ML19225B099
| ML19225B099 | |
| Person / Time | |
|---|---|
| Site: | 05000112 |
| Issue date: | 06/22/1979 |
| From: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Jischke M OKLAHOMA, UNIV. OF, NORMAN, OK |
| References | |
| NUDOCS 7907230510 | |
| Download: ML19225B099 (1) | |
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NUCLEAR REGULATORY COMMISSION
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A.4 LINGTON, T E X AS 7bO12 g
June 22, 1979 In Reply Refer To:
Docket No. 50-252 The University of New Mexico ATIN:
Dr. G. a. Uhan Reactor Supervisor Chemical & Nuclear Engineering Department Albuquerque, New Mexico 87131 Centicaen:
The enclosed Information Notice No. 79-16 is forwarded to you for in-formation. No sptaific action is requested and no written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, f
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Karl V. Seyfrit Director
Enclosure:
1.
IE Information Notice 79-16 2.
List of IE Inforcation Netitas issued in 1979 410 187 790733,,
UNITED STATES NUCLEAR REGULATORY C0tDlISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C.
20555 IE Information Notice No. 79-16 Date:
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 Bulletins (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 water reactors and pres-surized water reactors supplied by vendors other than Babcock and Wilcox.
No specific action or written response to this Information Notice is required.
If you desire additional information regarding this matter, cor. tact the Director of the app ~ rop ~riate NRC Regional Office.
Enclosures:
1.
IE Bulletin No. 79-05 with Enclosures 2.
IE Bulletin No.79-05A with Enclosures 3.
IE Bulletin No.79-05B 7 997 OS O 4i0 I88
U:lITED STATES
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il0 CLEAR REGULATORY CO:0!ISSIOil 0FFICE OF IflSPECTIO 1 Ai!D EtlFORCEf!Ef4T WASHIt;GT0tl, D.C.
20555 April 1, 1979 st IE Bulletin ?!o. 79-05
%a f!UCLEAR It!CIDEilT AT THREE f1ILE ISLAtlD 0i Description of Circumstances:
b a'l On 11 arch 28, 1979 the Three T4ile Island fluclear Power Plant, Unit 2 5
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 14ile Island event, some of the following may have contributed.)
For all Babcock and.Ui_lcox 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 i;RC staff of a postulated severe feedwater transient related to Babcock and Milcox 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 determine whether any significant deviations from E
expected performance occurred.
If any significant deviations are found, provide the details and an analysis of the significance and any corrective actions taken.
This material may be identified by i
reference if previously submitted to the flRC.
4.d 3.
Review the actions required by your operating procedures for coping with transients.
The items that should be addressed include:
32
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IE Bulletin No. 79-05 April 1, 1979 Page 2 of 3 RecoLnition of the possibility of forming voids in the primary a.
cool c.t system large enough to compromise the core cooling capab ili';y.
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b.
Operator action required to prevent the formation of such voids.
?j 00erator action required to ensure continued core cooling in c.
t.he event that such voids are formed.
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4.
Review the actions requested by the operating procedures and the la trainir.g iristructions 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-nent such as t'.e 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 :to' assure that undesired pumping of radioactiae 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.
(. _
The detailed results of these reviews shall be submitted within ten (10) days of the receipt of this Bulletin.
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IE Bulletin No. 79-05 April 1, 1979 Page 3 of 3 Reports should be submitted to the Director of the appropriate NRC Regional Office and a copy should be torwarded to the NRC Office of Inspection and Enforcement, Division of Reactor Construction Inspection, 20555.
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Washington, D.C.
g For all other operating reactors or reactors under const.. tion, this
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Bulletin is for information purposes and no report is requested.
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Approved by GAO, B180?25 (R0072); clearance expires 7-31-80.
Approval E
was given under a blanket clearance specifically for identified generic problems.
Enclosures:
1.
Preliminary Motifications Three Mile Island -
PMO-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
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5 L
4i0 i91 Page 1 of 3 EVALUATION OF FEED','ATER TRANSIENT Aw A loss of offsite power occurred at Davis-Besse on November 29, 1977, q
[.1 which resulted in shrinkage of the primary coolant volume to the degree that pressurizer level indication was lost.
A recomrendation to convey this information to certain hearing boards resulted in the attached ii discussion and evaluation of the event.
This discussion includes a d
review of a loss of feedwater safety analysis assuming forced flow, 3r 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
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EXCERPT FROM MEMO?J.SDUM ENTITLED " CONVEYING NEU INFORMATION TO LICENSIIC
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BOAFDS - DAVIS-BESSC UNITS 2 6 3 AND MIDLCD UNITS 1&
2", DArdo JANUARY 8, 19 79 FROM J. S. CRE SW.LL TO.T.F.
STREETER.
s 3.
