ML19225B103
| ML19225B103 | |
| Person / Time | |
|---|---|
| Site: | 05000360 |
| Issue date: | 06/22/1979 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Selby J Battelle Memorial Institute, PACIFIC NORTHWEST NATION |
| References | |
| NUDOCS 7907230520 | |
| Download: ML19225B103 (1) | |
Text
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SUIT E 202, WALNUT CRE E K PL AZ A 4%
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WALNUT CREEK, CALIF ORNI A 9450G ls-June 22,1979 Docket
- a. 50-360 Battelle flemorial Institute Pacific I!orthwest Laboratories Battelle Blvd.
Richland, WA 99352 Attention: lle. J. fl. Selby, Molager Radiation Standards and Engin:ering Gen tlemen:
The enclosed Information flotice No. 79-16 i:. forwarded to you for informa tion.
tic pecific action is requested and no written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely,
[$ 0, 1dl. ---
R. H. Engelken Director
Enclosures:
1.
IE Information ilotice 79-16 2.
List of IE Information flotices Issued in 1979 Q-410 099 99q230 G A
s UNITED STATES NUCLEAR REGULATORY CO., MISSION OFFICE OF IflSPECTION AND EriFORCEMErlT WASHINGT0fi, D.C.
20555 June 22, 1979 IE Information flotice fio. 79-16 fiUCLEAR If!CIDENT AT THREE MILE ISLAND Description of Circumstances:
On March 28, 1979, the Three tiile Island Nuclear Power Plant, Unit 2 experienced core damage which resulted from e 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 Inspectinn and Enforcenent Bulletins (IEBs) 79-05,79-05A and 79-05B to ir. form 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 water reactors and pres-surized water reactors supplied by vendors other than Babcock and Wilcox.
fio specific action or written response to this Information flotice is required.
If you desire additional information regarding this matter, contact the Director of the appropriate f;RC Regional Office.
Enclosures:
1.
IE Bulletin fio. 79-05 with Enclosure 2.
IE Bulletin fio,79-05A with Enclosure 1 901 03 01 8 3.
IE Bulletin fio.79-05B with Enclosure 410 100
IE Information Notice No. 79-16 June 22, 1979 Page 1 of 2 LISTING OF IE INFORMATI0b NOTICES ISSUED IN 1979 Information Subject Date Issued To Notice No.
Issued 79-01 Bergen-Paterson Hydraulic 2/2/79 All power reactor Shock and Sway Arrestor facilities with an OL or a CP 79-02 Attempted Extortion -
2/2/79 All Fuel Facilities Low Enriched Uraniua 79-03 Limitorque Valve Geared 2/9/79 All power reactor Limit Switch Lubricant facilities with an OL or a CP 79-04 Degradatien of 2/16/79 All power reactor Engineered facliities with an Safety Features OL or a CP 79-05 Use of Improper Materials 3/21/79 All power reactor in Safety-Related Components facilities with an OL or CP 79-06 Stress Analysis of 3/23/79 All Holders of Safety-Related Piping Reactor OL or CP 79-07 Rupture of Radwaste Tanks 3/26/79 All power reactor facilities with an OL or CP 79-08 Interconnection of 3/28/79 All power reactor Contaminated Systems with facilities with an Service Air Systems Used OL and Pu Processing As the Source of Breathing fuel facilities Air 79-09 Spill of Radioactively 3/30/79 All power reactor Contaminated Resin facilities with an OL 79-10 Nonconforming Pipe 4/16/79 All power reactor Support Struts facilities with a CP 79-11 Lower Reactor Vessel Head 5/7/79 All holders of Reactor Insulation Support Problem OLs and cps 410 101
IE Information Notice No. 79-16 June 22, 1979 Page 2 of 2 LISTING OF IE IfiFORf1ATION fiOTICES ISSUED Ifi 1979 Information Subject Date Issued To flotice f:o.
Issued 79-12 Attempted Damage to flew 5/11/79 All Fuel Facilities Fuel Assemblies Research Reactors, and Power Reactors with an OL or CP 79-13 Indication of Low tlater 5/29/79 All fiolders of Reactor Level in the Oyster Creek OLs and cps Reactor 79-14 fiRC Position of Electrical 6/11/79 All Power Reactor Cable Support Systeus Facilities with a CP 79-15 Deficient Procedures C/7/79 All holders of Reactor OLs and cps 410
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UlITED STATES l
f1VCLEAR REGULATORY COMMISSION f
0FFICE OF INSPECTION AND ENFORCEMEf!T i
WASHINGT01, D.C.
20555 April 1,1979 IE Bulletin f;o. 79-05 f;UCLEAR IflCIDENT AT THREE f1ILE ISLAND Description of Circunstances-On March 28, 1979 the Three flile Isiarid Nuclear Power Plant, Unit 2 l'
exper ienced core damage which resulted from a series of events which were initiated by a loss of feedwater transient.
Several aspects of the l
incident may have general applicability in addition to apparent generic j
applicability at operating Babccck and Milcox 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 f
sequences in the Three Mile Island event, some of the following may have contributed.)
