Text

Notice of OMB Review of Info Collection & Solicitation of Public Comment
	ML20199E986
Person / Time
Issue date:	11/19/1997
From:	Shelton B; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
To:	;
References
	NUDOCS 9711240061
	Download: ML20199E986 (107)
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U. S. NUCLEAR REGULATORY COMMISSION Documents Containing Reporting or Recordkeeping Requirements: Office of Management and Budget (OMB) Review AGENCY: U. S. Nuclear Regulatory Commission (NRC)

AC TION: Notice of the OMB review of information collection and solicitation of public comment.

SUMMARY

The NRC has recently submitted to OMB for review the following proposal for the collection of information under the provisions of the Paperwork Reduction Act of 1995 (44 U.S.C. Chapter 35).

1. Type of submission, new,' revision, or extension: Revision

2. The title of the information coliectien: 10 CFR 50.55a," Codes and Standards; Amended Requirements"

. f 3.. The form number if applicable: Not applicable .L a

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-4. How often the collection is required: The ASME has set a frequency for h

conducting these activities with its attendant recordkeeping based on f, 9711240061 971119 l PDR ORG CUSOMB !)/y (1

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. operating history and the need for component functionality. The

< - frequency is dependent on the safety function of the compenent - The information is generally not submitted to the NRC, but is retained by the l'

licensees to be made available to the NRC in the event of an NRC audit. Reporting requiremer.ts consist of one-time relief requests or technical specification amendments.

, i S. Who will be required or asked to report: Nuclear power plant licensees.

6. An estimate of the number of responses: The requirements will apply to licensees and applicants for nuclear power plant licenses. Because no applicants for construction permits or operating licensees are expected, r

the reports will apply to the 109 nuclear power plants with operating licenses.

7. The estimated number of annual respondents: 109.

8. An estimate of the total number of hours needed annually to complete the requirement or request: implementation of later Code edition and addenda for ASME Boiler and Pressure Vessel Code (BPV Code)

Section XI and OM Code activities is estimated to result in a (1) one-

. time recordkeeping burden of 48,502 hours0.00581 days 0.139 hours 8.300265e-4 weeks 1.91011e-4 months (445 hours0.00515 days 0.124 hours 7.357804e-4 weeks 1.693225e-4 months /pl ant) and (2) one-time reporting requirements of 328 hours0.0038 days 0.0911 hours 5.42328e-4 weeks 1.24804e-4 months (3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months / plant) for a total L of 18,830 hours0.00961 days 0.231 hours 0.00137 weeks 3.15815e-4 months . The estimated total annual industry increase in l.

recordkeeping burden is 13,512 hours0.00593 days 0.142 hours 8.465608e-4 weeks 1.94816e-4 months annually (124 hours0.00144 days 0.0344 hours 2.050265e-4 weeks 4.7182e-5 months / plant). Due u

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_ to elimination of certain ASME OM Code reporting requirements, the estimated total industry annual reporting burden will decrease by 4,245 hours0.00284 days 0.0681 hours 4.050926e-4 weeks 9.32225e-5 months _ annually (39 hours4.513889e-4 days 0.0108 hours 6.448413e-5 weeks 1.48395e-5 months / plant).

9. An indication of whether Section_3507(d), Pub. L.104-13 applies:

Applicable

10. Abstract: The Nuclear Regulatory Commission (NRC) regulations-require that nuclear power plant owners (1) construct Class 1 Class 2, and Class 3 components in accordance with the rules provided in Section 111, Division 1 " Requirements for Construction of Nuclear Power Plant Components," of the A..n .ican Society of Mechanical Engineers (ASME) Boiler and Pressuru Vessel Code (BF'V Code), (2) inspect ;

4 Class 1, Class 2, Class 3, Class MC (metal containment) and Class CC j oncrete containment) components in accordance with the rules provided in Section XI, Division 1, " Requirements for Inservice inspection of Nuclear Power Plant Components," of the ASME BPV f Code, and (3) test u ss 1, Class 2, and Class 3 pumps and valves in accordance with the rules provided in Section XI, Division 1, of the ASME BPV Code. Every 120 months licensees are required to update their inservice inspection (ISI) and inservice testing (IST) programs to meet the version of Section XI incorporated by reference into the-regulations in effect 12 months prior to the start of a new 120-month interval.

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e 3 Ff Submiti by (insert date 30 cayt after publication in the Federal RegiateI), comments that

. address'the following questions:

1. Is the proposed collection of information necessary for the NRC to -

properly perform its functions? Does the information have practical utility? .

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2. Is the burden estimate accurate?-

3. Is there a way to enhance the quaPty, utility, and clarity of the information to be collected? ,

-4. .Ho'v can the burden of the information collection be minirtized,-

including the use of automated collection techn! ques or other forms o' information technology 7 A copy of the supporting statement may be viewed free of charge at the NRC Public Document Room, 2120 L Street, NW (lower level), Washington, DC, OMB clearance package's are available via the NRC's interactive rulemaking website through the NRC home

page (http://www.nrc. gov). This site provides the availability to upload comtnents as files (any format), if your web browser supports that function.' For information about the

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interactive rulemaking website, contact Carol Gallagher, (301) 415-5905: e-mail

- CAG@nrc. gov.

. Comments and questions should be directed to the OMB reviewer by (insert date 30 days L

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l after publication in the Federal Realt ig):

Norma Gonzales Office of information and Regulatory Affairs (3150 0011)

NEOB-10202 Office of Management and Budget Washington, DC 20503 -

Comments Comments can also be submitted by telephone at (202) 395-3084.

The NRC Clearance Officer is Brenda Jo. Shelton, (301) 415-7233.

Dated at Reckville,' Maryland, this / ^ day of t 1 /u 1997, For the Nuclear Regulatory Commission.

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[Brenda4Shelteni-NRC Clearance Officer Information and Records Management Brar SS Office of the Chief Information Officer 4

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PAPERWORK REDUCTION ACT SUZMISSION Please read the instructions before completng this form. Fir additional f:rms cf assistance in completing this form, contact your agenc/s Paperwork Clearance Officef. Send two copies cf this f;rm, th2 collection instrument t) be reviewed, th] Supporting Statement, and any additional documentation to: Omce of Information and Regulatory Affalta, Omce of li,anagement and Budget. Docket Library, Room 10102,72517th, Street NW, Washington, DC 20503. j

1. Agency / Subagency originating request 2. OMBcontrolnumber U.S. Nuclear Regulatory Commission V a. 3150- 0011 b.None 1 Type of information collection (check one) 1 Type of review requested (check one) _

a. New collection V a Regutar c. Delegated V b. Revtsion of a currenty approved collection b. Emergency. Approval requested by (data):

i

c. Extension of a currentty approved collection 5. Will this information collection t.t e a a.Yes

- significant economic impact on '

d. Retratatement, without change, of a previously approved conection for which approval has expired substantial number of small entities? V b,No

e. Reinstatement, with change, of a previousy approved cotiection for which approval has expired Requested a. Three years from approval date 0 expiration date

f. Existing collection in use without an OMB control number V b. Other (Specify): 9/30/2000

7. Title 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities

8. Agency form number (s) (Fsppecable)

Not applicable D. #(eywords I corporation by reference, Nuclear power plants and reactors, Reporting and recordkeeping requirements

10. Abstract The proposed amendment would incorporate by reference into NRC regulations recordkeeping and reporting requirements oflater ASME Code editions and addenda, impose the expedited implementation of Appendix VII UT performance demonstration and require certain modifications to Section XI of the ASME OM Code.

1 Affected public (u. a av wy we vges onws he wat we v) 12. Obugation to respond (uare pwnwy we v emt as omer: met appy we v)

a. IndMduate or households d. Farms _

a Voluntary 1 b. Business or other for-profit _

e. Federal Covemment b. Required to obtain or retain benefits

c. Nobfor.orofit institutions f. State. Local or Tnbal Govemment P c. Mandatory

13. Annual reporting and recordkeeping hour burden 14. Annual reporting and recordkeeping cost burden te mousamte oracaers)

a. Number of respondents 109 s. Total annualized capital /startup costs 0

b. Total an tual responses 7.948 b. Total annual costs (O&M) 0

1. e. Total annualized cost requested 0 nggf,these r sponses qg n 0

c. Total annual hours requested 5.573.397 me 0

d. Current OMB Inventory 5,573.397

f. Explanation of difference

e. Defferenca O

1. Program change

f. Explanahon of difference

1. Program change 2. Adjustment

2. Adjustmer

15. Pur' pose of information couection

jrequency of recordkeeping or reporti check a# mat appry)

(Mark Prtwy wm 7'*and a somers that apply we TJ J a.Recordkeeping b. Third-party disclosure

a. App! lenten for benefits

e. Program planning ce rnanagement 7 c. Reporting _ _

b. Program evaluation f. Research 1. On occasion 2. Weeidy 3. MontNy

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c. General pumose statistica

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4. Ouarterty F. Semi-annually 7 6. Annually d Audit 7. Biennially 7 8. Other (desertbe) [nc time

17. Statistical methods 18. Agency contact (person who can best answerquestfons regardng the content ortNs submrssion)

Does tNs information col:ection employ stat:stical methods?

ame: Wally Norris Yes Q No Phone: 301-415-6796 OMB 831 Tw *= w ,.s *8 *F= 10/95

19.C:rtific ti:n for P:perw:rk R:ducti:n Act Submitti:ns On behalf of this Federal agency,I certify that the collection of infunnation encompassed by this ret i t e.nplies with 5 CFR 1320.9.

NOTil: ' Die text of 5 LIR 1320.9, asti the related provisions of 5 CFR 1320.8 (b)(3), appear at the end of the instructions. The certijlcation is a be made uith reference to those regulatoryprovisions as stiforth in the instructions.

The following is a surnmary of the topics, regarding the proposed collection of information, that the certification covers:

(a) It is necessary for the proper performance of agency functions; (b) It avoids unnecessary duplication; (c) It reduces burden on small entities; (d) 11 uses plain, Olierent, and unambiguous terminology that is understandable to respondents; (c) Its implementation will be consistent and compatible with cunent reporting and recordkeeping practices; (f) It indicates the retention periods for recordLt ping requirements; (g) It informs respondents of the infonnation called for under 5 CFR 1320,8 (b)(3):

(i) Why the infonnation is being collected, (ii) Use ofinformation; (iii) Durden estimate; (iv) Nature of resporae (voluntary, requited for a benefit, or mandatory);

(v) Nature of extent of confidentiality, and (vi) Need to display currently valid OMB control number; (b) It was devek ped by an omce that hss planned and allocated resources for the emeient and cifective manage-ment and use of the infonnation to be ce!!ected (see note in item 19 of the instructions);

(i) It uses effective and emeient statistical survey m:thodology; and (j) It maken appropriate use ofinfonnation technology.

If you are nnab!c to certify compliance with any of these provisions, identify the item below and explain the reason in item 18 of the Supporting Statement.

Date Sgnature of Authortzed Agency OffcLal Date Sywhre o' Sensor Oficial or desgnes f ,,,

B N 4 't Nef Informa' ion OfN e' f 77

\ 10/95 OMB8H

SUPPORTING STATEMENT FOR PROPOSED RULE 10 CFR PART 50.55a, CODES AND STANDARDS (OMB Clearance No. 3150-0011)

DESCRIPTION Oc THE INFOR*MIJQN COLLECTION The proposed amendment to 10 CFR 60.55a would require licensees to implement the 1995 Edition with the 1996 Addenda for (1)Section XI, Division 1. Class 1, Class 2, Class 3, Class MC, and Class CC components; (2) the

Code for Operation and Maintenance of Nuclear Power Plants"(OM Code) Class 1 Class 2, and Class 3 pumps and valves; and (3) Appendix Vill,

Performance Demonstration for Ultrasonic Exainination Systerns," to Section XI, Division 1. In addition, the modification for containment iso!ation valve inservice testing has been deleted.

A. JUSTIFICAT!ON

1. Need for and Practical Util tv of the Collection of Informati20 NRC Regulations in 10 CFR $ 50.55a incorporate by reference Division 1 rules of Section ill, " Rules for Construction of Nuclear Power Plant Components," and Division 1 rules of Section XI. " Rules for Inservice inspection of Nuclear Power Plant Components," of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (BPV Code). These sections of the BPV Code set forth the requirements to which ituclear power plant compone"'s are designed, constructed, tested and inspected. This proposed amendment would also incorporate by reference into Section 50.55a the ASME Operation and Maintenance Code (OM Code). The OM Code sets forth inservice testing (IST) requirements for pumps and valves. Section Ill,Section XI, and the OM Code all contal.1 recordkeeping requirements. In general, Secilon 111 records are needed to provide documentation that construction procedures have been properly implemented, and Section XI records are needed to document the plans for and results of inservice inspecilon (ISn. The OM Code records are needed to document the plans for and results of IST. The records developed are generally not collected by the NRC, but are retained by the licensee to be made available to the NRC in the event of an NRC audit. A description of each Code requirement along with its burden is listed under " Estimate of Burden."

ASME Secuon Ill, Subsection NCA, NCA 3290, ' Owner's Responsibility for Records,*

gives the authority to the Owner for designating the construction records to be maintained. ASME Section XI, Subsection IWA, lWA-6310,

Maintenance of Records," 4 requires that each licensee maintain ISI records and reports for the service lifetime of the component or system. Finat ,/ the ASME OM Code, Subsection ISTA, ISTA 3.3.1,

" Maintenance of Records," toquires that each licensee maintain IST records and reports for the service lifetime of the co:nponent or system.

2. Aaenev Use oilDf0fma1120 The records and reports are generally historicat in nature and provide data on which future activities and actions can be based. The practical utility of the information collection for NRC is that appropriate records and reportn sre availab'e for auditing by NRC inspection personnel to determine whether (i) ASME Code provisions for IC A l

construction, ISI, and IST are being prc perly imp:emented in accordance with $ 50.55a of the NRC regulations, (ii) specific enforcement actions are necessary, and (iii) tu notify :

. other licensees of potential prcblems or take action on potential generic problems on a class of components.

3. Reduction of Burden Throuah information Technoloav The records document the various plant specific construction, ist, and IST programs.

The NRC has no objection to the use of new information technologies and encourages their use.

4. Effort to Identifv Duolication and to.Use Similar information The NRC references ASME national consensus standards as a general practice to avoid duplication of these requirements. Therefore, the amendment does not duplicate the information collection requirements contained in any industry codes or standards or generic NRC or other Federal agency regulatory requivments. The NRC uses the information collection requirements specified in the ASME Code in lieu of developing specific Information collection requirements.

5. Effort to Reduce SmalLBusiness Burdem The proposed amendment will have no impact on the paperwork burden of small companies. The amendment to Section 50.55a affects only the licensing and operation of nuclear power plants. The companies that own such plants do not fall within the scope of the definition of "small entities" set forth in the Regulatory Flexibility Act in the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121. Since these companles are dominant in their service areas, the proposed amendment does not fallin the province of this Act.

6. Consecuer ces To Federal Proaram Activities if the Collection is not Conducted or is Conducted Less Freauentiv The information is generally not collected, but is retained by the licensee to be, made available to the NRC in the event of an NRC audit. These records document licensee implementation of ASME Code provisions for construction, ISI, and IST activities.

Performance of these activities ensures that safety-related systems will continue to perform their intended functions. The ASME has set a frequency for conducting these activities with its attendant recordkeeping based on operating history and the need for component functionality, if the information collection was tot conducted or was conducted less frequently (i.e., the inspections were conducted less frequently), a safety-related component or system may not be able to perform its intended function which then

" ay have an impact on public health and safety,

7. Circumstances Which Justifv Variation from OMB Guidelines The record retention periods for the information required by the ASME Codes are generally based on the service lifetime of the applicable component or system. Such lifetime retention of records is necessary to document historicalinformation on the design, examination, and testing of components and systems for evaluating degradation IC - 2

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throughout their service lifetime.

8. f,onsultations Outside the NRC The NRC staff contacted personnel from Brookhaven National Laboratory, Upton, Long Island, New York, and Idaho National Engineering, Laboratory, Idaho Falls, Idaho to obtain their views on the collection of information. The rule w!Il be published in the Endtral Reaister for comment.

9. Confidentiality of Information Propriety or Confidentialinformation is protected in accordance with 10 CFR 2.790 of the NRC's regulations. However confidentialinformation is not anticipated.

10. Justification for Sensitive Questions This regulation does not request sensitive information.

11. Estimated Annualized Cost to the Federal Govemment NRC inspection personnel who periodically audit nuclear power plant quality assurance records, or ISI and IST programs, would include audits of the records for proper preparation and mrsintenance. The increase in inspection costs is estimated to be four hours per plant when the activity is performed as part of a normal quality assurance audit or ISl/IST program inspection. The cost of this inspection time is fully recovered from fees charged to NRC licensees pursuant to 10 CFR Parts 170 and 171.

Presently, licensees must request NRC staff approval to defer Section XI Code repair for Class 3 moderate energy piping systems. Approximmely a total of 1,100 staff-hours are required to review these requests for approval to defer repair. The cost of this review time is fully recovered from fees charged to NRC licensees pursuant to 10 CFR Parts 170 and 171. Voluntary implementation by licensees of Code Case N-513,

Evaluation '

Criteria for Temporary Acceptance of Flaws in Class 3 Piping," and Code Case N-5231,

Mechanical Clamping Devices for Class 2 and 3 Piping," will obviate the need for licensecs to request NRC staff approval (i.e., the staff would not be required to expend 1,100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to review submittals).

12. Estimate of Burden

a. Number and Type of Respondents The recordkeeping and reporting requirements of Section 50.55a, through incorporation by referance of the ASME Codes, will apply to licensees and applicants for nuclear power plant licenses. Because no applicants for construction permits or operating licensees are expected, the requirements will apply to the 109 nuclear power plants with operating licenses.

b. Ettima_ted Hourg Section 50.56a specifies that the Code edition and addenda to be applied to IC - 3

reactor coolant pressure boundary, and Quality Group B and Quality Group C components must be determined by the provisions of paragraph NCA-1140 of Subsection NCA of Section til of the ASME Code. NCA 1140 specifies that the owner (or designee) shall establish the ASME Code edition and addenda to be included in the Design Specifications, but that in no case shall the Code edition and addenda dates established in the Design Specifications be earlier than three years prior to the date that the nuclear power plant construction permit is docketed. NCA 1140 further states that later ASME Code editions and addenda may be used by mutual consent of the Owner (or designee) and Certificate Holder. The earliest Section lit addenda that is addressed in the proposed rule is the 1989 Addenda. Since the last plant was docketed in October 1974 (Palo Verde Plants 1,2,3), there is no plant under construction for which implementation of the Section lli addenda specified in the proposed rule will be a requirement. Individual plants may implement these improved rules on a voluntary basis, but unless they make that choice, no additional information collection burden is incurred.

Nuclear power plants are required to update their inservice inspection and inservice test programs by incorporating into successive 120 month inspection intervals requirements of the latest edition and addenda of Section XI that have been incorporated by reference as of 12 months prior to the start the next 120 month inspection interval. On this basis, many plants may at one time be required to implement the revisions contained in the Section XI, Division 1, addenda and edition specified in the final rule. The number of plants that could implement the specified addenda will grow gradually as each plant updates its inservice inspection program at the 10 year interval. Therefore, conservatively, the total number of plants that may ultimately be required to implemed. the specified edition and addenda is 109. The revisions in the Section XI edition and addenda affected by the final rule that significantly affect recordkeeping requirements are addressed below. Each of these changes are contained in the 1995 Edition with the 1996 Addenda (the edition or addenda which contained the Code change is given in parentheses).

Section XI Recordkeeoino Burden:

(i) Table IWA 1600-1 (1991 Addenda) references a revised ASME N626 specification which requires that Authorized Inspection Agencies be accredited by ASME. It is estimated that the records associated with this change will result in an average of 10 person hours (p brs) per plant per year The recordkeeping burden is estimated to be 1,090 p-hrs /yr (i.e.,

10 p-brs/ plant yr x 109 plants). This estimate is based on discussion with an authorized nuclear inspection (ANI) organization, but the impact has been assigned to the owners who ultimately pay for ANI services. (Ref. Table 1) ,

(11) IWA 2210 (1990 Addenda) improves visual examination requirements and requires calibration records for light meters and test charts. Based on discussion with licensee personnel, it is estimated that the records associated with this change will result in an averst;e of 1 p br per plant per year. The recordkeeping burden is estimated to be 109 p brs/yr (i.e.,1 p-br/ plant-yr x 109 plants). (Ref. Table 1)

IC - 4 l

- (iii) IWA 2322 (1991 Addenda) requires that before the near distance test chart is used for the first time an optical comparator or other suitable instrument be used to verify the height of a representative lower case character. It is estimated that the records associated with this change will result in an average of 2 p hrs at each plant. The recordkeeping burdsn is estimated to ,

be 218 p-hrs (i.e.,2 p-hrs / plant x 109 plants). (Ref. Table 3) ;

(iv) IWA-4130 (1989 Addenda) requires more detail to be documented in repair plans, it is estimated that the records associated with this change will result in an average of 1 p br for each repair operation. Based on discussions with ,

licensee personnel, an average of 100 repair plans per plant per year is assumed. Therefore, the recordkeeping burden is estimated to be 10,900 p brs/yr (i.e.,100 p-hrwplant yr x 109 plants). (Ref. Table 1)

(v) IWA 4340 (1991 Addenda) eliminates a surface examination for certain repair removal cavities. Records will decrease approximately 16 p-brs per plant per 10 yr ISI interval because of the elimination of a need to submit a-relief request. The decrease in recordkeeping burden is estimated to be 174 p-brs/yr (i.e.,16 p hrs x 109 plants /10 yr interval). (Ref. Table 1) ty. %hle IWB 25001 (1994 Addenda) requires an estimated 2 p brs for each gnt per 10-year ISI interval for records associated with additional pump and valve intemal surface visual examinations. The recordkeeping burden -

is esilmated to be 22 p-brs/yr (i.e.,2 p-hrs x 109 p!snts/10 yr interval). (Ref.

