ML19208D801
| ML19208D801 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 08/16/1979 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| NUDOCS 7909290575 | |
| Download: ML19208D801 (20) | |
Text
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CENTRAL FILES
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PDR: liq ARKANSAS POWER & LIGHT COMPANY POST OFFICE BOX 551 LITTLE ROCK, ARKANSAS 72203 (501)371-4000 August 16, 1979 2-089-10 D
@M Mr. K. V. Seyfrit, Director a b Liu O Mice of Inspection & Enforcement o
U. S. Nuclear Regulatory Commission o
JU (D
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[ a Region IV J
611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011
Subject:
Arkansas Nuclear One-Unit 2 Docket No. 50-368 License No. NPF-6 Response to IE Bulletin 79-06B (File:
2-1510)
Gentlemen:
~
Attached is Revision 3 to our response to IE Bulletin 79-06B.
This revision supercedes the attachment to our letter of August 1,1979.
Very truly yours, b CwLY 0 V David C. Trimble, Manager Licensing Section DCT:JTE:pw cc: Mr. W. D. Johnson U.S. Nuclear Regulatory Commission P. O. Box 2090 Russellville, Arkansas 72801 Ignacia Villalva U.S. Nuclear Regulatory Commission Washington, D. C.
20555 Mail Stop:
Bethesda 242 1055 020 7*"
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59 Question:
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Review the description of circumstances describ::d in Enclosure 1 of
.5-IE Bulletin 79-05 and the preliminary chronology of the TMI-2
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^/28/79 accident included in Enclosure I to IE Bulletin 79-05A.
Ta5 a.
This review should be directed toward understanding: (1) the
-5 extrene seriousness and consequences of the s'multaneous MM blocking of both auxiliary feedwater trains at the Three liile 45 Island Unit 2 plant and other actions taken during the early Jg phases of the accident; (2) the apparent operational errors
- Mi which led to the eventual core damage; (3) that the potential 75 exists, under certain. accident or transient conditions, to iE have a water level in the pressurizer simultaneously with the T
reacter vessel not full of water; and (4) the necessity to 3
systematically analyze plant conditions and parameters and 3
take appropriate corrective action.
-?:5 3
b.
Operational personnel should be instructed to:
(1)notover-jj ride automatic action of engineered safety features unless T
continued operation of engineered safoty features will result
- ig in unsafe plant conditions (see Section 6e.); and (2) not make s
operaticnal decisions based solely on a single plant parameter F
indication when one or more confirmatory indications are
_3 ava il able.
=
c.
All licensed operators and plrnt c
.;cment and supervisors with operational responsibilities shall participate in this review and such participr'icn shall be documente:' in plant reccrds.
Response: ~
a We have reviewed, in detail, Enclosure 1 of IE Bulletin 79-05 and
~
the sequcnce of events included in Enclosure 1 to IE Bulletin g
79-0 5A.
We have worked closely with NRC, SD.', and' Met-Ed in gather-ing, analyzing, and dissea'nating all data we have been able to obtain to date to assure that our understanding of the incident is as accurate as possible.
+
Our reviews have been speci,"cally oriented toward those areas of 3
I concern addressed in IE Bulletin 79-06B and have included presenta-
=
tions to the operations staff by NRC, I&E Region IV personnel, and 1
NRC Operator Licensing Branch personnel, in addition to thorough 5
reviews and discussion by plant staff personnel.
All licensed
".i operators and plant management / supervisors with cperational respon-3 sibilities have participated in these reviews.
Documentation as to 4
their participation has been kept.
i 9
Additional operator instruction and guidance addressing the specific w
concerns of item 1.b of IE Eulletin 79-06B are being revised into plant prccedures as detailed in our response to item 6.
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2 Question:
2)
Review the actions required by your operating procedures for coping with transients and accidents, with particular attention to:
Recognition of the possibility of fonning voids in the pri-a.
r:nry coolant systen large enough to ccmpromise the core cooling capability, especially natural curculation capabili ty.
b.
Operation action required to prevent the fornation of such voids.
Operator action required to t. lhance core cooling in the eve.nt c.
such voids are fonred (e.g., rerm 4 venting).
