ML19305A492

Rept on Reactor Accident at TMI, Submitted by Danish Delegation

ML19305A492
Person / Time
Site:	Crane
Issue date:	07/16/1979
From:	ENVIRONMENTAL PROTECTION AGENCY
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592, NUDOCS 7908220161
Download: ML19305A492 (35)

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{{#Wiki_filter:- o -,,.., 9, pod NRC TRANSLATION 592 i REPORT ON THE REACTOR ACCIDENT AT DE AMERICAN "DIREE MILE ISLAND" NUCLEAR POWER PLANT NEAR HARRISBURG, PENNSYLVANIA M 1(g Submitted by the Danish delegation sent under the direction of the National Environmental Protection Agency (NEPA] with representatives r.f the NEPA, the Nuclear Plant In pectorate and the Riso Research Facility. 3 May 1979 l 1 1 850 035 7908220161 1

TABLE OF CONTENTS Pag 2 1. In troducti on ---- - ----- - ----------- - ----- ------------------ - 1 2. Data on the NPP----------------------------------------------- 3 3. Technical Description of the Course of the Accident--------- 6 3.1. Course o f the Event--------------------------------------- 6 3.2. Plant Condition After the Accident-------------------------- 12 4. Heal th Physics Conditions------------------------------------ 13 4.1. Overview of Dose Rates and Doses---------------------- - ---- 13 4.2. Types of Isotopes Released and Contamination----------------- 14 4.3. Dose Burdens and Consequences----------------------------- 14 5. Readiness----------------------------------------------------- 16 ~ 5.1. Em'e1 gen ck P lann'ind ----------------------------------------- 16 5.2. Local F.me rgency Planning------------------------------------ 17 5.3. Readiness to Take Measurements ---------------------------- 19 5.4. Sequence of Events in an Emerg'ency ------------------------ 19 5.5. Discussion-------------------------------------------------- 21 6. The American AEtho2itiesIHa5dliid of th'e Three Mile' Island ~~ ~ ~ Accident------------------------------------------------------ 22 6.1. The Consequences of the Accident for Other Light Water Reactors, Especially B.6 W Plants--------------------- - --- 24 F i gure s 1 - -- -------- - ----- - --- --- --- -- --- --- --- -- - ---- ----- 27-32 h. (\\ D S ogh '.t 850 037 ui l BAlHM 4,.

9 4 l l Foreword After its departure for the USA for the period April 17-April 25, 1979 in connection with the reactor accident at the "Three Mile Island" NPP, the Danish delegation issued the following report to the NEPA on 3 May 1979, to be forwarded to the Minister of Environmental 1 Protection. Through contact with the American authorities and by visiting the site of the reactor accident, the delegation was charged with obtaining information on the event and the American authoritie_s' prel_imi_ nary _ _ evaluation of it. ,It should be pointed out that the American authorities emphasited that l their studies and examinations concerning the reactor accident are still by no means conclusive and that their evaluation and conclusions i therefore are only preliminary in character. l With the exception of what is stated concerning the NRC's actions on April 27, the collection of data to use for the present report was completed when the delegation departed from Washington on Wednesday, April 25, 2100 local time. H. Hagen - H.C. Mortensen - Aksel Olsen (head) l Torben Petersen - H.P. Ryder - P.B. Suhr i l 2 e st9+%0-850 036 l l [

~ 1. Introduction At the talks in Stockholm on April 9, 1979 bctween representatives of the Swedash Department of Industry and Department of Agriculture and the Danish Ministry of Environmental Protection and the Ministry of Trade, it was agreed that health, environmerital and safety risks attendant on various forms of energy production should be researched jointly. As an initial step, representatives for the authorities of the two countries were to travel to the USA in order, by making , contact with the American authorities and visiting the site of the reactor accident, to collect data on the event as well as obtain the preliminary evaluations of the American authorities on it. A report was to be written on this trip as soon as possible. The report deadline was later set by the two governments at May 3, 1979. Accordingly, the Danish and Swedish delegations visited the USA from April 17 to April 26, 1979. The Danish delegation consisted of two representatives of, respectively, the NERA, the Nuclear Plant Inspectorate and the Riso Research Facility under the direction'of a representative of the NEPA. The Swedish delegation consisted of representatives of the National Nuclear Power Inspectorate and the Nuclear Radiologic Protection Institute. The study was organi:ed after coordination between the two delegations and with the aid of the ambassadors of the two countries in Washington. The delegations jointly collected data on the reactor accident and the preliminary views of the American authorities on it, including - the technical conditions surrounding the course of the accident; - measures taken at the plant to limit the accident; - the eherg"en~c'y conditioniahd measures [taken] ot5ide of the plant ~ ~ ~ ~ ' ' ' in order to protect the population in the vicinity. The delegations held talks with representatives of the federal authorities the Nuclear Regulatory _ Commission (NRC) and Environmental Protecti.on Agency (EPA) as well as representatives of the Emergency Agency of the State of Pennsylvania, the Pennsylvania Emergency Management Agency (PEMA). NRC is the Federal authority which is responsible for the licensing of nuclear power plants and for monitoring to ensure that they operate in accordance with the licensing provisions and prevailing rules with regard to safety and environmental protection; NRC is independent of the Department of Energy (DOE), the Energy Ministry, which handles the research and economic aspects. NRC is also the primary agency gp,, @" 1 sso w &L9 4 2 )(

~ responsible for (issuing] guidelines for the planning of readiness (stand-by] measures to protect the population in case of an accident at a nuclear facility. EPA is the federal authority in the area environmental protection. With regard to readiness in case of an accident at the nuclear facility, 'dA is also responsible for setting guidelines for implementing measures to protect the population on the basis of expected radiation doses. In addition, EPA advises on appropriate measures to protect against ionizing radiation in case of an accident and aides the agencies responsible..for readiness in formulatihg readiness plans. The PEMA is the agency directly under the governor which is the responsible coordination agency for planning and execution of the general. disaster _ emergency (plans] of the State.of Pennsylvania..., _ _ In this capacity the PEMA affords support and advice in emergency planning and in cases of disasters to the state offices (counties) and comunities (municipalities). The delegations made visits as follows': - Wednesday, April 18, the delegations received an introductory 1 i l briefing at the NRC; then the delegations held a conference among themselves to coordinate the work. l - Tuesday, April 19, the delegations together with representatives for a number of other countries.at a whole-day meeting at NRC obtained more detailed data and preliminary evaluations from a number of leading researchers of the NRC who were all directly involved in NRC's work on the accident situation. - Friday, April 20, the delegations paid a visit to Middletown and the "Three Mile Island" NPP. In this process they received a briefing from a leading representative for the group of NRC experts who were continuing to consider (means of] at+empting to bring the plant down to safe state, as well as a short briefing from representatives of the PEMA. - Monday, April 23, the two delegations attended a public hearing which the Sub-Committee for Nuclear Legislation, under the Senate Connaittee for the Environment and Public Facilities headed by Senator Gary Hart, held concerning the reactor accident. Here = ~ e -~ tes't'imony wa's' EiFir~d from theTGoversoF oV P'~nnsy1Fania andTuanairsment representatives of the companies which owned the "Three Mile Island" l NPP, as well as representatives for both organi:ations for the nuclear industry and those for consumer and environmental interests. - Tuesday, April 24, the delegations held discussions with representatives of the NRC on the technical problems, and with representatives of the EPA on the problems related to protecting the population against i i 2 g o@D Sti-+_a ~ 850 039 l

