ML20196F925

Safety Evaluation Re Thrid 10-year Interval Inservice Insp Program Relief Requests for Plant

ML20196F925
Person / Time
Site:	FitzPatrick
Issue date:	11/25/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20196F922	List: ML20196F918 ML20196F925
References
NUDOCS 9812070160
Download: ML20196F925 (48)

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j NUCLEAR REGULATORY COMMISSION

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t WASHINGTON, D.C. Samas anni SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PLAN REQUESTS FOR RELIEF POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 i

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1.0 INTRODUCTION

Inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1, l

2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and l

Pressure Vessel (B&PV) Code and applicable addenda as required by 10 CFR 50.55a(g),

i except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(6)(g)(i).10 CFR 50.55a(a)(3) states that attematives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed altematives would provide an acceptable level of quarty and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first ten-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the James A. FitzPatrick Nuclear Power Plant third ten-year inservice inspection (ISI) interval is the

.1989 Edition.

2.0 EVALUATION On January 26,1998, the Power Authority of the State of New York (the licensee, also known as the New York Power Authority) submitted the third ten-year interval inservice inspection program plan requests for relief for the James A. FitzPatrick Nuclear Power Plant. The licensee submitted additional information on September 21,1998, in response to an NRC request dated August 20,1998. The licensee has requested that staff review certain requests s

Enclosure i

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for relief for the 1998 fall outage. These requests for relief are discussed evaluated below.

The remaining requests for relief will be addressed in a separate safety evaluation.

The staff, with technical assistance from its contractor, the Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its third ten-year interval inservice inspection program request for relief for FitzPatrick. Based on the results of the review, the staff adopts the contractor's conclusions and recommendations presented in the attached Technical Letter Report (TLR).

Request for Relief No.' 1: ASME Code,Section XI, IWA-4700(a) requires that a system hydrostatic test be performed in accordance with IWA-5000 after repairs by welding on the pressure retaining boundary.

Pursuant to 10 CFR 50.55a(a)(3), the licensee has proposed to use Code Case N-416-1. The licensee stated.

The alternative examination req.Jrements will be implemented as defined by ASME Section XI Code Case N-416-1, Altemative Pressure Test Requirements for Welded Repairs or Installation of Replacement items by Welding, Class 1,2, and 3 Section XI, Division 1 with the following exceptions:

(1) Currently Draft RG 1.147 Rev.12, dated May 1997, imposes conditions in addition to those conditions specified in the Code Case, Additional surface examinations should be performed on the root (pass) layer of butt and socket welds of the pressure retaining boundary of Class 3 components when the surface examination method is used in accordance with Section 111.

In their September 21,1998 response to the NRC request for additional information, the licensee stated:

The Authority confirms and commits that a surface examination of the root pass weld layer for welded repairs and replacements in Class 3 systems, if required by ASME Section til for final acceptance, will be performed. The requirement to perform root pass weld layer inspections has been incorporated in Administrative Procedure AP-19.02, ' Post-Work Pressure Testing and Visual Inspection Requirements'.

The NRC staff understands that the licensee will perform a surface examination of the root pass weld layer for welded repairs and replacements in Class 3 systems, if a surface examination of the completed weld is required by ASME Section ill for final acceptance.

(2) Code Case N-416-1 requires NDE to be performed in accordance with the methods and acceptance criteria of the applicable Subsection of the.1992 Edition of Section Ill. Currently the volumetric examination is designated as radiography.

The Authority intends to utilize, radiographic or ultrasonics as the volumetric method, PT or MT as the surface method, the tachniques and Acceptance

. Standards of these examinations will be in accordance with the 1989 Edition of Section XI or the 1992 Edition of Section 111. Visual examination will be conducted in accordance with the 1989 or 1992 Edition of Section XI.

l Section XI c / the Code requires a system hydrostatic test to be performed in accordance with IWA-5000 after repairs by welding on the pressure-retaining boundary. The licensee proposed l

to implement the attemative to hydrostatic pressure tests contained in Code Case N-416-1 for Code Class 1,2, and 3 repairs / replacements. In addition, the licensee will supplement the pressure test with an additional surface examination on the root pass layer of Class 3 repair /

replacement welds or welded areas.

Code Case N-416-1 specifies that nondestructive examination (NDE) of the welds be performed in accordance with the applicable subsection of the 1992 Edition of Section Ill. In lieu of a hydrostatic test performed at pressure (s) above normal operating test pressure (s), this Code Case allows a VT-2 visual examination to be performed at nominal operating pressure and temperature in conjunction with a system leakage test, in accordance with paragraph IWA-5000 of the 1992 Edition of Section XI. The licensee is proposing to perform the VT-2 visual examination in accordance with the requirements of the 1989 Edition of Section XI.

Comparison of the system pressure test requirements of the 1992 Edition of Section XI to those of the 1989 Edition, the latest Code edition referenced it' 10 C ~R 50.55a, shows that:

The test frequencies and pressure conditions are unchanged; The hold times either remained the same or increased; The terminology associated with the system pressure test requirements for all three Code classes has been clarified and streamlined; and

+ The NDE requirements for welded repairs remain the same.

Hydrostatic testing only subjects the piping components to e small increase in pressure over the design pressure and, therefore, does not present a sigrdtNnt challenge to pressure boundary integrity. Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leak detection during the examination of components under pressure rather than as a measure of the structuralintegrity of the components.

Following welding, the Code requires volumetric examination (depending on wall thickness) of repairs or replacements in Code Class 1 and 2 piping components, but may only require a surface examination of the final weld pass in Code Class 3 piping. There are no ongoing NDE requirements for Code Class 3 components except for VT-2 visual examination for leaks in conjunction with the 10-year hydrostatic tests and the periodic pressure tests. However, the staff determined that the examinations required by Code Case N-416-1 are commensurate for Class 3 systems when 1) a surface examination is performed on the root pass layer of butt and socket welds, and 2) a system pressure test at nominal operating pressure is performed.

Considering the previous acceptance of Code Case N-416-1 by the NRC at FitzPatrick for the second 10-year ISI interval, and the revision of the attemative to include the performance of supplemental surface examination on the root pass for Class 3 systems, it is concluded that the licensee's proposed attemative will provide an acceptable level of quality and safety for the third 10-year ISI interval by providing reasonable assurance of structuralintegrity. Therefore, the staff concludes that the licensee's proposed attemative, to use Code Case N-416-1 with a l

4 supplemental surface examination on the root pass layer of butt and socket welds for Class 3 components, is authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of the Code Case is authorized for the current interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this Code Case, the licensee should follow all provisions in Code Case N-416-1 with limitations issued in Regulatory Guide 1.147, if any.

Request for Relief No. 2: ASME Code,Section XI, Examination Category F-A, item Nos.

F1.10-F1.70 require visual examinations (VT-3) each inspection interval, as defined by Figure I

IWF-1300-1, of Class 1,2, and 3 component supports. Supports selected for examination l

shall be as specified in IWF-2510.

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j Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N-491-1, for the l

examination of Class 1,2,3, and MC Component Supports.

The licensee has proposed the use of Code Case N-491-1. A similar Code Case N-491 was l

approved for use Regulatoiy Guide 1.147, inservice inspection Code Case Acceptability l

ASME Section XI, Division 1, Revision 11, dated October 1994. The licensee has performed a l

review / comparison of Code Case N-491 and N-491-1. The comparison revealed that the only change from Code Case N-491 to N-491-1 is paragraph 1220, Snubberinspection Requirements, which specifies, "The inservice inspection requirements for snubbers shall be in accordance with the Section XI Edition and Addenda specified in the Owner's inservice Inspection Program", when previously, Code Case N-491 stated, " inservice inspection requirements for snubbers shall be in accordance with the requirements of IWF-5000."

This change does not affect the selection or examination rules specified in Code Case N-491-

1. Therefore, based on the conclusion that Code Case N-491-1 is essentially identical to Code Case N-491 for component supports, the staff concludes that the licensee's proposal to use Code Case N-491/ provides an acceptable level of quality and safety and is authorized pursuant to 10 CFR 00.55a(a)(3)(i). The use of the Code Case is authorized for the current interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this Code Case, the i

licensee should follow all provisions in Code Case N-491-1 with the limitations issued in Regulatory Guide 1.147,if any.

Request for Relief No. 3: ASME Code,Section XI, Table IWB-2500-1, Examination Category B-P, Table IWC-2500-1, Examination Category C-H, and Table IWD-2500-1, Examination Categories D-A, D-B and D-C, require system hydrostatic testing of pressure-retaining components in accordance with IWA-5000 once each 10-year interval.

Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee has requested authorization to use Code Case N-498-1, Attemate Rules for 10-YearHydrostatic Pressure Testing for Class 1, 2, and 3 Systems,Section XI, Division 1.

i The Code requires the performance of a system hydrostatic test once per interval in j

accordance with the requirements of IWA-5000 for Class 1,2, and 3 pressure-retaining systems. In lieu of the Code-required hydrostatic testing requirements, the licensee has i

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. requested authorization to use Code Case N-498-1, Altemative Rules for 10-Year System Hydmstatic Testing for Class 1, 2, and 3 Systems, dated May 11,1994.

The system hydrostatic test, as stipulated in Section XI, is not a test of the structuralintegrity of I

the system but rather an enhanced leakage test. Hydrostatic testing only subjects the piping components to a smallincrease in pressure over the design pressure; therefore, piping dead l

weight, thermal expansion, and seismic loads present far greater challenges to the structural l

integrity of a system. Consequently, the Section XI hydrostatic pressure test is primarily l

regarded as a means to enhance leak detection during the examination of components under pressure, rather than as a method to determine the structuralintegrity of the components. In addition, the industry experience indicates that leaks are not being discovered as a result of l

hydrostatic test pressures causing a preexisting flaw to propagate through the wall. In most cases leaks are being found when the system is at normal operating pressure.

Code Case N-498, Altemative Rules for 10-Year System Hydrostatic Testing for Class 1 and 2 Systems, was previously approved for general use for Class 1 and 2 systems in Regulatory Guide 1.147, Rev.11. For Class 3 systems, Revision N-498-1 specifies requirements identical to those for Class 2 components (for Class 1 and 2 systems, the attemative requirements in N-498-1 are unchanged from N-498). In lieu of 10-year hydrostatic pressure testing at or near the end of the 10-year interval, Code Case N-498-1 requires a VT-2 visual examination at nominal operating pressure and temperature in conjunction with a system leakage test performed in accordance with Paragraph IWA-5000 of the 1992 Edition of Section XI.

Class 3 systems do not normally receive the amount and/or type of nondestructive examinations that Class 1 and 2 systems receive. While Class 1 and 2 system failures are relatively uncommon, Class 3 leaks occur more frequently and are caused by different failure mechanisms. Based on a previous review of Class 3 system failures requiring repair, the most common causes of failures are erosion-corrosion (EC), micro biologically-induced corrosion (MIC), and general corrosion. In general, licensees have implemented programs for the I

prevention, detection, and evaluation of EC and MIC; therefore, Class 3 systems receive inspection commensurate with their functions and expected failure mechanisms.

System hydrostatic testing can entail considerable time and radiation dose. The safety assurance provided by the enhanced leakage gained from a slight increase in system pressure during a hydrostotic test may be offset or negated by the necersity to gag or remove Code safety and/or relief valves (placing the system and the plant ir, an off-normal state), erect temporary supports in steam lines, and expend resources to set up testing with special i

equipment and gages. Therefore, performance of system hydrostatic testing represents a considerable burden for the licensee.