Resconse:
2a) Tne Arkansas Nucler One -Unit 2 (ANO-2) emergency procedures will be r:odified as a result of the Three Mile Island - Unit 2 (IMI-2) incident, to recognize the possibility of fonning steam voids in the Reactor Coolant Systen ("CS).
The operator's need to rronitor the RCS parameters to detect condi tions at or near saturation wil d be eghasized.
Na cural circulation is established by tennination of forced flaw (tripping :the Reactor Coolant Ptnps).
Procedures shall recuire verification of natural' circulatiorr and heat ~ rejection to tne Steam Generator secondary by confinning that:
RCS Ih's stabilize af ter coastdcnn of RGs then tend to decrease.
RCS Delta T tends to decrease (Th and Tc to converge) wi th 2
decreasing decay heat load af ter RCP coastdown.
Continued denand exists for feed flow to the steam genera-tors to naintain steam generator level.
Continued denand exists for turbine bypass, atrrnsphere dtnps, or safety valves operation to limit secondary pressure.
Voids in the RCS may be recognized by:
1)
Oscillations in RCP amperage and P
2)
Incore thermocouples indicating superheat 3)
Occillations in nuclear instrurrent (due to reduced shielding as a result of the voids).
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fonnation of such stean voids by:
44 5.7 siE 1)
Req. iring the operator to check the reactor coolant 5;25 pressure and teuperature during recovery fr.cn a reactor M
trip and other transients in order to achieve and naintain 343:
subcooling of the reactor coolant in the hot and cold legs.
During " followup actions" the operators will take steps g.=g
~9 to nnintain at least 50F subcooling in the RCS.
J@#=-
2)
Providing for tennination of operation of Engineered Safety
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Features (ESF) systens only when the condi tions described jEjj in the response to Otestions 6.b.1 and 6.b.2 are net. This EE will aid in the prevention of stean void fonnation and in Mi stean vr 'd elimination should they be fonmd and ensure con-
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tinued core cooling.
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2c) The ANO-2 energency procedures wilI be nudified to ensure that the core is cooled in the event that such voids are fonned by:
Ni 1)
Providing for tennination of operation of the ESF Systens when autocatically actuated by low pressure conditions onlp when the condi tions described in the response to Otestions 6.b.1 and 6.b.2 are net. This will aid in both the prevention of stean void fonnation and in stean void elimination should they be fonned, thus, ensuring core cooling.
Je 2)
Providing for the continued operation of at least one reactor coolant pum per loop to assist in core cooling during accident r=
condi tions if, it_ is advantageous for those accident condi tions.
p 3)
Providing for operator action to open Pressurizer BI,S vents in
-....A case stean voids are fonned in the RCS and RG flow and heat
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rejection to the stean generators is inadequate, a:
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Question:
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Review the containment isolation initiat-esign and procedures, and prepare and imple:r.ent all changes nos ery to permit contain-e:1++
asM ment isolation thether manual or autenatie of all lines whose sgg isolation does not degrade needed safety features or cooling
.;;sjQ capability, upon auta.atic initiation of safety injection.
