1CAN119402, Withdraws Exemption Request from Requirements of 10CFR50, App J & Reapplication for Exemption from 10CFR50,App J Concerning Type a Test Interval & Frequency & Associated TS Change Request

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Withdraws Exemption Request from Requirements of 10CFR50, App J & Reapplication for Exemption from 10CFR50,App J Concerning Type a Test Interval & Frequency & Associated TS Change Request
ML20078F716
Person / Time
Site: Arkansas Nuclear Entergy icon.png
Issue date: 11/08/1994
From: Mims D
ENTERGY OPERATIONS, INC.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20078F717 List:
References
1CAN119402, NUDOCS 9411150355
Download: ML20078F716 (12)
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Text

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m sne November 8,1994 1CAN119402 U. S. Nuclear Regulatory Commission Document Control Desk Mail Station PI-137 Washington, DC 20555

Subject:

Arkansas Nuclear One - Unit 1 Docket No. 50-313 License No. DPR-51 Withdrawal of Submitted Exemption Request from the Requirements of 10CFR50 Appendix J and Reapplication for Exemption from 10CFR50 Appendix J Concerning Type A Test Interval and Frequency and Associated Technical  ;

Specification Change Request Gentlemen:

1 By letter dated August 30,1994 (ICAN089405), Entergy Operations submitted an exemption request characterized as a cost beneficial licensing action (CBLA) which would have delayed performance of the next integrated leak rate test (ILRT) at Arkansas Nuclear One, Unit 1 (ANO-1) until after the completion of the staffs proposed rulemaking on Appendir J. After discussions with the NRC staff, Entergy Operations agrees that the August 30, 1994, exemption request could be considered a " performance-based" request per SECY-94-036,

" Staff Plans for Revising 10CFR50, Appendix J, ' Containment Leakage Testing,' and for Handling Exemption Requests," and, as a result, would not receive review priority by the NRC staff. Therefore, Entergy Operations withdraws the above dated exemption request.

The administrative technical specification change request included in the August 30, 1994, letter, however, is necessary in order to implement the supplemental exemption request described below. The technical specification change request is reiterated in its entirety in this submittal for convenience.

Since the current ILRT schedule would require the performance of ILRTs in the next two successive ANO-1 refueling outages, Entergy Operations hereby applies, in accordance with 10CFR50.12 and 10CFR50.90, for an exemption from 10CFR50 Appendix J, " Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," regarding integrated leak rate testing intervals and frequencies, and for a technical specification change necessary to implement the exemption at ANO-1. This exemption and change would provide a one-time interval extension for the next Type A test from the currently scheduled 34 months to 53 months. This extension would result in the delay of the ILRT currently scheduled for the twelflh refueling outage (lR12) until the thirteenth refueling outage (IR13) in order to correspond with the last refueling outage in the current ANO-l inservice inspection / inservice 3

. 9411150355 941108 1 PDR ADOCK 05000313 P PDR 1

U. S. NRC November 8,1994 1CANI19402 Page 2 testing (ISI/IST) interval. The exemption would also extend the 10-year ILRT interval to correspond to the end of the ISI/IST interval.

The exemption request has been found to meet the requirements of sections (a)(2)(ii) and (a)(2)(vi) of 10CFR50.12. Additionally, the proposed technical specification change has been

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i evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for these determinations are included in the attached submittal.

Entergy Operations requests that the effective date for this technical specification change and exemption be the date of NRC approval, but not later than December 30,1994, in order to 7

facilitate scheduling for the IR12 outage. Although this request is neither exigent nor emergency, your prompt review is requested.

l Very truly yours, k

wig 1 .Mims Dire or Licensing DCM/jd 1 Attachments I

l l

l l

U. S. NRC November 8,1994 1CANI19402 Page 3 cc: Mr. Leonard J. Callan Regional Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One 1448 S. R. 333 Russellville, AR 72801 Mr. George Kalman NRR Project Manager Region IV/ANO-1 & 2 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-H-3 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Eugene Imbro Director, CBLA Staff U. S. Nuclear Regulatory Commission NRR Mail Stop 9-A-1 One White Flint North 11555 Rockville Pike Rockville, MD 20852 Ms. Greta Dieus Arkansas Department of Health Division of Radiation Control and Emergency Management 4815 West Markham Street Little Rock, AR 72205

9 ATTACHMENT 10 1CANI19402 PROPOSED TECHNICAL SPECIFICATION AND RESPECTIVE SAFETY ANALYSES IN THE MATTER OF AMENDING LICENSE 'NO. DPR-51 ENTERGY OPERATIONS, INC.

