ML20052F301
| ML20052F301 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Browns Ferry |
| Issue date: | 11/03/1981 |
| From: | NRC |
| To: | |
| Shared Package | |
| ML19219B002 | List: |
| References | |
| FOIA-81-417 NUDOCS 8205120328 | |
| Download: ML20052F301 (61) | |
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n POSITION PAPER ON SCRAM DISCHARGE VOLUME (SDV) INDEPENDENCE FROM VENTING FOR SAFETY FUNCTION BACKGROUND The most significant feature that contributed to the Browns Ferry Unit 3 (BF3) partial scram failure which is ccmron to most BWRs, is the inadequate hydraulic coupling between the SDV headers and the instrumented volume (IV). This finding is shared by the NSSS supplier (GE), the BWR owners and NRR, IE and AE0D staff.
To improve the overall design of the SDV system and particularly the hydraulic coupling of the SDV headers and the IV, an NRR task force with IE participation has been working with a subgroup of the BWR owners to develop revised SDV system design and safety criteria. This effort has identified two major differences in opinion between the task force and owners subgroup. This paper is to document our position and reasoning for one of these differences.
POSITION Our position on the design criteria for SDV to IV hydraulic coupling is that the SDV system design shall function independent of its vent system given up to a maximum possible inleakage into the SDV headers.
(System function is defined as: The SDV shall have sufficient capacity to receive and contain watar exhausted by a full reactor scram without adversely affecting scram performance.) Our position varies from the owners subgroup by the phrase that requires independence from vents. The owners subgroup, while not ruling out systems independent of SDV vent, would prefer to permit as an option, a design which precludes loss of system function due to a single active failure
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2 in the SDV vent. AE0D has recommended independence from vents.
JUSTIFICATION Our position is based on several points:
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GE has analyzed the system extensively end has recommended a system configuration which incorporates. an improved hydraulic coupling b,:t ieen the IV and SDV.
In our opinion this improvement would permit operation of the SDV without reliance on its vent and is due to system difficulties from vent interfaces. The proposed GE design is to attach directly to the SDV header a large diameter instrument volume.
This assures immediate hydraulic communication betwee' the SDV beede s and the IV.
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Further, the GE recommendation does not use a previously untried system configuration and thus decreases the probability of unobserved adverse i'
ef fects, e.g., water hammer. This design is currently in place at 5 operating reactors.
It has been subject to in:trumentation damage, but these difficulties are not the result of the SDV-IV hydra J11c coupling arrangement and are overcome by other agreed upon design cr.teria.
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The allowance of credit for vents would open to interpretation many hydraulic design assumptions. The adequacy of such a design would be subject to substantial uncertainties.
The owners subgroup recommended design criteria is to allow for individual plant flexibility in design.
For example, some BWR plants may wish to replace the 2 inch diameter pipe from the SDV header to the IV with a larger pipe, and an improved vent system. - This has an advantage, in that, it increases the volume available for scram exhaust. The disadvantages be in the fact that it is an unproven more complex design. For the entire range of possible operating conditions of this new system design, there is substantial uncertainty in pre-
r-3 dicting system response.
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We'_ve mada estimates of the pipe size and protective system adequacy for the range of current slopes and lengths from the SDV headers to the IV.
The maxi-mum possible inleakage without rod motion is 5 gallons per minute (GPM). Summed e
over all scram discharge lines into a scram discharge header for a large BWR this amounts to about 465 GPM. For the slope of 1/8" per foot driving head, s
a pipe of about 8 inches in diameter would be required to pass the total inleakage. This is about the practical' limit for pipe size in this application.
Further, for a pipe length of about 150 ft which is the maximum it would take about 40 seconds for the inleakage to reach the IV assuming a functioning vent. At this time, inleakage may be sufficient to adversely affect scram so that the protective function response would be marginal. This calculation shows that even with an adequate vent function major system configuration changes may be required and that testing system response would be needed because of relative large uncertainties and small available margins.
Another potential source of uncertainty arises from hydrodynamic forces generated during system operation. Water hammers which damage system components and structures have been observed on SDV systems. A new design with larger pipes may also be subject to flow induced hydrodynamic forces that could damage the system and preclude its function. This also demonstrates the need for testing of system performance.
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. To test system response would require an extensive, elaborste test program.
For example, a combination of variable inleakage rates into each SDV header over.the range of possible inleakage would be required with measurement of protective instrument response and system parameters important to system function including associaged hydraulic forces on system components and structures to assure the absence of water hammer. The evaluation of such a complex test program would require a longer period of time than is con-sidered prudent for such an important system and the. risk of failure of such an undertaking are undue from a regulatory view point.
CONCLUSION A new SDV system design, which relies on a vent system, unnecessarily complicates system function and somewhat decreases the probability of success.
For such a vital safety function as provided by the SDV, the incumbant disadvantages and delays associated with potential designs which use the vent are not apprnpriate particularly when a solution sucn as proposed by GE is readily available and tested. Therefore, the independence of the~SDV system from vent performance supplies a ready resolution to the hydraulic coupling problem.
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PREFACE The findings, recorrendations, and conclusions contained in this report are, for reasons of timeliness, based on information gathered through infomal channels between the Tennessee Valley Authority, the General Electric Company, and the US NRC Headquarters and Regicnal offices. To i
the extent possible, the information used in the recort has been verified by cross checking with other aurces. The findings contained in this' recort, including the underlying causes of the partial scram failure which occurred at Browns Ferry Unit No. 3 (BF-3) on June 23, 1980, relate most d:rectly to the Browns Ferry reactor. However, similarities among boiling _
water reactor facilities leads us to believe that the findings and rec-cmmendations may be broadly and generically applied to most if not all operating BWRs. To this end, we recommend that a plant-by-plant review, not possible in this investigation, be undertaken by others, to assess the icplicability of these findings and recommendations to other ShRs and to provide analysis and evaluation of plant-unique design problems not un-covered in this investigation. Additionally, the scope of our investi-gation and recommendations was intentionally limited so as to address only the specific, direct and underlying causes of the partial scrara failure at 3F-3. We have not, therefore, taken the broader view, as could be taken by those most directly involved in the ATWS issue. We do believe, I
however, that some of the information presented in the report can be useful to those involved in this important generic concern. Finally, this in-vestigation was not able to pinpoint a single crecise root cause(s) wnich led to the BF-3 partial scram failure event, beyond to say it was caused by water in the scram discharge volume. However, we believe that, in totality, the various possible cause mechanisms discussed in this report include the actual, albeit, indeterminable root cause(s) of the event.
As a footnote, the writeri wish to acknowledge the invaluable and timely information provided by the SF-3 resident inspectors, James Chase and Robert Sullivan, without whose cooperation, timely issuance of this report would not have been possible.
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Ct TABLE OF CONTErlT5 Page PREFACE................................................................
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INTR 000CTI0ft........................................................
I 2 EVE!!T SEQUENCE......................................................
3 3 CESIGil AND CPERATION OF THE 3ROWNS FERRY UtlIT 3 SCRAM SYSTEM........
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? CAUSES INVESTIGATED.................................................
11 5 EVENT SEQUENCE ANALYSIS.............................................
14 6 SCRAM DISCHARGE VOLUME / SCRAM INSTRUMENT VOLUME INSPECTIONS AND TESTS.........................................................
19 7 PREVIOUS BWR EXPERIENCE OF FAILURE TO FULLY INSERT..............,...
23 3 FINDINGS............................................................
24 9 RECOMMENDATIONS....................................-................
35 10 CONCLUSIONS
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40 LIST OF FIGURES Ficure 2-1 Control Rod Posi tions Before Fi rst Manual Scr am...................
43 2-2 Control Rod Positions After First Scram...........................
44 2-3 Control Rod Positions After Second Scram..........................
45 24 Control Rod Position After Third Scram...........................,
45 3-1 Control Rod Drive.................................................
47 3-2 S c r am E l e c t r i c a l D i a g r am..........................................
48 3-3 Control Rod Scram Grout Assignment................................
49 3-4 S cr am V a l ve Arr an gemen t...........................................
50 3-5 Scram Volume Drain Arrangement....................................
51 LIST OF TA8LES Table 2-1 Event Secuence Recorder Printout.................................
41 5-1 Scram Discharge VolL.,e C*ain Time and Total Positiens Inserted....
