ML20210P111

SER Accepting Evaluation of Third 10-year Interval Inservice Insp Program Requests for Relief for Plant,Unit 3

ML20210P111
Person / Time
Site:	Crystal River
Issue date:	08/05/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20210P078	List: ML20210P074 ML20210P111
References
NUDOCS 9908120073
Download: ML20210P111 (36)

Text

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NUCLEAR REGULATORY COMMISSION WASHINGTON. O.C. 3000H001 SAFETY. EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO Th'E THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PLAN REQUESTS FOR RELIEF .

i FLORIDA POWER CORPORATION I CRYSTAL RIVER UNIT 3 DOCVET NO. 50-302 '.

1.0 INTRODUCTION

Altematives to Code requirements may be used by nuclear licensees when authorized by the U.S. Nuclear Regulatory Commission if the proposed alternatives to the requirements are such that they are shown to provide an acceptable level of quality and safety in lieu of the Code requirements [10 CFR 50.55a(a)(3)(i)), or if compliance with the Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety [10 CFR 50.55a(a)(3)(ii)].

A licensee may also submit requests for relief from certain Code requirements when a licensee has determined that conformance with certain Code requirements is impractical for its facility

[10 CFR 50.55a(g)(5)(iii)). Pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant such relief and may impose such attemative requirements as it determines are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

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 constructicn of the components. The regulations require that inservice examination of components and system pressure tests be conducted during the first ten-year interval and in 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) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The Code of record for Crystal River, Unit 3 (CR 3),

third 10-year inservice inspection ISI interval,is the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code.

2.0 EVALUATION ,

By [[letter::3F0498-03, Forwards ten-year ISI Program Manual for Third Insp Interval for Crystal River,Unit 3.Manual Updated to Meet Requirements of 1989 Edition of ASME Boiler & Pressure Vessel Code IAW ASME Section Xi,Except Where Specific Relief Is Requested|letter dated April 16,1998]], and revised November 30,1998, Florida Power Corporation (the licensee or FPC) submitted its third 10-year interval ISI program plan requests for relief for CR 3. Idaho Engineering and Environmental Laboratory (INEEL), evaluated the information 9908120073 990005 ENCLOSURE PDR

g. ADOCK 05000302 PDR

, i provided by the licensee in the April 16,1998, submittal in support of its third 10-year interval ISI program plan requests for relief. The NRC staff reviewed the changes and additional relief requests submitted in the November 30,1998, submittal, as well as the licensee's May 12, 1999, response to the staff's request for additional information. Based on the results of the review, the staff adopts the contractor's conclusions and recommendations presented in the attached Technical Letter Report with the exception of the conclusions for Requests for Relief Nos. 98-002-11, 98-003-II, and 98-005-PT.

l The information provided by the licensee in support of the requests for relief from Code requirements has been evaluated and the basis for disposition is documented below.

Raouest for Relief No. 98-001-II: ASME Code, Examination Category B-F, item B5.10 requires 100% volumetric and surface examination of pressure-retaining dissimilar metal welds, as )

defined by Figure IWB-2500-8, each inspection interval.

Pursuant to 10 CFR 50.5kia(a)(3){i), the licensee proposed to perform the volumetric examination of the outside surface from the nonle bore on the following reactor pressure I vessel (RPV) nonle to-safe end butt welds.

Location Identification #

"W" axis of Reactor Vessel B1.6.1 (UT)

B1.6.2 (PT)

"Y" axis of Reactor Vessel B1.6.3 (UT)

B1.6.4 (PT)

The hconsee stated:

" Florida Power Corporation will perform a full volumetric examination of both Core Flood Nonle to Safe End welds from the inside diameter using the enhanced ultrasonic technique as described in Babcock and Wilcox Owners Group (BWOG) Topical Report BAW-228P, Revision 1. This ultrasonic examination technique has been qualified to detect outside diameter surface cracking."

The Code requires 100% volumetric and surface examination of the subject noule-to-safe end welds. The licensee has proposed to perform an ultrasonic examination of the outside surface from the inside diameter of the noule in lieu of the Code-required surface examination. The technique proposed by the licensee was demonstrated to the NRC by the BWOG on August 11, 1993. The BWOG prepared Topical Report BAW-2228P, Revision 1. This report describes a modified fracture mechanics analysis applied to postulated outer surface, semi-elliptical circumferential reactor vessel noule-to-pipe weld flaws. It also demonstrates that adequate margins exist between the revised allowable limiting flaw size and the minimum flaw that was detected and sized during the demonstration. Therefore, the subject report provides a basis for substituting the ultrasonic examination for the surface examination.

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. 4 The NRC has reviewed BWOG Topical Report BAW-2228P, Revision 1, and accepted the results in a Safety Evaluation Report dated March 21,1996. The NRC concluded that the enhanced ultrasonic technique is an acceptable attemative to the Code-required surface examination provided that the following conditions are met:

1) stresses for the subject nozzle-to-pipe welds are bound by the input stresses under normal / upset and emergency / faulted conditions shown in Tables 1 to 4, and

2) the material for the core flood nozzles (nozzles of interest) is either A508 Class 2 carbon steel or SA 336 (316) stainless steel.

The licensee has confirmed that the input stresses under normal / upset and emergency? faulted conditions shown in BAW-2228P, Revision 1, apply to CR-3, and that the material is A508 Class 2, carbon steel. Therefore, the staff determined that the licensee's proposed volumetric examination in lieu of the required surface examination provides an acceptable level of quality and safety. Therefore, the licensee's proposed attemative examination is authorized pursuant to 10 CFR 50.55a(a)(3)(l).

Request for Relief No. 98-002-II: ASME Code, section XI, Examination Category B-L-1, item B12.10 requires 100% volumetric examination of pump casing welds, as defined by Figure IWB 2500-16, each inspection interval. Examinations are limited to at least one pump in each group of pumps performing similar functions in the system.

i The licensee proposed to perform VT-2 visual examinations, VT-3 visual examinations, and surface examinations on 25% of the upper and lower pump case scroll welds on pump RCP-1 A in lieu of the Code required volumetric examination. The licensee stated:

" Visual (VT-2) examinations will be performed during system leakage tests conducted prior to plant startup following each refueling outage. VT-3 examination will be conducted on the intemal surface of at least one pump if disassembled for maintenance during the interval, as required in Category B-L-2. j l

"A surface examination will be performed on 25 percent of the length of both the upper l and lower pump case scroll welds on one pump (RCP-1 A) during the third inspection interval. The 25 percent segment selected for inspection will be a segment that has not been previously inspected during the first or second inspection intervals."

The Code requires 100% volumetric examination of the pump cas!ng welds on at least one ,

pump in each group of pumps performing similar functions in the system each inspection l interval. The licensee proposed to perform a VT-2 visual examination during the system leakage tests conducted prior to startup following each refueling outage in accordance with Examination Category B-P requirements, a VT-3 visual examination on the internal surface of at least one pump in accordance with Examination Category B-L-2 requirements, and a surface examination on 25% of the length of both the upper and lower pump case scroll welds on pump RCP 1 A during the interval.

The design of the reactor coolant pumps makes field disassembly a difficult task, resulting in significant radiation exposure to conduct the required examinations. The pump casings are constructed of 316 stainless steel and have an excellent service record, with no instances of

4 pump casing weld degradation being reported by the nuclear industry. No unacceptable indications were noted during examinations performed during the second inspection interval at CR-3. Therefore, the staff finds that meeting certain ISI requirements would result in hardship on the licensee without a compensating increase in the level of quality and safety. The alternative examinations proposed by the licensee will provide adequate assurance of pressure boundary integrity. Thus, the proposed altemative is acceptable pursuant to 10 CFR 50.55a(a)(3)(li).

Recuest for Relief No. 98-003-11: ASME Code,Section XI, Examination Category F-A, item F1.30 requires a 100% VT-3 visual examination of weld and mechanical connections at intermediate joints in multi-connected integral and non-integral supports, as defined by, Figure IWF-1300-1. The licensee proposed to perform an attemative VT-3 visual examination on 10%

of the Reactor Pressure Vessel Support Skirt B1.12.1.

The licensee stated:

"VT-3 examination will be performed on 10% of the interior of the reactor vessel support at three positions along the circumference of the support. The same segments of the support skirt, located 120' apart, that were previously examined in the first and second inspection intervals, will be examined in the third inspection interval."

The Code requires that a VT-3 visual examination of 100% of the weld and mechanical connections at intermediate joints in multi-connected integral and non-integral supports be performed each inspection interval. The licensee has proposed that a VT-3 visual examination be performed on 10% of the interior of the reactor pressure vessel support at three positions along the circumference of the support.

Visual examinations during the first and second intervals indicated no degradation of the same areas of the support skirt covered by this request. In order to perform the required examination coverage, significant preparation work involving erection of temporary lighting, scaffolding, and surface preparation would be required. This work would result in significant radiation exposure, estimated by the licensee to be approximately 30 person-REM.

Considering (1) the amount of radiation exposure that would occur inspecting 100% of the welds associated with the interior of the reactor vessel skirt support, (2) that reactor vessel skirts have no histories of failure, and (3) that a sample of 10% of the interior weld surfaces of the reactor vessel support skirt will provide reasonable assurance of structural integrity, the staff finds that performing a VT-3 visual inspection of 100% of the interior reactor vessel support skirt weld surface would result in hardship on the licensee without a compensating increase in the level of safety or quality. Thus, the proposed altamative is acceptable pursuant to 10 CFR 50.55a(a)(3)(ii).

Recuest for Relief No. 98-004-il:- ASME Code,Section XI, Exarnination Category B-0, item B14.10 requires volumetric or surface examination on 10% of peripheral control rod drive (CRD) housing welds, as defined by Figure IWB 2500-18, each inspection interval.

Pursuant to 10 CFR 50.55a(a)(3)(1), the licensee proposed to perform volumetric or surface examinations on three welds of each of the six CRD mechanism (CRDM) housings that are removed from the reactor vessel head for maintenance during the third inspection interval.

l The Code requires 100% volumetric or surface examination of the welds on 10% of the peripheral CRD housings. Review of the licensee's documentation shows that there are 24 peripheral CRDMs on the reactor vessel head. Therefore, the licensee is required to examine the welds of three of the peripheral CRD housings. Each CRD housing has five welds, two welds below the CRDM flange connection and three welds above it. Consequently, the licensee is required to examine at least 15 welds. The licensee has proposed to perform volumatric or surface examinations of three welds on each of eix CRDM housings that will be removed from the reactor vessel head for maintenance during the interval. Therefore, the total number of welds to be examined will be 18, an increase of 20% over the required examination sample.