ARKANSAS NUCLEAR ONE, UNIT ONE DOCKET NO. 50-313 l

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' Attachment to I ICANI19402 Page 1 of 8  !

DESCRIPTION OF PROPOSED CHANGES

" Type A Tests" are defmed in Appendix J Section II.F as " tests intended to measure the primary reactor containment overallintegrated leakage rate."

Exemption is requested from the following portion of Appendix J,Section III.D.l.(a) for Type A test intervals:

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" Type A tests shall be performed, at approximately equal interuls during each 10-year service period. The third test of each set shall be conducted when the plant is shutdown for the 10-year plant inservice inspections."

The exemption will allow the interval between the second and third ILRT of the second period to be extended from 34 months to 53 months, while at the same time extending the end of the ILRT period to encompass the last refueling outage in the current ANO-1 ISI/IST interval, scheduled to begin September 20,1996 (IR13). The current ILRT schedule would require a test during IR12 to meet "approximately equal" interval requirement of Appendix J and an additional ILRT during 1R13, while the plant is shutdown for completion of the "10-year plant inservice inspections." Completion of these two ILRTs would result in the performance of four ILRTs during the second 10-year service period.

ANO-1 Technical Specifications (TS) 4.4.1.1.4 is being revised to reference 10CFR50, Appendix J directly, rather than to paraphrase the regulation, and to allow approved exemptions to the ILRT frequency requirements.

BACKGROUND The reactor building is a Seismic Class 1, fully continuous, reinforced prestressed concrete structure in the shape of a cylinder with a shallow domed roof and a flat foundation slab. The cylindrical portion is prestressed by a post tensioning system consisting of horizontal and vertical tendons. The dome has a 3-way post-tensioning system. Hoop tendons are placed in 3-240 degree systems using three buttresses as anchorages. The foundation slab is conventionally reinforced with high strength reinforcing steel and is founded on bedrock. A continuous access gallery is provided beneath the base slab for installation of vertical tendons.

A welded steel liner is attached to the inside face of the concrete shell to insure a high degree ofleak tightness. The base liner is installed on top of the structural slab and is covered with concrete. The structure provides shielding for both normal and accident conditions. The reactor building completely encloses the entire reactor and reactor coolant system (RCS).

This insures that an acceptable upper limit for leakage of radioactive materials to the environment will not be exceeded even if gross failure of the RCS occurs. Further information on the reactor building design can be found in Section 5.2 of the ANO-1 Safety Analysis Report (S AR).

The intent of the Type A test (i.e., ILRT) is to determine that the total leakage from containment does not exceed the maximum allowable leakage rate (La ) as specified in ANO-1

I Attachment to i ICAN119402 Page 2 of 8 TS 4.4.1.1.1, and the ANO-1 SAR. The ANO-1 reactor building design maximum allowable leakage rate (La) measured in weight percent /24 hours at the design pressure of 59 psig (ANO-1 TS conservatively require the ILRT to be performed at design pressure instead of peak accident pressure) is 0.200%/ day. ANO-1 TS 4.4.1.1.5 and 10CFR50, Appendix J require the measured Type A test acceptance criteria to be less than or equal to 75% of La , or 0.150%/ day, to allow for deterioration of leakage paths between tests. The maximum allowable leakage rate provides an input assumption to the calculation requi ed to ensure that the maximum allowable offsite dose during a design basis accident does not exceed that specified in 10CFR100.

The safety aspects of this request are similar to those approved on September 30,1594, for a one time ILRT frequency and interval extension exemption, and a corresponding technical specification amendment (amendment 196) at the Peach Bottom Atomic Power Station, Unit 2.

Type A Testing History Six Type A tests have been conducted at ANO-1, with five being conducted at design pressure, 59 psig. Two different testing methods were employed in the performance of these tests; the mass point leakage rate method and the total time leakage rate method. Only the total time method was utilized for the preoperational ILRT. Trends of previous test results indicate that the proposed 19 month extension would not jeopardize the ability of the reactor building to maintain the leakage rate at or below the required Type A limits. The results of the individual ANO-1 Type A tests follow:

Preoperational Type A Test 3 The preoperational Type A test was successfully performed on November 8-13,1973 with the following results: 1) A total time leakage rate of 0.0815%/ day at a full test pressure of 59 psig, and 2) A total time leakage rate of 0.0292%/ day at a reduced test pressure of 30 psig and a calculated acceptance limit of 0.107%/ day. The ILRT test report was submitted to the NRC by letter dated March 8,1974 (ICAN037404).

