ML19312D597
| ML19312D597 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 03/07/1980 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Howell S CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| References | |
| NUDOCS 8003250056 | |
| Download: ML19312D597 (9) | |
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March 7, 1980 Doeket Nos.:
50-329/330 Mr. S. H. Howell Vice President Consumers Power Company 212 West Michigan Avenue Jackson, MI 49201
Dear Mr. Howell:
SUBJECT:
SUPPLEMENTAL 10 CFR 50.54 REQUESTS REGARDING B&W SYSTEM S We are continuing our review of your December 4,1979 reply to our 10 CFR 50.54(F) requests of October 25, 1979 regarding the design adequacy of Babcock & Wilcox (B&W) Nuclear Steam Supply Systems utilizing once-through steam generators for Midland Plant, Units 1 and 2.
We find that we need additional information regarding the changes and studies proposed in Appendix F of your reply.
The information needed is listed in Enclosure 1.
We would appreciate your reply by April 3, 1980, consistent with our scheduled meetings with ACRS on April 8 (Subcommittee) and April 10 (Full Committee) on this matter.
Should you have questions or need clarification for these requests, please do not bestitate to contact us.
Sincerely, 1
S-L. S. Rubenstein, Acting Chief Light Water Reactors Branch No. 4 Division of Project Management
Enclosure:
Request for Additional Information cc w/ enclosure:
-See next page.
8'O0325o056
O Consumers Power Company 0
- ccs:
Michael I. Miller, Esq.
Isham, Lincoln & Beale Suite 4200 One First National Plaza
- Chicago, Illinois 60603 Judd L. Bacon, Esq.
Managing Attorney Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201 Mr. Paul A. Perry Secretary Consumers Power Company 212 W. Michigan Avenue Jackson, Michigan 49201 Myron M. Cher ry, Esq.
One IBM Plaza Chicago, Illinois 60611 Mary Sinclair.
5711 Summerset Drive Midland, Michigan 48640 Frank J. Kelley, Esq.
Attorney General State of Michigan Environmental Protection Division 720 Law Byilding Lansing, Michigan 48913 Mr. Wendell Marshall Route 10 v,
Midland, Michigan 48640 t
Grant J. Merritt, Esq.
Thompsen, Wielsen, Klaverkamp & James
- 4444 IDS Center-80 South Eighth Street Minneapolis, Minnesota 55402
- Mr. Don van Farowe, Chief Division of Radiological Health Department of Public Health P. 0. Box 33035 Lansing, Michigan 48909 1
Consumers Power Company ces (continued):
Resident Inspector / Midland NPS c/o U.S. Nuclear. Regulatory Comission P. O. Box' 1927 Midland, Michigan 48640 William J. Scanlon, Esq.
2034 Pauline Boulevard Ann Arbor.,-Michigan 48103 O
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F.1 Your discussion in Appendix F of the pre-THI-2 changes for Midland states s
-that newer control systems hardware (non-nuclear instrumentation [NNI]/
integrated control system [ICS]) using dual auctioneered power supplies for logic modules rather than individual power supplies are being used.
.For this modification, provide the logic and/or your failure mode and a.
effects analysis that shows how systems will respond to ' failure in the power supply and input parameters.
Also provide your design criteria for the ICS with respect to these types of failures.
Infor-mation'in the FSAR may be referenced or supplemented as appropriate for this response, b.
Operating events at several plants with B&W NSSS designs (including Rancho Seco in March 1978; Oconee Power Station, Unit 3 on November 10, 1079;'and the Crystal River Station on February 26, 1980) have occurred which resulted in loss of power to the ICS and/or NNI sys-tem.
The loss of power resulted in control system malfunctions, feedwater perturbations, and significant loss of or confused infor-mation to the Operator.
NUREG-0600 also discusses LER 78-021-03L on Three Mile Island, Unit 2 whereby the RCS depressurized and safety injection occurred on loss of a vital bus due to inverter failure.
Discuss the extent'to which these events would have been mitigated or precluded by the changes incorporated into the Midland design.
Include a response to action items 1 to 3 required of near-term
. licensees in IE Bulletin 79-27 and identify corrective actions you consider appropriate as a result of the Crystal River event.
F. 2
.We are~ concerned that an instability in the ICS could iead to transients initiating with plant parameters more severe than those assumed for the safety analysis or significantly increase ~the number of challenges to the 1
protection system during early plant life.
In this regard:
The Midland ICS includes a significant difference from ICS designs of a.
other plants in its evaporator steam development.
Describe all studies and tests which have been and will be conducted to establish stability;and reliability of the Midland ICS design.
b.
Operating experience at the Crystal River plant has indicated a control instability for the integrated control system when bringing the plant up to. power with a pump out of service.
Specify your criteria ~and describe Midland design features to preclude this type of' instability.
Describe your design criteria, features, and operational requirements c.
.for the ICS and its supporting systems to preclude instabilities when:
(1) switching from manual-te automatic control and vice versa
'or (2) switching from one operating mode for process steam to another i
mode.
1
F.3 Experiece at operating B&W plants have indicated that the dynamics associated with main feedwater term nation and steam generator pressure control following s reactor trip can Nad.to overcooling of the primary system.
