ML093380358

Petition for Decommissioning of Unit 1 & Suspension of Units 2 & 3.Alleges That Potential Consequences of Accident Could Be Enormous Due to Failure of Util to Comply W/Basic Safety Requirements

ML093380358
Person / Time
Site:	Indian Point
Issue date:	09/13/1979
From:	Weiss E UNION OF CONCERNED SCIENTISTS, Sheldon, Harmon, Roisman & Weiss
To:
Shared Package
ML093380359	List: ML093380358 ML093380362
References
FOIA-2023-000150 NUDOCS 7910180121
Download: ML093380358 (27)

Text

1,1377 40b (ISNRCD UNITED

-STATES OF AMERICA s p BEFORE THE NUCLEAR REGULATORY COMMIISSIOq E

O.,ob"j S.CIIO 2

Section$0' vi c UNION OF CONCERNED SCIENTISTS' PETITION F A'

DECOMM4ISSIONING OF INDIAN POINT UNIT 1 A14D SUSPENSION OF OPERATION OF UNITS 2 & 3 I.

INTRODUCTION

1. This petition to the Nuclear Regulatory Commission (NRC) is brought by the Union of Concerned Scientists (UCS).

The peti tion seeks immediate action to relieve the undue risk to public health and safety posed by the Indian Point nuclear power plants.

It is brought before the Commission rather than the staff for the reasons discussed below.

2. The Indian Point nuclear power plants are located in the most densely-populatedmetropolitan area of the Eastern United States.

The site in Buchanan, New York, is located less than 30 miles north of New York City. The site was chosen in the 1950's when there were essentially no criteria governing the acceptability of sites and designs for nuclear power plants.

3. Indian Point Unit 1, a pressurized water reactor manu factured by Babcock & Wilcox, was announced in February 1955 and its construction permit application was filed in March 1955.

The construction permit was issued in May 1956. However, Unit 1 never received a full-term operating license.

It operated on the basis of a provisional operating license from Mlarch 1962 until it was ordered shut down'in October 1974.

All fuel has been unloaded and the licensee, Consolidated Edison Company of New York, claims that no decision on future operation has been made.

/

908

1 11

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4. Indian Point Units 2 and 3 were originally conceived of as twins.

They are also pressurized water reactors, but were manu factured by Westinghouse. Unit 2 was announced in November 1965, received its construction permit in October 1966 and its operating license in October 1971, but did not begin commercial operation until August 1973.

Unit 3 was announced in April 1967, received its construction permit in August 1969 and its operating license in December 1975, and began commercial operation in August 1976.

5. The NRC has never determined what the consequences would be of a so-called Class 9 accident - especially a core meltdown with breach of containment - at the Indian Point site. Conformance with NRC regulations does not guarantee that such an accident will not occur; it is an attempt only to reduce the probability of having one. However, NRC does not presently have either an estimate of the probability of a catastrophic accident or an estimate of the conse quences of such an accident at this site.

There are two separate sets of circunstances which make this particularly significant for Indian Point. First, neither Unit 2 or 3 meets current NRC regula tions. They could not receive operating licenses if their applica tions were being reviewed today. Second, the location of the Indian Point plants in metropolitan New York presents the potential for enormous consequences to the densely-settled population.

6. This petition requests action by the Nuclear Regulatory Commission in three general areas.

First, the provisional operating license for Indian Point Unit 1 should be revoked and the plant decontaminated and decommissioned.

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7. Second, operation of Indian Point Units 2 and 3 should be immediately suspended because their known safety deficiencies pre clude operation without undue risk to the health and safety of the public.

These include a number of safety problems in Unit 2 which were identified and corrected in Unit 3 during the review of that plant but were never corrected for the earlier plant at the same site.

They also include safety deficiencies common to both units.

8. Third, the Commission should determine the potential con sequences of a Class 9 accident, especially a core meltdown with breach of containment, at the Indian Point site.

The Commission should then decide whether Ithose potential consequences are so severe as to render the Indian Point site an unsuitable location for a nuclear power plant.

9. Units 2 and 3 should not resume operation unless and until there is a favorable determination on the site suitability question discussed above and unless and until:

a) Unit 2 is backfitted to incorporate all safety-related changes inco rporated in Unit 3 prior to licensing of that plant; b) the known safety deficiencies described below are remedied; c) it is demonstrated that specific design features in each plant provide a rational basis for continued operation in the face of each applicable unresolved safety problem listed by the staff in NUREG-0410, the 1978 NRC report to Congress; and d) it is demonstrated that specific design features in each plant provide a degree of protection equivalent to that which I

- 4 would be provided by conformance with each Regulatory Guide currently applicable to new pressurized water reactors.

