ML20125A649
| ML20125A649 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 10/21/1991 |
| From: | NORTHEAST UTILITIES |
| To: | |
| References | |
| NUDOCS 9212090007 | |
| Download: ML20125A649 (14) | |
Text
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00T-21-91 11011 15:22 GEft FACIL LICE?Gil10 FAX 110. 2036655896 P.02 4!s Figure 7.1
.r_oo u m n a m w n nurre 1splosentation Date 10/21/91 JC0f _ 2-91 1 Estpiration Date
~Fnd of evela 11 I
1 Subject Main stean Line_Ereak inside containreant Unit _ (
Date _ 10/19/91 Rettuasted ny (NUP00 Director) _ 3.s. Kennan 4IM
/ O/w/9s
'Date Concurronce p.ager,AsaponsibleOr5anisation nkh/
M 7AM2 C.uar.u.,
Date ger, Responsible'0rganisation MoA h
16/m Com urem e:
Date
- Ranager, bis Or5anisation C.nemen.,,' '
O wum Dato Manager, Generation Facilities 1.icensing concurrence:
Date PORC Heating i Approva11 Date NUPOC Director
- JC0 is implenented upon final approval of the NUPOC Director.
Lower Partion To Be Used For Closure of JC0 Date Manager, Responsible organisation i
Conc ur.ne.
Date Manager, Responsible Organization l
Date I
Concurrence Manaler, Responsible organisation Date Manager, Generation Facilities Licensing Concurroncet
, Date
'PORC Meeting (
Approdal:
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Data RUPOC Director OriSinal CFL File; NUP0C Director; Technical Manager (s);
Distribution:
Manager, CFLt NKl/5WR3 Nuclear Records L
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Millstone Unit No. 2 k
Justification for Continued Operation #2 91-1 l
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I Main Steam line Break Inside Containment i
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g TABLE OF CONTENTS 1
Intfoduction..........................
1
Pilnt'ConditionDiscovered 1
Identification of Potential Adverse Effects on Safety 2
Justification for Continued Operation..............
2 1.
Procedure Changes and Operator Training........
3 Analysis of Operator Action i
4 111.
Additional Compensatory Action 6
Additional Considerations....................
7 1.
Containment Structural Evaluation for > 54 psig....
7 EEQ Evaluation for > 54 psig and > 286'F 111.
Independence of Feedwater Control Valve and 8
Feedwater Block Valve 9
Conclusion.........................
10 References
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_a GEN. FOIL LICENS!!G FAX n 203EEE5596
?. 05 OCT-2H1 !G 15:24 l IKTR000CT10ff This document provides the justification for Millstone Unit No. 2 to continue to'op'erate following the identif1;ation that the containment structure is outside of its design basis for certain postulated main steam line breaks (M$LBs) inside of containment.
This hypothetical condition is based u)on strict licensing requirements which include, but are not limited to, hypotie.
sizing a guillotine HSLB from 1007, power with the feedwater control valve failing in the open position and no operator intervention for the first 10 minutes of the transient.
This condition was analyzed and resulted in a ct.lculated peak containment pressure and temperature in excess of 90 psig and 400'F respectively. This is in contrast to the Millstone Unit No. 2 technicalis specification Section 5.2.2 which states that the reactor containrent designed and shall be maintained for a maximum internal prossure of 54 psig e.nd a temperature of 289'F.
PLANT CONDITION DISCOVERED As a result of the Steam Generator Replacement Project, new MSLB analyses were being performed to assess the impact of the new steam generators on the in performing these analyses, it was discovered that containment response, the conditions assumed for the current design basis containment response for a In particular, it was identified that a full power MSLB are not limiting.
case with a smaller break size and off site power available will result in a higher peak containment pressure and temperature.
This condition was evaluated under REF 9148 (MP2) and was determined to be reportable as a condition outside the design basis of the plant.
