ML20037A378
| ML20037A378 | |
| Person / Time | |
|---|---|
| Site: | Oconee, Crane  |
| Issue date: | 09/25/1972 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20037A372 | List: |
| References | |
| NUDOCS 8001100656 | |
| Download: ML20037A378 (10) | |
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6.3 Emergency Core Cooling System 6.3.1 Ceneral l
.t The Atomic Energy Commission recently reevaluated the theoretical l
i and experimental bases for predicting the performance of emergency,
t core cooling systems (ECCS),' including new information obtnined from industry and'AEC research programs in this field. As a result.of this reevaluatien, the Commission has developed interim acceptance criteria for emergency core tcoling systems for light-water power reactors. These criteria are described in an Interim Policy Statement issued on June 25, 1971, and published in the Federal Register on June 29,1971- (36 F.R. 12247).- By letter dated July 9, 1971, the Division of Reactor Licensing informed the applicant of the additional information that would be required-for our evaluation of the perfornance of the Three Mile Island Unit 1 ECCS.in accordance with the Interim Policy Statement.
The applicant provided a revised -
analysis of the Three Mile Island Unit 1 ECCS performance in ' report.
BAW-10034 titled "Multinode Analysis of B&W's 2568-MWt Nuclear' Plants
,During a Loss-of-Coolant Accident," dated October 1971, including Revisions 1, 2 and 3.
The analysis was performed using the B&W Evaluation Model in conformance with the Interim Policy Statement, Appendix A, Part 4.
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.P m The analysis was performed assuming the occurrence of a loss-of-coolant accident during operation at 102% of a nominal power level Rt.
of 2568 MW thermal as compared to g lower power level of 2535 MW thermal requested by the appl'icant.
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i 6.3.2
System Description
i The Three Mile Island Unit 1 emergency core cooling system consists -
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of a high pressure injection system, an injection system-employing core flooding tanks, and a low pressere injection system with external (to the containment) recirculation capability. Various combinations of these. systems arc
- employed to assure core cooling.
for the complete range of break sizc-;.
The high pressure injection system includes three pumps, each capable of delivering' 500 gpm at 600 psig reactor vessel pressure and dischstges to the reactor coolant inlet lines. One pump will provide the required
. minimum flow for the high pressure injection system. The high pressure injection pumps are located in the auxiliary building adjacent to the conteinment. A concentrated boric acid solution from the boric acid water storage tank is provided to the suction side of the high pressure pumps during ECCh operation. During normal reactor operation, the high pressure injection system recirculates reactor coolant for purification and for supply of seal water to the reactor coolant circulation pumps. The high pressure injection system is initiated at a low reactor cooit.nt system pressure of 1500 psig or 500 psig or a reactor bu-tiding pressure of 4 psig. Auronatic j
actuation switches the system from normal to emergency operating mode. One of the three high pressure pumps is normally in operation.
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The system is designed to withstand a single failure of an active-component-without a loss of function.
The two core flooding tanks are located in the' containment.outside of the secondary shield.
Each accumulator has a total v. lume of 1410 ft with a nominal store'd ' borated' water ' volume of '1040 'f t
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pressurized with nitrogen'to 600 psig.
Each; accumulator is connected to a separate reactor vessel core flooding nozzle by a flooding'line -
incorporating two check valves and a motor. operated normallyfopen-stop valve adjacent to the tank.'
The core flooding tanks'will-therefore inject water'. automatically. whenever the pressure in th'e
, primary system is-reduced below the core' flooding tank pressure of-600 psig.
The low pressure injection system includes two Epumps each capable of delivering 3000 gpm at 100 psig reactor vessel pressure, arranged:
to-deliver water to the reactor vessel through two ' separate injection -
lines. One low pressure injection ~ pump is-capable of removing the heat energy generated af ter a loss-of-coolant accident.
- The low pressure injection system pumps initially l take their suction from the borated water storage tank and,later, during' recirculation from the reactor building emergency sump.-
The recirculation system components are redundant so as to withstand a single failure of an.
active or passive component without lo'ss of function at the. required flow.
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-The low pressure injection system is' actuated on a low reactor
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coolant system pressure of 1500 psig or 500.psig, or a high reactor
. building pressure of.4 psig.
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t All of the-ECCS aubsystems can accomplishLtheir function when If operating on emergency (onsite)- power as' wcil as offsite power.
there is a loss of normal power sources. the engineered safeguards power line is connected to the emergency diesel generators.which l
have a startup time of 10 seconds or less. The pumps and valves of the injection system will be-energized at less than 100% voltage; and frequency to achieve the ' design injection flow rate within 25
. seconds.
6.3.3 Perfoqmance Evaluation 6.3.3.1 General We have developed a set of conservative assumptions and procedures to be used in conjunction with the Babcock and Wilcox developed codes to analyze the ECCS functions. The c.ssumptians.and procedures used by B&W in analyzing the performance of the Three Mile Island Unit 1 ECCS are described in Appendix A, Part 4 of the Interim Policy Statementypublished in the Federal Register on December 18, 1971 (F.R. Vol. 36, No. 244). Report BAW-10034 "Multinode' Analysis of B&W's 2568 MWt Nuclear Plants During a Loss-of-Coolant Accident,"
i October 1971, covers the-performance of cores for which the fuel.
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are pressurized and the peak linear heat rate is 18.15 KW/ft.
t From this analysis the 8.55 ft cold leg split is the limiting case accident with a peak clad temperature of 2177'F.
For comparison, the peak linear heat rate for Three Mile Island Unit 1 is 17.63 KW/ft and the core power is 2535 MWt.
