ML19225B101

Forwards IE Info Notice 79-16, Nuclear Incident at Tmi. No Action Required

ML19225B101
Person / Time
Site:	05000072, University of Utah
Issue date:	06/22/1979
From:	Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	Turley R UTAH, UNIV. OF, SALT LAKE CITY, UT
References
NUDOCS 7907230514
Download: ML19225B101 (1)

Text

{{#Wiki_filter:- I J h o. UNITED ST ATES f1 SIC y{ g,., cf',g NUCLEAR REGULATORY COMMISSION REGION IV }Kh,7$ % W,p-{d, fj 611 RY AN PLAZA DRIV E, SUIT E 1000 0, ' Al'9lyS / ; AR LINGTCN, TE X A5 76012 't,.....f June 22, 1979 In Reply Refer To: Docke t No. 50-72 50-407 The University of Utah ATTN: Dr. Richard E. Turley Reactor Adainistrator College of Engineering Salt Lake City, Utah 84112 Centlemen: The enclosed Infernatiou Notice No. 79-16 is forwarded to you fcr in-formation. No specific action is requested and no written response is required. this If you desire additional information regarding this matte.r, please contact office. Sincerely, / p .,/ p 1/f AA l y' Karl V. Seyfrit / Director

Enclosure:

1. IE Information Notice 79-16 2. List of IE Information Notices Issued in 1979 410 125 l907~9 s,

UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 IE Information Notice No. 79-16 D' ate: June 22, 1979 Page 1 of 1 NUCLEAR INCIDENT AT THREE MILE ISLAND Description of Circumstances: On March 28, 1979, the Three Mile Island Nuclear Power Plant, Unit 2 experienced core damage which resulted from a series of events which were initiated by a loss of feedwater transient. The seriousness of this incident makes an under-standing of its causes important to research and experimental facilities. This notice transmits copies of Inspection and Enforcement Bulletias (IEBs) 79-05, 79-05A and 79-05B to inform you of the details as known at the time the bulletins were issued. Enclosures 1 and 3 of IEB 79-05 and Enclosure 2 of IEB 79-05A have been deleted from this transmittal. IEB's similar to the 79-05 series were issued to licensees with boiling.ater reattors and pres-surized water reactors supplied by vendors other chan Babcock and Wilcox. No specific action or written response to this Information Notice is required. If you desire additional information regarding this matter, contact the Director of the app ~ropriate NRC Regional Of fice.

Enclosures:

1. IE Bulletin No. 79-05 with Eaclosures 2. IE Bulletin No. 79-05A with Enclosures 3. IE Bulletin No. 79-05B 410 126

UNITED STATES x' NUCLEAR REGULATORY COMMISSION 0FFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 April 1,1979 P IE Bulletin No. 79-05 5 6 NUCLEAR INCIDENT AT THREE MILE ISLAND [N Description of Circumstances: M G On March 28, 1979 the Three Mile Island Muclear Power Plant, Unit 2 h experienced core damage which resulted from a series of events which were initiated by a loss of feedwater transient. Several aspects of the incident may have general applicability in addition to apparent generic applicability at operating Babcock and Wilcox reactors. This bulletin is provided to inform you of the nuclear incident and to request certain actions. Actions To Be Taken By Licensees (Although the specific causes have not been determined for individual sequences in the Three Mile Island event, some of the following may have contributed.) For all Babcock and.Wi,1cox. pressurized water reactor facilities with an operating license: 1. Review the description (Enclosure 1) of the initiating events and subsequent course of the incident. Also review the evaluation by the NRC staff of a postulated severe feedwater transient related to Babcock and Wilcox PNRs as described in Enclosure 2. 'hese reviews should be directed at assessing the adequacy of your reactor systems to safely sustain cooldown transients such as these. 2. Review any transients of a similar nature which have occurred at your facility and determine whether any significant deviations from 50 expected performance occurred. If any significant deviations are found, provide the details and an analysis of the significance and j any corrective actions taken. This material may be identified by 9 reference if previously submitted to the NRC. 6tJ 3. Review the actions required by your operating procedures for coping with transients. The items that should be addressed include: ) ~ SL i 410 127

IE Bulletin No. 79-05 April 1, 1979 Page 2 of 3 Recognit;cn of the possibility of forming voids in the primary a. coolant system large enough to compromise the core cooling capability. E, b. Operator action required to prevent the formation of such voids. ] D Operator action required to ensure continued core cooling in. c. .m the event that such voids are formed. s 1 4. Review the actions requested by the operating procedures and the y training instructions to assure that operators do not override automatic actions of engineered safety features without sufficient cause for doing so. 5. Review all safety related valve positions and positioning require-ments to assure that engineered safety features and related equip-ment such as the auxiliary feedwater system, can perform their intended functions. Also review related procedures, such as those for maintenance and testing, to assure that such valves are returned to their correct positions following necessary manipulations. 6. Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the containment totassure that undesired pumping of radioactive liquids and gases will not occur inadvertently. In particular assure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.. List all such systems and indicate: Whether interlocks exist to prevent transfer when high a. radiation indication exists and, b. Whether such systems are isolated by the containment isolation signal. 7. Review your prompt reporting procedures for NRC notification to assure very early notification of serious events. 9N The detailed results of thesc reviews shall be submitted within ten (10) days of the receipt of this Bulletin. [f s b a1 410 128

t IE Bulletin fio. 79-05 April 1, 1979 Page 3 of 3 \\ Reports should be submitted to the Director of the appropriate !!RC Regional Office' and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Construction Inspection, 20555. y l'ashington, D.C. 1 For all other operating reactors or reactors under construction, this e Bulletin is for information purposes and no report is requested. N Approved by GAO, B180225 (R0072); clearance expires 7-31-80. Approval was given under a blanket clearance specifically for identified generic problems. y

Enclosures:

1. Prelininary ?!otifications Three Mile Island - Pt:0-67 and 67A, B, C, D, E,F,G 2. Evaluation of Feedwater Transients w/ attachment 3. List of IE Bulletins issued ~ in last 12 months ~'>+ a2. G ,b b k 9 4i0 129

Page 1 of 3 EVALUATION OF FEED'.!ATER TRANSIENT D D, A loss of offsite power occurred at Davis-Besse on l'ovember 29, 1977, D which resulted in shrinkage of the primary coolant volume to the degree that pressurizer level indication was lost. A recommendation to convey this information to certain hearing boards resulted in the attached discussion and evaluation of the event. This discussion includes a review of a loss of feedwater safety analysis assuming forced flow, which predicts dispersed primary system voiding, but no loss of core cooling. During the Three Mile Island event, however, the forced flow appears to have been terminated during the transient.

