ML19257C425
| ML19257C425 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 01/17/1980 |
| From: | Dieterich R SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | CALIFORNIA, STATE OF |
| Shared Package | |
| ML19257C422 | List: |
| References | |
| NUDOCS 8001290100 | |
| Download: ML19257C425 (14) | |
Text
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January 17, 1980 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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SACRAMENTO MUNICIPAL UTILITY DISTRICT
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Docket No. 50-312
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(Rancho Seco Nuclear Generating Station) )
LICENSEE'S ANSWERS (SET NO. 1) TO CALIFORNIA ENERGY COMMISSION'S SECOND SET OF INTERROGATORIES TO THE SACRAMENTO MUNICIPAL UTILITY DISTRICT 5.
INTERROGATORY:
What is the expected frequency of feedwater transients at the Rancho Seco facility?
Ex pla in.
(a)
Quantify the anticipated increase in reactor trips that will be experienced due to the changes in reactor system pressure setpoints (high pressure trip and pressurizer relief valve actuation), and the addition of anticipatory reactor trips (on turbine trips and loss of f eedwater).
(b)
Since Rancho Seco received its operating license, how many feedwater transients have occurred without reactor trip and how many have resulted in a reactor trip?
Please list' all feedwater transients which have occur-red at Rancho Seco, stating for each its cause and whether AFW and/or HPI was required.
(c)
How would SMUD response to each transient listed in subpart ( b), above, have been different if procedures instituted since TMI had been in effect?
Explain.
ANSWER:
It may not be possible, and in any event Licensee has not attempted, to predict the number of feedwater transients that will take place at Rancho Seco in the future.
Howeve r, in the fif ty-seven months since the f acility went into commercial operation, there have bean only ten loss of feedwater transients.
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Assuming, as is reasonable to anticipate, that the rate of feed-water transients will not increase, there will be an average of two feedwater transients per year at Rancho Seco.
(a)
A tabulation of the increase in reactor trips since the TMI accident resulting from lowering the high pressure tripi setpoint and adding the anticipatory reactor trips on loss of feedwater and turbine trip is given in Attachment III to a letter. dated November 16, 1979, from Licensee to the NRC.
Addition of the anticipatory reactor trip on turbine trip should result in an increased number of reactor trips since, prior to this modification, the transient resulting from a turbine trip would not necessarily have caused a reactor trip.
It is not possible, however, to quantify this increase because one cannot determine the number of ';urbine trips that will take place in the future, or ascertain how many of those would have resulted in a reactor trip absent the modification.
Addition of the anticipatory trip on loss of feedwater should not result in an increase in the number of reactor trips.
(b)
Information on those feedwater transients resulting in power reductions of greater than 20% is contained in the monthly operating reports submitted by Licensee to the NRC.
For those f eedwater transients that did result in a reactor trip, the in-formation requested in this Interrogatory is contained in Attach-ment I to the above-referenced letter of November 16, 1979, from Licensee to the NRC.
All transients listed in Attachment I required AFW, which was provided in every instance.
Only the transients of March 20, 1978g and January 5, 1979,resulted in 18fkS. 225 HPI activation.
(c)
As far as plant response is concerned, the existence of an anticipatory reactor trip on loss of feedwater would have resulted in a reactor trip earlier in each transient than what was experienced.
In terms of operator response, an auxiliary feedwater flow indicator was installed in the control room at Rancho Seco following the TMI accident.
With the availability of this new instrumentation, operating procedures have been modified to require the operator to verify auxiliary feedwater flow f ollowing a loss of main f eedwater, in addition to the always required verification of startup of both auxiliary feedwater pumps. Therefore, the operators would have complied with this additional procedure in each feedwater transient.
7.
INTERROGATORY:
Discuss specific procedures and/or modifications wnich increase the frequency of initiation of the auxiliary feed-water system.
ANSWER:
None of the procedures or modifications which have been introduced at Rancho Seco since the TMI accident increases.the frequency of initiation-of the auxiliary feedwater system.
8.
INTERROGATORY:
Identify the most severe over-cooling event which has occurred at Rancho Seco.
During this event was operator action required to maintain adequate core cooling and stable conditions?
If so, describe.
ANSWER:
The most severe over-cooling event at Rancho Seco occurred on March 20, 1978, and was initiated by the failure of the non-nuclear instrumentation.
This event is described in Reportable Occurrence 78-01, reported to the NRC on March 20, 1978, and in a follow-up report dated March 31, 1978.
No operator action was required to maintain adequate core cooling, 18!h1226 but operator action was required to limit the cooldown and prevent overfilling the pressurizer.