10.

Some of the safety issues raised by this petition are not unique to the Indian Point nuclear power plants. However, the magnitude of the consequences that could result from an accident at this site are believed to be unique'. This petition will demon strate that the Indian Point plants represent a clear and present danger to public health and safety. UCS believes that it is urgent that the NRC give this petition a priority at least as high as that accorded license applications for new plants.

II.

DESCRIPTION4 OF THE PETITIONER

11.

The Union of Concerned Scientists is a non-profit, public corporation which conducts scientific and technical research con cerning advanced technologies. The organization grew out of an informal faculty group at the Massachusetts Institute of Technology in the late 1960's.

It has grown into a coalition of scientists, engineers and other professionals concerned about the health, safety, environmental and national security problems posed by this country and abroad. UCS has published many technical reports on various aspects of nuclear technology. UCS maintains professional staff in Cambridge, Massachusetts, and Washington, D. C., and has a current public membership of over 85,000 sponsors who have made financial contributions to its work. Over 10,000 of these sponsors reside within 60 miles of the Indian Point nuclear power plants.

III.

JURISDICTION

12.

This petition is brought before the Cormission pursuant 1

-5 to the authority granted to it in 42 USC §2233(d), 2236(a), 2237 and 10 CFR S§2.204, 2.206(c)(1), 50.54, 50.100 and 50.109. Furthermore, this petition invokes the inherent supervisory authority of the Com mission to oversee all aspects of the regulatory and licensing pro cess and its "overriding responsibility for assuring public health and safety in the operation of nuclear power facilities."

In the Matter of Consolidated Edison Co. of N. Y., Inc.

(Indian Point, Units 1, 2 and 3).

CLI-75-8, NRCI 7518, 173, 1975.

13.

The inherent authority of the Commission has been exer cised on a number of occasions, despite the absence of express pro cedural authorization for Commission oversight or review in the regulations. Petition for Emergency and Remedial Action, CLI-78-6, 7 NRC 400 (1978); see also, U. S. Energy Research and Development Administration (Clinch River Breeder Reactor Project), CLI-76-13, NRCI, 76/8, 67, 75-76; Consumers Power Co.

(Midland Units 1 and 2),

CLI-78-38, RAI-73-12, 1084.

This authority is necessary for the Commission to carry out its mission to see that "public safety is the first, last, and a permanent consideration in any decision on the issuance of a construction permit or license to operate a nu clear facility."

Power Reactor Development Corp. v. International Union, 367 U.S. 396,402(1961).

14.

The Commission's inherent authority is explicitly recog nized in 10 CFR §2.206(c) (1).

10 CFR §2.206(a) and §2.206(b) pro vide a mechanism for petitions requesting show cause orders to be filed with the Director of Nuclear Reactor Regulation or the Director of Inspection and Enforcement, as appropriate, and reviewed

0

-6 sua sponte by the Commission.

However, §2.206(c) (1) states:

This reviewing power does not limit in any way either the Commission's supervisory power over delegated Staff actions or the Commission' s power to consult with the Staff on a formal or informal basis regarding the institution of pro ceedings under this section.

15.

In this case, it is clearly necessary for the Commission itself to take action.

The facts relied on are part of long standing staff policies and practices to (1) allow plants to hold provisional licenses, and in som~e cases to operate on them for many years, when those plants could not meet the requirements for a full term license; (2) fail to determine the consequences of Class 9 accidents; (3) permit plants to go into operation and continue to operate despite the existence of known safety defects and unresolved safety issues; and (4)*require one plant to change its design in order to meet minimum safety requirements while ignoring similar or identical plants with the very sane defects, and in this case, even on the same site.

16.

It would be futile to refer this petition back to the staff for action because it is, regrettably, the staff's failure to take action that is directly responsible for the conditions alleged.

IV.

STATEMENT OF THE CASE A. The Potential-Consequences of a Serious Reactor Accident at Indian Point Could Be Enormous

17.

lNearly ten percent of the population of the United States lives within 60 miles of the Ind.ian Point plant.

Despite the mag nitude of the population at risk, the NRC has never determined what the consequences could be of a serious accident at this site.

18.