a+x+x
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Response
122 39 We have reviewed the Contairtnent Isolation Actuation System (CIAS)
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design and procedures and have listed below the automatically actuate f{
valves which erovide penetration isolation (valve numbers in paren-sai thesis):
25
=3 Ca teaory I
.h 1)
Chrical and Volume Control Systen. Letdown (2CV-4821-1 and gg 2CV-4823-2)
MME]
Ca tecory 11 aC 2)
Chilled Water Supply to Ccntainment Coolers (2CV-3852-1)
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3)
Chillti Water Supply fro: Contairmnt Cuclers (2CV-3850-2 g3,3 and 2CV-3851-1) 55 4)
Component Cooling Water to Reactor Coolant Pump Coolers (2CV-5236-))
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- 5). Component Cooling Water fron Reactor Coolant Pui p Coolers
' ~' (2CV-5254-2 and 2CV-5255-1)
=
Catecnry III 6)
Containment Vent Header (2CV-2400-2 and 2CV-2401-1) h 7)
Reactor Coolant System and Pressurizer Sample (2SV-5833-1 and 2SV-5843-2) 8)
l'itrogen Supply to Safety injection Tanks (2CV-6207-2)
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Quench Tank Liouid Saaple (2SV-5S78-1 and 2SV-5871-2) 10)
Safety Injection Tank Sample (2SV-5876-2) m.s 185 jl3 11)
Quench Tank Makeup Water Supply (2CV-4690-2) a+m 21 12)
Containment Sump Drain (2CV-2060-1 and 2CV-2061-2) mas 13)
Containment Purge Inlet (2CV-3289-1, 2CV-8284-2 and W
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Contairaent Purge Outlet (2CV-8291-1, 2CV-8286-2 and jg 2CV-8285-1) i=
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Low Pressure flitrogen Supply (2CV-6213-2)
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16)
P,eactor Drain Tank Drain (2CV-2202-1 cnd 2CV-2201-2) h55 Cateaory IV gg 17)
Reactor Coolant Pump Controlled Bleedof,f (2CV-4847-2 and 2CV-4846-1) 7
~;E 18)
Steam Generator Sample (2CV-5852-2 and 2CV-5859-2)
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Air Particulate Monitor in Hydrogen Purge System
==Ei (2SV-S231-2, 2SV-8273-1 and 2SV-6271-2)
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20)
Air Particulate Monitor in Containment Atmosphere Sample
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(2SV-8261-2, 2SV-8265-1 and 2SV-826- ?)
I 21)
Fire Water"$upply '(2CV-3200-2)
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Item 1 (Category I) above is isolated upon receipt of a Safety 75E Injection Actuation Sicnal (SIAS) Een Reactor Coola or a Containment Isolation Actuatic 1
Signal (CIAS).
SIAS is generated JF ea pressur is less than or equal to 1740 psia or then Containment Building press Xi is greater than or equal to 18.4 psTE.
CIAS is generated tchen Contai ment Building pressure is greater than or equal to 18.4 psia.
ME Items 2 through 21 isolate upon receipt of a CIAS.
The valves
=
noted in Items 2 through 5 (Category II) are normally open during
==W 9h power operation since they are in systens which provide support to needed systems within-the Contain;;.ent Building.
The valves noted i
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i Items 6 through 16 (Categcry III) are nonnally closed during power cperation and are only opcacd periodically by specific manual operatio
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i i.e. there is no automa tic opening of any of these valves.
The valves noted in Items 17 through 21 (Category IV) are mrmally open during
+=s power operation but are not necessary to be open following receipt of as a SIAS.
Eih Since Items in Category II are providing support to systems within the Containment Building, the valves should stay open upon receipt 3gg of-a SIAS to prevent unnecessary equipment damage.
The systems e;
represented in Category II contribute to a " normal", orderly cool-M down following rerMpt of a SIAS.
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im Items i, Category III are normally closed during power operation a
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and specific manual operation is required to open them.
Further-
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more, to cause. full opening of the penetration, specific manual operation of at least two valves is required, each of which reo,uires dij=5 a specific and deliberate action.
Based on this fact, no changes to
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the Containment Building Isolation System are needed.
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RM items in Ca' tegory IV are nonnally open during power operation
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and specific manual operation is reovired to close these valves following receipt of a SIAS.
Each of the Category IV systems
+==i-MM were reviewed and it has been verified that a direct connection bet.:een the Containment Building atmosphere and the Auxiliary ij2]
Building atmosphere or the environmant does not exist while these
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penetrations are open.
Based on this fact, no changes to the
- l Containment Building Isolation System are needed.
45 mes However, to further increase the margin of safety, a iesign
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sij change is being evaluated for items in Category III and IV to add a SIAS to those valves.
This design change will provide an additional degree of assurance that no release path to the
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environs exists upon receipt of a SIAS without a concurrent CIAS.
.6is 1;M These modifications resulting from this evaluatiw. will be imple-mented during the first available plant outage to cold shutdown
===E of sufficient curation to accumodate the edification following j@]
canpletion.of the design change package, bu'. no later than the
=23 tirst retueling outage.
