ML20236K380
| ML20236K380 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco, 05000000 |
| Issue date: | 06/20/1986 |
| From: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Heltemes C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| Shared Package | |
| ML20236K265 | List:
|
| References | |
| FOIA-87-377 NUDOCS 8708070097 | |
| Download: ML20236K380 (7) | |
Text
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'o UNITED STATES
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g NUCLEAR REGULATORY COMMISSION REGION V g
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g 1450 6AARIA LANE, SUITE 210 WALNtJt' CRE E K, CALIFORNI A 94596
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h JUN e 0 $6 MEMORANDUM FOR:
C. J. Haltemes, Jr., Director, Of fice for Analysis and Evaluation of Operational Data FROM:
J. B. Martin, Regional Administrator
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR FIRST QUARTER CY 1986 Your memorandum of June 16, 1986 forwarded a draft Commission Paper with enclosures, subject as above, for our review, comment and concurrence.
We have reviewed the draft Commission Paper, and particularly the enclosures as they relate to Region V facilities.
Enclosures 1 and 2 provide markups and rewrites of portions of the enclosures which reflect our comments. These comments were discussed by Jess Crews of my staff with Jack Crooks of your staff by telephone on June 20, 1986.
With those comments provided in EnclosuresJ d 2, we concur in the issuance of the A0 Report to the Congress as dr ed.
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loss of ac power to the remaining inplant leads.
The subsequent 4-minute loss l
of inplant electric power started the emergency diesel generators (which by design did not load), de-energized all safety-related pumps and motors, signifi-l cantly reduced the number of control room instrument indications available for operators to diagnose plant conditions, produced spurious indications of safety injection system actuation, and caused the NRC Emergency Notification System (ENS) phone on the operator's desk to ring spuriously.
Restoration of inplant l
electric power was delayed by an unexpected response of an automatic sequencer' (that should have established conditions for remote-manual access to offsite l
power still available in the switchyard).
The temporary total loss of steam generator feedwater was the direct result of the loss of ac power to the two main feedwater and one auxiliary feedwater cump motors, and the designed 3-minute warm-up period of the steam powered auxiliary feedwater pump. The loss of the feedwater pumps, in combination with the fail-ure of five feedwater check valves to close (one at the discharge of each feed-water pump and one in the feedwater line to each of the three steam generators),
allowed loss of inventory from all three steam generators arid the partial voiding of the long horizontal runs of feedwater piping within the containment building.
The sut, sequent automatic start of feedwater injection by the steam powered auxiliary feedwater pump did not result in the recovery of steam generator level i
I because the auxiliary feedwater being injected into the feedwater lines was flowing backwards through the failed check valves to the ruptured feed heater in the condensate system.
i l
Later, operators isolated the feedwater lines upstreas of the failed check valves, i
as required by procedure, unknowingly initiating the process of refilling the i
feedwater lines in the containment building.
As the auxiliary feedwater pumps l
refilled the feedwater piping to the steam generators, conditions were being i
established for a phenomenon that can generate destructive forces greater than 150,000 pounds-force. Since the feedwater piping to the steam generators had drained because of the failed check valves, the pipes contained water and steaa i
l at high temperature and pressure from the steam generators. As the auxiliary l
feedwater system filled the piping with relatively cold water, an instability occurred at the steam / water interface, which created a slug of water in the I
steam space. The slug accelerated at great speed, as steam was condensed in front of the slug, until it encountered an obstruction or a change of direction in the piping, such as at an elbow or closed valve. Upon contact, the slug imparted its energy to the piping with the force of a hammer low, i.e., a
/
condensation-induced water hammer.
Because of the long (203 feet) horizontal layout of the feedwater piping to the B steam generator and other sustaining conditions, this piping experienced the water hammer.
