ML20125C733
| ML20125C733 | |
| Person / Time | |
|---|---|
| Issue date: | 12/26/1979 |
| From: | Ross D NRC - TMI-2 BULLETINS & ORDERS TASK FORCE |
| To: | Liebler G C-E OPERATING PLANTS OWNERS GROUP |
| References | |
| NUDOCS 8001110446 | |
| Download: ML20125C733 (26) | |
Text
,
/Ett))
'. i., J..i.lL 4iUiW COL 4.
,j v.wr:r. Tor o c. ::,st
. ) l e.i a DEC 2 6 93 Mr. G. E. Liebler, Chairman Combustion Engineering Owners' Group Florida Power & Light Company l
P. O. Box 013100 Miami, Florida 33101
Dear Mr. Liebler:
SUBJECT:
EVALVATION OF OPERATOR GUIDELINES FOR SMALL-BREAK LOSS-OF-COOLANT ACCIDENTS IN C-E DESIGNED OPERATING PLANTS Our letter of June 5,1979 (Robert W. Reid to all operating Combustion Engineering plants) requested that operating plants with C-E-designed reactors develop guidelines for the preparation of operating procedures to cope with small-break LOCA's.
In response to this request, the C-E i
Owners' Group submitted report CEN-114-P (Amendment 1P) which included the'se guidelines.
In response to our requests for additional information and to issues raised during our meeting of October 30, 1979, the guidelines were subsequently modified. The modified guidelines were submitted by your letter dated November 8, 1979.
In my letter to you dated November 14, 1979, we approved the modified guidelines for all C-E operating plants except for a plant having high pressure safety injection pumps with a 2400 psi shutoff head.
In your letter of December 13,1979 (see Enclosure 1) you provided modified guidelines for this class of plant. Subsequent to that, we held discussions with members of the C-E Owners' Group and C-E to clarify certain matters.
We have now completed our review of the modified guidelines. Our supplemental evaluation is provided as Enclosure 2 to this letter.
The supplemental eval-uation in Enclosure 2, together with the evaluation provided in my letter to you dated November 14, 1979, comprise the bases for our approval of the guidelines for this class of plant. The November 14, 1979 letter is provided as Enclosure 3 to this letter.
The November 14, 1979 letter contains a number of provisions which licensees are required to meet in implementing the guidelines. These provisions are equally applicable to those licensees that develop procedures from the revised guidelines.
90022193 8001110 d
f r.
Mr. G. E. Liebler All licensees with C-E-der <n cd reactors are expected to ;.rocee.' 5.ith the development of small break LUCA emergency procedures and operator training.
As indicated on Page 5 of Enclosure 6 to the Darrell G. Eisenhut letter dated September 13, 1979 to all operating nuclear pcuer plants, these procedures and elated operater training are to be implemented by Dacember 31, 1979.
Sincerely,
. f.419 D. F.( Ross, 'J r., Director Bulletins and'0rders Task Force
Enclosures:
As stated ccj See attached lists 1
90022194
E CCLOM"': 1 Deceraber 13, 1979 D.
4 aod F. Ross, Jr.
Bulictins and Orders Task force Of fice of HU: lear Reactor Regulation U.S. Nuclear Regulatory Co::ntssion
Subject:
Additional post-LOCA Guidance for Plants with High Pressure Safety In.iection Pumps with a 2400 psi shut-off Head
Reference:
HRC letter from Dr. D. F. Ross, Jr. to Mr. G. E. Liebler, dated
^
November 14, 1979
Dear Dr. Ross:
Your ruferenced letter forwarded the NRC evaluation of the LOCA guideline for CE designed plants.
That evaluation concluded that the guidelines are acceptable for CE operating plants having high pressure safety in,iettion pums with shut-off heads less than 1000 psi. However, the NRC has not yet deterr.ined that the guidelines are acceptable for a plant (Maine Yankee) having high pressuru safety inje: tion pu ps wit.h shutoff heads of 2400 psi.
Tne HRC is centerned with potential events in which water could be discharged through the safety valves while the operator is atte:npting to achieve a RC5 f~luid condit. ion of at least 500 below saturation.
De question is whether or ii not 500 of subcooling can be achieved, at which point the operator is allowed i
to st.op high pressure f'lw to the RCS, before the pressure in the RCS reaches 1l')
the set 00 int of the safety valves and water is discharged through these valves.
The shutoff head of the V, sine Yankee pump (2425 psi) is belcw tne setpoint of the safety valves (2500 psi) but is above the setpoint of the PORY (2400 psi) i end this needs to be addressed.
i l
To evaluate the NRC contem, rcpresentztive non LOCA events Mich depressurire I
the RCS were studied.
The events chosen were a failure of the reactor coolant I
pressure regulating system and a stearn line break.
Non-LOCA events result in the highest RCS repressurization due to high pressure pump action.
These events essentially represent a "zero break" LOCA case.
In this study it was assu:ned
[
that all four F.CP's were tripped follosing SIS actuation and that the resultinc
{
flw coastd:vn causes loss of pres:urizer sprays.
De study cor. fir cd that 500 l
of subcooling is achieved prior to reaching the setpcin!. of the PORV.
If no i
operator action is taken the high pressure purps could cause the PORV's to lift.