A nuclear power plant contains several tons of radioactive material, much of which is gaseous and, if released, could be borne

0-7-

0 away by the wind. The consequences of this kind of accident have been detailed by the Nuclear Regulatory Commission. The Reactor Safety Study, WASH-1400, also known as the "Rasmussen Report," was published by the NRC in October 1975.

19.

The NRC has since repudiated the probability estimates contained in WASH-1400. We wish to make it clear that UCS does not endorse the WASH-1400 consequences model. To the contrary, later discussion will indicate why WASH-1400 seriously underestimated the maximum potential consequences at Indian Point. However, for present purposes, we use Rasmussen's figures to yield an estimate of the pos sible consequences of a core meltdown with breach of containment.

20.

The consequences of the most serious accident analyzed in WASH-1400 were as follows:

Fatalities (from acute radiation sickness) 3,300 Fatalities (from radiation induced cancers) 45,000 Non-fatal Illnesses 285,000 Genetic Defects (in first generation born after the accident) 5,100 Property Damage

$14 billion Land Area Requiring Decontamination 3,200 square miles Area Requiring Relocation of Population 290 square miles Prepared from results in Reactor Safety Study, Tables 5-7 and 5-8, M~ain Report, pp. 84-85.

This number assumes continuing appearances of genetic defects for 30 years.

In fact, genetic defects would continue to appear for 4 to 5 generations.

8--

21.

The Reactor Safety Study also described some of the procedures that would be needed in order to decontaminate areas affected by an accident. Removal of radioactive material from hard surfaces could require replacement of roofing materials, sand blasting of walls and pavements or resurfacing of pavements. De contamination of land areas could require removal and disposal (probably burial) of vegetation and surface soil or deep plowing.

It is inconceivable that such measures would be feasible for a sig nificant portion of the metropolitan New York area.

22.

As serious as these predicted consequences are, there are a number of reasons why-the actual consequences could be far worse in the case of the Indian Point plants.

One of the most signifi cant is that the number of casualties described above assumes that a massive evacuation has taken place within hours of the accident.

The Reactor Safety Study calculations are based on the assumption that all people within five miles of the reactor could be evacuated in a few hours along with most of the people downwind for a distance of 25 miles within a 45-degree sector. This evacuation model is clearly not applicable to the Indian Point plants. In fact, a high NRC official acknowledged that the Reactor Safety Study's evacua tion model "does not, and was not intended, and does not today, reflect NRC's recommendation to State and local governments for emergency planning."

Moreover, the Reactor Safety Study itself Reactor Safety Study, Appendix VI, p. 11-19.

See, The Risks of Nuclear Power Reactors, Union of Concerned Scientists, 1977.

Ben Rusche, former Director of the NRC Office of Nuclear Reactor Regulation, Testimony before the California State Energy Re sources Conservation and Development Commission, August 23, 1976, p. 25.

00

-9 stated that for New York and other major metropolitan areas, "there is no presumption that the population...could be moved in less than 1 week."

Therefore, the consequences of a catastrophic accident at Indian Point would be significantly worse than the consequences re ported in the Reactor Safety Study.

23.

Chairman Hendrie has recently conceded that in the light of the inability to evacuate the vicinity of the plant, "special provisions" may have to be taken for Indian Point.

(See Nucleonics Week, Hay 17, 1979).

However, no specification of what these pro visions might be and when they might be implemented has been forth coming.

24.

The Commission can no longer hide behind the fiction that an accident resulting in releases of radiation to the public can never occur. To the extent that the Reactor Safety Study could ever have been relied on to support such an assertion, it can no longer be so used. On January 18, 1979, in its policy statement repudi ating WASH-1400, the NRC stated:

"The Commission does not regard as reliable the Reactor Safety Study's numerical estimates of the overall risk of reactor accident."

25.

The Commission acted none too soon in disclaiming reliance on the Reactor Safety Study. The accident at Three !iile Island less than three months later proved baseless the claim that all signifi cant accident sequences had been identified and protected against.

26.

The Indian Point reactors represent a clear and present danger to the health, safety and well being of millions of people.

Under these circumstances, it is necessary for the Commission to Reactor Safety Study, Appendix VI, p. 11-6

10 address itself to the question of whether the Indian Point site is suitable as a location for nuclear power plants.

B. The Unit 1 "Provisional" Operating License Makes a Mockery of Law and Basic Safety Requirements

27.

In 1962, Indian Point Unit 1 received a provisional opera ting license pursuant to a since-repealed regulation, 10 CFR 50.57.

