ML20076K713
| ML20076K713 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 09/14/1983 |
| From: | Weiss E HARMON & WEISS, UNION OF CONCERNED SCIENTISTS |
| To: | NRC COMMISSION (OCM) |
| References | |
| NUDOCS 8309150241 | |
| Download: ML20076K713 (9) | |
Text
.
bey,ED NET UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
' I4 R2:56 In the Matter of
)
METROPOLITAN EDISION COMPANY
)
k7f.[ff"'
)
Docket Not157-$89 -
(Three Mile Island Nuclear
)
(Restart)
Station, Unit No. 1
)
)
UNION OF CONCERNED SCIENTISTS' REPLY TO STAFF AND LICENSEE RESPONSES TO THE COMMISSION ORDER OF AUGUST S, 1983 1.
The Staf f and Licensee's View on the Scope of the Proceeding Is Inconsistent With Their Positions During the Restart Proceeding and With the Rulings of the ASLB and Is Incorrect.
All parties agree that, in order to be admitted and litigated in the restart proceeding, a contention was required to have a " nexus" with the TMI-2 accident and the lessons learned therein.
Indeed, UCS originally suggested this standard, which was adopted by the Licensing Board.
During the course of the proceeding it was never suggested that the only way in which a nexus could be catablished was by postulating a small break LOCA or loss of feedwater.
On the contrary, the UCS contentions that are directly relevant to BM-83-47, Contentions, 3,
5 and 14 were explicitly admitted by the ASLB based upon a " nexus", established largely through NRC's own
" lessons learned" documents, of a different and somewhat broader nature.
As we have fully described in our initial response to the Commission, that nexus inheres in the fact that i
8309150241 830914 PDR ADOCK 05000289 C
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. the TMI-2 accident disclosed a past failure to properly appreciate the role of equipment previously treated as unrelated to safety in the causation and mitigation of accidents.
No party claimed that this did not meet the " nexus" test.
No party objected to UCS's evidence which discussed functions of the PORV beyond those required for post-LOCA decay heat removal.
It is only now that the Staff and Licensee claim such matters to be beyond the scope of the Restart proceeding.
These are clearly the most belated post hoc rationalizations constructed for the purpose of preventing consideration of the consequences of a steam generator tube rupture at THI-l and are therefore of minimal persuasive value.
Moreover, neither the Staff nor Licensee mentions UCS Contention 14 (general safety system classification) or the Staff testimony on that issue.
As we have pointed out, need for the PORV for rapid depressurization after steam generator tube rupture would require classification of the component as safety grade within the four corners of the Staff's own testimony in this case.
UCS Response to Commission Order of August 3, 1983 at 10-11, (hereinaf ter "UCS Response").
It should also be noted that the Licensee is incorrect in asserting at page 5 of its response, that the only issues i
i raised relating to operability of the PORV were concerned with the role of the PORV in causing, aggravating and mitigating a LOCA.
UCS' testimony and proposed findings and the decisions of the boards dealt also with the pressure - control
. functions of the PORV during low temperature conditions and inadequate core cooling conditions.
UCS Response at 5-6.
In addition, both UCS and the Commonwealth of Pennsylvania attempted to demonstrate to the Licensing Board that rapid depressurization via the PORV is needed for steam generator tube ruptures--precisely the matter at issue here.
UCS
- Response at 6-8.
See also Tr. 8276, where the Commonwealth's attorney stated:
I would like to make clear that our concerns are somewhat different from the need to go to hot shutdown or cold shutdown articulated by the Board.
Our concern is with the view of the steam generator tube rupture LOCA and the reliability of systems used to reduce the pressure below the steam generator safety valve setpoint.
The the Staff and Licensee's objections, compounded by the Licensee's mischaracterization of a tube rupture as an event beyond the design basis, led the ASLB, which was, by its own admission, confused, to erroneously exclude this evidence.
This circumstance should not be used as a bootstrap to justify exclusion of this evidence by the Commission.
The Commission orders cited by the Licensee on pages 3 and l
4 of its response cannot be read as support for restricting the l
scope of this proceeding to the degree they urge.
Neither decision explicates the substance of the " nexus" requirement at all.
On the contrary, the Commission took no action during the pendency of this case to suggest that the ASLB was wrong in admitting UCS contentions 3, 5 and 14.
i
. Perhaps most remarkable is the Licensee's circular argument that, since the Staff never analyzed tube ruptures in connection with Class 9 accident issues, and the Board accepted the Staff's analysis, the tube rupture is therefore beyond the scope of the proceeding.
