ML19209C043
| ML19209C043 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 10/09/1979 |
| From: | YANKEE ATOMIC ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML19209C039 | List: |
| References | |
| TASK-07-04, TASK-7-4, TASK-RR NUDOCS 7910110431 | |
| Download: ML19209C043 (29) | |
Text
Docket No.'50-29 REPORT ON FAILURE EFFECTS OF EQUIPMENT NOT QUALIFIED FOR ACCIDENT llELATED ENVIRONMENTS Yankee Atomic Electric Company Rowe, Massachusetts October 1979
'f 7 91 01 10 64 1130 144
CONTENTS
-S E C_T_I O_N.
TITLE PAGF I.
Introduction 2
II.
License Background 3
III.
Phase-I 5
a-Procedure 5
b.
Areas of Concern 7
1.
P: ?ssurizer PORV Control 7
2.
Main Feed Control 17 3.
Automa tic Rod Control m
4.
Steam Generator Vent and Steam Dump System 23 c.
Summary of Results 25 IV.
Phase I I.
Objective 25 a.
b.
Results to Date 25 11JJ 145
I.
INTRODUCTION This report presents the results of work performed at Yankee Nuclear Power Station to determine the impact on the safety of the plant of failures of equipment subject to, but not qualified for, accident related environments.
This work is being conducted in two phases as follows:
Phase-I -
This phase surveys the specific systems identified in IE Information Notice 79-22.
Phase I has been completed and the survey results as well as a description of the survey procedure is included in this report.
Phase _-II - This phase includes further studies to identify other germane systems, if any, that are not specifically identified in IE Information Notice 79-22.
In addition, this phase addresses the impact of erroneous information displayed to plant operators that may result from failures to equipment not qualified for accident related environments.
The objectives of this phase and results to date are inc1'uded in this report.
1130 146..
II.
LICENSE BACKGROUND The NRC letter from the Director of Nuclear Reactor Regulation of September 17, 1979 requests, within 20 days, information of affirva.on which will enable the staff to determine, in light of the concerns discussed in that letter, whether or not Yankee's operating license should be modified, suspended or revoked.
The safety analysis for Yankee Nuclear Power Station may not exist in the same format as those recently licensed nuclear plants.
The following brief is intended to assist the staff with their determination by describing the evolution of the Yankee license.
The Yankee Nuclear. Power Station (YR), license No. DPR 3, submitted a Final Hazards Summary Report (FHSR) on September 15, 1959 in support of its operating license application.
Accident analyses were performed by Westinghouse to support the application.
Subsequently, in December, 1963 the licensed power level was increased to 600 Megawatts thermal and substantiating accident analyses were performed in support of this power
' increase.
In the interim, several proposed changes have been submitted, and approved, each with substantiating analyses.
On January 3,- 1974, a completely rewritten Final Safety Accident Report (FSAR) was submitted to the AEC but to date it has not been reviewed as a supporting document to the operating license.
In 1970 a review of the small break LOCA analysis by Yankee Atomic Electric Company, Nuclear Services Division (NSD), resulted in modification to the Safety Injection System.
llb
)k7 In 1974, Yankee performed an evaluation in accordance with the requirements of 10 CFR 50.59 Appendix K.
Currently, with a number of other older plants, YR is being reviewed by the NRC under the Systematic Evaluation Program.
As part of the Systematic Evaluation Program, the environmental qualification of equipment will be considered in the evaluation of the design basis events as tabulated in NUREG 0485 dated November 17, 1978, page 1-4.
This review has been underway for about two years and is due to continue about a similar length of time.
1130 148 III.
PHASE-I (Systems identified in IE Information Notice 79-22)
A.
PROCEDURE Phase I of the Yankee survey was directed to the specific systems identified in IE Information Notice 79-22.
The influence of each applicabic system on accident mitigation was considered.
When a system / environment interaction was determined to have the potential for degradation of the accident analysis, further investigation continued.
Such potential interactions are identified in Table No. 1.
In those systems where potentials did exist, nonsafety grade components exposed to accident environments were examined for their qualification to operate in that environment. Equipment not qualified was identified and sppears in the results of this report as " Areas of Concern".
