ML20076E069
| ML20076E069 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 08/09/1991 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076E067 | List: |
| References | |
| NUDOCS 9108150061 | |
| Download: ML20076E069 (5) | |
Text
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! j 3,s ( I UNITED STATES ni NUCLEAR REGULATORY COMMISSION n
'oM I [f' WASHINGTON. D.C. W4 s *~.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.16 3 TO FACILITY OPERATING LICENSE NO. DPR-50 METROPOLITAN EDISON COMPAhT JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTIIIC COMPANY GPU NUCLEAR CORPORATION THREE MILE ISLAND NUCLEAR STATION, UNIT No. 1 DOCKET NO. 50-289
1.0 INTRODUCTION
By 'etter dated July 9, 1991, as supplemented August 6, 1991, the GPU Nuclear CorJoration (the licensee) submitted a request for changes to the Three Mile Island Nuclear Station, Unit No. 1 (TMI-1) Facility Operating License.
The requested changes would raise the limit on primary-to-secondary (P/S) leakage allowed in the Once Through Steam Generators (OTSGs) from 0.1 gallons per minute (gpm) to 0.2 gpm for the remainder of fuel cycle 8.
Refueling Outage 8R is scheduled to start on September 27, 1991.
Specifically, the licensee's July 9, 1991, letter proposed a temporary change to Lictase Condition 2.e.(8)2, adding a footnote to allow leakage of up to 0.2 gpm for the remainder of the cycle.
The license condition specifies that, if the P/S leakage rate exceeds the baseline leakage rate established during the OTSG hot test program by more than 0.1 gpm, the facility shall be shut down and leak tested.
It further specifies that, if any increased leakage above the baseline leakage rate is due to defects in the fre'espan of the-tube (s), the leaking tube (s) shall be removed from service. During the current operating cycle at TMI-1, which started in early Marsh 1990, the OTSG P/S leakage rate has risen steadily from about 0.007 gpm to about 0.08 gpm above baseline as of the end of July 1991.
If the observed rate of change of P/S leakage continues (approximately 0.01 gpm every 60 days),
the licensee estimates that the rate may exceed 0.1 gpm above the baseline before the planned start of the 8R Refueling Outage. The reasons why the licensee considers it acceptable to operate above the present limit for a limited period of time ( a maximum of 7 weeks), and the NRC staff's evaluation, are presented below. The August 6,1991, letter provided clarifying information that did not change the initial proposed no significant hazards consideration determination.
2.0 EVALUATION The 0.1 gpm'(above baseline) leakage limit corresponds to the expected leakage rate during normal operation through the largest through-wall circumferential tube crack that can be tolerated without rendering the leaking tube vulnerable to rupture in the event of a postulated Main Steam Line Break (MSLB).
This limit was imposed on TMI-1 specifically because of the occurrence of 9108150061 910809 PDR ADOCK 050002fl9 p
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f 2-circumferential tube cracks initiating from the primary side (i.e., the inner diameter of the tubes) in the early 1980's.
The 0.1 gpm limit at TMI-1 is more conservative than the corresponding limit in the Babcock & Wilcox (B&W) Plant Standard Technical Specifications (NUREG-0103).
The P/S leakage limits specified in NUREG-0103 are 500 gallons per day (gpd) per steam generator (or approximately 0.35 gpm) and 1.0 gpm total leakage from all steam generators.
The 500 gpd limit is based on the expected leakage from largest axial tube crack that would not be expected to fail during an MSLB.
The 1.0 gpm limit also applies to TMI-1 and is to assure that P/S leakage during a design-basis MSLB will not result in offsite radiation doses exceeding the limits in 10 CFR Part 100.
During the past several' years there have been many inservice inspections of OTSG tubes at TMI-1.
These inspections utilize the eddy current testing (ECT) technique.
The licensee's and the NRC staff's review of the inservice inspection results indicates that the primary-side circumferential cracks have not experienced significant growth in recent years.