This regulation stated on its face that it provided an "intermedi ate procedure" prior to issuance of a full-term operating license in a case wh~re all of the safety findings required for a full-term license could not be made. A provisional license was limited to 18 months, although "upon good cause shown," this could be extended.

28.

The Federal Register notice of February 11, 1960, accom panying the proposal of this version of 10 CFR 50.57 clarified further that the intention of the regulation was to permit tempo rary operation pending complete approval of the full-term license application because of certain "practical problems" and the need in some cases to obtain actual operating experience prior to issuing the license. -This was, after all, just the beginning of the civil ian reactor program.

But-the notice clearly reflects the under standing that the licensee will be actively pursuing its full-term license:

"Under the proposed amendment,. after the completion of construction or the conclusion of preliminary testing.under the provisional operating license or both, the applicant would move for issuance of a final operating license for the full term of years requested." (25 Fed. Reg. 1225, Feb. 11, 1960).

29.

Contrary to the clear intention of the regulations and of the Atomic Energy Act, the provisional operating license for

11 Indian Point 1 was simply routinely extended by the staff in September 1963, August 1964, October 1965, May 1967, and November 1968.

Nothing approaching "good cause" was ever shown.

Nor, during all these years, had the licensee even applied for a full-term license.

30.

Finally, by a letter dated September 22, 1969, Peter Morris, then Director of the AEC Division of Reactor Licensing informed the licensee, Consolidated Edison, of a provision in the regulations, 10 CFR §2.109, which would free them of the necessity of applying for further extensions of the provisional operating license.

All the licensee had to do was formally apply for a full-term license, and "the existing license will not be deemed to have expired until the application has been finally determined."

(A copy of Morris's letter is attached.)

Thus, with the active help of the AEC staff, Consolidated Edison had found a gaping loophole in the Commission's regulations.

31.. A decade later, Indian Point 1 still has a provisional 6perating lic ense.

The plant has not operated since 1974 because Con Ed has been unwilling to install an adequate emergency core cooling system and undertake the other modifications ordered by the Commission. Con Ed is neither actively pursuing a license for Unit 1 nor pursuing a plan for decommissioning the facility.

32.

It should be noted that the only time Congress has explicitly granted the ABC or NRC the authority to issue provi sional or temporary operation in advance of the definitive safety findings necessary for a full-term license, it did so in a very limi ted and circumscribed way.

During the Arab oil embargo, Congress

12 passed an amendment to the Atomic Energy Act, 42 Usc §2242, to last only 18 months.

This provision, which expired on October 30, 1973, allowed temporary licenses to be granted upon a series of detailed findings, including the finding that the power was "essential" for "the adequacy and reliability of the power supply..."

The section goes on:

"(c)

The hearing on the application for the final operating license... shall be concluded as promptly as poss-ible.

The Commission shall vacate the tem porary operating license if it finds that the appli cant is not prosecuting the application for a final operating-license with due diligence." 42 USC §22-42(c)

(emphasis added).

33.

Although 42 USC §2242, quoted above, does not apply di rectly to Indian Point 1, the interpretation of the Atomic Energy Act which underlies it clearly does apply.

Central to the Con gressional language-in the section is the principle that operation without a full-term license is to be strictly limited, and must be contingent on the diligent pursuing of a full-term license.

Other wise, the strict safety provisions of the Act and regulations could be circumvented and frustrated, in precisely the way they have been circumvented in the case of Indian Point 1.

34.

Indian Point 1 cannot avoid the force of this. logic simply because it received a provisional operating license not specifically authorized by any act of Co ngress.

On the contrary, by the very nature of its limited scope, 42 USC §2242 indicated the clear intention of Congress.to preclude the automatic renewal of provisional licenses for plants such as Indian Point Unit 1.

35.

Nor can the issue be avoided on the ground that Indian Point 1 is not operating.

The appropriate way to deal wit h a plant

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which has outlived its useful life is to decontaminate and decommission it.

The Commission's regulations recognize this.

(See 10 CFR §50.82).

The irradiated facility cannot simply be permitted to remain in regulatory limbo. The Commission should revoke the provisional operating license for Indian Point 1 and order Con Ed to present a plan for decontaminating and decommis sioning the plant.

C. Safety Deficiencies Identified During the Review of Unit 3 Were Never Corrected for Unit 2

36.

