ML20148L556

Requests Relief from Certain ASME Code Section XI Requirements for Plant Second Inservice Insp Plan,Per Encl Detailed Requests.Southwest Research Inst Program Plan for Second Insp Interval For..., Also Encl.Fee Paid

ML20148L556
Person / Time
Site:	Calvert Cliffs
Issue date:	01/14/1988
From:	Tiernan J BALTIMORE GAS & ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
Shared Package
ML20148L561	List: ML20148L556 ML20148L643
References
RTR-REGGD-01.147, RTR-REGGD-1.147 NUDOCS 8801280182
Download: ML20148L556 (31)

Text

B A LTIMORE GAS AND ELECTRIC CHARLES CENTER R O. BOX 1475 BALTIMORE, MARYLAND 21203 ,

1 JOSEPH A.TIERNAN Vict PatsIDENT January 14, 1988 NUCLEAR ENERGY U. S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTENTION: Document Control Desk SUBJ ECT: Calvert Cliffs Nuclear Power Plant Units Nos. 1 & 2; Docket Nos. 50 317 6 50-318 Inservice Inspection Procra d gqu s lo1 Relief and Inservice Inspection Plans ENCIDSURES : (1) Relief Requests for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 (2) Program Plan for the Second Inservice Inspection Interval for Calvert Cliffs Nuclear Power Plant, Units 1 and 2 Gentlemen:

This letter forwards several related submittals required by 10 CFR 50 and ASME Code Section XI. These submittals are applicable to the second Inser-vice Inspection interval of Calvert Cliffs Units 1 and 2. This submittal does not address the Inservice Testing Program for pumps and valves; submittals pertaining to pump and valve testing have been filed separately.

As provided by 10 CFR 50.55 a(g)(6)(1), we are requesting relief from ASME Code Section XI requirements that have been determined to be impractical.

'I In accordance with 10 CFR 50.55 a(g)(5)(iii) and (iv) and NRC Staff Guid-1 ance letter dated November 24, 1976, the details for each exemption request are provided in Enclosure 1. These requests are based on experience gained in the course of inspections to date. Additional requests will be submit-ted as needed.

In addition to the above relief requests, two specific requests which apply to portions of the second inspection interval were previously filed and approved. These were non-generic, one time requests which identified specific requirements which were impractical. By letter dated March 26, 1987, from Mr. A. C. Thadani to Mr. J. A. Tiernan, relief was granted concerning the requirement to show that primary stress limits were satis-fled for a Calvert Ciffs Unit 1 main steam pipe area with reduced vall thickness. By letter dated October 31, 1987, from R. A. Capra to J. A.

Tiernan, relief was granted from the requirement to perform hydrostatic

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Document Control Desk January 14, 1988 Page 2 testing of an unisolatable portion of replaced steam generator blowdown piping. These two relief provisions are retained for the applicable periods.

During the second Inservice Inspection interval we intend to implement certain NRC approved Code Cases. The proposed change to Footnote 6 of 10 CFR 50.55(a) (ref: FR Doc. 87-14599 Filed 6-25-87) would incorporate NRC Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability -

ASME Section XI Division 1," as a reference which identifies the Code Cases acceptable to the NRC for implementation in the ISI program of light-water-cooled nuclear power plants. According to the supplementary information accompanying the proposed rule making, Code Cases listed in Regulatory Guide 1.147 may be used without specific requests to the NRC.

Once the proposed rule is issued in final form, as is expected by March 31, 1988, we intend to adopt the following Code Cases which are approved in Regulatory Guide 1.147.

{ code Case N-307-1 "Revised Ultrasonic Examination Volume for Class 1 Bolting, Table IWB-2500-1, Examination Category B-G-1, When the Examinations are Conducted From the Center-Drilled Hole" Code Case N-408 "Alternative Rules for Examination of Class 2 Piping" l

Code Case N-416 "Alternative Rules for Hydrostatic Testing of Repair or Replacement of Class 2 Piping" l Code Case N-424 "Qualification of Visual Examination Personnel"

, A copy of each is included in Table 6 of Enclosure 2 for your reference.

1 As required by ASME Code Section XI in the 1983 Edition with Addenda l through Summer 1983 subparagraph IWA-1400(c), Enclosure 2, Program Plan for the Second Inservice Inspection Interval for Calvert Cliffs Nuclear Power Plant. Units 1 and 2 is submitted herewith. This plan outlines the minimum criteria which will be applied to Calvert Cliffs Units 1 and 2 Inservice l Inspection Programs throughout the second interval. This plan does not

address the Inservice Testing Program for pumps and valves; that program l has been filed separately.

l The Program Plan is responsive to ISI examination requirements invoked by

! 10 CFR 50.55a and is in keeping with our commitment to adhere to ASME Code l rules. This Plan reflects our minimum obligations. We fully expect that these minimum requirements will be exceeded as has occured throughout our first inspection interval. Long Term ISI Examination Schedules for each of the Calvert Cliffs Units will be prepared in accordance with the ISI Program Plan. Following completion and review, the Long Term Examination Schedule summary will be filed under separate cover, i

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Document Control Desk January 14, 1988 Page 3 Should you, in the course of your review have^any questions regarding ,

this submittal, please do not hesitate to contact us. (

We have determined that this request constitutes an amendment for Calvert~

Cliffs Unit.Nos. 1 and 2', pursuant to 10 CFR 170.21. Accordingly, Baltimore Gas & Electric Check No. 1914614 in the amount of $150.00 is enclosed.

Very truly yours, j YO JAT/LMD/j af Enclosures cc: D. A. Brune, Esquire (w/o enclosures)

J. E. Silberg, Esquire (w/o enclosures) r S. A. McNeil, NRC T. Foley/D.C. Trimble, NRC (w/o enclosures)

W. T. Russell, NRC R. A. Capra, NRC (w/o enclosures)

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ENCII)SURE ONE 1

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RELIEF REQUESTS FOR CALVERT CLIFFS NUCLEAR POWER FIANT, UNITS 1 AND 2 l

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O ENCIDSURE ONE Table of Contents Summary of Relief Requests . . . . . . . . . . . . . . . . . . . . 1 Relief Request Nurnber One . . . . . . . . . . . . . . . . . . . . . 2 Relief Request Number Two . . . . . . . . . . . . . . . . 5 Relief Request Number Three . . . . . . . . . . . . . . . . . . . . 8 Relief Re post Number Four . . . . . . . . . . . . . . . . . . 10 Relief Request Number Five . . . . . . . . . . . . . . . . . . . . 14 Attachment (1)

Attachment (2)

Attachment (3)

Attachment (4)

Attachment (5)

Attachment (6)

Attachment (7)

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SUMMARY

OF RELIEF RIMUFSTS Relief ASME Requirement Proposed Request Section XI For Which Relief Reason for Alternate Number Reference Component Is Reauested Relief Reauest Examination 1 IWB-2500-1 42" and 30" Surface Difficult access Examine using UT Cat. B-J Reactor Coolant Examination. from the outside near outside sur-Pipe Welds in surface and prohibitive face technique from Reactor Vessel radiation environment. inside surface.

Annulus.

2 IWB-2500-1 Pump Case Welds Volumetric of Complex pump configura- Hydrostatic test, Cat. B-L-1, and Internals of case welds and tion and cast stainless surface, and visual B-L-2 Reactor Coolant visual of material. examinations of Pumps. internals. outside surface of one pump.

3 IVB-2500-1 Welds in CRD Surface examina- Portions of all CRD Surface examination Cat. B-O Housings. tion of 10% housing welds on five CRD housing (three) peri- inaccessible due to welds to compensate pheral CRD to closure head con- for inaccessible housings. figuration. portions.

4 IWA-5000, Portions of Class 2 Class 2 portions Perform hydrostatic IWC-5000 Class 2 HPSI, Hydrostatic cannot be isolated pressure tests to Aux HPSI, and Pressure Test from Class 1 due to Class 1 hydrostatic LPSI. every 10 years. check valve isolations. pressure require-ments.

5 IWA-5200, Class 3 Portions Hydrostatic Test These systems not System inservice IWD-5200 of Component every 10 years. practically isolated. pressure testing Cooling Water, annually in lieu of Service Water, hydrostatic test.

and Salt Water Cooling.

.s ENCLOSURE ONE Page 2 of 15 RELIEF REQUEST NUMBER ONE:

1. COMPONENT FOR WHICH RELIEF IS REOUESTED:

A. Name and Number Calvert Cliffs' Reactor Coolant System 42" and 30" piping welds located in the reactor vessel cavity annulus. The following welds are affected:

UNIT 1 Line. Weld Tyne 42-RC-11 1 Nozzle-to-Transition Piece 42-RC-11 2 Transition Place-to-Pipe 42-RC-11 2 LD-1 Longitudinal Seam 42-RC-11 2 LD-2 Longitudinal Seam 42-RC-12 1 Nozzle-to-Transition Piece 42-RC-12 2 Transition Piece-to-Pipe 42-RC-12 2 LD-1 Longitudinal Seam 42-RC-12 2 LD-2 Longitudinal Saam fs 30-RC-11A 12 LU-1 Longitudinal Seam 12 LU Longitudinal Seam 12 Elbow-to-Transition Piece 13 Transition Piece-to-Nozzle 30-RC-11B 12 LU-1 Longitudinal Seam 12 LU-2 Longitudinal Seam 12 Elbow-to-Transition Piece 13 Transition Piece-to Nozzle 30-RC-12A 12 LU-1 Longitudinal Seam 12 LU-2 Longitudinal Seam 12 Elbow-to-Transition Piece 13 Transition Piece-to-Nozzle .

30-RC-12B 12 LU-1 Longitudinal Seam 12 LU 2 Longitudinal Seam 12 Elbow-to-Transition Piece 13 Transition Piece to-Nozzle

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ENCLOSURE ONE Page 3 of 15 UNIT 2 Line Veld Tvoe 42-RC-21 1 Nozzle-to-Transition Piece 42-RC-21 2 Transition Piece-to-Pipe 42-RC-21 2 LD-1 Longitudinal Seam 42-RC-21 2 LD-2 Longitudinal Seam 42-RC-22 1 Nozzle-to-Transition Piece 42-RC-22 2 Transition Piece-to-Pipe 42 RC-22 2 LD-1 Longitudinal Seam 42-RC-22 2 LD-2 Longitudinal Seam 30-RC-21A 12 LU-1 Longitudinal Seam 12 LU-2 Longitudinal Seam 12 Elbow-to-Transition Piece 13 Transition Piece-to Nozzle 30-RC-21B 12 LU-1 Longitudinal Seam l 12 LU-2 Longitudinal Seam 12 Elbow-to-Transition Piece 13 Transition Piece-to-Nozzle 30-RC-22A 12 LU-1 Longitudinal Seam 12 LU-2 Longitudinal Seam O- 12 13 Elbow-to-Transition Piece Transition Piece-to-Nozzle 30-RC-22B 12 LU 1 Longitudinal Seam 12 LU 2 Longitudinal Seam 12 Elbow-to-Transition Piece ,

13 Transition Piece-to-Nozzle O

ENCLOSURE ONE Page 4 of 15

- m B. Function The reactor coolant system piping transfers reactor coolant from the reactor vessel outlets to the steam generator (S/G) inlets (42") and from the S/G outlets to the reactor coolant pumps (30") and from the coolant pumps to the reactor vessel inlet (30").

C. Code Class Current ISI Class: Class 1 Original Design: B31.7, Class 1 (1969)

II. CODE REOUIREMENT R OM WHICH RELIEF IS REOUESTED:

ASME Code Section XI, 1983 Edition with Addenda through Summer 1983, Category B-J, Item B9.11 and B9.12 requires that these welds receive both a surface and volumetric examination. The surface examination is impractical to perform.

