ML20004F216

Safety Evaluation Supporting Amend 6 to License DPR-22

ML20004F216
Person / Time
Site:	Monticello
Issue date:	06/03/1981
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20004F212	List: ML20004F211 ML20004F216 ML20004F217
References
NUDOCS 8106160624
Download: ML20004F216 (15)

Text

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!4UCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555

@..d / SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 6 TO LICENSE N0. DPR-22 NORTHERN STATES POWER COMPANY DOCKET NO. 50-263 MONTICELLO NUCLEAR GENERATING PLANT The Commissien has reviewed and evaluated the Inservice Inspection Program (excluding testing of pumps and valves) for the Monticello Nuclear Generating Plant, and finds it in compliance with Paragraph (g) of 10 CFR 50.55a,

" Inservice Inspection Requirements." The Inservice Inspection Program was submitted by Northern States Power Company (licensee) en March 15, 1978, and revised August 28, 1978, January 5,1979, February 26, 1979, July 27, 1979, March 5,1980 and July 16, 1980. Additional information was supplied in letters dated January 20, 1981, and February 20, 1981.

Pursuant to 10 CFR Part 50.55a(g)(6)(1), relief may be granted from specific requirements stated in the ASME Boiler and Pressure Vessel Code,Section XI,.

1974 Edition including Addenda through Summer 1975, which we have concluded to be impractical for the facility because of component or system design, geometry, or materials of construction.

In some cases, relief may be granted only after performing the alternative inspection requirements which the staff deems necessary.

Such relief and alternative requirements are authori:ed by law and will not endanger life or property or the common defense and security and are othe mise in the public interest giving due consideration to the burden upon the licenzee that could result if the requirements were imposed on the facility.

In some cases, relief should not be granted because of the factors stated in the evaluation of the specific request.

I.

Relief Recuest Evaluations A.

Reactor Vessel Welds in Core Region (Examination Category 3-A, Item 31.1); in Shell & Sottom Head (Examination Category 3-3, Item 21.2); Standby Liquid Control No::le to Vessel Welds (Examination Category 2-0, Item 31.4); Integrally Welded Vessel Stabilizer Lugs (Examination Category S-H, Item 31.12); and Stan'dby.'

Liquid Control No::le to Safa End Velds (Examination Category 3-F, Item Bl.6); (Relief Requests 15,16, and 19).

Code Recuirement For core, shell, and bottom head welds, volumetric examination of 10% of each longitudinal weld and 5% of each circusferential weld.

For the no::le to vessel welds, volumetric examination of weld, base metal and inside radius.

For the vessel stabili:er lugs, volumetric examination of 100% of the welding to the vessel.

For the no::le to safe end welds, volumetric and surface examination of the circumfeence of 100% of the welds.

810 6.16 0 Q L

P00R ORIGINAL.

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Code Deviation Recuest Delete volumetric examination from the following welds:

Examination Catecorv Weld Identification 8-A VLAA-1,2; VLSA-1,2; VCBA-2 B-B VLSA-1,2; VCBS-3; HMAB-9,10 B-D CPAD-1 B-F CPAF-2 B-H 4 lugs

. Reason for Recuest The design of the reactor internals and the external, biological shield and vessel insulation prevents access to these welds. The Monticello RPV was constructed with a 2'3 " thick biological shield wall surrounding it, with the exception of the top eight or so feet.

Between this wall and the reactor vessel shell is a space of approxi-mately 1 fo'ot that houses the thermal insulation.

The only access areas to the reactor vessel is:

1. at the. top eight feet above the biological shield wall, 2. through openings in the wall at dach nozzle location and two inspection ports below the skirt weld, 3. by the control red drives under the rea.ctor head, and 4. from the vessel inside diameter.

There are no no::les in the core region for access to Category B-A welds.

The area above the biological shield wall and at the no::le opening is further obstructed by non-removable insulation. A good portion of the vessel insulation was not designed to be removed and there--

fore it was installed prior to the installation of the piping, electrical conduits, vessel stabilizers, duct,, work, etc.