The licensee occasionally removes CRDM housings for maintenance by disconnecting the housings at the flange assembly. The three welds above the flange assembly are removed from the reactor head with the CRDM housing. The licensee is then able to perform the required examinations in an area with lower radiation. The licensee has determined that this procedure reduces the radiation dose by12 REM. The three welds included on the removal portion of the CRD housing are representative of the two welds below the CRDM flange connection. Based on the 20% increase of Code item B14.10 welds to be examined, and the ,

reduction in radiation exposure to inspection personnel, the staff determined that the licensee's l proposed altamative provides an acceptable level of quality and safety. Therefore, the licensee's proposed attemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Reauest for Relief No. 98-005-il:

Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee proposed to use Code Case N-524, Altemative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping. Code Case N-524 was approved for general use by reference in Regulatory Guide (RG) 1.147, l Revision 12 dated May 1999. The licensee must follow the conditions, if any, specified in the RG.

Reauest for Relief No. 98-006-II: ASME Code,Section XI, requires examination of integrally- l welded attachments as specified for Examination Categories C-C, D A, D-B, and D-C. The Code stipulates volumetric or surface examinations, as appropriate, and the extent of examinations.

Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee has proposed to use pc.1 ions of Code Case N 509, Altemative Rules for the Selection and Examination of Class 1,2, and 3 Integrally Welded Attachmentsin lieu of the requirements of the Code for Class 2 and J' integrally-welded attachments. The licensee stated:

"The following altemative examination requirements will be implernented as defined by ASME Code Case N-509, exclusive of the selection criteria which is defined below:

"ASME Code Case N-509 and Table 25001, Examination Category C-C, shall be used for Class 2 integrally welded attachments to piping as defined by Examination Category C-C of ASME Section XI, Article IWC.

"ASME Code Case N-509 and Table 2500-1 Examination Category D-A, shall be used for Class 3 integrally welded attachments to piping as defined by Examination Categories D A, D B, and D-C of ASME Section XI, Article IWD.

"In addition to the requirements of ASME Code Case N 509, the extent of scheduled examination shall be a minimum of 10% of the total population (no thickness exemption) of integral attachments in non-exempt Code Class 2 and 3 systems."

The Code requires examination of integrally-welded attachments, as specified, for Examination Categories C-C, D-A, D-B, and D-C. The Code stipulates volumetric or surface examinations, as appropriate, and the extent of examinations. 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 2 and 3 piping. The licensee has also committed to supplement the Code Case with a minimum examination sample of 10% of all integral attachments to non-exempt Class 2, and 3 systems. ,

The Code does not require examination of Class 1 integrally welded attachments in the third or

, fourth intervals and the licensee intends to use the provision of the Code for Class 1 attachments, while proposing to use Code Case N-509 as an alternative for Class 2 and 3 attachments. Code Case N-509 is written for Class 1,2, and 3 integrally welded attachments, and requires that Class 1,2, and 3 integrally welded attachments be examined each inspection interval. While Code Case N-509 significantly reduces the number of welded integral attachments to be examined, it requires a sample of Class 1 integrally welded attachments to be continued to be examined in later intervals. Consequently, the licensee's proposal reduces the Code Case N 509 examination sample size by not including Class 1 components, which reduces the effectiveness of the Code Case examination.

The staff determined that Code Case N-509 must be implemented in its entirety to be acceptable, and concludes that the licensee's proposed altemative, to implement Code Case N 509 for Class 2 and 3 welded integral attachments only, is non-conservative and does not provide an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is not authorized. j Reauest for Relief No. 98-007-11: Request for Relief 98-007-ll was withdrawn by the licensee in its letter dated May 12,1999. '

Reauest for Relief No.98-008 il: ASME Code,Section XI, Examination Cate2ory B-G-1, item B6.10 requires a surface examination of all RPV closure head nuts each 10-year interval.

Pursuant to 10 CFR 50.55a(a)(3), the licensee proposed to implement the VT-1 visual examination requirements of the 1989 Addenda of Section XI for the RPV closure head nuts.

The licensee stated: .

" Relief is being requested to substitute a visual (VT 1) examination for the required surface examination of each RPV closure head nut during the inspection interval.

"A visual examination will be performed on all RPV closure head nuts once during this inspection interval, as allowed in the 1989 Addenda of Section XI."

The licensee has requested relief from performing the Code-required surface examination of the RPV closure head nuts. As an altamative, the licensee proposed to perform a VT-1 visual examination as allowed in the 1989 Addenda to the 1989 Edition of ASME XI. In the 1989

Edition, Examination Category B-G-1 requires that all items be examined using VT-1 )

examination and/or volumetric examination (as applicable), with the exception of the surface examination for RPV closure head nuts and closure head studs (when removed).

Typical indications that would require corrective action on RPV closure head nuts are associated with degradation mechanisms such as boric acid attack, corrosion, or handling mechanisms such as galled threads and deformation. Typical surface examination procedures and techniques are not qualified to identify these forms of degradation.

Article IWB-3000, Acceptance Standard, IWB 3517.1, Visual Examination, VT-1, describes conditions that require corrective action prior to continued service of bolting and associated nuts. IWB-3517.1 requires crack like flaws to be compared to the flaw standards of IWB-3515.

Because the VT-1 visual examination acceptance criteria includes evaluation of crack-like indications and other relevant conditions requiring corrective action (i.e., deformed or sheared threads, localized corrosion, deformation of part, and other degradation mechanisms), the staff concluded that the VT-1 visual examination provides a more comprehensive assessment of the condition of the closure head nut, and is better suited to identify typical indications of degradation. As a result, the staff determined that VT-1 visual examinations provide an acceptable level of quality and safety. Therefore, the licensee's proposed altomative, to use VT-1 visual examinations, is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Raouest for Relief No. 98-009-11 The ASME Code,Section XI,1989 Edition Tables IWB-2412-1, IWC-2412-1, IWD-2412-1, IWE-2412-1, and Table 2410-2 from Code Case N 491, require the following percentage of  !

examinations for all components of Tables IWB-2500-1, lWC-2500-1, IWD-2500-1,  !

IWE-2500-1, IWF 2500-1 and Table-2500-1 of ASME Code Case N-491 that are not allowed to i be deferred:

Inspection Period, Minimum Maximum i Calendar years Examinations Examinations Service within interval Comoteted. % Credited. %

3 16 34  ;

7 50 67 10 100 100 The licensee has proposed to use the attemative requirements of Code Case N-598 during the third 10 year inspection interval which allows 50% of the required examinations to be completed by the end of the first period and 75% by the end of the second period as stated in the following table:

Inspection Period, Minimum Maximum Calendar years Examinations Examinations Service within interval Comoleted. % Credited. %

3 16 50 7 50 75 10 100 100

, .9 8-FPC stated that this relief would allow CR-3 to examine more welds during period 2, where there are two refueling outages, rather than during period 3 where there is only one refueling outage scheduled. FPC stated that the same number of examinations would be conducted during the 10 year inspection interval so this approach will provide an acceptable level of quality and safety as defined in 10 CFR 50.56a(a)(3)(i).

I The staff has evaluated FPC's proposed use of Code Case N-598 during the third 10-year inspection interval of CR-3. The purpose for adopting the code case is to perform more inspections during the second period of the interval, which has two planned outages, and take maximum credit for the percentage of inspections permissible by the code case in order to reduce the number of ISI related critical path items during the third period of the inspection interval, which has only one planned outage. As seen from the above tables, the 1986 Section XI Code allows a maximum credit of 67% of inspections in each Examination Category of the Code by the end of second period, where as, the code case would allow a maximum credit of 75% inspections in each examination category during the same period. The staff, therefore, focused its evaluation on any resulting compromise of safety due to the increase in maximum permissible percentage allowed under Code Case N 598. The staff obtained clarification from the licensee that as a result of the requested 8% increase in inspection that would be credited in the second period of the current interval, there will be no increase in time lapse between successive inspections of components in each examination category beyond 10-inspection years. The staff believes that the resulting change in percentage of inspections credited in the second period of the third interval has no impact on safety. Based on industry records pertaining to ISI of plant components, the 1998 Edition of ASME Code,Section XI, has also incorporated the alternative requirements on percentages specified in Code Case N-598.

The staff finds that the attemate provisions of Code Case N-598 would provide an acceptable level of quality and safety. Therefore, the licensee's proposed attemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of Code Case'N-598 is authorized for the third 10-year interval at CR-3 or until such time as Code Case N 598 is approved for general use by reference in a future revision of Regulatory Guide 1.147. After that time, if the licensee intends to continue to implement Code Case N-598, the licensee must follow the conditions, if any, specified in the RG.

Reauest for Relief No. 98-010-II. Revision 1:

ASME Code,Section XI, requires testing of all snubbers and relief valves under the rules in IWA-4000 and IWA 7000.

The licensee proposed to use the requirements identified in Code Case N-508-1 for the purpose of testing snubbers and relief valves that are rotated from stock and installed on components (including piping systems) within the Section XI boundary.

The staff reviewed the November 30,1998, submittal, and determined that additional information was necessary in order to complete the review. The information requested for Relief Request 010-ll was (1) provide a detailed description of the snubber and pressure relief valve replacement plans utilizing the Code Case, (2) provide a clarification of FPC's intention to comply with "same design and construction," as indicated in item (a) of the code case, (3) provide a clarification of what would be considered repair or replacement of both removed mechanical and hydraulic snubbers, as far as the use of an NIS-2 form is concemed, (4) provide a discussion of the potential impact on the required plant snubber service life

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monitoring program due to the use of the Code Case, and, (5) provide a discussion of the traceability of removed snubbers from their supported systems, as a result of uslig the code case, and its impact on the ability to trace a potential affected system, which has previously contained such a removed snubber that eventually falls a functional test.

