS At the request of the C-E Owners Group on Post-ThI Efforts, Combustion Engineering conducted a review of potential control systems interactions during high energy pipe break events.
C-E initially est 'lished a matrix of high energy pipe break events and control functio..s witnin C-E's ability to properly evaluate.
A list of separate systems that should also be con-sidered was developed and forwarded to the utilities participating in this effort along with a list of control function and events under consideration by C-E.
These lists are attached.
In the time available, C-E reduced this matrix so include only those systems and events which required further evaluation.
Some of thesc events were further eliminated by individual utilities on a plant specific basis.
A gen-eral description of the procedure used by C-E to reduce this matrix is listed below.
I.
An initial review of each postulated Control Function failure for each pipe break was completed and served as the basis for consideration.
Where a postulated failure could potentially increase the severity of a high energy pipe break, the fo lowing criteria were employed to resolve the concern:
1.
Is the postulated Control Function failure mode credible?
2.
Is the Centrol Function Equipment (Sensor, Cables, etc.)
in a location which could be impacted by the environment?
3.
Is the Control Function Equipent (Sensor, Cable, etc.)
qualified to operate properly in the postulated environment?
4.
Where the postulated Control Function failure is credible, could its impact potentially affect the conclusions pre-
.nted in the SAR? Consideration ~s such as Maximum Conn.;l tunction capabilities, and delayed, but proper operator action were employed in this effort.
It should be noted that the limited time available did not allow for ex-tensive anafysis,
Prudent engineering judgement was utilized to eliminate those events /interacticas which did not change the conclusions of SAR analyses.
Extensive evaluations involving the Auxiliary Feedwater system and otner long term cooling mechanisms have not been performed.
Auxiliary feedwater is being eveluated under Bulletins and Orders and Lessons Learned (fiUREG 0578).
This decision was made in order to concentrate on those items felt to be of greater significance in the short time available for assessment of control system high energy pipe break interactions..
In several cases, mcst notibly the PORV failure in the open nosition, no specific failure mechanism has been identified.
The only manner for such
)\\M
'6
a failure to occur would be for power to be inadvertently applied to the valve solenoid and not be removed.
Part of C-E's short term recommendations are for utilities to evaluate whether or not a failure mechanism of this type is credible.
The potential adverse impact of high energy pipe breaks on reactor coolant pumps was considered.
Both the seized shaft and the simultaneous three or four pump loss of flow were eliminated from consideration based on judge-ment thht these failures are not considered credible within the time' frame limited by operator action (30 minutes) due to environmental impact alone.
The impact of other potential loss of flow events (e.g. one or two pump loss of flow) during high eneray pipe breaks was reviewed and it was judged that the resulting raoid reactor trip was sufficient to ensure that the conclusions of the SAR would not change.
Each Utility further evaluated and eliminated items from the matrix (Attachment 2) based on current operating procedures or specific equipment configurations, locations, or levels of qualification. details potential event / interaction scenarios.
AP&L has not identified any changes that need to be made to emergency procedures as a result of our preliminary investigation.
1134
^27 MATRIX OF EVEtiTS/C0tlTROL rU!1CTI0ftS FOR FURTHER C0iSIDERATI0f! At4D ACT10ft Pice Break LEA Control Function SLB FWLB Ejection S8LOCA LBLOCA t
Pressurizer I
Level X
Pressurizer
{
Pressurc Pilot Operated Relief Valves X
X CEA Fosition X
X X
X Feedwater j
Flow l
X X
Bo ron Concentration Turbine I
i l
Control X
l l
Steam
{
i, i
Bypass X
i 1
Stean Dump Uostream of X
X l
i *, S ' !
J j
Steam Dump I
Downstream of i
X
{
MSIV l
Steam Gen.
Blowdown Condenser Reactor I
I Coolant Flow l
1 *7i
" 7 f)
{ l,,) ' r
?J
DEhCRIPTION OF REMAINING EVEllTS AND CONTROL FUNCTIONS 1.
Assessment of Control System Failures on Steam Line Break Event A.
Sequence of Events for Generic SAR Steam Line Break at Full Power, Inside or Ou'. side _C_ontainment 1.
