ML20215E777

Safety Evaluation Re Inservice Insp Program & Requests for Relief from Certain Inservice Insp Requirements

ML20215E777
Person / Time
Site:	Grand Gulf
Issue date:	10/06/1986
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20215E767	List: ML20215E763 ML20215E777
References
TAC-61929, TAC-62047, NUDOCS 8610150506
Download: ML20215E777 (6)

Text

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pun SEP 2 01966 d q{o, E\ . 'i UNITED STATES

NUCLEAR REGULATORY COMMISSION Enclosure j j WASHINoTON. D. C. 20555

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_ SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACT

_RELATED TO THE INSERVICE INSPECTION PROGRAM AND

_FROM CERTAIN INSERVICE INSPECTION REQUIREMENTS MISSISSIPPI POWER AND LIGHT COMPANY GRAND GULF NUCLEAR STATION UNIT 1 DOCKET NO. 50-416 A. INTRODUCTION The Technical Specification for the Grand Gulf Nuclear Station Unit 1 (GGNS-1) 3 components shall be performed in accordance with Sect Boiler CFR and Pressure Vessel Code and applicable Addenda as recuired b 50.55a(

Commission.g) except where specific written relief has been granted by the this Code Section, consequently certain requirements of n addenda of Section XI are impractical to perform because of the plants' design, component geometry, and material of construction Regulation 10 CFR 50.55a(g)(6)(i) authorizes the Commission to grant relief from those requirements upon making the necessary findings.

The Nuclear Regulatory Commission had by letter dated July

, 1986 22 a provided Safety Evaluation Report (SER) related to the first ten year Inservice Inspection requirements. Plan and requests for relief from certain inservice inspection plan for implementation.The Commission at that time accepted the inservice inspe j Code requirements were impractical to perform and pursuant to 1 50.55a(g)(6)(i) that the granting of the requested relief is authorized and would not endanger life or property, and is otherwise in the public interest.

During the period that this SER was b~eing prepared Mississippi Po Company (MP&L), submitted by letters dated July 14, 23, and August 15

, 1986 determined to be impractical to perform at the Grand G Unit 1.

B. REQUESTS FOR RELIEF B .1. 0 Welds

_B-G-2, and Flange Bolting, Category B-0, Item 61 Item B7 1 ff)ko g g %

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B .1.1 Code Requirement Welds in Control Rod Drive Housings, Item B14.10.

The welds in 10% of the peripheral control rod drive housings shall be surface or volumetrically examined in accordance with Figure IWB-2500-18 during each inspection interval. -

formed at or near the end of the inspection interval.The examinations may be per-Bolts, Studs, and Nuts in Reactor Vessel, Item B7.10. The surf aces of all bolts, studs, and nuts 2 in. or less in diameter in the the reactor first inspection vessel interval.

shall be visually examined (VT-1) during Bolting may be examined (a) in place under the or (c) when tension, (b)iswhen bolting the connection is disassembled, removed.

B .1. 2. Code Relief Request Relief is requested to exempt from inservice inspection, the peripheral CRD housing welds (tube-to-tube, tube-to-flange), the eight (8) bolts associated with each flange of 193 CRD housings, and the four (4) bolts associated with each flange of 58 incore housings .

B.I.3. _ Proposed Alternative Examination The welds and bolting will be subject to a system leakage test after each refueling and a hydrostatic test once each ten year in-l service inspection interval per the requirements of IWB-5221 and IWB-5222 of ASME Section XI.

8.1.4. Licensee's Basis for Requesting Relief The weld areas and bolting are not accessible for inspection unless the control rod drive support structure is removed. A 360 surface examination required by the Code cannot be accurately accomplished housings. from the outside due to interference from adjacent CRD would recuire that the control rod drive mechanism be remove could result in damage to the drive.