The licensee responded to the request for additional information in a letter dated May 12,1999. 1 In their response, the licensee stated that the Code Case replacement plan would be to s' elect a 1 anubber or pressure relief valve for testing pursuant to the CR-3 anubber or pressure relief valve testing program. The licensee would have the option of using previously installed snubbers or relief valves, of the same design and construction, from spares. The required testing and removal / replacement rotation would be controlled using a CR-3 maintenanc,e work  ;

request (WR). The WR would be the method of tracking the required inspections and tests, as well as_ documenting the serial number and location of both the removed and replacement 4 anubber or relief valve if the removed snubber or relief valve is found to be visually rejectable or have indications of failure, or requires repair or replacement activity, then the Code Case would not apply and the requirements of Section XI, IWA-4000 or IWA-7000 would apply.

FPC defines "same design and construction" as a functionally identical snubber of the same manufacturer or different manufacturer models which are functionally identical with only minor configuration or end attachment differences that can be installed without requiring a design or configuration change, same type (i.e., hydraulic), same size or capacity, and the same operational and design criteria. For relief valves, "same design and construction" is a functionally identical relief valve of the same manufacturer or different manufacturer models I which are functionally identical and exhibit the same technical and physical characteristics as the original, or is manufactured to the same national standard (s) governing all critical attributes of the item. This includes the same description, manufacturing controls, same size or capacity, and the same operational and design criteria. Because of this functionalidentity pertaining to both the snubbers and the relief valves, the safety characteristics of the proposed replacement l plan remains unchanged and thus is acceptable to the staff. '

The licensee also clarified that any repair or replacement of a removed snubber or relief valve would be performed in accordance with IWA-4000 or IWA 7000 and the applicable Section XI documentation would be generated. FPC has a 100% population of hydraulic anubbers and has no plans to use mechanical snubbers. Repairs under Section XI consist of welding, brazing, or metal removal and would not generally be associated with a hydraulic snubber.

Replacements will be limited to the snubber itself and any load bearing members or parts of the snubber. This criteria would include "hard parts'such as spherical bushings, snubber housings, or tie rods, and would exclude " soft parts'such as gaskets, seals, retaining rings, washers, wear shoes, shims, or hydraulic fluids. Because repairs and replacements, as described, preserve the functional characteristics of the affected parts, and thus do not degrade safety, the proposed replacement plan is acceptable to the staff.

Traceability of removed snubbers is maintained within the computerized work control system by l the use of WRs. WRs will record the removed and replaced snubber serial numbers along with the applicable work functions and completions. All planned and scheduled work, as well as the recording of the associated work completions, is also tracked by both procedural and applicable work request sign off steps. Snubbers are tracked by unique identifiers including serial number and location (unique tag / mark numbers). By searching on these unique identifiers in the computerized work control system, work associated with a snubber and/or snubber location can  !

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be determined. Snubbers are also tracked by individual serial number in the snubber Service Life Monitoring Program independently from the Section XI Repair / Replacement Program.

Additionally, all individual snubber and location information, service life, maintenance and testing history of a snubber will be maintained in a separate snubber data management system.

This allows the tracking and trending of inspection and maintenance results and the individual traceability to a location or a snubber, including the visual, functional, rebuilding and removal / replacement information and results. Based on the above, the staff concurred with the licensee that using the Code Case will provide sufficient capability of information retention for the existing plant snubber service life monitoring program and thus will provide an adequate level of quality and safety. The staff finds that the proposed use of Code Case N-508-1 provides an acceptable level of quality and safety. Therefore, the licensee's proposed '

altemative is authorized pursuant to 10 CFR 50.55a(a)(3)(i). The use of Code Case N 508-1 is authorized for the third 10-year interval at CR 3 or until such time as Code Case N-508-1 is approved for general use by reference in a future revision of RG 1.147. After that time, if the licensee intends to continue to implement Code Case N 508-1, the licensee must follow the conditions, if any, specified in the RG.

Raouest for Relief No. 98-001-PT:

Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee requested authorization to use Code Case N-4981, Attemative Rules for 10-Year Hydrostatic Pressure Testing for Class 1,2, and 3 Systems, Section X/, Division 1. Code Case N-498-1 was approved for general use by reference in RG1.147, Revision 12 dated May 1999. The licensee must follow the conditions, if any, specified in the RG.

Reauest for Relief No. 98-002-PT:

Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the attemative requirements contained in Code Case N-416-1 in lieu of the required Code examination. Code Case N-416-1 was approved for general use by reference in RG 1.147, Revision 12, dated May 1999. The licensee must follow the conditions, if any, specified in the RG.

Reauest for Relief No. 98-003-PT:

Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the altamative requirements  ;

contained in Code Case N-522, Pressure Testing of Containment Penetration Piping. Code i Case N-522 was approved for generai use by reference in RG 1.147 Revision 12, dated May

- 1999. The licensee must follow the conditions, if any, specified in the RG. i Raouest for Relief No. 98-004-PT: l ASME Code,Section XI, lWA-5250(a)(2) requires that if leakage occurs at a bolted connection in ASME Section XI components, the bolting shall be removed, VT-3 examined for corrosion, and evaluated in accordance with IWA-3100.

Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the altamative requirements contained in Code Case N 566, Corrective Action for Leukage identified at Bolted Connections.

The licensee stated:

" Florida Power Corporation will comply with the requirements of ASME Code Case N-566." '

in accordance with IWA 5250(a)(2) of the 1989 Editloo 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's proposed I altemative is to implement the requirements of Code Case N 566, Corrective Action for Leakage Identified at Botted Connections.

Code Case N-566 requires that leakage be stopped, and the bolting and component material be reviewed for joint integrity. 'However, the Code Case does not define specific and detailed requirements to be included in the " review" to determine joint integrity.

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Code Case N 566 also requires that, if the leakage is not stopped, the joint integrity be evaluated in accordance with IWB-3142.4, Acceptance by Analytica/ Evaluation. This type of Code evaluation should include consideration of the number and condition of bolts, leaking medium, bolt and component material, system function, and long-term leakage monitoring.

However, an IWB-3142.4 evaluation presupposes that the botting has been examined, and all unacceptable relevant conditions accounted for in the analytical evaluation. Since the bolting is not being removed, the Code Case does not provide a comprehensive evaluation process for ensuring joint integrity. Consequently, the staff determined that Code Case N-566, as written, does not provide an acceptable level of quality and safety. Therefore, the licensee's proposed altemative is not authorized.

Reauest for Relief No. 98-005-PT: i ASME Code,Section XI, IWA 5242(a) requires that insulation be removed from pressure-retaining bolted connections for VT-2 visual examination in systems borated for the purpose of controlling reactivity.

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Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use Code Case N-533, Altemative Requirements for VT-2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted Connections,Section XI, Division 1.

The licensee stated:

1

" Florida Power Corporation will comply with the requirements of ASME Section XI Code Case N-533. The following requirements will be met for systems borated for the purpose of controlling reactivity, a) "A system pressure test and VT-2 visual examination will be performed each refueling outage without removal of insulation, b) "Each refueling outage the insulation will be removed from the bolted connection, and a VT-2 visual examination will be performed. The connection is not required to be pressurized. Any evidence of leakage will be evaluated in accordance with IWA 5250."

The Code requires the removal of all insulation from pressure retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. As an altemative, the licensee has proposed to use Code Case N-533, which includes a system pressure test and subsequent direct visual examination with the insulation removed. This direct visual examination is performed without pressuriang the system, and both of these examinations are required once each refueling outage.

Code Case N-533 does not provide details of the examination parameters for the system l pressure test or VT-2 visual examination with insulation in place. The staff determined that the system pressure test and corresponding VT-2 visual examination with the insulation in place is to be performed with a minimum 4-hour hold time after attaining a test pressure of not less than the nominal operating pressure associated with 100% rated reactor power. The 4-hour hold allows time for leakage to penetrate the insulation, providing a means of detecting any significant leakage with the insulation in place. In its May 12,1999, response, the licensee committed to perform visual examination with a minimum 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time. Therefore, the licensee's proposed attemative, to use Code Case N-533, with the additional condition that the test pressure would be maintained for a minimum 4-hour hold time, provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(1). The use of the Code is authorized for the current interval or until such time as the Code Case is published in a future revision of RG1.147. At that time, if the licensee intends to continue to implement Code Case N-533, the licensee should follow all provisions in Code Case N-533 with limitations issued in RG1.147, if any.

Reauest for Relief No. 98-001-SS ASME Code,Section XI, Article IWF 5000, requires all safety-related snubbers to be examined and tested in accordance with the 1988 Addenda to ASME/ ANSI OM 1987, Part 4. Section 2.3.2.2 of Part 4 requires snubbers to be examined every 18 months (* 25%) and functionally tested every refueling outage.

The licensee requested relief from the snubber visual examination and functional testing requirements of the Code. Specifically, relief was requested from performing visual examinations every 18 months (* 25%) and functional testing at least every refueling outage assuming an 18 month refueling interval.

FPC proposed that the required Inservice examination and testing of snubbers be based upon a refueling interval of 24 months and allow the frequency of snubber visual examinations to be extended up to 48 months (* 25%) contingent upon the results of the examinations, consistent with the provisions of Generic Letter (GL) 90-09, "Altemative Requirements for Snubber Visual inspection Intervals and Corrective Actions."

FPC has implemented the use of the provisions of GL 90-09, based upon a continuing refueling interval of 24 months since the 1992 refuel;ng outage (RFO-8). The required Inservice examination and testing of snubbers has been conducted in accordance with the requirements of ths 1983 ASME Section XI Code approved for FPC as part of the Second 10-Year Interval inservice Inspection Program.

I

For the Third 10-Year Interval ISI Program, the ASME B&PV Code,1989 Edition,Section XI, Article IWF-5000, implements the requirements of Part 4 of the first Addenda to ASME/ ANSI OM-1987 (published in 1988). The schedule for functional testing is specified to be at least each refueling outage and the schedule for visual examinations specifies a maximum interval of 18 months (* 25%). These schedules assume that the refueling intervals will not exceed 18 months. However, the refueling interval for CR-3 is 24 months. Recognizing the need for an attemative visual inspection schedule that would generally allow the performance of visual inspections and corrective actions during plant outages, the NRC developed GL 90-09.

GL 90-09 provides an altamate acceptable examination and testing schedule while maintaining the same confidence level as the previously existing schedule. The attemative inspection schedule specified in GL 90-09 is based on a fuel cycle of up.to 24 months and on the number of unacceptable snubbers found during the previous inspection interval in proportion to.the sizes of the various snubber populations or categories.

