ML19339D110
| ML19339D110 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 02/09/1981 |
| From: | Basdekas D NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Tourtellotte J NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| Shared Package | |
| ML19339D106 | List: |
| References | |
| NUDOCS 8102170266 | |
| Download: ML19339D110 (3) | |
Text
{{#Wiki_filter:l /yto* *0 - t ?' g UaHTED STATES a 1 NUCLEAR REGULATORY COMMISSION n ! e,l* j WASHINGTON, D. C. 20555 4...../ m3w MEMORANDUM FOR: Jame:, P. Tourtellotte, Esq. Assistant Chief Hearing Counsel, ELD FROM: Demetrios L. Basdekas Reactor Safety Engineer, RSR, RES
SUBJECT:
SAFETY IMPLICATI0flS OF CONTROL SYSTEMS AND PLANT DYNAMICS, AND THEIR RELEVANCE TO THE TMI-l RESTART ASLB HFARING This is in response to Mr. Cutchin's request of February 3,1981
- o provide a written explanation of how my views, presented in documents provided to the Board, have direct application to THI-1.
I have read pages 11,027-11,030 of the hearing transcript and I believe that the Board desires an explanation specifically focused on my statement that "Even though [the issue of the effects on safety of the control systems] has been treated as a generic issue, it applies directly to the TMI-1...." I believe that the center of the Board's question is the word directly. The following explanatory remarks are intended to answer the Board's question on this point. The fact that a Failure Modes and Effects Analysis (FMEA) has been performed for the Integrated Control System (ICS) by Babcock & Wilcox does not mean that it has been an effective one in identifying important weaknesses in the THI-1 (a sister plant of THI-2) control systems. The recommendations I make in my memorandum to Dr. Ailearne of September 4,1979 (Reference No.1 in my memo to you of October 10, 1980) with respect to follow-up effort to complete the FMEA with the objective of accounting plant-unique features, applies directly to the TMI-1, in that the B&W performed FMEA was never extended, as it should have been, to include the TMI-l plant design features of its control systems and plant dynamics, which are unique to it. An evanple of lack of such effectiveness is on page 4-32 of B&W-1564 (copy attached). In Item 1-30 it is stated that no effect is expected for the case of steam-generator-level loss of control. This statement is not correct. The implications of this and related failures in the main feedwater control system are discussed in documents No. 12 and 16 on the list of documents I supplied to Mr. Cutchin of your office on October 30, 1980. Control system and other "non-safety" system failures on the secondary side may result in a rapid overcooling of the primary system subjecting the reactor vessel to a pressurized thermal shock that would threaten its very structural integrity. During our meeting in your office with NRR representatives sometime in early September 1980, I mentioned, 8302170 M.
s 2 If8 9 rg;; as an example, that my understanding was that TMI-1 was one of two plants in the country that did not have a " safety grade" main feedwater pump trip function on reactor / turbine trip. Furthermore, I believe that testimony which had been prepared by t.1e staff on this generic issue for the TMI-l Restart' Hearing needed to be challenged. I am addressing this point in the fourth paragraph of my memorandum to you dated October 10, 1980. My use of the word ~ directly was intended to point the direct applicability of my concerns on the subject issue to the TMI-1 in terms of this generic concern, in view of its specific similarities to TMI-2, and the specific points I have discussed earlier. I hope that this discussion is responsive to the. Board's question, and I request that you make a copy of this memorandum available to the Board. hus.~b$ b, thu(c/ Demetrios L. Basdekas Reactor Safety Engineer Plant Instrumentation, Control & Power Systems Branch, RSR, RES
Enclosure:
As stated cc: W. S. Farmer L. H. Sullivan l l l l l ~.,
Table 4-3. (Cont'd) c SHEET & FAILURE ITEM NO. INPUT NODE EFFECTS ON NSS REACTOR TRIP RDi4RK$ I-26 0% No effect if MfWBV is open. If MrWBV is closed. Possible RC pressure The H/A stations can (continued) the Loop A S.U. valve goes 801 open, causing trip te used to control the switch from 5.U. to Main for feedwater flow level af ter a trip if indication. Subsequen'.ly, the S.U. valve on necessary. Loop A will cycle between 50% and 80% open until level reaches the high level limit (IW 17.6). 1 27 Startup Feedwater Same as Loop A. Flow (Loop 8) 1-28 Temp. Compensated 1001 This failure could cause an undesired Probable on high RC Flow. Loop A reratioing of feedwater flow and very likely RC pressure. a reactor trip on RC pressure. Control after reactor trip is not changed.
- w e
c-> N 0% Icedwater flow will reratio, with SG-A going Probable on hig* on the low level limit.and the SG-B feed RC pressure flow limited only by BTU limits. For initial load of 100%, there is a net reduction in feedwater flow.and the reactor trips on high pressure. Control after reactor trip is not changed, 1-29 Teg. Compensated Same as for Loop A. i I RC Flow. Loop 8 i !l' 1 30 SG-A. Operate 252." (High) Loop A feed flow is reduced until SG-A reaches Yes Level the low level limit. The net loss of feedwater [ flow causes heatup of the primary and reactor ty trip on high pressure. Control after reactor g trip is not changed. fr-0." No ef fect, escept that $G-A loses the protection Not expected, go of having a high level limit. I a =08x 3}}