First Periodic Type A Test The first periodic Type A test was successfully performed on March 10-12, 1978 with the following results: 1) A calculated mass point leakage rate of 0.009%/ day and a 95% upper confidence limit (UCL) of 0.014%/ day, and 2) A total time leakage rate of 0.012%/ day and a 95% UCL of 0.069%/ day. Since this test was performed at a reduced pressure of 30 psig, the allowable leakage rate was reduced to 0.054%/ day. The ILRT test report was submitted to the NRC by letter dated July 8,1978 (ICAN077803).

Attachment to 1CANI19402 l Page 3 of 8 l Second Periodic Type A Test The second periodic Type A test was successfully performed at design pressure on February 24-25, 1981 with the following results: 1) A calculated mass point leakage rate of 0.038%/ day and a 95% UCL of 0.046%/ day, and 2) A total time leakage rate of 0.034%/ day and a 95% UCL of 0.092%/ day. The ILRT test report was submitted to the NRC by letter dated May 22,1981 (ICAN058109).

Third Periodic Type A Test The third periodic Type A test was successfully performed at design pressure on December 14-17, 1984 with the following results: 1) A calculated mass point leakage rate of 0.041%/ day and a 95% UCL of 0.046%/ day, and 2) A total time leakage rate of 0.031%/ day and a 95% UCL of 0.103%/ day. This test corresponded to the end of the first ILRT 10-year testing period. The ILRT test report was submitted to the NRC by letter dated March 21, 1985 (ICAN038507), as supplemented by Sampling Method Meets NRC Guidelines|letter dated July 15,1985]] ((ICAN078504).

Fourth Periodic Type A Test The fourth periodic Type A test was successfully performed at design pressure on November 8-9, 1988 with the following results: 1) An as found calculated mass point leakage rate of 0.081%/ day and a 95% UCL of 0.084%/ day,2) An as leR calculated mass point leakage rate of 0.077%/ day and a 95% UCL of 0.080%/ day,3) An as found total time leakage rate of 0.090%/ day and a 95% UCL of 0.ll8%/ day, and 4) An as len total time leakage rate of 0.086%/ day and a 95% UCL of 0.114%/ day. The ILRT test report was submitted to the NRC by letter dated March 21,1989 (ICAN038913).

Finh Periodic Type A Test The finh periodic Type A test was performed at design pressure on April 13-16,1992 with the following results: 1) ihe as found leakage rate was greater than La due to the local leakage rate adjustments specified by NRC IE Information Notice No. 85-71 from leakage at one penetration,2) An as len calculated mass point leakage rate of 0.0738%/ day and a 95%

UCL of 0.0844%/ day, ') An as found total time leakage rate of 0.131%/ day without including the one penetraton with a large leakage rate, and 4) An as len total time leakage rate of 0.0738%/ day and a 05% UCL of 0.1248%/ day. The ILRT test report was submitted to the NRC by letter dated J.me 25,1992 (ICAN069203).

The failure of the as found ILRT leakage rate to meet the 75% La criterion aner local leakage rate adjustments was due to the inability of reactor building purge penetration V-1 to maintain test pressure between valves CV-7402 and CV-7404 during the Type C test. The as found leakage rate at this penetration was not measurable because the penetration was unable to maintain test pressure and assumed to be greater than L a . The as len leakage after repairs (replacement of 35 seal adjustment bolts on valves CV-7402 and CV-7404) was not t

detectible. The local leak rate tests were performed prior to the ILRT. As discussed in the i

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Attachment to ICANI19402 Page 4 of 8 next section of this submittal, the purpose of the ILRT, the testing of the passive containment structure, was successfully fulfilled by the fifth periodic ILRT.

DISCUSSION OF CIIANGE Factors affecting leak tightness of the reactor building may be categorized as: 1) active components which are leak rate tested by Type B and C tests, and 2) passive components which constitute the reactor building structure and are tested during the Type A test.

Active Components The purpose of containment leak testing is to detect any containment leakage resulting from active or passive failures in the containment isolation boundaries before an accident occurs.

The existing Type B and C testing programs are not being modified by this request and will continue to effectively detect reactor building leakage caused by the degradation of active reactor building isolation components (e.g., valves) as well as sealing material within reactor building penetrations. The only potential failures that would not be detected by Type B and C testing are mechanical failure of the reactor building shell, penetration guard pipes, or welds between pipes and the reactor building shell. Only these reactor building structural failures could possibly be afTected by the propo. sed extension in Type A testing frequency.

