ML19317F488
| ML19317F488 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/24/1972 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19317F486 | List: |
| References | |
| NUDOCS 8001140719 | |
| Download: ML19317F488 (16) | |
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SUPPLEFINT NO. 1
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_ SAFETY EVALUATION
.BlTj!E, 1
DIVISION OF REACTOR LICENSING L
U. S. ATOMIC ENERGY COMMISSION IN THE MATTER OF DUKE POWER COMPANY OCONEE NUCLEAR STATION UNIT 1 4
DOCKET NO. 50-269 M 2 41972 1
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_T_ABLE OF CONTENTS f.a gg 1.0.
INTRODUCTION.
I 1.1 Cencral'......
1 1.2 Recent Experimental information.............
2 2.0 Dr.SCRIPTION OF EMERCENCY CORE COOLING SYSTEM.
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PERFORMANCE ANALYSIS OF EMERGENCY COOLING SYSTEM.....
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. Gent?fal...
8 1.2 Analynis of the Blowd own Period.............
9 1.1 Analyain of the Refill and Reflood Period..
9 J.4 Results....
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4.0 CONCLUSION
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-i tith 0yUCTION 1.1
_ General The Safety Evaluation by the Division of Reactor 1,1 censing dated December 29, 1970, included a description of the Oconce Nuclear Station Units 1, 2, and 3 emergency core cooDing system (ECCS) and our evaluation of the perfor mance
. analysis of this system for the spectrum of break si zes up to and including the double-ended severance of the large st pipe of the reactor coolant pressure boundary.
This evaluation was based upon ECCS analyses perforced by the applicant
,and reported in the Oconee Nucicar Station operating licen se application.
These analyses were performed using computer codes developed by B&W for analysis of large PWR react ors having safety injection systems.
Subsequently, the Atomic Energy Commission has re evalu-ated the theoretical and experimental basca for predi cting the I
performance of emergency core cooling systems
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, including new information obtained from industry and AEC resear h c programs in this field.
As a result of this reevaluation, the Com-
- mission has developed interim acceptance criteria for emer-gency core cooling systems for light-water power reacto rs.
These criteria are described in an Interim Policy Stat ement issued on June 25, 1971, and published in the Federal
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1,tyl ntg on. lune 29, 1971, (36 F.R. 1224 7).
By letter dated i
.luly 9,1971, the Division of Reactor Licensing inf ormed the applicant of the additional information that would be required for our evaluation of the performance of the Oconee Unit No.1 ECCS in accordance with the Interim Policy Statecent. The applicant provided a revised analysis of the Oconee Nuclear Station Unit 1 performance in a report titled "Multinode Analysis of B&W's 2568-MWt Nuclear Plants During a Loss-of-Coolant Accident" dated October 1971. The applicant also pro-vided a supplement to this report, identified as Supplement 10 t o the Oconce FSAR and dated December 17, 1971, that discusses I he analysis of ECCS using unpressurized fuel pins in Oconee llalts 1 and 3.
The analysis was perf ormed using the B&W Evaluation Model in conformance with the Interim Policy State-ment, Appendix A, Part 4.
The analysis was perforced assuming the occurrence of a loss-of-coolant accident during operation at 102% of the requested power icvel of 2568 MW thermal.
1.2 f(ecent Experimental Information Small-scale experiments have been conducted by the Aero-jet Nuclear Corporation (formerly Idaho Nuclear Corporation) under contract to the U. S. Atomic Energy Commission as part of the reactor safety research and development work being carried out at the National Reactor Testing Station, prin-cipally to assist in the development of analysis methods to be
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_3 used in the deaf gn and execution of the LOFT Project. Du ring the past several years tests under this program have been per-formed to investigate the phenomena of blowdown of heated high pressure water from:
(1) a simulated reactor vessel with and without internals, (2) a nimulated reactor primary system with a vessel and singic operating loop.
(3) a single loop system with an electrically-heated simulated reactor core, and (4) a single loop, electrically-heated core system with accumulator ECC injection.
The results of some of these tests (LOFT Semiscale series 845-851) conducted in late 1970 and early 1971 showed that the at that time analytical technique (RELAP-3 code) used by ANC
. for blowdown analysis did not accurately predict the phenomena
' that occurred during blowdown af ter the cold ECCS water was
' introduced.. The analysis had assumed that uniform and instantaneous mixing of the cold injection water and the hot
- residual fluid took place 'in the appropriate zones of the Semiscale system. The test showed that mixing is incomplete.
'In addition, the analysis did not predict that the cold ECCS 4
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water would be ejected from the vessel after injection.
This phenomenon was observed in several cold leg Sendscale tests; the performance of the ECCS was satisfactory for the hot leg tests.