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1 INTRODUCTION On June 23, 1980, the Browns Ferry 3 reactor experienced a partial f ailure of the scram system, while shutting down for a scheduled maintenance of the feedwater system. The reactor had been brought down to aporoximately 35%
power by reducing recirculation flow and by manual insertion of control rods. The subject event occurred when the control room operator initiated a manual scram to make the reactor subcritical which was the next step in the normal shutdown evolution. After manual scram actuation, the control rods on the West side of the core were observed.to be fully inserted. Hcw-ever, th? control rods on the East side of the core did not fully insert.
Yost of the East side rods came to rest in notch positions ranging between 00 anc 36 sfter all East side rod motion had ended.
Three additional scrarm and ahcut la minutes were required to achieve full insertien of the cartially withdrawn East side control rods. After all rods were com-oletely inserted, the operators resumed normal shutdown ocerations.
On July 2,1980, a team of NRC Feadquarters representatives from IE, NRR, and AE00 went to the Browns Ferry-site to gather detailed infcrmation on the event, the scram system design and operation, and the results of scram system tests which alraady had been performed by TVA personnel. With this initial direct contact at the plant, an independent investigation of the event cause and the recommended corrective actions was begun by the Office for Analysis and Evaluation of Operational Data (AECD). Over the next several days, additional. equipment testing was performed on the SF-3 scram system. Testing and analysis was also being conducted during this time by General Electric in San Jose, California to support TVA activities at the plant. During this period, AE00 continued to obtain, analyze, and evaluate information as it evolved from these and other sources to continue'its investigation.
The purpose of this report is to provide the analysis, evaluation, findings, o
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and recommendations which flowed from the investigation of the SF-3 event by the AE00, US NRC. Section 2 of the. report contains an event sequence.
Section 3 provides a description of the design and operation of the BF-3 scram system. Section 4 discusses the possible causes of the event which were investigated and the conclusions in each case. Section 5 provides an event sequence analysis; Section 6 orovides a summary of the tests and inscections performed at SF-3 which support the event sequence analysis-and some of the findings. Previous coerating experience and investigation findinos are contained in Sections 7 and 8, respectively. Specific recommendations to correct the deficiencies discussed lin the findings are provided in Section 9.
The conclusicns of this investigation are given in I
Section 10.
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2 EVENT SEQUENCE On June 28, 1980, power was being reduced by the control room operator at the Browas Ferry Unit 3 nuclear reactor in preparation for a scheduled shut-down for feedwater system maintenance. By 0131 hours0.00152 days <br />0.0364 hours <br />2.166005e-4 weeks <br />4.98455e-5 months <br />, the reactor power had been brought to 390 Ne via decreased recirculation flow and manual con-trol rod insertion. The operating personnel then initiated a manual reactor scram to complete insertion of the remaining control rods (which were at the positions shown in Figure 2-1 at the time) and thereby bring the reactor to a subcritical state.
Irredi:tely after decressing the manual scran buttons, the operat.,rs placed the reactor mode switch in the SHUTDOWN mode. Control room personnel ob-served that the blue scram lights for all. control rod drive scram inlet and outlet valves were illuminated, indicating that all scram valves were open.
Control rod position indication also showed that all of the rods on the West side of the core were fully inserted (except for one which had stopped at-position "02").
However, position indication showed that 75 rods on the East side of the core were not inserted fully. The East side control rods came to rest at oositions ranging from 46 to CO withdrawn with an average of about 23 positions withdrawn (position 48 corresponds to fully withdrawn). Rod position indications following the first manual scram are shown in Figure 2-2.
At this time IB rcds on the East side were fully insarted. As estimated by the LPRM readings, power level on the East side of the core following the first scram appeared to be less than two percent.
Following scram, the Scram Instrument Volume began to fill and the Scram In-strument Volume Hi level Scrim (level switches) actuated. This occurred some-what sooner than expected at about 19 seconds. The Hi Level scram condition was subsequently bypassed by the operator (as allowed in SHUTOOWN mode), to permit reactor protection system reset which occurrec 4 minutes and 31 seconds following the first scram.
One minute and 33 seconds later a second manual scram was initiated by the i l
O operator. The time following reset allowed partial drainage of the East and West Scram Discharge Volumes. Rod positions following the second scram are shown in Figure 2-3.
After this scram, 33 rods on the East side fully inserted. The second manual scram was reset af ter 59 seconds and the scram discharge volume was allowed to drain for 53 seccnds at which time a third manual scram was actuated. Upon completion of this scram, 47 rods on the East side were fully inserted. Rod positions following the third manual scrim are shown in Figure 2-4 The third scram was reset after 3 minutes and 26 seconos. The scram discharge level bypass switch was returned to normal 2 minutes and 40 seconds later. This action initiated a fourth, automatic scram due to 3 Scram Instrunent Volume Hi Level scram condition which had not cleared. At this time all rods on the East side were fully inserted.
A detailed sequence of events as provided by the event sequence recorder is shown in Table 2-1.
The total elapsed time between the initial scram and
- inal inserticn of all rods was la minutes 2 seconds. At this time the coerators continued normal shutdown operations.
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3 DESIGil Afl0 OPERATIOff CF THE BROW!!S FERRY UilIT 3 SCRAM SYSTEM
'Jecharsical and Hydraulic Desicn of the Scram System On a GE SWR, such as Browns Ferry Unit No. 3, the Control Rod Drive (CRD) and its associated Hydraulic Control Unit (HCU) provide the means by which each individual control rod can be rapidly inserteo upward into the core during a reactor scram. A simplified drawing of the CRD mechanism is shown in Figure 3-1.
During periods of no rod motion, the collet fingers are spring loaded into a groove on the index tube to hold the drive stationary against the force of gravity. High pressure cooling water is applied below the drive piston and equalized without CRD motion via controlled inleakage past the CRD seals and into the reactor. A CR0 temperature proce is provided internally to monitor each CRD to detect CRD heat-up should cooling water flow be interrupted or should excess leakage of high temperature RCS water fi n cut through the drive, drive insert line and scram outlet valve. Scram outlet valve leakage into the scram discharge volume on the order of 0.1 ccm would raise the probe temper-ature to the alarm setpoint of about 350 F.
At SF-3, water exhausted from the CRDs is routed to either an East or West header scram discharge volume. The scram discharge volume (SDV) is sized to provide a volume of approximately 3.3 gallons per CR0 (approximately 600 gallens total). The SDV volume is sized to limit the total amount of hot reactor water leakage past the seals during a reactor scram (maximum volume requirement) while providing enough free space at atmospheric pressure so that back pressure on the CRDs does not increase so rapidly as to impede rod insertion speed (minimum volume requirement).
In particular, the system design results in a pressure in the SDV imediately following full insertion rod motions of less than 65 psig.
Low pressure in the SDV is necessary to assure that technical specification scram speeds and full-in rod motion are achieved. The volume of water exhausted through the scr, outlet valve of a single normal drive for a full stroke is about 0.75 gallons, not including seal leakage and Jypass flow. The'1eakage 3
e and bypass flow for a single drive can be in excess of 5 gallons per minute.
Normal scram time. from full out to 90 percent insertion is less than 3 seconds.
Although the SDV is sized for a volume of approximately 3.3 gallons per drive and the drive stroke (without bypass) is cnly approximately 0.75 gallons, only a sinole reactor scram is nonnally possible with respect to the scram discharge volume. Leakage of reactor water past the seals fills the SDV rapidly as long as the scram outlet valves are open which would be the case without an RPS reset. This leakage occurs even on rods that are fully inserted. The leakage is an average of 2 gpm to 3 gpm per CRD. Thus, from this curce alone, the 3.3 gallons per drive of free volume available in the SDV is filled anc pressurized within two minutes. Thus, more than one scram would be possible only if the operator were able to reset the scram (closing the scram octlet valves) well within this time period. Without an early reset, the 50V would be filled and the SDV would have to be drained to attempt a rescram if rod motion is to be produced.
The East and West SDV headers are each provided with a vent line and vent valve.
Each header drains via a separate drain line into a scram instrument volume (SIV) where level monitoring instruments are located. The SIV, in turn, has drain piping and a drain valve.
During normal operation, the vent valves of the East and West SDV headers and the crain valve of the SIV are open. These valves are kept open to allow the leakage past the scram outlet valves entering the SDV to drain continuously into the SIV so that no build-up of water in the SDV occurs which could prevent 4
3 reactor scram. These valves close curing control rod scram inserticn to contain and limit the reactor water released through the scram outlet valves.
During a scram, inflow of water to the SDV normally ccntinues after control rod insertion is canpleted due to leakage past the CRD seals. Leakage continues until the scram is reset or until the SDV pressure eouilibrates with reactor pressure.
