ML19225B097
| ML19225B097 | |
| Person / Time | |
|---|---|
| Site: | Texas A&M University, 05000128 |
| Issue date: | 06/22/1979 |
| From: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Berg R TEXAS A&M UNIV., COLLEGE STATION, TX |
| References | |
| NUDOCS 7907230507 | |
| Download: ML19225B097 (1) | |
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UNITED STATES t
y $,;#rg' 'g NUCLEAR REGULATORY COMMISSION gpg 4
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June 22, 1979 In Reply Re f er To :
Docket No. 50-59 50-12P.
Texas AMI Universit/
ATTN:
Dr. R. R. Terg, Director Office of University Research College Station, Texas 77843 Gentlemen:
The enclosed Information Notice No. 79-16 is forwarded to you for in-No specific action is requested and no written response is required.
formation.
this If you desire additional information regarding this matter, please contact office.
Sincerely, f
fl, 0,/
b)/
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Karl V! Seyfrit /
Director
Enclosure:
1.
IE Infcrmation Notice 79-16 2.
List of IE Information Notices Issued in 1979 790723;;;
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AL9 ENF0kCEMENT 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 m kes 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 'Se 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, contact the Direccor of th e appropriate NRC Regional Office.
Enclosures:
1.
IE Bulletin No. 79-05 with Enclosures 2.
IE Bulletin No.79-05A with Enclosures If.; I)ff I.hs 'A~, 4 _
3.
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UNITED STATES
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fiUCLEAR REGULATORY COMMISSION OFFICE OF IftSPECTION AND ENFORCEMEtiT WASHINGTON, D.C.
20555 April 1,1979
?
2 fiUCLEAR INCIDENT AT THREE MILE ISLAND U
Description of Circumstances:
5 On March 28, 1979 the Three Mile Island t!uclear Power Plant, Unit 2 14 experienced core damage which resulted from a series of events whi S were initiated by a loss of feedwater transient.
Several acpects of the incident may have general applicability in addition to apparent generic applicability at operating Babcock and Wilcox reactors.
This balletin is provided to inform you of the nuclear incident and to request certain actions.
Actions To Be Taken By Licensees (Although the specific causes have not been determined for individual sequences in the Three Mile Island event, some of the following may have contributed.)
For all Babcock and Wilcox pressurized water reacto.- 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 fiRC staff of a portolated severe feedwater transient related to Babcock and 'ilcox 11Rs as described in Enclosure 2.
These reviews should be cirected at assessing the adequacy of yaur 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 expected performance occurred.
If any significant deviations are 4
found, provide the details and an analysis of the significance and lg any corrective actions taken.
This material may be identified by j
reference if previously submitted to the ?!RC.
j 3.
Review the actions required by your operating procedures for coping 5
n with transients.
The items that should be addressed include:
J 3L h
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IE Bulletin fio. 79=05 April 1,1979 Page 2 of 3 Recognition of the possibility of forming voids in the primary a.
coolant system large enough to compromise the core cooling capabili ty.
=~
b.
Operator action required to prevent the formation of such voids.
U d
Operator action required to ensure continued core cooling in h
c.
the event that such voids are formed.
d 4.
Review the actions requested by the operating procedures and the la training instructions to assure that operators do not override automatic actions of engineered safety features without sufficient cause for doing so.
5.
Review all safety related valve positions and positioning require-cents to assure that engineered safety features and related equip-ment such as the auxiliary feedwater system, can perform their intended functions.
Also review related procedures, such as those for maintenance and testing, to assure that such valves are returned to their correct positions following necessary manipulations.
~
6.
Review your c;e;ating modes and procedures for ail 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 parcular assure that such an occurrence woul? 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 isolctit a signal.
7.
Review your prompt reporting procedures for f;RC notification to assure very early notification of serious events.
!il The detailed results of these reviews shall be submitted within ten (10) days of the receipt of this Bulletin.
d 01 f
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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 forwarded to the MRC Office of Inspection and Enforcement, Division of Reactor Construction Inspection, Washington, D.C.
20555.
q A
For all other operating reactors or reactors under construction, this 1
Bulletin is for information purposes and no report is requested.
y e
Approved by GAO, B180225 (R0072); clearance expires 7-31-80.
Approval was given under a blanket clearance specifically for identified generic
{
problems.
9
Enclosures:
1.
Preliminary Notifications Three Mile Island -
PNO-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 mcnths Il R
h 1.
N 410 253
Page 1 of 3 EVALUATION OF FEEDWATER TRANSIENT 1
i' A loss of offsite power occurred at Davis-Besse on November 29, 1977, 9
which resulted in shrinkage of the primary coolant vclume to the degree a
that pressurizer level indication was lost.
A recommendation to convey
]~
this information to certain hearing boards resulted in the attached discussion and evaluation of the event. This discussion includes a o
review of a loss of feedwater safety analysis assuming forced flow, kj 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.NDUM ENTITLED " CONVEYING NEU INFOPOL\\ TION TO LICE!; SING BOARDS - DAVIS-3 ESSE UNITS 2 6 3 AND MIDLO D UNITS 1 6 2",
DATED JAN'.L\\RY 8, 1979, FROM J.S. CRES'nELi TO J.F. STREETER.