A very thorough review was performed, using drawings, sketches, and previous examination reports, to try and locate weld areas that possibly could be inspected.

It was concluded that some of the vessel welds appear to be close enough to no::1.e openings for perform-ing the examinations provided the insulation can be removed.

An attempt will be made to remove or modify the insulation and to examine some of these welds during the 1981 refueling outage.

In addition, during the upcoming outage we are hopeful that radiation levels will permit access to the bottum of the vessel (between the control red drives, skirt, and vessel). This would allow accessi-bility to a le*ger portion of the skirt to vessel weld and to the bottom head aeiidional and dollar plate welds.

Each of the welds that can be examined will be sketched to show the examination amount, extent and' location.

In addition, each area l

will be scheduled for examination during the next ten year interval.

l The design of the Monticello reactor vessel internals does not allow i

internal access

• .o the vessel welds either in or below the core regica, 5"It does give internal access to the top 20 ft. of the reactor vessel.

This access allows one ci;crmferential and two l

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longitudinal welds to be examined that were thought to be inaccessible from the outside of the vessel.

The Standby Liquid Control nozzle and safe end protrudes through a hole in the vessel thern.31 insulation with a minimum of clearance.

It appears, from an early sketch, that the bottom of the weld is next to and just cutside the insulation.

Tha insulation around this no :le was not designed with removable panels. With the present design the nozzle to safe end weld is 10C% accessible for ' ultrasonic examination from one side, and appreximately 50-75% accessible for a liquid penetrant examination.

The volumetric and surface examinations required on this weld will be performed to the extent possible during the 1981 refueling out-age.

During this outage an attempt will be made to modify the insulation to allow 10C% coverage of the weld for both examinaticn methods.

In addition, a sketch will be made that will identify any areas that ramain inaccessible.

ItshouldbenotedthattheMonticelloreactorvesselwasfabrickted and subject to as-built inspection under very demanding specifica-tions.

Because the site was inaccessible to a river barge of the capacity necessary to transport a fully assembled vessel, the vessel was assembled at the site frca shop-fabricated subassemblies. All requirements of Section III of the ASME Soiler and Pressure Vessel Code, 1965 Edition, including Addenda through Summer 1955, were satisified just as if the vessel were shop fabricated.

Additional requirements more stringent than those required by the Code were applied by General Electric due to the unique circumstances surround-ing the vessel fabrication.

Refer to Volume VII of the Monticello Final Safety Analysis Report, " Pressure Vessel Design Report" for details concerning vessel fabrication and inspection.

In addition, it should be noted that based on analysis of the dosimeter removed thereactorvesselwasestimatedtobeonly1.23x10'QvelatT/4of from the reactor vessel, the =aximum neutron fluence nvt at the end of c? signed life (40 years).

Based on the stringent construct-ion requirements, the inability to examine these welds during service is not considered to significantly reduce the vessel integrity.

Staff Evaluation The design of the biological shield, vessel insulation, and reactor internals prevents access from all of the code required examinations of the reactor vessel welds.

Imposition of the exact code require-ments would subject the licensee to extreme hardships in necessitating the removal of portions of the concrete biological shield and the permanently installed in ulation to perform the required examinations of the welds frem the vessel's outside surface.

The code requires that IC% of each longitucinal weld and 5% of each circumferential weld be examined during the inspection interval.

To compensate for those welds which are inaccessible for volumetric examination, the licensee has agreed to examine all accessible areas 3

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P00R ORIGINAL of circumferential or longitudinal welds.

Table I is a summary of each Category B-A and S-S weld, the length of each weldi and the length believed to be accessible for examination. The licensee has indicated that in excess of 150 feet of longitudinal and circumfer-ential welds may be examined by the end of the Interval.

The ASME code requires that only 40 feet be examined during the same time period.