For this reason, the NRC staff considers it unlikely that these primary-side cracks are contributing to the slowly increasing P/S leakage rate now being observed at TMI-1.
Peripheral tubes in the " lane" and " wedge" areas of OTSGs manufactured by B&W, including the OTSGs at TMI-1, are subject to high cross-flow velocities, which makes those tubes more vulnerable to high-cycle fatigue failure.
TMI-l experienced such a failure at the very beginning of the current fuel cycle on l
March 6, 1990. The tube that failed was in the lane / wedge area and was judged to fail as the result of high-cycle fatigue.
In that event, P/S leakage was observed to dramatically increase from barely detectable to approximately 1.4 gpm in a matter of minutes. Through-wall fatigue cracks are known to propagate to failure over extremely short time periods, typically ranging f rom about l~ hour to 3 days.
The slow evolution of P/S leakage rate at TMI-1 indicates that the leakage is not related to fatigue failure.
The licensee
. believes that leakage through mechanical plugs and/or expansion joints is the likely cause of.the observed leakage trend.
The TMI-1 OTSGs have approximately 3,300 tube plugs installed, some of which are fabricated from Inconel 600.
Industry experience has shown that Inconel plugs of the type installed at TMI-l are susceptible to primary water stress corrosion cracking.
Should a leaking plug be the source of the presently observed leakage, the NRC staff 4
noted that the tube would fill with water, precluding a high energy plug top release-event such as occurred at North Anna, Unit 1 (See NRC Bulletin 89-01).
The NRC staff also notes that any tube. leak associated with cracks in the kinetic expansion joints does not pose a tube integrity concern since the constraint provided by the tube sheet precludes any potential for gross rupture of the tube.
Although the exact source of the present leakage is not known, past inspection results at-TMI-l and the slow evolution of P/S leakage to date indicate that the current leakage does not reflect a potential vulnerability to tube ruptures during normal operation or postulated accidents.
The proposed temporary limit of 0.2 gpm for the remainder-of the present operating cycle is still a more
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. conservative limit than has been imposed at most other PWRs in the Uaited States and will continue to provide reasonable assurance that the plant will be required to _ shut _down before the source of the leakage may potentially impair OTSG tube int e g ri t y.-
The NRC' staff's conclusion is based upon the behavior of the P/S leakage over
-the past year or so as indicated by data provided by the licensee (i.e., a slow and somewhat predictable rate of change). Any significant change in the leak rate would indicate that perhaps new mechanisms not previously considered have
-developed and would be of concern to the staff.
In response to the staff's concerns, the licensee, by letter dated August 6, 1991, committed to augment the normal OTSG grab samples, which are already being taken more of ten than required by plant procedures, with additional vigilance in monitoring the main condenser offgas radioactivity detector (RMAS). This instrument is-a continuous monitor with alarms and would give the plant operators almost instantaneous indication of a tube leak or a dramatic change in the P/S leakage rate.. The licensee's August 6 letter also committed to shutting the plant down even before reaching the temporary limit of 0.2 gpm if the operators project that the sudden change in the. RMAS reading may lead to an increas e in the P/S leakage rate that may exceed the 0.2 gpm limit in a short time period.
In a telephone conversation on August 9, 1991, the staff discussed with the licensee additional details regarding how the RMAS readings would be used by plant operators to establish leak rate trends and project that the temporary 0.2 limit may be reached before the end-of the fuel cycle.
Recent experience exists for TMI-l as a result of the tube leak that occurred in the "A" OTSG on March 6, 1990.
During the propagation of that leak, RMAS readings increased from_5,000 counts / minute (epm) to 50,000 cpm in about 30 minutes._ The leak rate was later determined to be about 1.2 gpm.
During the August 9 conversation, the licensee agreed to provide a letter to the staff advising them of the' administrative procedures or guidelines that have been established to assist plant operators in translating RMAS trends-into significant changes in P/S leakage rate that would indicate _that a reactor shutdown is appropriate.