As noted'above, Indian Point Units 2 and 3 were origi nally conceived of as twins. However, the designs of Units 2 and 3 differ in ways that have a significant effect on the risk to public health and safety qreated by operation of each unit. Some of the design changes that were made to Unit 3 appear to have been made voluntarily by Consolidated Edison or its vendors.

Others were ordered by the staff during its review of the operating license application for Unit 3. The basis for each design change ordered by the staff w a s a determination that, absent the changes, opera tion of Unit 3 would pose.undue risk to public health and safety.

37.

There were no changes to the regulations in 10 CFR be tween issuance of the operating licenses October 1971 and December 1975 that could account for the staff's determinations that features of the Unit 2 design were unacceptable for Unit 3. One possible explanation is that the staff pursued enforcement of the Commission's regulations more vigorously on Unit 3 than was its custom four years earlier during the operating license review for Unit 2, Whatever the reasons, the fact remains that in its review of Indian Point

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Unit 3, the staff determined that the original proposal was unacceptable.

38.

At a minimum, Indian Point Unit 2 requires immediate backfitting to incorporate the changes made to Unit 3 as a result of the staff review of that Unit. In the language of 10 CFR 50.109, such modification will provide substantial, additional protection which is required for the public health and safety. There is no rational basis for this disparate treatment of two plants at the same site. Nor is there justification for further delay; the NRC knows what has to be done now. Indian Point Unit 2 should be ordered to cease operation pending the required modifications.

39.

Since not all of the design changes made to Unit 3 were the result of staff orders the Commission should, in addition, assess whether other design changes made voluntarily to Unit 3 should also be backfit on Unit 2. The staff should identify for the Commission all the safety-related design differences between Units 2 and 3, and, for those which were not ordered by the staff, discuss whether the Commission should require Unit 2 to be backfit with those design features.

40.

The following three examples of safety-related design differences between Indian Points 2 and 3 do not represent a com plete list of such differences.

Diesel Generator Buildings

41.

General Design Criterion 17 of 10 CFR Part 50, Appendix A, requires an onsite emergency power system which meets the single failure criterion. The purpose of the requirement is to provide a backup source of electric power to. the safety systems which must

15 operate to prevent a meltdown. The backup electrical source is needed in the case of a loss of offsite power which is a fairly common occurrence.

42.

At Unit 3, the three emergency diesel generators are housed in a reinforced concrete building designed to withstand earthquakes and tornado missiles.

In addition, each diesel genera tor is housed in its own separate concrete room.

This design is intended to provide protection against earthquakes, external mis siles, internal explosions and fires.

This is an example of an aspect of the design of Indian Point Unit 3 that apparently was not the result of specific order by the staff.

43.

By way of contrast, the diesel generators for Unit 2, at the very same site, are housed in a sheet-metal structure which does not meet seismic criteria and could not withstand the missiles gen erated by the design basis tornado.

Furthermore, the Unit 2 diesels are housed in a common room without adequate separation between the emergency generators so that such accidents as a crankcase explosion 6r fire could damage redundant generators.

44.

There can be little question that the diesel generators in Unit 2 do not meet GDC 17.

The vital onsite power supply for Unit 2 is vulnerable to disabling damage.

This condition poses a threat to public health and safety.

Battery System

45.

In order to provide an acceptable degree of independence for redundant safety power supplies, the staff prohibits any auto matic transfer switching between redundant safety systems.

(See

S 0

16 Regulatory Guide 1.6).

The review of Indian Point Unit 3 identified such unacceptable interconnections and the staff accordingly required Unit 3 to eliminate the automatic switching between redundant d-c power supplies.

The Safety Evaluation Report for Unit 3 states:

"We concluded that such a design could unduly compromise the independence of redundant safety systems."

Therefore, a third battery was provided at Unit 3 to allow elimina tion of automatic switching between redundant batteries.

46.

Despite this, no effort was made to correct the same deficiency at Unit 2. Unit 2 has only two batteries to supply con trol power for the three diesel generators and the three sets of safety equipment; two sets of safety equipment must function to cope with reactor accidents. Unit 2 is susceptible to a failure that could lead to a meltdown accident.

Auxiliary Feedwater System

47.

By the time Indian Point 3 was licensed, the staff was beginning to recognize that the auxiliary feedwater system is extremely important to safety and it was classified as an engi neered safety feature, in the same category as the emergency core cooling system. Auxiliary feedwater is the only way tQ remove decay heat from the reactor during the initial phase of cooling following either a normal shutdown or an accident other than a large loss of coolant accident.

48.