10.
INTERROGATORY:
Describe any atypical welding materials which may have been used in the Rancho Seco reactor pressure vessel.
If the safety analyses of the operating limits of the primary system have been chanaed due to such materials, discuss the analyses performed, including the assumptions, conclusions and uncertainties of any such analyses.
Indicate whether under the limiting conditions of a severe over-cooling transient, any technical specification would be exceeded which relates to the integrity of the pressure vessel.
Are there any modifications which could be made to Rancho Seco which would eliminate or reduce the risks associated with this concerr.?
ANSWER:
There is a slight possibility that weld material containing higher than typical concentrations of silicone and lower than typical concentrations of nickel may have been used in a vertical seam belt line weld in the Rancho Seco reactor vessel.
When this possibility was discovered, more restrictive normal heatup and cooldown limits were imposed to preserve the forty-year life of the plant.
These limits were incorporated into the Rancho Seco Technical Specifications.
The vendor.
Babcock and Wilcox ("B&W"), performed an analysis of this possible atypical weldment (B&W Report " Evaluation of the Atypical Weldment", BAW 1566, August 1979) and concluded that the heatup and cooldown limitations could be relaxed.
The NRC Staff later performed a safety evaluation of the reactor vessel and, on December 12, 1979, it agreed with B&W that the limitations could be relaxed.
The cooldown limitations imposed as a result of the possibility of an atypical weldment are based on a maximum cool-down rate of 100 degrees F. per hour during normal operating conditions; therefore, the 100 degrees F. cooldown rate has 18dbS. 227.
become a limiting condition of plant operation.
By definition, the Technical Specifications cooldown rate limit of 100 degrees F. per hour would always be exceeded in the event of a " severe ove r-cooling transient".
It must be noted that the ability of the reactor pressure vessel to withstand a severe over-cooling transient is the same whether or not an atypical weldment exists; in either case, isolated instances of rapid depressurizations and cooldowns are acceptable; in either case, analyses must be performed af ter each transient to datermine the fatigue cycle usage produced by the cooldown.
The only distinction between a typical and an atypical weldment lies in the long-term life of the reactor vessel, and not in the vessel's ability to handle over-cooling transients.
Based on extensive analysis by B&W and the NRC Staff, Licensee believes there is no risk to the integrity of the Rancho Seco vessel during transients on account of the possible atypical'weldment, and knows of no plant modifications that should be made on account of this alleged concern.
20.
INTERROGATORY:
Are there any valves critical to safety-related tunctions at Rancho Seco that cannot be remotely monitored and controlled from the control room?
Explain.
ANSWER:
In the remote event of a specific size small break in a specific location in the primary system, together with a loss of off-site power and the failure of one diesel generator to start, operator action is required to open manually the two (out of four) high pressure injection valves powered f rom the failed diesel ger.arator.
The operator is then also required to open one manual valve in a line which cross-connects the two high i 8fk3-228 pressure lines.
All other valves which are safety related and require operation following an accident are automatically operated with position indication available in the control room.
25.
INTERROGATORY:
What is the data logging / display capability of the plant computer / printer?
Can the output of the system keep pace with an information flow as great as that of TMI?
Indicate capacity limits, mechanisms for handling temporary information overloads, limits of overload capacity, and logic for handling conditions which exceed overload capacity.
ANSWER:
The main computer includes the following print / display equipment:
(1)
Alarm typer - nominal 150 characters per second (cps) print rate, used to print analog and digital alarms.
(2)
Utility typer - nominal 150 cps print rate, used to print groups of analog / digital point values on request.
(3)
Line printer - nominal 150 cps print rate, used to print out large groups of calculated results.
(4)
Log typer - nominal 25 cps print rate, prints hourly, daily demand performance logs.
(5)
Digital display - displays value of one analog (or calculated) point selected by the operator.
(6)
Analog trend recorders - capable of charting up to 6 point values as selected by the operator.
The Rancho Seco print devices (items 1, 2 and 3 above) have been upgraded to faster models than those used at TMI-2.
The TMI-2 typers are capable of printing 15 cps whereas those at Rancho Seco can print at 150 cps.
However, current system sof tware limitations reduce Rancho Seco's effective print rate' to 40-45 i Offk. 9 9 0 cps.
Therefore, Rancho Seco's computer can provide printed information approximately three times f aster than what was available at TMI-2.
Licensee does not have knowledge of the rate of data input to the TMI-2 computer, hence it is unable to determine whether the Rancho Seco computer would be able "to keep pace with an information flow as great as that of TMI".
