ML20214X368

Revised Chernobyl Nuclear Reactor Accident & Implications Upon Fort St Vrain

ML20214X368
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Site:	Fort Saint Vrain
Issue date:	11/26/1986
From:	GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER, PUBLIC SERVICE CO. OF COLORADO
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ML20214X361	List: ML20214X358 ML20214X368
References
TAC-60400, NUDOCS 8612110086
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t CHERNOBYL NUCLEAR REACTOR ACCIDENT AND ITS IMPLICATIONS UPON FORT ST. VRAIN PREPARED BY GA TECHNOLOGIES INC.

1 AND PUBLIC SERVICE COMPANY OF COLORADO REVISED NOVEMBER 26, 1986 4

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FOREWORD On May 9,1986, Public Service Company of Colorado (PSC) submitted a white paper entitled, "Chernobyl Nuclear Reactor Accident and Its Implications-Upon Fort St. Vrain" to the NRC. The white paper was i

prepared in response to a verbal request from the NRC staff that PSC assess the ' possibility of a " graphite fire" occurring at Fort St.

Vrain (FSV). (Ref. 1)

The May 1986 paper was based on the very limited information j

available at the time of the Chernobyl accident.

In the last 6 j

months much additional information about the accident has become available from various sources, most notably the Soviet report

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presented at the IAEA meeting in Vienna, August 25 through 29, 1986.

In a letter dated September 16, 1986, the NRC requested that PSC 3

-review the official (IAEA) translation of the Soviet report on the

~Chernobyl accident and use its information, as appropriate, to update the May 1986 white paper (Ref. 2). PSC has reviewed this report and several other sources of information on the Chernobyl accident.

While nothing in these sources has changed the conclusions of the May 1986 white paper regarding graphite oxidation, these sources have provided new insights and indications of lessons to be learned from the accident.

Many of the insights and lessons to be learned are generic in nature, i.e., they are applicable to all reactor types.

~

l Based on discussions with the NRC staff on October 17, 1986, this revised white paper focuses primarily on those aspects of the i

accident related to the use of graphite in FSV as the major core structural material.

In addition to addressing the issue of air J

ingress and graphite oxidation, the revised paper examines steam ingress and generation of the steam-reaction products, hydrogenandcarbonmonoxide(watergas) graphite t

4 Studies of graphite at GA Technologies Inc., Oak Ridge, Brookhaven, j

and Argonne National Laboratories have shown that self-sustained 4

burning of large pieces of nuclear grade graphite like those used in U.S.

graphite-moderated reactors is very difficult to attain.

i Extremely high temperatures and/or an oxygen-enriched atmosphere are required, or the graphite must be broken into very small pieces, as i

may have occurred to some extent at Chernobyl (Ref. 3).

Information j

in the Soviet report and subsequent analyses indicated that graphite oxidation was not the sole, or even a major, contributor to fission product releases. The initial steam explosion and the subsequent t

heatup caused by the trapping of core decay heat by the tons of boron j

carbide, dolomite, clay, sand, and lead used to smother the core were i

the ' major contributors to short and longer term fission product releases.

j 1

i As in the May 1986 white paper, the term " graphite fire" has not been used in this analysis. The concept of a free-burning fire or open i

flames with respect to a graphite mass in a confined volume having a i

restricted source of oxygen is inappropriate.

The term " graphite oxidation" is therefore used in this analysis.

1 3 1 a

No credible mechanism could be identified that would cause a sufficient number of failures to result in significant FSV graphite oxidation.

However, to respond to the basic intent of the NRC's May 1986 inquiry, this paper postulates sufficient failures of the largest redundant reactor pressure boundaries to achieve a worst-case graphite oxidation accident scenario.

It should not be concluded that a significant graphite oxidation scenario could realistically occur.

Similarly, a credible mechanism could not be identified that would cause a detonable mixture of hydrogen and carbon monoxide (water gas) and air to exist within the prestressed concrete reactor vessel (PCRV) cavity. However, for the purpose of assessing PCRV structural

margin, it was postulated nonmechanistically that such a mixture was present and was detonated.

It should likewise not be concluded that a

significant air and water gas detonation scenario could realistically occur.

I 11

CONTENTS FOREW0RD.......................................................

i t

1.

INTRODUCTION...............................................

1 2.

DESIGN DIFFERENCES:

FSV VERSUS RBMK.......................

3 2.1.

Fuel Design.........................................

3 2.2.

Primary Coolant.....................................

3 2.3.

Reactor Xinetics.....................................

4 I

2.4.

Primary System Enclosures...........................

4 2.5.

Core Cooling Capability.............................

5 2.6.

Secondary Containment...............................

6 2.7.

Conclusion...........................................

6

(

3.

AIR INGRESS AND GRAPHITE 0XIDATION.........................

7 3.1.

FSV Licensing Basis.................................

7 3.2.

Double DBA-2.........................................

8 4.

STEAM INGRESS AND WATER GAS GENERATION.....................

14 4.1.

FSV Licensing Basis................................

14 4.2.

Water Gas Detonation.................................

14 5.

CONCLUSIONS...............................................

16 6.

REFERENCES.................................................

17 i

1.