Attachment:

Discussion and Evaluation of Davis-Besse Transients ' C. Di: D 'r 3 )!3 ( 1 m heu 3lL 410 130

EXCERPT FROM MEMORANDUM ENTITLED " CONVEYING NEi! INFOPJIATION TO LICENSING BOARDS.- DAVIS-BESSE UNITS 2 6 3 AND MIDLu:D UNITS 1 & 2", DATED JANUARY 8, 1919, fro:! J.S. CRES~n' ELL TO J.F. STREETER. w 3. Inspection and Enforcement Report 50-346/7S-06 doct:aented that f pressurizer level had gone of fscale for approxinately five minutes during the Nove_:b'er 29, 1977 loss of off site pouar event. h There are scue indications that.other BW plants cay have prob-L le=s maintaining pressurizer level indications during transients. In addition, under certain conditions such as loss cf f eeduster R at 100% power with the reactor coolant pt=ips running the pres-( suricer may void completely. A special analysis has been per-9 forned concecning this event. This analysis is attached as. Because of pressuriier level etaintenance prob - lems the sizing of the pressurizer may require fur ther review. s Also noted during the event uns the fact that Tcold vent off-scale (less than 5200F). In addition, it was noted tha t the nakaup f1cu monitoring is limited to less than 160 gpa and that makeup flow may be substantially greater than this value. Tnis information should be exanined in light of the require-ments of CDC 13. DISCUSSION AND EVALUATION T51eevent at Davis Besse which resulted in loss of prdssurizer level indication has been revicued by NRR and the conclusion was reached that no unreviewed safety question existed. The pressurizer, together with the reactor coolant makeup system, is designed to maintain the primary system pressure and water level within their operational limits only during normal operating conditions. Cooldosm transients, such as loss of of fsite power and loss of feed-water, saceti=es result in prinary pressure and volume changes that are beyond the ability of this systen to control. The analyses of and experience with such transients show, however, that they can be sustained eithout compronising the safety of the reactor. The principal concern caused by such transients is that they night cause voiding in' the primary coolant systea that would Icad to loss of ability to ade-quately cool the reactor core. The cafety evaluation of the loss of of f site pcuer transient shows that, though 1cvel indication is lost, some uster recains in the pressurizer and the pressure does not decrease belou abcut 1600 psi. In order for voiding to occur, the pressure t:ust q decrease below the saturation pressure corresponding to the systen tengerature. 1600 psi is the saturation pressure corresponding to i; 605 F, which is also the caxiaue allowable core outlet temperature. Voiding in the pri=ary systen (excepting the pressurizer) is precluded ( in this case, since pressure does not decrease to saturation. e J e .C

Section 3.. The safety analysis for more severe cooldotn transients, such as the loss of f eedwater event, indicates that ~ to less than the syste= voluae exclusivethe water voluae could decrease of the pressurizer. During such an event, the emptying of the pressurizer would be followed by p' a pressure reduction below the saturation point and the foraation of 1] small voids throughout much of the priasry system. This vould not h result in the loss of core cooling becavse ~ over a larga volume and forced flow would preventthe voids would be dispersed [' them from coalescing sufficiently to preveat core cooling. The high pressure coolant injection pumps are started auccmatically when the pricary pressure decreases below 7 1600 psi. The re fo re, any pressure reduction which is suf ficien t to allow voiding vill also result in water inj ection which g vill rapidly restore the pricary water to normal levels. For these reasons, we believe that the inability of the pressurizer. and normal coolant nekeup systen to control some transients does not provide a basis f or requiring more capacity in these systens. General-Design Criterion 13 of Appendix A to 10 CFR 50 requires instrumentation to monitor variables over their anticipated ranges for " anticipated operational occurrences". Such occurrences are specifically defined to include loss of all offsite power. The fact that T cold goes'off scale at 520 F is not considered to be a deviation from this requirement because this indicator is backed up by wide range tenperature indication that extends to a low licit of 50 F. Seither do we consider the takeup flow monitoring to deviate since the aoount of cakeup flow in excess of 160 gpm does not a significant factor in the course of these occurrences. ppear to be a The loss of pressurizer unter level indication could be consi'dered to deviate f ron CDC 13, because ceans of determining the prinary coolantthis level indication provides the principal inventory. However, provision of a level indication that would cover all anticipated occurrences may not be practical. As discusced above,. the loss of feedvater event can lead to a comentary condition wherein no neaningful level exists, because the entire pri=ary system contains a. steam water mixture. It should be noted that the introduction to Appendix A (last paragraph) recognizes that fulfill ent of some of the criteria cay not always be _7 approp ria t e. This introduction also states that departures from the Criteria must he identified and justified. The discussion of CDC 13 in the Davis Besse FSAR lists the water level instrumentation, but e does not nention the possibility of loss of water level indication 3 t during transients. This apparent osiscion in the safety analysis vill be subj ected to further review. G

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4 m_ e e 410 132 m_ .m ~ ~ *

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UflITED STATES ( ' flVCLEAR REGULATORY C0 OilSSIO 1 0FFICE OF IllSPECTI0tl Arid ErlFORCEMEtiT WASHIflGT0:1, DC 20555 IE Bulletin flo. 79-05A [ Date: April 5,1979 y Page 1 of 5 7d f!UCLEAR IflCIDErlT AT THREE MILE ISLA?!D - SUPPLEMEilT b i Description of Circumstances: $u Preliminary information received by the flRC since issuance of TE D Bulletin 79-05 on April 1,1979, has identified six potential auman, design and mechanical failures which resulted in the core damage and radiation releases at the Thrce Mile Island Unit 2 nuclear plant. The information and actions in this supplement clarify and extend the o. iginal Bulletin and transmit a preliminary chronology of the TMI accident t_hrough the first 16 hours (Enclosure 1). 1. At the time of the initiating event, loss of feedwater, both of the auxiliary feedwater trains were valved out-of-service. 2. The pressurizer electromatic relief valve, which opened during the initial pressure surge, failed to close when the pressure decreased below.sthe actuation level. 3. Following rapid depressurization of the pressucizer, the pressurizer level indication cay have lead to erroneous inferences of high level in the reactor coolant system. The pressurizer level indication apparently led the operators to prematurely terminate high pressure injection flow, even though substantial voids existed in the reactor coolant system. 4. Because the containment does not isolate on hich pressure injection (HPI) initiation, the highly radioactive water from the relief valve discharge was pumped out of the containment by the automatic initiation of a transfer pump. This water entered the radioactive waste treatment system in the auxiliary building where soi.ie nf it overflowed to the floor. Outgassing from this water and discharge through the auxiliary building ventilation system and filters was Q the principal source of the offsite release of radioactive noble D gases. ai 8 5. Subsequently, the high pressure injection system was intermittently ?! operated attempting to control primary coolant inventory losses through the electromatic relief valve, apparently based on {- pressurizer level indication. Due to the presence o f steam and/or noncondensible voids elsewhere in the reactor coolant system, I this led to a further reduction in primary coolant inventory. 410 133 -._ n: \\