Double-ended steam line break occurs 2.
Reactor trip on low steam generator pressure 3.
MSIS initiates to isolate the steam generators 4
RCS temperature decreases due to excessive steam removal 5.
Total reactivity increases due to moderator cooldown effect 6.
MSIVs close 7.
Pressurizer empties 8.
Low pressurizer pressure initiates SIAS 9.
MFIVs close 10.
Safety injection boron reaches core 11.
Affected stean generator empties, terminating cooldown effect, the transient reactivity reaches peak and decreases gradually due to boron injection 12.
Limited or no post-trip return-to-power 13.
Go fuel in DUB B.
Steam Line Break With PORV Control System Failure 1
Significant Interaction Effects:
a.
Increased Containment Pressure A stuck open PORV in combination with a stear line break has n7t b.
been analyzed.
2.
Assumptions Stean line break (large break inside containment for Item 1.A above, a.
any size or location for Item 1.B above),
b.
Inadvertently PORVs open and rem' in open a
PORV Block valve also fails to close when required c.
d.
Initial condition:
full power 1 1 h
~ 2.9 3.
It must be emphasized that no mechanism has been identified for the PORV to inadvertently open and remain open since its signal to open comes from safety grade equipment and the Dresser valves and solenoids are qualified for an environment in excess of 400 F.
P00RDUINAL
4 Sequence of Events Large steam line break occurs inside containment.
a.
Reacter trip occurs on steam generator low pressure within 5 seconds.
b.
Snould the adverse environment cause the PORV to inadvertently c.
open and then renain open, the following steps may also occur.
It should be noted that no mechanism has been identified wiiich would cause this to occur.
d.
Stean from PORV fills quench tank and bursts rupture disk re-leasing steam to the containment and causing additional containment pressurization.
Mass removal via PORV causes additional void formation within the e.
C.
Steam Line Break With Feed.iater Flow Control System Failure 1.
Significant Interaction Effects Steam generator filling - causing potential piping structural a.
problems 2.
Assumptions Small steam'linb break inside containment that does not cause a.
an immediate reactor trip b.
Feedwater flow exceeds steam flow due to failure of steam generator level instrument, indicating flow c.
SAR conservatism i.
no operator action within 30 minutes 3.
Sequence of Events Small steam line break occurs which does not cause en ir.r.ediate a.
reactor trip b.
Steam generator level instrument fails, causing an increase cf feedaater flow in excess of steam flow Stear generator begins to fill causing increasedmoisture con:ent c.
'; steam d.
- no operator action occurs undefined piping structural problems
.ould result It should be emphasized that this event can be prevented by e.
prompt operator action.
Safety grade steam generator lesel in-strumentation exists, enabling comparison with control grade level instruments of the feed system 1134
~31
P00R 03G M D.
Steam Line Break With Failure of Main Steam Paths Downstream of MSIV's 1.
Significant Interaction Effects a.
Increase post-trip return-to-power 2.
Assumptions a.
Large steam line break inside containment b.
MSIV on unaffected steam generator fails to close.
This sequence of events is pertinent only if this assumption is nade, c.
Downstream of MSIV's main steam paths fail open d.
Initial condition:
full power e.
SAR conservatisms i.
end of cycle core ii.
the most reactive CEA stuck out iii.
steam bicwco:n througn steam line break 3.
The number of failures which must occur during this ecent are significant.
First there must be the large break.
Then the MSIV on the opposite steam generator must fail to close. Tr.ere is a stuct rod on reactor trip.
Then steam paths downstream of the ' git"s
.m t be affected.
These include turbine control valves +nd steau du
- and bypass valves. The prob'bility of this evc nt orem rirq is.uch less than 10-0 per iecctor,ecr 4.
Sequence of Events ll3h 2
a.
Large steam line break inside containmeat b.
Reactor trips on low steam generator pressure trip signal c.
MSIV on unaffected steam generator fails to close on MSIS d.
Main steam paths downstream,of MSIV open or fail to close due to control system malfunction caused by adverse environment following large steam line break.
e.
Open main steam paths increase the steam bloudown and increase moderator cooldown effect which adds positive reac-tivity to core.
