ML20205B509

SER Accepting Relief Requests 12R-24,Rev 0 & 12R-34,Rev 0, Related to Second 10-year Interval Inservice Insp for Byron Station,Units 1 & 2

ML20205B509
Person / Time
Site:	Byron
Issue date:	03/26/1999
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20205B496	List: ML20205B490 ML20205B509
References
NUDOCS 9903310273
Download: ML20205B509 (13)

Text

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

l' g

9 k UNITED STATES g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. *aana anny l

'+4 . . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULAILQN RELATED TO THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION I RELIEF REQUEST NOS.12R-24. REVISION O AND 12R-34. REVISION O COMMONWEALTH EDISON COMPANY BYRON STATION. UNITS 1 AND 2 DOCKET NOS. STN 50-454 AND STN 50-455

1.0 INTRODUCTION

Inservice inspection (ISI) of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in acordance with Section XI of the ASME Boller and Pressure Vessel Code (Code) and applicable addenda as required by 10 CFR 50.55a(g), except where specific written relief has been granted by ti.e Commission pursuant to 10 CFR 50.55a(6)(g)(i).10 CFR 50.55a(a)(3) states that attematives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (i) the proposed attematives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. Tiw Code of record for Byron Station, Units 1 and 2, second 10-year ISI interval, is the 1989 Edition of Section XI of the ASME Code.

2.0 EVALUATION By letter dated October 22,1998, Commonwealth Edison Company (Comed or the licensee) submitted its Second 10-Year Interval ISI Program Plan Requests for Relief Nos.12R-24,

- Revision 0 and 12R-34, Revision 0, for Byron Station, Units 1 and 2. The Idaho National l Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by

. the licensee in support of its Second 10-Year Interval ISI Program Requests for Relief Nos.

12R-24, Revision 0 and 12R-34, Revision 1, for Byron Station, Units 1 and 2. Based on the 9903310273 990326 PDR ADOCK 05000454 0 peg

I results of the review, the staff adopts the contractor's conclusions and recommendations i presented in the Technical Letter Report (TLR) attached.

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

Request for Relief No.12R-24, Revision 0: ASME Code,Section XI, Examination Category B-A, item B1.30 requires a 100 percent volumetric examination of the reactor vessel shell-to- l flange weld once each 10-year inspection interval. At least 50 percent of the shell-to-flange I weld shall be examined by the end of the first inspection period, and the remainder by the end  ;

of the third inspection period. However, if partial examinations are conducted from the flange l face, the remaining volumetric examinations required to be conducted from the vessel wall may be performed at or near the end of each inspection interval. I Paragraph IWB-2420(a) states that the sequence of component examinations established  ;

during the first inspection interval shall be repeated during each successive inspection interval, I to the extent prackal.

Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed to perform the required reactor pressure vessel (RPV) shell-to-flange weld examination during the third period of the second ,

interval. Relief is requested from repeating the sequence of examinations established in the I first ISI interval as required by IWB-2420(a). The licensee stated:

i

" Relief is requested from the requirements of IWB-2420(a) to repeat the sequence of examinations as established in the First inspection Interval and from Table IWB-2500-1, Examination Ca'.sgory B-A, Note (4) to require a volumetric examination from the flange face."

Paragraph IWB-2420(a) requires that the sequence of component examinations established during the first inspection interval be repeated during each successive inspection interval, to the extent practical. The licensee is requesting to defer the reactor vessel shell-to-flange weld examination to the end of the second interval to coincide with the examination of other RPV shell and nozzle-to-vessel welds.

During the first period of the first ISI interval, the RPV shell-to-flange weld received a partial examination from the flange face. This partial examination allowed the licensee to defer the remaining volumetric examination of the shell-to-flange weld until the third period. The licensee states that performing a partial examination of the subject weld from the flange face in the first period of the second interval creates personnel safety and radiation exposure concems.

Specifically, the manual scanning from the flange face requires personnel to position themselves under a suspended RPV head that is used to shield them from the significant radiation dose. The dose rates were measured to be 1.5 to 2.0 REM /hr at the center of the flange without using the RPV head for shielding and 0.5 to 1.0 REM /hr with the RPV head as shielding. Therefore, imposition of the Code requirements would result in a burden.