1 t
90022195
- w.
- .
_ q..-7
- * *
- i".?y,l,.r
l?mre is apprwirutely 5 minutes between the point in tirm that 500 sub-cooling is reached and the setpoint of the PORV is reached.
Additionally, when the PORY setpoint (2400 psi) is reached, the pressuri:er is not water solid.
The stuoy also indicates that there is an additionel 5 minutes be-fore the pressurizer is filled solid assic.ing two HPSI perps aru operating.
This is judged to be sufficient tim frarre for the operator to take action.
The action the operator takes in this situe. tion is as follows:
Preferred Action - Tenninate high pressure pu:np fw to the reactor coolant system.
Alternative Action - Prevent operation of the PORV's by:
(a) shutting down strea;n block valves.
01.
(b) position PORV's ' control switch to prevent autonatic i'
opening.
As t. result of the above evaluation it has been concluded that en additional precaution should be added to the CE Post LOCA Guidelines in order to incor-potate applicability to the Maine Yankee Plant.
That precaution fem the attachment to this letter. The additfon of this precaution should result in a determination that the LOCA guidelines are acceptable for develcping oper-ating procedures for Maine Yankee.
i If you should have any questions regarding this guidance; please feel free to I
contact rne or Mr. R. T. Harris of our Technical Advisary Comittee at (203) l 666-6911, extension 5519.
Very truly yours, C-E OWNERS GROUP yet 7%
George E. Liebler Chai rr.an
-l 4 ^
l 3~'
f' D"*
cc c
m jj il hl 90022196 l
I i
ia l
ADDITIONt.i. UJIChlCE FOR PLMITS_ VITH HIGt PRESS SfJETY 11;XC 10?! HMS UITH A N00 PSI SHlfi-L HEAD Add the additional precaution to the Post LOCA t
i 1
Guideline: es follows:
12.
An SIAS can be generated by events other than a LOCA, such as a failure of the reactor coolant pressure regulating system, tinued operation of high head-high pressuru injection pur.ps can Con-cause the RCS to repressurize to the setpoint of the PORI"s.
Opening of the PORV's should be prevented by:
(a) operating the 515 to rufntain RCS pressure below the FORY setpoint (2400 psi) while raintaining at least 503F subcooling.
.L ;,
E S i (b) position the P0F,V control switch to prevent auto.ratic ope ils of the PORV's.
i E
1 (c) shut the PORY block valves to negate consequences of PORY
~
opening.
1 e
e
- . p-b a
~ L
..b[ '
5':.,
- b p
d'.
90022197 k,
E*.
Il, '.
- f.,
a Ib g, -
['Z
.C EVALUATION OF SMALL-BREAK LOCA GUIDELINES FOR C-E C 1.j_.lL/NTS
(
HAVING SAFETY IlklECTION PUMPS WITH SHUT 0FF HEADS GREATEl, ;] iCM PSI INTRODUCTION By letter dated November 14, 1979, we approved the guidelines for deveicping small-break LOCA procedures in C-E operating plants.
This approval was limited to plants having safety injection (SI) pumps with shutoff heads less than 1600 psi; therefore, to support the implementation of these guidelines at plants having SI pumps with shutoff heads greater than 1600 psi, (i.e., Maine Yankee), the C-E Owners Group sub-mitted additional information in a letter dated December 13, 1979.
In addition, on December 13 and 14,1979, we held discussions with C-E personnel regarding our con-cerns associated with plants having SI pumps with high shutoff pressures.
EVALUATION To evaluate the potential of lifting the PORVs in C-E designed plants, all of which have a setpoint of 2400 psi, prior to satisfying the 50'F subcooling criterion at plants having SI pumps with a 2425 psi shutoff head (i.e., Maine Yankee), two non-LOCA events were analyzed:
(i) failure of the pressure regulating system, and (ii) a steam,line break.
The maximum calculated hot leg temperature for these events was 540*F.
Since this temperature is significantly below 612 F, the saturation j
temperature at 2400 psi with a 50" subcooling margin, our subcooling criterion should
]
be satisfied prior to lifting the PORVs. This calculated temperature of 540 F was i
based on the dynamic conditions prevailing during the pressurizer refill portion of the transients.
On December 13 and 14,1979, the staff discussed a more conservative steady state analysis with C-E personnel wherein no credit is taken for the dynamic effects of cold feedwater and SI flow.
Such an analysis simply considers natural circulation in the primary system transferring heat to the steam generator (SG) whose tempera-ture corresponds to the saturation pressure of the SG safety valves.
C-E stated that under these conditions, their analyses show that the maximum hot leg tempera-ture would be 580'F.
Since this temperature is also below the 612 F cited above, the 50 F subcooling criterion would also be met without exceeding 2400 psi, the PORV setpoint.
, CONCLUSIONS Based on the results of the analyses that show that the 50*F criterion will be met prior to lifting the PORVs, we find that the LOCA guidelines are acceptable for C-E plants having SI pumps with shutoff heads of 2425 psi (i.e., Maine Yankee). This approval, however, is contingent upon receiving documentation from the C-E Owners Group of the analyses showing that the 50 F subcooling criterion can be met without exceeding 2400 psi.
b 90022198
_ -. ~. -
~.
