ML16245A686
| ML16245A686 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley, Davis Besse, Oconee, Surry, Crystal River, Maine Yankee, FitzPatrick  |
| Issue date: | 05/14/1979 |
| From: | Mcdermott R Office of Nuclear Reactor Regulation |
| To: | Ross D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML16245A393 | List: |
| References | |
| FOIA-79-98 NUDOCS 7906250106 | |
| Download: ML16245A686 (7) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D. C. 20555 MAY 1 4 1979 NOTE TO:
Denwood F. Ross, Deputy Director, A41 Division of Project Management
)
5 Se'e-A to%01 THRU:
Donald J. Skovholt, Assistant Director for Quality Assurance &
Operations, Division of Project Management Walter P. Haass, Chief, Quality Assurance Branch
(
c Division of Project Management FROM:
Robert J. McDermott, Quality Assurance Branch, Division of Project Management
SUBJECT:
SAFETY ASSESSMENT OF OPERATING B&W REACTORS A. Introduction This memorandum is in response to your May 8, 1979 notel to me regarding the safety assessment of B&W licensed reactors. Your note, I believe, was prompted by notes dated April 23, 24, and 25, 1979 that I forwarded to D. Eisenhut, DOR. This information was forwarded to D. Eisenhut based upon discussions I had with you regarding safety concerns for the continued operation of the licensed B&W reactors. As you know, it was at your directive that this information was channeled to D. Eisenhut, DOR.
In attempting to respond to your May 8, 1979 note, I feel it is appropriate and useful to offer some background information and my perspectives which follow:
- 1. Soon after the TMI-2 incident occurred (early April 1979) I was verbally informed that I was assigned to a task group which was chaired by Steve Varga, DPM, established for the purpose of evaluating licensee's responses to IE Bulletin 7905 (B&W licensed facilities). I participated on that task force for a period of approximately two weeks.
During this time period, I and other members of the task group completed preliminary evaluations of the licensee's responses to Bulletin 7905 and a subsequent bulletin that was issued to holders of operating licenses for B&W reactors (7905A).
It was during this period of time that, based on my individual review of the substance of the responses coupled with my personal knowledge of the technical aspects of the B&W reactors, my concern for the continued safe operation of B&W facilities began.
INote to R. McDermott from D. Ross 7906250/06
Denwood F. Ross
-2
- 2. My position relative to the continued use of nuclear power for both the short and long term, considering all alternate energy sources currently available, is that it is necessary. This position is predicated on the grounds that both the construction and operation of the facilities is conducted in a manner that provides reasonable assurance for the health and safety of the public from nuclear risks.
It was in this spirit that I utilized resources available to me to promptly identify items which I considered to be of potential concern to NRR management. My actions were also prompted by my preliminary review of the bulletin responses, and my objective was to obtain information I believed relevant to reaching a prompt decision regarding the continued safe operation of licensed B&W reactors. My activity was accomplished in the background of numerous ongoing activities within the NRC staff relating to the TMI-2 event which I believed or perceived to represent an enormous burden on top NRC management and the staff in general.
These numerous activities included special requests and inquiries from members of the press, Congress, Commissioners, the ACRS, staffing at the TMI site, support activities, etc.
- 3. My perception at the time the below listed information was being developed was that the B&W licensed reactors would be permitted to continue to operate (or restart and operate) for some extended.
period of time until management and staff resources could be made available using the existing organizational structure and available resources. Items which I considered to be important for immediate consideration included:
- a. A complete understanding of the design and operation of main and auxiliary feedwater systems for all B&W licensed reactors with the exception of TMI 1 & 2.
- b. Mechanistic ways which pressurizer code safeties or power operated relief valves could be actuated. My concern here was related to failure to reseat that could result in small breaks (steam or water side) that were below the lower bounds of the B&W generic loss of coolant accident analysis.
Information supplied in the memos from R. McDermott to D. Eisenhut dated April 23 and 24, 1979 identified several potential problem areas with auxiliary feedwater systems at B&W reactors. Of particular note and concern was the fact that most of the reactors may not have enough installed auxiliary feedwater capacities (gpm) to satisfy the assumptions used in the B&W generic analysis 2for small break loss of coolant accidents (i.e., B&W assumes 500 gpm per steam generator with auxiliary feedwater flow to each steam generator in the small break loss of coolant analysis). Additionally, my initial review of the information obtained from the licensees relating to the auxiliary feedwater systems was that in several instances for demand events 2Assuming single active failures.
Denwood F. Ross
-3 requiring auxiliary feedwater,*operator action would be required to initiate auxiliary feedwater flow. There also existed concern for the auxiliary feedwater systems at all facilities except Davis Besse 1 regarding the interconnection of the auxiliary feedwater system with the integrated control system whereby injection of auxiliary feedwater into the steam generator would be prevented by malfunctions or failures occurring within the integrated control system.
