ML19323A576
| ML19323A576 | |
| Person / Time | |
|---|---|
| Site: | Washington Public Power Supply System |
| Issue date: | 03/25/1980 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Strand N WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| References | |
| NUDOCS 8004210408 | |
| Download: ML19323A576 (6) | |
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Docket Nos... 190 and 50-513 R 2 5 1980 Mr. N. O. Strand Managing Director Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352
Dear Mr. Strand:
SUBJECT:
SUPPLEMENTAL 10 CFR 50.54 REQUESTS REGARDING B&W SYSTEM SENSITIVITY FOR WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PROJECTS 1 & 4 (WNP 1 & 4)
We are continuing our review of your December 3,1979 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 WNP 1 & 4.
We find that we need additional information regarding the changes and studies proposed in Appendix F of your reply. The information need is listed in the Enclosure.
Please soply the requested information within sixty (60) days after receipt of this letter.
Please contact us if you desire c!w riscussions or clarification of the information requested.
Sincerely, Gk b.f Olan D. Parr, ief Light Water Reactors Branch No. 3 Division of Project Management
Enclosure:
l Request for Additional l
Information l
cc w/ enclosure:
l See next page 80 04 21 g4 og
l MAR 25 rg Mr. N. O. Strand cc:
Mr. B. D. Redd Jerome E. Sharfman United Engineers & Constructors, Inc.
Atomic Safety and l
30 South 17th Street Licensing Appeal Board Phi 1#delphia, Pennslylvania 19101 U. S. Nuclear Regulatory Commission Nicholas S. Reynolds, Esq.
Washington, D. C.
20555 DeBevoise & Liberman 1200 Seventeenth St., N. W.
Resident inspector /WPPSS NPS Washington, D. C.
20036 c/o U. S. NRC P. O. Box 69 Mr. E. G. Ward Richland, Washington 99352 Senior Project Manager Babcock & Wilcox Company P. O. Box 1260 Lynchburg, Virginia 23505 Robert Lazo, Esq., Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Dr. Donald P. deSylva Associate Professor of Marine Science Rosenthiel School of Marine and Atmospheric Science University of Miami Miami, Florida 33149 Or. Marvin M. Mann Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Richard S. Salzman, Chairman Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Comission Washington, D. C.
20555 Dr. John H. Buck Atomic Safety and Licensing Appeal Board U. S. Nuclear Regulatory Comission Washington, D. C.
20555 I
I e
EllCLOSURE REQUEST FOR ADDIT!0flAL INFORMATI0fl s
F.1 Your discussion in Appendix F of the pre-TMI changes for WNP 1/4 states that newer control systems hardware (non-nuclear instru-mentation (NNI)/ integrated control system (ICS)) using. dual auctioneer, power supplies for logic modules rather than individual power supplies are being used, a.
For this modification, provide the logic and/or your failure mode and effects analysis that shows how systems will respond to failure in the power supply and input parameters. Also provide your design criteria for the NNI and ICS with respect to these types of failures.
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, 1979; and the Crystal River Station on February 26,1980) have occurred which resulted in loss of power to the ICS and/or NNI system. The loss of power resulted in control system malfunctions, feedwater perturbations, and significant loss of or confused information 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 WNP 1/4 design. Include a response to action items 1 to 3 required of near-term licensees.in Bulletin 79-27 and items 2, 4, 5 and 6 of Enclosure 3 of letter dated March 6,1980 to all
., operating B&W Reactor Licensees pertaining to the Crystal River event.1 F.2 We are concerned that control system reponse could lead to transients initiating with plant parameters more severe than those assumed for the safety analysis or significantly increase the number of challenges to the protection system during early plant 1tfe. In this regard:
a.
Operating experience at the Crystal River plant has indicated a control system response problem when bringing the plant up to power with a pump out of service. Specify your criteria and describe WNP 1/4 design features to preclude this type of response problem.
b.
Describe your design criteria, features, and operational' l
requirements for the ICS and its supporting systems to preclude L
control response problems when switching from manual to auto-l matic control and vice versa.
F.3 Experience at operating B&W plants have indicated that the dynamics associated with main feedwater termination and steam genrator pressure control following a reactor trip can lead to overcooling of the primary system. Discuss your criteria and the adequacy of your existing and proposed design features and dianges to preclude this overcooling situation.
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i 2-F.4 Discuss the advantages and disadvantages, if any, of a control independent of the ICS to terminate main feedwater flow following a reactor trip.