The results of previous inspections have been reviewed and are provided below. These results show that FPC has maintained a high level of operability within the examination and testing program.

Visual Examination Results e Mid-Cycle Outage 8: 100% of as found safety-related visual examinations were performed with no recorded failures (267 inspections).

• RFO-8: 100% of as-found safety-related visual examinations were performed with no recorded failures (267 inspections).

• RFO-9: 100% of as-found visual examinations were performed with no recorded failures (255 inspections due to snubber reduction).

• RFO-10: As a supplemental scope in support of functional testing and removal of snubbers under a modification,56 as found visual examinations were performed with no recorded failures.

Functional Testina Results e RFO-8: 16% of the population were functionally tested (44 anubbers). No functional failures were noted.

• RFO-9: 10% of the population were functionally tested (27 anubbers). No functional failures were noted.

• RFO 10: 10% of the population were initially functionally tested. Due to one functional failure (evaluated as improper cleanliness during maintenance rebuild-lint trapped in the poppet valve), an expanded sample was tested for a total of 41 anubbers. No other functional failures were noted.

FPC proposed to implement an inspection and testing schedule in accordance with the provisions of GL 90-09, based on a refueling interval of 24 months (* 25%) to demonstrate the

, required functional and operable integrity of snubbers. The licensee stated that the use of this examination and testing program, based on a 24 month refueling outage interval that incorporates NRC GL 90-09 provisions, would substantially reduce non beneficial work activities and personnel radiation exposure, provide a continued acceptable level of quality and safety, and have no adverse impact on public health and safety.

I 1

The ASME Code,1989 Edition,Section XI, Paragraph IWF-5000 requirements for visual inspection and functional testing of snubbers are contained in Paragraph IWF 5300,

" INSERVICE EXAMINATIONS AND TESTS." Paragraph IWF-5300 defers these requirements I to the first addenda to ASME/ ANSI OM-1987, Part 4, which was published in 1988. l GL 90-09 provides an altamative schedule for snubber visual inspections that maintains the same confidence level as the Code required inspection levels and allows for inspections and corrective actions during plant outages. The GL guidance on the inspection interval is based on the number of unacceptable snubbers identified in the last inspection in proportion to the size of the various snubber populations and categories. The Code-required interval is based only on the number of unacceptable snubbers identified during the last inspection without regard to the population. Data provided by the licensee indicates that following this inspection program since 1992 has not resulted in an increase in the failure rate of snubbers The staff has determined that the visual inspection schedule included in GL 90-09 is an acceptable attemative to the Code requirements.

The staff finds that compliance with certain ISI requirements would result in hardship on the licensee without a compensating increase in the level of quality and safety, in that a plant shutdown would be required in order to conduct the inspection and testing. The attemative  ;

inspection and testing of safety-related snubbers will provide adequate assurance of snubber '

functionality. Thus the proposed alteinative is acceptable pursuant to 10 CFR 50.55a(a)(3)(li).

3.0 CONCLUSION

The staff concludes that the licensee's proposed alternatives contained in Requests for Relief l Nos.98-001 il,98-004-ll,98-005-11,98 008-11,98-009-ll,98-010 II, and 98-005-PT provide an l acceptable level of quality and safety. Therefore, the licensee's proposed attematives are authorized for the third interval pursuant to 10 CFR 50.55a(a)(3)(1). Requests for Relief Nos.

98-009-11,98-010-11, and 98-005-PT that contain Code Cases N-598, N-508-1, and N-533 are Authorized until such time as the Code Cases N-598, N-508-1, and N-533 are published in a future revision of Regulatory Guide 1.147. At that time, if the licensee intends to continue to implement Code Cases N-598, N-508-1, and N-533, the licensee should follow all provisions in the subject Code Cases with limitations issued in RG 1.147, if any.

The staff concludes that for Requests for Relief Nos. 98-002-II,98-003-II, and 98-001-SS, l certain Isis required by Code Section XI would result in hardship for the licensee without a compensating increase in the level of quality or safety. The attemative contained in Requests for Relief Nos. 98-00211,98-003-ll, and 98-001 SS are authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

For the attematives contained in Requests for Relief Nos. 98-006-11 and 98-004-PT, the licensee's proposed attematives do not provide an acceptable level of quality and safety.

Therefore, the alternatives contained in Requests for Relief Nos. 98-006-ll and 98-004-PT are not authorized. Request for Relief No. 98-007-11 was withdrawn by the licensee in its letter dated May 12,1999.

^

V in May 1999, the NRC approved for general use, by reference in RG 1.147, Revision 12, certain Code Cases approved by the ASME. Requests for Relief 98-005-ll,98-001-PT,98-002-PT and 98-003 PT requested authorization for Code Cases covered by RG 1.147, Revision 12, and, thus, do not require further NRC authorization. FPC must follow the conditions, if any, specified in the RG for these Code Cases.

Principal Contirbutor: Thomas K. McLellan Prakash Patnaik Amold Lee Leonard A.Wiens Date: August 5, 1999 -

Attachment:

INEEL Technical Letter Report

TECHNICAL LETTER REPORT ON THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM REQUESTS FOR RELIEF EDE FLORIDA POWER CORPORATION CRYSTAL RIVER. UNIT 3 DOCKET NUMBER: 50-302

1. INTRODUCTION By [[letter::3F0498-03, Forwards ten-year ISI Program Manual for Third Insp Interval for Crystal River,Unit 3.Manual Updated to Meet Requirements of 1989 Edition of ASME Boiler & Pressure Vessel Code IAW ASME Section Xi,Except Where Specific Relief Is Requested|letter dated April 16,1998]], the licensee, Florida Power Corporation, submitted its third 10-year interval inservice inspection (ISI) program, with associated requests for relief from the requirements of the ASME Code,Section XI, for Crystal River, Unit 3, third ISI interval. 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 Florida Power Corporation 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 Crystal River, Unit 3, third 10-year ISI interval, which began August 14,1998, is the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code.

A. Reauest for Relief No. 98-001-11. Examination Cateoorv B-F. Item B5.10. Pressure-Retalnino Dissimilar Welds Code Reauirement: Examination Category B-F, item B5.10 requires 100% volumetric and surface examination of pressure-retaining dissimilar metal welds, as defined by Figure IWB-2500-8, each inspection interval.

Licensee's Proposed Altemative: In accordance with 10 CFR 50.55a(s)(3)(i), the licensee proposed to perform the volumetric examination of the outside surface from the nozzle bore on the following RPV nozzle-to-safe end butt welds.

Location Identification #

~

'W' axis of Reactor Vessel B1.6.1 (UT) 81.6.2 (PT)

Y axis of Reactor Vessel B1.6.3 (UT)

B1.6.4 (PT)

Attachment

The licensee stated:

" Florida Power Corporation will perform a full volumetric examination of both Core Flood Nozzle to Safe End welds from the inside diameter using the enhanced ultrasonic technique as described in Babcock and Wilcox Owners Group (BWOG)

Topical Report BAW-228P, Revision 1. This ultrasonic examination technique has been qualified to detect outside diameter surface cracking."

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

"As an attemative to performing the surface examination on the core flood nozzles, FPC will perform an automated volumetric examination of the outside surface from the inside diameter of the nozzle. BWOG Topical Reports, BAW-2228 A a,nd BAW-2228P, Revision 1, describes a modified fracture mechanics analysis applied to postulated outer surface, semi-elliptical circumferential reactor vessel nozzle to pipe weld flaws. It demonstrates that adequate margins exist between the revised bounding allowable limit flaw size and the minimum flaw that was detected and sized during the Babcock and Wilcox Nuclear Technologies (BWNT) demonstration.

"FPC has previously reviewed both Topical Report BAW-2228-A and Topical Report BAW-2228P, Revision 1, for applicability to Crystal River Unit 3. FPC confirmed that the input stresses under normal / upset and emergency / faulted conditions shown in BAW-2228P, Revision 1, are applicable to CR-3 and that the materialis A508 Class 2, carbon steel.

"The proposed altamative examination provides an acceptable level of quality and safety as defined by 10CFR50.55a(a)(3)(i).

"The NRC has reviewed the BWOG Topical Report, BAW-2228P, Revision 1, and issued a Safety Evaluation Report, dated March 21,1996."

Evaluation: The Code requires 100% volumetric and surface examination of the subject nozzle-to-safe end welds. The licensee has proposed to perform an ultrasonic examination of the outside surface from the inside diameter of the nozzle in lieu of the Code-required surface examination. The technique proposed by the licensee was demonstrated to the Nuclear Regulatory Commission (NRC) by the Babcock and Wilcox Owner's Group (BWOG) on August 11,1993. The BWOG prepared Topical Report:

BAW 2228P, Revision 1. This report describes a modified fracture mechanics analysis applied to postulated outer surface, semi-elliptical circumferential reactor vessel nozzle-to-pipe weld flaws. It also demonstrates that adequate margins exist between the revised allowable limiting flaw size and the minimum flaw that was detected and sized during the demonstration. Therefore, the subject report provides a basis for substituting the ultrasonic examination for the surface examination.

The NRC has reviewed the BWOG Topical Report: BAW 2228P, Revision 1, and accepted the results in a Safety Evaluation Report dated March 21,1996. The NRC concluded that the enhanced ultrasonic technique is an acceptable attemative to the Code-required surface examination provided that the following conditions are met:

3

1) stresses for the subject nozzle-to. pipe welds are bound by the input stresses under normal / upset and emergency #sulted conditions shown in Tables 1 to 4',and

2) the material for the core flood nozzles (nozzles of interest) is either A508 Class 2 cart >on steel or SA 336 (316) stainless steel.

The licensee has confirmed that the input stresses under normal / upset and emergency #aulted conditions shown in BAW-2228P, Revision 1, apply to Crystal River, '

Unit 3, and that the materialis A508 Class 2, carbon steel. Therefore, based on the licensee's confirmation of the above conditions, the licensee's proposed volumetric examination in lieu of the required surface examination should provide an acceptable level of quality and safety. Therefore, it is recommended that the proposed altemative examination be authorized pursuant to 10 CFR 50.55a(s)(3)(i).

