ML19319D599
| ML19319D599 | |
| Person / Time | |
|---|---|
| Site: | Crystal River, Rancho Seco  |
| Issue date: | 05/08/1972 |
| From: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19319D589 | List: |
| References | |
| NUDOCS 8003170768 | |
| Download: ML19319D599 (10) | |
Text
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RANCHO SECO UNIT 1 6.3 EMERGENCY CORE COOLING SYSTEM 6.3.1
- Gene ral The Atonic Energf Cc nission recentiv reevaluated the theoretical and experimental bases for predicting the perfor-mance of emergency core cooling systems (ECCS), including new infor=ation obtained frem industry and AEC research orcerans in this field.
As a result of this reevaluation, the Co==is-sion has developed interin acceptance criteria for emercencv core cooling systems for light-water acuer reactors. Thes e criteria are described in an Interin Poliev Statenent issued en June 25,'1971, and published in the Faderal Re21s ter on June 29, 19 71, (36 F.R. 12247).
By letter dated Augus t 11, 1971, the Divisien of Reactor Licensinz inforned the annlicant of the additional infernation that would be recuired f or our evaluation of the performance cf the Rnnche Seco Unit 1 ECCS in accordance with the Interi: Policy Statement.
The acn11-cant provided a revis ed analysis cf the Rancho Seca Unit 1 ECCS performance in a report titled "Multinode Analysis of B&W's 2568-Mh't Nuclear Plants During a Less-of-coolant Acci-dent" dated October 1971 and revised on January 12, 1972 The applicant also provided in Amendment 9, dated March 10, 1972, to his FSAR a supole= ental analysis of tha ECCS fer a core pcwer of 2772 Mi thermal. The analysis was cerfer cd usinc the 3&W Evaluatien Medel in confermance with the Interin 8008170.7 h
, On the basis of our evaluation of the B&W analyses described in 6.3.3.1 above, we have determined that the emergency. core cooling system is acceptable and will pro-vide adequate protection for any less-of-coolant accident.
s 3-(5) The peak te=perature reached in the transient for the n
limiting 8.55 f t' cold leg split occurs about 30 seconds afcer the break.
Bas u our review of "W1 inode Analysis of 36W's 2568 Wt Nuclear Plants During a Loss-of-Coo 1&t Accident" BAW-10034, October 1971, we have concluded that the applicant has evaluated the refill and reflood events in an acceptable manner.
6.3.3.4 Results The applicant has calculated the following tenperatures for Arkansas Nuclear One, Unit 1 at 102.* of a necinal pcuer level of 2568 Mit:
Cold Leg ?ipe B reaks Peak Clad T a,paratures ( *7)
( Area)
(?/;e 3rcak) l 8.55 fe2 (Double Ended) 2052 i
8.55 f t" (Split) 2177*
2 3.0 ft (Split) 1652 0.5 ft" (Split) 1614 Hot leg 14.1 ft (3plit) 1621
- Limiting case.
The total core metal-water reaction is less than 1% for each of the asst =ned pipe breaks.
j
Policy Statement, Appendix A,' Part 4.
The analyris was per-forned-assuming the occurrence of a less-of-coolant accir'ent during operation at 102'! of the requested pcuer level of 2772 Mi thermal.
_6.3.2 SYSTEM DESCRIPTION The Rancho Seco Unit 1 emergency cote cooline systen con-sists of a high pressure injection system, an injection sys ten enplcying core flooding tanks, and a icw pressure injecticn system with external (to the containment) recirculatien cara-bility. Varicus ce=binaticus of thes e sys tems ar2 2:alcyed to assure core cooling ic the cc=plete range of b re.u. sizes.
The hi$ pressure injection sys ter includes three cuarc,
ench capable of delivering a mini uc of 500 tpn at 300 rs 12 l
reactor vessel crasaure and cischarges to the reacter coolant inle lines. One pump will provide the required ninin= #1cw i
i for the high pressure injection systan. The high pressure injection pu=es are located in the auxiliarv huildina adiacent to the centainment.
A concentrated boric acid solution from the boric acid water storage tank is provided to the suction
- side of the hizh pressure pu=ps during ECCS coeration. During nc=al'rasetcr c:aratien, the high prescur: injection sys tem recirculates reactor coolant for curificaticn and for sucolv af seal water to the reactor coolant circulatten :=rs. The
_..c_-.____.._
~.