~ 1 ionizing radiation in connection with the reactor accident. 'Ihen the delegations held a joint coordination meeting. - Wednesday, April 25, the delegations first met in Harrisburg with the leadership of the PEMA, who gave a more detailed briefing on the emergency measures [taken] in connection with the_ reactor accident. Then the delegations met at "Three Mile Island" with a representative of the Metropolitan Edison Utility Company, which operates the NPP. In consideration of the work being done' at the plant, it was not possible to visit the control room of Reactor Unit 2 which suffered the accident. During meetings with the American authorities, the delegation s were received everywhere with a great deal of kindness and courtesf. It should be pointed out that the American authorities emphasized that their studies and analyses in connection with the reactor accident were by no means complete and that the evaluations and conclusions of the authorities were thus only of a preliminary nature. A final statement by the NRC inspection division on the course of the accident at the plant can be expected by August 1979. This report will cover the conditions at the reactor plant up to the o night of Wednesday, March 28 and the conditions surrounding the leakage until the afternoon of Friday, March 30. In addition, President Carter organized a commission which at the latest by October / November 1979 is supposed _to issue a report on the reactor accident covering the activities of the authorities before, during and after the accident. 2. Data on the NPP The data given below are taken from the final safety report for the Three Mile Island Nuclear Station, Unit 2 (DOCKET-50320-80,1974-04-04). It consists of two units, both with pressure water reactors manufactured by Babcock $Wilcox (B4W): 2 Unit 1: 2535 MW (thermal), 819 MW (electrical) Unit 2: 2772 MW (thennal), 906 MW (electrical) Prior to the accident, it was possible to operate the reactor at full power for a short ti:ne since it went into commercial operation on l December 28, 1978. In early February, the production of elec y l 3 850 040

was 1,544,515 MWh, corresponding to approximately 60-70 full-power days. Ownership The plant is owned by the Metropolitan Edison Company, the Jersey Central Power and Light Company and the Pennsylvania Electric Company. These three companies all belong to the General Public Utilities Corporation, which overall has a production capacity (as of 1972) of 5,900 Mio. Metr.golitan Edison Company is responsible for the design, construction, operation and maintenance of the plant. Safety Certifications Unit 1 Unit 2 Building permit 18 May 1968 4 November 1970 Operating permit 19 April 1974 8 February 1978 Commercial operation October 1974 23 December 1978 Area The plant is located in Pennsylvania and is installed on Three Mile Island in the Susquehanna River (Figures 1 and 2). The actual area of the plant occupies 81 hectares of the island's total area of approximately 190 hectares. The entire island as well as another 130 h'icitares on 'nei~gfibiring~ ~ ~ islands and the areas closest to the river are the property of the ~ plant owner. A site sketch is shown in Figures 3 and 4 From this we see'that the plant is primarily cooled with the aid of cooling towers. General Data on Pressure Water Reactors The pressure water reactor is presently the most common type of reactor in the West. In the USA, besides B6W there are two other manufacturers, ~, Westinghouse and Combustion Engineering. At the present time there are i 9 B6W reactors in operation in the USA, and 22 B6W reactors are under construction or on order. In Europe the firm Babcock Brown Boveri GmbH is building a plant of this type in West Germany. 4 9 2 850 041 l ~

In all 93 pressure water reactors are in operation in, among others, the USA, Sweden, Germany, France, the USSR and Japan. In Europe this type of reactor is also built by Kraftwerk Union (Germany) or Framatome (France). These builders utili:e the same main principles but differ with regard to technical implementation. Construction and Operation of the B8W Plant . In pressure water reactors heat is transferred from the reactor to the steam turbine through two water circulation loops, a " primary" loop and a " secondary" loop. Figure 5 shows the dimensions in the primary system. Figure 6 shows a simplified diagram of the plant. D e normal operating pressure in the primary loop is so high (157 atm overpressure) that; the. water can_be, heated, to_approximately 345*C without _ boiling. The normal water temperature at the outlet of the reactor core is approximately 320*C. The primary loop water (a total of 340 m ) transports the heat'which 3 is generated by the reactor fuel to two separate steam generators; from here the water is pumped back to the reactor tank by the main ~ circulating pumps (two pumps at 6 MW per steam generator). The pressure in the primary loop is regulated by means of a pressurizer. The regulation is carried out by means of electrical heating elements and a sprinkler cooling system.. At the top of the pressuri:er there is a pressure relief valve and, connected in series with it is a remote-controlled isolation valve. In addition, there are two safety valves which are designed to protect the reactor against overpressure. In the given case these valves direct water from the primary loop downwards to a condensation tank at the bottom of the reactor contain-ment. D e cerlensation tank is equipped with three safety valves and a rupture disk which bursts when there is an overpressure. Water from the tank can therefore be_ led _out.into the reactor containment either via the safety valves or a. burst rupture disk. In the steam generators (secondary side) the steam which drives the turbines is produced. The steam is condensed in the condensator and returned partly by a condensate pump and partly by two so-called feedwater pumps to the steam generators. During this process the water passes through a clarification plant (deminerali:er]. B8W uses a steam generator, the design of which differs from those of the other manufacturers in that the B6W steam generator contains considerably less water on the secondary side. In order to ensure that feedwater is supplied to the steam generator secondary side in case the feedwater pumps fail, there are three auxiliary feedwater pumps, one turbine drive and two electric drives (Translator's Note: word " drive" is uncertain}. In addition, these pumps and steam generators have isolation valves, compare Figure 6. 0Qhk l .,P 850 042 + y w--w--r -ww-,