B. Reauest for Relief No. 98-002-II. Examination Cateaorv B-L-1. Item Number B12.10. l Pressure-Retainina Welds in Pumo Casinas Code Recuirement Examination Category B-L-1, item B12.10 requires 100% volumetric examination of pump casing welds, as defined by Figure IWB-2500-16, each inspection interval. Examinations are limited to at least one pump in each group of pumps performing similar functions in the system.

Licensee's Proposed Altemative In accordance with 10 CFR 50.55a(s)(3)(i), the licensee proposed to perform VT-2 visual examinations, VT-3 visual examinations, and surface examinations on 25% of the upper and lower pump case scroll welds on pump RCP-1A in lieu of the Code-required volumetric examination. The licensee stated:

" Visual (VT-2) examinations will be performed during system leakage tests conducted prior to plant startup following each refueling outage. VT-3 examination will be conducted on the intemal surface of at least one pump if disassembled for maintenance during the interval, as required in Category B L-2.

"A surface examination will be performed on 25% of the length of both the upper and lower pump case scroll welds on one pump (RCP-1A) during the third inspection interval. The 25% segment selected for inspection will be a segment that has not been previously inspected during the first or second inspection intervals."

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

" Conducting volumetric examination of reactor coolant pump casings would involve significant radiation exposure without an increase in quality or safety.

"The pump casings are 316 cast stainless steel. There have been no reported instances of pump casing weld degradation in the nuclear industry."

'The altomative examinations were performed during the second inspection and no unacceptable indications were noted."

Tatdes provided in the Topical Report are not included with this evaluation.

I Evaluation: The Code requires 100% volumetric examination of the pump casing welds )

on at least one pump in each group of pumps performing similar functions in the system each inspection interval . The licensee proposed to perform a VT-2 visual examination during the system leakage tests conducted prior to startup following each refueling outage in accordance with Examination Category B-P requirements, a VT-3 visual examination on the intemal surface of at least one pump in accordance with Examination Category B L-2 requirements, and a surface examination on 25% of the length of both the upper and lower pump case scroll welds on pump RCP-1 A during the interval.

The licensee requested relief under 10 CFR 50.55a(s)(3)(i), which requires that a proposed altemative be presented that provides an acceptable level of quality and safety, i.e. a level of quality and safety that is equivalent to the protection provided by the Code-required examinations. The INEEL staff does not believe the licensee's proposed attemative provides an equivalent means of examination as compared to the Code's volumetric examination requirements. Therefore, it is recommended that the licensee's proposed attemative not be authorized.

C. Reauest for Relief No. 98-003-il. Examination Catenorv F-A. Item Number F1.30.

Reactor Vessel Sucoort Skirt Code Reauirement: Examination Category F-A, item F1.30 requires a VT-3 visual examination of weld and mechanical connections at intermediate joints in multi-connected integral and non-integral supports, as defined by Figure IWF-1300-1.

Licensee's Proposed Attemative: In accordance with 10 CFR 50.55a(a)(3)(i), the licensee proposed to perform an altamative VT-3 visual examination on Reactor Pressure Vessel Support Skirt B1.12.1. The licenses stated:

"VT-3 examination will be performed on 10% of the interior of the reactor vessel support at three positions along the circumference of the support. The same segments of the support skirt, located 120' spart, that were previously examined in the first and second inspection intervals, will be examined in the third inspection interval."

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

" Relief is requested based on the radiation conditions below the reactor vessel which would result in significant radiation exposure to personnel. In order to perform a 100% visual examination, significant preparatory work involving erection of temporary lighting, scaffolding, surface preparation, etc. would be required.

This work would be conducted around the entire circumference of the reactor vessel support skirt interior. Based on previous exposure levels, it is estimated that radiation dose for all personnel involved could exceed 30 person REM.

" Previous visual examinations performed during the first and second inspection intervals indicated no degradation of the support skirt. The locations selected (approximately 120' apart) are the least restrictive in providing access to the support skirt weld area without creating a potential for damage to the incore guide j tubes. Examinations performed at these locations will adequately reveal the i l

condition of the reactor vessel support skirt without subjecting personnel to unnecessary radiation exposure.

l

"The proposed altemative examination provides an acceptable level of quality and safety as defined by 10CFR50.55a(s)(3)(i)."

Evaluation: The Code requires that a VT-3 visual examination of weld and mechanical connections at intermediate joints in multi-connected integral and non-integral supports each inspection interval. The licensee has proposed that a VT-3 visual examination be performed on 10% of the interior of the reactor pressure vessel support at three positions along the circumference of the support..

The licensee has requested relief under 10 CFR 50.55a(a)(3)(i), which requires that a proposed altemative be presented that provides an acceptable level of quality arid safety, i.e. a level of quality and safety that is equivalent to the protection provided by the Code-required examinations. The INEEL staff does not believe the licensee's proposed altemative provides a means of examination that is essentially equivalent to the F-A Code requirements. Furthermore, licensees must make a reasonable effort to maximize examination coverage. This includes consideration and use of additional examination techniques (e.g., remote /robotic visual examination). The licensee has not provided any information describing additional examination techniques that may have been considered to maximize examination coverage. Therefore, it is recommended that the licensee's proposed altamative not be authorized.

D. Reauest for Relief No. 98-004-il. Examination Cateoorv B-0. Item Number B14.10.

Welds in CRD Housino Code Reauirement Examination Category B-0, item B14.10 requires volumetric or surface examination on 10% of peripheral CRD housing welds, as defined by Figure IWB-2500-18, each inspection interval.

Licensee's Proposed Alternative In accordance with 10 CFR 50.55a(s)(3)(i), the licensee proposed to perform volumetric or surface examinations on three welds of each of the six CRDM housings that are removed from the reactor vessel head for maintenance during the third inspection interval.

Licensee's Basis for Prooosed Attemative (as stated):

"There are 24 peripheral Control Rod Drive Mechanisms on the CR-3 reactor vessel head. Examination of 10% of these CRDM's, would require three CRDM l housings receive either a volumetric or surface examination. Each CRDM housing I contains five Category B-O welds, three above the reactor vessel to CRDM mating flange and two below. A total of 15 welds would be examined in accordance with the current Code requirements. The proposed altemative examination of performing a volumetric or surface examination on six CRDM housings removed from the reactor vessel head would yield a total of 18 welds examined in the third inspection interval. This is a not increase above minimum code requirements of l three welds or 20%. I "By performing the CRDM weld examination on drive assemblies removed from the reactor vessel head for scheduled maintenance, the occupational exposure to personnelis greatly reduced. A saving of approximately 12 REM would be achieved over the ten year inspection interval. .

l I

"The proposed altamative examination provides an acceptable level of quality and safety as defined by 10CFR50.55a(a)(3)(i) and 10CFR50.55a(a)(3)(ii)."

Evaluation The Code requires 100% volumetric or surface examination of the welds on 10% of the peripheral CRD housings. Review of the licensee's documentation shows that there are 24 peripheral CRDMs on the reactor vessel head. Therefore, the licensee is required to examine the welds of three of the peripheral CRD housings. Each CRD housing has five welds, two welds below the CROM flange connection and three welds above it. Consequently, the licensee is required to examine at least 15 welds. The licensee has proposed to perform volumetric or surface examinations of three welds on each of six CRDM housings that will be removed from the reactor vessel head fof maintenance during the interval. Therefore, the total number of walds to be examined will be 18, an increase of 20% over the required examination samA.

The licensee occasionally removes CRDM housings for maintenance by disconnecting the housings at the flange assembly. The three welds above the flange assembly are removed from the reactor head with the CRDM housing. The licensee is then able to perform the required examinations in an area with lower radiation. The licensee has determined that this procedure reduces the radiation dose by12 REM. Based on the 20% increase of Code item B14.10 welds to be examined, and the reduction in radiation exposure to inspection personnel, the INEEL staff believes that the licensee's proposed altemative provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed attemative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

E. Reauest for Relief No. 98-005-II. Use of Code Case N-524. Altemative Examination Reauirements for Lonaitudinal Welds in Class 1 and 2 Pinina Code Reauirement Examination Categories B-J, C-F-1, and C-F-2, items B9.12, C5.12, C5.22, C5.52, and C5.62, require surface and volumetric examination of Class 1 and 2 longitudinal piping welds. Items 89.22, C5.42, and C5.82 require surface examination of Class 1 and 2 longitudinal piping welds. The examination volume / surface area of each longitudinal weld includes 12 inches (for Class 1) or 2.5t (for Class 2), measured from the intersection of the circumferential weld.

Licensee's Proposed Attemative: In accordance with 10 CFR 50.55a(s)(3)(i), the licensee proposed to use Code Case N-524, Altemative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping. The licensee stated:

"The following altamative examination requirements will be performed as defined in ASME Code Case N-524.

a) "When only a surface examination is required, examination of longitudinal piping welds is not required beyond those portions of the welds within the examination boundaries of intersecting circumferential welds.

b) "When both surface and volumetric examinations are required, examination of longitudinal piping welds is not required beyond those portions of the welds within the examination boundaries of intersecting circumferential welds provided the following requirements are met:

1

1) "Where longitudinal welds are specified and locations are known, examination requirements shall be met for both transverse and parallel flaws at the intersection of the welds and for that length of longitudinal weld within the circumferential weld examination volume;

2) "Where longitudinal welds are specified but locations are unknown, or the existence of longitudinal welds is uncertain, the examination requirements shall be met for both transverse and parallel flaws within the entire examination volume of intersecting circumferential welds."

Licensee's Basis for Proposed Altemative (as stated): -

'FPC is requesting relief from the above stated requirements based on ASME Code Case N-524 which has been issued by the American Society of Mechanical Engineers and has been included in the 1995 Addenda of Section XI. Code Case N 524 defines altamative examination requirements that may be applied to surface and volumetric examination of longitudinal welds in Class 1 and Class 2 piping and provides an acceptable level of quality and safety as defined by 10CFR50.55a(s)(3)(i).

" Code Case N 524 directs examination efforts at the high risk area at weld intersections. By eliminating the low risk areas of longitudinal welds from examination, the time requirements and radiation exposure of personnel are sigrdficantly reduced. The expected dose savings is estimated to be 1.6 to 2.0 man-Rom per weld with a projected outage savings of 12.8 to 16.0 man Rem based on examination of eight longitudinal welds. Compliance with the existing ASME Sechon XI requirements, in lieu of Code Case N-524, would result in unnecessary exposure without a compensating increase in the level of quality or safety.