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17 the interim, procedures will be implemented, by August 24, 1979, 4R requiring verification that the above Category III valves are closed gig upon a SIAS and. manual closing of the above Category IV valves upon
~=t2 a SIAS.
These procedures will remain in effect until implementa-3 tion of the above design change.
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Question:
ma=
4)
For facilities for which the auxiliary feedwater system is not
== 2 automatically initiated, prepare and implement immediately proce-
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dures which require the stationing of an individual (with no other Mn assigned concurrent duties and in direct and continuous comnunica-GE; tion with the control room) to prcmptly initiate adequate auxiliary
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feedwater to the steam generator (s) for those tr.ansients or acci-lMEEi dents, the consequences of which can be limited by such action.
c==.=.
g1=3 Resconse:
is This questiun is not applicable to Arkansas Nuclear One-Unit 2 l.pj}M (AND-2) as the Energency Feedwater System (Eftl) is designed to gg; automatically initiate as part of the Engineered Safety Featurt.
=ma Actuation System (ESFAS).
The system is designed to the latest pg33 revision (Rev.1) of Branch Technical Position ASS 10-1.
5g JE Those Design Basis Events which will cause automatic Emergency
_.;gg Feedwater actuation are:
56M Steamline Break (Inside Containment)
Steamline Break (Outside Containment)
?f!
Loss of Main Feedwater 7557 The variables which are monitored to indicate the above events are:
Steam Generator Pressure
. Steam Generator Level
- 1 Further-informaticrr on the AND-2 EFW System is presented in the ANO-2 Final Safety Analysis Report (FSAR) Sections 7 1.1.11.8. P nd 10.4.9, and in Volume IX, Respanse to NRC Qtu t ons 020 'G, 020.54,
- s i
222.22, and 222.90.
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_ucstion:
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5)
For your facilities, prepare and implement imediately procedures 2E; which:
ET?
Ujg a.
Identify those plant indications (such n valve discharge T
piping temperatur~e, valve position indiction, or valve discharge f3 relief tank temperature or pressure indication) which pla'nt operators may utilize to determine that pres.surizer power j:]
operated relief valve (s) are open; and EE M
b.
Direct the plant vperators to manually close the power operated
- g=
relief block valve (s) when reactor coolant system pressure is reduced to below the set point for normal automatic closure of
==
t=3 the power operated relief valve (s) and the valve (s) remain g
stuck open.
3 F.escans e:
- x 3-Question 5 is not applicable to Arkansas Nuclear One-Unit 2 (ANO-2) as the AND-2 design does not include power operated relief valves
=-3 on the pressuriser.
Overpressurization of the. Reactor Coolant System is precluded by r.sans of safety valves and the Reactor
= = ~
Protective System (RPS).
LX 7
inforc? tion on the safety valve' is presented in Sections 5.2.2 and 5.5.10 and Chapter 5A of the AND-2 Final Safety Analysis Report MS^hSyrmatin n tne Rn is pnsentce in section 7.2 of the
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6)
P.eview the action directed by the operating procedures and training instructions to ensure that:
Operators do not override autonatic actions of engineered a.
safety features, unless continued operation of engineered safety features will result in unsafe plant conditions.
For
~
example, if continued operation of engineered safety featpres would threaten reactor vessel integrity then the HPI should be secured (as noted in b(2) below).
b.
Operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been autanatically actuated because of low pressure condition, it must re ain in operation until either:
(1) Both low pressure injection (LPI) pumps are in operation and flowing for 20 minutes or longer; at a rate which would assure stable plant behavior; or (2)
The HPI system has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the_ saturation temperature for the existing RC5' pressure.
If 50 degrees subcooling cannot be maintained after HP1 cutoff, the HPI shall be reactivated. 'The degree of subcooling beyond 50 degrees F end the length of time HP1 is in operation shall be limited by the pressure /
tenperature considerations for the vessel integrity.
Operating procedures currently, or are revised to, specify c.
that in the event of HPI initiation with reactor coolant pumps (RCP) operating, at least. one RCP shall remin operating in each loop as long as the pump (s) is providing forced flow.
d.
Operators are provided additional information and instructions to not rely upon pressurizer level indication alone, but to also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water, inventory in the reactor primary system.
