ML19225B106
| ML19225B106 | |
| Person / Time | |
|---|---|
| Site: | Aerotest |
| Issue date: | 06/22/1979 |
| From: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Newacheck R AEROTEST OPERATIONS, INC. |
| References | |
| NUDOCS 7907230527 | |
| Download: ML19225B106 (1) | |
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UNITED STATES
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'g NUCLEAR REGULATORY COMMISSION s
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1990 N. CAllFORP I A OULEVARD
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6" SUITE 202, WALNUT CREE K PL AZA
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WALNUT CHEE K, CALIFORNIA 94536 June 22,1979 Docket No. 50-228 Aerotest Operations, Inc.
3455 Fostoria Way San Ramon, California 94583 Attention:
R. L. Hewacheck President Gentlemen:
The enclosed Information t!otice No. 79-16 is forwarded to you for information.
No specific action is requested and no written response is required.
If you desire additional information regarding this matter, please contact this office.
Sincerely, h)C('f.-), ( c. _
/
R. H. Engelken Director
Enclosures:
1.
IE Information Notice 79-16 2.
List of IE Information Notices Issued in 1979 410 07J p230 sa
3 UtilTED STATES flVCLEAR REGULATORY COMMISSI0il 0FFICE OF IllSPECTIO;i A;iD ENFORCEMEliT WASHIi!GTON, D.C.
20555 June 22, 1979 IE Information flotice fio. 79-16 fiUCLEAR IllCIDEllT AT THREE MILE ISLAi!D Description of Circumstances:
On March 28, 1979, the Three Mile Island fluclear 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 nakes 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 reactcrs supplied by vendors other than Babcock and Wilcox.
iio specific action or written response to this information ilotice is required.
If you desire additional information regarding this matter, contact the Director of the appropriate fiRC Regional Office.
Enclosures:
1.
IE Bulletin fio. 79-05 with Enclosure 2.
IE Bulletin fio,79-05A with Enclosure 3.
IE Bulletin i!o.79-05B with Enclosure 7907020186 410 074
e IE Information Notice No. 79-16 June 22, 1979 Page 1 of 2 LISTING OF IE INFORMATION NOTICES ISSUED IN 1979 Information Subject Cate Issued To Notice No.
Issued 79-01 Bergen-Paterson Hydraulic 2/2/79 All power reactor Shock and Sway Arrestor facilities with an OL or a CP 79-02 Attempted Extortion -
2/2/79 All Fuel Facilities Low Enriched Uranium 79-03 Limitorque Valve Geared 2/9/79 All power reactor Limit Switch Lubricant facilities with an OL or a CP 79-04 Degradation of 2/16/79 All power reactor Engineered facilities with an Safety Features OL or a CP 79-05 Use of Improper Materials 3/2I/79 All power reactor in Safety-Related Components facilities with an OL or CP 79-06 Stress Analysis of 3/23/79 All Holders of Safety-Related Piping Reactor OL or CP 79-07 Rupture of Radwaste Tanks 3/26/79 All power reactor facilities with an OL or CP 79-08 Interconnection of 3/28/79 All power reactor Contaminated Systems with facilities with an Service Air Systems Used OL and Pu Processing As the Source of Breathing fuel facilities Air 79-09 Spill of Radioactively 3/30/79 All power reactor Contaminated Resin facilities with an OL 79-10 Nonconforming Pipe 4/16/79 All power reactor Support Struts facilities with a CP 79-11 Lower Reactor Vessel Head 5/7/79 All holders of Reactor Insulation Support Problem OLs and cps 410 075
IE Information flotice fio. 79-16 June 22, 1979 Page 2 of 2 LISTIttG OF IE IriFOR1ATION fiOTICES ISSUED Ill 1979 Information Subject Date Issued To flotice fio.
Issued 79-12 Attempted Damage to flew 5/11/79 All Fuel Facilities Fuel Assemblies Research Reactors, and Power Reactors with an OL or CP 79-13 Indication of Low tlater 5/29/79 All Holders of Rcactor Level in the Oyster Creek OLs and cps Reacter 79-14 fiRC Position of Electrical 6/11/79 All Power Reactor Cable Support Systems Facilities with a CP 79-15 Deficient Procedures 6/7/79 All holders of Reactor OLs and cps 410 076
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UllITED STATES NUCLEAR REGULATORY COMMISSIO:
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OFFICE OF INSPECTI0il AND EriFORCEMENT WASHINGTO:1, D.C.
20555 April 1,1979 IE Bulletin fio. 79-05 4
fiUCLEAR INCIDENT AT THREE f1ILE ISLAtlD 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.
Several aspects of the f
incident may have general applicability in addition to apparent generic r'
applicability at operating Babcock and Milcox reactors.
This bulletin is provided to inforn you of the nuclear incident and to request certain actions.
Actions To Be Taken By Licensees (Although the specific causes have not been deternined for individual
(
sequences in the Three Mile Island event, some of the following may have contributed.)
For all Babcock and Wilcox pressurized water reactor facilities with an operating license:
1.