At the time Unit 2 was licensed for operation, auxiliary feedwater was not even being reviewed by the staff. The Safety Evaluation Report for Unit 2 contains no discussion of this system.

0 17 By way of contrast, Unit 3 was required to meet the single failure criterion and modifications had to be made to ensure that a break in the steam pipe to the turbine-driven auxiliary feedwater pump would not result in total loss of auxiliary feedwater due to failure of the two redundant motor-driven pumps which are located in the same room as the turbine-driven pump. The failure to require the auxiliary feedwater system in Unit 2 to meet criteria applicable to safety systems poses a threat to public health and safety.

D. Safety Deficiencies Common to Both Units 2 and 3 Must Be Corrected Fire Could Render Redundant Safety Systems Inoperable

49.

Electrical control, instrumentation, and power systems are a basic element of nuclear plant design. The thousands of electrical cables running through the plant are the central nervous system that controls the operation of all equipment, including the safety systems which must operate to mitigate the consequences of accidents.

50.

Fire is a clear threat to these cables, as the accident at Browns Ferry graphically demonstrated. A fire which damages cables can render safety systems inoperable. Therefore, the electrical cables in a nuclear plant must be designed, installed, and protected so that a single fire cannot destroy the cables con trolling redundant safety systems, wiping out all primary and backup systems at once.

This is required by General Design Cri terion 3 of 10 CFR Part 50 and 10.CFR 50.55a(h).

In short, safety systems are only as reliable as the electrical systems which control them*.

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51.

Following the fire at the Browns Ferry plant in March 1975, NRC undertook a re-evaluation of fire protection provisions in nuclear plants.

The NRC staff has required improvements in such areas as administrative controls and control of ignition sources.

However, tests conducted as part of the post-Browns Ferry-Fire Protection-Research Program have disclosed that the five-foot physical separation requirement of Regulatory Guide 1.75 is in adequate to prevent the spread of fire from one set of cables to the other.

In addition, tests on mineral wool blankets proposed for fire retardants have shown them to act as wicks in some cases, sprinkler systems have failed, and at least some "fire-retardant" cable coatings have burned.

52.

Four and a half years after the Browns Ferry accident, the NRC still permits plants to operate which it knows remain vulnerable to a destructive fire.

Indian Point Units 2 and 3 are among these.

53.

The NRC staff has concluded that fire protection at both units is inadequate.

At Unit 3, the staff concludes that modifica tion of the fire protection systems may be sufficient to preclude fire damage to redundant safety systems.

However, for Unit 2, changes to the safety systems themselves are needed to assure that a fire will not lead to a meltdown accident.

The staff has deter mined that, these changes involve the installation of an alternate shutdown cooling method, which is required "because of a few specific plant locations where the staff does not have reasonable assurance that postulated fire will not damage both redundant div isions of shutdown (cooling) systems."

(SECY-79-112, page 11).

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54.

In summary, a fire which the NRC concedes to be possible could disable the redundant safety systems for both Indian Point Units. This constitutes a present threat to public health and safety.

Serious Unresolved Safety Problems

-Exist at Both Units 2 and 3

55.

For many years, the NRC staff followed the practice of categorizing its most serious unresolved safety problems as ligeneric" and, having done so, simply ignoring them in the context of the proceedings to license individual power plants. This practice was finally unequivocally rejected by the NRC Appeal Board in two cases. Gulf States Utilities Co. (River Bend Station, Units 1 and 2),-ALAB-444, 6 NRC.760 (1977) (Construction permit); Virginia Electric and Power Co. (North Anna Nuclear Power Station, Units 1 and 2) ALAB-491, 8 NRC 245 (1978) (Operating license).

56.

Now, before a plant is permitted to begin operation, the NRC staff is required to identify all unresolved generic safety problems which apply to the plant, and to show either how they have been satisfactorily resolved on a plant-specific basis or, if they have not, to provide the specific justification for permitting the plant to operate. Despite this rigorous test for new operating licenses, the NRC has failed to formally face up to the existence of numerous of these safety problems in currently operating plants.

57.

In its report to Congress (NUREG-0410, January 1, 1978),

the NRC identified 133 unresolved safety issues affecting reactor safety or the licensing process. The staff subsequently

0 20 identified more than half of the 133 unresolved safety issues as directly applicable to the type of nuclear plants used at Indian Point Units 2 and 3.

58.

No evaluation has ever been performed specifically for the designs of Indian Point Units 2 and 3 to demonstrate why opera tion should be permitted in the face of such serious unresolved safety issues. As noted above, if the plants were not yet licensed, precisely such an evaluation would be required before licenses could be issued.