Inspection and Enforcement Report 50-346/78,06 doct:aented that pressurizer level had gone of f scale f or approxbately five ninutes during the Noveab'er 29, 1977 loss of off site power event.
[:9 Tnere are scoe indications that.other Tim plants may have prob-W lens maintaining pressurizer level indications during transients.
g In addition, under certain conditions such as loss of f eeduater q
100% power with the reactor coolant pt:aps running the pres-y at surizer 12ay void completely.
A special analysis has been per-9 forned concerning this event.
This analysis is attached as.
Because of pressuriier level etaintenance prob--
lems the sizing of the pressurizer nay require furchic review.
s Also noted during the event was the fact that Tcold went off-scale (less than 520oF).
In addition, it was noted that the reakaup flea monitoring is 1bited to less chan 160 spa and that nakeup flow nay be substantially greater than this value.
Tnis infomation should be exacined in light cJ. the require-cents of CDC 13 DISCUSSION AND EVALUATION Thie event at Davis Besse which resulted in loss of urdssurizer level indication has been reviewed by URR and the conclu' was reached that no unreviewed safety question existed.
The pressuriser, together with the reactor coolant makeup systen,1, designed to paintain the primary system pressure and water level v. thin their operational limits only durin; normal operating conditions.
Cooldown transients, such as loss of of fsite power nnd loss of feed-water, socetices result in pricary pressure and volume changes that are beyond the ability of this systea to control.
The analyses of and experience with such transients show, however, that they can be sustained ithout compro2ising the saf ety of the recctor.
The principal concern caused by such traasients is that they night cause voiding iri the pricary coolant systea that vould lead to loss of ability to ade-3 quately cool tha reactor core.
The safety evaluction of the loss of
- of f site power transient shows that, though level indication is lost, some unter recains in the pressurizer ar.d the pressure does not decrease belou about 1600 psi.
In order for voiding to occur, the pressure cust
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decrease below the saturation pressure corresponding to the systen 9
tengerature.
1600 psi is the saturation pressure corresponding to
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605 F, which is also the caxiouc allowable core outlet temperature.
Voiding in the pri=ary system (excepting the pressurizer) is precluded f,_
in this case, since pressure does not decrease to saturation.
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Section 3.
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The safety analysis for more severe cc oldoen transients, such au the loss of feedwater event, indicates that the rater voluae could ' decrease to less than the syste= volume exclusive of the pressuriscr.
During such an event, the emptying of the pressurizer would be followed by
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a pressure reduction below the saturstion point and the formation of small voids throughcut much of the primary system.
This vould not i
result in the loss of core cooling because the voids would be dispersed over a large vclume and forced flou would prevent thea from coalestin sufficiently to prevent core cooling.
The high pressure coolant o
injection pumps are started autcoatically when the pricary pressure J
'j decreases below 1600 psi.
Therefore, any pressure reduction which is suf ficient to allow voiding vill also result in water inj ection which gj will rapidly restore the primary water to normal levels.
For these reasons, we believe that the inability of the pressuriacr' and normal coolant nekeup systen to control some transients does not provide a basis f or requiring core capacity in these systens.
Gene: al Desigr Criterion 13 of Appendix A to 10 CFR 50 requires instru_untation to monitor variables over their anticipated ranges for " anticipated operational occurrences".
Such occurrences are specif).cally der'ned to include lons of all offsite power.
The fact that T cold goes ff scale at 520 F 1: not cons.dered to be a deviation frba this requit ent because this indicator is backed up by wide range tenperature dication that extends to a low limit of 50 F.
Neither do uc cont i-; c the tahcup flow monitoring to deviate cince the amount of cakeup tiow in excess of 160 gpa does not a significant factor in the course of these occurrences. ppear tc be a
The loss of pressurizer water level indication could he consi'dered to deviate fren GDC 13, because this level indication provides the principal ceans of deternining the primary coolant inven to ry.
However, provision of a level indication that would cover all anticipated occurrences may not be practical.
As discussed above,. the loss of feedvater event can lead to a comentary condition wherein no meaningful level extr.ts, because the entire pricary systes contains a. steam varer mixture, It should be noted that the introduction to Appendix A (last paragrapih recognizes that fulfill:ent of some of the criteria nay not always be y'
appropriate.
This introduction also states that departures from the Criteria cust he identified and justified.
The discussion of CDC 13 3
in the Davis Besse FSAR lists the water level instrumentatic', but does not r.ention the possibilic; of loss of water 1cval ind. ation i
during transients.
This apparent osiscion in the safety an _ynis
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vill be subj ected to further review.
