ML20086H731
| ML20086H731 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 12/05/1991 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 91-693A, NUDOCS 9112090323 | |
| Download: ML20086H731 (13) | |
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. i VIHOINIA UI.UCTHIC AND POWUH COMI%NY
- Hirummn,Vinomia cauci i
December 5, 1991 U.S. Nuclear Regulatory Commission Seria! No. 91693A Attention: Document Control Desk NL&P/JBL: R4 Washinginn, D.C. 20555 _ Docket No. 50 338 .
License No.
NPF 4 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY ,
NORTH ANNA POWER STATION UNIT 1 STEAM GENERATOR OPERATING CYCLE EVALUAT.lRN
.SMM14ARY RESPONSE TO REQUEST _EQR ADDITIONAL INFORMATION By letter dated November 7,1991, the NRC requested we provide additional inf_ormation regarding our August 6,1991 submittal which supports _a full cycle of operation for the North . Anna Unit 1 steam generators for the current operating cycle.
-The questions were direc.ted toward the Westinghouse Electric Corporation report, WCAP-13034, entitled " North Anna Unit 1 Steam- Generator _ Operating Cycle Evalut lion." This WCAP report provided detailed justification for an 18 month cycle of operation for Nonh Anna Unit 1 Cycle 9. Subsequently, we notified you of our revised outage schedule and requested approval to operate until April 18,1992 (i.e.,
-approximately a 13 month operating cycle). We also met with you on December 2, 1991 to respond to each of your questions in the request for additionalinformation.
We have concluded that there are no safety issues or concerns with respect to the current cycle of operation of North Anna Unit 1 for an additional three months without a mid-cycle outage for a total operating cycle of 13 months. The basis for this conclusion is summarized below. 1) Extensive inspections were performed and conservative plugging criteria-were consistently used during the.1991 refueling / tube inspection
-- outage. 2) A conservative analysis was provided which supports a full 18-month cycle of operation. '3) Very restrictive operating limits have been administratively imposed on allowable primary to-secondary leakage limits. 4) Station _ operators are well trained in leakage detection and actions to be taken in the event of unusual primary-to-secondary leakage. 5) State-of-the an N-16 primary to secondary leakage monitors and other redundant. leak detection systems are available. 6) During the eight months of operation in the current cycle, monitoring has indicated only very low leakage rates which is consistent with the past three operating cycles for North Anna Unit 1.
There were two additional issues discussed in our December 2,1991 meeting which have since been evaluated, We had previously committed to cease power operation within two hours of determining that none of the -Technical Specification required primary-to secondary leakage detection systems are available. Because of the concerns stated during the meeting, we additionally commit to reduce power to less b1120903iY m 1203 DR ADOcK 0500 3 [\
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than 50% within 90 minutes and below Mode 1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of determining no aval!able primary-to-secondary leakage detection systems. We will enter Mode 3 in
- the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> if leakage monitoring continues to be unavailable. The other issue was with regard to our steam line break analysis, ~ We have evaluated the radiological consequences of the maximum calculated integrated post accident primary to secondary leakage rate relative to the 10 CFR Part 100 allowable off-site dose limits and the GDC 19 allowable control room dose limits The postulated 49 GPM leakage rate during a steam line break did not result in any dose consequence in excess of these limits Even though we conclude that the consequence of this special steam line break analysis is within our design basis requirements, we commit to further impose an administrative limit on the primary coolant specific iodine activity of 75% of the specified Technical Specification limits.
A more detailed _ discussion of these issues and our operating cycle basis is provided in Attachment 1. Detailed written responses to your November 7,1991 request for additional information will be provided for the priority questions by December 9,1991 and for the remaining questions by December 17.1991.
Should you have any additional questions, piease contact us.
Very truly yours, m
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W. [.. Stewart Senior Vice President - Nuclear Attachment ec: U.S. Nuclear Regulatory Commission Region il 101 Marietta Street, N.W.