To help ensure the integrity of the vessel, the vessel materials are presently evaluated for radiation damage by use of the requirements in Regu'atery Guide 1.99, " Effects of Residual Elements on Predicted Radiation Canage to Reactor Vessel Materials." To ensure that stresses in the reactor vessel remain within acceptable limits, the Monticello Nuclear Generating Plant is subject to pressure / tempera-ture limitations for reactor coolant system heatup and cooldown operations, and inservice leak and hydrostatic tests in the plant Technical Specifications.

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The licensee has indicated that Category S-H welds may be accessible for examination pending the results of the next refueling outage.

We will review and evaluate this request for relief after the examinatiens have been attempted and the extent and methods of examinatien possible are kncwn.

We conclude that the augmented examinations fer certain welds, the ongeing surveillance program of the reacter vessel materials in the beltline regien, and the ves'sel design are adequate for providing an acceptable level of safety and assurance that the vessel's structural integrity will not be ccepremised during the inspection period by granting relief frem the examinations of 6 elds discussed above.

S.

Control Rod Drive and 3cttcm Head Dr.:in Vessel Penetrations (Examinatien Category S-E, Item 91.5).

Relief Request 17.

Code Recuirement The area surrounding each penetration shall be visually examined for evidence of leakage during the pressure test.

Code Deviation Recuest Visually inspect areas below penetrations during pressure test.

Reason for Recuest The design of the vessel, the biological shield, and vessel insulation I

prevents access to these areas that are directly adjacent to the vessel penetrations.

Sta,'? Evaluation l

The relief request evaluation in I.a. cescribes in detail the design of the vessel and insulation which makes these examinations 4

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TABLE I MONTICELLO REACTOR VESSEL WELD SUMM. iY EXAM WELD P.EQ'D EXAM ACCESSISLE*

CATEGORY NUMBER LOCATION LENGTH LENGTH LENGTH REMARKS Bl.1 B-A VLAA-1 TOP 27" 11' 1.1' 6"

ATTEMPT 2ND INTERVAL LONG VLAA-2 TOP 27" 11' 1.1' 6"

ATTEMPT '81 OUTAGE VL3A-1 SOT 117" 11' 1.1 NONE VL3A-2 207 117" 11' 1.1' NONE CIRC VCSA-2 57' 2.85' NONE 21.2 3-3 VLAA-1 30T 8'8" 11' 1.1' 4'

ATTEMPT 2ND INTERVAL LONG VLAA-2 BOT 8'8" 11' 1.I' NCNE 4'

ATTEMPT '8; CUTAGE Vt3A-1 TOP 15" 11' 1.1' VL3A-2 TCP 15" 11' 1.1 '

NCNE VLC3-1 l

4' AriEMPT '81 OUTAGE VLCS-2 11' 1.1' 4'

ATTEMPT '81 CUTAGE 11' 1.1' 5'

SCHEDULED '81 V LDB-1 11' 1.1' 5'

57" EXAMINED VLDB-2 CIRC VCSS-1 57' 2.85' 6'

ATTEMPT 'al CUTAGE 6'

ATTEu?T 2ND INTERVAL VC38-3 57' 2.85' NCNE 57' 2.85' 16' ATTEMPT '81 CUTAGE VCSS-4

'LOSURE HD HMCS-1 C

7'

.7' ALL 7' EXAMINED MERIDCNAL HMCB-2 7'

.7' ALL 2.5' EX AMINED 7'

.7' ALL 7' EXAMINED HMCS-3 7'

.7' ALL 7' SCHEDULED '81 HMC3-4 7'

.7' ALL 7' EXAMINED HMC3-5 7'

.7' lLL 7' SCHEDULED '31 HMC3-6 CIRC HCC3-2 25' 1.25' ALL 25' EXAMINED 30TTCM HD HMAB-1 6'2"

.6' 2'8" ATTEMPT 2ND INTERVAL 6'2"

.6' 2'8" ATTEMPT 2ND INTE?, VAL MERID0NAL HMAS-2 6'2"

