ML20082B959
| ML20082B959 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 07/09/1991 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20082B953 | List: |
| References | |
| NUDOCS 9107170204 | |
| Download: ML20082B959 (6) | |
Text
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. (8) Repaired Steam Generators in order to confirm the leak-tight integrity of-the Reactor Coolant System, including the steam generators, operation of the facility shall be in accordance with the following:
- 1. Prior to initial criticality, GPU. Nuclear Corporation shall submit to NRC the results of the steam generator hot test program and a summary of its management review.
- 2. GPU Nuclear Corporation shall confirm baseline primary-to-secondary leakage rate established during the steam generator hot test program. If leakage exceeds the baseline leakage rate by more than 0.1 gpm*,_the facility shall be shut down and leak tested. If any increased leakage above baseline is due to defects in the tube free span, the leaking tube (s) shall be removed from service. The baseline leakage shall be re-established, provided that the leakage limit of Technical Specification 3.1.6.3 is not exceeded.
- 3. GPU Nuclear Corporation shall complete its post-critical test program at each power range (0-5%, 5%-50%, 50%-100%) in conformance with the program described in Topical Report 008, Rev. 3, and shall have available the results of that test program and a summary of its management review, prior to ascension from each power range and prior to normal power operation.
- 4. Gpu Nuclear Corporation shall conduct eddy-current examinations, consistent with the extended inservice inspection pian defined in Table 3.3-1 of-NUREG-1019, either 90 calendar days af ter reaching full power, or 120 calendar days after exceeding 50%
power operation, whichever comes first. In the event of plant operation for an extended period at less than 50% power, GPU Nuclear Corporation shall provide an assessment at the end of 180 days of operation-at power levels between 5% and 50%, such assessment to contain recommendations and supporting information as to the necessity of a special eddy-current testing (ECT) shutdown before the end of the refueling cycle. (The NRC staff will evaluate that assessment and determine the time of the next eddy-current examination, consistent with the other provisions of the license conditions.) In the absence of such an assessment, a special- ECT shutdown shall-take place before an additional 30 days of operation at power above 5%.
- 1f leakage exceeds the baseline _ leakage rate-by more than 0.2 gpm during the remainder of the Cycle 8 operation, the facility shall be shutdown and leak tested. Operation at leakage rates of up to 0.2 gpm above the baseline leakage rate shall be acceptable during the remainder of Cycle 8 operation. Af ter the 9R refueling outage, the leakage limit and accompanying shutdown requirements revert to 0.1 gpm above the baseline leakage rate.
Amendment No. 103 Nf[ h P
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., C311-91-2066 Page 1-
- 1. Technical Specification Change Request No. 210 GPU 14uclear requests that the following revision be made to the Facility Operating License:
Replace page 6,
- 11. Reason for Change Based on current performance, it is anticipated that the OTSG primary-to-secondary leakage rate could exceed the previously established baseline leakage rate by more than 0.1 gpm (i.e., the license limit) prior to the planned shut'iown from Cycle 8 operation, which is currently scheduled for September 27, 1991. Should this occur, the current license condition requires the facility to be shutdown and leak tested.
This TSCR requests a temporary change to the license limit from 0.1 gpm to 0.2 gpm in e.ncess of the baseline leakage rate through the end of the current plant operating cycle. This would avoid the need for_ a major plant shutdown in addition to the Cycle 9 refueling outage shutdown during a relatively short time frame. The basis for this temporary change request is presented in Section III and clearly establishes that the plant can be operated safely with the' revised primary to secondary OTSG 1eakage limit.
The specific s.hange to the facility operating license is as follows:
page 6 has been revised to reflect the temporary change in Section 2(c)(8)(2). A note has been added at the bottom of the page reflecting the new tcmporary limit (0.2 gpm above baseline) to be in effect during the period applicable to this proposed change. After the Cycle 9 refueling outage the leakage limit will revert to 0.1 gpm above baseline, which is the, limit in effect prior to the proposed implementation of this temporary change request.
111. Safety Evaluation The current facility operating license SectionL2(c)(8)(2) states that if the primary-to-secondary (P/S) leakage rate exceeds the baseline leakage rate by more than 0.1 gpm, the facility shall be shutdown and leak tested. If any increased leakage above baseline is due to defects in the tube free span, the leaking tube (s) shall_be removed from service. The baseline leakage shall be re-established, provided that the Technical Specification (TS) 3.1.6.3 is not exceeded.