Giving consideration to the minimal amount of increased assurance provided by hydrostatic test pressures versus system leakage test pressures, the assurance of leak-tight integrity provided by the proposed attemative, and the hardship associated with performing the hydrostatic test, the staff finds that compliance with the Section XI hydrostatic testing requirements results in hardship and/or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, the staff concluded that the use of Code Case N-498-1 for Code Class 1,2, and 3 systems is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

The use of this Code Case is authorized for the current interval at the James A. FitzPatrick i

. l Nuclear Power Plant, or until the Code Case is approved for general use by reference in l

Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

Request for Relief No. 4: ASME Code,Section XI, requires examination of integrally-welded attachments as specified for Examination Categories B-H, B-K, C-C, D-A, D-8, and D.C. The Code stipulates volumetric or surface examinations, as appropriate, and the extent of i

examinations.

Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee has proposed the use of Code Case N-509,

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Attemate Rules 16r the Selection and Examination of Class 1, 2, and 3 Integrally Welded l

Attachments,Section XI, Division 1, in lieu of the requirements of the Code for Class 1,2, and j

i 3 integrally-welded attachments. The licensee stated:

i The following attemative examination requirements will be implemented as defined by ASME Section XI Code Case N-509, Altemative Rules for the Selection and Examination of Class 1,2, and 3 Integrally Welded Attachments l

Section XI, Division 1. In addition to those conditions specified in Code Case N-l 509: A minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined, with the following exception:

1) Exarnination Category and item No.(s) for Class 3 Integrally Welded Attachments are defined in accordance with the ASME Section XI,1989 Edition, Article IWD, Table IWD-2500-1.

In the September 21,1998 response to the NRC request for additional information, the licensee stated:

The Authority confirms and commits to include a minimum of 10% of the total number of non-exempt piping, pump, and valve integral attachments distributed among all Class 1,2, and 3 systems.

The licensee has proposed to apply the requirements of Code Case N-509 as an attemative to the Code requirements for the examination of integrally-welded attachments on Class 1,2, and 3 piping and components. The licensee's altemative includes a commitment to supplement the Code Case with a minimum sample of 10% of the total number of non-exempt piping, l

pump, and valve integral attachments distributed among all Class 1,2, and 3 systems. The l

licensee has also proposed that the examination category and item no.(s) for Class 3 integrally

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i welded attachments be defined in accordance with the Examination Category and item No.(s) i l

for Class 3 Integrally Welded Attachments defined in accordance with the ASME Section XI, 1989 Edition, Article IWD, Table IWD-25001.

Table 2500-1, in Code Case N-509, classifies the welded attachments under Pressure Vessels, Piping, Pumps, and Valves. Whereas Table IWD-2500-1, Examination Category D-A, item Nos. D1.20 - D1.60 in ASME Section XI,1989 Edition, classifies the welded attachments under Component Supports and Restraints, Mechanical and Hydraulic Snubbers, Spring Type Supports, Constant Load Type Supports, and Shock Absorbers Considering that the licensee proposed attemative will supplement the Code Case with a minimum sample of 10% of the i

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. total number of non-exempt piping, pump, and valve integral attachments distributed among all Class 1,2, and 3 systems, the staff concludes that the proposed use of the 1989 Edition, Article IWD, Table IWD 2500-1 for the classification of exarnination category and item numbers for Class 3 integral attachments provides an acceptable level of quality and safety by providing assurance of structuralintegrity. The Code examir*.ation requirements are based on sampling to ensure the detection of service-induced degradation, extending the sampling philosophy to the integral attachment welds will provide an equivalent level of quality and safety. Therefore, the staff concluded that the proposed altemative is authorized pursuant to j

10 CFR 50.55a(a)(3)(i) for the current interval at James A. FitzPatrick Nuclear Power Plant, or until Code Case N-509 is approved for general use by reference in Regulatory Guide 1.147.

After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

Request for Relief No. 5: ASME Code,Section XI, Examination Category B-J requires that volumetric and/or surface examinations be performed on longitudinal welds as defined by i

Figure IWB-2500-8. The examination includes at least a pipe-diameter length but not more l

than 12 in. of each longitudinal weld intersecting the circumferential weld required to be examined by Examination Categories B-J and B-F. Examination Categories C-F-1 and C-F-2 requires 100% volumetric and/or surface examination, as defined by Figure IWC-2500-7, -12 and -13, for 2.5t of each longitudinal weld intersecting circumferential welds examined.

In accordance with 10 CFR 50.55a(a)(3), the licensee proposed the use of Code Case N-524

'Altemative Examination Requirements forLongitudinal Welds in Class 1 and Class 2 Piping" in lieu of the Code requirements.

ASME Section XI requires the examination of one pipe diameter, but not more than 12 inches, of Class i longitudinal piping welds. For Class 2 piping welds, the length of longitudinal weld required to be examined is 2.5 times the pipe thickness. These lengths are measured from the intersection with the circumferential weld. The licensee's proposed attemative is to examine only the portions of longitudinal weld within the examination area of the intersecting circumferential weld in accordance with Code Case N-524, Altemative Examination Requirements for Longitudinal Welds in Class 1 and Class 2 Piping.

Longitudinal welds are produced during the manufacture of the piping, not in the field as is the case for circumferential welds. Consequently, longitudinal welds are fabricated under strict manufacturing standards, which provides assurance of structuralintegrity. These welds have also been subjected to the preservice and initial inservice examinations, which provide i

additional assurance of structuralintegrity. No significant loading conditions or material degradation mechanisms have been identified to date that specifically relate to longitudinal seam welds in nuclear plant piping. The most critical region of the longitudinal weld is the portion that intersects the circumferential weld. Since this region will be examined during the examination of the circumferential weld, the staff concludad that the licensee's proposed altemative provides an acceptable level of quality and safety. The staff also concludes that the licensee's proposed altamative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the current interval at James A. FitzPatrick Nuclear Power Plant, or until Code Case N-524 is i

approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee j

must follow the conditions, if any, specified in the regulatory guide.

j Request for Rolief No. 6:

. Q.qge Raouirement:

IWA-4800 The records required by IWA-6000 shall be completed for all repairs.

lWA4210(c)

The Owner shall prepare inservice inspection summary report for l

Class 1 and 2 pressure retaining components and their supports.

lWA4220(c)

Inservice inspection summary reports shall be required at the completion of each inspection conducted during a refueling outage. Examinations, tests, replacements, and repairs conducted since the preceding summary report shall be included.

l IWA-6220(d)

Each summary report shall contain the following:

(1) refueling outage number (when applicable);

(2) Owner's Report for Inservice inspection, Form NIS-1; and (3) Owners Report for Repair or Replacement, Form NIS-2.

l IWA-6230 Within 90 days of the completion of the inservice inspection i

conducted during each refueling outage, the Owner shall file ISI Summary Reports with the enforcement and regulatory i

authorities.

Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N-532, "Altemative Requirements to Repair and Replacement Documentation Requirements and inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000, Division 1," as altemative requirements to repair and replacement documentation requirements and inservice summary report preparation and submission as required by IWA-4000, and IWA-6000. The staff reviewed the proposed attemative documentation requirements of Code Case N-532 and determined that although the required forms have changed, the information required by the Code is available. Code Case N-532 would require preparation of the Repair / Replacement Certification Record, Form NIS-2A. The completed i

form NIS-2A shall be certified by an Authorized Nuclear Inservice inspector (ANil) as defined in l

ASME Code,Section XI, IWA-2130 and shall be maintained by the Owner. Furthermore, the Owners Activity Report Form, OAR-1 shall be prepared and certified by an ANil upon completion of each refueling outage. The OAR-1 form snail contain an abstract of applicable l

examinations and tests, a list of item (s) with flaws or relevant conditions that require evaluation I

L to determine acceptability for continued service, and an abstract of repairs, replacements and I

corrective measures perforTned as a result of unacceptable flaws or relevant conditions.

I Hence, the information provided in the documentation pertaining to the use of Code Case N-532, can be used in the same manner to assess the safety implications of Code activities I

performed during an outage. A review using the information as prescribed by the Code Case

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will, therefore, provide the same or improved level of quality and safety as reviews that may be conducted using the Code reporting requirements. Therefore, the staff concludes that the use L

of the licensee's attemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the current l

interval at James A. FitzPatrick Nuclear Power Plant, or until Code Case N-532 is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

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Request for Relief No. 7 This relief request was withdrawn by the licensee in its letter dated September 21,1998.

Request for Relief No. 8: Examination Category C-H, All Pressure Retaining Components The licensee in its letter dated September 21,1998 stated, that this relief was not required for this outage and it will provide additional information conceming this request for relief within 90 days of its September 21,1998 response to the NRC request for additional information. This relief will be addressed in a separate safety evaluation after the licensee's submittal.

Request for Relief No. 9, Rev.1: 10 CFR 50.55a(g)(6)(ii)(B), requires that the containment sections (IWE and IWL) of the subject plants' ISI programs are developed in accordance with the requirements of the 1992 Edition,1992 Addenda of Section XI. IWA-2210 requires specific minimum illumination and maximum direct examination distance for visual examinations as specified in Table IWA-2210-1. IWA-2211 requires that VT-1 examinations l

be conducted to detect discontinuities and imperfections on the surfaces of components, including such conditions as cracks, wear, corrosion or erosion.

lWA-2213 requires that VT-3 examinations be conducted to determine the general mechanical and structural condition of components and their supports by verifying parameters such as clearances, settings, and physical displacements; and to detect discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear, or erosion. VT-3 includes examinations for conditions that could affect operability or functional adequacy of snubbers and constant load and spring type supports.

IWA-2300 requires in that personnel performing nondestructive examinations be qualified and certified under specified conditions as listed in lWA-2300.

l In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed the use of the ASME Section XI 1989 Edition visual examination requirements for use with Subsections IWE. The l

licensee stated:

The ASME Section XI 1989 Edition (no addenda) VT-1, VT-3 and Supplemental Examination (surface and volumetric per IWE) requirements, for ' Examination Methods'(IWA-2200) and ' Qualifications of Nondestructive Examination Personne!' (IWA 2300) will ba applied to IWE required examinations. These are l

the same requirements that apply to all other safety related components in the ISI program and will be supplemented with indoctrination overview training by FitzPatrick's site Level lli (who has attended an EPRI IWE training course) covering the IWE examination requirements prior to personnel performing IWE VT-1 and VT-3 examinations. The results of examinations performed under this proposed alternative are considered equivalent to the results that would be obtained under the 1992 Edition of ASME Section XI.

Recently an EPRI training course was given to (licensee) Engineering /ISI personnel and [ licensee) QC Level 11 and ill examiners. This two-day course provided an overview and guidance for containment inspections, coatings inspections and engineering evaluations. This ensures that QC and i

i Engineering personnel have received specific training on containment inspections.

l This attemative will apply until the beginning of Refuel Outage (RO14 currently scheduled to begin in October 2000.

To comply with the expedited containment examination requirements of 10 CFR l

50.55a(g)(6)(ii)(B), licensees must perform visual examinations on Class MC and Metallic i

Liners of Class CC Concrete Components of Light-Water Cooled Plants per the requirements of IWE, and visual examinations on Class CC Concrete Components of Light-Water Cooled Plants per the requirements of IWL. These examinations are to be performsd to the requirements of the 1992 Edition 1992 Addenda of ASME XI. The licensee has proposed the l

use of IWA-2200 and IWA-2300, VT-1 and VT-3 requirements from the 1989 Edition of ASME l

Section XI in lieu of the 1992 Edition,1992 Addenda requirements.

The staff has performed a review of the subject requirements found in the 1989 Edition and the 1992 Edition,1992 Addenda of ASME XI. The essential differences found during the review consisted primarily of the methods used to determine if satisfactory lighting and distances existed to perform the examinations. The 1992 Code requires that a quantitative i

approach is used to measure illumination levels, whereas the 1989 Code requires that a 18%

neutral gray card with a 1/32 inch black line be used to determine acceptable examination conditions.

While the methods for determination of acceptable illumination are different from 1989 to 1992, the staff determined that both methods provide an acceptable approach to illumination determination by providing adequate conditions for acceptable examinations. Differences were also noted in the methods used to perform visual acuity examinations of NDT personnel, however, the requirement for visual acuity exists in both the 1969 and 1992 Codes.