Review the description (Enclosure 1) of the initiating events and subsequent course of the incident. Also review the evaluation by the f!PsC staff of a postulated severe feedwater transient related to Babcock and Wilcox PWRs as described in Enclosure 2.
These reviews should be directed at assessing the adequacy of your reactor systems to safely sustain cooldown transients such as
- these, r.:
2.
Review any transients of a similar nature which have occurred at F"
your facility and determine whether any'significant deviations from I
expected performance occurred.
If any significant deviations are found, provide the details and an analysis of the significance and any corrective actions taken.
This naterial may be identified by reference if previously submitted to the NRC.
3.
Review the actions required by your operating procedures for coping h
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with transients.
The items that should be addressed include:
410 077 s 6 af -,'
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IE Bulletin ilo. 79-05 April 1, 1979 Page 2 of 3 n
a.
Recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling capability.
a b.
Operator action required to prevent the formation of such voids.
+
c.
Operator action required to ensure continued core cooling in the event that such voids are formed.
4.
Review the actions requested by the operating procedures and the f
training instructions to assure that operators do not override automatic actions of engineered safety features without sufficient cause for k
I 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 systcm, can perform their intended functions.
Also review related procedures, such as those
[
for maintenance and testing, to assure that such valves are returned g
to their correct positions following necessary manipulations.
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6.
Review your operating nodes 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:
a.
Whether interlocks exist to prevent transfer when high radiation indication exists and, b.
Wbcther such systems are isolated by the containment isolation signal.
7.
Review your prompt reporting procedures for tiRC notification to assure very early notification of serious events.
(
The detailed results of these reviews shall be submitted within ten (10) days of the receipt of this Bulletin.
i 1
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uu 078 s:
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IE Bulletin Mo. 79-05 April 1, 1979 Page 3 of 3
[N Reports should be submitted to the Director of the appropriate NRC
[~
Regional Office and a copy should be forwarded to the NRC Office of
~
Inspection and Enforcement, Division of Reactor Construction Inspection, Washington, D.C.
20555.
For all other operating reactor or reactors under construction, this L-Bulletin is for information purposes and no report is requested.
,]
N Approved by GAO, B18C225 (R0372); clearance expires 7-31-80.
Approval was given under a blanket clearance specifically for identified generic l
problems.
,I
Enclosures:
1.
Preliminary Hotifications - Three Mile Island PMos 67, 67A, B, C, D, E, F and G i
2.
Evaluation of Feedwater Transients with Attachment 3.
List of IE Bulletins issued in the past 12 months 7
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1 7
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410 079 c....
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ENCLOSURE 2 t
tvaluation of Feedwater Transient
- 7 A loss of offsite power occurred at Davis-Besse on November 29, 1977, which resulted in shrinkage of the primary coolant volume to the degree that pressurizer level indication was lost.
A reconmenda tion to convey this information to certain hearing boards resulted in the attached discussion and evaluation of the event.
This discussion includes a i,
review of a loss of feedwater safety analysis assuming forced flow, which predicts dispersed primary system voiding, but no loss of core cooling. During the Three Mile Island event, however. the forced flow appears to have been terminated duri.ig the transient.
Attachment Discussion and Evaluation of t
Davis-Besse Transient t
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E7CERPI FROM MEMORAUDUM Erf1TLED " CONVEYING I;EU INFORMATION TO LICEMSI!:0
' I E0ARDS - DAVIS-BESSE UNITS 2 & 3 AMD MIDUd:D UNITS 1 6 2",
DATED JANUARY 8, 1979, FPOM J.S. CREST.' ELL TO J.F.
STREETER M
t e7
..y 3.
Inspection and Enforcement Report 50-346/78-06 documented that pressurizer level had gone of f scale f or approxt ately five uinutes during the November 29, 1977 loss of off site power event.
There are some indications that other B&U plants nay have prot-O lens caintaining pressurizer level indications during transients.
7 in addition, under.certain conditions such as loss of feedwater at 100% power with the reacter coolant pumps running the pres-surizer may void completely.
A cpecial analysis has been per-Il formed concerning this event.
This analysis is attached as.
Because of pressurizer level raintenance prob-7 lems the t.izing of the pressurizer'nay require further review.
Also noted during th:* cven t var, the face that T en 1 A_w.s. >f,c..
scale (less than 5200F).
In addition, it was noted that the i
nakeup flou monitoring is linited to less than 160 gpa and that nakeup ficw may be substantially greater thnu this value.
I Tuis information should be e>:amined in light _of the require-t'ents of CDC 13.
DISCUSSION AND EVALUATEON
((
The event at Davis Besse which resulted in loss of pressurizer level indication has been revie"ed by URR and the conclusich was reached that no unreviewed safety question cxisted.
The pressurizer, together with the reactor coolant enkcup systen, in designed to naintain the primary syster 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, sometimes result in primary pressure and volume changes that are beyond the ability of this system to control. The analyses of and experience with such transients show, however, that they can be sustained without comproaising the safety of the reactor.
The principal concern caused by such transients is that they night cause voiding in the primary coolant system that would lead to loss of ability to ade-r
[
quately cool the reactor core.
