ML20138J963
| ML20138J963 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/07/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20138J955 | List: |
| References | |
| NUDOCS 9705090218 | |
| Download: ML20138J963 (7) | |
Text
pug p
UNITED STATES j
j NUCLEAR RE2ULATORY COMMISSION 2
WASHINGTON, D.C. 20005 4001
~%
/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO.189 TO FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS. INC.
ARKANSAS NUCLEAR ONE. UNIT N0. 1 DOCKET NO. 50-313
1.0 INTRODUCTION
On April 8, 1997, Entergy Operations Inc. (the licensee) reported to tiie staff in a teleconference that Arkansas Nuclear One, Unit 1 (ANO-1) was not in compliance with Technical Specifications (TSs) Section 4.18.5.b and a reactor shutdown would be required in accordance with TSs 4.0.3 and 3.0.3.
The issue was related to in service steam generator tubes that contain intergranular attack (IGA) that is believed to exceed the TS repair limit. The licensee requested that the Nuclear Regulatory Commission (NRC) exercise discretion not to enforce compliance with the actions required in TS 4.18.5.b.
By letter dated April 9,1997, the licensee submitted its formal request for a Notice of Enforcement Discretion (N0ED) pursuant to NRC's policy regarding exercise of discretion for an operating facility, as described in Section VII.C, of the General Statement of " Policy and Procedures for NRC Enforcement Actions" (Enforcement Policy), NUREG-1600.
By letter dated April 11, 1997, the staff documented the issuance of the N0ED to ANO-1. The N0E0 had been issued verbally on April 9,1997, after the staff concluded that the licensee's technical basis for the request was satisfactory.
Subsequent to the issuance of N0ED, the licensee submitted an exigent TS amendment to TS 4.18.5.b on April 11, 1997. The proposed change would added to TS 4.18.5.b the following statement:
"(t)ubes with intergranular attack within the upper tubesheet with the potential of through-wall depths greater than the plugging limit may remain in service for the remainder of cycle 14."
Cycle 14 is scheduled to end in spring 1998.
Following discussions with the NRC staff, the licensee submitted a supplement to the exigent amendment request on May 2, 1997.
TS 3.1.6.3.0 requires the reactor shutdown if steam generator tube leakage exceeds 500 gallons per day.
The supplement reduced the permissible tube leakage to 144 gallons per day for the duration of cycle 14. Tiie supplement adds a more conservative requirement to.the TSs to compensate for continued operation with tubes that may include IGA flaws that exceed TS limits. The supplement does not' change the scope of the notice and does not change the significant hazards evaluation that was published in the Federal Register on April 15, 1997.
9705090218 970507 PDR ADOCK 05000313 PDR h
I
~
. g 2.0 DISCUSSION The ANO-1 TSs require that tubes having degradation greater than 40% through wall be repaired or removed from service. During the steam generator tube inspection in refueling outage (IR13) in December 1996, the licensee used the bobbin probe to size the depth of indications in the upper tubesheet that they attributed to IGA.
Prior to the inspection, the licensee used Electric Power Research Institute guidelines to qualify an eddy current (EC) sizing technique i
specifically for measuring the depth of IGA indications. As a result of inspection, the licensee has left about 470 IGA indications in service because their depths were measured by EC at less than the TS repair limit of 40%
through-wall. The licensee stated that the 470 IGA indications were located in the region of the tubes insis; the upper tubesheet. The IGA indications with a depth greater than or equal to 40 percent through-wall, as measured by the qualified sizing technique, were removed from service.
During the outage, the licensee removed three tubes containing a total of 11 IGA indications for destructive examinations. The tibes were selected on the basis of indications that would have required repair or were near the repair limit. The licensee concluded based on EC measurements that the degradation 4
in the three pulled tubes bounded the degradation of the tubes left in service i
with IGA in the upper tubesheet. After the burst test in the laboratory, the licensee compared the actual depths of degradation measured by destructive examinations to the depths measured by their eddy current technique during the inspection. The comparison yielded a systematic non-conservative bias of 3%
to 50% for the IGA patches on these three tubes.
The discrepancy in the IGA measurements of the bobbin probe raised the concerns that some of the 470 indications remaining in service may contain indications exceeding the TS repair limit of 40% through wall. The proposed j
TS amendment would give AN0-1 an one-time authorization to operate with tubes having indications exceeding the TS limit for the remainder of the cycle.
ANO-1 uses two once through steam generators (model 177) fabricated by Babcock and Wilcox.
Steam generator A contains 285 IGA indications and steam generator B contains 185 IGA indications in the upper tubesheet, j
3.0 EVALUATION The staff focused its review on whether the tubes containing the 470 indications would maintain adequate structural and leakage integrity during the remainder of cycle 14, given that some of these indications may exceed the TS repair limit.
. Regulatory Guide-(RG) 1.121 specifies that tube structural integrity may be demonstrated by subjecting the tube to the larger of three times the normal operating differential pressure or 1.4 times the main steam line break j
differential pressure. Tube leakage integrity may be demonstrated by subjecting the tube to the differential pressure the tube would experience under a postulated main steam line break.