Considering that the 1989 Code requirements for visual examination are being used for all of the other inservice inspection examination requirements, including the Reactor Pressure Vessel visual examinations, the staff concludes that the licensees proposed attemative provides an acceptable level of quality and safety and is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Request for Relief No.10: The 1989 Edition of Section XI, IWA-5250(a)(2) requires that if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100.

In accordance with 10 CFR 50.55a(a)(3), the licensee has proposed an attemative to the requirements of IWA-5250(a)(2). The licensee stated:

The ISI Program at JAF will specify conducting visual (VT-3) examination in-place on bolted connections when leakage occurs. Evidence of degradation to the bolted connection or bolting shall require the bolt closest to the source of the leakage to be removed visually examined (VT-3) and evaluated to !WA-3100.

When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA 3100.

. In the September 21,1998 response to the NRC request for additional information the licensee also stated:

An engineering evaluation will be performed to identify the most prudent course of action to determine the condition of the bolting and/or the root cause of the leak when leakage is detected at a bolted connection.

A situation may be encountered that involves a leaking joint fo!!owing the complete replacement of bolting materials. The root cause of the leak may be thermal expansion of the piping and bolting materials due to system heatup. In such a case, re-torquing the joint bolting usually stops the leak. Removal of the joint bolting to evaluate it for corrosion would be a hardship and unwarranted if the bolting material is new or underwent VT-3 examinations prior to installation.

ASME Section XI interpretation XI-1-92-01 recognizes this situation as one in which the requirements of IWA-5250(a)(2) do not apply. Other situations which could also result in hardship include:

bolted connections which require a specific torque pattem, primary containment isolation valves which require stroke time testing and local leak rate testing following reassembly, components which would require extension disassembly to comply with the requirements of IWA-5250(a)(2) due to equipment clearances IWA-5250(a)(2) does not include other factors which may indicate the condition of the bolted connection. Other factors that will be considered in the Authority's evaluation of leakage at bolted connections will include, but not necessarily be limited to:

bolting materials corrosiveness of the process fluid leakage location history of leakage at the connection or other system components visual evidence of corrosion at the connection (while connection is assembled) service age of the bolting materials.

Training on the use of the above leakage evaluation criteria, as well as the appropriate corrective actions to be taken if the evaluation is inconclusive or identifies bolting degradation, will be provided to appropriate ISI, Operation Review Group, and Quality Assurance personnel prior to the use of the proposed attemative. The leakage evaluation criteria and appropriate corrective

I l actions will be formally proceduralized within 6 months following completion of the up coming refuel outage.

in a letter dated November 3,1998, the licensee further described the VT-3 examination process that will be utilized as part of this request for relief. The licensee stated:

Relevant conditions required to be identified during the VT-3 examination include erosion, corrosion, wear, and crack-like orlinear flaws. Localized general corrosion that reduces the bolt or stud cross sectional area by more than 5% is recorded in the VT-3 non-destructive examination report.

In accordance with the 1989 Edition of ASME XI, if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100. In lieu of this requirement, the licensee has proposed to conduct a visual (VT-

3) examination in-place on bolted connections when leakage occurs. An evaluation of the botting will be performed to determine its susceptibility to corrosion. This evaluation considers a number of parameters, including botting materials, the potential for corrosion, and visual evidence of corrosion with the bolting in place. If evidence of degradation exists, the bolt closest to the source of the leakage will be removed, visually examined (VT-3), and evaluated to IWA-3100. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100.

lWA-3100 invokes the use of subparagraphs IWB-3000, IWC-3000, IWD-3000 for Class 1, 2, and 3 pressure retaining components respectively. However, none of these subparagraphs provide an acceptance criteria for VT-3 examinations. Therefore, the ability to perform a meaningful evaluation on the bolting without an applicable acceptance criteria may be inappropriate. The staff determined that a VT-1 visual examination utilizing the acceptance criteria defined in IWB-3000 provides a more appropriate method of examination of the subject botting than a VT-3 visual examination. However, the licensee provided a description of the VT-3 examination process that will be utilized as part of this request for relief. The VT-3 examination will include the identification of erosion, corrosion, wear, and crack-like or linear flaws. Also, localized general corrosion that reduces the bolt or stud cross sectional area by more than 5% will be recorded in the VT-3 examination report. The VT-3 visual examination as stated by the licensee closely parallels the Code VT-1 visual examination standards described in IWB-3517.1 conceming degradation related to leakage at bolted connections.

Similar requests for relief have been approved with the condition that a VT-1 visual examination be performed using the acceptance criteria for VT-1 examinations. Based on the determination that the licensee's VT-3 visual examination closely parallels the VT-1 visual examination standards conceming degradation related to leakage at bolted connections, and that the evaluation process proposed by the licensee provides a sound engineering approach, it is concluded that an acceptable level of quality and safety is provided, but for an interim period.

in order to clarify the acceptance standards used for future examinations and to promote i

consistent use of the subject attemative, the staff has determined that VT-1 visual examinations should be used for all bolting at the James A. FitzPatrick Nuclear Power Plant. It is recognized that imposition of this requirement could negatively impact the current outage, but should be implemented for future inspections. Therefore, the staff concluded that the a

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. licensee's proposed altemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for 1 year from the date of the letter forwarding this safety evaluation.

Request for Relief No.12: ASME Code,Section XI, Table IWC-2500-1, Examination Category C-H, requires a VT-2 visual examination during the system leakage test (IWC-5221) once each period and during the system hydrostatic test (IWC-5222) once each interval for all Class 2 pressure retaining components.

Pursuant to 10 CFR 50.55a(a)(3), the licensee proposed to implement Code Case N-522, Pressure Testing of Containment Penetration Piping. The licensee stated:

The following altamative examination requirements will be implemented as defined by ASME Section XI Code Case N-522, Pressure Testing of Containment Penetratiors Piping,Section XI, Division 1. In addition to the requirements specified in Code Case N-522, the following conditions shall be met:

The test should be conducted at the peak calculated containment pressure and the test procedure should permit the detection and location of through-wall leakage in Containment isolation Valves (CIVs) and pipe segments between the CIVs.

The Code requires that a VT-2 visual examination be performed during system pressure testing for all Class 2 pressure-retaining piping, including those segments that penetrate primary containment. As an attemative, the licensee proposed to implement the requirements of Code Case N-522, Pressure Testing of Containment Penetration Piping. Code Case N-522 specifies that 10 CFR 50, Appendix J testing may oe used as an attemative to Section XI pressure tests, for certain containment penetration piping.

The subject piping is fabricated and designated as Class 2 only because it penetrates the primary reactor containtr.ent and is considered an extension of the containment vessel. Since the piping on either side of these penetrations is non-classed, the requirements of Appendix J are more appropriate than those of Examination Category C-H. Appendix J pressure tests verify the leak-tight integrity of the primary reactor containment and of systems and components that penetrate containment by local leak rate and integrated leak rate tests.

The Class 2 containment isolation valves (CIVs) and connecting pipe segments must withstand the peak calculated containment intemal pressure. The pressure retaining integrity of the CIV's and connecting piping and their associated safety functions may be verified with an Appendix J test conducted at the peak calculated containment pressure. The licensec has proposed to perform the Appendix J testing at no less than the peak calculated containment pressure and will use procedures and techniques capable of detecting and locating through-wall leakage in the pipe segments between the CIV's.

Appendix J, Option A-Prescriptive Requirements, requires that three Type A tests be performed at approximately equal intervals during the 10 year ISI interval, with the third test being done while shutdown for the 10-year plant ISI. Option A also requires Type B and C

. 4 tests be performed during each refueling outage, but in no case at intervals greater than 2 years. This is more frequent than the period:c pressure tests required by ASME Section XI.

Appendix J, Option B-Performance Based Requirements, allows a licensee to perform Type A, B, and C tests at frequencies related to the safety significance and historical performance of the system's isolation capabilities. This could, in effect, allow only one test to be performed during the 10-year ISI interval. However, the staff's position, as stated in Regulatory Guide 1.163 Performance-based Containment Leak-Test Program, is that the licensee is to establish test intervals of no greater than 60 months for Type C tests because of uncertainties (particulariy unquantified leakage rates for test failures, repetitive / common mode failures, and aging effects) in historical Type C component performance data. While this five-year limit results in an increased time between testing over that required by Section XI (forty months), it is believed that Appendix J tests are more appropriate and provide reasonable assurance of j

the continued operability of containment penetrations. Therefore, the staff determined that the test frequencies associated with Appendix J, Option A (Type A, B or C) or Option B (Type C)

Tests are appropriate for the corresponding pressure test frequencies required by the Code.

The licensee has chosen to use Option B for its Appendix J Tests. However, the licensee has stated in a telephone conversation with the staff that the examination of the subject components will be performed only when repair or replacement occurs. Therefore, the staff recognizes that the licensee test frequencies may exceed the 60 month interval for a Type C test as stated in Regulatory Guide 1.163. The staff concluded that the licensee's proposed test frequency is inappropriate due to uncertainties with repetitive / common mode failures, and aging effects in Type C components.

Based on the above analysis and information submitted, the staff concludes that an acceptable level of quality and safety will be provided by the licensee's proposed attemative for a period of 60 months. Therefore, the licensee's proposed attemative to implement Code Case N-522 is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for a limited time not to exceed 60 months from the date of the letter forwarding this safety evaluation, or until such time as Code Case N-522 is published in a future revision of Regulatory Guide (RG) 1.147. At that time, if the licensee intends to continue to implement Code Case N-522, the licensee should follow all provisions in Code Case N-522 with limitations issued in RG 1.147, if any.

Request for Relief No.13: ASME Code,Section XI, IWA-4400(a) requires that all welding shall be performed in accordance with Welding Procedure Specifications that have been qualified by the Owner or repair organization in accordance with the requirements of the codes specified in the Repair Program in accordance with IWA-4120.

In accordance with 10 CFR 50.55a(a)(3)(i), the licensee has proposed to use Code Case N-573, Transfer of Procedure Qn!!! tation Records Between Owners. The licensee stated:

The following attemative te: in 1 requirements will be implemented as defined by ASME Section XI Code Caw h-573, Transfer of Procedure Qualtr. cation Records [PQR] Between Owners,Section XI, Division 1.

1.

[The licensee) will perform a technical review of the supplying Owner's PQR 4

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. 2.

The supplying Owner will state in writing that the PQR was pe formed under an acceptable Nuclear Quality Assurance program that meets ASME Section XI, lWA-1400 and that it was performed in accordance with ASME Section IX.

3.

[The licensee) will generate a picensee) WPS [ welding procedure specification) using the Variables establishoc'in the supplied PQR(s). [The licensee) PQR's may supplement these or other Owner supplied PQR's.

4.

The WPS will be approved and signed by [the licensee).

5.

The WPS will be demonstrated successfully by [the licensee) by completing a welder performance qualification test using the parameters of the [ licensee) WPS.

6.

[The licensee) will not transfer the supplied PQR to any other Owner.

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[the licensee) will document the use of this Code Case on the appropriate NIS-2 form.

I IWA-4400(a) requires that all welding shall be performed in accordance with Welding Procedure Specifications (WPS) that have been qualified by the Owner or repair organization in accordance with the requirements of the codes specified in the Repair Program, per IWA-4120. The licensee has proposed the use of Code Case N-573, Transfer of Procedure Qualification Records Between Owners. This Code Case essentially allows the use of a welding or brazing procedure qualification record (PQR) qualified by sne owner to be used by another owner for the development of the WPS. The specific requirements listed in Code Case N-573 shall be met by the Owner that performed the procedure qualification, and by the Owner intending to use the PQR. These requirements are:

(a) The Owner that performed the procedure qualification test shall certify, by signing the PQR, that testing was performed in accordance with Section IX.

(b) The Owner that performed the procedure qualification test shall certify, in writing, that the procedure qualification was conducted in accordance with a Quality Assurance Program that satisfies the requirements of IWA-1400.