Resoonse:
6a) A caution note will be added to the ANO-2 emergency procedures instructing the operators not to override the automatic actions of the Engineered Safety Features (ESF) without first determining the consequences of that override and consulting with the shift supervi sor.
The procedures will also be modified to add clarifying steps to aid operators in recognizing a spurious actuation and provid'e for orderly ten.,ination of the sparious actuation.
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-9 Gb)
The AH0-2 cmcrgency procedures will be modified to specify ggi the following actions.
If the High Pressure Safety Injection J._=g (HPSI) systen has been auto..atically actuated because of a low gg pressure condition it must renain in operation until:
ess g
1)
Low Pressure Safety Injection (LPSI) is in progress with a flow rate in excess of 2000 gp;a and the situation has been
==
y=a stable for 20 minutes; or,
==
M 2)
The HPSI system has been in operation for 20 minutes, and i=
all hot and cold leg temperatures are at least 50F below
=GE the satsration temperature for the existing RCS pressure, igg If 50F degrees subcooling cannot be maintained after HPSI T;;;
cutoff, the HPSI shall be reactivated; or, W5
" LEE 3)
The RCS pressure returns to nor:aal operating pressure with 45 the temperature in the hot and cold legs being controlled gg with at least 50F subcooling by an operable steam gen-ge erator; or,
- - -i!H iEE 4)
Unless continued op2 ration would result in an unsafe
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plant condition.
2 6c)
The AMO-2 " Loss of Coolant /RC Pressure" emergency procedure will be i.?
revised to specify:
MM
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For break sizes exceeding HPSI Capacity:
2 If HPSI initiatien automatically occurs because m;r
.of low Reactor Coolant pressure, anc' the reactor 2
coolent pr.'ps are in operation, then at least one 7=~i RCP/ loop will renain-in operation until Lou 3g Pressure Safety Injection flow is established
}
and verified.
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For break sizes within HPSI Capacity:
=
If HPSI initiation automatically occurs because of Ji low Reactor Coolant pressure and the reactor j;
coolant pumps ere in operation, then at least ai one RCP/ loop will remain in operation until j
LPSI injection or decay heat is established and gg verified or continued operati'on of the RCPs would create an unsafe plant condition.
=g
--iE5E 6d) AHO-2 plant procedures are being revised to require the operators 7
to monitor RCS pressures and temperatures following transients 5M to assure that adequate cargin to saturation conditions is T5 maintained.
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'~lo fessen the adiguity of this parameter (margin to saturation) 2jff
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special scales for RCS temperature and pressure indicators and recorders are being constructed which correlate saturation
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pressures and temperatures (reduced by SOF) to the current 75; indicated temperature and pressure scales.
With these special fil.
scales, the operator can rapidly assess core conditions relative
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..w. 3)! Questinn: C 7) Review all safety-related valve positions, positioning requirenents W: and positive controls to assure that valves re.nain positioned (open sj or closed) in a conner to ensure the proper operation of engineered
- g safety features.. Also review related procedures, such as those for Fi maintenance, testing, plant and system startup, and supervi,sory.