Equipment whose qualifications could not be determined was considered not qualified.
Equipment whose failure mode could not be determined was considered to fail in the mode most adverse to accident mitigation.
If several items of nonqualified equipment were exposed to the same environment from any one accident, all are considered to fail in the most adverse mode.
Each " Area of Concern" was reviewed for its ef fects on accident mitigatio".
The impact of the concerns on the safety analysis appears in the evaluation for each " Area of Concern".
1130 149 TABLE-I SYSTEMS SPECIFIED IN IE INFORMATION NOTICE 79-22 Pressurizer Main Feed Reactor Steam Generator Acc ide nt PORV Control Control Vent System System System Inc. Stm. Dump X
X Steam Break Inside Vapor Container Steam X
X Break Outside Vapor Container Small LOCA X
X Large LOCA X
X Feedline Break Insidc X
X Vapor Container Feedline Break X
Outside Vapor Container 1130 150
III.B AREAS OF CONCERN B.1.
PRESSURIZER POWER OPERATED PELIEF-VALVE-CONTROL SYSTEM
System Description
The pressurizer power operated relief valve control system consists of the pressurizer pressure detector (PR-PD-6) and transmitter along with cabling necessary to control the operation of the solenoid relief valve (PR-SOV-90).
This equipment is located inside the vapor container.
The cabling exits the vapor container at the electrical penetration and is routed to the control room and switchgear room which are devoid of high energy piping. The control pdrtions of this system are located in the control room and switchgear room.
Concerns The investigation uncovered several concerns in this area as follows:
1.
.The pressurizer pressure transmitter which generates the actuation signal which opens the PORV on main coolant system high pressure is located inside the vapor container and is subject to environments which are related to accidents.
Since the transmitter is not qualified for this environment, it is assumed to fail by generating a high pressure signal.
Our concern,is the impact to the core resulting from a stean lir.e, feed line, or main coolant system break coincident with an open PORV. ))}]
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2.
In the above scenarios, the impact on the pressure of the vapor container is a concern.
Safety Evaluation Although this pressure transmitter is not environmentally qualified, Yankee has done a failure mode analysis of this transmitter and concludes that it will fail in the safe direction (low pressure).
PR-PD-6 is a Bourdon Pressure Tube Linear Variable Differential Transformer (LVDT) type transmitter; therefore, adverse environmental conditions would have no effect on the mechanical elements of the transmitter. The electrical signal generating circuit, including the resistance circuits, wiring, primary and secondary transfonner windings, and the temperature compensating circuits were evaluated for all combinations of open and short circuit failure and in no case could increase in output signal to the trip circuit be determined feasible.
This evaluation verified automatic trip system " fail safe" operation. As further assurance, Yankee has placed an order for an environmentally quali1fied pressure transmitter which will be installed at the refueling outage following transmitter delivery.
Nonetheless, Yankee has looked at the potential prob 1 cms caused by the opening of the solenoid relief valve.
This review looked at two di f fe re nt analyses.
The first is concerned with the temperature and pressure effects on the vapor container vessel.
The second addresses the e ffects on the core and reactor coolant system.
1c2_
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B.1.1 Af fect s on V.C.
Pressure Section 402 of the Yankee Rowe's Final Hazards Summary Report describes the vapor container design criteria.
The design pressure of the vapor container is greater than the calculated pressure rise following the complete severance of a 20" main coolant line with two open ends and the simultaneous rupture of one secondary main steam line.
This analysis was redone in 1971 (see Proposed Change No. 96, dated August 6, 1971).
A licensing analysis of the containment p re s su re transient resulting from a main steam line rupture only, inside containment has not been performed.
Similarly, the containment pressure transient resulting from a main steam line rupture ins ide containment followed by opening of the solenoid relief valve has not been determined.
However, scopirg analyses, as part of the YR SEP effort, to determine containment response to a main steam line rupture have been performed by YAEC.
These analyses, based on RELAP4 blowdown analysis and CONTEMPT-LT026 containment analysis, indicate that the design containment pressure of 34.5 psig is not exceeded for the most severe steam line ruptura.
