ML20031D632
| ML20031D632 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/02/1981 |
| From: | Blanchard D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20031D553 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8110130562 | |
| Download: ML20031D632 (15) | |
Text
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TESTIMONY OF DAVID BLANCHARD WITH RESPECT TO BOARD QUESTION 2 On May 2,1979, a loss of feedwater transient occurred at the Oyster Creek boiling water reactor.
In the sequence of events which took place following this transient, water level over the top of the core decreased. The core never became uncovered and fuel damage did not occur. The Licensing Board has asked:
Did the facts learned from the loss-of-feedwater event at Oyster Creek on May 2, 1979, suggest any measures, other than those included in Amendment 30 to the Big Rock Point Technical Specifications which would be important in preventing a severe loss-of-feedwater accident?
Could an accident which might occur from this cause threaten the Licensee's ability to maintain the spent fuel pool in a safe condition?
The purpose of this testimony is to explain the sequence of events which occurred at Oyster Creek, the measures which have been taken at big Rock Point Plant to prevent an accident occurring due to a similar sequence of events, and to show that the proposed replacement of spent fuel storage racks i
will have no effect on either the likelihood or consequences of such an accident at the Big Rock Point Plant.
Big Rock Point and Oyster Creek are both non-Jet pumpi General Electric boiling water reactors. Tnere are similarities and differences between the two designs which have a bearing on the events which occurred at Oyster Creek.
It is appropriate, therefore, to describe the design differences between the two reactors.
Oyster Creek Design (Figure 1)
The Oyster Creek reactor has moisture separatirg and ateam drying s-equipment in the upper portion of the reactor vessel.
Water which is miO981-2127a123 8110130562 811005 PDR ADOCK 05000155 T
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separated from the steam in this equipment is recirculated to the core via an annular region along the inner surface of the reactor vessel. This recirculation return annulus is separated from the region of the reactor vessel in which the core is located by a steel cylinder. This cylinder is referred to as the core shroud. The annulus between the core chroud and the reactor vessel is closed at the bottom and thus is not free-communicating w!.th the lower plenum of the reactor vessel. Recirculating water is returned to the lower plenum via five external recirculation loops. Each of these loops consists of a 26" diameter line and a single variable speed recirculation pump. Valves are present upstream and downstream of the pump to permit a pump to be isolated. There is also a 2" bypass line around the pump discharge isolation valve. This discharge bypass line provides a flow path to enable an Isolated pump to be started while mininizing the amount of cold water which is introduced to the core. At the time of the May 2, 1979 transient, one of the five recirculating loops was out of service; that is, the pump discharge isolation valve was clo:ed. The pump suction valve and discharge bypass valve in toe out-of-service loop were both open.
The Oyster Creek design provides for emergency cooling through the use of two isolation condensers. The isolation condensers are tanks partially filled with water and vented to the atmosphere. Steam is drawn from the upper portion of the reactor vessel and passed through tube bundles submerged in water inside the tanks. The steam condenses in the tube bundles and condensate returns to the reactor core by entering one of the recirculation loops upstream of tha pump and its suction isolation valve.
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1 Big Rock Point Desian (Figure 2) l The Big Rock Point reactor design incorporates an external steam drum. A steam and water mixture is carried from the reactor vessel to the steem drum by six 14" diameter pipes, called risers, which enter the bottom of the steam drum. Within the steam drum, extensions of the riser pipes carry i
the steam and water mixture to the upper half of the steam drum where moisture separting and steam drying equipment, equivalent to that in the upper portion I
of the Oyster Creek reactor vessel, is located. Water which is separated from the steam enters the steam drum lower portion outside the riser extensions, i
This water is returned to the core via four 17" diameter downcomer pipes which exit the bottca of the steam drum. The downcomers are in two pairs. In each l
pair, the two downcomers combine into a single 24" diameter line whfch forms the suction side of a recirculating pump. The two recirculating pumps discharge into the lower reactor vessel plenum through 20" diameter pipes.