" Longitudinal welds are produced during pipe fabrication, as opposed to circumferential welds which are field produced. The ASME Code contains requirements for characteristics and performance of materials and components, and for exs,mination of longitudinal piping welds during fabrication. Additionally, the ASME Code specifies the minimum chemical and physical properties of the material to ensure structuralintegrity of the longitudinal welds 6t the time of pipe fabrication.

"The preservice, and inservice examinations conducted during the first inspection interval provide assurance of longitudinal weld structural integrity for the service life of the plant to date.

" Experience within the United States reveals that ASME Code longitudinal welds have not experienced degradation warranting inservice examination beyond that required to meet the circumferential weld examination requirements. To date, no significant loading conditions or material degradation mechanisms have become evident that specifically relate to longitudinal seam welds in nuclear power plant piping, if degradation of a longitudinal weld was to occur, it is expected to be located at an intersechon with a circumferential weld, which is inspected in l

8-accordance with Code Case N-524. Therefore, the health and safety of the public will continue to be maintained while implementing the altamative examination requirements of Code Case N-524."

Evaluation: 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 i licensee's proposed altamative is to examine only the portions of longitudinal weld contained within the examination volume of the intersecting circumferential weld. This altemative is contained in Code Case N-524, Attemative 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 are circumferential welds. Consequently, the welds are fabricated under the strict guidelines specified by the manufacturing standard, which provides assurance of structural integrity. These welds have also been subjected to the preservice and initial inservice examinations, as applicable, which provide additional assurance of structural integrity. To date, no significant loading conditions or material degradation mechanisms have been identified that specifically relate to longitudinal seam welds in Class 1,2, and 3 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 licensee's altamative provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed altamative be authonzed pursuant to 10 CFR 50.55a(a)(3)(i). The use of Code Case N-524 should be authorized for the third 10 year interval at Crystal River Unit _

3, 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.

F. Reauest for Relief No. 98-006-11. Use of Code Case N 509. Attemate Rules forthe Selection and Examination of Class 1. 2. and 3 Intearallv Welded Attachments Code Reauirement: The Code requires examination of integrally-welded attachments as specified for Examination Categories C-C, D A, D-B, and D-C. The Code stipulates volumetric or surface examinations, as appropriate, and the extent of examinations.

Licensee's Proposed Altemative: Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee has proposed to use portions of Code Case N-509, Altemate Rules forthe Selection and Examination of Class 1, 2, and 3 Integrally Welded Attachments in lieu of the requirements of the Code for Class 2 and 3 integrally-welded attachments. The licensee stated:

"The following altemative examination requirements will be implemented as defined by ASME Code Case N-509, exclusive of the selection criteria which is defined below:

"ASME Code Case N 509 and Table 2500-1, Examination Category C-C, shall be used for Class 2 integrally welded attachments to piping as defined by Examination Category C-C of ASME Section XI, Article IWC.

I I

9 "ASME Code Case N-509 and Table 25001 Examination Category D A, shall be used for Class 3 integrally welded attachments to piping as defined by Examination Categories D-A, D B, and D-C of ASME Section XI, Article IWD.

, "In addition to the requirements of ASME Code Case N-509, the extent of scheduled examination shall be a minimum of 10% of the total population (no thickness exemption) of integral attachments in non-exempt Code Class 2 and 3 systems."

1 1.icensee's Basis for Proposed Altemative (as stated): I l

"FPC is requesting relief from the above stated requirements based on ASME Code Case N-509 which has been issued by the American Society of Mechanical Engineers and has been included in the 1995 Addenda of Section XI.

"At CR-3 components supports are selected for examination per ASME Code Case N-491 which was adopted by the 1990 Addenda to ASME Section XI, Subarticle IWF-2510. Application of Code Case N-509 defines attemative examination requirements that may be applied to ASME Code Class 1,2 and 3 integrally welded attachments. The extent of examination as stated in Note 5 of Table 2500-1, Examination Categories B-K, C-C, and D-A cf Code Case N-509 reduces the required examinations to 10% of the integral attachments associated with component supports selected for examination under the 1990 Addenda to Secticn XI, paragraph IWF-2510.

"CR-3 will schedule a minimum of 10% of the total population of integral attachment to piping in non-exempt Code Class 2 and 3 systems. This approach is conservative and would provide an acceptable level of quality and safety.

" Examination of some Class 2 integrally welded attachments can result in radiation exposure of 1.6 to 2.0 mrem per examination. By limiting examinations to 10% of the total population of integral attachments to piping in non-exempt Code Class 2, and 3 systems, the time requirements and the radiation exposure of personnel would be reduced. Therefore compliance with the existing ASME Section XI 1989 requirements would result in higher exposure without a compensating increase in the level of quality or safety.

" 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 l 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 altamative examination requirements of Code Case N-509."

i Paragraph 6.22 of the Program Plan states:

I i

"The examination requirements of the 1989 Edition of ASME Code,Section XI will be used for all non-exempt integral attachments on pumps and valves during the third inspection interval.

"ASME Code Case N-509 requires that integrally welded attachments for vessels be examined during each inspection interval and allows, in note (4), only one of multiple vessels to be selected for examination. For piping Code Case N-509 only requires a sample of 10% of the welds associated with the component supports selected for examination under the 1990 Addenda. FPC will examine a minimum of 10% of the total number of Class 2 integrally welded attachments."

Evaluation: The Code requires examination of integrally-welded attachments, aE specified, for Examination Categories C-C, D-A, D-B, and D.C. The Code stipulates volumetric or surface examinations, as appropriate, and the extent of examinations. The licensee has proposed to apply the requirements of Code Case N-509 as an altamative to the Code requirements for the examination of integrally welded attachments on Class 2 and 3 piping. The licensee has also committed to supplement the Code Case with a minimum examination sample of 10% of all integral attachments to non-exempt Class 2, and 3 systems.

The Code does not require examination of Class 1 integrally welded attachments in the third or fourth intervals. However, Code Case N-509 is written for Class 1,2, and 3 integrally welded attachments, and requires that Class 1,2, and 3 integrally welded attachments be examined each inspection interval. While Code Case N-509 significantly reduces the number of welded integral attachments to be examined, a sample of Class 1 integrally welded attachments will continue to be examined in later intervals. Reducing the Code Case N-509 examination sample by not including Class 1 components reduces the effectiveness of the Code Case. It is the INEEL's opinion that Code Case N-509 must be implemented in its entirety. The INEEL staff believes that the licensee's proposed altamative, to implement Code Case N-509 for Class 2 and 3 welded integral attachments only, is non conservative and does not provide an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed attemative not be authorized.

G. Reauest for Relief No. 98-007-II. Examination Cateoorv B-D. Item Numbers B3.90 and B3.100. Full Penetration Wolds of Nonles in Vessels Code Reauirement: Section XI, Table IWB-2500-1, Examination Category B-D, items B3.90 and B3.100 require that, for reactor pressure vessel (RPV) noule welds and inner radius sections, at least 25% but not more than 50% (credited) of the nonles be examined by the end of the first inspection period and the remainder by the end of the inspection interval.

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

licensee proposed to perform the requirod examinations by the end of the inspection interval. The licensee stated:

" Florida Power Corporation will perform the examinations, as required by Category )

B-D on all required reactor vessel Nonie-to-Shell welds and their Nonie-Inside-Radius sections by the end of the inspection interval."

The welds and examination areas are listed below.

WeldID ltem No. Last Exam Next Examination B1.4.3A - Nozzle-to-Vessel B3.90 1996 2006 B1.4.38 - Inside Radius 83.100 1996 2006 B1.4.7A - Nozzle-to-Vessel B3.90 1996 2006 Proposed 81.4.78 - Inside Radius 83.100 1990 2006 Proposed 81.4.4A - Nozzle-to-Vessel 83.90 1996 2006 81.4.48 - Inside Radius 83.100 1996 2006 B1.4.5A - Nozzle-to-Vessel 83.90 1996 2006 51.4.58 - Inside Radius B3.100 1996 2006 B1.4.6A - Nozzle-to-Vessel B3.90 1996 2006 Proposed

. 81.4.68 - Inside Radius 83.100 1990 2006 Proposed 81.4.6A - Nozzle-to-Vessel B3.90 1906 2006 B1.4.68 - Inside Radius 83.100 , i 2006 81.4.1 A - Nozzle-to Vessel 83.90 1996 2006 81.4.18 - Inside Radius 83.100 1996 2006 81.6.1 - Nozzle-to-Safe End 1996 2006 81.4.2A - Nozzle-to Vessel 83.90 1996 2006 B1.4.18 - Inside Radius 83.100 1996 2006 B1.6.3 - Nozzle-to-Safe End 1996 2006 Licensee's Basis for Proposed Altemative (as stated):

"The Crystal River Unit 3 reactor vessel has a total of eight nozzles that are classified as Category B-D. The Code requires that a minimum of 25% of these nozzles be examined by the end of the first period of the third interval. The reactor vessel design only permits examination of the two outlet nozzles without removal of the Core Support Assembly (CSA). This examination is also limited to scanning from the nozzle bore due to interference with the CSA.

"Although the scheduled welds and inside radius sections are to be examined with automated ultrasonic equipment, an additional four person-rom of radiation exposure and one to two [ cubic] feet of radioactive waste is expected.

"Six ci the eight reactor vessel nozzle to shell welds and their inside radius sections were last examined in 1996 with acceptable results. Supplemental ruminations were also performed on the two outlet nozzles during the 1996 semination. This examination consisted of scanning the weld from the reactor meal shell inside diameter. No indications associated with material degradation I were identified during these examinations. The acceptable results of these past l examinations are considered to be representative of the present reactor vessel i nozzle weld conditions at Crystal River Unit 3."

i

. Evaluation: The Code requires examination of at least 25%, but not more than 50%

(credited) of RPV nozzles and associated inside radius (IR) sections and nozzle safe ends during the first inspection interval. The licensee has proposed to defer the required examinations to the end of the third 10 Year interval. This proposed attemative is similar to Code Case N-521, Attemative Rules IbrDeferralofInspec6ons of Nozzle-to Vessel Wolds, inside Radius Sec60ns, and Nozzle-to-Safe End Welds of a Pressunzed Water Reactor Vessel.