Although the LOFT Sendscale tests in this series have provided information for evaluation of the adequacy of analy-tical models, the results of these tests cannot be applied di rectly to describe the performance of pressurized water reactors following a loss-of-coolant accident because the test loop used was not designed so as to properly scale parameters af fecting system performance. These include (1) the elevation head of the inlet annulus water, (2) the ratio of steam bubble diameters to the width of the vessel inlet annulus, (3) multiple flow loops, (4) relative loop and core resistances, (5) containment back pressure, (6) surface to volume ratios, (7) pump flow resistance, (8) s team generator model, (9) core heat rate, and (10) core internals.
Although the results of the small LOFT Semiscale experi-ments would not be expected to describe the performance of larAc power reactors, we have taken into account the results of these tests _ in establishing the acceptability of PWR interim evaluation models listed in Appendix A of the Com-minulon's policy statement by including the conservative p
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nsnumption that all of the water injected by the accumulators during blowdown is lost. ' Another consideration that led to this conservative assumption was the inadequacy of the cur-rently'uned calculational techniques to predict accumulator water behavior during blowdown. As further experimental information or improved calcelational techniques become avail-abic, this conservative assumption will be reevaluated.
2.0 DESCRIPTION
OF EMERGENCY CORE COOLING SYSTEM The Oconee Unit 1 emergency core cooling system (ECCS)
J consists of a high pressure injection sys tem, an injection system employing core flooding tanks, and a low pressure injection system with external (to the containment) recir-culation capability.
Various combinations of these systems are employed to assure core cooling for the complete range of break sizen.
The high pressure injection sys tem includes three numps, cach capable of delivering 450 gpm at 585 psig reactor vessel pressure and discharges to the reactor coolant inlet lines.
One pump will provide the required minimum flow.
The high pressure injection pumps are located in the auxiliary building adjacent to the containment.
A concentrated horic acid solu-tion from the boric acid water storage tank is provided to the auction side of the high pressure pumps during ECCS operation.
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During normal reactor operation, the high pressure injection l
e ctor coolant for purification and for system recircu ates r a supply of seal water to the reactor coolant circulation pumps.
The high pressure injection system is initiated at a low reactor coolant system pressure of 1500 psig or a reactor Automatic actuation switches the building pressure of 4 psig.
One of the system from normal to emargency operating mode.
The three high pressure pumps is normally in operation.
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 con-Each accumulator tainment outside of the secondary shield.
3 with a minimum stored borated has a total volume of 1410 f t water volume of 1040 ft pressurized with nitrogen to 600 psig.
3 Each accumulator is connected to a separate reactor vessel core flooding nozzle by a flooding line incorporating two check valves and a motor operated normally open stop valve The core flooding tanks will therefore adjacent to the tank.
inject water automatically whenever the pressure in the primary system is reduced below the core flooding tank pres-sure of 600 psig.
The low pressure injection system includes two pumps plus a spare pump each capable of delivering 3000 gpm at 100 psig N 6
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he reactor vesset pressure arranged to deliver water to t One low reactor vessel through two separate injection lines.
rgy pressure injection pump is capable of removing the heat ene accident.
generated af ter n loss-of-coolant The low pressure injection system pumps take their (initially) and nuction f rom the borated water storage tank The recirculation system the reactor building emergency sump.
d a single f ailure components are redundant so as to withstan f function at of an active or passive component without loss o the required flow.
The low pressure injection system is actuated on a low tr reactor coolant system pressure of 500 psig or a high reac o buildJ ng _ pressure of 4 psig.
All of the ECCS subsystema can accomplish their function ffsite when operating on emergency (onsite) power as well as o llf there is a loss of normal power sources the engi-puwer.
hydro neered safeguards power line is connected to the Keowee d in 23 unit which will start up and accelerate to full spee The pumps and valves of the injection system seconds or less.
to will be energized at less.than 100% voltage and frequency d
achieve the design injection flow rate within 25 secon s.
1.0 PERFORMANCE ANALYSIS OF EMERGENCY CORE COOLING SYSTEM
'i. I Ceneral 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 assumptions and procedures used by B&W in analyzing the per-formance of the Oconee Unit No. 1 ECCS are described in Appendix A, Part 4 of the Interim Folicy Statenent published 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," October 1971, covers the performance of cores for which all fuel pins are pressurized. In addition, Supplement 10 of the FSAR pre-sents the B&W LOCA analysis for cores having unpressurized pina as will be the case for Oconee Units 1 and 3.
Unit I will have unpressurized and pressurized (a mixture) pins for the first two cycles and Unit 3 will have a mixture for the first cycle only. The analysis for the core with a mixture of pressurized and unpressurized pins resulted in a heat rate limit of 17.4 KW/f t, for the 102% power case to meet the 2300*F maxinnim cladding temperature criteria. The applicant sub-l mitted an analysis in Supplement 10 to support his claim that the 17.4 W/f t. limitation would not result in power penalty K
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and that there would be adequate margin below this limit through core life. For comparison, the analysis reported in BAW-10034 is based upon an 18.15 gw/f t peak linear heat rate for cores with pressurized pins only. The 8.55 ft cold Icg split in the limiting case accident with a peak temperature of 2284*F in the case of the mixed core and 2177'F in the case of the pressurized pin only case.