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l A pressure. difference of at least 550 psi must be applied between the CRD insert and withdraw lines to drive the rods in during a scram. The pressure difference applied at the beginning of a scram is provided by the 1500 psia scram accumulator and atmospheric pressure in the empty SDV. As CRD scram
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insertion progresses, pressure losses in the driving fluid due to line losses reduce the insert line pressure to below reactor coolant system pressure.
At that time, the ball check valve, integral to the CRD, allows reactor coolant system water to come in under the piston to complete the scram, before any
.significant build-up in scram discharge volun.e pressure due to filling frcm-leakage and bypass flow.
RPS Electrical Design A simplified schematic Jf the electrical components of the Reactor Protection System (RPS) is shown in Figure 3-2.
It is divided into two independent trip channels A and B. -Each of the channels can be tripped (de-energized) by either the. manual rcram relays or the two subchannel relays. The subchannel relays-are de-energued and opened whenever any one.of a variety of trip conditions exist in the reactor or associated equipment. The automatic logic can be described as "one-out-of-two taken twice." For purposes of analysis of the Browns Ferry event, the automatic trip logic will not be discussed because this event occurred first with a manual scram.
I With reference to Figure 3-2, both scram solenoid valves A and B must change position to provide a scram. Electrically, this recuires a trip of both channel A and channel S.
De-energizing the two scram solenoids e,hanges the air flow.
from the control air supply to the vent path. For manual scrams, a separate scram button is provided on the control panel for, each channel.
'A manual scram is initiated by depressing both the channel A scram button and the channel 3 4
scram button. Because of the power recuirements of 185 separate scram solenoid valves on each channel, each channel is divided electrically into 4 separate scram groups. Centrol rods associated with the HC'Js from the four groups are distributed randomly througnout the core as shown in Figure 3-3.
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Scram Operation r
The Reactor Protection System performs its design functicn by de-energizing the 370 scram solenoid air supply valves (2 for each centrol rod drive HCU),
de-energizing the two scram discharge volume (SOV)-air supply solenoid valves, and energizing the four backup scram solenoid valves in the air supply lines as shown in Figure 3-4 Scram insertion is achieved for each individual control rod by opening the scram inlet andascram outlet valves. This applies 1500 psi accumulator pressure t
to the " insert" Side of the control rod drive piston and vents the " withdraw" sice of the piston to the SDV which is at atmospheric pressure.
4 For normal unscrammed conditions, the scram inlet and cutlet valves are held shut by control air pressure applied through the energized scram air supply valves (S39A and 539B in Figure 3-4).
The SDV vent and SIV drain valves are held open by air pressure appiled through the energized discharge volume air supply i
valves (S37A ano 5378). The air header which supplies control air to all of the 372 air supply valves (370 scram, 2 vent / drain) is pressurized through de-energized backup scram valves (S35A and $358, 57CA and S708). The SDV vent and SIV drain valves can be operated manually from the controi room.
A scram signal de-energizes both air supply valves for each cod, de-energizes the scram discharge volume air supply valves, anc energizes the back-up scram valves, thus venting air pressure from the scram inlet and outlet valves and the SOV and S:V valves. This causes the scram valves to open and the SUV vent and SIV drain valves to close.
In the event the individual control rod air supply valves should fail to change position -(i.e., mechanical bind-up, etc.),
the back-up scram valves which were energized and vented air to depressurize the air supply header assure ooening of the scram valves. Thus, even if an air supply valve failed to shift, that rod woula still scram. A check valve is
- provided arcund the downstream back-up scram valve in the air supply line so the upstrern valve can assist in the air header venting or assume venting in case the downstream valve fails.
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Physical-Layout of the Scram System Hycraulic Comocrents at Browns Ferry Unit 3 At Browns Ferry Unit ilo. 3, the HCUs for all of the CRDs are physically arranged i
in rows on the " East" and " West" sides of the reactor vessel, outsice the drywell; l
and inside the reactor building. The CRDs on the West side of the core are controlled by the West side HCus and the CRDs on the East side of the core are controlled by the East side HCU;. Drives along the interface centerline, f.
between the East and West sides of the core, are alternately routed to the East and West headers. A simplified diagram of the physical arrangement of the HCUs, scrmn discharge volume, and vent and drain system is shown in Figure 3-5.
t The WCus on each side of the reactor are arrange] in a rowr.
- 1mer.i ne / acove the a rows of HCUs are tuo crcss connected " race
- rack" shaceo huaders f aori-cated with 6" piping into which the discharge fr.9 A;ch scr;m outlet valve is otoed. The two connected G" headers on the ! st side compris2 the Elst bank scram discharge volume (SOV) and the two connected 6" headers on the West side i
ccmprise the West back scram dischargo volume. Each HCU insert and w thdraw line is conriected to the C'.Os celow the Tctor vessel witn 3/4" Senacule 30 l
piping through which the scram inlet and scram outlet water flows (and water for normal rod drive motion). These lines average over 50 feet in length.
The lines from the HCU scram outlet vcive to the SDV are fabricated with 3/4" 5checule 30 oicing and are approximately 10' in length.
The Scram Instrument Volume (SIV) is located on tne West side of the reactor at one end of the West side HCUs (and SDV).
It is ccnfigured as a 12" diameter 10' high vertical cylirder. Single f? oat-type level switches are installed to monitor the 3 gallon and 25 gallon levels. Four float-type level switches are provided at the 50 gallon level for the purpose of ~ initiating a reactor. scram -
(SIV Hi Level Scram) before the 50V begins to fill beyonc tne point where I
- omplete control rod insertion would be prevented.
At 2 owns Ferry Unit 3, the East bank and West bank SDV sach drain via 2" I
schedule 160 pipe to a single SIV located cn the West side. The drain line for the West bank is approximately 15' long while that from the East bank is approximately 150' long.
In each run, the total elevation fall in the line
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1 is approximately'l' 7".
On the East bank run this is an average 0.13" fall pe. foot of horizontal run.
The drain line piping at the bottom of tne scram instrument volume and the vent line piping at the high points of the slightly inclined East bank and West bank 50V headers are routed down to the Clean Radwaste Drain (CRW) piping physically located in the reactor building floor. The CRW system is a closed drain' system which discharges underwater in the Reactor Building Equipment Drain Sump at the lowest elevation in the reactor building. Many other equipments are drained l
and vented by this system.
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CAUSES INVESTIGATED Immediately following the event at Browns Ferry 3, all aspects of the scram system were investigated in an effort to find the cause. The Reactor Pro-tection System (RPS), the air system, mechanical aspects of the CRD and various valves, the CRD and HCU hydraulics, and the possibility of air-in l
the hydraulic system were considered. Finally, attention was focused on the East bank Scram Discharge Volume.
4 Electrical Investications Following the first manual scram, the operators verified that the blue scram lights were illuminated for all control rods. Both the scram inlet and the outlet valve stem position switches must show an open valve position to illuminate these lights.
Illumination of these lights for all CRCs would 1
indicate th_at the electrical portion of the RPS had successfully generated a scram signal to coen all scram solenoid valves and that all scram valves
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had actually ocened for all control rods.
The Reactor Manual' Control System (RMCS) which is designed to control only one control rod at a time was reviewed to determine if there could have been possible interference with the scram function.
It was determined that postulated gross failure of the RMCS and initiation of multiple control rod drive withdrawal signals would not prevent insertion during scram since upward scram forces are more than three times the magnitude of the with-drawal forces under these conditions.
By use of reference drawings, hydraulic control units from each of the four rod scram groups. vere verified to be randomly positioned on both sides of the core as shown in Figure 3-3.# Control rod electrical signals to a group 1 red on the E3st side of the core and a group 1 rod en the West side of the core would be identical. The red inserticn pattern during the event shows that on the East side a number of rods from each electrical group did not completely insert while on the West side, rods from all electrical groups did completely insert.
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Based on this analysis it was concluded that the failure of rods to
. fully insert only on the East side was not causec by any electrical mal-function in the RPS trip logic.
TVA test (entitled SMI 150) was performed to verify that the response times for the scram actuating relays te fully de-energize were within technical specifications. Verification that they were, eliminated the concern that I
an electrical croblem delayec opening of the East side scram valvts which in turn resulted in partial insertien.
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A test of the voltage on all channel A and - h mnel 3 scram gecuor showed that all went to zero following a manual scram and all returnec to 125. AC when reset. This test was run to verify the requirements of US f;RC IE Bulletin 80-17. A visual and electrical search of the scram circuitry cabinets for spurious voltage sources and loose wires (that might have provided a path for alectrical power) to prevent dropout of the s c n relays was cerformeo by TVA. Mone was found.