"Ihe reactor is equipped with various safety systems which in case of a failure or accident are supposed to bring the reactor to a safe shutdown condition. During shutdown (scram] control rods are inserted into the reactor core; in doing so the chain reaction is stopped. In doing so the generation of heat is reduced rapidly to approximately 5% and is then reduced in such a way,that the output after six hours is less than 1%. To a certain extent cooling is required subsequently for a fairly long time. To cool the reactor core in case the normal cooling system fails, there are three types of emergency cooling systems: a high pressure system and low pressure system, both based on pumps, as well as a low pressure system which is based on a pressurized tank. The plant's primary system is mounted in a cylindrical air-tight tank of pre-stressed concrete which is constructed to [ withstand] an over-pressure of 4.3 atm. The containment is equipped with various cooling and purification systems, including equipment to remove hydrogen (a recombination device). In case of an accident, the valves in all passages to the containment wall are supposed to remain blocked unless they are part of the safety system. This isolation is carried out automatically when there is elevated pressure (0.3 atm) in the building. i j In an auxiliary building outside of the reactor containment there is a tank to store radioactive water and gas from the reactor containment. This building is not air-tight, but is equipped with a filter ventilation system which in particular is designed to hold up particulate radio- _ctive substances and iodine. 3. Technical Descrintion of the Course of the Accident 3.1. Course of the Event The sequence of events described below is based on a detailed list of events compiled by the NRC and supplementary data collected during the visit. The times given are in local time. There is a t.ine difference ~of six hours between Denmark and Harrisburg. l The initiating event occurred at 0400 on March 28, 1979 while the l reactor was operating at 98% of full output. The regional NRC office in Philadelphia was notified at 0745 hours, C==I and NRC's alert center in Washington (Incident Response Center) was Q alerted at 0810 hours. g c==2 1. The feedwater supply to the steam generators (secondary side) was Q interrupted and the condensate pump shut down. The cause of this c=u is unknown. Immediately thereafter the feedwater pumps shut down. With the failure of the feedwater supply, the turbine is supposed g 6 g Ccs) Cco] @b 850 043

automatically ; hut off and the steam from the steam generators is sent directly into the condensator; this was done. 2. Due to'the inadequate supply of feedwater to-the steam generators, the pressure in the reactor primary cooling system began to rise because the heat was no longer being removed on the appropriate scale. This pressure rise activated. the safety shutdown [ scram system], and the reactor shut down, as it was supposed to, with the automatic insertion of the control rods. Imediately prior to this, the pressuri:erW11ef valve _ opened;__this caused the [ pressure in the primary system to begin to drop. Steam and'tihen later a mixture of water and steam flowed through the relief valve towards the condensate tank at the bottom of the reactor cordaiident'. ~ 3. Within 30-40 seconds of the failure of the normal feedwater system, auxiliary pumps were supposed to start automatically and establish rr. auxiliary water and feedwater supply. The pumps started up properly, but then two valves, in violation of the safety rules and perhaps in connection with an earlier routine inspection of the auxiliary feedwater pumps, (two words inserted, illegible], the emergency feedwater supply was not established. 4. The pressure, drop,_in the primary system was supposed to orevent the relief. valve from closing _ automatically..For unknown reasons. however, the valves remained open and the pressure in the primary system continued to drop. It has been established that the operators were apparently not aware that the valves were still open. This was due to the fact that there was no direct indication in the control room of the valve positions - the instruments only show whether control current is being sent to the valves. After several minutes, the operators began to doubt whether the. valve was actually closed. Therefore they attempted to determine from a data prinout the temperature at the valve outlet._. The, temperature was elevated, and the. operators. _ j interpreted this to mean that the valve was leaking and therefore did did not close the isolation valve in order to avoid losing the option of automatic pressure regulation. 5. During normal operation the pressure in the reactoFiT~1sratsi.~ ~ ~ Approximately two minutes after the shutdown of the feedwater suppiv. pressure in the reactor dropped _to _110 ata due to_~the open relief valve..At this pressure the high pressure emergency--- l cooling system was automatically activated. The high pressure l emergency cooling system pumped water into the reactor, and the l water level in the pressuri:er began to rise. l 6. At some time ir. the period 4-11 minutes after the shutdown of the i feedwater supply, the operator noted that the pressuri:er was l completely filled with wate, and he therefore shut off the high RAS D 850 044 g a. ._.,e

j pressure emergency cooling systen in order to maintain the option of pressure regulation. 7. Eight minutes after the beginning of the accident, the auxiliary feedwater supply to the two steam generators was established with ) opening of the two isolation valves, compare item 3. Due to a ) snail leakage in one steam generator, it was blocked. ) Approximately 11 minutes after the beginning of the cecident, the ~i 8. 1 water level in the pressurizer had sunk so much that the operator restarted the high pressure emergency cooling system. During the l periodic use o.f the high pressure emergency cooling system as well 1 as with the continued blow-off through the still open valve, the _ operator was then able to regulate the water level in the pressurizer. In this process the pressare in.the primary system was stabilited. 9. After approximately 15 minutes the rupture disk on the condensation tank was burst due to an overpressure; then water and steam flowed out into the reactor containment. After as little as 71/2 mint $tes, a pump which was designed to pump water out of the containment into a tank in the auxiliary building was started automatically. This may have been due to the fact that the safety valves of the :endensation tank were opened prematurely so that water was forced into the containment. Since the containment was not sealed at this time, because there was no 0.3 atm over-pressure in the containment, water from the containment was automatically pu= ped over into a tank in the auxiliary building. Due to a leak in the system and/or an overflow from the tank, approximately 40 m3 of water spilled onto the floor in the auxiliary building; at this time radioactive noble gases and iodine were released into the building's air filtration system. The filtration system is designed to hold up iodine and particulate radioactive materials, but not to stop noble gases (mainly xenon and krypton) which were thus released into the environment.

10. Due to the open relief valve, tihe pressETe'~condiMons 'iM 'the

reactor were such that the water in the reactor system was very close to the boiling point. This caused either a lack of water or the generation of steam bubbles at the main coolant pump inlet. The steam bubbles recondensed in these pumps where the pressure was higher, and this caused the development of vibrations in the ^ pumps (cavitation). Vibrations in the pumps can be noted in the 4' control room. Approximately i I/4 hours after the beginning of the accident, the vibrations were so strong that the operator g shut down the pumps in the one loop and approximately 1 3/4 hours 7,, after the start of the accident he shut down.those in the other loop. In this process most the heat removal via the steam generators g;, 1 i was stopped, and the temperature in the reactor began to rise again. g (u_d a gg ca 1 fa, 850 045 1 .f

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11. In approximately 14 minutes the temperature above the reactor core rose to more than 330*C. Above this point the metering instruments can no longer track the temperature.