For all Labcock and Wilcox 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 flRC staff of a postulated severe feedwater transient related to Babcock and Wilcox PWRs as described in Enclosure 2.
These 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 deternine uhether any'significant deviations from expected performance occurred.
If any significant deviations are found, provide the details and an analysis of the significance and any corrective actions taken.
This naterial m y be identificci by reference if previously subnitted to the fiRC.
3.
Review the actions required by your operating procedures for coping with transients.
The itcms that should be addressed include:
(
410 103
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,, ::- ~ ;
IE Bulletin No. 79-05 April 1, 1979 l
Page 2 of 3
[m
'l a.
Recognition of the possibility of forming voids in the primary coolant system large enough to comprcoise the core cooling capability.
3 Y
b.
Operator action required to prevent the formation of such voids.
.J c.
Operator action required to ensure continued core cooling in the event that such voids are forned.
4.
Review the actions requested by the operating procedures and the training instructions to assure that operators do not override automatic actions of engineered safety features without sufficient cause for i
I doing so.
5.
Peview all safety related valve positions and positioning require-nents 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 naintenance and testing, to assure that such valves are returned g
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 to assure 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 systens and indicate:
a.
Whether interlocks exist to prevent transfer when high radiation indication exists and, l
b.
Whether such systems are isolated by the containment isolation signal.
J 7.
Review your prompt reporting procedures for NRC notification to assure very early notification of serious events.
The detailed results of these reviews shall be submitted within ten (10) days of the receipt of this Bulletin.
24.
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410 104
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IE Bulletin No. 79-05 April 1, 1979 L
Page 3 of 3 L
f M
Reports should b' submitted t.: the Director of the appropriate NRC Regional Office anc a copy should be forwarded to the NRC Office of Inspection and Enforcenent, Division of Reactor Construction Inspection, Washington, D.C.
20555.
For all other operating reactor or reactors under construction, this Bulletin is for information purposes and no report is requested.
Approved by GAO, B180225 (R0072); clearance expires 7-31-80.
Approval vias given under a blanket clearance specifically for identified generic problems.
Enclosures.
1.
Preliminary Notifications - Three Mile Island PN0s 67, 67A, B, C, D, E, F and G g
2.
Evaluation of Feedwater Transients with Attachment 3.
List of IE Bulletins j
issued in the past 12 nonths t
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6 9
C AIC 105
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EtiCLOSURE 2 Lvaluation of reedwater Transient 7
l2 A loss of offsite po',;er occurred at Davis-Besse on ibvember 29, 1977, i
which resulted in shrinkage of the prinary coolant volume to the degree that pressur;zer level indication was lost.
A reconmendation to convey this information to certain hearing boards resulted in the attached discussion and evaluation of the event.
This discussion includes a
,,a review of a loss of feedwater safety arialysis assuming forced flow, tihich predicts dispersed primary systen voiding, but no loss of core cooling. During the Three flile Island event, however, the forced flow appears to have been terminated during the transient.
/ttachment Discussion and Evaluation of Davis-Besse Transient t
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EXCEnpr FROM ME".ORAUDUM FNTITLED "CO.WEYINO NEW IN' ORMATION TO LICEMSI!:G i
E0ARDs - davis-BEssE U:!ITS 2 6 3 AND MIDMND UNITS 1 5 2",
DATED JANUARY 8, 1979, FRO:I J.S. CREsa-m TO J. I'. STREETER tt'm r~
3.
Inspection and Enforcement Repor t 50-346/78-06 doctraented that
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pressurizer levei had gone off ocale f or approximately five
- j),
tainuten during the Uoveabbr 29, 1977 loss of offsite power event, There are some indications that other E6U plants nay have prob-j le.as raintaining pressurizer 1cvel indications during transients.
in addition, under certain conditions cuch as loss of feedwater at 100% psuer with the reactor coolant pumps running the prea-y surizer raay void completely.
A special analysis has been per-
' y forned concerning this event.
This analysis is attached as.
Because of pressur er 1cvel caintenance prob-1 ens the sizing of the pressurie.
ay require further revicu.
Also noted during the event van rho f.w r that Ten M '*:n u ff scale (less than 5200F).
In addition, it was noted that the nakeup fica nonitoring is linited to less than 160 gpa and that nakeup flew ray he substantially nreater thnu _this value.
This in form.ation sheuld be e>.aained ir light of the requirc-rients of COC 13.
,DISCL,S }IOS AND EVALU.*. TION
((
1he event at Davis Eesse which resulted in loss of pressuriner level indication has been reviewed by URR and the conclusiob was rcached that no unreviewed snfety question existed.
The pressurizer, together with the reat_ tor coolant makeup systen, io desi;;ned to nulntain the primary system pressure and water level within their operational limits only during nor=M operating conditions.
Cooldown transients, such as loss of of f site power and loss of feed-water, so~et imes result in primary pressure and volume changes that are beyond the ability of this system to control. The analyses of and experience with such transients show, however, that they can be i
sustained without compromising the cafety of the reactor.