Tha safety evaluation of the loss of offaite power transient shows that, though level' indication is lost, co=a water renains in the pressurizer and the pressure does not decrease
[
be'cw about 1600 psi.
In order for voiding to occur, the pressure must dc case below the saturation presoure corresponding to the systen teng,rature.
1600 psi is the saturation pressure corresponding to 605 F, which is alse the maxinua ellowable core outlet temperature.
Voiding in the primary system (excepting the pressurizer) in precluded
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in thin case, since presaure dona r.ot decrease to saturation.
Attachment
.f (Enclosure,2)
A.
(Page 1 of 2) 410 081-v
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i Section 3.
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- t l
f The safety analysis for core severe cooldown transients, cuch as the i
loss of feedvater event, indicatca that the water volume could decrease Q
to les's than tha system volune exclusive of the pressurizer.
During such an event, the captying of the pressurizer would be followed by a pressure reduction below the saturation point and the forcation of small voids throughout much of the prl=ary system. This would not result in the loss of core cooling because the voids would be dispersed over a large voluma and forced flow would prevent them from coalescing i
sufficiently to prevent core cooling.
The high pressure coolant injection pumps are started automatically when the pri:ury pressure I.
i decrease below 1600 psi..
Therefore, any pressure reduction which is sufficient to allow voiding vill also result in water inj ection which will rapidly restore the primar-f water to normal levels.
For these reasons, we believe that the inability of the pressuricer and normal coolant makeup systen to control some transients does not provide a basis for requiring rore capacity in these systems.
Cencral-Design Criter_ on 13 of Appendix A to 10 CFR 50 req 61res instrumentation to monitor variables over their auticipated rangen for " anticipated operational occurrences".
Such occurrences are specifically defined to include loss of all offsite power.
The fact that T cold goes off scale at 520"F is not considered to be a deviation l
/,
5 from this requirecent becaun this indicator is backed up by vide t
range terparature indication that extends to a low limit of 50"F.
Neither do va consider the takeup flow ronitoring to deviate since the amount of makeup flow in excess of 160 gpm does not appear to be a significant factor in the course of these occurrences.
The loss of pressurizer v.ater level indication could be considered to deviate f rom GDC 13, because this level indication provides the principal tacans of determining the priaary coolant inventory.
However, provision of a level indication that vould cover all anticipated occurrences nay not be practical. As discussed above, the lons of feedwater event can lead to a momentary condition wherein no meaningful level exists, because the entire primary systen contains a steam water nixture.
It should be noted that the introduction to Appendix A (last paragraph) y 9
recognizes that fulfillment of some of the criteria may not always be appropriate.
This introduction also states that departures f ron the Criteria uust be identified and justified.
The discussion of GDC 13 in the Davis Besse FSAR lists the varer level instrumentation, but does not cention the possibility of loss of water level indicati' n
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g during transients.
This apparent omission in the cafety analysis vill be subjected to further review.
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UNITED STATES
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NUCLEAR REGULATOP.Y CO:'. MISSION
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OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, DC 20555 APRIL 5, 1979 IE Bulletin 79-05A F
NUCLEAR INCIDENT AT THREE MILE ISLAND - SUPPLJMENT A
Description of Circumstances:
y Preliminary information received by the NRC since issuance of IE i
Bulletin 79-05 on April 1,1979 has identified six potential human, I..
design and mechanical failures which resulted in the core damage and radiation releases at the Tnree Mile Island Unit 2 nuclear plant. The j
information and actions in this supplement clarify and extend the original Bulletin and transmit a preliminary chronology of the TMI accident t
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.
j 2.
The pressurizer electroratic relief valve, which opened during the initial pressure surge, failed to close when the pressure decreased below the actuation level.
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3.
Following rapid depressurization of the pressurizer, the pressurizer level indication ray have lead to erroneous inferences of high l
level in the reactor coolant system.
The pressurizer level indication apparently led the operators to prematurely terminate high pressure injection flow, even though substantial voids existed in the reactor coolant system.
4.
Because the containment does not isolate on high pressure injection (HPI) initiation, the highly radioactive water fro, the rclief valve discharge was pumped out of the containment uy 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 I
through the auxiliary building ventilation system and filters was I
the principal source of the offsite release of radioactive noble
_c m
gases.
5.
Subsequently, the high pressuie in#ection system was interuittently operated attempti.9 to control princry coolant inventory losses through the elec'.romatic relief valve, apparently based on pressurizer levr.1 indication.
Due to the presence of steam and/or noncondensible voids elsewhere in the reactcr coolant system, this led to a further reduction in princry coolant inventory.
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r IE Bulleun 79-05A April 5,1979 s
Page 2 of 5 l
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6.
Tripping of reactor coolant pumps during the course of the transient, to protect against pump damage due to pump vibration, led to fuel damage since voids in the reactor coolant system prevented natural ci rcul ation.
]
Actions To Be Taken by Licencees:
i FOR ALL Babcock and Uilcox pressurized water reactor facilities with an operating license (the actions specified below replace those specified i
in IE Bulletin 79-05):
j t
1.
(This item clarifies and expands upon item 1. of IE Bulletin 79-05.)