In the subsequent period the reactor core temperature was such that the fuel elements were damaged. 21/3 hours after the start of the accident, the pressurizer isolation valve was closed. During the periodic use of the emergency cooling system and with the opening and closing of the remote-controlled isolation valve, an attempt was made to control the pressure in the reactor. Approximately 3 hours after the isolation valve was closed for the first time, i.e., approximately 51/3 hours after the start of the accident, the conditions were stabili:ed to such an extent that the core again had adequate cooling.

12. Approximately 7 1/2 hours after the start of the accident, it was decided to reduce the pressure in the reactor in order to be able to shift to the normal shutdown cooling system which can only be used at low pressure.

In the subsequent hours the con-ditions were such that the cooling again was not satisfactory and the core remained overheated, with new fuel damage resulting.

13. After 8 3/4 hours, the pressure had dropped so much that the low pressure emergency cooling system was started up.

Ten hours after the start of the accident, the cooling of the reactor had been improved. so much that the temperature above the reactor core could again be measured.

14. After approximately 16 hours, i.e., at approximately 2000 hours on Wednesday, stable cold conditions were established. A main circulation pump was started. The steam generator and the turbine condensor were put into operation. The temperat 2re and pressure were kept at approximately 160*C and 70 atm The temperature measurements of various spots in the core, acwever, continued to indicate higher temperatures, locally approximately 310*C.

15. For all of Thursday, March 29, this conditions was kept unchanged.

On Thursday afternoon a coolant water specimen indicated a high level of radioactivity (0.8 C1/ml); this confirmed that there was extensive damage to the fuel. p

16. On the night of Friday, March 30, the operators proceeded to reduce the temperat'tre, but they were frustrated. The reason for this was that it was assumed that there was a gas bubble at the top of the reactor tank at that time.

It was thought that the w - bubble had a volume of approximately 30-40 m3 4 850 046

s The bubble which was assumed to be all at the top of the reactor tank probably contained hydrogen formed during the reaction between l fuel cladding and the water at a time during the accident when the fuel temperature was extraordinarily high. In addition, it probably contained steam, radioactive and non-radioactive gases from the fuel and small quantities of hydrogen and oxygen which were formed during irradiation of the water (radiolysis). Ort that day and the next, the posttilated bubble gave rise to a number of thoughts concerning the possible risk of explosion or a risk of the bubble expanding due to the continued drop in pressure and thus reducing the cooling of the core. Withi~ view to,' among other' things,liuously removed 'from' the prim the removal of the bubble from thepri5ary's~ystemTwateiwaiionti system via the so-called "let down line" and sent to the tank for radioactive water and air in the auxiliary building. j On March 30, the pressure in this tank temporarily reached a level close to the maximum operating pressure. In order to reduce this i l pressure and thus make it possible to continue removing gas from the primary system, connections were established with the containmenc l so that the radioactive gas and water could be pumped into it from l the above-mentioned tank in the auxiliary building. The work on l this project, however, entailed a series of releases of radiqactive gases into the environment on March 30. For a brief time on Friday, it was thought tn.t the bubble was growing, but later it was found that it had again shrunk. On Saturday, March 31 at approximately 1600 hours the bubble was 3 considered to have decreased in si:e to approximately 13 m, and l on Monday, April 2 it again decreased significantly in size. y The method which was used to assess the si:e of the bubble was relatively uncertain. There are indications that the bubble may the entire time have been considerably smaller than initially indicated or that it mainly originated from-(consisted of] steam.' ~

17. Approximately 5 hours after the accident began, the reactor con-(c@$,

tainment was automatically sealed off since the pressure rose above l 0.3 atm (overpressure]. During the accident radioactive gases and g~ hydrogen were released into the containment. At approximately l 1400 hours on Wednesday, March 28 there was a brief pressure rise b l to approximately 2 atm. This was pressumeably due to the ignition of hydrogen generated in connection with the overheating of the core. The pressure transient caused the reactor containment sprinkler system to go into operation. The system pumped approximately 20,000 liters of water containing NaOH into the containment. NaOH is added 10 850 047 O -m-

to the sprinkler water in order to bind any iodine which may be present. On Saturday, March 31 approximately 1.5% hydrogen was measured in the containment atmosphere. On Monday, April 2 the concentration was 2.3%. The hydrogen content in the containment represents a risk since a hydrogen concentration in the atmospheric air of 4% is f'.ammable, and between 8% and 20% there is the risk of an explosion.

18. On Friday, March 30 water samples were taken from the reactor primary cooling system and gas specimens were collected from the containment in order to provide a basis, for planning subsequent measures. The taking of these specimens was relatively; difficult and entailed radiation doses to 2 employees of, respectively, 3.1 and 3.4 rem.

19. After the above-mentioned gas releases on March 30, more work was done on establishing safer methods of handling the radioactive gases in the auxiliary building.

At the same time an attempt was made to establish new cooling systems which were to be connected to the secondary side of the reactor steam generators in order to ensure stable long-term cooling of the reactor core. In this process the circulation of water in the primary cooling system outside of the containment it-self was avoided; the use of the normal long-term cooling system is supposed to have caused this. The new secondary cooling system is supposed to ensure an operating pressure above the primary system pressure so that leakage from the primary to the secondary system is avoided.

20. On April 5 it was no longer possible to observe the assumed bubble.

The speed with which it disappeared indicates that its size was less than originally assumed and that it consisted partly of steam from the area in the core where the high temperature was measured.

21. On April 5, 2 hydrogen recombiners were connected in order to remove the hydrogen in the containment. Using one of them, the hydrogen concertration was reduced on April 17 from approximately 2

2.3% to approximately 1.36%. ?-

22. On April 16 an increase was noted in the release of radioactive iodine from the plant, probably because the iodine filters in the a :

auxiliary building's ventilation system were at the saturation point and therefore began to loose their effectiveness. Even though the quantity of iodine released was small, the first filter was replaced on April 20. 850 048gb u

23. With the cooling method discussed in item 14, on April 23 the water temperature of the primary system dropped to approximately l

75'C and the highest measured temperature in the reactor core j had dropped to approximately 135'C. The plan subsequently on May 2 was to shut down the main cooling pumps and to shift to t cooling by natural circulation between the core and the steam generators. Until the new cooling systems mentioned in item 19 can be' connected to the steam generators, the heat will continue to be removed from it using the main condensors.

