ML19225B100
| ML19225B100 | |
| Person / Time | |
|---|---|
| Site: | 05000192 |
| Issue date: | 06/22/1979 |
| From: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| To: | Gloyna E TEXAS, UNIV. OF, AUSTIN, TX |
| References | |
| NUDOCS 7907230512 | |
| Download: ML19225B100 (1) | |
Text
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'o UNITED STATES
-f4 SIC NUCLEAR REGULATORY CGMMISSION f 'li
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':g, June 22, 1979 la Re ply Re f e r To :
Docket No. 50-192 The University of Texas ATTN:
Dr. E. F. Cloyna, Dean College of Engineering Austin, Texas 78712 Centlemen:
The enclosed Inform cion Notice No. 79-16 is forvarded to you for in-forcation.
No specific action is requested and no written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely,
/
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Karl V. Seyfrit f Director Ecclosure:
1.
IE Information Notice 79-16 2.
List of IE Inforcation Notices Issued in 1979 s.
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UNITED STATES NUCLEAR REGULATORY COMMISSION 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 ISLul)
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, contact the Director of the appropriate NRC Regional Of fice.
Enclosures:
1.
IE Bulletin No. 79-05 with Enclosures 2.
IE Bulletin No.79-05A with Enclosures 3.
IE Bulletin No.79-05B 70,070'30 410 157
(
UlITED STATES f;UCLEAR REGULATORY COMMISSI0il 0FFICE OF INSPECTIO1 AfD Ef1FORCEMENT WASHIt!GTO:1, D.C.
? 555 April 1, 1979 1
fiUCLEAR IfCIDEtlT AT THREE MILE ISLAND n+
Description of Circumstances.
y On March 28, 1979 the Three Mile Island t!uclear Power Plant, Unit 2 b
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 Hilcox reactors.
This bulletin is provided to inform you of the nuclear incident and to request certain actions.
Actions To Be Taken By Licensees (Although the specific causes have not been determined for individual sequences in the Three Mile Island event, some of the following may have contributed.)
For all Babcock and Wi.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 f?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 fram expected performance occurred.
If any significant deviations are 4
found, provide the details and an analysis of the significance and y
any corrective actions taken.
This material may be identified by b
reference if previously submitted to the itRC.
d 5
3.
Review the actions required by your operating procedures for coping 5
with transients.
The items that should be addressed include:
b EL 410 158 we*
w-ww -
o.
IE Bulletin No. 79-05 April 1,1979 f
Page 2 of 3 Recognition of the possibility of forming voids in the primary a.
ccolant system large enough to compromise the core cooling capability.
3 b.
Operator action required to prevent the for.mation of such voids.
a l
Operator action reouired to ensure continued core cooling in c.
the event that such voids are formed.
'j I
4.
..eview the actions requested by the operating procedures and the L
training instructions to assure that operators do not override automatic actions of engineered safety features without sufficient cause for doing so.
5.
Review all safety related valve positions and positioning require-ments to assure that engineered safety features and related equip-ment such as the auxiliary feedwater system, can perform their intended functions.
Also review related procedures, such as tho:e for maintenance and testing, to acsure 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 radioactive liquids and gases will not occur inadvertently.
~
In particular assure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.. List all such systems and indicate:
Whether interlocks exist to prevent transfer when high a.
radiation indication exists and, b.
Whether such systems are isolated by the containment isolation signal.
7.
Review your prompt reporting procedures for NRC notification to 3
assure very early notification of serious events.
[j s
The detailed results of these reviews shall be submitted within ten 9
(10) days of the receipt of this Bulletin.
b b1 4i0 159 m
IE Bulletin To. 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,
[".3 b'ashington, D.C.
20555.
For all other operating reactors or reactors under construction, this j
Bulletin is for information purposes and no report is requested.
J Approved by GAO, B180225 (R0072); clearance expires 7-31-80.
Approval N
was given under a blanket clearance specifically for identified generic y
problems.
Enclosures:
1.
Preliminary Notifications Three Mile Island -
PHO-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 r
s:
0
- s a
i A
N 410 160 Page 1 of 3 s
EVALUATION OF FEEDWATER TRANSIENT y
i A loss of offsite power occurred at Davis-Besse on November 29, 1977, which resulted in shrinkage of tha primary coolant volume to the degree s
that pressurizer level indication veas lost.
A recommendation to convey this information to certain hearing boards resulted in the attached 4
discussion and evaluation of the event.
This discussion includes a g
review of a loss of feedwater safety analysis assuming forced flow,
$2 which predicts dispersed primary system voiding, but no loss of core cooling.
During the Three Mile Island event, however, the forced flow appears to have been terminated during the transient.
Attachnent:
Discussion and Evaluation of Davis-Besse Transients
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1.
4iU i 6i op mm
EXCERPT FROM MD'0?J.NDUM ENTITLED " CONVEYING NEW INFOR3L\\ TION TO LICENSING BOARDS - DAVIS-BESSE (TNITS 2 6 3 AND MIDU.ND UNITS 16 2",
DATED JANUARY 3, 1979, FROM J.S. CRESWELI. TO J.F. STREETER.
s f
3.
Inspection and Enf orecaent Report 50-346/78--06 doct:acnted that pressurizer level had gone of f scale for approxinately five i
minutes during the Novemb'er 29, 1977 loss of of f site povar event.
y There arc scoe indications that.other B&W plants may have prob-c)
Ic=s maintaining pressuriser level indications during transients.
)$
In addition, under certain conditions such as loss of f eeduater 100% power with the reactor coolant pt:aps running the pres-at suricer may void completely.
A special analysis has been per-9 formed concerning this event.