TS 3.1,6.3 states that if the P/S leakage through the steam generator tubes exceeds one (1) gpm total '.? both steam _ generators, the reactor shall be' placed in cold shutdowr utnin.3F hours of detection.
This request for an operating licen arefMme: dces not change t'ae one (1)_gpm limit imposed by TS 3.1.F..t
.. C311-91-2066 Page 2 Tempcrarily increasing the facility operating license leak rate limit from 0.1 gpm to 0.2 gpm above baseline during the time period remaining in Cycle 8 operation will not increase the risk of an OTSG tube rupture nor impact the safe operation of TMI-1 during this time period.
The P/S leak rate has increased steadily from 0.007 gpm at the beginning of the current operating cycle in March 1990, to a level approaching the limit today. The specific source (s) of the observed P/S leakage cannot be detected during plant operation; however, leakage through plugs and/or expansion joints is believed to be the cause of the current observable increased leak rate. Increased leakage from plugs and joints constitutes an increase in baseline leakage. Although it is possible that a tube may have a small leak, it is unlikely. Based on TMI-1 accumulated cycles ,
(flow induced vibration) any leaking tube subject to failure would have been expected to fail. Any thru-wall crack in a tube is expected to be stable during normal operation, as discussed below.
The TMI-1 steam generators have approximately 3,300 plugs installed, many of which are leak-limiting mechanical (rolled or ribbed) plugs. As such they are not a leak-tight design. Accumulated leakage through some of these plugs is possible, due to the P/S pressure differences during plant operation. The steam generators were also previously subjected to .
kinetic expansion to repair tube defects within the upper tube sheet. L The kinetic expansion joint is also a leak-limiting one and leakage through these joints is also possible. Should a crack propagate to a ,
tube f ailure within the tube sheet, the leakage would be limited by the crevice opening between the tube and the tube sheet. The parted tube would still be captured by the tube sheet and, as.such, it would not affect the integrity of the adjacent tubes in the steam generators.
Previous analysis prepared by GPUN and reviewed by the NRC (see NUREG 1019, Supplement 1) reported on the stability of tube crack growth -
during normal operating conditions, including plant normal heat-up and cooldown cycles. Low cycle loading from start-up and cooldown is the '
only significant contributor to stable crack growth. However, during ;
Cycle 8 operation the tubes have not undergone cooldown or startup since March 1990. Thus, should a small P/S leakage result from a small crack in an unplugged tube, this small crack, if growing at all, would be growing very slowly in a relatively stable fashion. It would not propagate during a short period of time to a tube rupture failure at normal operating load or cooldown load conditions. Therefore, a crack associated with a primary-to-secondary leakrate of less than or equal to 0.2 gpm would not be expected to rapidly propagate under normal operating conditions or during cooldown low-cycle loading conditions.
Peripheral or lane and wedge tubes, which are subject to very high cross ,
flows and fluid velocities, are vulnerable to high cycle fatigue which could cause tube failure if a defect-were present. During March 1990, TM1-1 experienced its first tube f ailure during normal operation shortly l af ter restart from the Cycle 8 refueling outage. 'During this experience there was no precursory indication or warning from P/S leakage of a potential tube failure. An evaluation of this incident showed that if a lane tube has a small defect it will propagate rapidly to a gross failure
t I311-91-2066 .
Page 3 (NRC SER related to Amendment No. 153). This high cycle fatigue failure !
typically occurs shortly after plants return to powee following an !
outage. j The current leak rate trend is stable. As such, the trend is not evidence of a rapidly propagating thru-wall crack. Any tube with a !
thru-wall crack subject to rapid f ailure would have aircady ruptured due l to high cycle fatigue during the current fuel cycle B operation since l that tube would have already experienced billions of cycles. Therefore j the proposed leakage limit of 0.2 gpm does not challenge the integrity of :
the OTSG tubes, and the increased leakage is most likely caused by a f breach in a barrier or component such as a plug. !
The new temporary 0.2 gpm leak rate is well within and is still more !
restrictive thrn the Standard Technical Specification (STS) value of ;
0.347 gpm limit leakage from each individual steam generator ;
(NUREG-0103). Although not specifically approved for TMI-1, the STS is ;
industry wide guidance and has been found acceptable for use at other B&W niants by the NRC. The limit was intended to ensure that leaking steam >
generator tubes would not burst under the spectrum of normal operating !
and upset conditions. !