With removal of the drive, a small amount of reactor water would escape to the CRD cavity area, possibly causing contamination of personnel and equipment. The time frame associated with the CRD support structure removal and CRD mechanism would be approximately six (6) man-hours per drive.

received by personnel in this interval cannot justify the inspectionDosage process to possibly find a fault which could be discovered by Manual (Technical Specification) limits in effect. excessive leakage in t Request for exemption for the following reasons: from inservice inspection should be granted i

1.

The peripheral CRD housing welds have been examined by radio- l graphy and liquid penetrant methods and have been hydrostatic {

tested in accordance with ASME Section III Code requirements.

D

3 2.

All incore and CRD housing bolting has been examined in accordance with the requirements of ASME Section III, which exceed the Section XI (VT-1) visual examination requirements. ~

3.

The welds and bolting will be subject to a system leakage test (IWB-5221) each refueling outage and hydrostatic test (IWB-5222) once each ten-year inservice inspection interval per the require-ments of ASME Section XI.

4.

If the welds and/or bolts fail while in operation, the maximum leakage rate, by calculation, housing tube-to-flange weld. will occur at the peripheral CRD is 681 gpm. The maximum calculated leak rate

" exemptions by makeuBy criteria established in Subarticle IWB-1200, for GGNS is 878 gpm,p capacity " the normal makeup capability which exceeds the calculated maximum leakage.

5.

Leak detection is provided with the leakage detection system, with continuous monitoring in the control room 6..

The CRD housing supports would prevent ejection of the housings in case of total failure of the welds or bolts.

7.

Removal of the control rod drive support structure would result in hardships quality with no compensatory increase in the level of and safety.

B.1.5 Evaluation Relief was requested and granted to eliminate examination of the control roc drive the preservice and in-core housing weld and flange bolting during examination.

The primary basis for establishing the structural integrity of these components is, therefore, the construc-tion code examinations and system leakage and hydrostatic tests ASME Code Section III was the construction code and, as a result, the .

peripheral penetrant CR0 housing welds were examined by radiography and liquid methods.

Section III was also completed. Examination of the bolting in accordance with s

The licensee has shown that failure of a CRD housing and/or bolt would result inofathe capacity leak rate from makeup the system which would be less than the system. i In addition, CRD housing supports wouldorprevent welds ejection bolts. During of the housings in case of total failure of the operation continuously earlier stage. in the control room s,o leaks can be detected at anleak d Further examination of the CRD housings or bolts in accordance with Removal of the CRD support strthe Code would require retroval of the .

personnel radiation exposure. ucture and driver would result in some l l

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4 Relief inations is warranted would be impractical. on the basis that conducting the pro posed exam-was established through the construction examinationsThe preservice reliability during ISI.and leakage tests canHydrostatic be used to continually the bolting capability of thewould system.not result in a leakage erate makeup above th'In any eve If it becomes necessary to remove the CRD support re in structu conduct surface examination of the accessible welds and visual examination of the bolting.

o CR0 mechanism were removed for other purposes, Correspondingly, if a of the housing welds should be conducted. volumetric examination This evaluation 1-00008 that was grantedisbyidenticalthe Commission to the July evaluation 22 request No. of relief B.1.6.

the licensee's basis for the request has been e. modifi d, 1986 except that Conclusion Based on the above evaluation '

and bolting discussed above, a,dherence tos the is impractical. Code requirem It is further concluded that the proposed sonable assurance of structural erval. reliability rea- during This evaluating conclusion letter of July 22, Relief is identical to the conclusion 1986,Request I-00008 and statedonin the Commissi n reached relief. except for the licensee's basis f or the The staff is valid and does not changeagrees that thetheprevious modifiedconclusio basis for the request relief is granted as requested provided: ns. Therefore, (a) are performed as required by the Code andthe Cod (b) the Code required examinations are performed on the accessible areas if the CR0 support structure or CRDs are B.2 removed for other maintenance activities.

vessels, Category B-0, Items B3.90 of Nozzles and in B3.100 Relief R B.2.1 Code Reauirement  ;

ASME Section XI Code requires vessel nozzle examination t consist of a full thickness examination of theo attaching weld radius, and IWB-2500-7, 1977a portion of the~ nozzle bore in accordance , inner g.

with Fi edition through 1979 Summer Addenda B.2.2 .