This analysis is attached as.
Because of pressurifer level. maintenance prob-lems the sizing of the pressuri:ce nay require further review.
s Also noted during the event was the fact that Tcold vent off-scale (less than 520cF).
In addition, it wa s noted that the takaup fica monitoring is limited to less than 160 gpa and that makeup flow may be substantially greater than this value.
Tnis infornation should be exacined in light of the require-nents of CDC 13.
DISCUSSION AND EVALUATION Thie event at Davis Besse which resulted in loss of nrdssurl=cr level indication has been reviewed by URE and the conclusion was reached that no unreviewed safety question existed.
The pressurizer, together with the reactor coolant makeup system, ie designed to maintain the primary syste:a pressure and water level within their operational limits only durin3 normal operating conditions.
Cooldown transients, such as loss of of fsite power and loss of feed-water, soaetites result in pri=ary 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 without compromising the safety of the reactor.
The principal concern caused by such t casients is that they might cause voiding 16 the pricary coolant systea that would lead to loss of ability to ade-3
~
quately cool the reactor core.
The safety evaluction of the loss of f
of fsite pcuer transient shows that, thcugh level indication is lost, so=e water recains in the pressurizer and the pressure does not decrease A
below about 1600 psi.
In order for voiding to occur, the pressure cust decrease below the saturation pressure corresponding to the systea x
tengerature.
1600 psi is the saturation pressure corresponding to 605 F, which is also the caxi=un allowable core outlet temperature.
5 Voiding in the pri=ary systcc (excepting the pressuricer) is precluded 9
in this case, since pressure does not decrease to saturation.
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Settion 3..
~
The safety analysis for more severe cooldoen transients, such as the loss of feedwater event, indicates that the water voluae could decrease to less than the syste= volume exclusive of the pressurizer.
During such an event, the emptying of the pressurirer vould be followed by 3
a pressure reduction below the saturation point and the formation of small voids throushout much of the primary system.
This vould not g
~
result in the loss of core cooling because the voids would be dispersed j
over a larga volume and forced flow would prevent them from coalescing 4
sufficiently to prevent core cooling.
The high pressure coolant injection pumps are started aucc2atically uhun the pricacy pressure i
decreases below 1600 psi.
f, The re fo re, any pressure reduction which is sufficient to allow voiding vill also result in water injection which vill rapidly restore the primary water to.norcal icvels.
For these reasons, we believe that the inability of the pressuri:cr and normal coolant nckcup systes to control some transients does not provide a basis f or requiring more capscity in these systens.
General. Design Criterion 13 of Appendix A co 10 CFR 50 requires instru=2ntation to monitor variables over their anticipated ranges for " anticipated operational occurrences".
Such occurrences are specif'ically defined to include loss of all of fsite power.
The fact that T cold goes'off scale at 520 F is not considered to be a deviation from this requirement because this indicator is backed up by wide range temperature indication that extends to a low limit of 50 F.
Neither do ve con ~ sider the takeup ficw nonitoring to deviate since the amount of cakeup flow in excess of 160 gp2 does not a a significant factor in the course of these occurrences. ppear to be The loss of pressuricer water level indication could be considered to deviate from CDC 13, because this level indication provides th2 principal ceans of determining the primary coolant inve n to ry.
However, provision of a level indication that would cover all anticipated occurrences may not be practical.
As discusned above, the loss of feedvater event can lead to a comentary condition wherein no meaningful level exists, because the entire priuary system contains a.stea= vater mixture.
It should be noted that the introduction to Appendix A (last paragraph) recognizes that fulfill:ent of some of the criteria may not always be t
y approp ria te.
This introduction also states that departures from the Criteria must he identified and justified.
The discuccion of CDC 13 g
in the Davis Besse FSAR lists the water level instrumentation, but p
does not nention the possibility of loss of water level indication 5M during transients.
This apparent osiscion in the safety analysis 3
vill be subjected to further review.
J s
a._
410
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i UlITED STATES
(
NUCLEAR REGULATORY C0:01ISSIO:1 0FFICE OF IllSPECTIO 1 Atl0 EtlFORCEf tEflT WASHIllGT0:1, DC 20555 IE Bulletin flo.79-05A f~
Date: April 5,1979
?
Page 1 of 5 flVCLEAT IflCIDEllT AT THREE MILE ISLAt 0 - SUPPLEMEllT
}4 1
Description of Circumstances:
Preliminary information received by the flRC since issuance of IE Bulletin 79-05 on April 1,1979, has identified six potential human, design and mechanical failures which r asulted 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 3ulletin and transmit a preliminary chronology of the TMI accident through the first 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> (Enclosure 1).
1.
At the time of the initiating event, loss of feedwater, both of the auxiliary feedwater trains were valved out-of-service.
2.
The presrurizer 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 system.
The pressurizer level indication apparently led the operators to prematurely tenninate high pressure injection flow, 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 D
through the auxiliary building ventilation system and filters was
$]
the principal source of the offsite release of radioactive noble gases.
(:
v 5.
Subsequently, the high pressure injection system was intermittently 3
operated attempting to control primary coolant inventory losses j
through the electromatic relief valve, apparently based on y
pressurizer level indication.
Due to the presence of steam and/or noncondensible voids elsewhere in the reactor coolant system, I this led to a further reduction in primary coolant inventory.
4)d bk
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April 5,1979 Page 2 of 5 6.
Tripping of reactor coolaat pumps during the course of the transient, T
to protect against pum, d' age due to pump vibration, led to fuel 1
damage since voids in the reactor coolant system prevented natural circulation.
Actions To Be Taken by Licensees:
j j
For all Babcock and Wilcox pressurized water reactor facilities with an e
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 Be, P tin 79-05.)