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UtlITED STATES
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' tiUCLEAR REGULATORY C0:01ISSIO::
0FFICE OF IllSPECTIOil AtID EllFORCEttEllT WASHIf!GTON, DC 20555 IE Bulletin flo.79-05A Date:
April 5,1979 cc Page 1 of 5
}c f!UCLEAR IllCIDEilT AT THREE MILE ISLAf!D - SUPPLEMENT
/
h Description of Circumstances:
?Jv Preliminary information received by the f;RC since issuance of IE D
Bulletin 79-05 on April 1,1979, has identified six potential human, design and mechanical failures which resulted in the core damage and r:Jiation releases at the Three Mile Island Unit 2 nuclear plant. The information and actions in this supplement clarify and extend the original Bulletin and transmit a preliminary chronology of the TMI accident t_hrough 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 time of the initiatii., 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.,the actuation level.
3.
Following rapid depressurization of the pressurizer, the pressurizer level indication may have lead to erroneous inferences of high level in the reactor coolant sys tem.
The pressurizer level indication apparently led the operators to prematurely terminate high pressure injection flow, even though substantial voids exi; ted in the reacts coolant system.
4.
Because the containment does not 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 floor.
Outgassing from this water and discharge
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through the auxiliary building ventilation system and filters was f:
the principal source of the offsite release of radioactive noble
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gases.
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5.
Subsequently, the high pressure injection system was intermittently
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operated attempting to control primary coolant inventory losses y
through the electromatic relief valve, apparently based on
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pressurizer level indication.
Due to the presence of steam and/or A
noncondensible voids elsewhere in the reactor coolant system, I this led to a further reduction in primary coolant inventory.
410 195
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April 5, 1979 x
Paga 2 of 5 6.
Tripping of reactor coolant pumps during the course of the transient, f
to protect against pump damage due to pump vibration, led to fuel damage since voids in the reactor coolant 'iystem prevented natural circulation.
9 Actions To Be Taken by Licensees:
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For all Babcock and Wilcox pressurized water reactor facilities with an d
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 toward
~ understanding the sequence of events 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 (Enclnsure 1) which have occurred at your facility (.ies).
If any significant deviations from expected performance are identified ia your revicw, provide details and an analysis of the safety significance together with a description of any corrective actions taken.
Reference may be made to previous infomation provided to the f!RC, if appropriate, in responding to this item.
3.
(This item claHcies 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:
Recognition of the possibility of forming voids in the primary a.
coolant system large enough to compromise the core cooling capability, especially natural circulation capability.
b.
Operator action required to prevent the format-ir i of such voids.
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c.
Operator action required to enhance core cooling in the event 1
such voids are for 2d.
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Date: April 5, 1979 Page 3 of 5 4.
(This item clarifies and expands upon item 4. of IE Bulletin 79-05.)
nr 7.J P.eview the actions directed by the or: rating procedures and training W
instructions to ensure that:
ib Operators do not override uutcaatic actions of engineered k
a.
safety features.
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b.
Operating procedure; currently, or are revised to, specify 1,
that if the high pressure injection (HPI) system has been automatically act.uated because of low pressure condition, it must remain in operation until citi er:
(1)
Both low pressure injection (L.PI) 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 sys:em has been in operation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the saturation temperatore for the existing RCS pressure.
If 50 degree subcooling cannot be maintained af ter HPI cutoff, the HP; shall be reactivated.
Operating procedures currently, or are revised tu, specify c.
that in the event of HPI initiation, with reactor coolant pumps (RCP) operating, at least one RCP per loop shall remain ope ra ting.
d.
Operators are provided additional information 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 revises item 5. of IE Bulletin 79-05.)
y b
Verify that emergency feedwater valves are in the open position in accordance with item 3 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 encineered safety features.
Also review related procedures, suci as those for maintenance and testing, to ensure that such valves are returned to their correct positions 7
following necessary manipulations.
410 197 w em-
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Date:
Aprii 5, 1979 Page 4 of 5
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nr 6.
Review the containment isolation initiatic design and procedures, W
and prepare and implement all changes necessary to cause containment
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isolation of all lines whose isolation does not degrade core cooling capability upon automatic initiation of safety injection.
7.
For manual valves or manually-operated motor-driven valves which could defeat or compromise the flow of auxiliary feedwater to the D
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 uhen heat removal from the primary system is through the steam generators.
When two inde-pendent 1003. capacity 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 steam generators for cooling within the next.12.. hours.
When 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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.
In particular, ensure that such an occurrence would not be caused 3
by the resetting of engineered safety features instrumentation.
List all such systems end indicate:
J a.
Whether interlocks exist to prevent transfer when high radiation indication exists, and J
b.
Whether such systems are isolated by the containment isolation 1
signal.
410 198 O__
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April 5,1979 Page 5 of 5 10.
Review and modify as necessary your naintenance and test procedures to ensure that they require:
y d
Verification, by inspection, of the operability of redundant i
a.
safety-related systems prior to the removal of any safety-related system from service.