III. BASIS FOR RELIEF:

In order to perform the required surface examination, the examiners '

must gain access to the reactor vessel annulus area housing these reactor coolant piping welds. This area is very difficult to enter, f"T provides marginal room for mobility, and has high radiation.

U IV. ALTERNATIVE EXAMINATIONS:

As an alternate to performing a surface examination, a 45-degree shear wave UT examination of the outside surface will be performed by utilizing mechanized ultrasonic techniques from the inside of the pipe. This method of examination has been qualified for the detection of unacceptable outside surface flaws through the use of a mock up with induced cracks ranging from 1/2 the maximum to the maximum allowable Codo flaw depth. The use of this examination method will cause a significant reduction in radiation exposure.

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ENCIDSURE ONE Page 5 of 15 I

RELIEF REQUEST NUMBER 'IVO:

1. COMPONENT FOR WHICH RELIEF IS REOUESTED: )

A. Name and Number Calvert Cliffs Unit 1 Reactor Coolant Pumps (RCP's) #11A, #11B,

1. 12A, and #12B, and Calvert Cliffs Unit 2 Reactor Coolant Pumps

1. 21A, #21B, #22A, and #228. All pumps are identical in design and function and are Byron-Jackson Type DFSS Reactor Coolant Pumps, Serial Numbers 691N-(437 through 44, Size 35 x 35 x 43.

B. Function Each Calvert Cliffs unit has four reactor coolant pumps which are welded to the 30" recirculation loop. These pumps function during normal reactor operation to provide forced recirculation through the core.

C. Code Class Current ISI Class: Class I original Design: ASME Ccde Section III, 1965 Edition with addenda through Winter 1967, Class 1.

O II. CODE REOUIREMENT FROM WHICH RELIEP IS REOUESTED:

ASME Code Section XI, 1983 Edition with Addenda through Summer 1983, Examinction Categories B L-1 and B-L-2, requires volumetric examina-tion of casing welds anc visual examination of internal pressure boundary surfaces of one pump casing in each of the pump groups performing similar system functions during each inspection interval.

These examinations are impractical for the Reactor Coolant Pumps at Calvert Cliffs Units 1 and 2.

III. BASIS FOR RELIEF:

A. The design configuration of each pump, as shown in Attachment (1), corresponds to a Type E pump illustrated in Figure NB 3442.5-1 (1977 Edition, ASME Code Section III). No practical technique exists to perform Inservice Inspection Radiographic Examination (RT) or Ultrasonic Fxamination (UT) of this type Pump.

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ENCIDSURE ONE Page 6 of 15 B. The presence of the diffuser vanes precludes conventional RT.

The vanes and radiation field prevent placement of the Rf film cassettes inside the pump. Placement of the film on the outside of the pump is feasible, but there is no radiographic source suitable for placement inside the pump. Standard gamma sources are of limited use for penetrating the thick casting, and back-ground radiation from the inside surface of the pump impairs film sensitivity. The Miniature Linear Accelerator (MINAC) was considered, but the Type E pump design precludes positioning of the accelerator inside the pump. Double wall radiography utiliz-ing the MINAC may be useful for a pcrtion of the casing welds.

This technique has not been qualified and may not be adequate.

C. The coarse grain structure inherent in thick stainless steel castings precludes the use of conventional UT. Developments in ultrasonic techniques to date have not provided a method to examine thick stainless steel castings; ultrasonic examination would be preferred over the difficulties and dangers of thick wall radiography.

D. The pump casing is fabricated from cast stainless steel (ASTM A351, Grade CF8M). The material is essentially a cast type 316 stainless steel. This material is widely used in the nuclear industry and no industry failures of this type material in reactor coolant pumps have been noted. The presence of delta p ferrite (typically 15% or more) imparts increased resistance to t intergranular stress corrosion cracking (IGSCC). The delta ferrite also improves resistance to pitting corrosion.

E. Report Number ERP-06-102, Revision O. August 1983, prepared for the Electric Power Research Institute by NUTECH Engineers, Incorporated, concludes that:

1. Based on the generic pump casing analysis, there is justifi-cation for the extension of the pump casing examination up to 15 years.

2. Plant unique analysis probably will show greater margins of safety.

3. The fracture mechanics analysis shows that large, final flaw sizes can be tolerated in the pump casing before fracture is predicted

4. The recent 10 year Inservice Inspection of several pump casings (Type F) indicates no detectable flaw growth from base line inspections, which corroborates the. above analyti-cal conclusion.

F. Pump disassembly for a limited visual examination of the interior pressure boundary surfaces of a reactor coolant pump is of little merit. Over 700 manhours and over 20 person / rem is estimated for

ENCLDSURE ONE Page 7 of 15

() disassembly, visual inspection, and reassembly of one reactor coolant pump. Additional manhours and person rem will be expended by Radia-tion Protection personnel providing direct coverage of this work.

Most of the work would be performed under full face mask conditions.

The radiation exposure cannot be justified considering the limitations of the internal visual examination.

IV. ALTERNATIVE EXAMINATIONS:

A. One pump interior will be inspected to the extent practical (in recognition of the vanes therein) only if any pump be disassem-bled for any other reason.

B. The reactor coolant pumps shall be hydrostatically tested per the requirements of ASME Code Section XI.

C. A surface examination of one RCP in each unit shall be performed once each interval on the exterior casing weld surface areas by the liquid penetrant method.

D. A visual examination of one RCP in each unit shall be performed once each interval on the exterior pump case surfaces.

This exemption was approved for use during our initial Inservice Inspection inte rval, copies of the approval letters, dated September O. 18, 1985, from H. R. Denton to Mr. A. E. Lundvall and November 6, 1985, from Mr. E. J. Butcher to Mr. A. E. Lundvall are included in-Attachment (2).

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'J RELIEF REQUEST NUMBER THREE:

I. COMPONENT FOR WHICH RELIEF IS REOUESTED:

A. Name and Number Peripheral Control Rod Housings (28)

B. Functi2D The reactor vessel head contains 85 control rod housings which serve as an extension of the pressure boundary in which control rod extension shafts can be raised and lowered. Each housing extends through a penetration in the reactor vessel head and is welded on the head inside surface with a 'J' groove type weld.

The housing contains only one full penetration circumferential butt veld which is shop fabricated and examined prior to assembly into the reactor vessel head. When installed, peripheral housing welds extend partially into the head itself.

C. Code Class Current ISI Class: Class 1 Original Design: ASME Code Section III, 1965 Edition f-- with Addenda through Winter 1967,

( j Class 1.

II. CODE REOUIREMENT M ON WilICll RELIEF IS REOUESTED:

ASME Code Section XI, 1983 Edition with Addenda through Summer 1983, Examination Table IWB 2500 1, Examination Category B 0, requires a volumetric or surface examination to include 100% of the welds in 10%

of the peripheral Control Rod Drive Housings during each inspection interval.

Relief is requested from the Code requirement to examine 100% of 10%

of the peripheral Control Rod Drive (CRD) Housing welds.

III. BASIS FOR RELIEF:

A 100% examination of these welds is impractical due to design config-uration, accessibility limitations and material of construction.

Reference Attachment (3) CE Drawing #233-415.

Ultrasonic examination will not provide meaningful results due to the geometric configuration of the joint and the material properties l (Inconel to-Stainless Steel welds). Radiographic examination cannot ,

be performed due to the design configuration and accessibility. l Therefore, a surface examination has been elected as the method of examination,

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ENCIDSURE ONE Page 9 of 15 Three of the 28 peripheral CRD Housing welds should be examined to meet the Code. However, only a portion of each weld is accessible for examination since the welds are partially obstructed because they extend into the Closure Head itself.

IV. ALTERNATIVE EXAMINATIONS:

In order to meet the intent of the ASME requirements, portions of additional CRD Housing welds will be examined to satisfy the equiva-lent of 100% of three welds. This will be done by examining 75% of three welds and 50% of two welds.

This exemption was approved for use during our first Inservice Inspec-tion interval. A copy of the approval letter, dated May 11, 1987, from Mr. R. A. Capra to Mr. J. A. Tiernan is included in Attachment (4).

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ENCLOSURE ONE Page 10 of 15 RELIEF REQUEST NUMBER FOUR:

I. Comoonent For Which Relief is Reauested:

A. Name and Number Calvert Cliffs piping associated with the High Pressure Safety Injection (HPSI), Auxiliary HPSI, and Low Pressure Safety Injec-tion (LPSI) Loop Isolation MOV's to the Reactor Coolant System, as shown on Attachment (5). The following lines are affected:

UNIT 1 EgQN IQ LINE NOS.

1-SI-118 1-SI-615-MOV 6"CC-13-1001 1-SI-616-MOV 2"CC-13-1019 1-SI-617 MOV 3"CC 13 1014 2"CC-13-1005 2"CC-6-1002 1-SI-128 1-SI-625 MOV 6"CC-13-1002 1-SI-626-MOV 2"CC-13 1018 1-SI 627 MOV 3"CC-13-1015 2"CC-13-1006 O 2"CC-6-1004 1-SI 138 1-SI-635 MOV 6"CC-13-1003 1-SI-636 MOV 2"CC-13 1016 1 SI-637 MOV 3"CC-13-1021 2"CC-13-1007 ,

2"CC-6 1005 1-SI-148 1-SI-645 MOV 6"CC-13 1004 1 SI 646 MOV 2"CC-13-1017 1-SI-647 MOV 3"CC-13-1020 2"CC-13 1008 2"CC-6-1006 l

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ENCLOSURE ONE Page 11 of 15 UNIT 2 IRQH IQ LINE NOS.

2 SI-118 2 SI-615-MOV 6"CC-13 2001 2-SI-616-MOV 2"CC-13 2019 2-SI-617 MOV 3"CC 13 2014 2"CC-13-2005 2"CC-6 2002 2-SI-128 2 SI-625-MOV 6"CC-13-2002 2-SI-626-MOV 2"CC-13-2018 2-SI-627-MOV 3"CC-13-2015 2"CC 13-2006 2"CC-6-2004 2-SI-138 2-SI 635-MOV 6"CC-13-2003 2-SI-636-MOV 2"CC 13-2016 2-SI-637 MOV 3"CC-13-2021 2"CC-13-2007 2"CC-6-2005 2-SI-148 2-SI-645 MOV 6"CC-13 2004 2 SI-646-MOV 2"CC-13 2017 2-SI-647-MOV 3"CC-13-2020

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(,, ) 2"CC-6 2006 B. Function The Safety Injection systems supply emergency core cooling, in the unlikely event of a loss-of coolant incident, to limit fuel rod damage and fission product release, and ensure adequate shutdown margin regardless of temperature. The systems also supply continuous long term post incident cooling of the core by recirculation of borated water from the containment sump.

C. Code Class Current ISI Class: Class 2 Original Design: B31.7, Class 2 (1969) )

II. CODE REOUIREMENT Mt0M WHICH RELIEP IS REOUESTED:

ASME Code Section XI, 1983 Edition with Addenda through Summer 1983, i requires hydrostatic testing of all Class 2 piping and components as I set forth in Articles IWA-5000 and IWC 5000. The test requirement for. j Class 2 piping and components is 1.25 times system pressure P,y, for i systems with Design Temperatures above 200 F. The system pressure

,-_ P,y shall be the lowest pressure setting among the number of safety or ks- i

ENCLOSURE ONE Page 12 of 15

() relief valves provided for overpressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure Pd shall be substituted for Psv' III. BASIS FOR RELIEF:

A. The listed portion of Class 2 piping from HPSI, Aux. HPSI, and LPSI Loop Isolation MOVs to the RCS is isolated from the RCS by two check valves. The higher system pressure test requirements of these portions of Class 2 systems cannot be accomplished because of the lack of positive isolation from the Class 1 system in the test direction.