Review any transients similar to the Davis Besse event (Enclosure 2 of IE Bulletin 79-05) and any others which contain similar elements from the enclosed chronology (Enclosure 1) which have occurred at your facility (ies).
If any significant deviations from expected performance are identified in your review, provide details and an analysis of the safety significance together with a description of any corrective actions taken.
Reference may be made to previous information provided to the NRC, 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:
7 a.
Recognition of the possibility of forming voids in the primary 1
coolant system large enough to compromise the core cooling capability, especially natural circulation capability.
b.
Operator action required to prevent the formation of such voids.
c.
Operator action required to enhance core cooling in the event 5.
such voids are formed.
410 165 6
,e
(
IE Bulletin No.79-05A Date: April 5, 1979 s
Dage 3 of 5 4.
(This item clarifies and expands upon item 4. of IE Bulletin 79-05.)
r h
Review the actions directed by the operating procedures and training h
instructions to ensure that:
a.
Operators do not override automatic actions of engineered safety features.
f
(
b.
Operating procedures currently, or are revised t specify p
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 degrees below the saturation temperature for the
,r existing RCS pressure.
If 50 degree subcooling cannot
(
be maintained afte-HPI cutoff, the HPI shall be reactivated.
Operating procedures currently, or are revised to, specify c.
that in the event of HPI initiation, with reactor coolant pumps (RCP) operating, at least one RCP per loop shall remain opera ting.
d.
Operators are provided additional 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.)
Verify that emergency feedwater valves are in the open position in N
accordance with item 8 below.
Also, review cll safety-related d
valve positions and positioning requirements to assure that 9
valves are positioned (open or closed) in a manner to ensure the 0
proper operation of engineered safety features.
Also review related procedures, such as those for maintenance and testing,
[3 to ensure that such valves are returned to their correct positions 4
following necessary manipulations.
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Date:
April 5, 1979 Page 4 of 5 w
6.
Review the contai:. ment isolation initiation design and procedures,
)#
and prepare and implement all changes necessary to cause cor. ainment isolation of all lines whose isolation does not degrade core cooling b
capability upon automatic initiation of safety injection, y
i 7.
For manual valves or manually-operated motor-driven val.as which could defeat or compromise the flow of auxiliary feedwater to the steam generators, prepare and implement p"ocedures 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 procedares which assure that two independent steam generator auxiliary feedwater flow paths, each with 100% flow capacity, are operable at any time when heat removal from the primary system is thro"gh the steam generators.
When two inde-pendent 1003 capacity flow vths are not available, the capacity shall be restored within 72 r., es 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 cooling 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 r.uximum 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.
'n In particular, ensure that such an occurrence would not be caused
?
by the rosetting of engineered safety features instrumentation.
List a
all such systems and indicate:
j a.
Whether interlocks exist to prevent transfer when high radiation 0
indication exists, and J
b.
Whether such systems are isolated by the containment isolation 1
signal.
4iU 167
('
Date:
April S,1979 Page 5 of 5 10.
Review and modify as necessary your maintenance and test procedures to ensure that they require:
Ol?
Verification, by inspection, of the operability of redundant a.
g safety-related systems prior to the rer. oval of any safety-related system from service.
g M
b.
Verification of the operability of all safety-related systems 1
when they are returned to service following maintenance or testing.
L A means of notifying involved reac.:or operating personnel c.
whenever a s' aty-related system is removed from and returned to service.
- 11. All operating and maintenance personnel should be made aware of the extreme suriousness and consequences of the simultaneous blocking i
of both auxiliar.y feedwater trains at the Three ftile Island Unit 2 I
plant and othe. actions taken during the early phases of the accident.
12.
Review your prnpt reporting procedures for f;RC not.Y ' :ation to assure very early notification of serious events.
For Babcock and Wilcox pressurized water reactor facilities with an x
operating license, respord to Items 1, 2, 3, 4.a and 5 by April 11, 1979.
Since these items are s..stantially the same as those specified in IE Bulletin G-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.
3 P,eports should be submitted to the Director of the appropriate f!RC Regional Office and a copy should be forwarded to the fiRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, DC 20555.
For all other reactors with an operating license or construction pennit, this Bulletin is for information purposes and no written response is required.
Approved by GAO, B 180225 (R0072); clearance expircs 7-31-80.
Approval 9
was given under a blanket clearance specifically for identified generic
]9 probl ems.
Enclosures:
y 5,
1.
Preliminary Chronology of TriI-2 3/38/79 Accident Until Core Cooling Restored.
g-2.
List of IE Bulletins issued in last 12 months.
t) ) Q
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Date: April 5, 1979 Page 1 of 3 T
PRELIf4IflARY CHRONOLOGY OF TMI-2 3/28/79 ACCIDEflT.
[
UNTIL CORE C00 lit!G RESTORED
~
f TIME (Approximate)
EVEtiT l
4 about 4 Ali Loss of Condensate Pump b
~
(t = 0)
Loss of Feedwater Turbine Trip t = 3-6 sec.
Electromatic relief valve opns {_2255 psi) to relieve pressure in RCS t = 9-12 sec.
Reactor trip on high RCS pressure (2355 psi) t = 12-15 sec.
RCS pressure decays to 2205 psi (relief valve should have closed) t = 15 sec.
RCS hot leg temperature peaks at 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 rapidly t = 1 min.
Steam Generators A and B secondary level very low - <irying out over next couple of minutes.
~1 t = 2 min.
ECCS initiatign (HPI) at 1600 psi Il bi t = 4 - 11 min.
Presso m c lev _1 off scale - high - one 4,
'wy tripped at about 4 min.