The forces from the water hammer displaced the 10-inch diameter feedwater piping, distorted its original configuration, c=:cd = 90-Sch crack, and damaged pipe hangers and snubbers. 4*-seeend, th: One-half ir.ch thick-j pip %g w:: irreversibly d=:ged--the T-inch crc k, 30 perc-hrough th all-at-places 7-4ndicate; beu clete the pipe had ccme te sp'ittiv en,-
Outside the containment building, the forces associated with the water hammer were enough to stretch 10 one-half-inch diameter bolts holding the bonnet on a 4-inch bypass check valve by about one-half inch. All of the bolts were stretched into an hour glass shape.
The steam and water from the check valve body to bonnet interface had sufficient force to blow away the insulation from 1
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l Fut.ure reports vill be made as appropriate.
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86-2 Loss of Integrated Control System Power and Overcooling Transient
)
1 The following information pertaining to this event is also being reported con-
)
currently in the Federal Register.
Appendix A (see the third general criterion) j of this report notes that major deficiencies in design, construction, use of, or management controls for licensed facilities or material can be considered an j
i abnormal occurrence.
I Date and Place - On December 26, 1985, Rancho Seco Nuclear Generating Station, I
located in Sacramento County, California, experienced a loss of de power within the integrated control system (ICS) while the plant was operating at 76 percent power.
Following the loss of ICS dc power, the reactor tripped on high reactor coolant system (RCS) pressure followed by a rapid overcooling transient and l
automatic initiation of the safety features actuation system'on low RCS prissure.
l The overcooling transient continued until ICS dc power was restored 26 minutes i
after its loss. The significance of the event is that a nonsafety related sys-tem failure initiated a plant transient which could have been more severe under other postulated scenarios.
i j
The Rancho Seco Nuclear Generating Station, operated by the Sacramento Municipal i
Utility District (SMUD), is a Babcock & Wilcox (B&W)-designed pressurized water l
reactor.
l Nature and Probable Consequences - At 4:14 a.m. on December 26, 1985, the plant was operating at 76 percent power, when a loss of ICS de power occurred as a l
l result of a single failure.
The loss 6f de power to the ICS (a nonsafety-related i
system) caused a number of feedwater and steam valvbs to reposition automatically l
and also caused the loss of remote control of the affected valves from the con-trol room.
In addition, the main feedwater (MFW) pump turbines slowed to minimum speed and the auxiliary feedwater (AFW) pumps started. The immediate result was a reactor coolant system (RCS) undercooking condition that resulted in the reac-tor tripping on high pressure.
The reactur trip was followed by an overcooling condition that resulted in safety features actuation and excessive RCS cooldown.
The transient was initiated by the failure of a single module in the nonsafety-related ICS (i.e., the spurious tripping of the power supply module that inter-rupted all +/-24 Vdc power).
The most probable cause of this failure was a demon-weaknes; that opparently acdc the circaR-susceptible tc creatic cpcrc-
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igh reci;tance ccnnection (i.e. a
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moqitor which = pen}d the deMgn tecaknen end caused the module to tWfh 6en66 bad crimp connection in the wiring between the +24 Vdc bus and the power supply 4Adefvolta96 and m+ctt Upt all de ppwcf.
The operators were not immediately able to restore de power within the ICS.
As a result, nonlicensed operators were sent to isolate the affected steam and feedwater valves locally with handwheels.
During the first 7 minutes of the incident, the excessive steam and feedwater flows resulted in a rapid RCS cool-down of over 100*F.
The pressurizer emptied and a small bubble formed in the reactor vessel head.
The RCS cooldown continued and the RCS depressurized to l
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l loss of ac power to the remaining ir. plant loads.
The subsequent 4-minute loss of inplant electric power started the emergency diesel generators (which by design did not load), de energized all safety-related pumps and motors, signifi-cantly reduced the number of control room instrument indications available for operators to diagnose plant conditions, produced spurious indications of safety injection system actuation, and caused the NRC Emergency Notification System (CNS) phone on the operator's desk to ring spuriously.
Restoratica of inpla l
electric power was delayed by - -~
- - - - - - of an autcmatic sequen er y
i pthat should have established conditions for remote-manual acces,s to offsite v
l g ggp c]
M powerstillavailableintheswitchyard%
g The temporary total loss of steam generator feedwater was the direct result of l
l the loss of ac power to the two main feedwater and one auxiliary feedwater pump l
motors, and the designed 3-minuta warm-up period of the steam-powered auxiliary feedwater pump.
