ML19224C233
| ML19224C233 | |
| Person / Time | |
|---|---|
| Site: | Crane, Farley  |
| Issue date: | 04/24/1979 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| References | |
| NUDOCS 7906290483 | |
| Download: ML19224C233 (21) | |
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. -.,a Lh'MS"2 e ;,.us a S : 15 Alabama Power re ::u:. yn e x::c :,=em April 24,1979 NRC II 3ulletins Nos.
79-06 A & 06 A Rev.1 Mr. Ja:ns ?. O'Reilly U. S. Nuclear Regulatory Cc==ission Region II 101 Mariet ta S tree t, N. W.
.7 Suite 3100 Atiar u, Georgia 30303
Dear Mr. O'Reilly:
Alab. a ?cuer Conpany has prepared the enclosed response to ite s
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l th: ugh 12 of the referenced NRC II Sulletins for the Joseph M.
Farle-Nuclear Plant.
Iten 13 of the bulletins requested that proposed dcsign changes and Technical Specifications be sub=itted within thirty (30) days of receipt of the Sulletins.
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Proposed changes to the Joseph M. Farley ?lant and Technical Speci-
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fications vill be sub=itted to NRC Nuclear Reactor Regulation by (w_
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May 14, 19 79.
If you have any questions, please advise.
Yours very truly,
.:N
- o. oL. Clayten, Jr. 4 -
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t FI.C/0DK/==b Enclos ure
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Mr. R. A. Tho=as Mr. G. F. TrowbrEdge fg._
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5W A?CO Response to 70-06A
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11etins 79-05,79-05A, 79-C6,79-05A and 79-06A, Rev. I have been
.ev' awed by APCo with input fran 'Jestinghouse, 3echt21 and Southern
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Cc: any Services Incorporated.
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A training program for all licensed operators and plant anagement and supervisors with aperational responsibilities is being developed.
This training program will include, but will not be li=1ted to the follcwing
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items:
p.
E.
a Review of the description of the circunstances described in p.
enclosure 1 of IE 3ulletin 79-05.
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b.
Preliminary chronology of the TMI-2, 3/29/79, accident in-
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cluded in enclosure 1 cc IE Sulletin 79-05A.
Any procedural (Administrative or Operating) changes that
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c.
are cade as a result of the above review.
d.
Any design changes that are made as a result of the above review.
A re-ecphasis of existing procedures (Administrative or e.
Operating) which are related to the TMI-2 accident or applicable IE 3ulletins.
This training program will be directed toward understanding:
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a.
The extre=e seriousness and consequences of the s1=ultaneous blocking of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident.
. _ 7; b.
The apparent eperattonal errors which led to the eventual core da= age.
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That the potential exists, under certain accident or transient conditions, to F,ve a water level in the pressurizer simultaneously with the reactor vessel not
-" 3-full of water.
d.
The necessity to syste=atically analyze plant conditions and rx..._
parameters and take appropriate corrective action.
}5}{n This training program will instruct operating personnel to:
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a.
Not override auto =atic action of engineered safety features
,--.T unless continued operation of engineered safety features will result in unsafe plant conditions as described by pro-cedures or the actuation is known to be spuriously initiated, and
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APCO Response to 79-06A b.
not make operational decisions based solely on a single 2-1.
pl int parameter indication when one or more confirmatory 1- :ications is available.
This training prrgram will be c acpleted prior to the plant return-ing to power operation.
(The plant is presently in " ode 5 for re-fueling.)
Participation in the program will be documented in plant training records.
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APC0 Response to 79-06A
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2.
The actions required by our operating procedures for coping with trar.sients and accidents have been reviewed with particul." attention
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i-T. t a.
Recognition of the possibility of for=ing 'coids in the pri-r '- ~
=a ry coolant syste= large enough to compromise the core -
cooling capability, especially natural circulation capability.
b.
Operator action required to prevent the for=ation of such voids.
c.
Operator action required to enhance core cooling in the event such voids are fo rmed.