The following discussion of overload capacity assumes, for illustration purposes, that the Interrogatory is restricted to the Rancho Seco computer's handling of alarm printouts.
The processing of alarm messages makes use of two core memory buffers and one disk memory buffer.
This arrangement provides a tempo-rary storage area, apart from the central computer, where messages are kept until the relatively slow output device can i
process them.
By means of this arrangement, the central computer is freed to perform other functions.
Buffering is done in two steps:
(1) The originating program assembles pertinent informa-tion and stores this information in either the main core memory buffer, which holds twenty such messages, or in a similar backup memory buffer, if the main buffer is busy or full.
This process takes a few microseconds.
(2) Information in these core buffers is transferred, as alarms occur, to the disk memory buffer at a rate which is limited by the disk's transfer rate.
This procest takes a few milliseconds per alarm.
The disk memory buffer can contain up to 1,365 alarm inputs, which are awaiting output, at any one time.
Licensee is currently implementing another computer system which acts as a backup to the main plant computer and operates independently.
This second computer monitors only the 18250- 230 most critical plant parameters, and since it is of an advanced design, it can process alarm messages at a much higher rate on the order of two or more alarms per second with essentially unlimited overload storage.
26.
INTERROGATORY:
Is the capacity of the rad waste treatment equipment adequate to handle radioactive releases of gases and liquids equivalent to those experienced at TMI?
Explain.
ANSWER:
Rancho Seco's radioactive waste treatment equipment does not have the capacity to treat radioactive releases of gases and liquids " equivalent to those experienced at TMI".
However, in the event of a comparable accident, tha containment isolation system at Rancho Seco would isolate any such releases, so that the gases and liquids which were released outside the containment during the Three Mile Island accident would be retained in the containment building at Rancho Seco.
Any such materials would then be treated in preparation for eventual disposal, in a similar manner to what is presently being done at TMI.
35.
INTERROGATORY:
If SMUD denies in whole or in part any of the requested admissions filed by CEC to SMUD, dated December 21, 1979, please provide the basis for each such denial.
ANSWER:
The following answers provide the basis for some of the denials asserted in " Licensee's Answers to California Energy Commission Requests for Admissions to Sacramento Municipal Util-ity District".
The second number identifying each answer below corresponds to the number of the admission requests denied.
ANSWER 3 5-7 :
Based upon information provided by the NRC Staff 183E 231 in NUREG-0560, " Staff Report on the Generic Assessment of Feedwater Transients in Pressurized Water Reactors Designed by the Babcock and Wilcox Company", Licensee believes that the design of the auxiliary feedwater systems at Rancho Seco and TMI-2 are not substantially identical.
For example, the Rancho Seco auxiliary feedwater system utilizes one motoradriven pump with a capacity of 840 gpm and one motor and turbine tandem driven pump also with a capacity of 840 gpm.
The TMI-2 system utilizes two motor driven pumps, each with a capacity of 470 gpm and one turbine driven pump with a capacity of 940 gpm.
I ANSWER 35-10:
The power operated relief valve and two code safety valves at Rancho Seco discharge into an 8,450-gallon pressurizer relief tank which can then be emptied into a 34,000-gallon reactor coolant drain tank.
Licensee understands that these three valves at TMI-2 discharge into a reactor coolant drain tank Licensee believes has a capacity of 7,500 gallons.
This tank can then be emptied into three reactor coolant bleed tanks which have a volume unknown to Licensee.
ANSWER 35-15 :
Rancho Seco has primary system vents at the top of each control rod drive housing, on each incore instrument cube, and at the high point on each hot leg.
In addition, the power operated relief valve, safety valves, and pressurizer vent and sample line may be used to vent the primary system.
ANSWER 3 5-17 :
Had the same failures and operator responses occurred at Rancho Seco as at TMI-2, radioactivity would not have been released to the auxiliary building because of a different containment isolation system at Rancho Seco.
ANSWER 3 5-18 :
The failure of the power operated relief valve to close was one of several events leading to the accident at TMI-2.
Howeve r, this was not a significant contributing f actor to the accident.
The plant is capable of responding to such an event, as demonstrated by other documented cases where power operated relief valves have failed to close and the event has been mitigated successfully.
ANSWER 35-19:
The pressurizer capacity is adequate to accommodate a pressure drop to the initiating point of high pressure injection.
ANSWER 3 5-20 :
The elevation of the steam generator relative to the reactor vessel is adequate to achieve natural circulation and was therefore not a significant contributing factor to the accident at TMI-2.