Table 1) ,

(vii) IWB-4300 (1989 Addenda) requires an estimated 4 p hrs for records for each pressurized water reactor (PWR) plant in conjunction with each series of steam generator sleeving operations during any refueling outage. The additional records include the Sleeving Procedure Specification, procedure qualification, performance qualification for personnel, location records, and examination records. lf sleeving operations are performed an average of three times each ten year interval for each PWR plant, the recordkeeping burden is estimated to be 86 p hrs /yr (i.e.,72 PWR plants /3 times in 10 years x 4 hrs each) (Ref. Table 1)

(vili) IWB-1220, lWC-1220, and IWD 1220 (1991 Addenda) each give an '

exemption for inaccessible integral attachments. Record expenses will be reduced about 16 p-hrs per plant per 10 year ISI interval since it will no

. longer be required to document these inaccessible integral attachments in requests for relief. The decrease in recordkeeping burden is estimated to be 174 p-hrs /yr (i.e.,16 p-bra x 109 plants /10 yr interval). (Ref. Table 1) ,

(lx) IWC 5222(e) (1991 Addenda) exempts open-ended lines from hydrostatic tests. Records will decrease about 16 p hrs per plant per 10-yr ISI interval because of the elimination of the need for a relief request. The decrease in

, recordkeeping burden is estimated to be 174 p-hrs /yr (i.e.,16 p brs x 109

- plants /10 yr interval). (Ref. Table 1)

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(x) IWD-2420 (1991 Addenda) adds successive examination requirements for Class 3 components. Records willincrease about 8 p brs per plant per year.

The recordkeeping burden is estimated M be 872 p brs/yr (i.e.,8 p brs/ plant.-

yr x 109 plants). (Ref. Table 1)

(xi) IWA 5221, Table IWB-25001, IWB-5200, Table IWC 2500-1, IWC 5200, and IWD-5240 (1993 Addenda) have all been revised to stipulate a system leakage test"in lieu of a system hydrostatic test during each 10 par ir.terval. Records will decrease about 16 person hours per boiling-water ;

reactor (BWR) plant per 10-year interval through the eHmination of the need ,

for a relief request (Note, the cost decrease applies only to BWR plants !

which encounter problems witn obtaining the Code-required pressure for hydrostatic testing of Class 2 portions of the main steam system). The decrease in recordkeeping burden is estimated to be 59 p-hrs /yr (i.e..

16 p-hrs /10 x 37 BWR plants). (Ref. Table 1)

(xil) IWF 1230 (1990 Addenda) exempts examination of inaccessible supports.

Eliminating the need for a relief request is estimated to save 16 person-hours per plant per 10-year interval. The decrease in recordkeeping burden is estimated to be 174 p hrs /yr(i.e.,16 p brs/10 x 109 plants). (Ref, Table 1)

(xiii) IWF 2430, IWF 2510, and Table IWF 25001 (1990 Addenda)- The exemption for supports of multiple components allowed under previous versions of IWF 2510(b) has been deleted. However, this change does not increase the number of supports required to be examined. In conjunction volth the deletion of the IWF-2510 exemption, Table IWF 25001 adopts for the first time representative sampling (i.e., grouping) which reduces the number of supports required to be examined by over 100. Even though the ,

adoption of representative sampling is considered an improvement over present procedures in that there is more assurance that defective supports will be detected, the ASME added the provisions of IWF 2430(c) and (d) which would require that if the examinations performed under IWF-2430(a) and (b) result in the detection of a large number of defective supports, additional examinations may be required. The reduction in the number of examinations attained through sampling is estimated to save 12 p-brs in recordkeeping per plant per year. Records associated with possible ,

additional examinations could add 8 p-hrs per plant per year which gives a net decrease of 4 p hrs in recordkeeping per plant per year. Thus, the recordkeeping burden is estimated to decrease by 436 p-brs/yr (i.e.,

4 p-hrs / plant yr x 109 plants). (Ref. Table 1)

In addition, the following recordkeeping requirements are incurred through the modifications to Section XI:

(i) 50.55a(b)(2)(xv) requires that licensees of all PWRs perform a vo!umetric examination of the Class 1 High Pressure Safety injection System within 6 months of the final rule, and per the standard ASME Section XI ISIinterval schedule thereafter. The one-time burden at 6 months is estimated to be 144 p hrs /yr (i.e.,2 p-brs/ plant x 1 examination 6 months after final rule 1C - 6 1

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published x 72 plants). (Ref. Table 3). The continuing burden associated with routine 120-month ISI interval examination is 43.2 p-hrs /yr (i.e.,2 p-brs/ plant x 3 examinations /10-year interval x 72 planto). (Ref. Table 1) t (ii) 50.55a(b)(2)(xvi) permits licensees to voluntarily adopt the provisions of Code Case N 513 for temporary acceptan, . of a flaw in certain Class 3 piping, item 2.0(d) of the Code Case requires a flaw evaluation to be performed. In addition, item 2.0(e) of the Code Case allows the licensee to perform a flaw growth analysis to establish the allowable time for temporary operation. Periodic examinations of no morq than 90-day intervals shall be conducted to verify the analysis, it is estimated that each licensee will apply the Code Case 20 times each year. The increase in burden is estimated to be 2180 p brs/yr (i.e.,20 occurrences x 1 p-br/ flaw evaluation flaw growth analysis x 109 plants). (Ref. Table 1)

(iii) 50.55a(b)(2)(xvi) also permits licensees to voluntarily adopt the provisions of Code Case N-5231 for temporary use of mechanical clamping devices for Class 2 and Class 3 piping. Section 9.0 of the Code Case requires the Owner to prepare a plan for monitoring defect growth, and perform periodic examinations of no more than 90-day intervals to verify the analysis, it is estimated that each licensee will apply the Code Case 20 times each year.

The increase in burden is estimated to be 2180 p-hrs /yr (i.e.,20 occurrences x 1 p-br/ flaw evaluation-flaw growth analysis x 109 plants). (Ref. Table 1)

Accendix Vill Recordkeeoino Burden:

(i) Appendix Vill, Article Vill 5000 (1996 Addenda) requires that qualification records be kept. The records will be generated when the qualification activities are performed. A conservative estimate is that ten percent of the total initial Appendix Vill qualification costs per plant will apply to records.

The costs are equivalent to an average per plant total of 260 person-hours (p-brs) for Appendix Vill records. The recordkeeping burden is estimated to be a one-time total of 28,340 p-hrs (i.e.,260 p-brs/ plant x 109 plants). (Ref.

Table 3)

Reoortino Reaulrements Asicciated with Imolementation of Later Editions and Mdenda of Section XI: ,

(i) 50.55a(b)(2)(xiv) requires that licensees define the Class 2 piping subject to volumetric and surface examination in the ASME Code required Preservice Inspection, and submit it for approval by the NRC prior to implementation.

The estimated burden to prepare the submittal for this one-time reporting burden is 218 p brs/yr (i.e.,2 p-brs/ plant x 109 plants). (Ref. Table 4)

(ii) Because licensees presently must request NRC approval to defer repair of moderate energy Class 3 piping, those licensees that voluntarily implement Code Case N-513 and Code Case N 523-1 would no longer be required to l request approval to defer repair. All 100 licensees are expected to implement the Code Cases approximately 20 times each year. The estimated decrease in burden which would result from licensees not having i

IC - 7 l

to prepare and submit each request is 4,300 p-hrs /yr (i.e.,2 p-brs/ request x 20 request / year x 109 plants). Because relief requests have traditonally been treated as exemptions, no burden for relief requests has been included in the Part 50 extension approval. Therefore, no burden reduction can be claimed. (Ref. Table 3)

QM. Code Recordkeeoina Burden:

(i) Table ISTB 4.7.1 1 (1994 Addenda) requires more accurate pressure instruments for the comprehensive and preservice pump tests. Additional records would be required for the procurement and periodic calibration of these instruments. The burden is estimated at one p-br per plant per instrument per year. Assuming three new instruments per plant, it is estimated that the increased burden would be 327 p-brs/yr (i.e.,3 instruments x 1 p-brs/yr x 109 plants). (Ref. Table 1)

(ii) ISTB 5.2.2(b) and Table ISTB 4.1-1 (1994 Addenda) have eliminated the requirement for quarterly measurement of vibration and either flowrate or pressure for standby pumps. This would result in fewer test records and a decrease in burden estimated at 2,180 p-brs/yr (i.e.,10 standby pumps x

% p br/ test x 4 tests /yr x 109 plants) (Ref. Table 1)

(iii) Appendix 1,1.3.7(a) (1994 Addenda) changes the test frequency for containment vacuum breakers from 6 months to 2 years or during a refueling outage, whichever is sooner. Assuming 2 vacuum breakers per PWR, the estimated reduction in recordkeeping requirements is 54 p-hrs /yr (i.e.,1.5 less tests /yr x % p br/ test x 72 PWR plants). (Ref. Table 1)

(iv) Appendix 1,4.1.2(a) and 8.1.2(a) (1994 Addenda) allow air or nitrogen to be substituted at the same temperature without the additional altemate test media requirements. This will result in fewer records. Assuming two correlation evaluations per plant, the estimated decrease in recordkeeping requirements is 872 p-hrs /yr (i.e.,2 X 4 p-brs/ evaluation X 109 plants). (Ref.

Table 1)

Egportina Reauirements Associated with Imolementation of Later Editions and Addenda of the OM Code:

OM Code:

(i) ISTA 3.2.1 (1990 Edition) does not include the existing Section XI requirement for preparing and submitting a summary report for Class 1 and Class 2 pump and valve tests to the NRC. The decrease in burden is estimated to be 4,360 p-brs/yr (i.e.,40 p-hrs / plant / year x 109 plants). (Ref.

Table 2)

(ii) ISTB 3.2 and 4.3 (1994 Addenda) require bypass / test loops to accommodate within 120% of design fl6w when used for the comprehensive or Group A tests. For the purpose of this analysis, it is assumed that all PWRs would have to modify the test loops in the containment spray system IC - 8

i or prepare and submit a relief request to the NRC for approval. The <

estimated burden to prepare a relief request is 18 p-hr per PWR per ten-year inspection interval. This gives an increased burden of 115 p hrs /yr (i.e.,16 p-brs/10 x 72 plants). (Ref. Table 1)

In addition, ine following recordkeeping requirements are incurred through the modifications to Code Case OMN 1 ano Appendix 11 of the OM Code:

l (i) 50.55a(b)(3)(iii)(A) requires that the adequacy of the initial test interval for certain electric operated valve assemblies be evaluated between 5 and 6 years after implementation of Code Case OMN 1. The Code Case is a voluntary alternative, and this would be a one time burden occurring 5 to 6 i years after the final rule is issued. Assuming that half of the plants choose to implement the Code Case, the estimated increase in recordkeeping burden is 5,500 p-brs/yr (i.e., i p-br/ evaluation x 100 motor-operated valves -

x 55 plants). (Ref. Table 3)

(ii) 50.55a(b)(3)(iv)(B) requires trending and evaluation of test data to support changes in the check valve test frequency. This one-time evaluation is to be performed at a maximum of 3 years afterimplementation of Appendix 11. On average, there are 260 safety related check valves per plant. The time required for trending and evaluation of test data is estimated at 1 p br/ valve.

Assuming that one-half of the plants implement the optional appendix, and assuming that all of the evaluations are performed in the same year, the burden is estimated to be 14,300 p-hrs /yr (260 check valves x 1 p br/ evaluation x 55 plants). (Ref. Table 3)

In addition, the following one-time reporting requirement is incurred through the modification to the OM Code:

(i) 50.55a(b)(3)(v) requires that a licensee voluntarily choosing to use Subsection ISTD for the examination of snubbers may do so after processing a one time plant technical specification change. It is estimated that one-half of the plants will choose to implement Subsection ISTD. The estimated burden to prepara a technical specification change is 110 p brs/yr (i.e.,2 p-brs/ plant x 55 plants). (Ref. Table 4) r L

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Table 1. AEME BPV Code Section XI and A&ME Oh! Code Annual Recordkeeping Burden Retention b necelan XI Number of Annual Total Annual Cuet or OM Plants A5ected llcrdkping Annual (8131/hr) Period * !

Code Reddan Annunty HrvPlant flours !

IWA 16031 Table 109. 10 1,090 142,790 Lifebme !

(1991 Addenda) 109 109 14,279 Lifetime IWA 2210(1990 Addenda) 1 10,900 1,427.900 Lifetime i IWA.4130 (1989 Addenda) 109 100 IWA 4340(1991 Addenda) .II 16 174 22.794 N'A TAllLE LWD 25001 11 2 22 2,882 Lifetime 22 4 86 11,266 Lifctime IWil 4300(1989 Addenda)

IWil-1220,IWC 1220, A 11. 16 174 22,794 N!A . .

IWik1220 *

(1991 Addenda)

!WA 5221. Table IWB-2500 37 .l.6 59 7,729 N'A 1,IWB 5200, Table IWC.

2500-1. IWC.5200, IWD.

55240(1993 Addenda)

IWC 5222(e)(1991 11 16 174 22,794 NA '

Addenda) 109 8 872 114,232 - Lifetime IWD-2420(1991 Addenda) 109 .l .6 174 22,794 N/A >

IWF.1230 (1990 Addenda)

IWF.2430, IWF.2510, & 109 4 436 57,116 Lifetime Table IWF.2500-1 (1990 Addenda)

. . -7 .. y gn ~ + ,

U.._t i SectionXISubtotalL A09 m/[.109 - 1,888 1.5$7.329f Table ISTB 4 7.1 1 109 3 327 42,837 Lifetime (1994 Addenda) 109 20. 2180 285,580 N/A ISTD S.2.2(b) A Table 4.11(1994 Addenda)

Appendix 1,1.3.7(a) 72 0.75 54 7,074 N'A (1994 Addenda)

. Appendix 1,4.1.2(a), 8.1.2(a) 109 8 872 114.232 N/A (1994 Addenda)

,. - = . m. ~ . n .. , u ,

je - . OM Code SuMenll.-- .109[ [ $25,5l. ,, L-2.779 E 364,049.W t ,

50.55a(bX2Xxv) 72 0.60 43 2 5659.2 Lifetime 50.554bX2Xxvi) 109 20 2,180 - 285,580 Lifetime 50.55a(bX2Xxsi) 109 20 2,180 285,580 fRetime g

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J. - 50.55a Subtotel' 1109 D - = ~ '40.6 ~ < 4,40) ' s $76,819 --

VW+ G D; a . caw +T,my:% W smyy5#+ns .5xm m En;.;;uc,p M yf:nM ;

iddelIXl 4LOMiUdissdfigMM005djgd6MMig4dr d13 $1GsdO, ,770072 dyngd$$@j ,

'lifethne means the hiettme of the component or systeen

+ A negathe number bulicates a reduction in recordlieeping burden IC - 10 B

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Tande 2, AS%lE BPV Code heetion XI and ASNIE 0%1 Code Reporting RequirenwntsSection XI Number of Annual Total A:uiual Cost Retention or 0%I Plants Affected Rc rdLping Annual (5131/ht) Period' Code Revliion Annustly firs /liant flours ISTA 3.2 I (iV90 Adderala) 109 -40 -4,360 571,160 Lifetime ISTil 3 2 anJ 4.3 (1994 72 1.6 115 15,091 Lifetime Addenda) p , m em.,c w u r m w::, m e.nav ,nyms-n w mn.muemunow n e -nvry,+m' en LQM Ukihta(La , ' [ hh;il09dba,h639hdjAhdiMMM)4 tit:p; fdidd

'I alAc 3. AS%IF flPV Code Section XI and AS%IE 0%1 Code One-Thne Recordkeeping Iturden Sutton X1 Numler of RcrdLping Total Annual Cost Hetention or 0%I 11 ants Affected lits /liant Annual ($lJi/hr) PerkNI' Code Reitsket Annually llours 109 2 218 28,$58 Lifetime IWA 2322(1991 Addeuda)

Arpendix Vi!!, Article Vill. 109 260 28,340 3,72 2,$40 Lifetime 5000 (1989 'htough 1996 Addenda)

Atio4 XI Subtotal . 109 262. ; 28,$$8 3,741,091

$0.55a(bX2 K n) 72 2 144 18.864 Lifetime 50 55a(bK3Xiii)( A) $5 100 $,500 720,500 Lif-time

$0 $5mbx3xidl3) $$ 260 14,300 1,873,300 Lifetime

~ $0,$$a bubtotal . 5$ - : 362 :19,944 , .2,612,664. ,

pg,m r w wve:~mw:r~qmy - y~~,.m;~rry m y muypm w #rg m; Llotalt ,L;.. M,iaddi109 slemn3432CLAB.502;a:16.3$3,762 L;.,a .m

lafetune means the hfettme of the etunpiment or system

. A negethe number hullesten a reducthm in recordkeeping twden l

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Table d. ASME BPV Code nectke XI and ASME OM Code One-Time Reporting Requirements j l

Sectkm XI Naumber of Reimrting Total Annual Coat Re' ision l or OM Planes Affected firs'I'lant Annual (5131/hr) Period * .

Code Rettekm Annually llours 50.55a(by2Xxit) - 109 2 218 28.558 lifetime 50.55albX3Xt) 55 2 110 14A10 1.ifetime go fN^mY p1nnnp,ntmmi.n.

,w w. .w. nig?,t;nnnin:.gm..npymmtyn,u.e,mygym>pernvrirewomw1m,pv,;,q;ne.

.ung -, u. .w, _q s _, .

gy g _. 2 u,.ty

%%N~M' &&L,d2t%idM1Macuh;'.A,%13sl)a;Ja2%gra.;D,z w a2,9110a;tdQJu!&

244;f;& 4 gig;jeg .

lJfethne neeans the lifethme of the cmapanent or syntese - >

. A nesetive nonber indicates a reducthm in recordlweping burden i

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From Table 1, the totalindustry increase in the recordkeeping burden is 13,512 p-brs per year, resulting in an average annual of approximately 124 p-brs per plant. The largest contribution to the increase is the additional recordkeeping required in IWA 4130 for repair activities.

The change in reporting requirements (Table 2) resulting from modifications to the OM Code would decrease the annual reporting requirements burden by 4,245 p-brs a year or 38.4 p brs/ plant per year. The totalindustry burden for recordkeeping and reporting requirements will increase by 7,251 p-brs a year or 67 p-brs/ plant per year [11,496 p brs (Table 1) 4,245 p brs (Table 3)).

The one time recordkeeping burden (Table 3) resulting from changes to the Section XI requirements would be 28,820 p-brs, or an average of 262 p-brs/ plant. The increase is almost entirely associated with the one-time mandatory cost for Appendix Vill implementation for the qualification and performance demonstration for ultrasonic testing, An additional 19,944 p brs would be added by the proposed NRC modifications to Section XI and OM Code requirements. The totalindustry burden for one time recordkeeping burden would be 48,502 p-brs, or an average of 624 p brs/ plant.