. high pressure injection system is initiated at a icw reactor coolant system pressure of 1600 psig or a reactor building pressure of 4 rsig. Autenatic actuation switches the sys ten from normal to emergency operating mode.
One of the three high pressure pumps is normally in operation. The system is designed to withstand a single f ailure of an active co=nonent without a loss of function.
The two core flooding tanks are located in the centain-ment outside of the secondary shield. Each accunulator has a total volume of 1410 f t with a neminal stored borated wa:ar volume of 1040 f pressurized with nitrogen to 600 nsiz.
F.ach accumulator is connected to a separate reacecr vessel core flooding nos:le by a flooding line incornorating two check valves and a =cter operated normally cpen s tcp valve adj acent to the tank. The core flooding tanks will therefore inject water autcratically whenever tha cressure in the ori-cary system is reduced below the core flooding tank oressure of 600 psig.
The low pressure injection system includes two puens each capable of delivering 3000 gpm at 100 psig reactor vessel 1
pressure arranged to deliver water to the reactor vess'el i
i I
through two separate injecticn lines. One icw cressure in.iec-tien pump is capable of removing the heat enarty generated af ter a loss-of-coc la..t accident.
[-
-m 3
4-The icw pressure injection system pu=ps initially take their suction from the borated water storage tank and later, during recirculation frem the reactor building emergency su:.
The recirculatica system cc=penents are redundant so as to withstand a single f ailure of an active or passive compcnent l
without loss of function at the required ficv.
The icw pressure injecticn system is actuated en a icw reactor coolant systen pressure of 1600 psig or a high reactor building pressure of 4 psig.
All of the ECCS subsystems can accc:plish their function when operating en emergency (casite) power as well as cifsite
- pcwer, if there is a less of nornal pcwer sources the engi-neered safeguards power lina is connected to the emergen;f diesel scncrators which have a s tartup time of 10 seconds or less. The pumps and valves of the injection systa= will b e energized at less than 100% voltage and frequency to achieve the design injectica flew race within 25 seconds.
6.3.3 PERFORMANCE EVALUATICN 6.3.3.1 General We have developed a set of conservative assumptiens and procedures to be used in conjunction with the Babccck and Wilcox developed codes to analy:e the ECCS functiens. The assur:ptions and procedures used by 35W in analyzing the per-formance of the Rancho Seco Unit 1 ECCS are described in
=
m
. Appendix A, Part 4 of the Interim Policy Statement published in the Federal Recister en Decenber 18, 19 71 ( F. R. Vol. 36, No. 244). Report 2L'-10031 "Ifultinede.d.alysis of S &W's 2563 SMt Nuclear Plants During a Less-of-Coolant Accident," October 1971, covers the perfor=ance of cores for which the fuel pins are pressurized and ~ the peak linear heat rate is 13.15 KW/ft.
In addition, Amend:ent'9 of the FSAR presents the 35W LOCA analysis for a core thermal pcwer cf 2772 SM and a peak linear o
heat rate of 19.6 KW/ft.
From this analysis the S.55 f t' cold leg split is the limiting case accident with a peak clad tenperature of 2231*F.
6.3.2.2 Analysis of the 31cudeen Pericd 2e applicant usad the CRAFI and T112TA 1-3 co puter cedes f or the analysis of the blevdcun phase of the transient.
Using these codes, and the evaluatica =cdel specified in Appendix A, Part 4, of the Interi Policy S tatc=2nt, the applicant provided the reevaluatien of the ECCS performance in co=pliance with the Com=issien's Interin Policy Statement.
For the blewdown portion of the accident, we have con-cluded that the applicant's analyses as reported in 3AW-10034 and Amendment 9 of the 75AR ccaform to the requirasants Ope:1-fled in the Commission's Interin Policy Statement, Appendix A, Pa rt 4 a
e
W p:
fs 6.1.3.3 Analvsta of the Refill and Reficed Period The applicant has censidered the thermal behavior of the core during the refill and reflood nortion of the icss-of-coolant accident, which is explained as follows:
(1)
The vessel refill is provided initially by the core flooding tanks, and later by the pumping sys tems, and is assu=ed to start at the end of the bicwdown period. The reactor vessel is assumed to be essentially dry at the end of the bicwdewn period, as a result of the conserva-tive assumption in Appendi:: A, Part 4, of the Interia Policy Statement that water injected by the ccre ficodinc tanks prior to the end of blowdern is ejected frc: the crimary system.