The blowdown analysis is very conservative since it is based on pure-steam blowdown which yields maximum containment pressure and temperature conditions.
,,n J l33 r-
_9_
The added effect of a consequential opening of the pressurizer solenoid relief valve is believed to be minimal due to its small size (throat area of.601'in;
), which relates to a capacity of approximately 20 lbsm-sec~ of saturated steam at 2400 psia.
For a main feedwater line break inside containment, a maximum of one steam generator can blowdown to containment.
Thus, three steam generators would remain to allow damping of the transient resulting from the single steam generator blowdown.
Blowdown from the other three steam generators through the break is prevented in the following two ways:
1.
Blowdown via reverse flow through the intact feedwater lines
~
is prevented by check valves located in each of the four main feedwater lines, 2.
Blowdown via steam flow from the intact steam generators to the steam generator with the ruptured feedwater line and then out the break is prevented by Non-Return valves in each of the main steam lines connected to the four steam generators.
The resulting blowdown to the containment is therefore limited to that of one steam generator, which will consist of subcooled, two phase and steam blowdown, plus subcooled feedwater.
In comparison to the assumcd pure steam blowdown during a steam line rupture, the containment pressure transient resulting from a 1130 154 feedwater line break is expected to be significantly less severe due to the blowdown fluid energy. The re fore, the containment transient resulting from a main steam line rupture will bound the feedwater line rupture containment response. The added effect of a consequential npening of the pressurizer solenoid rate relief valve wil? be mininal and again bounded by the steam line rupture followed by opening of the PORV containment response which, as stated above, will not exceed containmant design pressure.
B.1.2 A f fec t s on Core and -Main Coolant Sys ten -
B.1.2.1 Stean Lin_e or Feed Line Brea_k Only Submittals dated July 2, 1973 and September 12, 1973 provide analysis on the affects of a main steam line break outside of containment on the core and reactor coolant system.
A steam line break inside the vapor container at Yankee Rowe can blowdown a maximum of one generator while a steam line break outside the vapor container could blowdown more than one generator.
For this reason the af fects of the outside break are more severe.
A licensing analysis for inside steam line breaks has not been performed since they are considered less severe than the outside breaks on the core and main coolant system.
1i3 155 A feedwatre line break analysis has not been performed for Yankee Rowe.
Each of the main feedwater lines feeding YR's four steam generators contains a check valve inside containment.
Therefore, a main feedwater line break outside containment will result in only minimal steam generator blowdown due to the quick closing of the main feedwater check valves.
Hence, the NSSS trancient will be minimal and will resemble a simple loss of feedwater event.
As previously stated, the most limiting feedwater line break will result in blowdown of only one steam generator.
Three other steam generators will remain intact and will feel the effects of the rupture only by not having feedwater supplied to them due to the diversion of the feedwater to the break. Thus, we have essentially two simultaneou's transients:
e.
1 1.
A partial loss of heat sink due to steam generator fluid loss from the steam generator connected to the ruptured feedwater line.
(this is a maximum loss in heat sink of 25 percent) 2.
A loss of feedwater event due to feedwater being diverted out the break.
130 i56 As stated, this event has not been analyzed for Yankee Rowe, but the RCS response is not expected to be outside criteria such as minimum DNBR > 1.30 and maximum RCS pressure < 110 percent of design for the following reasons:
1.
The plant will be tripped on a number of conditions including:
a) low steam generator water level in the intact steam generators b) high pressurizer water level 2.
Three steam generators are more than adequate to maintain the plant in a safe condition after trip with subsequent reinitiation of feedwater after isolation of the break.
This is readily accomplished by isolating the broken feed line from outside the containment.
3.
Comparison with later plant feedwater line rupture analyses have been made which indicate that the response of YR should be bounded by these analyses for the following reasons:
a) YR has a larger pressurizer steam volume per Mw and per RCS volume than do later plants, and hence, RCS 113C 157 pressure transients will be less severe than is expected for later plants, b) YR has a larger initial steam generator liquid inventory at normal water level and at the -low st eam generator water level per Mw and per volume than most later plants, and hence, a larger heat sink.
c)
RCS operating pressure is 2015 psia which is 485 psia less than the design pressure of 2500 psia, thus providing significant margin to design criteria, whereas, later plants typically operate at approximately 2200 psig.
d) YR has four steam generators and thus a feedwater line break can result in a maximum loss of 25 percent of the secondary heat sink.