3.
Inspection and Enf orcement Report. 50-346/78--06 document ed that y
pressurizer level had gone of f scale f or approxicately five 9
minutes during the Nove=be' r 2), 1977 loss of offsite power event.
- j Tacre are sc2e indications that.other B&W plants any have prob-Ic=s naintaining pressurizer level indications during transients.
O In addition, under certain conditions such as loss of feeduater d
100% power with the reactor coolant pu:2ps running the pres-O at surizer may void completely.
A special analysis has been per-g forned concerning this event.
This are. lysis is attached as.
Because of pressurizer level maintenance prob-lets the sizing of the pressurizer may require further review.
s Also noted during the event was the fact that Tcold vent off-scale (less than 520oF).
In addition, it was noted that the makeup flev monitoring is limited to less than 160 gpa and that makeup flow tiay be substantially greater than this value.
This infornation should be exacined in light of the require-ments of CDC 13.
DISCUSSION AND EVALUATION Thie event at D2vis Besse which resulted in loss of uressoriter level indication has been reviewed by URR and the conclusion was reached that no unreviewed safety question existed.
The pressurizer, together with the reactor coolant makeup system, is designed to maintain the primary syste= pressure and water level within their operational limits only during normal operating conditions.
Cooldown transients, such as loss of offsite power and loss of feed-water, saceti=es result in pri=ary pr ssure and volume changee that are beyond the ability of this sys tea to control. The analyses of and experience with such transients show, however, chat they can be sustained without compromising tac t2 fety of the reactor.
The principal concern caused by such transients is that they night cause voiding in the pricary coolant systea that would lead to loss of ability to ade-
'3 quately cool the reactor core.
The safety evaluation of the loss of t!
. of fsite pcuer transient shows that, thcugh level indication is lost, 5]
some water rcaalas in the pressurizer and the pressure does not decrease 3]
below about 1600 psi.
In order for voiding to occur, the pressure cust decrease below the saturation pressure corresponding to the systen
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tecoerature.
1600 psi is the saturation pressure corresponding to il 605 F, which is also the =aximur allowable core outlet temperature.
A Voiding in the pri=ary systcc (excepting the pressurizer) is precluded
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in this case, since pressure does not decrease to saturation.
e 41 f) lic IU L o, J
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Section 3. (
The safety analysis for more severe cooldoen transients, such as the loss of feedvatar event, indicates that the water voluae could decrease to less than the syste= volume exclusive of the pressurizer.
During such an cvent, the emptying of the pressurizer vould be followed by I
a pressure reduction below the saturation point and the fornation of small voids throughoyt much of the primary system.
This vould not u
result in the loss of core cooling because the voids would ba dispersed q
over a larga volune and forced flow would prevent them from coalescing
@i sufficiently to prevent core cooling.
The high pressure coolant g
injection pumps are started autcmatically when the pricary pressure a
decreases below 1600 psi.
Therefore, any pressure reduction which is sufficient to allev voiding vill al o result in water inj ection which 3,
will rapidly restore the pricary va..r to.nornal levels.
For these reasons, ve believe that the inability of the pressuri:cr.
and nornal coolant nckcup systen to control some transients does not provide a basis f or requiring more capacity in these systens.
General-Design Criterion 13 of Appendix A to 10 CFR 50 recuires instrunantation to monitor variables over their anticipated ranges for "anticipa ted operational occurrences".
Such occurrences are specif' cally defined to include loss of all of fsite power.
The fact i
that T cold goes off scale at 520 F is not considered to be a deviation fr~o= this requirement because this indicator is backed up by vide range temperature indication that extends to a lov limit of 50 F.
Neither do ve consider the takeup flow monitoring to deviate cince the caount of cakeup flow in excess of 160 gpa does not appear to be a significant factor in the course of the;e occurrences.
The loss of pressurizer water level indication could be considered to deviate from GDC 13, because this level indication provides the principal ceans of deternining the pri=aty coolant inventory.
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 r.o meaningful level exists, because the entire primar' system contains a. steam water mixture.
It should be noted that the introduction to Appendix A (Inst paragraph) recognises that fulfill =ent of some of the criteria aay not always be
]
appropriate.
This introduction also states that departures from the Criteria cust he identified and justified.
The discussion of CDC 13 d;
in the Davis Besse FSAR lists the water level instrumentation, but j
does not nention the possibility of loss of water level indication Q
during transients.
This apparent osission in the safety analysis j
vill be subjected to further review.
me t
- e 410 256 ame m
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U:lITED STATES
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flVCLEAR REGULATORY C0:dISSIO 1 0FFICE OF IfiSPECTI0tl A!D EflFORCEMEflT WASHIflGTO:1, DC 20555 IE Bulletin flo. i9-05A h
Date: April 5,1979 Page 1 of 5 flUCLEAR IfiCIDEflT AT THREE MILE ISLAt:0 - SUPPLEMEflT Description of Circumstances:
f Preliminary infomation received by the flRC 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 radiation 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 initiating event, loss of feedwater, both of the auxiliary feedwater trains were valved out-of-service.