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3:
y b.
Verification of the operability of all safety-related systems U
when they are returned to service following maintenance or testing.
qj 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 fiile 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 r
assure very early notification of serious events.
k 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 sutaitted to the Director of the appropriate ?!RC Regional Of fice and a copy should be forwarded to the f;RC Office of Inspection and Enforcement, Division of Reactor Operatinns Inspection, Washington, DC 20555.
For all other reactors with an operating license or construction permit, this Bulletin is for infomation purposes and no written response is required.
3 Approved by GAO, B 180225 (R0072); clearance expires 7-31-80.
Approval was given under a blanket clearance specifically for identified generic J
probl ems.
W d
Enclosures:
g 1.
Preliminary Chronology of Tf1I-2 3/38/79 3
Accident Until Core Cooling Restored.
g 2.
List of IE Bulletins issued in last 12 months.
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IE Bulletin flo.79-05A Date: April 5,1979 Page 1 of 3 5
PRELIMIt1ARY d
CHR0:10 LOGY OF TMI-2 3/28/79 ACCIDEtiT.
r UitTIL CORE C00 lit;G RESTORED
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TIME (Approximate)
EVErlT n
u about 4 AM Loss of Condensate Pump L
(t
- 0)
Loss of Feedwal.er 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 611 degrees F, 2147 psi (450 psi over
~~ =
saturation) t = 30 sec.
All three auxiliary feedwater pumps running at pressure (Pumps 2A and 28 started at turbine trip).
flo flow was injected since discharge valves were closed.
t = 1 min.
Pressurizer level indication begins to rise rapidl; t = 1 min.
Steam Generators A and B secondary level very low - drying out over next couple of minutes.
m t = 2 min.
ECCS initiation (HPI) at 1600 psi t = 4 - 11 min.
Pressurizer level off scale - high - one HPI pump manually tripped at about 4 min.
30 sec.
Second pump tripped at about g
10 min. 30 sec.
a t = 6 min.
RCS flashes as pressure bottoms out at
^
1350 psig Hot leg temperature of 584 degrees F).
410 200
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Date:
April 5,1979 Page 2 of 3 TIfiE EVEtiT 4
t = 7 min., 30 sec.
Reactor building sump. pump came on.
t = 8 min.
Auxiliary feedwater flow is initiated by opeiling closed valves gj p
t = 8 min.18 sec.
Steam Generator B pressure reached minimum a
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 flPI 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 condition at about 1015 psig and about 550 degrees F.
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 pumps in loop B 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 BEGIflS !! EAT UP TPRISIEr1T - liot leg temperature begins to rise to 620 degrees F (.off scale within 14 minutes) and cold leg temperature drops to 150 degrees F.
(llPI 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 y
operator after S.G.-B isolated to prevent leakage j
t = 3 hou s RCS pressure increases to 2150 psi and
'.o electromatic relief valve opened x
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 5
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 Il psi -
(I' RCS pressure 1750; containment pressure increases from 1 to 3 psig 410 201 h
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Date: April 5, 1979 Page 3 of 3 TIf4E EVENT Y
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t = 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> n
Peak containment pressure of 4.5 psig pl 9
t = 5 - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> RCS pressure inc eased from 1250 psi to 4
2100 psi 3
t = 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> rt Operator opens electromatic relief valve to 2
depressurize RCS to attempt initiation of 3d RHR at 400 psi 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 discharge 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 af ter 500 gal. of NaOH injected (about 2 minutes of operation) 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 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 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 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 Now (4/4)
High radiation in containment All core thermocouples less than 460
)
degrees F.
Using pressurizer vent valve with small
'd makeup flow ijj Slow cooldown RB pressure negative s
b-N2L 410 202 re we
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, =
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.c UNITED STATES liUCLEAR REGULATORY COMMISSIO.'t
(
OFFICE OF INSPECTION AND ENFORCEf!ENT A
SASHINGTON, DC 20555 APRIL 21, 1979 l
!!UCLEAR INCIDENT AT THREF_ HILE ISLAND - SUpplEMEhT
[
.'m e
Descriptica of Circumstances:.
-e.
e
.g.
.c-Continued NRC evaluation of the nuclear incident at Three flile Islarid r,
Unit 2 has identified ceasures-in addition to those discussed in IE
. 1 r-
~Bullatin 79-05 and 79-05A which should be acted upon by licensees with
. L reactors designed by B&L. As discussed in Item 4.c. of Actions to be taken by Licensees in IES79-05A, the preferred mode of core cooling' following a trancient or accident i s te provide forced flow using reactor coolant pumps.
f[ 1..' It appears that natural circulation was not successfully achieved upon securing the raatter ccarlacpcmps-during the first two hours of the Three lille Island (THI) No. 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 it.terference with natural
... circulation, the operator should ensure that the primary system is subcooled, and remains,subcooled, before any attempt is made to establish N
'. natural circulation.