REF 9148 and should be referenced for a has been attached to this JC0 (Attachment 1) ding previous analysis of MSLBs detailed discussion of the background surroun inside containment.
REF 9148 raises the possibility that the analyzed MSLB location and assumed single f ailure are nonconservative.
Additional analyses have confirmed that the current design basis analysis assumptions with respect to power level, break size and single ' failure are not Hmiting.
For the case of a 1.0 ft8 MSLB at full power with a single failure of the feedwater centrol valve closure signal upon turbine trip, the peak containment pressure is predicted to exceed the containment design pressure of 54 psig.
Further, a peak con-tainment pressure in excess of 90 psig was predicted for a double ended (6 ftt) MSLB at full power with the single failure of the feedwater control l
valve to close.
l Zero power and other single failures were evaluated and shown to give an acceptable containment response.
IDENTIFICATION OF POTENTIAL ADVERSE EFFEcT ON Sjffly In REF 9148, the impact of the more limiting conditions for M5LBs were evaluated.
In sumary, it was found that a MSLB at hot full power was (nore limitin than zero power for containment response.
In addition, it wa.
"'g"6Wthreak size _andJA1_11miting single failure were
OCT-2Mi G 15:N SER MIL LICEEN3 F E E. 2:2 E EE36 F ;E
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. for the single failure in which the feedwater control valve was operable but a failure in the turbine trip closure signal was assumed, the limiting break For these conditions peak containment pressure was predict-size was 1.0 ft'.
Since the ed to be 68 psig with a peak containment air temperature of 419'F.the p desigt pressure is 54 psig,le the air temperature exceeded the containment the design pressure.
Whi design temperature, it was concluded that because of condensation effects and the containment structure did not exceed the containment design thermal lag, ture, further, based upon engineering judgment tLt the short-basis tempera it was term temperature peak will be offset by condensation and themal lag,fication concluded that the equipment on the llectrical Environmental Quali (EEQ) Haster List would continue to be operable.
For the failure of the feedwater control valve to close on demand, the limit-ing break size is a double ended break, for this case, it was estimated that An the peak containment pressure would be on the order of 70 to 60 psig.2 additional parametric study (Attachment (ABB) shows that with credit for operator action at ten minutes to terminate feedwater, the predicted peak containment pressure is 93 psig and the peak In this case, the long tem tempera-containment air temperature was 427'F.
ture could also exceed the current qualification temperature of EEQ equipment and thermal lag could not be credited to offset the high temperature.
In spite of exceeding the containment design pressure for a design basis "5LB, It it is expected that there will be no significant radioactivity re the limits provided in 10CFR100.
JUSilflCATION FOR CONTINUr0 OPERATION 1.
Procedural Chances and Ooerator Training A dedicated licensed operator shall be stationed at CO 5 and will perform the following actions after a reactor trip:
OBSERVE a valid reactor trip.
a.
CLOSE both main feedwater block valves, 2 fW 42A and 2 FW 42B.
b.
tiqu The following step is a contingency action to prevent containment overpressurization, although due to time limitations, no credit l
has been taken for this action in the main steam line break l
l evaluation, If either main feedwater block valve f ails to close and the asso valve fails to close, then SECURE all c.
ciated feedwater control l
condensate pumps.
The dedicated operator shall have no concurrent duties with the main feedwater control valve isolation following a reactor trip.
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OCT-2tf91 tion,15:25
_ GEN. FACIL LICENSING l FAX NO.l2036655899-P. 07 l
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The above _ operator actions are nasont.ble and can be performed in the time frame evaluated in the main steam line break analysist The dedicated operator will be stationed - at the feedwater c..itrol a.
I station, and The required operator actions are simple and require no diagnosis.
1-b.
The above operator actions are cons 4 stent with the existing Standard Post The existing contingency for excessive TripActionprocedure(EOP2525).feedwater flow is to manually close th Accomplishing this existing contingency action early does not effect any other preceding actions since th9se actions are completely independent of the main feedwater action step.
will be provided to each shift via a training guide and a Trainin briefin by the Shift Supervisor concerning the dedicated operator duties and Act ons.