1 HPI 3C M-pump tripped at about
?
10.
t = 6 min.
RCS flashes as ;ressu're bottoms out at 1350 psig Hot leg temperature of 584 degrees F).
'. : 9
. _ _ _. =.
April 5, 1979 Page 2 of 3 TIf!E EVENT t = 7 min., 30 sec.
Reactor building sump. pump came on.
I d
t = 8 min.
Auxiliary feedwater flow is initiated j
by opening closed valves g
e, t = 8 min. 18 sec.
Steam Generator B pressure reached minimum 1
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 BEGINS HEAT UP TRANSIENT - Hot leg temperature begins to rise to 620 degrees F (.off scale within 14 minutes) and cold leg temperature drops to 150 degrees F.
(HPI water) t = 2.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Electromatic relief valve isolated by operator after S.G.-B isolated to prevent
(
leakage y
I3 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 y
electromatic relief valve opened Y
A'
?
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 3
1-t = 3.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> I*
RC drain tank pressure spike of 11 psi -
RCS pressure 1750; containment pressure
(
increases from 1 to 3 psig 3
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I I
.m
IE Bulletin t'o.79-05A Date: April 5,1979 Page 3 of 3 TIME EVENT w
]
- 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 /> RCS pressure increased from 1250 psi to 2
2100 psi S
u t = 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Operator opens electromatic relief valve to depressurize RCS to attempt initiation of k
RHR at 400 psi 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 ctopped af ter 500 gal. of NaOH injected (about 2 'ninutes 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, hat 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 thennoccuples less than 460 degrees F.
. ]
Using pressurizer vent valve with small 1
makeup flow h
Slow cooldown G
RB pressure negative
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r UNITED STATES NUCLEAR REGULATORY CO?. MISSION OFFICE OF INSPECTION AND ENFORCEMENT b'ASHIriGTON, DC 20555 APRIL 21, 1979
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7 IE Bulletin 79-053 fj i
NUCLEAR INCIDENT AT THREE. MILE ISLAND - SUPPLEMENT g
De3criptien of Circumstances:.
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.v Id Continued NRC evaluation of the nuclear incident at Three Mile Islan'd Y
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&L. ' As discussed in Item 4.c. of Actions to be taken by Licensees in IEB;79-05A, the preferred mode of core cooling' following a transient or accident is to provide forced flow using' reactor coolant pumps.
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It appears that natural circulation was not successfully achieved upon securing the raattomcciantpcmps%cing the first two hours of the i,
Three Mile Island (TMI) 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 coolant system. To avoid this potential for interference with natural
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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 Wilcox reactor systems is enhanced by maintaining a relatively high water level on the secondary side of the once through steam generators (OTSG).
It is also promoted by injection of auxiliary feedwater at the upper nozzles in the OTSGs. -The integrated Control Systen automatically sets the OTSG level setpoint. to 50% on the operating range when all reactor coolant pumps (RCP) are secured.
- However, in unusual or abnornal situations, manual actions by the operator to increase steam generator level will enhance natural circulation capability in anticipation of a possible loss of operation of the reactor coolant pumps.
As stated previously, forced flow of primary coolant through the core is preferred to natural circulation.
g Other means of reducing the possibility of void formation in the reactor d
coolant system are:
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A.
Minimize the operation of the Power Operated Relief Valve (PORV) on j
the pressurizer and thereby reduce the possibility of pressure j) reduction by a blowdown through a PORV that was stuck open.
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IE Bulletin 79-053 April 21, 1979 Page 2 of 4 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. a This bulletin addresses, among other things, the means to achieve these i objectivas. 1 ~. %_ j Actions To Ba Taken by Licensees: For all Babcock and Hilcox pressurized water reactor facilities with an ~ ,i ~ - ...w. -- P I operating license: =.._. (Underlined sentences are modifications to,.and k .._W'._. r ssparsade, IEB-79-05A).
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d '.I .l. Develop procedures and train operation personnel on methods of -..s.i establishing and maintaining natural circulation. The procedures -W..u._ and training must include means of monitoring heat removal efficiency T-. 9,% f - by available plant. instrumentation. The procedures must also contain c a method of assuring that the primary coolant system is subcooled by " ' '. ', ~ ' at least SD"F before natural circulation is initiated.
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In the event that these instructicns incorporate anticipatory fillino -. ;....' 'f of the OTSG prior to securing the reactor coolant pumps, a detailed ' R"Y.,- analysis should be done to provide guidance as to the expected system ~~ The instructions should include the following precautions: response. 1. -~n m a. maintain pressurizer level s'ufficient 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 .7,, satisfy the subcooling criterion for natural circulation; caintain pressure '- temperature envelope within Appendix G limits c. for vessel integrit3 4 - 's, 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. y =7 . 2. Modify the actions required in Item 4a and 45 of IE Bulletin 79-05A l to take into acccunt vessel integrity considerations.