Proposed NRC modifications to Section XI and the OM Code requirements which would require licensees to submit to the NRC one-time requests for approval to use altematives would require 328 p-hrs or 4 p-brs/ plant for preparation of the requests (Table 4).

c. Estim.ated Costs of the Information Collution Re.quirements It is estimated that the annual costs to the industry resulting from the increase in annual information collection requirements required by the referenced Codes in the proposed amendment to Section 50.55a is a total of $1,214,003/ year [13,512 hours0.00593 days 0.142 hours 8.465608e-4 weeks 1.94816e-4 months x 5131/hr (Table

1) 4,245 hours0.00284 days 0.0681 hours 4.050926e-4 weeks 9.32225e-5 months x $131/hr (Table 2)). One-time costs are estimated at 56,396,730 [48,502 hours0.00581 days 0.139 hours 8.300265e-4 weeks 1.91011e-4 months x $131/hr (Table 3) + 328 hours0.0038 days 0.0911 hours 5.42328e-4 weeks 1.24804e-4 months x $131/hr (Table 4)).

d. Engrd Retention Penod Section XI, Division 1, IWA-6000, " Records and Reports," stipulates requirements for ISI records and reports and identifies those records that must be maintained for the service lifetime of the component or system as follows:

Index to record file

. Preservice and inservice inspection plans e Preservice and inservice inspection reports

. Repair records and reporis

Replacement records and reports

. Nondestructive examination procedures

Nondestructive examination records

Pump records and reports

. Valve records and reports

+ Pressure test pucedures

Pressure test records IC - 13

.-. _ v

P ASME OM Code requirements for inservice examination and testing records and reports '

are given in ISTA 3, " Records" (previously covered by IWA-6000). Records identified in ISTA 3 that mut,t be maintained for the service lifetime of the component are as follows: ,

Index to record file l

+ Preservice and inservice test plans

+ Pump records

. Valve records L

Lifetime retention of the ASME BPV Code Section XI and ASME OM Code records is necessary to ensure adeqvate historical information on the design, examination, and testing of components and systems to evaluate degradation of these components and systems throughout their service lifetime. The recordkeeping requirements in later Codes '

are essentially the same type of documents that are currently required, and the ISTA '

requirements reduced the number of records required by IWA of the 1989 Edition of Section XI, ;

13. OtherAdddional Costs .

i As discussed under 1. 'Need for and Practical Utility of the Collection of Information,"

ASME Sections lil and XI, and the ASME OM Code each contain requirements governing licensee maintenance of construction, ISI, and IST records and reports, respectively, for the service hfetime of the component or system. Licensees preserve the records in '

storage facilities that provide protection from hazards such as winds, floods, fires, and '

environmental conditions such as adverse humidity conditions. The costs associated with the records storage facilities is not known by th( NRC and would likely be incurred by licensees in the course of doing business.

14. Reasons for Chanae in Burden The change in burden results from a change in ASME BPV Code and ASME OM Code recordkeeping requirements effected by the addenda and editions that are being incorporated by reference through this proposed amendment into the NRC regulations, the imposition and expedited implementation of Appendix Vill UT performance demonstration, and the NRC modifications to requirements contained in Section XI and the OM Code.

15. Publication for Statistical Use This information will not be published for statistical use,

16. Payment or Gifts To Resoondents Not applicable.

' 17, Exceotion To The Certification Statement Not applicable, i

IC - 14 l !

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.- i l

18. Reasod Not nimalavina the Funiration Date :

1 Not applicable.

The requirement will be contained in a regulation. Amending the Code of Federal Regulations to display information that, in an annual publication, could become out of date would confuse the public. j i B. COLL FCTION OF INFORMATION EMPLOYING STATISTICAL METHODS Statistical methods are not used in the collection of the required information. ,

i I

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[7590-01 P) -l Nuclear Regulatory Commission l 10 CFR Part 50 !

J RIN 3150 AE26 l Industry Codes and Standards; Amended Requirements

' AGENCY: Nuclear Regulatory Commission. I

.. .i ACTON: Proposed rule. .

t

SUMMARY

The Nuclear Regulatory Commission (NRC) regulations require that nuclear power plant owners (1) construct Class 1, Class 2, and Class 3 components in accordance with the rules provided in Section Ill, Division 1, ' Requirements for Construction of Nuclear Power Plant Components," of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPV Code), (2) inspect Class 1, Class 2, Class 3, Class MC (metal containment) !

and Class CC (concrete containment) components in accordance with the rules provided in Section XI, Division 1,

Requirements for Inservice inspection of Nuclear Power Plant Components," of the ASME BPV Code, and (3) test Class 1, Class 2, and Class 3 pumps and valves in accordance with the rules provided in Section XI, Division 1, of the ASME BPV Code.

The NRC proposes to amend 10 CFR 50,55a to revise the requirements for construction, r inservice inspection (ISI), and inservice testing (IST) of nuclear power plant components. For construction, the proposed rule would permit the use of Section Ill, Division 1, of the ASME BPV Code,198g Addenda through the 1996 Addenda, for Class 1, Class 2, and Class 3 components J

with six proposed limitations and a modification.

i s i OGC 97- 004691

For ISI, the proposed amendment would require licensees to implement Section XI, Division 1, of the ASME BPV Code,1995 Edition with the 1996 Addenda, for Class 1, Class 2, and Class 3 components with five proposed limitations. Licensees would be permitted to implement: Code Case N-513 which addresses flaws in low and moderate energy Class 3 piping; Code Case N C 3 which addresses the temporary use of mechanical clamps in Class 2 and 3 piping; and Subsection IWE and Subsection IWL,1995 Edition with the 1996 Addt,nda.

The proposed rule would expedite implementation of Appendix Vi!I, " Performance Demonstration for U!trasonic Examination Systems," to Section XI, Division 1, with three proposed modifications. An expedited implementation schedule would also be required for a proposed modification to Section XI which addresses volumetric examination of the Class 1 high pressure safety injection (HPSI) system in pressurized water reactors (PWRs).

For IST, the proposed amendment would require licensees to implement the 1995 Edition with the 1996 Addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for Class 1, Class 2, and Class 3 pumps and valves with one limitation and one modification.10 CFR 50.55a has been clarified with respect to which pumps and valves are to be included in a licensee's IST program. Licensees would be permitted to implement: Code Case OMN-1 with one modification in lieu of stroke time testing; Appendix 11 (which is an attemative to the check valve condition monitoring program provishns contained in Subsection ISTC of the OM Code) with three proposed modifications; and Subsection ISTD for the IST of snubbers. Finally, based upon supporting information received since the last rulemaking, the modification presently in 6 50.55a for containment isolation valve inservice testing has been deleted.

2

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

The Statement of Considerations concludes by clarifying the NRC position regarding ASME Code Interpretations, and discussing NRC Direction Setting issue Number 13 (DSI 13) ,

with regard to NRC endorsement of industry codes and standards.

DATES: Submit comments by Untert date 90 devs aftttou_blication in the Federal Realster). !

Comments received after this date will be considered if it is practical to do so, but the l Commission is able to ensure consideration only for comments received on or before this date.

ADDRESSES: Comments may be sent to: Secretary, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001. A' 'N: Rulemaking and Adjudications Staff. Hand ciellver comments to 11545 Rockville Pike, RockvHle, Maryland,20552, between 7:30 am and 4:15 pm on Federal workdays.

You may also provide comments via the NRC's inteiactive rulemaking website through the NRC home page (http://www.nrc. gov). This site provides the availability to upload comments as files (any format), if your web browser supports that function. For information about the interactive website, contact Ms. Carol Gallagher, (301) 415 5905; e-mail CAG@nrc. gov.

Single copies of this proposed rulemaking may be obtained by written request or telefax to 301-415-2260 or from Frank C. Chemy, Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone: 301-415-6786, or Wallace E. Norris, Division of Engineering Technology, U.S.

Nuclear Regulatory Commission Washington, DC 20555-0001, Telephone: 301-415-6796.

Cerisin documents related to this rulemaking, including comments received, may be examined at the NRC Public Document Room,2120 L Street NW. (Lower Level), Washington, DC. These .

3 1

, , . _ . , . . . . , _ . . . - _ . . . . _ , , , . . _ . . . . _ . , . . . . . . , . , _ . _ . , , . . , , , , . . ~ . m, . . . . .. ,_, .

i same documents may also be viewed and downloaded via the interactive rulemaking website as established by NRC for this rulemaking.

FOR FURTHER INFORMATION CONTACT: Frank C. Chemy, Division of Engineering Technology, Office of Nuclear Rt gulatory Research, U.S. Nuclear Regulatory Commission, {

V!ashington, DC 20555-0001, Telephone: 301415-6786, or Wallace E. Norris, Division of Engineering Technology, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone: 301415-6796.

SUPPLEMENTARY INFORMATION:

i

1. Background

2. Summary of Proposed Revisions to $ 50.55a 2.1 List of Each Revision and Implementation Schedule 2.2 Discussion 2.3 120 Month Update 2.3.1 Section XI 2.3.1.1 Class 1,2, and 3 Components, including Supports 2.3.1.2 Limitations:

2.3.1.2.1 Engineering Judgement 2.3.1.2.2 Quality Assurance 2.3.1.2.3 Class 1 Piping ;

2.3.1.2.4 Class 2 Piping 2.3.1.2.5 Reconciliation of Quality Pequirements 2.3.2 OM Code 2.3.2.1 Class 1,2, and 3 Pumps and Valves 2.3.2.2 Background - OM Code 2.3.2.3 Clarification of Safety-Related Valves 2.3.2,4 Limitation:

2.3.2.4.1 Quality Assurance

2.3.2.5 Modification:

2.3.2.5.1 Stroke Time Testing 2.4 Expedited implementation !

2.4.1 Appendix Vill 2.4.1.1 Modifications:

2.4.1.1.1 Appendix Vill Personnel Qualification 2.4.1.1.2 Appendix Vill Specimen Set Cracks ,

2.4.1.1.3 Appendix Vill Specimen Set Microstructure 2.4.2 Generic Letter on Appendix Vill 4

, --- , . . , , , n ., ----s---,. --

re vr,-, -- - . . ~ , - - <-n. +

2.4.3 Class 1 Piping Volumetric Examination 2.5 Voluntary implementation 2.5.1 Section ill 2.5.1.1 Limitations:

2.5.1.1.1 Engineering Judgement i i

2.5.1.1.2 Section lli Materials 2.5.1.1.3 Weld Leg Dimensions 2.5.1.1.4 Seismic Design i 2.5.1.1.5 Quality Assurance )

2.5.1.1.6 Independence of Inspection ,

2.5.1.2 Modification:

2.5.1.2.1 Applicable Code Version for New Construction 2.5.2 Section XI 2.5.2.1 Subsection IWE and Subsection IWL 2.5.2.2 Flaws in Class 3 Piping; Mechelcal Clamping Devices 2.5.3 OM Code -

2.5.3.1 Code Case OMN-1 2,5.3.2 Appendix ll 2.0 3.3 Subsection ISTD 2.5.3.4 Containment Isolation Valves 2.6 ASME Code Interpretations 2.7 DSI-13 2.8 Steam Generators

3. Finding of No Significant EnvironmentalImpact 4, Paperwork Reduction Act Statement

5. Public Protection Notification

6. Regulatory Analysis-

7. Regulatory Flexibility Certification

8. Backfit Analysis

9. List of Subjects in 10 CFR Part 50

10. Part 50 Domestic Licensing of Production and Utilization Facilities

1. Background

The NRC is proposing to amend 10 CFR 50.55a, which defines the requirements for applying industry codes and standards to nuclear power plants. Section 50.55a presently requires that nuclear power plant owners (1) construct Class 1, Class 2, and Class 3 components in accordarce with the rules provided in the 1989 Edition of Section lit, Division 1,

Requirements for Construction of Nuclear Power Plant Components,* of the American Society of Mechanical Engint,ers (ASME) Boiler and Pressure Vessel Cade (BPV Code), (2) Inspect Cluss 1, Class 2, and Class 3 components in accordance with the rules provided in the 1989 5

Edition of Set, tion XI, Division 1, ' Requirements for Inservice Inspection of Nuclear Power Plant Components," of the ASME BPV Code with certain limitations and modifications, (3) Inspect Class MC (metal containment) and Class CC (concrete containment) components in accordance with the rules provided in the 1992 Edition with the 1992 Addenda of Section XI, Division 1 with certain modifications, and (4) test Class 1, Class 2, and Class 3 pumps and valves in accordance with the rules provided in the 1989 Edition of Section XI, Division 1, of the ASME BPV Code with certain limitations and modifications. Every 120 months licensees are required to update their ISI and IST programs to meet the version of Section XIincorporated by reference into S 50.55a and in effect 12 months prior to the start of a new 120-month interval.

The NRC proposes to amend 10 CFR 50.55a to revise the requirements for construction, ISI, and IST of nuclear power plant components. For construction, the proposed rule would permit the use of Section 111, Division 1, of the ASME BPV Code,1989 Addenda through the 1996 Addenda, for Class 1 Class 2, and Class 3 components. Six proposed limitations to the implementation of Section ll1 are included which address the issues of engineering judgement, Secticn 111 materials, weld leg dimensions, seismic design, quality assurance, and independence of inspection. A modification has been included addressing the applicab!e Code version for new construction.

For ISI, the proposed amendment would require licensees to implement Section XI, Division 1, of the ASME BPV Code,1995 Edition with the 1996 Addenda, for Class 1, Class 2, and Class 3. Five proposed limitations to the implementation of Sect:on XI are included which address the issues of engineering judgement, quality assurance, Cless 1 piping, Class 2 piping, and reconciliation of replacement items. Licensees wot'd be permitted to implement Code Case N-513 which addresses flaws in low and moderate energy Class pipi and Code Case N-523 6

1 which addresses the temporary use of mechanical clamps in Class 2 and 3 piping. Licensees would also be permitted to implement Subsection IWE and Subsection IWL,1995 Edition with the 1996 Addenda.

The proposed rule would expedite implementation of Appendix Vill,' Performance Demonstration for Ultrasonic Examination Systems," to Section XI, Division 1. Three proposed modifications to the implementation of Appendix Vill are included to address the issues of personnel qualification, specimen set cracks, and specimen set microstructure. An expedited implementation schedule would also be required for a proposed modification to Section XI which addresses volumetric examination of the Class 1 high pressure safety injection (HPSI) system in pressurized water reactors (PWRs),

For !ST, the proposed amendment would require licensees to implement the 1995 Edition with the 1996 Addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) for Class 1, Class 2, and Class 3 pumps and valves.10 CFR 50.55a has been clarified with respect to which pu 1ps and valves are to be included in a licensee's IST program. A proposed limitation is included which addresses the issue of quality assurance (QA).

A proposed modification to the implementation of the OM Code is included which addresses stroke time testing, Licensees would be permitted to implement Code Case OMN-1 with one modification in lieu of stroke time testing. In addition, Appendix 11 to the OM Code is an alternative to the check valve condition monitoring program provisions contained in Subsection ISTC of the OM Code. Three pioposed modifications to the implementation of Appendix 11 are included which supplement the appendix check valve condition monitoring program. Licensees would be permitted to use Subsection ISTD for the IST of snubbers. Finally, based upon 7

4

supporting information received since the last rulemaking, the modification presently in 9 50.55a for containment isolation valve inservice testing has been deleted.

The mechanism for endorsement of the ASME standards, which has been used since the first endorsement in 1971, has been to incorporate by reference the ASME BPV Code rules hto 6 50.55a. The regulation identifies which editions and addenda of the BPV Code have been approved for use by the NRC. On August 6,1992 (57 FR 34666), the NRC published a final rule in the Federal Register to amend 10 CFR Part 50, " Domestic Licensing of Production and Utilization Facilities." This final rule amended S 50.55a to incorporate by reference the 1986 Addenda,1987 Addenda,1988 Addenda, and 1989 Edition of Section Ill, Division 1, and the 1986 Addenda,1987 Addenda,1988 Addenda, and 1989 Edition of Section XI, Division 1, of the BPV Code, with specified modifications. The amendment imposed an augmented examination of reactor vessel shell welds. The amendment also separated the requirements for IST of pumps and valves from those for ISI of other components by placing the requirements for Inservice testing in a separate paragraph. For IST of pumps and valves, the regulation, throu9h its incorporation by reference of the 1989 Edition of Section XI, endorsed Part 1, " Requirements for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices," Part 6,

" Inservice Testing of Pumps in Light Water Reactor Power Plants," and Part 10, " Inservice Testing of Valves in Light Water Reactor Power Plants," of ASME/ ANSI OMa-1988 to ASME/ ANSI OM-1987.

On August 8,1996 (61 FR 41303), the NRC published a final rule in the Federal Register to amend 10 CFR 50.55a to incorporatu by reference for the first time ASME Section XI, Division 1, Subsection IWE," Requirements for Class MC and Metallic Liners of Class CC Components of Light Water Cooled Power Plants," and Subsection IWL, " Requirements for Class CC Concrete 8

r i

i Components of Light Water Cooled Power Plants." Subsection IWE provides criteria for visual inspection of the surface of metal containments, the steelliners of concrete containments, pressure-retaining bolts, and seats and gaskets. Subsection IWL provides criteria for visual i

inspection of concrete pressure refaining shells and shell components and for the examination of unbonded post-tensioning systems. l

2. Summary of Proposed Revisions to 9 50.55a !

The revisions to 9 50.55a which would result from adoption of the 1,989 Addenda through the 1936 Addenda have been divided into three groups based on the proposed implementation schedule (i.e.,120-month update, expedited, and voluntary). For each of these groups, it is i indicated in parentheses whether or not particular items are considered a backfit under 10 CFR ,

50.109 as discussed in Section 8. Backfit Analysis. This section provides a list of each revision and its implementation schtdule, followed by a discussion of the proposed revisions.

2.1 List of Each Revision and implementation Schedule 120-Month Update (in accordance with 9 50.55a(g)(4)(1) and 9 50.55a(f)(4)(1))

Section XI (Not A Backfit) l e Class 1. 2, and 3 Components, including Supports

= Limitations o Engineering Judgement o Quality Assurance o Class 1 Piping i o Class 2 Piping i o Reconciliation of Quality Requirements l

OM Code (Not A Backfit)

= Class 1,2, and 3 Pumps and Valves e Clarification of Safety-Related Valves

= Limitation o Quality Assurance 9

3 I

e Modification . ,i o Stroke Time Testing ;

Expedited implementation (after 6 months from the date of the final rule - Backfit) t Section XI ;

e Appendix Vill (including three modifications) ,

o Personnel Qualification l o Specimen Set Cracks :

L

o. Specimen Set Microstructure '

e Class 1 Piping Volumetric Examination Voluntary implementation [may be used when final rule published) ;

Section lli(Not A Backfit) e - Class 1,2, and 3 Components

. Limitations - '

o Engineering Judgement o Section ill Materials o Weld Leg Dimensions .

o Seismic Design o Quality Assurance o Independence of Inspectior:

. Modification o Applicable Code Version for New Construction Section XI (Not A Backfit) - u

e Subsections IWE and IWL,1995 Edition with the 1996 Addenda e Flaws in Class 3 Piping; Mechanical Clamping Devices o Limitation m Scope OM Code (Not A Backfit) o Code Case OMN 1 i l

o Limitstion on Length of Test Interval e Appendix ll (including three modifications)

!- o Valve Opening and Closing Functions o 1. imitation of Length of Initial Test interval o Condition Monitoring Program

= Subsection ISTD e Containment isoletion Valves l

2.2 Discussion 2.3 -120 Month Update 1

2.3.1 Section XI ;

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. . _ _u.-_m,._. _ . . . _ . _ . _ _ _ _ _ . . ~ _ _ . . _ - . _ . . _ _ _ _ _ . _ - . _ . . . , . - , . . _ _

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i 2.3.1.1 Class 1,2, and 3 Components, including Supports Section 50.55a(b)(2) together with $ 60.55a(g)(4) of the proposed rule would require that licensees implement the 1995 Edition with the 1996 Addenda of Section XI, Division 1, for Class 1 Class 2, and Class 3 components and their supports. Five prop. sed limitations would be included to address NRC positions on the use of Section XI.

2.3.1.2 Limitations 2.3.1.2.1 Engineering Judgement The first proposed limitation to the implementation of Section XI would address an NRC position with regard to the Foreword in the 1992 Addenda through the 1996 Addenda of the BPV Code. That Foreword addresses the use of " engineering judgement" for ISI activities not specifically considered by the Code. Proposed paragraph 50.55a(b)(2)(xi) would require that i

when a licensee reibs en engineering judgement for activities or evaluations of components or l

systems within the scope of 9 50.55a that are not directly addressed by the BPV Code, the licensee must receive NRC approval for those activities or evaluations pursuant to 10 CFR 50.55a(a)(3).

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2.3.1.2.2 Quality Assurance The second proposed limitation to the implementation of Section XI pertains to the use of i

NQA 1 v.ith Section XI.Section XI references the use of either NOA-1 or the Owner's Appendix I B Quality Assurance Program (10 CFR Part 50, Appendix B,

Quality Assurance Criteria for Nuclear Power Plants and Fuel Processing Plants") as part of its individual requirements for a 11 w

QA program At present,6 50.55a endorses the 1989 Edition of the ASME Code which ,

i i

references NOA 1 i979 for Section XI. The 1996 Addonda of the ASME Code references NQA 1 1992 for Section XI.

The NRC has reviewed the requirements of NCA 1,1986 Addenda through the 1992 Addenda, that are part of the incorporation by reference of 6ection XI, and has determined that by itself, NQA 1 would not adequately describe how to satisfy the, requirements of 10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," sinco there are various aspects of operational phase QA and administrative controls which are not addressed by NQA 1.

