ML19319D597
| ML19319D597 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, Arkansas Nuclear  |
| Issue date: | 05/08/1972 |
| From: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19319D589 | List: |
| References | |
| NUDOCS 8003170766 | |
| Download: ML19319D597 (10) | |
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,i ARKANSAS NUCLEAR ONE - UNIT 1 6.3 EMERGENCY CORE COOLING SYSTEM 6.3.1 General The Attsr::ic Encrgy Co==1ssion recently reevaluated the theoretical and experimental bases for predicting the perfor-mance of emergency re cooling syste=s (ECCS), including new infor=atien obtained from industry and AEC research progrs=s in this field. As a result of this reevaluation, the Co==1s-sion has developed interin acceptance criteria for emergency core cooling systems for li$t-water power reactors. Thes e criteria are described in an Interim Policy S tatement issued on June 25, 1971, and published in the Federal Regis ter en June 29,1971, (36 F.2.12247). By letter dated Augus t 11, 1971, the Divisicn of Raactor Licensing inforced the applicant of the additional information that would be required for our evaluation of the performance of the Arkansas Nuclear One, Unit 1 ECCS in accordance with the Interi: Policy S tatement.
The applicant provided a revised analysis of the Arkansas Nuclear One, Unit 1 ECCS performance in a report titled "Multinode Analysis of 3&W's 2568-MWt Nuclear Plants During a Less-of-Coolant Accident" dated October 1971 and revised en January 12, 1972 The analysis was perfor=ed using the 35%
Evaluation Model in conformance with the Interim Policy S tate-ment, Appendix A Part 4 The analysis was perferred assuming 8003170756 l
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coolant systen pressure of 1500 esig or a Teactor building pressure of 4 psig.
Automatic actuation switches the svs tem fron normal to energency operating node. One of the three high pressure punos is normally in operation. - The syster is designed to withstand a single failure of an active component without a loss of function.
I The two core flooding tanks are located in the contain-nent outside of the second ry shield.
Each accuculator has a total volume of 1410 ft with a ncminal stcred borated water volune of 1040 f t pressurized. '.ch nttrogen to 600 psig.
Each accumulator is connected to a secar te rece:or vessel coro floodine noczle by a ficcdinz line incer:oratinz two check valves and a nocer eneratec narrally onen s eco valve adjacent t o th e t ank. The ccre floodin: tar.ks will therefore inject water automatically whenever the pressure l
in the primar/ ays ten is reduced belov
,e core '1codiaz tank' pressure of 600 psig.
.The low' pressure injection sys te= includes two p"nos each capable of delivering 3000 gpm at 100 psig reactor vessel pressure arrcnzed to delivar water to the reactor vessel through two separate injection lines. One low pressure injec-tion puma is capable of removing the heat ener27 2enerated af ter.a loss-of-coolant accident.
. the occurrence of a less-of-coolant accident during operation at 102% of the requested pcwer leve. of 2568 :G therral.
6.3.2 SYSTEM DESCRIPTION The Arkansas Nuclear One, Unit 1 emergency core cooling system consists of a high pressure injection sys te=, an injec-tien system employing core flooding tanks, and a low cressure injecticn system with external (to the containment) recircula-tion capability. Varicus combinations of these systems are e= ployed to assure core cocling for the ec=clete range of break sizes.
The high pressur2 injecticn 4 is :em includes three ou=cs,
each capable of delivering a mini =u: of 500 gpm at 600 psig reacter vessel pressure and discharges te the reactor coolant inlet lines.
One pu=p will provide the required minime= flew for the high pressure injection sys tem. The hign pressure injection pumps are located in the auxiliary building adjecent to the containment. A concentrated boric acid solution from the boric acid water s torage tank is provided to the suction side of the high pressure pumps during ECCS operation. During normal reactor operation, the high pressure injection sys tem recirculates rasctor coolant for purification and for sucoly of seal water to the reactor ecolant circulation pumps. The high pressure injection system is initiated at a lcw reactor L
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.~ The low pressure injection system pu=ps initially take their suction f rom the b', rated water storage tank and later, during recirculation from the reactor building energency su=p.
The recirculation system components are redundant so as to withstand a single failure of an active or passive component without loss of function at the required flow, i
The low pressure injection system is actuated on a low reactor coolant system pressure of 1500 psig or a high reactor building pressure of 4 psig.
All cf the ECCS subsysta=a can acco=plish their function when operating cn e=ergency (onsite) power as well as offs 1:e p cwer.
If there is a less of ner:21 pcwer sources the engi-neered safeguards pcwer line is connected to the emergency diesel generators which have a star:cp time of 15 seccnds or less. The pumps and valves of the injection systa= will be energized at less than 100% vol: age and frequency to achieve the design injection flow rate within 25 seconds.