Industry experience indicates that 97% of the failures associated with Type A tests are found to be due to Type B and C tested penetrations (Draft NUREG 1493, " Performance-Based Containment Leak Test Program"). The local leak rate testing frequencies of these penetrations are not affected by this proposed change. Therefore, continued overall leak tightness of the active reactor building components can be assured by the existing Type B and C testing program. At ANO-1, the only ILRT failure was due to leakage through an active 1 component. .

l Passive Structure j l

Two mechanisms could adversely affect the passive structural capability of the reactor building. The first is deterioration of the structure due to pressure, temperature, radiation, chemical or c ther such efTects. Secondly, modifications can be made to the structure which, if not carefully controlled, could leave the structure with reduced capability.

Absent actual accident conditions, structural deterioration is a gradual phenomenon requiring l periods of time well in excess of the proposed 19 month interval extension. Other than accident conditions, the only pressure challenge to the reactor building is the ILRT itself. l 10CFR50 Appendix J Section V. A. requires a general inspection of the accessible interior and exterior surfaces of the reactor building structures and components to be performed prior to any Type A test to uncover any evidence of structural deterioration which may afTect either the reactor building structural integrity or leak tightness. At ANO-1 there has been no evidence of structural deterioration that would impact structural integrity or leak tightness.

Attachment to 1CANI19402 Page 5 of 8 Modifications which would alter the passive reactor building structure are infrequent and i would receive extensive review to ensure reactor building capabilities are not diminished. The ANO design change and 10CFR50.59 programs have been demonstrated effective in providing a high quality oversight of such safety significant modifications.

Entergy Operations has reviewed modifications installed since the performance of the last l ILRT which could affect reactor building integrity, as well as those scheduled for completion in IR12. Modifications installed in IRll received appropriate Type B or C tests after completion, or did not affect the passive components of the reactor building. Appropriate l modifications to be installed in IR12 will also receive Type B or C tests.

Risk Imnget Assessment The risk impact of reactor building structural leakage is measured by a pathway created for radionuclides in the event that the reactor building is challenged, such as in a loss of coolant accident (LOCA) or severe accident. Such leakage does not create any new accident scenarios, nor does it contribute to the initiation of any accident.

From a risk standpoint, the purpose of Appendix J leak testing is to detect any reactor building leakage resulting from failures in the reactor building isolation boundary before an l accident occurs. Such leakage could be the result of leakage through reactor building penetrations, through airlocks, or through reactor building structural faults. The Appendix J Type B and C tests, which are unaffected by this proposed change, will continue to detect leakage through reactor building valves, penetrations, and airlocks. The only potential failures which would not be detected by Type B and C testing are mechanical failures of the containment shell (i.e., degradations or modifications to the containment shell), penetration guard pipes, or welds between pipes and the reactor building shell. Thus, the only potential effect of the proposed change to the Type A test frequency is the probability that reactor building structural leakage would go undetected between tests.

The reactor building structure is passive. Under normal operating conditions, there is no significant environmental or operational stress which could contribute to its degradation. A review of modifications for potential effects to the reactor building structure is discussed in the preceding section. Passive failures resulting in significant reactor building structural leakage are therefore extremely unlikely to develop between Type A tests. No such failures have occurred at ANO-1.

Postulated reactor building failure under severe accident conditions is primarily due to phenomenological effects associated with severe accidents. Such phenomenological effects were considered as part of the ANO-1 IPE. None of the identified reactor building failure mechanisms for severe accidents would be impacted by the proposed increase in the testing interval.

Attachment to ICANI19402 Page 6 of 8 Dose Reduction By omitting the ILRT scheduled for 1R12, dose savings would be achieved by eliminating contamination and reducing exposure associated with venting and draining and from setups and restorations ofinstrumentation required to perform the ILRT.

BASIS FOR EXEMPTION The proposed interval extension meets the criteria for special circumstances as described in 10CFR50.12(a)(2)(ii) and (vi).

50.12(a)(2)(ii) Application of the Regulation is not Necessary to A_q_hieve the Underlying Purnose of the Rule The underlying purpose of 10CFR50, Appendix J, is still achieved if a fourth ILRT is not conducted in the second period. Appendix J states that the leakage test requirements set forth in this appendix provide for periodic verification by tests of the leak tight integrity of the primary reactor containment. The appendix further states that the purpose of the tests are to assure that leakage through the primary reactor containment shall not exceed the allowable leakage rate values as specified in the technical specifications or associated bases.

Section III.D.l.(a) states that a set of three periodic tests shall be performed at approximately equal intervals during each 10-year period and that the third test shall be, conducted when the plant is shutdown for the 10-year plant inservice inspections. This exemption would provide a one time 19 month extension to the interval between ILRTs, thus causing unequal intervals, would extend the ILRT 10-year period to match the inservice inspection interval, and would eliminate the need to perform a fourth ILRT during the second ILRT period. The methodology, acceptance criteria, and technical specification leaxage limits for the performance of the Type A test will not change.