My initial review of information relating to mechanistic ways in which pressurizer code safety valves or power operated relief valves could open (i.e., system pressure reaching valve set points) disclosed that there were several plant transients initiated by malfunctions or failures in the secondary or balance of plant portion of the facility that would result in lifting of PORV or code safeties. Additionally, directives included in IE Bulletin 7905A imposed requirements for the plant operator to establish procedures to assure continued operation of the high pressure coolant injection pumps for a 20-minute period. This latter fact, coupled with the fact that there are several plant transients initiated by secondary system malfunctions that would automatically start Hpsi pumps and my perception that some operators would explicitly follow the prescription as outlined in the bulletin, heightened my concern because I was convinced that pressurizer code safety valves or PORV's would open in considerably less time than 20 minutes. An added concern is that available information obtained from the valve manufacturers for the code safeties and relief valves was that they stressed that the valves were only designed for steam service and that the effects on the valves from passing 2-phase or solid water through the valves were not known. The above stated concerns are related again to the possibility of creating small steam side or water side breaks that are smaller than those addressed in B&W generic analysis.
B. Summary of Information Provided to Date on B&W Reactors Several items of potential concern were contained in enclosures to my notes to D. Eisenhut dated April 23, 24, and 25, 1979. A summary of each item is provided below.
April 23, 1979 note to D. Eisenhut -
Subject:
Information Applicable to B&W Reactors.
The enclosure contained seven items of potential concern as listed below.
Item Summary of Possible Common Mode Failures of Auxiliary Feedwater Systems Observed in Operating PWRs.
(4/22/79)
This issue was highlighted for management attention because there have been at least seven reported events where common mode failures have been reported to the NRC. It is my personal opinion that the common mode failures of auxiliary feedwater systems
- (assuming single failure)
Denwood F. Ross
-4 that have not been reported to the Commission would be many times the reported number. This is due in part, I believe, to the current wording of the technical specifications relating to LCO's (for the auxiliary feedwater systems) and the wording relating to the reporting requirements contained in the technical specifications. It should be noted that six of the seven common mode failures that have been reported to date resulted from system interactions with systems that are normally considered to be non-safety grade.
Item 2 - Available Information of Pressurizer Safety Valves for B&W Licensed Reactors.
(4/23/79)
This information was provided to management primarily for the reason of identifying break area size that would result if a pressurizer safety valve fully opened and failed to reclose. In all plants reviewed (Crystal River, Arkansas 1, Rancho Seco, Oconee 1-3, and Davis-Besse 1) the equivalent break size area for a stuck open safety valve would be significantly less than the smallest break assumed in the B&W generic loss of coolant analysis. B&W's smallest break assumption is.05 ft2. This was and is of concern because I believe the break size at TMI-2 was also significantly less than.05 ft2.
Item 3 - Comments on Auxiliary Feedwater System Capacities.
(4/22/79)
This information was provided for management's attention to highlight that four of the five plants reviewed (Crystal River 3, Rancho Seco, Arkansas 1, and Davis-Besse)may not have the capacity (gpm) equivalent to that assumed in the B&W generic loss of coolant accident analysis. 3 Item 4 -
Summary of Available Information on Pressurizer Code Safeties.
(4/22/79)
This information was provided to communicate my findings relating to information obtained from the manufacturers of the pressurizer code safety valves and their concerns related to 2-phase flow or solid phase flow through the code safeties. Davis-Besse 1 code safeties were supplied by Crosby Company and Dresser supplied safety valves for Oconee, Rancho Seco, Crystal River and ANO-1.
Crosby representatives have stated in our communications with them that they believe.that their valves will sustain damage on mixed (2-phase) flow or solid water flow. Both valve manufacturers stressed that valves are only designed for steam service and that they believe some damage may result from 2-phase or solid water flow through the valves. This item is of potential concern if Hpsi pumps are operated for a 20-minute period when 2-phase or solid water flow may be passing through these valves.
3 If a single active failure in the auxiliary feedwater system is assumed.
Denwood F. Ross
-5 Item 5 - Time for HPI to Lift Pressurizer Code Safeties. (4/23/79)
This information was provided to indicate if the bulletin directive (7905A) was followed explicitly (i.e., if Hpsi pumps started auto matically, the operators should allow continued operation for a minimum of 20 minutes), it would be likely or probable that the pressurizer code safeties would lift in significantly less time than 20 minutes and that the possibility exists for mixed (2-phase) or solid water flow through the valves. See Item 4 above for potential safety concerns.
Items 6 and 7 - Comments on Oconee Feedwater Systems and Rancho Seco Feedwater Systems. (4/22/79)
This information was provided to identify potential areas of concern relating to the reliability of feedwater systems at Oconee and Rancho Seco. The summary provided in each of these documents highlighted several areas of potential concern relating to the reliability of these systems. This information was also provided to assure that management was aware of the assumptions relating the auxiliary feedwater flow rates that were utilized by B&W in their generic LOCA analysis.