B. In addition, the portions of piping listed below cannot by hydro-statically tested due to the inability to align the charging pumps to pressurize this piping and the operability requirements of these systems when the RCS is pressurized.

UNIT 1 IEQM IQ LINE NOS.

1-SI-114 1-SI 615-MOV 6"CC 13 1001 1-SI-124 1-SI 625-MOV 6"CC 13-1002 p)

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1-SI-134 1-SI-635-MOV 6"CC 13 1003 1 SI-144 1-SI 645 MOV 6"CC-13 1004 UNIT 2 FROM IQ LINE NOS.

2-SI-114 2-SI 615-MOV 6"CC-13-2001 2-SI 124 2-SI-625-MOV 6"CC-13 2002 2-SI-134 2-SI 635 MOV 6"CC-13-2003 2-SI-144 2-SI-645-MOV 6"CC 13-2004 i

IV. ALTERNATIVE EXAMINATIONS:

A. Excluding the piping listed in III.B, the remaining piping will 1

be hydrostatically pressure tested to the requirements of IWB-5000 for Class 1 piping. This piping can be pressurized via alignment of the charging system to the Aux. HPSI header, i

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ENCLOSURE ONE Page 13 of 15

( B. For the portions of piping which cannot be hydrostatically pressure tested, as listed in III.B. a leakage. test will be performed each refueling cycle, in accordance with Technical Specification 6.14. In this test the piping vill-be pressurized i to LPSI pump discharge pressure and a VT 2 examination for leakage will be conducted. In addition, welds will continue to be selected and examined per Section XI. Article IWC 2000.

This exemption was approved for use during our initial Inservice Inspection interval. A copy of the approval letter, dated November 14, 1985, from Mr. H. R. Denton to Mr. A. E, Lundvall is included in Attachment (6).

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ENCLOSURE ONE Page 14 of 15 RELIEF REQUEST NUMBER FIVE:

1. COMPONENT mR VHICH RELIEF IS REOUESTED:

A. Name and Number Calvert Cliffs' piping associated with the Component Cooling, Service Water, and Salt Water Cooling Systems and currently -

classified as ASME XI Class 3.

B. Function The Component Cooling and Service Water Systems remove heat from various auxiliary systems. Items cooled by Component Cooling Water include the letdown and shutdown cooling heat exchangers; reactor and steam generator supports; RCP, HPSI and LPSI seals and coolers; and containnent penetrations. The Service Water System removes heat from turbine plant components, blowdown recovery heat exchangers, containment cooling units, spent fuel pool cooling heat exchangers, and emergency diesel generator heat exchangers. The Salt Water Cooling System provides the cooling medium for the component cooling and service water heat exchang-ers.

C. Code Class Current ISI Class: Class 3 Original Design: B31.1 (1967)

II. CODE REOUIREMENT WOM WHICH RELIEF IS RFDUESTED:

ASME Code Section XI, 1983 Edition with Addenda through Summer 1983, requires hydrostatic pressure testing of all Class 3 systems in accor-dance with Subarticles IWA 5200 and IWD 5200. Paragraph IWD-5223(a) specifies that the hydrostatic test pressure shall be at least 1.10 times the system pressure, for systems with design temperatures of 200 F or less.

III. BASIS FOR RELIEF: ,

Hydrostatic pressure on isolated portions cannot be achieved because on the main headers of these systems, butterfly valves are installed, )

and a sufficient seal cannot be obtained.  !

I IV. ALTERNATIVE EXAMINATIONS: '

A system inservice pressure test will be performed on an annual basis i for portions of these systems outside of containment and on a refuel- i ing outage basis for those portions located inside containment.

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ENCIASURE ONE Page 15 of 15 This exemption was approved for use during our initial Inservice Inspection interval. Copies of the approval letters, dated January 24, 1983, from Mr. R. A. Clark to Mr. A. E. Lundvall and December 13, 1982, from Mr. R. A. Clark to Mr. A. E. Lundvall, are included in Attachment (7).

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REACTOR COOIANT PUMP O Figure NB-3442.5-1 1977 Edition, ASME Code Section III )

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NB.3000 - DESIGN NB-3442.6-NB.3442.7 e

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Q Section A - A 1.w FIG. NB.3442.7(a)-1 .

AXIALLY SPLIT CASING VOLUTE PUMP, FIG. NB 3442.5-1 TYPE E PUMP TYPE G q, ' rA I N cre tesi .

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% bA l $ * $ 1 FIG. NB 3442.7(a)-2 AXIALLY SPLIT CASING VOLUTE PUMP.

FIG. NBJ442.6(a).1 TYPE F PUMP TYPE G NB 3442 6 Design of Type F Pumps , NB 3442.7 Design of Type G Pumps *3 (c) Type F pumps are those having radially split, (a) Type G pumps are those having axially split, axisymmetric casings with either tangential or radial single, or double volute casings, as illustrated in Figs.

outlets as illustrated in Fig. NB.3442.6(a) 1. The NB-3442.7(a) 1 and NB 3442.7(a)-2.

basic configuration of a Type F pump casing is a shell .(b) Manufacturers shall review examination re-with a dished head attached at one end and a bolting quirements for compatibility, flange at the other. The inlet enters through the (c) An acceptable method orcalculating the stress dished head and the outlet may be either tangent to in highly stressed sections of the pump case, such as the side or normal to the center line of the casing.

Variations of these inlet and outlet locations are uit is neogmud mat oGn acceptable procedures may exist which also consutute adequate design m:2ods and atis not the Permitted. ,  ;,i,,u,, ,, p,,ws, i3,, ,it,,,,,,, ,,,3,e, p,,yun o, is,y

,} (b) The design of Type F pumps shall be in can de shon to have ban uustactory by actual wme's y/ accordance with this Subarticle, upenence.

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O ATTACllMENT (2)

LETTERS FROM TIIE NRC TO BALTIMORE CAS & ELECTRIC COMPANY September 18, 1985 i

November 6, 1985 ,

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHING TON. D. C. 20555 *

'+4 .',,.*' September 18, 1985 Docket Nos. 50-317

• and 50-318 Mr. A. E. Lundvall, Jr.

Vice President - Supply Baltimore Gas & Electric Company P. O. Box 1475 Baltimore, Maryland 21203

Dear Mr. Lundvall:

The NRC has provided relief from a requirement of the ASME Boiler and Pressure Vessel Code,Section XI, which BG&E has detennined to be impractical in accordance with your application dated February 4,1985 as supplemented by your letters dated May 31, 1985 and June 24, 1985.

The code relief, granted in accordance with 10 CFR Part 50 Section 50.55a(g)(6)(1), relates to the requirement for 100% volumetric examination of reactor coolant pump casing welds. You have proposed an acceptable, alternate fonn of examination consisting of: (1) pump interior examination O to the extent practical in the event that a pump is disassembled, (2) hydrostatic testing, and (3) surface examination of one reactor coolant pump's casing welds, per unit, and a 100% visual examination of this pump's exterior.

A copy of our Safety Evaluation is enclosed. -

Sincerely, p

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. i od . e , ctor Office of Nuclear Reactor Regulation

Enclosure:

Safety Evaluation rR AINING & TE_CH'!ICtt SERVICES cc w/ enclosure:

See next page iT n.mr 2: =- k e ,

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Mr. A. E. Lundvall, Jr.  ;

Baltimore Gas & Electric Company Calvert Cliffs Nuclear Power Plant cc:

Mr. William T. Bowen, President Regional Administrator, Region !

Calvert County Board of U.S. Nuclear Regulatory Comission

. Comissioners Office of Executive Director Prince Frederick, Maryland 20768 for Operations 631 Park Avenue D. A. Brune, Esq. King of Prussia, Pennysivania 19406 General Counsel Baltimore Gas and Electric Company Mr. Charles B. Brinkman P. O. Box 1475 Manager - Washington Nuclear Operations Baltimore, Maryland 21203 Combustion Engineering, Inc.

7910 Woodmont Avenue George F. Trowbridge, Esq. Bethesda, Maryland 20814 Shaw, Pittman, Potts and Trowbridge 1800. M Street, NW Mr. J. A. Tiernan, Manager Washington, DC 20036 Nuclear Power Department Calvert Cliffs Nuclear Power Plant Mr. R. C. L. Olson, Principal Engineer Maryland Routes 2 and 4 Nuclear Licensing Analysis Unit Lusby, Maryland 20657 Baltimore Gas and Electric Company Room 720 - G&E Building Mr. R. E. Denton, General Supervisor P. O. Box 1475 Training and Technical Services Baltimore, Maryland 21203 Calvert Cliffs Nuclear Power Plant Maryland Routes 2 and 4 Resident Inspector c/o U.S. Nuclear Regulatory Commission Combustion Engineering, Inc.

P. O. Box 437 ATTN: Mr. R. R. Mills, Manager Lusby, Maryland 20657 Engineering Services P. O. Box 500 Mr. Leon B. Russell Windsor, Connecticut 06095  !

Plant Superintendent l Calvert Cliffs Nuclear Power Plant Department of Natural Resources Maryland Routes 2 and 4 Energy Administration, Power Plent Lusby, Maryland 20657 Siting Program

• ATTN: Mr. T. Magette Bechtel Power Corporation Tawes State Office Building ATTN: Mr. D. E. Stewart Annapolis, Maryland 21204 Calvert Cliffs Project Engineer 15740 Shady Grove Road Gaithersburg, Maryland 20760 1 Mr. R. M. Douglass, Manager Quality Assurance Department l Baltimore Gas and Electric Company Fort Smallwood Road Complex P. O. Box 1475 Baltimore, Maryland 21203 l

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! NUCLEAR REGULATORY CChMMISSION

$ :I WASH mo T ON, 0. c. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REL ATING TO REQUEST FOR RELIEF FROM RADIOGRAPHIC AND VISUAL IN5PECTION OF REACTOR COOLANT PUMP CA5ING5 BALTIMORE GA5 AND ELECTRIC COMPANY CALVERT CLIFF 5 NUCLEAR POWER PLANT, UNIT NOS. 1 AND 2 DOCKET N05. 50-317 AND 50-318

Background

Section XI of the ASME Code requires examinations of reactor coolant pumps during each 10-year interval of plant operation. By letter dated February 4  ;

1985. Baltimore Gas and Electric Company submitted requests for relief from the requirements for Calvert Cliffs Units 1 and 2 and provided infomation in support of the requests. Pursuant to 10 CFR 50.55a(g)(6)(1), this information, together with supplemental infomation in BG8E's letters dated May 31, 1985 and June 24, 1985, was evaluated to determine if the requirement l is impractical to perfom on the component and if the necessary findings can  !

be made to grant relief as requested.

Relief Recuest ASME Code Section XI 1974 Edition with Addenda through Sumer 1975 ,

examination categories B-L-1 and B-L-2 require 100% volumetric, examination of casing welds and visual examination of the internal pressure bou'ndary .

surfaces of one pump casing in each of the pump groups perfoming similar system functions each inspection interval. The licensee has found this requirement to be impractical and has requested relief. Alternative examinations have been proposed.

Code Class Current ISI Class: Class 1. ,

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Function Each Calvert Cliffsiknit has four reactor coolant pumps which are welded to the 30" recirculation loop. fhese pumps function during normal reactor operation to provide forced recirculation through the core. All pumps are identical in design and function and are Byron-Jackson Type DFSS.