The pri=ary indication of '<oid for=ation in the pri=ary coolant syste=
is when the pressurizer pressure falls below the Hot Leg saturation pressure. Mcwever, during a loss of coolant event, void for=ation in the pri=ary coolant syste= would be expected with two exceptions:
1) the loss of coolant is being caused by a stack open pressurizer relief or saf:ty valve which closes er is isalated before the syste= depressurizes to Hot Leg saturation, or 2) the reactor coolant syste= reaches an equilibriu= pressure above Hot Leg saturation, when the High Head Safety Inj ection (HHSI) flow equals the break flow.
5=
In these two specific cases, confir=ation of no voids in the syste=
will be apparent by the pressure in the pressurizer when co= pared against Hot Leg saturation pressure.
In the re=aining cases the engineered safeguard systens have been designed to cope with voiding.
Thus, it is not necessary to be able to recognize void for=ation in these cases.
As a result of this review, the following actions will be taker prior
~ -~L to returning to power operation:
a.
E=ergeacy Operating Procedure 1 - Loss of Reactor Coolant will be revised to incorporate additional procedural guidance for the z_._._
operator for those accidents involving a loss of coolant through
_ _ _ cr a stuck open pressurizer relief-or safety valve which subsequently m;.:gy reseats or is isolated before Hot Leg saturation conditions occur or through a s=all pipe break in which HHSI flow and break flow equilibrate at a pressure above that corresponding to Hot Leg saturation conditions.
This is to ensure that for those accidents where voiding can be crevented,it will.
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APC0 Response to 79-05A
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E=ergency Operating Procedures will be revised to prevent the formation of voids or to enhance cooling in the event T_:
voids are for:ed are described in response to Iten 6, 7 and
{-52 2-12.
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Revised Emergency Operation Procedures as described above adequately desc:ibe the necessary operator actious to enhance core cooling if the primary coolant systen has been voided due to either a loss of coolant, loss of secondary ecolant, or steam generator tube rupture acc id ent.
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APCO Response to 79-06A MW 3.
Lcw pressuri:er level histables will be placed in the tripped condition 7* ;-
when these bistables are required to be operable by technical specifi-
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cations. Lcw pressuri:er levei bistables will be returned to their r
nornal operating positions during applicable pressuri:er pressure channel
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surveillance tests.
Energency Operating Procedures will be changed to require the operator to nanually initiate Safety Injection upon low
--r pressurizer pressure if not autenatically initiated.
A design change is being processed to delete the autenatic initiation of Safety Injection on coincident Lew Pressurizer Pressure and Level and to add :he autenatic initiation of Saf aty Injection on 2/3 Low
-zz Pressurizer Fressure signals.
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4 APCO Response to 79-06A
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The containcent isolation system is designed to minicize post accident
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laahage f rom containment penetrations in syste=s whi are not required
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to operate after an accident.
Each of these contain=en: penetrations 2:1.
is provided with a double barrier so that no single failu.e or mal-function results in a loss of isolation or excessive leakage.
The contain=ent isolation consists of Phase A and Phase 3 isolation.
Phase A isolates all non-essential process lines, but does not affect sa f ety inj ec tion contain=ent spray, component cooling, steas and feed-water lines.
Phase A isolation is initiated by (automatic or =anual) safety injection initiation and also can be initiated manually.
Phase cr 3 isolat, all re=aining process lines (except safety injection, con-tain=ent spray lines and auxiliary feedwater) and is initia ed by auto-catic or canual initiation,of containment spray.
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Contain=ent isolation is such that it is not reset by the elimination or resetting of the initiating signals, for example by resetting safety injection.
Containment Isolation can only be reset by =anual controls on the main control board.
Verification of contain;ent isolation upon autematic or =anual initiation of safety injectica is required by E:ergency Operating Procedures.
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APC0 Response to 79-06A T c1osure 1
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5.
The Farley Auxiliary Feedwater Syste:t is cc= prised of two =otor
- rre driven and one turbine driven auxiliary feedwater pu=ps, all o'
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which are autocatically started and sealed in upon receipt of their
. ;#ii actuation signals.
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A?C0 Response to 79-06A 5.
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6.
Approved plant operating irocedures list parameters to be conitored 2:1 21-.
which indicate that a pres.turizer pcwer operated relief is open or
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leaking. These procedures equired isolation of the relief. To
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consolidate these parametete and to restate required :peracor action Energency Operating Procedure (EOP-1) Loss oi; Reactor Coolant is being cevised to fully identify those plant indications (which include:
valve discharge piping temperature, valve position indication, and pressuriser relief tank temperature, pressure and level indication) which Plant Operators may utill:e to deterrine that a prersuri:er power operated relief valve is open or leaking.