4  :

h Examination Category B-A, item B1.30 of the Code allows licensees to perform the RPV 4 shell-to-flange weld examination in conjunction with the nonle examinations of Examination 4 Category B-D. The licensee was authori::ed to use Code Case N 521 in an SE dated

January 13,1998; this Code Case allows the deferral of nonle-to-vessel welds, inside radius i sections, and noule-to-safe end welds to the end of the interval. Code Case N-521 stipulates i that (a) no inservice repairs or replacements by welding have aver been performed, (b) no j identified flaws or relevant conditions exist that require successive inspections in accordance l with IWB-2420(b), and (c) the unit is not in the first inspection interval. In addition, the NRC l staff has added a provision that the period of time between successive examinations does not i exceed the 10-vsar interval requirements. The licensee stated that these criteria are met for

.' both Byron Stat. ion RPVs.

! The licensee has volumetrically examined all of the subject shell-to-flange welds during the third period of the first ISI interval. Therefore, less than 10 years will have expired before the second l Interval examinations are performed (except as allowed by IWA-2430). The subject welds will i be examined along with the other RPV shell and nonie-to-vessel welds using the same automated equipment, personnel, and procedures. This proce<.,s improves the reliability and j reproducibility of the examinations and, therefore, provides reasonable assurance of structural

] integrity for the RPV shell-to-flange. Further, requiring the licensee to manually perform the

first period examinations on only the RPV shell to-flange weld would result in hardship or

j. unusual difficulty without a compensating increase in the level of quality and safety. Therefore, the licensee's proposed alternative contained in Request for Relief No.12R-24, Revision 0 is j authorized pursuant to 10 CFR 50.55a(s)(3)(li).

j Request for Relief No.12R-34: ASME Code,Section XI, IWA-5242(a) states o1at, for systems j borated for the purpose of controlling reactivity, insulation shall t>e removed from pretsure-j retaining bolted connections for a direct VT-2 visual examination.

Pursuant to 10 CFR 50.55a(a)(3)(l), the licensee proposed to perform VT-2 visual examinations j' with insulation in place for the following valved

$ 1(2)Sl8948A-D 1(2)RC8002A D l 1(2)Sl8949A-D 1(2)RC8003A-D

! 1(2)RC8001A-D 1(2)RH8701 A-B The Code requires that, for systems borated for the purpose of controlling reactivity, the insulation is to be removed from pressure-retaining bolted connections and, for Class i systems, the connection is to be VT-2 visually examined during each refueling outage. .The licensee is proposing to perform the Code-required inservice leakage test fer the subject Class 1 valves at the required frequancy (each refueling outage) without removing insulation from the bolted connection. A minimum 4-hour hold time at norrnal operating pressure will be maintained before the VT-2 visual examination. If evidence of leakage !s detected, the insulation will be removed and an evaluation performed.

The licensee maintains that alloy A-286 bolting material in the subject valves is essentially immune to wastage or erosion-corrosion problems associated with boric acid attack. In C - . _- . . .

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

addition, it has been determined that stress corrosion cracking is only a concem for A-286 materialif the botting materialis loaded to 100 kai or higher. For the subject Clacs 1 valves, the licensee states that none of the bolting is loaded to more than 65 ksi. In an NRC SE dated

. October 2,1998, the NRC reviewed and authc-ized this proposed attemative for Byron Station's sister plant, Braidwood Station.

During the staff's review of the relief, the concem was related to other susceptible carbon steel components (integral to the safe operation of the valves) that might be subjected to boric acid attack in the event of undetected slow leakage. These components could potentially be located where degradation would not be detectable while the valve is insulated. The licensee clarified that the subject Class 1 valves are not insulated past the valve body. Furthermore, the licensee clarified that the valve packing, yo,ke, and stem are all exposed and that any leakage from this area would be detected. Therefore, other carbon steel components are not exposed to undetected boric acid attack.

Based on the resistance of A286 bolting materials to boric acid attack, and the licensee's commitment to remove the insulation if any leakage is detected, the staff concludes that the licensee's proposed alternative provides an acceptable level of quality and safety. Therefore, the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

3.0 CONCLUSION

S The staff concludes that the Code requirements contained in Request for Relief 12R-24 would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The licensee's proposed alternat've provides reasonable assurance of inservice structural integrity and is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

For Request for Relief 12R-34, the staff concludes that the licensee's proposed attemative provides an acceptable level of quality and safety and is authorized pursuant to 10 CFR 50.55a(a)(3)(i).

Attachment:

Technical Letter Report .