- 11. Analvsjs of Goerator_. Action Immediate corrective action (as detailed above) involves positioning an operator at the main control board dedica -
' ' closing the feedwater Valve-8'ock Valve) upon block valve (also titled the Feedwater Regul into procedures.and the a reactor trip, This will be incorporat.u Based dedicated operators will be provided training-for this action.-
upon demonstr?tions on the Millstone Unit No. 2 specific simulator, it is closure - cac. be initiation of feedwater. block. valve Since the stroke concluded that reliably. accomplished in less than fifteen seconds. time for valve within 25 seconds i.fter a reactor trip car, be credited as a back up_
to a failure of the clonte of the feedwater control valve.
Parametric study #5 of Attachment 2 includes : case in-which feedwater isolation was assumed within 25 seconds after a reactor-trip. This casein addit is a full. power -case with --# break size: of _- 6.0 ft8. -
esse-assumes that the initial. containment pressure is 2.1.psig and 5%.
feedwater flow through the bypass valve is= maintained.throughout t For this case, a - peak-containment Since this is-- tho' limiting case for -the-failure of the -
accident.
back up-to predicted.
feedwater control valve, credit for this operator actin % al the feedwater control valve failure will maintain containment pressure below the containment design pressure.
Credit for this-olerator action will also provide ' adequate mitigation for.
the failure of tm turbine trip signal-to close the feedwater-- control-For this sinDie. failure, the 1 miting condition is a small MSLB.
' Assuming a : fdlure of the turbine trip signal to close: the -feedwater -
. valve.
control valve,. the Main Steam Isolation Signal will not be generated for 140 seconds,- thus :feedwater isolation (closure of the fwdwater control t
' trip valve) will not occur-until 150 seconds However :since a reac or l
.10Jredicted to occur:at 10 accords, crediting. operator action-to c ose
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d H 1 MON 15:26 GEN. FACIL LICENSING FAX NO. 2036655%6 P.08 i
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4 that for this case feedwater would be isolated within 35 seconds of the MSLB event. This would result in the peak containment pressur: eemaining below the design pressure.
lif. [lbditional.Camens_atino FEters i
A.
Evaluation of probability of OccurrntJt
.nt sequence, a pro 5abilistic To evaluate the likelihood of %e n~,.e overall 1robability of a main risk assessr nt wn performd.
steam line bh.* inside conteinment with a faslure of the feedwater The control valve i. conservatively estimated to be 10 s per year.
short-term corrective measures (remote manual closure of the feedwater block valves by a dedicated operator) reduces the proba-bility of this event (masn steam line 'feak concurrent with failure to isolate feedwater) to about C per year.
Alternative corrective measures involving new automatic functiont (as discussed inSectionC)reducesthisprobabilitytoabout10'7peryear.
It should be noted that the overall probability discussed above applies only to n scenario that could lead to exceeding the contain-ment design pressure and does not consider the angnitude of the pressure res>onse, nor does it include the likelihood of a subse-quent loss 0 containment integrity. These types of events would be 3
of an even smaller probability.
B.
Feedwater Block Valve Caoabilities The operator action discussed above results in closure of both feedwater block valves (MOV 2-FW 42A/B).
These calves are non-QA, are not sowered from a vital power source, and are not currently in As a result of this the Nort1 east Utilities (NU) GL 89-10 progra*
ich has been applied concern, the NU GL 89 10 MOV Test Program ap, to these valves and the results indicate that an analytical approach field testing provides sufficient confidence that the ba-Led b alve will operate s::ccessfully and can thus be credited to at.tuator The analytical and testing minimize the consequences of a MSLB.
efforts undertaken to address the capabilities of the MOVs (2-FW-42A and B) is described below.
1.
