ML19224B834
| ML19224B834 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee, Crane |
| Issue date: | 05/08/1979 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7906270032 | |
| Download: ML19224B834 (27) | |
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WASHINGTON, D. C. 20555 t
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May 8, 1979 NOTE T0:
S. Varga FROM:
D. Ross, Jr.
SUBJECT:
CE RECOMMENDATIONS ON BULLETIN RESPONSES, AS SENT T MAINE YANKEE (4/20/79)
The subject recomendations (six copies enclosed) should be reviewed along with your review of CE bulletin responses.
Please determine the following:
1.
To what extent has M-Y passed through the CE recomendations?
2.
Is the general quality of the CE work better than that given by M-Y?
3.
Should we get similar info frcm W? B&W?
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D'. Ross, Jr.
Enclosure :
As stated cc:
J. Heltemes R. Reid A. Thadani N. Mosely, IE T. Novak E. Case Z. Rosztoczy
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,Combuttion Engineering. Inc.
Tel. 203/ 683-19 1 s
1000 Promect Hiii Acac Teler 9 9297 G
j Yhncser. Connecticut 06035 7:7 COMBUSTION hid ENGINEERING April 20, 1979 NFS-MY-79-18 Mr. D. E, hbody Maine Yankee Atomic Pcwer Company 20 Turnpike Road h'es tboro,.vassachusetts 01531
Subject:
I 6 E Bulletin 79-063 Question Response
Reference:
C-E Letter, Serial NFS-MY-79-16 dated April 20, 1979
Enclosure:
(1) Response to Questions 2, 6, and 11 of IEE Bul'Mtin 79-06B, dated April 14, 1979
Dear Mr. > body:
Enclosure (1) is provided for your infcmation and use in accordance~ with the commitment of Reference (A). This info =ation package provided by this letter is as complete as possible considering the limited tine available for prepara-tion and may be used by Maine Yankee Atcmic Power Company as a basis for your response to the NRC questiens posed in 16E Bulletin 79-063 relating to the
'IMI-2 event.
Combastion Engineering plans to provide, in the near future, additional infoma-tion on the referenced cuestions as well as scme of the other questions presented in this b_11etin.
Further infomation will be based on additional clarification and understanding of the circumstances surrounding the events at BlI-2 and a continued review of the applicability of the recogni::cd areas of concern to C-E NSSS designs.
Any feedback on the content of the enclosure as well as your independent assess-ment of the situation will assist C-E in providing the most accurate and timely information possibic.
In addition, continued interfac.ing betwee.. our
.mpanies will allow C-E to best include your interests in our discussions with the NRC if and when we are called to attend another NRC meeting en D!I-2 related matters.
This infonation is provided under the terms and conditions of General Service Agreement. 'ihe precise arrangement associated with cost sharing an.1 billing will be the subject of future correspondence.
If you wish to discuss the answers provided as Enclosure (1), please do not hesitate to contact me as socn as possible.
Very truly yours, CCMEUSTIO
'3IqiERING,INC.
t -ra NuclearServiceSiteManag.g-C. Bc.ghter uB:llG cc: Distribution
--_.SPCMSE TO QUESTION #2 RE Both cendensible and non-cendensible void generation is possible within the RCS.
Condensible or steam voids are produced when the system pressure is reduced to the saturation condition of the core outlet temperature.
This mechanism has the potential for significant void generation and inhibiting core cooling.
There are several sources for non-condensible voids.
There is an insignificant amount of dissolved cas in the RCS coolant during normal operation.
(30-50cc H2/Ka H20, 400ft3 0 STP)
Even with a significant depressurization, the quantity c' gas available frem this source would have a negligible impact en system natural circulation capability.
Follcuing a LOCA event, there is the possibility of generating a significant non-condensible void consisting of nitrcaen frcm the SIS tan's and nydrogen from clad oxidation which has t'n'e potential to significantly reduce or prevent natural circulaticn core cooling.