IE Bulletin fio. 79-05A Date: April 5, 1979 Page 2 of 5 6. Tripping of reactor coolant pumps during the course of the transient, f to protect against pum.p damage due to pump vibration, led to fuel damage since voids in the reactor coolant system prevented natural b circulation. [ Actions To Be Taken by Licensees: j R For a'l Babcock and Wilcox pressurized water reactor facilities with an operating license (the actions specified below replace those specified L -in IE Bulletin 79-05): 1. (This item clarifies and expands upon item 1. of IE Bulletin 79-05.) In addition to the review of circumstances described in Enclosure 1 of IE Bulletin 79-05, review the enclosed preliminary chronology of the TMI-2 3/28/79 accident. This review should be directed touard understanding the sequence of ever.;s to ensure against such an accident at your facility (ies). 2. (This item clarifies and expands upon item 2. of IE Bulletin 79-05.) Review any transients similar to the Davis Besse event (Enclosure 2 of IE Bulletin 79-05) and any others which contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (.ies). If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a description of any corrective actions taken. Reference may be made to previous information provided to the f!RC, if appropriate, in responding to this item. 3. (This item clarifies item 3. of IE Bulletin 79-05.) Review the actions required by your operating procedures for coping with transients and accidents, with particular attention to: 6 Recognition of the possibility of forming voids in the primary a. coolant system large enough to compromise the core cooling u capability, especially natural circulation capability. Y b. Operator action required to prevent the formation of such ~ voids. i. y c. Operator action required to enhance core cooling in the event 1 such voids are formed. 410 134 27

/ IE Bulletin No. 79-05A ( Date: April 5, 1979 Page 3 of 5 (This item clarifies and expands upon itern 4. of IE Bulletin 79-05.) 4. g m Review the actions directed by the operating procedures and training y$ instructions to ensure that: Operators do not override auto.Tatic actions of engineered n. a safety features. d Ei b. Operating procedures currently, or are revised to, specify n that if the high presrure injection (HPI) system has been automatically actuated because of low pressure condition, it must remain in operation until either: (1) Both low pressure injection (.LPI) pumps are in operaticn and flowing at a rate in excess of 1000 gpm each, and the situation has been stable for 20 minutes, or (2) The HPI system has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure. If 50 dagree subcooling cannot be maintained af ter HPI cutoff, the HPI shall be reactivated. Operating procedures currently, or are revised to, specify c. that in the event of HPI initiation, with reactor coolant pumps (RCP) operating, at least one RCP per loop shall remain opera ting. d. Operators are provided additional infomation and instructions to not rely upon pressurizer level indication alone, but to also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water inventory in the reactor primary system. 5. (This item rev ses item 5. of IE Bulletin 79-05.) Verify that emergency feedwater valves are in the open pasition in S accordance with item 8 below. Also, review all safety-related valve positions and positioning requirements to assure that i valves are positioned (open or closed) in a manner to ensure the proper operation of engineered safety features. Also review i related procedures, such as those for maintenance and testing, = to ensure that such valves are returned to their correct positions 4 following necessary manipulations. ^

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( IE Bulletin tio. 79-05A Date: April 5, 1979 Page 4 of 5 ~ v 6. Review the containment isolation initiation design and procedures, y and prepare and implement all changes necessary to cause contain:aent isolation of all lines whose isolation does not degrade core cooling n capability upon automatic initiation of safety injection.

j 4

7. For manual valves or manually-operated motor-driven valves s hich ? could defeat or compromise the flow of auxiliary feedwater to the

(

steam generators, prepare and implement procedures which: { a. require that such valves be locked in their correct position; or b. require other similar positive position controls. 8. Prepare and implement immediately procedures which assure that two independent steam generator auxiliary feedwater flow paths, each with 100% flow capacity, are operable at any time when heat removal from the primary system is through the steam generators. When two inde-pendent 100% capacity flow paths are not available, the capacity shall be restored within 72 hours or the plant shall be placed in a cooling mode which does not rely on steam generators for cooling within the nexts 12., hou rs. t! hen at least one 100% capacity flow path is not available, the reactor shall be made subcritical within one hour and the facility placed in a shutdown codling mode which does not rely on steam generators for cooling within 12 hours or at the maximum safe' shutdown rate. 9. (This item revises item 6 of IE Bulletin 79-05.) Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping of radioactive liquids and gases will not occur inadvertently. .I In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation. List all such systems and indicate: J i a. t!hether interlocks exist to prevent transfer when high radiation indication exists, and J b. t!hether such systems are isolated by the containment isolation i signal. 410 136 eeu a

IE Bulletin flo. 79-05A Date: April 5,1979 Page 5 of 5 10. Review and modify as r.ecessary your maintenance and test procedures to ensure that they require: D Verification, by inspection, of the operability of redundant .-a a. ) safety-related systems prior to the removal of any safety-g related system from service. 'd i b. Verification of the operability of all safety-related systems ]$ when they are returned to service following maintenance or testing. L A means of notifying involved reactor operating personnel c. whenever a safety-related system is removed from and returned to service. 11. All operating and maintenance personnel should be made aware of the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three tiile Island Unit 2 plant and other actions taken during the early phases of the accident. 12. Review your prompt reporting procedures for fiRC notification to assure very early notification of serious events. For Babcock and Wilcox pressurized water reactor facilities with an operating license, respond to Items 1, 2, 3, 4.a and 5 by April 11, 1979. Since these items are substantially the same as those specified in IE Bulletin 79-05, the required date for response has not been changed. Respond to Items 4.b through 4.d, and 6 through 12 by April 16, 1979. Reports should be submitted to the Director of the appropriate f!RC Regional Office and a copy should be forwarded to the fiRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, DC 20555. For all other reactors with an operating license or construction permit, this Bulletin is for information purposes and no written response is required. Approved by GAO, B 180225 (R0072); clearance expires 7-31-80. Approval was given under a blanket clearance specifically for identified generic problems. g g m

Enclosures:

.t. 1. Preliminary Chronology of Till-2 3/33/79 Accident Until Core Cooling Restored. g ~. 2. List of IE Bulletins issued in last 12 months. . i n 1 77 4 i u 1J/ - ~ ~ ~ - = = = = -, _ _ = = _ - - - -. 2

IE Bulletin No. 79-05A i Date: April 5,1979 Page 1 of 3 w-PRELIMINARY CHRONOLOGY OF TMI-2 3/28/79 ACCIDENT, GI UNTIL CORE COOLING RESTORED g q 3 TIME (Approximate) EVENT ~ y about 4 AM Loss of Condensate Pump ga (t = 0) Loss of Feedwater Turbine Trip t = 3-6 sec. Electromatic relief valve opns (.2255 psi) to relieve pressure in RCS t = 9-12 sec. Reactor trip on high RCS pressure (2355 psi) t = 12-15 sec. RCS pressure decays to 2205 psi (relief valve should have closed) { t = 15 sec. RCS hot leg temperature peaks at ^"ec-611 degrees F, 2147 psi (450 psi over sa tura tion) ~ t = 30 sec. All three auxiliary feedwater pumps running at pressure (Pumps 2A and 2B started at turbine trip). No flow was injected since discharge valves were closed. t = 1 min. Pressurizer level indication begins to rise rapidly t = 1 min. Steam Generators A and B secondary level very low - drying out over next couple of minutes. nr i m t = 2 min. ECCS initiation (HPI) at 1600 psi jj u t = 4 - 11 min. Pressurizer level off scale - high - one 'J HPI pump manually tripped at about 4 min. c 30 sec. Second pump tripped at about 10 min. 30 sec. hF t = 6 min. RCS flashes as pressure bottoms out at 1350 psig Hot leg temperature of 584 degrees F). 4\\3 \\3no 2

7 IE Bulletin tio. 79-05A ( '- Date: April 5, 1979 Page 2 of 3 ~ TIf4E EVEtiT ( 72 t = 7 min., 30 sec. Reactor building sump pump came on. d t = 8 min. Auxiliary feedwater flow is initiated

p by opening closed valves

{j t = 8 min. 18 sec. Steam Generator B pressure reached minimum O L t = 8 min. 21 sec. Steam Generator A pressure starts to recover t = 11 min. Pressurizer level indication comes back on scale and decreases t = 11-12 min. flakeup Pump (ECCS HPI flow) restarted by operators t = 15 min. RC Drain / Quench Tank rupture disk blows at 190 psig (.setpoint 200 psig) due to continued dis. charge of electromatic relief valve t = 20 - 60 min. System parameters stabilized in saturated 7' - " condition at about 1015 psig and about 550 degrees F. t = 1 hour, 15 min. Operator trips RC pumps in Loop B t = 1 hour, 40 min. Operator trips RC pumps in loop A t = 1-3/4 - 2 hours CORE BEGIfiS HEAT UP TRANSIEtti - Hot leg temperature begins to rise to 620 degrees F (off scale within 14 minutes) and cold leg temperature drops to 150 degrees F. (HPIwaterl t = 2.3 hour Electromatic relief valve isolated by s operator af ter S.G.-B isolated to prevent leakaga 'T t = 3 hours RCS pressure increases to 2150 psi and C electromatic relief valve opened j 5: m t = 3.25 hours RC drain tank pressure spike of 5 psig [> t = 3.8 hours RC drain tank pressure spike of 11 psi - [- RCS pressure 1750; containment pressure ( increases from 1 to 3 psig .n i gg [4 \\ d i J/

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/( IE Bulletin tio. 79-05A Date: April 5,1979 Page 3 of 3 TII'E EVEflT r t = 5 hours Peak containment pressure of 4.5 psig t = 5 - 6 hours RCS pressure increased froro 1250 psi to ? 2100 psi t = 7.5 hours Operator opens electromatic relief valve to I depressurize RCS to attempt initiation of L RHR at 400 psi t = 8 - 9 hours RCS pressure decreases to about 500 psi Core Flood Tanks partially aischarga t = 10 hour 28 psig containment pressure spike, containment sprays initiated and stopped after 500 gal of t;aOH injected (about 2 minutes of operation) t = 13.5 hours Electromatic relief valve closed to repressurize RCS, collapse voids, and start RC pump t = 13.5 - 16 hours RCS pressure increased from 650 psi to 2300 psi t = 16 hours RC pump in Loop A started, hot leg temperature decreases to 560 degrees F, and cold leg temperature increases to 400 degrees F. indicating flow through steam generator Thereafter S/G "A" steaming to condensor Condensor vacuum re-established RCS cooled to about 280 degrees F., 1000 psi flow (4/4) High radiation in containment All core themocouples less than 460 degrees F. 3 Using pressurizer vent valve with small S makeup flow Slow cooldown jj a RB pressure negative T j e 3L N

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.c UNITED STATES fiUCLEAR REGiltATORY COMMISSION \\ OFFICE OF Ill5PECTION AND Et'FORCEttEtlT WASHIriGTON, DC 20555 APRIL 21, 1979 y IE Bulletin 79-053 ?!UCLEAR INCIDiNT AT THPIE. MILE ISLAND - SUFFLEMENT b Bescriptica of Circumstances:. I f._ ..c. c ^ Continued tlRC evaluation or the nuclear incident at Three Mile Islan'd 5 -/ - - -Unit 2 has identified ceasures-in addition to those discussed in IE i, Bullatin 79-05 and 79-05A which shou'Id be acted upon by licensees with reactors designed by 8&L_ -As discussed in Item 4.c. of Actinns to be taken by Licensees in IES 79-05A, the preferred mode of core cooling-following a trancient or-accident is to provide forced f' low using' reactor coolant pumps. It appears that natural circulation was not successfully achieved upon a securing the reactor ccolarrt pcmprduring the first two hours of~ tha Three Mile Island (TliI) tio. 2 incident of March 28, 1979. Initiation of natural circulation was inhibited by significant coolant voids, possibly aggravated by release of noncondensible gases, in the primary f coolant system. / To avoid-this potential for interference with natural ( ..cf rculation, the operator should ensure that the primary system is x subcooled, and remains subcooled, before any attempt is made to establish - ]. natural circulation.~ flatural circulation in Babcock and Wilcox reactor systems is enhanced by maintaining a relatively high water level on the secondary side of the cnce through steam generators (OTSG). It is also promoted by injection of auxiliary feedwater at the upper nozzles in the OT5Gs. Tne integrated Control System autcmatically sets Me OTSG level setpoint. to 50% on the operating range when all reactor coolant pumps (RCP) are secured.