J
(
UNIT [" "1
[V.
gj NUCLEAR REGULATORY COMMISSION
, b.,,...d s
W ASHirJGTO N. D. C. 20555 7s G.!ad /
- f mu m I
Mr. G. E. Liebler, Chairman D,e ]D m]D
.q_
Combustion Engineering Owners Group Florida Power & Light Company u
com om 1 R
P. O. Box 013100 Miami, Florida 33101
Dear Mr. Liebler:
SUBJECT:
EVALUATION OF OPERATOR GUIDELINES FOR SMALL-BREAK LOSS-0F-COOLANT ACCIDENTS IN.C-E DESIGNED OPERATIN3 PLANTS Our letter of June 5,1979 (Robert W. Reid to all operating Combustion Engineering plants) requested that operating plants with C-E designed reactors develop guidelines for the preparation of operating procedures to cope with small-break LOCA's.
In response to this recuest, the C-E Owners Group submitted report CEN-ll4-P (Anendment IP) wnich included said gifidelines.
In response to our requests for additional infccmation and to issues raised during our meeting of October 30, 1979, the guide-lines were subsequently modified. The modified guidelines were submitted by your letter to D. F. Ross dated November 8,1979.
We have completed our review of the modified guidelines, and are attaching hereto as Enclosure 1 a copy of our evaluation.
As stated in our evaluation, we have concluded that the guidelines submitted by your November 8,1979 letter are acceptable for use in developing operat-ing procedures to cope with small-break LOCA's in C-E operating plants having high-pressure safety injection purnps with shut-off heads less than 1600 psi. Although the guidelines were based on a reference plant having 200 psi safety injection tanks and 1300 psi high-pressure safety injection pumps, you have stated that they are applicable to all operating C-E plants, including those with 600 psi safety injection tanks and those witn 2400 psi high-pressure safety injection pumps. However, we have net as yet determined that the guidelines are acceptable for a plant having high.-pressure safety injection pumps with a 2400 psi shut-off head. Our concern is related to the potential events in which water could be discharged through the safety valves while the operator is attempting to reach a condition of at least 500 F below saturation. A copy of the approved guidelines, subject to acceptably incorporating those revisions required by Enclosure 1, is attached hereto as Enclosure 2.
I Those licensees with C-E designed reactors for whien these guicelines are approved may now proceed with the development of small-break LOCA emergency procedures and operator training.
In developing these procedures, eacn licensee must account for the effects of specific design characteristics at its ;1 ant.
As indicated on Page 5 of Enclosure 6 to the Darrell G. Eisenhut letter cated Septemcer 13, 1979 to all operating nuclear power plants, these prececures anc relatec cperator training are to be implemented by December 31, 1979.
90022199
'. s "J79 Mr. G. E. Liecle-In implementing thesc ; rocedures, each licensee shall provice:
)
(1) The instrument uncertainties involved with HPI terminaticn criteria to indicate that the criteria will assure subcooled coce:ticr.5.
(2) Adequate assurance that the HPSI pumps will not ce run deacheaded in the recirculation mode and that minimum flow require ents will be met.
(3 ) An indication of the typicality of the analyses doctnented in CEN-ll4-P ( Amendment IP) and in the modified guidelines shown in relative to its own plant.
Licensees will also be required to implement emergency procedures covering the extended loss of all feedwater, (including pressure vessel integrity considerations), and to revise emergency procedures for initiating and moni-toring natural circulation, including provisions for p1 ant cooldo,in. These procedures will be based on guidelines which the C-E Owners Group are develop-ing under " inadequate core cooling."
As part of our audit program, we expect to examine the procedures at a lead C-E operating plant initially, and at other C-E operating plants at a later date to assure that the procedures were developed in accordance witii the aoproved guidelines.
We also plan to check out some of the procedures at a C-E simulator on a schedule to be developed later.
It should be noted however, that our audit program need not impede progress toward implementing the procedures and associated training by December 31, 1979.
Sincerely,
/
i- %,c / 7 g(
D.'.
oss, Jr., Director Bulletins & Orders Task Force
Enclosures:
As stated cc: See attached lists 90022200 J
E N C L 0.5 U R E 1
e 90022201 l
=
1 i
Evaluation of Combustion Engineering Post-LOCA Opercting Guidelines i
Introduction By letter dated June 5,1979, the staff requested that all operating CE plants provide guidelines for the preparation of operational procedures for the recovery of plants following small LOCA's.
The guidelines were to cover both short-term and long-term situations and follow through to a stable condition.
Recognition of the event, precautions, actions, and prohibited actions were to be included also. CE submitted CEN-il4-P-(NP), " Review of Small Break Transients in Combustion Engineering Nuclear Steam Supply Systems" in July,1979 and CEf;-il5-P(NP),
" Response to NRC IE Bulletin 79-06C Items 2 and 3 for Combustion Engineering 1
Nuclear Steam Supply Systems" in August,1979.
CEt'-114-P(NP) was submitted l
in response to our request for information while CEN-ll5-P(NP) revised this response to account for the impact of RCP operating requirements.
i Summary Descriotion:
CE Post-LOCA Operating Guidelines The guideline submitted by CE is preceded by a bases section which supplies I
background material for the information presented in the guideline.