Licensee Basis for Relief Request A. The design configuration of the pump corresponds to a Type E pump illustrated in Figure NB-3442.5-1 (1977 Edition, ASME Code Section 111).

No practical technique currently exists to perform Inservice Inspection Radiographic Examination (RT) or Ultrasonic Examination (UT) cf this pump type, o B. The presence of the diffuser vanes precludes conventional RT. The vanes U prevent placement of the RT film cassettes inside the pump (as does the radiation field in terms of radiographic film and personnel radiation exposure). Placement of the film on the outside of the pump is feasible,

- but there is no radiographic source suitable for placement inside the pump. Standard isotopic radiation sources are too weak to penetrate the thick casting and background radiation from the inside surface of the pump would diminish sensitivity. Special strong isotopic sources would be impractical to handle and position inside the pump due to personnel radiological exposure from the radiographic source itself.

The recently developed Miniature Linear Accelera.or (MINAC) was con-sidered, but the Type E pump design precludes positioning of the accelerator inside the pump. Double wall radiography utilizing the MINAC has also been considered with some hope of attaining meaningful radiographs of a portion of the casing welds. This technique has.not been qualified to date and appears to be some time off, if at all possible.

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U C. The coarse grain structure inherent in thick stainless steel castings precludes the use of conventional UT. Future developments in. ultra-sonic techniques may provide a method to examine thick stainless steel casting and,if developed, this would be preferred over the

. difficulties and dangers of thick wall radiography. We are hopeful that the Ultrasonic Data Recording and Processing System (UDRPS) technology may provide some breakthrough in stainless steel casting UT. ,

D. The pump casing is fabricated from cast stainless steel (ASTM A351, Grade CF8M). The material is essentially a cast-type 316 stainless steel. This material is widely used in the nuclear industry and no 1 industry failures of this type material in reactor coolant pumps have been noted. The presence of delta ferrite (typically 15% or more) imparts increased resistance to intergranular stress corrosion cracking (IGSCC). The delta ferrite also improves resistance to pitting corrosion.

E. Report Number ERP-06-102, Revision 0, August 1983, prepared for the Electric Power Research Institute by NUTECH Engineers, Incorporated, l

concludes that:

1. Basedonthegenericpumpcasinganalysis,thereis'jostification for the extension of the pump-casing examination up to 15 years.

2. Plant-unique analysis will show greater margins of safety.

3. The tearing modulus analysis shows that large, final flaw sizes can be tolerated in the pump casing before fracture is pre-dicted.

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4. The recent 10 year Inservice Inspection of several pump casings (Type F) indicates no detectable flaw growth from base line inspections, which-corroborates the above analytical conclusion.

F. Pump disassembly for the sole purpose of conducting a very limited visual examination of the interior pressure boundary surfaces of a reactor coolant pump is fruitless, particularly in light of the manhours and radiation exposure that would be expended. The pump has an as-cast surface texture for the most part.

G. Over 1,000 manhours and over 50 person tem are estimated to disassemble, visually inspect, and reassemble one reactor coolant pump. The manhour estimate is based only on on-site outage work performed by Maintenance, Operations, and Nondestructive Testing personnel. The estimate does not include engineering time or pre-outagejobplanning. Additionally, manhours and person rem will be expended by Radiation Protection personnel providing direct coverage. The time required to perform the disassembly and ir.5pu tion

, would be approximately 2 weeks of critical path time. Most of the work would be performed under full face mask conditions.

Alternate Examinations Proposed by Licensee ,

A. The pump interior will be inspected to the extent practical (in recognition of the vanes therein) should the pump be disassembled for any other reason.

B. The reactor coolant pumps shall be hydrostatically tested per the requirements of ASME Code Section XI.

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l C. A surface examination of one RCP in each unit shall be performed on the exterior casing weld surface areas by the liquid penetrant method. Also, the pump selected shall receive a 100% visual examina- l tion of the exterior pump case surfaces.  !

The proposed additional exa'minations will identify flaws that may have propagated or originated at the pump outer surface since preservice examination. Since the coa scceptance standards for allowable sur-face flaw indication length is significantly less than that allowed for a subsurface flaw, the pump surfaces represent the more critical site for flaw location.

Staff Evaluation and Conclusion The need for this relief was recognized during the initial Inservice O Inspection program development. At that time the NRC Resident Inspector requested that the relief request submittal be delayed in hope that techniques might be developed and qualified by the end of the first 10 year interval. It is now apparent that no such technique applicable

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to the pumps will be available before the first interval concludes.

During operation, the condition of the pumps is monitored for abnormalities.

Each RCP has vibration monitoring instrumentation. The reac' tor coolant flow is monitored and displayed in the control room. When flow is reduced to 95% >

of design, the reactor is automatically tripped. The Reactor Coolant System is monitored for impact due to loose parts or foreign objects. In addition to the above, the reactor coolant pump's motor current is monitored. The RCP motors also have high vibration alarms.

Considering the pump design, materials of construction of the pump casing,  :

and the radiation levels associated with performing the required examinatio.:.

the staff finds the examinations impractical to perform. In lieu of the i

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O volumetric examination of the pump casing weld and visual inspection of the internal surfaces, the licensee has comitted to perform a surface examination of the welds. In addition to the surface examination, a visual inspection of )

the casing exterior surface will be perfonned d[rity the hydrostatic test of f the reactor coolant system. In the event that the pump has to be disassembled for operational or maintenance purposes, the required visual inspection of the internal surfaces will be performed.

We conclude that conducting a 100% volumetric examination of pump casing welds is impractical. Moreover, the alternate surface and visual examinations which will be perfonned on the pump casing will provide adequate assurance of its structural integrity and therefore relief from the volumetric examination of the casing weld and visual inspection of the internal surfaces may be granted.

Therefore,inaccordancewith10CFR50.55a(g)(6)(1),wefindtherelief requested may be granted. The relief is authorized by law and will not endanger life or property or the comon defense and security and is otherwise in the public interest giving due consideration to .he burden upon the

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licensee that could result if the requirements were imposed on the facility. ,

Principal Contributor:  ;

B. Turov11n, DE D. Jaffe, DL Date: September 18, 1985 l

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• UNITED STATES 9,,

NUCLEAR REGULATORY COMMISSION

[ - W ASHINGTON, D. C. 20555 5

\q / Novenber 6,1985

< IEM3.Elil"' 3"ICAL SERVICES ir >:-. ll-l&

3= LN:ma y Docket Nos. 50-317 and 50-318 AGS-TRO _ ' .-

W.'.1' TO CLA y Wy _ .Q !'c.'2 ICQ'D PE-FCM Mr. A. E. Lundvall, Jr. PE-IFM pottog._gp Vice President - Supply FILE: 60. & 1, o A Baltimore Gas & Electric Company P. O. Box 1475 , , _ m Baltimore, Maryland 21203 j:9 ,r ,h, rarm'Ar g;qv,r#

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Dear Mr. Lundvall:

Your letter dated October 15, 1985 states that the reactor coolant pumps will be examined in accordance with ASME Code Section XI 1974 edition with Addenda through 1975, Article IWA-5240, during the hydrostatic test. This inspection, together with the visual inspection of the pump casing of one pump per unit (in accordance with IWA-2210) is herein interpreted as sufficient to meet the "visual" inspection requirements for the reactor coolant pump O

casings as contained in our safety evaluation dated September 18, 1985 concerning relief from certain ASME Boiler and Pressure Vessel Code Requirements Sincerely.

Edward J. Butcher, Acting Chief Operating Reactors Branch No. 3 Division of Licensing ,

cc: See next page O

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Mr. A. E. Lundvall, Jr.  ;

Baltimore Gas & Electric Company Calvert Cliffs Nuclear Power Plant cc:

Mr. William T. Bowen, President Regional Administrator, Region I Calvert County Board of U.S. Nuclear Regulatory Comission Commissioners Office of Executive Director Prince Frederick, Maryland 20768 for 0 mrations 631 Par ( Avenue D. A. Brune, Esq. King of Prussia, Pennys1vania 19406 General Counsel Baltimore Gas and Electric Company Mr. Charles B. Brinkman P. O. Box 1475 Manager - Washington Nuclear Operations Baltimore, Maryland 21203 Combustion Engineering. Inc.

7910 Woodmont Avenue George F. Trowbridge, Esq. Bethesda, Maryland 20814 Shaw, Pittman, Potts and Trowbridge 1800 M Street, NW Mr. J. A. Tiernan, Manager Washington, DC 20036 Nuclear Power Department Calvert Cliffs Nuclear Power Plant Mr. R. C. L. Olson, Principal Engineer Maryland Routes 2 and 4 ,

Nuclear Licensing Analysis Unit Lusby, Maryland 20657 Baltimore Gas and Electric Company

- Room 720 - G&E Building Mr. R. E. Denton, General Supervisor P. O. Box 1475 Training and Technical Services Baltimore, Maryland 21203 Calvert Cliffs Nuclear Power Plant Maryland Routes 2 and 4 O Resident Inspector Lusby, Maryland 20657 c/o U.S. Nuclear Regulatory Comission Combustion Engineering, Inc.

P. O. Box 437 ATTN: Mr. R. R. Mills, Manager Lusby, Maryland 20657 Engineering Services P. O. Box 500 Mr. Leon B. Russell Windsor, Connecticut 06095 '

Plant Superintendent Calvert Cliffs Nuclear Power Plant Department of Natural Resources Maryland Routes 2 and 4 Energy Administration Power Plant Lusby, Maryland 20657 Siting Program ' -

ATTN: Mr. T. Magette Bechtel Power Corporation Tawes State Office Building i ATTN: Mr. D. E. Stewart Annapolis, Maryland 21204 Calvert Cliffs Project Engineer 15740 Shady Grove Road Gaithersburg, Maryland 20760 Mr. R. M. Douglass Manager Quality Assurance Department Baltimore Gas and Electric Company Fort Smallwood Road Complex P. O. Box 1475 '

Baltimore, Maryland 21203 O

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O ATTACIMENT (3)

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O AITACIIMENT (4) i LETTER FROM Tile NRC TO l 1

l BALTIMORE CAS & ELECTRIC COMPANY l

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May 11, 1987 4

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• UNITED STATES NUCLEAR REGULATORY COMMISSION

, . WA$mOTON, D. C. 20$55 )

O [.D(/.... May 11, 1987 Docket Nos. 50-317 ,

and 50-318 Mr. J. A. Tiernan -

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Vice President-Nuclear Energy Baltimore Gas & Electric Company P. O. Box 1d75 Baltimor,e, MD 21203

SUBJECT:

RELIEF FROM THE 1974 ASME CFDE SECTION XI REQUIREMENTS FOR CLASS 1 AND 2 BOLTING AND CONTROL R0D DRIVE HOUSINGS The Commission staff has completed their review of your request for ASME Code update and relief as provided in your submittal dated October 2,1986. This submittal requested that the Section XI reouf rements of the ASME Code be updated from the 1974 Edition to the 1977 Edition or later approved editions for Class 1 and 2 boit ing and for control rod drive housings. These inservice inspections were performed based upon the 1977 ASME Code requirements during the Fall 1986 Unit I refueling outage, s

Your supplemental letter of December 4,1986, requested Comission approval J of an alternative examination method for the control rod drive housings from that described in IVB-2600,Section XI of the 1977 ASME Code as detemined that these reouirements were impractical to perform (you had the as were 1974 ASME Code requirements) for the control rod drive housings.

The 1977 ASME Code requires that 100% of the welds on 10% of the peripheral control rod drive housings be examined. This submittal stated that 100% of the welds head could not be examined as these welds extended into the reactor vessel itself. Instead, you proposed to perform an equivalent, alternative examination by inspecting 75% of the welds on three control rod drive housings and 50% of the welds on two control rod drive housings.