- ihe principal 3
concern caused by nuch transients is t.ut they :.ight cause voiding in the primary coolant systen that would lead to loss of ability to ad_
. j, quately cool the reactor core. The safety evalention of the loss of of fsite power transient shous that, though level' indication is lost,
i
, l y
coma water recains in the pressurizer and the pressure doun not decrease
.s below about 1600 psi.
In order for voiding to occur, the pressure must decrease below the saturation pressure corresponding to the cystem tegerature.
1600 psi is the saturation pressure corresponding to 605 F, which is also the maxinua ellowable core outlet temperature.
Voiding in the primary-system (excepting the pressurizer) in precipagd
- 1 in thin case, since pres,ure does not decrease to saturation.
Attachment (Enclosure,2) l (Page 1 of 2) 410 107
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Sectic, 3.
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The cafety analyais for core severe coaldown transiento, cuch as the i
loas of feedvater event, indicaten that the water volume could decrease
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to less than the. cyaten voltr..e exclualve of the preasurizer.
During such an event, the captying of the preanuriner would be followed by a pressure reduction below the caturation point and the forcation of naall voida throughout rauch of the primary systna.
This would not result in the loss of core cooling becauae the voids would be dispernad av r a large volume and forced flou would prevent them f rom coalescing i
sufficiently to prevent core coolin",.
The high pre.saure coolant injection pumps are started autcoatically when t he prinary pressure
.I.
decreases below 1600 psi.
'ib refore, any pressura reductior which Lc
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cufficient to allow voiding util also result in water inj ectica uhtch will rapidly rentore the primar'f water to norral levels.
i
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For these reasona, we believe that the inability of the pressuricer
{
and normal coolant nakuup cystua to control some transients does not i
provide a bani.2 for requirin?. tore capacity in thc3+- systems.
Ceneral. Design Criterion 13 of Appendi:: A to 10 CFR 50 requires inntrementa'-lun to nonitor variables over their anti.cipated range'.
for " ant.lcip ited operational occurrences".
Such occurrences are ep?cifically defined to includa loss el all o f fsite power.
'ine fact that Y cold goes of f scale at 520"F in not considored to be.. deviation i
/
f rma this requiretent bntause this todicator 3n backed up by ef de 5
r:mne te,perature indication that cytends ta a 30w 11mit of 50 P.
Neither do vu consider the uakuup flow roaitoring to deviate cince the ar.ount of makeup flou Jn extens of 160 gpm does not appear to he a significant factor in the cource of thes" occurrences.
The loss of pressurin r v.ater Jevel indication could be considered to deviate from GDC 13, because this '.evel Indication providea the principal t;eans of determining tha primary 'oolant inventory.
However, provision of a level indication that sivuld..over all anticipated occurrences may not be practical. As discussed ebeve,. the loss of feedwater event can lead to a co ntotary condition vheref n no neaningful level exists, because the entire primary synten contains a ntnaa water mixture.
It chould be noted that the introduction to Appendix A (last paragraph) recognizes that fulfill =ent of come of the critoria may not always bu nppropriate.
This introduction also state 9 that departures f ron the Criteria uust be identified and juntf fled.
The discusaton of GDC 13 in the Davis Eenne FSAR liuto the water level instrenentation, but docq not cention the possibility of loss of water level indication
,1 during transients.
This apparent ominston in the saf ety analysi'.
vill be subj ected to further review.
(page 2 ol'2) 410 108 h
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UNITED STATES
/
NUCLEAR PJGULATORY COMPISSION
\\
OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 APRIL 5, 1979 IE Bulletin 79-05A NUCLEAR INCIDEfiT AT THREE MILE ISLAND - SUPPLEMENT
[r1 Description of Circumstances:
Preliminary information received by the NRC since issuance of IE Bulletin 79-05 on April 1,1979 has identified six potential human, I.
design and mechanical failures which resulted in the core damage and radiation releases at the Three Mile Island Unit 2 nuclear plant.
The I
information and actions in this supplement clarify and extend the original Bulletin and transmit a preliminary chronology of the TMI accident through the first 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (Enclosure 1).
1.
At the tire of the initiating event, loss of feeduater, both of the auxiliary feedwater trains were valved out of service.
2.
The pressurizer electromatic relief valve, v'hich opened during the initial pressure surge, failed to close when the pressure j
decreased below the actuation level.
/
3.
Following rapid depressurizatiori of the pressurizer, the pressurizer i
level indication may 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 i ot isolate on high 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 some of it overflowed to the fluor.
Outgassing from this water and discharge through the cuxiliary building ventilation system and filters was the principal source of the offsite release of radioactive noble 3-x gases.
5.
Subsequently, the high pressure injection system was intermittently operated attempting to control primary coolant inveritory losses through the electromatic relief valve, apparently based on pressurizer level indication.
Due to the presence of steam and/or noncondensible voids elsewhere in the reactor coolant system, this led to a further reduction in prinary coolant inventory.
1 410 109
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IE Bulletin 79-05A April 5,1979 Page 2 of 5 l
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6.