F.5 Specify the extent to which control limitations such as valve and pump speed responses affect main feedwater response during l
startup from the manual to the automatic operational mode.
F.6 State the design objectives of the improved auxiliary feedwater control system. Also indicate whether it will:
a.
Initiate for all loss of MFW events,either total or partial and at what lower limit; b.
Initiate on loss of offsite power; c.
Preclude overcooling or undercooling of the primary system even with a single failure in the system (e.g., failures in input, power, valves);
l d.
Interact in any adverse fasion with the Feed-Only-Good-Generator interlock.
F.7 For your intended revision to the AFW intiation 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.
F.8 In addition to the improved FOGG logic to be provided as part of your revised AFW evaluations, identify those events and combin-ations 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 studies and tests perfomed to assure proper integration and interaction of the F0GG interlock with other systems.
F.9 You state that you are censidering changes to improve the algorithm used for AFW flow control to limit primary cooldown rate following AFW actuation. Describe how these changes would provide the capability to distinguish in a positive manner between transients and accidents.. Also describe how two-phase level during swell from depressurization affects level detection and how this will be treated.
F.10 The modifications, reconsnendations, 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 potential system interaction effects might result. Therefore, a complete and integrated review
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3 of the primary and secondary system should be performed to i
assure that no significant adverse interactions result from the modifications that are ultimately made. Describe your plans and schedules with regard to perfoming such a comprenensive integrated evaluation of these changes, based upon conservative and
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realistic analyses and simulator comparisons as appropriate.
l l
F.11 Provide the following analyses:
l a.
Overcooling event initiated by steam pressure regulator or throttle valve malfunction resulting in increased steem flow.
b.
Overcooling event initiated by feedwater system malfunctions i
t that result in decreased feedwater temperature.
For these analyses, assume no beneficial operator action before
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10 minutes. Also, only qualified safety systems should be assumed for mitigation. Identify which safety and nonsafety grade systems are consdiered to operate during this transient and specify the part I
each of these systems take in the transients. Identify the signals acting upon these systems during the transients.
l The analyses should be performed for a period of at least 10 i
minutes after transient initiation. If existing analyses which are presented for a shorter duration are utilized for this response, i
then confim that during the time now shown out to 10 minutes:
j (1) No operator action is required or assumed, l
(2) No changes in operating systems are required.
(3) No significant changes result out to 10 minutes, such that extrapolation from the results presented is considered i
valid.
I F.12 You have stated during related' meetings with NRC and with ACRS subcomitteesthat the analyses presented in your current 50.54(f) i response were not necessarily selected to represent the worst case. Provide your recommendations as to what criteria, i
assumptions, and experience should be recognized in defining the l
worst case for design purposes.
l F.13 To prevent automatic tripping of the reactor coolant pumps due to ESFAS initiated by overcooling events, you state that the WNP 1/4 pump trip logic will include coincidence circuitry sensing pump motor nurrent. This input is intended to actuate on degraded pump current indicative of signific' ant RCS,oid formation characteristic of a LOCA; but for overcooling events, the extent of void formation shouTd 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 development program for this circuitry, including that phase directed to the distinction of a valid motor current signal. What criteria will distinguish a l
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, valid signal? How will the system be verified in an actual nuclear power plant or under realistic conditions? Provide your i
current schedule for this program.
F.14 After the PORV closed during the transient at Crystal River Unit 3 on February 26, 1980, the reactor coolant system pressure increased from approximately 1400 psi to 2400 psi in less than 3 minutes.
The last 600 psi (from 1800 to 2400 psi) of this increase occurred in less than 1 minute. This caused lif*.ing of the code safety valves.
Operating guidelines for B&W supplied plants typically recornend tennination of high pressure injection when hot and cold leg temper-l 0
atures are at least 50 F below the saturation temperature of the existing reactor coolant system pressure and the action is necessary I
to prevent the indicated pressurizer level from going off scale.
In view of this characteristic of rapid depressurization, what operator action, and basis thereof, is proposed to reduce the potential for lifting of the WNP 1/4 code safety valves?
F.15 It is our understanding that the B&W 205 plants operate with a considerably smaller water inventory in the steam generator than the B&W 177 plants. Explain what effect this has on the sensitivity of the 205 plants to both undercooling and overcooling events.
Include the impact of MFW and AFW response times and reliabilities in your evaluation, t
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