2.
The pressurizer electromatic relief valve, which opened during the initial pressure surge, failed to close when the pressure decreased below,the actuation lavel.
3.
Follo'.ving rapid depressurizatit n of the pressurizer, the pressurizer level indication may have lead t.o erroneous inferences of high level in the reactor coolant system.
The pressurizer level indication apparently led the opcrators to prematurely teminate high pressure injection flo.1, even though substantial voids existed in the reactor 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 through the auxiliary building ventilation system and filters was
?
the principal source of the offsite release of radioactive noble gases.
Gis 5.
Subsequently, the high pressure injection system was intermittently h
operated attempting to control primary coolant inventory losses
]
through the electrcmatic relief valve, apparently based on 3
pressurizer level indication.
Due to the presence of steam ar.d/or
+-
noncondensible voids elsewhere in the reactor coolant system, I this led to a further reduction in primary coolant inventory.
77 0nu
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Date:
April 5, 1979 Page 2 of 5 6.
Tripping cf reactor coolant pumps during the course of the transient, I
to protect against r. ump damage due to pump vibration, led to fuel e
damage since voids in the reactor coolant system prevented natural circulation.
Actions To Be Taken by Licensees:
13 Fo. 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 Bullr tin 79-05, review the enclosed preliminary chronolony 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 ar.y others which contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (.ies).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a description of any corrective actions taken.
Reference may be made to previous infomation provided to the !!RC, if appropriate, in responding to this item.
3.
(This item clarifies item 3. of IE Bulletin 79-05.)
Review the actions required by your operating procedures for coping with transients and accidents, with particular attention to:
a.
Recognition of tie possibility of foming voids in the primary coolant system large enough to compromise the core cooling j
capability, especially natural circulation capability.
1 1
b.
Operator action required to prevent the fomation of such J
voids.
E J,
J c.
Operator action required to enhance core cooling in the event 1
such voids are formed.
410 258 CT
(-
Page 3 of 5 Date: April 5, 1979 4.
(This item clarifies and expands upon item 4. of IE Bulletin 79-05.)
g d
Review the actions directed by the operating procedures and training d
instructions to ensure that:
lq a.
Operators do not override automatic actions of engineered 9
safety features.
I; n
b.
Operating procedures currently, or are revised to, specify
[
that 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 degreee below the saturation temperature for the existing RCS pressure.
If 50 degree subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated.
c.
Operating procedures currently, or are revised to, specify that in the event of HPI initiation, with reactor coolant pumps (RCP) operating, at least one RCP per loop shall remain opera ting.
d.
Operators are provided additional 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.)
S Verify that emergency feedwater valves are in the open position in 2
accordance with item 8 below.
Also, review all safety-related A
valve positions and positioning requirements to assure that d
valves are positioned (open or closed) in a marmer to ensure the i
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 o
following necessary manipulations.
410 259
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As il 5,1979 Page 4 of 5 pr-6.
Review the containment isolation initiation design and procedures, q
and prepare and implement all changes necessary to cause containment E
isolation of all lines whose isolation does not degrade core cooling f
capability upon automatic initiation of safety injection.
7.
For manual valves or manually-operated motor-driven valves which h
could defeat or compromise the flow of auxiliary feedwater to the 4
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 icmediately procedures which assure that two independent steam generator auxiliary feedwater flow paths, each with 100% flow capacity, are operable at any time when heat removal f rom the primacy system is through the steam generators.
When two inde-pendent 100% capac.ty 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 nextm12 hours.
When 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 co6 ling mode which does not rely on steam generators for cooling witnin 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 t 'd liquids ot.t of the prim?ry containment to assure that undesirec pumping of radioactive liquids and gases will not occur inadverteatly.
In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.
List j
all such systems and indicate:
a.
Whether interlocks exist to prevent transfer when high radiation indication exists, and 2
b.
Whether such systems are isolated by the containment isolation 1
signal.
410 c o O-m
April 5, 1979 Page 5 of 5 10.
Revieu and modify as necessary your maintenance and test procedures to ensure that they require:
f O
Verification, by inspection, of the operability of redundant f
a.
safety-related systems prior to the removal of any safety-related system from :,ervice.
7 b.
Verification of the operability of all safety-related systems 3
when they are returned to service following maintenance or testing.
L A means of notifying involved reactor operating personnel c.
whenever a safety-related system is removed from and returned to service.
- 11. All operating and maintenance personnel should be made aware of the extreme seriousness and consequences of the simultaneous blocking of both auxiliary feedwater trains at the Three tiile Island Unit 2 plant and other actions taken during the early phases of the accident.
- 17..