!l D
In addition to the review of circumstances described in Enclosure 1 g
of IE Bulletin 79-05, review the enclosed preliminary chronology of
[
the TMI-2 3/28/79 accident.
This review should 'f directed toward i'
understanding the sequence of events to ensure against such an accident at your facility (ies).
2.
(This item clarifies and expands upon item ?. of IE Bulletin 79-05.)
j
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Review any transients sirilar 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 fiRC, if appropriate, in responding to this item.
3.
(This item clarifies item 3. o IE Bulletin 79-05.)
c Review the actions wquired by your operating procedures for coping U
with transients and accidents, with particular attention to:
j a.
Recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling l
capability, especially natural circulation capabilit,.
'. g b.
Operator action required to prevent the formation of such voids.
p c.
Operator action required tc enhance core cooling in the event such voids are forned.
g
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410 084
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I IE Bulletin 79-05A April 5,1979 k
Page 3 of 5 i
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4.
(This item clarifies and expands upon item 4. of IE Bulletin 79-05.)
Review the actions directed by the operating procedures and training instructions to ensure that:
)
Operators do not override automatic actions of engineered a.
safety features.
b.
Operating procedures currently, or are revised to, specify 1
that if the high pressure injection (HPI) system has been l
l automatically actuated because of low pressure condition, it rust remain in operation until either:
f, (1) Both low pressure injection (LPI) pumps are in operation I
and flo. ling at a rate in excess of 1000 gpn each and the I
situation has been stable for 20 minutes, or i
(2) The HPI system has been in operation for 20 ninutes, and all hot and cold leg temperatures ere at least 50 degrees below the saturation temperature for the existing RCS pressure.
If 50 degree subcooling cannot be naintained af ter HPI cutof f, the HPI shal' be reactivated.
c.
Operating procedures currently, or are revised to, specify that in the event of HPI initiation, with reactor coolant punps (RCP) operating, at least cne RCP per loop shall remain operating.
d.
Operators are provided additi
'l information and instructions to not rely upon pressurizer level indication alone, but to also exanine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g., water inventory in the reactor primary system.
t 5.
(This item revises item 5. of IE Bulletin 79-05.)
[7 -
Verify that emergency feedwater valves are in the open position in accordance with item 8 below.
Also, review all safety-related valve positions and positioning requirements to assure that valves are positioned (open or closed) in a manner to ensure the proper operation of engineered safety features.
Also review related procedures, such as those for maintenance and testing, to ensure that such valves are returned to their correct positions Q
following necessary manipulations.
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IE Bulletin 79-05A April 5,1979
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page 4 of 5 m
6.
Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to cause containment isolation of all lines whose isolat:on does not degrade core cooling capability upon automatic initiation of safety injection.
7.
For manual valves or nanually-operated motor-driven valves which could defeat or conpromise the flow of auxiliary feedwater to the steam generators, prepare and implement procedures which a.
require that such valves be locked in their correct position; f
or t
b.
require other similar positive position controls.
8.
Prepare and implement ir. mediately procedures which assure that two independent steam generator auxiliary feedwater flow paths, each with 1005 flow capacity, are operable at any time when heat removal from the primary system is through the steam generators.
When two inde-pendent 100% capacity flow paths are not available, the capacity shall be restored within 72 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 made which does not rely on steam generators for cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Knen at least one 100X 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 maximum safe shutdown rate.
9.
(This item revises item 6 of IE Bulletin 79-05.)
Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and liquids out of the primary containment to assure that undesired pumping of radioactive liquids and gases will not occur inadvertently.
7 In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation.
List all such systems and indicate:
a.
Whether interlocks exist to prevent transfer when high radiation indication exists, and 4
b.
Whether such systems are isolated by the containment isolation signal.
h.
~ QMI
IE Bulletin 79-(
April 5,1979 Page 5 of 5 l
,3 10.
Review and modify as necessary your maintenance and test procedures to ensure that they require:
Verification, by inspection, of the operability of redundant a.
safety-related systems prior to the removal of any safety-related system from service.
b.
Verification of the operability of all safety-related systems if.
when they are returned to service following maintenance or testing.
I A means of notifying involved reactor operating personnel c.
whenever a safety-related systen is reroved from and returned j
to service.
h 11.
All operating and maintenance personnel should be nade aware of the 5
extreme seriousness and consequences of the sinultaneous blocking
(
i of both auxiliary feedwater trains at the Three Mile Island Unit 2 plant and 6ther actions taken during the early phases of the accident.
12.
Review your prompt reporting procedures for MRC notification to assure very early notification of serious events.
For Babcock and Wilcox pressurized water reactor facilities with an operating license, respond to Items 1, 2, 3, 4.a and 5 by April 11, 1979.
Since these items are substantially the same as those specified in IE Bulletin 79-05, the required date for response has not been changed.
Respond to Items 4.b through 4.d, and 6 through 12 by April 16, 1979.
Reports should be submitted to the Director of the appropriate MRC Regional Office and a copy should be forwarded to the MRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, DC 20555.
For all other reactors with an operating license or construction permit, this Bulletin is for information purposes and no written response is required.
7 u
Approved tzy GAO, B 180225 (R0072); clearance expires 7-31-80.