However, in unusual or ab.non al situations, manual actions by the operator to increase steam generator level will enhance natural circulation capability in anticipation of a possible loss of operation of the reactor ceolant pumps.

As stated previously, forced flow of primary coolant through the core is preferred to natural circulation. Other means of reducing the possibility of void formtion in the reactor coolant systen are: 1q k A. Minimize the operation of the Power Operated Relief Valve (PORV) on 7 the pressurizer and thereby reduce the possibility of pressure O reduction by a blowdown through a PORV that was stuck open. gf t. e i e \\.

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IE Bulletin 79-033 April 21, 1979 Page 2 of 4 B. Reduce the energy input to the reactor coolant system by a prompt reactor trip during transients that result in primary system pressure increases. 'Dy This bulletin addresses, among other things, the means to c.chieve these cbjectives. g 2 4 . -. (., Actions To Ba Taken by 1.icensees: Q-T C ' For all Babcock and Hilcox pressurized water reactor facilitier with an n operating license: (Underlined sentences are codifications to,.and J b .p,;7._. 5 superseda, IG-79-05A). .1~ ~ [?.l. - Develop procedures and train operation personnel on methods of '.. 5.4 establishing and maintaining natural circulation. The procedures E cc._- -and training must include means of monitoring heat removal efficiency T . M " by available plant. instrumentation. The procedures must also.contain %.F a method of assuring that the primary coolant system is subcooled by '. 7,- ~ ' at least 50 F before natural circulation is initiated. 5 -. .@.y In the event that these instruct. ions incorporate anticipatory fillino ' f. ~~ - -j. of the OTSG prior to securing the reactor coola it pumps, a detailed f;' c ~ analysis should be done to provide guidance as to the expected system 1. response. The instructions should include the following precautions: 2 ..,, : c a. maintain pressurizer level sufficient to prevent loss of level ~ indication in the pressurizer; ~ b. assure availability of adequate capacity of pressurizer heaters, for pressure control and maintain primary system pressure to satisfy the subcooling criterion for natural circulation; ( maintain pressure'- temperature envelopr-within Appendix G limits c. for vessel integrity. u. ~ Procedures and training shall also be provided to maintain core cooling in the event both rain feedwater and auxiliary feedwater are lost while {. in the natural -irculation core cooling mode. g c- .. 2. P,odify the actions required in Item 4a and 4b of IE Bulletin 79-05A 'c; to take into account vessel integrity considerations. y ~ "4. F.eview the action directed by the operating procedures and -f training instructions to ensure that: ~ .- a. M u Operators do not override automatic actions of engineered a. safety features, unless continued operation of engineered [ -(gf ~ T_'~ [y ~ _ m. m I il 1 [r L.9 i '!.2,. t.~.'... l v, 1 1 .u. f 6 =.

IE Bulletin 79-053 [ April 21,1979 ( Page 3 of 4 A r safety features will result in unsafe plant conditions. For 2 examole. It continued operatiqn or engineered safety features. Mould threaten reactor vessei integrity tnen the HPI shouid be g secured (as notec in ot2) below). b ~.z.G

b.. Operating procedures. currently, or are revised to, specify that

-+-:;u k . %, e'1 if the-Mgtr pressure injection. ? actuated because of low pressure (HPI) system has been automatically z

j operation -until either:.

condition, it nust re: rain in- .?. R J.W ' C. - (1)< Both low pressure injection. (LPI) pumps are in operation .. _- n and ficwing at a rate in excess of 1000 gpm each and.the _ ,Jg ~.. situation-has-been stable for 20 minutes, or if -: (2) The HPI system has been in operation for 20 minutes, and TC. W G.', alt hot and cold leg temperatures are at least 50 degrees below the-saturation temperature for the existing.RCS . n:s ~ pressure. '. If 50 degrees subccoling cannot be maintained .' 'g,J. : :.. cfter HPI cutoff, the HPI shall be reactivated.

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The decree of subccolina-beycnd 50 degrees F and the length of time ~ . 4-t liPI is an oneration shall be limited by the pressure / _temaera ture considerations for the vessel in teority." 3. Following detailed analysis, describe the modifications to design and procedures-which you have implemented to assure the reduction of the likelihood of automatic actuation of the pressurizer PORV during ~ anticipated transients. This analysis shall include coasideration of a rodification of the'high pressure scram setpoint and the PORV opening setpoint such that reactor scram will preclude opening of EW in-Enclosure 1.the PORY for the spectrum of anticipated transients discusse Changes developed by this analysis shall not fo;- these anticipated transients. result in increased frequency of pressurize 4. Provide procedures end training to operating personnel for a prompt canual trip of the reactor for transients that result in a pressure increase in the reactor coolant system. These transients include: u-a. loss of nain feedwater .. a w" b. turbine trip -?'&y ,,f."- r. P.ain Steam Isolation Valve closure c. ~ %Ig

d.. Loss of offsite power 4d e.

Low OTSG level y f. low pressurizer level. l s. 41n i43 e ..==-=........z---~---=..==---- - - -=.. -.. . ~.. _

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IE Bulletin 79-053 April 21,1979 Page 4 of 4 5. Provide for flRC approval a design revies and schedule for impimntat of a safety grade autcmatic anticipatory reactor scram for loss of feed-water, turbine trip, or significant reduction in steam generator level %.jj 6. The actions ret;uired in item 12 of IE Bulletin 79-05A are modiffed as O ~ - u follcws: .-N O RavicN ycur piv...,at reporting proceduras for fiRC notification to assure-j. P '~ that NRC is notified within one hour of the tima ' ".M _a controlled or excected condition or operation. the reactor is not in cidI' ' > ' 2.1@ < .'. ra i n ta i n ed w i t.1 iiH C.an ocen con:nnucus comunication channel shall be L YW Jil2F 7. Procose chances, as recuired, tt$' those technical specifications which, must be rodtried as a result or your 1rnolementinn the above items Response schedule for BaW designed f'acilities: For Items I, 2, 4 and 6, alt facilities with an enerating license a. ] respond within 14 days of receipt of this Bullet 6. 3, b. For Item 3, all facilities currently operating, respond within 24 hours. operating, respond before resuming operation.All facilities with [%- .For Iterns 5 and 7, all facilities with an operating license respond c. in 30 days. Of fice and a copy should be for.iarded to the flRC Off Enforcamnt, Division of Reactor Operations Inspection, Washington, D. C 20.2 3. For all other pcwer reactors with an operating license or construction pernit, this Bulletin is for infonration purposes and no written response is required. Approved by GAO, B'IS0225 (R0072); clearance expires 7/31/80. was given under a blanket clearance specifically for identified generic [ Approval problers, d - b .f S N,