Review your prompt reporting procedures for fiRC notification to assure very earl; notification of serious events.
Fo.- Babcock and Wilcox pressurized water reactor facilities with an operating license, respond to Items 1, 2, 3, 4.a and 5 by April 11, 1979.
Since these items are substantially the same as those specified in IE Bulletin 79-05, the required date for response has not been changed.
Respond to Items 4.b through 4.d, and 6 through 12 by April 16, 1979.
Reports should be submitted to the Dir.ctor of the appropriate f;RC Regional Office and a copy should be forwarded to the f;RC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, DC 20555.
For all other reactors with an operating license or construction permit, this Bulletin is for infomation purposes and no written response is required.
3 Approved by GAO, B 180225 (R0072); clea"ance expires 7-31-80.
Approval b
was given under a blanket clearance spf :ifically for identified generic d
probl ems.
a'
Enclosures:
O 1.
Preliminary Chrcnology of Tfil-2 3/38/79 5
Accident Until Core Cooling Restored.
g' 2.
List of IE Bulle'; ins issued in last 12 months.
p,r iI dO
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Date: April 5, 1979 Page l'of 3 r
PRELIMINARY CHRONOLOGY OF TMI-2 3/28/79 ACCIDENT.
UNTIL CORE COOLING RESTORED j
d TIME (Approximate)
EVENT A
n i
e, about 4 AM Loss of Condensate Pump L
(t = 0)
Loss of Feedwaixc 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 2B started at turbine trip).
No flow was injected since discharge valves were closed.
t = 1 min.
Pressurizer level indication begins to rise rapidly t = 1 min.
Steam Generators A and B secondary level very low - drying out over next couple of minutes.
t = 2 min.
ECCS initiation (HPI) at 1600 psi t = 4 - 11 min.
Pressurizer level off scale - high - one 1
HPI pump manually tripped at about 4 min.
6 30 sec.
Second pump tripped at about 10 - '
30 sec.
I t = 6 mir..
.eahes as pressure bottoms out at 1350 psig Hot leg temperature of 584 degrees F).
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IE Bulletin tie.79-05A Date: April 5, 1979 Page 2 of 3 TIf4E EVEtlT
{
o t = 7 min., 30 sec.
Reactor building sump pump came on.
{
t = 8 min.
Auxiliary feedwater flow is initiated 5
by opening closed valves jj 3
t = 8 min.18 sec.
Steam Generator B pressure reached minimum 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.
Makeup Pump (ECCS HPI flow) restarted by operators t = 15 min.
RC Drain / Quench Tank rupture disk blows at 190 psig (setpoint 200 psig) due to continued discharge 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 BEGIf15 HEAT UP TPJdlSIEf1T - Hot leg temperature begins to rise to 620 degrees F (off scale within 14 minutes) and cold leg temperature drops io 150 degrees F.
OlPIwater) 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 1
leakage El n
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 h
electromatic relief valve opened j.'
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 d
1 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 410 263 N
IE Bulletin flo.79-05A Date: April 5, 1979 Page 3 of 3 TIME EVENT 1
t = 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Peak containment pressure of 4.5 psig t = 5 - 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 3
RCS pressure increased from 1250 psi to z
2100 psi Q
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 rtCS to attemf r initiation of D
RHR at 400 psi t = 8 - 9 hourc 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 /> w
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 1. cop A started, hot leg temperature decreases to 560 degrees F, and cold leg temperature increases to 400 degrees F.
indicating flow through steam generator Thereafter S/G "A" steaming to condensor Condensor vacuum re-established RCS cooled to about 280 degrees F.,
1000 psi flow (4/4)
High radiation in containment All core thermocouples less than 460 degrees F.
l' Using pressurizer vent valve with small makeup flow r
Slow cooldown RB pressure negative
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II d1 410 264 m
.c UNITED STATES til' CLEAR REGULATORY COP. MISSION OFFICC 0F Ill5PECTION AND Et'FORCEMEllT WASHIriGTON, DC 20555
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APRL. 21, 1979
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IE Bulletin 79-053 tlUCLEAR IhCIDiNT AT THREE. MILE ISLAMD - SUPPLEMENT
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Descriptien of Circumstances:.
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.5 Continued NRC evaluation of the nuclear incident at Three Mile Islan'd Unit 2 has identified measures-in addition to those discussed in IE
. L Bulletin 79-05 and 79-05A which should be acted upon by licensees with reactors designed by B&'A_ ~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 is to provide forced flow using' reactor coolant pumps.
A L..
It appears that natural circulation was not successfully achieved upon securing the reactor ccalant-pumps-during the first two hours of the Three Mile Island (TMI) No. 2 incident of March 28, 1919.
Initiation
~..
of natural circulation was inhibited by significant coolant voids, possibly aggravated by release of noncondensible gases, in the prirary
["
'.. circulation, the oparatur should ensure that the primary system is coolant system. To avoid this potential for interference with natural N
subcooled, and remains subcooled, before any attempt is made to establish
- JJ natural circulation.~
Natural circulation irr Babccck and Milcox reactor systems is enhanced by maintaining a relatively high water le. vel 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 1evel e cint to 50% on the operating range when all reactor coolant pumpe (RL, are secured.