The guide-line itself is split into four sections:
Symptoms, Immediate Actions, Follow-Up Actions, and Precautions.
The Symptoms are a list of indications which an operator is expected to utili:e in confirming that a small break loss-of-coolant accident has occurred.
Low pressurizer pressure, high containment sump level, high containment pressure cr temperature, safety injection actuation, and high or low pressurizer level are among the symptoms provided to the operator to assist in the identificatien of this accident.
A diagnostics cnart has been appended to the LCCA guidelines to clarify symptoms and to channel the operator's actions into the correct procedure.
90022202
I=ediate.::ir.: =re th:se acti:ns which are re:.ir-d :s ;' 2:5 :he ;;r.: in a s de
~
cendi-icn. ~hese steps are distingui:hed fr:m subseque. : pr::i:ral ::t :- by a re-quire ent for remerization. An cpera:Or mast know these sce;s.-ith:u reference to a procedure, thereby ensuring that there is no delay ir. achie. i.; a scie ::r.dition.
The guidelines require that the reactor be tripped; standa_M p:3:- rip a::icns be carrie, out (plant specip.c); sa.,ety injection se :nitiate, (1 no: au ::atica2,y y
actuated); reacter coolant pumps be tripped after SIAS acraa i:n er 1cw F.CS pres-sure; auxiliary feedwater flow be established if main feedwa:er is nc available; verification that the CIAS and SIAS signals have properly acr22:ed; the SIS be operated to maintain a 50 F subcooling margin and indicated pressuriner level; and the break be located and isolated if possible.
em e
. 3, D
D
~
ol\\_A.
wo o
Tollcu-Up Actions are actions required to place the plant in a stable condition.
The previous procedural steps (Immediate Actions) ensured tha: -he reac:cr was in a safe ccndition, that the core rerains covered by ECCS opera-1:n, and that escap-ing radioactivity is isolated by CIAS.
The ner: steps are air.ed at bringing the plant to a lower mode of operation, cold shutdown.
The Follow-Up Actions recuire a plant cooldown within one-hour using the steam du ps er t rbi.e bypass system.
Tne cooldown is centinued via a number of alternative paths sud as 1 erg-tem. recircula-tion, initiation of shutdown cooling, continued use of the stes. d = s and e.ergency feed, or, as a last resort, cperting of the power operated relief valves.
Tne Drecautions section lists warnings which the opera:Or rast observe :: ensure plant safety. For example, the cperator is warned *ha; pressr iter level may not always be a tnie indicator of fluid inventer / and that pri.rarf syste. te.perarare must be monitored when establishing auxilia-y feedwater to pre' cent excessive ecci-down rates. A total cf eleven ?nscautions have been included f:r i:ple a.tatien by the licensees in the appropriate procedural locaticns.
90022203
Evaluation The NRC staff reviewed the post-LOCa operating gu'.t;' r:. th respect to tne following critical operator actions:
D 1.
Reactor coolant pump trip 2.
Safety injection termination criteria 3.
Verification of safety systems actuation t
4.
Verification of a heat sink.
l During our review, the staff identified modifications to be made to the guide-1 lines to enhance the directions to the operator.
These modifications were subsequ,ently incorporated in the guidelines via revisions issued on Nove:6er 8, 1979 The criteria for tripping the reactor coolant pumps are consistent with the requirements of IE Bulletin 79-06C. All operating reactor coolant pumps are stopped after an SIAS caused by low reactor coolant system pressure and after it has been verified that the reactor has been shutdown for at least five seconds.
We conclude that this criterion is acceptable subject revising "Immediate Action" item 3 of the guidelines to be consistent with the above wording.
The criterion for terminating safety injection flow is based on the establishment U
and maintenance of a 50 F subcooling margin along with an indication of pressurizer level.
The staff concurs that these criteria are sufficient for ensuring that safety injection can be terminated without concern for detrimental voiding in the primary system.
We conclude that tnis criterion is acceptable for those plants with low-head HPSI pumps K 1600 psi).
90022204
As part of his inmediate actions, the coerator is directed to terify the reactor trip, safety injection actuation, adequate auxiliary feedwater flow (if main feedwater is not available), and containment isolation actuation.
We concur that these actions are sufficient to ensure minimum safeguarcs and heat sink availability needed to mitigate small break LOCAs.
The staff noted that the guidelines are based on obtaining at least minimum safeguards operation to mitigate.small break LOCAs.
We require each licensee to extend the emergency procedures to cover the loss of all feedwater.
Procedures for this degraded condition should also take into account pressure vessel integrity considerations. The Owners Group has committed to prepare guidelines for oper,ational procedures regarding the loss of all feedwater as part of its effort on the issue of inadequate core cooling.
The staff also requires that the emergency procedures include instructions for monitoring and initiating (if lost) natural circulation for small break LOCAs where heat removal by the steam generators is required.
A separate guideline has been received on natural circulation cperation.
The staff, upon completion of its evaluation, will require that the natural circulation guideline be appended to or referenced by the appropriate emergency procedures.
The staff requires that each licensee provide procedures for ecoling down the plant under natural circulation conditions.
