ML15245A123
| ML15245A123 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco  |
| Issue date: | 08/21/1979 |
| From: | Ross D NRC - TMI-2 BULLETINS & ORDERS TASK FORCE |
| To: | DUKE POWER CO., FLORIDA POWER CORP., SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| References | |
| NUDOCS 7909270496 | |
| Download: ML15245A123 (6) | |
Text
ocket File NRC PDR M. Fairtile Local PDR oss ORB Reading T. ovlek NRR Reading J: etemes V Stello S.
Isrel AUG 2 979 D. E'isenhut I Matthews Z. Rosztoczy R.
Vo1her P. Norian T. J. Cyrter B. Sheron W. Russell W. Jensen Docket Nos: 50-269, 50-270 50-287 G,
G0, Vlssing 50-302,U -3 2, 50-313 D. Dilanni 50-346 C. Nelson D. Garner TO ALL BABCOCK & WILCOX OPERATING PLANTS (EXCEPT THREE MILE ISLAND, UNITS 1 & 2)
SUBJECT:
IDENTIFICATION AND RESOLUTION OF LONG-TERM GENERIC ISSUES RELATED TO THE COMMISSION ORDERS OF MAY 1979 The purpose of this letter is twofold. First, I would like to discuss some of my concerns regarding the present working relationship between the NRC, the five Babcock & Wilcox (RB&W) utilities under the long-term provisions of the Commission Orders of May 1979, the B&W Company, and the B&W Owners' Group. Secondly, I want to identify to you what I summarize are the long-term generic issues related to the Commission Orders which must be resolved.
The eight letters listed in Enclosure 2 illustrate the manner in which the various parties (licensees, B&W, and NRC) communicated in the first four months following the Three Mile Island, Unit 2 (TMI-2) accident. For generic matters, we interacted directly with B&W. In general, the time response to our requests was satisfactory.
Now, however, when we interface directly with the B&W Owners' Group (TMI-2 Follow up Subcommittee), there is considerable uncertainty, in my mind, concerning the ability of the.Owners' Group to respond promptly to generic matters. For example, I sent a letter to B&W (Reference 5-) on July 12 requesting continued generic analyses on reactor vessel brittle fracture conditions. Subsequently, I was advised by Mr. J. H. Taylor (B&W), by letter dated August 3 (Reference 8), that such matters should be referred to the Owners' Group and not to B&W. However, the Owners' Group indicated in meetings with us on July 19 and August 9, 1979, that requests for work should be made to individual licensees, as the Owners' Group has no authority to make commitments involving utilities' resources. At our August 9 meeting with the Owners' Group, I was informed that no action has taken place on my July 12 letter. This has led me to realize that the present organizational interactions are not resulting in prompt decisions and that some remedial measures are necessary.
My review of the charter of the Owners' Group, which is Enclosure 3 to the meeting summary of our meeting with the Owners' Group on July 19 (Reference 10), confirmed that the Owners' Group is indeed not empowered to make commitments for individual member utilities. This appears to be the crux of the problem. The purpose of having generic discussions with the Owners' Group is to expedite resolution of C~e
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common issues. At the present time, I am concerned with the schedule and implementation of the long-term requirements of the Commission Orders. As you know, the Orders do not require that we have generic resolution on these matters, although it is certainly efficient for all parties to do so. These Orders do require, however, that the long-term actions be accomplished "...as promptly as practicable." Mr. Harold R. Denton advised the Commission, during a briefing on the Oconee restart, that the "long-term" was in the time frame of three to six months. Therefore, what is necessary is that future generic discussions result in agreed-upon resolution of issues which can be promptly implemented. This implies to me that representatives of each utility partic ipating in such discussions must be authorized to reach agreements and make commitments for each utility. I am not sure of the future role of the Owners' Group under these conditions. I hope each licensee will consider the points I have raised concerning this matter, clarify the role of the Owners' Group, and advise the NRC of the actions which will.be taken to rectify the situation.