INTRODUCTION The accident at the Chernobyl RBMK nuclear power plant.in the Soviet-Union resulted in Nuclear Regulatory Commission inquiries at the time of - the accident concerning the possibility of graphite oxidation at 1

Fort St. Vrain (FSV). Graphite oxidation was previously addressed and reviewed thoroughly during the licensing of FSV.

The FSV-construction permit and operating license were issued based on the NRC's conclusion

"...that the potential amount of air ingress and (graphite) oxidation would be small and would not appreciably affect the radiological consequences." (Ref. 4)

In May 1986, Public Service Company of Colorado (PSC) and GA Technologies Inc. (GA) conducted another review of FSV in light of the ' Soviet. accident.

This review's purpose. was to determine if graphite oxidation, such as experienced at Chernobyl, could occur at FSV and how such a postulated reaction could be extinguished.

To have graphite oxidation, sufficient oxygen and heat must be

_ present in the prestressed concrete reactor vessel (PCRV). The PCRV provides both primary and secondary containment of the reactor. To achieve this oxidation in the inert helium atmosphere of the core, the PCRV must be breached in such a manner as to create an airflow path. Only one combination of failures can be postulated which would establish the " chimney effect" flow conditions necessary to obtain significant oxidation.

This chimney effect can only be established by the simultaneous failure of both the primary and secondary closures of a _ top PCRV penetration, accompanied by a similar failure of both closures of a bottom PCRV penetration. This incredible accident, well beyond the FSV design basis and FSAR licensing basis, was developed to consider massive graphite oxidation as a worst-case bounding scenario for air ingress into the PCRV.

This postulated event is essentially the occurrence of two simultaneous Design Basis Accident No.

2 (DBA-2),

Rapid Depressurization/ Blowdown, each of which has an estimated median frequency of 1 x 10E-9 per year, per Revision 4 of the Updated FSAR Section 14.11.1.4.

The Soviet report on the Chernobyl accident presented at the IAEA conference in Vienna, August 25 through 29, 1986, (pg. 21 of Ref. 5) indicated that a mixture of gases containing hydrogen and carbon monoxide capable of explosion upon mixing with air was formed during the accident. This gas mixture (known as ' water gas") was formed as a result of steam-graphite reactions and zirconium-water reactions that occurred following pressure tube failure.

Although it is not clear whether a water gas-air detonation occurred at Chernobyl, PSC and GA have recently conducted another evaluation of FSV to determine if detonable water gas mixtures could be formed in the FSV PCRV and whether PCRV integrity could be retained in the event of a water gas detonation.

1

A wide range of accidents involving steam ingress into the PCRV and steam-graphite reactions was addressed during the licensing of FSV.

Steam generator tube failures,

in combination with a variety of protective action failures, were considered. None of these accident scenarios resulted in PCRV overpressurization and potential air ingress.

In no case was a mechanism identified to obtain the significant air ingress required to create a detonable air-water gas mixture. However, for the purpose of assessing PCRV structural

margin, an evaluation of the effects of detonation of a stoichiometric air-water gas mixture within the PCRV cavity has been conducted.

Based on available information and this review, PSC and GA conclude that the RBMK reactor design aspects that were a major contritutor to the Chernobyl accident are not present at FSV. The nonmectanistic accident scenario necessary to permit significant air ingress at FSV is highly incredible.

The air ingress, and thereby the graphite oxidation, could be halted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by flooding the Reactor Building with firewater up to the bottom PCRV penetration. Offsite dose consequences would still be within 10CFR100 guideline 3.

No scenario that results in a detonable mixture of air and wa:er gas within the PCRV cavity could be identified. However, even if such a mixture is assumed to exist and to be detonated, resulting stresses are well within the capability of the PCRV structure, anc PCRV integrity is retained.

l i

2

2.

DESIGN DIFFERENCES:

FSV VERSUS RBMK Although both reactors have a graphite moderator, there are significant fundamental differences between the FSV high temperature, gas-cooled reactor (HTGR) and the Chernobyl pressure tube-type heterogeneous uranium-graphite boiling-water reactor (RBMK).

These include the fuel design, the primary coolant, reactivity control, the primary system enclosure, the core cooling capability, and the secondary radioactivity barriers.

Based upon the available information, it is clear that many of the design features which led to the accident at Chernobyl are not present at FSV.

2.1.

FUEL DESIGN The FSV HTGR utilizes ceramic-coated TH/UC2 and THC2 fuel particles embedded in a graphite matrix within graphite fuel blocks, which are cooled by helium flowing through channels in the block. The particle coatings are the primary barrier to fission product release.

The fuel particles are thermally coupled with the graphite moderator, resulting in a core that heats up slowly during power or flow transients.

This slow thermal response, in combination with a large margin between normal peak operating temperatures and the temperatures at which particle coating failures occur, contributes to the basic safety of the FSV fuel design.

(Ref. 6, 7 and FSAR Section 3.6)

The RBMK utilizes zirconium alloy clad U02 fuel elements within stainless steel and zirconium alloy pressure tubes located within the central holes of graphite columns. The fuel elements are cooled by light water that boils as it passes through the core in the pressure tubes.

