ML20206B603

Forwards Comments Re Implication of Chernobyl Reactor Accident.Design Differences Between Fort St Vrain & Chernobyl Preclude Accident Similar to Chernobyl from Occurring at Fort St Vrain

ML20206B603
Person / Time
Site:	Fort Saint Vrain
Issue date:	04/01/1987
From:	Heitner K Office of Nuclear Reactor Regulation
To:	Robert Williams PUBLIC SERVICE CO. OF COLORADO
References
TAC-60400, TAC-63576, NUDOCS 8704090248
Download: ML20206B603 (7)

Text

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' April 1,;1987 A

-Docket No. 05-267: l l

Mr. R. O. Williams, Jr. .

Vice President, Nuclear Operations Public Service Company of Colorado Post Office Box 840 -

i- Denver, Colorado 80201-0840

SUBJECT:

RESPONSE OF PUBLIC SERVICE COMPANY OF COLORADO (PSC)'TO

! REPORT ON CHERNOBYL ACCIDENT e

h The staff has reviewed your submittal dated December 4,'1986 (P-86641),

transmitting an updated version of PSC's previous report, "Chernobyl Nuclear Reactor Accident =and Its Implications.Upon Fort St.-Vrain." In reviewing.

your report, the staff has drawn two major. conclusions. First, the design differences between Fort St. Vrain and Chernobyl preclude an accident -

similar to that at Chernobyl from occurring at Fort St..Vrain. Second, when-hypothetical accidents beyond the design basis of Fort St. Vrain are examined, j the plant still appears to have considerable margins of safety.

Our coments are enclosed.

i Sincerely, 4

l original signed by

! Kenneth L. Heitner, Project Manager

Standardization and Special i Projects Directorate Division of PWR Licensing B Office of Nuclear Reactor Regulation f~

Enclosure:

As stated cc w/ enclosure:

See next page

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April'1, 1987

% * . . . + .o' Docket No. 05-267-Mr. R. O. Williams, Jr.

Vice President, Nuclear Operations Public Service Company of Colorado Post Office Box 840 Denver, Colorado 80201-0840

SUBJECT:

RESPONSE OF PUBLIC SERVICE COMPANY OF COLORADO (PSC) TO REPORT ON CHERN0BYL ACCIDENT The staff has reviewed your submittal dated December 4, 1986 (P-86641),

transmitting an updated version of PSC's previous report, "Chernobyl Nuclear Reactor Accident and Its Implications Upon Fort St. Vrain." In reviewing your report, the staff has drawn two major conclusions. First, the design differences between Fort St. Vrain and Chernobyl preclude an accident similar to that at Chernobyl from occurring at Fort St. Vrain. Second, when hypothetical accidents beyond the design basis of Fort St. Vrain are examined, the plant still appears to have considerable margins of safety.

Our comments are enclosed.

Sincerely, Mi .%L Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosure:

As stated i cc w/ enclosure:

See next page

3

's Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain cc:

.Mr. D. W. Warembourg, Manager Albert J. Hazle, Director Nuclear Engineering Division Radiation Control Division Public Service Company Department of Health of Colorado 4210 East lith Avenue P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein,'14/159A Mr. R. O. Williams, Acting Manager GA Technologies, Inc. Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 92138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L..Brey, Manager Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Public Service Company of Colorado Denver, Colorado 80201' 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission Mr. R. F. Walker P. 0. Box 840 Public Service Company of Colorado Platteville, Colorado 80651 Post Office Box 840 Denver, Colorado 80201-0840 Kelley, Stansfield & 0'Donnell Public Service Company Building Connitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-D Denver, Colorado 80211 Regional Aaministrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011 Chairman, Board of County Connissioners of Weld County, Colorado Greeley, Colorade 80631 Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 i Denver, Colorado 80202-2413 j 1

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COMMENTS BY THE OFFICE OF NUCLEAR REACTOR REGULATION

, IMPLICATIONS OF CHERN0BYL. REACTOR ACCIDENT FOR FORT ST. VRAIN.