Suite 2900 Atlanta, Georgia 30323 -
Mr. M. S. Lesser NRC Senior Resident inspector North Anna Power Station l
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ATTACHMENT 1 s
STEAM- GENERATOR OPERATING CYCLE EVALUATION
SUMMARY
REPORT NOr i ANNA POWER: STATION UNIT 1
' VIRGINIA ELECTRIC AND POWER- COMPANY
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- i ATTACHMENT
- 1-NORTH ANNA POWER STATION
-UNIT 1 STEAM GENERATOR OPERATING CYCLE EVALUATION Executivelummary The process for the North Anna Unit 1 operating' cycle evaluation has been extremely detailed considering:f! eld NDE results, laboratory analysis, structural mechanics,
. specific characteristics of the degradation mechanisms, statistical methods, and flow induced vibration analyses. -In order to establish the approariate operating interval for the unit,1the condition of the steam generators was est mated for the end of the ,
proposed cycl _e of operation (September 1992) based on prior inspection results and -
then the capability of theLsteam generators was determined with respect to the
- guidelines of NRC Regulatory Guide 1.121.
lThe _Westingh6use evaluation report provides a technical basis for full cycle operation of North Anna Unit 1.' It presents a strong case relative to tube burst by incorporating the uncertainties relative-to eddy. current and crack growth. The most significantry' degraded population of tubes considering each limiting degradation mechanism was analyzed. The evalua_ tion provides a probabilistic assessment with respect to vibration-einduced propagation of circumferential cracks considering both turbulence and t fluidelastic components:and their effect relative to crack angle and depth. For tube
)vibraton, the evaluation identifies the periaheral zone between the tubesheet and first tube support plate and a very limited popu ation of tubes at the top support plate as the
-.most limiting areas of the steam generator tube bundle.
North Anna Unit 1: has extremely low, administratively controlled, primary-to secondary.
-leakage rate: limits and operating constraints.- The N 16_ monitors are the best equipment available to monitor such leakage. Should a tube in the peripheral zone '
between the tubesheet and first tube support plate develop a circumferential corrosion - 4 crack of sufficient size, tho evaluation indicates that turbulence would be the predicted initial' propagating ' mechanism. In the event -that a crack were propagated by turbulence. *he evaluation.in_dicates that the increa_se in leakage and propagation rate would prov.Je sufficient respons_e time to take appropriate operator actions. For the top support plate, the small population of susceptible tubes. and the lack of any
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circumferantial . degradation at elevations above the fifth support plate in Unit 1 make it extremely unlikely that tube: failure would occur. Corrosion ' cracking, which is not influenced by flow-induced vibrationo is not expected to grow to sufficient size to cause-any concern. But should this occur, leakage would be detected with the aforementioned instrumentation.
~ By letter dated October 11,1991, we informed the NRC of a revised outage schedule i
for North _ Anna Unit 1 to minimize the effects of the long steam generator replacement outage on the overall nuclear outage plan for our four operating units. The revised schedule included substantially shortened operating cycles for both th current (Cycle i
a n
. 9) and the next (Cycle 10) operating cycles for North Anna Unit 1. Cycle 9 was shortened from 18 to approximately 13 months.
At our November 18,1991 meeting whn NRC, we presented our basis for continued operation of Unit 1 through April 1992. We also presented the impact to Virginia Power from a mid cycle outage. At our December 2,1991 meeting, we provided our response to the NRC staff's request for additionalinformation.
Based on the results of our operating interval assessment and in consideration of the NRC's questions and concerns, we find no technical reason to preclude a 13-month cycle of operation and we conclude that the mid cycle outage for North Anna Unit 1 is not necessary,
Background
During the 1991 refueling outage for North Anna Power Station Unit 1, an extensive eddy current inspection program was conducteo on the tubes in each of the three steam generators. - The results of the inspections were classified as Category C 3 for each of the unit's steam generators, i.e., greater than 1% of the inspec:ed tubes were defective and required plugging. Based on thepe inspection results, NRC approval for
- resumption of power operation was required in accordance with North Anna Unit 1 Technical Specification Table 4.4-2. In addition, a tube segment was pulled from B" steam generator to further examine circumferentially oriented indications at the tube to tube support plate intersections.
On February 26,1991, the inspection results, our technical evaluation, and our basis for restart and operation of the unit were discussed in detail at a meeting with the NRC.
Later_ that day, by letter dated February 26, 1991, Virginia Electric and Power Company requested NRC approval for restart of North Anna Unit 1. NRC approval for resumption of power operation was granted by letter dated March 7,1991. However, the NRC's startup approval stipulates that we must either implement a mid-cycle inspection of the steam generator tubes (i.e., perform the next inspection of the steam generator tubes at a frequency not to exceed 10 months from startup of the unit) or provido the additionalinformation necessary to justify a full cycle of operation.