"?] Tests Various tests were run on the CR0s on the East side. to verify that CRD seal integrity, friction and scram times were within allowable limits.
CRD seal integrity was measured via a stall-test. Results of these tests-did not indicate any unusual amount of flow during stall conditions and, consecuently, the.CRD seals were judged to be intact. Stall tests could also have provided a means of detecting scram outlet valve leakage. How-ever, this test was not done. Friction tests and single rod scram tests-also showed no anomalies.
Non-Condensible Gas in the Hydrsulic System The effects of air or nitrogen in the CRD Hydraulic system were considered.
Ucon cuestioning, GE CR0 experts stated that non-condensible. gas in the hydraulic system would only cause problems with normal insert and withdraw metions but would not cause oroblems with scrim insertion. This is because stepping the rods' requires intricate timing of rod motion and latching i
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whereas scramming is a single motion. During a stepping function any non-condensible gas would undergo compressions and expansions much different fecm the behavior of the non-compressible liquid.
The presence of nitrogen gas in the SDV prior to scram would be no different than the presence of air which is there routinely. Following initiation of the scrim, the vent valve closes slowly enougn to allow a good portion of the non-condensible gas in the 50V to be vented.
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EVENT SEQUENCE ANALYSIS As discussed previously, the control rod drive HCU exhausts are partitioned into East and West bank scram discharge headers. Control Rod Drives which discharge into the East header are located on the East half of the core while CRDs which exhaust into the West headers are positioned on the West side of the core.
The most notable observation of the control rod positions after the first manual scram was that all of the control rod drives exhausting into the West header inserted full-in (except for the CRD at position 30-23 which inserted to within one notch of full-in) while the control rods exhausting into the East bank header inserted an average of only 20 positions. This CR0 insertion pattern provides strong evidence that the fundamental cause* of the extensive f ailure-to-fully-insert of the CRDs on the East side of the core was hydraulic in nature. More specifically, the rod pattern resulted from an inability of the East header CRDs to exhaust properly for some reason.
With respect to possible multiple scram outlet valve failures, all of the East header scram discharge valves were observed by the control room operators to have opened upon manual scram actuation. Additionally, ali of the manual isolation valves on the scram discharge lines of the East header HCDs were inspected by the licensee immediately upon shutdown, with each founc to be fully open. Accordingly, the remaining possible hydraulic causes could have been blockages in most of the CRD scram exhaust discharge lines or inadequate free volume (or high back pressure) in the East header SDV. Subsequent scram testing of numerous East header CRDs which failed to fully insert demonstrated, however, that no blockages existed in the CRD exhaust lines. Excessively rapid buildup of back pressure in the East bank SDV, due to multiple CR0 seal failures, could also be postulated as a mechanism which could inhibit full-in control rod motion. However, stall tests performed on the East bank CRDs,
- See Section a for a discussion of other possible causes investigated.
o together with individual rod and full core scram tests performed prior to restart, demonstrated that an excessively rapid increase in SDV back pressure resulting from multiple CRD seal failures was not the cause of the partial scram failure. Accordingly, it was concluded that, for some reason, the East bank SDV had inadequate free volume available to accept the full scram discharge from all East bank CRDs exhausting into the East header. Thus, the observec East bank control rod insertion behavior can best be exDlained on the basis that the East header SDV was at least partially filled with water when the operator manually scrammed the reactor.
As discussed in Section 3, adequate free volume must be available in both the East and West headers to acccrm:cdate water exnausted during control rod scram insertion. Furthemore, water must be exhausted into the SDV with low back pressure on the drive piston to assure that technical specification scram speeds are met. A recuction in the free volume in the SDV could tend to in-crease back pressure on the drive pistons too fast which could then increase the time required to complete scram insertion. Complete rod insertion would still be possible, however, even for significant reductions in the available free volume in the SDV as demonstrated in recent single CR0 scram test simula-tions performed by GE. The GE tests showed that for a 40". decrease in the available SDV, a control rod can still fully insert over a broad range of seal leakage values. For a 70". reduction (i.e.,1.0 gal / drive remaining) in available scram discharge header free volume, the rods could still fully insert if seal leakage rates were small enough. For a reduction in SDV of this magnitude, however, increasing seal leakage rates can cause the CRD travel (number of positions inserted) to decrease. The tests show that drive travel decreases to only 36 positions (out of Ja, when a 70". reduction is coupled with a seal leakage rate of 8.9 gpm. The GE test cases run for 55f. recuction in free volume.(.5 gal / drive remaining) showed that even with no seal leakage, the drive would insert only 28 positions and decreased to 22 positions for 5.2 gpm and 18 positions for S.9 gpm leakab.
Finally, as expected, the tests showed that the U<Ds would not insert at all if there were no free volume in unich to exhaust (0.0 gal / drive) regardless of seal leakage. Thus, 1,_
L these tests clearly demonstrate that CRD travel during scram insertions can be sharply reduced if the amount of available exhaust volume is reduced suffi-ciently.
Since the manual rescrams (scrams 12 and 3) occurred with East bank scram dis-charge volume almost full of water on each occasicn, these later scrams can be used as models for back-checking the cause of the observed East bank control rod insertion behavior during the first scram. That is, the-fullness of the East bank SDV during the first scram can be qualitatively and somewhat quan-titatively inferred by comparing it with rod motions during the later scrams.
The available free volumes in the East bank SDV for each of the later scrams can be calculated oy multiplying the drain times discussed in Section 2 by the East bank scram discharge volume drain rates discussed in Secticn 6.
The amount of free volume wnich would have had to have been available during these later scrams can also be calculated from the observed rod motions during these scrams together with the GE test results. Comparing the volumes calculated both ways can then be used to show whether or not the observed rod motions during each scram were consistent with the amount of discharge volume made available by the drains between scrams. Once these are shown to be consistent, cne can infer the limited amount of free volume which must have been present in the East bank SDV during the first scram. The East bank drain times, total number of positions inserted, and average number of positions inserted per rod used in this analysis are shown in Table 5-1.
The drain times between the first and second manual scran was 93 seconds and i
between the second and third manual scra+; the drain t).tc was 53 seconds.
Tests at Browns Ferry show that the nc.r'a, irait rate fcr the East SDV is aoout 11.6 gpm when East anc West scram cnarge volumes are draining simultaneously. Thus, by multiplying this normal crain rate times the drain-time between scrams, we can calculate approximately how much water cculd have drained out (free volume made available) of the East bank header curing
, l
i i
the periods between scrams. Multiplying, one finds that about 18 gallons would have been made available during the first drain (between scrams 1 and
- 2) for the second scram while about 10.2 gallons would have been made avail-able during the second drain (between scrams 2 and 3) for the third scram.
On the other hand, frcm the GE tests and the average rod motion given in Table 5-1 to a first approximation and assuming no CR0 seal leakage, an average of.18 gallons per drive was available for tne second scran anile about.07 gallons per drive was available on average for the third scram.
Thus, for 93 drives, to a first approximation and given no seal leakage, a total of about 17 gallons of free volume was available in the East SDV for the second scram wnile about 7 gallons was available for the third scram.
However, if every East bank CRD were assumed to have a seal leakage of 5 gpm,*
frcm the GE test results the required volume per drive would have had to have been no more than about 20 percent more than the above values. That is, about 20.3 gallons of free volume would have had to have been available during the second scram and about 9 gallons for the third scram.
Comparing the results of the above calculations, it could be concluded that the East SDV was draining normally between scrams one and two, and two and three and that the average rod insertion during the second and third scrams was the amount which one would expect for the amount of wlume made available by the drain. Thus, the insertion behavior of the East bank control rods logically could be explained on the basis of a virtually filled SGV during the second and third scrams.
This same approach can now be used to infer the cause of limited control rod motion during the first manual ram. From Figures 2-2 and 2-3, the average control rod insertion during the first scram was 20 positions. From this value we would infer (using the GE test results; that there was an average of cnly.35 gallons per drive available (or about 33 gallons total) in the
' Conservative based on CRD maintenance recommendations.
East scrsn volume during the first manual scram. This assumes no seal leakage. With a 5 gpm seal leakage, we would infer that only about.45 gallons per drive (or about 42 gallons total) had been available.