(c) The Owner accepting the completed PQR shall accept responsibility for obtaining any additional supporting information needed for WPS development.

(d) The Owner accepting the completed PQR shall document, on each resulting WPS, the parameters applicable to welding. Each WPS shall be supported by all necessary PQR's (e) The Owner accepting the completed PQR shall accept responsibility for the PQR.

Acceptance shall be documented by the Owner's approval of each WPS that references the PQR.

. (f) The Owner accepting the completed PQR shall demonstrate technical competence in application of the received POR by completing a performance qualification test using the parameters of a resulting WPS.

(g) The Owner may accept and use a PQR only when it is received directly from the Owner that certified the PQR.

i (h) Use of this Code Case shall be shown on the NIS-2 form documenting welding or brazing.

The staff concluded that qualification of a procedure for the purpose of joining materials by either welding or brazing may be performed by any Owner provided the applicable requirements for procedure qualification are maintained. The staff also determined that Owners may use procedures qualified by other Owners provided the conditions / requirements listed in Code Case N-573 are met. The licensee revised the altamative to comply with requirements specified in Code Case N-573. Therefore, the staff concludes that the licensee's proposed altemative provides an acceptable level of quality and safety by providing j

reasonable assurance of structuralintegrity. The use of this altemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the current interval at James A. FitzPatrick Nuclear Power Plant, or until Code Case N-573 is approved for general use by reference in Regulatory Guide 1.147.

j After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

3.0 CONCLUSION

The staff concludes that for Requests for Relief Nos. 1,2,4,5,6,9,12, and 13 the licensee's proposed altematives provide an acceptable level of quality and safety; therefore, the licensee's proposed altamatives are authorized for the third interval pursuant to 10 CFR 50.55a(a)(3)(i).

In addition, the staff concluded for Request for Relief No.10 that the licensee's proposed attemative provides an acceptable level of quality and safety, but for an interim period. In order to clarify the acceptance standards used for future examinations and to promote consistent use of the subject attemative, the staff has determined that VT-1 visual examinations should be used for all bolting at the James A. FitzPatrick Nuclear Power Plant in the future. The staff recognizes that imposition of this requirement would negatively impact the current outage. Therefore, the staff concluded that the licensee's proposed altemative contained in Request for Relief No.10 is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for a limited time, not to exceed 1 year from the date of letter forwarding this safety evaluation.

For Request for Relief No. 3, the staff concluded that the Code requirements would result in a hardship without a compensating increase in the level of quality and safety and the licensee's proposed attemative provides reasonable assurance of structuralintegrity of the subject components. Therefore, staff concluded that the licensee's proposed attemative is authorized pursuant to 10 CFR 50.55a(3)(ii).

Relief requests nos.1 - 6, and 13 relate to application of various Code Cases. As noted above, the use of these Code Cases is authorized for the current interval, or until such time as

- a Code Case is published in a future revision of Regulatory Guide 1.147. At that time, if the

t 17-licensee intends to continue to implement those Code Cases, the licensee should follow all provisions in the Code Cases with limitations issued in Regulatory Guide 1.147, if any.

Relief request no.12 proposes use of Code Case N-522 as an alternative to the ASME Code requirements. For this relief request, the staff concluded that the licensee's proposed alternative provides an acceptable level of quality and safety and is authorized pursuant to 10 CFR 50.55a(a)(3)(i) for a limited time not to exceed 60 months from the date of the letter fonwarding this safety evaluation, or until such time as Code Case N-522 is published in a l

future revision of Regulatory Guide (RG) 1.147. At that time, if the licensee intends to continue to implement Code Case N-522, the licensee should follow all provisions in Code Case N-522 i

with limitations issued in RG 1.147, if any.

Request for Relief No. 7 was withdrawn by the licensee in letter dated September 21,1998.

As stated in the letter dated September 21,1998, the licensee has determined that Request for Relief No. 8 will not be required for the current refueling outage. Therefore, additional information for completion of the review of this request for relief will be provided at a later date.

Principal Contributor: T. McLellan Date:

November 25, 1998

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TECHNICAL LETTER REPORT ON THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM EXPEDITED REQUESTS FOR RELIEF FOR NEW YORK POWER AUTHORITY JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NUMBER: 50-333 1.

INTROD11CTION By [[letter::JPN-98-002, Forwards Third ISI Interval Insp Program Plan for JAFNPP & Requests for Relief from Snubber Insp Requirements of ASME Section Xi,Article IWF-5000|letter dated January 26,1998]], the licensee, New York Power Authority, submitted their third interval 10-year inservice inspection (ISI) program, including requests for relief for the James A. FITZPATRICK (JAF) Nuclear Power Plant. The licensee requested that the review of certain requests for relief be expedited for implementation in an upcoming outage. Additionalinformation was submitted September 21,1998, in response to the Nuclear Regulatory Commission (NRC) request for additional information (RAl). The Idaho National Engineering and Environmental Laboratory (INEEL) staff's evaluation of the subject requests for relief is in the following section.

2.

EVALUATION The information provided by New York Power Authority in support of the requests for relief from Code requirements has been evaluated and the bases for disposition are documented below. The Code of record for the James A. FitzPatrick Nuclear Power Plant, third 10-year ISI interval, which began September 28,1997, is the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code.

A.

Reauest for Relief No.1. Use of Code Case N-416-1. Alternative Pressure Test Reauirements for Welded Reoairs or installation of Reolacement Items by Weldina.

Section XI. Division 1 Code Reauirement: IWA-4700(a) requires that a system hydrostatic test be performed in accordance with IWA-5000 after repairs by welding on the pressure retaining boundary.

Licensee's Proposed Alternative: Pursuant to 10 CFR 50.55a(a)(3), the licensee has proposed to use Code Case N-416-1. The licensee stated:

"The alternative examination requirements will be implemented as defined by ASME Section XI Code Case N-416-1, Alternative Pressure Test Requirements for Welded Repairs or Installation of Replacement items by Welding, Class 1,2, and 3 Section XI, Division 1 with the following exceptions:

"(1)

. Currently Draft RG 1.147 Rev.12, dated May 1997, imposes conditions in addition to those conditions specified in the Code Case, Additional surface examinations should be performed on the root (pass) layer of butt and socket welds of the pressure retaining boundary of Class 3 components when the surface examination method is used in accordance with Section Ill."

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, In their response to the NRC request for additional information the licensee stated:

l.

"The Authority confirms and commits that a surface examination of the root pass l

weld layer for welded repairs and replacements in Class 3 systems, if required by ASME Section lli for final acceptance, will be performed. The requirement to perform root pass weld layer inspections has been incorporated in Administrative Procedure AP-19.02, ' Post-Work Pressure Testing and Visual Inspection Requirements'."

The licensee provided clarification to this statement during a conference call on September 30,1998. The licensee confirmed that a surface examination of the root pass weld layer for welded repairs and replacements in Class 3 systems will be l

performed, if a surface examination of the completed weld is required by ASME Section ill for final acceptance.

"(2)

Code Case N-416-1 requires NDE to be performed in accordance with the methods and acceptance criteria of the applicable Subsection of the 1992 Edition of Section Ill. Currently the volumetric examination is designated as radiography.

"The Authority intends to utilize, radiographic or ultrasonics as the volumetric method, PT or MT as the surface method, the techniques and Acceptance Standards of these examinations will be in accordance with the 1989 Edition of Section XI or the 1992 Edition of Section Ill. Visual examination will be conducted in accordance with the 1989 or 1992 Edition of Section XI.

Licensee's Basis for Proposed Alternative:

'The need for relief from the hydrostatic test requirement for weld repairs and welded pipe replacements is sometimes unexpected as a result of system conditions requiring immediate repair, or repairs required prior to startup as identified during outage related Inspections. As a consequence, immediate communication with NRC for testing relief on a case-by-csse basis is necessary to avoid exceeding limiting conditions for operation or startup delays. This places an unnecessary burden on the Authority's JAF Plant and NRC resources.

" Draft RG 1.147, Rev.12, dated May 1997 includes Code Case N-416-1, Alternate Pressure Test Requirements for Welded Repairs or Installation of Replacement items by Welding, Class 1,2, and 3 Section XI, Division 1", this code case has not been published in Regulatory Guide 1.147, inservice Inspection Code Case Acceptability ASME Section XI, Division 1", however, the NRC staff has approved it's use at JAF during the 2*' ISI interval and other nuclear stations."

Evaluation: Section XI of the Code requires a system hydrostatic test to be performed in accordance with IWA-5000 after repairs by welding on the pressure-retaining boundary. The licensee proposed to implement the alternative to hydrostatic pressure tests contained in Code Case N-416-1 for Code Class 1,2,

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3-and 3 repairs / replacements. In addition, the licensee will supplement the pressure test with an additional surface examination on the root pass layer of Class 3 repair /

replacement welds or welded areas.

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Code Case N-416-1 specifies that nondestructive examination (NDE) of the welds be performed in accordance with the applicable subsection of the 1992 Edition of Section Ill. In lieu of a hydrostatic test performed at pressure (s) above normal operating test pressure (s), this Code Case allows a VT-2 visual examination to be I

i j

performed at nominal operating pressure and temperature in conjunction with a i

system leakage test, in accordance with paragraph IWA-5000 of the 1992 Edition of l-Section XI. The licensee is proposing to perform the VT-2 visual examination in accordance with the requirements of the 1989 Edition of Section XI. Comparison of the system pressure test requirements of the 1992 Edition of Section XI to those of the 1989 Edition, the latest Code edition referenced in 10 CFR 50.55a, shows that:

1

• The test frequencies and pressure conditions ere unchanged;

• The hold times either remained the same or increased;

• The terminology associated with the system pressure test requirements for all three Code classes has been clarified and streamlined; and

• The NDE requirements for welded repairs remain the same.

Hydrostatic testing only subjects the piping components to a small increase in pressure over the design pressure and, therefore, does not present a significant challenge to pressure boundary integrity. Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leak detection during the examination of components under pressure rather than as a measure of the structuralintegrity of the components.

Following welding, the Code requires volumetric examination (depending on wall thickness) of repairs or replacements in Code Class 1 and 2 piping components, but may only require a surface examination of the final weld pass in Code Class 3 d

piping. There are no ongoing NDE requirements for Code Class 3 components i

except for VT-2 visual examination for leaks in conjunction with the 10-year i

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. hydrostatic tests and the periodic pressure tests. However, the INEEL staff believes that the examinations required by Code Case N-416-1 are commensurate for Class 3 systems when 1) a surface examination is performed on the root pass 1

layer of butt and socket welds, and 2) a system pressure test at nominal operating l

pressure is performed.

Considering the previous acceptance of Code Case N-416-1 by the NRC at FitzPatrick for the second 10-year ISI interval, and the commitment to perform i

supplemental surface examination on the root pass for Class 3 systems, it is concluded that the licensee's proposed alternative will provide an acceptable level i

of quality and safety for the third 10-year ISI interval. Therefore, it is recommended that the licensee's proposed alternative, to use Code Case N-416-1 with a supplemental surface examination on the root pass layer of butt and socket welds for Class 3 components, be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of the Code Case should be authorized for the current interval or until such time as the Code Case is published in a future revision of Regulatory Guide 1.147.

At that time, if the licensee intends to continue to implement this Code Case, the licensee should follow all provisions in Code Case N-416-1 with limitations issued in Regulatory Guide 1.147, if any.

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B.

Reauest for Relief No. 2. Use of Code Case N-491-1. Altemative Rules for Examination of Class 1. 2. 3. and MC Comoonent Sucoorts of Liaht-Water Cooled Power Plants Code Reautrement: Examination Category F-A, item Nos. F1.10-F1.70 require l

visual examinations (VT-3) each inspection interval, as defined by Figure IWF-1300-1, of Class 1,2, and 3 component supports. Supports selected for examination shall be as specified in IWF-2510.

l Licensee's ProDosed Alternative: In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N-491-1, for the examination of Class 1,2,3, and MC Component Supports.