E periodic (e.g., daily / shift checks,) surveillance to ensure that jg such valves are returned to their correct positions following 5 necessary manipulations ar.d are maintained in their proper posi- }} tions during,all operational modes. Mg Resconse: am M We have cenpleted a review of the Eng*,neered Safety Feature (ESF) valves and their positioning requirements. The ESF systems are: y a) Contairment Isolation System (CIS) ~ hiy b) Containment Spray System (CSS) } c) Containment Cooling System (CCS) d) Safety Injection Systca (SIS) e) Penetration Roor. Ventilation System (PRVS) )$ f) Ibin Steam Isolation Sys te:a (l' SIS) g)~ Emergency Fee 6ater Syster (EFS) =; U h) Chemical and Volur.e Control System (CVCS) (c,') i) Diesel Fuel Oil and Starting Air System j) Emergency Boration Systcas rg [{@ (y k) Service Water b--] O Based on this review, and our review of related procedures, we have Og concluded that our procedures are adequate to ensure that valves 'in =, Ig ESF systems are uaintained in their, proper position, or are capable O; of being properly positioned in the event of an Engineered Safety Feature Actuation Signal (ESFAS). 5 7= The procedures reviewed are summarized as follows Maintenance - Pr'ior to taking an ESF systen out of service, the Control Roon must be notified as i equired by procedure. The 73 r.edundant train of the affected ESF system will be inspected to verify operability prior to taking the aforementioned system out of service. The inspection uill include checking, control board indi-cations, liOV status, alann status, and verification that the last 1055 0.32
== surveillance test was within the surveillance interval and demon-s trated operability. The out-of-service syston includes couponents for which maintenance is to be performed as well as the valve (s) used, to isolate the couponent for maintenance. Tags are placed on the affected out-of-service equipment, both at the equipment proper and at I4 tor Control Center (MCC) breakers, if applicable. I.djitionally, the out-of-service equipment is entered into the station log. Following coupletion of maintenance, and renoval of out-of-service tags, the systen is re-aligned to its proper configuration by the operator, the Control Roca is notified of systen return to service, and entry is made in the station log. Surveillance tests are perfonned to verify the operability of the affected equipment. Testino - All ESF systems are required by ASME Section XI and/or /.im-% Technical Specifications to be tested to ensure operability. Test frequencies very according to the conpogent being tested, and the reesca for testing. Upon cc7:pletion of ESF systen testing, for t;hatever reason, the subject system is verified as required by procedure to be properly aligned to allow the systea to perform i ts safety funct. ion _ The.varification of lineup is done-by the operator using sign-offs in the procedure. During our revicu, all manually operated valves were found to be procedurally required to be in their correct position. The pro-ccdu es further reluire the system lineups to be verified correct price to declaring the systen operable. However, several of these valves in systeas not classified as AS",E Codes 1, 2, or 3 were not subject to the " Category E" listing (i.e., required to be locked, sealed ~or otherwise secured in their pecper position). Thesa. val ves, in the !)iesel Fuel Oil Syste:a and Diesel Starting Air Sysicm, though not classified as Class 1, 2, or 3 will be added to the 'lategory E" prondare list and 7.s suc ae required to be and will be locked, sealed, or other.a se secui ed in their proper position during operation, ti !s further assuring proper valve positioning of all safety-related valves in their associated system. These procedural changes will be implemented by June 1,1979. Thus, based on procedural controls and this review, we feel assured that all safety-rested valves are positioned in, or are capable of being positiened in thei-ESF position upon receipt of an ESFAS, thereby ensuring.the required re.sponse of systens to postulated events. Startup - All safety-related systens are required to be operable (valves in the correct position) prior to and/or during plant start-up as appropriate. L bd] [Ii W ud) lfU l l, PElb 1055 033
g _Qtto_s_ tion:
- NiN 8)
P,cview your operating modes and procedures for all systems designed ~
- .4'l to transfer potentially radioactive gases and liquids out of'the 5=2 primary containment to assure that undesired pumping, venting or W;g other" release of radioactive liquids and gases will not occur inadvertently.