.6' 2'8" ATTEMPT 2ND INTERVAL HMAB-3 6'2"

.6' 2'8" ATTEMPT 2ND INTERVAL HMAS-4 6'2"

.6' 2'8" ATTEMPT 2ND. INTERVAL HMA8-5 6'2"

.6' 2'8" ATTEMPT 2ND INTERVAL HMAB-6 6'2"

.6' 2'8" ATTEMPT '81 CUTAGE HMAB-7 6'2"

.6' 2'8" ATTEMPT '31 CUTAGE HMAB-8 l

HMAB-9 14'5"

1. 5 '

NONE l

HMAS-10 13'8" 1.4' NONE 44' 2.2' ALL ATTEMPT '81 CUTAGE CIRC HCAS-1 81.12 E-H l

LUGS 4 LUGS 4 LUGS 4 LUGS ATTEMPT '81 CUTAGE l

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l Length depends on the results obtained curing the 1E81 outage.

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?00R BRIGINM impractical. However, since the vessel pressure test hold time is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and the ifcensee nas committed te inspectica of the areas

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below these penetrations for evidence of leakage,~ any loss of leak-tight integrity should be detectable. We agree that the alternate test and inspection proce' dure proposed by the licensee provides an acceptable level of safety and therefore grant relief from the visual inspection requirement of Category B-E.

C.

Closure Head Flange Leakage Sensors (Examination Category B-E, Item 81.5).

Relief Request 18.

' Code Recuirement Th'e area surrounding each penetration shall be visually examined for evidence of leakage during the pressure test.

Code Deviation Recuest Nozzles will be visually inspected only if area insulation is removed for other maintenance or inspection activities.

This relief request applies to nor:les N-13 and N-14.

Reason for Recuest There are two seal leak detector no::les; N13 and N14.

Nc::le 13 has an o'pening on the vessel flange face between the inner and outer "0" rings, and no::le 14 has an opening between the outer "0" ring and the outer gasket. The inner "0" ring seal would have to leak for N13 to see liquid level and pressure buildup, and both "0" ring seals would have to leak for N14.

The seal leakage is monitored through N13, and the line from N14 is run to a point fr. side the drywell and capped off.

The reactor vessel head seal leak detection system is designed to monitor reactor vessel head seal integrity.

This is accomplished by detecting ifquid level buildup and pressure buildup in tne drain line between two metal "0" rings whicii comprisa the reactor vessel head seal. The reactor vessel head seal leak detection system con-siste of a closed chamber located in the head seal drain line. A float type level switch is mounted in this chamber. A pressure switch and a pressure indicator are also included as part of the system.

The only control included in the vessel head seal leak detection system is the reactor flange drain valve switch, wnic.': operates to energize and de-energize valves CV-2369 and CV-2370. With the switch in the CLOSE position, SV-2369 is energized to open CV-2369 and SV-2370 is de-energi:ed to close CV-2370.

Local contacts light the CV-2369 red indicator lamp and the CV-2370 green indicator lamp, indicating the seal leak detection system is in service. To drain the system, the handswitch is placed in the CPEN position de-ener-gizing SV-2369 to close CV-2369 and energizing SV-2370 to open CV-2370. allowing any leakage collected in the ficat cage to drain to the Drywell Equipment Drain Sump.

The facility to crain the reactor vessel head-seal leak detection system provides a convenient method of determining the approximate leak rate.

Draining and 6

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returning the system to service permits recording the time required to refill the closed chamber and subsequent calculation of the leak rate.

The reacter vessel insulation prohibits access to the N13 and N14 nozzle area, and was not designed to be removed. The insult.tfon would receive considerable damage by removal and replacement.

Removable insulation has not been considered due to the fact that the seal leak detectors do not see reactor pressure anc it would therefore rerve ro purpose.