Thus, the fuel elements are not thermally coupled with the graphite moderator in the RBMK core, and the thermal response time of the core is similar to that of light water reactors. The Soviet report indicates that conditions for steam-zirconium and other exothermic chemical reactions were created during the Chernobyl accident. The FSV helium coolant is chemically inert.

2.2.

PRIMARY COOLANT The FSV HTGR coolant is helium that is circulated downward through the core to the steam generators, located beneath the core, before returning to the top plenum through the helium circulators. These circulators are driven by steam during power operation, and auxiliary steam or water from numerous sources during shutdown cooling.

The RBMK coolant is light water that is boiled as it passes through the core in the pressurized metallic tubes.

The steam produced in the reactor is fed to the turbine via a separator drum, and its condensate is directed to the reactor circulation loop.

Thus the RBMK reactor is a direct cycle system with multiple potential pathways for air ingress into the core and for fission product escape to the atmosphere.

3

2.3.

REACTOR KINETICS Based on the Soviet report, the major design-related contributor to the Chernobyl accident was the positive void coefficient of reactivity that resulted in a steep power excursion. Core and coolant conditions at the time of the accident were such that small changes in coolant quality resulted in large void changes.

Evaluations of the accident funded by the Department of Energy, as described in Ref.

3, indicate that a very large, rapid reactivity insertion occurred. The power excursion caused rapid fuel

failure, fuel-coolant and cladding-coolant reactions, and a steam explosion that caused failure of multiple pressure tubes, core disassembly, and-destruction' of the reactor enclosure and Reactor Building. The positive void coefficient at Chernobyl results from the fact that the core is ovennoderated; i.e., thermal neutron leakage is so small that water acts as a neutron absorber when compared to graphite.

Hence, coolant voiding improves the core neutronics and results in a positive reactivity insertion.

In FSV the fuel and isothermal coefficients of reactivity are negative over the entire range of operation. The two phase steam-water coolant in the RBMK core, in a combination with other factors, can result in a positive reactivity insertion when primary coolant is lost or void fraction increased. The use of the single phase helium coolant in FSV prevents any similar positive reactivity insertion

~

upon loss of primary coolant inventory or pressure.

In fact, there would be a slightly negative reactivity insertion in the event of the loss of the helium.

The FSV core is undermoderated, so only the addition of steam to the core can contribute positive reactivity.

However, the rate of reactivity addition resulting from the worst-case steam ingress scenario described in FSAR Section 14.5 is very slow, and the total reactivity insertion associated with that scenario is smaller than that associated with the rod withdrawal accidents evaluated in FSAR Section 14.2.2.

Hence, the reactor kinetics characteristics of FSV preclude the type of positive reactivity insertion experienced at Chernobyl.

2.4.

PRIMARY SYSTEM ENCLOSURES l

The entire FSV helium primary coolant system is enclosed within the structurally redundant steel-lined PCRV that operates at a pressure i

of 700 psia.

The PCRV is designed to withstand over 1700 psig without failure and has multiple plant protective system features and safety valves that prevent pressure from exceeding the Reference i

Pressure of 845 psig. The PCRV liner serves as the second barrier to fission product release (the ceramic carbon fuel particle coatings serving as the primary barrier to fission product release).

PCRV i

penetrations are each provided with two structurally independent, i

redundant closure seal assemblies. The primary closures, located at l

the PCRV interior, in effect serve as an extension of the PCRV liner t

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as the second fission product barrier. The secondary closures are located at the PCRV exterior.

The space between the primary and secondary closures is filled with pressurized helium gas.

In conjunction with the PCRV prestressed concrete, the secondary closures serve as the third barrier to fission product release and perform a ~ function that is largely analogous 'to that of an LWR containment building.

The Chernobyl core is contained within a low-pressure reactor enclosure in a concrete well, confined by metal lower and upper supports and a metal cylindrical jacket. The reactor space is filled with a mixture of helium and nitrogen with surrounding spaces filled with backup nitrogen.

In the RBMK design, the boiling water coolant passes through the low-pressure (0.08 MPa) enclosure space to and from the turbines.

It also features a 0.25 MPa pressure suppression type containment designed to accommodate the failure of only a

' limited number of pressure tubes. (Ref. 5) 2.5.

CORE COOLING CAPABILITY The FSV plant has two main cooling loops, either of which is capable of emergency cooling with the primary system pressurized or depressurized.

Each loop has two helium circulators, which can be operated by cold reheat steam, bypass flash tank steam, auxiliary boiler steam, feedwater, condensate, and firewater.

For credible accidents, one of the four helium circulators provide ample flow for core heat removal via the steam generators. Core heat removal can be accomplished via the feedwater, condensate, or firewater systems through the steam generators. Electrical power to provide this core cooling function would be available from any of the five off-site transmission lines, or either of the two emergency diesel generators.

The ceramic-coated uranium-carbide fuel and the graphite core's high thermal inertia permits a 1 1/2 hour delay before resumption of core cooling is needed to prevent fuel damage after a loss of forced helium circulation.

The reactor plant cooling water and firewater systems can remove decay heat via the PCRV liner cooling tubes, if the steam generators or helium circulators cannot be used.