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"4. Review the action directed by the operating procedures and 5 training instructions to ensure that: j Operators do not override automatic actions of engineered O a. safety features, unless continued operation of engineered ( ,'e m m.A.. !aC - - e ..mm.e- ^ e m..
i IE Bulletin 79-058 April 21,1979 Page 3 of 4 f _ safety features will result in unsafe plant conditions. For example. if continued operation or engineered safety features would tnreaten reactor vessel 7 5 integrity tnen the HPI should be h t _ secured (as noted in D 2 below
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Operating procedures. currently, or are revised to, specify that +-@ N -? if the-higtr ;r. essure injection. 5 actuated because of low pressure (HPI) system has been automatically - E, ', c N .f* _' operation unt-il either:. condition, it must remain in-3, , ; y(jf, C - m (1) Both 1. low pressure injection. (LPI) pumps are in operation ~. L . N3. Jp._ and ficwing rt a rate in excess of 1000 gpm each and.the. situation har been stable for 20 minutes, or
- -y' (2)
TM HPI system has been in operation for 20 minutes, and. -A-MT9. ; all. hot and cold leg temperatures are at least 50 degrees below the-saturation temperature for the existing.RCS ' ". u.. pressu re. - If 50 degrees subcooling cannot be maintained '. g,;: after HPI cutoff, the HPI shall be reactivated. p..3-The decree of subccolino-beyond 50 degrees F and the length of time ..M c- -- HPI is in cocration snall be limited by the pressure / _temcerature considerations for the vessel in teari ty. " 3. Following detailed analysis, describe the modifications to design and procedures-which you-have implemented to assure the reduction of the likelihood of automatic actuation of the pressurizer PORV during anticipated transients. This analysis shall include consideration of a imdification of the'high pressure scram setpoint and the PORV opening setpoint such that reactor scram will preclude opening of the PORV for the spectrua of anticipated transients discussed by ban irrEnclosure 1. Changes developed by this analysis shall not result in increased frequency of pressurizer safety valve operation for these anticipated-transients. 4. Provide procedures and training to operating personnel for a prompt manual trip of the reactor for transients that result in a pressure increase in the reactor coolant system. These transients include: + a. loss of main feedwater ~ ~.. ~ u b. turbine trip 3.' y' y T-Main Steam Isolation Valve closure d c. ~ -- %. ] T
- d... Loss of offsite power U
e. Low OT5G level L g f. Tow pressurizer level. i _.Q4-r-4 g e . =. ::- :2:.. ;:: ~? A - - ~ ' ~ ' ~~ ~ ~ ~ ' ' ~ ~ ~
IE Bulletin 79-05a April 21,1979 Page 4 of 4 5. Provide for NRC approval a design review and schedule for impleirentation of a safety grade autcmatic anticipatory reactor scram for loss of feed-i. water, turbine trip, or significant reduction in steam generator level. ] 6. The actions required in item 12 of IE Bulletin 79-05A are modiffed as o follows: s$& that NRC is notified within one hour of the time the re f i~ G, ( 3-T
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jZ; a controlled or expected condition of coeration. 1 ' Q@E t ' -cM an open continucus. comunication channel shall be establistied andFurther, at that t 1. s '. traintained witn NRC. ?.=SD jiila ~ 7. precose chances. as reouired, to those technical rDecifications which, '. ~. r rust be modified as a result of your 1molemengry the above items. E- . Response schedule for BaW designed facilities: For Items 1, 2, 4 and 6, all facilities with an operating license 7 _ _._ a. respond within 14 days of receipt of this Bulletin. ' ~ '. b. For Item 3, all facilities currently operating, respond withirr 24 hours. operating, respond before resuming operation.All facilities with .For Items 5 and 7, a'll facilities with an operating license respond c. ~ ~ in 3D days. Officu and a copy should be forwarded to the NRC Offic r.nforcement, Division of Reactor Operations Inspection, Washington 23555. , D. C. Fo. all other pcwer reactors with an operating license or construction periit, this Bulletin is for information purposes and no written response is required. Approved by GAG, B180225 [R0072); clearance expires 7/31/80. c Approval 3 was given under a blanket clearance specifically for identified generic q problers. y 3' t! a ~ __. ~.. ~ ~