This guidance should utilize data obtained during_the March 1990 event. The licensee _ agreed to provide this information to the staff within two weeks of-issuance of this amendment.
Based on the above staff evaluation and on-the licensee's commitment in its August 6 letter, the staff concludes that the licensee's proposed change to License Condition 2.c.(8)2 in the July 9, 1991, letter will continue to provide l
adequate-assurance of,.blic health and safety and is, therefore, acceptable.
In 3.0 ' EXIGENT CIRCUMSTANCES-l
.The licensee's July 9, 1991, request stated that expeditious approval of the L
proposed change was required to avoid unnecessary shutdown of the plant.
Based on the trend of leakage, the'value of P/S leakage was projected to reach the present limit by July 31, 1991. The staff considers the request to be valid L
and-that shutting the plant down if the 0.1 gpm limit is exceeded would not-necessarily be appropriate.
The staff also considers that the licensee took reasonably timely action in making this request and that exigent actions on
. the part of the NRC staff are appropriate and pursuant to 10 CFR 50.91(a')(6).
The 0.1 gpm limit was not, in fact, reached on July 31 but will likely be reached within the near future.
4.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The Connission's regulations in 10 CFR 50.92 state that the Commission may make a final determination that the license amendment involves no significant hazards consideration if operation of the facility, in accordance with the amendment, would not:
1.
Involve a significant increase in the probability or consequences of any accident previously analyzed.
Changing the leak rate limit above baseline from 0.1 gpm to 0.2 gpm does not challenge the integrity of the OTSG tubes because of the stable nature of tube. cracks expected during normal operation if they exist.
The consequence of a tube rupture is bounded by the previous analysis in the TMI-l FSAR for a double ended tube rupture.
As a result, tube integrity is unaffected.
Thus there is no increase in the probability or consequences of an accident previously analyzed.
2.
Create the possibility of a new or different kind of accident from any accident previously analyzed.
OTSG tube rupture and an MSLB accident are the only accidents requiring consideration by this change.
Increasing the leak rate limit from 0.1 gpm to 0.2 gpm will not affect the structural integrity of the tubes. No other tube failure mechanisms are created by this change. This change revises an administrative restriction on plant operation and does not affect any safety system. Therefore, the possibility of a new or different kind of accident from any accident previously evaluated is not created.
3.
Involve a significant reduction in a margin of scfety.
The plant license precludes operation in excess of 0.1 gpm above the baseline leak rate.
If after shutting down the source of-leakage cannot be located, it is permitted to reestablish a new baseline under these circumstances.
However, under no circumstance may the leakage limit of 1 gpm (TS Section 3.1,6.3) be exceeded for both steam generators.
This request does not-change'the 1 gpm limitation.
Also, the observed leakage is believed to be from multiple sources and not indicative of rapid tube failure, based on t.he current leak rate trend. Therefore, there is no significant reduction in the margin of safety.
Accordingly, the NRC staff concludes that the proposed amendment involves n:
L significant hazards consideration, i
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5.0 STATE CONSULTATION
s In accordance with the Commission's regulations, the Pennsylvania State official-was_ notified of the proposed issuance-of the amendment. The State official had no comments.
6.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10-CFR Part 20. The NRC staff-has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents ti,t may be released'offsite, and that_there is no significant increase in individual or cumulative occupational radiation
. exposure.- The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no'public. comment on such' finding-(56 FR 33311). Accordingly, the amendment meets the eligibility criteria for categorica1' exclusion set forth in 10 CFR 51.22(c)(9)._ Pursuant to 10 CFR.51.22(b) no environmental impact st.atement or environmental assessment need be prepared in connection wit,h the issuance of the amendment.
7.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above, that:
(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
Principal Contributor: Emmett Murphy Date: August 9, 1991 s