Tripping of rcactor coolant purrps during the course of the transient, to protect against pump damage due to pump vibration, led to fuel damage rince voids in the reactor coolant system prevented natural circulaticn.
Actions To Be Taken by Licensees:
~
i.
FOR ALL Babcock and Wilcox pressurized water reactor facilities with an operating license (the actions specified below replace those specified e
in IE Bulletin 79-05):
[
1.
(This iten clarifies and expands upon item 1. of IE Bulletin 79-05.)
i In addition to the review of circunstances described in Enclosure 1 i
of IE Bulletin 79-05, review the enclosed prelininary chronology of l
the TMI-2 3/28/79 accident.
This review should be directed toward understanding the sequence of event: to ensure against such an accident at your facility (ies).
2.
(This item clarifies and expands upon item 2. of IE Bulletin 79-05.)
l
\\
Review any transients similar to the Davis flesse event (Enclosure 2 of IE Bulletin 79-05) and any cthers which contain sinilar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (ies).
If any significant devi_.tions fron 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 iten 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.
a.
Recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling capability, especially natural circulation capability, b.
Operator action required to prevent the formation of such voids.
ua c.
Operator action required to enhance core cooling in the event f
such voids are forced.
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410 110
'. G.. J :M
r IE Bulletin 79-05A April 5,1979 page 3 of 5 t
4.
(This item clarifies and expands upon item 4. of IE Bulletin 79-05.)
7 Review th' actions directed by the operating procedures and training 1
! a instructions to ensure that:
4 Operators do not override automatic actions of engineered a.
I safety features.
b.
Operating procedures currently, or are revised to, specify 1
that if the high pressure injection (HFI) system has been automatically actuated because of low pressure condition, J
it rust remain in operation until either:
(1) Both 10.1 pressure injer tion (LPI) pumps are in operation I
and flowing at a rate in excess of 1000 gpm each and the l
situation has been strble for 20 ninutes, or (2) The HPI system has been in operation for 20 ninutes, and all hot and cold leg temperatures are 9t least 50 degrees below the saturatio.i temperature for the existing RCS pressure.
If 50 degree subcooling cannot be naintained after HPI cutof f, 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 operating.
d.
Operators are provided additional information and instructions to not rely upon pressurizer level indication alone, but to also exani Te pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water inventory in the reactor primary systen.
5.
(This item revises item 5. of IE Bulletin 79-05.)
Verify that emergency feeduater valves are in the open position in accordance with item 8 below. Also, review all safety-related valve positions and positioning requirements to assure that valves are positioned (open or closed) in a manner to ensure the proper operation of engineered safety features.
Also review related procedures, such as those for maintenance and testing, to ensure that such valves are returned to their correct positions following necessary manipulations.
410 111
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- E Bulletin 79-05A April 5,1979 Page 4 of 5 7
6.
Review the containrent isolation initiation design and procedures, and prepare and irplerent all chances necessary to cause containment j
isolation of all lines whose isolation does not degrade core cooling
('.
capability upon automatic initiation of safety injection.
7.
For nanual valves or nanually-operated motor-driven valves which could defeat or compromise the flow of auxiliary feed'.iater to the j
stean generators, prepare and ir plement procedures which:
4 i
a.
require that such valves be locked in their correct position, or b.
require other sinilar positivc position controls.
8.
Prepare and ir plement irmediately procedures which assure that two l
independent steam generator auxiliary feedwater. low paths, each with 100% flow capacity, are operable at any tim when heat removal from the primary systen is through the steam generators.
When two inde-pendent 100:' captcity flow paths are not available, the capacity shall be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant shall be placed in a cooling mode which does not rely on stean generators for cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
\\
Wnen at least one 100~ capacity flow path is not available, the reactor shall be made subtritical within one hour and the facility placed in a shutdown cooling node which does not rely on steam generators for cooling within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or at the maximun safe shutdown rate.
9.
(This iten revises item 6 of IE Bulletin 79-05.)
Review your operating modes and procedures for all systems designed to transfer pctentially radioactive gases and liquids out of the primary containment to assure that undesired pumping of radioact e liquids and gases will not occur inadvertently.
L'e n
In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.
List all such systens and indicate:
a.
Whether interlocks exist to prevent transfer when hign radiation indication exists, and w
b.
Whether such systems are isolated by the containment isolation signal.
410 112
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IE Bulletin 79-05A April 5,1979 Page 5 of 5 ll l
7m 10.
Review and modify as necessary your mairitenance and test procedures to ensure that they require:
j Verification, by inspection, of the operability of redundant a.
safety-related systems prior to the removal of any safety-related system from service.
b.
Verification of the operability of all safety-related systems d.
when they are returned to service following maintenance or testing.
l A means of notifyir.g involved reactor operating personnel f
c.
whenever a safety-related system is removed from and returned i
to service.
l
- 11. All operating and maintenance personnel should be nade aware of the f
extreme seriousness and consequences of the simultaneous blocking
.l of both auxiliary feedwater trains at the Three tlile Island Unit 2
~
plant and other actions taken during the early phases of the accident.