This procedure will direct the Plant Cperator to manually close the motor operated isolation valves for the power operated relief valves when reactor coolant system pressure is reduced belcw the pcwer operated relief valve actuation setpoint for nor=al auto =a, tic closure and the valve (s) remains open.
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APC0 Respcnse to 79 _
Inclosure 1
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The Farley Nuclear Plant E=ergency Operating Procefures have
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been reviewed and modificationt will be =ade to = ore clearly g ~.~
guide the operator in terminating safety injection.
This will avoid potentially adverse plant conditions under certain cases of continued operation of the safety inj ectivu system identified below.
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In the case of LCCA's where pressurizer level can be increased by HHSI or secondary side line breaks which lead to pri=ary syste= heatup transients, continued operation of the HHSI pu=ps result in lif ting of the pressurizer power operated relief valves
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and/or safety valves.
Specific guidance has been developed to
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terminate safety injection prior to this occurrence, with the plant in a stable condition.
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Similar criteria and guidance will be incorporated for ter=ination
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of safety injection following secondary side breaks whi_n lead to a pri=ary syste= cooldown.
In this case continued operation of HHSI pu=ps would lead to conditions which potenti ally could exceed i
reactor vessel pressure /te=perature criteria.
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Following a steam generator tube rupture, criteria and guidance have been prepared for termination of safety injection to reduce the quantity of primary coolant which passes to the secondary side of the steam generator.
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In these cases specified abcve, the criteria and guidance for
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terminating Safety Inj ection are based on the plant being in a
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stable, controlled subcooled condition prior to safety inj ection t e r=1na tion.
Should a subsequent system disturbance occu, speci-fic criteria for reinitiation of safety injection will be provided.
The criteria to allow termination of HHSI specify conditions which
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result in at least 500F subcooling, though no specific subcooling criterion is included.
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We have chosen parameter values consistent with a highly subecoled
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state in the pri=ary system.
For exa=cle, following a s=all break
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LOCA, the criteria for terminating high head safety inj ection flow
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are:
1)
Wide range RCS pressure >2000 psi, and
- 2) Wide range RCS pressure increasing, and
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Adequate stable level in at least 1 steam generator
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to assure a sufficient heat sink, and e
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AFC0 Response to 79-06A
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Pressurizer level >50%.
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These provide for a subcooled pri=ary side, stable pri=ary side conditions, -- x and a heat sink via the steam generator.
Additional clarification with respect to conditions under which c.
the reactor coolant pu=ps are to be =anually tripped will be provided in E=ergency Operating Procedures.
These conditions will include verification that HMSI pu=ps are operatings and that the reactor coolant syste= pressure is decreasing and is below 1550 psig which is less than safety injection actuation.
In addition structions are provided to direct the operator to
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trip the pu=ps on isolation of co=ponent cooling water to the pu=ps resulting from Phase 3 isolation.
d.
The E=ergency Operating Procedures will be revised to direct the operat.c to cons. der the follcwing para =eters as the bases for operator action:
Wide range RCS te=perature and pressure Steam pressure Steam generator water level Contain=ert pressure RWST water level Auxiligry Feedwater Flow Pressuri:er water level Accu =ulator Level
~~"
?JiR Pu=p Flow Boron Inj ection Tank Flaw Contain=ent Spray Flew Spray Additive Tank Outlet Flow The operating personnel will be instructed in the necessity to =onitor all the para =eters listed above and do not solely rely upon one para-
=eter to facilitate the early detection of co= pound accidents or engineered safeguards syste=s =alfunction(s).
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O APLO Response to 79-06A
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Valve positions for safety related syste=s are specified in the FNP r:..
Operating Procedures.
System Operating Procedures have valve check
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lists that establish by sign-of f the initial valve positions for the
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start of system operation.
The body of these procedures is written such that startup of the system and/or subsequent shutdown will =ain-tain the required valve position for the existing mcde and/or return the system to the initial conditions.
f All engineered safeguards syste=s have Surveillance Test Procedures which require verification by sign-off correct valve align =ent for required ficw paths at an interval established by Plant Administrative Procedures and Technical Specifications.
