ML19319D604
| ML19319D604 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/08/1972 |
| From: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19319D589 | List: |
| References | |
| NUDOCS 8003170772 | |
| Download: ML19319D604 (10) | |
Text
-
m CRYSTAL RIVER UNIT 3 6.3 EMERGENCY CORE COOLING SYSTEM 6 3.1 Cene ral The ^.tcaic Energy Cornission -recently reevaluated the theoretical and experimental bases for predicting the per-formance of emergency core cooling systems (ECCS), including new information obtained from indus try and AEC research pro-grams in this field.
A.s a result of this reevaluation, the Cormission has developed interim acceptance criteria for emer-gency core cooling sy'ite=s - for light-water power reactors.
These criteria are described in an Interim ?olicy Statenent issued on June 25, 1971, and published in the Federal Retister on June 29, 1971, (36 F.R. 12247).
By letter dated August 9,1971, the Division of Reactor Licensing infomed the applicant of the additienal infor=stien that would be required for our evaluation of the perfor=ance of the Crys tal River Unit 3 ECCS in accordanct with the Interi Policy S tate =ent.
The applicant provided a revised analysis of the Crys tal River Unit 3 ECCS performance in a report titled "Multinode Analysis of B&W's 2568-hWe Nuclear ?lants During a Less-of-Coolant Accident" dated October 1971 and revised on January 12, 1972.
The cnalysis was performed using the 3&J Evaluation Mcdel in conformance with the Interim Policy Statement, Appendix A, Part 4 The analysis was performed assining the cccurrence of 8 %s 12 o 'y),7
n m
. a loss-of-coolant accident during operation at 102% of a nnninal power icyc1 of 2568 Mi thermal even though the i
reques ted power icvel is slightly less than this, i.e., 2544 MJ thermal.
6.3.2 SYSTEM DESCRIPTION The Crystal River Unit 3 emergency core cooline sys ten consists of a high pressure injection system, an injection system employing core flooding tanks, and a low pressure in-jection system with external (to the containment) recircula-tion capability. Various corhinations of these sys tw.s are a
enployed to assure core cooling for the ccenleta rance of
~
break sizes.
The high pressure injection sys ten includes three nunos, each capahic of delivering a tdnimum of 500 2nm at 600 nsin reactor vessel pressure and discharges to the reactor ecolant inlet lines. One pu=p vill provide the required minimun flew for the high pressure injection sys tem.
The high pressure t
injection pumps are located in the auxiliary building adjacent to the containment. - A concentrated boric acid solution from 1
the boric acid water storage tank is provided to the suction
. side of the high pressure punns during ECCS oneration.
Durina normal reactor operation, the high pressure injection s/ stem recirculates reactor coolant for curificatien and for suctiv of seal water to the reactor cociant circulation m=es.
The
=
a
m
. h tRh pressure injection sys tem is initiated at a low reactor coolant system pressure of 1500 psig or a reactor buildin2 pressura of 4 psig.
Autcmatic actuation switches the systen fron normal-co emergency operating mode.
One of the tnree high pressure pumps is normally in one'_ cien. Tha sybtem is designed to withstand a single failure of an active ecnnonent without a loss of function.
The two core flooding tanks are located in the contain-ment outside of the secondary shield.
Each accumulator has a total volume cf 1410 ft with a nc=inal stored horated water 3
volume of 1040 ft pressurleed with nitrogen to 600 psi.
Each accumulator is connected to a separate reactor vessel core ficoding nozzle by a ficocing line incornoratinz two check valves and a motor operated normally open s ton valve adjacent to the tank.
The core ficodinz tanks will therefore inject water autocatically whenever the cressure in the pri-mary system is reduced below the core flooding tank nressure of 600 psig.
The low pressure injection system includes two pumps each capable of delivering 3000 gpm at 100 psie reactor vessel pres-1 sure arranged to deliver water to the reactor vessel thrcunh
- two separate injection lines.
One lev oressure injection numn is. capable ~ o# removing the heat enerzy generated af ter a less -
of-coolant accident, s
w 3
-4w
- an-#MPe = =t4W _b
-e W
n.
~)
4 The low pressure injection system pumps initially take their suction from the borated water s torage tank and later, during recirculation from the reactor building energency sump.
The recirculation system components are redundant so as to withstand a single failure of an active or passive co=ponent without loss of function at the required flow.
The. low pressure injection s stem is actuated on a low reactor coolant system pressu: of 500 psig or a high reactor building pressure of 4 psig.
All of the ECCS subsystems can accccplish their function when operating on e=ergency (cnsite) pcwer as well as of fsite power.
If there is a loss of nornal power sources the engi-
. neered aafeguards power line is connected to the energency
. diesel acnerators which have a startup time of 10 seconds or less. The pu=ps and valves of the injection system will be i
1 l
energized at less than 100% voltage and frequency to achieve i
I the design injection flow race within 25 seconds.
6.3.3 PERF010fANCE EVALl'ATION 6.3.3.1 General We have developed a set of conservative assu=utions and procedures to be used in conjunction with the 3 becch and
~
Wilcox developed codes to analyze the ECCS functions. The
~
mi e.
W g-r y.
m -
w..
4
~
. ~. - -..