12.
Review your prompt reporting procedures for I!RC notification to assure very early notification of serious events.
Foi Babcock and Wilcox pressurized water reactor facilities with an operating license, respond to Iterns 1, 2, 3, 4.a and 5 by April 11, g
1979.
Since these items are substantially the some 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 f!RC Of fice 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, q
Enclosures:
f 1.
Preliminary Chronology of Tl11-2 3/38/79 Accident Until Core Cooling Restored.
2.
List of IE Bulletins issued in last 12 months.
410 ii3 3,
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IE Bulletin 79-05A April 5,1979 PRELIMIi'ARY I
CHRONOLOGY OF TMI-2 3/28/79 ACCIDEilT UNTIL CORE C00LIi;G RESTORED TIf!E (Approximate)
EVEflT about 4 AM Loss of Condensate Pump 1.
(t = 0)
Loss of Feedwater Turbine Trip t = 3-6 sec.
Electromatic relief valve opens (2255 psi) f to relieve pressure in RCS j
l t = 9-12 sec.
Reactor trip on high RCS pressure (2355 psi) t = 12-15 sec.
RCS pressure decays to 2205 psi r
(relief valve should have closed) t = 15 sec.
RCS hot leg temperature peaks at 611 degrees F, 2147 psi (450 psi over s
saturation) t = 30 sec.
All three auxiliary feedwater punps running at pressure (Pumps %A and 2B started at turbine trip). Ho 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 l
very low - drying out over next couple of minutes.
r t = 2 nin.
ECCS initiation (HPI) at 1600 psi t = 4 - 11 min.
Pressurizer level off scale - high - one l
HPI pump manually tripped at about 4 min.
30 sec. Second pump tripped at about 10 min. 30 sec.
O
^
t = 6 min.
RCS flashes as pressure bottoms out at 1350 psig (Hot leg temperature of 584 degrees F) t = 7 min., 30 sec.
Reactor building sump pump came on.
A10 114 c:s
t '
TIME EVEllT t = 8 nin.
Auxiliary feedwater flow is initiated
[
by opening closed valves j
y t = 8 min. 18 sec.
Steam Generator B pressure reached minimum t = 8 min. 21 sec.
Steam Generator A pressure starts to recover i
t = 11 min.
Pressurizer level indication comes back on scale and decreases g
(.
t = 11-12 min.
Makeup Pump (ECCS HPI flow) restarted by l
operators t = 15 min.
RC Drain / Quench Tank rupture disk blows at l
190 psig (setpoint 200 psig) due to continued E
discharge of electromatic relief valve
}
t = 20 - 60 min.
System parameters stabiliznd in saturated condition at about 1015 psig and about 550 degrees F.
P t = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 15 min.
Operator trips RC punps in loop B l
t I,
t = 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, 40 min.
Operator trips RC pumps in Loop A t = 1-3/4 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> CORE BEGII;S HEAT UP TRAilSIEl!T - Hot leg temperature begins to rise to CPO degrees F (off scale within 14 minutes) and cold leg temperature drops to 150 degrees F.
(HPI water) t = 2.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Electromatic relief valve isolated by operator af ter S.G.-B isolated to prevent leakage t = 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> RCS pressure increases to 2150 psi and (T
electromatic relief valve opened 4
t = 3.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> RC drain tank pressure spike of 5 psig i.
t = 3.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RC drain tank pressure spike of 11 psi -
RCS pressure 1750; containment pressure increases from 1 to 3 psig t = 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Peak contairrent pressure of 4.5 psig 1_
t = 5 - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> RCS pressure increased from 1250 psi to
/
to 2100 psi S.,
col unn 410 115
\\ *
-.....;4 n:.::
i i TIliE EVENT t = 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Operator opens electromatic relief valve to depressurize RCS to attempt initiation of k'
RHR at 400 psi i
t = 8 - 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> RCS pressure decreases to about 500 psi Core Flood Tanks partially discharga L?
t = 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 28 psig containment pressure spike, containment sprays initiated and stopped after 500 gal. of d
ita0H injected (about 2 ninutes of operation)
F i
t t = 13.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Electromatic relief valve closed to repressurize RCS, collapse voids, and start RC pump i
t = 13.5 - 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> RCS pressure increased from 650 psi to 2300 psi j
t = 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> RC pump in Loop A started, hot leg temperature decreases to 560 degrees F, and cold leg teirperature increases to 400 degrees F.
i indicating flow through steam generator j
i Thereafter S/G "A" stecning to condeasor
/
Condensor vacuun re-es tablished
(
RCS cooled to about 280 degrees F.,
1000 psi flor (4/4)
High radiaticn in containment All core thermocouples less than 460 degrees F.