As ststed in our February 26,1991 letter, we committed to provide the NRC with the results and assessments of the analysis on the tube segment pulled from the "B" steam generator, in addition, we identified that assessments must be made of the potentia!
for crack propagation of circumferential cracks cue to tube vibration, the potential for burst pressure reduction due to combined circumferential and axial cracks, and the potential leakage during a postulated steamline break for the projected end-of-cycle crack distribution.
On August 6,1991, we submitted a r9 port to the NRC containing the results and logic of our evaluations. In early November,1991, after review of this repcrt, the NRC staff l
requested additional information in the form of a list of 66 questions regarding the report and its conclusions. Meethgs were held with the NRC on November 18,1991 and December 2,1991 to review tnese questions and present our responses.
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4 Extent of Insoection Program During the 1991 refueling outage, an inspection program was conducted which relied on diverse inspection methods and which focused on areas of the steam generator tubing which had previously experienced tube degradation (i.e., the tube to tube support plate intersectior e and the top of the tubesheet region). The scope of this program was conservative with respect to the requirements of the North Anna Unit 1 Technical Specifications and NRC Regulatory Guide 1.83. The eddy current examination methods utilized include bobbin coil testing,8x1 probe testing, rotating pancake coil testing, and profilometry, in addition, a video scan of the tubesheet (inside the hot leg and cold leg channel head) coincident with a secondary side hydrostatic pressure of approximately 400 psig was performed to identify any leaking tubes or plugs before and after the other tube examinations were performed. Due to the nature of degradation previously found in the North Anna Unit 1 steam generators, extensive supplementalinspections were performed. The bobbin coil probe was used to examine the full-length (from the inlet tube end to tube end on the outlet side) of 100% of the available tubes.
Another of the examination methods, the 8x1 probe, was used as a screening tool for -
the tube support plate areas on the hot leg side. The 8x1 probe was used because it proved useful in identifying volumetric indications previously seen in the tube suppor1 plate regions. The rotating pancake coil (RPC) was then used to disposition distorted indications (Dis) and possible indications (Pis) identified by the 8x1 probe examination of these areas. In addition, the RPC probe was used to examine other suspect areas such as the WEXTEX expansion zone at the top of the tubesheet (hot leg side) for circumferential indications.
Comparisons of RPC and 8x1 probe detection capability were performed. Due to the geometry differences between expansion transitions and the dented tube support plate locations, the 8x1 probe performance was considered more applicable to dented intersections. The major differences between the 8x1 and RPC at support plates is that the 8x1 is expected to be less sensitive for detection of short cracks, such as those less than 45. The 8x1 pt- e is very reliable for crack lengths of 75 and higher. Both pulled tube and laborat y tube data bases have been utilized to develop the estimated detection threshold used in the steam generator operating cycle evaluation.
Tube R11-C14 from the "B' steam generator was identified during the inspection as one of 92 tubes with potential circumferentially oriented degradation. This tube was cut below the second support plate on the hot leg side and the lower segment was removed for examination to characterize the corrosion cracking suspected to exist at the first support plate. The results of the examination of the pulled steam generator tube were discussed with the NRC in conference calls held on March 8,1991 and April 26, 1991, by our letter to the.NRC, Serial No.91-267, dated May 7,1991 and discussed in detail at our update meeting on May 30,1991 at the NRC offices in White Flint.
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Tube Burst Caoabiktv Assessment in prov!ous years, cssessments of tube integrity for North Anna woro performed tot axial cracks. The results of those assessments havo been previously presented to the NHC. Our current efforts have focused on circumferential cracking at the WEXTEX transition and at the tube support plate elevations. Structural evaluations of WEXTEX area degradation t, ave boon performed and presented as early as 1989.
As part of our ovaluation, we have conducted an assessment of estimated tubo integrity for the end of the operating eve!o. Not only were the average tubo conditions assessed, but the expected populsion and condition of the most significantly degraded tubes were dolormined. Tho assessment methodology emoloyed to determine expected end of cycle tubo degradation was encompassed by the estimated threshold of detection, estimated growth rate, end oddy current uncertainty as required by NRC Regulatory Guido 1.121.