In accordance with the provisions of 10 CFR 50.55a(g)(4)(iv), the staff has detemined that your requests for ASME Code update of the Section XI requirements for Class 1 and 2 bolting and for the control rod drive housings is acceptable and this partial ASME Code update is hereby approved.

In additien, the staff has reviewed the requirements to examine 100% of the welds on three peripheral control rod drive housings and has determined this requirenent to be impractical due to the physical configuration of these housings and the reactor vessel head itself. The Comission finds that the alternative examination method proposed in your December 4,1986 submittal is '

acceptable continoent upon your performance of 100% weld examinations on the peripheral control rod drive housings at times when this examination is .

i physically feasible (e.g., if the control rod drive housing was physically O renoved from the reactor vessel head). This contingency shall be applicable to all future 10-year inservice inspection intervals for each unit.

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t The Conmission hereby grants this relief from the weld examination requirements for the peripheral control rod drive housings pursuant to 10 CFR 50.55a(g)(6)(1) and finds that this relief is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

Our Safety Evaluation is erclosed.

Sincetely, U 0 - [W Robert A. Capra, Acting Director Project Directorate I-1 Division of Re:.ctor Projects, I/II

Enclosure:

Safety Evaluation cc w/ enclosure:

See next page s.

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Mr. J. A. Tiernan Baltimore 4as & Electric. Company Calvert Cliffs Nuclear Power Plant cc: -

Mr. William T. Bowen, President Regional Administrator, Region I Calvert County Board of U.S. Nuclear Regulatory Commission Commissioners Office of Executive Director Prince Frederick, Maryland 20768 for Operations 631 Park Avenue -

D. A. Brune, Esq. King of Prussia, Pennys1vania 19406 General Counsel Baltimore Gas and Electric Company P. O. Box 1475 Baltimore, Maryland 21203 Jay E. Silberg Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.W.

Washington, DC 20037 Mr. M. E. Sowman, General Supervisor Technical Services Engineering Calvert Cliffs Nuclear Power Plant MD Rts 2 & 4, P. O. Box 1535 Lusby, Maryland 20657-0073 Resident Inspector i c/o U.S. Nuclear Regulatory Commission j P. O. Box 437 O- Lusby, Maryland 20657-0073 Bechtel Power Corporation ATTN: Mr. D. E. Stewart Calvert Cliffs Project Engineer 15740 Shady Grove Road Gaithersburg, Maryland 20760 Combustion Engineering, Inc.

ATTN: Mr. W. R. Horlacher, III Project Manager P. O. Box 500 ,

1000 Prospect Hill Road I Windsor, Connecticut 06095-0500 -

Department of Natural Resources Energy Administration, Power Plant Siting Program ATTN: Mr. T. Magette Tawes State Office Building Annapolis, Maryland 21204 O

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&3 Ih 1 -[ t, E UNITED STATES NUCLEAB REGULATORY COMMISSION WASHINGTON, D. C. 20555 AQ'e LA..'CQrl  ;.. .

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION BALTIMOFE GA5 & ELECTRIC COMPANY CALVERT CLIFF 5 NUCLEAR POWER PLANT, UNITS 1 AND 2 DOCKET NOS. 50-317 AND 318 RELIEF FROM INSERVICE INSPECTION REOUIREMENTS OF '

SECTION XI 0F THE ASME CODE

1.0 INTRODUCTION

The Technical Specifications for the Calvert Cliffs Units 1 and 2 require that inservice examination of ASME Code Class 1, 2 and 3 components shall

- be perfomed in accordance with Section XI of the ASME Code as required by 10 CFR 50.55a(g)(4) except where specific written relief has been granted by the Commission. Some plants were designed in confomance to early editions of this Code Section, consequently certain requirements of later editions and addenda of Section XI are impractical to perform because of the plant's design, component geometry, and material of construction. Paragraph 10 CFR 50.55a(g)(6)(i) authorizes the Commission to grant relief from those requirements upon making the necessary findings.

In a letter dated October 2, 1986, as supplemented December 4, 1986, the Baltimore Gas & Electric Company (BG&E), the licensee, identified

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A specific ASME Code reouirements that BG&E detennined to be imoractical to perform at Calvert Cliffs and requested relief from these requirements.

The staff has evaluated the licensee's supporting technical justification and finds it to be acceptable.

2.0 EVALUATION OF RELIEF RE0 VESTS The licensee requested relief from specific inservice inspection (ISI) requirements and provided supporting technical infomation. The staff

. reviewed this information as related to the existing design, geometry and materials of construction of the components.

A. . Relief Request No.1, Examination Categories B-G-1, B-G-2 and C-D, ASME Code Class 1 and 2 Bolting.

Code Recuirements: ASME Section XI, 1974 Edition including Addenda through Suniner 1975, requires the following:

1. Class 1 Bolting (a) B-G-1 Volumetric examination is required on pressure-retaining bolting that is 2 inches and larger in diameter.

(b) B-G-2 Visual examination is required for pressure-retaining bolting that is smaller than 2 inches  !

ha in diameter, i

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/'N U 2. Class 2 Bolting: ,

C-D Visual and either surface or volumetric examina -

tions are required for pressure-retaining bolting exceeding 1 inch in diameter.

Code Relief Recuest: The licensee proposed to meet the require-ments of the 1977 Edition of Section XI and later editions and addenda of ASME Section XI, in which Examination Categories B-G-1, B-G-2 and C-D are redefined. Category B-G-1 is redefined as pressure-retaining Class 1 bolting, larger than 2 inches in diameter. Category B-G-2 is redefined as pressure-retaining Class 1 bolting, 2 inches and smaller in diameter. Bolting that is exactly 2 inches in diameter is shifted from Category B-G-1 to B-G-2. Similarly, Category C-D is redefined as pressure-retaining Class 2 bolting exceeding 2 inches in diameter. The licensee proposed to adopt the definitions set forth in the later editions and addenda of Section XI Code to define the boundaries for Categories B-G-1, B-G-2 and C-D.

Basis for Relief Later editions and addenda of the Section XI Code are approved for use, as per paragraph (g) of 10 CFR 50.55a of the Code of N Federal Regulations. Paragraph g(4)(iv) allows the adoption of portions of later approved editions and addenda to the Code provided that all related requirements of the respective editions and addenda are met. The licensee feels that the above stated adoptions are in compliance with the stated regulations.

Staff Evaluation Paragraph 10 CFR 50.55a(g)(4)(iv) states: "Inservice examinations of components, tests of pumps and valves, and system pressure tests, may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in paragraph (b) of this section, s cject to the limitations and modifications listed in paragraph (b) of this section and subject to Comission approval. Portions of editions or addenda may be used provided that all related requirements of the resoective editions or addenda are met." , .

The licensee intends to use provisions from the 1977 and later approved Section XI ASME Code editions and addenda. Even though the extent and method of examinations have been reduced, other licensees with ISI programs based on the later ASME Code documents are following these requirements pursuant to 10 CFR 50.55a(g)(41 The staff has detennined that the licensee's proposal confonns to the requirements of the regulation that "all related requirements O

1 1

1 of the respective editions or addenda are met." Therefore, the staff concludes that the licensee's proposal is acceptable.

B. Relief Request No. 2, Examination Category B-0, Peripheral Control Rod Drive Housings Code Reauirements: Article IWB-2600 of ASME Section XI, 1974 Edition incluoing Addenda through Sunrner 1975 requires a volumetric examination to include 100% of the welds in 10% of the peripheral

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control rod drive (CRD) housings during each inspection interval.

Code Relief Recuest: In the October 2, 1986 submittal, the licensee proposed to meet the requirements of Article IWB-2600 of the 1977 Edition of Section XI and later editions and addenda of the Section XI Code, which require surface or volumetric examination of 100% of the welds in 10% of the peripheral CR0 housings. The licensee proposed to perfom surface examinations, as per later editions of the code, rather than volumetric examinations.

On December 4,1986, the licensee modified this relief request after detemining that these requirements were impractical due to difficulties experienced in the perfomance of the surface examinations on the Unit 1 peripheral CRD housings.

h' There are 28 peripheral CRD housings in the installed configuration.

After removal of the reactor vessel (RV) head shroud and insulation, the licensee attempted to inspect 100% of the welds on three peripheral CRD housings and detemined that only part of the CRD housing welds could be examined as the welds extend into the RV head itself.

An alternative CR0 housing surface examination was conducted by inspecting 75% of the welds on three CRD housings and 50'; of the welds on two CRD housings.

Basis for Relief (1) Later editions and addenda of the Section XI Code are approved for use, as per paragraph (g) of 10 CFR 50.55a of the Code of Federal Regulations. Paragraph g(4)(iv) allows the adoption of portions of later approved editions and addenda to the code provided that all related requirements of the respective editions and addenda are met. The licensee feels'that the above stated adoptions are in compliance with the stated regulations.

. O

4 (2) Volumetric examination of these welds is impractical due to design configuration, accessibility and materials of construction as described in CE Drawing No. 233-412, Weld Details, which has been provided to the staff.

(a) Ultrasonic examina' tion will not provide meaningful results due to the geometeric configuration of the joint and mater-ial properties (inconel-to-stainless steel welds).

~

(b) Radiographic examination cannot be performed due to the design configuration and accessibility.

(3) Proposed alternatives to the requirements of Section XI of the ASME Code which are determined to be impractical, may be permitted by 10 CFR 50.55a(g)(6)(1) if the proposed alternatives are detennined to be authorized by law and will not endanger life or property or the comon defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

Staff Evaluation The licensee's letter of October 2,1986 proposes to use provisions from the 1977 and later approved Section XI ASME Code editions and addenda. The staff has determined that this r posal does conform with the requirements of 10 CFR 50.55a(g)(4)( v that "portions of editions or addenda may be used provided that all related requirements of the editions or addenda are met."

In addition, the staff has determined that the requirement to examine 100% of the welds on three peripheral CRD housings is impractical due to the physical constraints of the installed configuration of the housings and the reactor vessel head. The alternative examination method as proposed in the licensee's December 4,1986 submittal has been detennined to be acceptable pursuant to 10 CFR 50.55a(g)(6)(i) and contingent upon the licensee's performance of 100% weld examinations on the peripheral CRD housings in the event this examination is feasible (e.g., if the CRD housing is physically removed from the reactor vessel head). This contingency shall be applicable to all future 10-year inservice inspection intervals for each unit.

3.0 CONCLUSION

The staff has completed the review of the licensee's letters dated October 2, 1986 and December 4,1986 based on the provisions of 10 CFR 50.55a(gM6)(i).

The staff concludes that the licensee's proposal to update to the 1977 Section XI requirements of the ASME Code for Class 1 and 2 boltino and CR0 housings meets the provisions of 10 CFR 50.55a(g)(4)(iv), is acceptable, O and therefore, the licensee shall be granted relief to update to the requirements of the 1977 and later editions and addenda of Section XI of the ASME Code for the Class 1 and 2 bolting and the CRD housing

e 1

O examinations. In addition, the staff has determined that the*1977 ASME Code (and 1974 ASME Code, too) requirement to examine 100% of the welds on three peripheral CR0 housings is impractical and that this relief shall be granted contingent upon the performance of these 100% CRD housing weld inspections when physically possible. This granting of relief is authorized .

by law and will not endanger life or property or the connon defense and security and is otherwise in the public interest, considering the burden that could result if these ASME Code Section XI requirements for the CRD housing examinations were imposed on the facility. Therefore, in accordance witK 10 CFR 50.55a(g)(6)(1), this relief is granted.

Date: May 11, 1987 Principal Contributors M. Hum, S. McNeil i.