Plant Surveillance Test
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Procedures are written in such a =anner as to return system valves to the required position for the existing mode.
This is accomplished by sign-off steps in the p'rocedure.
Req uir ement s for positive controls (locks on valves, locked out pcwer to MCV's, etc.) are specified in the Technical Specifications and in plant procedures. The implementations of positive controls is accomplished by sign-off steps in the appropriate plant procedures, e.g. Unit Operating Procedures, System Operating Procedures and Sur-veillance Test Procedures. These precedures have been revie,ed within the past six souths in accordance with Plant Administrative irocedures.
It has been the Plant's policy to perform valve check lists or ficw path verification checklist following maj or outages, significant main-tenance in areas centaining safety related equip =ent, or prior to re-turning a safeguards system to service from =aintenance or off-normal operation.
This policy will be incorporated into Plant Ad=inistrative Procedures.
In addition, shift relief require =ents will include a walk-down of.=ain control room control boards to check proper ulign=ent of ra=otely-operated equipment.
For additional controls relative to system status during raintenance 2-activities refer to the response provided for item 10.
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APCO Rasponae to 79-06A 9.
Farley syste=s designed to transfer potentially radioactive gases or liquids out of the contain=ent consist of the following:
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Contain=ent Su=p Pt=: ping System a.
No interlock exists to preclude liquid transfer when g_
high radiation indica tica exists.
b.
The system is isolated by the contain=ent isolation signal. The system is no t =ade operable until con-tain=ent isolator is reset and individual isolation valves are positively opened by the operator.
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Surveillance testing required by technical specifications c.
verifics operability of the feature described in ite= b.
Energency Operating Procedure 1 - Loss of Reactor Coolant will be revised to caution the operator not to =anually reopen the containment su=p pu=p discharge isolation valves when high
-tivity conditions exist in contain=ent until the ef fect
.i pt=: ping the contain=ent su=p have been analy:ed.
2.
F.eactor Ccolant Drain System s.
No interlock exists to preclude liquid transfer when high radiation indication exists.
b.
The system is isolated by the containment isolation signal. The system is not =ade operable until con-tain=ent isolator is reset and individual isolation valves are positively opened by the operator, Surveillance testing required by technical specifications c.
verifies operability of the feature described in item b.
E=ergency Operating Procedure 1 - Less of Reactor Coolant will be revised to caution the operator not to =anually reopen die reactor coolant drain *nnk pu=p discharge isolation valves when high activity conditions exist in
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the reactor coolant system until the effects of pu= ping
..x the coolant have been analyzed.
3.
Contain=ent Purge System TT c '
The system is isolated by high radiation in the purge T-"
a.
exhaust line. The system is not =ade operable until
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the isolation signal is reset and positive action by the operator is initiated to start the system.
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A?CO Response to 79-06A ue b.
The system is isolated by the contai _ ant isolation z'n.
signa!. The system is not =ade operable until con-r zi tainmant isolator is reset and positive action by the r:1-operator is initiated to start the system.
'ri c.
Surveillance testing required by technical specifications verifies operability of the feature described in ite=s a an d b.
4.
Containment Post Accident Venting System This system is designed to be used as a backup system for the redundant hydrogen rece binars located in the contain=ent.
- -z The Post Accident Venting System containcent isolation valves are locked closed to preclude operation unless required by failure of the Hydrogen Recc biners folicwing a LOCA.
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APCO ?esponse to 79-06A 2
10.
Ad=inistrreive Procedure 52, Equip =ent Status Control and Maintenance Authoriration, and Ad=inistrative ?rocedure 16, Conduct of Operaticas -
Operations Gro.p, has been reviewed with respect to the require =ents for the following:
a.
Verification, by test or inspection, of the operabili y of redundant safety-related syste=s prior to re= oval of any safety-related system frem service.
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b.
Verification of the operability of all safety-rela + ed L-"
syste=s uhen they are returned to service followi e maintenance or testing.
c.
Explicit notification of involved reactor operational personnel whenever a safety-related system is recoved from and returned to service.
The following actions with respect to the above review require =ents will be taken prior to leaving Mcde 5:
a.