The downcomers and associated pumps thus form two recirculating loops.
Isolation valves exist upstream and downstream of the two recirculating pumps as in the Oyster Creek design. A bypass line is 5" in diameter. In this i
design, the reattor vessel and risers are equivalent to the region inside the core shroud at Oyster Creek; the downcomers and the steam drum region outside the riser extensions is equivalent to Oyster Creek's downconer annulus.
Big Rock Point uses an emergency condenser for emergency cooling.
This unit is equivalent to the isolation condensers at Oyster Creek. A single tank exists containing two independent tube bundles.- Steam is drawn from the upper portion of the steam drum and is condensed within the tube bundles. The
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condensate is returned via natural circulation to a point at the bottom of the 3
miO981-2127a123
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steam drum rather than into the downcomer pipes as is done at Oyster Creek.
Condensate return flow from either emergency condenser tube bundle can therefore flow through any of the four downcomers.
May 2, 1979 Oyster Creek Event The event which occurred at Oyster Creek May 2, 1979, began with a reactor scram. Following the reactor scram a complete loss of feedwater occurred. Feedwater is the primary source of nakeup to the primary coolant system replacing steam flow from the reactor to the turbine generator. The main steam isolation valve was closed and the isolation condensers were placed in service to remove dacay heat.
In accordance with operating procedures in effect at that time, the recirculating pump discharge isolation valves in the two loops receiving condensate from an isolation condenser were closed. At a later time, the remaining two discharge isolation valves, those in the recirculating lines not receiving condensate from the isolation condcasers, were also closed. This latter action was apparently taken in anticipation of restarting the recirculating pumps.
(Interlocks prevent a pump from being started unless the discharge isolation valve is closed and only the bypasa line is open.)
At this point, the condensate from both isolation condensers was entering recirculating lines upstream of the closeo valves. This condensate was, in effect, backing up into the annulus region, thereby providing an indication of increasing reactor water level on instrumentation which sensed level in the annulus. The flow path between the annulus and lower reactor OV 4
miO981-2127al23
O plenum was limited to that provided by the five two inch bypass lines (pump suction isolation valves in all loops remained open).
Natural circulation flow from the annulus region through the limited available flow path initially could not return as much water to the reactor vessel as was being removed by the isolation condensers in the form of steam.
Accordingly, a redistribution of water inventory occurred from the region within the core shroud to the annulus. Eventually, a low level alarm was received on an instrument which sensed level in the region inside the core shroud. At a later time, one recirculaiing pump was started and its dischiege isolation valve opened. This caused the inventory to redistribute again and the water level within the core shroud region to increase.
The action of closing the valves in the recirculation loops essentially isolated existing level instrumentation from the core region inside the shroud.At no time during this event did the water level decrease below the top of the core. No fuel damage occurred. However, if one or more of the two inch bypass valves had been closed, it is possible that the redistribution of water inventory could have been sufficient to decrease the water level within the care shroud to below the top of the fuel while reactor vessel level instrumentation falsely indicated sufficient water inventory above the core.
Consumers Power Company's Initial Review Since Big Rock Point uses external recirculation loops and an emergency condenser, similar to Oyster Creek, Consumers Power Company
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evaluated the possibility of a similar event occurring at Big Rock Point.
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O During May,1979, the initial evaluation identified that a similar occurrence might happen if a loss of feedwater event were followed by complete isolation of both external recirculating loops (ie, closing both 24" suction valves, either 24" suction valve concurrently with the opposite 20" discharge and 5" bypass valve, or both 20" discharge valves and both 5" bypass valves). Under these circumstances, decay heat would cause boiling within the core region.
Steam would be withdrawn from the steam drum and condensed in the emer;ency condenser. Condensate returning from the emergency condenser would enter the steam drum and be unable to return to the lower plenum due to isolation of the i
external recirculating loops.
