ML19270H172
| ML19270H172 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/24/1979 |
| From: | Mattson R Office of Nuclear Reactor Regulation |
| To: | Myers H HOUSE OF REP., INTERIOR & INSULAR AFFAIRS |
| Shared Package | |
| ML19259C524 | List: |
| References | |
| TASK-TF, TASK-TMR NUDOCS 7906230131 | |
| Download: ML19270H172 (9) | |
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Dr. Henry R. Myers, Special Consultant Subcommittee on Energy and the Environment Comittee on Interior & Insular Affairs United States House of Representatives Washington, D. C.
Dear Dr. Myers:
Pursuant to our conversation earlier today involving my testimony at Mr. Udall's hearing on Monday, May 21, I have summarized and enclosed'the information I have available on when the 1978 Michelson report became available to the staff and the technical significance of the concerns it raised. Please note that there is an ongoing NRC investigation of this matter. Please contact either the Comission or me if you desire further information prior to the completion of the investigation.
Sincerely,
.%%m Roger J. Mattson, Director Division of Systems Safety Office of Nuclear Reactor Regulation
Enclosures:
- 1) Summary of NRR information 5/23/79
- 2) Eisenhut 4/17/79. transmittal of Michelson rpt
- 3) Draft of Michelson rpt - B&W reactors
- 4) Draft of Michelson rpt - CE reactors
- 5) Figures associated with draft Michelson rpts
- 6) Draft material supplied by Ebersole to staff (undtd)
- 7) Reactor Systems Branch Review Reminder of 1/10/78
- 8) Letter from Roy to Mattson 4/26/79
- 9) Appendix 5 of Volume II of 5/7/79 letter from Taylor to Mattson c
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STATUS OF KNOWLEDGE IN THE OFFICE OF NUCLEAR REACTOR REGULATION AS OF MAY 23, 1979 CONCERNING THE JANUARY 1978 REPORT BY C. MICHELSON ENTITLED " DECAY HEAT REMOVAL DURING A VERY SMALL BREAK LOCA FOR A B&W 205 - FUEL ASSEMBLY PWR" As far as we can ascertain at this time, the NRC staff first received a copy of the January 1978 Michelson report early in April, 1979 (enclosures 2,3,4,5).
The report was not formally transmitted from TVA but informally provided by Dr. Michelson upon request of the staff. More recently, it has been learned that a copy of the Michelson report was formally transmitted to B&W by TVA in April 1978, and copies apparently were available to the ACRS. This action is presently being investigated by the NRC's Office of Inspecter and Auditor.
It is clear however, that the January 1978 report had not received formal NRC staff review prior to the TMI-2 accident.
We have also ascertained within the past few days that two handwritten documents which were apparently drafts of the material which eventually became the January 1978 Michelson report were informally provided to a member of the NRC staff in late 1977 or early 1978 by Jesse Eberscle.
Mr. Michelson's supervisor at TVA and a member of the ACRS (enclosure 6).
The staff member recalls discussing the general areas of natural circula-tion and the effects of noncondensibla gases in about that time frame with Mr. Ebersole. He does not recall responding formally to the handwritten material provided by Mr. Ebersole.
In January 1978, that same staff member originated a Reactor Systems Branch review reminder (enclosure 7) which in part treats the concerns raised in Mr. Michalson's report of January 1978.
From our review of the letters between TVA and B&W now available to us (enclosure 8), the Michelson report apparently was not considered by TVA ENCLOSURE 1
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to identify any specific safety problems, but rather to identify a number of concerns regarding core c6 cling.during very small break LOCAs.
Exchange of technical information, including concerns such as in the Michelson report, is frequently carried out between the vendors and the customers without NRC involvement.
If the concerns identified in the January 1978 Michelson report were subsequently determined by B&W or TVA to involve defects which could create a substantial hazard, then they should have reported them to NRC.
Since TVA apparently did not know initially if any safety problem existed, and B&W, in a letter to TVA on January 27, 1979, subsequently asserted that none existed *, neither organization apparently believed it was necessary to report these findings to NRC since such reporting did not occur.
However, NRC's Office of Inspector and Auditor is conducting an independent investigation in order to determine if failure to report the Michelson conclusions to NRC by either organization constituted a violation of the requirements of 10 CFR 21 (Reporting of Defects and Noncompliance).
The January 1978 report by C. Michelson documented concerns regarding the ability of the core to remain covered for breaks in the B&W primary coolant system smaller tnan those breaks nomally analyzed for licensing applications.