Staff Evaluation The Northern States Pcwer Company has a system designed to monitor the reacter vessel head seal integrity and-leakage. Nc::le 14 is outside the vessel head 0-rings and no::le 13 is located between ~

the vessel head 0-rings. Any leakage of the vessel flange 0-rings would be detected by the vessel head seal leak detection system.

1 The licensee has agreed to inspect the area around each nc::le if the insulation is removed for maintenance er inspection activities.

In addition, the staff requirt.1 that the vessel head seal be hycro-statically tested once per inspection interval. The staff concluces that the combination of the leak detection system design and hydre-static testing provide reasonable as,surance that the integrity of these nc::1ss will be maintained.

Therefore, relief frem the visual inspection requirements of nozzles 13 and 14 may be granted; j

0.

Piping Socket Welds (Examination Category 3-1, ! tem B4.8) and Valve Sennet Solting (Examination Catagory S-G-2, Item 36.9).

Relief Request No. 22.

Code Recuirement Category S-J:

Surface examinations performed during each inspection interval shall include all of the area of 25% of the circumferential Joints including the adjoining one-foot sections of icngitucinal joints and 25% of the pipe branch connection joints of welds six inches in diameter and smaller.

Category S-G-2: A visual examination shall be conducted en all pressure retaining bolting two-inch diameter or smaller.

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Reason for Reouest These two, 2-inch decin lines are reading in excess of 2 R/hr. The location of these lines prevents the use of snielding or distance to provide any significant reduction in radiation exposure to cer-sonnel. We have estimated that exposure to inspection and insulating personnel would be in excess of 1 man-rem for the examination of appecxigely fcur sccket welds anc the bolting of four valves.

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o Staff Evaluation The ifcensee has not demnnstrated that the performance of the code required examinatio'ns is impractical due to design, geometry, or materials cf construction and therefore relief cannot be granted at this time.

E.

Visual Examination of.lecirculation Pumps Casings (Examination Category 8-L-2, Item 85.7).

Relief Request 41.

Code Recuirement

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The internal pressure boundary surfaces of. one pump in each group of pumps performing similar functions in the system shall be visually examined once each interval.

Coce Deviation Reauest The visual ' examination of the recirculation pumps internal pressure surfaces will not be scheduled as required by code.

Reason for Recuest Oisassembly of the recirculation pumps for the sole purpose of visual examination of the casing internal pressure surfaces requires many man-hours of skilled maintenance personnel.

Increased radiation exposures result from this activity.

The probability of pump failure is increased by unnecessarily disassemcling the units.

Deferring the examination has no effect on integrity of the pum:s.

The internal pressure surfaces of these pumps will be visually examined when the pumps are disassembled for,. maintenance.

Staff Evaluation The disassembly of the reactor recirculation pumps to inspect the internal pressure retaining surfaces is a major effort in terms of man-hours and personnel exposure. An estimated 614 man-hours and 30-50 man-ren would be required for a visual examination.

The licensee has committed to a' visual examination if the pump is dfsassembled l

I for maintenance.

In addition the following parameters are monitored I

by the ifcensee for each of the recirculation pumps:

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1.

Pump flow I

2.

Differential pressure across pump 3.

Pump motor vibration 4.

Pump oil level 5.

Pumo_ discharge temperature 6.

Puro seal circulation pressure (first and second stage)

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POOR ORIGIN 7.

Temperature of the motor thrust bearing (upper and lower face) 8.

Temperature of the motor guide bearing (upper and lower) 9.

Temperature of the motor-windings 10.

Temperature of the first and second stage seal cavity 11.

Temperature of the seal closed cooling water 12.

Temperature of the motor closed cooling water Deviaticns in some of the above acnitored parameters may indicate degradation of pump ccmpenents. These cegraded components could affect pump performance to the point that maintenance would be required long before degradation of the internal pressure boundar9 surface is noticeable.

We conclude' that the effort required for dissasembly of the pumps and the associated radiation exposure solely for a visual examina-tion is impractical and h: poses an undue burden on the licensee.