10 CFR 50.34(b)(6)(ll) requires that "The information on the controls to be used for a nuclear power plant or a fuel reprocessing plant shall include a discussion of how the applicable requirements of Appendix B will satisfied." This information is required to be submitted to the NRC as part of the Final Safety Analysis Report (FSAR). Standard Review Plan (SRP) 17.2,

" Quality Assurance During the Operations Phase,' states that "The QA program descripticn presented in the FSAR must discuss how each criterion of Appendix B will be met." Furthct, the SRP states *The acceptance criteria include a commitment to comply with the regulatory ,

positions presented in the appropriate issue of the Regulatory Guides including the requirements of ANSI Standard N45.2.12 and tne Branch Technical Position listed in subsection V of SRP

- Section 17.1. Thus, the commitment constitutes an integral part of the QA program description and requirements." *.iie NRC has determined that the provisions of NQA 1,1986 Addenda through the 1992 Aridenda, would not sat'sfy the criteria specified in SRP 17.2 for describing how the requirements of Appendix B will be satisfied for operational activities. There are numerous areas where American National Standards Institute (ANSI) standards or NRC regulatory 12 4

o positions, which have been long standing comerstones of an Owner's QA Program, are either nonmandatory or missing altogether from the NOA 1 provisions. However, the Owner'sSection XI QA Program, which has been approved by the NRC, is adequate. Thus, the Commission has determined that the requiroments of NQA 1,1986 Addenda through the 1992 Addenda, are acceptable for use in the context of Section XI, as permitted by IWA-1400, provided the licensee utilizes its 10 CFR Part 50, Appendix B, CA program in conjunction with Section XI. Changes to a licansee's Q/. program shall be made in accordance with 10 CFR 50.54(a). Further, where NQA 1 and Section XI co not address the commitments contained in the licensee's Appendix B QA program,descriptiota, such comm!tments shall be applied to Section XI activities. Proposed S 50.55a(b)(2)(xii) contairs the requirement addressing licensee's commitments related to Section XI.

2.3.1.2.3 Class 1 Pipir g t

The third propoa.ed li,,1itation to the implementation or Section XI would require licensees to hse the rules for Section XI IWB-1220, " Components Exempt from Examination," that are cont ined in the 1989 E dition in lieu of the rules in the 1989 Addenda through the 1996 Addenda.

These aler Code adden da contain provisions of Code Cases N-198-1," Exemption from Examination for ASME Class 1 and Class 2 Piping Located at Containment Penetrations;" N-322, "P.xamination Requirements for Integrally Welded or Forged Attachments to Class 1 Piping at Containment Penetratiom;" and N-324, " Examination Requirements for Integrally Welded or Forged Attachments to Class 2 Piping at Containment Penetrations;" which were fo ;nd to be unacceptable. Because the NRC had previously determined the Code cases to be unacceptable, they were not endorsed in any revision of Regulatory Guide 1.147, " Inservice Inspection Code Case Acceptability- ASME Section XI, Division 1." The provisions of Code Case N-198-1 were 13

determined by the NRC to be unacceptable because industry experience has shown that welds in service-sensitive BWR stainless steel piping, many of which are located in Containment Penetrations, are subjected to an aggressive environment (BWR water at reactor operating temperatures) and w'll experience intergranular Stress Corrosion Cracking. Exempting these welds from examination could result in conditions which reduce the required margins to failure to unacceptable levels. The provisions of Code Cases N *M ,1d N-324 were determined to be unacceptable because some important piping was exemp;ed kom inspection. Access difficulties was the basis in the Code cases for exempting these areas from examination, but the NRC

' developed the break exclusion zone design and examination criteria utilized for most containment penetration piping expecting not only that Section XI inspections would be performed but that augmented inspections would be performed. These design and examination criteria are contained in Branch Technical Position MEB 3-1, an attae.hment of NRC Standard Review Plan 3.6.2," Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping " Thus, proposed S 50.55a(b)(2)(xlii) would require licensees to use the rules Mr IWB-1220 that are contained in the 1989 Edition in lieu of the rules in the 1989 Addenda through the 1996 Addenda.

2.3.1.2.4 Class 2 Piping The fourth proposed limitation to the implementation of Section XI, contained in S 50.55a(b)(2)(xiv), would confine implementation of Section XI IWC-1220, " Components Exempt from Examination," IWC-1221, " Components Within RHR (Residual Heat Removal), ECC (Emergency Cool Cooling), and CHR (Containment Heat Removal) Systems or Portions of Systems," and IWC-1222, " Components Within Systems or Portions of Systems Other Than RHR, ECC, and CHR Systems," 1989 Addenda through the 1996 Addenda. T% provisions of l

14 l

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Code Case N-408-3, "Altemative Rules for Examination of Class 2 Piping," were incorporated into Subsection IWC in the 1989 Addenda. These provisions contain rules for determining which Class 2 components are subject to volumetric and surface examination. The NRC had previously determined that the provisions of the Code Case were acceptable if the licensee defined the Class 2 piping subject to volumetric and surface examination and received approval prior to !mplementation. Approval was required to ensure that safety significant components in the Residual Heat Removal, Emergency Core Cooling, and Containment Heat Removal systems are not exempted from appropriate examination requirements. Thus, the requirements contained in IWC-1220, IWC-1221, and lWC-1222,1989 Addenda through the 1996 Addenda, for determining the enmponents subject to examination and establishing examination requirements for Class 2 piping may be used if the licensee defines the Class 2 piping subject to volumetric and surface examination, and submits this information to the NRC for approval pursuant to S 50.55a(a)(3).

2.3.1.2.5 Reconciliation of Quality Requirements The fifth proposed limitation to the implementation of Section XI addresses reconciliation of replacement items [9 50.55a(b)(2)(xx)(A)) and the definition of Construction Code

[S 50.55a(b)(2)(xx)(B)). Changes to IWA-4222, Reconciliation of Owner's Requirements,"in the 1995 Addenda would oermit a replacement item produced at a facility not having a 10 CFR Part 50, Appendix B qualified program to be used in safety-related epplications. With regard to the definition of Construction Code, a new definition of Construction Code appeared in IWA-9000,

" Glossary,"in the 1993 Addenda. Due to the changes made in IWA-4200 in the 1995 Addenda, the change in definition could result in standards being utilized which do not contain any QA requirements, or contain QA requirements that do not fully comply with Appendix 8. Thus, when 15

implementing the 1995 Addenda through the 1996 Addenda,6 50.55a(b)(2)(xx)(A) would require reconciliation of replacement items to the original QA requirements. Section 50.55a(b)(2)(xx)(B) would require a licensee to reconcile replacement items to the Construction Code and to the QA requirernents as described in the Owner's QA program.

Section XI Article IWA-4000 provides rules and requirements for the repair and replacement of pressure retaining components and their supports. Versions of IWA-4000 previous to the 1995 Addenda permitted a licenses to purchase a replacement item to the standards of the original Construction Code or a later version, provided that the technical requirements of an item such as design and fabrication, as well as the nontechnical requirements (idantified as administrative requirements in IWA-4222) such as QA and Authorized Inspection of the later version were reconciled with those of the original Construction Code and Owner's Requirements, Reconcillation ensures that the replacement item meets ce,tain standards of quality so that it is satisfactory for the specified design and operating conditions. In the 1995 Addenda, the provisions of Code Case N-554," Alternative Requirements for Reconcdiation of Replacement items," were incorporated into an extensive rewrite of IWA-4200. As a result of these changes to IWA-4200, specifically IWA-4222(a)(2), the nontechnical requirements for Cla:s 1,2, and 3 safety-related replacement items would no longer need to be reconciled which may result in noncompliance with 10 CFR Part 50, Appendix B. NRC regulations require that any item which performs a safety-related function must meet Appendix B. Appendix B invokes, among other things, contmls on suppliers of safety-related items. By not requiring reconciliation of the administrative requirements, the provisions in IWA-4222(a)(2) of the 1995 Addenda through the 1996 Addenda, would allow vendors having a OA program which does not meet Appendix B to be utilized, and may result in noncompliance with Appendix B. These deficiencies could be resolved if the Code provided for commercial grade item dedication in accordance with

'6

10 CFR Part 21," Reporting of Defects and Noncompliance." However, IWA-4222 does not address commercial grade dedication. In addition, it should be pointed out that a separate Code Case which provides an attemative for a specific provision in IWA 4200, Code Case N-567, "Altemative Requirements for Class 1,2, and 3 Replacement Components," was modified to require the reconciliation of nontechnical requirements before the Code Ccse was approved.

Therefore, .an inconsistency exists between the Code and a Code Case. Thus, when implementing the 1995 Addenda through the 1996 Addenda, S 50.55a(b)(2)(xx)(A) would require reconciliation of replacement items to the original CA requirements.

The provisions of the Code in IWA-4222(a)(2) discussed above address newly manufactured replacement parts. A further limitation on the use of Article IWA-4200 in the 1995 Addenda through the 1996 Addenda is contained in 6 50.55a(b)(2)(xx)(B), IWA-4222(b)

. addresses the use of items from a facility which was shutdown or for which construction was halted. IWA-4222(b) permits the use of either the administrative requirements of the Construction Code of the item being replaced or the administrative requirements of the Construction Code of the item being used for replacement. However, the definition of

" Construction Code" was changed in the 1993 Addenda. In versions of Section XI previous to the 1993 Addenda, Construction Code was defined in IWA-9000, "Giossary," as "the body of technical requirements that govemed the construction of the item." included in the body of technical requirements that governed the construction of the item was a requirement to reconcile the Owner's specificatica requirements, which :ncluded NRC regulatory requirements and apolicable Owner design and procurement specifications that invoke technical and nontechnical requirements (e.g.,10 CFR Part 50, Appendix B), in the 1993 Addenda, the definition became nationally recognized Codes such as ASME, Specifications such as the American Society of Testing and Materials (ASTM)land des!gnated Code Cases. Either definition of Construction 4

17 i

- . - - _. _______-___a

t Code would include the original Construction CodeHor the design and construction of piping, such as B31.1," Power _ Piping," and B31.7, " Nuclear Piping," and those for the design and construction of storage tanks, such as the American Petroleum Institute (API) 620, ? Design and_

Construction of Large, Welded, Low-Pressure Storage Tanks," and API 650, " Welded Steel ;

Tanks for Oil Storage." However, many of these standards utilized for construction do not contain any QA requirements, or they contain QA requirements that do not fully comp;y with

- Appendix B. Therefore, in order to satisfy Appendix B, QA requirements similar to or meeting Appendix B were invoked in thu Owner's original procurement documents. Thus, when

. implementing lWA-4200 (including subparagraphs IWA-4221, IWA-4222, IWA-4223, IWA-4224, and IWA-5224), $ 50.55a(b)(2)(xx)(B) would require a licensee to reconcile replacement items to the Construction Code and to the QA requirements as described in the Owner's QA program.

2.3.2 OM Code (120-Month Update) 2.3.2.1 Class 1,2, and 3 Pumps and Valves l

The proposed amendment to 9 50.55a(f)(4) would require that IST of pumps and valves 4 be performed in accordance with the ASME

Code for Operation and Maintenance of Nuclear j Power Plants"(OM Code). A proposed new section, S 50.55a(b)(3), would specify the editions j

and addenda of the OM Code that have been incorporated by reference into S 50.55a. f Paragraph 50.55a(b)(3) together with 9 50.55a(f)(4) of the proposed rule would require that licensees implement the 1995 Edition with the 1996 Addenda of the OM Code. Existing _

6 50.55a(f)(1) has been modified to clarify which pumps and valves are to be included in the IST program. One proposed limitation to implementation oi the OM Code addressing QA, and one proposed modification bf the OM Code addressing stroke time testing have been included.

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. . ._ ~- - - -

2.3.2.2 Background OM Code i Until 1990, the ASME Code requirements addressing IST of pumps and valves were

- contained in Section XI Subsections lWP (pumps) and IWV (valves). The provisions of IWP and IWV were last incorporated by reference into $ 50.55a in a final rulemaking published on August 6,1992 (57 FR 34666). In 1990, the ASME published the initial edition of the OM Code which provides rules for IST of pumps and valves. The requirements contained in the 1990 Edition are identical to the requirements contained in the 1989 Edition of Section XI Subsections lWP (pumps) and IWV (valves). The ASME Board on Nuclear Codes and Standards has transferred responsibility for rules on IST from Section XI to the OM Committee. As such, the Section XI rules for inservice testing of pumps and valves that are presently incorporated by reference into NRC regulations are no longer being updated by Section XI.

The ASME 1990 Edition of the OM Code consists of one section (Section IST) entitled

" Rules for Inservice Testing of Light Water Reactor Power Plants." This section is divided into four subsections, ISTA, " General Requirements," ISTB, " Inservice Testing of Pumps in Light-Water Reactor Power Plants," ISTC, " Inservice Testing of Valves in Light-Water Reactor Power Plants," and ISTD, " Examination and Performance Testing of Nuclear Power Plant Dynamic Restraints (Snubbers)." The IST of snubbers is govemed by plant technical specifications and, 4

thus, has never been included in S 50.55a. Therefore, this proposed rule only requires implementation of Subsectior:s ISTA, ISTB, and ISTC. However, S 50.55a(b)(3)(v) would permit licensees to implement Subsection ISTD of the 1996 Addenda by making a change to their technical specifications in accordance with applicable NRC requirements.

19

2.3.2.3 Clarification of Safety-Related Valves The existing S 50.55a(f)(1) has been interpreted by some licensees to mean that all

, safety-related pumps and valves regardless of ASME Code Class (or equivalent) were to be included in the IST program. The NRC proposes to modify this paragraph to clarify that the provisions of 9 50.55a(f)(1) apply only to pumps and valves in steam, water, air, and liquid radioactive waste systems that perform a function to shut down the reactor, maintain tne reactor in a safe shutdown condition, mitigate the consequences of an accident, or provide overpressure protection for such systems.

2.3.2.4 Limitation d 2.3.2.4.1 Quality Assurance The limitation to the implementation of the OM Code pertains to the use of NQA-1,

" Quality Assurance Requirements for Nuclear Facilities,"with the OM Code. The OM Code ,

references the use of either NOA 1 or the Owner's Appendix B Quality Assurance Program as part of it Individual requirements for a QA program. At pcesent,9 50.55a endorses NOA-1-1979 for the OM Code. The 1996 Addenda also endorses NOA 1-1979. Thus, the 1996 OM Code has not endorsed a later version of NQA-1. Because this rulemaking wou'a incorporate the OM Code by reference into 6 50.55a for the first time, a limitation is include d to aFJress the same issues discussed previously in the Section XI section on QA.

The NRC has determined that the provisions of NQA-1,1979 Addenda, wc aid not adequately describe how to satisfy the requirements of Appendix B as satisfied by S 50.34(b)(6)(ii). Further, there are various aspects of opera +ional phase QA and administrative 20

controls which are not addressed by NQA-1. There are numerous areas where American National Standards Institute (ANSI) standards or NRC regulatory positions, which are specified in SRP 17.2, are either nonmandatory or missing altogether from the NQA 1 provisions. However, the Owner's QA Program, which has been approved by the NRC, is adequate. Thus, the NRC has determined that the requirements of NOA 1 1979, that are part of the incorporation by reference of the OM Code, is acceptable for use in the context of the OM Code, as permitted by ISTA 1.4, provided the licensee utilizes its 10 CFR Part 50, Apoendix B, QA program in conjt...: tion with the OM Code. Changes to licensee's CA program shall be made in accordance with 10 CFR 50.54. Further, where NQA 1 and the OM Code do not sodress the commitments contained in the licensee's Appendix B QA program description, se 1 commitments shall be applied to OM Code activities. Proposed 6 50.55a(b)(3)(1) addres. 1see's commitments related to the OM Code.

2.3.2.5 Modification 2.3.2.5.1 Stroke Time Testing Pry.osed S 50.55a(b)(3)(ii) would require that the stroke time testing requirement of Subsection ISTC of the OM Code applicable for motor-operated valves (MOVs) be supplemented with programs that licensees have previously coriimitted to perform, prior to issuance of this amendment to S 50.55a, for demonstrating the design basis capability of MOVs. Stroke time testing of MOVs has been specified in ASME Section XI and is currently required by S 50.55a(f).

This same testing is required by the OM Code. This testing is a useful tool and complements other tests used to verify MOV functiori. Variation in measured stroke times can indicate valve

. degradation. Addiilonally, periodic stroking provides valve exercise and some measure of on-demand reliability. However, as discussed in NRC Generic Letter (GL) 89-10 " Safety-Related 21 l

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

^

i h

? Motor-Operated Valve Testing and Surveillance" dated June 28,1989, it is now recognized that

- the stroke' time' testing alone is not sufficient to provide assurance of MOV capability under -

' design-basis conditions.

1 Subsequent to licensees implementing programs pursuant to GL 89-10, the NRC issued t

Generic Letter 96-05, " Periodic Verification of Design-Basis Capability of Safety-Related

. Motor-operated Valves," on September 18,1996. This generic letter requested licensees to ,

establish a program, or to ensure the effectiveness of their current program, to verify on a periodic basis that safety-related motor-operated valves continue to be capable of performing

their safety functions within the current licensing bases of the facility. Prior to issuance of this s rule, licensees have made licensing commitments pursuant to GL 96-05 that have been reviewed

. by the NRC staff. Most licensees have committed to parti::ipate in the Joint Owners Group (JOG) Program on MOV Periodic Verification. The JOG program includes three phases:

(1) licensees will establish an interim static diagnostic testing program developed by JOG with a test freq'uency based on margin and safety significance; (2) JOG will coordinate a dynamic

- testing program over the next 5 years that includes approximately 150 MOVs with participating licensees each testing a few MOVs three times over this interval; and (3) based on the results of the dynamic testing program, JOG will establish a long-term periodic test program. Proposed

$ 50.55a(b)(3)(ii) wouid require that licensees supplement the stroke time testing requirements of

- the OM Code with these commitments. .

f1 in .

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.... n ~ -

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l l2Ai Expedited implementationi 2A.11 Appendix Vill- -

fThe proposed rule would require that licensees expedite implementation of mandatory -

l Appendix Vill, " Performance Demonstration for Ultrasonic Examination Systems," to Section XI,

~

1995 Edition with the 1996 Addenda. Three proposed modifications would be included to address NRC positions on the use of Appendix Vill, Licensees would be required to implement -

Appendix Vill, including the modifications, for all examinations of the pressure vessel, piping, nozzles, ano oolts and ' studs which occur after 6 months from the date of the final rule. -The proposed rule would not require any change to a licensee's ist schedule for examination of these -

components, but would require that the provisions of Appendix Vlli be used for all examinations after that date rather than the ultrasonic testing (UT) procedures and personriel requirements - ;

iD presently being utilized by licensees.'

~ Appendix Vill provides the requirements for performance demonstration for ultrasonic l

testing (UT) procedures, equioment, and personnel used to detect flaws and size flaws, its requirements are applicable to all UT performed for Class 1, Class 2, and Class 3 items (i.e., '

- t- - _

. reactor vessel, nozzles, piping, and bolting and studs). These requirements are also to be utilized when implementing the augmented inservice Inspection program for reactor vessel shell -

welds presently required by $ 50.55a(g)(6)(ii)(A). The NRC has reviewed the 1995 Edition with the 1996 Addenda of Appendix Vill and has determined that the provisions contained in this i

it

? appendix should be used with three modifications (addresseu colow). This mandatory appendix

~.would normally be adopted as part.of the routine 120-month update specified in 9 50.55a(g)(4),-

but because of the importance of the Appendix Vill program, the NRC has determined that its 3 requirements should be implemented after 6 months from the date of the final rule. The s

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N

t performance demonstration requirements in Appendix Vill would substantia!!y improve the ability.

- of an examiner to detect and characterize flaws in examined components. UT procedures and personnel requiremente are presently' contained in Section XI but, as detailed in the documented evaluation required by 9 50.109(a)(4), personnel qualified to Appendix Vill are significantly better at detecting flaws. The industry's Performance Demonstration Initiative (PDI) established a process in accordance with Appendix Vill for reactor vessel, nozzle, piping, and bolting examinations. PDI has received considerable support from the industry, and every licensee has contributed financially. The majority of the cost of PD1 was in setting up the samples, which has

' been completed. Proposed S 50.55a(g)(6)(ii)(C)(1) would require licensees to utilize the improved requirements in Appendix Vill for all examinations of reactor vessels (includirig nozzles), piping, and bolting performed after 6 months from the date of the final rule. To date, the PDI program has qualified over 300 individuals for piping and five teams for vessel

' vaminations.' Thus, the NRC does not believe that a 6-month implementation period would result in hardship.