There are several indications or rethods available to the plant operators to determine the presence of voids in tne RCS. A reliable indicaticn of steam voids is a coroarisen of the system saturaticn cressure as derived from the core cutlet temperatore with the pressurizer pressure with prc::er account for measurenent errors and,RCP coeraticn.
Indications of voids, both condensible and non-condensible, are also obtainable fren ficw derived information.
Increasing core tT, erratic steam generator d/p, erratic RCP rotor current and RCP vibration monitoring instrumentation provide prompt indication of system voids.
With system intecrity kncun and ficws to and frcm the system kncwn, a ceasure of void ray be obtained by chargin; a known arcunt of rass into the RCS and monitoring the corresponding change in pressurizer level.
To prevent void generation within the RCS, it is of caramount imocrtance to maintain systen pressure above the saturaticn pressure asscciated with coolant temperatures.
To facilitate operations, the saturr. tion curve should be prcminently displayed in the control recn and consideration will be given to oroviding an on-line monitoring and alarm of minimum systen sub-cooling.
Operator actions and transients which result in depressurization will be reviewed with the emanasis an understanding resultant conseauences and the necessity for conductina sucn operaticns in a controllable nanner.
Plant coerations to be revieoed incluce pressurizer spraying, venting and de-cassing; cvercnoling of the RCS by over-feeding of the stean generators with the resultant loss of pressurizer pressure and pressurizer inventory control and HPSI pump cut-off when normal pressurizer pressure control is not available.
The presence of voids in the RCS is indicative of a loss of prenure control and inventory situation.
The initial ocerator actions are twofold. The first basic consideration is the cacability to re-astablish pressure control and increase overall system pressure.
This will also tend to collapse steam voids and to limit the amount nf ncn-condensible gases released frem solution. Operator actions to be taken depend upon whether or not system integrity can be raintained.
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RESP 0MSE TO QUESTICil #2 (Cent.)
Inherent in the effort to restore pressure control is the check and isolation of paths that could be the cause of loss of primary system inventory.
Pressure control is restored by fluid make up by the CVC5 and by the pressurizer control system, if primary systen inventory can be restored.
If the pressurizer cannoc be restcred, pressure control for snall break LOCA's is provided by the HPSI pumos while maintaining the forced circulation and heat transfer to the secondary system.
^
Once basic pressure control has been re-established and the steam voids condensed, any non-condensible gases can be removed by appropriate venting and degassing precedures.
Concurrently, the second basic consideration is directed towards maintaining or re-establishing the secondary heit sink.
Central to returning the bulk RCS fluid to a sub-cooled condition is the ability to maintain or re-establish RCS forced circulation and primary-to-secondary heat trans fer.
Steam voids swept into and condensed in the steam generators f acilitates the stabilization of loco concitiens.
Relative to core decay heat removal recuire ents, sucstantial forced circulatien is provided by the RCP's even coerating in a cavitating
- mode, (Item 6.c provides guidelines for continuet operation of RCP's).
Concurrently, the third basic consideration is that there are multiple parameters available for use in determinaticn of leak locations and the possibility of subsequent isolation.
For example, status of the PORV's can be determined frca an evaluation of the folicwing indications:
- PORV/ Safety Valve Discharge Temperature
- Quench Tank Level
- Quench Tank Temperature
- Pressuri:cr Pressure
- Pressurizer Level Another examole wculd be the assessment of the charging / letdown system through an evaluation of the following parameters:
- Volume Control Tank Level
- Letdown Valve Position
- Letdown Line Temperature
- Charging Flow
- Charging Pump Instrumentation If the leak path is isolable, pressurizer inventory is restored by the CVCS charging sys tem.
Pressurizer heaters are used to restore system pressure and in conjunction with pressurizer spray used to control system pressure wnile loco temperatures are maintained in a subcooled conditico with the secondary plant heat removal systems.
If the 515 was activated, the guidance for securing the SIS is found in Questicn 6.
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RESPONSE TO OUESTION #2 (Cont.)
If the leak path is not isolable, actuation of the safety injection system (either manually or automatically) will provide additional fluid inventory makeup and some measure of pressure control.
