ML19312C368

Suppl 1 to SER for Facility

ML19312C368
Person / Time
Site:	Oconee
Issue date:	03/24/1972
From:	US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML19312C365	List: ML19312C364 ML19312C368
References
NUDOCS 7912130898
Download: ML19312C368 (14)

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SUPPLEMENT NO. 1 i

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I SA_FETY EVALUATION a

'l BY THE DIVISION OF REACTOR LICENSING i

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_U. S. ATOMIC ENERGY COMMISSION

__IN THE MATTER OF 1

DUKE POWER COMPANY 1,

_OCONEE NUCLEAR STATION UNIT 1_

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TABLE OF CONTENTS pag

1.0 INTRODUCTION

I 1.1 General......

1 1.2 Recent Experimental Informstion.

2 2.0 l)ESCRIPTION OF EMERCENCY CORE COOLING SYSTEM.......

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'I. 0 l'ERFORMANCE ANALYSIS OF EMERGENCY COOLING SYSTEM...

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3.1 General.....

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3.2 Analysis of the slowdown Period.....

9 3.3 Analysis of the Refill and Reflood Period......

9 3.4 Results.

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4.0 CONCLUSION

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1.0 I NTROD_UCTION 1.1 General The Safety Evaluation by the Division of Reactor Licensing dated December 29, 1970, included a description of the Oconee Nuclear Station Units 1, 2, and 3 emergency core cooling system (ECCS) and our evaluation of the performance analysir of this system for the spectrum of break sizes up to l

and including tric double-ended severance of the largest pipe of the reactor coolant pressure boundary. This evaluation was based upon ECCS analyses pertorned by the applicant and reported in the Oconee Nuclear Station operating license a:iCtration. Theoe analyses were performed using computer

.3<:en developed Fy 36W for analysis of large PWR reactors having safety injection systems, Subsequently, the Atomic Energy Comission has reevalw-ated the thenretical and exp+.rimental bases for predicting the performance of emergency core cooling systems, including new information obtained from industry and AEC research programs in this field. As a result of this reevaluation, the Cons mission has developed interim acceptance criteria for emer-gency core cooling 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

f used in the desJgn 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, l

(2) a simulated reactor primary system with a vessel and cingle operating loop.

(3) a single Icop 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 analytical technique (RELAP-3 code) used by ANC at that time for blowdown analysis did not accurately predict the phenomena that occurred during blow own 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 i

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. water would be ejected from the vessel af ter injection.

This phenomenon was observed in several cold leg Sendscale tests; the performance of the ECCS was satisf actory for the hot leg tests.

Although the LOFT Sendscale testr, 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 italet 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) steam 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 large 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 Coce mission's policy statement by including the conservative i

L=l fa 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 used calculational techniques to predict accumulator water behavior during blowdown. As further experimental information or improved calculational techniques become avail-able, this conservative assumption will be reevaluated.

2.0 DESCRIPTION

OF PMERGENCY CORE COOLING SYSTEM The Oconee Unit 1 cmergency core cooling system (ECCS) consists of a high pressure injection system, 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 sizes.

The high pressure injection sys tem includes three numps, each 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 boric acid solu-tion from the boric acid water storage tank is provided to the suction side of the high pressure pumps during ECCS operation.

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. 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 coolant system pressure of 1500 psig or a reactor building pressure of 4 psig. Automatic actuation switches the i

system 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 failure of an active component without a loss of function.

The two core flooding tanks are located in the con-tainment outside of the secondary shield. Each accumulator has a total volume of 1410 ft3 with a minimum stored borated water volume of 1040 f t 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 normally open stop valve adjacent to the tank. The core flooding tanks will therefore 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 i

a, spare pump each capable of delivering 3000' gpm at 100 psig f.

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. 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 take their suction f rom the borated water storage tank (initially) and the reactor building emergency sump. The recirculation system

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components are redundant so as to withstand a single failure of on active or passive component without loss of function at the required flow.

The low pressure injection system is actuated on a low reactor coolant system pressure of 500 psig or a high reactor building pressure of 4 psig.

All of the ECCS subsystens can accomplish their function when operating on emergency (onsite) power as well as of fsite power.

If there is a loss of normal power sources the engi-neered safeguards power line is connected to the Keowee hydro unit which will start up and accelerate to full speed in 23 seconds or less. The pumps and valves of the injection system will he energized at less.than 100% voltage and frequency to achieve the design injection flow rate within 25 seconds.

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, 3.0 PERFORMANCE ANALYSIS OF EMERCENCY CORE COOLING SYSTEM 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 assumptions and procedures used by B&W in analyzing the per-formance of the Oconee Unit No.1 ECCS are described in i

I Appendix A, Part 4 of the Interim Policy Statetent published in the Federal _R_egister on December 18, 1971 (F.R. Vol. 36, No. 244).