A post -trip return-to-power is more severe under these conditions.
PVDAORGNa E.
Steam Line Break With Atmospheric Dump Valve Control System Failure 1.
Significant Interaction a.
Post-acciden+ controlled cooldown 2.
Assumptions a.
Steam line break outside containment and upstream of I;51V b.
"tmosphe is dump valves on opposite stean line open and renain open*
c.
SAR conservatism i.
no operator action within 30 minutes 3.
Sequence of Events A steam line break outside 'of containment but upstream of the a.
MSIV occurs b.
Reactor trip on low steam generator pressure Atmospheric dump valves upstream of MSIV's opdn and remain open c.
due to control system failure
- The failure mechanism identified is a failure of the input signals that would cause the valve to open if operating in the automatic mode.
Although no cperator action is assumed for 30 minutes prompt operator action to shut the open val.e would mitigate any effects of this event.
bd 33
d.
If no operator action takes place there would be the potential for dry-out and depressurization of both steam generators Failure to shut atmospheric dump valves could inhibit a con-e.
trolled plant cooldown by limiting the ability of the auxiliary feed pumps to deliver to the steam generator (s)
P0B DML II.
Assessment of Impact of Control Systeu Failures on Feed Line Creal E.yn-.
S r.1 CEA Ejection A.
SSR Feed Line Break 1.
Sequence of Events i',ain feed line break occurs downstream of reverse flow check a.
valve, discharging main feed and steam generator fluid b.
RCS heatup dua to loss of subcooled feed ficw Reactor trip cccurs on steam generator low,>ater level or c.
high pressurizer pressure.
Turbine trip occurs on reactor trip d.
Rapid RCS heatup and pressurization due loss of heat transfer as the ruptured steam generator empties Depressurization of the ruptured steam generator initiates M515 and e.
isolates the intact generator f.
RCS pressurization terminates, with opening of primary relief /.
safety valves and decreasing core heat flux RCS cooldown begins, controlled by the main steam safety val.es g.
h.
Auxiliary feed is initiated automatically or by operator action ll)4 3t
B.
Feed Line Break With RCS Inventory Control Failure 1.
Significant Interaction Effect Increased RCS pressurization due to liquid filled pressurizer a.
2.
Assumptions a.
Small feed liae break inside containment b.
Adverse environment impacts pressurizer level instrument causing indication to fail low which causes the control system to increase inventory (and pressurizer level) c.
Initial conditions i.
102% power ii.
steam bypass control system in manual mode iii.
beginning-of-cycle core param2ters d.
Analysis conservatisms i.
no operator action for at least 30 minutes ii, no credit for stnam generator low water level trip in ruptu.ed unit until empty iii.
heat transfer in ruptured steam generator ;nstantaneouslj terminated on emptying iv.
failure of the feed line reverse flow chece v;? /e, if the break occurs upstream of the valve 3.
Sequence of Events a.
Feed line break in containment b.
Pain feed spillsfrom break Adverse containment environment causes pressurizer level indication c.
to fail low causing RCS ii.ventory to increase d.
Reactor trip occurs on oteam generator low water level on high pressurizer pressure.
Turbine trips on reacto.r trip RC'S heatup results from rapid decrease in SG heat tran sfer due e.
to loss of fluid from the ruptured steam generator f.
Pressurizer relief and/or safety valves open 1134 35
g.
Potential for pressurizer to fill with liquid exists due to high level in pressurizer prior to heatup.
Relief / safety valve relief capacity reduced by liquid discharge h.
Extent of increased RCS pressurization is deperl dent on tir.ie of pressurizer filling relative to the rapid heatup C.
Feed Line Break With PORV Control Failure 1.
Significant Interaction Effects a.
A failed open PORY in combination with a feed line break has not been analyzed 2.
Assumptions a.
Feed line break inside containment b.
PORV's inadvertently open and remain open c.
PCRV block valve also fails to close' when required d.
No operator action until 20 minutas 3.
PORV would not be expected to remain open due to actuation malfunction since Dresser valves and solenoids are qualified for 0
temperatures in excess or 400 F 4.
Sequence of Events a.
Feed line break occurs inside containment b.