(Alternatively, steam could be directed to the main condenser in which case the total inventory in the primary system would decrease if loss of feedwater occurred.) A redistribution of water inventory could therefore occur until the
- tater level in the steam drum had risen to the j
top of the riser extensions. At this point, condensate could return to the j
reactor vessel bv flowina de": the risers. A preliminary evaluation of the volume available within the steam drum below the top of the riser extensions indicated that water level within the reactor vessel might be decreased below the top of the core in such an event. Consumers Power Company therefore recognized the importance of avoiding this type of occurrence.
Big Rock Point operating procedures were reviewed subsequent to this preliminary evaluation to identify all statements requiring that recirculating pumps be removed from service or that loop isolation valves be closed. Pro-
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J cedures were clarified to ensure that one recirculating loop would remain in service during all reactor power operations.
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O In addition, Consumers Power Company submitted proposed changes to the Big Rock Point Technical Specifications on September 28, 1979. These changes required that one recirculating loop (50% of total recirculating f %
area) be left open at all times during reactor power operations. These proposed Technical Specification requirements were the same as those which had been implemented at Oyster Creek (accounting for differences in design of the two plants). The reactor vessel low level alarm set point was also established as a safety limit for Big Rock Point. This set point is equivalent to the low level alarm point within the core shroud which was made a s-fety limit at Oyster Creek.
The proposed Technical Specifications change was approved by NRC and issued as Amendment No. 30 to facility operating license DPR-6 on October 30, 4
1979. No modifications were made to the recommended changes committed by Consumers Power Company. The NRC Safety Evaluation Report accompanying Amendment No. 30 stated that the additional Technical Specification requirements proposed by Consumers Power Company were the same as those which had been identified as necessary by NRC. Since that time the Big Rock Point Plant has operated with the new Technical Specifications requirements in effect.
Subsequent Detailed Evaluation The possibility of an event similar to that at Oyster Creek was subsequently evaluated in more detail as part of the Big Rock Point Probabilistic Risk Assessment (PRA). This evaluation pointed out two major differences between the Big Rock Point and the Oyster Creek reactors with 7
miO981-2127a123
' O respect to this event. The first difference is that the Big Rock Point j
reactor vessel level instrumentation provides indication of water level directly above the core whether or not the recirculation loop valves were closed. While isolation of these loops at Oyster Creek could preclude accurate measurement of actual water level in the core, the level instrumentation at Big Rock Point is independent of the position of the valves in these loops. Second the PRA evaluation included consideration of a source of makeup water not considered in the initial evaluation described above.
This is the control rod drive (CRD) system. The CRD system is a hydraulic j
system including two positive displacement pumps. Under normal operation, one of these pumps is always aperating providing a small flow of water to the i
reactor vessel; the second pump is on standby and starts automatically upon reactor trip. Water addition provided by operation of th. CRD pumps is sufficient to keep the core covered and re-establish consnunication of the i
water in the drum with the vessel by filling the drum and over flowing the risers. CRD water was available during the Oyster Creek event but not in sufficient quantities to overcome water losses in the core region due to decay heat.
(The Oyster Creek decay heat loads are approximately seven times larger 1
than those which occur at Big Rock Point due to the difference in operating power levels at each of the plants.) Maximum decay heat loads at Big Rock Point are sufficiently small to allow CRD pump water addition to makeup for these losses.
Taking these differences into account, an event at Big Rock Point similar to that which occured at Oyster Creek can be expected to evolve as follows. The reactor is assumed to trip an a result of conditions arising V
8 miO981-2127a123
O from either the initiating transient, the total loss of feedwater or the closure of the recirculation loop valves.
(A trip signal is generated if the valve in both loops are at least 10 percent closed.) The second CRD pump would start automatically, increasing the flow. CRD pump flow would be directed to the reactor vessel via the cooling flow to the control rod drives and to the downcomers via an intertie with the reactor cleanup system. Makeup water would thus be provided to both the reactor vessel and the steam drum.