24. Parallel with the installation of more backup systems to ensure the long-term cooling the reactor core, the operators are proceeding with building systems, including a storage tank, which.are, required to remove.the radioactive material from the primary cooling system, containment and auxiliary buiiding.

~ 3.2. Plant Condition After the Accident It is the view of the NRC staff that the outer fuel elements in the I reactor core are relatively undamaged but that the center and other portions are heavily damaged and heavily deformed in terms of geometry i l without, however, melting having occurred. This view is based on available data concerning the release of fission products, hydrogen t production, temperature and other measurements in the reactor core during and after the accident, as well as an estimate of the period for which and the extent to which the reactor core was exposed during the accident. The reactor primary cooling system contains a large percentage of the more volatile fission products (for example, iodine and cesium), but only limited quantities of the less volatile ones (for example, strontium) and traces of uranium. When the reactor core is removed, it will probably be possible to clean the primary system without significant problems. c::=0 h e containment holds large quantities of radioactive fission gases c' and iodine, the latter partly in the air and in the water in the containment sump and partly deposited on the walls and equipment. At the present time it is uncertain whether there are also significant l quantities of the more long-lived radioactive isotopes, especially c% ' l cesium, the presence of which may make clean-up very difficult, g In the auxiliary building outside of the containment, there is a tank b with primary system water which contains _ radioactive _ gases and iodine. _ In addition, the floor and filter are contaminated. _ Clean-un here can probably be carried out without significant difficulty, but it l may be a time-consuming process. l l l 12 850 049 e r 4 .i

During the Senate hearing on April 23, 1979, plant representatives declared that their insurance would cover damages at the site up to 300 million dollars. They stated. as well that it would probably be possible to put the damaged reactor back on line within 2-3 years. 4. Health Physics 4.1. Overview of Dose Rates and Doses The first leak connected with the accident probably occurred due to the pumping of water from the reactor containment into the auxiliary building after the first fuel element damage (compare Section 3.1, item 9). The leakage on the following days originated mainly from the auxiliary building. The release of radioactivity took place via j the building ventilation system. The ventilation system has a built-in filter which holds up iodine and particulate radioactive material. Thus the leakage is limited to a large extend to noble gases. On Wednesday, March 28 at 1045 hours, a radiation level of 3 mrem per 1 hour (mrem /hr) was measured at the ground surface approximately 500 m away from the plant, but the level approximately 1,500 m away was 1 mrem /hr. During Wednesday and Thursday, the measurements continued, among others , from aircraft. Outside of the plant's environs,. radiation levels of up to 4 mrem /hr at the ground surface was measured. On Thursday morning the highest measured radiation level at the ground surface outside of the immediate vicinity of the plant was 0.15 mrem /hr. On Friday between 0800 and 1500 hours, radioactivity again escaped from the auxiliary building into the atmosphere (compare Section 3.1, item 16). Just outside of the plant area, 20-2S mrem /hr were measured, but elsewhere the level was at most a few mrem /hr. Helicopter measurements showed that the cloud was following the river towards the northwest and that 3-10 km away from the plant it could no longer be detected. On Friday afternoon and Saturday morning, the highest radiation level in the area of the plant was 2 mrem /hr measured from an aircraft. On Sunday the level at the ground surface along the cloud's main axis 2 dropped to 0.6 mrem /hr at a distance of approximately 500 m and to l 0.06 mrem /hr at a distance of 3-5 km away from the plant. Over the next few days the radiation level dropped.to virtually the normal background level. Aside from a few leaks due to work operations which caused brief radiation levels of up to 1.5 mrem /hr at the ground l N surface, the radiation level remained for the most part at the background

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i All of these radiation levels are instantaneous values and they lasted for limited periods of time. The values are therefore no direct j expression of the impact on the environment which. occurred. The biological damage is detensined from the collective doses which are covered.in Section 4.3. 4.2. Types of Isotopes Released and Contamination It was found that the most important isotopes in the release were xenon-133 (half-life S.3 days), xenon-135 (half-life 9.2 hours) and traces of radioactive iodine, mainly _ iodine-131 (half-life 8.1 days). The most important radiation source was thus the noble gas xenon, which cannot i be incorporated by living organisms. Separate analyses of samples of milk from the area were taken. In some of them traces of radioactive iodine were found. The highest measured concentration. was 41 picoeurie/ liter (pCi/1). For comparison, it can be mentioned that Professor R. Linnemann from the Institute of Radiology of the University of Pennsylvania declared that in. Pennsylvania up to 2-300 pCi/1 were measured in Pennsylvania due to Chinese tests of nuclear weapons. The authorities do not consider it necessary to take special measures until the level. exceeds 12,000 pCi/1. Cesium was also found in :ntik, but at a concentration which is considered to originated entirely from past nuclear weapons tests. In addition to milk tests, samples were also taken of vegetation, soil and water. In individual cases the soil and vegetation samples exhibited traces of iodine. In addition, ce:ium was found in concentrations corresponding to those origanting from past nuclear weapons tests. Therefore, apparently no cesium was released in connection with the accident. After March 30 several hundred em3 of slightly contaminated water from a tank at the plant were dumped into the river. Water samples from c: ::--S the river exhibited not radioactive iodine, q Approximately 2 weeks after the accident, there was a small leakage of iodine from the auxiliary building. The increase was blamed on the filters in the ventilation system gradually becoming saturated. After g%, the filters were replaced, the leakage was again reduced. 4.3. Dose Burdens and Consecuences The point which is decisive for any biological damage is We cumulative dose. A hypothetical person who spends 24 hours in i day out_o_f, doors _. at a spot where the measured dose rates will have received the cumulative doses given in the table below. l 14 850 051 e-w ..,,,w w w

Period Estimated Cumulative Dose (area). Dose (ares) 3/28-3/29. 45 45 3/29-3/31 29 74 3/31-4/1 8 82 4/1 -4/2 0 82 . 4/2 -4/3 1 83 4/3 -4/4 0 83 4/4 -4/5 1 84 4/5 -4/6 2 86 4/6 -4/7 0 86 The cumulative dose toliisliinagina'y person'(due to the accident) is ~ ~~ ^ r thus approximately 90 mrem. A representative person who remained within a radius of 16 km from the reactor would have received approximately 11 mrem, while a representative person within 80 km of the plant would have receive approximately 1.6 mrem. For the sake of comparison, we should mention that the background radiation in the area is approximately 120 mrem /yr. The EI5ulative doseTtolhe 'popuiation within a' radius of 80 km (approximateip" ~ '~ ~ 2 million people) is estimated at approximately 3600 man-res*. On the basis of these figures, it can be expectsd that there will occur less than 1 fatal case of cancer (0.7) and less than 1 non-fatal case of cancer (0.7) over a period of approximately 30 years. Individual employees at the plant received larger doses than people in the area. Three individuals received approximately 4 rem, three received app _roximately.1 rem,_ while the rest received less than I rem. The International Radiologic Protection Commission recommends that [the Bose to] employees should be kept under 5 rem /yr.