Natural circulation in' Babcock and Milcox 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 promoted by injection of auxiliary feedwater at the upper nozzles in the OT5Gs.
The integrated Control System automatically sets the OTSG level setpoint to 50% on the operating range when all reactor coolant pumps (RCP) nre secured.
However,,
in unusual or abnon.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 reactcr coolant pumps.
As stated previously, forced flos of primary coolant through the core is preferred to r.atural circuiation.
y Other means of reducing the possibility of void famation in the reactor coolant system cre:
7 p
A.
Minimize the operation of the Pcwer Operated Relief Valve (PORV) on the pressurizer end ther ey reduce the possibility of pressure 2
y reduction by a blowdown through a PORV that was stuck open.
c.D 1.
r
~7 5555b5.=.nh-!!5.$5.-T-h5b:
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IE Bulletin 79-033 April 21,1979 Page' 2 of 4 S.
Reduce the energy input to the reactor coolant system by a prompt reactor trip during transients that result in primary system pressure.
increases.
N This bulletin addresses, among other things, the means to achieve these a
J objactives.
R 1
d
.J._'f Actions To Ba Taken by Licensees:
- f4 5
,. c.
0-7 M For all Babcock and Hilcox pressurized water reactor facilities with an
__i(,f._.. x s uparsada, IEB-79-05A).(Underlined sentences are modifications to,.and L-N-
operating license:
k
, + a..
3-l[.
- 1. - Develop procedures and train operation personnel on methods of
. 5. 4.
establishing and maintaining natural circulation.
The procedures W'ec.
and training must include means of monitoring heat removal efficiency. '
T,. Z ~~
by available plant. instrumentation.
The procedures must also.contain '
EF -
a trethod of assuring that the primary coolant system is subcooled by
~f: ' ~'
at least 50*F before natural circulation is initiated.
- X -
- @y. - :
In the event that these instruct. ions incorporate anticipatory fillino
-.; c.f. '
of the OTSG prior to securing the rea^'or coolant pumps, a detailed
- W>~ T.. ~analysis should ba done to provide guidance as to the expected system response.
The instructions should include the following precautions:
.s, w
s; ",
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
.;.g.y.,
satisfy the subcocling criterion for natural circulation; s
caintain oressure - temperature envelope witt.in Appendix G limits c.
u.
for vess i integrity.
..A,,
Procedures and training shall also be provided to maintain core cooling in the event both main feedwater and auxiliary feedwater are lost while in the natural circulatico core cooling mode.
Modify the actions required in Item 4a and 4b of IE Bulletin 79-05A t
w
~.. 2.
'[
to take into account vessel integrity considerations.
e
~
"4.
F.eview the action directed by the operating procedures and
~~
training instructions to ensure that:
o a
Operators do not override automatic actions of engineered a.
safety features, unless continued operation of engineered
- qf-3.a,,
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,(
April 21,1979
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Page 3 of 4
(
f safety _ features will result in unsafe plant conditions.
5 For 2
example, if cont:nued operacion or engineered safety feature' would tnreaten reactor vessel integrity toen the HPI should be p
s.
secured (as noted in o(2) celow).
.~
~
~
- .t. : %
v.
b.
Ocerating procedures. currently, or are revised to, specify the-9.* M M
if the-Mgtr pressure injection WL actuated because of low pressure (!!PI') system has been automtically - 2 operation until either:.
cendition, it must remain in-
?,,
4.1,' C. - t:
(1)e sa Both, low pressure injection. (LPI L
and ficwing at a rate in excess o)f 1000 gpm each ut.a.th punps are in operation
- g~ ',
situation has-been stable for 20 minutes, or sa :-:
(2)
~.~2C- -
The HPI system has been in operation for 20 minutes, and TE X.',
all. hot and cold leg ter.'-
atures are at least 50 degrees
.c-below the saturation temp.ature for the existing.RCS "9.
pressure.-. If 50 degrees subcooling cannot be maintained
[.2J.::
~
after HPI' cutoff, the llPI shall bc reactivated.
[3.-
The daaree of subecolino beyond 50 degrees F and the length lif time ~
.:Qe.
11PI is in operatica snall be limited by the pressure /
..--. 6 temperature considerations for the vessel intearity."
~
3.
Following detailed analysis, describe the modifications to design and procedures-which you-have imp' emented to assure the reduction of the likelihood of automatic actuation of the pressurizer PORV during
~
anticipated transients.
This Enalysis shall include consideration of a rmdification of the'high r ressure scram setpoint and the PORV opening setpoint such that rcactor scram will preclude opening of EW irrEnclosure 1.the PORY for the spectrum of anticipated transients discuss 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 ranual trip of the reactor for transients that result in a pressure increase in the reactor coolant system.