The flow provided by the two CRD pumps is sufficient to replace the liquid inventory boiled off as a result of. decay heat.
If the standby CRD pump were out of service or were to fail to start for some reason, water level above the core would decrease as it did at Oyster Creek. Existing level instrumentation would indicate this drop in water level, however,giving the operator sufficient information to take corrective action to terminate the transient.
This may be accomplished in any number of ways including restarting the feedwater pumps or the CRD pumps to add water to the reactor vessel, opening the recirculation loops to provide steam drum water to the vessel, or manually actuating the Reactor Depressurization System to provide core spray cooling.
Any of these actions would provide sufficient cooling to the core to prevent fuel damage.
The results of this evaluation were reported to the NRC Staff in the report of the PRA submitted March 31, 1981.
Potential Consequences of a Similar Event The principal concern arising from the event at Oyster Creek was the possibility of damage to the reactor core. If the water level were to drop DO 9
miO981-2127a123
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below the top of the active fuel region in the Big Rock reactor it is possible that damage could occur which would release 'ilssion products to the coolant.
Since in this particular event no postulated break in the primary coolant system is involved, these fission products would remain in the primary coolant system unless the reactor depressurizing system (RDS) were activated.
(Actuation of RDS would not be necessary to recover from such an event since establishing a flow path by opening valves in the recirculating loop is sufficient to terminate the event and ensure continued ability to cool the core. This is precisely the way plant personnel recove i from the loss of feedwater event at Oyster Creek.)
The radiation levels created by the released fission products might preclude access to the containment. A detailed evaluation of the radiation levels within the containment which might result if the fission products are retained in the primary coolant system has not been performed and it is possible that access might not be precluded or might be limited for only a brief period.
If RDS is actuated, the resulting effect is the same as a loss of coolant accident, and access to the containment would be dependent on the extent of fuel damage sustained during the event. The spent fuel pool makeup capability described in my testimony addressing Christa Maria Contention 8 and O'Neill Contention III E-2 would enable the spent fuel pool to be maintained in a sate condition even if access to the containment were to be precluded.
The proposed increase in the amount of spent fuel in the spent fuel pool would have no effect whatever on the probability of loss of feedwater transient similar to that experienced at Oyster Creek, on the likelihood of
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isolation of the recirculation loops, nor on the subsequent sequence of x_
10 miO981-2127a123
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O events. Ability to maintain water level in the spent fuel pool is assured even if reactor core damage due to such an event were to occur, and thus the additional spent fuel stored in the Big Rock Point Spent Fuel Pool would also have no effect on the consequences of such an event.
Conclusion The changes made in Technical Specifications at Eig Rock Point following the event at Oyster Creek are the same as those Technical Specifications changes made at Oyster Creek. These requirements assure adequate communication between the steam drum and lower reactor vessel plenum to prevent a redistribution of core inventory which might uncover the reactor core. In addition the design differences between Big Rock Point and Oyster Creek described above provide additional assurance that the Big Rock Point core will be cooled adequately in the event of an occurrance similar to that at Oyster Creek. Accordingly, loss of feedwater events, such as occurred at Oyster Creek, are not likely to result in fuel failure, abnormally high radiation levels, or adverse environment inside containment. Thus, such events would have little or no effect on the ability to operate or maintain the spent fuel storage pool.
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11 miO981-2127a123
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g3 LICENSING BOARD QUESTION 1 O
A.
THE BOARD QUESTION Has the proper operation of any of the valves mentioned in Items 5 and 6 on Page 4 of the Safety Assessment (viz, Valves CV/4096, CV/4097, CV/4027, CV/4105 and MO/7050) been relied upon to mitigate the results of an accident in the spent fuel pool?
If so, how would a failure of the type experienced with these valves affect the results of such an accident?
B.