The testing history, structural capability of the reactor building, and the risk assessment discussed previously establish that ANO-1 has had acceptable reactor building leakage rates, that the structural integrity of the reactor building is assured and that there is negligible risk impact in extending the Type A test interval.

Thus, there is significant assurance that the extended interval between Type A tests will continue to provide periodic verification of the leak tight integrity of the reactor building while reducing occupational exposure and the cost associated with performing two ILRTs.

50.12(a)(2)(vi) Presence of Material Circumstances not Considered when the Regulation was Adopted There are material circumstances not considered when the regulation was adopted. The bervfit of time has provided experience and information which provide a better perspective

l Attachment to  !

1CAN119402 )

Page 7 of 8 about containment integrity. Two important material circumstances are testing fdstory and the  ;

development of probabilistic risk assessments (PRAs).

I Since the promulgation of 10CFR50, Appendix J, in 1973, more than 20 years of nuclear power plant operating history has been obtained. Entergy Operations' personnel have performed a review ofindustry data and did not find any instances where a Type A test failed to meet Appendix J acceptance criteria as a result of a containment structural leak not due to initial fabrication or a plant modification. This additional operating history provides a significant indicator that containment structural integrity (passive) is not a safety concern.

Plant specific PRAs were not available and therefore were not considered when the regulation requiring compliance with Appendix J (10CFR50.54(o)) was adopted. Overall plant risk due to reactor building leakage is relatively small given the small probability of reactor building leakage itself. The predominant contributor to degraded reactor building integrity is the phenomenological effects of a severe accident, and not preexisting reactor building isolation conditions.

DETERMINATION OF NO SIGNIFICANT IIAZARDS CONSIDERATION An evaluation of the proposed change has been performed in accordance with 10CFR50.91(a)(1) regarding no significant hazards considerations using the standards in 10CFR50.92(c). A discussion of these standards as they relate to this amendment request follows:

Criterion 1 - Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

The proposed change revises Technical Specification 4.4.1.1.4 to reference the testing frequency requirements of 10CFR50, Appendix J, and to state that NRC approved exemptions to the applicable regulatory requirements are permitted. The current requirements of TS 4.4.1.1.4 paraphrase the requirements of Section III.D.I.(a) of Appendix J. The proposed administrative revision simply deletes the paraphrased language and directly references Appendix J. No new requirements are added, nor are any existing requnements deleted. An approved exemption to Section III.D.l.(a) of Appendix J would not necessarily affect the requirements of TS 4.4.1.1.4, unless the proposed clarification phrase permitting the use of approved exemptions is added. Any specific changes to the requirements of Section III.D.I.(h) will require a submittal from Entergy Operations under 10CFR50.12 and subsequent review and approval by the NRC prior to implementation. The proposed change is stated generically to avoid the need for further TS changes if different exemptions are approved in the future.

The proposed change, in itself, does not affect reactor operations or accident analysis and has no radiological consequences. The change provides clarification so that TS changes will not be necessary in the future to correspond to applicable NRC approved exemptions from the

Attachment to

-' 1CAN119402 Page 8 of 8 requirements of Appendix J. Therefore, this change does n91 involve a significant increase in the probability or consequences of any accident previously evaluated.

Criterion 2- Does Not Create the Possibility of a New or DilTerent Kind of Accident from any Previously Evaluated.

The proposed change provides clarification to a specification which paraphrases a codified ,

requirement. Since the proposed amendment would not change the design, configuration or method of operation of the plant, it would nE create the possibility of a new or different kind of accident from any previously evaluated.

Criterion 3 - Does Not Involve a Significant Reduction in the Margin of Safety.

The proposed change is administrative and clarifies the relationship between the requirements of TS 4.4.1.1.4, Appendix J, and any approved exemptions to Appendix J. It does not, in itself, change a safety limit, an LCO, or a surveillance requirement on equipment required to l operate the plant. The NRC will directly approve change proposed exemption to III.D.I.(a) of Appendix J prior to implementation. Therefore, this change does not involve a significant reduction in the margin of safety.

The Commission has provided Guidance in 51 FR 7750 dated March 6,1986, concerning the application of these 10CFR50.92 standards by providing examples of amendments which are likely to involve ng significant hazards considerations. The proposed amendment most closely

, matches example (i): "A purely administrative change to the technical specifications; for example, a change to achieve consistency throughout the technical specifications, correction of an error, or a change in nomenclature."

{ 1 Therefore, based upon the reasoning presented above and the previous discussion of the amendment request, Entergy Operations has determined that the requested change does n91 involve a significant hazards consideration.

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