- However, in unusual or abnormal situations, manual actions by the operator to increrse steam generator level will enhance natural circulation capability in anticipation of a passible loss of operation of the reactor coolant pumps.
As stated previously, forced flow of primary coolaat through the core is preferred to natural circulation.
5 f 'her means of reducing the possibility of void formtion ia the reactor h
toolant system are.
y d;
A.
Minimize the operation of the Power Operated Relief Valve (PORV) on i
the pressurizer and thereby reduce the possibility of pressure J
reduction by a blowdown through a PORV that was stuck open.
d
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IE Bulletin 79-053 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.
f This bulletin addresses, among other thirm.
ans to achi2ve these
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objectives.
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Actions To Be Taken by Licensees:
a 3 ~, -:..
d For all Babcock and Milcox pressurized water reactor facilities with an
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operat'ing license:
(Underlined sentences are modifications to,.and U
E '._.esspeesade, IE3-79-05A).
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1.
Develop procedur's and train operation personnel on methods of Z.. 5.0 establishing and maintaining natural circulation.
The procedures
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_ and training must include means of conitoring heat removal efficiency '
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by available plant. instrumentation.
The procedures must also.contain 3 % _2 a rmthod of assuring that the primary coolant system is subcooled by 7,[
at least 50*F before natural circulation is initiated.
i...
'c -l:ji c. -
In the event that these instruct. ions incorporate anticipatory fillino f.
of the OT5G prior to securing the reactor coolant pumps, a detailed W ?.,. analysis should be done to provide guidance as to the expected system response.
The instructions should include the following precautions:
2
.s
,. ~ "
a.
maintain pressurizer level sufficient to prevent loss of level indication in.the pressurizer; b.
assure availabil'ity of adequate capacity of pressurizer heators, for pressure control and maintain primary system pressure to n a,_
satisfy the subcooling criterica for natural circulation;
(
caintain pressure'- temperature envelope within Appendix G limits c.
for vescel integrity.
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 circulation core cooling mode.
9
...2.
Modify the acticns required in Item 4a and 4b of IE Bulletin 79-05A t
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to take into account vessel integrity considerations.
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"4.
F.eview the action directed by the operating procedures and 7
training instructions to ensure that:
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Operators do not override automatic actions of engineered a.
safety features, unless continued operation of engineered
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IE Bulletin 79-053 Apri1 21,1979 Page 3 of 4 f
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_ safety features will result in_ unsafe plant conditions.
For examcle, if continued operation or engineered safety features - -
would tnreaten reactor vessel intecrity tnen the HPI should be g
f secured as noted in O(2 Delow).
~
",h b.
Operatirrg procedures. currently, or are revised to$ specify that
-e.-e,* q
,e if the-higtr pressure injection.
3 actuated because of low pressure (HPI) system has been automatically -N
(,, b
-pf operation until either:.
condition, it must remain in-
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(1). Both, low pressure injection. (LPI) pumps are in operation
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and ficwing at a rate in excess of 1000 gpm each and.the' -
situation-has-been stable for 20 minutes, or s
,f-C (2)
The HPI system has been in operation for 20 minutes, and..
.. (m -
all. bot and cold leg temperatures are at least 50 degrees
-ic u below the saturation terperature for the existing.RC5
.-gi::
pressure.'. If 50 degrees subcooling cannot be maintained after HPI cutoff, the HPI shall be reacti
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vated.
The decree
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of subcoolino beycnd 50 degrees F and the 'ength of time HPI is in operation snall be limited by the pressure /
temperature consideraticas for the vessel in teari ty."
3.
Following detailed analysis, describe the modifications to design and
(
procedures-which you-have imples.nted to assure the reduction of the likelihood of automatic actuation of the pressurizer PORV during anticipated transients.
This analysis shall include consideration of a modification of the'high pressure scram setpoint and the PORV opening setpoint such that reactor scram will preclude opening of the PORY for the spectru:n of anticipated transients discussed by BaW in--Enclosure 1.
Changes developed by this analysis shall not result in increased frequency of pressuri2.er safety valve operation for these anticipated transients.
4.
Provide procedures end training to operating personnel for a prompt nanual trip the reactor for translents that result in a pressure increase in tne reactor coolant system.
These transients include:
'.. 1 a.
loss of main feedwater
~
i b.
s-5 Main Steam Isolation Valve closure c.
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- d. _.. Loss of offsite power
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Low OT5G 1evel L
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Tow pressurizer level.
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IE Bulletin 79-053 April 21,1979 Page 4 of 4 5.
Provide for NRC approval a design review and schedule for impleimntation of a safety grade autcoatic ant.icipatory reactor scram for loss of feed-water, turbine trip, or significant reduction in steam generator level.
7 5
6.