SAR analyses have shown acceptable results for 2-loop plants in which a 50 percent heat sink loss is possible.
Based on the reasons cited above, it is firmly believed that a main feedwater line rupture will not result in violation of basic crit 2 cia of minimum DNBR > 1.3 and maximum RCS pressure < 110 percent of design.
B.1.2.2 Main Steam and PD,[V The effects n the core ar.4 ma.
coolant ;ystem of the opening of the PORV as a consequence of a steam line rupture have not been previously addressed, llowever, we have 7m.,.
examined this postulated sequence and have concluded the following at this time.
A steam line rupture event inside containment followed by opening of the PORV should be bounded by the opening of the PORV without a main steam line rupture for the following reasons:
1.
The steam line rupture will result in cooldown of the RCS which results in RCS shrinkage and pressure decay.
Adding a PORV opening event to this previous event will result in a further rapid RCS pressure decay.
Therefore, ECCS water will be injected into the RCS at an earlier tire into the combined event and at a greater rate than if the event were only a PORV opening.
2.
Three intact steam generators are more than adequate to maintain RCS pressures low enough to allow adequate ECCS per fo rmanc e.
B.1.2.3 Small LOCA and PORV Proposed Change No. 145 dated January 6, 1977 and subsequent supplements provide analysis on a spectrum of LOCA's and Yankee review has concluded that the failure of the solenoid relief valve without any other event is bounded by these analyses.
1 g, iY The failure of the sole
'i operated relie f valve, in addit on to 1 previously occurring small LOCA, has not been i
previotis1 v addressed. This could possibly result in a cold Icg break occurriny; at the same time as a hot leg break.
It is Yankee's best engineering judgement that the operation of the se'enoid relief valve following a small cold Icg break will result in increased core cooling.
The main coolant system is further depressurized by the opening of the solenoid relief valve and more ECCS flow to the core and out the loop 1 hot leg and pressurizer surge line is achieved.
The failure of the PORV af ter a small hot leg break has not been previously analyzed but Yankee's best engineering judgement believes th a t this would be similar to a slightly larger hot leg break which is bounded by the existing LOCA analysis.
B.I.2.4 Larce-lor.' ynd-PORV The fe ~ ' re if the PORV rfter either a large hot or cold leg br~ak has not been previously analyzcd but Yankee believes it will have an insignifice.. af fect sirce main coolant pressure is extremely low and very little steam flow would pass through the PORV since diff erential pressure is almost zero.
I130 160 B.1.2.5 Feedwater Line Break-and PORV The primary NSSS response to a feedwater line rupture is a partial loss of heat sink (one steam generator) and a possible RCS heatup prior to reactor trip.
Subsequent to reactor trip, RCS conditions will stabilize at typical post-trip condition since three steam generators are more than adequate to cool the plant down from hot full power conditions to typical post-trip ccnditions.
The impact of a consequential PORV opening during this event is expected to be very similar to opening of the PORV from full power initial conditions, which has been previously determined to be bounded by existing small break LOCA analysis.
The major lifference is the availability of only three steam generators versus the four assumed in typical LOCA analyses for YR.
As prev;,:us1" stated, tnis ef fect is minimal since three steam generators are adequate to. maintain RCS.pressr es lot enou gh to assure efficient ECCS operation, and ther c re, ad-quate core cooling.
B.2.
MAIN FEED CONTROL SYSTEM
System Description
Flow schematics of the main feedwater sy' tem are shown on drawing s
numbers M-14 and M-15.
1130 161 The feedwater control system is designed to automatically maintain the proper water level in the four steam generators under all operating conditions.
It also has provisions to allow the system tc be manually operated.
Each steam generator level control system is separate from the others, i.e., one can be on manual control while the others are in automatic and vice versa.
The Control System is a three element system which compares steam flow, feed flow, and steam generator level.
Density compensation for level and pressure compensation for steam flow is provided.