Licensee's Response at 4-5.
The Board's acceptance of the Staff's testimony is at most a compounding of its earlier erroneous ruling.
More likely, the Board never even considered the implications of the Staff's failure to analyze tube ruptures in the context of Class 9 accidents.
The Board's decision on that issue contains no indication that the matter was raised or considered.
Finally, the accuracy of the Licensee's unqualified assertion that TMI-2 did not involve a steam generator tube rupture is far from certain.
We have enclosed memoranda prepared by the Staff in June of 1979 indicating that there was, in fact, a primary to secondary leak in steam generator B at TMI-2 during the accident.
2.
Even If Only Small Break LOCAs Can Be Considered, a Steam Generator Tube Rupture Is a Small Break LOCA.
The Licensing Board's ruling excluding evidence on tube breaks, although far from a model of clarity, was clearly premised on the conclusion that a tube rupture is a small break LOCA.
Tr. 7986, 8277.
Now the Licensee wishes to have it both ways, seeking to have the Commission exclude this evidence on the opposite grounds: that a tube rupture is not a LOCA.
The Commission's rules define a loss-of-coolant accident as follows:
i
-S-Loss-of-coolant accidents (LOCA's) are hypothetical accidents that would result from the loss of reactor coolant, at a rate in excess of the capability of the reactor coolant makeup system, from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system.
10 CPR 50.46(C)(1).
A tube break falls squarely within that definition.
The Commission's decision in the ECCS rulemaking cited by Licensee does not hold that a tube rupture is not a LOCA.
It simply upholds, in a single uninformative sentence, exclusion of evidence specifically dealing with tube ruptures from the ECCS rulemaking proceeding.
3.
The Issues Raised Have Important Safety Significance.
The Licensee's argument on this point is almost entirely inapposite.
It totally ignores the well-established standards for reopening a record and argues instead that the Commission cannot order this evidence considered unless it could find that 4
the concerns expressed in BM-83-47 warrant the immediately effective suspension of the TMI-l operating license.
Licensee's Response at 6-8.
Such a suggestion is absurd.
The record must be reopened if the matters raised are relevant, have important significance and have the potential for affecting the outcome of this proceeding.
The fact that TMI-l I
is under a shutdown order does not change these standards, which are designed in order to ensure that the adjudicatory 1
record is accurate and coniplete.
The Commission has made its i
( view quite clear--a view heartily endorsed by the Staff and Licensee--that the status of the shutdown order is legally independent of the ongoing adjudication.
Moreover, the Commission is considering immediate effectiveness on its own, not subject to the rules governing adjudicatious.
The Licensees' position, if adopted, would establich a principle allowing an erroneous decision to be reached by the adjudicatory boards if the issue involved were not, by itself, sufficient to compel shutdown.
The Commission can surely not sanction that outcome.
The Staf f's position on the safety significance of these issues is mystifying.
Indeed, the Staff discusses the matter as if it were a bystander and not the originator of BM-83-47.
Staff counsel takes no position on the need for the PORV in TMI-1 for tube rupture; he simply parses the language of the Board Notification and urges us to find comfort in the fact that BM-83-47 "does not indicate that a safety-grade PORV is the only acceptable means" for rapid depressurization and that some Combustion Engineering plants have safety-grade spray systems to perform this function.
Staff Response at 9.
The Staff fails to suggest any other means of rapid depressurization for TMI-1.
We are apparently intended to conclude that there is no " grave" threat to public safety based on the inference that, if there were, the Staff would have already ordered corrective action.
Id. at 9-10.
No reason is suggested as to why such action has not been ordered.
In any a
. case, the Staff's failure to act immediately is hardly dispositive on the question of the safety significance of the issue.
It is the province of the boards to determine ti.e sufficiency of the Staff's action on the basis of a rational record.
UCS believes that the relevance of BM-83-47 to the issues litigated in the Restart proceeding is clear, as is the safety significance of these issues.
The record should be reopened Respectfully submitted, 9.1/a bsd Ellyn R. Weiss HARMON & WEISS 17 25 I Street, N.W.
Suite 506 Washington, D.C.
20006 (202) 833-9070 Dated: September 14, 1983
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NN 2 6 DocketNo.50=l320 1979 MEMORANDUM FOR: Mitchte Rogovin, Director, NRC/TMI Special Investigation Group FROM: Roger J. Mattson, Director, Division of Systens Safety
SUBJECT:
TMI-2 STEN 4 GENERATOR B Enclosed is a current evaluation of the cause of the primary to secondary leak that occurred in Steani Generator a at TMI-2 during the accident.