N 3M In particular, ensure that such an occurrence would not be' caused gg. by the resetting of engineered safety features instrucentation. 55 List all such systems and indicate: N: h" nether interlocks exist to prevent transfer when high a. "M radiation indication exists, and Mni iM b. Whether such systems are isolated by the containment isolation (Q s ig nal. 95E 5 The basis on W1ich continued operability of the above features c. M is assured. 33 _ Response: ...M M The systems designed to transfer potentfally-radioactive g ses cnd qq ~ liquids out of the Contairmant Building at A:0-2 do nots autow.ticall; 3 discharge under any conditior.s. The systems that require spicific M canval 07eration are as follows (valve numbers in parenthesis): 1) Chr.iical cnd Volume Control Sy: tem Letdein (2CV-4821-1 and 2CV-4023-2) 5 2) 'Cor.tainment Vent !!?adce (2CV-24CO-? and 2CV-2401-1) 3) Reactor Coolant Syste., and Pressuri.2:r Sample (2SV-5833-1 and s 2SV-5843-2 ) O, Q 2 4) Quench Tank Liquid Sample (2SV-587S-1 and 2SV-5871-2) rg& LD (g. c =y (q 5) Safety Injection Tank Scmple (2SV-5876-2) bd Q 6) Contair. ment Sump Drain (2CV-2060-1 and 2CV-2051-2)
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- 7) ~ Containment Purge Inlet (2CV-8289-1, 2CV-8284-2 and
...g b 2CV-8283-1) - m= JEk 8) Containment Purge Outlet (2CV-8291-1, 2CV-82SS-2 and 75 2CV-8285-1) gg-eg 9) P.eactor Drain Tank Drain (2CV-2202-1 and 2CV-2201-2) e= ? 5.5 10) Peactor Coolant Pu.np Controlled 81eedof f (2CV-4847-2 and 5 ~ 2CV-4846-1) En ~5 11) Hydrogen Purge Systc.n and Air Particulate Monitor R= (2SV-8231-2, 2SV-0273-1 and 2SV-0271-2) a, 3 e 1055 034
-, = t55 3 12) Contoinment Atanphere Sraple and Air Particulate lionitor g (2SV-8261-2, 2SV-8265-1 and 2SV-8263-2) w.g All of the above syste.ns receive a CIAS to close. None of the L-5+ systens will reopen following resetting of the CIAS without 7== specific operation. C WEE Iten 1 receives a SIAS as well as a CIAS to close. RR Items 2 through 9 are nomally closed during power operation and M require specific manual operation to open. They receive a CIAS M755 to close and vill not open following resetting of the CIAS with-M out specific manual operation. 5ME M Itcas 10 through 12 are nomally open during power operation. They jb receive a CIAS to close and rtill not open follo ing resetting of the TE CIAS without specific manual operation. .Mi !=8 Items 2 (Containment Vent Header) and 8 (Containment Purge Outlet) have a high radiation release interloc.' to close the systeas auto-s catically. Iteas 11.(Hydrogen Purge Systea and Air Perticulate. l'anitor) and 12 (Contair.nent Atcosphere Sempic and Ai = o "ticulate c l'onitor) have radiatio 1 alanas in the system which will annunciate in the Control P. con to alert operator of a high radiction release. All of the above listed valves are periodically surveillance tested as required by the A::0-2 Technical Specifications per Section 4.0.5 (AS:12 Section XI testinal anc Section 3/4.6.1 (Appendix "J" to a 100FP.50 testing). The CIAS is verified opreble per Section 3/4.3.2 of the Technicel Specifications. The radiction aonitoring instru-
== nentaticr. is verified cp.t.?ble per Section 3/4/.3.3 of the ff!O-2 Technical Specifications. B. sed on the above, the inadvertent release of radioactive 2:w: or lipids by automatic netns following resetting of the CIAS is not pn,io'e. T :cific n3nual oparation of the systems would be re-9 .! ta open the contcinment penetrations. Furthercore, automatic for discharging radioactive gsses or liquids does not occur i
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p.-ior to initiation of CIAS. Transfer of contaminated fluids requires specific manual operation. l_k 3=m
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Question: 9) Review and modify as necessary your maintenance and test procedures to ensure that they require: Verification, by test or inspection, of the operability.of a. redundant safety-related systems prior to the removal of any sa fety-rela ted syste:n frr.n service. b. Verifipition of the operability of all safety-related systems when they are returned to service following maintenance or testing. Explicit notification of involved reactor operational personoc1 c. whenever a safety-related syste:n is removed from and returned to service. IIes ponse: Prior to initiating raintenance on safety-related systems, the re-dundant systen will be inspected to verify operability. The inspection will include checking control bmd indications,. MOV status, alam status, and verificatici that tha last surveillance test denonstrated operability and was witnin the sur.reillance interval. Plant Quality Control procedures recuire that a Jeb Order be issued for any r.ain-tenance of s:fety-rela ted ("Q") syste,s. All Job Orders require authorization by the effected unit's Shift Supervisor prior to work C or.. enc i ng. The Joti Order fom is currently toing ravised to spccifically iden-tify all Pre-mintenance and Post-caintenance requirecents, and to include verification that those rquirenents are cet pricr to de-clcring a systen or ccmponent CPE:".SLE after uaintenance. The revised Job Order foru will b2 developed and inple.aented before the AUD-2 core is made critical following our current outage. 