2.4.1.1 Modifications 2.4.1.1.1 Appendix Vill Personnel Qualification The first proposed modification of Appendix Vill relates to its requirement that ultrasonic examination personnel meet the requirements of Appendix Vil," Qualification 0: Nondestructive Exam; nation Personnel for Ultrasonic Examination," to Section XI. Appendix Vll first appeared in Section XI in the 1988 Addenda and was incorporated by reference into S 50.55a in a final rule published on August 6,1992 (57 FR 34666). The NRC believes that the requirement in Appendix Vll-4240 for personnel to receive a minimum of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of training on an annual basis t>

inadequate. Proposed 9 50.55a(b)(2)(xvii) would require that all personnel qualifiad for 24 i

performing ultrasonic examinations in accordance with Appendix Vill receive 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months of annual training which includes laboratory work and examination of flawed specimens. Signals can be difficult to interpret, and as detailed in the regulatory analysis for this rulemaking, experience and studies indicate that the examiner must practice on a frequent basis to maintain the capability for proper interpretation. In addition, these studies hav'shown that this capability begins to diminish within approximately 6 months if skills are not maintained. Thus,10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of annual training is not sufficient practice to maintain skills. The NRC believes that a mirdmum of 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months of annual training, not 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , is required to maintain an examiner's abilities in this highly specialized skill area. The NRC expects that licensees would distribute the training over the cotrse of the year to ensure that interpretation skills do not diminish.

2.4.1.1.2 Appendix Vlli Specimen Set Cracks The second proposed modification of Appendix Vill would require that all flaws in the specimen sets used for performance der- nstration for piping, vessels, and nozzles be cracks.

For piping, Appendix Vill requires that all of the flaws in a specimen set be cracks. However, for vessels and nozzles, Appendix Vlli would allow as many as 50% of the flaws to be notches. For the purpose of demonstrating nondestructive examination (NDE) capabilities, notches are not realistic representations of service induced cracks. An inspector cannot properly interpret service induced cracks by qualifying with specimens containing notches. Notches are easier to detect than flaws because notches have a higher amplitude and simpler signal characteristics.

Notches are easier to interpret and, in fact, the probability of detecting notches can be much higher than the probability of detecting cracks under similar conditions. In addition, Appendix Vill provides a screening test that uses a relatively small sample size containing few flaws. If some of the flaws are replaced by notches that are unrealistic, the screening test becomes ineffective.

25

Because of these considerations, the flaws in the specimen sets utilized for piping by EPRI for the PDI are all cracks. The regulatory snalysis for this rulemaking contains a detailed discussion of the importance of using cracks in the specimens. Thus, proposed 9 50.55a(b)(2)(xiii) would require that all flaws in the specimen sets used for performance demonstration be cracks.

2.4.1.1.3 Appendix Vill Specimen Set Microstructure The third proposed modification of Appendix Vlli would require that all specimens for single-side tests contain microstructures like the components to be inspected and flaws with non-optimum characteristics consistent with field experience that provide realistic challenges to the UT technique. Appendix Vlli does not distinguish cpecimens for two-sided examinations from those used for single sided examination.

Appendix Vill was originally developed usinq UT lessons leanted from two-sided examinations of welds. This UT experience provided the input for designing specimens and selecting, locating, and characterizing flaws. Studies have shown that defect characteristics such as shape, size, depth, tilt angle, skew angle, roughness, and crack tip affect thy probability of detecting a particular flaw. For example,it was demonstrated in one particular study (Reference 22 in the documented evaluation) that a particular flaw was over three times more reflect:ve in one direction, thus easier to detect, than in the opposite direction. Specimens designed for two-sided examination may not have defects which are uppropiiate for single-sided performance demonstration;l.e., the specimens may not adequately test an examiners proficiency in detecting flaws. Therefore,in order to proceed with the effort of qualifying UT systems (equipment, procedures, and personnel) for single-sided examinations, proposed S 50.55a(b)(2)(xx) would require the industry to develop sets of specimens that contain 26

microstructures similar to the types found in the components to be inspected and flaws with non-optimum characteristics, such as skew, tilt, and roughness, consistent with field experience that provide realistic challenges for single sided performance demonstration.

2.4.2 Generic Letter on Appendix Vill A draft generic letter was published in the Federal Register (61 FR 69120) for public comment on December 31,1996, to alert the industry to the importance of using equipment, procedures, and examiners capable of reliably detecting and sizing flaws in the performance of comprehensive examinations of reactor vessels and piping. The generic letter stated that even though the need for improvement clearly existed, the staff had reached the conclusion that immediate backfitting of Appendix Villin advance of this proposed rulemaking was not warranted. This conclusion was based on consideration of defense-in-depth measures Code margins in component design, leakage monitoring systems, and also that Anpendix Vill was s.iready being applied to selected piping subject to intergranular stress corrosion cracking. The NRC received 16 comment letters on the generic letter.

The comments generally were very similar and can be summarized in the following five items: (1) it is inappropriate to request licensees to voluntariiy commit to a program in a generic letter, (2) the urgency for licensee's to voluntarily commit to implementing Appendix Vill is incora 3nt with the statement in the generic letter that a safety concem does not exist that would warrant immediate backfitting in advance of the rulemaking; (3) the performance-based qualification program of Appendix Vill should be approved an attemative to the current ASME Code, and Appendix Vllt as implemented by PDi should be recognized as an acceptable attomative for Appendix Vill; (4) the NRC should provide guidance on incorporating Appendix Vill 27

and/or PDI into plant specific ISI programs; and (5) the generic letter would request that licensees update their _UT ISI and augmented inspection commitments to a Code edition not yet referenced in the regulations.

With regard to the first comment, the NRC disagrees that it is inappropriate to request licensees to voluntarily commit to a program in a genericletter. This is one mechanism available to the NRC for alerting licenseos, for example, to degraded conditions which may unacceptably affect the function of safety-related components. The second commer,t takes the generic letter statement out of cor' text. What the generic letter actually stated was that a safety concem did not exist to warrant immediate backfitting in advance of the rulemaking because of defense-in-depth measures, Code margins in design, and that Appendix Vill was already being applied to selected piping subject to intergialular stress corrosion cracking. Tha NRC strongly disagrees that Appendix Vlil and Appendix Vill as implemented by PDI should be attematives to the present Cods rules. As detalled in the document 3d evaluation for backfitting gendix Vlli, it has been demonstrated that exeminers previously considered qualified under Section XI generally have marginal UT skills. This was evident from the discourrgingly low percentage of examiners initially satisfying the screening criteria for detecting 'iaws under the PDI presgram. Comment four regarding guidance on incorporating Appendix Villinto present ISI programs, and comment five regarding Code edition are automatically resolved in a miemaking format.

At the time the generic letter was issued, this proposed rulemaking was still under development. The purpose of the generic letter was to alert the industry to the (1) generally poor performance in detecting fl'.ws and (2) the Commission's intent to endorse Appendix Vill via rulemaking. Publication of a final rule would obviate the need for the generic letter.

1 28 ]

i

2.4.3 Class 1 Piping Volumetric Examination A proposed modification of Section XI would require licensees of pressurized water reactor plants to supplement the surface examination of Class 1 High Pressure Safety injection Systems (HPSI) piping as required by Examination Category B-J of Table IWB-2500-1 for nominal pipe sizes (NPS) between 4 (inches) and 1% (inches), with a volumetric (ultrasonic) examination. This requirement is proposed because (1)inside diameter cracking of HPSI piping in the subject size range has been previously discovered (as detailed in NRC Generic Letter ,

85-20, "High Pressure injection /Make-Up Nozzle Cracking in Babcock and Wilcox Plants," and in NRC Information Notice 97-46, "Unisolable Crack in High-Pressure Injection Piping,"), (2) failure of this line could result in a small break loss of coolant accident while directly affecting the system designed to mitigate such an event, and (3) volumetric examinations are already required by the Code for Cicss 2 portions of this system (Table IWC-2500-1, Examination Category C-F-1) within the same NPS range, Thus, not only are the requirements between Class 1 and Class 2 inconsistent (with the Class 1 portions being subject to less stringent testing requirements as compared with Class 2 portions of the same type of piping), but operating experience has shown that these reactor coolant pressure boundary (RCPB) pipe examinations need to be more comprehensive. Proposed S 50.55a(b)(2)(xv) would require licensees to supplement the Section XI required surface examination for the Class 1 portion of the HPSI system with volumetric examination in order to ensure the integrity of the reactor coolant pressure boundary as required by General Design Criteria (GDC) 14,10 CFR Part 50, Appendix A, or similar provisions in the licensing basis for these facilities, and Criteria 11 and XVI of 10 CFR Part 50, Appendix B. Licensees would be required to perform the volumetric oxamination during any ISI e

program inspection of the HPSI system performed after 6 months from the date of the final rule.

l Utilization of licensee's existing ISI schedules will result in the volumetric examinations being l

29

f f

implemented in a reasonable period of time while not impacting lengths of outaget, or requiring facility shutdown solely for performance of these' examinations.

2.5 Voluntaryimplementation 2.5.1- Section ill J The NRC has' reviewed the 1989 Addenda,1990 Addenda,1991 Addenda,1992 Edition, 1992 Addenda,1993 Addenda,1994 Addenda,1995 Edition, and 1996 Addenda of Section ill, Division 1, for Class 1, Class 2, and Class 3 components, and has determined that they are acceptable for voluntary use with six proposed limitations. In addition,6 50.55a would be modified to ensure consistency between 9 50.55a and NCA-1140.

The version of Section lit utilized by licensees is chosen prior to construction. Section 50.55a permits licensees to use the original construction code during the operational phase or voluntarily update to a later version which has been endorsed by 6 50.55a. Accordingly, the proposed limitations to Section lli become effective only when a licensee voluntarily updates to a later version. The modification would only apply to a applicant for a new construction permit.

2.5.1.1 Limitations 2.5.1.1.1 Engineering Judgement The first proposed limitation to the implementation of Section til would establish an NRC

~ restriction with regard to the Foreword in the 1992 Addenda through the 1996 Addenda of the BPV Code. That Foreword addresses the use of" engineering judgement

for construction

activities not specifically considered by the Code. Proposed paragraph 50.55a(b)(1)(i) would 30

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

^

require that when a licensee reres'on engineering judgement for activities .or evaluations of components or systems _within the scopc of $ 50.55a that are not directly addressed by the BPV -

Code, the licensee must' receive NRC approval for those activities or evaluations pursuant to

+

6 50.55a(a)(3). _

i 2.5,1,1.2 Section 111 Materials The second proposed limitation to the implementation of Section til pertains to a .

reference to' Section 11 " Materials," Part D, " Properties." Section 11, Pa'i D, contained many printing errors in the 1992 Edition. These errors were corrected in the 1992 Addenda. Proposed S 50.55a(b)(1)(ii) would require that Section 11,1992 Addenda, be applied when using the 1992 Edition of Section 111. The limitation is necessary to ensure that users of the Code use the design stresses intended by the ASME Code.

2.5.1.1.3 Weld Leg Dimensions The third proposed limitation to the implementation of Section lit would correct a conflict in the design and construction requirements in Subsection NB (Class 1 Components),

Subsection NC (Class 2), and Subsection ND (Class 3) of Section 111,1989 Addenda through the 1996 Addenda of the BPV Code. Two equations in NB-3633.4(c)(1), Footnote 11 to Figure NC-3673.2(b)-1, and Figure ND-3673.2(b)-1 were modified in the 1989 Addenda and are no longer in agreement with Figures NB-4427-1, NC-4427-1, and ND-4427-1. This change results in a different weld leg dirnension depending on whether the dimension is derived from the text or calculated from the figures. Thus, to ensure consistency, proposed S 50.55a(b)(1)(iii) would 31

^

y- ,

f

y [ require that licensees use the 1989 Editiori for the above referenced _ paragraphs and s .-

L lieu of the 1989 Addenda through the 1996 Addenda.

2 5.1.1 AiSeismic Design '

j The fourth proposed limitation to the implementation of Section til pertains to new -

r3quirements for piping design evaluation contained in the 1994 Addenda through the 1996

-t

' Addenda of the BPV Code. The NRC has determined that changes to subarticles NS-3200, -1 4

" Design by Analysis," NS 3600, " Piping Design " NC-3600, Piping Des 19n," and ND 360 Design,' of Section 111 for Class 1,2, and 3 piping design evaluation for reversing dynam c

"(e.g., earthquake and other similar type dynamic loads which cycle about a mean value) a unacceptable The new requirements ats based on the premise that loads such as earthqu >

loads are not capable of producing collapse or gross distortion of a component. The-requirements, in pad, are based on General Electric evaluations of the test data perform

. sponsorship of the Electric Power Research institt.te (EPHI) and the NRC. However, NRC I

evaluations of the data do not support the changes and indicate lower margins than those estimated in earlier evaluations.' The ASME has established a special working group to reevaluate the bases for the seismic design for piping.L Thus, in proposed $ 50.55a(b)(1)(iv),-

licensees would be permitted to use articles NB-3200, NS-3600, NC-3600, and ND-3600, in the '

1989 Addenda through the 1993 Addenda, but woula be prohibited from using these ~

. requirements in the 1994 Addenda through the 1996 Addenda.

~

M 32

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g .

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1 2.5.1.1.5 Quality Assurance The fifth proposed limitation to the implementation of Section 111 pertains to the use of NQA-1, " Quality Assurance Requirements for Nuclear Facilities," with Section Ill. Section ill references NQA 1 as part of its individual requirements for a QA program by integrating portions of NQA-1 into the QA program defined in NCA-4000," Quality Assurance." At present,9 50.55a endorses the 1989 Edition of the ASME Code which references NOA-1-1986 for Section 111. The 1996 Addenda of the ASME Code references NQA-1-1992 for Section Ill.

The NRC has reviewed the requirements of NOA-1,1986 Addenda through the 1992 Addenda, that are part of the incorporation by reference of Section Ill, and has determined that the provisions of NQA-1 are acceptable for use in the context of Section lit activities. Portions of NQA-1 are integrated into Section 111 administrative, quality, and technical provisions which provide a complete QA program for design and construction. NQA 1 by itself would not adequately describe how to satisfy the requirements of 10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants." The additional criteria contained in Section lil, such as nuclear accreditation, audits, and third party inspection, establishes a complete program and satisfies the requirements of Appendix B (i.e., the provisions of Section 111 integrated with NQA-1). Because licensees may voluntarily choose to apply later provisions of Section lit, proposed S 50.55a(b)(1)(v) contains a limitation which would require that the edition and addenda of NOA 1 specified by NCA-4000 of Section ill be used in conjunction with the administrative, quality, and technical provisions contained in the edition of Section ill being utilized.

33

-2.5.1.1.6 Independence of inspeetion The sixth proposed limitation to the implementation of Section til would prohibit licensees from using subparagraph NCA-4134,10(a)," Inspection,"in the 1995 Edition through the 1996 Addenda. Prior to this edition and addenda, NCA-4134.10(a) required that the provisions of NQA 1, " Quality Assurance Program Requirements for Nuclear Facilities," Basic Requirement 10,

" Inspection," and Supplement 10S-1," Supplementary Requirements for Inspection," be utilized without exception, in the 1995 Edition, NCA-4134.10(a) was modified so that paragraph 2 of Supplement 10S-1 and the requirements for independence of inspection were no longer required.

Supplement 10S-1,2.1, states that " Inspection Personnel shall not report directly to the immediate supervisors who are responsible for performing the work being inspected."

Subparagraph 2.2 states "Each person who verifies conformance of work activities for purposes of acceptance shall be qualified to perform the assigned task." By exempting Supplement 10S-1 paragraph 2 from the requirements of NCA-4134.10, Section Ill could promote noncompliance with 10 CFR 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," Criterion 1," Organization." This criterion requires that persons performing QA functions repori to a management level such that authority and organizational freedom, including sufficient independence from cost and schedule when opposed to safety considerations, are prov*H. Thus,in proposed 6 50.55a(b)(1)(vi), licensees would be permitted to use the provisions contained in NCA-4134.10(a), in the 1989 Addenda through the 1994 Addenda, but would be prohibited from using these provisions in the 1995 Edition through the 1996 Addenda.

34

2.5.1.2 Modification

- 2.5.1.2.1 Applicable Code Version for New Construction The proposed modification of Section lit addresses a possible cratflict between 11CA-1140 and S 50.55a for new construction. NCA-1140 of Section lit requires that the length of time between the date of the edition and addenda used for new construction and the docket date of the nuclear power plant be no greater than three years. Paragraph 50.55a(b)(1) requires that the edition and addenda utilized be incorporated by reference into the regulations. The possibility exists that the edition and addenda required by the ASME Code to be used for new construction would not be incorporated by reference into 9 50.55a. In order to resolve this possible discrepancy, the NRC proposes to modify exist:ng $$$ 50.55a(c)(3)(1),50.55a(d)(2)(i), and 50.55a(e)(2)(i), to permit an applicant for a construction permit to use the fatest edition and addenda wh5 has been incorporated by reference into S 50.55a(b)(1)If the requirements o h i ASME Code and the regulations cannot simultaneously be satisfied.

2.5.2 Section XI (Voluntary implementation)

Licensees would be permitted to update from the 1992 Edition with the 1992 Addenda of Subsection IWE and Subsection IWL to the 1995 Edition with the 1996 Addenda. In addition, licensees could implement Code Case N-513, Evaluation Criteria for Temporary Acceptance of Flaws in Class 3 Piping," and Code Case N-523-1, " Mechanical Clamping Devices for Class 2 and 3 Piping."

35

-4 I

2.5.2.1 Subsection IWE and Subsect'en IWL r

Many of the provisions in Section XI Subsection (WL, " Requirements for Class CC ,

Concrete Components of Light Water Cooled Power Plants," pertaining to the inspection of the ;

tendons of concrete containments were based on guidance contained in Regulatory Guide 1.35, i

" Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments." A final rule

- published on August 8,1996 (61 FR 41303) incorporated by reference the 1992 Edition with the 1992 Addenda of Subsection IWE," Requirements for Class MC and Metallic Liners of Class CC

- Components of Light Water Cooled Power Plants." and Subsection lWL At that time, there were several key positions in the regulatory guide addressing the trending of prestress losses,

- unanticipated tendon elongation, grease leakage, and excessive water in the sampled sheathing filler grease not addressed in Subsection IWL because the ASME Code committees had not yet completed consideration of these positions. Due to the importance of these positions, the final rule addressed them in paragraphs 50.55a(b)(2)(ix)(A) through 50.54(b)(2)(ix)(D)(2), in addition, the final rule contained $ 50.55a(b)(2)(ix)(E) which addressed the occurrence of-

degradation in lanccessible areas of containments.

1 i

Since publication of the 1992 Addenda, the ASME Code committees have completed their consideration of those regulatory guide positions. Most have been incorporated into subsequent edition and 'addenoa, and the 1995 Edition with the 1996 Addenda addresses all of

the modifications listed above except grease leakage and degradation in inaccessible areas.

.Thus, licensees would be required to utilize the modifications presently in 9 50.55a addressing -

- grease leakage and degradation in inaccessible areas. The NRC has determined that the provisions contained in Subsection'lWE and Subsection IWL,1995 Edition with the 1996 Addenda Code, in conjunction with the modifications, would be acceptable.

~

36 4-.-.,- .-,ry. _m-. , ,--r , , .- w,, ,.m,-

. The final mie published on August 8,1996 (61 FR 41303) incorporated Subsection IWE and Subsection lW1. into 6 50.55a for the first time. The final rule contained a requirement for licensees to develop and implement a containment ISI program within five years. Each plant had a pre-existing ISI program to address Class 1, Class 2, and Class 3 components. The rule left it to the licensee's discretion whether to have two separate ISI programs, or merge the containment ISI program with the pre-existing program.

It has been over a year since the final rule was issued, and some licensees have begun the development of a containment ISI program to comply with the required 5 year implementation period. This containment ISI program will be based on the 1992 Edition with the 1892 Addenda n required by the final rule. However, other licensees have indicated that they will request NRC approvai pursuant to $ 50.55a(a)(3) to use later editions and addenda of Subsection IWE and Subsection IWL before this proposud rule becomes final. Thus, to provide flexibility,

$ 50.55a(b)(2)(vi) has been modified. Licensees wouldi be permitted to implement either the presently required 1992 Edition with the 1992 Addenda, or the latest containment examination proviolens; i.e.,1995 Edition with the 1996 Addenda.

for those licensees implementing the 1992 Edition with the 1992 Addenda, all of the modifications contained in paragraphs 50.55a(b)(2)(ix)(A) through 50.55a(b)(2)(ix)(D)(3) must be applied as presently required by $ 50.55a. Licensecs wishing to implement the 1995 Edition with the 1996 Addenda would be required to apply paragraphs 50.55a(b)(2)(ix)(A),

50.55a(b)(2)(ix)(D)(3), and 50.55a(b)(2)(ix)(E). Paragraph $ 50.55a(b)(2)(ix) would thus be modified. According to S 50.55a(g)(6)(ii)(B)(1), the containment examinations performed during the 5-year implementation period are those examinations which are required by Subsection IWE during the first period of what will be the first containment inspection interval. (Since Subsection 37

4 IWL is based on a 5-year schedule, standard Section XI periods do not apply for the examination

- ol concrete containments and their post-tensioning systems). With completion of the Crst period examinations, the second period of the first containment ISIinterval would begin. The end of the third period completes the first containment ISI interval, a containment ISI 120-month update has been completed, and the second containment ISI interval would begin.