PrincipalContributor: T.'McLellan Dated: March 26, 1999 l

l l

l

1 TECHNICAL LETTER REPORT 1

ON SECOND 10-YEAR INTERVAL INSERVICE INSPECTION REQUESTS FOR RELIEF 12R-24. REV. 0 AND 12R-34. REV. 0 EQB COMMONWEALTH EDISON COMPANY BYRON STATION. UNITS 1 AND 2 DOCKET NUMBERS: 60 454 and 50 455

~

1. INTRODUCTION By letter dated October 22,1998, the licensee, Commonwealth Edison Company, submitted Requests for Relief 12R-24, Rev. O, and 12R-34, Rev. O, seeking relief from the requirements of the ASME Code,Section XI, for the Byron Station, Units 1 and 2, second 10-year inservice inspection (ISI) Interval. The Idaho National Engineering and 4

Environmental Laboratory (INEEL) staffs evaluation of the subject requests for relief are in the following section.

2. EVALUATION The information provided by Commonwealth Edison Company 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 Byron Station, Units 1 and 2, second 10-year ISI interval, which began in September and August of 1996, respectively, is the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code.

A. Reauest for Relief No.12R-24. Rev. O. Examination Cateaorv B-A. Item B1.30. Deftmil of Volumetric Examination' on the RFV Shell-to-Flance Weld i

Code Reautrement: Examination Category B-A, Item B1.30 requires a 100% volumetric examination of the reactor vessel shell-to-flange weld once each ten-year inspection i

interval. At least 50% of the shell-to-flange weld shall be examined by the end of the first inspection period, and the remainder by the end of the third inspection period.

. However, if partial examinations are conducted from the fiange face, the remaining 9

volumetric examinations required to be conducted from the vessel wall may be performed at or near the end of each inspection interval.

Paragraph IWB-2420(a) states that the sequence of component examinations established during the first inspection interval shall be repeated du;ing each succe taive inspection interval, to the extent practical.

j ATTACHMENT 1

4

}

t

r ,

Licensee's Proposed Attematly.g: In accordance with 10 CFR 50.55a(a)(3)(ii), the licensee proposed to perform the required RPV shell-to-flange weld examination during the third period of the second interval. Relief is requested from repeating the sequence of examinations established in the first ISI interval as required by IWB-2420(a). The licensee stated:

• Relief is requested from the requirements of IWB-2420(a) to repeat the sequence of examinations as established in the First inspection interval and from Table IWB-2500-1, Examination Category B-A, Note (4) to require a volumetric examination from the flange face."

Licensee's Basis for Pr.qggssd Attemative (as stated):

" Relief is requested to defer the entire reactor vessel shell-to-flange vessel weld examination to the end of Byron Station's second ten-year inspection interval to coincide with the examinations of other reactor vessel shell and nozzle-to-vessel welds. The automated equipment necessary for the vessel shell and nozzle-to-vessel examinations can also be used to satisfy the examination of the shell-to-flange weld. Mobilizing this automated equipment tu perform a partial exam! nation in the first period would constitute an economic and schedule hardship. In the First inspection Interval, this weld was examined twice (in the First Period and again in the Third Period). The first time was a partial examination from the flange face performed manually. The second time was a complete examination from the vessel interior performed with automated equipment. This relief requests that the examination only be performed at the end of the interval.

" Byron Station baueves that performing a partial examination of the reactor vessel shell-to-flange weld in the first period of the second inspection interval would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety for the following reasons:

a) Personnel Safety: Volumetric examination of these welds from the flange would typically be performed manually, requiring inspection personnel to position themselves under a suspended reactor vessel head. The reactor head is used as shielding for dose reduction. This situation is a potential safety hazard, which can be avoided by deferring the examination of 100% of this weld to the end of the interval, b) Radiation Exposure: As mentioned above, this inspection is perfomied in a radiation area and significant shielding (the reactor head) is necessary for dose reduction. During refueling outage B2R07, dose rates at the center of the flange were measured at 1.5 to 2.0 REM per hour. Even with the reactor head as shletding, the dose rates ranged from approximately 0.5 to 1.0 REM per hour, i

c) The cost of scheduling automated equipment to perform an examination that would not meet the requirements of ASME Section XI would be an unnecessary burden on station resources.

"From an industry perspective, the deferral of Byron Station's shell-to-flange volumetric examinations to the end of the second inspection interval will not decrease the level of quality and safety because Pressurized Water Reactor (PWR) vessels similar to Byron Station's have been operating for over 20 years with no recorded inservice induced flaws or potential degradation mechanisms.