Approval was given under a blanket clearance specifically for identified generic 7,
problens.
- . q r
Enclosures:
e 1.
Preliminary Chror. ology of TMI-2 3/38/79 Accident Until Core Cooling Restored.
2.
List of IE Bulletins issued in last 12 months.
410 087 q, ;:. ~~;
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IE Bulletin 79-05A
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April 5,1979 t
PRELIMINARY CHRONOLOGY OF Till-2 3/28/79 ACCIDEi!T
.I UNTIL CORE C00LIf!G RESTORED r
[.7 TIME (Approximate)
CVENT about 4 AM Loss of Condensate Pump 1.
(t = 0)
Loss of Feedwater Turbine Trip t = 3-6 set..
Electromatic relief valve opens (2255 psi) i to relieve pressure in RCS j
t = 9-12 sec.
Reactor trip on high RCS pressure (2355 psi)
L = 12-15 sec.
RCS pressure decays to 2205 psi (relief valve should have closed)
L t = 15 sec.
RCS hot leg terperature 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). !!o flow was injected since discharge valves were closed.
t = 1 nin.
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 5
minutes.
y 1
t = 2 nin.
ECCS initiatioh (HPI) at 1600 psi a
t = 4 - 11 min.
Pressurizer level off scale - high - one flPI pump canually tripped at about 4 min.
30 sec.
Second pump tripped at about 2
10 min. 30 sec.
- c t = 6 min.
RCS flashes as pressure bottoms out at 1350 psig (Hot leg temperature of 584 degrees F) t = 7 min., 30 sec.
Reactor building surp pump came on.
410 088 f
m;m
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Tlf:E EVE!!T t = 8 nin.
Auxiliary feeduater flow is initiated y
by opening closed valves q.;
t = 8 min. 18 sec.
Steam Generator B pressure reached minimum t = 8 nin. 21 sec.
Stea:n Generator A pressure starts to recover a
t = 11 nin.
Pressurizer level indication comes back on scale and decreases j
g.
t = 11-12 min.
Ilakeup Pump (ECCS HPI flow) restarted by operators t = 15 nin.
RC Drain / Quench Tank rupture disk blows at I
190 psig (setpoint 200 psig) due to continued I
discharge of electromatic relief valve I
t = 20 - 60 min.
System parameters stabilized in saturated i
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 nin.
Operator trips RC punps 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 f
t = 1-3/4 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> CORE BEGII;5 HEAT UP TRANSIEf;T - Hot 1 -
temperature begins to rise to 620 dc.mes F (off scale within 14 ninutes) and cold leg temperature drops to 150 degrees F.
(HPI water) t = 2.3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> Electromatic relief valve isolated by operator af ter S.G.-B isolated to prevent leakage t = 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> RCS pressure increases to 2150 psi and f
electromatic relief valve opened 7
t = 3.25 1.ours RC drain tank pressure spike of 5 psig r'
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; conteinment pressure increases from 1 to 3 psig t = 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Peak 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
[
to 2100 psi col unn 00 089
I ;
(
i TIME EVENT t = 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Operator opens electromatic relief valve to depressurize RCS to atter.pt initiation of RHR at 400 psi i
t = 8 - 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> RCS pressure decreases to about 500 psi Core Flood Tanks partially discharge t = 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 28 psig containr.ent pressure spike, containment sprays initiated and stopped after 500 gal. of 1.
Na0H injected (about 2 minutes of operation) l--
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 k'
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
iG hours RC pump in Loop A started, hot leg temperature e
decreases to 560 degrees F, and cold leg
{
temperature increases to 400 degrecs F.
indicating flow through steam generator i
Thereafter S/G "A" steaming to ccndensor Condensor vacuum re-established RCS cooled to about 280 degrees F.,
1000 psi No.s(4/4)
High radietion in containnent All co-e thercocouples less than 460 degrees F-Using pressurizer vent valve with small nakeup fic.v Slow cooldown RB pressure negative e:e t
d ATu j
\\ l
UNITED STATES fiUCLEAR REGULATORY COD 11SSIO:1 0FFICE OF IflSPECTION AfD EtiFORCEfullT WASHIf!GTO:!, DC 20555 April 21, 1979 IE Bulletin 79-05B fiUCLEAR IfiCIDEf!T AT THREE MILE ISLAllD SUPPLEMEUT Description of Circumstances:
Continued I;RC evaluation of the nuclear incident at Three flile Island Unit 2 has identified measures in addition to those discussed in IE Bulletin 79-05 and 79-05A which should be acted upon by licensees with reactors designed by B&U.
As discussed in Item 4.c. of Actions to be taken by Licensees in IEB 79-05A.
the preferred mode of core cooling following a transient or accident is to pro-vide forced flow using reactor coolant pumps.
It appears that natural circulation was not successfully achieved upon securing the reactor coolant pumps during the first two hours of the Three Mile Island (TMI) f;o. 2 incident of March 28, 1979.
Initiation of natural circulation was inhibited by significant coolant voids, possibly aggravated by release of non-condensible gases, in the primary coolant system.