As licensees have begun developing their containment ISI programs, the NRC has received requests to clarify the implementation schedule for ISI of concrete containments and

. their post-tensioning systems. The current wording of $ 50.55a(g)(6)(ii)(B)(2) requiring licensees

to implement 'the inservice examinations which correspond to the number of years of operation which are speelfied /n Subsection /WL'has created confusion regarding whether the first examination of concrete is required to meet the examination schedule in Section XI, Subsectiun IWL, IWL-2410, which is based on the date of the Structural integrity Test (SIT), or may be performed at any time between September 9,1996 and September 9,2001, According to S 50.55a(g)(6)(ii)(B)(2) of the final rulemaking, the first examination of concrete may be performed at any time between September 9,1996, and September 9,2001. The date of the first examination of concrete is not conditional upon compliance with Subsection (WL-2410 or the SIT. The purpose of the italicized words is to maintain the present 5-year schedule for examination of the post-tensioning system as operating plants transition to Subsection IWL. For operating reactors; there is no need to repeat the 1,3,5-year implementation cycle.

Section 50.55a(g)(6)(ll)(B)(2) also stated thehhe first examination performed shall serve the same purpose for operating plants as the preservice examination specified for plants not yet in operatior The affected plants are presently operating, but they will be performing the examination of concrete under Subsection lWL for the first time. Because the plants are 38

operating, a Section XI preservice examination cannot be performed. Therefore, the first concrete examination is to be an inservice examination which will serve as the baseline (the same purpose for operating plants as the preservice examination specified for plants not yet in operation). With completion of this first examination of concrete, the second five-year Subsection IWL ISI period would begin. Likewise, examinations of the post-tensioning system at the n* year (e.g., the 15th year post-tensionh.g system examination), if performed to the requirements of Subsection IWL, ere to be performed to the ISI requirements, not the preservice requirements.

The NRC has also been requested to clarily the schedule for future examinations of concrete and their post-tensioning systems at both operating and new plants. There is no requirement in Subsection IWL to perform the examination of the concrete and the examination of the post-tensioning system at the same time. The examination of the concrete under Subsection lWL and the examination of the liner plates of cooc ele containments under Subsection IWE may be performed at any time during the 5-year expedited implementation. This examination of the concrete and liner plate provides the baseline for comparison with future containmer't ISI. Coordination of theco schedules in future examinations is left to each licensee.

j Now plants would be required to follow all of the provisions contained in Subsection IWL, i.e.,

j satisfy the preservice examination requirement 1 and adopt the 1,3, 5-year examination schedule ISI schedula.

2.5.2.2 Flaws in Class 3 Piping Proposed S 50.55c(b)(2)(xvi) would permit licensees to use Code Case N-513,

" Evaluation Criteria for Temporary Acceptance of Flaws in Class 3 Piping," and Code Case 39

N-523-1, " Mechanical Clamping Devices for Class 2 and 3 Piping."Section XI contains repair methods for pipes with a flaw exceeding acceptable limits. These repairs restore the integrity of the flawed piping. There are certain cases, however, where a section XI Code repair may be impractical for a flaw detected during plant operation (i.e., a plant shutdown would be required to effect the Code repair). For many safety-related piping systems, immediate repair is required regardless of plant status. However,it has been determined that under certain conditions, temporary acceptance of flaws, including through-wall leaking, of low and moderate energy Class 3 piping is acceptable provided that the conditions are met, and the repair is effected during the next outage. At present, licensees must request NRC staff approval to defer Section XI Code repair for these Class 3 moderate energy (200 'F,275 psig) piping systems. The NRC has reviewed Code Case N-513 and Code Case N 523-1 and has deterniined that Code Case N 523-1 is acceptable. Code Case N-513 is acceptable except for the scope and Section 4.0.

Section 1.0(a) of the Scope to Code Case N 513 limits the use of the requirements to Class 3 piping. However, Section 1.0(c) would allow the flaw evaluation criteria to be applied to all sizes of ferritic steel and austenitic stainless steel pipe and tube. Without some limitation on the scope of the Code Case, the flaw evaluation criteria could be applied to components such as pumps and valves, original construction deficiencies, and pressure boundary leakage; applications for which the criteria should not be utilized. Thus, the NRC has determined that the Code Case shall not be applied to: (1) components other than pipe and tube, such as pumps, valves, expansion joints, and heat exchangers; (2) the discovery and repair of flaws or deficiencies remaining from original construction; (3) leakage through a flange gasket; (4) threaded connections employing nonstructural seal walds for leakage prevention (through seal weld leakage is not a structural flaw, thread integrity must be maintained); and (5) degraded 40

Pocket welds. A proposed limitation would be added in $ 50.55a(b)(2)(xvi)(B) which would preclude the use of Code Case N-513 for these applications.

The first paragraph of Section 4.0 of Code Case N 513 contains the flaw acceptance criteria. The criteria provide a safety margin based on snVee loading conditions. The second paragraph of Section 4.0, however, would permit a reduction of the safety factors based on a detailed engineering evaluation. No criteria or guidance is given forjustifying a reduction, or limiting the amount of reduction. The acceptance criteria of the first paragraph are based on so md principles. The second paragraph would allow ever finer calculation until the avalfable margins became unacceptablylow. A limitation would be added in proposed 9 50.55a(b)(2)(xvi)(A) requiring that when implementing Code Case N-513, the specific safety factors in the first paragraph of Section 4.0 be wtisfied. The use of Code Case N 513, with the limitations, and Code Case N 5231 would obviate the need for licensees to request approval for deferring repairs, thus saving 14RC ar.J licensee resources.

2.5.3 OM Code (Voluntary implementation) r Licensees would be permitted to implement Code Case OMN-1 Irilieu of stroke time testing cs required in Subsection ISTC. Licensees would also be permitted to implemer;t Appendix il as an attemhtive to the condition monitoring program provisioris contained in Subsectio'n ISTC. However, licensees choosing to implement Appendix 11 would be required to apply the three proposed modifications to Appendix ll to supplement check valve condition monitoring, in addition, licensees would be permtMed to use Subsection ISTD for the IST of snubbers.

41 s

l l

. 2.5.3.1 Code Case OMN 1 ,

L i An attemative to the provisions contained in $ 50.55a(b)(3)(ii) is included in proposed ;

$ 50.55a(b)(3)(lii) which would permit licensees to voluntarily implement ASME Code Case OMN 1,Altemative Rules for Preservice and Inservice Testing of Ciertain Electric Motor Operated Valve Assemblies In LWR Power Plants? The NRC has determined that for motor-operated valves, Code Case OMN 1 is acceptable in lieu of subsection ISTC, except for leakage rate testing (ISTC 4.3) which must continue to be performed. As indicated in Attachment i to GL 96-05, the Code case meets the intent of the gene',c letter, but with certain limitations which were discussed in the generic letter. The NRC supports the OMN 1 maximum mutor operated valve test interval of 10 years based on current knowledge and experience, but believes it prudent to require that licensees evaluate the inSrmation obtained for each motor operated valve during the first five years of use of the Code case, or three refueling outages (whichever is longer) to validata assumptions made in justifying a longer test interval. These limitations on the use of OMN-1 would be added to the rule as a modification in $ 50.55a(b)(3)(lii)(A). Thus, Code Case OMN 1 la acceptable in lieu of Subsection ISTC, other than leakage rate testing requirements, with the modification that five years or three refueling outages (whichever is longer) from initial implementation of Code Case OMN-1, the adequacy of the test interval for each motor operated valve must be evaluated and adjusted as necessary. ,

in addition, as noted in GL 96-05, licensees are cautioned when implementing Code Case OMN 1 that the benefits of performing a particular test should be balanced against the potential adverse effects placed on the valven or ystems caused by this testing. Code Case OMN-1 specifies that an IST program should consist of a mixture of static and dynamic testing. While there rnay be benefits to performing dynamic testing, there are also potential detriments to its use i

42

.. - - - . , .w,- ,,- w & *w, .,,y-w, , , ,y.-n-e pw,y y-,,,-. , . ,yw y- apy,vvyc-

i (i.e., valve damage). Licensees should be cognizant of this for owch MOV when selecting the !

appropriate method or combination of methods for the IST program. ,

2.5.3.2 Appendix il <

Paragraph ISTC 4.S.5 of Subsection ISTC permits the owner to use Appendix ll, ' Check Valve Condition Monitoring Program," of the OM Code, as an alternative to the testing or ,

examination provisions of ISTC 4.5.1 through ISTC 4.5.4. If an owner elects to use Appendix 11, the provisions of Appendix ll become idandatory. However, upon reviewing the appendix, the NRC has determined that the requirements in Appendix ll must be supplemented. The first area that the NRC believes requires supplementation is the demonstration of acceptable valve performance, Appendix ll requires no testing or examination of the check valve obturator movement to both the open and closed positions. Testing or examination of the check valve obturator in one direction only cani.M assure the unamblguous detection of a functionally .

degraded check valve. The valve obturator must be tested nr examined in both the opening and closing directions to ascess its condition and confirm acceptable acceptable performance.

Proposed 6 50.55a(b)(3)(iv)(A) would require bi-directional testing of check valves.

Length of test intervalis the second area of Appe . dix 11 where the NRC believes the rules must be supplemented. Appendix ll was first incorporated into the OM Code in the 1996 Addenda. Thus, the operating experience database does not yet exist to support long term test intervals for the condition monitoring concept. Under the current check valve IST program, most velves are tested quaterly during plant operation. The interval for certain valves has been extended to .a eling outages. Under the appendix, a licensee would be able to extend the Interval without limit. A policy of prudent and safe interval extension dictates that any additional P

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interval extension must be limited to one fuel cycle, and this extension must be based on sufficient experience to justify the additional time. Interval changes or e'"ensions must be l Justified and limited within the existing performance and experience database. Condition f monitoring and the current experience riata base may qualify some valves for an inillal extension to evt1/ (.ther fuel cyc!s, while trending and evaluation of the data rey dictate that the testing

.nterval for some valves be reduced. Extensions of IST Intervals must consider plant safety and be supported by trending and evaluating both generic and plant specific performance data to ensure the component is capable of performing its intended function over the entire IST interval.

Proposed 9 50.55a(b)(3)(iv)(B) would limit the time between the initial test or examination and second test or examination to two fuel cycles or three years (whichever is longer), with additional extensions limited to ene fuel cycle, and the total interval would be limited to a maximum of 10 years. An extension or reduction in the interval between tests or examinations would have to be supporied by trending and evaluation of performance wta.

The final crea in Appendix 11 which the Commission believes should be supplemented is the requirement applicable to a licensee who discontinues a constion monitoring program. A licensee who discontinues use of Appendix it, under IST 4.5.5 is required to return to the requirements of IST 4.5.4. However, the NRC believes the requirements of IST 4.5.1 through IST 4.5.4 must be also met. Hence, if the monitoring program is discontinued, proposed 9 50.55a(b)(3)(lii)(C) would require a licensee to implement the provisions of IST 4.5.1 through IST 4.5.4. ,

44 .

i 2.5.3.3 Subsection ISTD

. The IST of dynamic restraints or snubbers is govemed by plant technical specification and, thus, has never been included in 9 50.55a. However, the NRC has reviewed Subsection ISTD,1995 Edition with the 1996 Addenda, and has determined that the provisions for iST of snubbers are an acceptable attemative to the requirements contained in the plant teen.11 cal specifications. Subsection ISTD,1996 Addenda, includes new provisions for service life monitoring of snubbers. The new provisions require that the service lives of snubbers be predicted and evaluated to ensure that the service life will not be exceeded before the next scheduled refueling outage. These new provisions simply formalize preventativo maintenance practices presently found in most plants. Because the IST of snubbers is govemed by plant ,

technical specifications, Subsection ISTD is not included in the proposed mandatory i requirements of the m 3 making, but licensees mcy choose to voluntarily implement Subsection ISTD,1995 Edition with the 1096 Addendo. by processing a change to their technical specifications. This proposed modification is contained b $ 50.55a(b)(3)(v).

2.5.3.4 Containment isolation Valves Tha proposed amendment would delete the existing modification in S 50.55a(b)(2)(vii) for IST of containment isolation valves (CIVs), wiiich was added to the regulations in a rulemaking effective on Augest 6,1992 (57 FR 34666). That rulemaking incorporated by reference, among other things, the 1989 Edition of ASME Section Xt. Subsection IWV that endorsed Part 10 of ASME//,NSI OMa 1988 for valve inservice testing. A modification to the testing requirements of Pwt b related to CIVs was included in the rulemaking indicating that paragraphs 4.2.2.3(e) and 4.2.2.3(f) of Part 10 were to be applied to CIVs. As noted in the " Supplementary Information" for 45

l the August 6,1992 rulemaking, the ASME Operations and Maintenance (OM) Committee had initiated action to: (1) perform a comprehensive review of OM Part 10 CIV testing requirements and acceptance standards; and (2) develop a btsis document that would provide, as a minimum, a documented basis for not including the requirements for analysis of leakage rates and corrective actions in Part 10 for those CIVs that do not provide a reactor coolant system pressure ;

Isolation function. The NRC made a commitment via the Supplementary information to reevaluate the need for the modification to Section XI, Subsection lWV, following review of this OM Committee basis r5 cument. This basis document was transmitted to the NRC in d letter 1 from Steve Weinman, Secretary, OM Committee, to Eric S. BecNord, Director, Office of Nuclear Regulatory Research, dated February 16,1994. The NRC has determined that the requirements of 10 CFR 50, Appendix J, ensure adequate identhication analysis, and corrective actions for leakage monitoring of CIVs, and that the existing modification in $ 50.55a(b)(2)(vil) should be deleted. The regulatory analysis for this proposed rule contains a detailed discussion of the basis document findings and the NRC staff evaluation. .

2.6 ASME Code Interpretations The ASME issues Interpretations to clarify provisions of the BPV and OM Codet.

Requests for interpretations are submitted by users, and after appropriate committee deliberations and balloting, responses are issued by the ASME, Generally, the NRC agrees with l these interpretations. When the NRC incorporates by referance specific editions and addenda into its regulations, the NRC has a certain understanding of those editions and addenda.

Because an Interpretation is issued subsequent to issuance of the provision to which it refers, the lmerpretation may arfect that understanding. Whlie the NRC acknowledges that the ASME is :

l

. the official interpreter of the Code, the NRC will not accept ASME interpretations that, in NRC's l

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l opinion, are contrary to NRC requirements or may adversely impact facility operations.

Interpretations have been issued which in some cases, conflicted with or were inconsistent with NRC reouirements. These resulted in enforcement actions. Of particular concern are Code j l

interpretations that may be implemented following initiation of enforcement action by the NRC.

ASME Code Interpretations were discussed in Part 9900, Technical Guidance, of the NRC Inspection Manual. Part 9900 provides that licensees should exerciso caution when applying j interpretations as they are not spec.ifically part of the incorporatiori by reference into S 50.55a and have not received NRC approval.

2.7 DSI 13 Since 1992, when the Commission last revised S 50.55a to endorse new ASME Code Editions and addenda (57 FR 34660), several decolopments have occurred which have raised some fundamentalissues with respect to the Commission's endorsement of ASME Codes. First, on October 21,1993, Entergy Operations, Inc. submitted a request that would relieve it from updating its ISI and IST programs to the last ASME Code edition and addenda incorporated by reference into 6 50.55a. The underlying premise of the request was that a licensee should not be required to upgrade its ISI and IST program without considering whether the costs of the upgrade are warranted in light of the increased safety afforded by the updated Code edition and audenda.

Though the request was later withdrawn, the underlying premise resulted in NRC reconsideration of the 120-month update. Requiring Code updates every 120-months is still under active consideration. However, the proposed rule has been prepared under the traditional approach; i.e., licensees would be required to update their iSI and IST programs every 120-months to the l

latest edition and addenda incorporated by reference into S 50.55a. If a decision is reached l

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.absequent to pubil:ation of the proposed rule that is adverse to this approach, this position will be corrected prior to publication of the final rule.

Second, the National Technology Transfer and Advancement Act of 1995, PL 104113, was signed into law on March 7,1996. The Act directs federal agencies to achieve greater reliance on technical standar:Is developed by voluntary consensus standards development organizations. Finally, the Commission commenced a Strategic Assessment and Rebaselining Initiative. One of the issues addressed in this effort was Direction Setting issue (DSI) 13, which raised the cuestion, *In performing its regulatory responsibilities, what consideration should the NRC glve to industry activities." A draft paper addressing DSI 13 was published for public comment on September 16,1996, after which the Commission held public meetings to facilitate understanding of the issues and receive commerts on the DS! 13 draft paper. Based on the public comments, the Commission has directed the NRC Staff to address how industry initiatives should be evaluated, and to evaluate severalissues related to NRC endorsement of industry codes and standards. As part of this evaluation, the Staff is addressing issues relevant to the NRC's endorsement of the ASME Code, including periodic updating, the impact of 10 CFR 50.109 (the Backfit Rule), and streamlining the process for NRC review and endorsement of the ASME Code.

2.8 Steam Generators A'SME Code requirements for repair of heat exchanger tubes by sleevlag were added to Section XI in the 1989 Addenda. Minimum Code requirements for tube sleeving was added to the Code so that licensees would not have to develop sleeving programs and have them approved by the NRC on a case-by-case basis. The NRC has reviewed the Code requirements 48

for sleeving and determined that they are acceptable. However, it should be recognized that there are other relevant requirements, and that a considerable amount of effort is presently being expended due to the number of occurrences of degraded steam generator tubing. For example, licensees are required by either 10 CFR 50.55a(f) or by the plant technical specifications to perform periodic inservice inspections and to repair (e.g., sleeving) or remove from service (by installing plugs in the tube ends) all tubes found to contain flaws exceeding the plugging limit (i.e., tube repair criteria). In addition, current technical specifications contain operationalleakage limits. Licensee's have frequently found it necessary to implement measures beyond minimum Code and technical specification requirements to ensure adequate tube integrity when significant degradation problems are encountered. Thus, the NRC determination that the sleeving requirements are acceptable should be kept in perspective.

3. Finding of No Significant EnvironmentalImpact Based upon an environmental assessment, the Comniission has determined, under the National Environmental Policy Act of 1969, as amended, and the Commission's regulations in Subpart A of 10 CFR Part 51, that this rule, if adcpted, would not have a significant effect on the quality of the human environment and therefore an environmentalimpact statement is not required.

The proposed rule is cne part of a regulatory framework directed to ensuring pressure boundary integrity and the operatior al readiness of pumps and valves. The proposed rule incorporates provisions centained in the BPV Code and the OM Code for the construction, inservice inspection, and inservice testing of components used in nuclear power plants, has been I

{ 49 l

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i updated to incorporate improved technology and methodology. Therefore, in the general sense, the proposed rule would have a positive impact on the environment. ,

t The proposed rule would impose the Section XI 1995 Edition with the 1996 Addenda. As ,

most of the technical changes to this edition /addeads merely incorporate improved technology and methodology, imposition of these requirements is not expected to either increase or ,

decrease occupational exposure. However, imposition of paragraphs IWF 2510, Table IWF 2500-1, Examination Category F A, and IWF-2430, would result in fewer supports being -

examined which would decrease the occupationa! exposure compared to present support inspection plans. It is estimated that an examiner receives approximately 100 millirerrs for every 25 supports examined. Adoption of the new provisions is expected to decrease the total number of supports to be examined by approximately 115 per unit per interval. Thus, the reduction in occupational exposure is estimated to be 460 millirems per unit each inspection interval or 50.14 ,

rems for 109 units.

. The proposed rule would impose Appendix Vill to Section XI,1995 Edition with the 1996 Addenda, BPV Code, for the first time and would expedite its implementation. Appendix \All provides rules for the performance demonstration of ultrasonic examination systems, procedures, an'd personnel, implernentation of this appendix should result in a decrease in occupational exposure. Appendix Vill qualified procedures and personnel should reduce repeat ultrasonic testing (UT), which could reduce occupational exposure. In addition, flaws should be detecteG at an earlier stage of growth resulting in less extensive repair operations, which could further reduce occupationc! exposure.

50 r.---.n , -. r w- ..~- - .,

l The pro' posed rule would incorporate by reference into the regulations tiin 1995 Edition !

with the 1996 Addenda of the OM Code. Imposition of the OM Code is not expected to either increase or decrease occupational exposure. The types of testing associated with the 1995 ,

Edition with the 1996 Addenda of the OM Codo are essentially the same as the OM standards contained in the 1989 Edition of Section XI referenced in a final rule published on August 6,1992 (57 FR 34666).

t Actions required of applicants and licensees to implement the proposed rule ere of the same nature as those applicants and licensees have been performing for many years.

Therefore, this action should not increase the potential for a negative environmental impact.

The NRC has sent a copy of the Environmental Assessment and the proposed rule to every State Liaison Officer and requested their comments on the Environmental Assessment.

The environmental assessment is available for inspection at the NRC Public Document Room, 2120 L Street NW (Lower Level), Washington, DC. Single copies of the environmental assessment are available from Frank C. Chemy, Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone: 301-415-6786, or Wallace E. Norris, Division of Engineering Technology, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone:

301-415-6796.