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-... -.. -.--.-.- -.-.--- _.-- - -..-- -~_.--- _

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Licensee's Basis for Prooosed Alternative:

At JAF during the 2"a Ten-Year Interval, component supports were selected for examination per ASME Code Case N-491, ' Alternative Rules for Examination of i

Class 1,2,3 and MC Component Supports of Light Water Cooled Power Plants,Section XI, Division 1', which is approved for use in Regulatory Guide 1.147, Rev

11. During the 3* Ten-Year Interval JAF intends to implement ASME code Case N-491-1, ' Alternative Rules for Examination Class 1,2,3 and MC Component j

Supports of Light-Water Cooled Power Plants,Section XI, Division 1'. A detailed i

review and comparison of Code Cases N-491 and N-491-1 reveals the only j

change is Paragraph 1220, Snubber inspection Requirements which specifies, the inservice inspection requirements for snubbers shall be in accordance with i

the Section XI Edition and Addenda specified in the Owner's Inservice inspection Program, when previously Code Case N-491 stated, inservice inspection 3

requirements for snubbers shall be in accordance with the requirements of IWF-j 5000. This change does not effect the selection or examination rules specified in Code Case N-491-1. The snubber inspection program at JAF will be conducted in accordance with Section XI Edition specified for the JAF Inservice Inspection Program along with ASME Code Cases and/or relief requests. ASME Code Case N-491-1 will be implemented at JAF for selection and examination of Class l

1,2,3 and MC Component Supports during the 3* Ten-Year Interval.

" Draft RG 1.147 dated May 1997, includes Code Case N-491-1, 'Altemative Rules for Examination of Class 1,2,3 and MC Component Supports of Light-water Cooled Power Plants,Section XI, Divicion 1', this Code Case has not been published in Regulatory Guide 1.147, inservice inspection Code Case Acceptability ASME Section XI, Division 1", however, the NRC staff has approved its use at other nuclear stations."

Evaluation: The licensee has proposed the use of Code Case N-491-1. Code Case N-491 has been approved for use by the NRC in Regulatory Guide 1.147, l;

Inservice Inspection Code Case Acceptability ASME Section XI, Division 1, Revision 11. The licensee has performed a review / comparison of Code Case N-491 and N-491-1. The comparison revealed that the only change from Code Case N-491 to N-491-1 is paragraph 1220, Snubberinspection Requirements, which specifies, "The inservice inspection requirements for snubbers shall be in accordance with the Section XI Edition and Addenda specified in the Owner's inservice inspection Program", when previously, Code Case N-491 stated,

" inservice inspection requirements for snubbers shall be in accordance with the requirements of IWF-5000'. This change does not effect the selection or examination rules specified in Code Case N-491-1. Therefore, based on the conclusion that Code Case N-491-1 is essentially identical to Code Case N-491 for component supports, and that Code Case N-491 was previously approved for use

6-l by the NRC in Regulatory Guide 1.147, it is determined that the licensee's proposal i

to use Code Case N-491-1, provides an acceptable level of quality and safety and should be authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of the Code Case should be authorized for the current interval or until such time as the Code j

Case is published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement this Code Case, the licensee should follow all provisions in Code Case N-491-1 with the limitations issued in Regulatory Guide 1.147,if any.

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Reauest for Relief No. 3. Reauest for Authorization to Use Code Case N-498-1.

Alternate Rules for 10 Year Hydrostatic Pressure Testina for Class 1. 2. and 3 Systems.Section XI. Division 1 i

Code Reauirement: Table IWB-2500-1, Examination Category B-P, Table i

IWC-2500-1, Examination Category C-H, and Table IWD-2500-1, Examination Categories D-A, D-B and D-C, require system hydrostatic testing of pressure-retaining components in accordance with IWA-5000 once each 10-year interval.

l Licensee's Prooosed Alternative: Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee has requested authorization to use Code Case N-498-1, Alternate Rules for 10-Year Hydrostatic Pressure Testing for Class 1,2, and 3 Systems,Section XI, Division 1.

Licensee's Basis for Prooosed Alternative:

"The hydrostatic test requirement results in unusual difficulties without a compensating increase in the level of quality and safety. The difficulties are associated with the installation of blank flanges to isolate the tested portion from connecting systems, the removal of check valve internals, and when necessary the erection and removal of scaffolding, the removal and replacement of insulation, the removal and restoration to service (along with retesting) of l

electrical components, the setup and removal of the testing equipment, and the return of the system to its normal configuration.

" Draft Regulatory Guide 1.147, dated May 1997, includes Code Case N-498-1,

' Alternative Rules for 10-Year Hydrostatic Testing for Class 1,2, and 3 Systems, i

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.Section XI, Division 1', this Code Case has not been published in Regulatory Guide 1.147, inservice inspection Code Case Acceptability ASME Section XI, Division 1", however, the NRC staff has approved its use at other nuclear stations."

Evaluation: ihe Code requires the performance of a system hydrostatic test once per interval in accordance with the requirements of IWA-5000 for Class 1,2, and 3 j

pressure-retaining systems. In lieu of the Code-required hydrostatic testing requirements, the licensee has requested authorization to use Code Case N-498-1, Attemative Rules for 10-Year System Hydrostatic Testing for Class 1,2, and 3 Systems, dated May 11,1994.

l The system hydrostatic test, as stipulated in Section XI, is not a test of the structuralintegrity of the system but rather an enhanced leakage test.' Hydrostatic j

testing only subjects the piping components to a small increase in pressure over the design pressure; therefore, piping dead weight, thermal expansion, and seismic loads present far greater challenges to the structural integrity of a system.

Consequently, the Section XI hydrostatic pressure test is primarily regarded as a means to enhance leak detection during the examination of components under pressure, rather than as a method to determine the structural integrity of the l

components. In addition, the industry experience indicates that leaks are not being discovered as a result of hydrostatic test pressures causing a preexisting flaw to propagate through the wall. In most cases leaks are being found when the system is at normal operating pressure.

Code Case N-498, Altemative Rules for 10-Year System Hydrostatic Testing for Class 1 and 2 Systems, was previously approved for general use for Class 1 and 2 systems in Regulatory Guide 1.147, Rev.11. For Class 3 systems, Revision N-498-1 specifies requirements identical to those for Class 2 components (for Class 1 l

and 2 systems, the alternative requirements in N-498-1 are unchanged from N-498). In lieu of 10-year hydrostatic pressure testing at or near the end of the 10-year interval, Code Case N-498-1 requires a VT-2 visual examination at nominal i

1. S. H. B_ush and R. R. Maccary, ' Development ofIn-Service Inspection Safety Philosophy for l

U.S.A. Nuclear Power Plants,' ASME,1971 i

) operating pressure and temperature in conjunction with a system leakage test performed in accordance with Paragraph IWA-5000 of the 1992 Edition of Section XI.

Class 3 systems do not normally receive the amount and/or type of nondestructive examinations that Class 1 and 2 systems receive. While Class 1 and 2 system failures are relatively uncommon, Class 3 leaks occur more frequently and are caused by different failure mechanisms. Based on a previous review of Class 3 system failures requiring repair, the most common causes of failures are erosion-8 corrosion (EC), micro biologically-induced corrosion (MIC), and general corrosion.

in general, licensees have implemented programs for the prevention, detection, and evaluation of EC and MIC; therefore, Class 3 systems receive inspection commensurate with their functions and expected failure mechanisms.

System hydrostatic testing entails considerable time, radiation dose, and dollar resources. The safety assurance provided by the enhanced leakage gained from a slight increase in system pressure during a hydrostatic test may be offset or negated by the necessity to gag or remove Code safety and/or relief valves (placing the system, and thus the plant, in an off-normal state), erect temporary supports in steam lines, and expend resources to set up testing with special equipment and gages. Therefore, performance of system hydrostatic testing represents a considerable burden for the licensee.

Giving consideration to the minimal amount of increased assurance provided by hydrostatic test pressures versus system leakage test pressures, and the hardship associated with performing the hydrostatic test, the INEEL staff finds that compliance with the Section XI hydrostatic testing requirements results in hardship and/or unusual difficulty without a compensating increase in the level of quality and safety. Therefore, it is recommended that the use of Code Case N-498-1 for Code Class 1,2, and 3 systems be authorized pursuant to 10 CFR 50.55a(a)(3)(ii). The

2. Documented in Licensee Event Reports and the Nuclear Plant Reliability Data System databases.

I 1

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.g.

use of this Code Case should be authorized for the current interval at the James A.

FitzPatrick Nuclear Power Plant, or until the Code Case is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

D.

Reouest for Relief No. 4. Use of Code Case N-509. Attemate Rules for the Selection and Examination of Class 1. 2. and 3 Intearally Welded Attachments.

Section XI. Division _f.

Code Reouirement: The Code requires examination of integrally-welded I

attachments as specified for Examination Categories B-H, B-K, C-C, D-A, D-B, and D-C. The Code stipulates volumetric or surface examinations, as appropriate, and the extent of examinations.

l l

ensee ha proposed th u e of Co e as N-509 Alte nate R e or the Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments, Section X/, Division 1, in lieu of the requirements of the Code for Class 1, 2, and 3 integrally-welded attachments. The licensee stated:

"The following alternative examination requirements will be implemented as defined by ASME Section XI Code Case N-509, Alternative Rules for the Selection and Examination of Class 1,2, and 3 Integrally Welded AttachmentsSection XI, Division 1. In addition to those conditions specified in Code Case N-509: A minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined, with the following exception:

"1)

Examination Category and item No.(s) for Class 3 Integrally Welded Attachments are defined in accordance with the ASME Section XI,1989 Edition, Article IWD, Table IWD-2500-1."

In their response to the NRC request for additional information the licensee stated:

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' "The Authority confirms and commits to include a minimum of 10% of the total number of non-exempt piping, pump, and valve integral attachments distributed among all C'.ss 1, 2, and 3 systems."

Licensee's Basis for Proposed Attemative:

"At JAF, for the 3rd Ten-Year Interval component supports si;all be selected for examination in accordance with ASME Code Case N-491-1, which has also been requested as a relief, (Reference Relief Request 2).

Application of Code Case N-509 defines alternative examination requirements that may be applied to ASME Code Class 1,2, and 3 integrally weed attachments. The extent of examination as stated in Note 5 of Table 2500-1, Examination Categories B-H, B-K, C-C, and DA of Code Case N-509 requires examination of integral attachments associated with component supports selected for examination under the 1989 ASME Section XI, paragraph IWF-2510. Code Case N-509 has been issued by the American Society of Mechanical Engineers and has been includcd in the 1995 Addenda of Section XI.

" Industry experience in the United States has also shown that ASME Code integral attachment welds have not experienced degradation that would warrant continued examination to the extent required by the 1989 Edition of ASME Section XI. To date, no significant loading conditions or known material degradation mechanisms have become evident that specifically relate to integral attachment welds in nuclear power plant piping. Should a service induced defect be detected in these welds, ASME Code Case N-509, specifies examination expansion criteria to ensure degradation in other attachment welds would be detected.

Therefore, the health and safety of the public will continue to be maintained while implementing the alternative exammation requirements of Code Case N-509. The 1989 Edition of Section XIinspection requirement for Class 2 and 3 integrally welded attschments result in unusual difficulties without a compensating increase in the level of quality and safety. The difficulties and cost associated with the increased radiation exposure. Additional, activities included in this effort are crection and removal of scaffolding, the removs! and replacement of insulation.

" Draft RC 1.147 dated May 1997, includes Code Case N-509, Alternate Rules for the Selecuon ano Examination of Class 1,2, and integrally Welded AttachmentsSection XI, Division 1, this code case has not been published in Regulatory Guide 1.147, inservice Inspection Coda Case Acceptabilit" ASME Section XI, Division 1", however, the NRC staff has approved its use at other nuclear stations."