These transients include:
-d-a.
loss of main feedwater a
- c; b.
turbine trip n:.
a
'.E,. ' b Main Steam Isolation Valve closure p
c.
1
~
ng- }
d.
e.
Iow OT5G level
. l f.
109 pressurizer level.
l n
s:-
s s., 1 410 205
~~..U_^"
...l.4 ~*
- ..c -
IE Bulletin 79-053 April 21,1979 Page 4 of 4 s
5.
Provide for ilRC approval a design review and schedule for implesren of a safety grade automatic anticipatory reactor scram for loss of feed-water, turbine trip, or significant reduction in steam generator level U
6.
The actions required in item 12 of IE Bulletin 79-05A are modiffed as O
hi
~ ^
follcws:
f,$
U
- y....
Review ycur pregt reporting proceduras for flRC notification to assure-5.
m G:---?~
that tiRC is notified within one hour of the time the reactor is C
"%'J a controlled or expected condition of operation.
-uR R. *t l' WD Eccen contnnucus: comunicacion channel shall be established andFurther L
- a
% Q~'..<~
raintained w1th HRC.
fli:F 7.
prooose chances as recuired, t
'i -
reust te rediried as a result of your amolemontino the above itemsth n.
. _ f]j:.
P.esponse schedule for B&W designed facilities:
.a.
4
- a..ror Items 1, 2, 4 and 6, all facilities with an operating license respond within 14 days of receipt of this Bulletin.
~'
b.
..?: ~ - c.
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 />.
operating, respond before resuming operation.All facilities wit k:ke-
.For Items 5 and 7, all facilities with an operating license respond c.
~~
in 30 days.
~
- Rcports should be submitted to the Director of the appropriate f;RC Regiona Office and a copy should be forsarded to the llRC Office of Inspection and Enforce =nt, Division of Reactor Operations Inspection,1:ashington _
20555.
, D. C, lor all other.pcuer reactors with an operating license or construction permit, this Bulletin s for information purposes and no written response is required.
' - /gproved by GAO, B180225 (R0072); clearance expires 7/31/80.
~
i was given under a blanket clearance specifically for identified generic Approval T
- problems,
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Co.'iTIUDING REVIEM OF THE SEQUENCE OF EVENTS LEADIffG
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-2 UN PARCH ZU,1979 SHGM5 TilAT ACTIOri CAU DE TAY.Ed TO PHOVIDE ASSUR iHAT THF. PILOT-OPERATED RELIEF VALVE (PORY) t PLA*TTS MILL R3T BE ACTUATED BY A7iTICIPATED TRAll5IEMIS 1riHICil FX(E A SIGTIIFICA?fr PROBASILITY OF UCCURRING_ IN T1tESE PLMTS GCCURRED OR
!XTT DEGRADE Tl!E SAFETT OF Ti!E AFFECTED PLNITS HITH RESPECT TO 7;1E A.TTICIPATED TRAN51EifiS OF COHCEiFi ARE:TO T;GrmL, UPS f
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':TIONS PROPOS,ED 1,, h TIESTRECTIiG IUACTOR POWER TO A VALUE DIICH IT THE:PORV.
FGITTTS REMAIHED AT-THEIR-CURRENT-vat UES.THE. REACTOR FR
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T65URE MG ACTUATION OF THE PORT.LG'!EP.IMG TiiE THE DESICli THE SETPOINT FUR PORY ACTUATIori REmINED AT THEIR CURR
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AOJn'ITIP3 THE HIdi PRE 5SURE TP.IP AND THE PORY SETP PGT! ACTUATIO,'l FOR THE CLASS OF A'iTICIPATED EVENTS OF CONCERN PEE 55UP.E OF THE REACTOR nEMAlt!ED AT IT5 CURREnr VALUE.
THE DESICal
'1; NiALYSIS OF THE INFACT OF T1!ESE VARICUS ALTERNATIVES iviD TH iG A55LQING THAT THE PORY WILL riot ACTUATE FOR THE
!? CONCEFJi HAS DEEM C0?C3LETED,.
THE RESULTS 5HOW THAT:
3.
LWERIris Tid flIGH PRESSURE REACTOR TRIP SETPOINT FROM ?
3, 2J55 PSIG TO 2.IGO PSIG Y,,
' N.
W AGD
. l.3 1
~ ;PAISING THE SETPOIrrT FOR THE PILOT OPERATED RELIEF VALVE
],]FE1_2255 PSIG TO 2450 PSIG i0 VIDE 5 TNE REQUIRED ASSURNiCE.
THIS ACTION HAS THE FURTHER ADVit!TAGES O a s.