The actual pressure maintair.ed under these conditions will be a function of SIS fleurate and the size of the break.
Forced RC5 circulation and heat transfer to the secondary system must be maintained to reduce RCS temperatures. Throttlino of the HPSI injection flow rate will decrease system pressure and must be cerforred in an crderly and controlled fashion with corresponding reductions in plant temperatures.
Note: Some of the instrumentation described in the previous res;cnses may not be qualified nor have redundancy for these specific accident conditions.
The o;erational personnel will be instructed on the qualification level and on alternate instrumentaticn for accident concitions.
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RESP 0 IGE TO QUESTION 6a:
The operating procedures and training instructions have been reviewed to provide that autcnatic actions of engineered safety features will not be overridden by operator action, unless the continued operation will result in unsafe plant conditions or until the plant is in a safe and stable con-dition.
In some circumstances continued unattended oceration of engineered safety features may result in undesirable plant conditions.
The circunstances which the oporator must be alert to and the basis for taking control in those circunstances are described belcw.
In ceneral, it should be deter-mined by the operator that the function being performed by the ccnponent or systen, which is to be stocped or controlled by tne oDerator, is performed by alternate means prior to taking over autcmatic actions of engineered safety features.
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There are tuo basic types c/ engicieered safety feature actuations; those that initiate active systens designed to nitigate an event and those tnat provide an isolation function and have a passive role follcaing the initial action.
The primary active systems are the safety irjection, containment spray and emergency feedwater systems.
The safety injection system can safely renain in operation for extended periods when actuated by a low pressure condition.
However, should it 0
remain in cperation while the reacter coolant system is cooled belcw 250 there is a potential for overpressurizing tne reactor vessel if pCS oressure is too high.
Question 6 (b) provides information en tne basis for termin-ating high pressure safety injection ficw.
The containment scray system actuated on a high contain ent pressure sigr.al can operate continuously. Hcwever, the consequence of continued or lennthy operation may jecpardize the operation of equicment which would be desire-able if not necessary to minimize the effects of an event.
The continuatien of spray with operation of reactor coolant pumps could make the coolant pump motors inoperable. The continued o;:eration of the containcent spray systen will introduce several hundred thousand gallons of water into the containrent with potential flooding of useful if not necessary equipment.
Early consid-eration should be given to termination of the containment sprays if contain-ment pressure has returned to norral or below.
The use and availability of emergency feedwater is essential fcr all oper-ational events with the exception of a major loss of ccolant accident.
Feed would normally be provided to both steam generators.
Isolation of a stean generator nay be desireable if high activity is observed in steam releases.
If isolation of a steam generatcr is desireable (e.g. for stean generator tube leak), it is necessary to stco feedwater to the gen-erator to prevent lif ting of the safety valves and loss of feed supply.
In the event of a steam line break the steam cener3 tors should not be fed until the cocidown transient has been terminated and reactor coolant temperature is above 4000F V the boron concentration is sufficient to preclude criticality.
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RESPONSE TO QUE5 TION 6a (cont.)
The principal enqincered safety features signals which provide isolation functions are Containment isolation (CIAS) and mainsteam isolation (ftSIS).
The containraent isolation system is discussed in Item 3 of 79-063.
The mainstean isolation signal is actuated by high containment pressure or low steam generator pressure will isolate the main condenser and allow steam venting to the atnosphere via relief valves or dump valves. To minimize radioactivity releases, it is desireable to utilize the main condenser through the use of the turbine bypass system.
The decision to take this
~
action would be based on the availability of secondary systems and offsite
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RESPONSE TO CUESTION 6b:
Operating proc w es and traininj instructions have been reviewed to ensure that
. HPI system, automatically or manually actuated, will remain in ^, ration until the functions it is intended to perform (pressur
.ontrol, core cooling, reactivity control, inventory control) a rt -
..) for by other systems.
Item 6b of IE Bulletin No. 79-069 prc,. wes specific guidance on the use of the HP! system for inventory and pressure control. We concur with this guidance and have exoanded it to include the functions of core cooling and reactivity control.