The above calculaticnal results snow that the partial scram failure during the first scram can most easily be explained by having an initially partly filled East bank SDV. Similarly, the subsequent CRD failures-to-fully-insert are explainable cased on a partly fillec scram discharge volume.
It should be pointed out, however, that there was considerable spread among
~
the control rods in the number of notches inserted after the first scram.
The variaticn from rod-to-rod could be explained by CRD-to-CRD differences in such parameters as seal leakage (which significantly effects number of notches inserted), control rod drive friction, nitrogen accumulator pressures, etc.
Finally, evidence that the East bank scram discharge volume was initially partly filled with water can be fcund in the elapsed time to activate the.
SIV Hi level scram switches following the first manual scram. Reactor scrams at SF-3 prior to tne June 23, 1980 event resulted in time delays frcm reactor scram actuation to SIV Hi Level scram actuaticn ranging from 42 to 54 seconds.
The first manual scram from the June 28 event had a delay of only 19 seconds.
For a normally empty SDV and SIV, the time delay would represent the time it takes for water exhausted from the CRDs to enter and begin to fill the SDV, i
travel down the SDV-to-SIV drain lines, and fill the SIV to the 50 gallon level.
If water were already in the East SDV, water exhausted from the CRDs wculd almost immediately start to push water out of the East SDV and into the crain line. This wculd cause the SIV to fill more rapidly. Thus, an elapsed time of cnly 19 seconds to actuate the SIV Hi Level scram switches provides important evidence that the East SDV was already almost completely filled with water at the time of the first manual scram.
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6 SCRAM DISCHARGE VOLUMEhCpAl1 INSTRUMENT VOLUME It45PECTich5 Al40 TE5T5 Followi:rg the partial scram failure event at BF-3, TVA, with the assistance of GE, embarked on an extensive inspection and test program. These inspections
{
and testsfwere performed to try to pinpoint what caused substantial water to'be present,in the East bank scram discharge volume on June 23, 1980, s
wnile the scram instrument volume level switches were indicating both headers ;
were empty. The inspect.on program included physical examinations of the drain ano vent piping, the scram discharge and instrument volumes, as well as the drain and vent valves. These inspections were performed in an attempt to detem%e if a vent or drain line blockage had caused the East bank scram
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discharge volume to not drain properly. Additionally, drain tests were per-formed:en both the East and West headers to establish the drain characteristics of these ccmponents. The following paragrachs summarize the results of these inspections and tests.
l Inscections, The T' drain line between the East bank scram discharge volume and tha scram instrument volume were checked for blockages. The drain piping was cut at several locaticns. Metal tape was then inserted through the drain piping seg-ments. These inspections uncovered no obstructions in the piping between the Sg and SIV which could have impeded normal draining of the SDV. A fiber optics inspection of the inside of the SDV at the low point of the 6" diameter l
SDV (where the 2" drain line ccnnects to the 6" SDV) revealed no foreign objects. which cquic have block'ed water from draining out of the 6" SOV into be 2" drain line. The vent piping which cross-connects the high points of thf East bank scram discharge header was also cut, flushed, and inspected.
No. obstruction was. fou,nd in these vent lines which could have impeded or prevented normal drairiing of the East SDV.
Foll6 wing tlie event, 'the vent valve on the East header was removed and a
, e, vacuu4pymp connected to the Clean Radioactive Waste (CRW) side of the vent
-line. Eight 13) inches cf' mercury was indicated by the 1.35 CFM vacuum pump
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M.
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i gauge after a few minutes of. pumping but fell off sharply to 2 inches shortly thereafter. Neither the validity of this vacuum reading nor the reason for the apparent and brief vacuum pull could be determined by TVA.
Several days _later when the 1.25 CFM vacuum pump was reconnected to the same East header common vent pipe, no vacuum could be drawn after the vent line was flushed. A test of the East header vent valve itself showed that it was operable.
The scram instrument volume was also visually examined with a boroscope by inserting it through the vent and drain line penetrations. No obstructions were found which could have prevented draining into or out of the instrument volume.
An inspection of the 6" East bank SDV and drain line showed that they sloped continuously downward toward the instrument volume, with the exception of a localized 3/4" rise in drain line at the expansion loop in the steam vault.
This might have been' a loop seal of greater depth when the steam vault was hot during normal reactor power operation. The overall drop in the drain line between the East SDV and instrument volume was determined to be l' 7" over its 150 ft. length. From the inspections discussed above, TVA was not able to locate a blockage, loop seal, valve maloperation, or other impediment to draining which could be described as the root cause for holding water in East SDV.
Scram Discharge Volume Vent and Drain Tests TVA performed a series of drain tests on both East and West SDV headers cver a period of several days immediately following the partial scram failure event.
The purpose of these tests was to determine the effects of a restricted vent path en East and West bank SDV drain capabilities and to quantify the normal drain characteristics of the SDV. Special test procedures were written for these tests.
^O e
l Typically, these tests involved initially filling the East and West SDV discharge headers and scram instrument volume tank with room temperature demineralized water. During filling operations, the East and West header vent valves were kept open and tne scram instrument volume drain valve was kept closed. Normal drain times and drain rates for the East and West SDV headers and scram instrument volume were then determined by recording the elapsed time necessary to empty these volumes with the vent and drain valves open. Vacuum hold tests (simulating vent line blockages) were performed to determine the drain capabilities of the headers with the vent valves closed.
Water level in the SIV and SDV was monitored by ultrasonic equipment and I
verified by a clear tygon (manometer) tube attachea to the scram discharge volume headers. Clearing times of the 50, 25, and 3 gallon level switches attached to the SIV tank were also recorded during the tests.
Summary of Test Results Scram Discharge Volume Vacuum Hold Tests East Header With tne West header drained to empty, the East header was allowed to drain into the SIV with the East header vent valve and SIV drain valve closed and the West header vent valve open. For this condition (which s c 3d a biccked East header vent), water drainea from the East SDV into c 5IV tank at a rate of only 0.6 gpm.
West Header For this test, the East header was first drained to empty by opening its associated vent valve together with the SIV drain valve. The West header was then allowed to drain into the SIV with the West header vent valve and SIV drain valve closed. For this condition (which simulated a blocked West header vent), water drained from the West SDV into the SIV tank at a rate of about 3.2 gpm.
i East and West Headers For this test, both the East and West headers were allowed to drain simultan-j ecusly into the SIV tank with their respective vent valves closed and the SIV tank drain valve closed. After an initial water surge, the combined drain rates of the two headers into the SIV tank was 0.6 gpm.
Scram Discharge Volume Drsin Tests These tests were performed to dettrmine the drain times and drain rates of l
the SDV and SIV during normal draining (open vent and drain) conditions.
Drain tests were performed for both East and West headers draining at the same time. The system was first filled with the SDV vent valves open ano the SIV drain valve closed. At time zero the drain valve was opened.
Ultrasonics indicated that the West header emptied after abcut 9% minutes
,vhile the East header enptied censiderably later at about 25 minutes.
Additicnally, the 50 and 25 gallon switches in the scram instrument volume cleared at about 914 minutes and 101/4 minuter., respectively. The SIV 1
3 gallen switch cleared after 11 minutes and '!0 seconds had elapsed. Sased cn the volumes associated with the SDV headers, these tests showed the
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average drain rate (with both SDV headers draining together) of the East SDV header to be 11.6 gpm while the average drain rate of the West SDV header was shown to be about 35 gpm. The average drain rate for the SIV based on clearing of the SIV level switches was 24.5 gpm. However, this drain rate was with the East SDV header still draining into the SIV at an average rate of 11.6 spm. That is, the SIV drained 24.5 gpm faster than the East SDV drained.
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7 PREVIOUS BWR EXPERIElCE OF E ILURE To FULLf Ifl5ERT A review of previous BWR exoerience was cerformed with respect to failure to fully insert control rods and problems with the SDV. The sources of in-formation used were ilUREG-0640 and computer searches of LERs. Computer searches via the NRC LER system and the Oak Ridge Nuclear Safety Information Center data base revealed no later events more significant than those re-ported in flUREG-0640.
!'ost instances of failure of rods to fu? t insert resulted in a number of rods latching in position 02 (position 00 is fully inserted). Up to the time of publication of flVREG-0640 in April of 1978, 12 scram events where some rods failed to fully insert were tabulated. These events in general involved a relatively small number of CR0s, between 2 and 15. However, one event at Dresden 2 in flovember of 1974 involved 96 rods, flinety-three stopped at position 02, one at position 04, and two at oosition 06. The only cause recorted for the failure of rods to fully insert was damaged stoo piston seals. Stop piston seal damage can cause excessive leakage past these seals during a scram which could be large enough to fill (and
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pressurize)the discharge volume in advance of the control rods reaching their full-in position.