Visual examinations will be conducted if the pumps are disassemcled for other reasons. Other parameters of pump operation will aid in detecting degradation of pump component!.

Relief is therefore granted from the visual examination requirement.

F.

Visual' Examination of Recirculation Valves (Examination Category S-M-2, Item S6.7), Relief Request 42.

Code Recuirement Visual excmination of internal pressure boun'Eary surfaces en valves exceeding 4 in. nominal pipe size, ence each interval.

One valve in each group of valves of the same constructional design (e.g., glebe, gate, or check valve), manuf acturir.g methed, and manuf acturer that performs similar functions in the system shall be examined.

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l Code Deviation Recuest t

l The visual examination of tne recirculation valve internal pressure l

surfaces will not be scheduled as required by the code.

Reason for Recuest Dissassembly of the recirculation valves for the sole purpose of visual examination of tne internal pressure surfaces requires many man-hours frem skilled maintenance personnel and increased radiation exposures result frem this activity.

The probability of l

l valve failure is increased by unnecessarily disassembling the units.

Deferring the examination has no effect on the integrity of the valves. The internal pressure surfaces will be visually examined

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when thelumps are disassembled for maintenance.

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P00R ORIGINAL Staff Evaluation The disassembly of these valves to visually inspect the pressure retaining surfaces would require approximately 266 man-hours of labor and 13-25 man-rem in radiation exposure. The licensee has r:cmmitted to a visual examination if the valves are disassembled for maintenance.

In addition, periodic hydrostatic testing as required by the code provides further assurance that the structural integrity of the valve bodies will be maintained.

Relief is therefore granted from the visual examination requirement of the code.

G.

Pressure Testing of all Class 1, 2, and 3 Components (Relief Requests 30 and 31).

Code Recuirement

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IWB-5000: Class 1 components shall be subject to a hydrostatic pressure test at or near the end of each inspection interval. The pressure shal be 1.10 X P, at 100*F.

IWC-5000: Class 2 components shall be subjected :o a hydrostatic pressure test at or near the end of each inspection interval. The pressure shall be 1.25 X P at 100*F.

D IWD-5000: Class 3 ccaponents shall be subjected to a system pressure test of 1.10 X P

• O Code Deviation Recuest The test pressure requirements will not be met an certain components.

Reason for Recuest The code does not s ecognize that non-isolable junctions of components with different design pressures or different ASME classes exist (i.e.,

pump suction and discharge lines, piping upstream and dcwnstream of restr' cting orifices, etc. ).

Pressurizing components to the require-i ments of the code may result in over pressurizing the non-isolable components.

Staff Evaluation Because of the design of some systems, isolation of different ASME classes cannot be made at the class boundary.

The licensee has committed to hydrostatically test the non-isolable portions of these systems at the 1cwer design pressure and to conduct pressure tests during system inservice tests.

The staf.f concludes that the alternate testing program preposed by the licensee is adequate to previce evidence of leakage and, therefore, grants relief from code pressure test requirements for these functions.

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P00R 0RIGINAL H.

Reactor vessel Safe End Welds and Piping Welds (Examination Categories B-F, B-J, B-K, B-F, C-F, C-G).

Relief Request No. 21.

Code Recuirement The ultrasonic examination requirements of Appendix I of Section XI apply to Class 1 and 2 l'erritic vessels 2-1/2 inches and greater in thickness. Where Appendix I is not applicable, the provisions of Article 5 of Section V shall apply.

Code Deviation Recuest The rules of Appendix III, including Supplement 7, of the Winter 1975 and Summer 1976 Addenda to ASME Section XI will govern the ultrasonic examination method for the inspection of pipe welds and welds of,,

components fabricated from pipe components.

Reason for.Recuest The use of side drill holes (instead of slots) to establish a distance amplitude correction curve (DAC) for pipe weld inspections, as recuired by Appendix I of Section XI and Article 5 of Section V, results in an excessive instrument gain setting which greatly impairs the inspector's ability to detect and to interpret indications by producing a lower signal-to-noise-ratio and reduces the range of useable DAC.