O

O ATTACIMENT (5)

CALVERT CLIFFS UNIT 1 SKETCll 1 SAFETY INJECTION SYSTEM O

and CALVERT CLIFFS UNIT 2 SKETCll 2 SAFETY INJECTION SYSTEM O

O O O

, CLASS 1 CLASS 2 m p=-CC-6-1002 1-SI-617-Nov I

' Aux w sI SI TANK 1-sI-215 3 -CC-13-1014 2*-CC-13-iOO5 " 1-sI-616-NOV 2,"-CC 13-1019 HPsI

( 1-sI-217 1-sI-st4e 1-sI-118 1-5I-114 g_st.61'Her 4 'd '/ 6'-CC-13-iO01

'd WI 1-sI-627-n0V 2'-CC-6-1004 Aux r sI SI TANK t-sI-2s 3 ~

1-sI-62tHer 2"-CC-13-iOO6 2' 1018

$ 1-sI-227 1-sI-624-Nov i-sI-128 1-sI-124 M = 'A '/ 6*-CC-13-iOO2

'A  %@ wsI 1-sI-62!Her w r

~

5 2*-0C-6-1005 @ i-sI-637e E! SI TANK  % Aux r sI 8 1-sI-235 3'-CC-13-iO21 2'-CC-13-iOO7 cc 2'-CC-13-1016

- 1~8I-133

%@ wsI 1-SI N 1-SI-237 1-sI-634-MOV g_gy_ g g gg

{ ' LPsI g 4 'd '/ 6"-CC-13-iOO3 cr 2"-CC-6-1006 1-SI-647-MOV SI TAE aux y sI 1-sI-26 3'-CC-13-iO20 14I "

2'-CC-13-1008 -

msI 2'-CC-13-1017 .

-1 I-143 1-sI-247 1-sI-644-ier g_gy_gg i-SI,i44

1. LPsI k 6. CC-13-iOO4 TEST ERIIEENT (I2-5000) m CLASS i CLASS 2 m TEST EQUIEENT (IWC-5000) 1.02 X 1001 OP PESSLE AT 500' F 1.25 X DESIGN PESSLK AT 100' F CALVERT CLIFFS UNIT i SKETCH i

O O O

, CLASS 1 CLASS 2 - 2'-CC-6-2002 @ 2-sI-617-NDv i SI TANK N Aux r sI M -215 4 2-sI41m 2'-CC-13-2005 sameen - - 2*- C-13-2 9

- 2~SI~if3 I 2-sI-614-nov 2-sI-217 @ os m M-m mm 4 N / 6*-CC-13-2001

'M WI 2-sI-627-Mov 2'-CC-6-2004 Aux msI SI TANK 2 M 6 M 2'-CC-13-2018 rs M 2-sI-227 2-sI-62 m 2-sI-12s 2-sI-124 2-sI-625-Nov

/ '/ 6'-CC-13-2002 M LPs1 2-sI-637e 5 2'-CC-6-2005 Aux rsI dN SI TANK 3'-CC-13-2021 8 2-sI-235 2'-CC-13-2007 @_ 2-sI-636-Nov cc 2'-CC-13-2016

- SI-133 N rsI 2-SI-237 2-sI-634-MDv 2-s -134

{ 2-SI-138 p 2-SI-635-Mov

< < M '/

N LPSI E 6"-CC-13-2003 2'-CC-6-2006 2-sI-60-MOV SI TANK Aux e sI 2-sI-245 3'-CC-13-2020 2*-CC-13-2008 2-51 6 2*-CC-13-2017 t wsI 2-sI-247 2-sI-64 N 2-SI-148 2-sI-144 2-st e k

4 V 6. CC-13-2004 LPSI TEST EQUIRBENT (I2-5000) m CLASS i CLASS 2 m TEST EDUIROGE (IWC-5000) 1.02 X 1001 OP PEsSLNE AT 500' F 1.25 X DESIGN PEsSLNE AT 100' F CALVERT CLIFFS UNIT 2 SKETCH 2

l l

lO ATTACHMENT (6)

LETIER FROM THE NRC TO BALTIMORE CAS & ELECTRIC COMPANY O

November 14, 1985 0

f.0*J / j jpato Coh

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A( Eb*.* (d /sv y

47 (g UNITED STATES NUCLEAR REGULATORY COMMISSION g E- WASHINGTON. D. C. 20555 s.,...../ November 14, 1985

~

Docket Nos. 50-317 ff}.c.

D/d3 b and 50-318 Mr. A. E. .Lundvall, Jr.

Vice President - Supply l Baltimore Gas & Electric Company l

P. O. Box 1475 Baltimore, Maryland 21203 1

Dear Mr. Lundvall:

The NRC has provided relief from a requirement of the ASME Boiler and Pressure Vessel Code,Section XI, which BG&E has determined to be impractical in accordance with your application dated August 28, 1985. l The code relief, granted in accordance with 10 CFR Part 50, Section 50.55a(g)(6)(1), l '

relates to the requirement for pressure testin with the High Pressure Safety Injection, Auxiliary (HPSI)g of Class HPSI, and Low2 piping Pres- associ sureSafetyInjection(LPSI)LoopIsolationHOVstotheReactorCoolantSystem.

You have proposed acceptable, alternate testing as described in the enclosed I

Safety Evaluation.

We have determined that the testing for which relief has been requested and approved is impractical and pursuant to 10 CFR 50.55a(g)(6)(1), that the granting of this relief is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest. In making this determination, we have given due consideration to the burden that could result if these requirements were imposed on your facility.

1 Sincerely,

• ~

f Harold R. Denton, Director Office of Nuclear Reactor Regulation

Enclosure:

Safety Evaluation g . .

_.....I?m,,gy,77ggg cc w/ enclosure See next page m _ n . , ,,, ,

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/ Pr-7C:1 pz_Im vn e i_w FILE: _.

$A.lk

Mr. A. E. Lundvall, Jr.

Baltimore Gas & Electric Company ,Calvert Cliffs Nuclear Power Plant CC:

Mr. William T. Bowen, President Regional Administrator, Region I l Calvert County Board of U.S. Nuclear Regulatory Comission l Commissioners Office of Executive Director  :

for Operations I Prince Frederick, Maryland 20768 631 Park Avenue i D. A. Brun,e, Esq. King of Prussia, Pennysivania 19406 General Counsel Baltimore Gas and Electric Company Mr. Charles B. Brinkman P. O. Box 1475 Manager - Washington Nuclear Operations Baltimore, Maryland 21203 Combustion Engineering, Inc.

7910 Woodmont Avenue Bethesda, Maryland 20814 l George F. Trowbridge, Esq.

Shaw, Pittman, Potts and Trowbridge 1800 M Street, NW Mr. J. A. Tiernan, Manager Washington, DC 20036 Nuclear Power Department Calvert Cliffs Nuclear Power Plant i Mr. R. C. L. Olson, Principal Engineer Maryland Routes 2 and 4 l Nuclear Licensing Analysis Unit Lusby, Maryland 20657 Baltimore Gas and Electric Company Room 720 - G&E Building Mr R. E. Denton, General Supervisor P. O. Box 1475 ' Training and Technical Services 1 Baltimore, Maryland 21203 Calvert Cliffs Nuclear Power Plant j Maryland Routes 2 and 4 I (C] Lusby, Maryland 20657 l Resident inspector  ;

c/o U.S. Nuclear Regulatory Commission Combustion Engineering, Inc.

P. O. Box 437 ATTN: Mr. R. R. Mills, Manager ,

Lusby, Maryland 20657 Engineering Services 1 P. O. Box 500 l Mr. Leon B. Russell Windsor, Connecticut 06095  ;

Plant Superintendent (

Calvert Cliffs Nuclear Power Plant Department of Natural Resources Maryland Routes 2 and 4 Energy Administration, Power Plant Lusby, Maryland 20657 . Siting Program ATTN: Mr. T. Hagette Bechtel Power Corporation Tawes State Office Building l ATTN: Mr. D. E. Stewart Annapolis, Maryland 21204 Calvert Cliffs Project Engineer 15740 Shady Grove Road Gaithersburg, Maryland 20760 Mr. R. M. Douglass, Manager Quality Assurance Department Baltimore Gas and Electric Company Fort Smallwood Road Complex P. O. Box 1475 Baltimore, Maryland 21203 O

l

\

$a ato oq'o, UNITED STATES j NUCLEAR REGULATORY COMMISSION l WASHINGTON, D. C. 20555 5 j

% * +o O 1 1

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION _

RE0 VEST FOR RELIEF FROM INSERVICE PRESSURE TEST RE0VIREMENTS

. BALTIMORE GAS AND ELECTRIC CALVERT CLIFFS NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-317 AND 50-318 INTRODUCTION The Technical Specifications for the Calvert Cliffs Nuclear Power Plant Units 1 and 2 state that inservice examination of ASME B&PV Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission. The Examination Program is based upon the requirements of the 1974 Edition with the Addenda through the Sumer of 1975. Certain requirements of this Edition and Addenda of Section XI are impractical to perform on older plants because of the plants' design, component geometry, materials of construction or the need for extensive temporary modifications and the resultant substantial radiation exposure to p1 ant personnei.

By letter dated August 28,1985, the 8sitimore Electric Company requested relief from the pressure test inspectico requirements of the Code for sections of pipes determined to be impractical to perform these tests.

Requests for Relief Relief is requested for Class 2 piping from the High Pressure Safety Injection (HPSI), Auxiliary HPSI,' and Low Pressure Safety Inspection (LPSI) Loop Isolation MOVs to the Reactor Coolant System (RCS).

The following lines are affected:

j

1. Y

i O -

Ci O~

1 UNIT 1 UNIT 2 .

FROM TO LINE N05. FROM TO -

LINE NOS.

1-51-118 1-SI-615-M0V 6"CC-13-1001 2-51-118 2-615-MOV 6"CC-13-2001 1-SI-616-MOV 2"CC-13-1019 2-SI-616-MOV 2"CC-13-2019 1-SI-617-MOV 3"CC-13-1014 2-SI-617-MOV 3"CC-13-2014 2"CC-13-1005 2"CC-13-2005 2"CC-6-1002 2"CC-6-2002 1-51-128 1-SI-625-MOV 6"CC-13-1002 2-51-128 2-SI-625-MOV 6"CC-13-2002 1-SI-626-MOV 2"CC-13-1018 2-SI-626-MOV 2"CC-13-2018 1-SI-627-MOV . 3"CC-13-1015 2-SI-627-MOV 3"CC-13-2015 2"CC-13-1006 2"CC-13-2006 ,

2"CC-6-1004 2"CC-6-2004 L

1-51-138 1-SI-635-MOV 6"CC-13-1003 2-51-138 2-SI-635-MOV 6"CC-13-2003 1-SI-636-MOV 2"CC-13-1016 2-SI-636-MOV 2"CC-13-2016 1-SI-637-MOV 3"CC-13-1021 2-SI-637-MOV 3"CC-13-2021 2"CC-13-1007 2"CC-13-2007 2"CC-6-1005 2"CC-6-2005 l

1-51-148 1-SI-645-MOV 6"CC-13-1004 2-SI-148 2-SI-645-MOV 6"CC-13-2004 1-SI-646-MOV ' 2"CC-13-1017 2-SI-646-MOV 2"CC-13-2017 1-SI-647-M0V 3"CC-13-1020 2-SI-647-MOV 3"CC-13-2020

! 2"CC-13-1008 2"CC-13-2008 2"CC-6-1006 2"CC-6-2006

3 J ISI Code Class 2 Requirements ASME Co'de Section XI requires hydrostatic testing of all Class 2 piping and components as set forth in Articles IWA-5000 and IWC-5000. The test pressure requirement for Class 2 piping and components is 1.25 times the design pressure when tested at a temperature not less than 100*F.