Steam generator fluid and/or main feed spill from break RCS heatup and pressurizatidn results from loss of feed ficw c.
d.
FORV opens on high pressure and fails to reclose due tc adverse environment 1134 36
Reactor trip occurs on high pressurizer pressure.
on reactor trip f.
RCS depressurization occurs if PORV's fail to reclose 9
Mass removal via PORV causes void formation within RCS h.
Feed line break in combination with a failed open PORV has not been analyzed D.
Feed Line Break With Feedwater Control Failure 1.
Significant Interection Effects Overfilling of the steam generator (s) causing potential structural a.
problers 2.
Assumpticns a.
Small feed lire break inside containment b.
Feed control in automatic mode c.
Adverse envircnment causes steam generator level indication to fail Icw which causes the feed control syst; to increase feed flow above the steam flow d.
No operator action for 30 minutes i134 37
3.
Sequence of Events a.
A small feed line break occurs inside containment b.
Main feed spills from break Steam generator level instrument fails indicating low and causes c.
increased feed flow in excess of steam flow d.
(team generator begins to fill causing increased moisture content of steam If no operator action occurs undefined structural problems e.
could result f.
It should be emphasized that this event can be prevented by prompt operator action.
Safety grade level instrumentation exists to compare to control grade instruments.
The feed system can then be controlled manually 300ROR!GNAL E.
Feed line Break with Atmosphei :c Steam Dump Control Failure 1.
Significant Interaction Effects a.
Controlled plant cooldown 2.
Asstmptions Feed line break outside containment and downstreaa of reverse a.
flow check valve b.
Adverse environment impacts the atmospheric steam dump control on unaffected steam generator causing an uncontrolled steam release upstream of the MSIV's c.
!!o operator action until 30 minutes.*
j j } ;'j
}g
- ine failure mechanisn identified is a failure of the input signals that would cause the valve to open if operating in the automatic mode.
Although no cperator action is assumed for 30 min..tes, prompt operator action to shot the open valve $:ould mitigate any effects n tHs event.
3.
Sequence of Events a.
Feed line break occurs outside of containment downstream of check valve b.
Steam generator fluid and/or main feed spill from break Reactor trip occurs on steam generator low water level or high c.
pressurizer pressure.
Turbine trip occurs on reactor trip d.
Steam generator pressure increases following turbine trip Environment could cause atmospheric dump valves upstream of t.
MSIV in unaffected steam generator to open and remain open f.
If no operator actio'n takes place there would be a potential for dry out and depressurization of both steam generators 9
Depressurization of both steam generators may limit the ability of the auxiliary feed pumps to deliver to the steam generator (s).
F.
CEA Ejection '.-lith Failure of FWCS A feedwater control system malfunction during a CEA ejection cculd prc-duce effects similar to those described for the other events in this section.
S
III.
Potential Effect of Reactor Regulating System During High Energy Pipe 3reak Events A.
CEA position malfunctions due to steam and feedline breaks and CEA ejection 1.
Significant interaction effect:
a.
Potentially higher reactor power levels prior to reactor trip than presently analyzed 2.
Assumptions a.
Small high energy pipe break inside containment b.
Reactor regulating system in automatic mode c.
Adverse environment results in a low indicated power level from the ex-core sensor input to the Reactor Regulating System causing CEAs to be withdrawn 3.
Sequence of events a.
High energy pipe break inside containment of a small enough size where i mediate reactor trip does not occur b.
Control grade ex-core sensor indication fails lo.; Jue to adverse environmental impact c.
Reactor regulating system causes CEAs to be wi:r.dri..n d.
Reactor power exceeds the power previously assum.ed during t're transient e.
Reactor trip occurs due to high energy pipe break at conditicr.s not considered in present analyses UJ j
j
..m m
I l J 'r
.u
3.
Small Break LOCA With CEA Control System Malfunction.
1.
Significant interaction effects Potential exists for increasing power.
This would cause pressure a.
to remain above low pressurizer pressure trip for a longer period than previously assumed 2.
Assumptions a.
Small break LOCA inside containment b.
CEA control system in automatic mode Adverse environment impacts CEA control system or related sensors c.
resulting in consequential failure d.