A dedicated ACM diesel generator can power the electrical equipment needed for PCRV liner cooling. Establishment of PCRV liner cooling can be delayed 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> following a permanent loss of forced circulation.

These various forced circulation and PCRV liner core cooling methods provide a wide range of diverse cooling that could be used in the event of an emergency.

In contrast, the RBMK reactor Emergency Core Cooling System (ECCS) must be automatically operated very quickly, to provide water to prevent fuel melting, massive overheating and seal failure in the fuel elements.

The RBMK ECCS consists of several complicated subsystems, of which a number must operate to ensure core cooling.

5

Based on the exceptionally long time span before fuel failure, FSV has a considerable advantage in its emergency cooling requirements over the RBMK reactor.

2.6.

SECONDARY CONTAINMENT Both the RBMK and FSV plants are housed in a Reactor Building.

However, the FSV secondary penetration closures on the PCRV provide a second independent closure to provide radioactivity retention. Thus, FSV has a secondary containment built into the massive PCRV that protects the ceramic fuel. Both primary and secondary penetration closures are constructed to the requirements of ASME Section III.

The FSV Reactor Building also confines and delays release of radioactive material.

It features a filtered ventilation system for fission product cleanup.

A louver system protects the building against overpressurization.

2.7.

CONCLUSION There are many more differences than similarities between the FSV HTGR and the Chernobyl RBMK. Many of the design features that played a major role in the Chernobyl accident are not present at FSV. The FSV plant has clear and well-recognized design safety features that were included in the conceptual design phase. All major coefficients of reactivity are negative over the entire range of operation.

Prevention against graphite oxidation by air or steam was a major consideration in the selection of ceramic coated high-temperature fuel, the helium coolant, and the PCRV design with two independent closures at each penetration.

6

3.

AIR INGRESS AND GRAPHITE OXIDATION 3.1.

FSV LICENSING BASIS The potential consequences of air ingress into the PCRV were considered during the design and licensing of FSV.

The licensing basis presented in the FSAR considers a wide spectrum of incidents and accidents.

Of these

events, DBA-2 (Rapid Depressurization/ Blowdown), has the largest potential for air ingress and graphite oxidation. 08A-2 is described in FSAR Section 14.11; it consists of a

hypothetical simultaneous failure of the two independent closures in the PCRV bottom head access penetration.

An analogous failure for an LWR would involve simultaneous failure of the primary coolant boundary and the containment, which is a scenario considered to be beyond the design basis for an LWR.

The FSAR provides a discussion of credibility of this accident. The accident is considered to be entirely hypothetical, because the PCRV penetrations have been designed to prevent such failures. Design featuras incorporated to achieve this objective include redundant closures designed and fabricated to ASME Section III requirements to withstand up to 2.1 times Reference Pressure (2.5 times normal operating pressure), two independent means of transferring the pressure loads from the primary closures to the PCRV concrete, and structural independence of the redundant closures.

Nevertheless, in DBA-2 a rapid, uncontrolled primary system depressurization (about 2 minutes in duration) is postulated due to the hypothetical simultaneous failure of the two independent closures in the PCRV bottom head access penetration.

Failure of the closures of this penetration results in the largest possible PCRV depressurization area and rate.

The reactor coolant, with its circulating activity, is released from the PCRV to the Reactor Building, along with a fraction of the removable plated-out fission products.

The results of the FSAR analysis of this event show that adequate primary circuit cooling would be maintained following the depressurization by use of the circulators on either steam or water turbine drive with the reduced coolant density.

No damage to the circulators would occur during the depressurization. Maximum fuel temperature following the depressurization remains below that at which, according to the conservative fuel failure model in the FSAR, rapid fuel particle coating degradation is expected to occur.

Section 14.11.2.3 of the FSAR discusses the potential for ingress of air into the PCRV following the DBA-2 blowdown.

The primary air ingress mechanisms are breathing due to PCRV cooldown and diffusion of air through the open penetration.

Analyses of these phenomena indicate that the resultant graphite oxidation rate is less than 3 lb per hour, and the heat generated from the reaction is less than 1% of the reactor decay heat.

In summary, the consequcnces of air ingress and graphite oxidation were considered in the FSV licensing basis.

No accidents that led to 7

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significant air ingress were identified. Two design basis accidents, the permanent Loss of Forced Circulation of the helium coolant (DBA-

1) and the rapid depressurization of the PCRV (DBA-2), were selected as being representative of accidents that, although they are highly improbable, have the potential for significant consequences. The ability of the plant and its engineered safety features to control the possible escape of fission products from the facility were assessed.

The calculated potential consequences of the design basis accidents chosen exceed those of all other accidents considered. The accident resulting in the largest air ingress, DBA-2, is a highly improbable accident. (Ref. 8) 3.2.

DOUBLE DBA-2 The accident analyses in the FSV FSAR indicate that the potential for air ingress into the PCRV is small.

However, nonmechanistic hypothetical accident scenarios beyond the design basis have been considered in light of the Chernobyl accident. The purpose of these evaluations was to determine whether any remotely plausible scenario that involves significant graphite oxidation exists.