's. 2 Analysis of the Blowdown Period The applicant used the CRAFT and THETA 1-B computer codes for the analysis of the blowdown phase of the transient.
Using these codes, and the evaluation model specified in Appendix A, Part 4, of the Interim Policy Statenant, the applicant provided the reevaluation of the ECCS performance in comp]1ance with the Commission's Interim Policy Statement.
For the blowdown portion of the accident, we have con-cluded that the applicant's analyses as reported in BAW-10034 and Supplement 10 of the FSAR, conf orm to the requirements npecified in the Commission's Interim Policy Statement, Appendix A, Part 4.
3.3 Analysis of the Refill and Reflood Period
_ - 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, and later by the pumping systems, and is assumed to start at the end of the blowdown period.
The reactor vessel is assuned to be essentially dry at the end of the blowdcun period, as a result of the con-servative assumption in Appendix A, Part 4, of the interim Policy Statenant that water injected by the core flooding tanks prior to end-of-blowdevn is ejected f rom the primary system.
(2) 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 utarts.
The end of blowdown is 14.6 seconds after rupture for the 8 55 ft cold Icg double ended break and reflood (to the bottom of the core) is complete about 23 seconds after rupture The end of blowdown is 18.7 seconds after rupture for the 8.55 f t2 cold leg split and reflood is complete about 26 seconds after rup tu re.
(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 re-flood rate H
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Increanea and peaks Ln ahont H.S secondo af ter the end of blowdown at about 11 to 12 inches per second, then decreases rapidly levelir g of f at about 5.5 inches per second about 10 seconds af ter the end of blowdown. At 10 neconda. after the end of blowdown, the water covers about 12 inches of the core for the case of a double ended cold leg break and 20 inches of the core for the case of a 8.55 ft cold leg split.
(4) The amount of steam generated in the core together with the steam flow path resistance governs the rate of steam flow. The steam flow path is assured to be only through the vent valves within the reactor vessel and no credit 18 taken for steam flow around the loop. The s team flow resistance also limits the rate of 11guid rise in the core, but the annulus water level continues to increase until the liquid level reaches the inlet nozzle. Core flood tanks and low pressure injection system water is piped directly to the reactor vessel with no intervening reactor coolant system piping.
(5) The peak temperature ' reached in the transient for the 2
limiting 8.55 f t cold leg split occurs about 30 seconds af ter the break.
Based on our review of "Multinode Analysis of B&W's 2568 MWL Nucicar Plants During a Loss-of-Coolant Accident" BAW-10034, October 1971, and Supplement 10 to the FSAR we have
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ennelnded that the applicant has evaluated the refill and retlood events in.an acceptabic manner.
l.4 Itenu l t s The appilcant has calculated the following temperatures for oconec Unit No. I at 102% of a nordnal power level of 2568 MWt:
Cold I.cg Pipe Breaks _
Peak Clad Temperatures (*F) 1 Area)__
_(Type Break)
_P_ressurized Pins Unpressurized Pins 8.55 f t (Double Ended) 2052 2072 8.55 l't (Split) 2177*
2284*
3.0 f t (Split) 1652 l'662 0.5 f t (Split) 1614 1561 Ilot leg L4.1 f t (Split) 1621 1605
- l.imiting case.
The total core metal-water reaction is less than 1% for
-each of the assumed pipe breaks.
4.0 CONCLUSION
S On the basis of our evaluation of the additional B&W analyses, described in 3.1 above, we conclude that our accept-ance criteria, as described in the Commission's Interim Policy Statement have been met:
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(1) The maximina calculated fuci element cladding temperature does not exceed 2300'F.
(2) The amount of fuel element cladding that reacts chemi-cally with water or steam does not exceed 1% of the total acount of cladding in the reactor.
(3) The eniculated clad temperature transient is terminated at a time when the core geometry is still amenable to cooling, and before the cladding is so embrittled as to f all during or af ter 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.
These are the same acceptance criteria that we stated on pages 42 and 43 of our Safety Evaluation on Oconee Unit 1.
Tne results of the applicant's analyses for a loss-of-coolant accident initiated at a core power icvel of 2568 MWt show that the acceptance criteria are met on the basis of analyses perforned in accordance with an acceptable evaluation model given in the Interim Policy Statcaent.
On the basis of our evaluation of the additional B&W analyses described in 3.1 above, we have determined that the f-tr
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conclusion that the emergency core cooling system is accept-able and will provide adequate protection for any loss-of-coolant accident, as set forth on page 43 of our Safety Evalu-ntion dated December 29, 1970, remains applicable for the i
.0conee Nuclear Station reactors for core powers up to 2568 me..
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