Code Case N-521 states that examination of RPV nozzles, IR sections, and nozzle-to-safe end welds may be deferred provided (a) no inservice repairs or replacements by welding have ever been performed on any of the subject areas, (b) none of the subject areas contain identified flaws or relevant conditions that currently require successive inspections in accordance with IWB 2420(b), and (c) the unit is not in the first interval.

The licensee has not confirmed that all the above conditions have been met. An additional requirement imposed by the NRC is that all subject areas be scheduled for examination such that the new sequence of examinations will not exceed 10 (Code) years between examinations. It appears that the scheduled sequence of examinations will exceed ten years for two of the inner-radius examinations. Therefore, the licensee's proposed attemative does not provide an acceptable level of quality and safety, and it is recommended that the proposed altamative not be authonzed.

H. Recuest for Relief No. 98-008-II. Examination Catenorv B-G-1. Item B6.10. Reactor Pressure Vessel Closure Head Nuts Code Reauiremont Examination Category B-G-1, item B6.10 requires a surface examination of all RPV closure head nuts each 10 year interval.

Licensee's Proposed Altemative: In accordance with 10 CFR 50.55a(a)(3), the licensee proposed to implement the VT-1 visual examination requirements of the 1989 Addenda of Section XI for the RPV closure head nuts. The licensee stated:

" Relief is being requested to substitute a visual (VT-1) examination for the required surface examination of each RPV closure head nut during the inspection interval.

"A visual examination will be performed on all RPV closure head nuts once during tnis inspection interval, as allowed in the 1989 Addenda of Section XI."

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

"The ASME Code Committee has recognized that a surface examination is an excessive examination technique to be applied to RPV nuts. The 1989 Addenda

^

of Section XI has changed the requirement from a surface examination to a Visual VT-1 examination.

" Extensive cleaning of these nuts is required for a sudace examination to be performed. This excessive cleaning results in, additio.",41 radioactive waste, addebonal costs and the inefficient use of available manpower resources. The cleaning materials and the wet magnetic particle test solution results in ' mixed waste', i.e., contaminated and hazardous material that can not be economically processed. The unnecessary generation of ' mixed waste' is not consistent with the Radioactive Waste Management Program at CR-3. To continue to perform the

)

required surface examinations of these nuts will not provide any potential increases in plant safety margins and the additional costs for these examinations are no longer warranted based on the changes now published in the 1989 Addenda of ASME Section XI."

Evaluation: The licensee has requested relief from performing the Code-required surface examination of the RPV closure head nuts. As an altamative, the licensee proposed to perform a VT-1 visual examination as allowed in the 1989 Addenda to the 1989 Edition of ASME XI. In the 1989 Edition, Examination Category B-G-1 requires ,

that all items be examined using VT-1 examination and/or volumetric examination (as applicable), with the exception of the surface examination for RPV closure head puts and closure head studs (when removed). -

The 1989 Edition of the Code, item B6.10, does not provide acceptance criteria for surface examination of RPV closure head nuts. (At that time the acceptance criteria were in the course of preparation). Typicalindications that would require corrective ,

action on RPV closure head nuts are associated with degradation mechanisms such as {

boric acid attack, corrosion, or handling mechanisms such as galled threads and deformation. Typical surface examination procedures and techniques are not qualified to identify these forms of degradation.

Article IWB-3000, Acceptance Standard, IWB-3517.1, Visual Examination, VT-1, describes conditions that require corrective action prior to continued service of botting and associated nuts. IWB-3517.1 requires crack-like flaws to bem, wad to the flaw standards of IWB-3515. Because the VT-1 visual examination n.reptance ci ieria includes evaluation of crack-like indications and other relevant conditions ree Jiring corrective action (i.e., deformed or sheared threads, localized corrosion, def ,rmation of part, and other degradation mechanisms), it can be concluded that the VT. i visual examination provides a more comprehensive assessment of the condition of the closure head nut. As a result, the INEEL staff believes that VT-1 visual examinator'. provides an acceptable level of quality and safety.

Based on the comprehensive assessment that the VT-1 visual examination provides, and considering that the 1989 Addenda and later editions of the Code require only a VT-1 visual examination on reactor pressure vessel closure head nuts, it is concluded that an acceptable level of quality and safety will be provided by the proposed altamative. Therefore, it is recommended that the proposed attemative, VT-1 visual examination, be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

I. Reauest for Relief No. 98-001-PT. Use of Code Case N-498-1. Attemative Rules for 10-Year Svstem Hydrostate Testina tbr Class 1. 2. and 3 Svstems j Code Reauiremont Table IWB 2500-1, Examination Category B-P; Table IWC-2500-1, Examination Category C-H; and Table IWD 25001, Examination Categories D A, D-B, l and D-C require system hydrostatic testing of pressure-retaining components in accordance with IWA-5000 once each 10-year interval.

Licensee's Prooosed Attemative: Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee requested authorization to use Code Case N-498-1, Attemate Rules for 10-Year i

14 Hydmstatic Pressute Testing for Class 1, 2, and 3 Systems,Section XI, CWvision 1. The licensee stated:

"FPC will perform inservice leak tests or functional leak tests in accordance with the requirements of ASME Code Case N-498-1."

pcensee's Basis for Proposed Attemative (as stated):

"The system hydrostatic test required by ASME Section XI is not a test of structural integrity of the system but rather an enhanced leakage test. Piping and components are designed for a number of loadings that are postulated to occur under the various modes of plant operation. Hydrostatic testing subjects the piping and components to only a small increase in pressure over the design pressure and, therefore, does not present a significant challenge to pressure boundary integrity. Piping dead weight, thermal expansion, and seismic loads typically present a far greater challenge to the structural integrity of a system and are not part of the loading imposed during a hydrostatic test. Hydrostatic testing is primarily a means to enhance leak detection.

"In lieu of hydrostatic pressure testing at or near the end of the inspection interval, ASME Code Case N-498-1 requires a VT-2 visual examination at nominal operating pressure and temperature.

Evaluation The Code requires a system hydrostatic test once per interval in accordance with the requirements of IWA-5000 for Class 1,2, and 3 pressure-retaining systems. In lieu of the Code-required hydrostatic testing, the licensee has requested authorization to use Code Case N-498-1, Altemative Rules for 10-Year System Hydrostatic 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 structural integrity of the system but rather an enhanced leakage test.8 Hydrostatic 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 structuralintegrity of the 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 on Class 1 and 2 systems in Regulatory Guide 1.147, Rev. 9. For Cisss 3 systems, Revision N-498-1 specifies requirements identical to those for Class 2 components (for Class 1 and 2 systems, the

~

altamative 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 S. H. Bush and R. R. Maccery, ' Development ofin-Service Inspection Safety Philosophyfor U.S.A.

NuclearPowerPlants," ASME,1971.

L

N-4981 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. Common causes of failure are erosion-corrosion (EC),

microbiologically-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. .

Considering that Code Case N-498 was found to be an acceptable sitemative for Class 1 and 2 systems, and that Class 3 systems receive inspections commensurate with their function and expected failure mechanisms, the licensee's proposed altemative, to use Code Case N-498-1, should provide an accel: table level of quality and safety.

Therefore it is recommended that the licensee's proposed altamative be authorized pursuant to 10 CFR 50.55a(s)(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-498-1 with limitations issued in Regulatory Guide 1.147, if any.

J. Reauest for Relief No. 98-002-PT. Use of Code Case N-416-1. Altemate Testino for '

Class 1. Class 2. and Class 3 Welded Repaired / Replaced Components Code Reauirement Section XI, Table IWA-4700(a) requires that a system hydrostatic test in accordance with IWA 5000 after repairs by welding on the pressure-retaining boundary.

Licensee's Proposed Altemative: Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the altamative requirements contained in Code Case N-416-1 in lieu of the required Code examination. The licensee stated:

"The following altemative examination requirements will be implemented as defined by ASME Section XI Code Case N-416-1, 'Altemative Pressure Test Requirement for Welded Repairs or installation of Replacement items by Welding, Class 1,2, and 3 Section Xi'.

"A system leak test with associated VT-2 examinations will be conducted in lieu of performing the hydrostatic pressure test.

"NDE shall be performed in accordance with methods and acceptance criteria of the applicable sub-section of ASME B&PV Code, Section lit,1992 Edition.

" Prior to or immediately upon retum to service, a visual examination (VT-2) shall be performed in conjunction with a system leakage test, ASME B&PV Code,1992 Edition,Section XI, in accordance with paragraph IWA-5000, at nominal operating' pressure and temperature.

"Use of this Code Case shall be documented on an NIS-2 Form.'

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

"FPC is requesting relief from the above stated requirements based on ASME Code Case N-4161 which has been issued by the American Society of Mechanical Engineers.

" Piping components are designed for a number of loadings that would be postulated to occur under the various modes of plant operation. Hydrostatic testing only subjects the piping to a small increase in pressure over operating pressure and does not present a significant challenge to pressure boundary integrity. Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leakage detection during the examination of components under pressure, rather than solely as a measure to determine structural integrity of the component.

Evaluation: Section XI of the Code requires a system hydrostatic test in accordance with IWA-5000 after repairs by welding on the pressure-retaining boundary. The licensee proposes to implement the attemative to hydrostatic pressure tests contained in Code Case N-416-1 for Code Class 1,2, and 3 repairs / replacements.

Code Case N-416-1 requires a) NDE shall be performed in accordance with the methods and acceptance l criteria of the applicable Subsection of the 1992 Edition of Section Ill.

b) Prior to or immediately upon retum to service, a visual examination (VT-2) shall be performed in conjunction with a system leakage test, using the 1992 Edition of Section XI, in accordance with para. IWA-5000, at nominal operating pressure and temperature.

c) Use of this Code shall be documented on an NIS 2 Form.

The NRC has imposed an additional requirement for Class 3 butt and socket welded joints, that a surface examination on the root pass layer be performed when a surface examination on the completed weld is required to satisfy Section lil requirements.

The licensee has committed to comply with the requirements found in Code Case N-416-

1. However, the licensee has not committed to perform a surface examination on the root pass layer of Class 3 butt and socket welds when a surface examination on the completed weld is required in accordance with Section ill requirements. Therefore, it is recommended that the licensee's proposed attemative not be authorized.

K. Raouest for Relief No. 98-003-PT. Use of Code Case N-522. Pressure Testino of Containment Penetration PiDin0 Code Reouirement ASME Section XI, Table IWC-2500-1, Category C-H, requires pressure testing of all Class 2 pressure-retaining components in accordance with IWC-5221 and IWC-5222.