Conservative models ware developed for each aircumferential degradation mechanism. A model for circumferential outside diameter stress corrosion cracking (ODSCC) at the support platos and for circumferential primary water stress corrosion cracking (PWSCC) at the WEXTEX fransition was utilized, incorporating the unique characte;istics of growth, eddy currr uncertainty, and detection threshold for each.
Results of recent work with the Wes*gneuso Owner's Group WEXTEX subgroup have boon utilized in det lin0 those models. As a result of this assossment, we concluded that all tul with projected end of-cycle cracks using a 95% cumulativo probability level) coulo httand three tinnes the normal operating condition pressure loadings.
Shortly after submittal of the August 6,1991 operating cycle evaluation report, we completed an evaluation concerning the optimum timing for replacement of the existing steam generators at Unit 1. t3ased upon this evaluation, the desired operating interval for the current cycle has been reduced from 18 to 13 months. From the 18-month crack anglo projections, the predicted end of cycle angle is dominated by the threshold of detection and oddy current uncertainty. Growth rate accounts for approximately 40% of 'ho total crack angle. The impact of further reducing the operating interval fron.13 months to 10 months reduces the largest expected crack angle by less than 10%
Wo have also assessed the occurrence of combined cracks (axial and circumferential) at the support plate and the expansion transition. Based on the RPC inspection data as well as the pulled tube examination results, it is judged that the location of axial and circumferential cracking at the same elevation will be separated such that mixed modo cracking will not occur at the tube support plate elevations. Relative to the WEXTEX expansion transition region, RPC testing results indicated the presence of axial cracks on only four tubes. None of these tubes had both axial and circumferential orientations of cracking ', the WEXTEX transition region.
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Tube Vibration Annivseg Vibration analyses were performed for postulated circumfwential degradation that could occur in the unit. The most limiting areas that exist in the steam generator were determined to be the top of the tubesheet, the bottom of the first support plate, and a limited number of unsupported tubes at the top (seventh) support plate. For the tubeshoot to first support plato area, the peripheral tubes are subject to the highest vibration loadings. Potential crack propagation by two mechanismo, turbulence and fluidolastic excitation, was reviewed for the lower region of the steam generator.
The first mechanism, turbulence, is characterized by substantially lower amplitudes and loadings. As a turbulence propagated crack is advanced, it has the potential to reach an extent where fluidelastic excitation may become the driving mechanism.
Fluidelastic excitation is characterized by higher loadings and amplitudes, providing the driving force to rapidly propagate an oxisting crack as cited in NRC Bulletin 88 02.
From our evaluation, turbulence is the expected initial activating mechanism, providing the initial leakage which permits timely shutdown of the unit with substantial margin.
However, based on conservative projections for the end of c)cle cracks and their associated distributions, no tubes am predicted to exceed the through wall angular threshold for crack propagation by turbulence.
For the top support plate, the most limiting tubes in service from the NRC Bulletin 88-02 ovaluation were identified (reference WCAP 12351 entitled, North Anne Unit 1 Eveluation for Tube Vibration induced Fatigue, dated 1989). The majority of circumferential Indications from the last inspection are located in the lower tube support plate areas. Based upon past experience, it is extremely unlikely that a circumferentialindication would exist at the top support plate location. Even if this low probability event were to occur, North Anna has implemented a leakage monitoring program capable of timely chutdown of the unit.
Leak Before Break Evaluation At North Anna, the leakage limits itave been reduced to compensate for the lower leakage rates typically observed from stress corrosion cracks (SCC). It is not unusual for short length SCC to result in little or no leakage during plant operation. The principal causes for low leakage are associated with the presence of ligaments or the influence of support structures. However, these factors, which influence the leakrate, also contribute to the structural capability of the tubing. This strength contribution more than compensates for the reduced leukage from a leak before break point of view.
In the pulled tube, R11 C14, negligible leakage was present although a throughwall crack of 70 90 was present. The morphology of the ODSCC crack observed in this location revealed a highly irregular or Intenocking crack face. While low leakage for this morphology can be expected for such small crack angles, the crack opening area increases with crack length and would be expected to leak at greater than 50 GPD y before reaching the Regulatory Guide 1.121 structural limit. Should tube vibration occur, the vibration induced propagation is predicted to enhance crack leakage.