Plant coerating procedures related to safeguard systems are guided by the following policy.
Regardless of the cause of system actuation, the operator does not alter the system operation until it is demonstratec that other systems and eouipment are providing the functions that the safeguard system is intended to perform.
Reactor coolant system parareters must be within specification, and the operator must demonstrate control of the non-safeguard systems.
Safeguard systems, when no longer required, are realigned for automated response.
Operating procedures for the HPI system have been revieled and revised as necessary to reflect the following specific guidance.
Guidance for each of the systems main functions of pressure control, core cooling, reactivity control, and inventory control is provic ed.
M w.wdp of t. Sh dtk > STAk i 4,,gytstet.isEtna M M N d]N (pressure contr:1) gy w9M; rw S
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The HPI system will remain in coe.at(ion for 20 r.inutes and until all RCS hot and cold temperatures are at least 50 degrees below the saturaticn temperature for the existing RCS pressure.
If 50 degrees of subcooling cannot be maintained af ter HPI cutoff, the HP! shall be reactivated.
In providing tne above, minimum Pressure-Temperature operating restrictions related to RCS integrity take precedence and shall not be violated by HPI system operation.
Ficure 1
is provided in the c::erating procedures to facilitate the implementation of the above criterion.
(core cooling)
The HPI system will remain in operation until the operator demonstrates that core cooling is provided by the steam generators or the shutdown cooling system.
Forced or natural circulation cooling of the core, with the steam generators acting as the heat sink, is demonstrated by having RCS and secondary system tem::eratures and pressures stable, within specification, functioning and controllable.
Forced circulation cooling is not considered in effect unless reactor coolant temperatures are essentially equal, are not fluctuating, and are either decreasing or not changing.
Natural circulation cooling is not considered in effect unless T -Tc = aT is less than SCCF and g
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7 RESPONSE TO OUESTION 6b (Cont.)
T is stable or decreasing.
There must be feed fisw to the steam gen-ehatorsandfeedficwoutofthesteamgenerators.
I (reactivity control)
The HPI system will remain in operation until the ocerator demonstrates that reactivity control ir provided by other systems.
Reactivity control is provided by other systems if core power is within s::ecification and stable, control rods are fully inserted er shown to be operable, T is ave stable and within scecification for coolant bcron concentration c 1 finali), if the normal CVCS baration path is cemonstrated to be o,2rable.
(inventory control)
The HPI system will re ain in operation until the ocerator c onstrates that RCS inventary control is provided by other systems.
Inve.ntory control is normally assessed by reading pressurizer level.
To demonstrate inventory control, the operator will rely on pressurizer level but will confirm the functioning of this instrumentation by ;:erforming the folicwing operations, a)
Initiate cha gir; pump ficw anc demonstrate that the pressurizer level instrumentation responds as expected.
(typical)
No. of charging pumps level change 0 inches /hr.
1
+ 21-inches /10 min.
2
+ 42-inches /10 min.
3
+ 63-inches /10 min.
b) Activate pressurizer heaters and demonstrate that the pressurizer pressure and temperature instrumentation respond as expected.
(typical)
+ 15 psi / min.
c) Activate pressurizer spray and demonstrate that pressurizer pressure instrumentation responds as expected.
(typical)
- 26 psi / min.
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-s-RESP 0:15E TO QUEST!0:16C:
If the possibility of bulk void generation in the RCS exists, ther RCPs should be in operation to ensure adequate system circulation to cemit the necess3ry core heat removal by the secondary plant systems. Accordingly the plant operators have been instructed to not trip more than one RCP 0
in ea".h loop if the core outlet coolant temperature is less than 20 F subcooled regardless of the status of the RCP rronitoring instrumentation.
If RCP's are off initially, one RCP in each loop shculd be started.
Some plant emergency procedures, for example, 5tcam Generator Tube Leak /
Rupture, require isolation of a steam generator and securing forced circulation to the affected steam generator.
The essential requirement of establishing adequate subccoling in the RCS takes crecedence and is to be confirmed prior to carrying out any such follow-up actions.