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FINDINGS 4
1.
The cartial failure to scram at BF-3 on June 2S,1980, was acoarently due to the cresence of water in the East scram discharae volume header.
As succorted by the tests, inspections and analyses discussed in Sections 4 and 5 of this report, the !pparent cause of the extensive failure of con-trol rods to fully insert on the East side of the core was the presence of water in the East scram discharge volume. header.
1 5
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t
- 24 1
~~
e 2.
The SF-3 scram instrument volume Hi level scram function did not and does not orovide protection against the accumulation of water in the East scram discharge volume header (with attendant loss of East bank scram function) even for normal venting anc craining conditions.
Drain rate tests perforce: at SF-3 shcw that water drains out of the scram instrument volume tank censiderably faster than water drains into it frcm the East bank scram discharge volume header even for normal, free, unobstructed venting and draining. Based on the tests, the average drain rate of the SIV is approximately 35 gpm while the average drain rate of the East bank scram discharge volume header is approximately 11.5 gpm. For these drain characteristics, water will drain out of the SIV leaving it virtually empty while water may still be present in the East bank SCV. This actually occurred in the East header drain tests. During the test, the SIV ematied about 20 minutes before the East header fully drained.
With these relative draining characteristics, if water were to leak into the SDV faster than 11.6 gpm, water would accumulate in and fill the. East header (since water is being added faster than it can drain cut). At the sine time, the water draining out of.P2 East header (i.tk., at 11.6 gpm) will not accumulata in the SIV since the SIV drains at a faster rate (i.e.,
35 gon). This process would result in water filling the East header with-out an automatic SIV Hi Level scram ever occurring. We have also found that water drains out of the SIV so raoidly that the SIV Not Dralneo alarm would not alarm in the control rcom. Thus, there wculd be neither centrol rcom ir.dication that water is filling an East SDV nor automatic reactor scram actuation to provide protection against partial loss of scram capability.
In view of the above, with regard to the SIV hi Level automatic scram function, we have found that continuous autcmatic protecticn against filling the East bank SOV (with subsequent cartial loss of scram function) never did and still does not accear to exist at SF-3.
Furthermore, any SWR with a SIV normal drain rate significantly faster than its SDV normal drain rate also would be without autcmatic protection against filling of the 50V. Although r
l 4
l cot verifies by test, it is likely that the BF-3 West SDV header also would be in this category.
]
The loss of automatic scran function can be explained in hydraulic head terms. The SIV is a high cylindrical tank and the 50 gallen SIV Hi Level scran is located over S' above the bottom of the tank. Thus, it is
[
necessary to build up a head of 3' in the SIV before the Hi Level Trip switches can actuate.
If the drain line from the SIV to CRW is a relatively short line (as is the case at BF-3) an 3' driving head, would result in a fairly raoid drain rate. On the other hand, the SDV header is a horizontal Dipe with a small sloce. Even when filled, the maximun head of water that can be developed above the SDV drain (at 3F-3) is approximately 2h'.
- Thus, P
even with a relatively short drain line between the SDV and the SIV, the flow rate in this line would normally be low because of the low head.
i Actually the SDV header drain and the SIV drain are the same size for SF-3, but the SDV drain is considerably Icnger. As a result,the lower available hydrostatic head in combination with the higher fluid flow resistance re-suits in a much slower drain rate for the East SDV header than for the SIV.'
Such an arrangement can never detect accumulation of water in the SDV.
de e
. f-
,, ~ - -.,
3.
A sinole blockage in the West header vent or drain line could comoletely disable the automatic reactor orotection function installed to orotect against a loss of scram capability for all control rods.
For plants like BF-3 which have one 50V which normally drains significantly slower than the SIV, it is possible to comoletely disable the protection provided by the SIV Hi Level scram for both the East and West 50/ by costulating a blockage on the f aster draining SDV. Reduced flow frcm a blockage on this faster draining header SDV, when combined with the normally slower jraining header flow, may total less than the scram instrument volume drain rate which would then result in the SIV emotying with both SDVs still containing water. This would be a serious and undetectable condition if water inleakage were to subsequently develop into both SDV headers such as to keep the headers full at all times. For such a situaticn, there would be no automatic scram to protect against a total loss of scram function due to CRD water inleakage since the SIV water level would never rise to actuate the SIV Hi Level scram switches.
- 27
4 With the current scram discharge volume / scram instrument volume design, a blockage in the SDV vent or drain path can cause a cartial loss of scram caoability and disable the orotection function installed to orevent
.i t,.
As discussed in the previous sections, a blockage in the SDV header vent or drain path will drastically reduce the drain rate of the scram discharge volume. Water leaking past the scram outlet valves (or from other sources) could then cause the scram discharge volume to fill. Since the CRD tamperature,
probes would allcw about.1 gpm of undetected leakage, as much as 9 gpm could leak into the SDV header undetected from all CRDs. Thus, given a partially blocked West header drain, for example, the West header could easily start to fill with water, leaking in undetected through the West sice CRD scram outlet valves. At the same time, since the drain rate of the West header with a drain line blocked could now be substantially less than the SIV drain rate, water would not accumulate in the SIV. Therefore, the SIV Hi Level scram switches would not actuate to prevent filling of the header.
Thus, with the present SDV/SIV and Hi Level scram arrangement, a single failure such as a blockage of a SDV drain or vent can help initiate a partial loss of scram capability and disable the protective' function designed to prevent the loss.
l 28 -
1
5.
There are numerous actual and cotential mechanisms for introducing and retaininc water in the SDV with no accumulation in the SIV.
Review of the vent and drain paths for the scram discharge volume and the scram instrument volume has shown that there are numerous actual and potential mechanisms which could slow or even stop SDV drainage into the SIV. Since the SIV would still maintain a high drain rate, it would be possible for the SDV to retain water while SIV instrumentation indicates
- emoty, a
Possible sources of water are: water from the previcus scram; multiple scram outlet valve leakage; or injection frem SDY flush lines.
Pechanisms which retard free draining of water out of the SDV include:
a blockage in the vent piping; a plugged SDV-to-SIV drain line; s closed SDY vent valve; a vacuum held in the SDV by a 1c00 seal screwhe e in the vent line; vent line siphon effects from water in the SDV vent line; venting to the closed CRW system in the Reactor Building Crain Sump below wat.er without vacuum breakers; vacuum effects frcm fluid ficws through the CRW piping system; Vacuum effects from condensing hot water ir. SDY frcm the previous scram.
Venting of the SDV to atmospheric pressure while the SIV drains into the closed CRW drain system (which could be pres 3urized above atmospheric pressure) could also inhibit draining of the SDV headers if there is insuff'cient downward slope in the SDV drain line, Since the CRW exhausts under water in the Aeactor Building Drain Sump and non-condensible gases are present in the fluids draining through the CRW drain systen, there is a possibility for pressure to build up in the CRW drsin system. This cressure, in conjunction with a small loop seal in the drain line from the SDV to the SIV, could hold up water in the SDV even if the SDV were vented directly to atmosphere..
6.
The current scram discharoe volume / scram instrument volume desian results in' the automatic Hi Level scram (safety) function beino directly dependent on the nonsafety-related raactor buildino Clean Radioactive Waste drain system.
For the scram instrument volume Hi level scram switches to activate, water must accumulate in the scram instrament volume. For water to be able to accumulate in the SIV, it must be able to drain at an adequate rate from the SDV into the SIV. However, from the drain rate tests performed at BF-3, imorocer venting of the SDV can sharply or totally prevent water from draining out of the SDV. Pry er draining of the 50V is directly dependent on the venting function provided by the reactor building Clean Radioactive Waste drain system (a required systems interaction). Accordingly, we would conclude that operability of the SIV Hi Level scram function is dependent on the venting provicert by the nonsafety-related reactor building CRW system. Unanticicated adverse venting behavior of the CRW system, which rasults in reduced venting of air back into the SDVs, can result in the holdup of water in the SOV with little or no accumulation of water in the SIV. This dependancy appears to be particularly inapprupriate if not unacceptable for a reactor protection function which is intended to prevent the loss of reactor scram capability.
7.
The BF-3 cartial scram failure event, together with recent events at other BWRs, have shown that float-tyce water level monitoring instruments have a significant degree of unreliability.