Using pressurizer vent valve with small nakeup flow Slow cooldown RB pressure negative k:-
a
. _]
]
O A
-A (s
410 116
~
S. f.
UtilTED STATES
(
IluCLEAR REGULATORY COMMISSIO:t OFFICE OF IfiSPECTI0il Ai!D Ef1FORCEME!!T WASHIfiGTO:1, DC 20555 Apri l 21, 1979 IE Bulletin 79-05B fiUCLEAR It!CIDEi!T AT THREE MILE ISLA!!D - SUPPLEMEilT Description of Circumstances:
Continued i;RC evaluation of the nuclear incident at Three Mile Island Unit 2 has identified measures in addition to those discussed in IE Bulletin 79-05 and 79-05A which should be acted upon by licensees with reactors designed by B&W.
As discussed in Item 4.c. of Actions to be taken by Licensees in IEB 79-05A.
the preferred mode of core cooling following a transient or accident is to pro-vide forced flow using reactor coolant pumps.
It appears that natural circulation was not successfully achieved upon securing the reactor coolant pumps during the first two hours of the Three Mile Island (TMI) fio. 2 incident of March 28, 1979.
Initiation of natural circulation was inhibited by significant coolant voids, possibly aggravated by release of non-condensible gases, in the primary coolant system.
To avoid this potential for interference with natural circulation, the operator should ensure that the
[
primary system is subcooled, and renains subcooled, before any attempt is made to establish natural circulation.
i flautural circulation in Babcock and Wilcox reactor systems is enhanced by maintaining a relatively high water level on the secondary side of the once through steam generators (OTSG).
It is also prorioted by injection of auxiliary feedviater at the upper nozzles in the OTSGs.
The integrated Control System automatically sets the OTSG level setpoint to 503 on the operating range when all reactor coolant pumps (RCP) are secured.
However, in unusual or abnormal situations, manual actions by the operator to increase steam generator level will enchance natural circulation capabilit.,
n anticipation of a possible of operation of the reactor coolant purps.
As sta ted previously, forced flov-of primary coolant through the core is preferred to natural circulation.
r means of reducing the possibility of void formation in the reactor coolant
. tem are:
A.
Minimize the operation of the Power Operated Relief Valve (PORV) on the pressurizer and thereby reduce the possibility of pressure reduction by a blowdown through a PORV that was stuck open.
k!0 lll
IE Bulletin 79-05B 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.
This bulletin addresses, among "ther things, the means to achieve these obj ec tives.
Actions To Be Taken by Licensees:
For all Babcock and Uilcox pressurized water reactor facilities with an operating license:
(Underlined sentences are modifications to, and supersede, IEB-79-05A).
1.
Develop procedures and train operation personnel on methods of establishing and maintaining natural circulation.
The procedures and training must include means of r:onitoring heat removal efficiency by available plant i ns trumenta tion.
The procedures must also contain a method of assuring g
that the primary coolant systen is subcooled by at least 50 F before natural circulation is initiated.
In the event that these instructions incorporate anticipatory filling of the OTSG prior to securing the reactor coolant pumps, a detailed analysis should be done to provide guidance as to the expected system response.
The instructions should include the following precautions:
maintain pressurizer level sufficient to prevent loss of level a.
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, and maintain pressure - temperature envelope within Appendix G limits c.
for vessel integrity.
Procedures and training shall also be provided to uaintain core cooling in the event both main feedwater and auxiliary feedwater are lost while in the natural circulation core cooling mode.
2.
Modify the actions required in Iten 4a and 4b of IE Bulletin 79-05A to take into account vessel integrity considerations.
"4.
Review the action directed by the operating procedures and training instructions to ensure that:
Operators do not override automatic actions of engineered a.
safety features, unlejs__ aLntinued 02 era tion of eng_ineered
(~
410 118
IE Bulletin 79-05B April 21,1979 Page 3 of 4 For safety _ fea tures will result in unsafe plant condi tions.
example, i f con ti nued opera tion of eng ineered saf ety_fe_atur;es_
would threaten reactor vessel integrity then the HPI should be secured (as noted in R2) below).
Operating procedures currently, or are revised to, specify that b.
if the high pressure 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 operation 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 belou the saturatior, temperature for the existing RCS If 50 degrees subcooling cannot be maintained pressure.
af ter HPI cutof f, the HP1 shall be reactivated.
The_de, gree of subcooling_ beyond 50_deg_rees F and the length of time HPI is in opera tion shall be limi ted by_ the_ pressure /
temperature considerations for the vessel integrity."
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 antici-pated transients.
This analysis shall include consideration of a modifi-cation of the high pressure scram setpoint and the POVR opening setpoint such that reactor scram will preclude opening of the PORV for the spec-trum of anticipated transients discussed by BM! in Enclosure 1.
Changes developed by this analysis shall not result in increased frequency of pressurizer safety valve operation for these anticipated transients.
4.
Provide procedures and training to operating personnel for a prompt manual trip of the reactor for transients that result in a pressure increase in the reactor coolant systen. These transients include:
a.
loss of main feedwater b.
turbine trip main Steam Isolation Valve closure c.
d.
lou OTSG level f.
low pressurizer level.
410 119
IE Bulletin 79-05B April 21, 1979 Page 4 of 4
(
5.
Provide for NRC approval a design review and schedule for implementation of a safety grade automatic anticipatory reactor scram for loss of feed-water, turbine trip, or significant reduction in steam generator level.