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~ EXTRACT OF BfM COV.UNICATION - RECEIVED BY ilRC Enclosure iMTRODUCT-TOM 0 Page 1 of 4 [ C0hTIfl0ING PiVIE'd 0F THE SEQUENCE OF EVENTS LEADING TO T11E \\ -2 Ott MARCH Z8,,1979 SHG4S TitAT ACTIOri Cart DE TAKEN TO PHOVIDE ASSURAtu,E THAT TliE PILUT-OPERATED RELIEF YALVE (FORY) toutlTED ON Tile PRE FLATfTS t!Ill NGT BE ACTUATED BY A7iTICIPATED Trail 51EMIS WilICi FAYE A SIGi/IFICAUT PFG3ABILIT'l UF OCCURRING _ IN T1tESE PUViTS. !;UT DEGPADE THE SAFETT OF THE AFFECTED PL7JIT5 HITH RESPErr TO TilEIR RESP THIS ACTIOTI ?!UST" To tiORmL; UFSET OR ACCICENT COMDITIor:S MOR LEAD TO UriREVI< UED SAFETY COUCE TFl A.TTICIPATED TRN1SIENTS OF COHCER?i ARE: i l.0'SS bi~ EXTERTtAl-ELECTRICAL LOA 0 W r. 2 . TURBINE TRIP.. .-N
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LOSS OF FAIN FEEENATER d ? f4 LOSS OF CONCENSER VACULW ZMTE 5.. IHADVERTENI CLOSURE OF FjIn STFRt ISOLATION VALVES (MSIV) - .t =:2h3 ; ' A fiUxaER OF ALTERHATIVES VERECONSIDERED Irl DEVELOPING SELO'J_.IfiCLUDIliG: 't.;M MISTRICTING NEACTOR POWER TO A VALUE WICH MOULD ASSURE NO ACTUA TV TIE PORY. . _.. POIIiTS REPAINED AT-THEIR-CURRENT-VALUES.THE. REACTOR Phw EC ~..: 2_. LOWERING THE ffIGH TRES50RE REACTOR TRIP SETPOINT TO A V 'AS5URE NO ACTUATION OF THE PORT. THE DESICli PRESSURE OF THE REACTOR AUD T,'F SETPOINT FOR PORY ACTUATIOil PENAINED AT THElO CURREfiT VALUES. ~ LG',ERING THE HIGH PRESSURE RE5CTOR TRIP SETPOINT AND ADJUSTIffG TH OPEPATIHT PRESSURE (AND TEMPEPATURE) 0F THE REACTOR TO AS .~
- ACTUATILTl AND TO PROVIDE ADEQUATE IMRGIN TO ACC05ciODATE V
. OPEPATIftG PRESSURE,. THE SETFOIflT FOR FORY ACTUATION RErmINED AT ITS . CURRElfT VALUE. THIS ALTERNATIVE WOULO REQUCE liET ELECTRICAL OUTPUT 4. AGI 5TIP;3 THEHIbl PRE 5SURE TRIP ANO THE PORY SETPOINTS TO A55URE tio PG'S ACTUATIO71 FOR THE cud 5 OF AtiTICIPATED EVEtlTS OF CONCERM THE UES1 Cal PF255URE OF THE REACTOR FlMINED AT IT5 CURRENT VALUE FJI NiALYSIS OF THE IFPACT OF THESE VARICUS ALTERNATIVES Ari TO A55LGIING THAT THE PORY t4ILL TKTr ACTUATE FOR THE CLASS OF CONCEFil HAS BEE?f COMPLETED,. THE RESULTS 5HOW THAT: ~ LO'ERItts TITE HIGH PRESSURE REACTOR TRIP SETPOINT FRO?t .. 3 ~.' 1, 2555 PSIG TO 2390 PSIG y
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- P.AISING THE SETFOINT FOR THE PILOT OPERATED RELIEF VALVE kl^
"h ,'FRC't 225S PSIG TO 2450 PSIG FD0VIDES TliE REQUIRED ASSURANCE.THIS ACTION HAS THE FURTHER AUVAtlTAGES OF: l.-=.. a^& & 7=g = =mp-r. ., =.=... F. .;.;.L 4_.: 2. ~ ~~.- - W @Z TJMi. M 5 MW-f H " ~.T R ... Z T.T.. ~ ~~ ~ = 'f3E5?isc T 1 ! J i,0
EXTPACT OF B&W ComuNICATION - RECEIVED BY HRC 4/20j79 Page 2 of < 1. RECUCIMs T E FRD3ABILITT OF PORl/ mig A5HE CODE PRESSURIZER SAFETY V ACTtnTIO4 FOR OT11ER UICREASItXI PRES 5URE TRAHSIEriFS. 7 ErTSE50/IbS P5E550RE RELIEF C PhCITY -FOR ALL HIGd P . G jj 3 h ELUfU!ATIirei THE PossIa1LITY OF IffTRODUCIffG UrlREVIEUED S i' 4.: NEDucNi9 THE i!E AT E4ICH THE STENT SYSTEM IIEAT SIUr. ROU J ~ 3 THE EYEhT EMERGENCY FEEI7 DATER FLC4 WERE DELAYED-L A $L? GARY OF Tile Il' FACT OF DIE PROPOSED SRTFOItrT C TPalLSIEIITS IS GIVES IN TABLE 1. ~ 87.J PINiTS ARE CURRENTI.Y CAPAULE OF R IBACX TO.15'C 047 F LOAD.OR TRIP OF THE TURBUiE. THIS CAPABILITY REQUIRES ACTUATIUti 0F Tile PILOT CPERATED RELIEF VALYES. THE CAPABILITY IllCREASES TllE RELIABILITY OF POWER nFTER THESE TRANSIENTS. SUPPL't TO THE SYSTEM BY RETUR 1,' ORE f]UICXLY NR BE TRIFFED FOR THESE EVEUTS:THE ACTION PROPO5ED ABOV .. fiOTE: - The effect of changing the reactor coolant system pressure trip setpoint upon peak pressurizer pn'ssure is typified by the attached figure 1. which was developed by. B&W for a loss of feedwater transient. .. = s. b h a s . ~_ l ^-
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TABLE 1 Fr - 1sure 1 -r SL%RY OF PROTECTIOtt AGAIrf5T PORY ACTUAT10:1 / PEDVIDED B7 PROPOSED SETP0lHT DIANGES FOR ALL ( N1TICIPATED TRANSIENIS EXTRACT OF B&W CCF,1JHECATION - RECEIVED BY tmC 4/?_0p9. c 1.
- 1;TICIPATED TRMtSIEifr5 mfIOl HAVE OCCURRED AT BW FLAHIS A:1D MIIGl NOULU. 7 r.DF4.L7 ACTIVATE FORY AT THE CURRENT SETPOInr (2255 PSIC):
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TUR3IME TRIP.'.. i7 =
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~ I ti. LCSS OF EITERHAL ELECTRICAL LDAD ' b - m. -: f. . ~M C ~.