MATERIAL FACTS AS TO WHICH THERE IS NO GENUINE IF2 SUE TO BE HEARD.
1.
Valves CV/4096 and CV/4097 are supply ventilation valves to containment.
They are in series.
(Affi-davit of Thomas Bordine on Board Question 1 at p. 1).
2.
Valve CV/4097 has experienced repetitive leakage rate failures during tests; Valve CV/4096 has not.
(Affidavit of Thomas Bordine on Board Question 1 at p. 2).
3.
Valve CV/4097 was modified in February and March of 1979.
Since that time no significant leakage has occurred through Valve CV/4097.
(Affidavit of Thomas Bordine on Board Question 1 at p. 2).
1 4.
A release to the environment through valves CV/4096 or CV/4097 from a spent fuel accident would require the unlikely simultaneous failure of both ve1ves together with containment pressurization.
(Affidavit of Thomas Bordine on Board Question 1 at p. 2).
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6 m
1 0
5.
Modifications related to Valves CV/4096 and CV/4097 are scheduled to be made which will assure i'nde-pendency and redundancy of simultaneous failure and thus minimize the likelihood of simultaneous I
failure.
(Affidavit of Thomas Bordine on Board Question 1 at p. 2).
6.
Valve CV/4027 is on a line through which water flows from the spent fuel pool surge tank and the spent fuel pool cooling system to the radwaste system.
(Affidavit of Thomas Bordine on Board Question 1 at p. 4).
fg 7.
Valve CV/4027 has experienced leakage rates in excess LJ of Technical Specification limits during several tests.
(Affidavit of Thomas Bordine on Board Question 1 at p. 3).
l 8.
There are valves in the same line upstream from CV/4027 which are maintained in a closed position which would prevent releases from the spent fuel pool through this line.
(Affidavit of Thomas Bordine on Board Question 1 at p. 3).
9.
If any radioactive liquids due to an accident in the spent fuel pool were somehow to reach and leak through CV/4027 t. hey would go to the radwaste system and therefore there would be no uacontrolled release to the environment.
(Affidavit of Thomas Bordine C) on Board Question 1 at p. 4).
2 F'/T 10.
Valve M0/7050 is the Main Steam Isolation valve (Affidavit of Thomas Bordine on Board Question 1 at p. 4).
11.
The main steam system, of which M0/7050 is a part, does not connect with spent fuel pool related piping or with the atmosphere above the spent fuel pcol, so failure of M0/7050 would not affect the results of an accident in the spent fuel pool.
(Affidavit of Thomas Bordine on Board Question 1 at p. 4).
12.
Valve CV/4105 is a control valve that provides demineralized water to the spent fuel pool and surge tank.
(Affidavit of Thomas Bordine on Board O
Question 1 at p. 5).
13.
No repetitive failures have occurred with Valve CV/4105.
(Affidavit of Thomas Bordine on Board Question 1 at p. 5).
14.
CV/4105 is located in a 2 inch line filled with water which is pumped into containment.
(Affidavit of Thomas Bordine on Board Question 1 at p. 5).
15.
Because numerous check valves and manual valves exist which prevent backflow out of containment through this line, Valve CV/4105 does not need to close to prevent any release of radioactive material from the spent fuel pool to the environment.
(Affidavit of Thomas Bordine on Board Question 1 at pp. 5-6).
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16.
Valve CV/4105 is not relied upon to add makeup water to the spent fuel pool in the event of a reactor accident which would preclude access to the containment.
(Affidavit of Thomas Bordine on Board Question 1 at p. 6).
C.
DISCUSSION As the affidavit of Thomas Bordine explains, valves CV/4096 and CV/4097 are the only valves mentioned in Board Question 1 whose proper operation is relied upon to mitigate the results of an accident in the spent fuel pool.
These valves are in series; only one of them has experienced repetitive leak rate failures; and this valve has been fixed.