Tnese procedures should address boration control and monitoring, cooldown of the pressurizer, and acequate criteria for monitoring coolant system temperatures to ensure that voids co not form in the primary system which could inhibit acecuate heat removals As in 90022205
t D M
}D ?]D*
,5 DJu aju u j
\\
i t e case of loss of all feedwater, the Combustion Engineering Gnners Group has h
committed to prepare guidelines for operational procedures regaroing cooldown under natural circulation conditions as part of its effort on inadequate core cooling.
Conclusions Based on our review, we conclude that the small-break loss-of-coolant accident operating guidelines submitted by the Combustion Engineering Owners Group on November 8,1979 are acceptable for C-E plants having high-pressure safety injec-tion pumps with shut-off heads 1600 psi or less.
Accordingly, said guidelines can be used for developing operating procedures for coping with smal'l-break loss-of-coolant accidents for such plants, provided that the licensees imple-ment the requirements noted above when developing their procedures.
Our 1
acceptance of these generic guidelines notwithstanding, each licensee must account for the effects of specific plant design parameters (e.g., differences in the shut-off pressures of high-pressure safety injection pumps, differences in the design pressure of the safety injection tanks), when translar.ing these j
guidelines into plant specific operating procedures.
90022206
ENCL 05URE E
90022207 i
I
P.O. Box 52910C Miami, FL 33152 November 8, 1979 Dr. Denwood F. Ross, Jr.
Director Bulletins and Orders Task Force Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D..C.
20555
Subject:
Transmittal of Revised Post-LOCA Guidelines
Reference:
(A) NRC letter from Dr. D. F. Ross, Jr. to Mr. G. E. Liebler, dated October 19, 1979 (B) IE Bulletin 79-06C, dated July 26, 1979 (C) NUREG-0578, July 1979
Dear Dr. Ross:
Reference A requested additional infomation regarding the guidelines presented in CEN-114 Revision A and CEN-115 for loss of coolant accidents.LOCA).
Questions regarding those guidelines were further discussed in a meeting with the NRC staff on October 30, 1979, and a number of revisions were agreed upon.
This letter transmits those revised Post-LOCA guidelines.
These guidelines are being sub-mitted for your approval on behalf of the Combustion Engineering Owners Group so that they may be incorporated into utility procedures in accordance with Reference B and the schedule presented in Reference C.
It should be noted tnat these guidelines do not necessarily reflect the preferred i
actions of our vendor, Combustion Engineering.
Combustion Engineering's ore-ferred actions remain as stated in CEN-il5.
The NRC staff has specifically re-quested that the guidelines for P.JP operation be revised to incorporate tne RCP operating requirements stated in IE Bulletin 79-06C (Reference A, Item I.6.E).
Combustion Engineering has been unable to identify a transient analyzed in Chapter 6 or 15 of the FSAR that will result in violation of acceptance criteria, provided the RCP's are not tripped until the rods have been fully inserted for 5 seconds. The enclosed guidelines have therefore been revised to reflect the staff's request.
If you should have any questions regarding these guidelines, please feel free to contact me at (305) 552-3811.
Very truly yours, iERS GROUP enepww George :. Liebler
'"'i"*'"
90022208
"'. POST LOC A GUi3EL:"E3 2 hsss for Post-LOCA Ooeratina Guidel.ines provided below is a general description of plant res:enses to large and small break LOCA's.
This is intended to supply background raterial for the information presented in the guidelines.
A small break LOCA is characteri:ed by:
a)
A slow loss of RCS pr. essure during the short term (10'to 30 minutes) and equilibrium cressure above
- 300 psia in the long term (30 to 480 minutes) resulting from matching safety injection flow and flow from the break, b) A loss of RCS inventory during the short term followed by a refilling of the RCS during the long term.
't) Core cooling is initially by the steam generator (s) and flow from the break and later by the shutdown cooling system.
The break does not always (depending on siz,e) provide the necessary heat removal yet.,
decletes RCS inventory.
Breaks in RCS piping less unan 2 inches in diameter fall into this category.
The steam generators provide cooling for forced or natural circulation o'f the RCS, if inventory is depleted, in a boiloff and reflux mode.
The shutdown cooling system is used after the RCS has been refilled and pressure control is provided by the HPSI pumps and the charging pumps.
A general description of small break LOCA operations follows:
Initially, the plant is hot and pressurized. A small break LOCA results in a slow loss of RCS inventory and a decrease in pressure.
Low pressuri:er pressure initiates a SIAS which automatically actuates the 515.
The reactor is tripped.
The operator stops the reactor coolant pumps.
Auxiliary feedwater is established to the steam generators.
Steam cump is provided manually using atmospheric dump valves or turbine bypass valves, or autcmatically by the steam generator dump and bypass system or by steam generator relief vaives.
90022209
'inis value is typical, it may vary for specific designs.
D"D WWW g r b
wu' d @l/h,
i For very small breaks, the stcr yne
- i-s the ria i r.e t-sink, and additional heat is removed with the er ' N : rcugh tne treai.. Continued reactor coolant pump operation durin; :r:: ce-icd could aid heat removal by the steam generators.
However, fcr s EI' hot leg breaks, reactor coolant pump operation will result in a higher :..o-:hase mixture level in the reactor vessel and hot leg piping.