Basis for Relief Request A. The listed portion of Class 2 piping from HPSI, Aux. HPSI, and LPSI -

Loop Isolation MOV to RCS cannot be isolated from the RCS.

Licensee's Proposed Alternative Tests The licensee proposes to perform a hydrostatic pressure test of the above

\ listed piping, excluding the piping listed in the relief request Item B.

below, to the pressure test requirements of IWB-5000 for Class 1 piping.

This piping can be pressurized via alignment of the charging system to the Aux. HPSI header.

. B. The portions of piping listed below cannot be hydrostatically tested due

-to inability to align charging pumps to pressurize this piping and the on .oility requirements of these portions when the RCS is pressurized.

UNIT 1 FROM 0 LINE NO. l 1-51-114 1-SI-615-MOV 6"CC-13-1001 -

1-51-124 1-SI-625-H0V 6"CC-13-1002 1-51-134 1-SI-635-MOV 6"CC-13-1003  ;

q 1-51-144 1-SI-645'-MOV 6"CC-13-1004 l

V 1 l

l _

1 UNIT 2 FROM T0 0 LINE NO.

l 2-S1-114 - 2-SI-615-MOV 6"CC-13-2001 2-51-124 2-SI-625-MOV 6"CC-13-2002 2-51-134 2-SI-635-MOV 6"CC-13-2003 2-SI-144 2-SI-645-MOV 6"CC-13-2004  ;

i i

Licensee's Proposed Alternate Tests l The following tests and examinations are proposed in lieu of hydrostatic '

testing for proving.the integrity of the piping listed in Item B. above.

1. Each refueling cycle, a leakage test of this piping is performed in accordance with Technical Specification 6.14. In this test the piping listed in Item B. is pressurized to LPSI pump discharge pressure and a visual examination for leakage is conducted.

2. Welds will be selected and examined per Section XI, Article IWC-2000.

Evaluation -

u.

The section of piping upstream of check valves SI-118,-128,-138,-148, for Units 1 and 2, cannot be tested at a pressure of 1.25 times design pressure l without making extensive temporary modifications to keep the valves closed.

The modifications would require: (1) disassembly of the valves, (2), welding ,

of temporary blocks (on the downstream side) inside the valve bodies to hold )

a "jack screw" type arrangement to keep the valve closed, (3) removal of the l temporaryblockingdevicesfromthevalvesaftertestingand(4) performing necessary nondestructive. testing to assure the integrity of the valve bodies O

I...

O before returning them to service. Without the. temporary modifications to the l check valves, the Class 1 system downstream of the check valves would be )

pressurized to the test pressure of the Class 2 system. This pressure exceeds the Class 1 hydrostatic pressure requirements.

Conclusion l Based upon a review of the system design, the basis for relief request, and the ,

licensee's proposed alternate tests, the staff concludes that relief granted l from examination and pressure test requirements and alternate methods imposed through this document give reasonable assurance of the piping pressure boundary integrity, that granting relief where the Code requirements are impractical is authorized by law and will not endanger life or property, or the comon defense

)

and security, and is otherwise in the public interest considering the burden that could result if they were imposed on the facility. Therefore, in accordance with 10 CFR 50.55a (g)(6)(1), relief is granted.

Principal Contributor:

8. Turovlin  !

Date: November 14, 1985

O ATTACHMENT (7) l I

1 l

LETTERS FROM Tile NRC l

TO 1 l

I BALTIMORE GAS & ELECTRIC COMPANY O l December 13, 1982 January 24, 1983 0

~

Q u

'. !j;p*"c o Uf41TED STATES g NUCLEAR REGULATORY COMMISSION

.i .l,{ . . ,. J WASHINGTON. D. C. 20555 O N.-[,, ...([ g c ., i:

^'"',# DEC 131982

-.r . .

.s . ... -. ~ _, _ ,

Docket tios. 50-317 and 50-318 '

I Mr. A. E. Lundvall, Jr.

Vice President - Supply . I Baltimore Gas & Electric Company

. P. O. Box 1475-Baltimore, Maryland 21203 ,

Dear P.r. Lundvall:

The !!RC has completed its review of the Inservice ' Inspection (ISI) Program for Calvert Cliffs Units 1 and 2. In the course of our review,.:e have granted relief relating to the following ASME Boiler and Pressure Vessel '

. Code,Section XI requirements: ,

'e Inspection of the Seal Weld in the Closure Head,

-e Inspection of the Primary tiozzle-to-Vessel Welds and the Nozzle Inside Radiused Section, o Inspection of the Reactor Ve.ssil Cladding, e Repair of an Arc Strike, Class 2 Pipe, e ' Pressure Test Hold Time, ,

e . Class 3 System Pressure Tests, e InserviceLeakTests(HydrostaticTesting)'fortheSaltWaterCooling - -

)

. System and Service W,ater Main, Headers.'- ,

e Ultrasonic Examination Techniques, o Repair and Hydrostatic Testing for Small Steam and Feedwater Piping, Class 2.

In the course of reviewing these requests for relief, we have found that changes have been needed in these requests to meet our requirements. These changes have been discussed with and agreed to by your staff.

The above requests for relief were submitted pursuant to 10 CFR Part 50, O sec4easossets)(s)(4v)4" '4cet4ome d ted oece=8er s. 1978 serch 29

, , . , , . ~ _ _ . - . . . , . _ _ _ . .. -. _ _ _ . _ _ . _ y __ w ,

s -

Mr. A. E. Lundvall, Jr. .- 2-O -

1980, November-19,7980'arrd Raf 29,'1981. 7heseTequests for relief are - -

herein granted per 10 CFR Part 50, Section 50.55a(g)(6)(i).

Copies of the Safety Evaluation and Federal Register Notice are enclosed.

Sincerely,

[.

Robert A. Clark, Chief

~

Operating Reactors Branch #3 '

Division of Licensing

Enclosures:

As stated cc: See next page O ..

' , \' ,

y e

~

4 8 5

9 e

0 -

4

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

'~ Baltim' ore Gas and Electric Company i

~

.es A. Biddison, Jr. Ms. Mary Harrison, President veneral Counsel .

Calvert County Board of County Commissioners Baltimore Gas and Elec.tric Company- .

Prince-Fredwick, MD 20768 .

P. O. Box 1475 Sc1timore, MD 21203 U. S. Environmental Protection Agency ,

Region III Office George F. Trowbridge, Esquire Attn: Regional Radiation Representative Shaw, Pittman, Potts and Trowbridge Curtis Building (Sixth Floor) 1800 M Street, N. W. Sixth and Walnut Streets Washington, D. C. 20036 Philadelphia, PA 19106 Mr. R. C. L. Olson, Principal Engineer Mr. Ralph E. Architzel Nuclear Licensing Analysis Unit R.esident Reactor Inspector Baltimore Gas and Electric C.ompany NRC Inspection and Enforcement Room 922 - G&E Building P. O. Bos 437 P. O. Box 1475 Lusby, MD 20.657 Baltimore, MD 21203 Mr. Charles Be Brinkman Mr. Leon S. Russell Manager - Washington Nuclear Operations Plant Superintendent Conbustion Engineering, Inc.

Calvert Cliffs Nuclear Power Plant 4853 Cordell Avenue, Suite A-1 Maryland Routes 2 & 4 Bethesda, MD 20014

. Lusby, MD 20657 Mr. J. A. Tiernan, Manager

htel Power Corporation Nuclear Power Department pdtn:

s Mr. .J. C. Ventura Calvert Cliffs Nuclear Powe'r Plant Caivert Cliffs Project Engineer Maryland Routes 2 & 4 15740 Shady Grove Road Lusby, MD 20657 Gaithersburg, MD 20760 -

Mr. W. J. Lippold, Supervisor Combustion Engineering, Inc. Nuclear Fuel Management 4ttn: Mr. P. W. Kruse, Manager Baltimore Gas and Electric Company ,'~

Engineering Services Calvert Cliffs Nuclear Power Plant P. O. Box 500 P. O. Box 1475

' Windsor, CT 06095 ' Bal timor~e, . Maryland 21203 Director,' Department of State Planning Mr. R. E. Denton, General Supervisor Training & Technical Services

• 6 30r We'st Preston Street Baltimore, MD 21201 Calvert Cliffs Nuclear Power Plant Maryland Routes 2 & 4 Lusby, MD 20657 Mr. R. M. Douglass, Manager Quality Assurance Department -

Baltimore Gas & Electric Company Administrator, Power Plant Siting Program .

Fort Smallwood Road Complex Energy and Coastal Zone Administration P. 0.' Box 1475 Department of Natural Resources tinore, MD 21203 Tawes State Office Building Annapolis, m 21204 -

r. S. M. Davis, General Supervisor ,

Operations Quality Assurance Calvert Cliffs Nuclear Power Plant Regional Administra' tor Maryland Routes 2 & 4 Nuclear Regulatory Commission, Region I i.usby,MD 20657 - Office of Executive Director for Operations 631 Park Avenue

. [

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~g UNITED STATES NUCLEAR REGULATORY COMMISSION

$. . g

.,. ,; WASHINGTON, D. C. 20555

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O ..... / ~

~ -

SAFETY' EMUATION' BY THE OFFICE'0F HUCCEAR REACTOR REGULATION RELATED TO REQUESTS FOR RELIEF FROM INS'ERVICE INSPECTION REQUIREMENTS BALTIMORE GAS AND ELECTRIC COMPANY CALVERT CLIFFS UNITS NO. 1 AND *:NO. 2 DOCKET NOS. 50-317 AND 50-318

~

INTRODUCTION ,

Technical Specification 4.0.5 for the Calvert Cliffs Unit Nos. I and 2 nuclear plants states that inservice examination of ASME Code Class 1, 2, and 3 components shall be performed in accordahce with Section XI of the ASME Soiler and Pressure Vessel Code and applicable addenda as re-quired by 10 CFR 50.55a(g) except where specific written relief has been granted by the Comission. Certain requirements of later editions and tidenda of Section XI are impractical to perform on older plants because, ,

of the design, component geometry, and materials of construction. Thus,  !

10 CFR 50.55a(g)(6)(i) authorizes the Comission to grant relief from those l O -

requirements upon making the necessary findings.

v...

By letters dated December 5, 1978, March 29, 1980, November 19, 1980 and May.29, 1981, Baltimore Gas and Electric Company (BGLE) submitted requests for relief from certain Code requirements determined to be impractical to perform on the Calvert Cliffs Unit Nos. 1 and 2 nur. lear plants during the inspection interval. Additional information concerning these requests for relief was submitted by BG&E letters dated July 22, 1982, August 30, 1982 and October 29, 1982. The programs are based on the requirements of the 1974 Edition .through Sumer 1975 Addenda of Section XI of the, ASME Code. .

EVALUATION - _ ..

Requests for relief from the requirements of Section XI which have been i determined to be impractical to perform have been reviewed by the NRC  !

staff's contractor, Science Applications, Inc. The contractor's evalua-  !

tions are presented in the Technical Evaluation Report (TER) attached.

One request for relief, involving the repair of an are strike on Class 2 piping, was not reviewed in the TER. This requsst was granted as.shown .

in Table 2. The staff has reviewed the TER and agrees with the evaluations and recomendations except as indicated. A sumary of the determinations made by the s'taff is presented in the following tables;.

O sso, n 4

I - .

s . .

O: O .