6.3.3.2 Analysis of the Blowdovu Period The applicant used the CRAFT and THETA 1-B computer codes for the analysis of the blowdown phase of the transient. Using these I
codes, and the evaluation model specified in Appendix A, Part 4, of the Interim Policy Statement,.the applicant provided the re-evaluation of the FCCS performance in compliance with the Commission's Interim Policy Statement.
For the blowdown portion of the accident, we have concluded that the-applicant's analyses as reported in BAW-10034 conform to the require-ments specified in the Commission's Interim Policy Statement, Appendix A, Part 4.
6.3.3.3 Analysis of the Refill and Reflood Perio(
The applicant has considered the thermal behavior of the core during the refill and reflood portion of the loss-of-coolant accident, which is explained as follows:
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The vessel refill is provided initially by the core flooding tanks, start at the and later by the pumping systems, and is assumed to,s
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4 end of the blowdown period.- The reactor vessel ~is assumed to be essentially dry at the end of'the1 blowdown period, asia result of the' conservative assumption in Appendix A,,Part 4, of the Interim Policy Statement that water injected by. the core
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flooding tanks prior to the end of blowdown is ejected from the primary system.
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No heat transfer in the core is assumed until the level of water reaches the bottom of the core, at. which" time refill'is considered complete and the core reflood starts..
The end of blowdown is 14.6 seconds after rupture'for the 2
8.55 ft cold leg double-ended b'reak end refill to the bottom I
of the core is complete about 23 seconds after. rupture. -The end of blowdown is 18.7 seconds after rupture for_the 8.55 ft cold leg split and reflood is complete about 26 seconds after rupture.
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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'll to 12 inches per second, then decreases rapidly. leveling off at about 5.5 inches per second about 10 seconds after the end of blow-down. At 10 seconds after the end of blowdown, the water covers.
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i about 12 inches of the core for the case of a dcubl e-ended cold leg break and 20 inches of the core for the case 5 of a 8.55 ft2 cold leg split.
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The amount of steam generated in the core togeth er with the steam flow path is assumed to be only through the v ent valves within the reactor vessel and no credit is taken f or steam flow around the loop.
The steam flow resistance also limits the rate of liquid rise in the core, but the annulus water le vel continues to increase tatil the liquid level reaches th e inlet nozzle.
Core flooding tanks and low pressure injection system wat er is' piped directly to the reactor vessel with no intervening reactor coolant system piping.
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The peak temperature reached in the transient for the limiting 8.55 ft cold leg split occurs about 30 seconds after the break 3
i Based on our review of "Multinode Analysis of B&W's 2568 MWt Nuclear Plants During a Loss-of-Coolant Accident" BAW-10034, Revision dated May 1972, we have concluded that the applicant h as evaluated the refill and reflood events in an acceptable m j
anner.
6.3.3.4 Results l
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The applicant has calculated the following temperatur s f 4
e or Three Mile Island Unit 1 at 102% of a nominal power level of 2568 H Wt (requested power level is 2535 MWt):
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Cold Leg Pipe Breaks P_eak Clad Temperatures (*F)
(Area)
(Type Break) 8.55 ft (Double Ended) 2052 8.55 ft (Split) 2177*
2 3.0 ft (Split) 1652 0.5 ft (Split) 1614 r
Hot Leg 2
14.1 ft (Split) 1621 i
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- Limiting case.
The total core metal-water reaction is less than 1% for each of the
. assumed pipe breaks.
6.3.3. $ Small Break Analysis We have reviewed the applicant's analysis of the conse-Zen quences of small breaks requiring the operator of the emergency core cooling system.
The peak clad temperatures calculated by the applicant for this class cf breaks is less than 1100*F.
While we are in the process of performing a more detailed review of small break analyses for PWR plants on a generic basis, we have concluded that the information presently available on the Three Mile Island Unit 1 application provides adequate assurances of acceptable ECCS performance for small break accidents.
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. 6. 3.3.h Core Flooding' Tank I.ine Break Analysis
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We have reviewed the applicant's analysis of the consequences.-
of a double ended break in the line connecting the -core flooding tank to the reactor vessel.
The applicant provided a revised analysis for the Three Mile Island Unit 1 in report BAW 10034, 1972.
The evaluation model is similar
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Supplement 1,' dated August to that in the AEC Interime Acceptance Criteria for Emergency Core ' Cooling Sys tems, Appendix' A,,l' art 4, for B&W-internals vent valve plants with some modifications as requested for
..3 consideration of the break in the core flooding tank line.
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- 1, The analysis shows that the peak' cladding temperature is.
755*F for a nominal power 1cvel of 2568 Wt. 'A potential for t
metal water reaction or fuel pin swelling or ' rupture does not We conclude that
' exist because of the low clad temperature.
the AEC Interim Acceptance-Criteria have been met.
.s 6.3.4
-Conclusions On the basis of our evaluation of the additional B&W analyses,.
described in 6.3.3.1 above, we conclude that our acceptance criteria,
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as described in the Commission's Interim Policy Statement-have been -
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i met:
The maximum calculated fuel element cladding temperature does 1.
1 not exceed 2300*F.
I The amount of fuel' element cladding that reacts chemically withi 2.
i water or steam does not exceed 1% of the total amount of cladding in the reactor.
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3.. The calculated clad ; temperature 1 ransient ~is terminated ati a t
d time when the core geometry is still amenable to cooling,.an I
before the. cladding is so embrittled as to fail during or after t
quenching.
The core temperature is reduced and decay heat is removed for 4.
an extended period of time, as required by the 'long lived
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radioactivity remaining in the core.
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The results of the applicant's analyses for a loss-of-coolant' accident-
'l initiated at'a core power Icvel of 2568 MWt show that the acceptance criteria are met on the' basis of ana1 ses performed in accordance with-5
. an acceptable evaluation model givera 'in.the Interim Policy Statement.
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 provide adequate protection for.'any loss-of-coolant accident.
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