AP-52, Equipnent Status Centrol and Maintenance Authoriration, will be revised to require that prior to re=oving a safety-related system from service for any reason other than scheduled surveillance testing that the operability of the redundant safety-related system will be verified by one of tha following =ethods:
1.
Placing the redundant safety-telated system in service for equip =ent required to function during normal operation.
2.
Testing the redundant safety-related systes for fluid syste=s required to re=ove decay heat fro = the reactor under accident conditions. These syste=s consist of auxiliary feedwater and residual heat re= oval.
=
3.
Inspecting the redundant safety-related system. This co==itsent does not apply when a safety-related system is declared inoperabio due to a =alfunction which requires going under a limiting condition for operation to repair
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er to instru=entation which is placed in the tripped con-
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dicien when recoved from service.
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AP-32, Equip =ent Status Control and Maintenance Authorization presently requires that the operability of safety-related systets be verified prior to their being returned to service following taintenance or testing (excluding surveillance testing) by performance of applicable surveillance test pro-cedures. This requirement will be reviewed and discussed in the training progra= described in item one (1). above.
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APCO ?.csponse to 79-06A
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AP-16, Conduct of Opera tion - Operations Group, will be c.
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changed to nake it clear that the Plant Operator at-the-
, _ _ _ ;2.
Controls is requit ed to log re= oval and return to service
! __;.1 of all safety-rela tad systa=s and equip = cat.
A?-52, 1_
Equip: cat Status Con trol and Maintenance Authorizatio,n, will be revised to require explicit notification of tne cperator at the controls by the Shif t Forenan whenever a safety-related system is re=oved fro = service or found to be inoperable and retuned to service. shift turnove r and relief require:ents vill insure that awareness of I
out-of-service safety-related syste=s is =aintained until
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the sjstem is declared operable.
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w A?CO Response to 79-06 A 1.'.. ihe Parley Nuclear Plant E=ergency Operating ?rocedures reference the E erScacy Inplementing Procedures that require NRC notification as required by the Farley Nuclear Plant Energency Plan,Section III, C, 7.
Prior te raLarn to power operations, the appropriate E=er;ency Imple=enting Procedure vill be revised to require NRC notification within one (1) hour of the ti=e the reactor is not in a controlled or e cpected condition of operation.
Upon no tification, a continuous co==unication channel shall be es tablished and _.aintained with the NRC.
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A?CO ?,asponse to 79-06A 12.
Operating precedures have been reviewed with respect to dealing 27 vith significant amounts of non-ccadensible gases that may be generated during a transient or other accident conditions and would re.af n inside the reactor coolant system or be released to the containnent.
The engineered safeguards are designed and analy:ed to seet the 11=1ts of 10 CFR 50.46 shich require that the hydrogen generation from clad water reaction in a LOCA be li=ited to less than 1% of the clad metal, and nevhere exceed 17% of the clad thickness.
The modes for renoving hydrogen from the reactor coolant syste are:
1.
Hydregen can be stripped from the reactor coolant to the
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pressuri:er vapor space by pressuri:e: spray operatien if a reactor coolant punp is opera ting.
2.
Hydrogen in the pressurizer vapor space can be vented by power operated relief valves to the pressurizer relief tank or directly to the Vola c Control Tank via the vapor space sanple line.
3.
Hydrogen can be remov-d from the reactor coolant syste= by ta l=scovn line and stripped in the Volu=e Control Tank h
where it enters the vaste gas system. The vaste gas systa=
storage capacity is 8 tanka at 600 cuaic feet each. The vaste gas system has two independent co= pressors and hydrogen re-co=h ine rs.
4.
In the event of a LOCA, hydrogen would vent with the steam to the contain=en.t.
If for some reason a non-condensible gas bubble becomes situated
' = n-so=ewhere in the pri=ary coolant syste=s, there are many optiens for continued core cooling and re ving the bubble.
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With a gas bubble located in the upper head, several methods of core cooling are unaffected. The steam generator can be used to remove decay heat using reactor coolant pu=p forced flev or natural circulation. The safety injection system can be used to cool the core while venting through the pressurizer power operated 264 020 c
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J2C0 Pasponse to 79-06A
...=-a relief valve.
Core cooling by any of these rethods can proceed indefinitely if the pri=ary coolant pressure is held constant.