- Moreover, O
both valves are scheduled to be modified in the forthcoming refueling outage (prior to installation of tha proposed racks) to ensure that they are independent and redundant, and therefore the likelihood of a simultaneous failure, which would be necessary to allow a release of radioactive materials through these valves to the environment, will be minimized.
The affidavit of Thomas Bordine fully answers Board Question 1 and therefore summary disposition is appropriate. OO
STATE OF MICHIGAN
)
)
SS.
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COUNTY OF JACKSON
)
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD IN THE MATTER OF
)
CONSUMERS POWER COMPANY
)
Docket No. 50-155 OLA (Big Rock Point Nuclear Plant)
) (Spent Fuel Pool Expansion)
AFFIDAVIT OF THOMAS C. BORDINE My nace is Thomas C. Bordine.
My business address is 1945 W. Parnall Road, Jackson, Michigan.
I am employed by Consumers Power Company as a Staff Licensing Engineer for the Big Rock Point Plant.
I have a Bachelor of Science Degree in Mechanical Engineering from Wayne State University, Detroit.
I joined Consumers Power Company in the Gas Engineer-ing Department of the Livonia, Michigan district office in June 1967 and remained at that location until December 1975 (excepting two years military leave in 1968-1970).
My primary responsibilities as Systems Planning Designer included directing the updating of the Gas Network Flow Analysis, as well as the formulation of proposed system networks.
In January of 1976, I transferred to the Big Rock Point Plant, Charlevoix, Michigan as Quality Assurance Engineer.
In that capacity, my responsibilities as a staff member of the site Quality Assurance Departruent included conducting QA Program Audits and Surveillances; performing quality reviews of O
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procedures, design and procurement documents; and monitoring and performing final close-out reviews of the site corrective action system documents.
.. In June 1977, I was assigned the position of n)
Quality Assurance Superintendent, Big Rock Point Plant.
(_
My responsibilities included organizing and directing the activities of the Plant Quality Assurance and Quality Control staff; providing assistance to the plant technical staff to more effectively implement the QA Program; analyz-ing the effectiveness of the QA Program and recommending program improvements to QA Department Management; reporting the status of QA Program implementation at the plant to QA Department Management; and representing the QA Program and QA Department to the NRC I/E Branch personnel.
During my assignment at Big Rock, I also participated in and com-pleted, in April 1980, an administrative certification
()
program for Senior Reactor Operator Equivalency.
The program included extensive training in plant systems, con-trols and licensing requirements as well as reactor start-up qualification at the General Electric Boiling Water Reactor Simulator.
l In August 1980, I was assigned to work at the Institute of Nuclear Power Operations, Atlanta, Georgia in a one-year on-loan capacity.
My primary responsibility at the Institute was to establish a process for the development and coordination of nuclear operations management criteria.
Since August 1981, I have been assigned as Big Rock Point Staff Licensing Engineer at the Corporate Office in Jackson, Michigan.
I am responsible for the coordination and response
()
of licensing activities related to the Big Rock Point Plant which include the spent fuel expansion project.
1
\\ I am the author of the documents entitled "Testi-O mony of Thomas C. Bordine Concerning O'Neill Contention II-B", " Testimony of Thomas C. Bordine Concerning O'Neill Contention II-C", and " Testimony of Thomas C. Bordine Concerning Licensing Board Question No.
1".
I believe that my educational background and work experience qualifies me to respond to those contentions and question.
Where state-ments of fact are made in this testimony, they are based on my personal knowledge.
Where results of calculations performed by Consumers Power Company are discussed, I have personally reviewed and verified the calculations.
I swear that this affidavit and the testimony attached hereto are true and correct, to the best of my O
knowledge and belief.
fYsm mN Thomas C.
B6rdine SUBSCRIBED AND SWORN TO before me this 2nd day of October 1981.
l A AtL Y bl/k(b/fj)
Notay? Fublic Jackson County, Michigan My Commission expires March 26, 1983
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