The radiological effects have been evaluated in the TMI-1 FSAR Section i 14.1.7.10 for a steam generator tube rupture. The analysis has been ,
wrformed for the complete severance of a steam generator tube. Some of l t.e release of radioactive noble gases and iodine would be released to i th atmosphere through the condenser off-gas system. The previous !
an lysis assumed that a double-ended rupture of one steam generator tube i oc urr, with unrertricted discharge from each end. l, The TM1-1 FSAR MSLB radiological consequences are based on a 1 gpm tube leak, and 1% failed fuel, which is extremely conservative considering .
past cycle fuel performance. A MSLB accident with a 0.2 gpm tube leak is :
bounded by the previous analysis. Since the proposed 0.2 gpm initial leak rate is only a small fraction of that previously analyzed in the TM1-1 FSAR, the environmental dose contribution from the initial leakage of 0.2 gpm would be a very small fraction of 10 CFR 100 Timits even
^
during design basis accidents.
This change in P/S leakage limit will not affect nuclear safety or safe )
plant operations. The P/S leakage will be monitored and, if it exceeds :
0.2 gpm, the reactor will be shutdown. The existing relatively ' table !
rate of increase in- P/S leakage is most likely from multiple sources. l Even if the leakage is from a single tube, this small crack will be '
stable and is not expected tn grow to a tube failure during.the remainder '
of cycle 8 operation. Therefore, the integrity of the tube is not -
affected by this change from 0.1 to 0.2 gpm leak rate above the baseline. !
Although there will-be a slight change in the secondary water chemistry i due to the P/S. leakage-increase and leakage of boric acid into the secondary side may affect the pH, the pH will be maintained to within'the secondary chemistry specification by increasing morpholine and ammonia concentration. With the pH maintained within the specification, there '
are no unexpected erosion / corrosion concerns in the secondary system. ,
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. t I111-91-2000 !
Page 4 1
- In conclusion, temporarily changing the operating license condition leck ;
rate limit will not impact plant safety. The increased leak rate !
ooserved is most likely a change in the baseline rate and not indicative of a tube failure. This change does not impact the capability to .
withstand a MSLB accident. [
The proposed 0.2 gpm above baseline limit is more restrictive than the STS limit of 0.347 gpm which is intended to ensure that the tube will not !
burst under normal and upset conditions. I i
The TMl-1 Technical Specification (3.1.6.3) limit of I gpm total P/S f leakage which bounds the proposed administrative limit of 0.2 gpm above
- bascline, was intended to ensure that the off-site doses from a MSLB .
accident would be a small fraction of 10 CFR Part 100 limits. This temporary change will not exceed the TMl _1 Technical Specification limit i or 10 CFR Part 100 limits. '
Operation uf the plant during the short time period covered by this TSCR with an increased leak rate up to 0.2 gpm above baseline does not present i any significant increase in off-site radiological consequences and would f' remain a small fraction of 10 CFR Part 20 limits.
IV. No Sionificant Hazards Consideration GPU Nuclear has determined that this License Amendment request poses no significant hazards as defined in 10 CFR 50.92 in that operation of 1MI-1 in accordance with the proposed amendment will not:
- 1. Involve a significant increare in the probability of occurrence or consequences of an accident previously evaluated.
Changing the leak rate limit above baseline from 0.1 gpm to 0.2 gpm does not challange 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-1 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 evaluated.
- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated, t 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.
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.C311-91-2066 page 5 I
- 3. Involve a significant reduction in a margin of safety.
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 re-establish a new baseline under these circumstances. However, under no circumstance may the leakage limit of 1 ppm (TS Section 3.1.6.3) be exceeded for both steam generators.
This TSCR does not change the 1 gpm limitation. Also, the observed I leakage is believed to be from multiple sources and not indicative l of rapid tube failure, based on the current leak rate trend. i Therefore, there is no significa't reduction in the margin of l safety. I V. Implementation l This is a temporary change request covering the period to the end of the--
current operating cycle, which is currently scheduled for !
September 27, 1991.
It is requested that the amendment authorizing this-TSCR bc issued on an i expedited basis, and be effective upon issuance so that operation without interruption can be continued for the remainder of the current operating cycle.
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