The letter to each licensee which lifted the Commission Order, requested that each licensee submit, within 30 days, its proposed schedule for the completion of the long-term items of the Order. In general, I find these letters lacking in specificity. It is my intention to define for you what I see as the long term issues, related to the Orders, which must be resolved. Several of the long-term issues are generic in nature to all of the B&W operating plants. provides a listing, status summary, and required action on each of the Order-related generic matters. Additional generic matters developed as part of the recommendations of the Lessons Learned Task Force (NUREG-0578) and other issues which may develop based upon our assessment of the other PWP designs, will be the subject of future correspondence. In addition, plant-specific items related to the Commission Orders will also be the subject of future correspondence. Each licensee should review Enclosure 1 and provide the requested information for items 1, 2, 4, 5, 7, and 8 by the dates indicated.
We will proceed on the basis that any subsequent information requests or staff positions will be sent to each affected licensee. If the licensees decide to respond.collectively and generically,'that will be satisfa tory.
Sincerely, D. F. Ross, Jr., Director Bulletins and Orders Task Force Office of Nucl ar Reactor Rnolation
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ENCLOSURE 1 ITEMS RELATED TO THE LONG-TERM PORTION OF COMMISSION ORDERS GENERIC TO ALL B&W OPERATING PLANTS Direct Requirements of the Commission Orders:
- 1. Failure mode and affects analysis of the integrated control system.
B&W has indicated that this report will be available for our review by August 20, 1979. By August 31, 1979, each licensee should endorse this report, or indicate the degree to which it is not applicable. Following our staff review of this report, any system or procedural changes necessary will be sent to each licensee.
- 2. Continued operator training and drilling.
Each licensee shall document the steps it has taken to insure that continued operator training and drilling incorporates the necessary lessons learned from TMI-2 and assures a continuing high state of preparedness. This shall be submitted to the NRC by September 21, 1979. Pending Commission action regarding improvements in the Operator Licensing Program, this requirement may be keyed to an upgrade in the initial training and requalification program by licensees.
- 3. Upgrade of the anticipatory reactor trip to safety-grade.
Each licensee has submitted a preliminary design for implementing a safety grade reactor trip upon loss of main feedwater and/or turbine trip. The staff is evaluating these proposals at the present time. Staff comments will be issued to each licensee by August 31, 1979. In light of the recent failure of the control-grade trip at ANO-1, accelerated installation schedules should be developed.
- 4. Auxiliary/emergency feedwater system reliability upgrade.
The long-term provisions of the Orders vary on this requirement. We believe that the most efficient way to fully define the needed improvements is to perform the AFW/EFW system reliability study discussed in our July 19 and August 9, 1979 meetings with the Owners' Group. By August 17, 1979, we expect a letter from B&W outlining in detail the scope of the study and the schedule for completing the study. By August 31, 1979, each licensee will submit a letter to the NRC committing to the proposed schedule and study, or provide an alternative. The study for the lead plant (tentatively Rancho Seco) will be available for our review in draft form by September 17, 1979. The studies for the remaining plants will be available in draft form by October 22, 1979. The final report will be published by December 3, 1979.
-2 Requirements Developed During Our Staff Evaluations of Licensees' Compliance with the Commission Orders:
- 5. A detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with extended loss of all feedwater.
This issue was identified in the staff evaluations for Rancho Seco, Davis Besse 1, and Crystal River 3. However, it is also applicable to Oconee and Arkansas Nuclear One 1. Our request for additional information on this subject was sent to Mr. J. H. Taylor (B&W) from Mr. D. F. Ross (NRC) by letter dated July 12. In a letter from Taylor to Ross dated August 3, 1979, B&W stated:
"Prior to responding to your letter (dated July 12), we feel it is essential to have discussions with our utility customers. Following this discussion, we will provide you with a schedule." We desire this schedule from the B&W utilities by August 31, 1979.
Note:
It appears to us that the concern is valid for Davis-Besse, but to a lesser degree due to the significantly lower shutoff head of the HPI pumps.
- 6. PORV and safety valve lift frequency and mechanical reliability.
This item is discussed in Section 8.4.6 of NUREG-0560 and endorsed in the staff's evaluation for each plant. This requirement has been superseded in scope and schedule by recommendation 2.1.2 of NUREG-0578. Licensees will be directed by letter to take further action on this,.matter in the near future.
- 7. Small Break LOCA Analysis.