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FEtSER'lIHG PRESSURE RELIEF ChP5 CITY.FOR ALL HIC
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ELTHlitATING T1tE PosS!31LITT OF IliTROD'!CING UtinEVIE Q@i] cfEliT EMERCEllCT FEED:ATER FLC'ri L:
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LOl:.U.OR TRIP OF TriE TunaurE.
THIS CAPABILITY REQUIRES ACTUATIOri 0F Tile P CFERATED RELIEF VALYES,.
THE CAPABILITY IllCREASES THE P.ELIABILITY OF PO SUFFL't TO THE S'/ STEM BY P~TURTlIlLG.THE UHITS TO PO AFTER THESE TRAH5IEUTS-ETR BE TRIPPED FOR THESE EYEUTS:THE ACTION PROF 05ED ABO
.1:\\ -l
._ liOTE:
- ThF effect of changing the rnactor coola5t system pressure trip setpoint upon peak pressurizer pressure is typ1 fled by the attached figure 1. which was developed by-B&W for a loss of feedwater transient.
, 9.
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I'JWICIPATED TRi3SIEi;T5 NHIC?f HAVE OCCURRED AT BLM FLNirS A'lD itIIICH
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f ETICIPATED TRNfSIErfTS Hi!ICH HAVE NOT OCCURRED AT BsW PLAN T' a03ABILTIT E'iE?TT3) NiD 11H101 MOULD NOFX4LLY ACTUATE PORY AT THE CU1RENC Su ruIliT (2255 'PSIG):
A.-
SOME CGilTROC RCD GROUP HITHURA'lALS (MODERATE TO IIIGli REACTIVITY
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NUCLEAR REGULATOBY COMMISSION
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' April 20,.1979 x
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'" - For.crrble Victor G111=nkv K - A.*.:tL g Ch.nimn -
L U. 5. Np: lear RegclMory Coc=:icnicn
,, Kathington, DC 20355
Dear Dr. Gilinsky:
~
'b.is ? 2tter is in responce to youth of April 18, 1979 Lhich re ;uested 9.at the ACRS notify the 'Cor.niscionees inco5iately if pe believe any of our oral recoc:n:.ndation= of Ap:11 17? should te acted upon b= fore otte ne:!: regularly scheduled :merity., at which we could urepsto n for:e1
. let't cr.
Toa Coe.mittee discussed this topic by conference telephona cc11 an' April 19 a.d offers the following. ce=mnts.
\\
20.1 of the rececendstfons r:ude by the ACRS in its' r.,ecting with the cc:u.icciences on April 17, 1979, ere generic in nature and coply to all F.Gn.,Nanc. verc intended to require ij=edicto chpn2es in oportting pro-ceduros or pinnt radificatiens of operating F,ms.
Such ch mes chould be rade onIV af ter study of their effects on overall safety.
Su;h stud-ie.s chatu.c be c.3de by the licensees and their suppliers or consu.lters.c i.md trf the IG.C Staf f.
TnO Co=tittpe believes that. th2se stuSies that0 d b2 b-pcn in the naar future on a time scale that will not divert the
.NR" staif or the industry representatives from their tasks relating to the ccoldown of Tnree Mile Irland Unit 2.
nawever, the Cemitteo be -
. lievts thht-it 1:ould be rossible arri desirc.ble to initiate irr.m3iately a curvey of opcrating p: redures for achiev.ing naturh1 circulation,, in-.
c1t'dirrj thc. cf.de when ofFn'ite pr. :'r is lost, an-3 the ro'l e of the pre.r Ourl2.=r htaters in su E,pYdcedtros.
q Ist its ocetlng ca I@ril 16 and 17i.1979, the co= Itteo discusse.d.trith ra
~
' the ERC S-iff the r. utter of natural circulation for the Shree MD.o Ic.
.L' le.4 Unit 2 plant.
Tne cr H ttee bellCves that this raattor is receiv-Ing careful attention by the Nac staff and the 21cencee.
la-
..m.
n To ED3 for Appropriate Action.
Distribution:
~DR, OlA.
Rapifcxcd to Epo, PA, E.. case.
79-1117..
L_
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?gth (A6,g a.S fJUCLEAR REGULATORY COT.R'.10SION
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April 18,1979
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P.IMCPJ2% FOR:
Chairman Hendric y
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Cc.missic:n Gilinsky M
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.cc=, ~miener Bradford L
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R. F. Fraley, E:cecutive Director Adv}sory Ccimittee on Reactor Sa T gcards
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. Attached for your inforr.2 tion and use. is a copy of :t'ne recc=eM2-tions of the Adviso..y Committee on Reactor Safeguardo Valch were orally presented to and discussed with you on April 17,1979 re--
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gardirs the recent accident at the Three Mile Island Nuclear Sta-tion Unit 2.