Discuss your criteria and the dequacy of your existing and proposed design features and changes to preclude this overcooling situation.
F.4 Your response states that you intend to bring together information from B&W and your own evaluations of B&W operating plant experience coupled ~
with the ICS-FMEA and a B&W review of overcooling transients to identify the changes which may significantly decrease the frequency of upsets to feedwater.
State when.this review and analysis of the MFW system will be performed and how the recommendations-and studies proposed in your response are likely to be affected by your results.
F. 5 Other applicants responding to our October 25, 1979 requests have con-sidered the following additional items to further decrease undesirable perturbation in the once-through steam generator:
Increased demineralized water makeup capacity to the condenser a.
hotwell during runback following a turbine trip.
b.
Increased bypass capability around the condensate polishers with fast acting valves.
Discuss these and like considerations which you have given to the Midland design and their effectiveness.
F.6 Discuss the advantages and disadvantages, if any, of a control independent of the ICS to terminate main feedwater flow following a reactor trip.
F.7 Specify the extent to which control limitations such as valve and pump speed responses effect main feedwater stability, particularly:
(1) during startup from the manual to the automatic operational mode or (2) during automatic switchover from one process steam mode to another.
F.8 State the design objectives of the revised auxiliary feedwater control system.
Also indicate whether it will:
. Initiate for all loss of MFW events, either total or partial, hnd at
'a..
what lower limit; b.~
Initiate on SIAS; c.
Initiate on loss of offs'ite power;
' d.-
Preclude overcooling or undercooling of the primary system even with a single failure'in the system (e.g., failures in input, power, valves);
and.
i o
- 5 e.-
Interact in any adverse fashion with the Feed-Only-Good-Generator interlock.
Also, describe h'ow you will demonstrate that the dynamic response has been achieved.
F. 9 For your intended revision to the AFW initiation logic, identify the signals (e.g., generator level, no feedwater flow, loss of pump suction pressure, SIAS, and loss of steam flow to pumps) that will be used to initiate AFW and justify their use.
Also, update your response to our request 031.51 to identify the type and characteristics of the revised
-transmitters selected for the reverse feedwater-flow monitoring system.
F.10 You state that changes to the Midland auxiliary feedwater configuration since the TMI-2 accident will include:
i-a.
Modification of the AFW pump suction piping.from one interconnected system for both Midland units to two systems operating independently to supply AFW to each unit, and b.
The addition'of redundant flowpaths from the discharge of each AFW pump to each steam generator.
Provide a simplified diagram illustrating the previous and revised con-
' figurations.. Include a table denoting valve positions during normal and abnormal operating conditions.
Specify your schedule for completion of the details of the revised Midland'AFW design.
F.11 In the event of a steam line break upstream of the MSIVs accompanied by failure of a MSIV in.the intact line, reliance is placed upon reliable but non-safety grade circuits and downstream valves to isolate steam flow except for residual flows associated with turbine gland sealing, etc.
Describe the behavior of the revised F0GG interlock during this accident scenario, including.the significance, if any, of the residual steam flow
' limits on the F0GG. system.
~
F.12
.In addition to tne FMEA'for the revised F0GG interlock to be provided as t
part of your revised AFW evaluations, identify those events and combina-tions of events which have been.and will be evaluated to assure that no confused or inadvertent inputs (such as from a previously unrecognized event or event combination) can lead to a-malfunction or undesirable operation of the F0GG system.
Also describe any sttidies and tests performed to assure proper integration and interaction of the F0GG interlock with other systems.
i F.13 Describe the results of your design evaluation' studies for the several
. process steam operating modes-performed'to determine whether any unique opportunities for operator or equipment errors (such as improper system
. alignments, including misalignments between circuitry and hardware) or
' adverse interactions unique to a given mode exist which could. lead to overcooling or overpressurization transients <or accidents more severe than
4 those for which the protection systems have been analyzed and designed.
Identify the maximum potential contribution of the process steam to the sensitivity of overcooling events for the Midland plant, whether as a result of heat extraction through the tertiary heat exchangers or via any control system change-influenced by the Dow use of steam.
F.14 You state that you are currently investigating modifications to the AFW level control system to limit primary cooldown rate following AFW actua-tion.
Describe how the control modifications under consideration would provide the capability to distinguish in a positive manner between
~ transients and accidents with regard to SG level setpoint control.
Also describe how two phase level during swell from depressurization affects level detection and how this is treated in the analyses.
F.15 The modifications, recommendations, and studies you present to reduce sensitivity are in the direction of additional automation of the plants.
While this approach leaves the operator free to verify system performance and should improve the control of transients, we are concerned that poten-tial system interaction effects might result.
Therefore, e complete and integrated review of the primary and secondary system should be performed to assure that no significant adverse interactions result from the modifications that are ultimately made.
Describe your plans and schedules with regards to performing such a comprehensive, integrated evaluation of these changes, based upon conservative and realistic analyses and simu-lator comparisons as appropriate.
F.16 Provide the following analyses:
Overcooling event initiated by steam pressure regulator malfunction a.
resulting in increased steam flow.