0 If the core outlet temperature is more than 20 F subccoled, it is permissi-ble to trip an;. or all RCPS if the ronitoring instrumentation and cump operating precedures indicate such action is desirable.
However, tne manual trip of all FCPs is to be conducted in an oroerly manner.
One RCP may be tripped in each loop concurrently.
The remaining RCP in each loop may not be tripped until the Procedure for Entering Natural Circulacion has been implemented.
If all of the RCPs trip automatically, the oceratcrc ore to take immediate steps to verify the existance of adequate natural circulation.
An indication of this condition is the ability of t'ie core aT to stabilize within minutes.
TAVG should be controllable by reans of the secondary plant heat sink.
The ability of the system to stacilize with RCS temperatures subcooled is to be clearly ascertained by closely monitoring pressurizer pressure anc core outlet temperatures.
Concurrently wi th this evaluation the cause of tne autonatic trip sncula be determined and corrected, if at all possible, to cermit restarting cne RCP in eacn icca regardless of the status of the RCP ronitorina instrumentation if inade;uate natural circulation is indicated.
ine remaining P"Ps snould rem,ain available for operation, if needed to provice core cooling.
$h Nerdon CL ed A a g chh h4 254 149
.. RESPONSE TO CUESTION 6C:
If the possibility of bulk void generation in the RCS exists, then at least one RCP should be in operation to ensure adequate system circulation to permit the necessary core heat removal by the secondary plant systems.
Accordingly the plant operators have been instructed to trip no more than U
two RCPs if the core outlet coolant temperature is less than 20 F subcoolec regardless of the status of the RCP mo. 'toring instrumentation.
If RCP's are of f initially, prf RCP should 5e st ted.
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If the core outlet temperature is more than 20 F subcooled, it is permissable to trip any or all RCPs if the monitoring instrumentation and pump operating procedures indicate such action is desirable. Hnwever, tne manJal trip of all RCPs is to be conductec in an orderly manner. Two RCPs may be tripped concurrently.
The remaining pr.P may not be tripped until the Procedure for Entering Natural Circulation has been implemented.
If all of the RCPs trip automatically, the operators 3re to take immediate steps to verify the existance of adeauate natural circulations. An indication of tnis condition is the ability of the core ai to stabilize within minutes. TAVG should be controllable by means of the secondary plant heat sink.
The ability of the system to stabilize with RCS temperatures subcooled is to be clearly ascertained by closely monitoring pressurizer pressure and core outlet temperatures.
Concurrently, with this evaluation the cause of the automatic trip should be determined and corrected, if at all possible, to permit restarting one RCP regardless of the status of the RCP monitoring instrumentation if inadequate natural circulation is indicated.
The remaining RCPs should remain available for operation, if needed, to provide core cooling.
Maine Yankee Section 6C
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RESP 0flSE TO QUESTI0fl 6d:
The following information identifies for the operator the diversity and redundancy of indications available to him as an aid in keeping Reactor Coolant System (RCS) inventory within acceptable limits.
During normal operation, the plant control systems are designed to keep plant parameters within narrow operating ranges.
When these operating ranges are exceeded, the plant safety systems are cesianed to autonatically keep plant parameters within acceotable linits for a sufficient period to allow for timely required operator acticn (e.g., from not less than 10 minutes for initiating auxiliary feed-water, to well over an hour to suitch to shutdown cooling).
Tnese time periods are adequate for the operator 1) to assess the diverst and redundant indications described belcw and 2) to verify the r.ecessity for operator action before that action is taker..
The purpose of raintainin; sufficient RCS inventory is to assure adequate cooling capab.lity for the core and oressure contr:1 of the system.
The major fac; ors which impact RCS inventory are plant responses which have the potential for forming voids in the RCS.
The response to Question 2 addresses the diagnosis, prevention, and mitigation of voids in the RCS.
In su=ary, the prinary autenatic actions to prevent and mitigate the formation of voids in the RCS are based upcn redundant pressurizer pressure indi:ation which initiate a reactor trip and SIAS on respective low pressurizer pressure setooints.