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Overall, leak before break can be expected to a high probability for OD or ID corrosion cracks approaching accident condition tube burst limits and is essentially assured under tu:)e vibration conditions. The North Anna Unit 1 N 16 fonkage monitoring system and implemoriting action steps, including the 50 GPD shutdown limit, would lead to plant shutdown in response to increasing leak rate trends well before tube rupture. l Conservative Leakaae Limits and Ooerating_ Constraints North Anna has aggressively investigated each of the principal degradation mechanisms by conducting tube removals and destructive examinations. Five tubes with G support plate elevations have been removed and analyzed over the last 6
. years. This has permitted better understanding of the mechanism, facilitated revisions '
to the eddy current rules to enhance crack detectability, and development of leakage models to explain the lower than expected leak rates observed.
Axlal cracking occurring at the tube support plates has been extensively reviewed.
Actual leakrates experienced in 1984 and 1985 were compared to the predicted leakrates from our leakage models in 1989 and found to be consistent with the predictions. The morphology of the cracks from other pulled tubes were also reviewed and from all mechanisms and it is expected that tube leakage would occur prior to challenging tube integuty. Programmatic reductions in the allowable primary to-secondary leakrates have been implemented, reducing the allowable leakage by a factor of 10 over the last 4 years.
By license amendment dated Decsmber 12,1988, Specification 3.4.6.3 was added to the North Anna Units 1 and 2 Technical Specifications to implement more stringent primary to secondary leakage limits and Technical Specification 3.4.6.4 was added to establish surveillance instrumentation requirements necessary to assure compliance with those leakage limits. The applicability of these Specifications is MODE 1 above >
50% power. The Specifications were added to ensure prompt operator action in response to a potential tube leak at the t p tube support plate due to fatigue.
In 1989, more conservative primary to secondary leakage limits were administratively implemented. These administrative limits were implemented to address a concern for prompt operator action as a result of any tube leak indication. The administrative controls limit primary to secondary leakage to 150 GPD total from all steam generators or 50 GPD from an individual steam generator. Originally, the administrative controls required that if either of these limits were excocded, power would be reduced to less than 50% power within 90 minutes, in response to recent concerns of circumferentially oriented tube degradation at the ,
lower tube support plates (and as discussed in our February 26,1991 letter), these administrative controls and required actions have been made trore conservative. The leakage limits previously established are maintained. However, the operating limitations are now applicable to operation at all power levels in MODE 1. If either limit is exceeded or will apparently be exceeded, then the operaters are required to reduce Page 6 of 10
p'ower to less than 50% power within 90 minutes and bolnw MODE 1 within two (2) hours from detection.
Following our submittal of the August 6,1991 report, we added an additional requirement to trip the reactor in the event the leakage in an individual steam generator exceeds 100 GPD increase in 30 minutes. Additionally, in regard to the case of no primary to secondary leakage detection instruments available, we will immediately suspend any power increases in progress (if applicable), reduce power to less than 50% power within 90 minutes, and below Mode 1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of deterrnining no available primary 1, secondary leakage detection systems. We will enter Mode 3 in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> if nn method of primary to secondary leakage monitoring continues to be available.
Lenhag9Alec1!0lLand Respansp_Capabihty At North Anna, an N 16 radiation monitoring system was installed in 1987. This system, as currently configured, has a monitor on each main steam line and a monitor on the main steam header. This configuration provides a redundant means of determining the operating leakage from the steam generators. Additionally, the N 16 monitoring system provides real time leakage determination. The system readout permits small changes in the leaktate to be observoa. Alarm setpoints have been incorporated to alert the operators if a significant step change occurs. Other backup or confirmatory leak detection methods include air ejector, steam generator blowdown, main steam radiation monitors, and grab sample chemistry and air ejector analyses.
Through increased frequency simulator training, control room operators have been made intimately aware and instructed to respond conservatively to indications of abnormal primary to secondary leakage indications. In addition, on-shift control room operators record and trend primary to secondary leakage from each N 16 radiation monitoring system and the condenser air ejector exhaust radiation monitor.