Report BAW-100a4 "Multinode Analysis of B&W's 2568 MWt Nuclear Plants During a Loss-9f-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 pins as will be the case for oconee Units 1 and 3.

Unit 1 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 maximum cladding temperature criteria. The applicant sub-mitted an analysis in Supplement 10 to support his claim that the 17.4 KW/f t.1. imitation would not result in power penalty I

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and that there would be adequate margin below thi s limit through core life.

For comparison, the analysis reported in BAW-10034 is based upon an 18.15 KW/ft peak linear heat rate for cores with pressurized pins only.

The 8.55 ft cold Icg split is the limiting case accident with a peak temperature of 2284'F in the case of the mixed core and 2177'F i the case of n

the pressurized pin only case.

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3.2 Analysis of the Blowdown Period The applicant used the CRAFI and THETA 1-B c omputer codes for the analysis of the blowdown phase of the t ransien t.

Using these codes, and the evaluation model s pecified in Appendix A, Part 4, of the Interim Policy Statement, the applicant provided the reevaluation of the ECCS p erformance in compliance with the Commission's Interim Policy St atement.

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, conform to the re quirements specified in the Commission's interim Policy Stat

ement, Appendix A, Part 4.

3.3 Analysis of the Refill and Reflood Period The applicant has considered the thermal behavi or of the core during the refill and reflood portion of the l oss-of-coolant accident, which is explained as follows:

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. (1)

The vessel refill is provided initially by the core flooding tanks, and later by the pumping systens, and is assumed to start ac the end of the blowdown period. The reactor vessel is assuned to be essentially dry at the end of the blowdown period, as a result of the con-servative assumption in Appendix A, Part 4, of the Interim Policy Statement that water injected by the core j

flooding tanks prior to end-of-blowdown is ejected from 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 starts.

The end of blowdown is 14.6 seconde af ter rupture for the 8.55 ft cold leg 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 af ter rupture for the 8.55 fc cold leg split and reflood is complete about 26 seconds after s

rupture.

(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 enour.h of the core is covered to generate substantial amounts of steam. The re-flood rate t-i i

a increaucu 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 blowdown.

t At 10 seconds after the end of blowdownthe water covers about 12 inches of the core for the case of a d ubl o

e ended cold leg break and 20 inches of the core for thecase of a 8.55 ft cold leg split.

(4)

The amount of steam generated in the core t ogether with the steam flow path resistance govern s the rate of steam flow.

The steam flow path is assumed to be only th rough the vent valves within the reactor vessel and no credit is taken for steam flow around the JoopThe steam flow resistance also limits the rate of liquid rise in the core, but the annulus water icvel continues to increase until the liquid level reaches the inlet nozzle.

Core flood tanks and low pressure injection syste j

m water is i

piped directly to the reactor vessel with no int ervening reactor coolant system piping.

(5)

The peak temperature ~ reached in the transi ent for the limiting 8.55 ft cold leg split occurs about 30 seconds after the break.

Based on our review of "Multinode Analysis of B&W' s 2568 WWt Nuclear Plants During a Loss-of-Coolant Ac id c ent" BAW-10034, October 1971, and Supplement 10 to th t

e FSAR we have i

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concluded that the applicant has evaluated the refill and rei tood events in an acceptabic manner.

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!!psults The applicant has calculated the following temperatures f or Oconce Unit No. 1 at 102% of a nominal power level of 2568 MWL:

_ Cold 1,eg Pipe Breaks Peak Clad Temperatures (*F) 1 Area)

(Type Break)

Pressurized Pins Unpressurized Pins 8.55 ft (Double Ended) 2052 2072 8.55 ft (Split) 2177*

2284*

3.0 ft (Split) 1652 1662 0.5 ft (Split) 1614 1561 Ilot leg 14.1 ft (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 maximum calculated fuel elecent 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 amount of cladding in the reactor.

(3)

The calculated clad temperature transient is terminated at a time when the core geometry is still amenable to i

cooling, and before the cladding is so embrittled as to fail 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.

i These are the sane acceptance criteria that we stated on pages 42 and 43 of our Safety Evaluation on Oconee Unit 1.

j The results of the applicant's analyses for a loss-of-coolant accident initiated at a core power level of 2566 MWt show that the acceptance criteria are met on the basis of analyses performed in accordance with an acceptable evaluation model given in the Interim Policy Statement.

On the basis of our evaluation of the additional B&W analyses described in 3.1 above, we have determined that the i

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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-ation dated December 29, 1970, remains applicable for the Oconee Nuclear Station reactors fc. : ore powers up to 2568 MWt.

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