Since each PWR vesselin operation is representative of the operating conditions throughout industry, continued inspection of these vessels ensures that any potential degradation mechanism will be detected.

" Performing all the automated reactor vessel examinations during a single refueling outage improves consistency of the examinations by utilizing the same equipment, personnel, and procedures. Moreover, this improves the reliability and reproducibility of the examinations. Based on lack of any previous indications in the flange-to-shell weld, requiring the inspection of only the flange-to-shell weld during the first period would constitute a safety, exposure and schedule hardship without a compensating increase in quality or safety.

"In addition, Byron Station believes that deferral of the examination of the reactor vessel shell-to-flange weld to the end of the second inspection interval will

provide an acceptable level of safety and quality for the following reasons

a) The purpose of Table IWB-2500-1, Examination Category B-A, Note (4) is to permit the licensee to combine the examinations of the flange-tc-shell from the flange surface and the vessel-to-nozzle welds, since both examinations could use automated scanning equipment.

b) Both of Byron Station's reactor vessel shell-to-flange welds were 100%

volumetrically examined during the Third Period of the First inspection Interval. The 100% coverage obtained by this examination was independent of the First Period volumetric examination from the flange face. No indications or relevant conditions were discovered that required successive inspections in accordance with Paragraph IWB-2420(b). No inservice repairs or replacements by welding have ever been performed on any of the reactor vessel welds at Byron Station.

c) The performance of these examinations in the First inspection Interval is

. such that no more than ten years will transpire until the Second Inspection Interval examinations except as allowed by IWA-2430. This provides a reasonable assurance of operational readinoss."

Evaluation: Paragraph IWB-2420(a) requires that the sequence of component l

examinations established during the first inspection interval be repeated during each b

i

o .

)

i j

e successive inspection interval, to the extent practical. The licensee is requesting to i defer the reactor vessel shell-to-flange weld examination to the end of the second ,

3 interval to coincide with the examination of other RPV shell and nozzle-to-vessel welds. I

! During the first period of the first ISI interval, the RPV shell-to-flange weld received a partial examination from the flange face. This partial examination allowed the licensee to defer the remaining volumetric examination of the shell-to-flange weld until the third

) period. The licensee states that performing a partial examination of the subject weld 1

l from the flange face in the first period of the second interval creates personnel safety J l

and radiation exposure concems. Specifically, the manual scanning frein the flange face requires personnel to position themselves under a suspended RPV head that is used to shield them from the significant radiation dose. The dose rates were measured to be 1.5 to 2.0 REM /hr at the center of the flange without using the RPV head for shielding and 0.5 to 1.0 REM /hr with the RPV head as shielding. Therefore, imposition of the Code requirements would resuit in a burden.

Examination Category B-A, item B1.30 of the Code allows licensees to perform the RPV shell-to-flange weld examination in conjunction with the nozzle examinations of Examination Category B-D. The licensee was authorized to use Code Case N-521 in an SE dated January 13,10g8; this Code Case allows the deferral of nozzle-to-vessel welds, inside radius sections, and nozzle-to-safe end welds to the end of the interval.

Code Case N-521 stipulates that (a) no inservice repairs or replacements by welding have ever been performed, (b) no identified flaws or relevant conditions exist that require successive inspections in accordance with IWB-2420(b), and (c) the unit is not in the first inspection interval, in addition, the NRC staff has added a provision that the period of time between successive examinations does not exceed the 10 year interval requirements. The licensee stated that these criteria are met for both Byron Station reactor pressure vessels.

The licensee has volumetrically examined all of the subject shell-to-flange welds during the third period of the first ISI interval. Therefore, no more than 10 years will transpire i

. . . - . . . -..-.-.- - - - - .- - . - - - - .~ - - ..- - -

, n before second interval examinations are performed (except as allowed by IWA-2430).

The subject welds will be examined along with the other RPV shell and nozzle-to-vessel welds using the same automated equipment, personnel, and procedures. This process improves the reliability and reproducibility of the examinations, therefore providing 4

reasonable assurance that the structural integrity of the RPV shell-to-flange is being maintained. Further, requiring the licensee to manually perform the first penod examinations on only the RPV shell-to-flange weld would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. For these reasons, it is .scommended that the proposed relief be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

B. Reauest for Relief No.12R-34. lWA-5242(at System Pressure Tests for insulated Bolted Connections Code Reauirement: IWA-5242(a) states that for systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure-retalning bolted connections for a direct VT-2 v:sual examination.