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EXTRACT OF Bn COW.U.'ilCATION - RECE!VED BY tlRC Enclosure .I_.4T__ROI7d CT 104 Page 1 of 4 k' CO?frIfl0ING PlVIEM OF THE SEQUENCE OF EVEllTS LEADING TO -2 Utf FARCH ZU.1979 SHCM5 THAT ACTIOTi Carl RE TAKEri TO PHO TH.AT THE PILOT-OPERATED RELIEF VALVE (PORY) t;OUrlTED UTI THE PRE 55 FAYE A SIGNIFICA?fr PROBABILIT'l UF DCCURR !7TT TIEGRADE Ti!E SAFETT OF THE AFFECTED PLNITS WITH THIS ACTIOil th]ST~ TO tor'WL; UPSET OR ACCICDfT C0:tDITIOT:s NUR LEAD TO UNHEVIEWED SAFET 7.u2 /LTTICIPATED TRAfiSIENTS OF C0!!CERM ARE-h L ' LOS$ GF NXTERTh ELECTRICAL LOAD 2 ki 7 ',. 2 .TUR31NE TRIP.. J. d m-J 3. LOSS OF PAIN FEEC7AATER ~' ~ f !}. 1.055 0F CCMcE715ER VACUUM. ~.- ~ ] - M 5.. IHADVETITEfff CLOSURE OF l'AIN STEAM ISOLATION V!n.VES ,.L ark.' - : 3 A frGEi? 0F ALTEiL~TIVES WERE~CoiGIDERED Irl DEVELOPI 2 I' UIJ 1NCLUDIim: 'i l FiESTR CTI?iO fNACTOR f'O'JER TO A VA UE WilCH MOULD Th~ THE; P0rTf FGIITT5 REPAIHED AT-THEIR-CURREftT-VALUES. Tile. REACTOR PhviE 2. T55URE NO ACTUATIOM OF THE FORY.1.0"ERIMG Tile THE SETPOIllr FOR PORY ACTUATION RE!?. TINED AT LGUERUiG THE HIGH PRESSURE RE5CTOR TRIP SETPOI L. 07EPATITET PRESSURE (NiD TEM'EPATURE) 0F T!!E REACTOR OPERATIfrG PRES $UREACTIGTIU'l AND TO PROVICE ADEQU THE SETFOIriT FOR P0HV ACTUATION REFAIMEU AT CURRElif VALUE ITS THIS ALTERNATIVE 5?OULD REQUCE HET ELECTRICAL O 5. MENTTIPG THE_HICH PRESStfRE TRIP Ario THE PORY SET WP7 ACTUSTIO?! FOR TiiE CLASS OF ANTICIPATED EVEtirS OF C PRES 50RE OF THE REACTOR REicIt!ED AT ITS CURRErlf VALUE THE UESIGN

fi A?t LYSIS OF THE IIGACT OF T1!ESE VARICUS ALTERNATIVES AND THEIR iG A55URIfiG TE'.T THE PORY MILL tiGT ACTUATE FOR THE C

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(- i. REdjCN;:i TdE PE03ABILIT7 0F bORY Mi ACTIMTIC?1 FOR OTHER Ims/ssNG PRE 55U S RIZER SnFETf VALE { F J'S E. T Fr~ESER'/I?iG Press!" HELIEF CAP 5 CITY.FOR ALL HIch PRESSURE TPRIS ITS_ 3. ELMITMTBG THE M5IalLITT OF IliTRODUtilf0 UTiREVIEUED SAFETY CortCErug

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E5DUC TID THE fiME AT E4TCH THE STENT SYSTEff IIEfJ SU ~ QlE E7EhT EMRGEllCY FEED:ATEit FLC'ri KERE DELAYEO. - ( .s ED S$TPOIliT CbAt GES Off ALL RITIC PR TP I 5 IS V t Ti BL BW F1.Alf75 NRE CURRENTLY CAPA5LE OF R BACT TG 75% OF FULL '0W _Qg3 TRIP OF THE TURBUTE. THIS CAPABILITY TiEthIfi$S ACTUATI. 0 em* cD RELIce VALYES. THE CAPABILITY It! CREASES TF P.ELI A SUFFLY TO THE SYSTEM BY P~ TURTLING.THE UtiITS TO P011ER GEi E E TJ F TH E (f T .J' ... liOTE: .e. The effect of changing the reactor coolant system pressure trip setpoint upon peak pressurizer pressure is typitied by the attached figure 1. which was developed by ~ caw for a loss of feedwater transient. -h, h 4 _ .= e I --f 3 g ,.l :-.} ? \\ $ .li-3 410 146 'f ~. ... _. __ _. p... 1 ~>..- 2 ~ x s. ,e 9 ZT J

[ TAS!.E 1 SL%RY OF PROTECTIO.t AGAIrf5T PORY ACTUATIG;I f PIXP/IDED B7 PRDFOSED SETP0l?iT Dil1NGES FOR ALL ( NTTICIPATED TRA1151Etif5 EXTfP.CT OF BR CQPMJtlRCATIO:1 - PECE[VED BY tlRC 4/20p9. 3 T j'MICIPATED TRNESIEITT5 k"r!IO! HAVE OCCURRED AT B&M FLNIIS A'lD }illICU HOULD..C' H 65$d.L7 ACTIVATE P0hY AT T11E CURRENT SETPOINT (2255 PSIC)- I xn .2 ,9 Tlid31EE TRIP ,~:J.rX. 1 .c -~ -.'C< b LUSS OF EITi.RfAL ELECTRICAL LDAD . 16,' n.W! G ~ .,.~. :. L ~ h t.~: LCSS'0F MAIM FEErMATER. ', D.' LG35 OF CCMUEnsEftvACutM . -.' y -' g7 p -E - }:.- ? INMTERTEIIT CLOSURE OF ?tSIV h.TICIPATED TRAMSIENTSimIGFHAVE OCCURRED AT UaV PLA iTS NiD hila! '- + - L'0ULD R0rT,LLY ACT1! ATE FORY AT Tile PROPOSED SETForrrr (2 '.0 PSIG): ~ ' z: M. ,...i'.- .'i R!Tr[CIPATED TRNTSIEriTS.UilIOi HAVE NOT OCCURRED AT D7N PLAflT ~ 1 P' R03A3ILITT E'iEtiT3) NiD 11HIG1 UGULO NORFALLY AcrUATE PORY AT THE CURRENT SETFOIliT (2255 'PSIC): A :.* ~- SGhE CGftTRCL RUD GROUP HITilDRNIALS (MODERATE To llIGf t REACTIVITY .Eddul GROUPS NOT OTIIERUISE PROTECTED BY !!IGH FLUX TRIP)