Sne.ciaLSigamline Break Anq!ysjs in addition to the general tube integnty assessments previously discussed, analyses were performed to assess the potential for crack propagation by vibration during a main steamline break event. For these analyses, no significant crack propagation is expected dunng such events.
An operational leak rate in excess of 50 GPD per steam generator will result in l immediate action to shut down the unit, effectively ending the operating cycle. Thus, primary-to-secondary leakage in the event of a main steamline break is limited by the analysis that establishes the leakage rate of the most limiting crack. This potentialleak rate is expected to be less than 9.5 GPM. Based on the operational leakage limit controls, it is expected that no other cracks would have propagated to the point that i they would contribute more than a negligible amount of "1e overallleakage rate.
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The radiological consequences of the potential 9.5 GPM post accident primary to-secondary leakage rato results were evaluated relative to the off sito doso assessment and control room habitability calculations. The increased leakage rato did not result in any increaso in the dose consequences at the sito boundary in excess of thoso previously ovaluated and approved by the NRC for North Anna. It was concluded that no unroviewod safety question existed.
Consldoration was also given to the control room operator doso consequences, it was observed that the limiting doso to control room operators as a result of the revised leak rato remalnod the thyroid doso. Although the skin doso appeared to increase, additional conservatism in the now analysis accounted for this change. Tho wholo l Dody doso also increased. However, the whole body dose consequences were not :
significant (in absolute terms) and thyroid dose is clearly the limiting dose to control ,
room operators when compared to the GDC 19 allowable limits. Therefore, it was determined that the Increased leakage rato did not result in increasing the limiting dose consequences to control room operators in excess of that previously evaluated and approved by the NRC and it was concluded that no unroviewod safety question existed.
However, to further substantiate our conclusions, an overly conservative approach was also ovaluated in that the cumulativo potentialleakage from a postulated main steamline break for the projected ond of cycle (i.e.,18 months) crack distribution was assessed. This approach was evaluated to bound potential leakage in the unlikely event that a large aopulation of tubes have through wall or near through wall cracks that do not leak curing normal operating conditions, but may leak during a main steamline break event. Therefore,it is assumed that the 50 GPD limit is ignored and that the end of cycle crack distribution models are then applied. Further, an overly conservative assumption is made that all cracks instantaneously becomo through-wall cracks. The integrated leakage rate from this approach is calculated to bo 49 GPM.
Although this scenario is considered to be overly conservative, the radiological consequences of the assumed 49 GPM integrated post accident primary to secondary leakage rate into the affected steam generator after the main steam line break were ovaluated relative to the 10 CFR Part 100 allowable off site dose limits and the GDC-19 allowable control room dose limits for North Anna Unit 1. Two specific casos woro examined as part of the evaluation. The first case was performed consistent with the Standard Review Plan guidelines with respect to requirements for iodine spiking. This case looked at the doses resulting from a MSLB assuming a pre accident spike to the 60 pCl/ gram dose-equivalent lodine 131 short term Technical Specification limit for full power operation as shown in Technical Specification Figure 3.41. The dose results of this case are within the 10 CFR Part 100 and GDC 19 limits. However, the thyroid dose at the exclunion area boundary is slightly greater than 10% of the Part 100 limits.
A second case was examined in order to reduce the calculated exclusion area boundary thyroid doso below 30 Rem (10% of the limit). This case was also examined consistent with the Standard Review Plan guidelines with the exception that a pro-accident lodine spike of 45 pCl/ gram (75% of the short term North Anna Unit 1 Technical Specification limit for full power as shown in Technical Specification Figure 3.41) was assumed. The dose results of this case are within the Part 100 and GDC-Pago 8 of 10
19 limits and the off site doses are less than a small fraction (< 10%) of the 10 CFR Part 100 limits.
The integrated leakage rate is calculated on an 18 month operating cycle. This adds an additional conservatism to this assessment. Even though we contend this special steam line break scenario is overly conservative, we commit to administratively control the North Anna Unit 1 coolant activity to less than 75% of the Technical Specifications limit (0.75 pCl/ gram dose equivalent lodine 131 for normal operation as shown in Technical Specification 3.4.8.a with a corresponding 25% reduction in the short term Technical Specification limits as shown in Figure 3.41) until the steam generators are replaced. This will ensure that the off site doses associated with this special steam line break analysis would remain below a small fraction of the 10 CFR Part 100 limits.