Licensee's Proposed Altemative: In accordance with 10 CFR 50.55a(a)(3)(l), the licensee proposed to perform VT-2 visual examinations with insulation in place for the following valves:

1(2)Sl8948A-D 1(2)RC8002A-D 1(2)Sl8949A-D 1(2)RC8003A-D 1(2)RC8001A-D 1(2)RH8701A-B The licensee stated:

"For ASME Code Class 1 systems borated for the purpose of controlling reactivity, a system inservice leakage test shall be performed in accordance with the frequency required in table IWB-2500 (each refueling outage), without the removal of insulation from the bolted donnections. The requirements for inservice leak tests shall be augmented with a minimum 4-hour hold time at system normal operating pressure prior to the VT-2 visual examination to allow for leakage propagation from the insulation. If evidence of leakage is detected on the above listed valves, the insulation shall be removed and the evaluations for corrective measures performed in accordance with IWA-5250 (as medified for Byron by Relief Request 12R-12).

"For the valves listed above, the insulation shall be removed from the bolted

connections and a VT-2 examination shall be conducted, with the system

depressurized on a once per 10-year interval frequency. These examinations

)

i

. . ., , , , _m.. - . _ . .. .

,-v---

> ~

shall be distributed throughout the inspection interval. If evidence of leakage is detected evaluations for corrective measures performed will be performed in accordance with IWA-5250 (as modified for Byron by Relief Request 12R12). ,

These inspections shall be implemented through application of the Byron Station '

predefined surveillance program to assure they are performed within the  ;

prescribed time periods. I "The referenced, Class 1 valves, are included in the scope of Byron ASME Code j Relief Request 12R-11, Revision 2, currently in review with the NRC, which stipulates an inspection frequency of every refuel cycle with the system depressurized for Class 1 valves. The 10-year frequency of this relief request (12R-34) supercedes 12R-11 for these valves only.

"Regardless of whether a component is scheduled for examination or not, any I evidence of leakage will result in evaluations for corrective measures in accordance with IWA-5250 (as modified for Byron Station by Relief Request 12R-12)."

Licensee's Basis for Prooosed Alternative (as stated): l

• Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed altemative would provide an acceptable level of quality and safety.

Specifically, relief is requested from the requirement to remove insulation for the Class 1 components listed above for VT-2 on the frequency specified in Table IWB-2500, Category B-P, (each refueling outage). The following supports a reduced inspection frequency: '

"1. ASME Code Section XI paragraph IWA-5242(a) requires the removal of insulation from pressure retaining bolted connections for W-2 visual l examinations when the system is borated for the purpose of controlling )

reactivity. Paragraph IWA-5242(a) requires this for all bolting, regardless I of mCerials of construction, when the system is not borated for the purpose of controlling reactivity, insulation removalis not required by paragraph IWA-5242(a) for VT-2 visual examinations.

Comed believes that by the ASME Code only requiring insulation removal for borated systems, the intent of this requirement is to address early detection of boric acid wastage of the botting materials. In either borated or non-borated systems, if leakage results in wetting of the bolting material, the required VT-2 visual examinations would not be effective at detection of incipient stress corrosion cracking if it occurs; only the volumetric examinations of IWB-2500 Category B-G-1 would be effective.

For this reason, if the bolting material of construction is resistant to boric acid wastage, there is no reduction in margin of safety if the required VT-2 visual examinations are performed without insulation removal each refueling the same way as is performed for other Class 1 non-borated systems. The proposed attemative provision of inspecting these

I components on a once per 10-year laterval basis will provide sufficient 3 assurance that these highly corrosion resistant components have not degraded.

"2. For the valves listed above both the stud / bolt material and closure nut j material utilize SA-453 Grade 660 Class B. Also known as alloy A-286, the nominal composition of this ferrous alloy is 25Ni-15Cr-2Ti-AI.

l According to Reference 1', the 'high chromium content of alloy A-286 gives it similar resistance to general corrosion in boric acid as possessed

, by stainless steel, which is essentially immune to the wastage or erosion-j corrosion problems.' fieference 1 determines that for A-286 material, 4

stress corrosion vacking is only a concem if bolting material is loaded to

100 kai or higher. For the valves listed above, review of the installation i procedures shows that none of the bolting is loaded to more than 65 ksi.

l Therefore, stress corrosion cracking is not a concem. Also, Reference 1

states that a review of available NRC Public Documentation revealed no reports of failures of alloy A-286 used for extemal reactor vessel bolting 4

service in B&W units over a service period of greater then 20 years.