This item is discussed in Section 8.4.2 of NUREG-0560 and endorsed in the staff's evaluation for each plant. Most of this work has been completed for the B&W plants. However, additional information is still required before the staff can issue its evaluation (NUREG-0565 - "Staff Report on Generic Evaluation of Small Break Loss-of-Coolant Accident Behavior for Babcock & Wilcox Operating Plants"). Attachment A to this enclosure is a listing of the specific information needed. We plan on issuing NUREG-0565 in late September 1979. By August 31, 1979, provide a schedule for the submission of items 1 through 5 of Attachment A such that the information will be received in time to support the publication of NUREG-0565.
- 8. Analysis for Loss of Feedwater and Other Anticipated Transients.
This item is discussed in Section 8.4.1 of NUREG-0560 and endorsed in the staff's evaluation of each plant. Some of this work has been completed; however, the scope and schedule of this requirement has been superseded by recommendation 2.1.9 of NUREG-0578. In a meeting with the staff on August 9, 1979, B&W and the B&W Owners' Group presented a program by which they intend to satisfy this requirement. Subject to incorporation of the comments given by the staff at the August 9 meeting and additional comments discussed with B&W by phone (Z. Rosztoczy (NRC) and E. Kane (B&W))
on August 14, 1979, the staff expects the proposed program and schedule for completing this item will be acceptable. By August 31, 1979, each utility should provide a written program outline and schedule for completion of this item.
LATIlMENT A TO ENCLOSURE 1 LISTING OF OUTSTANDING ITEMS RELATED TO B&W SMALL BREAK ANALYSIS
- 1. Requests made at a meeting in Bethesda, April 26, 1979:
A. Provide a benchmark analysis of sequential auxiliary feedwater flow to the steam generators following a loss of main feedwater. This analysis was provided in a letter from J. Taylor (B&W) to R. Mattson (NRC) dated June 15, 1979.
However, in this analysis the TRAP-2 code with a 6 node steam generator model was utilized. All small break analyses presented to the NRC have been performed using the CRAFT-2 code with a 3 node steam generator model.
We require a benchmark analysis for sequential auxiliary feedwater flow also be performed using CRAFT-2 with a 3 node steam generator representation.
B. Provide justification of relief and safety valve flow models used in the CRAFT-2 code.
A. Provide justification that the 3 node steam generator model used in the CRAFT-2 analysis of small breaks is adequate for the prediction of steam generator/heat transfer.
B. Provide the reactor system response to a stuck open PORV for the case of a small break which causes the reactor system to pressurize to the PORV setpoint.
- 3. Regarding the presence of noncondensible gases within the reactor coolant system following a small break LOCA:
A. Provide the sources of noncondensible gases in the primary system.
B. Discuss the effect of noncondensible gases on:
(1) condensation heat transfer, (2) system pressure calculations and (3) natural circulation flow.
C. Describe any operator actions and/or emergency procedures necessary to preclude introduction of significant quantities of noncondensible gases into the primary system.
D. Describe operator actions to be taken in the event of a significant accumulation of noncondensible gases in the primary system.
- 4. Provide a CRAFT-2 simulation for the first three hours of the TMI-2 accident.
The first 20 minutes of this analysis was provided in the "Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" (May 7, 1979).
We require that the analysis be extended for a period of three' hours in order to evaluate the ability of the CRAFT-2 code to evaluate the sequential reactor coolant pump trips and the subsequent period in which natural circulation was lost in the primary system. The analysis should include at least curves for the following paramaters:-pressure, temperature, void fraction, and flow in the reactor coolant loops.
0 ATTACHMENT A to ENCLOSURE 1 page 2
- 5. Perform an evaluation of the recent Semiscale small break experiment (S-07-10B) with your small break computer program. This request was sent to D. Holt (Chairman, B&W Owners' Group Subcommittee on TMI-2 Follow-up) from D. Ross on July 16, 1979. Copies of this letter were sent to all B&W Licensees.
- 6. Pretest calculations of the Loss of Fluid Test (LOFT) small break tests shall be performed as means to verify the analyses performed in support of small break emergency procedures and in support of an eventual long term verification of compliance with Appendix K to 10 CFR Part 50. This item is discussed in recommendation number 2.1.9 of NUREG-0578.
The items 1-6 in this attachment appear to be the type of information that in the past would be generated by B&W and sent to us.
However, in consideration of the revised working relationship, we require each utility to separately be responsible for supplying the above information.