6.3.3 PERFORMANCE EVALUATION 6.3.3.1 General We have developed a set of conservative assu=ptions and procedures to be used in conjunction with the Sabccck and Wilcox developed codes to analyze the ECCS functions. The assumptiens and procedures used by 3&W in analyzing the
S performance of the Arkansas Nuclear One, Unit 1 ECCS are described in Appendix A, Part 4 of the Interim Policy State-cent published in tha federal Ragister en Dacceber 13, 1971 (F.R. Vol. 36, No. 244). Report 3AW-10034 "<inode Analysis 1
of BfN's 2568 W: Nuclear Plants During a Loss-of-Coolant Accident," October 1971, covers the performance of cores for which the fuel pins are pressurized and t.he peak linear heat a
rate is 13.15 W /ft. From this analysis the 8.55 f t" cold leg split is the limiting case accident with a peak clad tempera-ture of 2177'F.
For ccmparison, the peak linaar heat rate for Arkansas Nuclear One, Unit 1 is 17.va W/f t.
1 6.3.3.2 Analysis of the 31cwdewn Period The appli cant used the CRAIT and OETA 1-3 co=puter codes for tha analysis of the blowdown phase of the transient.
-l Using these codes, and the evaluation model specified in j
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Appendix A, Part 4, of the Interim Policy Statement, the applicant provided the reevaluation of the ECCS performance in i
i compliance with the Commission's Interim Policy Statement.
f For the blowdown portion of the accident, we have con-cluded that the applicant's analyses as reported in BAW-10034 conform to the requirements specified in the Comcission's Interim Policy Statement, Appendix A, Part 4 I
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. 6.3.3.3 Analysis of the Refill and Reflood Period The applicant has considered the ther.al behavior of the core during the refill and reficed pertien of the less-of-coolant accident, which is explained as follows:
(1)
The vessel refill is provided initially by the core i
flooding tanks, ~ and later by the cunning sys ters, and is assured to start at the end of the blowdewn eeriod. The reactor vessel is assu=ed to be ess entially drv at the end of the blewdewn period, as a result o# the conserva-tive assunpticn in Appendix A, ? art I, of the Interin Policy Statement that water injected by the cara ficoding tanks prior to the end cf blevdcun is ejected #re :he prinary syn ten.
( 2). No heat transfer in the core is assured until the la ral of water reaches the bottet of the core, at which time i
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refill is considered cc=plete and the cora reficod starts.
The end of blowdown is 14.6 seconds af ter rupture for the 8.55 ft~ cold leg double-ended break and refill to the botton of the core is complete about 23 seconds after runture. The end a
of blevdcun is 13.7 seconds after ructura for the 3.55 f "t cold leg split and reflood is co=plete abcut 26 seconds after rupture.
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, (3)
The reflood of the core is characterized initially by a rapid liquid level rise both in the core and in the vessel annulus until enough of the core is covered to generate subs tantial amounts of s team. The core reflooding rate increases and peaks in about 8.5 seconds af ter the end of blowdown at about 11 to 12 inches per second, then decreases rapidly leveling off at about 5.5 inches per second about 10 seconds after the end of blowd own. At 10 seconds after the and of blowdcun, the water covers abcut 12 inches of the core for the case of a double-ended cold leg break and 20 inchen of the core 2
for the case of a 8.55 f t cold leg split.
(4) ' The a ount of steaa generated in the core together wita the s t,an fice path res is tance gove rns th e rate o f s team flow. The steam flow path is assu=ed to be only through the vent valves within the reactor vessel and no credit is taken for steam flew around the loop. The s tea: 11cw resistance also limits the rate of liquid rise in the core, but the annulus water level continues to increase until the liquid level reaches the inlet nozzle.
Core flooding tanks and icv pressure injection system water is piped directly to the reactor vessel with no intervening reactor coolant system piping.
m lD 9-6.3.4 CONCLUSICNS 9
On the basis of our evaluation of the additional 35*4 analyses, described in 6.3.3.1 above, we conclude that our acceptance criteria, as described in the Co ission's Interim Policy Statement have been met:
- (1) The maxt=um calculated fuel t.lement cladding temperature does not exceed 2300*F.
(2) The amount of fuel ele =ent cladding that reacts chemi-cally with water or steam dces not exceed l'! of the total amount of cladding in the reactor.
( 3)
The calculated clad temperature transient is terninated at a cine when the core geonatrv is a till amenable to cooling, and before the cladding is so embrittled as to fail during or aftar quenching.
(4)
The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long lived radioactivity remaining in the core.
The results of the applicant's analyses for a loss-of-i coolant accident initiated at a core power level of 2568 Mit i
show that the acceptance criteria are net en the basis of analyses perfor=ed in accordance with an acceptable evaluacicn model given in the Interim Policy State =ent.
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