. )

Evaluation: The licensee has proposed to apply the requirements of Code Case N-509 as an alternative to the Code requirements for the examination of integrally-welded attachments on Class 1,2, and 3 piping and components.

The licensee has committed to supplement the Code Case with a minimum sample of 10% of the total number of non-exemr* oiping, pump, and valve

ntegral attachments distributed among all Class 1,2, and 3 systems. The licensee has also proposed that the examination category and item no.(s) for Class 3 integrally welded attachments be defined in accordance with the Examination Category and item No.(s) for Class 3 Integrally Welded Attachments defined in accordance with the ASME Section XI,1989 Edition, Article IWD, Table IWD-2500-1.

Table 2500-1, in Code Case N-509, classifies tha welded attachments under Pressure Vessels, Piping, Pumps, and Valves. Whereas Table IWD-2500-1, Examination Category D-A, item Nos. D1.20 - D1.60 in ASME Section XI, 1989 Edition, classifies the welded attachments under Component Supports and Restraints, Mechanical and Hydraulic Snubbers, Spring Type Supports, Constant Load Type Supports, and Shock Absorbers. Considering that the licensee has committed to supplement the Code Case with a minimum sample of 10% of the total number of non-exempt piping, pump, and valve integral attachments distributed among all Class 1,2, and 3 systems, the INEEL staff believes that the prcposed use of the 1989 Edition, Article IWD, Table IWD-2500-1 for the classification of examination categor/ and item nos for Class illintegral attachments provides an acceptable level of quality and safety.

Also, considering that most of the Code examination requirements are based on sampling to ensure the detection of service-induced degradation, extending the sampling philosophy to the integral attachment welds wd.' provide an equivalent level of quality and safety. Therefore, the INEEL staff believes that the proposed alternative should be authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the current interval at James A. FitzPatrick Nuclear Power Plant, or until Code Case N-509 is approved for general use by reference in

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. Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

E.

Reauest for Relief No. 5. Use of Code Case N-524. A/remate Examination Reartirements for Loncitudinal Welds in Class 1 and 2 Pioina.Section XI.

Division 1

.Qode Reauirement: Examination Category B-J requires that volumetric and/or surface examinations be performed on longitudinal welds as defined by Figure IWB-2500-8. The examination incledes at least a pipe-diameter length but not more than 12 in. of each longitudinal weld intersecting the circumferential weld required to be examined by Examination Categories B-J and B-F.

Examination Categories C-F-1 and C-F-2 requires 100% volumetric and/or surface examination, as defined by Figure IWC-2500-7, -12 and -13, for 2.5t of each longitudinal weld intersecting circumferential welds examined.

l Licensee's Proposed Alternative: In accordance with 10 CFR 50.55a(a)(3), the licensee proposed the use of Code Case N-524 in lieu of the Code requirements.

Licensee's Basis for Proposed Alternative.:

"1.

Longitudinal welds are fabricated during original manufacturing under controlled shop conditions, which produce higher quality and more uniform residual stress patterns.

"2.

Longitudinel piping welds undergo heat treatment in the shop, which enhances the material properties of the weld and reduces the residual stresses created by welding.

"3.

Results of previous weld inspections throughout the industry indicate that longitudinal welds have not been a safety concern, nor has there i

been any evidence of longitudinal weld defects compromising safety at j

nuclear power plants.

j h l "4.

Longitudinal welds have not been shown to b6 susceptible to any particular degradation mechanism.

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"5.

The only areas of a longitudinal weld which may be considerad suspect l

are the ends of the weld where it is adjacent to the field fabricated l

circumferential weld. These areas fall within the volumetric examination boundaries of the adjacent circumferential weld.

"6.

The man-rem exposure and cost associated with the inspection of longitudinal welds is dependent on the time it would take to removelreinstall insulation and interferences, locate the weld, prepare the weld for examination and perform the examination.

"7.

Based on the above arguments, there is little, if any, technical benefit to l

performing inservice inspections on longitudinal piping welds. In addition, l

there are substantial radiation exposure and cost considerations associated with these inspections.

Evaluation: ASME Section XI requires the examination of one pipe diameter, but not more than 12 inches, of Class 1 longitudinal pip:ng welds. For Class 2 piping welds, the length of longitudinal weld required to be examined is 2.5 times the pipe thickness. These lengths are measured from the intersection with the circumferential weld. The licensee's proposed alternative is to L

examine only the portions of longitudinal weld within the examination area of the intersecting circumferential weld in accordance with Code Case N 524, i

Altemative Examination Requirements for Longitudinal Welds in Class 1 and Class 2 Piping.

Longitudinal welds are produced during the manufacture of the piping, not in the field as is the case for circumferential welds. Consequently, lengitudinal welds are fabricated under strict manufacturing standards, which provides i

assurance of structuralintsgrity. These welds have also been subjected to the

[

preservice and initial inservice examinations, which provide additional I'

assurance of structuralintegrity. No significant loading conditions or material l

degradation mechanisms have been identified to date that specifically relate to t

j longitudinal seam welds in nuclear plant piping. The most critical region of the l

longitudinal weld is the portion that intersects the circumferential weld. Since

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i

! this region will be examined during the examination of the circumferential

- weld, the licensee's alternative provides an acceptable level of quality and safety. Therefore, the proposed alternative should be authorized pursuant to i

t 10 CFR 50.55a(a)(3)(i) for the current interval at James A. FitzPatrick Nudear i

Power Plant, or until Code Case N-524 is approved for general use by l

referenca in Regulatory Guide 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

l F.

Reauest for Relief No. 6. Use of Code Case N-532. A/temative Reauirements to Renair and Reolacement Documentation Reauirements and Inservice Summarv Reoort Pre:3aration and Submission as Reauired bv IWA-4000 and IWA-6000.Section XI. Division 1 Code Reauirement:

IWA-4800 The records required by IWA-6000 shall be completed for all repairs.

l IWA-6210(c) The Owner shall prepare inservice inspection summary report for l

Class 1 and 2 pressure retaining components and their supports.

l IWA-622O(c) Inservice inspection summary reports shall be required at the i

completion of each inspection conducted during a refueling outage. Examinations, tests, replacements, and repairs conducted since the preceding summary report shall be included.

IWA-6220(d) Each summary report sha. contain the following:

l (1) refueling outage number (when applicable);

l (2) Owner's Report for Inservice Inspection, Form NIS-1; and i

(3) Owner's Report for Repair or Replacement, Form NIS-2.

IWA-6230 Within 90 dsys of the completion of the inservice inspection conducted during each refueling outage, the Owner shall file ISI i

Summary Reports with the enforcement and regulatory i.

authorities.

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. Licensee's Proposed Afternative: In accordance with 10 CFR 50.55a(a)(3)(i),

the licensee proposed to use Code Case N 532 as alternative requirements to repair and replacement documentation requirements and inservice summary report preparation and submission as required by IWA-4000, and IWA-6000.

The licensee stated:

"As an alternate to the requirements of IW/. 4800, IWA-6000, and IWA-7528(8), JAF willimplement ASME Code Case N-532, ' Alternative Requirements to Repair and Replacement Documentation Requirements and inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000, Division 1'."

Licensee's Basis for Proposed Alternative:

"JAF feels that the summary report required by IWA-6000 does not contain the information necessary to assure compliance with Code requirements, and therefore does not provide a compensation increase in the quality and/or safety at JAF.

"The summary report does not furnish evidence of compliance with the ASME Boiler and Pressure Vessel Code,Section XI, inspection Program B, percentage requirements as mandated by IWB-2412 IWC-2412, and IWD-2412.

" Class 3 components are excluded from the summary report Submittal.

"Both a Final Report and Summary Report must be prepared, reviewed and approved in order to comply with Sub-articles IWA-6220 and IWA-631 O respectively.

"The preparation, review, approval and certification of each record and report, within the time frame of 90 days following completion of each refueling outage, increases substantially the costs associated with inservice inspection activities, and pute an unreasonable time constraint on JAF without an increase in assurance of Cods compliance.

" Draft RG 1.147 dated May 1997, includes Code Case N-532, Alternative Requirements to Repair and Replacement Documentation requirements and Inservice Summary Report Preparation and Submission l

as Required by IWA-4000 and IWA-6000,Section XI Division 1. This Code Case has not been published in Regulatory Guide 1.147, ' Inservice inspection Code Case Acceptability ASME Section XI, Division 1',

however, the NRC staff has approved it's use at other nuclec-eations."

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l Evaluation: The staff reviewed the proposed alternative documentation requirements of Code Case N-532 and determined that although the required forms have changed, the information required by the Code is available. Code l

Case N-532 would require preparation of ti. Repair / Replacement Certification l

l Record, Form NIS-2A. The completed form NIS-2A shall be certified by an 1

l Authorized Nuclear Inservice inspector (ANil) as defined in ASME Code, l

Section XI, IWA-2130 and shall be maintained by the Owner. Furthermore, the Owner'S Activity Report Form, OAR-1 shall be prepared and certified by an l

ANil upon completion of each refueling outage. The OAR-1 form shall contain an abstract of applicable examinations and tests, a list of item (s) with flaws or l

relevant conditions that require evaluation to determine acceptability for j

continued service, and an abstract of repairs, replacements and corrective measures performed as a result of unacceptable flaws or relevant conditions.

l l

Hence, the information provided in the documentation pertaining to the use of Code Case N-532, can be used in the same manner to assess the safety implications of Code activities performed during an outage. A review using the information as prescribed by the Code Case will, therefore, provide the same or improved level of quality and safety as reviews that may be conducted using the Code reporting requirements. In addition, more detailed information may be requested by the staff if it is deemed necessary. Therefore, the use of this alternative should be authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the current interval at James A. FitzPatrick Nuclear Power Plant, or until Code Case N-532 is approved for general use by reference in Regulatory Guide 1.147. After that time, the licensee must follow the conditions,'if any, i

specified in the regulatory guide.

i G.

Reouest for Relief No. 7. NUREG-0619, inspection Intervals Withdrawn by licensee in letter dated September 21,1998

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. H.

Reauest for Relief No. 8. Examination Cateaorv C-H. All Pressure Retainino Comoonents The licensee stated in their RAI response that they will provide additional 1

information concerning this request for relief within 90 days of the Response to the NRC request for additionalinformation dated September 21,1998. This request for relief was to be expedited, but will now be evaluated at a later j

i date.

1.

Reauest for Relief No. 9. Rev.1 lWE Reauirements for Class MC and Metaffic l.iners of Class CC Components. IWA-2210. Visual Examination. IWA-2211.

VT-1 Examination. IWA-2213. VT-3 Examination. lWA-2300. Qualifications of Nondestructive Examination Personnel Code Reauirement: As required by 10 CFR50.55a(g)(6)(ii)(B), the containment sections (IWE and IWL) of the subject plants' ISI programs are developed in accordance with the requirements of the 1992 Edition,1992 Addenda of Section XI. lWA-2210 requires specific minimum illumination and maximum direct examination distance for visual examinations as specified in Table IWA-2210-1. lWA-2211 requires that VT-1 examinations be conducted to detect j

discontinuities and imperfections on the surfaces of components, including such conditions as cracks, wear, corrosion or erosion. IWA-2213 requires that VT 3 examinations be conducted to determine the general mechanical and structural condition of components and their supports by verifying parameters such as clearances, settings, and physical displacements; and to detect i

discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing partr, debris, corrosion, wear, or erosion. VT-3 includes examinations for conditions that could affect operability or functional i

adequacy of snubbers and constant load and spring type supports. IWA 2300 l

requires in that personnel performing nondestructive examinations be qualified and certified under specified conditions as listed in IWA-2300.

i l

l Licensee's Prooosed Alternative: In accordance with 10 CFR 50.55a(a)(3)(i),

the licensee proposed the use of the ASME Section XI 1989 Edition vist'al examination requirements for use with Subsections IWE. The licensee stated:

"The ASME Section XI 1989 Edition (no addenda) VT-1, VT-3 and Supplemental Examination (surface and volumetric per IWE) requirements, for ' Examination Methods' (lWA-2200) and ' Qualifications of Nondestructive Examination Personnel' (IWA-2300) will be applied to j

IWE required examinations. These are the same requirements that apply l

to all other safety related components in the ISI program and will be supplemented with indoctrination overview training by FitzPatrick's site Level lli (who has attended an EPRI IWE training course) covering the IWE examination requirements prior to personnel performing IWE VT-1 and VT-3 examinations. The results of examinations performed under this proposed alternative are considered equivalent to the results that would be obtained under the 1992 Edition of ASME Section XI.