1.0 INTRODUCTION

Like the. Chernobyl RBMK-1000 reactor.- the Fort St. Vrain (FSV) High' Temperature Gas Cooled Reactor contains a large amount of graphite, which is used as a moderator. Because of this similarity. the licensee for FSV, the Public Service Company of Colorado, was recuested to evaluate the implications of the Chernobyl accident.for FSV. The final version of the licensee's report (Reference 1) was submitted by letter dated December 4, 1986 (P-86641). It has three main sections entitled,

" Design Differences, Air Ingress and. Graphite Oxidation, and Steam Ingress and Water Gas Generation". The staff's comments and conclusions

on these matters follow.

2.0 COMMENTS 2.1 Design Differences The licensee identified six major differences between the FSV reactor and the RBMK-1000 reactor which would preclude an accident similar to that at Chernobyl from occurring. These are: fuel design, primary coolant, reactor kinetics, primary system enclosures, core cooling capability, and secondary containment. Each of these differences is discussed below:

2.1.1 Fuel Design The RBMK reactor has fuel elements that are made of the same materials as those in pressurized water reactors; i.e., uranium dioxide pellets inside of zirconium allg tubes. This fuel loses its integrity at temperatures above 1300 F. The heat generated in these fuel elements is carried away by water.

l The fuel in the FSV reactor is essentially silicon carbide clad uranium carbide particles which are embedded in a graphite matrix. fhisfuel is expected to lose its integrity at temperatures above 2900 F. The hest generated in this fuel is carried away by helium. As a result of ttis difference, the licensee concluded that an accident directly cwparable to the Chernobyl accident cannot occur at FSV.

2.1.2 Primary Coolant The primary coolant in the RBMK reactor is water, whereas the primary coolant in the FSV reactor is helium. The loss of water from the RBMK reactor can cause a positive reactivity excursion to prompt critical and a concurrent loss of cooling of the metallic fuel elements. The loss of helium from the FSV reactor results in negative reactivity and a loss of cooling of the graphite fuel elements. However, the graphite fuel elements have a much higher heat capacity and melting i

temperature. Thus, in the FSV reactor the loss of helium coolant does not cause an immediate cooling problem as in the RBMK reactor, and the FSV reactor will not go prompt critical-under this situation. As a i

r cult, the licensee concluded that a Chernobyl type accident cannot i

happen at FSV.

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1 2.1.3 Reactor Kinetics-With both water and graphite in the core, the neutrons in the RBMK c reactor are overmoderated to the point where it has a positive void:

coefficient of reactivity at low power levels. This effect resulted in the reactivity excursion which-resulted in the Chernobyl accident.

The FSV core,'.during normal operation, has essentially no water in it-and is designed to be undermoderated. Thus, the loss of its helium coolant, which is also a neutron moderator, would result in a negative, not positive, change in reactivity. . Although the' addition of water into the FSV core would increase its reactivity, in order to get an' effective amount of water into the core, it has to be absorbed by the graphite, a slow process. .And, it is physically impossible to get water into the core rapidly enough to cause a significant increase in

{ reactivity.

In summary, the kinetics of the RBMK reactor, with its water / graphite

moderator, led to a positive reactivity excursion at low power. The

' FSV reactor is undermoderated and cannot. undergo a reactivity excursion of the magnitude experienced at Chernobyl.

. 2.1.4 Primary System Enclosures RBMK reactors are inside of concrete lined wells which do not have 3

pressure containing domes on top of them. In contrast, the FSV-reactor is inside a prestressed, reinforced, concrete pressure vessel l

which is capable of containing high pressures. Thus, the licensee i concluded that the release of a large amount of radioactivity from FSV i is much less likely than from RBMK, under accident conditions.

2.1.5 Core Cooling Capability In the event of an accident, the emergency core cooling system for the RBMK reactor, because of its water cooled fuel elements, must be automatically operated in seconds to prevent massive overheating, seal 4

failure, and fuel melting. In contrast, the licensee stated that the

large heat capacity of the graphite in the Fort St. Vrain reactor l precludes a rapid heatup. Emergency cooling is not needed for about

one hour and a half. Thus, significant core damage of the FSV reactor 3

is much less likely than significant core damage of the RBMK reactor, j when normal core cooling is interrupted.