The concerns primarily focused on the lack of documented information which confimed that the consequences of breaks presently considered for licensing applications conservatively bound the consequences of smaller breaks.
- In a more recent submittal; letter, J. H. Taylor to R. J. Mattson, dated May 7,1979, transmitting " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" (Volumes I and II),
B&W confimed the conclusions of their January 27, 1979. letter with more detailed evaluations and analyses and updated the B&W evaluation of the Michelson report (Volume II, Appendix 5; enclosure 9).
Y-3 The basis for Michelson's concerns were hand-calculated steady-state mass and energy balances.
It did not adount for detailed effects (e.g.,
transient terms and geometry) usually modelled in the more sophisticated computer codes used for licensing.
The four most significant items of the report which had a direct bearing on the eventr, at TMI-2 as well as the behavior of B&W plants to small breaks are as follow::
1.) Very small break plant response differs significantly from plant response to small breaks described in Safety Analysis Reports.
2.) Natural circulation plays an important role in core cooling following very small breaks and could be interrupted because of steam formation in the hot leg piping.
3.) Pressurizer level indication is not a correct indication of system water inventory, and 4.) Small break isolation by operator action causing system repressuri-zation with subseugent relief and/or safety valve failure.
The report brought attention to the fact that very small breaks in the primary coolant system behave differently than small breaks previously analyzed and therefore provided an indication that different emergency procedures might be needed for very small breaks. The January response by B&W to the Michelson report did not address this question and no changes were made in the emergency operating procedures. The May 1979 submittal confimed the behavior of the plants to very small breaks as described by Michelson and provided guidelines for the preparation of emergency proce-dures in the event of very small breaks. These guidelines are presently being adapted as emergency procedures for the various operating B&W plants.
4 The report also brough attention to the importance of natural circulation for very small breaks. From this,_ coupled with the thermal-hydraulic behavior of the Three Mile Island plant, it was learned that previous modeling representations were not sufficient, and that additional nodaliza-tion in the pressurizer and steam generator models was needed to more accurately represent the expected system behavior.
As a result of these model changes, analyses have confirmed Michelson's prediction that natural circulation could be interrupted. However, these analyses also showed that core uncovery does not occur for any of these very small breaks.
In the TMI-2 accident, the pressurizer level indicated the nressurizer was full of liquid.,The operators mistakenly interpreted that to mean the system was full of water, and shut off the high pressure ECC injection. The indication of a full pressurizer while other par ts of the primary system may be voided could also occur for small breaks analyzed for licensing. As stated in the May 1979 submittal, additional operating procedures will be given to all plant operators such that system pressure will be a main measurement system for inventory determination.
In addition, hot and cold coolant loop temperatures would also be used.
In the TMI-2 accident the power-operated relief valve on the pressurizer failed open during overpressure. Subsequen.t isolation of this failed valve with an upstream block valve resulted in break isolation.
While the events in the TMI-2 accident did not follow the sequence postulated by Michelson, both valve failure and " break" isolation did occur.
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5 The isolation of a break
- is not specifically considered in safety analyses.
B&W stated that the isolation of a,small break and subsequent repressuri-zation does not produce a less safe condition than not isolating a break.
This is because any repressurization that results in relief and/or safety valve opening or failure is bounded by small break analyses sizes slightly larger than the valve opening size. The staff agrees in principle with this explanation. Howeven', we will require all applicants and licensees to analyze very small breaks which exhibit repressurization with subse-quent pressurizer valve failure a's part of their evaluation of plant response to small breaks.
The significance of the Michelson report findings to conclusions regarding the acceptability of consequences due to small breaks in B&W plants will be addressed in detail in a staff report to be issued shortly. This is one of the considerations requiring resolutfor before restart of the presently shutdown B&W plants.
The key conclusions of the staff evaluation of the January 1978 Michelson report and the B&W response to the report are as follows:
1.) The overall behavior of the plants to small breaks was shown to be, for the most part, consistent with the behavior as predicted by Michelson and, within the expected accuracy the B&W analyses sub-stantiate Michelson's hand calculation results; 2.) This behavior did not result in unacceptable consequences and the core 3 not calculated to uncover for the small break accident scenarios postulated by TVA (Michelson);
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6 3.) Applicants and licensees will be required to include, as part of their ongoing re-evaluation of plant response to small breaks a) analyses of breaks which exhibit repressurization with subsequent pressurizer valve failure, and b) documentation of analyses and data which support the conclusion that steam condensation-induced structural loadings are bounded by the large break LOCA structural loadings.
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