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y nL' R.;P. Fraley Executive Director
. Attach :ent-Reccmmandations of the NRC Advicory Co.rrnittee on Reactor Safeguards Re. the 3/28/79..ccident -
at The Three tiile Island Nuclear Station Unit 2 g_.
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'P.EC&_WGTICNS OF Tric h"JCLEAR RS30EKIDRY COTISSION ADVISORY COM!iIT d REWR SAFEI30ARDS R33ARDING TPE t'. ARCH 28,.1979 ACCIDEST AT TcIE 'mRES HILE ISUJin tr.. EAR STATION UNIT 2 7 -1 presel.ted orally to, arr3 discussed with, the NRC
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~.1'-j Com:nissioners during the ACRS-Co.=nissioners F:ecting
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on April 17; 1979:- mshington, D. C.
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Estural circulation Is an important rode of reactor cooling, both as
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a planned process and as a process that rnay be used under abnormal circu= stances.
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%e Cem.aittee believes that greater understanding oc h
-' 1 : 'this rode of cooling'. Is._ required and that. detailed analyses.should -
E C, by dev-loped by licensees or-their suppliers.
Tne analyses should be W-supported, as necessary, by - experiment.
Procedures should be ' de--
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veloped for initiating ~ natural circulation in a safe manner and for providing the operator with assura:re that circulation has, in fact, been established.
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.21s may require. Installation of instrumentation to measure or indicate flew at low water velocity.
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sha use of natural circulation for decay heat removal following a loss of offsite power sources requires the maintenance of a suitablo over-pressure on the reactor coolant system.
This overpressurei may be assured by placing the pressurizer heaters on a qualified onsite powr source with a suitable arrangement of heaters and power distri-y bution t
to provide -redundant: capability..
Presently ~ oparating ESR plants should b2 surveyed expeditiously to detennine whether such arrangpents can be provided. to assure this aspect of natural circula-tion capability.
Tae plant operator should be adequately ir formeil at all times con-cerning the conditions of reactor coolant system operation t.hich might affect the capability to place the system in the natural circu-lation rode of operatio.T or to sustain such a made.
Of particular importance is that information which-might indicate that the reactor coolant system is approaching the saturation pressum correspendityg-to the core exit temperature.
This impending loss of system over-pressure will signal to the operator a possible loss of natural circulation capability.
Fuch a warning may be derived from pressur-3f iter pressure instruments an6 hot leg temperatures in conjt' action with
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conventional steam tables..
A suitable display of this information should be provided to the plant operator at all times.
In addition,
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consideration should be given to the use of the flow exit tempera-tures from the fuel subassemblies, where available, as an additional M
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'Ihe exit temperature of coolant fre.n the core is currently measured by ther:cccuples in many hgs to 'detersine core performance.
We.
ccanittee recem ends that these temparature measurenents, as currently
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available,.be used to guide the operator concerning core status. ne 5
range of the information displayed ar 1 recorded should include tha
'I full capability of the thermocouples.
It is also recom..crded that E
ether existirx3 Inst =entation be exatained for ii s passible use in whting operating a;: tion during a. transient.
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_ - _;wi'th rega rd to the. definition and imple: entation of instre nantatiert
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course of a serious accident.
This should include improved sampling
.d. precedure.2 under accident conditions and techniques to help previde 1[M improved guidance to offsite authorities, shoult1 this be needed.
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_...Cortnittee recomends that a phased Implementation approach b e - em-played so that techniques can be adopted shortly after they are J r.-J' 27- {. judged to be appropriate.
ne%Cas reccc. mends that a high priority be placed on the developnent
. and implementation of safetv research on the behavior of light water
- reactors during anomalous transients.
ne NRC may find it appropriate to develop a capability to si:nulate a wide range of pastulated tran-nient cr.d accident conditions in ordar to gain ir reased insight into ceasures which can be taken to improve reactot sa fety.
The ACRS uis%s to reiterate its previous reccamendations tat a high priority be given to research to improve reactor safety.
- Consideration should be given to the desirability of ' additional.
~. ' equip.ent status r>anitoring on.various engineered safeguards features
- and their supportirr3 services to help assure their availability at all times.
%e ACRS is continuirrg its review of the implications of this accident a:vd hop' to provide further advice as it is developed.
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IE Information Notice 79-16 June 22, 1979 LISTING OF IE INFORMATION NOTICES ISSUED IN 1979 Information Subject Date Issued To Notice No.
Issued -
79-01 Bergen-Paterson H draulic 2/2/79 All power reactor 3
Shock and Sway Arrestor facilities with an Opera ting License (OL) or a Construc-tica Permit (CP) 79-02 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 Operatit.g Licer.;e (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 with an Operating License (OL) or a Construc-tion Permit (CP)
Enclosure Page 1 of 2 410 2i6
b 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 Radioactively 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 2 410 217