Licensee's Proposed Attemative: Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the attemative requirements contained in Code Case N 522, Pressure Testing of Containment Penetration Piping. The licensee stated:

"The following attemative examination requirement will be performed as defined in ASME Code Case N-522.

"FPC will comply with 10CFR50, Appendix J, Option B testing requirements."

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

" Florida Power CeiTee'. tion is requesting relief from the above stated requirements based on ASME Code Case N-522 which allows Appendix J testing of the subject penetrations as an altamative to the Code required Category C-H pressure tests.

" Appendix J leakage tests are performed at a test pressure of 54.2 psig, which'is I the peak calculated containment intomal pressure for the design basis loss of coolant accident. The containment design pressure is 55 psig. It is expected that leakage due to through wall flaws would be greater than the Appendix J leakage acceptance criteria. Thus, Appendix J testing should be acceptable for locating and detecting through wall flaws.

"The ASME Code Committee evaluated the proposed attemative testing requirements and determined them to be acceptable for ensuring the required integrity of the subject Class 2 penetrations. Pursuant to the requirements of 10CFR50.55a(s)(3)(i), the implementation of Appendix J requirements ensures an I acceptable level of quality and safety and does not decrease the margin of public health and safety."

Evaluation The Code requires a VT 2 visual examination during system pressure testing for all Class 2 pressure-retaining piping, including those segments that penetrate primary containment. As an altamative, 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 altamative to Section XI pressure tests for certain containment penetration piping.

Appendix J contains two options for examination requirements. 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 during shutdown i for the 10-year plant ISI. Option A also requires Type B and C tests during each j refueling outage, but in no case at intervals greater than 2 years. This is more frequent i 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. The staff's position, as stated in Regulatory Guide 1.163, Perfbrmance-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 (particular1y unquantified leakage rates for test failures, repetitive / common mode  !

failures, and aging effects) in historical Type C component performance data. While this l 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. j i

The licensee has provided no information concerning the Option B inspection frequency at Crystal River-3 for the subject examinations. Therefore, it is unclear whether the licensee's proposed test frequencies will exceed the 60-month interval for a Type C test as stated in Regulatory Guide 1.163. Therefore, it is recommended that the proposed altemative not be authorized.

L Reauest for Relief No.98-004 PT. Use of Code Case N-566. Correctim Action fbr Leakaae Identr6ed at Bolted Connectrons Code Reauirement Section XI, IWA-5250(a)(2) requires that if leakage occurs at a bolted connection in ASME Section XI components, the bolting shall be removed, VT-3 examined for corrosion, and evaluated in accordance with IWA-3100. -

Licensee's Proposed Attemative: Pursuant to 10 CFR 50.55a(a)(3)(i), the licensee proposed to use the attemative requirements contained in Code Case N-566, Correctim Action for Leakage IdentiRed at Bolted Connections. The licensee stated:

" Florida Power Corporation will comply with the requirements of ASME Code Case N-566."

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

" Florida Power Corporation is requesting relief from the above stated requirements based on ASME Code Case N-566. Code Case N-566 specifies attemative corrective action when leakage is detected at bolted connections during performance of ASME Section XI pressure tests.

"The requirements of IWA-5250(a)(2) could cause a plant transient. Some restrictions with IWA-5250(a)(2) are:

1. "The requirement does not allow engineering judgment of tightening undertorqued bolts.

2. "The requirement for bolt removal is without regard to amount of leakage.

3. The Code does not require the Owner to stop leakage.

4. VT-3 inspection of the botting is not necessarily going to stop the leak.

5. "lWA-3100 does not provide an acceptance standard for VT-3 bolt examinations. l

6. "The recdrement increases radiological dose to workers for leaks that l may not have any operational or structural impact.

7. "The leakage will likely increase, if system conditions allow, when removing one bolt at a time.

"The ASME Code Committee evaluated the proposed attematives contained in Code Case N-566 and determined that they are acceptable for corrective action for leakage at bolted connections. Code Case N 566 resolves the implementation problems associated with IWA-5250(a)(2) and allows two altematives as corrective action for leaks at bolted connections. The first option is to stop the leak, and review the botting and component material forjoint integrity. This review will ,

document inspechon results of the bolted connection for corrosion and other signs of degradation. The second option is, if the leakage can not be stopped, the joint

shall be evaluated in accordance with IWB-3142.4 forjoint integrity. This evaluation shall consider the number and condition of bolts, the leakage medium, the bolt and component material, the system function, and the monitoring of leakage.

"The implementation of Code Case N-566 pursuant to 10CFR50.55a(a)(3)(i) ensures an acceptable level of quality and safety and does not decrease the  !

margin of public health and safety. Implementation of the altamative corrective l measures will reduce costs, personnel radiation dose, and outage time." i Evaluation In accordance with IWA-5250(a)(2) of 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's proposed altemative is to implement the requirements of Code Case N 566, Conective Action forLeakage Identined at Bolted Connections.

Code Case N-566 requires that leakage be stopped, and the bolting and component material be reviewed forjoint integrity. However, the Code Casa does not define specific and detailed requirements to be included in the " review" to determine joint l integrity. '

Code Case N-566 also requires that, if the leakage is not stopped, the joint integrity be evaluated in accordance with IWB-3142.4, Acceptance by AnalyticalEvaluation. This  !

type of Code evaluation should include consideration of the number and condition of bolts, leaking medium, bolt and component material, system function, and long term leakage monitoring. However, an lWB-3142.4 evaluation presupposes that the bolting 1 has been examined, and all unacceptable relevant conditions accounted for in the l analytical evaluation. Since the botting is not being removed, the Code Casa does not l provide a comprehensive evaluation process for ensuring joint integrity. Consequently,  !

the INEEL staff does not believe that Code Case N-566, as written, provides an ,

acceptable level of quality and safety. Therefore, it is recommended that the licensee's '

proposed attemative not be authorized. j M. Reauest for Relief No. 98-005-PT. Use of Code Case N-533. Attemative Reauirements for VT-2 Visual Examination of Class 1 Insulated Pressure-Retainina Bolted Connections Code Reauirement IWA-5242(a) requires that insu'ation be removed from pressure-retaining bolted connections for VT-2 visual examination in systems borated for the purpose of controlling reactivity.

Licensee's Proposed Altemative: In accordance with 10 CFR 50.55a(s)(3)(i), the licensee proposed to use Code Case N-533, Allematin Requirements for VT-2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted Connections,Section XI, Dfvision 1. The licensee stated:

" Florida Power Corporation will comply with the requirements of ASME Section XI Code Case N-533. The following requirements will be met for systems borated for the purpose of controlling reactivity.

a) "A system pressure test and VT-2 visual examination will be performed each refueling outage without removal of insulation.

b) "Each refueling outage the insulation will be removed from the bolted connection, and a VT-2 visual examination will be performed. The connection is not required to be pressurized. Any evidence of leakage will be evaluated in accordance with IWA 5250."

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

" Florida Power Corporation is requesting relief based on ASME Section XI Code Case N-533 which specifies an attemative to the requirements of IWA-5242(a) which requires removal of insulation from Class 1 pressure retaining bolted connections to perform a VT-2 visual examination.

"The ASME Code Committee evaluated the proposed attemative contained in Code Case N-533 and determined that the altamative requirements for VT-2 visual examination of Class 1 insulated pressure retaining bolted connections was acceptable. Code Case N-533 resolves the implementation problem associated with IWA-5242(a) and allows altamative requirements, both of which will be met.'

The first requirement is to perform a system pressure test and VT-2 visual eixamination each refueling outage without removal of the insulation. The second requirement is to remove the insulation from each bolted connection and perform a VT-2 visual examination each refueling outage. The connection for this examination is not required to be pressurized, and any evidence of leakage is to be evaluated in accordance with IWA-5250.

"lWA-5242(a) requires that components be pressurized and insulation be removed (for systems that are borated for the purpose of controlling reactivity) during the visual examination VT-2. The insulation removal requirement causes ,

significant maintenance problems when the Class 1 system is at normal operating I pressure and temperature. Removal of insulation on hot components can also  ;

cause significant stress on the pressure boundary that may not have been '

analyzed. The additional effort requ: red to replace insulation on hot components will delay plant startup.

"The implementation of Code Case N-533 pursuant to 10CFR50.55a(s)(3)(i) -

ensures an acceptable level of quality and safety and does not decrease the margin of public health and safety. Implementation of the attemative requirements will reduce personnel radiation dose."

Evaluation: The Code requires the removal of allinsulation from pressure-retaining bolted connections in systems borated for the purpose of controlling reactivity when performing VT-2 visual examinations during system pressure tests. As an attemative, the licensee has proposed to use Code Case N-533, which includes a system pressure test and subsequent direct visual examination with the insulation removed. This direct visual examination is performed without pressurizing the system, and both of these examinations are required once each refueling outage.

I Code Case N 533 does not provide details of the examination parameters for the system pressure test or VT-2 visual examination with insulation in place. The INEEL staff believes that the system pressure test and corresponding Vi-2 visual examination with the insulation in place should be performed with a minimum 4-hour hold time after attaining a test pressure of not less than the nominal operating pressure associated with 100% rated reactor power. The 4-hour hold allows time for leakage to penetrate the insulation, providing a means of detecting any significant leakage with the insulation in place. The licensee has not provided specif'c details concoming the hold time associated with the proposed altamative. Consequently, the INEEL staff does not believe that the licensee's proposed use of Code Case N-533 provides an acceptable level of quality and safety. Therefore, it is recommended that the licensee's proposed altemative not be authorized. ~

3. CONCLUSION The INEEL staff evaluated the licensee's submittal and concludes that for Requests for Relief Nos. 98-001-11, 98-004-II, 98-005-II, 98-008-il, and 98-001 PT the licensee's proposed attematives will provide an acceptable level of quality and safety. Therefore, it is recommended that these proposed altamatives be authorized for the third interval pursuant to 10 CFR 50.55a(s)(3)(i). For Requests for Relief Nos. 98-002-11,98-003-11, 98-006-11, 98-007-II, 98-002-PT, 98-003-PT, 98-004-PT, and 98-005-PT, the licensee's altematives do not provide an acceptable level of quality and safety. Therefore, it is recommended that these altamatives not be authorized.

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