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50DERATOFI DILUTIO:1. .. u. -.S .~,y'....*- 1 n ~ . f. ~ ~8NTICIP5TED'TRAri3IEriTS FrIICH 11 AVE tt0T CccunRED AT DFM PLA.'ITS (LO 3. i . EVEliTS) NiG WiiIC11 UGULD ACTUATE TIIE PORV AT Tire PR0_PO5ED SETPOIrlT -N 4 } .y. c.; (2450 PSIG): A. t . s'- -)_ L l A 50^!E Cat.TROI. ROD GROUP UITHDTUNALS (HIGII' REACTIVIT'Y UCRTl! t;0T- " h... - :. T[{ OTHERHISE PROTECTED BY HIDI FI,UX TRIP).

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l E-1 g f UNJTED STATES TG r. 8[ 4 j "o NUCLEAR REGULATORY COMMISSION l M 9/gf y. ADV!3ony cc:a.uTTrs. on aEr.croa c;, recur no:;

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' - Fonorrble Victor Gili:=iy U -~. Actir:g Chniman - L U.. S. Up= lear Regulahory Cceniccion .., Kadireton, D;' 20355 Deac Dr. Gilinsky: ~ T:.is letter is in respn:m to your's of April 18, 1979 t. hic'n requested diat the ACRS notify the 'Cor.missionecc in:nzdiately if we believe any of our oral reeccm:ndation= of April 17?should te acted upon bafore our nedt regularly GehMuled :necting at cich ve could prepro n fo:rol .let'ter. Toa Ccrr.mittee discoscM this topic by conferen-e telephona call on' April 19 and offers the folicWing. c wnts. All of the reccc.endsti'ons r.ade by the ACRS in l's~ nectir4 With the x Ccr._Liccloncrs on April 17/ 1979, ere genaric in niture cr:1 cpply to all P.ac..Nanc. Vere intended to requi're ipediato, chj.nges in o,0crotirq pr.o-ceduros or plant :03Sifications of o.reratirs F.mc. Such changes rhould c b cede p:Gy after study of their effects on overall safety. su:h stud-its chatu.c be c.3de by the licensees and their suppliers or consulta:s.c im-i by the NRC Starf. Tne Co caitthe b21ieves that. th2se studies should b2 b.gra in the naar future on a tire c: ale that will not divert the - .NR" Staff or the industry representatives from theit tacks relating to the cooldown of Tnree Mile Irland Unit 2. nawever, the Cc:r.i tteo br_~ ~ lic n.s th'at it s.ould bc.possible arri desircSle to ini tJate icdiately a navey o.k. oparntirs procedures for achiev. ins natuchl circulatich,. in-cludirry thc. cA6e when ofE:~ite po.22r is lost, arr3 the role of the prer-curier heaters in suc5,pr6ceduros. ~.5 v Ist its ocet'Ing cn April 16 ar.S 17,1979, the co=:ittee discusse.d trith ' ' the F.RC Staff the tx tter of natural circulation for the 'three MD.o Is. ? le.,d Unit 2 plant-Tne ce"ttee bellCves that thic taatter is receiv-S ~ ing e reful attention by the lac Staff and the 11ccr_Oce. N z w To ED3 for A.ppropria te Action. Distribution: Chm, Crarc, PE, OGC, OCA, SECY, . ej SD2, OlA. Rapifaxcd to EDO, PA. E

case.

79-1117.. t^ I e 4iU l 't / w-ee m6ee e p.umm emus eme me--=W g 6*D' 4 "" g

s ( l s, i I IW:Mahle Victo:* Gilin=hy 2-Jeril ho,.19,rg \\, w .q.- - 4-_r,._.L, _... -. ei y u... n .-1._--~~.- ~ ~ - W.2 c.:ier.it:cdn oc. w._r.cWicns to L% Cor213sion en April 17 were [ r ,..;- rc: mtended to apply to.nu c.e nile Iclend Unit. 2. . 2. J ' N.

.1

{.'f.-S.) ddlf't t::: 'drit0-- a f'.:.rthe:**.Prt. CD thO Tdtters 65 our Fh.y 10, 1973 .JL c-:::t:::ir-- - L -..- ~ - - - -,I. 6* .%e * ' ;..;s k, *..., ' sincerely, ....i ^

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r g.,74 NUCLEAR REGULATORY cot.WISSION UNITED STATES ,g 2 ag s =- l y.. y, e.sg ADVISORY CO!.'.!.'.LTTEE 0:2 REACTO3 SAFCCUARDS ~ (.

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April 18,1979 ~ . s - o . r,

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..=l'?~: -}.6.r..', .Ccwi'micnec Bra 6fa d L ('- . ;. ' y-} ~ Corr:.issioner Ahearne ~' ' ?.7 "$m - FRCM: R. F. Fraley, Exccutive Director Mvisory Ccinnittee on Reactor Safegt.'.rds -e s. j,'f,.- Attached for your infor-2 tion and use. is a copy of :the reco.=enda-- tions of the I.dvisory Cor.,. ittee on Reactor Safeguards s'alch Ware orally presented to and discussed with you on April 17,1979 re-J garding the recent accident at the Three !!Ile Island Nuclear Sta-tion Unit 2. s ~ [ /)V .t, R.;F. Fraley Executive Director . Attach ent: Recermndations of the NRC Advisory Co.r:nittee on Reactor Safeguards Re. the 3/28/79 Accident - at The force tiile Island Nuclear Station Unit 2 To 1.- p