The actions required in item 12 of IE Bulletin 79-05A are modified as
(
follcws:
?
fh Revicw your prcept reporting proceduras for NRC notification to assure 5' d
%-f that NRC is notiffed within one hour of the tire the reactor is not in D
' * {.7'.
a controlled or expected conoition of operation.
2, J-_+ O an ocen continucus. comunication channel shal1 be established andFurther, at that t1 L
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- 73 s '.. raintained witn fac.
.?ei:F 7.
Procose chances. as recuired, to those technical specifications which.
must ce rediried as a result of your 1molemanting the above items.
['. _ Response schedule for BaW designed f'acilities:
For Items 1, 2, 4 and 6, all facilities with an operating license m,
a.
respond within 14 days of receipt of this Bulletin.
u.' '
~ b.
For Item 3, all facilities currently operating, respond within 24 h e.,rs.
All facilities with an operating license, not currently operating, respond before resuming operation. -
W.-
- c..For Items 5 and 7, all facilities with an operating license respond
- ~
in 30 days.
f '- Reports should be submitted to the Director of the appropriate NRC Region
- Office and a copy should be forwarded to the li.RC Office of Inspection and Enforcemnt, Division of Reactor Operaticas Inspection, Washington, D. C.
20555.
For all other pc12 reactors with an operating license or construction pemit, this Bulletin is for infomation purposes and no written response is required.
Igproved by GAO, B180225 [R0072); clearance expires 7/31/80.
was given under a blanket clearance specifically for identified generic 3
Approval problems.
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CONTIfl0ING PEVIEW OF THE SEQUENCE OF EVEtiTS LEADIftG TO TU
-2 U?f PAT!CH Z8,1979 SHCMS TIIAT ACTION Cart DE TAY.EN TO PROVIDE ASSURANCE THAT THF. PILOT-OPERATED RElsEF VALVE (PORY) fDUTITED UN Til
!WE A SIGIIIFICMG' PR03 ABILITY OF DCCURRING MIT DEGPAGE Ti!E SAFETT OF TTIE AFFECTED PLNITS HIT!) RESP THIS ACTION ?!uST-TFE NiTICIPATED TRA?isIEtrrs GF C0ilCEf0f ARE:TO i;0RML, UFSET OR M
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IHADVtsitNI CLOSURE OF MAIN STEA?! ISOLATION VALVES (?fSIV
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A IPJGER OF ALTER.%~TIVES VERE COMSIDERED Ill DEVELOPING Tile BELG'J.IliCLUDITG:
U,9 EESTRICTING REACTOR POWER TO A VALUE WiICH WOULO ASSURE rio ACRI
'i THE:PORY.
~ ~ PGIlliS RE?t4IHED AT-THEIMURREllT-VALUES.THE. REACTOR PRUTECT
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- 1a_T LO7ERING THE ffIGH PRES 50RE REACTOR TRIP SETPO To3URE NO ACTUATIOM OF THE FORT.
THE SETPOITIT F03 PORY T4TUATI0li RENAINED AT TilE LG'JERING TFE HIGH PRESSURE RE5CTOS TRIP SETPOINT N 07EPATITET PRESSURE (NiD TEMPERATURE) 0F THE REACTOR TO ACTUATIG't Aric TO PROVIDE AGEQUATE mRGIH TO ACCotet0DATE V OPEPATIfG PRESSURE.
THE SETFOINT FOR PORY ACTUATION RE. MAIMED AT ITS CURREliT VALUE.,
THIS ALTERNATIVE WOULD REOUCE HET ELECTRICAL OUTP 4.
AUPJ3TIris THk_HIGl PRE 5SURE TRIP NiG Tile FORY SETPOINTS TO AssunE r;D PGM ACTUATIO71 FOR THE CLASS OF RITICIPATED EVEtiTS OF CON PEESSURE OF THE PEACTOR REPAINED AT ITS CURRENT VALUE.
THE DESIGH
?ll AliALYSIS OF THE IFPACT OF THESE VARICUS ALTERNATIVES OF CONCEFJi HAS BEE?i COMPLETEDTG ASSLGIfiG THAT TH THE RESULT 5 5HOW THAT:
LO'ERING TFE HIGH PRESSURE REACTOR TRIP SETPOINT FRO?f
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FROVIDES TIIE REquIKED ASSURANCE.
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red.iCIfG TiriE PRD3ABILI.TT OF PORY MID A3HE CODE ACTUATIO.4 FOR OTilER IilCREASIrXI PRES 5URE TFIJiSIENTS.
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Er-dESER'1IbG PRESSURE RELIEF CAP 5 CITY.FOR ALL !
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EL1HIMATING THE POSSISTLITY 0F IlfrRODUCING UNREVIEUED M
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EN)UCNfG THE iINE AT E4ICH THE STENT SYSTEM HE 3
q THE EYENT ENERGENCY FEEC'<:ATER FLC*d WERE DELAYED L
h SGS. MART OF TdE IP.PNCT OF T)lE PROPOSED S$T TFJJISIE!US IS GIVES Iti TABLE-1, B21 FI/JfT3 ARE. CURRENT 7.Y CAPA5LE OF Rt IBACX TO.TS:t O LOAU.OR TRIP OF TrlE TURBINE.