If a lower syste= pressure is desired, a controlled depressurt:stion will allow the bubble to grow slowly until it uncovers the top of the hot legs.
This controlled depressuri:stion can be perforced in two ways:
1.
If the reactor coolant pu=ps can be operated, depressuri:ation can be performed with a steam bubble in the pressuri er.
Depressurization would be through the pressurizer power operated relief valve. Extra control is achieved with the pressurizer heaters and sprays if available. As the bubble grows to t! e top of the hot leg, s=' ll bubbles are carried through the syste=.
a Degassing is dcne with the spray line and/or the Che=ical and Volu=e Control System. The steam generators will carry away de-cay heat.
2.
If the reactor coolant pu=ps cannot be operated or their operation is undesirable, the pressurizer can be =ade water solid with the safety injection pu=ps running and the power operated relief valve open.
Depressuri:ation is controlled by judicious use of the various valves, lines and pt=ps available in the safety injection system and by adjusting the pressurizer relief valve. As the bubble grows to the top of the hot leg, it slides across the hot leg and up into the steam generators.
As depressurization continues the gas bubbles grow in the steam generators and upper head but the core re=ains covered and cooled by safety injection water. If there is enough gas, the pressurizer surge line would eventually be " uncovered".
So=e of the gas would burp into the pressuri er and out of the valve.
This burping process would continue until the system was at tha desired pressure. At that ti=e the current cooling mode could be continued or the system could be placed in an RJR mode.
1-Note that a gas bubble cannot be located in the steam generator
$I_x-with the reactor coolant pu=ps running. If a gas bubble for:s in the steam generator during natural circulation, the reactor
.--T-coolant pe=ps could be turned back on for degassing or safety injection flow could be initiated with the power operated relief valve open.
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Also note that the gas bubbles cannot uncover the core in the above depressurization schemes because it will always tend to float to the top of the system and it cannot co= press water.
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ApCO Respor.se to 79-06 A E=ergency operating procedures vill be developed to incorporate the =ethods of dealing with non-cendensible gases as described above. These procedures v111 reference existing syste= operating procedures (e.g. vaste gas system operation, CVCS lecdown/
charging and VCT degassing) as appropriate.
These procedures vill provide =ethods of calculating the volu=e of a gas bubble in the RCS in the highly unlikely event it s:tould occur.
A por t accident = iring sys tes is pravided to promote =1xing in the contain=ent. This system consists of four contain=ent
=ixing fans, two reactor cavity fans and associated ductwork, and is initiated on a safety injection signal.
The four contain=ent =ixing fans, located on the hatch covers above the reactor coolant pu=ps, are provided to circulate the contain=ent a tmos ph e re.
These fans take suction from the equip =ent ec=part-
=ents and the volu=e over the containment su=p and discharge upward, thereby establishing flow downward areund the periphery of the containment, through the lover contain=ent volu=e and upward through the steas generator co=part=ents.
The reactor cavity fans are provided to assure =ixing from this volu=e.
1he flow from these fans is discharged from the reactor cavity upward around the reactor vessel and outward through the incore instru=ent chase. Four con tain=ent air coolers located above the operating deck are turned on auto =atically on a safe ty injection signal and v111 pro =ote further =1xing in the contain-All of these fans are redundant Seia=ic Category 1 and
=ent.
are powered by the emergency power system.
These syste=s are described in FSAR section 6.2.5.2.4 The post accident contain=ent atmosphere sa=pling system is comprised of two independent trains consisting of sa=ple inlet lines, a re=ovable sa=ple container, a hydrogen analyzer and a sa=ple return line routed to contain=ent. This system is
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described in FSAR section 6.2.5.2.3.
w TV s redundant electric Hydrogen reco=hiners are located in con-tata=ent for use in re=oving hydrogen gas in the contain=ent atmosphere during post accident conditions.
These recombiners
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=eet all engineered safety feature require =ents and the controls and instru=entation for each are located on separate panels in the =ain con trol room. This system is described in FSAR section
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6.2.3.2.1.
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APCO Response to 79-06 A
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':he operations of syste=s and equip =ent related to c =bustible gas control in the centain=ent at=esphere are described in plant F=ergency Operating Procedures and Systes Operating Procedures.
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