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April 17, 1979 7 'RECG"'ENCATICNS O? WC ? UCLEAR RmUEATORY COcISSION ADVISOR'l COM CN REACICR SMEUARCS REARDING THE MARCH 28, 1979 ACCIDEC.T AT TdB TdRES HILS ISLCD UUCLEAR STATION UNIT 2 = presented orally to, and discussed with, the lac -' 'Eb ' 2. y Commissioners during the ACRS-Cc=missioner s Meeting i [ on April 17; 1979 - hwshirrgton, D. C. g G n. S Y_ L Natural circulatiqn is an important rode of reactor cooling, both as , 's 'j.. E o J.'. a planned process and-as a process that ray be used under abnormal f circumstances. Tae Ccamitte:: believes that greater understanding of '. ', ' 'this rede.of coolics ILrequired and that detailed analyses.should. ^' E'.f, by developed by licensees-cr-their suppliers. Tne analyses shotGd be W supported, as necessary, by experiment. Procedures should be ' de-- veloped for initiating' natural circulation in a safe manner -and for providir:g the operator with assuran::e that circulation has,- in fact, 1'.2 heen established..Tois may require. Installation of instrumentation to measure or indicate flew at low water velocity. g, ..L. t .- c'Iha use of natural circulat'on for decay heat recoval following a loss of offsite pawer sources reqcires the maintenance of a suitablo over-pressure on the reactor coolant sys tem. This overpressurei may be assured by placing the pressurizer heaters on a qualified onsite powr source with _a suitable arrange. ment of heaters and power distri-v bution to provide--redundant: capabilit.y. Presently ~ operating EWR t plants should b2 surveyed expeditiously to determine whether such arrangements can be provided, to assure this aspect of natural circula-tion capability. Tae plant oterator should b. adequately informed at all times con-cerning the conditions of reactor coolant system operation bhich might affect the capability to place the systen in the natural circu-lation rode of operatio:r or to sustain such a mode. Of particular importance is that information #11ch-might indicate that the reactor coolant system is approaching the saturation pressure correspondirrg-to the core exit temperature. Tnis impending loss of syste:a over-pressure will signal to the operator a possible loss of natural circulation capability. Such a warning may be derived from pressur-yD irer pressure instruments.and hot Icg temperatures in conjt'nction with conventional stea a tables.- A suitab'; display of this information should be provided to the plant m%ctor at all times. In addition, [ consideration should be given to the use of the flow exit tempe ra-- c~ . tures from the fuel subasse:-blies, eere available, as an additioncd. Indication of natural circulation- "7 ' ' C a 1 't i.U iaL esa ee se h- = =. -

2-The exit tmperature of coolant fro.n the core is currentlv measured by ther:ccouples in rany RlRs to ' deter $Ine core perfo rme.nce. 'ihe. y Cccittee reccer. ends that these temperature measurements, cs currently ~ ? available,.be used to guide the operator concerning core status. Tne N range of the information displayed ard recorded should include tha h-full capability of the-themoccuples. It is also recon'. mended that N cther existing instrir:cntation be exc;ained fo r its passible use in h r"-* ting operatir,9 action during a. transient. ? s i . ' _-fwi'th regard to the. definition and implementation of instrutnantation [3 m:r - The: ACRS rec m-.cnds thdt operating power reactors be given priority ..f-d thich provides additional inform.stion to help diagnose and follow the course of a seric"s accident. This should include improved sampling .9 : precedures under accident corditions and techniques to help previde ~ f.~ improved guidance to offsite authorities, should this be needed 'Ihe . L' _...Cormittee recommends that a phased Impler,entation approach be-ec- . u.- r J - played so that technicues can,be adopted shortly af ter they are . f t;.7 -jtr3ged to be appropriate. Tne'ICRS rcccc= ends that a high priority be placed on the develop ent and implcmentation of safety research on the behavior of light water , reac. ors during an =alous transients. Tne NaC may find it appropriate to develop a capability to simulate a wide range of pastulated tran-sient and accident conditions in ordar to gain increased insight into measures thich can be taken to improve reactor safety, h ACRS uishes to reiterate its previcus reccamandat.ons that a high priority be given to research to improve reactor safety.

Considera tion should be given to the desirability of

.di tional . ' equip.ent status :nonitoriag on.various engineered safoguards features - and their supporting services to help assure their availability at all times. 'lhe ACF.5 is continuing its revicW of the implications of this accident a:vd hope to provide further advice z s it is developed. .n 5 e ~ w.- 9 c B ~ ~ t. - a. ^ "..-A U t_ 1 ~_. 4 "~7 I U 224- -- -:. ----- 5 5 =. = ~- -.= r = = === r- = -==- = = = =.= = =E. = = -m a-- = ~ 7. ' C :^ :._ ~. ~~ - - - -.. - --:=...~--=.-~----

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IE Informatica Notice 79-16 June 22, 1979 LISTING OF IE INFORMATION NOTICES ISSUED IN 1979 Information Subj ect Date Issued To Notice No. Issued 79-01 Bergen-Paterson Hydraulic 2/2/79 All power reactor Shock and Sway Arrestor facilities with an Operating License (OL) or a Construc-tioa Permit (CP) 79-D: Attempted Extortion - 2/2/79 All Fuel Facilities Low Enriched Uranium 79-03 Limitorque Valve Geared 2/9/79 All power reactor Limit Switch Lubricant facilities with an Operating License (OL) or a Construc-tion Permit (CP) 79-04 Degradation of Engineered 2/16/79 All power reactor Safety Features facilities with an Operating License (OL) or a Construc-tion Permit (CP) 79-05 Use of Improper Materials 3/21/79 All power reactor In Safety-Related Components facilities with an Operating License (OL) or a Construc-tion Permit (CP) 79-06 Stress Analysis of 3/23/79 All Holders of an Safety-Related Piping Reactor Operating License (OL) or a Construction Pe rmi t (CP) 79-07 Rupture of Radwaste 3/26/79 All power reactor Tanks facilities witn an Operating License (OL) or a Construc-tion Permit (CP) Enclosure Page 1 of 2 s<m

IE Information Notice No. 79-16 June 22, 1979 79-08 Interconnection of 3/28/79 All power reactor Contaminated Systems with facilities with an Service Air Systems Used Operating License As the Source of Breathing (OL) and Pu Proces-Air sing fuel facilities 79-09 Spill of Rai oactively 3/30/79 All power reactor Contaminated Resin facilities with an Operating License (OL) 79-10 Nonconforming Pipe 4/16/79 All power reactor Support Struts facilities with a Construction Permit (CP) 79-11 Lower Reactor Vessel Head 5/7/79 All Holders of Reactor Insulation Support Problem Operating Licenses (OLs) Construction Permits (cps) 79-12 Attempted Damage to New 5/11/79 All fuel facilities, Fuel Assemblies research reactors, and power reactors with an Operating Licensee (OL) or a Construction Permit (CP) 79-13 Indication of Low Water 5/29/79 All Holders of Operating Level in the Oyster Creek License (OL) or Reactor Construction Permit (CP) 79-14 NUC Position of Electrical 6/11/79 All Power Reactor Cable Support Systems facilities with a Construction Permit (CP) and applicants 79-15 Deficient Procedures 6/7/79 All Holders of Reactor Operating Licenses (OLs) and Construction Permits (cps) Enclosure Page 2 of 2 410 155}}