To avoid this potential for interference with natural circulation, the operator should ensure that the
[
primary systen is subcooled, and remains subcooled, before any atteupt is made to establish natural circulation.
s flautural circulation in Babcock and Wilcox reactor systems is enhanced by maintaining a relatively high water level on the secondary side of the once through steam generators (OTSG).
It is also promoted by injection of auxiliary feedaater at the upper nozzles in the OTSGs.
The integra ted Control System automatically sets the OTSG level setpoint to 50; on the operating range when all reactor coolant pumps (RCP) are secured.
However, in unusual or abnormal situations, manual actions by the operator to increa steam generator level will enchance natural circulation capability in antici ation of a possible lors t
of operation of the reactor coolant pumps.
As stated previously, forced flow of primary coolant through the core is preferred to natural circulation.
Other means of reducing the possibility of 5/oid formation in the reactor coolant system are:
A.
Minimize the operation of the Power Operated Relief Valve (PORV) on the pressurizer and thereby reduce the possibility of pressure reduction by a blowdoso through a PORV that was stuck open.
m 0H
IE Bulletin 79-05D April 21, 1979 page 2 of 4 B.
Reduce the energy input to the reactor coolant system by a prompt reactor trip during transients that result in primary system pressure increases.
This bulletin addresses, among other things, the means to achieve these objectives.
Actions To Be Taken by Licensees:
For all Babcock and Milcox pressurized water reactor facilities with an operating license:
(Underlined sentences are modifications to, and supersede, IEB-79-05A).
1.
Develop procedures and train operation personnel on methods of establishing and maintaining natural circulation.
The procedures and training must include means of monitoring heat removal ef ficiency by available plant ins trumen ta tion.
The procedures nust also contain a method of assuring that the primary coolant system is subcooled by at least 50 F before natural circulation is initiated.
In the event that these instructions incorporate anticipatory filling of the OTSG prior to securing the reactor coolant pumps, a detailed analysis should be done to provide guidance es to the expected system The instructions should include the following precautions:
response.
maintain pressurizer level sufficient to prevent loss of level t
a.
indication in the pressurizer, b.
assure availability of adequate capacity of pressurizer heahrs, for pressure control and maintain primary system pressure to satisff the subcooling criterion for natural circulation, and maintain pressure - temperature envelope within Appendix G limits c.
for vessel integrity.
procedures and training shall also be provided ta maintain core cooling in the event both main feedwater and auxiliary feedwater are lost while in the natural circulation core cooling mode.
2.
flodify the actions required in Item 4a and 4b of IE Bulletin 79-05A to take into account vessel integrity considerations.
"4.
Review the action directed by the operating procedures and training instructions to ensure that:
Operators do not override automatic actions of engineered a.
safety features, unl s_s_ continued operation of_ engineered f
(
mwua
IE Bulletin 79-05B April 21, 1979 Page 3 of 4 safety f ea tures_piill _ result __in unsa f e. plant _ condi tions. _ For
.exapple, i f con ti nuqd opyra tiortof_qng i ng_ red _ sa f e ty_ f ea tures would threaten reactor vessel integrity _ then the llPI should be secured (asnotedinb(2)_beloE).
b.
Operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been automatically actua ted because of low pressure condition, it must remain in operation until either (1)
Doth low pressure injection (t PI) pumps are in operation and flowing at a rate in excess of 1000 gpm each and the situation has been stable for 20 minutes, or (2) The HPI systua has br en in operation for 20 minutes, and all hot and cold leg temperatures are at least 50 degrees below the saturation temper a ture for the existing RCS pressure.
If 50 degrees subcooling cannot be naintained af ter PPI cutof f, the HPI shall be reactivated. T_he degree of subcoolini ]eyond 50 decjrees F _and tJp__leng_th iof_ tim t
HPI is in opera tion _s_ hall _be l imi_ted by_ the pressure /
tcypqra ture _ considqrations f_or _the yessel_ integi:i ty. "
3.
Following detailed analysis, describe the modifications to design and procedures which you have implemented to assure the reduction of the i
likelihood o.
atoratic actuation of the pressurizer P03V during antici-pated transients.
This analysis shall include consideration of a nodifi-cation of the high pressure scran setpoint and the POVR opening setpoint such that reactor scram will preclude opening of the PORV for the spec-tru:n of anticipated transients discussed by Cf.W in Enclosure 1.
Changes developed by this analysis shall not result in increased frequency of pressurizer safety valve operation for these anticipated transients.
4.
Provide procedures and training to operating personnel for a prompt manual trip of the reactor for transients that result in a pressure increase in the reactor coolant systen. These transients include:
a.
loss of main feed. vater b.
turbine trip c.
main Steam Isolation Valve closure d.
low OTSG 1evel f.
low prest.urizer level.
4l0 093
IE Bulletin 79-05B April 21,1979 Page 4 of 4 5.
Provide for I;RC approval a design review and schedule for implementation of a safety grade automatic anticipatory reactor scron for lass of feed-water, turbine trip, or significant reduction in steam generator level.
6.
The actions required in item 12 of IE Bulletin 79-05A are modi fied as follows:
Review your prompt reporting procedures for ?!RC notification to assure that tlRC is notified within one hour of the time the reactor is no_t in a controlled or expected condition of operation.