I J
. The licensee performed pressure tests on the three removed tubes in accordance with RG 1.121 and reported that the tubes withstood pressures in excess of 10,000 psig without leaking or bursting. The three times normal operating differential pressure for ANO-1 is 3765 psig and the differential pressure tubes would experience under a postulated main steam line break is 2500 psid i
for AN0-1. The burst testing results indicate that substantial structural margin exists for pulled tubes with IGA indications.
In addition, in 1996, the licensee performed burst tests on laboratory prepared tubes in support of the IGA study. The burst test program consisted of nine tubes containing drilled through-wall holes up to 0.5 inches in diameter and one tube containing no defects. All tubes with the laboratory defects were tested within a simulated tubesheet. A test bladder was inserted into the tube specimen and placed over the drilled hole to prevent leakage so i
that a burst test could be conducted. Nine of the specimens burst at pressures greater than 10,941 psig.
Each tube burst outside the tubesheet 3
within the non-defected portion of the tubes. One tube reached a pressure of 9,577 psig but did not burst due to test bladder leakage.
The burst test results indicated that the tubesheet provides sufficient support to preclude tube burst within the tubesheet.
The licensee compared the IGA indication data between cycle 12 and cycle 13 and found that IGA indications exhibited little or no growth.
The licensee stated that review of tubesheet IGA eddy-current data prior to cycle 12 confirms the same observation.
In addition, during May 1996, tubing in steam generator B was subjected to a differential pressure of about 2100 psig for several hours as a result of a feedwater transient. The structural and
. leakage integrity of the tubes were maintained during the event.
4 As mentioned above, the licensee tested pulled tubes with IGA indications that bounded the indications left in service. The test results showed that the pulled tubes satisfied the margins of RG 1.121.
In addition, the existing IGA indications showed little or no growth. The staff concludes that the licensee has demonstrated that the IGA indications left in service would not i
significantly affect the structural or leakage integrity of the tubes for the remainder of the cycle.
2 In addition, the licensee stated that the worse case scenario resulting from the continued operation with the existing IGA flaws would be the development of a primary-to-secondary leak. To compensate for this leakage concern, the licensee added more conservative primary-to-secondary leakage criterion to the TSs.
Permissible TS leakage criterion was decreased from 500 gallons per day to 144 gallons per day for the duration of cycle 14. TSs require reactor shutdown within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the leakage criterion is exceeded. Additionally, the licensee revised procedure A0P 1203-023 to require reactor shutdown if confirmed tube leakage exceeds 100 gallons per day. The more stringent procedural limit is intended to provide assurance that the TS leakage limit is never exceeded, f,
l i
i L
[ '
i I
ANO-1 has several monitors to detect steam generator tube leakage as a.part of the defense-in-depth measures. The licensee uses monitors to detect radiation i
levels in the condenter off-gas and N-16 gamma levels in the secondary i
systems. The main steam high range radiation monitors provide input to the j
safety parameter display system (SPDS) for display in the front of the control room. The SPDS display will flash to alert the operators when a i
parameter causes the system to alarm such as during tube leakage. The control l
room'also has alarms from other monitors to notify the operators if activity (leakage) is detected. The licensee has trained the operators to respond to i
primary-to-secondary tube leaks and ruptures. The training enables the operators to perform timely diagnosis and to take corrective actions as j
necessary to shut down the plant.
4.0 TECHNICAL CONCLUSION The staff concludes that the licensee has provided information on the tubesheet IGA left in service that demonstrates that the structural and leakage integrity of the tubes in AN0-1 steam generators will be maintained for the remainder of cycle 14. The staff also concludes that tube leakage or burst would not be expected even in the unlikely event of a main steam line break.
Staff approval of the TS change to continue operation with IGA indications for the duration of cycle 14 is based on the conclusion that the inherent structural support provided by the tubesheet would preclude catastrophic tube H
failure. Should tube leaks develop in this region, the staff concludes that existing monitoring capabilities and the TS imposed conservative leakage
-limits minimize the likelihood for continued operation with undetected or i
excessive primary-to-secondary leakage.
]
5.0 EXIGENT CIRCUMSTANCES
The Commission's: regulations,10 CFR 50.91, contain provisions for issuance of amendments.when the usual-30-day public notice period cannot be met. One type of special exception is an exigency. An exigency is a case where the staff and licensee need to act promptly and the staff has determined that the amendment involves no significant hazards considerations.
Under such circumstances, the Commission notifies the public in one of two ways: by issuing a Federal Reaister notice providing an opportunity for hearing and allowing at least two weeks for prior public comments, or by issuing a press release discussing the proposed changes, using the local media.
In this case, che Commission used the first approach.
The licensee's initial application was noticed in the Eggen1 Reaister on April 22, 1997, at which time the staff proposed a no significant hazards consideration determination.
In the application, the licensee requested that the amendment be processed under exigent circumstances for the following reason. During the IR13 refueling outage, an eddy current technique was used for the satisfactory completion of the ANO-1~ steam generator inspection
- - 1 surveillance. The technique used had been qualified per Appendix H of the EPRI "PWR Steam Generator Tube Examination Guidelines." This technique was used to depth size all intragranular attack flaws within the upper tubesheet.