Control system causes CEA to withdraw Standard LOCA licensing assumptions e.
?.
Sequence of events Small creak LOCA occurs inside containment a.
b.
CEA control system in automatic r' ode Adverse environment caused by rupture potentially causes excore c.
power indication to indicate low power level d.
Should CEAs begin to withdraw, the magnitude of the overpower excursion prior to scram would be increased.
This could produce a higher prinary system pressure which could then delay reactor trip and SIAS and result in higher peak clad temperatures 1134 11
ARKANSAS NUCLEAR ONE-UNIT 2 PLANT SPECIFIC INFORMATION The anal ys e s provided by Combus tion Engineering (Attachment 3) apply to ANO-2 except as noted in this Attachment.
Due to plant specific operating procedures and /o r equ i pme n t configura-tions we have been able to eliminate some items identified in A t t achme n t s 2 and 3.
I t en numbe rs in this Attachment match those in Attachment 3.
All interactions identified in Attach-ment 2 will be addressed.
Iten I.b - St eam Line Break Wi th PORV Control Sys t en Fa ilure There is no PORV installed on the ANO-2 pressurizer.
This precludes any interaction in this case.
ITEM I.c - S t e am Line Break wi th Feedwa t er Cont rol Syste,m Failure The ANO-2 Reac t or Pro t ec t ive Systen includes a high steam generator level trip and its accompanying alarm would alert the operator of a high stern generator level condition.
He could then look at his safety grade s t e am genera tor level indication and manually control feedwater.
Additionally, if steam generator level were allowed to increase beyond the point initiating a reactor trip, s t e rn pr e s su r e could be expected to decrease as the generator approached a filled condition.
The decreasing s team pressure would ini tiate safety grade main steam isolation actuation signal which a
s ecures no rmal feedwater to the steam generators.
The ANO-2 operators have traditionally been trained to react to feed-water c o r.t r ol sys t em mal funct i on such as this and, since warning is provided, could be expected to properly respond to this casualty.
I t en I.d - St eam Line Break Wi th Failure of Main Steam Paths Downstream oT Tp ?'s As stated in the assumptions li s t'ed in I t en I.D.
2 of Attachment 3,
this sequence of events is only a problem if it is asuumed that the Main Stemn Isol a t ion Val ve (MSIV) on the unaffected s team generatc r f ails to close.
The ANO-2 FSAR (Table 10.3-4) states that a failure of a MSIV to close is very improbable because:
(1)
Redundant solenoid valves are provided on the air supply and exhaust lines to the air cylinders on each of the MSIV's.
Redundant powe r supplies and 1134 A2
MSIS signals to these solenoid valves preclude the possibility of a si.,
e electrical failure resulting in the failure of the MSIV to close.
(2)
The MSIV's are designed such that they will close with the air cylinder only or the springs only.
The reliability of these valves precludes this postulated inter-action.
I t em I.e - S t e am Li ne Break Wi th Atmo spher i c Dump Va l ve Control System Failure As stated in the assumptions of Attachment 3 for this interaction, this sequence of events is only a problem if the atmospheric dump valves on the opposite steam line open and remain open.
In this event, a Main Steam Isolation Signal (MSIS) would be initiated.
The atmospheric dump valves up s t r e am o f t he MS IV's receive this signal (MSIS) and close.
These valves are "Q" valves and their actuating devices are redundant, safety grade.
The reliabili ty of these valves preclud-s this postulated inter-action.
Itan II. B - Feed Line Break With RCS Inventory Control
_ Failure The ANO-2 Pressurizer level indication system is safety grade.
This precludes this interaction.
I t en II. C - Feed Line Break With PORV Control Failure There is no PORV installed on the ANO-2 oressurizer.
This precludes any interaction in this ccse.
Item II. D - Feed Line Break With Feedwater Control Failure See response to item I.c.
Item II. 5 - Feed Line Break With Atmospheric Pump Control Failure See response as Item I.E.
Itea III. Potential Effect of Reactor Regulations System During High Energy Pige Break Events.
The ANO-2 reactor regulating systen does not have an automatic CEA withdrawal feature.
This precludes any possible interaction.
l l 3'l
'3