3.2.1 Selection of Double DBA-2 Based upon these evaluations, it was determined that the largest air ingress one could plausibly postulate would result from a Double DBA-2 accident in which simultaneous failures of both independent closures in two separate PCRV penetrations occur (a quadruple failure initiating event).

Air ingress resulting from any other remotely plausible scenario would be bounded by this Double DBA-2 event.

For a Double DBA-2 to result in significant air ingress, one penetration failure must occur in the top of the PCRV, while the second failure simultaneously occurs in the bottom of the PCRV.

These failures could produce a flowpath

(" chimney") via which air could be drawn into the lower opening and convected through the steam generators and the reactor core to the reactor top plenum, where it is discharged through the upper opening.

If both openings occur in the top of the PCRV or both occur in the bottom, the necessary i

convective flowpath would not be established, and the consequences of the Double DBA-2 would be similar to those of the FSAR DBA-2.

In addition to the location of the double failures, the timing of the failures is important. The double failures must occur at essentially the same time.

The PCRV blowdown transient can be as short as 2 minutes. Once PCRV pressure has been reduced to atmospheric, a second penetration failure becomes even more incredible.

l The incredibility of this two penetration failure is confirmed by the l

NRC conclusion "...that a sudden total failure of both the inner and i

outer closures of any one of the PCRV penetrations would represent a conservative upper limit to the spectrum of hypothetical failures that might conceivably occur in the primary systein envelope."

(Ref.

8) 8

. _ _ _ _ ~

1 3.2.2.

Probability of Occurrence The median frequency of a Double DBA-2, with failures occurring simultaneously at the top and bottom of the PCRV, has not been specifically calculated but is certainly significantly less than the DBA-2 value of 1x10E-9 per year, with an uncertainty factor (P95/ Median) of less than 90 (FSAR Section 14.11.1.4). Even at the 1

j upper limit of the uncertainty band, the accident frequency is so i

small that the Double DBA-2 accident is considered to be highly improbable.

3.2.3.

Air Ingress Rate To calculate the rate of air ingress and graphite oxidation associated with a Double DBA-2, it was conservatively assumed that the upper.and lower PCRV access penetrations failed. The effective flow areas associated with these failed penetrations are the largest j

in the top and bottom head of the PCRV, 59 square inches and 87 square inches, respectively.

It was assumed that forced circulation is shut off or is not available.

It was assumed that the subsequent core heatup and cooldown would proceed as described in Section 14.10 1

of the FSAR, and that the PCRV liner cooling system would continue to e

operate.

1 The rate of air ingress would increase with time as the reactor core heats up. The results of the ingress analysis show that the maximum air ingress rate is loss than 400 scfm when maximum core temperatures j

are attained.

3.2.4.

Graphite Oxidation Rate s

Graphite will oxidize in the presence of air at elevated temperaturer.

Since graphite is porous, the overall oxidation process involves four. basic steps:

(1) transport of oxy surface, (2) diffusion of oxygen into the porous media, (3) gen to the chemical reaction, and (4) diffusion of the reaction products, carbon monoxide i

and carbon dioxide, out. At low temperatures, 600 degrees to 900 l

degrees F, steps (1) and (2) are fast relative to chemical reaction, and the oxidation occurs homogeneously throughout the material.

At high temperatures (above 1500 degrees F), the rate of reaction is fast relative to transport, and oxidation occurs primarily on the i

surface. The overall oxidation rate, therefore, is equal to the rate i

of transport of oxygen to the surface. At intermediate temperatures, simultaneous diffusion in depth and chemical reactions occur creating i

an oxidation " profile."

At high graphite temperatures, the l

predominate reaction is:

C + 1/2 02 --> CO + 52,000 Btu /lb mole of C i

Daring most of the Double DBA-2 core heatup, high graphite j

temperatures would prevail. Oxidation would occur primarily on the 9

l l

I

surfaces of the graphite and would be limited by the rate of air transport to the surfaces.

To calculate the graphite oxidation rate associated with the Double DBA-2 accident, it was assumed that all of the air entering the bottom of the PCRV is drawn upward through the core. Realistically, some air would be expected to flow around the outside of the core barrel and not come into contact with the graphite. Average initial core graphite temperatures were assumed to be about 1600 degrees F,

and the core support structures (posts, blocks, and bottom reflectors) were assumed to be initially at about 1400 degrees F.

At these temperatures, most of the incoming oxygen will be consumed by the core support structures.

As core heatup proceeds, the core support area temperature can approach 1800 degrees F, increasing the likelihood of preferential oxidation in that area.

If it is assumed that, for reasons discussed in Subsection 3.2.5, the airflow is terminated after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the following oxidation results are obtained. Core support post temperatures would reach about 1500 degrees F.

Their structural integrity would be maintained, with less than 0.33 inches being oxidized from the post radius. The core support post diameter is 6 inches, and each post bears an average compressive stress of 350 psi. Loss of 0.33. inches from the post radius results in an increase in the average compressive stress to 440 psi, well below the 8000 psi compressive strength of the material. The maximum oxidation rate in the center of the core, where temperatures would reach 4800 degrees F, would be about 0.1 weight percent (wt %) per hour (assuming that all oxidation is concentrated in the active core) or about 2.5 wt% in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Since fuel particle coating failure during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of this accident is expected to be only about 12%, the resulting release of fission products from the oxidized graphite is somewhat minimized.