Consecuently, for a break in the bottom of the hot leg, the break is covered longer by two-phase mixture, causing a larger loss of water inventory from the vessel.
This eventually results in a lower coolant level in the reactor vessel.
The result could be a higher clad temperature and a delay in refilli~ng the vessel.
The net effect of reactor coolant pump oper,ation during the initial period may be to i,ncrease the severity of the accident.
The NRC has therefore requested that the RCP ooerating requirements stated in IE Bulletin 79-06C be incorporated into the guidelines for operating plants following LOCs (MRC letter from Dr. D.F. Ross to G. E. Liebler, dated October 19, 1979).
Sulletin 79-06C directed to holders of optrating licenses to:
"Upon reactor trip and HPI initiation caused by low reactor coolant system pressure, imrrediately trip all operating RCP's."
This action shourd not result im the violation of acceptance criteria for transients or accidents in chapter 6 or 15 of the FSAR, orovided the RCP's are not tripped until rods have been fully inserted for 5 seconds.
This delay is to allow for the cecay of the heat flux following reactor trip before reducing forced flow.
The time necessary to refill the RCS and regain control of pressure and inven-tory depencs on break size, break location, and the number of HPSI pumps and charging pumps actuated.
With only one HPS-1 pump activated, and a break located on the bottom of the cold leg, it may take as long as 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to refill the RCS, With all injection pumps operable, the time is about I hour.
In the period of time it takes tne RCS to refill some voiding in the RCS will occur.
- This, condition can be recognized by indicatien thit RCS hot leg temperature or core thermocouple temperature is equal to the saturation temperature for the existine RCS pressure.
In this mode, decay heat is re oved by coiling in tne core and condensation in the steam generator.
In accition, heat is removed by flow frem the break.
The operator must ensure that tne SIS is providing ficw to the RCS, and the steam Senerators are removing heat. These actions will ensure adequate core cooling and eventually a subcooled c:ndition will be acnieved.
Once RCS oressure anc te perature are adequately re:v:st, the shutc0wn c cling system 90022210
'D @ r D
P D Wj s
n L%k 3
b d
is ? laced in coeration.
In the event that the feecwater su: ply :..c : tc generator is exhausted and the shutcown cooling system is inc:eraole, :,e 23.,
are opened to ensure that the flow from tne injection syster is suffi:ient cool the core.
The SIS will be realignec for cold leg inje:- :n :n'.y.
Cm.
flushing is from the cold legs through the core and out the PORV.
Simultaneous het and cold leg injection is used for botn small break anc large break LOCA's so the operator does not have to distinguisn between them at the time when simultaneous injection is required fer large breaks.
(For small breaks, the boron concentration remains low due to dispersal throughout the RCS, so hot ahd cold leg injection is not essential).
Reactor coolant system pressure is used to differentiate between smali and large break LOCA's.
However, the delineation between small and large breaks does not need to be precise since there is a range of intermediate breaks for whic,h either response will produce satisfactory results.
The guidelines take this 'into account with the decisions to be made af ter eight hours.
The large break LOCA is characterized by:
a) A rapid loss of RCS pressure in 10 seconds.to 3 minutes with equilibrium pressures below* 300 psia and, in the case of the largest breaks, the RCS pressure nearly equal to containment pressure.
b)
Core cooling is provided for by large flow from the injection system due to low RCS pressure.
The flow from the break provides sufficient heat removal.
Simultane'ous not and colo leg injection is required'to prevent po:'sible boric acid accumulation in the.
Core.
A general description of.large break LOCA cperations fo11cws:
Initially, the plant is hot and pressurized.
A large break LOCA results in a rapid loss of inventory and pressure.
Low aressurizer pressure initiates a SIAS which automatically actuates the SIS.
The reactor is. tripped.
Auxiliary feedwater is established to the steam generators.
Steam cumo is provided manually using atmospheric steam dum: valves or turbine by ass valves.
The major mechanism for heat removai is the flew from the 5:5 "This valve is typical, it may vary for specific cesigns.
90022211
nrough the core and out the break.
COntain9:nt : es;ure Ey be hign anc centainment isolation is likely.
Containment s: ray egy 2ve :een auto-matically activated.
D""""]D D
7D' i
o o Ju o.
o The SIS is aligned to provide simultaneous hot and celd leg injection which is sufficient to cool the core and flush the reactor vessel indefinitely.
For both large and small break LOCA's, continuec monitoring of conditions in the RCS and performance of safety systems should be dcne.
All available indications should be used to aid in diagnosing the event since the accident may cause irregularities in a particular instrument reading.
Regardless of the cause of actuation of a safety systen, tne automatic respcnse should not be altered until it has been demonstrated tnat otner systems and equipment are providing the functions that the safety system is intended to perf rm.