"Recently an EPRI training course was given to NYPA Engineering /ISI l

personnel and NYPA QC Level ll and til examiners. This two-day course l

provided an overview and guidance for containment inspections, l

cnatings inspections and engineering evaluations. This ensures that QC l

cnd Engineering personnel have received specific training on containment inspections.

This alternative will apply until the beginning of Refuel Outage (RO14 currently scheduled to begin in October 2000."

Licensee's Basis for Proposed Alternative (as stated):

" Requiring redundant administrative programs for IWE Visual and l

Supplemental examinations (surface and volumetric per IWE) in parallel l

with the existing programs for all other ISI components is a significant burden that has no benefit to quality and safety. Inspections performed to the present ISI Program and Code of record are essentially equivalent to that required by the 1992 Edition with 1992 Addenda of the ASME XI Code. Specific training has been provided to Authority Engineering /ISI personnel and OC personnel who oversee contract inspectors and l

provide review of inspection data sheets as required for all ISI i

examinations."

i Evatur%: To comply with the expedited examination of containment 4

requirements of 10 CFR 50.55a(g)(6)(ii)(B), licensees must perform visual examinations on Class MC and Metallic Liners of Class CC Concrete

)

Components of Light-Water Cooled Plants per the requirements of IWE, and

19-visual examinations on Class CC Concrete Components of Light-Water Cooled Plants per the requirements of IWL. These examinations are to be performed to the requirements of the 1992 Edition 1992 Addenda of ASME XI. The I

licensee has proposed the use of IWA-2200 and IWA-2300, VT-1 and VT-3 requirements from the 1989 Edition of ASME Section XI in lieu of the 1992 Edition,1992 Addenda requirements. The INEEL staff has performed a review of the subject requirements found in the 1989 Edition and the 1992 Edition, l

1992 Addenda of ASME XI. The essential differences found during the review consisted primarily of the methods used to determine if satisfactory lighting i

and distances existed to perform the examinations. The 1992 Code requires i

that a quantitative approach is used to ineasure illumination levels. Whereas the 1989 Code requires that a 18% neutral gray card with a 1/32 inch black line be used to determine acceptable examination conditions. While the methods for determination of acceptable illumination are different from 1989 to 1992, the INEEL staff believes that both methods provide an acceptable approach to illumination determinatice. Differences were also noted in the methods used to perform visual acuity examinations of NDT personnel, however, the requirement for visual acuity exists in both the 1989 and 1992 Codes. Considering that the 1989 Code requirements for visual examination are being used for all of the other inservice inspection examination requirements, including the Reactor Pressure Vessel visual examinations, the INEEL staff believes that the licensees proposed alternative provides an acceptable level of quality and safety and should be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

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J.

Reauest for Relief No.10. Paraaraoh IWA-5250(a)(2) Corrective Measures for Leakaae at Bolted Connections i

Code Reauiremen.t: The 1989 Edition of Section XI, IWA-5250(a)(2) requires i

that if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 i

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o

. 3 visually examined for corrosion, and evaluated in accordance with IWA-3100.

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Licensee's Proposed Alternative: In accordance with 10 CFR 50.55a(a)(3), the licensee has proposed an alternative to the requirements of IWA-5250(a)(2).

The licensee stated:

"The ISI Program at JAF will specify conducting visual (VT-3) examination in-place on bolted connections when leakage occurs.

Evidence of degradation to the bolted connection or bolting shat! require the bolt closest to the source of the leakage to be removed visually examined (VT-3) and evaluated to IWA-3100. When the removed bolt has evidence of degradation, al! remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100."

in their response to the NRC request for additional information the licensee also stated:

"An engineering evaluation will be performed to identify the most prudent course of action to determine the condition of the bolting and/or the root cause of the leak when leakage is detected at a bolted connection.

"A situation may be encountered that involves a leaking joint following the complete replacement of bolting materials. The root cause of the leak may be thermal expansion of the piping and botting materials due to system heatup. In such a case, re-torquing the joint botting usually stops the leak. Removal of the joint bolting to evaluate it for corrosion would be a hardship and unwarranted if the botting materialis new or underwent VT-3 examinations prior to installation. ASME Section XI Interpretation XI-1-92-01 recognizes this situation as one in which the requirements of IWA-5250(a)(2) do not apply. Other situations which could also result in hardship include:

bolted connections which require a specific torque pattern, primary containment isolation valves which require stroke time testing and local leak rate tasting following reassembly, components which would require extension disassembly to comply with the requirements of IWA 5250(a)(2) due to equipment clearances "lWA-5250(a)(2) does not include other factors which may indicate the condition of the bolted connection. Other factors that will be considered

. in the Authority's evaluation of leakage at bolted connections will include, but not necessarily be limited to:

1. bolting materials

2. corrosiveness of the process fluid 3.~ leakage location

4. history of leakage at the connection or other system components

. 5. visual evidence of corrosion at the connection (while connection is assembled)

6. service age of the bolting materials.

" Training on the use of the above leakage evaluation ~ criteria, as well as the appropriate corrective actions to be taken if the evaluation is inconclusive or identifies bolting degradation, will be provided to appropriate ISI, Operation Review Group, and Quality Assurance personr.el prior to the use of the proposed alternative. The leakage evaluation criteria and appropriate corrective actions will be formally proceduralized within 6 months following completion of the up coming refuel outage."

In a letter dated November 3,1998, the licensee further described the VT-3 examination process that will be utilized as part of this request for relief. The licensee steted:

" Relevant conditions required to be identified during the VT 3 examination include erosion, corrosion, wear, and crack-like or linear flaws. Localized general corrosion that reduces the bolt or stud cross sectional area by more than 5% is recorded in the VT-3 non-destructive examination report."

j i

Licensee's Basis for Proposed Alternative (as stated):

"Later Codes and Addenda (ASME Section XI 1992 Edition) have acknowledged this requirement as a significant burden that has no benefit to quality or safety and have modified IWA-5250(a)(2) as follows:

"If leakage occurs at a bolted connection on other than a gaseous system, one of the bolts shall be removed, VT 3 examined, and evaluated in accordance with IWA-3100. The bolt selected shall be the one closest to the source of leakage. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100

m l

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! "The degradation of bolting due to leakage from borated systems is not an occurrence that is considered a problem at Boiling Water Reactors (BWRs).

" Pressure testing at JAF will be conducted in accordance with Relief Request #3, which evokes the requirements of Code Case N-498-1.

Code Case N-498-1 requires that the visual examination (VT-2) be performed at nominal operating pressure but states that temperature shall not exceed limiting conditions for the hydrostatic test curve as contained in the plant Technical Specifications. JAF expects a certain amount of leakage during pressure testing from bolted connections (valves, flanges) during RCS pressure testing primarily due to the lower testing temperature. The majority of leakage identified during testing is from packing leaks but a small percentage is attributed to the CRD Flanges and other pressure retaining bolted connections. Usually this leakage is arrested as the plant heats up or additional torquing is performed to stop the leakage, in those cases where leakage is not arrested based on the above actions an evaluation is performed and, when necessary, correctivo measures taken.

" Removal and inspection of botting, without indication of bolting degradation, can result in a system or portions of a system being placed in a mode of inoperation, limited condition of operation (LCO), or place the plant in a shutdown condition. This constitutes a significant level of risk and burden on the plant without a compensating benefit to quality or safety."

Evaluation: In accordance with the 1989 Edition of ASME XI, if leakage l

occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100. In lieu l

of this requirement, the licensee has proposed to conduct a visual (VT-3) examination in-place on bolted connections when leakage occurs. An evaluation of the botting will be performed to determine its susceptibility to corrosion. This evaluation considers a number of parameters, including bolting materials, the potential for corrosion, and visual evidence of corrosion with the bolting in place. If evidence of degradation exists, the bolt closest to the l

source of the leakage will be removed, visually examined (VT-3), and evaluated to IWA-3100. When the removed bolt has evidence of degradation, 4

~

a;l remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100.

4 g

. IWA-3100 invokes the se of subparagraphs IWB-3000, IWC-3000, IWD-3000 for Class 1,2, and 3 pressure retaining components respectively.

However, none of these subparagraphs provide an acceptance criteria for VT-3 examinations. Therefore, the ability to perform a meaningful evaluation on the botting without an applicable ac:eptance criteria may be inappropriate. The INEEl. staff believes that a VT-1 visual examination utilizing the acceptance criter a defined in IWB 3000 provides a more appropriate method of examination of the subject botting than a VT-3 visual examination. However, the licensee provided a description of the VT-3 examination process that will be utilized as part of tnis request for relief. The VT-3 examination willinclude

~

the identification of erosion, corrosion, wear, and crack-like or linear flaws.

l Also, localized general corrosion that reduces the bolt or stud cross sectional area by more than 5% will be recorded in the VT-3 examinttion report. The VT-3 visual examination as stated by the licensee appears to closely parallel the Code VT-1 visual examination standards described in IWB-3517.1 concerning degradation related to leakage at bolted connections. Similar l

requests for relief have been approved with the condition that a VT-1 visual examination be performed using the acceptance criteria for VT-1 examinations.

Based on the determination that the licensee's VT-3 visual examination closely parallels the VT-1 visual examination standards concerning degradation related to leakage at bolted connections, and that the evaluation process proposed by l

the licensee provides a sound engineering approach, it is concluded that an acceptable level of quality and safety is provided.

I However, in order to clarify the acceptance standards used for future examinations and to promote consistent use of the subject alternative, the i

staff has determined that VT-1 visual examinations should be used for all botting at the James A. FitzPatrick Nuclear Power Plant. It is recognized that t

imposition of this requirement could negatively impact the current outage, but should be implemented for future inspections. Therefore, it is recommended i

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l 7 that the licensee's proposed alternative be authorized for a limited time, not to exceed one year from the date of this Safety Evaluation Report.

K.

Reauest for Relief No.12. Use of Code Case N-522. Pressure Testino of j

Containment Penetration Pioina.Section XI. Division 1 Code Reauirement: Table IWC-2500-1, Examination Category C-H, requires a VT-2 visual examination during the system leakage test (IWC-5221) once each period and during the system hydrostatic test (IWC-5222) once each interval for all Class 2 pressure retaining components.

l Licensee's Prooosed Alternative: In accordance with 10 CFR 50.55a(a)(3), the licensee proposed to implement Code Case N-522, Pressure Testing of Containment Penetration Piping. The licensee stated:

I "The following alternative examination requirements will be implemented as defined l

by ASME Section XI Code Case N-522, Pressure Testing of Containment Penetration Piping,Section XI, Division 1. In addition to the requirements specified in Code Case N 522, the following conditions shall be met:

"The test should be conducted at the peak calculated containment pressure and the test procedure should permit the detection and location of through-wall leakage in l

Containment isolation Valves (CIVs) and pipe segments between the CIVs."

Licensee's Basis for Prooosed Alternative:

l l

"The leakage testing requirements results in unusual difficulties without a compensating increase in the level of quality and safety.

"ASME Code Case N-522 recognizes and addresses this fact and proposes an j

alternative which maintains an acceptable level of quality and safety.