4. Paperwork Reduction Act Statement This proposed rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). This rule has been submitted to the -

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Office of Management and Budget for review and approval of the paperwork requirements, The public reporting burden for this information collection is estimated to average 67 person-hours per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. The U.S. Nuclear Regulatory Commission is seeking public comment on the potential impact of the information collections contained in the proposed rule and on the following issues:

1 1, is the proposed information collection necessary for the proper performance of the functions of the NRC, including whether the information will have practical utility?

2. Is the estimate of burden accurate?

3. Is there a way to enhance the quality, utility, and clarity of the information to be collected?

4. How can the burden of the information collection be minimized, including the use of automated collection techniques?

Send comments on any aspect of this proposed collection of information,8ncluding suggestions for further reducing the burden, to the Information and Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington DC 20555-0001, or by Internet electronic mall at BJSi@NRC. Gov; and to the Desk Officer, Office of Information and Regulatory Affairs NEOB-10202, (3150-0011), Office of Management and Budget, Washington DC 20503.

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Comments to OMB on the information collections or on the above issues should be submitted by (insert date 30 days after publication in the Federal Register). Comments received after this date will be considered if it is practical to do so, but assurance of consideration cannot be given to comments received after this date,

5. Public Protection Notification The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.

6. Regulatory Anal/sis The Commission has prepared a draft regulatory analysis on this proposed regulation.

The analysis examines the costs and benefits of the alternatives considered by the Commission.

The draft analysis is available for inspection in the NRC Public Document Room,2120 L Street NW (Lower Level), Washington DC. The Commission requests public comment on the draft analysir.. Single copies of the analysis may be obtained from Frank C. Cherny, Division of Engl-neering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone: 301-415-6786, Wallace E. Norris, Division of Engineering Technology, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone: 301415-6796.

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_ _ _ _ _ __ _ _ m . . __ _ -- __ . . _- __ _ _

7. Regulatory Flexibility Certi$ cation in accordance with the Regulatory Flexibility Act of 1980,5 U.S.C. 605(b), the Commission certifies that this rule will not, if promulgated, have a significant economic impact on e substantial number of small entities. This proposed rule affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the scope of the definition of "small entitles" set forth in the Regulatory Flexibility Act or the Small Business Size Standards set out in regulations issued by the Small Business Administration at 13 CFR Part 121.

8. Backfit Analysis The Nuclear Regulatory Commission (NRC) regulations,10 CFR 50.55a, requires that nuclear power plant owners (1) construct Class 1, Class 2, and Class 3 components in accordance with the rules provided in Section Ill, Division 1," Requirements for Construction of Nuclear Power Plant Components," of the American Society of Mechanical Engineers (ASME)

Doller and Pressure Vessel Code (BPV Code), (2) inspect Class 1, Class 2, Class 3, Class MC (metal containment) and Class CC (concrete containment) components in accordance with the rules provided in Section XI, Division 1, "Raquirements for Inservice Inspection of Nuclear Power Plant Components," of the BPV Code, and (3) test Class 1, Class 2, and Class 3 pumps and valves in accoroance with the rules provided in Section XI, Division 1. Licensees are required to update every 120 months to the version of Section XIincorporated by reference into 6 50.55a 12 rnonths prior to the start of a new ten year interval.

54 t

The proposed amendment to 6 50.55a would require licensees to update ISI in accordance with Section XI of the ASME BPV Code and IST in accordance with the ASME OM Code. Licensees would be required to implement the 1995 Edition with the %96 Addenda of (1)Section XI, Division i for Class 1, Class 2, Class 3, Class MC, and Class CC components; (2) the " Code for Operation and Maintenance of Nuclear Power Plants"(OM Code) for Class 1 Class 2, and Class 3 pumps and valves; cnd (3) Appendix Vill,

Performance Demonstration for Ultrasonic Examination Systems," to Section XI, Division 1. As permitted by 9 50.55a(s)(3),

licensees may voluntarily update to the 1989 Addenda through the 1996 Addenda of Section 111 of the BPV' >do, with limitation. in addition, the modification for containment isolation valve inservice testing that applied to the 1989 Edition of the BPV Code has been deleted. Licensees will continue to be required to update their ISI and IST programs every 120 months to the version of Section XI and the OM Code incorporated by reference and in effect at least 12 months prior to the start of a new 120 month interval.

The NRC position on the routine 120-month update to 6 50.55a has consistently been that 10 CFR 50.109 does not require a backfit analysis of the routine 120-month update to 9 50.55a. The basis for the NRC position is that, (1) Soction ill, Division 1, update applies only to new construction (i.e., the edition and addenda to be used in the construction of a plar.t are selected based upon the date of the construction permit and are not changed thereafter, except voluntarily by the licensee), (2) licensees understand that 9 50.55a requires that they update their inservice inspection program every 10 years to t',;c !st6# edition and addenda of Section XI that were incorporated by reference in 9 50.55a and in effect 12 months before the start of the next inspection interval, and (3) endorsing and updating references to the ASME Code, a national consensus standard developed by the participants (including the NRC) wah broad and varied interests, is consistent with both the intent and spint of the backfit rule (i.e., NRC provides for the 55

protection of the public health and safety, and does not unilaterally impose undue burden on i applicants or licensees). Finally, to ensure that any interested member of the pubi.. that rnay not have had an opportunity to pcrtic:pate in the national consensus standard process is able to communicate with the NRC, proposed rules are published in the Federal Register.

The provisions for IST of pumps and valves were originally contained in Section XI Subsections IWP and iWV Section XI,1989 Edition was incorporated by reference in the August 6,1992 rulemaking (57 FR 34666). The 1990 OM Code standards, Parts 1,6, and 10 cf ASME/ ANSI OM41987, are identical to Section XI,1989 Edition. This proposed amendment is an administrative changa simply referencing the 1995 Edition with the 1996 Addenda of the OM Code. Therefore, imposition of the 1995 Edition with the 1996 Addenda of the OM Code Is not a

- backfit.

Appendix Vill, " Performance Demonstra' ion for Ultrasonic Examination Systems,' to Section XI would be used to demonstrate the qualification of personnel and procedures 'or performing nondectructive examination of welds in components of systems that include the reactor coolant system and the emergency core cooling systems in nuclear power facilities.

Appendix Vill would greatly enhance the reliability of detection and sizing of cracks and flaws, and it delineates a method for qualification of the personnel and procedures. The appendix would normally be imposed by the 120-month update requirement, but because of its importance, implementation of Appenulx Villis being expedited by the rulemaking. Because of the expedited implementation schedule, the imposition of Appendix Vill is being considered a backfit.

Licensees would be required to imolement Appendix Vill, including the modifications, for all examinations of the pressure vessel, piping, nozzles, and bolts and studs which occur after 6 months from the date of the final rule. The proposed rule would not require any change to a 56

licensee's 181 schedule for examination of these components, but would require that the provisions of Appendix Vill be used for all examinations after that date rather than the UT procedures and personnel requirements presently being utilized by licensees.

The NRC has concluded, on the basis of the documented evaluation required by 6 5';.109(a)(4), that imposition of Appendix Vill, which would greatly enhance the overall level of assurance of the safety and reliability of ultrasonic examination techniques in detecting and slzing flaws, is necessary to bring the facilities described into compliance with GDC 14,10 CFR Part 50 Appendix A, or similar provisions in the licensing basis for these facilities, and Criteria il and XVI, of 10 CFR Part 50, Appendix 5 The modification to Section XI to require licensees to supplement the surface examination of the Class 1 portion (RCPB) of the HPSI system with volumetric examination would ensure the integrity of the reacter coolant system pressure boundary and maintenance of emergency core cooling system operability. The operability of this system is necessary to ensure the protection of the public health and safety, and the NRC has concluded, onthe basis of the documented evaluation required by 9 50.109(a)(4), that licensees must supplement the Section XI required surface examination for the Class 1 portion of the HPSI system with volumetric examination in order to ensure the integrity of the reactor coolant pressure boundary as required by GDC 14,10 CFR Part 50, Appendix A, or similar provisions in the licensing basis for these facilities, and Criteria 11 and XVI, of 10 CFR Part 50, Appendix B. Volumetric examination would be required during any ISI program inspection of the HPSI system performed after 6 months from the date of the final rule.

57

GDC 14,

Reactor coolant pressure boundary,"(RCPB) or similar provisions in the licensing basis for the, acilities, specify that the RCPB be designed, fabricated, erected, and tested so as to have an extremely low probability of abnormalleakage, or rapidly propagating failure, and of gross rupture. There has recent'y been an occurrence of Cross rupture in the Class 1 portion of a HPSI system, and a number of occunences of abnormalleakage in the RCPB in other plants.

imposition of Appendix Vlli and the HPSI volumetric examination is also necessary to bring the facilities described into compliance with Criteria !!," Quality Assurance Program," and Criteria XVI,

Corrective Actions," of Appendix B to 10 CFR Part 50. Criteria il requires, in part, that a QA program shall take into account the need for special controis, processes, test equipment, tools, and skills to attain the required quality and the need for verification of quality by inspection and test. Evidence indicater that there are shortcomings in the qualifications of personnel and procedures in ensuring the reliability of the examinations. These safety significant revisions to the Code include specific requiremer.ts for UT parformance demonstration, with statistically based acceptance critoria for blind testing of UT systems (procedures, equipment, and personnel) used to detect and size flaws. Criteria XVI requires that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformancer, are premptly identified and corrected, in analyzing the occurrences of pipe break and leakage, it is apparent that the RCPB is subject to certain types of degradation. Information gathered by the NRC staff Indicates that many licensees h::ve not reacted to this serious safety concern by performing more comprehensive examinations. The NRC believes that there is a 'assis for reasonably concluding that such degradation could occur in virtually all PWRs. Because of the serious degradation which has occurred, and the belief that additional occurrences of noncompliance with GDC 14, 58 i

l

I and Criteria ll and XVI will be reported, the NRC has determined that imposition of Appendix Vill and volumetric examination of the HPSI system 6 months after the final rule has been published under the compliance exception to S 50.109(a)(4)(l) is appropriate, therefore, a backfit analysis is not required and the cost-benefit standards of 6 50.109(a)(3) do not apply. A complete discussion is contained in the documented evaluation.

The rationale for application of the backfit rule and the backfit justification for the various items contained in this proposed rule are contained in the regulatory analysis and documented evaluation. The regulatory analysis and documented evaluation are available forinspection at the NRC Public Document Room,2120 L Street NW (Lower Level), Washington, DC. Single copies of the regulatory analysis and documented evaluation are available from Frank C. Chemy, Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone: 301415-6786, or Wallace E.

Norris, Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, Telephone: 301415-6796.

9. List of Subjects in 10 CFR Part 50 l

Antitrust, Classified information, Fire protection, incorporation by reference, Intergovemmental relations, Nuclear power plants and reactors, Penalty, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, as amended, and 5 U.S.C. 553, the NRC is proposing to adopt the following amendments to 10 CFR Part 50.

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10. PART 50 DOMESTIC LIC* NSING d OF PRODUCTION AND UTILIZATION FACILITIES 1 The authority citation for Part 50 continues to read as follows:

AUTHORITY: Secs.102,103,104,105,161,182,183,186,189,68 Stat. 936,937,938, 948,953,954,955,956, as amended, sec. 234,83 Stat.1244, as amended (42 U.S.C. 2132, 2133,2134,2135,2201,2232,2233,2238,2239,2282); secs. 201, as amended, 202,206,88 Stat.1242, as amended, 1244,1246 (42 U.S.C. 5841,5842,5846).

Section 50.7 also issued under Pub. L.95-601, sec.10,92 Stat. 2951 (42 U.S.C. 5851).

Section 50.10 also issued under secs. 101,185,68 Stat. 955, as amended (42 U.S.C. 2131, 2235); sec.102 Pub. L. 91 190,83 Stat. 853 (42 U.S.C. 4332). Sections 50.13,50.54(dd), and 50.103 also issued under sec.108, BB Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23, 50.35,50.55, and 50.56 also issued under sec.185,68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec.102, Pub. L 91 190,83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec. 204,88 Stat.1245 (42 U.S.C.

584'4). Sections 50.58,50.91, and 50.92 also issued under Pub. L.97-415,96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec.122,68 Stat. 939 (42 U.S.C. 2152).

Sections 50.00 50.81 also issued under sec.184,66 Stat. 954, as amended (42 U.S.C. 2234).

Aprendix F also issued under sec.187,68 Stat. 955 (42 U.S.C. 2237).

(b) The approved information collection requirements contained in this part appear in SS50.30, 50.33, 50.33a, 50.34, 50.34 a, 50.35, 50.36, 50.36a, 50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 59.60, 50.61, 50.62, 50.63, 50.64, 50 5, 50.66, 50.71, 60 1

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I 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.120, and Appendices A, B, E, G, H, I, J, K, .

M, N, O, Q, R, and 8 to this part l

i

2. Section 50.55a is amended by adding paragraphs (b)(1)(l), (b)(1)(ii), (b)(1)(lii),

(b)(1)(iv), (b)(1)(v), (b)(1)(v)(A), (b)(2)(xi), (b)(2)(xli), (b}(2)(xlii), (b)(2)(xiv), (b)(2)(xv), (b)(2)(xvi),

(b)(2)(xv!)(A), (b)(2)(xvi)(B), (b)(2)(xvi)(B)(.1), (b)(2)(xvi)(B)(2), (b)(2)(xvi)(B)(1), (b)(2)(xvi)(B)(4),

? (b)(2)(xvi)(B)(1), (b)(2)(xvil), (b)(2)(xvili), (b)(2)(xix), (b)(2)(xx)(A), (b)(2)(xx)(B), (b)(3), (b)(3)(i),

'(b)(3)(li), (b)(3)(li)(A), (b)(3)(lii), (b)(3)(lii)(A), (b)(3)(lii)(B), (b)(3)(lii)(C), (f)(3)(lii)(B), (f)(3)(lv)(B), ;

(g)(6)(A)(Q), (g)(6)(C)(1), and (9)(6)(C)(2), and revising paragraphs (b)(1), (b)(2), (b)(2)(iv)(A), !

- (b)(2)(lv)(B), (b)(2)(vi), (b)(2)(vill), (b)(2)(tx), (c)(2), (f)(1), (f)(3)(ill)(A), (f)(3)(iv)(A), (f)(4), (f)(5)(li),

(g)(1), (g)(3)(i), (g)(4), (g)(6)(li)(A)(1), (g)(6)(ii)(A)(2), (g)(6)(ll)(A)(Q), Footnote 4, Footnote 5, and Footnote 7, and deleting the requirements in paragraph (b)(2)(vii) as follows:

A 50.55a Codes and standards.

(b) The ASME Boiler and Prossure Vessel Code, and the ASME ode for Operation and Maintenance of Nuclear Power Plants, which are referenced in the following paragraphs, were approved forincorporation by reference by the Director of the Federal Register. A notice of any changes inade to the materir .ncorporated by reference will be published in the Federal j . Register. Copies of the ASME Boller and Pressure Vessel Code and the ABME Code for p _ Operation and Maintenance of Nuclear Power Plants may be purchased from the American .

L i l[

61

.. - . ~ , , . . ,- - -. ~- - - . - ~ . . . .-

l Society of Mechanical Engineers, United Engineering Center,345 East 47th Street, New York, j NY 10017. They are also available forinspection at the NRC Library, Two White Flint North, 11545 Rockville Pike, Rockville, Maryland 20852 2738.

l L

j (1) As used in this section, references to Section ill of the ASME Boller and Pressure -

Vessel Code refer to Section ill, Division 1, and include editions through the 1995 Edition and addenda through the 1996 Addenda, subject to the following limitations and modifications:

(i) Enaineerina ludaement. When a licensee relies on engineering judgment for activities N evaluet!ons of components or systems within the scope of 10 CFR 50.55a that are not directly addressed by the ASME Boiler and Pressure Vessel Code, the NRC must approve the activities ;

or evaluations pursuant to 10 CFR 50.55a(s)(3).

(ii) Section ill Materials. When applying the 1992 Edition of Section Ill, licensees shall apply the 1992 Edition with the 1992 Addenda of Section 11 of the ASMc Boller and Pressure Vessel Code. ,

L (111) Weld lea dimensions. When applying the 1989 Addenda through the 1996 Addenda l

of Section lit, licensees shall not apply paragraph NB 3683.4(c)(1), Footnote 11 to Figure NC-3673.2(b) 1, and Figure ND 3C73.2(b)-1, and shall continue to use the requirements in the l

1989 Edition for this paragraph and figures.

(iv) Seismic deslan. Licensees may use Articles NS-3200, NS-3600, NC-3600, and ND 3600 through the 1993 Addenda, subject to the limitation specified in (b)(1)(iii). Licensees shall not use the provisions in the 1994 Addenda through the 1996 Addenda for these Articles.

l i 62 w +w'+ s-*W' 'wv D e . , e pi e++ evrw-4-Wee'-qT**p'N9 Pt owwye

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I (v) Quality assurance When applying editions and addenda later than the 1989 Edition of Section Ill, the requirements of NQA 1,' Quality Assurance Requirements for Nuclear j Facilities,* 1G86 Edition through the 1992 Addenda are acceptable for use provided that both NQA 1 and the quality assurance provlsions specified in NCA-4000 are used in conjunction with

-the administrative, quality, and technical provicions contained in the edition and addenda of Section lli belr,g utilized.

(vi) Independence of inspection. Licensees shall not apply NCA-4134.10(a) of Section 111,1995 Edition with the 1996 Addenda, and shall use NCA 4134.10(a),1994 Addenda.

(2) An used in this section, references to Section XI of the ASME Boiler and Pressure Vessel Code refer to Section XI, Division 1, and include editions through the 1995 Edition and addenda through the 1996 Addenda, subject to the following limitations and modifications:

v (iv) Pressure-retalnina welds in ASME Code Class 2 oloina (apolies to Tables IWC-2520 or IWC-2520-1. CateaoEQ1.1. (A) Appropriate Code Class 2 pipe welds in Residual Heat Removal Systems, Emergency Core Cooling Systerns, and Containment Heat Reme al Systems, must be examined. When applying editions and addenda up to the 1983 Edition through the ,

Summer 1983 Addenda of Section XI of the ASME Code, the extent of examination for these systems must be determined by the requirements of paragraph IWC-1220, Tablo IWC-2520 -

Category C-F and C-G, and paragraph IWC-2411 in the 1974 Edition and Addenda through the Summer 1975 Addenda.

63 P

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(B) For a nuclear power plant whose application for a construction permit was docketed

- prior to July 1,1978, when applying editions e.nd addenda up to the 1983 Edition through the Summer 1983 Addenda of Section XI of the ASME Code, the extent of examination for Code ;

Class 2 pipe welds may be determined by the requirements of paragraph IWC 1220, Table t

IWC 2520 Category C F and C-G and paragraph IWC-2411 in the 1974 Edition and Addenda ihrough the Summer 1975 Addenda of Section XI of the ASME Code or other requirements the Commisalon may adopt.

P (vi) Effective edition and addenda of Subsection IWE and Subsection IWL Section XI.

l.icensees shall use either the 1992 Edition with the 1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection IWE and Subsection IWL as modified and supplemented by the requirements in 9 50.55a(b)(2)(ix) and $ 50.55a(b)(2)(x).

(vil) (Reserved)

(vlii)Section XI References to OM Pad 4. OM Part 6 andpM Part 10 (Table IWA 16001). When L 1. g Table IWA 1600-1, " Referenced Standards and Specifications"in the Section XI, Division 1,1987 Addenda,1988 Addenda, or 1989 Edition, the specified " Revision Date or Indicator" for ASME/ ANSI OM Part 4. ASME/ ANSI Part 6, and ASME/ ANSI Part 10 shall i

be the OMa 1988 Addenda to the OM 1987 Edition. These requirements have been incorporated into the 1990 Edition of the OM Code which is incorporated by reference in ,

paragraph (b)(3) of this section.

.4 64

.._........,_,,,,i._,.;,._._._,....._ ,._._.., , ..... ..,.. - , -..;.,....-..,_..._ ._ . . , . . , _ . , ,

1 (ix) Examination of concrete containments. Licensees applying Subsection lWL,1992 Edition with the 1992 Addenda, shall apply all of the modifications in this paragraph. Licensees choosing to apply the 1995 Edition with the 1996 Addenda shall apply subparagraphs (ix)(A), j (ix)(D)(3. ), and (lx)(E).

1 (xl) Enoineerino ludaement. When a licenses relles on engineering judgment for activities or evaluations of components or systems within the scope of 10 CFR 50.55a that are act directly addressed by the ASME Boiler and Pressure Vessel Code, the NRC must approve the activities or evaluations pursuant to 10 CFR 50.55a(a)(3).

(xii) Quality Assurance. When applying Section XI ed;tions and addenda later than the 1989 Edition, the requirements of NQA-1,

Quality Assurance Requirements for Nuclear Facilities " 1979 Addenda through the 1989 Edition are acceptable as permitted by IWA 1400 of Section XI, provided the licensee utilizes its 10 CFR Part 50, Appendix B, quality assurance program, in conjunction with Section XI requirements. Changes to licensee's quality assurance program shall be made in accorda%e with 10 CFR 50.54(a). In addition, where NOA-1 and Section XI do not address the commitments contained in the licensee's Appendix B quality assurance program description, such commitments shall be applied to Section XI activities.