With the exception of the steam generator level transmitters, steam flow transmitters, the steam generator pressure transmitters and the feedwater flow transmitters, all equipment is located inside the switchgear room or control room which are d,evoid of high energy lines.
One set of the steam generator Icvel transmitters, those for level control, are located in the vapor container.
The steam generator pressure transmitters and steam flow transmitters are located in the pump room area of the turbine building.
The feedwater flow transmitters are located by the feedwater control valves in the mezzanine level of the turbine building.
11;9 162 The o t iw r
't of : t i 'n u n,rator levol
.t' yo, ich ar.
tf'ty grade, and ;)rovide for :afety relateJ t:
are qe d i fied for varc t caso cent ai ne nt accident conditionn and are located in,ide th' vapor container.
~ihese transnittern generate signals 6:ich act uate a reactor trip on low.tean generator level.
Concerns Several concerns c.<isted ind vere.tvestigated.
The c o nc e r r.; were a.s foilows:
1.
The steam generator level control transnitters ars l o c a t e in the vapor container and a re subject to environments from clean or feed line bree' s inside the vapor container or fron the e nvi ro nme n t caused by a lors of coolant eidant.
These t rans rd t t e r s are not rj ua l i f ied nor is their failure unie proven.
The assumption the re f ore, is that either a steam or feed break inaide tho 'apor container or a los-of coolant accident will, in short time, cause the feedwater regulating valves to close shutting off flow to the steam generators with eventual a ttenuation of secondary heat sink.
2.
The feedwa ter flow t ra n s mi t,t e r s, stean generator presnare trananit!crn, and steam flov transnitters are located in ehe turbiu, Euii ling and are subject to environnoutt cans ni by breakn in the stean lines or feedlines.
The assumption is the
"~0 163 P00R ORIB;NAL
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samc as above for either a steam or feed line break outside the vapor container, that is, loss of secondary heat sink.
3.
If feedwater flow is not terminated to a steam generator which has suf fered a broken steam line, the vapor container may be overpressurized.
Safety Evaluation 1.
Steam generator low level reactor trip is a safety grade system and is qualified for environments more severe than would exist with a large steam line break inside the vapor container coupled with a small break loss of coolant accident.
The low level trip actuates an annunciator on the main control board.
Thus, operator notification of steam generator low level is assured and proper response ensures retention of secondary heat sink.
The existing large break loss of goolant analysis does not require secondary side cooling.
Small break LOCA's require secondary cooling which is provided by either the main feed system, auxiliary f'cedwater system, with safety grade steam generator level trip system providing the required level information to the operator.
2.
Since all steam generator level trips and level indications are available during outside steam or feedline breaks, the operator can respond to this transient and manual 19 restore steam generator level.
I13J i64 3.
A main steam line break causes a reactor scram which results in a turbine trip.
If power level is above 15 MWE, the turbine trip causes the main feed pumps to trip.
The operators can then determine the affected steam generator and initiate feed to the 3 unaf fected generators.
Yankee concludes that the worst case failure of non qualified components in the main feed or main feed control system will have no adverse impact on the existing safety analysis for the reasons mentioned above.
The auxiliary feed system has been reviewed by the staf f as parc of the TMI related review and as part of the SEP program.
As a point of interest, the existing auxiliary feed system although acceptable, is being upgraded to provide complete separation frcm steam and feed lines outside the vapor container. The backfit schedule provides for placing the order for two new feed pumps within 60 days of this report.
Eventually the system will be diesel generator powered with new diesels, the size of which will be determined at the conclusion of the SEP program.
It is intended that the entire system be of latest safety criteria.
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B.3 AUTOMATIC ROD CONTROL SYSTEM
System Description
The Yankee Rowe Rod Control System consists of control rods, their drive mechanisms, and associated instrumentation, cabling and control logic.
The automatic features associated with this system are scram capability which is caused by numerous conditions and rod insertion which is causad by high main coolant temperature.
There is no automatic rod withdrawal feature.
The control logic is located in the Switchgear Room and Control Room which are devoid of high energy piping systems.