59.

In addition, in the years since the applications for the Indian Point Units were filed, the staff's knowledge of nuclear plant designs improved and many unacceptable designs and operating practices were discovered. Operating experience also provided information indicating the need for design changes and operating restrictions. As a result, the staff developed and. promulgated many technical positions and Regulatory Guides. However, the staff's general practice was to apply these new requirements only to plants that had not received construction permits or, in a few instances, to plants that-did not hold an operating license. Again, there has been no systematic evaluation of the need to upgrade Indian Point to account for important safety lessons learned.

60.

The following examples are selected from the list of NRC-acknowledged unresolved safety problems. All apply to Indian Point Units 2 and 3. They are offered not as an exhaustive list See Appendix A to "Testimony of Michael B. Aycock, Lawrence P.

Crocker and Cecil 0. Thomas, Jr., relating to the Status of NRC Staff Activities Regarding Generic Safety Issues, September 27, 1978 submitted to ASLB in Dockets 50-556 and 50-557.

21 but only to illustrate the seriousness of the problems involved.

a) Post-Accident Monitoring Indian Point Units 2 and 3 do not have adequate, reliable instrumentation to monitor variables and systems affecting the integrity of the reactor core, the pressure boundary or the containment after an accident.

The accident at Three Mile Island (TMl) demonstrated graphically the inadequacy of post-accident monitoring, in terms of the parameters monitored, the range and accuracy of the instru mnentation, and the ability of the instrumentation to-survive the accident and post-accident environment.

For example, there is no way to directly measure the water level or temperature in the core after an accident.

The only temperature measurements at T1I were from non-safety grade equipment, some of which "luckily" survived the accident.

The accident demonstrated that without adequate reliable instrumentation, reactor operators cannot be expected to take proper corrective action in the plant or to give timely notice of the need to activate offsite emergency procedures.

b) Aging of Equipment Structures, systems and components important to safety must be qualified to demonstrate their ability to withstand natural forces such as earthquakes and the accident environment and still perform their safety functions.

In analyzing the ability of equipment to survive, insufficient account was taken of the effect of aging,,which is known to progressively weaken components.

Brand new equipment may have been tested, but no systematic effort was made to determine for how long the results would be valid.

Thus,

22 although Units 2 and 3 were essentially certified to be safe for their lifetimes when they received operating licenses, the Coininis sion cannot say with reasonable assurance that sufficient margin exists to maintain equipment qualification for several decades.

c) Asymmetric Loads on Reactor The' designer of Indian Point Units 2 and 3 did not adequately account for the effect of asymmetric loading resulting from a pipe break in the area between the reactor vessel anid the shield wall.

A pipe break in certain locations between the vessel and the shield wall would cause instantaneous extreme pressure differentials, causing forces which could tip the vessel, shearing the pipes and preventing cooling.

In addition, these forces could damage the fuel spacer grids and distort the fuel geometry.

The end result could be that all emergency core cooling systems would be rendered incapable of preventing core meltdown.

V.

RELIEF REQUESTED

61.

The NRC's obligation to ensure the safety of the public does not stop when a license is issued.

To the contrary, the United States Supreme Court has held, and UCS fully agrees, that "public safety is first, last and a permanent consideration in any decision on the issuance of a construction permit or a license to operate a nuclear facility."

Power Reactor Development Corp. v.

International Union of Electrical Radio and Machine Workers, 367 U.S. 396, 402, 81 S.Ct. 1529, 1532 (1961).

62.

Moreover, in the Power Reactor case, supra, the Supreme Court emphasized that, even after operation of a reactor is licensed,

23 the Commnission will retain jurisdiction "to ensure that the highest safety standards are maintained."

(367 U.S. 402, 81 S.Ct. 1532).

Indeed, it is precisely this assurance of continued vigilance after licensing on the part of the Commission, combined with the fact that permittees proceed at their own risk, which is the alleged justifi cation for issuing permits and licenses pending final resolution of outstanding safety issues.

If, after licensing, a grave safety problem is disclosed, the explicit promise of the Commission to continually assure the safety of operating reactors cannot be avoided.

63.

The facts outlined above demonstrate that despite their relative youth, Indian Point Units 2 and 3 are relics of the past.

They were licensed when less was-known about safety problems and when regulatory requirements were much less strict than today.