April 24, 1979 note to D. Eisenhut -
Subject:
Additional Information Applicable to B&W Reactors. The enclosure contained three items of potential concern as listed below.
Item 1 - Comments on Arkansas Unit No. 1 Feedwater Systems.
(4/24/79)
This information was provided to identify potential areas of concern relating to the reliability of feedwater systems at Arkansas Unit No. 1. The summary provided in the document highlighted several areas of potential concern relating to the reliability of this system.
Item 2 - Comments on Davis-Besse Unit No. 1 Feedwater Systems.
(4/24/79)
This information was provided to identify potential areas of concern relating to the reliability of feedwater systems at Davis-Besse Unit No. 1. The summary provided in the document highlighted several areas of potential concern relating to the reliability of this system.
Item 3 - Comments on Crystal River Feedwater Systems.
(4/23/79)
This information was provided to identify potential areas of concern relating to the reliability of feedwater systems at Crystal River.
The summary provided in the document highlighted several areas of concern relating to the reliability of this system.
Denwood F. Ross
-6 April 24, 1979 note to D. Eisenhut -
Subject:
Additional Information Applicable to B&W Reactors. The enclosure contained one item of potential concern as identified below.
Item 1 - Calculation of Time For Makeup Pumps to Lift Pressurizer Safeties for Davis-Besse. (4/24/79)
This information was provided because of the unique aspects of the Davis-Besse 1 high pressure coolant injection system designed to mitigate small breaks in the reactor coolant system. Based on my review of theHpsi system design of this facility (i.e., low head Hpsi pumps -%J 1650 psig shutoff head), the only mechanistic way to open to code safeties would be to operate the makeup pumps (these are not Hpsi pumps but the normal makeup pumps with a high shutoff head of 2774 psig.).
April 25, 1979 note to D. Eisenhut -
Subject:
Additional Information Applicable to B&W Reactors. The enclosure contained two items of potential concern as listed below.
Item 1 - Effective Break Size Calculations for TMI-2.
(4/25/79)
This information was provided to assure that management was informed that the b st estimate break size area for the TMI-2 event was 0.00729 ft, which is below the range of break size analyzed by B&W.
Item 2 - Comparison of Davis-Besse 1 ECCS to Other B&W Licensed Plants. (4/25/79)
This information was provided to assure that management was informed of the unique characteristics of the Davis-Besse 1 ECCS design.
C. Conclusions Your note to me dated May 8, 1979 regarding the safety assessment of B&W operating reactors requested that I answer the following two questions and to identify any other areas of potential concern I had:
A. Whether my concerns are being addressed by the staff, and B. Whether the residual uncertainty is, in my opinion, too great to resume operation.
In response to Item A, I believe that the staff is reviewing all of the items of potential concern identified to date by me in my notes to D. Eisenhut dated April 23, 24, and 25, 1979. That said, however, I have no current knowledge of the status of all the conclusions reached by the staff regarding my items of potential concern.
Denwood F. Ross
-7 In regards to Item B above, my personal opinions are as follows:
Currently licensed B&W facilities should be shutdown or remain shutdown
- until,
- a. An analysis for each facility has been submitted and reviewed by the staff that shows conclusively that the core can be adequately cooled by natural circulation. Analysis should include evaluation of natural circulation with only one coolant loop in service and confirmatory testing should be conducted.
- b. An analysis for each facility has been submitted and reviewed by the staff of small break loss of coolant accidents that is consistent with the capabilities of the emergency core cooling and auxiliary feedwater systems (as-built).
- c. Each licensee has proposed and the staff has reviewed design modifications that will substantially improve the reliability and automatic availability of auxiliary feedwater systems above the existing levels. This item is not applicable to Davis-Besse 1.
- d. Current directives which have been issued by the staff have been reviewed to assess the licensees' perception of the directives, the procedural implementation of the directives, and the anticipated operator response to anticipated operational occurrences in response to these procedures.
The technical basis for the actions required by the directives, particularly, continued running of Hpsi and reactor coolant system pumps, should be provided to the owners-operators of the B&W facilities.
In summary, I believe that the above items are those that are, in my opinion, necessary and sufficient to provide reasonable assurance for the protection of the health and safety of the public from nuclear risk,for the short-term (A;4 months) until longer term corrective actions can be taken.
This conclusion is based on my current understanding of B&W power plant designs that I reviewed and my concern that the continued use of nuclear power as a national energy source may be precluded if another TMI-2 incident were to occur.
Robert J McDermott Quality Assurance Branch Division of Project Management cc:. H. Denton E. Case R. Boyd R. Mattson V. Stello D. Skovholt W. Haass