Q 90022212 t
4 l
l e
c e
BREAK IDEtiTIFICATI0tt
'RESSURIZER PRESSURE DECREASitlG OR UrlEXPLAltlED CllAtlGE Ill PRESSURIZER LEVEL e
OBSERVE STFAot GEilERA10R PREhaVRES ABil0RilALLY L011 Itl ONE STEAll GEllERATOR PRESSURES IfflTIALLY tt01:t1AL OR OR B0Til STEAt1 GEllERATORS RISitlG lilEf1 SUBSEQUEllTLY DfCREASitlG AfTER A REACTOR T_ RIP...
r STEATI LitlE LOCA OR RUPIURE M
S.G.' TUBE RUPIURE 4
)
O
__M M
__ '{
OBSERVE C0t!IAlliittfil PitESSURE, OBSERVE C0tlTAltit1EllT PRESSURE, RADIAll0tt Att0 SUI 1P LEVEL RADI AT10tl Atl0 SullP LEVEL W
,.\\
O j
~
.. X.
A -_
..-..,6. -- _ }
g m
litchi ASillG tl0Rf1AL IrlCREASillG tl0llitAl.
Oo N
b VERIFY lilGil RADIAIlott AIR FJfCIOR u
DISCIIARGE
_ _ _ _._ __ ]
y l
SIEAtt ilREAK SIEAtt BREAK v
IriSIDE OUISil)E T
S G I'8Ilf C0lli Alitill flI C0!!IAIll!1EflT LOCA nigpiupt
Guicaline: for Ooeratin: riants Fo' mtin: L2C;'s Syectoms 4
Reactor coolant system leak exceeds :he ca:acity of
- - ::c n'.;
. a;i.;
oumps.
2.
A reactor trip may have occurred.
3.
The Safety Injection System (SIS) may have autome:ically ac:uate:.
4.
Any one or more of the following indications or alarms may be present.
a) Low pressurizer pressure b)
High containment pressure or tem:erature c)
High containment sump level d) High containment radiation y,
e) High or low pressurizer level WQ W
f)
High quench tank level
- 3 g) High quench tank temperature h) High quench tank pressure i)
T,y decreasing or at saturati'an temperature for RCS pressure.
Immediate cticns 1.
Trip the reactor if not already tripped and carry out standard post trip actions.
2.
Initiate safety injection if it has not already been actuated by the safety injection actuation signal.
3.
After an SIAS caused by low reactor coolan system pressure and after it has been verified that all rods nave been fully inserted for 5 seconds, stop all operating reactor cooiant pumps.
4.
If main feedwater is not available, immedia:ely establish or verify an auxiliary, feedwater flow of "gpm.
5.
If the containment isolation actuation signal (CIAS) is activated, ensure that the system has properly actua:ad.
6.
Ensure that the systems receiving an SIAS are pecperly actuated and that CIAS is actuated.
7.
After,any SIAS, operate the 515" untii RCS hot and coic leg temperatures 0
are at least 50 F below saturation :empera:ure for ne RCS pressure and a oressurizer level is incicate:, unless the cause of the 51A5 has 0
been verified to be an inadvertent 1:taa:icn.
If 50 F subcooling cannot be maintained after the system has :een s;;;;ec, tne ni;n pressure injection system must be restarted.
90022214
6.
Atte/pt to lo:3 e 3.d isolate int 5:.;r:e f tne leak.
- ri; itie le;,.
loca-icns incluce, bu, are rot li"i e:.,.ne p00.\\'s,
- ne l e.::wn li ne and secole lines.
Fcilow-US Actions d
I i
1.
Operate atmospheric steam dump valves (or turbine bypass valves if tne condenser is available) to maintain or re:uce plant temperature and reduce steam generator pressure below tne steam generator relief valve setpoints.
Begin a plant cooldown as soon as possible anc in any case within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2.
Manually align the safety injection and cnarging systems.to provide flow to the RCS hot and cold ' legs" twc hours after the LOCA**.
~
3.
If the pressure and inventory control with the SIS cannot be established after* eight hours and RCS pressure is less than* 300 psig, continue the hot and colc leg injection.
4.
If pressure and inventory control with the SIS are establisned after* eight
.. hours and RCS pressure is greater than* 300 psig, conduct one of the follow-ing activities.
The activities are listed in order of decreasing preference.
a)
RCS' pressure above* 300 psig indicates that the system has refilled and subcooling has occurred.
Verify this by checking Ine saturation pressure for the existing temperature.
Realign the SIS for cold leg injection.
Continue to maintain subcooling and reduce RCS pressure to the initiation pressure for shutdown cooling by reducing the flow delivered by the high pressure injection and charging pumps and by venting or isolating the safety injection tanks as necessary.
While recucing pressure and after shutdown cooling is initiated, maintain RCS oressure with tne energing pumps and/or the HPSI pumps to continue to maintain at least 500 sub-cooling, or-b)
Continue to remove decay heat using emergency feed and steam dump if adequate condensate is available and (a) cannot be implemented. or c) Open pressurizer power operated relief valves and align the 5:S for cold leg injection if (a) or_(b) carnet be,,imole,mented.
Inis value is typical, it may vary for soecific designs.
includes stopping charging pumps on sore piants 90022215
Erecautions 1.
Before restarting T.:?':.-:rc nat co:lin; v.ater services tc tne ;um::
has been restored.
2.
Pressuri:er level may nct always be a true indicator of RCS f'.uid invent:ry.
Pressurizer steam space ruptures, reference leg failures, and reference leg flashing may cause incications which are contrary to true concitions.
3.
All available indications silould be used to aid in diagnosing the event since the accident may cause irregularities in a particular instrument reading.