(xiii) plass 1 pipino. Licensees shall not apply IWB-1220, " Components Exempt from Examination," of Section XI,1980 Aud6Jda through the 1996 Addenda, and shall apply 65 l

IWB 1220,1989 Edition.

(xlv) Clasp _2 pipina. Prior to applying the provisions of IWC 1220, " Components Exempt from Examination," IWC 1221, " Components Within RHR, ECC, and CHR Systems or Portions of Systems," and IWC-1222, " Components Within Systems or Portions of Systems Other Than RHR, ECC, and CHR Systems," 1989 Addenda through the 1996 Addenda, licensees shall define the Class 2 piping subject to volumetric and surface examination, and submit this Information for approval by the NRC staff pursuant to 6 50.55a(a)(3) prio, to implementation.

(xv) plass 1 pjpina volumetric examinatipl. When perforrning weld examinations of High Pressure Safety injection Systems, as required by Table IWB-25001, Examination Category B-J, item Numbers B9.20, B9.21, and B9.22, alllicensees of pressurized water reactor facilities shall perform volumetric examination of the Class 1 portion of the system after(insert 6 months from the date of the final rule).

(xvi) Eaws in Class 3 pipina moderate enprgyl200 'F. 215_psig)_pjping. Licensees may use the provisions of Code Case N-513," Evaluation Criteria for Temporary Acceptance of Flaws in Class 3 Piping," Rev 0, and Code Case N-523-1," Mechanical Clamping Devices for Class 2 and 3 Piping." Licensees choosing to apply Code Case N-5231 shal: apply all of its provisiens.

Licensees choosing to apply Code Case N 513 shall apply all of its provisions subject to the following:

(A) When implementing Code Case N-513, the specific safety facters in paragraph 4.

must be satisfied.

66

3 l

r (B) Code Case N-513 shall not be applied to:

(1) Components other than pipe and tube, such as pumps, valves, expansion joints, and heat exchangers; (2) The discovery and repair of flaws or deficiencies remaining from orijinal construction; (3) Leakage through a flange gasket; (i) Threaded connections employing nonstructural seal welds for leakage prevention (through seal weld leakage is not a structural flaw, thread Integrity must be maintained); and (E) Degraded socket welds.

(xvii) Appendix Vill personnel avalification. All personnel qualified for performing ultrasonic examinations in accordence with Appendix '/Ill shall receive 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months of annual training that includes laboratory work and examination of flawed specimens. .

(xviii) Appendix Vill specimen set cracks. All flaws in the specimen sets used for performance demonstration for piping, vessels, and nozzles shall be cracks.

'(xix) Aooendix Vill specimen set microstructure All specimens for single-side tests shall contain microstructures of the type found in components to be inspected, and flaws with non-optimum characteristics consistent witn field experience that provide realistic challenges to the UT techniques, 67

().x) Reconciliation of ggality Rean1rements. The following limitations apply when implementing Section XI, IWA-4200,1995 Addenda through the 1996 Addenda:

(A) Licensees shall not apply IWA 4200, of Section XI,1995 Addenda through the 1996 Addenda, for reconciliation of the administrative requirements for replacement items, and shall reconcile the administrative requirements with the original Construction Code and the O.vner's requirements .as required by the 1995 Edition.

(B) Lkensees shall not apply the definition of Construction Code in IWA 9000,

" Glossary,' 1993 Addenda through the 1996 Addenda, and shall apply the definition r;f Construction Code in IWA 9000,1992 Et tion.

(3) As used in this section, references to the OM Code refer to the ASME Code for Operation and Maintenance of Nuclear Power Plants, and include addenda through the 1996 Addenda and editions through the 1995 Edition subject to the following lim;tations and modifications:

(i) Quality Assurance. When applying editions and addenda of the OM Code,1990 and later, the requirements of NOA 1, " Quality Assurance Requirements for Nuclear Facilities," 1979 Addenda, are acceptable as permitted by ISTA 1,4 of the OM Code, provided the licensee i

utilizes its 10 CFR Part 50, Appendix B, quality assurance program, in conjunction with the OM Code requirements, Changes to licensee's quality assurance program shall be made in accordance with 10 CFR 50,54(a). In addition, where NQA-1 and the OM Code do not address the commitmonts contained in the it.ensea's Appendix B quality assurance program description, l

such commitments shall be applied to OM Code activ.tles, 68

(ii) Stroke time testina. Licensees shall comply with the provisions on stroke time testing in OM Code ISTC 4.2,1995 Edition with the 1996 Addenda, and the programs developed under their licensing commitments for demonstrating design basis capability of motor-operated valves.

(iii) Code Case OMN-1. As an altemative to S 50.55a(b)(3)(li), licensees may use Code Case OMN-1, "Altemative Rules for Preservice and inservice Testing of Cericin Electric Operated Valve Assemblies !n LWR Power Plants," Rev. O,1995 Edition with the 1996 Addenda, in conjunction with ISTC 4.3,1995 Edition with the 1996 Addenda. Licensees choosing to apply the Code case shall apply all of its provisions.

(A) The adequacy of the test interval for each valve shall be evaluated and adjusted as necessary but not later than five years or three refueling outages (whichever is longer) from initial implementation of ASME Code Case OMN-1.

(iv) Appendix 11. The following modifications apply when implementing Appendix II,

" Check Veive Condition Monitoring Program," o' the OM Code,1995 Edition with the 1996 Addenda:

(A) Valve opening and closing functions must be demonstrated when flow testing or examination methods (nonintrusive, or disassembly and inspection) are used; (B) The niitiat interval for tests and associated examinations shall not exceed two fuel cycles or 3 years, whichever is longer, any extension of this interval shall not exceed one fuel cycle por extension with the maximum interval not to exceed 10 years; trending and evaluation of existing data shall be used to reduce or extend time the interval between tests.

69

4

'(C) If the Appen' dix 11 condition monitoring program is discontinued, then the requirements of ISTC 4.5.1 through 4.5.4 shall be implemented.

(v) Subsection ISTD. Licensees may use Subsection ISTD, OM Code,1995 Edition with the 1996 Addendt., by making a change to their technical specifications in accordance with -

applicable NRC requirements. Licensees choosing to apply the subsection shall apply all of its provisions.

(c)

(3) The Code Edition, Addenda, and optional Code Cases' to be applied to components of the reactor coolant pressure boundary must be determined by the provisions of paragraph

'NCA-1140, Subsect:on NCA of Section lit of the ASME Boiler and Pressure Vessel Code, but:

(1) the edition and addenda applied to a component must be those which are incorporated by reference in paragraph (b)(1) of this section, and, in case of conflict between (b)(1) and paragraph NCA-1140, the !stest edition and addenda incorporated by reference in (b)(1) shall be applied, (ii) the ASME Code provisions applied to the pressure vessel may be dated no earlier than the Summer 1972 Addanta of the 1971 edition, (iii) the ASME Code provisions applied to piping, pumps, and valves may be dated no earlier than the Winter 1972 Addenda of the 1971 edition, and (iv) ASME Code Cases' must have been determined suitable for use by the NRC.

~* * *

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. = . .

(2) The Code Edition, Addenda, and optional Code Cases' to be applied to the systems and components identified in paragraph (d)(1) of this section must be determined by the rules of paragraph NCA-1140, Subsection NCA of Section til of the ASME Boiler Vessel and Pressure Code, but (i) the edition and addenda must be those which are incorporated by reference in paragraph (b)(1) of this section, and, in case of conflict between (b)(1) and Paragraph NCA 1140, the latest edition and addenda incorporated by reference in (b)(1) shall be applied, (ii) the ASME Code provisions applied to the systems and components may be dated no earli- '980 Edition, and (iii) the ASME Code Cases' must have been determined suitable fo

Pr tiu NRC.

(e)

(2) The Code Edition. Addenda, and optional Code Cases' to be applied to the systems and components identified in parar ,ph (e)(1) of this section must be determined by the rules of paragraph NCA 1140, Subsection NCA of Section !!! of the ASME Boiler and Pressure Vessel Code, but: (i) the edition and addenda must be those whici are incorporated by reference in paragraph (b)(1) of this section, and, in case of conflict between (b)(1) and paragraph NCA 1140, the latest edition and adder

A incorporated by reference in (b)(1) shall be applied, (ii) the ASME Code provisions applied to the systems and components may be dated no earlier than the 1980 Edition, and (iii) the ASME Code Cases' must have been determined suitable for use oy the NRC.

(f) inservice testino reouirements. Requirements for inservice inspection of Class 1, Class 2, Class 3, Class MC, and Class CC components (including their supports) are located in

$ 50.55a(g).

(1) For a boiling or pressurized water-cooled nuclear power facility whose construction 71

I 1

l permit was' issued prior to January 1,1971, pumps and valves must meet the test requirementi (f paragraphs (f)(4) and (f)(5) of this section to the extent practical. Pumps and valves which are part of the reactor coc' ant pressure boundary must meet the requirements applicable to components wh!ch are classified as ASME Code Class 1. Other pumps and valves in steam, watcr, sir, and liquid-radioactive-waste systems that perform a function to shut down the reactor or maintain the reactor in a safe shutdown condition, mitigate the consequences of an accident, or provide overpressure protection for such systems (in meeting the requirements of the 1988 Edition, or later, of the Boller and Pressure Vessel or OM Code), must meet the test requirements applicable to compnents which are classified as ASME Code Class 2 or Class 3, (2) For a bolling or prescurized water-cooled nuclear power facility whose construction permit was issued on or after January 1,1971, but before July 1,1974, pumps and valves which are classified es ASME Code Class 1 and Cless 2 must be designed and be provided with access to enable the performance of inservice tests for operational readiness set forth in editions 8

of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda in effect 6 months prior to the date of issuance of the construction permit. The pumps and valves may meet the requirements set forth in subsequent editions of this code and addenda which are incorporated by reference in paragraph (b) of this section, subject to limitations and modifications listed therein.

(3)

(iii)(A) Pumps and valves, in facilities whose construction permit was issued before finsert effective date of the final rule), which are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in Section XI of editions of tne ASME Boller and Pressure Vessel C7de and Addenda8 applied to the construction of the 72

particular pump or va've or the Summer 1973 Addenda, whichever is later, i

(B) Pumps and valves, in facilities whose construction permit is issued on or after (insert r,'fective date of the final rule), which are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice iesting of the pumps and valves for assessing operational readiness set forth in editions and addenda of the ASME OM Code referenced in paragraph (b)(3) at the time the construction permit is issued.

(iv)(A) Pumps and valves, in facilities whose construction permit was issued before (insert effective date of rule), which are classified as ASME Code Class 2 and Class 3 must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in Section XI of editions of the ASME Boiler rad Pressure Vessel Code and Addenda' applied to the construction of the particular pump or valve or the Summer 1973 Addenda, whichever is later.

(B) Pumps and va}ves, in facilities whose construction permit is issued on or after finsert effective date of the finalf_ule), which are classified as ASME Code Ciass 2 and 3 must be designed and be provided with access to enable the performance of inservice testing of the pumps and valves for assessing operational readiness set forth in editions and addenda of the ASME OM Code referenced in paragraph (b)(3) at the time the construction permit is issued.

(4) Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves which are classified as ASME Code Class 1, Class 2 and Class 3 must meet the ine nvice test requirements, except design and access provisions, set forth in the

' ASME OM Code and addenda that become effective subsequent to editions and addenda 73

specified in paragraphs (f)(2) and (f)(3) of this section and that are incorporated by reference in paragraph (f) of this section, to the extent practical within the limitations of design, geometry and materials of construction of the components.

(g)

(1) For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1,1971, components (including supports) must meet the

- requirements of paragraphs (g)(4) and (g)(5) of this section to the extent practical. Components which cre part of the reactor coolant pressure bour.Jary and their supports must meet the requirements applicable to components which are classified as ASME Code Class 1. Other pressure vessels, piping, pumps and valves, and their supports in stemn, water, air, and liquid-radioactive-waste systems t.1st provide pressure boundary integrity for systems that perform a function to shut down the reactor or rnaintain the reactor in a safe shutdown condition, or mitigate the consequences of an accident, must meet the requirements applicable to components which are classified as ASME Code Class 2 or Class 3.

(3)

_ (i) Components (including supports) which are classified es ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice examination of such components and must meet the preservice examination requirements set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Addenda' applied to the 74

r , -

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~oonstruction of the_ particular component.~ ,

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(iii)-(v) [ Reserved)-

. (4) Throughout the service IMe of a boiling or pressurized water-cooled nuclear power - .

1 facility, components (including supports) which are [ classified as ASME Code Class 1, Class 2

and Class 3 must meet the requirements, except design and access provisions and preservice i examination requirements, set forth in Section XI of editions of the ASME Boiler and Pressure Vessel Code and Add 3nda that become effective subsequent to editior.s specified in paragraphs-- ,

(g)(2) and (g)(3) of this section and that are incorporated by reference in paragraph (b) of this t

. section, to the extent practical within the limitations of design, geometry and materials of construction of the components. Components which are classified as Class MC pressure

+

retaining components and their integral attachments, and components wHch are classified as c

Class CC pressure retalning components and their integral attachments must meet the requirements, except design and access provisions and preservice examination requirements,

-set forth in Section XI of the ASME Boiler and Pressere Vessel Code and Addenda that are incorporated by reference in paragraph (b), subject to the limitation listed in paragraph (b)(2)(vi) and the modifications listed paragraph (b)(2)(ix) and (b)(2)(x) of this section, to the extent .

practical within the limitation of design, geometry and materials of construction of the t

components.

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- (A)(1) All previously granted relieb under $ 50,55a to licensees for the extent of - ,

ic volumetric exemination of reactor vessel shell welds specified in item B1.10 of Examination

- Category B-A, " Pressure Retalning Welds in Reactor Vessel,"in Table IWB-25001 of Subsection 11WB in applicable edition and addenda of Section XI, Division 1, of the ASME Boiler and

% +

Pressure Vessel Code, during 'the inservice inspection interval in effect on September 8,19U2 are hereby revoked, subject to the specific modification in $ 50.55a(g)(6)(ii)(A)(3)(iv) for licensees that defer the augmented examination in accordance with $ _50.55a(g)(6)(ii)(A)(3).,

(2) All licensees shall augment their reactor vessel examination by imp'ementing once, as part of the inservice inspection interval in effect on September 8,1992, the examination requirements for reactor vessel shell welds specified in item B1.10 of Examination Category B-A,

" Pressure Retaining Welds in Reactor Vessel,"in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code, subject to ' ,

g $the conditions specified in 9 50.55a(g)(6)(li)(A)(3) and (4) The augmented examination, when not -

deferred in accordance with the provisions of 9 50.55a(g)(6)(ii)(A)(3), shall be performed in -- )

i4 accordance with the related procedures specified in the Section'XI edition and addenda .

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applicable to the inservice inspection !nterval in effect on September 8,1992, and may be used i W

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as a substitute for the reactor vessel shell weld examination scheduled for implementation during the inservice inspection intervalin effect on September 8,1992. For the purpose of this augmented examination, " essentially 100W as used in Table IWB-2500-1 means more than 90 percent of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry, (g . . .

(y) Licensees with fewer than 40 months remaining in the inservice inspection intervalin effect on September 8,1992 may extend that interval in accordance with the provisions of Section XI (1980 Edition) IWA 2430(d) for the purpose of implementing the augmented examination during that interval.

(yi) The deferred augmented examination shall be performed in accordance with the related procedures specified in the Section XI edition and addenda applicable to the inspectir.n interval in which the augmented examination is performed.

(6) Augmented examinations of reactor vessel shell welds that are performed in accordance with S 50.55a(g)(6)(ii)(A) after[nsert 6 months from the date of the final rule} must be performed in accordance with S 50.55a(g)(6)(ii)(C).

77

(C) Apolication of Appendix Vill to Section XI Examinations. (1) All reactor vessel-(including nozzles) ultrasonic exan,inations, all piping ultrasonic examinations, and ali botting 6itrasonic examinations performed after (insert 6 months from the date of the final rulel must be performed in accordance with Appendix Vill of Section XI, Division 1,1995 Edition with the 1996 Addonda of the ASME Boller end Pressure Vessel Code.

(h) 8 For ASME Code Editions and Addenda issued prior to the Winter 1977 Addenda, the Code Edition and Addenda applicable to the component is govemed by the order or contract date for the component, not the contract date for the nuclear energi system. For the Winter 1977 addenda and subsequent editions and addenda the method for determining the appliccble Code ,

editions and addenda is contained in Paragraph NCA-1140 of Section lli of the ASME Code.

7 i For purposes of this regulation the proposed IEEE 279 became "in effect" on August 30, 1968, and the revised issue IEEE-279-1971 became "in effect" on June 3,1971. Copies may be obtained from the Institute of Electrical and Electronics Engineers, United Engineering Center, 345 East 47th St., New York, NY 10017. Cop!es are available for inspection at the NRC Library, l Two White Flint North,11545 Rockville Pike, Rockville, Maryland 20852 2738.

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Dated at this- _ day of 1997._ ,

For the Nucle,yr Regulatory Commission.

f2

_ L Joseph Callan, >

Executive Director for Operations. .

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obtained from the Institute of Electrical and Electronics Engineers, United Engineering Center, 345 East 47th St., New York, NY 10017. Copies are available for inspection at the NRC Library, Two White Flint North,11545 Rockville Pike, Rockville, Maryland 20852-2738.

Dated at this day of 1997.

For the Nuclear Regulatory Commission.

L. Joseph Callan, Executive Director for Operations.

RECORD NOTE: A copy of this proposed rule was sent to IG for information en: July 7,1997.

Distribution:

SGEB r/f GSIB r/f Central File BMMorris Doc Name: a:\prorule5. wad OFC SGEB:DET GS*d:DET GSIB:DET 0:DET NAME WNonts FChemy FCoffman LShao DATE 10114197 10/14 197 10/14/97 10/14!97 OFC NRR' OE' CFO* ClO*

NAME $ Collins Jlieberman LJFunctes AJGalante by egnail TE 07I30197 07/14 197 07/17197 07/24/97 OFC tRM' OGC' ADM' AEoD*

NAME BShelton WJolmstead ELHalman DFRoss by e. mall DATE 07/24197 10/14/97 07/29/97 07/30197 OFC Rti!* D:RES EDO NAME ABBeach MRKnapp JLCallan DATE o7/30/97 10/16/97 / 19 7 OFFICE RECORD COPY RES FILE CODE:

80

_____________________________y

as a substitute for the reactor vsssel shell weld examination scheduled for implementation dudng the inservice inspection intervalin effect on September 8,1992. For the purpose of this augmented examination, " essentially 100W as used in Table IWB-2500-1 means more than 90 percent of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry.

g . . .

(v) Licensees with fewer than 40 months remaining in the inservice inspection intervalin effect on September 8,1992 may extend that interval in accordance with the provisions of Section XI (1989 Edition) IWA-2430(d) for the purpose of implementing the augmented examination during that interval.

(yj) The deferred augmented examination shall be performed in accordance with the related procedures specified in the Section XI edition and addenda applicable to the inspection interval in which the augmented examination is performed.

(s) Augmented examinations of reactor vessel shell welds that are performed in accordance with $ 50.55a(w(6)(ii);A) after { insert 6 months from the date of the final rule} must be performed in accordance with 6 LO.55a(g)(6)(ii)(C).

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7 (C). Apolication of App 9ndix Vhi to St.ction XI Examinations. (f) All reactor vessel (including nozzles) ultrasonic examinations, all piping ultrasonic examinations, and all botting ultrasonic examinations performed after [Lnsert 6 months from the date of the final ru!al must be .

performed in accordance with Appendix Vill of Section XI, Division 1,1995 Edition with the 1996 -

Addenda of the AGME Boller and Pressure Vessel Code.

(h) 8 For ASME Code Editions and Addenda issued prior to the Winter 1977 Addenda, the l . Code Edition and Addenda applicable to the component is govemed by the order or contract date for the component, not the contract date for the nuclear energy system. For the Winter 1977 addenda and subsequent editions and addenda the method for determining the applicable Code editions and addenda is contained in Paragraph NCA-1140 of Section ill of the ASME Code.

7

For purposes of this regulation the proposed IEEE 279 became "in effect" on August 30, 1968, and the revised issue IEEE 279-1971 became "in effect" on June 3,1971. Cooies may be obtained from the Institute of Electrical and Electronics Engineers, United Engineering Center, 345 East 47th St., New York, NY 10017. Copies are available for inspection at the NRC Library, 1

Two White Flint North,11545 Rockville Pike, Rockville, Maryland 20852-2738.

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- Dated at _ this day of 1997.

For the Nuclear Regulatory Commisslen, i

i

~

L Joseph Callan, Executive Director for Operations.

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ML20199E986

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ML20199E986

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