• Man-rem is an expression for the number of people who received [a dose of] radiation times the dose, measured in rem, received by these people.

l 850 052 4

The above-cited radiation doses are based on conditions before approximately April 10. 'the doses to the population after this time are estimated, on the basis of the available data, to be negligible. 5. Readiness 5.1. Emergency Plannihii ~~ l The guidelines for emeigencyjla'idiiiji"foEnu~c1'e'ariplants in the USA ~ are described in ths "Guidi ant 6eEEllit for Developieht and~!Ealua'tfon -~ of State and I.ocal Government Radiological Emergency Response Plans in Support of Fixed Nuclear Facilities", NUREG 65/111, published by the NRC in December 1974 With reference to these guidelines, the individual states which have l either expectations of having a nuclear facility within their boundaries or near them must develop emergene)/ plans to handle the case.of an., _. i l. accident at a-nuclear facility as part of its general disaster plans. With coordination provided by the states, emergency plans should be developed by the local authorities in cooperation with the plant owner. The plans should state the breakdown of responsibility and the functions ~~ ~ and procedures which are to be followed in the case of an emergency. ~~ The state and local authorities as well as the operators of nuclear plants must meet at least once a year in order to discuss and update the coordinated plans. l In addition, there is a report from a working group " Planning Basis l for.the Development of State and Local Government Radiological Emergency Plans in Support of Light Water Nuclear Power Plants", NUREG-0396, l issued by the NRC and EPA in December 1978. c===0 The working group recommends, among other things, that - emergency planning be carried out to handle a number of different c==. 3 accidents; g C22:2 - as a point of departure for determining the areas within which plans @rj,1 should be made for emergency measures, zones can be expected within g

1) 10 miles for inhalation doses from the passage of the cloud as well as whole-body doses from the passage of the cloud and precipitation, bjea y

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i 5.2. I,ocal Emergency Planning i The state of Pennsylvania maintains - the Pennsylvania Emergency Management Agency (PEMA) which is the guiding institution and is supported by a number of other authorities, including '- the Division of Environmental Resources (DER), the radiologic unit of which is the Bureau. of Radiation Protection (BRP). This unit estimates the radiation doses on the basis of measurements. The PEMA plans the state's general disaster emergency preparations and coordinates among the agencies. In doing so the PEMA aids the individual offices and communities with their planning and execution of emergency measures. T'he PEMA referi to I:he Giver'no(oi Pennsylvania. The Governor can only l order the take-over of the direction of the local, authority's emergency i measures if he declares a state of emergency. The PEMA carried out emergency planning for the TMI plant accident together with. the agency heads and held regular practice sessions. PEMA has an underground control room (Emergency Operation Center) to which the various agency representatives are summoned. These repre-sentatives have the right to make decisions within the agency's own area of responsibility. In a disaster situation 25-30 people will be present. The centrol room is the work place for PEMA officials during normal working hours. Outside of normal working hours, there is always an official on watch. It is_PEMA_'s_ experience that during normal working hours it is possible to summon the emergency service heads within 10-15 minutes. Outside of normal., working hours, for example.at 0300,. the PEMA leadership can be summoned within 20-30 minutes, while the other workers and agency representatives can be on the spot within 1-2 hours. From the control. room there are telex links with all the offices in the state to be used for general disaster planning and direct telephone lines to the reactor facility, and there is also a radio link between the plant and Dauphin County where Harrisburg is located. The general disaster emergency plans "are tested in staff exeMes ~ ~ approximately once a quarter. j With reagrd to alerting the public, it was_found that PEMA,_as_the_. _ _ director of general disaster emergency planning, has voluntary arrange-h l 1, tgh 850 054 --e-.- -w --,w-.--

ment with the state's local radio sta?.icus which in emergency situations immediately broadcast information anc messages from the PEMA or the Governor to the population. At night, in the given case sirens will be used as a precursor for radio messages. 'Ihe implementation of emergency measures is normally the responsibility of the individual agencies and communities which, hcwever, can request PEMA to provide support from the state's resources Each office has developed plans with resource outlines, including a chart showing traffic capacity. These traffic capacity calculations are coordinated among the offices. In addition, in each office in at least one town preparations are made for one or more reception stations (schools, gymnasiums, etc.) with the necessary facilities to house and feed evacuees. Copies of the office's and community's detailed plans are also available at PEMA. In the PEMA control room there is also a chart showing the road capacity and housing locations. l l Therefore, it is possible for PEMA (regardless of when an evacuation situation may arise in Pennsylvania) on short notice to coordinate l or direct the execution of an actual evacuation even if, of course, it is more difficult the more people there are involved. Among other things, it was noted that in 1972 PEMA had evacuated approximately 250,000 people in connection with a flood, including 80,000 people in 3 hours, and in 1977 they evacuated approximately 30,000 people in connection with a corresponding natural disaster. Under prevailing plans, it is supposed to be possible to evacuate an area of 5 miles in 2 1/2 hours. In the accident situation [under discussion here] PEMA in 3 days was able to develop detailed plans for evacuating an area of 20 miles on the basis of-the pre-e.tisting I plans. I Under this plan, it was considered possible to evacuate an area of 10 miles in 7 hcurs and and area of 20 miles in 10 hours. This evacuation sequence made no allowance for the fact that some of the population would have left the area.already. l It is noted that within 5 miles of the plant, there live 24,522 people l - 10 miles of the plant, there live 133,672 people - 20 miles of the plant, there live 636,073 people. 18 850 055 6 e m Y W