Concerne None Safety Evaluation Since there is no automatic rod withdrcwal capability associated with this system and because all control logic associated with rod withdrawal are remote from high energy line break environments, there is no equipment failure mode caused by high energy line break which can result in a reactivity insertion. Th e re fore, Yankee concludes that this system is not within the scope of concerns expressed in IE Information Notice 79-22.
I133 168 B.4 STEA'! CENERATOR-POWER OPERATED VEN r SYSTEM INCLUDINC STEAM DUMPS Description The Yankee steam generators are provided with spring actuated safety valves for overpressure protection.
In addition, a two-inch air operated atmospheric dump valve (TV-411) is provided for venting and to control reactor coolant temperature during low power physics testing.
This valve is intended to provide a limited amount of heat removal capacity.
This valve, which is manually isolated during normal operation, can be manually operated locally or remotely from the control room but it has no automatic features, thus, it is not subject to adverse operation as a result of environmental affects.
A 6" steam dump valve (PCV-402) provides turbine by pass from the 24" main steam line to the condenser.
This line is normally closed and is used during cooldowns, startups and certain transient conditions.
The steam dump valve (PCV-402) is an air operated valve which can be operated manually from the main control board, or operated automatically to limit upstream steam header pressure to 760 psig normally (adjustable). When open this valve will pass up to approximately 100,000 lbs. of steam per hour.
Drawing number M-20 shows a schematic of the main steam system.
TV-411 and PCV-402 are circl ':
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Concern 1.
The controls for PCV-402 are letated near the valve and would be subject to, but are not qualified for the environmental conditions resulting from rupture of the steam line.
The valve is postulated to fail by going to the open position.
This valve is not subject to any other accident environment such as LOCA or feedline break.
Safety-Evaluation 1.
The controls for valve (PCV-402) are pneumatic, except for trips from high VC pressure and low condenser vacuum.
An adverse environment around the controller would cause the valve to fail closed. No scenarios can be postulated which cause the valve to fail open.
Nonetheless, Yankee has looked at the potential failure of this valve.
Since steam break analysis which is part of the Yankee operating license analysis is based on a blowdown rate of 4,000,000 lbs/hr. and since PCV-4D2 is rated for only 100,000 lbs/hr, this event is considered to be bounded by the existing steam line rupture analysis.
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III.C Phase-I C.
Summary of Results Yankee concludes that the failure of non-environmentally qualified equipment in the system described does not result in scenarios which are unreviewed safety questions.
Yankee's best engineering judgement is that these new transients will not be limiting.
Phase II a.
Obiectives This phase will address the impac t of erroneous information displayed to plant operators from failures of unqualified equipment.
As part of YAEC's Systematic Evaluation Program ef fort detailed analysis for such events as steam line rupture and feedwater line rupture will be performed.
A review of other plant systems will be performed to ensure that all transients are bounded by analysis, b.
Results to Date The shutdown cooling system isolation valves have aircady been rev iewed.
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B.1 SilUTDOWN COOLING SYSTEM ISOLATION VALVES System Description _
The YR Shutdown Cooling System as shown on Dwg. M-5 is composed of piping. valves, cooling pump and heat exchanger.
The valves are located inside the vapor container while the key locked control switches are located in the Primary Auxiliary Building.
Concerns A steam generator blowdown line break in the Primary Auxiliary Building may create an environment which may possibly cause the non-environmentally qualified valve motor control circuits to malfunction resulting in the valves opening.
Opening either of the two suction or two discharge motor operated valves when the main coolant system is at operating pressure has the potential for creating a Loss of Coolant Accident outside the vapor container which is an unacceptable accident.
Safety Evaluation To preclude th is type accident, the power to the valve operators will be removed by disconnecting the motor lead cables as they leave the motor 4 4 n
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starters.
These valves are numbered SC-:OV-551, 552, S53, and 5:>i nJ will be le f t in the c lon ed pos i t i on tihen r.a i.a c oolant p re n s:u re i. r
'ove the de si ;n oporating pressut e of the shut dotm cooling system.
f Removal o f the motor lead cables does not create an unreviewe.1 sa fe ty question since there are no interlocks or cutomatic fea ture s associated with these val ves.
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