Critical parameters must be verified when one or.more confirmatory indications are available.
4.
When extablishing auxiliary feedwater flow to the steam generators, monitor primary system temperature and pressure to avoic exceeding a 1000F/ hour cooldown rate.
5.
Feedwater is normally provided to both steam generators.
Isolation of a single steam generator is mandatory if a steam generator tube rupture is detected in that generator to prevent lifting of the safety valves or reseat them if they have lifted.
This action will also reduce 9e amount of radioactivity released.
For small breaks in the RCS wnere steam generators are important for heat removal one steam generator must be used for this purpose even if primary to secondary leaks are detected.
6.
Continued lengthy operation of the containment soray may jeopardi:e the operation of equipment which would be. desirable or necessary to mitigate the consecuences of the event.
Early consideration should be given to termination of spfay operation.
If the containment pressure has returned to below tne actuation setpoint, the system may be stopped.
The system should be realigned for automatic actuation.
7.
Observe ali available indications to cetermine c nditions within the RCS.
__Use RCS hot leg temperature, RCS cold leg temperature, core exit thermo-couple temperature, and RCS oressure to cetermine if the RCS is subtcoled or saturated.
An increase in temperature above tne saturation temperature-f for the existing pressure is an ' indication cf voiding in tne RCS. A de-crease in operating RCP motor current or erratic pump t.P is also an indication of voiding.
If this occurs the operator must ensure that the RCP's are turned of f, the SIS is providing makeuo to the RCS, ano tha t.the steam generato.rs are removing heat from the RCS.
90022216
E.
itonitor refueling water tank level to verify tne Sr.i#.
recirculation.
If a recirculation ac uatien sigt.si '::2. : ::_
.=
operator must prevent the HPSI pumps ' rom coe stin; :: 's:: -.:
.-in-um flow conoitions.
If all HPSI pumps and cnarging ou :: are cre-t'n; and the HPS! pumos are delivering less than 30 gpm per 0;mp, turn :f f tne charging pumps one at a time and then HPS: punps one at a time un:il only one HPSI pump remains operating.
This will ensure that mir.imum ficw requirements will be met by the flow through the our.: to the RCS for the smallest break size that results in a SIAS.
9.
Monitor the auxiliary building radiation levels anc sump levels after an RAS to attempt to detect leakage from the SIS.
Even if leaks are cetected at least one high pressure safety injection pump must remain in 0:eration to provide flow to the RCS.
10.
If there is a high radioactivity level in the reactor coolant system, circulation of this fluid in the SCS may result in high area radioactivity readingt. in the auxiliary building.
The activity level of the RCS should be determined prior to initiating SCS flow.
- 11. Minimum Pressure - Temperature operating restrictions take precedence over requirements for operation of the high pressure injection or charging 0
system to achieve 50 subccoling during operation of the shutdown cooling system.
90022217 D)
- DD IID D **
M e MM)L f\\f A
es m
m-
. ~.
)
,,'DPr.20 a d3 ruiLq
- D r7 e'
e w
Mr. David Bixel "r. G. 'E. '. d e : : e,
F i t -r g r tiuclear Licensing '4niistrator Combusti:- E.;' earir; Owners Gr:_p Consumers Power,Connany Florica ::xer 1r: ' i g,. C ompany 212 West Michigan Avenue P. O. Scx 1310 Jackson, Michigan 49201 v j a:-i,. : : :- f 3 Mr. Willian Cavana;g.,111 Mr. Ken Morris
, Vice Chairman Executive Director of Generation.
Combustioc. Engineerin; Owners Group and Construction Omaha Public Pcwer District Arkansas Power & Light Company Fourth & Jones P. O. Box 551 Omaha, Nebraska 68105 Little~ Rock, Arkansas 72203
"" John Garrity, Chairman Mr. A. E. Lundvall, Jr.
GuWelines Subgroup Vice-P' resident - Supply Maine Yankee Atem1c Pcwer Company Baltimire Gas & Electric Company
.dison Crive P.. Box 1475 Augusta,. Maine 04336 Baltimore, Maryland 21203 Mr. Robert T. Harris, Chairman Mr. Theodore E. Short
nalys t s Subgroup Assistant General Manager Northeast Utilities Service Co.
Omaha Public Power District
. O. Box 270 1623 Harney Street Hartford, Connecticut 06101 Omaha, Nebraska 68102 Mr. David S. Van de Walle Mr. Robert H. Groce Consumers Power Compar.y Licensing Engineer 212 West Michigan Avenue Yankee Atomic Electric Company Jackson, Michigan 49201 20 Turnpike Road Westboro, Massachusetts 01581 Mr. William Szymczak Yankee Atcmic Power Company 25 Research Drive Mr. W. G. Counsil, Vice-P-resident Nuclear Engineering & Operatier Westboro, Massachusetts 01581 Northeast Nuclear Energy Company P. O. Box 270 Hartford, Connecticut 06101 Mr. J. T. Enos Arkansas Power & Light Company P. O. Ecx 551 Dr. P.obert E. Uhrig, Vice-Presid Little Rock, Arkansas 7220s Advanced Systems & Technology Flcrica Power & Lignt Company P. O. Box 529100 Miami, Florida 33152 90022218 s
.,_