As required by the technical specifications, all upper tube sheet IGA i
indications with a depth size of greater than the plugging limit as determined l
by the qualified sizing technique, were also removed from service by plugging.
During the steam generator inspections, three tube samples containing upper tubesheet IGA flaws were removed from the "B" steam generator and sent offsite to be analyzed for future development of an alternate repair criteria and to further support the qualified eddy current sizing technique employed during refueling outages. The preliminary destructive examination results were recently received by the ANO staff. This data arrived approximately 5 months j
after the resumption of operation following the steam generator inspections that occurred during 1R13. These results indicate that the flaw depths do not correlate well with the depths sized using the qualified eddy current technique. Upon further review, ANO has determined that the application of the sizir.g criterion is no longer valid. With the qualified sizing technique j
invalidated, there is a potential that tubes could have been left in service with indications that have through-wall depths greater than the plugging limit j
specified in the technical specifications. This would be considered a condition that is not allowed by the technical specifications. Prior to the i
l receipt of the preliminary destructive examination results, ANO had no reason to question the adequacy of the steam generator inspections that occurred during 1R13.
In order to continue plant operation in non-compliance with Technical Specifi ation 4.18, enforcement discretion was verbally requested by the licensee and received from the NRC on April 9, 1997.
Enforcement discretion was requested for a period of time necessary for the NRC to process this technical specification change which will allow continued operation in the current configuration for the remainder of the operating cycle. A Notice of Enforcement Discretion was issued for this purpose and limited to May 7,1997, after which time the actions of TS 3.0.3 are required to be followed (i.e.,
the reactor would be required to be shutdown). Accordingly, the licensee promptly submitted its amendment application and requested that the proposed technical specification change be considered under exigent circumstances as described in 10 CFR 50.91(a)(6).
Accordingly, pursuant to 10 CFR 50.91(a)(6), the Comission has determined that an exigent situation exists and that failure to act in a timely way will result in an unnecessary shutdown of the plant.
Further, the Comission has i
determined that the exigent situation is not due to the failure of the licensee to act in u timely manner.
6.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION The Comission's regulations in 10 CFR 50.92 state that the Comission may make a final determination that a license amendment involves no significant hazards considerations if operation of the facility in accordance with the amendment would not (1) involve a significant increase in the probability or
consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.
Operation of the facility in accordance with the proposed amendment will not involve a significant increase in the probability or consequences of an accident previously evaluated. A steam generator tube failure is a breach of the reactor coolant pressure boundary. This type of event is analyzed in the ANO-1 Safety Anklysis Report (SAR).
Proposed operation for the remainder of the fuel cycle with potential tube flaws that exceed 40% through-wall in the tube sheet region d~oes not involve a significant increase in the probability or consequence of an accident previously evaluated. Tube flaws in the tube sheet region can develop into primary system leaks during the current fuel cycle however the potential for a catastrophic tube failure, an accident analyzed in the SAR, is reduced by the fact that the tube sheet serves as a structural support for the tube segments which may have flaws exceeding 40%
through-wall.
Licensee test results verify that catastrophic tube failures are not likely to occur in the tube sheet region. Tube leaks that could develop in the tube sheet region would be detected during operation and the reactor would be shut down well before the leakage could challenge the accident evaluations described in the SAR.
Operation of the facility in accordance with the proposed amendment will not create the possibility of a new or different kind of accident from any accident previously evaluated.
This change does not introduce any new modes or methods of plant operation.
The design and purpose of the steam generators is not affected by the proposed changes.
In addition, a steam generator tube failure is already addressed in existing accident analysis.
Operation of the facility in accordance with the proposed amendment will not involve a significant reduction in a margin of safety. As noted above, operation with tube flaws in the tube sheet region does not significantly increase the probability or consequence of an accident previously evaluated.
As a result, operation with tube flaws in the tube sheet region that may 1
exceed 40% through-wall does not involve a significant reduction in a margin of safety.
Should these flaws develop into actual reactor coolant leaks, the reactor would be shut down before any safety margins could be significantly reduced.
Based on the above considerations, the staff concludes that the amendments meet the three criteria of 10 CFR 50.92. Therefore, the staff has made a final determination that the proposed amendments do not involve a significant hazards consideration.
7.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Arkansas State official was notified of the proposed issuance of the amendment. The State official had no comments.
i I
4 l -
l 1
8.0 ENVIRONMENTAL CONSIDERATION
i 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 i
Part 20 and changes surveillance requirements. The NRC staff has determined l
that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative i
occupational radiation exposure. The Commission has previously issued a i
proposed finding that the amendment involves no significant hazards i
consideration, and there has been no public comment on such finding (62 FR 19628). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51'22(c)(9).
Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
9.0 [Qt(CLUSION 4
The Commission has concluded, based on the considerations discussed above, 4
that:
(1) there is reasonable assurance that the health and safety of the i
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:
J. Tsao Date:
May 7, 1997 i
i i
i
-w-y,
-.,,e
%~