If no action were taken to terminate the airflow, graphite oxidation could continue indefinitely.

The term " burning" refers to rapid oxidation in which the heat of combustion exceeds the rate of cooling by airflow or radiation, which results in a self-sustaining reaction.

For an HTGR fuel block in an unrestricted airflow, this self-sustaining phenomenon occurs only above 1800 degrees F.

At this temperature, the rate of oxidation is limited by the rate of air transport to the graphite surfaces.

If the supply of air is stopped, oxidation will cease.

In the Double DBA-2, the amount of airflow is limited to about 360 scfm.

All oxygen which contacts the graphite will be consumed. This limits the total exothermic heat generated by the graphite oxidation to less than 5% of the core decay heat 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after reactor shutdown.

Therefore, the heat of combustion is much less than the decay heat, and contributes little to the core heat up.

10

3.2.5 Mitigating Action To stop the graphite oxidation resulting from a Double DBA-2, it is 4

necessary only to stop the flow of air through the PCRV.

The most direct and simple means of doing this would be to flood the lower 3 1/2 floors of the Reactor Buf1 ding. This action would create a water level high enough to seal off the bottom access penetration of the PCRV, thereby blocking the airflow path.

The volume of water required is about 2.5 million gallons. This water could be obtained from the storage ponds located outside of the plant, which have a minimum inventory of 20 million gallons, via one or more firewater pumps.

It is estimated that the lower 31/2 floors could be flooded in about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Water ingress into the PCRV resulting from liftoff of spray as the water level approaches the lower opening would be expected to be small relative to the water ingress scenarios already evaluated in the FSAR.

Reactor Building access for this action would be necessary only to run hoses and lines into the building. Dose rates in the Reactor Building 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following DBA-2 have been estimated to be about 26 Rem / hour, a value sufficiently low to allow access for this purpose.

Temperatures in the building at I hour would be less than 100 degrees F, also sufficiently low to allow access.

After the lower 31/2 floors of the Reactor Building are flooded and the flow of air through the PCRV is stopped, the reactor would be in a permanent loss of forced circulation (LOFC) condition, with an opening in the top of the PCRV.

Continued core cooling would be 4

provided by the PCRV liner cooling system. Analyses in Appendix D of the FSAR show that liner cooling flow may be restored within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> after LOFC initiation, while the liner temperature is low enough that the integrity of the liner and cooling tubes can be assured.

Access to the Reactor Building would be required to redistribute liner cooling flow to the PCRV top head liner. A review of the procedures for this operation indicates that all valves necessary to establish redistributed Loop II flow are located above the building flood level.

If the normal systems were not available for continued core cooling, back-up cooling with firewater could be established.

Procedure reviews indicate that all valves to which access is needed to establish back-up liner cooling on firewater are also located above this Reactor Building flood level.

All local operations necessary for reserve shutdown system actuation could also be conducted at locations above the flood level.

Analysis of the Reactor Building environment during the 24 to 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> time period after accident initiation indicate that dose rates would be about 21 Rem / hour (at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) to 1 Rem / hour (at 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />), and temperatures would be less than 100 degrees F,

provided that the Reactor Building louvers are opened.

The louvers can be opened remotely from the control room.

These conditions will allow personnel access to the Reactor Building to perform the operations necessary to ensure continued core cooling.

11

, _ _ ~ _ _ _ _....., _ _ _. _ _ _ _ _, _ _., _ _ _ _.. _. _ _. _ _ _ _ _ _ _ _ _ _ _, _., _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _. _ _ _ -. _ _ _ _ _

iif,

[ ',

/

/

,3.2.6.

Dose Consequences t-Offsite radiological dose consequences resulting from a Class 9 accident scenario such as a Double DBA-2 would be much more severe than those calculated for the FSAR design basis accidents except during the initial stage. Double DBA-2 doses have been estimated under the following assumptions:

1.

Accident occurs at 105% reactor power.

2.

Offsite doses resulting from the initial depressurization are the same as those for DBA-2 in the FSAR, using FSV FSAR assumptions, d

3.

Fuel failure is consistent with GA core heatup simulation test results.

Fission product release fractions are consistent with FSAR models.

4.

Offsite doses during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after depressurization result from the core heatup source term, and oxidation of fission product bearing graphite. Airflow rate is conservatively assumed to be 400 scfm for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, even though a maximum air flow rate of about 360 scfm was calculated.

5.

Air ingress into the PCRV after Reactor Building flooding is negligible.

6.

FCRV leakage rate from 24 to 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> is 6% per day, and from 120 to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> is 2% per day, based on expansion of gases remaining in the PCRV during heatup.

7.

Meteorology, breathing rates, etc., are the same as assumed for DBA-2 in the FSAR.

8.

With the Reactor Building Ventilation System in operation, release is elevated.

Without the ventilation system operating, release is at ground level.

9.

Ventilation filter efficiency is at FSAR values.