2.1.6 Secondary Containment j The secondary containment for the RBMK reactor was the reactor building; which was not designed to contain pressure. In addition to the high pressure prestressed concrete reactor vessel around the FSV reactor.

independent, secondary closures are provided on all penetrations to

contain' pressure. The FSV Reactor Building provides for a filtered i release of gaseous and particulate fission products. Thus, the licensee i concluded that the release of a large amount of radioactivity from FSV j is much less-likely than from RBMK, under this type of accident condition.

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2.1.7 Conclusions on Design Differences The staff has reviewed the licensee's comparison of the Chernobyl ~

RBMK-1000 reactor and FSV. Due to the major design differences

. between these plants, most notably the undermoderation of the FSV core, the staff agrees with the licensee's conclusion that a reactivity accident similar to that which occurred at Chernobyl is precluded at FSV.

2.2 Air Ingress and Graphite Oxidation In the portion of the report on air ingress and graphite oxidation, it was stated that the consequences of air ingress and graphite _ oxidation i were considered in the FSV licensing basis, and no accidents that. led

to significant air ingress were identified. However, in light of the Chernobyl accident, a hypothetical accident beyond.the design basis has been postulated and designated as " Double DBA-2" (double the Design-Basis Accident No. 2, which was considered in the basis for licensing).

The Double DBA-2 accident assumed that both. independent closures in one penetration on the top and both independent closures in another i at the bottom of the prestressed concrete reactor vessel (PCRV) penetration failed i simultaneously. Thus, 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 would be discharged through the upper opening. Graphite oxidation occurs when this air goes through the PCRV. It was also stated that the flow of air through the-PCRV could be stopped by flooding the lower 36 floors of the Reactor Building and that completion of this flooding would take about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Within

, this time span the core support posts would be oxidized less than 0.66

[ inches from their original diameters of 6 inches, and the increased compressive stress in the posts would be insignificant. The staff believes that the structural integrity of the core support posts would

not be affected as a result of 0.66 inches oxidation. - It was further l

! stated that the offsite dose consequences were within the 2-hour  ;

i exclusion area boundary and 30-day low population zone guidelines of 10 l CFR Part 100.

2.3 Steam Ingress and Water Gas Generation l 1

, In this section of the PSC report, it was stated that the consequences i of steam-graphite reactions and water gas production were considered in the FSV Ifcensing basis. In that evaluation, no accidents that led to significant air ingress and a potentially detonable mixture of air and water gas within the PCRV cavity were identified. However,'a hypothetical detonation resulting from an arbitrarily undiluted,

, stoichimetric air-water gas mixture (72% air and 28% water gas) within the PCRV cavity at atmospheric pressure was postulated. A finite element

computer code, DYNA-3D, was used to analyze the PCRV response due to the assumed detonation.

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,1 ~ The results of this analysis indicated that the maximum pressure at the cavity wall was only 900 psi. This pressure was well below the 1775 psi ultimate failure pressure of the PCRV. The staff has concluded that the 900 psi pressure generated from the hypothetical detonation does not constitute any threat to the structural integrity of the PCRV.

3.0 CONCLUSION

S In summary, the only significant similarity between the Cherncbyl and FSV reactors is that they both contain a large amount of graphite, which is used as a moderator. There are major design differences between these reactors, as discussed above, which preclude an accident similar to the Chernobyl accident from occurring at FSV.

Furthermore, based on the staff review, it is concluded that the structural integrity of the PCRV would be maintained during and after the assumed accident scenarios. Although the initiating events are beyond the plant's original design basis, the plant design appears to have an adequate margin of safety to withstand these events.

4.0 REFERENCES

1. Letter from H. L. Brey, Public Service Comp ny of Colorado to H. N. Berkow (NRC), dated December 4, 1986 P-86641).

2. Memorandum from D. M. Crutchfield to X. Heitner on " Comments on the Response of Public Service Company of Colorado to Report on Chernobyl Accident," February 6 1987.

Principal Contributors: E. Lantz, DPWRL-B J. Ma, DPWRL-B Dated: April 1, 1987 j

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