Further, at that time an open continuous communication channel shall be established and maintained wi th flRC.
7.
Propose changes, as required, to those technical _syecifications which must be r.odified as a resul t of your implementina the above items.
Response schedule for BP,W designed faciliLies:
a.
For Items 1, 2, 4 and 6, all facilities with an operating 1-icense respond within 14 days of receipt of this Bulletin.
b.
For Item 3, all facilities currently operating, respond within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
All facilities with an operating license, not currently operating, respond before resuming operations.
c.
For items 5 and 7, all facilities with an operating license respond in 30 days.
Reports should be submitted to the Director of the appropriate flRC Regional Of fice and a copy should be forwarded to the i;RC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washing ton, D.C. 20555.
For all other power reactors with an operating license or construction permit, this Bulletin is for information purposes and no written response is required.
Approved by GAO, B180225 (R0072); clearance expires 7/31/80.
Approval was given under a blanket clearance specifically for identified generic problems.
Extract of B&W Communication Received by t1RC
(
s10 w;4 s
EXTRACT OF B&W C0muNICATION - RECEIVED BY NRC DrTRUCUCT.IO1 M0m pm ] 0f 4
(
~T CO?fTINUING RE' LIER OF THE SEQUE??CE OF EVENTS LEADIllG TO THE INCIDErlT AT l-2 CN f4 ARC!! 28,1979 SHOWS TilAT ACTI0ti CNi BE TAKEN TO PROVIDE ASSutWiCE THAT THE PILOT-OPERATED RELIEF VALVE (PORY) i;00NTF.9 ON Tile PRE 550RIZEH OF BMI PL%iTS l'ILL NOT BE ACnJATED BY AtlTICIPATED TfMil5IEtiTS WilICil llAVE OCCUnnED OR HAVE a oIGilIFICMTT PROBABILITY OF GCCURRING IN T1(ESE PLAMIS. THIS ACTI0tltrJST NOT DEGRADE Tl!E SAFETY OF THE AFFECTED PLN1TS WITH RESPECT TO Ti!EIR RE5 TO TAmiAL, UPSE.T OR ACCIDErfT CONDITIONS NOR LEAD TO UNHEVIEWED SAFETY CONCERN 5.
THE NiTICIPATED TRANSIEUTS OF CONCERM ARE:
L LONS dE E;tTERtiAL ELECTRICAL LOAD 2.
TURBIME TRIP 3.
1.05S 0F PAIrl FEECMATER 4
LOSS OF CONDEUSER VACUUM
~
5.
IHADVERTEiTT CLOSURE OF PAIN STEAM ISOLATION VALVES (MSIV).
A MU?GER OF ALTEPJ!ATIVES UERE CONSIDERED IN DEVELOPING Tile ACT 3ELOM INCLUDING:
L rdSTRICTIid TEACTOR POWER TO A VALUE kNICH WOULD ASSURE NO AcruATION TiiE PORY.
Tile. REACTOR PROTECTION SYSTEH, DESIGN PRESSURE AND PORY SET-POINTS REtMINED AT THEIR CURRENT VALUES.
.2.
LOHERING THE flIGH PRES 50RE REACTOR TRIP SETPOINT TO A VM.UE mi[Cil ASSURE NO ACTUATI0tf 0F THE PORY.
THE DESIOi PRESSURE OF T!!E REACTOR Nia THE SETPOINT FDH PORY ACTUNOON REMAltiED AT TilEIR CURREriT VAltlES.
LOUERIt!G THE HIGil PRESSURE REAcr0R TRIP SETPOINT NfU ADJU
/
(
OPEP.ATING PRESSURE (AND TEXPERATURE) 0F Tire REACT 0H TO ASSURE ACTUATION NiG TO PROVIDE ADEQUATE IMRGIN TO ACC0hMODATE VNIIATI GPERATING PRESSURE.
Tile SEIPOItlT FOR PORY ACTUATION rem!NEO AT [T5 CURREKr VAtt/E.
THIS ALTERNATIVE WOULD REDUCE NET ELECTRICAL OUTPUT.
4.
ADdUSTI?G THE HIGH PRESSURE TRIP NiU THE FORY SETPOINTS TO ASSU POV ACTUATION FOR THE CLASS OF ANTICIPATED EVErlTS OF CONCCR PRESSURE OF THE REACTOR REMAINED AT ITS CURRENT VALUE.
THE DESIGN FIT EffALYSIS OF Tire INFACT OF THEsE VARIOUS ALTERNATIVES AND TO ASSURING THAT THE PORY WILL HOT ACTUATE FOR Tile CLASS OF CONCERM HAS BEEif COMPLETED. THE RESULTS SHOW.TilAT:
L' CT.!ERING THE HIGi P5 ESSURE REACTOR TRIP SET."0Itfr fROM 2355 PSIG TO_2200 PSIG AND PAISI?;G TifE SETPOINT FOR THE PILor OPERATED REllEF VALVE FRO?! P155 PSIG TO 2150 PSIfG.
Pr.0VIDES Tile REQUIRED ASSURNICE.