In case of an evacuation, a large portion of the population would be expected to use their own transport resources to flee. Therefore extra fuel for the filling stations is urgently required so that the evacuation cast take place. ~ In addition, evacuation can be carried out with school buses and private buses and there are 4 trains reserved with room for 1,000 people for this purpose. Fihally approximately 100 helico ' be requisitioned for transporting patients [to hospitals] pters can and for traffic control, along with a large number.of ambulances. It was found that there was no' hospital within the 5 mile area, there were 3 in the 10 mile area and 9 in the 20 mile area. Overall, there are approximately 4,000 patients in these hospitals. The number was reduced to approximstely 1,000 within 1-2 days, partly by refusing to admit new patients for non-emergency care and partly by sending home patients whose treatment had started and who did not absolutely require hospital care. 5.3. Readiness to Take 55asurements... No BRP takes measurements in the environment and estimates radiation doses and on this basis advises PEMA whether it should institute protective measures, for example advising the population to stay indoors or evacuation. ~ ~ ~ Fedeial agencTeY (DOE and ' EPA) provide advice on teams which, after an emergency is announced, can be sent to support the local ] agencies with regard to implementing measures and to giving advice on evaluating the measurement results as well as on the possible institution of, protective _ measures. 5.4 secuenie'of Eveits in an Eie~riency' ~-~ ~ At the present time there is no official outline of the sequence of { events for emergency measures. The list given below is there based on verbal statements by the representatives of the various agencies. March 28, approx. 0700 PEMA vas advised by the plant of an accident which would possibly entail a state of increased readiness in the immediate vicinity of the plant. PEMA immediately informed the DER and contacted Dauphin County as well as the other offices closest to the plant. In addition, PEMA called in all of the emergency planning officials, and Governor nornburgh was also advised. Finally, a telephone line was ,.850 056 g[7 a 19 l g

arranged with the 6 offices within the 10 mile area. approx. 0800 The situation was discussed with Governor Thornburgh against the background of the data obtained. from the plant. It was decided that PEMA should implement updating of the plans for evacuation.within the 20 mile area. 0915 EPA was advised by the NRC. In the afternoon representatives arrived from the Department of Energy (DOE) with a radiologic measurement team. In addiation Airborne Radiological Monitoring Systems from the Department of Defense with aircraft and helicopters (were sent]. I.ater on Wednesday representatives of EPA arrived with a radiologic monitoring team. On the basis of the measurements taken, it was simply decided to allow PEMA/ DER to main-l tain a state of increased readiness. March 30 With regard to emergency planning the situation first became critical when on the morning,of Friday! l March 30 there were several leaks of radio-activity into the air..These leaks came from the primary cooling water which was pumped into an auxiliary building frem the reactor contain-ment. 0840 The general alarm was sounded. This was due to a report that at a level of 600 ft. above the ~ top of the chimney ~ 1200 mre::i/hr was~ measured.' This report was never confirmed and was pre-sumeably in error. The combined emergency service heads were again summoned and Governor c==S Thornburgh was advised. Q b l l 0915 In the NRC control room in Washington, it was l decided to suggest to PEMA that they begin to g,* prepare an actual evacuation of the residents in the 10 mile area. However, Governor g Thornburgh was the one who was actually g supposed to make the final decision. 0925 The NRC recotanendation was endorsed by The~ ~ ~ ~ Chairman of the Commission, Hendrie. G3

rat, 20 850 057

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l 1115 The measurements which, PRP had_taken gave,no justification for proposing the institution of protective measures. Governor Thornburgh therefore contacted the NRC in Washington by telephone, and the NRC declared that the report of a measurement of 1200 mres/hr above the chey could not be confirmed. Then Governor Thornburgh decided, as a precau.ti.onary measure in case of large leaks, to advise that pregnant women and children ~under sch~osi age should leave the 5 mile area and that the population within the 10 mile tone should stay indoors until further noticM. In addition, all schools within the 5 mile area were supposed to remain closed until further notice. 1500 The leakage was greatly reduced when a large portion of the radioactive water in the auxiliary building was stored in a tank at the plant. 2200 NRC adviced that plans be developed for evacuation within the radious of 20 miles. April 3 During the course of Tuesday the situation improved sufficiently for Governor Thornburgh to decide to recind the' measures taken. 5.5. Discussion The recommendation that pregnant women and children under school age should leave the 5 mile area meant, in the estimation of PEMA, that . approximately 50% of the population within this :ene left the area , on their own initiative (approximately 13,000 people) and that approximately 1/3 of the population thin the 10 mile radius left the area at their own initiative as well (approximately 45,000 people). Various estimates indicate that a total of between 80,000 and 2C0,000 people left the area of their volition. This took place over a period of I-2 days. In this process there was no panic and, as far as PEFA is aware, there were no personal injuries or fatalities which could be attributed to the voluntary evacuation. = It is noted that evacuation of a large number of people is not such a rare event in Pennsylvania in connection with natural disasters, etc. With regard to the news media coverage of the events during the accident situation,. PEMA called attention to the fact that the local media gave Toot ~- infomation input and understood how the information reflected the actual situation to a large extend. However, it was characteristic 'o 850 058 f 22 n.g l v

~ that the further away one was from the area. of the events, the more dramatized was the media's depiction of the situation. ThePEMA'r'epresent'atives ~ stressed as well that..it._ was_, cons _idered that emergency preparations were made for a situation such as this. The actual problem (as in other emergency situations) was a lack of adequate and correct information. In addition, there was a desire to make it possible to make more extensive use of radio communications than were available. Finally, PEMA stated that', of course, it planned to' review 1*s t emergency plans for nuclear facilities in light of the experience gained from the 1MI accident. Finally it should be noted that EPA declared that there were no plans to change its guidelines for the institution of protective measures for the population on the basis of radiation doses. The American IutEo'r' ides' H' ndling of'the' Three Mile Is' land Accident 6. ~ ~ a I Immediately after the NRC received their report of the accident at Three Mile Island, inspectors were sent to the plant from NRC's regioaal inspection plant. The NRC HQ in Washington manned its so-called " Incident Response Center". Since the scope of the accident was known, NRC sent a larger staff headed by Director Harold Denton, who at the same time was designated as President Carter's personal representative. NRC's task on the spot was partly to ensure that control measures were carried out and partly to make certain that all activities designed to gain full control over the reactor were carried out to an extent which guaranteed the minimum possible risk of the release of radioactivity to the environment. In this regard the NRC staff on the spot attempted to assess the conceivable consequences.of the individual activities before they were permitted [to be released). This was done with the aid of NRC's own specialists in Washington.as well as specialists from universities and national research laboratories. Measurements of the radiation level in the plant environs were also carried out by EPA and DOE, the Food. and Drug Administration (FDA) and Pennsylvania's own emergency section, PEMA. Together with a large number of [one work unknown] and the other y American reactor manufacturers, DOE provided extensive technical assistance both with regard to advice and equipment, for example new lodine filters were flown in from California. oJ ~ 850 059