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-> ~ ; I- . y-- ' _ _ D. LGSS OF CC;;DEMSEft.VACU1It 6 ~ ,_---i.E IRAD'iERTE!ir CLOSURE -OF MSIV ~ 2.MMTTICIPATED TRAMSIENTSDiIDFHAVE OCCURRED AT BW PLleGS AND }illICH ~ l'CUI.D KORMALLY ACTUATE FORY AT Tile' PROPOSED SETPOITTT (Z".50 PSIG): ^ n_.,. t. O-TC > 3 RETICIPATED TRNiSIEriTS.15fIGi HAVE NOT OCCURRED AT BtN PLNiTS '(LOW P" R03 ABILITY EVEliTS) AND 11HI01 MOULO ff0RMALLY ACTUATE PORY AT THE CITR5.Enr xt evIlfT (2255 PSIG): ih* Sche CGNTROL RCD GROUP HITilDRAMALS (MODERATE TO !!IGli REACTIVITY . EGRTH ' CRotr/5 f?OT OTiiERMISE PRbTECTED BY HIQi FLUX TRIP m -- ~. D. KUDERATOR DILUTIO't. ~ T.'i "i 9 ~ 9 --[.- .. ~, f .h MTICIPATED TRMiSIENTS HHIOl HAVE ft0T OCCURRED AT DFM PLArtTS (LOW PROBA ~ i h .. ' EVEfiTS) AND UHIO! UOULD ACTUATE TIIE FORV AT THE PROPO5ED SETPOIrlT 4 ^ y 3 L) _(2450 PSIG): 7__ - ' ~ i L ^ ~ .3.,n-- A. ~ 'SG'IE C0trTROL ROC GOUP UITHDTUNALS (HIGil' REACTIVIT'Y UORTI! ?;0T- {~. ,.. g[. OTHERHISE PROTECTED BY HIDI FlpX TRIP). ~ ,y g-. cey -: ..cj -. __..__._ _ _ -_, q [. y - ;m9
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_ _ _ _. _ _.... '.. _j 1 L-1 _4 l PaSe 4 of 4 EXTPACT OF B&W COP.%UNICA ON - RECEIVED BY NRC 4/20/79 s I L.,..,l....l... .l i s... a s z. .. ~ ...[....t l e...._.j .l..
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///j;.,ht Rc.S_ g< P/ZE,;;;' Sus E 1 ... g.- .l'.L 4 s \\... 1....u. ~ ......,....,_. m m. m g y...... s._. s. s_. _._. ..l_..... }... ...g.... ....-.s.... g .. _. _.. <_._r _m 4 ..x Peak pressurizer pressure as a function of RCS pressure trip setpoint f.or.tial pressures. ini. loss of feedwater-transient for expected conditions and various a i Figure 'l 6 d. 1 ' ' + @S _ = .. = _..ei--.m.ee. N
s k- } g i ,., D UNITED oTATES ~ _ a R.' r. Ip/p~. j "n NUCLEAR REGULATORY COMMISS!QN J.'h9 ADWionY cc.:.wnre or nEAcroa sp.cccut,no: 4 ,f. ro.auncTou.c.c h s3 7,..- April 20, 1979
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[ a = l e i Y a a L: a v c.. .'*ar.orrble Victor Gllinskv d . A'.: ting Ch. nim, L. U.. S. Mr.= lear Regulatory Ccanissicn . Kashington, DC 20355 l
Dear Dr. Gilinsky:
~ Bis letter is in respana to yourn of April 18, 1919 t.hich requested that the ACRS notify the 'Cormissioners ircardiately if we believe any of our orcl recec::wndation= of April 17Jshould te acted upon bOfore our nes. regularly scheduled niecting at which ue could prepsro a for ol .let~ter. Tne Ccemittee oiscussed-this topic by conference relephona ec11 an' April 19 ad offers the fo13cwing. c=wnts. \\ All of the recoc;nendsti'ons r.ade by the ACRS in its meting with the Cc,.s.lecionats on April 17/ 1979, ete genaric in nature and cpply to all N?s..None were intended to recuire ipr.ediato, changes in oporoting pro-ceduros or plcnt Indifications of o_rerating F,cc. Such changes thould b ceee on3y after study of their eifects on overall safety, such stud-ie_s ch:uld be n.*:de by the licensees and their suppliers or consulta:s. imd by the IGC S* af f. The Cor alttpe believes that th2Se studies should M Mgun in the naar future on a tir,e scale that will not divert the .SRC Staff or the industry representatives from their tasks relating to tha cooldpwn of Otree Mile Irland Unit 2. Eowever, the Cc.w.i tteo ht~ lic*n.s th'at it would he passible arri desirable ta initiate ir.ediately a Inrvey of opcratir/3 procedures for achiedog naturil circulation,. Irr-- c3ndirr3 thc. c;de vder' ofEsHe p3 'ar is lost, and the role of the p'rer- - Nri:.=r heaters in su=5, p'f6:eduros. y! O
- ,t its tacat'ing cn April 15 and 17,1979, the Co
- ::Ittec discussed.trith the ERC St.aff the r. utter of natural circulation for the Wree MD.o Is.