The loss of the feedwater pumps, in combination vith the fail-ure of five feedwater check valves to close (one at the discharge cf each feed-water pump and one in the feedwater line to each of the thre,e steam generators),
allowed loss of ir.ventory from all three steam generators and the partial voiding i
of the long horizontal runs of feedwater piping within the containment t,uilding.
l The subsequent automatic start of feedwater injection by the steam powered auxiliary feedwater pump did not result in the recovery of steam generator level because the auxiliary feedwater being injected into the feedwater lines was flowing backwards through the failed check valves to the ruptured feed heater l
in the condensate system.
Later, operators isolated the feedwater lines upstream of the failed check vrives, as required by procedure, unknowingly initiating the process of refilling the feedwater lines in the containment building.
As the auxiliary feedwater purcps refilled the feedwater piping to the steam generators, conditions were being established for a phenomenon that can generate destructive forces greater than 150,000 pounds-force.
Since the feedwater piping to the steam generators had drained because of the failed check valves, the pipes contained water and steam at high temperature and pressure from the steam generators. As the auxiliary feedwater system filled the piping with relatively cold water, an instability occurred at the steam / water interface, which created a slug of water in the steam space. The slug accelerated at great speed, as steam was condensed in front of the slug, until it encountered an obstruction or a change of direction in the piping, such as at an elbow or closed valve.
Upon con, tact, the slug y
imparted its energy to the piping with the force of a hammergb ow, i. e., a p
condensation-induced water hammer.
y, Because of the long (203 feet) horizontal layout of the feedwater piping to the g
B steam generator and other sustaining conditions, this piping experienced the
/
water hainmer. The forces from the water hammer displaced the 10-inch diameter M-feedwater piping, distorted its original configuration, caused an 80-inch crack,
/
and damaged pipe hangers and snubbers.
In seconds, the one-half inch thick l %Wf piping was irreversibly damaged--the 80-inch crack, 30 percent through the wall at places, indicates how close the pipe had come to splitting open.
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Outside the containment building, the forces associated with the water hammer were enough to stratch 10 one-half-inch diameter bolts holding the bonnet on a 4-inch bypass check valve by about one-half inch. All of the bolts were stretched into an hour glass shape.
The steam and water from the check valve body te bonnet interface had sufficient force to blow away the insulation from 2
all the piping located 360 degrees around the check valve.
The significant steam and water leak from this check valve constituted the second leak in the
- event, The design of the steam system at Unit I has the three steam 1)nes joined into a common pipe (or steam header) inside the containment building without any valves to prevent simultaneous blowdown of all three steam generators should a leak in a steamline or a feedwuter line occur.
Hence, the leak from the B feed-water bypass check valve located outside the containment building communicated with all three steam generators, via the steam header and B feedring, and their steam inventories were vented via the leak to the atmosphere.
In addition, the auxiliary feedwater flow to B steam generator escaped from this leak instead of i
going to the steam generator.
Despite these problems, operators later succeeded in recovering level indication in the two steam generators not directly associated with the feedwater piping leak. With the reestablishment of steam generator levels, the operctors safely
/
brought the plant to a stable cold shutdown c ion without a significant release of radioactivity to the environment t d re'ex,isting primary to secondary y leak was not exacerbated) and without significant additional damage to plant '
equfpment.
Cause or causes - The snost significant aspect of the event ws.s that five safety-M hted feedwater systen check valves degraded to the point of inoperability during a period of icss than a year, without detection by the licentoa, and that their failure jeopardized the integrity of safety-related feedwater piping.
The root causes of the check valve failures were a combination of inadequate maintenan:c, inadequate intervice tetting, inadequate design, and inadequate consideration of the effects of reduced power operations.
Actions Take to Prevent Recurrence Licensee - The licensee has undertsken an extensive _ study (including testing programs) of the multipic fcilures associated with the event to determine root causes and effective cofrective acticns to preclude facurrence.