\\ Now NRC wishes to continue to carefully monitor the subsequent work in order to bring the DfI Reactor No. 2 to a safe condition and to have the plant subseque.ntly cleaned or disassembled. Together with this activity, NRC has initiated assessments of what occurred at TMI with a view to determining what conseguences it m_ight have over, respectively, the short and long. terms in order to ensure that corresponding accidents do not occur again. The preliminary results of this assessment have been presented to the independent technical agency, the Advisory Committee of Reactor Safeguards (ACRS) which to a large_ ext'enr_ has agreed with NRC's preliminary conclusions, including the institution of a number of administrative safety measures which are discussed below in Section 6.1. These safety measures, of course, apply first and foremost to B f, W reactors, but also are applicable to all other reactor plants. NRC and ACRS have also undertaken a preliminary evaluation of what should be done over the slightly longer term and here have pointed out technical improvements (among oth:.r things, measurement of the, water level in the reactor core) and emphasize the conditions which are analyted in the safety reports and the greater need for operator qualifications and training. The work done to clarify the accident, sequence, to analyze it and to draw the necessary conclusions is, however, so great in scope that it can be completed at the earliest within 1/2-1 year. On April 11, 1979 President Carter nam an independent commission which was assigned to study the conditions surrounding the accident and to issue a report within 6 months. In connection with the accident, the commission's task is to: - evaluate the technical sequence of events and the causes of them, - evaluate the role of the reactor owner, l - evaluate the readiness and actions of NRC and the other Federal, state and local authorities. - evaluate NRC's licensing and supervision practice in relation to TMI; - determine whether the public's right to information concerning the BtI accident was upheld and what should be done in a similar l situation in order to provide the public with precise, en=" rive and correct information, - develop apprcpriate recommendations based on the results of the g above-mentioned studies. 0 23 D m J 850 060 4 en-, r-um-e,--,w-n e- , -, -+ m-me

The Sub-Countittee for Control of Nuclear Legislation under the Senate Committee for Environment [ Protection) and Public Facilities headed by Senator Gary Hart has held 2 hearings up until now. Hearing No. 2 on April 23, 1979 was devoted to shedingJght on the following questions: - ne cause of the accident - The emergency plans of Pennsylvania - n e reactor's condition and recommissioning of the plant - Fixing of costs in connection with the accident. 6.1. The Consecuences of the Accident for Other Light Water Reactors, Especially B & W Plants In addition to analyting the activities in connection with TMI Reactor No. 2, NRC has, as mentioned above, evaluated the. accident in order to determine what technical or administrative measures can be taken 1 i in order to ensure that corresponding accidents do not occur at other t plants. In this connection, among other things, NRC focused on the following conditions of major importance for the ac:ident: - The auxiliary feedwater system was blocked - The pressuri:er relief valve was not closed again when the primary system pressure dropped back to normal - The pressurizer level indication is not a reliable reflection of the water level in the primary system when the primary system is holding water with a temperature corresponding to the boiling point and the relief valves are open at the same time 4 - The containment was not sealed when the high pressure emergency g cooling system was started up, and this made it possible subsequently for radioactive water to be pumped into the auxilia v building under M uncontrolled conditions g - The high pressure emergency cooling system was regulated by the operator solely on the basis of the pressuri:er level indication. This led to a further reduction in the amount of water in the primary f3 system W - The shutdown of the main coolant pumps in order to prevent cavitation g contributed to the damaging of the fuel while cooling by_ natural, circulation at this time was hampered due to the generation of steam,_ 6 3 in the primary system. (Eil., 24 l 850 061 f

In addition, NRC focused on five general conditions in B & W reactors which, it is assumed, contribute to these units levying greater requirements on the operating personnel than is the case with other reactors. These are: - The quantity of water contained in the seconda27 side of the steam generators is 3-8 times smaller than in other pressure water reactors. All other things being equal, this means that B $ W reactors are more sensitive 'co interruptions in the feedwater supply than pressure water reactors - B 6 W reactors also generally seem to be more sensitive to changes in the secondary system than other pressure water reactors - B 4 4 reactors are constructed in such a way that the rressuri:er relief valve opens under certain operating conditions without the reactor being automatically shut down - The level difference between the heat sink in the steam generator and the reactor tank may cause B 6 W reactors to have less effective natural circulation than otner pressure water reacrors - B 5 W reactors have no automatic shutdown [ scram mechanism] when there is a malfunction in the secondary loop. This.is consonant with the fact that t! ey are constructed so that the pressurizer relief valve is actuated more frequently than in other pressure water reactors. On the basis of the foregoing, NRC has required the 75nersi~of reactor facilities which use B 4 W reactors to study: - Whether steam pockets can develop in the primary system without being noticed - That the automatic high pressure emergency cooling system is not shut down until backup emergency cooling is established or all temperatures are at least 30*C below the blowing plate of the existing system pressure - That procedures exist for training operators in han<img a number of operating events which normally lead to pressure transients in the primary system - That at all times there are at least 2 independent pipes available for supplying emergency feedwater. h 0k s s 9 @e 850Obh ~

s As a continuation to the above-mentioned injunctions, NRC's technical staff adviced the commission to order the plant owners to shut down the B 6 W reactors currently in operation until the above-mentioned conditions were met in order to avoid the possibility of similar l ~ accidents and in order to allow technical changes and improvements ~ concerning, among other things, the feedwater system and the reactor scram system to be made. The cossaission began this adjustment on April 27 and ordered the staff to issue a formal instruction.to the plant owners to.this end. For all other pressure water and boiling water-reactors, on April 14, 1979 NRC ordered the plant owners to examine a number of corresponding conditions. Reports from the individual plants on these conditions should be ready within 10-30 days, after which a decision will be mado as to whether short-term or long-term measures are required. 850 063 9 4 1 o 'I 4 t.

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r %generat ' .) + to a T hG ) to,. N o / I e i m i / / Feedwater pump ' I h g) Air coolln' g g g A " scs Reactor NL e' l V / 6 ~~ i ~ i ation plant I Ileactor coro Hain coolant pumps h 'to a p and condensato pump s / f \\e/ 'Auxill'ary building / / Drain tank y g g /' \\ fii yyuuuususus suusu us, / / Emerg. cool. (gas)y tank'.mdr f' ~l Deca CC '+ (J water r/ sihmar/ra hs/M b $ "T-- h ~ l ~ 7.*S4.29 April 1979 fff.f 'yw w w w w u n u w w m 3 N v, O e 1 9 .}}