Additional indications to the operator for diagnosing possible void formation in the RCS include:
1.
Cold leg and hot leg temperatures near the saturation temperature of the RCS pressure.
2.
Changes in reactor coolant pump temperatures, currents or other parameters which indicate pump cavitation.
3.
Decreases in indicated steam generator ap.
4.
The absence of or unexplained changes in pressurizer level indication.
Additional indications to the operator for diagnosing possible leakage from the RCS which could lead to void formation in the RCS include:
1.
Increases in containm temperature and pressure.
2.
Increase or filling of the containment sump.
3.
Increase in indicated stean generator level not due to changes in feedwater or steam ficw rates.
4 Increase in the pressure and temperature of the *anks into which the RCS relief valves discharge.
4
x RESP 0tiSE TO CUESTI0rl 6d (Cont.)
5.
Valve alignments, pipe ficw rates, or pipe temperatures which indicate unintended RCS fluid flow from pipes connected to the RCS.
If pressurizer level is indicated, it shculd first be used as a diagnostic ef fort to determine if programed changes in pressurizer level and/or pressure show the formation or absence of void in the rest of the RCS, before it is used as an indication of inventory control (see section 6b).
When the operator has diagnosed the potential or actual for :ation of RCS voids, he should follow the prevention or mitigaticn actions identified in the response to Question 2.
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RESPONSE TO CUESTICN #11 In response to this question, capabilities are discusseo to 1.
Vent cccumulated hydrogen and/or other non condensible gases frcm the reactor coolant system I.RCS) 2.
degassify the reactor coolant i
I process the waste gas after removal from the reactor coolant 3.
i Possible additional equipment is also presented for consideration to more efficient 1/ venc and process the waste gas.
Proscr regard must be given to the additional equipment for the full consequences of inadvertent operatico.
j The guidelines are provided assuming that all coer=tiens are to be perfcrmed cutside the containment. Additional cperational flexibility may be available in scme situaticns if entry into the contairment is allowed.
Some tonsient or accident scenaries may also include the presence of signifkant quantities of radicactive fission gases along with the hydrogen in the reactor coolant.
The presence of the fission gases coJld limit cont 3inment entry and, in turn, the methods used to vent and process the hydrocen gas.
1.
Existing Systers A.
Venting the PCS - There are several potenti?.1 high points within the reactor coolant systen where non-condensiDie gases may accumulate.
These include the following:
a) Pressurizer b)
Reactor ccciant cu o seal cavities c) Shutdown cooling pi::ing at the reacter coolant system interface.
(applies to ftillstone II and St. Lucie el cnly) d ',
Reactor vessel head above the hot leg no::les.
e) Top of the steam generator U-tubes.
In order to remove accumulated gases from varicus high points, the following method are available:
a) The pressurizer steam space may be vented through the primary sample sy~ tem steaa space sample line.
This line penetrates the conLinment, the fiuid is cccled, and discharged to the volume control tank (VCT) and/or other reactor coolant quality waste tanks such as, the flash tank or equipment drain tank.
Each of these tanks vent to the gas waste management system (G'JIS) for storage of the gas, b) The reactor coolant cump seal cavity may be vented through the controlled bleedoff for each reactor ccolant purp.
This ficw is typically routed to the volume control tank or the reactor drain tank. These tanks can be vented to the G'JIS.
RESP 0ilSE TO 00ESTI0ft 411 (cont.)
c) The RCS pressurizer surge line and hot leg sample system connections may be used to partially vent the loops and reactor vessel head. The path for the venting is the same as that described for the pressurizer steam space in (a) above.
d) The reactor vessel head may be partially vented through the pressurizer as described in (a) or (c) above if the RCS fluid level pernits.
Ccepiete removal of gas accumulated in the reactor vessel head requires redesolving it in the ccolant.
e) Pressurizer motor operated relief valves (all plants except Arkansas) can be used to vent gas in the pressurizer (and through the pressurizer other portions of the RCS) to the quench (reactor drain) tank. Procedures to use these valves must recognize the following consequences:
(1) Operation of one relief valve will result in an RCS pressure (7 a.
decrease en the order of 900 osi/ minute.