,q n
. assumptions and procedures used by 36W in analyzing the per-formance of the ' Crystal River Unit 3 ECCS are described in Appendix A, Part 4 of the Interim Policy Statement published in the Federal _ Register on December 18, 19 71 ( F.R. Vol. 36, No. 244). Report BAW-10034 "Multinode Analysis of 3&W's 2568
>Mt Nuclear Plants During a Loss-of-Coolant Accident," October 1971, covers the performance of cores for which the fuel pins are pressurized and the peak linear heat rate is 18.15 IG/f t.
From this analysis the 8.55 f t' cold leg split is the limiting case accident with a peak clad temperature of 2177*F. For comparison, the peak linear heat rate for Crfstal River Unit 3 is 17.31 'G/ft and the core pcwer is 2514 >M:.
6.3.3.2 Analynis of the 31cudcun Period The applicant used the CRA?! and THETA 1-
'caputer codes for the analysis of the bicwdevn phase of the transient.
Using these codes, and the evaluation model soecified in Appendix A, Part 4, of the Interim Policy Statement, th e applicant provided the reevaluation of the ECCS performance in compliance with the Commission's Interim Policy S tatement.
For the blowdown portion of the accident, we have con-cluded that the applicant's analyses as repcrted in 3AN-10034 conform to the requirements specified in the Commission's Interim Policy Statement, Appendix A, Part 4
. ~..
. %)
6.3.3.3 Analysis of the Refill and Reflood Period The applicant has censidered the thernal behavior of the core during the refill and reflood portion of the less-of-coolant accident, which is explained as follows:
(1) The vessel refill is provided initially by the core flooding tanks, and later by the pu=oing systems, and is assumed to s tart at the end of the blewdcwn period. The reactor vessel is assumed to be essentially dry at the end of the blowdcun period, as a result of the conserva-tive assnention in Appendix A, Part 4, of the Interin Policy Statenant that water iajec:ed h / the core floodina tanks prior to the end of blowdewn H ejected fren the primary sys Na, (2)
No heat transfar in the core is racu=ed until the level of water reaches the botton of the core, at which time refill is considered complete and the core reficod starts.
The end of blowdcun is 14.6 seconds af ter ruoture for the 8.55 f t' cold leg double-ended break and refill to the bottom of the core is complete about 23 seconds after rupture. The end 9
of bicwdcun is 18.7. seconds af ter rupture for the 8.55 f t~
cold lec solic and reflood is co=clete about 26 s ?conds af ter rupture.
k._
r m
n
. ( 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 substantial amounts of steam.
The core reflooding rate increases and peaks in about 8.5 seconds after the end of blowdown at about 11 to 12 inches per second, then decreases rapidly leveling off at about 5.5 inches per sccend about 10 seconds after the end of blowdown. At 10. seconds af ter the end of blewdewn, the
- water covers about 12 inches of the core f or the case of a double-ended cold lag break and 20 inches of the cera for the case of a 8.55 ft cold ' leg s olit.
(4 The amount of s tea = generated in :h2 core together vitz.
the steam flev path resistance gevarns the rate of steam flcw. The steam flew path is assumed to be only through the vent valves within the reacter vessel and no credit is taken for steam ficw around the locp. The s teca ficw 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 tanka and low pressure injection systen water is piped direct!.y to the reactor vessel with no intervening-reactor coolant systen piping.
'm
3 n
. (5)
The peak temperature reached in the transient for the 2
limiting 8.55 f t cold leg split occurs about 30 seconds after the break.
Based on our review of "Multinode Analysis of B&W's 2568 FWt Nuclear Plants During a Loss-of-Coolant 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 Crystal R'ver Unit 3 at 102% of a nominal power level of 2568 FMt:
Cold Lee Pine Breaks Peak Clad Temoeratures ('F)
(Area)
(Tvee B reak) 8.55 ft (Double Ended) 2052 8.55 ft2 (Split) 2177*
3.0 ft (Split) 1652 0.5 ft (Split) 1614 Hot leg 14.1 ft (Split) 1621
- Limiting case.
The total core metal-water reaction is less than 1% for each of the assumed pipe breaks.
m
_9 6.3.4 C0!:CLUSID' S On the basis of our evaluation of ne additimal 3W
.inalyses, desc ribed in 6.3. 3.1 above, we cenclude -he our acceptance criteria, as described in the Conmissicn's Interim l'olicy S tatenent have been met:
( 1)
The naximun calculated fuel cle.ent cladding te: perature does not exceed 2300
- F.
(2)
The amount of fuel element cladding that reacts ch e~.1-cally with water or steam does not exceed 1% of the total tmount of cla@ing in the reactor.
"" e calculated clad tem,2:ature transient is terminated
( 3) i at a time when the core geenetry is still amenable to cooling, c.nl before the cladding 1.s no c=brittled as to fat' during or af ter quenching.
(4) The core temperature is reduced and decay heat is removed for an extended period of tine, as recuired by the lonc lived radioactivity remaining in the core.
The results of the applicant's analyses for c. loss-of-coolant accident initiated at a core power level of 2568 MJt show that the acceptance criteria are met on the basis of i
l analyses performed in accordance with an acceptabic evaluation model given in the Interin Policy Statement.
)
i i
N 1
m
. On the basis of our evaluation of the SW an lyses described in 6.3.3.1 above, we have deter =ined that th e energency core cooling system is acceptable and will pro-vide adequate protection for any loss-of-coolant accident.
l l
1 l