Staff Evaluation Appendix III, including Supplement 7 of the Winter 1975 Addenda and Summer 1976 Addenda, is not approved by tn? NBC in 50.55a(b) of 10 CFR 50.

However, the 1977 Edition through Summer 1978 Addenda are essentially the same as the earlier addenda.

Hence, the provisions of Appendix III specified by the licensee may be utili;ed.

For ferritic vessels greater than 2 inches in thickness, the require-ments of Article 4 of Section V in the Winter 1975 Addenda or Appendix I in the 1974 Edition of Section XI must be met since Appendix III applies only to piping welds.

The staff requires the following regarding DAC recording levels for piping welds:

1)

All indications at or above SC% DAC shall be recorded.

2)

All indications 10C% DAC or greater shall be recorded and evaluated in accordance with the rules of Section XI.

3)

Indications 20% of DAC or greater which are interpreted to be a crack must be identified and evaluated to the rules of Section XI.

4)

The owner shall evaluate and take corrective action for the disposition of any indication investigated and found to be otner than geometric in nature.

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POOR ORIGINM

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Class 1 and 2 Solts and Studs (Examination Categories B-G-1 and C-0),

Relief Request 24.

Code Recuirement

, Article 5 of Section V requires that calibration be established by a test bar of the same nominal composition and diameter as the pro-duction part and minimum of one-half of the length. A 3/8-inch (10mm) diameter Y 3-inch (76mm) deep flat bottom hole shall be drilled in one end of the bar and plugged with similar material to full depth. A distance amplitude curve shall be established by

' scanning frcm both ends of the best bar.

Code Deviation Recuest Use the alternate procedure for ultrasonic examination in Paragraph T525.1 for bolts and studs prior to threading (Sack Reflection. Procedure).

Reason for Recuest The variation in ultrasonic attenuation between bolts and/or studs diminishes the usefulness of a DAC generated from a particular specimen.

In addition, a procedure qualification test was performed comparing the ASME Section V requirements with the NSP Back Reflection Belt and Stud Examination Procedure. The results showed that the NSP procedure was sligntly more sensitive than the ASME Section V requirements.

t l

l The comparison was performed using a 2" dia. X 15" long test bar l

machined as a calibracion standard in accordance with ASME Section V, and another 15" long section of the same bar used to simulate a stud. The back reflection of the stud was set at 8C% F5H in accordance with the NSP procedure. The amplitude of the 3/8" dia. FBH (at 12") in the calibratien standard was then noted to be about 6 dB greater than the reporting level for the NSP procedure.

This result should be generally true regardless of bar length for diameters greater than 3/4" because the ratio between the area of the FSH and the 3/4" dia. transducer remains constant at 1/4.

Therefore, the FBH should produce an echo of at least 25% of the back return, depending on the relative finish of the reflecting l

surfaces.

In addition, the poorer the end reflecting surfaces the more sensitive the NSP procedure would be with respect to the Section V method. This tends to make the NSP procedure a more conservative approach to bolt and stud examination.

Staff EvA uation l

As an alternate procedure the licensee proposes examination of studs / bolts using the first back reflection. The procecures 12 w

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requires that the back reflection be set at 80% of screen height.-

Any discontinuity producing an indication 20% of the back reflection or greater is recorded.

In addition, any indication producing 50%

loss of back reflection that is not caused by bolt or stud geometry is recorded and evaluated:

The licensee's precedure will be used where the code requires ultrasonic examination of bolts / studs except for reactor vessel studs which will be examined per code.

The licensee provided data showing that this alternate procedure is as sensitive for flaw detection as the alternate cede procedure.

Paragraph IWA-2240 of Section XI permits the use of alternative examinations provided the results yield demonstrated equivalence or superiority.

Relief may therefore be granted for this deviatica from the code.

J.