Two hour doses at the Exclusion Area Boundary (EAB) (590 m radius) for this accident are the same as those given for DBA-2 in the FSAR:

2.5 rem whole body, 17.4 rem thyroid, and 4.8 rem bone. These doses are within 10CFR100 two-hour guidelines for the EAB.

12 i

I i

The following EA8 doses were calculated for this accident for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />'after the accident:

Whole Body Thyroid Bone With ventilation 43 rem 770 rem 7.2 rem Without ventilation 11 rem 1313 rem 8.9 rem The following 30-day offsite doses ' ere calculated for this accident w

for the Low Population Zone (LPZ) boundary (16,000 m radius):

Whole Body Thyroid Bone With ventilation 1.5 rem 32 rem 0.66 rem Without ventilation 0.3 rem 53 rem 0.78 rem 10CFR100 guidelines 25 rem 300 rem Additional dose accumulation from 30 to 180 days would be minimal relative to these values.

Thyroid doses with ventilation system operation are lower than those withcut ventilation due to the removal of iodine by the Reactor Building Ventilation System filters. Whole body doses, on the other hand, are dominated by noble gases, which are not remo'ved by the filters.

Operation of the Ventilation System forces the noble gases out of the Reactor Building at a faster rate, providing less time for radionuclide decay to reduce the offsite dose. Hence, whole body doses are higher with the Reactor Building Ventilation System operating.

It can be seen that, in spite of the extreme severity of this accident scenario, the offsite dose consequences are within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> EAB and the 30 day LPZ 10CFR100 guidelines.

13

4.

STEAM INGRESS AND WATER GAS GENERATION 4.1.

FSV LICENSING BASIS The potential consequences of steam ingress into the PCRV and subsequent steam-graphite reactions were also considered during the design and licensing of FSV. The steam-graphite reaction is highly endothermic. Thus, it is not of concern with regard to its thermal consequences. However, it is of concern due to its potential effects on the structural integrity of the core components and due to the production of hydrogen and carbon monoxide (water gas), which increases the PCRV internal pressure.

A wide range of steam generator tube rupture scenarios involving deuble-ended flow from an offset shear failure of an evaporator-economizer-superheater tube or subheader are considered in FSAR Section 14.5.

Scenarios range from normal protective action (reactor scram upon indication of high moisture in the PCRV, followed by dumping the inventory of the leaking steam generator) to scenarios involving dumping the contents of the wrong steam generator, moisture 1

detector failure, or no steam generator dump.

In addition, scenarios

]

involving multiple tube failures of up to all 54 tubes (18 j

subheaders) in a single module are evaluated.

In all cases, the peak PCRV pressure resulting from the steam-graphite reaction is less than the 812 psia setpoint of the PCRV relief valve rupture disk. Hence, no mechanism was identified for ingress of air to mix with the water gas in the PCRV, as could occur slowly if the relief valve failed to reseat properly.

Another steam ingress scenario developed during the initial licensing of FSV involved the occurrence of DBA-2 in combination with a pre-existing maximum tolerable steam leak.

Under these accident conditions, much of the water gas produced from the steam-graphite reaction would be expelled from the PCRV during the depressurization.

Following depressurization, however, a small amount of water gas, mixed with residual helium, would remain in the PCRV. The subsequent small air ingress described in Section 3.1 would result in a small amount of air-water gas mixture within the PCRV cavity extensively diluted with helium. This mixture would not be detonable.

The minimum air content for detonation of a nondiluted air-water gas mixture is about 40%.

(Ref. 9)

In summary, the consequences of steam-graphite reactions and water gas production were considered in the FSV licensing basis.

No accidents that lead to significant air ingress and a potentially detonable mixture of air and water gas within the PCRV cavity were identified, i

l 4.2, WATER GAS DETONATION Although no scenario has been identified that could result in a detonable mixture of air and water gas within the PCRV cavity, 14 i

4 analyses have nevertheless been conducted to assess PCRV structural margin with detonation.

It was arbitrarily assumed that an undiluted, stoichiometric air-water gas mixture (72% air, 28% water gas) exists within the PCRV cavity at atmospheric pressure.

This mixture is the most highly detonable mixture possible.

(Ref. 9)

Detonation of this mixture at the centerline of the PCRV lower cavity was modeled using the finite element code DYNA-30, which was developed at Lawrence Livermore Laboratory to model the effects of nuclear blasts on structures, and has been used at GA to study the effects of such blasts on missile silos. DYNA-30 is a hydrodynamic code that provides the capability of converting the increase in internal energy into potential energy by generating a steep pressure gradient or shock wave. The radial increase of internal energy due to the detonation of the assumed explosive mixture is modeled. At the inside liner of the wall, the shock wave transmits this energy to the wall structure, as modeled by the equation of motion for the wall.

The PCRV wall structure consisting of the internal liner, rebar, concrete wall, and steel prestressing tendons was modeled by an ensemble of interconnected flexible structural elements.