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The BF-3 partial scram failure event demcnstrated on several occasions a significant unreliability of float-type level switches. As shown on the event sequence recorder printcut (Table 2-1), several of the 50 gallon level instruments f ailed to activate on different occasions. Furthermore, during calibration testing of the SIV level switches following plant shut-down, both the 3 gallon and 25 gallon switches were found to be inoperable.
After the instrument taps were flusned o# residue, the switches operated satisfactorily. During drain rate testing of the BF-3 SDV, two of the four 50 gallon switches f ailed to activate twice in two drain tests. Additionally, inspections at Brunswick Unit No.1, following a reactor scram cn November 14, 1979, revealed incperable alarm and rod block level switches due to bent float rods. Other surveillances and inspections at Hatch Unit 1 on June 13, 1979, found two SIV Hi Level switches inoperable due to bent floats binding against the inside of the float chamber. These recent' experiences indicate a significant degree of unreliability of float-type level switches resulting from various causes.
as 8.
With the current BWR Reactor Protection System looic, the cresence of certain automatic scram ccnditions oreclude SDV draininc (scram reset) to cermit a rescram.
In order to drain the SDV for rescram following a scram actuation, it is necessary to recoen the SDV vent and drain valves and to reclose the scram inlet and outlet valves (RPS reset). This recuires the following steps:
- 1) clace the reactor mode switch in SHUT 00WN or REFL'EL; 2) actuate the DISCHARGE VOLUME HI WATER LEVEL BYPASS switch; 3) the reactor trip signal must clear or be bypassed in SHUTDOWN or REFUEL modes; and finally 4) reset the RPS. However, the following reactor trip functions cannot be
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bypassed by the operator in the SHUTDOWN or REFUEL mode:
Crywell High Pressure Reactor Vessel Low Level
- 'ain Steam Line Hi Radiation Neutron Monitor System Trip Reactor Vessel High Pressure Condensor low Vacuum
- Main Steam Line Isolation Valve Closure
- Thus,if any of the above trip conditions are cresent, resetting the RPS would not be possible.
For example, if a spurious MSIV closure event should occur at power with the 50V initially full of water, a reactor scram would occur with the control rods failing to fully insert.
If the MSIV closure trip (or Reactor Vessel High Pressure) condition persisted, then a rescram attempt would not be possible since it cannot be bypassed in SHUTDOWN or REFUEL modes. Thus, the trio condition itself would prevent the possibility of rescram. However, we do not consider that any codificaticn is recuired in the RPS trip / reset circuitry to enable the operator to reset the RPS in the presence of any automatic scram condition, since the capability to reset and rescram has not been defined as a required protective action.
- Depends on Reactor System Pressure Interlock setpoint. '
o 9.
If a scram condition exists which cannot be byoassed in SHUTOCWil or REFUEL mode, then failure (to close) of a 50V vent or SIV drain valve can result f
in an unisolatable blowdown of reactor coolant outside crimary centainment.
With the reactor in an unscramed state, the scram outlet valves provide both a reactor coolant. pressure boundary function and a primary containment isolaticn function. During a reactor scram, the scram outlet valves open (ona per control rod drive) and the SDV vent and SIV drain valves close.
Reactor coolant pressure boundary integrity and primary containment iso-lation functions are then transferred to the scram discharge volume vent and SIV drain valves which seal the SDV. We nave fcuno tnat tnere are no redundant isolation valves in the vent or drain lines to provide these isolation functions for a scram condition. The f ailure of any one of
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these valves in the coen position, therefore, could result in an uncontrolled blowdown of reactor water outside primary containment anc into the CRW drain lines if the operator could not reset the scram. The blowdown would ultimately discharge to a drain sump which is not designed to handle the heat loao or pressure buildup. With the present 3WR RPS design, the operator would be able to reestablish primary centainment isolation (with scram out-let valve closure) only if the RPS could be reset. However, if a reactor scram condition persists and it cannot be bypassed in SHUTDOWM or REFUEL mode (i.e., any of those listed in Finding 48) it would be impossible to reset the RPS to terminate the blowdown.
Thus, for example, a scram caused by spurious closure of the MSIVs with a failed open scram instrument volume drain valve would result in an uncor.-
trolled blowdown of reactor coolant outside primary containment and into the drain sump room which contains the engineered safeguard pumps which are reouired for mitigation. Slowdown would continue as long as the MSIV closure scram condition existed (MSIVs not reopened) since this trip cannot be bypassed in SHUTDOWN or REFUEL mode. That is, the scram outlet valves could not be reclosed to isolate tne blowdown until the MSIVs could be reopened. For events which result in scrams caused by conditions which cannot readily be cleared, uncontrolled blowdown into the reactor building (secondary containment) could be sustained for an indefinite peried of time with possible envirencental impact on the required mitigating features. l l
- 10. The energency coerating instructions at BF-3 did not include a crocedure or guidance for the coerator to follow in the event of a cartial or ccmolete scram failure.
The Browns Ferry plants, as perhaps do most (if not all) other BWRs (and probably all other LWRs), do not have emergency procedures for the operator to follow in the event of a partial or complete scram failure. We have found that, although control rcom coerators are trained to verify that the rods have fully inserted upon a sc.-am actuation, precedures do not exist for the operator's immediate or subsequent actions if full control rod insertion does not occur. Moreover, although operators are fully knowledge-able of the function and operation of the standby licuid centrol (poison) system, the plant does not have specific procedures which state when the alternate shutdown system must be actuated.
- J4 -
O 9 RECDMMENDATIONS 1.
The coerability of the Scram Instrument Volume Hi Level Scram, function should be indeoendent of the Scram Discharge Volume venting and draining recuirements.
The current BWR scram discharge volume / scram instrument volume design configuraticn requires proper venting of the SDV and proper ':DV-to-SIV draining to assure operability of the scram instrument volume :-:1 Level scram function. We recommend that the operability of the Hi Level scram be made independent of SDV venting or draining rer.;uirensnts. We make this recommendation because of Finding Nos.1 through 6 discussed in the previous section. That is, the hydraulic factors which control water level in the SDV and SIV should not be able to negate the response of the Hi level orotection function. We believe the acceptable configura-tion would be to place the SIV tcnk directly under the icw end of the 6" SDV header and to connect the top of the SIV tank to "
bottom at the low end of the SDV header by a short vertical 6" dia e.er pipe (rather than the current 2" diameter horizontal pioe). This arrangement should assure water soillage from the SDV directly down to the tank containing the level nonitoring instruments. Furthermore, it would not depend on venting or craining phenomena which are sensitive to Slot *ges. We also recommend two seoarate scram instrument volume tanks, one on each SDV header bank.
Separate instrument volumes, in immediate proximity to their rescective headers, should assure procer water spillage into the SIVs and orovide adecuate redundancy for protection against a total loss of scram cacability.
It is our firm belief that modifications which simply improve One venting of the SDV/SIV volume arrangement to assure operability of the SIV Hi Level scram functicn are not adequate. We recortnend that this uniquely imoortant safety functicn be made completely indeoendent of any vent or drain arrangements, thereby separating the water accumulation control and protection functions. We further recommend that in situ fill tests be performed to demonstrate that the operability of the protective Hi Level scram functicn is insensitive to the vent or drain arrangement for the design configuration finally installed.
-3~-
I O
2.
Scram instrument volume water level monitorinc instruments for the SIV Hi Level scram function should be both redundant and diverse.
It is recommended that diversity be added to the redundancy of SIV level monitoring instruments for the SIV Hi level scram function. Currently, there are redundant float-type level switches for each RPS channel for the Hi Level scram function. On several occasions recently, as discussed in Finding No. 7, more than one float-type level switch was observed to be inoperable at once. During and immediately following the BF-3 partial scram failure event, several float-type level switches in the instrument volume f ailed to actuate.
In view of these experiences, we recommend that diversity be included in the level mcnitoring function for the SIV Hi Level scram function. The important and unique protection provided by this trip function requires that the presence of water in the SIV be monitored continuously with extremely high reliability. We are recommending that diversity be added in order to assure this reliability.
Monitoring techniques, such a: di#farential pressure cells, ultrasonic detection or conductivity probes, may be considered along with others for this purpose.
e O
3.
All vent and drain caths from the scram discharce volume and scram instrument volume should have redundant automatic isolation valves.
As discussed in Finding No. 9, scrams which occur as a result of automatic reactor trip conditions which cannot be cleared or bypassed in REFUEL or SHUTDOWN modes can result in unisolatable reactor system blowdcwns out-side of primary centainment if the 50V vent or SIV drain valve fails to close. To protect against such occurrences, we recommend that redundant valves be placed on all vent and drain lines connected to these volumes.