On April 8, 1986, subsequent to several meetings with NRC staff and the Commis-sfon, the licensee submitted a comprehensive report (Ref.1) documenting the results of their investigations to that aate and providing some conclusions and corrective actions being implemented. The licensee committed to provide adds-tional information by April 30, 1986.
)
The licensee concluded that the most likely cause of the cable failure dich initiated the event was temperature-induced degradation due to the presence of' I
local heat sources such as hot pipe flanges.
Additionally, the licenses concipded that the failure of the five check valves was caused by (1) their proximity to turbulent flow, (2) the fact that the valves were oversized and therefore did not remain fully open in normal operation, (?) the design by which the valve disc was fastened to the valve hinge, and (4) extended reduced flow operation at 90% power which exacerbated the effects of the design deficiencies.
The licensee's actions described in the April 8,1986 report were extensive and l
included examinaticas and corrective actions in the creas of testing, procedures development, training, maintenance, quality assurance, emergency preparedness, post-trip review and safety review programs.
3 a
Future reports will be made as appropriate.
86-2 Loss of Integrated Control System Power and Overcoo'ing Transient The following information pertaining to this event is also being reported con-currently in the Federal Register.
Appendix A (see the third general criteriori) of this report notes that major deficiencies in design, construction, use of, or management controls for licensed facilities or material can be considered an abnormal occurrer.ce.
Date and Place - On December 26, 1985, Rancho Seco Nuclear Generating Station, located in Sacramento County, California, experienced a loss of de power within the integrated control system (ICS) while the plant was operating at 76 percent Following the loss of ICS dc power, the reactor tripped on high reactor power.
coolant system (RCS) pressure followed by a rapid overcooling transient and automatic initiation of the safety features actuation system'on low RCS prissure.
The overcooling transient cottinued until ICS de power was restored 26 minutes after its loss. The significance of the event is that a nonsafety related sys-tem failure initiated a p? ant transient which could have been more severe under other postulated scenarios.
The Rancho Seco Nuclear Generating Station, operated by the Sacramento Municipal Utility District (SMUD), is a Babcock & Wilcox (B&W)-designed pressurized water reactor.
Nature and Probable Consequences - At 4:14 a.m. on December 26, 1985, the plant was operating at 7G percent power, when a loss of ICS de power occurred as a result of a single failure. The loss of de power to the ICS (a nonsafety-related system) caused a number of feedwater and steam valves to reposition automatically and also caused the loss of remote control of the affected valves from the con-trol roorie.
In addition, the main feedwater (MFW) pump turbincs slowed to minimum speed and the auxiliary feedwater (AFW) pumps started.
Tne immediate result was a reactor coolant system (RCS) undersooling condition that resulted in the reac-l tor tripping en high pressure.
The reactor trip was followed by an overcooling condition that resulted in safety features actuation and excessive RCS cooldown.
i The transient wts initiated by the failure of a single module in the nonsafety-i related ICS (i.e.,, the spuricus tripping of the power supply module that inter-j rupted all +/-24 Vdc power).
The most probable cause of this failure was
- tn meratic nnera-design- =akness that ppean+1y mada t hs.,g4mit ;gcept 4 hl o one 49 vac ous enu-n e p;;r &nniy monitor j
tisn_if "cented s wststence"-treiween vaca_to-devM up, anu Life beveiopmenc vi a istgn re3i dent -cc ectinn U ^- a l
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/ bad crimp connection) in the wiring between the +24 Vdc bus and the power supply
/ monitor which 27"('" S:f p.c ' ir W caused the module E
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j The operators were not immediately able to restore de power witt; n the ICS.
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As c. result, nonlicensed operators were sent to isolate the a,fected steam and f
feedwater valves locally with handwheels.
During the firstJ minutes of the l
incident, the excessive steam end feedwater flows resulte M n a rapid RCS cool-j i
down of over 10G'F. The pressurizer emptied and a smal ubble formed in the l
reactor vessel head. The RCS cooldown continued and e RCS depressurized to l
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