THIS CAPABILITY REQUIRES ACTUATICfl OF THE PILO C'/ERATED RELIEF VALYES.
THE CAPABILITY INCREASES THE P.ELIABILITY OF pod' ER AFTER THESE TRRISIEliTS. SUPPLY TO THE SYSTEM BY RETURN REACTOR BE TR1PPED FOR THESE EVEllTS:THE ACTION PROPOS
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- The effect of changing the reactor coolant systera pressure trip setpoint upon peak pressurtzer pretsure is typified by the attached figure 1. which was developed by.
B&W for a loss of feedwater transient.
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A.TTICIPATED TFNISIENTS EXTPACT OF BatCQP)fJitLCATIO!i - RECEIVED BY llRC 4/20/79.
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N#ICIPATED R^5SIEIT5 WHIO! HAVE OCCURRED AT BLM FLNirs A'lD MIICil WOULD.
2 6- )ML7 ACTIVhTE PORY AT THE CURRE?fr SETP0trfr (2255 PSIC):
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TURSIEE TRIP '
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DiTICIPATED TRNiSIErf75.inlICH HAVE NOT OCCURRED AT DTN PLA!iIS }(
3 P' R03ABILITT E'iENTS) NiD HHICH UGULD riORR;LLY ACTUATE PORY AT THE CURPSTT St.uvINT (2255 PSIG):
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i UN:TEO 5i"ATES k.' I (f.,i' 1 NUCLEAR FtEGULATORY CGf.iMISSION
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ADVDORY CO:.NITTEE Ot; REAC Ort SportcuAno; n'.
., #y nuitttIGTON. D. c ie:,735 s,.s~
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>, April 20,.1979
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9 throrable victor cILinskv
- -. Actin r Chair =a.n-
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L U.. S. Up=1c2r Regclatory Ccc:nissicn
. Kashington, DC 20.555.
Daac Dr. Gilinsky:
1 bis letter is in respon== to yourn of April 18, 1979 tehich re,n.;e r.ed that the ACR.5 notify the 'Conniscioners 1:=a2diately if we believe any of our orcl rococ=wndation of April 17?should te acted ugun b:-fore our no:!: regularly 30hMtded meting at which ue could preparo a fonol let'ter.
Tne Cocaittee oiscussed this topic by conference telephona
(
call an' April 19 and o'.fers the following. c=wnts.
.\\
All of the recocandatf~ ors r;ude by the ACRS in its~ metirs vith tl e cor.siccioners on April 17/ 1979, ere generic in nature and cpply to all F,Gs.
J6ne vero intended to require i=.ediato, changes in operotirig pro-r:eduros or plant ndifications of operating IwRr.
Such changes should be cede on'ly after study of their effects on overall safety.
Such stud-ie.s sh:uld be made by the licensees and their suppliers or consultanrc imd by the MRC Staff Tne Com.ittpe helieves that th2se stu5ies should b2 bg.:n in the naar future on a tire scale that will not divert the
- NRC Staff or the industry representatives from their tacks reinting to the cc.oldpwn of Tntc-e fille Island Unit 2.
Rowever, the Cc:rer.i tteo be--
' lieves th'at it voeld be. pas.qible and desirchle to ini tf ate indiat'ely a nnvey of opcratirg procedures for achiedn3 natu:A1 circulatich,. in--
cludirG thc. Cf.de when off-'ite p3*-cr is lost, anS the role of the p'res s
curl 2.cr heaters in suc5 p'r~6 eduros.
mo 1 t its raa.ctIng en April 15 and 17,1979, the Co==lttee dir :dssed.trith s
~
' the ERC St.aff the r. utter of natural circulation for the Shree P.D o I _
12.4 Unit 2 plant.
ne ccer.ittee bellCvGs that this raatter is receiv-.
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Ing c reful atrcntion by the imC Staff and the licensee-N E.9 To ED3 for A.ppropriate Action-Di s trib uti.on :
Chm, Crars, PE, OGC, OCA, SEQ *, L "D2, OlA.
Rapifcxcd to E00. PA. E ; case.
79-1117..
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.t U.'JITED STATES E 'Nh d.n NUCLEAR REGULATORY cot.if.ilSSION
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ADVISO;lY CO!/.!/.LTTCE O!J REACTOR SAFCGUARDS x
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3pril 18,1979
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Chair::.an Herdelc 2'
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, F. Fraley, E:cecutive Director Advisory Cc;=rittee on Reactor Safeguards
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Attached for your infor:ution and use is a copy of :the recomenda--
tions of the Advisory Co.7.,. ittee on Reactor Safeguards which ware orally presented to and discussed with you on April 17,1979 re--
gardi.,g the recent accident at the Gree Mile Island Nuclear Sta-tion Unit 2.