In generating the concrete wall model, the conservative assumption was made to ignore the initial prestressing of the PCRV. The structural element's dimensions and physical properties were based on data provided in the FSAR.,

The results of this analysis, summarized in Ref. 10, showed that the 900 psi maximum pressure at the cavity wall, resulting from a hypothetical detonation, was well below the 1775 psi ultimate failure pressure of the PCRV. The stresses induced into the concrete were lower than their tested tensile and compression strengths. The conservatively calculated tendon tensile stress was less than 1% of the 168,000 psi tendon load after prestressing (FSAR Section 5.6).

The maximum liner hoop stress is well within its linear elastic range noted in FSAR Section 5.7.

Therefore, it is concluded that the PCRV possesses sufficient structural margin to withstand detonation of a non-mechanistically assumed worst case mixture of air and water gas within the PCRV cavity.

Based on the small radial displacement of the liner (0.10 inch maximum), low liner stress (well within elastic limit), and previous studies of the ability of the PCRV to maintafr. its integrity despite thermal degradation (see FSAR Appendix D.2.3), it is PSC's and GA's engineering judgment that the PCRV liner cooling system would adequately support core decay heat removal and maintain PCRV structural integrity following a worst case air-water gas detonation.

15

r-5.

CONCLUSIONS A comparison of the RBMK reactor design and FSV indicates that there are more differences than similarities between the designs. Many of the design features that played a major role in the Chernobyl accident are not present at FSV.

The consequences of air ingress into the PCRV and graphite oxidation were onsidered during the licensing of FSV. No accidents that led to significant air ingress were identified. The accident resulting in the largest air ingress, DBA-2, is highly improbable.

One can postulate scenarios beyond the design and licensing basis that cause significant air ingress at FSV, such as a highly improbable Double DBA-2.

The consequences of a Double DBA-2 accident, involving simultaneous failure of both closures of both the top and bottom PCRV access penetrations, have been assessed.

Simple, direct means exist to terminate airflow and graphite oxidation within at most 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by flooding the Reactor Building up to the level of the lower PCRV opening. During these 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, core support graphite oxidation and oxidation in the center of the core would be minimal.

Structural integrity of the core and core support structure would be retained.

Self-sustained graphite burning would not occur, and graphite oxidation would stop upon termination of airflow.

The offsite dose consequences of this scenario are still within 10CFR100 guidelines at the LPZ boundary, under the conservative FSV dose calculation assumptions in the FSAR.

The consequences of steam-graphite reactions and the formation of water gas were also considered during the licensing of FSV.

No accidents that lead to the creation of a detonable air-water gas mixture within the PCRV cavity were identified.

However, to assess PCRV structural margin, it has been arbitrarily assumed that a worst-case stoichiometric air-water gas mixture is detonated within the PCRV cavity.

Analyses indicate that the resulting stresses are well within the capability of the PCRV, and structural integrity is retained.

Based upon these evaluations, it is concluded that an accident directly comparable to the Chernobyl accident cannot occur at FSV due to fundamental design differences. Any remotely comparable accident is highly improbable and the dose consequences are within 10CFR100 guidelines.

The design margins of the PCRV are sufficiently large that the structure can successfully withstand a hypothetical worst-case detonation of a mixture of air and water gas within the PCRV cavity. Therefore, it is concluded that the risk to public health and safety from accidents at FSV remains small.

4 a

16

r 6.

REFERENCES 1.

PSC letter dated May 9, 1986, H.L. Brey to H.N. Berkow (USNRC),

Subject:

Chernobyl Nuclear Reactor Accident and Its Implications Upon Fort St. Vrain".

(P-86358) 2.

USNRC letter dated September 16, 1986, K. L. Heitner to R.0.

Williams,

Subject:

" Report on Chernobyl Accident".

(G-86502) 3.

" Analysis of Chernobyl Event", Work Shop Report, September 16, 1986, Dulles Airport Marriott, Depart. dent of Energy.

4.

" Safety Evaluation by the Division of Reactor Licensing USAEC in the Matter of Public Service Company of Colorado, Fort St. Vrain Nuclear Generating Station, Docket Nc 50-267", June 21, 1968, pp. 73-75.

5.

"The Accident at the Chernobyl Nuclear Power Plant and Its Consequences, Information Compiled for the IAEA Experts' Meeting, 25-29 August, 1986, Vienna".

USSR State Committee on the Utilization of Atomic Energy.

6.

PSC letter dated July 24, 1979, F. Swart to Themis P. Speis,

Subject,

" Fort St.

Vrain Fuel Particle Coating Failure".

(P-79157) 7.

PSC letter dated December 10, 1985, R. F. Walker to H. N. Berkow,

Subject:

" Confirmatory Action in Support of 35 Percent Power Restriction During EQ Schedule Extension Period", Attachment No.

3, "FSV Fuel Performance Under High Temperature Conditions".

(P-85460) 8.

" Safety Evaluation by the Division of Reactor Licensing USAEC in the Matter of Public Service Company of Colorado, Fort St.

Vrain Nuclear Generating Station, Docket No. 50-267", January 20, 1972, pages 33, 35.

9.

"HTGR Accident Initiation and Progression Analysis Status Report

- Phase II, Risk Assessment", GA-A 15000, April 1978, General Atomic Company, page 4-197.

10. GA Technologies, Doc. No. 909080, October 17, 1986 " Chemical Detonation Within Fort St. Vrain PCRV".

17