Original signed by Roi;er J. Mattson
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Roger J. Mattson, Director Division of Systens Safety
Enclosure:
As Stated cc:
R. Minogue, SD E. Case, HRR it. Denton, MRR r
D. Vassallo, NRR
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UNITED STATES
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c JUN 141979 NOTE TO: R.J.Mattson%
FROM:
R. J. Bosnak
SUBJECT:
TMI-2 STEAM GENERATOR B
Reference:
Note to R. J. Bosnak frcm R. J. Mattson of June 1,1979 on TMI-2 Steam Generators A baseline eddy current inspection of steam generator B at THI-2 during minor dents to 95% through wall penetrations.g with defects ranging from Decerrber 1977 revealed approximately 400 tube Thirty five tubes with defects greater than 40% of the wall thickness required plugging. A majority of the cddy current signals (ECT) were indicative of a dimple (ding) 1.e., reduction of inside diamater without detectable reduction in
.vall thickness. These defects presumably occurred during the fabrication process. Other ECT signals were indicative of scab type defects.
Circumferential cracks can initiate at such locations under conditions of high cycle fatigue.
It is surmized that excessive flow induced vibrations may have caused high cycle fatigue failures at other B&W plants notably Oconee 1, 2, and 3.
Concern about excessive tube vibration at TMI had been raised after the baseline inspection in De enber 1977.2 A test was designed to investigate the reduction in alternating stress by installation of tube sleeves at two locations of concern and addition of intemediate supports at two different locations at the upper most tube span.
It is possible that during the period between the Decenber 1977 baseline inspection and the accident at THI-2 in March 1979, circumferential cracks may have been initiated at locations of high flow induced vibrations.
It is estimated that a circumferential crack with a depth which had progressed to greater than seventy percent through wall would be unable to withstand the pressure and themal loads imposed during the March 1979 transients. Such a crack would then be expected to pop through the remaining wall resulting in primary to secondary leakage.
Contact:
J. R. Rajan, DSS:MEB, X27533/72 7 967/cw -
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R. J. Mattson The bending and thermal stre'sses in the tubes during pressure and thermal transients similar to those that occurred during the TMI-2 accident were evaluated by B&W recently.3 In the analysis of the postulated transient, primary system pressure builds to the
)
maximum value associated with the safety valve setpoint. Since primary flow is unavailable, a steam environment exists on the primary side of the tubes. On the secondary side, the steam generator is completely depressurized, boiled dry. A temperature differential of approximately 450cF may exist across the steam generator tube walls under such conditions.
Initiation of auxiliary feedwater flow to the steam generator at this stage results in a rapid cooling of the tubes which in turn produces significant tensile loads in the axial direction. The bending stress on ghe tube outer wall due to a temperature differential, AT, of 450 F across the tube wall can be as high as 80 ksi tension I
which will cause plastic deformation of a portion of the tube.
A circumferential crack of depth seventy percent through wall or more located in this region is likely to penetrate the tube wall and the crack opening is likely to increase resulting in a i
primary to secondary leak. During a decrease in aT across the tube wall and a consequent reduction in bending stress, the circumferential crack would tend to close up, resulting in either a decrease or complete stoppage of the primary to secondary leak, depending on the size of the crack.
Other possible sources of primary to secondary leakage were also examined. These include:
Leakage due to failure of the welds, attachments or other a.
modifications made in the TMI Unit 2 steam generators to install instrumentation for monitoring vibraticn flow and pressure data.4 i
t b.
Leakage as a result of other design modifications made in the steam generators for example: (1) Tube sleeving modifications, (2)
Lane flow blockers (3) Auxiliary feedwater nozzle modifications, and i
(4) Secondary side, lane tube stiffeners.
1 9
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JUN 141979 R. J. Mattson 3-c.
Failure of steam generator tube plugs installed after the initial baseline inspection of 1977.
d.
Leakage of the tubes due to other types of damage viz. wear, stress-corrosion cracking or erosion / pitting.
While the possibility of leakage dua to these mechanisms cannot be ruled out, the most probable cause appears to be high cycle fatigue cracking discussed earlier.
R. J. Bosnak, Chief Mechanical Engineering Branch Division of Systems Safety cc:
F. Schroeder, DSS J. Knight, DSS W. Minners, DSS J. Rajan, DSS H. Silver, DPit F. Cherny, DSS e
A
References:
1.