I r

" Draft RG 1.147 dated May 1997, includes Code Case N-522, Pressure Testing of Containment Penetration Piping,Section XI, Division, this Code Case has not been published in Regulatory Guide 1.147, inservice inspection Code Case Acceptability ASME Section XI, Division 1.

" Penetrations at JAF are currently tested in accordance with the Appendix J Testing Program."

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! In their response to the NRC request for additionalinformation the licensee also stated:

"The use of Code Case N-522 is required for those systems that have been classified as ISI Class 2 but that cannot be pressure tested under ASME Code Case N-416-1 (see Relief Request No.1).

"Many of thest, systems are air systems or are open to the atmosphere and the pressure test requirements of Code Case N-416-1 as stated in Relief Request No.1 do not apply. Two examples are provided below:

1.

" Portions of the Drywell, Inerting, CAD and Purge and Containment Differential Pressure System (system 27), which forms part of the containment, are classified as ISI Class 2 in accordance with ASME XI IWA-1320 of the ASME XI 1992 Edition through 1992 Addenda. If a piping replacement were to be completed between the containment and an AOV (e.g., AOV-112 and line number 24"-N-152A-8), use of Code Case N-416-1 would require augmented NDE (ASME lit) and pressure testing per ASME XI 1992 Edition. A system leakage test per paragraph IWA-5000 would be required. This section of piping is normally open to containment atmosphere and the required system leakage test could not be performed.

"In addition, the exemptions per IWC-5222 would require demonstration of an open flow path, which is not feasible for this air system. This would require having an air pressure source or flooding of containment to test this piping and also might require modification of the system.

"In this case, performance of the augmented NDE per ASME Code Case N-416-1 and the 10 CFR 50 Appendix J test required by Code Case N-522 is equivalent and ensures an acceptable level of quality and safety.

2.

"The RHR containment spray piping (line number 10"-W20-302-12A/B) is an open ended system in containment and is classified as ISI Class 2. If a piping replacement were to be completed between the containment and one of the containment isolation valves (e.g.,10MOV-3-A), use of Code Case N-416-1 l

would require augmented NDE (ASME 111) and pressure testing per ASME XI l

1992 Edition. A system leakage test per paragraph IWA-5000 would be required. This section of piping is normally open to containment atmosphere and the required system leakage test could not be performed without wetting the containment and electrical equipment inside the containment.

"In addition, the exemptions per IWC-5222 would require demonstration of an open flow path. This can be done but requires additional manpower and radiation exposure to line up the RHR system with an air test source.

Personnel would have to verify flow by checking the containment spray

, nozzles in containment in a radiation area thereby contributing to the increases in radiation exposure.

I "In this case, performance of the augmented NDE per ASME Code Case N-416-1 and the 10 CFR 50 Appendix J test required by Code Case N-522 is equivalent and ensure an acceptable level of quality and safety while reducing radiation exposure to plant staff."

Evaluation: The Code requires that a VT-2 visual examination be performed during system pressure testing for all Class 2 pressure-retaining piping, including those segments that penetrate primary containment. As an alternative, the licensee proposed to implement the requirements of Code Case N-522, Pressure Testing of Containment Penetration Piping. Code Case N-522 specifies that 10 CFR 50, Appendix J testing may be used as an alternative to Section XI pressure tests, for certain containment penetration piping.

The subject piping is fabricated and designated as Class 2 only because it penetrates the primary reactor containment and is considered an extension of the containment vessel. Since the piping on either side of these penetrations is non-classed, the requirements of Appendix J are more appropriate than those of Examination Category C-H.. Appendix J pressure tests verify the leak-tight integrity,f the primary reactor containment and of systems and components that penetrate containment by local leak rate and integrated leak rate tests.

The Class 2 containment isolation valves (CIVs) and connecting pipe segments must withstand the peak calculated containment internal pressure. The INEEL staff j

l believes that the pressure retaining integrity of the CIV's and connecting piping and their associated safety functions may be verified with an Appendix J test conducted at the peak calculated containment pressure. The licensee has proposed to perform j

the Appendix J testing at no less than the peak calculated containment pressure and l-l will use procedures and techniques capable of detecting and locating through-wall leakage in the pipe segments between the CIV's.

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,4,

i Appendix J, Option A-Prescriptive Requirements, requires that three Type A tests be performed at approximately equal intervals during the 10 year ISI interval, with the third test being done while shutdown for the 10-year plant ISI. Option A also l

requires Type B and C tects be performed during each refueling outage, but in no case at intervals greater than 2 years. This is more frequent than the periodic pressure tests required by ASME Section XI.

Appendix J, Option B-Performance Based Requirements, allows a licensee to perform Type A, B, and C tests at frequencies related to the safety significance and historical performance of the system's isolation capabilities. This could, in effect, allow only one test to be performed during the 10-year ISI interval. However, the staff's position, as stated in Regulatory Guide 1.163 Performance-based Containment Leak-Test Program, is that the licensee is to establish test intervals of no greater than 60 months for Type C tests because of uncertainties (particularly unquantified leakage rates for test failures, repetitive / common mode failures, and aging effects) in historical Type C component performance data. While this five-year limit results in an increased time between testing over that required by Section XI (forty months), it is believed that Appendix J tests are more appropriate and provide reasonable assurance of the continued operability of containment penetrations. Therefore, the INEEL staff believes that the test frequencies associated with Appendix J, Option A (Type A, B or C) or Option B (Type C) Tests are commensurate with the Code-required pressure test frequencies.

The licensee has chosen to use Option B. However, the licensee has stated in a telephone conversation with the Staff that the examination of the subject components will be performed only when repe!r or replacement occurs. Therefore, the INEEL staff recognizes that the licensee test frequencies may exceed the 60 month interval for a Type C test as stated in Regulatory Guide 1.163. The INEEL staff believes that the licensee's proposed test frequency is inappropriate due to uncertairities with repetitive / common mode failures, and aging effects in Type C components.

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, Based on the above analysis and information submitted, the INEEL staff believes that an acceptable level of quality and safety will be provided by the licensee's proposed alternative for a period of 60 months. Therefore, it is recommended that the licensee's proposed alternative to implement Code Case N 522 be authorized pursuant to 10 CFR 50.55a(a)(3)(i) for a limited time not to exceed 60 months from the date of the Safety Evaluation Report.

L.

Reauest for Relief No.13. Use of Code Case N-573. Transfer of Procedure Qualification Records Between Owners.Section XI. Division 1 i

l Code Reauirement: IWA-4400(a) requires that all welding shall be performed in accordance with Welding Procedure Specifications that have been qualified l-i by the Owner or repair organization in accordance with the requirements of the codes specified in the Repair Program in accordance with IWA-4120.

Licensee's Proposed Alternative: In accordance with 10 CFR 50.55a(a)(3)(i),

the licensee has proposed to use Code Case N-573, Transfer of Procedure Qualification Records Between Owners. The licensee stated:

"The following alternative testing requirements will be implemented as defined by ASME Section XI Code Case N-573, Transfer of Procedure Qualification Records Between Owners,Section XI, Division 1.

1.

"NYPA will perform a technical review of the supplying Owner's POR 2.

"The supplying Owner will state in writing that the POR was I

performed under an acceptable Nuclear Quality Assurance f'

program that meets ASME Section XI, IWA 1400 and that it I

was performed in accordance with ASME Section IX.

l-3.

"NYPA will generate a NYPA WPS using the Variables established in the supplied POR(s). NYPA POR's may supplement these or other Owner supplied POR's.

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"The WPS will be approved and signed by NYPA.

j 5.

"The WPS will be demonstrated successfully by NYPA by completing a welder performance qualification test using the parameters of the NYPA WPS.

6.

"NYPA will not transfer the supplied PQR to any other Owner.

7.

"NYPA will document the use of this Code Case on the appropriate NIS-2 form."

Licensee's Basis for Proposed Afternative-

"The basis for this relief is to implement ASME Code Case N-573, which j

eliminates the redundancy currently required by the Code for each i

organization to independently qualify all welding procedures even though they have met the qualif! cation process at another facility. Code Case N-573 recognizes and addresses this fact and proposes an alternative which maintains an acceptable level of quality and safety."

1 Evaluation: IWA-4400(a) requires that all welding shall be performed in accordance with Welding Procedure Specifications (WPS) that have been qualified by the Owner or repair organization in accordance with the requirements of the codes specified in the Repair Program, per IWA-4120.

The licensee has proposed the use of Code Case N-573, Transfer of Procedure Qualification Records Between Owners. This Code Case essentially allows the use of a welding or brazing procedure qualification record (POR) qualified by one owner to be used by another owner for the development of the WPS. The specific requirements listed in Code Case N-573 shall be met by the Owner that performed the procedure qualification, and by the Owner intending to use the PQR. These requirements are:

(a)

The Owner that performed the procedure qualification test shall certify, by signing the POR, that testing was performed in accordance with Section IX.

(b)

The Owner that performed the procedure qualification test shall certify, in writing, that the procedure qualification was conducted in accordance

s.

s with a Quality Assurance Program that satisfies the requirements of IWA-1400.

(c)

The Owner accepting the completed POR shall accept responsibility for obtaining any additional supporting information needed for WPS development.

(d)

The Owner accepting the completed POR shall document, on each resulting WPS, the parameters applicable to welding. Each WPS shall be supported by all necessary POR's (e)

The Owner accepting the completed POR shall accept responsibility for the POR. Acceptance shall be documented by the Owner's approval of each WPS that references the POR.

L (f)

The Owner accepting the completed POR shall demonstrate technical competence in application of the received POR by completing a performance qualification test using the parameters of a resulting WPS.

(g)

The Owner may accept and use a POR only when it is received directly from the Owner that certified the POR.

(h)

Use of this Code Case shall be shown on the NIS-2 form documen&g welding or brazing.

The INEEL staff believes that qualification of a procedure for the purpose of joining materials by either welding or brazing may be performed by any Owner provided the applicable requirements for procedure qualification are maintained. The INEEL staff also believes that Owners may use procedures qualified by other Owners provided the conditions / requirements listed in Code Case N-573 are met. The licensee has committed to comply with requirements specified in Code Case N-573. Therefore, the proposed i

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alternative provides an acceptable level of quality and safety and the use of 1'

this alternative should be authorized pursuant to 10 CFR 50.55a(a)(3)(i) for the j-current interval at James A. FitzPatrick Nuclear Power Plant, or until Code i

Case N-573 is approved for general use by reference in Regulatory Guide

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' 1.147. After that time, the licensee must follow the conditions, if any, specified in the regulatory guide.

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CONCLUSION l

The INEEL staff has reviewed the licensee's submittal and concludes that for Requests for Relief Nos. 1, 2, 4, 5, 6, 9,12, and 13 the licensee's proposed alternatives will provide an acceptable level of quality and safety. Therefore, it is recommended that these proposed alternatives be authorized for the third interval pursuant to 10 CFR 50.55a(a)(3)(i). For Request for Relief No.10 the licensee's proposed alternative will provide an acceptable level of quality and safety.

However, in order to clarify the acceptance standards used for future examinations and to promote consistent use of the subject alternative, the staff has determined l

that VT-1 visual examinations should be used for all bolting at the James A.

FitzPatrick Nuclear Power Plant. It is recognized that imposition of this requirement could negatively impact the current outage, but should be implemented for future j

inspections. Therefore, it is recommended that the licensee's proposed alternative be authorized for a limited time, not to exceed one year from the date of this Safety i

Evaluation Report. For Request for Relief No. 3, it is concluded that the Code requirements would result in a hardship without a compensating increase in the level of quality and safety. Therefore, it is recommended that the proposed alternative be authorized pursuant to 10 CFR 50.55a(3)(ii). Request for Relief No. 7 was withdrawn by the licensee in letter dated September 21,1998. As stated in the letter dated September 21,1998, the licensee has determined that Request for Relief No. 8 will not be required for the outage, therefore, additional information for t

completion of the review of this request for relief will be provided at a later date.

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