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R. ; F. Fralcy Executive Director
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Attachment:
Recc=endations of the Unc Advisory Co.r.vittee on Reactor Safeguards Re. the 3/28/79 Accident
- at We Three Mile Island Nuclear Station Unit 2
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l@ril 17, 1979 iPM.TCATICNS O? 'nic NUCLEAR RS3UENIORY CCFaISSION ADVISORY CMMI'1TEC Ci RFWICR SAFZ:3UARCS TO3ARDING THE & ARCH 28, 1979 ACCIDENT AT
'IMO 'nL9J.E MILE ISERO WCI. EAR STATION UNIT 2 2-y}5 presented orally to, arr3 discussed with, the NRC n
Cc=mincioners during the ACRS-Commi ssioners Meeting ia on April 17, 1979:- hwshington, D. C.
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. +. 4 L %. 6 Ltural circulatiqn ir an important rode of reactor cooling, both as iM
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a planned process and as a process that r:ay be uned under abnormal
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Th: Cc.mittee believes that greater understanding of k
'. 'this rede of cooliref. ILrequired and that detailed analyses.should.
- gf, b2 dev-loped by licensees or-their suppliers.
The analyses should be supported, as necessary, by experiment.
Procedures should bo ' de-
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veloped for initiating natural circulation in a safe manner and for providing the operator with assurance that circulation has, in fact, best established. This may require. Installation of instrumentation to 7 -2
, maasure or indicate flow at low water velocity.
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"Iha use of natural circulation for decay heat removal following a loss f
of offsite power sources requires the maintenance of a suitablo over-pressure en the reactor coolant systen.
This overpressure; rnay be assured by placing the pressurizer heaters on a qualified onsite Pow-r source with a suitable arrangement of heaters and power distri-bution to provide-redundant: capability..
Fresently operating F,G plants should b2 surveyed exp-M. itiously to detennine whether such arrar.genents can be provided, to assure this aspect of natural circula-tion ca,cability.
'Ihe plant operator should be adequately infon ed at all tir as con-cerning the conditions of reactor coolant system operation thich might affect the capability to place the system in the natural circu-lation pode of operation or to sustain such a r:>ade.
Of particular impartance is that information 5/nich-might indicate that the reactor coolant system is approa31og the saturation pressure corresponding' to the core exit temperature.
This impending loss of system over-pressure will signal to the opercror a possible loss of natural circulation capability.
Such a warning may be derived frca pressur-
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iner pressure instruments.and hot leg temperatures in conjunction with d
conventional - stcar:t tables..
A suitable display of this information d
should be provided to the plant operator at all times.
In addition, consideration should be given to the use of the flow exit tempera-
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tures frees the fuel subassemblies, cere available, as an additional 2
indication of natural circulation.
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We exit temperature of coolant frca the core is currently measured by therroccupies in r.any nas to ' determine core performance.
%c Cc,mittee reccc.. ands that these temp 2rature measurements, as currently.
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available,.be used to guide the operator concerning core status.
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G range of the information displayed and recorded should include tha d
full capability of ths-themoccuples.
It is also reco.r.rcrx3ed that other existing instru=antation be examined for its possible use in p
r"+tlng operating action during a. transient.
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.TT - Ihe. ACES reccc-. ends thdt operating power reactors be given priority p
jwi'th regard to the-definition and impler. entation of instrumentation g
- '. #.- which provides additional information to help dicgnose and follow the course of a serious accident.
This should include improved sampli.ng precedures under accident corditions ard techniques to help previde
, Im.croved guidance to offsite authorities, should this be needed
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.' _..comittee recommends that a phased implementation approach b e - e.-
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played so that technicues can be adopted shortly after they are 7.-
' jtxiged to be appropriate.
%e'Acas recce= ends that a high priority be placed on the develop-ent and LT.plcr.entation of safety resca:ch on the behavior of 1Ight water reactors during an :alcus transients.
'Ihe !!P.C may find it appropriate to develop a capability to simulate a wide rar.ge of postulated tran-y sient and accident conditions in ordar to gain increased insight into measurea-L*11ch can be taken to improve reactor safety.
'Ihe ACRS utshes to reiterate its previous reco.=2ndations that a high priority 7_
be given to res(arch to improve reactor safety.
Consideration should be giten to the desirabi1Ity or ' additional
-' equip. ant status monitoring on.various engineered safeguards. features ar.d their supporting services to help assure their availability at all times.
'Ibe ACRS is continuing its review of the implications of this accident and ho;a 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 Hydraulic 2/2/79 All power reactor Shock and Sway Arrestor facilities with an Operating License (CL) er a Construc-tion 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 Operating License (OL) or a Construc-tion Permit (CP) 79-04 Degradation of Engineered 2/16/79 All pawer reactor Safety Features facilities with an Operating License (OL) or a Construc-tion Permit (CP) 79-05 Use of Improper Matcrials 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 Permit (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 4!0 278
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 Pe rmit (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 react ors, 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 Suppor-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 1 n 7 7 r,
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