ANSWER 3 5-21:
Voiding in the primary system does not neces-sarily result in an inability to achieve natural circulation since the high pressure injection system can be used to raise pressure to a point where voids would condense and allow natural circulation to occur.
Licensee understands that the inability to achieve natural circulation at TMI-2 was due to the presence of non-condensible gases, and not because of voiding.
18 233 ANSWER 3 5-22:
The capacity of the feedwater side of th'e steam generators was not a significant contributing factor to the accident at TMI-3 Core cooling could have been maintained with the volume of water present on the feedwater side of the steam generator at TMI-2.
ANSWER 3 5-23:
The lack of direct indication of reactor coolant inventory apart from pressurizer inventory was not a significant contributing f actor to the accident at TMI-2.
The significant factor that the operators should have recognized was the lack of a subcooled condition in the primary system as indicated by reactor coolant temperature and pressure.
ANSWE R 3 5-24 :
The ability to vent gases from high points in the primary system would have facilitated natural circulation cooling at TMI-2 af ter a significant quantity of gases was present.
The non-condensible gases were generated during the period of core damage, therefore venting would not have mitigated the accident.
ANSWER 3 5-27:
The failure of the reactor containment to isolate led to the release of radioactivity.
The TMI-2 reactor coolant drain tank was not intended to operate as a primary fission product barrier.
ANSWER 3 5-31:
The hydrogen buildup at TMI-2 was within the re-actor vessel where the hydrogen is generated during the metal-water reaction.
A hydrogen recombiner was not requirld at TMI-2 1834 234
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since there was not a sufficient buildup of hydrogen in the containment for an explosion sufficient to exceed the containment de s ig n.
ANSWER 3 5-3 7:
Forced circulation cooling is available notwith-standing voiding or loss of pressure in the primary system.
ANSWER 3 5-38:
A loss of of f-site power together with a turbine trip at Rancho Seco would disrupt forced circulation, but only until temperature was reduced to a point where the decay heat system could provide forced circulation while still maintaining subcooling.
ANSWER 3 5-40 :
Forced circulation cooling is not unavailable because of voiding in the primary system.
If forced and natural circulation cooling are not available, natural condensation cooling and/or ECCS injection can provide core cooling.
ANSWER 3 5-4 2:
See Answer 35-40.
ANSWER 3 5-4 3 :
An analysis of boiling and venting in the primary system has been performed which shows it to be a reliable means of core cooling.
See Licensee's answer (in Set No. 3) to Interrogatory 11.
ANSWER 35-50:
Natural condensation cooling was assumed in the small break LOCA analysis performed in support of the licensing of Rancho Seco.
i856 235 ANSWER 35-53:
If the reactor coolant pumps are operating, the high pressure injection system will adequately maintain the reactor coolant inventory for core cooling during small breaks af any size.
See Licensee's answer (in Set No. 3) to Interrogatory 12.
ANSWER 3 5-5 8:
The Rancho Seco high pressure injection system is not expected to be activated automatically during feedwater transients not accompanied by a small break LOCA.
Typically reactor coolant system pressures do not drop to the low pressure trip point for safety features initiation following a feedwater transient.
37.
INTERROGATORY:
Provide the System Design Description together with operator procedures and equipment loading sequences for the diesel generators at Rancho Seco.
ANSWER:
The design description of the sources of auxiliary power includf.ig the diesel generators is included in Section 8.2.3 of the FSAR.
Included in this description is a table providing the equipment loading sequences for the diesel generators.
Operating Procedure A-31 provides procedures for operation of the diesel generators.
18S:2 236-UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
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EACRAMENTO MUNICIPAL UTILITY DISTRICT
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Docke t No. 50-312
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(Rancho Seco Nuclear Generating Station) )
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AFFIDAVIT OF ROBERT A.
DIETERICH County of Sacramento )
- SS.
State of California
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Robe rt A.
Dieterich, being duly sworn according to law, deposes and says that he is a Senior Nuclear Engineer in the Generation Engineering Department of the Sacramento Municipal Utility Dis trict; and that the answers contained in " Licensee's Answers (Se t No. 1) to California Energy Commission 's Second Se t of Interrogatories to the Sacramento Municipal Utility District" are true and correct to the bes t of his knowledge and belief.
Robert A.
Die te rich Sworn to and subscribed before me this day of January, 1980 Notary Public OFFICI AL SEAL PATRICIA K. GEISLER
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1834 237 wm aasue ou<cm
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My Commeron Emmres A emoer 22.1983
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