Letter from H. Silver to Metrcpolitan Edison Company - Summary of Meeting on Steam Generator Tube" Inspection June 30, 1978.
2.
Letter frcm Metropolitan Edison Company - Steam Generator Tube Sleeve Qualification Program, Dec. 22, 1977 to S. A. Varga, NRC.
3.
B&W Report of May 7, 1979, " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant Appendix 2 (Steam Generator Tube Thermal Stress Evaluation) 4.
B&W Report of Dec. 22, 1979 "Once Through Steam Generator Instrumentation Program for Three Mile Island #2."
Report No. 773570139 f
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of
)
)
METROPOLITAN EDISION COMPANY
)
)
Docket No. 50-289 (Three Mile Island Nuclear
)
(Restart)
Station, Unit No. 1)
)
)
CERTIFICATE OF SERVICE I hereby certify that copies of the UNION OF CONCERNED SCIENTISTS' REPLY TO STAFF AND LICENSEE RESPONSES TO THE COMMISSION ORDER OF AUGUST 5, 1983 have been served on the following by first class mail, postage prepaid this 14th day of September, 1983, or as otherwise indicated.
- Nunzio Palladino, Chairman Dr. Walter H. Jordan U.S.
Nuclear Regulatory Atomic Safety and Commission Licensing Board Panel Washington, D.C.
20555 881 West Outer Drive Oak Ridge, TN 37830
- Victor Gilinsky, Commissioner Dr. Linda W.
Little U.S. Nuclear Regulatory Atomic Safety and Commission Licensing Board Panel Washington,, D.C.
20555 5000 Hermitage Drive Raleigh, NC.
27612
- Frederick M. Bernthal, Commissioner Judge Gary L.
Milhollin U.S. Nuclear Regulatory 1815 Jefferson St.
Commission Madison, Wisconsin 53711 Washington, D.C.
20555
- Thomas Roberts, Commissioner
- *Ju dge Ga ry J. Edles, U.S.
Nuclear Regulatory Chairman Commission Atomic Safety and Washington, D.C.
20555 Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555
- James Asselstine, Commissioner U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Ivan W.
Smith, Chairman
- *Ju dge John H. Buck Atomic Safety and Licensing Atomic Safety and Board Panel Licensing Appeal Board i
U.S. Nuclear Regulatory Panel Commission U.S. Nuclear Regulatory Washington, D.C.
20555 Commission Washington,,
D.C.
20555
- Judge Christine N.
Kohl
- Counsel for NRC Staff Atomic Safety and Licensing Office of Executive Legal Appeal Board Panel Director U.S. Nuclear Regulatory U.S.
Nuclear Regulatory Commission Commission Washington, D.C.
20555 Washington, D.C.
20555
- Judge Reginald L. Gotchy
- Docketing and Service Atomic Safety and Licensing Section Appeal Board Panel Office of the Secretary U.S. Nuclear Regulatory U.S.
Nuclear Regulatory Commission Commission Washington, D.C. 20555 Washington, D.C.
20555 Mrs. Marjorie Aamodt
- * * *Geo rg e F. Trowbridge, Esq.
R.D.
95 Shaw, Pittman, Potts &
Coatsville, PA 19320 Trowbridge 1800 M Street N.W.
Washington, D.C.
20036 Douglas R.
Blazey, Esquire Public Information and Chief Counsel Resource Center Department of Enviro, Res.
1037 Maclay Street 514 Executive House Harrisburg, PA. 17103 Harrisbury, PA. 17120 Louise Bradford Three Mile Island Alert 325 Peffer Street
/
N,b 2 dea 4 ly)GI
!!arrisburg, PA 17102 Ellpn R. Weiss Jordan D.
Cunningham, Esq.
Fox, Farr & Cunningham 2320 North Second Street
- Hand delivered to 1717 H Harrisburg, PA 17110
- Street, N.W., Washington, D.C.
Dr. Judith H. Johnsrud
- Hand delivered to 4350 Dr. Chauncey Kepford East West Highway, Environmental Coalition on Bethesda, MD.
Nuclear Power 4 33 Orlando Avenue
John A.
Levin, Esq.
- Hand deliver?d to Assistant Counsel indicated address.
Pennsylvania Public Utility Commission Post Office Box 3265 Harrisburg, PA 17120 Ms. Ga il B. Phelps Michael McBride, Esq.
245 West Philadelphia Street LeBoeuf, Lamb, Leiby &
York, PA 17404 MacCrae 1333 New Hampshire Ave.
Suite 1100 Washington, D.C.
20036