6.
The actions required in item 12 of IE Bulletin 79-05A are modified as follows:
Review your prompt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condi tion of operation.
Further, at that time an_ open continuous comunication channel shall be established and maintai rd with NRC.
7.
Propose changes, as required, to those technical specifications which must i.e modified as a resul t of your implementing the above i tems.
Response schedule for B&W designed facilities:
a.
For Itens 1, 2, 4 and 6, all facilities with an operating license respond within 14 days of receipt of this Bulletin.
b.
For Item 3, all facilities currently operating, respond within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
All facilities with an operating license, not currently opera ting, respond before resuming operations.
/
c.
For Items 5 and 7, all facilities with an operating license respond
\\
in 30 days.
Reports should be submitted to the Director of the appropriate NRC Regional Of fice and a copy should be forwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D.C. 20555.
For all other power reactors witn an operating license or construction penait, this Bulletin is for information purposes and no written response is required.
Approved by GAO, B180225 (R0072); clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for identified generic probleus.
Extract of B&W Communication Received by NRC t10 iz0
EXTRACT OF-B&W C0mVNICATION - RECEIVED BY fiRC
!?fIROCUCT-103 Paga 1 of 4 E CONTIfl0ING REVIEM OF THE SEQUENCE OF EYENTS LEADITlG TO TIIE IrlCIDErlT AT i-2 Otl NARQ128,1979 SHfAIS TIIAT ACTION CAri UE TAKEN TO PH0 VIDE ASSURAUCE TMT THE PILOT-OPERATED RELIEF VALVE (PORY) IDUNTE9 OTI TiiE PRESSURIZER OF BN PLANTS HILL NOT DE ACTUATED BY ANTICIPATED TRAtl5IEff TS HUICH llAVE OCCURRED OR M7E A SIGNIFICA?TT PROBABILITY OF GCCURRING IN THE5E PLAtlIS.
Tills ACTIC?1 r;'J5T l;UT DEGRADE THE SAFET7 0F THE AFFECIED PLANTS WITil RESPECT TO TirEIR RESPO!isE TO F,0RMAL, UPSET 0 4 ACCIDENT CONDITIONS NOR LEAD TO UNHEVIEWED SAFETY CONCERNS.
THE ANTICIPATED Tf/11SIEfITS OF CONCERM ARE:
i LOSS 6E EXTERML ELECTRICAL LOAD 2.
TUP.3IME TRIP 3.
LOSS OF MI?! FEEDWATER 0
LOSS OF CONDEILSER VACUUH 5.
INADVERTE?fT CLOSURE OF ?>51N STEAM ISOLATION VALVES (MSIV).
A liLWRER OF ALTEPJIATIVES UERE CONSIDEREO IN DEVELOPING TifE ACTION SELG! INCLUDING:
L ET:5TP.ICTE0 REACTOR PCWER TO A VALUE WiICH WOULO ASSURE NO ACruATIO:10F THE PORY.
THE. REACTOR PROTECTION SYSTEH, DESIGN PRESSURE AND PORY SET POINTS REtMINED AT THEIR CURRENT vat.UES.
.2.
LOMERING THE IIIGH PRES 5URE REACTOR TRIP SETPOINT TO A VALUE Mi!CH WU ASSUV NO ACTUATION OF THE PORY.
THE DESIC1! PRESSURE OF THE iiEACf0R TJiD T;-iE S EtFultti FOR FORY ACTUATIoll REHAINED AT TllEIR CURREflT VALUES.
' [0MERING T!fE HIGH PRESSURE REAcr0R TRIP SETPOINT Anu ADJUSTIrl
(
(
OPERATIT?G PRES $URE (AND TEXPEPATURE) 0F Tl!E REACTOR TO ASSURE fl0 ACTI!ATIU3 AND TO PROVIDE ADEQUATE MARGIN TO ACCom0DATE VA9IATIONS G7ERTiTIMG PRESSURE.
THE SETPOINT FOR PURV ACTUA TION REPAINED Af [TS CURRENT VALUE.,
THIS ALTEFJIATIVE HOULD REQUCE fiET FLECTRICAL OUTPUT.
S.
ADiUSTING 1HE HIGi PRESSURE TRIP ANU THE FORY SETPOINIS TO ASSURE PG?.'t ACTUATIO?) FOR THE CLASS OF IUiTICIPATED EVENTS OT CONCERN THE UEStGN PRESSURE OF THE PIACTOR REHAINED AT ITs CURREtif VALUE.
"M AT'ALYSIS OF THE ItPACT OF THESE VARIOUS ALTERNATIVES AND T TO ASSU21 rig THAT THE PORY WILL fl0T ACTUATE FOR Tile CLASS OF ANTICIPATED TRA OF CONCEP11 HAS BEEif CCMPLETED.
THE RESULTS SHOW.TilAT:
LOMERING THE HIGi PRESSUTiE REACTOR TRIP SETPOIfff FROM 2255 PSIG TO 2200 PSIG AUD PAISING Tt!E SETPOIj'TT FOR THE PILOT OPERATED RELIEF VALVE FRC;t 2E55 PEIG TO_ Pf,50 PSJG Pr.OVIDES THE REQUIRED A550RA'!CE.