Redundant valves would also protect against equipment damage which might otherwise occur as a result of excessively sicw closure or delayed closure of one of'the isolation valves. These valves must be qualified and capable of closing against full reactor pressure, flow, and temperature ccnditions in case the lines are not isolated within normally soecified time limits. The vent and drain lines and drain supports must also be designed for the hydraulic loads and instabilities associated with the blowdchn of the high pressure / temperature reactor coolant to the drain system. Prolonged blowdown may be ruled out as a design basis with appropriate diverse isolation or other acceptable provisions. 310wdcwn instability due to isolation valve time delay is believed to be the cause of failure of the float-type level switches at Brunswick Unit No. 1.
o 4
Emergency operatine procedures and coerator trainino should be provided for complete and cartial scram failure conditions.
In view of Finding No.10, we recommend that emergency operating procedures and training be provided to control room operators to respond to partial or complete scram failure conditions. These procedures should include explicit statements regarding the conditions for which the standby liquid control system must be used. The procedures should incluoe cautions regarding operator actions which should not be taken which could result in a severe transient condition (e.g.,
main turbine trip) being created. The procedures shoulo provide guidance to the operator for starting up safety systems for standby readiness (e.g., HPCI on minimum flow) or for tripping other systems (e.g., re-circulaticn pumps). The order of operator actions (i.e., immediate, subsequent) should be considered, as well as when the operator should begin attempting to insert rods manually.
We believe that such operations (human factors) aspects can and should be implemented in the near term. Such procedures and training would assure, in the near term, the most 3ppropriate control room operator action during a scram failure event and wall in advance of any ATWS modifications which may be required in the long tem.
. l
5.
Consider modifying the SDV vent ani SIV drain arrangement to imorove scram discharge volume drain reliability.*
As discussed previcusly, scram discharge volume draining currently depends on uncertain vent and drain functions provided by the reactor building Clean Radioactive Waste drain piping, along with relatively small diameter, nonredundant vent and drain piping, which are susceptable to blockage. This current, relatively unreliable SDV venting arrangement could be improved by increasing the vent line size and by adding an alternate, reliable, and isolatable vent path. The alternate vent path could be installed with a check valve and air operated isolation valve to provide an alternate and isolatable path for air inleakage into the scram discharge volume. The alternate vent path could be vented either directly to the Reacter Building atmosphere or to a gas treatment system with a vacuum breaker. The check valve would provide automatic isolation of this redundant line upon pressuri:stien of the scram dis-charge volume during a reactor scram. The drain function cculd also be improved by providing a second drain line frcm the SIV to the CRW floor drain.
We believe that modificaticns, such as those described above, would help improve SIV drain reliability.
Improvements such as these wculd thus help to further reduce the number of challenges to the SI'V Hi level protective scram function.
- Although this recommendation is only for censideration, we do belie,e that it would further reduce the risks associated with loss of scram capability arising from water accumulation in the SDV. -
O 10 CONCLUSIONS
~
The Browns Ferry Unit 3 partial scram failure event which occurred en June 28, 1980, demonstrated that the present SWR scram system can be vul-nerable to loss of scram capability while operating at power. Furthermore, the event showed that the loss of scram capability can occur in a way which goes undetected by the operator and unprotected by the reactor protection system.
The information and analysis of the BF-3 partial scram failure, which is provided in this report, concludes that the cause of the loss of scram capability was the presence of water in the East scram discharge header, w
Furthermore, our analysis of the scram discharge volume / scram instrument volume design configuration, together witn its vent and drain characteristics, leads us to conclude that numerous actual and postulated mechanisms exist which can cause the scram discharge volume to fill undetected and without protection against such filling. Our analyses also show that certain scram events can result in an unisolated reactor coolant blowdown outside of primary containment following a single isolation. valve f ailure.
In view of these design deficiencies, we believe it necessary that modifica-tions be made to the scram discharge volume / scram instrument volume arrange-ment and isolation features. Our specific recommendations for change in the SDV/?IV design which flow #ren our findings have caer ;rovided in ;his re? ort.
We believe that these r contendations should ce corsidered along.sith those of others who are also reviewing the BF-3 event. We do believe, however, that the design changes described in the recommendations are necessary to adecuately reduce the risks associated with the present unreliability of the SWR scram system arising fecm undetected accumulation of water in the scram discharge volume.
Y
' Table 2-1 Event Sequence Recorder Printout 01 31 16 A034 Reactor Scram Manual 8 CY 5 34 A034
+
CY 6 44 A033 Reactor Scram Manual A 01 31 24 A035 Reactor Trip Actuator Al or A2 CY 3 33 A035 CY 3 39 A021 Reactor Low Water Level A CY 3 42 A023 Reactor Low Water level C CY 3 47 A021 Reactor Low Water Level D CY 3 47 A036 Reactor Trio Actuator 81 or 32 CY 3 56 A022 Reactor low Water Level B CY 5 11 A076 REPT C Trioped 01 31 34 A003 Discharge Volume High Water Level C' CY 3 42 A003 CY 5 01 A004 Discharge Volune High Water Level 3 01 31 37 A002 Discharge Volume High Water Level B CY 6 53 A002 01 31 40 A001 Discharge Volume High Water Level A i
CY 0 03 A001 CY 0 18 A106 Malfunction Bus Energized i-CY 0 33 A038 Turb. Stop Valve Closure Scram Trip A f
CY 0 33 A040 Turb. 51.60 Valve Closure Scram Trio C
{
'CY 0 33 A041 Turb. Stop Valve Closure Scram Trip D i
CY 0 34 A039 Turb. Stop Valve Closure Scram Trip'B CY 0 47 A043 Turb. Gen. Load Rejection Scram Trip B i
CY 0 47 A045 Turb. Gen. Load Rejection Scram Trip D CY 0 48 A042 Turb. Gen. Load Rejection Scram Trip A
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CY 0 43 A044-Turb. Gen. Load Rejection Scram Trip C ACS4 Turb. Tripped - Loss of Hydr. Trip Pressure 01 32 01 1021 Reactor Low Water Level A 4
.I CY l 12 N021
~
CY l 57 N023 Reactor Low Water Level'C CY 2 04.'l024 Reactor Lcw Nater Level D i
CY 3 35 ?!022 Reactor Low Water Level B 01 34 45 A053 IRM Uoscale Trip on Level.F-CY 4 30 A058 01 34 a8 A057 IRM Upscale Trio on Level 0 CY 7 36 A057 9 07 "n47 7v CY ? 13 A057 _
t Table 2-1 Event Sequence Recorder Printout i
t CY O'26 N058 IRM Upscale Trip on Level F l
CY 0 49 A056 IRM Upscale Trip on Level 3 CY 0 55 P:056 CY 1 01 A056
-CY l 16 N056-CV 1 41 A056.
CY l 48 N056 f
i CY l 56 A056 CY ? 21 N057 IRM Uoscale Trip on Level D i
CY a 14 N056 IRM Upscale Trip on Level 3 01 42 00 A035 Reactor Trip Actuator Al or A2 l
CY 9 23 A035 01 2 37 f:035 CY 6 35 NO35 CY 6 36 NO35 Reactor Trip Actuator 31 or 32 l
CY 3 05 NO34 Reactor Scram Manual 3 i
CY E 06 NO33 Reactor Scram Manual A 01 c517 A035 Reactor Trip Actuator Al or A2 N Y 6 47 A035
[
CY 47 A036
' Reactor Trip Actuator-31 or 32 01 45 36 N002 Discharge Volume High Water Level B l
CY 5 09 N002 CY 6 16 A002 0146 30 A031 Reactor Scram Manual 3 CY 9 48 A034 f
CY 9 45 A033 Reactor ?crin Manual A 01 47 43 NO35 Reactor Trip Actuator Al or A2 CY 2 37 NO35 g
CY-2 38 NO36 Reactor Trip Actuator 31 or 32 CY.3 05 NO33 Reactor Scram Manual A l
CY 3 05 N034 Reactor Scram Manual 3 01 57 04 N002 Discharge Volume High Water Level 3 CY 3 22 N002 CY a 03 N001 Discharge Volume High Water Level A 01.57 3a N004 Discharge Volume High Water Level D l
CY 4 07 N004 CY 4 19 NCO3 Discharge Volume High Water Level C 02 28 06 D SO TVA 3FUP -
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Figure 2 J Control Rod Positions Af ter Third Scrm I
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