Non-Welded Piping and Valve Supports (Examination Categories 3-K-2 and C-E-2)

Relief Request No. 23.

Code Recuirements The examination performed during each inspection interval shall cover all support ccmponents.

The areas of examination shall include support components that extend from the piping, valve, and pump attachment to and including the attachment to the supporting structure.

Code Deviation Recuest The Code requires all areas of the support ccmponent frem the piping, valve, and pump attachment to and including the attachment to the supporting structure.

Insulatier, will not be, removed for visual examination of these support components.

Reason for Recuest The general radiation background field for the inspection of Class 1 systems located within containment ranges from 30 to 400 mr and the Class 2 systems have permanent type of insulation (insulation not designed for removal and replacement). Acproximately 35 man-rem in personnel exposure would result from Class 1 examinations in the third inspection period.

It has been our experience that any loss of support capability or

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inadequate restraint can usually be detected througn the inspection of the uninsulated portion of the support and the surrounding insula-tion.

It is our contentien that the remeval and replacement of insulation for the sole purpose of inspecting Class 1 supports would result 'i undue radiation exposure to personnel without previding a significant increase in safety.

The governing cedes and regulations used in the design and construction of those systems that are now classified as Class 2 and 3 did not require provisiens for inscection access fE these systems.

Thus, it would be an uncue burden without compensating increase in safety to require insulation removal for support inspection.

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POORggjgjy Staff Evaluation The mechanical connections to pipe strap are exposed and the licensee has committed to visual examination of these connections.

The estimated exposure is 35 man-rem just to remove insulation to perform visual inspection of the pipe strap.

Possible damage to the pipe strap can be detected by noting damage to insulation and the licensee has committed to remove insulation around pipe support in the event that visual examination of the insulation indicates pipe strap damage or loss of support.

The staff finds the design of the supports and insulation make the exact code required examinations impractical, agrees that the above procedure will provide an adequate level of safety and, therefore, grants relief frem the visual inspection criteria of Category S-K-2 and C-E-2; Items 1.4, 2.2, 2.3 and 2.4.

II. Additional Relief Requests In addition to the relief requests evaluated in Section I, the hicensee submitted three requests for relief which involved updating to examination requirements c? the 1977 Edition through Summer 1978 Addenda of Section XI of the ASME Code.

Updating to the requirements of later NRC approved editions and addenda is permitted by 50.55a(g)(4)(iv), provided all of the related requirements of the respective editions or addenda are met.

We have evaluated the following relief requests submitted by the licensee and find them to be acceptable and in accordance wita 50.55a(g)(4)(iv):

Relief Recuest Examination Comoonent Identi fication Catecorv Examina:1on No. 20 8-I-1, Item 31.13 Visual exam of closure head cladding.

No. 36 All Class 1, 2 & 3 Holding time during hydrotest.

No. 37 All Class 2 & 3 Air &

Pressure testing with Nitrogen Systems contained fluid.

No. 38 All Class l'& 2 Austenitic stainless steel temperature during hydrotest.

III. Administrative Changes Table 3.2.G on pages 70 and 71 of the Technical Specifications was renumbered to Table 3.2.7 to correct duplication of numbering with Table 3.2.6 on page 60A. A minor editorial change was also made to Table 3.2.7.

We find these changes acceptable as a~dministrative changes.

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Summary Based on the evaluations of the requested relief from ASME Code requirements, we conclude that the Monticello Nuclear Generating Plant Inservice Inspection Program for last period in the first ten yea: interval meets the requirements of the 1974 Edition through Summer 1975 Addenda of the ASME Code to the extent practical and thus complies with 50.55a(gl in 10 CFR 50.

V.

Environmental Consideration We have determined that this amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involvesan action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement, or negative declaration and environ-mental impact appraisal need not Be prepared in connection with the'-

issuance of the amendment.

VI. Conclusion We have concluded, Based on the considerations discussed above, that:

(.1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (21 there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Commission's reou-lations and the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated: June 3,71981 O

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