Caution %d pwS approach is then require:1 since failure to close the valve in a timely fashicn or failure of the relief valve to close could lead to excessive depressurization and, in turn, to additional gas accumulation.
(2) Continued operation of these valves could result in exceeding quench tank capacity causing the rupture disc to burst. RCS fluid then has a direct path to the containment.
Confirmation and establishment of adequate reactor coolant makeup is required.
)
B.
Degassina the RCS - The existing systems typically have the folicwing capabilities to cegass dissolve hydrogen and/or non-condensible gases from the reactor coolant:
1.
Degassing via the pressurizer: With this method all pressurizer heaters are energized, spray ficw is adjusted to maintain a constant plant pressure, and the pressurizer is vented via the existing steam space primary semple connection to the VCT or other location. An example of the time required to degass the coolant is given on Figurc I (the time is dependent on the initial H2 concentra tion).
Table 1 provides examples for specific plant degas half-lifes of the reactor coolant.
2.
Degassing the RCS via the CVCS: With this method the reactor coolant passes through the letdeun line to the Volume Ccnvol Tank (VCT).
The VCT is vented directly to the GWMS.
Figure 2 indicates typical plaat capabilities for various letdcun rates.
Table 2 provides specific plant cegas half-life of the reactor ccolant.
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RESP 0MSE TO CUESTION E11 (cent.)
C.
!!aste Gas Processinn - The plant gaseous waste management systems and/or hycrogen recomoiner may be used to process or store the hydrogen and/or ncn-condensible gases collected frcm the RCS or containment 1.
Caseous waste management systens The approximate storage capacitics of the clant gasecus waste management systems (GWMS) are shown in Table 3.
The capacities of these systws are based uocn holding gases from normal sources for a ceriod of time such as 30, 60, or 90 days, before controlled release to the environment and represent the maximum alle. table volume of the vented and/or degassed H2 that could be stored outside the containrent.
The storage volume may, howevec, be limited by the technical specifications fcr the maximum curies of radioactive ncble gases.
The basis for this technical specification limit is typically one degassed reactor coolant volume of gas with the source of activity consistent with 1% failed fuel.
2.
Containment Hydrogen Reccmbiners The hydrogen in the containment building may, for all plants except Maine Yankee ard Fcrt Calhoun, be removed by use of the installed hydrogen recombiner. Table 4 contains scecific information regarding each plant rectroiner capabilities.
No special shielding v.0ulo be required for these plants which have recombiners since they are located insice ccntainment.
The hydrogen removal capability for those plants with recombiner can be excressed in a purification as a half-life cf acproximately four (4) days.
Portable recombiners (external to the containment) may be used on Maine Yankee and Fort Calhoun.
II.
Consideration of Additional Ecuicrent To enhance plant cacability for venting and storing abnormally large quantities of hydrogen and fission product gase, the following additional equipment shculd be considered:
Proper regard for the consequences of inadvertent cperation of any additional equipment should be considered also.
A.
RCS Hich Point Ventina 1.
Figure 3 shows a possible schere for venting the reactor vessel head of large quantities of free hydrogen gas.
This system provides the capability of venting the head to either the pressurizer or directly to the cuench tank.
Provisicns are also made to augment the vent rates for the a dsurizer steam space sampling line and the reactor drain / quench tank vent to G'.015.
Portions of this systen aculd have to be ASME Code Class 1 and Seismic Category I.
The system must also be designed to minimize interference with the reactor vessel head removal during refueling.
L
RESPONSE TO C'JES~IC." ell (cent.)
The consequences of inadvertent operation of such a system must be determined.
The design must incorporate features which minimize the consequences.
2.
Controlled venting of accurulated gases in the pressurizer to the quench tank could b9 facilitated with a vent patn from the pressuri:er spray line.
This will permit Figure 3 illustrates this mothed.
This rethod provices a controlled pathuay, which is remotely ccerable, for the collection of gases within contain-en and also ;ernits a scheduled venting to the G'."5, if desirable.