I j "3. "The valves listed above are among the highest radiation level

{ components in the Borated Bolting inspection Program. Insulation

! removal combined with scaffolding erection and inspection time are j

expected to contribute significantly (approximately 1.5 person-rem) to the overall dose received. As discussed above, there is no significant

increats in plant safety to be gained by performing VT-2 inspection of a these materials on an every refueling outage frequency, i

, "The following Byron Station bolting examination commitments and material l control programs in conjunction with the Proposed Altemative [ Provisions) 1 provide an acceptable level of safety and quality for bolted connections in  ;

systems borated for the purpose of controlling reactivity.

"In re?ponse to NRC Generic Letter 88-05, Byron Station has established a '

i program for Engineering to ! aspect all boric acid leaks discovered in the l containment building and to evaluate the impact of those leaks on carbon steel i

or low alloy steel components. Any evidence of leakage, including dry boric acid j crystals or residue, is 'nspected and evaluated regardless of whether the leak

{ was discovered at power or during an outage, lasues such as the following are j considered in the inspection and evaluation:

1) Evidence of corrosion or metal degradation

2) Effect the leak may have on the pressure bounda:y

? 1 Materials Handbook for Nuclear Plant Pressure Boundary Applications, EPRI TR 109668-SI, i WO4392-01, Final Report, Revision 0, December 1997 (EPRI Proprietary) f

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

3) Possibility of boric acid traveling along the inside of insulation on piping, and

4) Possibility of dripping or spraying on other components.

" Based on this evaluation, Byron Engineering initiates appropriate corrective actions to prevent reoccurrence of the leak and to repair, if necessary, any degraded materials or components.

"In summary, for systems borated for the purpose of controlling reactivity, when the bolting material is SA-453 Grade 660 and therefore immune to boric acid corrosion, Comed requests relief from the requirement of ASME Code Section XI paragraph IWA-5242(a) that insulation shall be removed from pressure retaining bolted connections for VT-2 visual examination. Volumetric examinations as applicable to IWB-2500 Categories B-G-1 will continue to be performed."

Evaluation: The Code requires that, for systems borated for the purpose of controlling reactivity, the insulation is to be removed from pressure-retaining bolted connections and, for Class 1 systems, the connection is to be VT-2 visually examined during each refueling outage. The licensee is proposing to perform the Code-required inservice leakage test for the subject Class 1 valves at the required frequency (each refueling outage) without removing insulation from the bolted connection. A minimum 4-hour hold time at normal operating pressure will be maintained before the VT-2 visual examination. If evidence of leakage is detectM , the insulation will be removed and an evaluation performed.

l The licensee maintains that alloy A-286 bolting material in the subject valves is essentially immune to wastage or erosion-corrosion problems associated with boric acid attack. In addition, it has been determined that stress corrosion cracking is only a concern for A-286 material, if the bolting material is loaded to 100 kai or higher. For the subject Class 1 valves, the licensee states that none of the bolting is loaded to more than 65 ksi. In an SE dated October 2,1998, the NRC reviewed and authorized this proposed attemative for Byron Station's sister plant, Braidwood Station. The INEEL staff concurs with this evaluation.

4 However, it is the INEEL's opinion that other susceptible carbon steel components I (integral to the safe operation of the valves) may be subjected to boric acid attack in the i

.r..

a o event of undetected slow leakage. These components could potentially be located where degradation would not be detectable while the valve is insulated. In a telephone conference call on March 18,1999, the licensee confirmed that the subject Class 1 valves are not insulated past the valve body. Furthermore, the licensee stated that the valve packing, yoke, and stem are all exposed and that any leakage from this area would be detected. Consequently,' other carbon steel components will not be exposed to undetected boric acid attack.

Based on the resistance of A286 bolting materials to boric acid attack, and the licensee's commitment to remove the insulation if any leakage is detected, it is concluded that an acceptable level of quality and safety is provided. Therefore, it is recommendad that the licensee's proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

3. CONCLUSION The INEEL staff evaluated the licensee's submittal and concluded that the circumstances stated in Request for Relief 12R-24 result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The proposed alternative will provide reasonable assurance of inservice structural integrity and should be authorized pursuant to 10 CFR 50.55a(a)(3)(ii). For Request for Relief 12R-34, the proposed attemative provides an acceptable level of quality and safety and should be authorized pursuant to 10 CFR 50.55a(a)(3)(i).

J I

b