NOV Fluid Conditions indicates that two MSL Break cases The new MSLB analysis provide bounding fluid conditions for the MOV's (2-FW-42 A and Specifically; i) large MSLBs resulting in a feed pump trip B).
prior to MOV actuation and ii) a smaller M$lB resulttng in no feed pump trip.
Fluid conditions resulting from these cases are discussed below.
l The large MSLB case results in a feedwater pump trip prior to MOV-2 FW 42A and B closure, due to a main steam line isolation signal. Thus, MOV upstream conditions at valve closure will be
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was determined to be 277 psig. Thus, MOV delta P = 538 psig -
277 psig = 261 psid for the large MSL break case.
The small break case does not result in a feed pump trip, however, steam generator backpressure is significantly higher.
HOV upstream conditions at MOV closure considers both conden-and 1022 psig 538 psig pressurest sate pump and feedpump MOV downstream ?ressure at MOV respectively (Reference 2). closure is conservatively assumed to be (low SG pressure, MSI setpoint). Thus. HOV Delta P = 538 psig
+ 1022 psig 465 psig = 1095 psid (for the small MSL break case).
The worst case fluid conditions resulting from a MSLB for MOVs 2 FW-42 A/B is 1095 psid differential pressure.
2.
M_ OV Thrust Calculations Valve actuator thrust evaluations undertaken to address this JC0 show that MOV's 2-FW-42A and B are capable of closure against worst case main steam line break fluid conditions (1095 psid as discussed above).
The original design require-This ments were used to determine the required stem thrust.89-078 073-EM will be documented in NUSCO QA calculation No.
In addition, valve structural integrity was evaluated with respect to the thrust requirements.
The analytical Rev. O.
approach discussed tbove is consistent with the NU GL 89-10 MOV It is noted that the original specification for 2 FW 42A and B required a valve closure at 1600 psid test program.
(Reference 3).
3.
MOV Stroke Time MOV's 2 FW 42A and B exhibit acceptable stroke speed character-istics to satisfy the 10 second stroke time required by this Specifically, i) the original specification (Reference 3)
JCO.
the motor required a maxinum stroke time of 10 seconds, ii) troke time operator vendor data (from LimitorQue) determined sto and iii) Millstone Unit No. 2 plant start-up tests measured the stroke time to be g seconds.
4.
MOV Field Testino Prior to mode 1 operation, valve diagnostic testing (VOTES) in accordante with established procedures develosed under the NU GL 89-10 MOV Test Program will be performec to confirm the Portions of this testing analytical approach described above.
include direct measurement of actuator thrust via strain gages and stroke time measurements.
l Analytical evaluations backed up by the field testing described
- - ICvnHaidm10ufficient confidence that MOV's 2-FW-42A and B
OCT,-21-91 MON 15:27 GDL FA0iL LICENSING FAX NO. 2035655996 P.10
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. will close in the event of feed cont.a1 valve failure during the spectrum of main steam line breaks considered by this JCO.
- C.
Alternative Corrective Action Modifications are contemplated during the current cycle that would i
preclude the need to credit operator action.
These modifications may include the following:
1.
Installation of a safety grade trip on the feedwater block valves 2 FW 42A and 2-FW.42B which will automatically close the valves on receipt of a high containment pressure signal.
The isolation signa'l shall originate from the Engineered Safety Actuation System (ESAS).
The proposed safety grade isolation signal for each motor operated valve shall not be comen to the safety grade isolation sidal for its in series control valve.
.lfestalling a safety grade trip to the feednter pumps, conden, 2) sate cumps and heater drains tank pump on receipt of a high conta' nment pressure signal. This option may not be as desir-able as Option 1.
for the small MSLB scenario where failure of the nonsafety 3) grade turbine trip signal will not result in the automatic closure of the feedwater control valves, the addition of another safety grade trip may remove the need for operator action given the f ailure of the nonsafety grade turbine trip signal. Due to the length of time required to generate a main l
steam isolation signal for the small MSLB scenario, reliance is placed on a nonsafety grade closure signal..To remove the reliance on this nonsafety grade trip signal, it may be desirable to add a safety grade trip signal on receipt of a high containment pressure signal.
additional corrective actions are also being consid-Longer term, cred.