This is seriously compounded by the fact that it is highly unlikely that the site would be approved today because of the proximity of ex tremely large numbers of people. -Despite this, the NRC has marched tesolutely "eyes front", not applying the lessons learned about safety to Indian Point.

64.

We have shown, in addition, that the problem is far from an abstract or theoretical one.

To the contrary, the concrete examples given provide clear evidence that Indian Point presents a serious threat to public health and safety.

65.

Therefore, the following relief is requested:

a) The provisional operating license for Unit 1 should be immediately revoked.

-24 b) Consolidated Edison should be ordered to submit a plan within.90 days for decontaminating and decommissioning Unit 1.

c) The Commission should order operation suspended at Units 2 and 3. These units should not be permitted to resume opera tion unless and until the Commission determines that 1) the site is suitable for nuclear power generation; 2) each applicable unresolved safety problem is addressed, and 3) the requirements of each Regula tory Guide are addressed.

d) In order to make these determinations, the Commission should establish a special Atomic Safety and Licensing Board to compile a record after adjudication hearings addressing the following questions:

1) What would the-consequences be of a Class 9 accident at Indian Point?

2) What specific offsite emergency procedures could feasibly be taken to protect the public in the event of such an accident and to what extent would-these measures mitigate the con

§equences of a Class 9 accident?

3) With respect to each applicable unresolved safety problem in NUREG-0410, what are the specific design features of Units 2 and 3 which compensate for the current absence of a solu tion to that problem and what is the current status of the generic study of the problem?

A) With respect to each Regulatory Guide applicable to pressurized water reactors, what are the specific design features which constitute conformance or provide an equivalent level of protection?

-25

5) What are the safety-related design differences between Units 2 and 3, distinguishing between those changes ordered by the staff and those made voluntarily?

e) Based upon the record compiled by the Atomic Safety and Licensing Board, the Commission should then decide whether the Indian Point site is suitable and, if so, which specific added safety features and off-site emergency measures are necessary to protect public health and safety.

These should be implemented before operation is permitted to resume.

In addition, resumption of operation should in no case be permitted until:

1) all design changes ordered by the staff to Unit 3 are backf it to Unit 2;

2) the Unit 2 diesel generators are housed in separate rooms in a building which can withstand earthquakes, missiles, explosions, and fires;

3) there is an acceptable degree of independence for redundant safety power supplies by the addition of a battery At Unit 2;

4) the auxiliary feedwater system for Unit 2 has been reviewed to determine its conformance with the requirements for a safety system and all necessary changes are made; and

5) the measures which the staff concedes are neces sary to provide adequate protection in the event of a fire are implemented for both units.

VI.

C0ONSIDERATION OF FACTORS OTHER THAN HEALTH AND SAFETY

66.

UCS recognizes that past commissions, when faced with the-discovery of previously undisclosed safety problems, have

26 balked at the prospect of shutting down operating reactors to correct those problems immediately. In doing so, the Commission and its staff appeared on occasion to accept the proposition that its mandate to protect public health and safety could be balanced against certain extrinsic economic factors.

UCS believes that balancing of purely economic factors against public safety is out side of the jurisdiction of NRC and would compromise its mandate.

67.

However, if the Commission determines that such matters as potential power supply deficiencies are legally relevant and can provide a reason for permitting operation in the face of the safety problems discussed in this petition, it must require the affected utility to provide evidence constituting a definitive showing on each of the following criteria:

a) that the utility is using all alternative sources of power available to it, including purchase power and deferral of routine maintenance shutdown of other capacity on its system; b) that the utility is using all means available to cut load, including load shedding techniques; c) that the risk to health and safety from loss-of-load is greater than the risk to public health and safety from a major nuclear accident; and d) that loss-of-load after all compensating measures have been adopted would, in fact, create health and safety problems of significant importance.

68.

In addition, if the affected utility reets all of the above-listed criteria, operation of the reactor in question should.

only be permitted during those periods of peak demand.

27 VII. REQUEST FOR DISQUALIFICATION

69.

Chairman Hendrie served as Deputy Director for the Division of Technical Review from 1972-1974.

In that position, he was at least partially responsible for the staff policy and prac tices which form the basis of this petition. In such circumstances, it would not seem appropriate for the Chairman to rule on the ques tion's raised herein.

By the Union of Concerned Scientists By their Attorney, Elly Weiss Sheldon, Harmon, Roisman & Weiss 1725 I Street, N.W., Suite 506 Washington, D. C. 20006 Telephone:

(202) 833-9070