3.
The shutdo..n ccoling system sucticn lines fer Millstone II and St. Lucie i ray recuire high reint vents :: ;reclude gas binding of the decay beat reToval system since tho picing elevaticn is greater than the hot leg elevation.
This venting could be vented to the pressurizer er tne reactor drain /cuench tar' A possible ret'.cd for venting tne piping is shc..n on Figure :.
4.
To prcvide additional storage ca:acity for vented and/cr cegassed H2 and fissicn gases, the centainrent itsel' is a cossible alternative.
RCS vents to the containnent wcuid occur via the quench ta9 rupture disc cnly if the ructure cisc is brcken fra additicnal e;uic~ent necessary).
Tne VCT can be vented to the containrent instead of the G'.l"S by the ad:ition of tem:trarv and/cr perrar.ent piping as shcun in Figure 5.
For ar.y venting operation, consideraticn must be given to:
a) The effects cf rapid decressurizaticn with saturation conditions and the potential gereraticn cf gas, b) ensuring that adecuate raketo to the FIS is available to compensate fcr any fluid lost during venting.
c) fer venting of the react:r coolant system cr "CT to the containment, due consideraticn must be given for retnccs to adecuantely cix the hydrogen g2s.
Generally, the hydrogen mixing is acco-plished by tha containment spray system, the containment fan c0clers, and containmer.: air circulatcrs wnich ;ermit convective mixing and prevents entra: rent.
Eacn plant must evalua.e tne discharge points of vented hydrogen (e.c. ructure disc area of the quench tank) to prevent local ccncentration of hydrogen.
254 156 e
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TABLE 1 Pressurizer Decassina Estimated Es tim.a ted Half Life Half Life Plant flame at 2000 PSIA at 1000 PSIA Arkansas (AP&L) 8.9 hrs.
8.1 hrs.
Baltimore (Calvert Cliffs #152) 10.3 hrs.
9.4 hrs.
Florida P&L (St. Lucie el) 10.3 hrs.
9.4 hrs.
(Wss" 50 #/hr)
Paine Yankee 17.4 hrs.
8.3 hrs.
t;ortheast Utilities (Millstone 42) 10.3 hrs.
9.4 hrs.
Omaha (Fort Calhoun) 6.0 hrs.
5.5 hrs.
Consumers Power (Palisades) 9.0 hrs.
8.2 hrs.
TABLE 2 Volume Control Tank n gassina e
Estimated P1 ant fiame Half Life (hoursi Letdown Flow 40 GP!1 128 GPQ Arkansas 28.4 8.9 Cal. Cliffs, St. Lucie, 27.4 8.5 Mills tone 23.6 7.4 Palisades 23.9 7.5 Maine Yankee 26.6 8.3 Omaha 16.0 5.0 254 157
TABLE 3 GWM5 Storage Canabilities Estimated Plant Storace Cacacitv Millstone 2 15,100 SCF Haine Yankee 3,800 SCF St. Lucie 1 4,200 SCF Omaha 12,270 SCF I
Arkansaa 22,800 SCF Palisades 8,775 SrF Calvert Cliffs 1 20,130 SCF Calvert Cliffs 2 20,130 SCF 1
k 1
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ye 6 -
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TABLE 4 Containment flydrogen Recombiners l'
Post LOCA*
Containment flydrogen Radiation Shielding Post-LOCA Integrated Cose**
P,l a n t_
Recombiner for Cembiner for Equioment quall'ication_
Arkansas 2 Yes flot required 3.3 x 10 rads 8
Calvert Cliffs 1&2 Yes flot required 1 x 10 rads 0
St. Lucie 1 Yes flot requir ad 1 x 10 rads Iiaine Yankee flone flot Applicable I
8 Millstone 2 Yes flot required 1 x 10 rads Omaha flone flot Applicable Palisades Yes flot required flot availaide i
In-Containment Thermal Recombiner
- All recombiner units qualified to 2x10 rads Flow rate is 100 - 120 scfm N
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