These actions may include replacement of the existing feedwater block valves with QA Category lE Safety Related valves and controls which will automatically close upon receipt of, most likely, a high containment pressure signal.
Also, the valves must l
be properly sized to close against the expected differential pres-suras as well as fast acting to limit the amount of feedwater to the steam generator, ADDITIONAL CONSIDERATIONS The above demonstrates that the containment and associated equipment is The actions specified ensure that during the most limiting MSLB oserable.
that containment pressure does not exceed 54 psig.
However, as an aided seasure of confidence for the protection of the public health and safety and thus the implementation of the JCO, the following analysis was conducted which provides reasonable assurance that even if containment pressure were to exceed its design basis of 54 psig, the containment and associated equipment would still be Gapable of perfonning their intended safety functions.
001-21-91.110f1 15:28 GEN. FMIL LICEllSillG FAX 110. 203%55890 P. !!
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1.
Gonhjngnt Structural Evaluttion foLJ_jii_pMg The Millstone Unit No. 2 containment consists of a prostressed, rein-forced concrete cylinder and dome connected to and supperted by a massive reinforced concrete foundation slab.
The containment was designed for an internal pressure of 54 psig, and was tested to 62 psig during the structural integrity test.
The working stress design method was used to
' design the containment structure for various load cases, including the case of a design pressure of 54 psig.
The containment structure wi.s checked for factored loads and load combinations, including the case with a 1.5 load factor on the design pressure, which corresponds to 81 psig.
The code requires that " strength be adequate to support the factored loads and that serviceability of the structure at the service load level beassured,"(ACI-31871CommentarySection9.1.1).
The ultimate capacity of containments has been studied and documented by many sources recently.
In general, the anticipated ultimate capacity of a containment structure has been found to be 2 to 2.5 timas design aressure. NUREG 1150 entitled " Severe Accident Risks: An Assessment for Fivo U.S. Nuclear Power Plants" studies the ultihlate ca)acity of typical containment structures.
Included in this study was Pion, which is a prestressed containment, with a design pressure of 47 psig.
The evalua-tion determined that a lower bound on the ultimate capacity was around 100psig(afactorof2).
These detailed studies have taken into account material strengths being higher than assumed, code allowables being conservative, as well as a detailed evaluation of structural behavior during beyond design basis events. A similar detailed study has not been performed for Millstone Unit No. 2, but the same factors which contribute to a lower bound ultirrste capacity of 2 - 2.5 times design exist in the Hillstone Unit No. 2 containment structure, This discussion on studies relative to ultimate capacity further substantiates that the containment can support the factored load case and beyond.
These postulated load cases are beyond the design basis of the containment structure, but within the overall load carrying capability of the structure.
11.
EE0 Evglpation for > 5LasigJnd > 289'E The new postulated main steam line break includes an initial 70-second duration peak of 412*f at 54 psig.
The corresponding saturated tempera-ture during this superheated temperature peak is 302'F.
The ECQ Master List equipment are cetermined to be qualifiable under these conditions by analysis which shows that equipment qualification tests are more severe than the temperature that the components will experience during the postulated steam line break.
The 412'F superheated steam will not heat equipment beyond.their quali-fled temperature because of the short-term period at the superheated conditions.
The tempe ature a component will reach is dependent upon the rate of heat transfer from the envirnnent to the surface of the compo-nent.
The rate of heat transfer will be dependent upon the saturated temperature of the environment for the short time that the conditions remain superheated.
A condensate la,1r will fctm on all component surf aces during the initial phase of the accident.
The temperature of this condensate layer will be less than or equal to the saturated
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, terrperaturc, therefore the temperature of the component will remain less than the saturated temperature while this condensate layer exists.
removal.