R 1ca4 Unit 2 plant. Le Cc"ttee belicycs that this astter is receiv-5 Ing c reful attention by the lac Staff anJ the 11cca ee. @d To ED3 for Appropriate Action. Distribution: chm, Cars, PE, OGC, OCA, set:Y, L a ~D2, OlA. Rapifcxcd to EDO, PA, E., case. 79-1117.. s-413
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- 45 CT...it:.ce'Ir om rececWions to the Comal:5sion c.m April 17 were N
r .....L m: mta:dc:d to apply to.'nt ce Mlle Icland L' nit. 2. .i O $~ -. c. .-... ~.. _. g yrhb pls.1 te write-a F -thse rep:2rt. en{ there-r:strern at our sty ig,1973 L Q.. c::ctu:T- . -- --.:. =,... _ .~ ~ Sincerely, .~' ~ af _._-G ~ * .~. _. - -f f ~ , -l ; / s'[n/fd*g . ~... Wr s . '.1.5 - Itu W. 'rbon ~ ~ ..;'-.f',.-
- Chairman g
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UNITED STATES 3 NUCLEAR REGULATORY COr.WISSION O9(<g[(*j 2
- 3VISORY CO?..i.'.lTTCE 0:2 REACTO3 SAF 'UARDS
\\@p f . O[%.. - WAssi.40ro 4. o. c. 2 ass 4 April 18,1979 ~
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m - N 7 hu +J 4 IEMCRAEUM EUR: Chair: nan liendric - o m . ^u... .E ' Co.missioner Gilinsky I'~ .,5:$[ 7-.. - Comissioner Kennedy. t iriec_ j ~...', ,Cc:=sissioner Bradford '? R".. - ',. '. ~~ Coctissioneir Ahearne -e ~ - 1 C- ~- GCM: R. F. Fraley, Executive Director Advisory Cec =aittee on Reactor Safeguards = _._.5, ~r. r'- Attached for your information and use. is a copy of :the reco=eh 72 tions of the Advisory Co..,. ittee on Reactor safeguards diich ware orally presented to and discussed with you on April 17,1979 re-- garding the recent accident at the Three Mile Island Nuclear Sta-tion Unit 2. \\ . :.r f( /;t ' R. : F. Fraley Executive Director . Attach: ent: Reccamndations of the NRC Advicory Comittee on Reactor Safeguards Re. the 3/28/79 Accident ' at nie Ihree Nile Island Nuclear Station Unit 2 . 44 .. ~.. s P k
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April 17, 1979 iPM.MENPATICNS O? WiC NUCLEAR RfnU[Nf0RY ComISSION ADVISORY COMMITIEC W REAC'IUR SAFD3UARCS R EARDING THE MARCH 28, 1979 ACCIDEC.T AT THE n!RES HILE ISER;D NUCLEAR STATION UNIT 2 .'Rv presented orally to, and discussed with, the NRC - ' - ' 9} .= Commissioners durirg the ACRS-Com:nissioners Meeting ' ;j en April 17, 1979:- Washington, D. C. 3 9 s J L 5,. u Natural circulatiqn is an impactant made of reactor cooling, both as . 9 J.~. a planned process and as a process that may be used under abnormal / circe = stances. "Ihe Ccmmitter believes that greater understanding of X - ^' .C. ' ',' 'this rede of cooling'.1s _ required and that. detailed analyses.should iE f be developed by licensees.-or their suppliers. W e analyses should be S-supported, - as necessary, by experiment. Procedures should be ~ de- ' ? :" veloped for initiatirx3 natural circulation in a safe manner and for providing the operator with assur.mee that circulation has, in fact, { _; bem established..This may require. Installation of instrumentation to measure or indicate flew at low water velocity. ..s. . Iha use of natural circulation for decay heat removal following a loss of offsite po.4er sources requires the maintenance of a suitable over-pressure on the reactor coolant sys ter.. This overpressure, may be accured by placing the pressurizer hea ters on a qualified onsite powr source with a. suitable arrange. ent of heaters and power distri-bution to provide -redundant. capability.. Presently ~ operating FdR plants should be surveyed exp Mitiously to detenaine whether such arrangements can be provided, to assure this aspect of natural circula-tion capability. Se plant operator should be adequately informed at all times con-cernIng the conditions of. reactor coolant system operation which might affect the capability to place the system in the natural circu-lation pode of operation or to sustain such a made. Of particular importance is that information which-might indicate that the reactor coolant system is approaching the satucation pressure corresponding-to the core exit temperature. This impending loss of system over-pressure will signal to the operator a possible loss of natural circulation capability. Such a warnirg may be derived from pressur-N iner pressure instruments.and hot leg tenperatures in conjunction with C conventional - steem tables.. A suitable display of this information O should be provided to the plant oper: tor at all times. In addition, [] consideration should be given to the use of the flow exit tempera-- i tures froco the fuel subassemblies, where available, as an additional 5:s Indication of natural circulation. 7' ] 1 p e 'I!> i C, j .=~ --- : _=E= -?---==- : gNN=T =35+'E^ ^ -W J.f % - - - = - - - - =
The exit temperature of coolant fro.n the core is currently measured by therroccuples in many nas to ' determine core perfomance. We. Cc,;nittee recommands that these temperature measurements, as currently R ,r e available,. be used to guide tre operator concerning core status. %e range of the Information displayed and recc W J should include tha g full capability of the thermocouples. It is also recommended that ] cther *existiry instnr:entation ce examined for its possible use in b;] r nmhting operating action during a. transient. kf - %e JCS reccc= ends that operating power reactors be given priority @k . __ -fwi'th regard to the. definition and Implementation of instrumentatica Ahich provides additional information to help diagnose and follow the ..j; course of a serious accident. This should include improved sampli.ng -' ;; procedures under accident conditions and techniques to help previde 7,. "[_-) Comittee recommendsImproved guidance to offsite authorities, should this be needed. The that a phased implementation approach be en- ~ D-1 played so that techniques can be adopted shortly after they are 2 .. J.3 -judged to be appropriate. ^ %e*ACRS recommends that a high priority be placed on the deve?opnent and implenentation of safety reseatch on the behavior of light water . reactors during anomalous transients. The NRC may find it appronriate ( . to develop a capability to simulate a wide range of postulated tran-sient and accident conditions in order to gain increased insight into measures which can be taken to impcove reactor safat. n e ACns uishes to reiterate its previous recom;nendations that a high priority bc given to research to improva reactor safety. Consideration should be given to the desirability of ' additional equip.ent status ponitoring on.various engineered safeguards features ._ and their supporting services to help assure their availability at all times. The ACRS is continuing its review of the implications of this accident and hora to provide further advice as it is developed. N q n Np aa h-M m L .) [. - i ' ^. " * = - *.=
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 (OL) or 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 power reactor Safety Featurer facilities wit 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 rmit (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 i85
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 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) 410 186 Enclosure Page 2 of 2}}