.b.
Overcooling event initiated by feedwater system malfunctions that result in decreased feedwater temperature.
~ For these analyses, assume no beneficial operator action befare 10 minutes.
Also, only qualified safety systems should be assumed for mitigation.
Identify which safety and nonsafety grade systems are considered to operate during this transient and specify the part each of these systems take in the transients.
Identify the signals acting upon these systems during the transients.
The analyses should be performed for a period of at least 10 minutes after transient initiation.
If exisiting analyses which are presented for a
-shorter dur tion are utilized for_this' response, then confirm that during i
the time n shown out to 10 minutes:
. (1) No operator action is required or assumed, 1
(2) No changes in operating. systems are required, t
-5 (3) No significant changes result out to 10 minutes, such that
~
extrapolation from the results presented is considered valid.
F.17' You have stated during related meetings with NRC and with ACRS sub-committees that the analyses presented in your current 50.54(F) response were not necessarily select 2d to represent the worst case.
Provide your recommendations as to what criteria, assumptions, and experience should be recognized in defining the worst case for design purposes.
F.18 Regarding your proposed changes to the pressurizer level indication, specify the new location of the instrumentation taps and revise FSAR Figure 3.8-73 accordingly.
Also provide or reference the relationship between " indicated" and " actual" level for the revised Midland design.
F.19 Provide additional detail regarding the safety sequence analyses to be performed by your contractor, EDS Nuclear.
Identify the 16 safety and operational sequence diagrams and the 15 auxiliary system diagrams to be used as the vehicle for this review.
Describe the end product of this study and describe how these results will be factored into the AT0G program.
F.20 Your reply notes that an in-depth reliability assessment is being performed on the Midland AFW systems:
a.
Studies performed by several operating nuclear plants have concluded that a significant improvement in reliability and plant availability results from addition of a second motor-operated auxiliary feedwater pump. We require that the benefits-from such an addition be included as part of the results of your reliability assessment of the Midland AFW system.
b.
Other than the auxiliary feedwater system, what Midland systems and changes will be the subjects of your reliability assessments? State your planned completion date for these analyses.
F.21
~0n what basis did you determine that pressurizer heater banks 5 and 6 alone will provide sufficient heating capacity if only these banks are uprated to safety grade?
Identify the limiting transient or accident which established the required heating capacity.
What provisions for equipment failure are provided in this selection?
F.22 You note that the existing pressurizer. heater low level interlock design is being reviewed to determine its adequacy in the event of loss of liquid inventory in the pressurizer.
Describe how energized pressurizer heaters fail when uncovered and provide justification that such failure would not threaten or cause. failure'of the-reactor coolant pressure boundary.
F.23 You state that a subcooling meter will be provided with redundant safety grade hot leg temperature and reactor coolant system pressure input.
. Clarify whether it is your intent to provide a subcooling meter which is itself safety grade. -If not, justify your position.
Specify the detec-tion and indicating range and sensitivity for this meter and its inputs.
r I
j.
F.24 You state that' the technical feasibility of providing a low flow indi-cation as a means of confirming core cooling during natural circulation modes of cooldown is being assessed. What criteria are being used for this assessment? What power requirements for this instrumentation are intended?
F.25 In view of the experience from the TMI-2 accident, justify your proposed use of non safety grade equipment (core exit thermocouples with the plant
- computer) as a means of determining adequate core cooling.
What physical or practical limitations, if any, preclude use of a safety grade system for this purpose? Your justification should be coupled with the fact that a positive, direct means for detection and removal of a gas bubble from the reactor vessel head is not yet included in your proposals.
Include in your discussion what backup is provided for operation when the plant computer is down.
Also, specify the range and sensitivity of the detec-tion and indication measurements.
F.26 To prevent automatic tripping of the reactor coolant pumps due to ESFAS initiated by overcooling events, you state tnat the Midland pump trip logic will include coincidence circuitry sensing pump motor current.
This input is intended to actuate on degraded pump current indicative of significant RCS void formation characteristic of a LOCA; but for overcooling events, the extent of void formation should not reach a point where degraded pump current will trip the pumps and undesirable pump trip will thus be avoided.
Describe the significant elements of the develop-ment program for this circuitry, including that phase directed to the distinction of a valid motor current signal.
What criteria will dis-tinguish a valid signal? How will the system be verified in an actual nuclear power plant or under realistic conditions? Provide your current schedule for this program.
F.27 After the PORV closed during the transient at Crystal River Unit 3 on February 26, 1980, the reactor coolant system pressure increased from approximately 1300 psi to 2400 psi in less than 3 minutes.
The last 600 ps! (from 1800 to 2400 psi) of this increase occurred in less than 1 minute.
This caused lifting of the code safety valves.
Operating guidelines for B&W supplied plants typically recommend termination of high pressure injection when hot and cold leg temperatures are at leas 50 F below the saturation temperature.of the existing reactor coolant system pressure and the action is necessary to prevent the indicated pressurize P level from going off scale.
In view of this characteristic of rapid repressurization, what operator action, and basis thereof, is proposed to reduce the potential for lifting of the Midland code safety valves?
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