THIS AfrION HrtS THE FURTllER AUVAtlTAGES OF:
.t
.._ i $ a
{
_ Q _.
l
}
EXTPACT OF B&W CO:HUNICATION - RECEIVED BY llRC Page 2 of 4 4/20/79 t
I L
11EDUCIt?G THE PR03AUILIT7 0F POR'/ AND A3HE CODE PRESSURIZER SAFETY VAlyc.
. i t
ACIlt"* TION FOR OTHER UICREASItXi PRESSURE TRANSIENTS.
j
~
2.
PRESERVING PRESSURE RELIEF CAPACITY.FOR ALL HICH PRESSURE TRANSIEllTS.
5L1HIFf571NG THE POSSIBlLITY DF IflTR000CING UNREVIEMED SAFETY CONCERUS.
4.
EEUUCnlG THE TIFE AT iEICl! TrlE STENT SYSTEtt ilEAT SIHr. WOULO BE LOST In '
THE E'/EST EMERGENCY FEEDWATER FLG'.4 WERE DELAYED.
h StWRARY OF Tile IMPACr OF Tile PROPOSED SETPOItrT CHANGES Oil ALL ANTICIPATE TR/UISIEffTS IS GIVEN IN TABLE T.
BUI PLAlUS ARE CURRErTTL1 CAPA9tE OF RthfBACK TO.15.#, OF FULL POWER UPO?! LOSS OF LOAD OR TRIP OF THE TURBINE.
TllIS CAPAUILITY REQUIRES ACTUATIOll OF Tile PII.0T-CPERATZO RELIEF VALYES.
TIIE CAPABILITY INCREASES TllE RELI ADILITY OF POWER SUPPL't TO THE SYSTEM DY PETURTHMG.711E UNITS TO POWER GEtlERATION tiORE QUICTl.Y AFTEP. THESE TPMISIENTS.
Tr!E ACTI0ll PROPOSED ABOVE WILL REQUIRE THAT Tile REACTOR '3E TRIPPED FOR DIESE EVEffiS;
[
't!OIE:
(
The effect o-f changing the reactor coolant. system pressure trip setpoint upon peak pressurizer pressure is typif f ed by the attached figure 1. which was developed by.
B&W for a loss of feedwater transient.
0 9
d e
e 6
3 g
.h.
s 410 122.
~.
84.
O
TABLE 'l Stht$RY OF PROTECTICM ACAltt3r PORY AcrUATIOri PRO'1IDED BY PROPOSED SETP0ltiT CliNIGES FOR ALL
(-
A.TTICIPATED TRMISIEUT5 EXTRACT OF B&W;CCF3UNECATION RECEIVED BY FRC 4/20/72 p
i.
NITICIPATED TRA?iSIEI!T5 MHIQI HAVE OCCURRED AT B2M PildlTS NID 1tilIOl W bkNAl.L7 ACTIVATE FOR' Y AT Tile _CURRFtfT SETFOIrfr (2255 PSIG):
'~
A.
TUR31E TRIP A
Less OF CTEEHAL ELECTRICAL LOAO C.
l_C55 0F NAIN FEEDMATEtt D.
LCSS OF CONDENSER VACULW 2.
IMADVERTEffT CLOSURE OF PGIV 2.
fJsTICIPATED TRA%IEriTS m1101 HAVE OCCURRED AT D3M PLANTS Nip WilICH L'0ULO t' oft,'.RL7 ACTUATE PORV AT Til$ PROPG5EO SE rPOIrlT (2450 PSIG):
COM A
TJirICIPATED TRNfSIErfr51n1I01 llAVE ?K)T OCCURRED AT B54 PLMITS (LCW PR03ASILIT( EVENTS) AND HillCH WOULD r.0RXALLY ACTUATE P0HV AT DIE CURMnT SETFOINT (EE i 'SIG):
A.
SOM CCilTRCL RCD GROUP HITHDRAVALS (MODERATE TO IIIGli REACTIVITY
. lf0RTH'GR0trP5 f;OT OmERWISE PRUTECTED BY HIGH FLU 7. TRIP).
~
8.
MODERATOR DILUTION.
9.
11.HCIPdTED TR!LSSIENTS MIICil llAVE 170T OCCURREO AT B&V PL EVERTS) AND U11IC11 UGULD ACTUATE T1!E PORY AT T1IE PROPO5ED SETPOIrlT
'(2450 P5IG):
(
A.
sow corrTs0t non cR0uF titruLnAlsAts (iircil nEACrIvrTy nonTii r:Or oTitERil: SE PP.0TECTED BY llIGF1 FLUX TRIP),
410 123 I.
c
o s
=
c=.
in Page 4 of 4 i~l EXTPACT OF B&W COMUNICATION - RECEIVED BY NRC
(
4/20/79 i
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~~
for a loss of feedwater transient for expected conditions and various
! -.=.
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