THIS ACTION HA5 TiiE FURTilER AUVNITAGES Or:
/1i 0 095
.c
= e.....
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r
-7.
- ' ~ '_-
l
}
EXTPACT OF B&W C0muMICATION - RECEIVED BY hRC P e 2 o' 4 4/20/79 e
t 3
'}.
ELECOCIt?G WE PROBASILIT7 0F POR'/ MiD A3HE CODE PRESSURIZER 3ATETY VALVE
~
ACT1% TION FOR OTHER UICREASIfD PRESSURE TP/J15IEriTS.
N.
FRESEir/ING PRESSURE RELIEF CkPkCITV FOR ALL HICJi PRESSURE TP#lSIEllTS.
~
ELUII?thTING THE POSSIB'ILIT/ DF IfffRODUCING UNREVIEMEO SAFETY C0tlCERiis.
4.
NEDUCDIG Tile THE AT ibi!CH THE STF_Nr SYSTEM llEAT SINr. Woulo UE LOST Irl '
THE EVEtiT EMERGEllCY FEEG;:hTER FLG9 HERE DELAYEU 3 StrdMARY 0" Tile IMPACT GF Tile PTOPOSED SETPourf CHNIGES 071 ALL AriTICIPATED TRAMSIEfiTS IS GIVEN Ill TABLE 7.
'BUI PIA?fT5 ARE CURREffT1.1 CAPASLE UF RthlBACX TU 15% OF FULL POWER UFO?! LOSS OF LOAD OR 'mIP OF THE TURB UIE.
THIS CAPAUILITY REQUIRES ACTUATI0tl OF Tile PILOT-CPERATED RELIEF VALVES.
THE CAPABILITY It! CREASES Tile PELIADILITY OF F0WER SUPPLY TO THE SYSTEM DY RETURIi1NG.' DIE UNITS TO POWER GEtlERATION DORE QU'CTLY AFTER THESE TRANSIENTS.
T11E ACTION PROPOSED ABOVE WILL REqulRE TilAT Tile REACTL ' BE TR1PPED FOR T15ESE EVEffiS-i
't!0TE:
s The effect of changing the reactor coolant system pressure trip setpoint upon peak pressurizer pressure is typified by the attached figure 1. which was developed by.
B&W for a loss of feedwater transient.
/
r.~
- 0 9 6.-.
.t 410.
+
,g
f'...
TABLE g SL'.%9T/ OF PROTECTION AGAIris T PORY ACTUAT r0TI
(
PROVIDED B7 PROPOSED SE'P0ltiT CHNIGES FOR ALL ANTICIPATED TANIS1E )TS EXTRACT OF B&WcCQ@Utl[CATIOli RECElVED BY tiRC 4/20/71
'h.
/bTICIPATED TRNASIEi;T3 miIC11 t! AVE OCCURRED AT D&W PildlTS /JiD 'cht!C11900LD I,DkNM.L7 ACTIVATE FORY AT THE CURRErfT SETP0f tfr (2255 PSIG):
A.
TUdBI32 TRIP L
tc55 0F EXTERi?d. ELECTRICAL LUA0 C.
f CSS OF Rh13 FEED 1lATER h.
LOSS DF CfX; DENSER VACUL?!
2.
CADVERTEffT CLOSURE OF N5I'1 2.
idfr:CIPATED TPRtsIEriTs h"rlICH !! AVE 6CcuRRED AT D?9 PLANTS Nip HilICH L'CU' D t ORP.*.LL7 ACTUATE PORV AT Tile PROPOSEO, SETPOITIT (2450 PSIG):
L
(
RCGI
- h JiHTICIPATED TRNiSIEriTS IDlICH llAVE il0T OCCURRED AT D5W PLNHd (LO '
P' R03ASILITY EVENTS) AND HiiICH ff0ULD f?ORMLLY ACTUATE FORY AT Ctr,tREnT SETFOINT (E255 PSIG):
A-S0hE CCitTRCL RCD GROUP WITHDMyALS (MODERATE TO IIIGO REACTIVITY
.FORTH GR0tTPS T;0T OTHEINI5E PROTECTED BY ifIG11 FLUX TRIP).
l B.
fGOEPATOR DIlifrION.
3"TICIP5TED TPMi3IEITIS MIIOl llAVE fiOT OCCURRED AT DFN PLNITS-(CUW T EVEff75) AND ll;IIC111TOULD ACTUATE TIIE PORY AT T1IE PROPO5ED SETPOIto (24S0 PSIG):
(
A.
SO'E C01(TROL R00 GROUP 1IITilDP.AJAt.s (liIGil nEACrIyITY !? ORT 11 r:0r OTHERiflSE PROTECTED BY HIGH FLUX TRIP),
7
I Enc'losure 1 c... -
t...
I..:
Page 4 of 4 EXTRACT 0 B&'d COMUNICATION - RECEIVED BY NRC
[
4/20/79 i
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i -V- ----m--
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a'-..
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Pea'< pressurizer pressure as a function of RCS pressure trip setpoint for a loss of feedwater transient for expected conditions and various initial pressures.
e.
Figu,re 1 f,,'
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