The motors The CAR fans are important for containment heat ascociated with these fans are enclosed, and have an internal cooling system. These motors were tested up to 80 psig with an associated saturated temperature of 324*F.
With an internal cooling system they
'would not be directly exposed to a postulated higher temperature.
Another item of particular interest for this event are the containment These have been tested to 70 psig which corresponds to a penetrations.
316'F saturation temperature.
A review of EEQ equipment in containment shows that all qualification testing was performed at a peak pressure of at least 69 psig with dura-Since this pressure and its' tions in the order of several minutes.
associated saturated temper 4ture of 315'F is more severe than the postu-lated peak saturated temperature of the MSLB, we concluded that all EEQ Master List ecuipment in containment could be qualifi64 for containment pressures up ;o 69 psig, and that this equipment is operable for the postulated transient, and the ensuing post accident operating time.
111. Lndenendence of fe@ater Control Valve and feedwater Block V To provide additional assurance on the reliability of using either the feedwater block valve or the feedwater control valve for each respective steam generator, an evaluation was conducted to determine the degree of electrical separation between the valves.
The following describes the electrical separation between the valves.
Each steam generator has a feedwater block valve and feedwater control for steem valve in series on the secondary side of the steam generator, generator "1"; feedwater block valve FW 42A is fed from HCC B12 Feedwater control tod from 4ky bus 24A via a lad center transformer.
valve FW-51A is powered from AC Fanel IAC-1 which bus 24C.
The common electrical power source is the NSST through breaker 52-A;02.
bus 24A for these two valves.
For stear generator "2", the feedwater block valve is FW-42 feedwater control valve is FW-518.
l from MCC 21 which is connected tc 4ky bus 24B via a load center trans-AC pan Feedwater control valve FW-518 is powered frez which in turn is fed from HCC 61 which is then powered. a a load former.
bus 240.
transformer from the 4kV this bus is connected to the NSST through breaker 52-A206.
The control power for both feedwater block valves is from' the control A loss of Nrmer located in each MCC for the individual valve.
feedwater block valve does not affect the tranz for an individual In addition, the control power for l
powe'
. dant feedwater control valve.
This scheme is powered from a re!
the turbine trip scheme was reviewed.
separate 24 vde battery for the turbine generator EHC control system.
l
4 Cy21991110t1115:30 GEN. FACIL LICENSING FAX NO. 2036655896 P.13
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electrical cables. for FW-42A - and - 551A are in The routing of the separate trays -and conduits except a single tray located in:the cable
.The electrical cables for -FW-42B and-FW-518 share three vault room.
trays located in the turbine building, tg.ThIdrobanility of ~a' consnon. mode failure of the cablesilocated in^ the-same cable tray during the short time interval of concern with-the hypothetical main steam line break is approximately s,10 while the probability of the failure of a 4ky bus is less than 10 The separation of the electrical power for control-and operation of tne valves to their respective 4ky buses and the probability of en event.
rendering either 4ky bus unavailable is low. ' Although the routing of the-cables for the redundant valves do share the same cable traysnin-some locations, the probability of a single event affecting all the c' ables.in the tray is also sufficiently low.
CONCWSION Based - upon-the information provided above, there is - a - reasonable : assurance that with the actHns of a~ dedicated operator or implementation-of one or more of the alternative corrective-actions, the containment pressure will remain If one-or below the design basis value during a main-steam line break event.
both of the first-two options-identified on page 6 is implemented, ;there would;
'be no further need for the dedicated operator.
Therefore, it 'is. our-determination that' the continued operation of. Millstone Unit -No. 2 for. the:
reminder of cycle 11 will not involve any undue risk to-the health and safety-of the public.
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OCijl-91;i10N 15:31 GEN. FACIL LICENSING l FeX.t10,-2038655896-PJ 14 -
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EfilEES 1.
NUSCO QA calculation 90 RPS 736 EM Rev. 5
/" 3.I h chtel specification 7604 M-2706, Rev 5.
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