1,' hen perfoming testing on a safety-rela ted system, A;'0-2 Tech Specs address operability of the redundnt system. Docenntation is re-quired, by a testing precedure and/or Job Order, b3 specific check-offs and signoffs that a " r ty-related system is returned to its e proper operable condition following testing of that system. All safety systems taken out of servicc are noted in the Station Log and noted on the safety system status board. Current oper-ating Procedures require operators, coming on shift, to read all log entries back to their previous shift or for the previous.7 days whichever is shorter and each shift is required to review the. station log, plant annunciators, system status board, and equipment tag out book. /_ToJJA To yum Q D ? Am d6 1055 036
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- 10) Review your prenpt reporting procedures for NRC noti.fication to
... = = =5.2 assure that NRC is notified within one hour of the the the reactor '~?E! is not in a. controlled or expected condi tion of operation. Further .5M at that tiue an open continuous comunication channel shall be .33 established and unintained wi th NRC. osg!. K Resoonse: =us It has been and will continue to be AP&L's policy to prc=ptly notify a=. M the Nuclear Regulatory Ccranission of any unusual event at Arkansas gg Nuclear Che. This policy is not limited to itens vhich are deened = =[ reportable per the Technical Specifications,or federal regulations.
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- 4E In the event of ar. energency situation, our procedure for imple-2 nunting the Fmergency Plan require early notification of the Naclear 75 Regulatory-Camission.
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To further clarify the above policy and procedure, v.e wil.1 nodify g the AT Aininistrative Controls Manual and Energency Procedure 1202.34 for Personnel Response, to include the following state-nunt. "Upon notification by the Shif t Supervisor of an event at APO, the Duty Evrgency Coordinator will assess the situation as, its-seriousness _ If the ~E assessannt indicates that the health and safety of the oublic might be endangered or there might be a potential for significant public interest (e.g., ~ radioactivi ty release, etc.), AP&L Management and NRC shall be irrmediately notified regardless of the reportability of the event as defined in the Techni-cal Specification or federal regulations." s Notification of NRC within one hour that such a condition exists, l + will be via the " hot line" phone which was recently installed by NRC. .=.. ?
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- 11) Review operating modes and procedures to deal with significant a:anunts of hydrogen gas that may be generated durimi a transient
/Mi .# accident that would either reaain inside the primary Jgt (,ts.. - be released to the containment. .p;.EJ' y;..:g !'es ponse: =ze (;y.(.h] Current Afir.2 caergency procedures (LOCA procedures) were reviewed gg crncernir.9 H concentration control in contain:..ent and were found 2 giM + o provide suf ficient guidance and procedural control to minimize =3 che potential danger of acc"aulation of significant quantities of M In addition, the Ai!0-2 Contairment Building is equipped with =Mi H7 Recombiners. Further discussion of this system is presented g i6 Section 6.2, of the Ai!0-2 Final Safety Analyses Report (FSAR). w-MS In the event of 11 accumulation in the R'ecctor Coolant System (RCS) we balieve cdditi$nal naasures pro /ided in our response to Question "M 2 t.bove are sufficient to preclude and/or control such an event. Ib + ever, the fo11 suing methods could be used to renove large nonconden- .] sible gas vc W;. M 1) Dissolve and/or suspend noncendensible gas by use of forced + flou (P.nctor Coolant Pumps) cnd/or adjustcents to P.CS tem- =i perature and pressure; 2) DeSas the pressuri;:er via the steam space sample line; 3)' Dep.s via le tdown through the vacuum degassifier T~ t (hot leg) semple lines to the + 4) Dzgas via the ho Volt.s Control Tank. = a v. . =;:: _ _.. 5. .- 2 42 _ _.b 7.=;g 5
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4. Egg Question: JE5 12) Proposed change:;, as required, to those technical specifications 145 which must be modified as a result of your implementing the above items.
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Response
si= g.gg Ho changes to the AM0-2 Technical Specifica'tions ware deaiied necessa 72; as a resclt of these responses s s 2.;;;;; v==
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