Code Relief Recuest Relief is requested from performing IWB-2500-7, 1977 xar-g.

the Code requ edition through 1979 Sumeer Addenda .

5 B.2.3 Proposed Alternate Examination Tre above cited 35 nozzle to vessel e examined welds in will b Summer 1983 Addenda, plus allEdition, and the referenced documents.

related requ ncluding IWB 3512 B.2.4 ~

Licensee's Basis for Requesting Relief to be examined during the ten s thatyear inspectioTh are required n interval. These techniques with a combination ofngproceduresexam manual preservice, automated systems for RPV examinatioSince the perform General on Electric 9 nozzles. and are currently scheduled to bns have been e used during RF01 reactor inspections helps to minimize for better data collection and analysis man sure and provide rem ex ems for tion of volume coverage exists.andnations, equipment a limita-available Due to nozzle geometr with the examination,y, thicknesses and multiple angles associated manual scans are required to supplement the automated examination in order to obtain e. Itfull coverag anticipated that the supplemental examinations require is would average of calibration and6 examination.

man-hours per nozzle (2 men x an 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />); thi s includes The examinations would require working from a ski the annulus between the RPV and biological shi ldp box lowere No.

1. ring and No. 4 ring regions and frome a for 2 region. ladde nozzles in the the changing of trIn addition r for nozzles in the No.

each inner radius examination requires terclockwise scan.ansducer we,dges to facilitate a clockwis approximately personnel. The 54 9 nozzles man-hours scheduled in examination timefor notRF01 i will a recent outage at the Susquehannaractor Nuclear in Pow I') examiners may be required to complete ant, afull vol of minimum ume examinations.

forming a full thichess examination iculties of byper-redLa er e ination volume. The newly defined examinati ucing the required exam-associated with examining ae full nozzle thicknes difficulties volumetric examinations to the volume specified .

Performance b of the 1983 the need for s Edition, Summer 1983 Addends y of IWB ASME 2500-7(b) S of manual scans. upplementing the automated ultrasonic examina The newly defined examination area includes thos area of examination that ise areas n omittedthat by ation. The by the 19 Addenda of ASME Section XI did not have any on, Summer 1983 identified during th+ preservice examinationrecordable indications l

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B.2.5. Evaluation Updating to the requirements of later approved editions and addenda '

is permitted by 10 CFR 50.55a(g)(4)(iv). Perfonning the examinations to the requirements of the earlier editions of the code would necessitate supplemental manual scans to obtain the required coverage. The manual scans would increase the radiation dosage accumulated by the inspectors without a compensating increase in the confidence of structural integrity of the examined joint. The later edition of the code rightfully recognized the extent of exposure without the compensating benefit and reoriented the required exam-ination volume to facilitate complete conformance by automated systems and to include those areas that by industry experience have the potential for crack initiation.

B.2.6. Conclusion Based on the above evaluation, it is concluded that for the welds discussed above, examinatisn to the requirements of the presently authorized Code is impractical due to the excessive radiation tc the inspectors without compensating benefit. It is further conc-cluded that the proposed alternate examination will provide reason-able assurance of structural reliability during this interval.

C.

SUMMARY

CONCLUSI0f, Based on its review of the inservice inspection relief requests, the staff concludes that relief granted from the Code examination and testing requirements and the proposed alternate methods of examination and testing give reasonable assurance of the piping and component pressure boundary and support structural integrity. The staff has dt:termined that the Code requirements are impractical and, pursuant to 10 CFR 50.55a(c)(6)(i), the cranting of the requested relief is authorized by law and will not endanger life or property, or the cormion defense and security, and is otherwise in the public interest considering the burden that could result if the require-ments were imposed on the facility.

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