T;,is administrative control provides additional conservatism to our operating cycle evaluation and will be in place as soon as possible and no later than January 9,1992 to support the additional period of operation.
Conclusion Based on the our operating interval evaluation, we believe that a mid cycle outage is not necessary. Four tiers of conservatism exist in our evaluation. They are:
- Conservative inspection
- Conservative analysis methodology
- Conservative, state of the art leakage monitoring and detection capability
- Conservative leakage limits and operating restrictions The first tier of conservatism is in the inspection progiam. A diverso inspection program was performed that utilized standard bobbin probe,8x1 probe, and rotating pancake coil probe inspection methods, Based upon thete inspections, corrective actions such as tube plugging or stabilizing have been performed. This re established tube bundle integrity for beginning the current cycle of operation, t The second tier deals with the analysis methods used to determine tube Integrity through the operating interval. The methodology employed is consistent with criteria contained in Regulatory Guide 1.121. An expected distribution of c' 'mferential Indications was developed for the next outage. Threshold of detet . uncertainty, and growth rates were incorporated to develop this distribution. Both , Jied tube and laboratory _ samples were used to evaluate the eddy current detection threshold.
Pulled tube crack morphologies and the observed NDE uncertainties were applied to establish crack growth and structuralintegrity for the end of cycle distribution of cracks.
To confirm the methods used, the 1989 inspection data was reviewed and an end of-cycle (1991) distribution was predicted. This predicted distribution was consistent with the actual distribution encountered during the 1991 outage. For the 1992 refueling outage, it is projected (using a 95% cumulative probability level) that all tubes are expected to meet three times normal operating differential pressure loadings.
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- Potontial tube vibration mechanisms were also investigated. The most limiting areas for potential vibration induced propagation of existing cracks were identified. The probability of having a tubo sufficiently degraded to propagate by vibration is '
extremely low, in tho event this were to occur, turbulence is the expected inillating vibration mechanism, providing adequate time to detect the leak and shut down the unit. For a few tubos at the top support plate, fluidolastic vibration is predicted to be the initiating mechanism. However, in accordance with the NRC Bulletin 88 02 analysis, only a small population of tubes are potentially susceptible (reference WCAP 12351). As no e reumferential degradation has been observed at the top two support plates and with very little observed at the fifth support plato, it is extremely unlikely that this would occur in these locations during the current operating cycle.
The third tier of conservatism deals with leakage monitoring and detection. North Anna Technical Specifications requiro that certain primary to secondary leak detection systems be operable during operation. One of the key systems used is N 16. This system permits determination of real time leak rates, The system has alert and alarm set points and is located in the Control Room. As a result, plant operators are better able to trend leakago and take prompt corrective action should an adverse trend develop.
Finally, we have imposed conservative leakage limits to provide high margin with
-respect to leak bofore break. Models were developed to predict the leakage and structural behavior of degraded' tubes. Leakage predicted by these models was consistent with leakage physically observed in the unit. The morphology of both the ODSCC and WEXTEX PWSCC is expected to result in leakage prior to challenging tube structurallimits. The leakage limit of 50 GPD established is 10% of the typical leakage limit established in most plant technical Specifications and one half the limit contained in the North Anna Technical Specifications. Additional response tirne requirements for leakage rato changes ranging from a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> shutdown period to a reactor trip have been incorporated into plant procedures, permitting shutdown prior to tube rupture, in conclusion, the analysis demonstrates that tube with circumferential cracks can withstand significant (both angular and through wall penotration) degradation without challenging structural limits. The established threshold of detection, uncertainty, and growth rates (established from tubo pull data) support full cycle operation. However Virginia Electric and Power Company has provided additional margin by reducing the cycle duration from 18 months to 13 months. Imposition of a 10 month inspection outago would reduce the expected largest crack angle by less than 10%
Conservative leakage limits have been imposed to provide high leak before break margin. Aspects of those leakage limits have been in place for over four years. Prior operating cycles have resulted in leakage well below the 50 GPD limit, with similar degradation and less extensivo inspection methods or scope. There are no technical or operational reasons to preclude a full 18 month cycle of operation. Additional margin is provided with the shortened 13 month operating cycle. As a result, no mid-cycle (10 month) inspection is warranted. L Page 10 of 10
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