(2)
Nn heat transfer in the core is assumed until the level of water reaches the bette of the core, at which time re fill is censidered complate and the core reficod starts.
The end of blowdown is 14.8 seconds after rupture for the 8.55 ft cold leg double-ended break and refill to the botton of the core is cceplete about 26 seconds af ter rupture. The end of blowdevn is 19.4 seconds af ter rurture for the 5.55 fr-cold les split and reflood is cenclete about 29 seconds af ter rupture.
. =.
~s (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 encuzh of the cera is covered to generate subs tantial amounts of s team. The core reflooding rate increases and peaks in abcut 11 to 12 seconds af ter the end of blowdewn at about 9 to 10 inches per second, then decreases rapidly leveling off at about 4 inches per second about 13 seconds after the end of b lowdown. At 13 seconds after the end of bicwdcwn, the water covers about 20 inches of th e cora "cr the case of a 3.53 ft cold leg split.
(4) The amount of steam generated in the cora together with tha.4 team flew pa:h resistance governa che rata e' s teen flow. The staan flow path is assumed to be oniv throuch the vent valves within the reactor vessel sad no credit is taken for steam fice around the loon.
The s team ficw resistance also limits the rate of liquid rise in the core, but the annulus water level continues to increase until the 13 ;,6 d level reaches the inlet nozzle.
Core flooding ta-ks and low pressure injection system water is piped directly to Obe re cor vessel with no intervening reactor coolant system 3
-a.g.
s_+,m e
3.
-Sbe (5) The peak te=perature reached in the transient for the 9
liniting 8.55 f t' cold leg split occurs about 33 seccnds af ter the break.
Based on our review of "Multinode Analysis of 36W's 2568 Frdt Nuclear Plants During a Loss-of-Coolant Accident" bah'-
10034, October 1971, and Amend:ent 9 to the FSAR we have cen-cluded that the applicant has evaluated the refill and reflood events in an acceptchie canner.
6.3.3.4 Results The applicant has calculated the f olleving temperatures for Rancho Seco Unit 1 at 102f. of a nc=inal pcwer level of 2772 SMt:
Cold lac Fire Breaks Peak Clad Terceratures ( *F)
(Arca)
(_Tvne 3rcak)
(
9 3.55 f t' (Double Ended) 2174 9
3.55 f t' (Split) 22S1*
9 3.0 ft' (Split) 1697 9
0.5 ft' (Split) 2054 Hot leg 14.1 ft-2 (Split) 1609
- Limiting case.
The total core metal-vater reaction is less than l'.
for each of the assumed pipe breaks.
w w
m y
=-
w-4 y-t-
-y
m.
. 6.3.4 CONCLt'SIONS On. the basis of our evaluation of the additional 36'a' l
analyses, described in 6.3.3.1 above, we conclude that our acceptance criteria, as described in the Commission's Interim Policy Statement have been met:
(1)
The maximu= calculated fuel ele =ent cladding temperature does not exceed 2300'F.
(2) The a= cunt of fuel element cladding that reacts chemi-cally with water or steam does not exceed 1*' of the tctal amount of cladding in.ti a reactor.
( 3) The calculated clad temperature transient is terninated at a time when the core gecretry is s till amenable to cooling, and b efore the cladding u so erbrittlec as to f ail during or af ter quenching.
(4)
The core temperature is reduced and decay heat is removed
]
for en extended pericd of time, as required by the long i
lived radioactivity remaining in the cere.
The results of the applicant's analyses for a loss-of-coolant accident initiate / at a core pcuer level of 2772 be ahew that the acceptance criteria are cet en the basis of analyses performed in accordance with an acceptabic evaluation model given in the Interis Policy Statement.
~_
m
- On the basis of our evaluation of the BW analyses described in 6.3.3.1 above, we have determined thet one emer-gency core coolin;; system is acceptable and ',ill provide ade-quate protection for any loss-of-coolant accident.