O Tant.E 1 CLASS 1 COMPONENTS LICENSEE PROPOSED IWD-2600 IWB-2500 SYSTEM OR AREA TO DE REQUIRED ALTERNATIVE RELIEF REQUEST ITEM NO. EXAM'. CAT. COMPONENT EXAMINED METl!0D EXAMINATION STATUS B1'. 2 i D-B Reactor 5% of Volumetric . Visual During Granted Provided i Vessel -

Circumfer- System Pressure Examination Sample Closure ential Weld Test & Cladding of Other Category B-D' licad (6-2090) Examination Weld be Inc,reased to Achieve Equivalent t

Sample Size

• Bl.4 .

B-D Reactor Nozzle-To- Volumetric: Volumetric: Granted Vessel Vessel 25% of Welds . 25% During

. Nozzles Welds And And Radiused 1st. Period.. .

None During

- Inside Sections Dur ,

i Radiused ing 1st Period, Second Period ..

Sections 50% by End of 100% During 3rd Second Period,. Period

-f, 100% by End of ---

. Interval 111. 14 ., B-I-1 Reactor Vessel Visual Visual Granted i Vessel Cladding Examination When Core I of Six Barrel is Patches Removed

. Distributed Evenly Over Three 40-month .

i . Periods

~

• Cor- rsations with representatives of BC&E indicated that this provision is acceptable.

8 0 O O I

~

TABLE 2 CLASS 2 COMPONENTS * ,

= f LICENSEE PROPOSED IWC-2520 SYSTEM OR AREA 10 BE REQUIRED ALTERNATIVE RELIEF REQUEST IWC-2600 EXAMINED METIIOD EXAMINATION STATUS ITEM NO. EXAM. CAT. COMPONENT (Repair) Shutdown Arc Strike Volumetric- None Granted C2.1 Cooling "

Repair Area Radiography (IWC-4000)/ 2-inch Cross s

Connect '

e I

i e

T . .

\

9 b

• Thi_ request for relief is not described in the TER _..d.is based upon a request dated May 29, 1981.

l I

I. s TABLE 3 -

l t

• =t o <

• e .

• me 9. e-CtASS-3 COMPONEffTS - - . .

(SEE TABLE 4)

O B

O e

4 e

S e

8' e

a.  ; ..

0 0

a e e

~

um , 9 He e

9 e

e 9

4 9

e v - , - - - - - - . - , - - - - . ,

( TABLE 4

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

PRESSURE TEST IWC-5000 & LICENSEE PROPOSED SYSTEM OR - IWD-5000 TEST ALTERNATIVE RELIEF REQUEST

. COMPONENT PRESSURE REQUIREMENT TEST PRESSURE STATUS Class 1, 2 & 3 Hold Time Shall be Perform Test in Approved Four Hours Accordance with '

. the 1977 Edition, Winter 1978 Addenda Class 3, Diesel Test P.ressure shall Monitor Critical Granted

• I Generator Com- be 1.10 Times the Parameters, weekly ponents System Design Pressure Load' Test, and In-service Leak Test Each Inspection Period O ;r. -

Salt Water The System Test Perform an In- Granted * -

Cooling Systems, Pressure Sna11 be service Leak Class 3 at least 1.10 Times .

Test Yearly on The System Design Above-Ground Pressure Portions to Verify System .

Integrity Service Water The System Test Perform an In- Granted

• System Main Pressure Shall be service Leak ,

Headers at least 1.10 Times Test Yearly to -

The System Design' Verify Sy' stem

• Pressure Integrity -

O .

. l

~

• By letter dated October 29, 1982, BG&E provided an appropriate basis for determining thtt the 1977 Edition Summer 1978 Addenda, is impractical for these pressure tests.

Accordingly, these requests for relief are granted.without additional provisions.

. , , , - . . . , . , , , . , - ., ..* . , , . ,-,,,-y., - , , - - - - - -. I m--a,

N -

TABLE 4

[}

- * . PR. ESSURE TEST

= (CONTINUED)'

IWC-5000' & LICENSEE PROPOSED SYSTEM OR IWD-5000 TEST ALTERNATIVE RELIEF REQUEST COMPONENT PRESSURE REQUIREMENT TEST PRESSURE STATUS Class 2 Steam The System Pressure Examine Components Granted

• And Feedwater Shall be at least Under Normal Opera-Piping 5-inch 1.25 Times The System ting Pressure Corre-

. Mominal Pipe Design Pressure sponding to 100% Rated Size And -

. Reactor Power; Perform Smaller That -

Liquid Penetrant Exam-Cannot be . inations on First And.

Isolated From Last Wel,d Pass; Volu-Steam Generator metric Examination of Secendary Side Welds Greater Than 1-inch After Repair Nominal Pipe Size.

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()+l.iditionalinformationcontainedintheBG&EletterdatedAugust 30, 1982 was .

considered which had not been reviewed in the TER (see NRC letter dated November 19, 1982.) .

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TABLE 5 O . .

_. . . ULT. RAS 0_NIC. EXAt1INATION TECHNIQUE LICENSEE PROPOSED f ,

SYSTEM OR ALTERNATIVE RELIEF REQUEST COMPONENT REQUIREMENT TEST PRESSURE STATUS Piping Welds Section XI, 1974 All Indications Granted With Edition, Appendix I Which Exceed 100% Additional or Article V of of Reference L'evel Requirement Section V h'ill be Evaluated That Indica-And All Indications tions 20%

Which exceed 50% or Greater of of Reference Level Reference Will be Rec'rdedo Level That

,- Are Inter-preted to be a Crack Must be Identified And Evaluated According to The Rules of Section XI*

1 O conversetions wit 8 rePrese tetives < rom satt iaeicetes thet 181,prov4sica is acceptable. ,

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C0i!CLUSI'ON The relief from 'the Code is based upon our review of the information sub-mitted by.BGinE to support the determination that compliance with the ASME Code inservice inspection requirements would be impractical for the facility.

We have determined that the inspections from which this relief is sought are impractical and. pursuant to 10 CFR 50.55a(g)(6)(i), that the granting of this relief is authorized by law and will not endanger life or property, or the comon defense and security, and is otherwise in the public interest.

In. making this determination, we have given due consideration to the burden that could result if these requirements were imposed on the facility. We have determined that the granting of this relief does not involve a signi-ficant increase in the probability or consequences.of an accident previously evaluated, does not create the possibility of an accident of a type dif-ferent from any evaluated previously, does not involve a significant reduction in a safety marain, and thus, does not involve a f,ignificant hazards considera-tion. Furthermore, we have determined that the granting of this relief from ASME Code requirements 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. We have concluded that the granting of this relief is insignificant from the standpoint of environmental impact and pursuant to 10 CF.R 51.5(d)(4) that neither an environmental impact statement nor a negative declaration and environmental impact appraisal needs to be prepared i ?. connection with this action.

Date: DEC 131982 ,

Principal Contributors:

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G.' Johnson ' '

D. Jaffe ,

At'tachment: SAI Technical Evaluation Report 7

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7590-01 UNITED, STAT _ES NU. CLEAR REGULATORY COMMISSION DOCKET NOS. 50-317 AND 50-318 BALTIMORE GAS AND ELECTRIC COMPANY NOTICE OF GRANTING OF RELIEF FROM ASME SECTION XI INSERVICE INSPECTION REQUIREMENTS

. The U. S.. Nuclear Regulatory Commission (the Comission) has granted relief from certain' requirements of the ASME Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components" to Baltimore Gas and Electric Company (the licensee), which re' vised the inservice inspection program for Calvert Cliffs Nuclear Power Plant, Units No. 1 and No. 2. The ASME Code requirements are incorporated by reference into the Comission's rules and regulations in 10 CFR Part 50. The relief is effective as of its O ..

date.of issuance.

The NRC has provided a relief from the ASME Boiler and Pressure Vessel Code,Section XI, regarding the requirements for:

. e Inspection of Seal W61d in Closure Head, .

e Inspection of Primary Nozzle-to-Vessel Wilds and Nozzle Inside Radiused Section, , ,, , , ,

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e I p t o'f Riact'or Ee's'sel C1'a'Eding ~N,[.~ ,7 " ','. . . ' . [ . ^ T

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e Repair of an Arc Strike, Class 2 Pipe, e Pressure Test Hold Time, ..

e Class 3 System Pressure Tests, ,

e Inservice Leak Test (Hydrostatic Testing) for the Salt Water Cooling System and Service Water Main Headers, ,

e Ultrasonic Examination Techniques, O e Repair and -Hydrostatic Testing for Small Steam and Feedwater Piping, Class 2.

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7590-01

..s ..-s The Commission has determined that the granting of,this relief will not result in any significant environmental impact and that pursuant to 10 CFR 151.5(d)(4) an environmental impact statement or negative declaration and environmental impact appraisal need not be prepared in connection with

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this action. .

For further details with respect to this action, see (1) the licensee's request for relief from code requirements dated December 5, 1978, March 29, 1980, November 19, 1980 and May 29, 1981 and additional information submitted

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by the licensee's letters dated July 22, 1982, August 30, 1982, October 29, 1982 and (2) the Comission's related Safety Evaluation. All of these items

. art. available for public inspection at the Comission's Public Document Room, 1717'H. Street, N.W. Washington, D.C' . 20555, and at the Calvert Co0nty Library,i Prince Frederick, Maryland. A copy of item (2) may be obtained upon request addressed to the U. S. Nuclear Reg 61atory Commission, Washington, D. C.  ;

20555, Attention: Director, Division of Licensing.

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Dated at Bethesda, Maryland this 13th. day of Dececher,1982..

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FOR.THE NUCLEAR REGULATORY COMMISSION'<.'. ' ' _ ,'

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Robert A. Clark, Chief '

Operating Reactors Branch f3' l Division of Licensing

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UNITED STATES NUCLEAR REGULATORY COMMISSION l'4 s - -

'A ASHIN G T ON. D. C. 20555 k,..;j.//: JAN 2 4 1983 Docket Nos. 50-317 '

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and 50-318 Mr. A. E. Lundvall, Jr. -

Vice President - Supply .

Baltimore Gas & Electric Company P. O. Box 1475 ,

Baltimore Maryland .21203

Dear Mr. Lundvall:

The NRC has provided relief from requirements of the ASME Boiler and Pressure Vessel Code,Section XI, which BG&E has determined to be imprac-tical, in accordance with your applications dated November 6,1981 and December 21, 1982. ,

The code relief, granted in. accordance with 10 CFR Part 50, Section 50.55a (g)(6)(i),relatesto(1)ExaminationofreactorVEsselclosureheadclad- -

ding, (2) Code Case N-210. "Exemption to Hydrostatic Tests after Repairs,"

j (3) Code Case N-307 for Centerdrilled Hole Ultrasonic Examination of Studs, (4) Increased inservice leak testing in lieu of hydrostatic pressure testing of the Class III component cooling water system, and (5) Hydrostatic testing of welds that cannot be isolated from the steam generators (Unit 2 only).

Copies of our Safety Evaluation and Federal Register Notice are enclosed.

Sincerely,

_t . .

bert A. lark, Chief Operating Reactors Branch f3 i Division of Licensing )

Enclosures:

1. Safety Evaluation ,

2. Federal Register Notice TRAINING & TECHNICAL SERVICES "i cc: See next page DATE RECEIVED- O! ,

vl t-T&TS ACTION NOTE / RETURN l t' .TS - --

f40TE RETAlH i Ac G YOUR COMMENTS _

PLEASE SEE ME

. OL&S - RESPONSE REQ'O 6 Y ILE .

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g %q, UNITED STATES *

3. # 'n j .;.1 NUCLEAR REGULATORY COMMISSION WAsHIN G TO N, D. C. 20$$5 0 *%,*...* SAF

/ 'ETY EVALUATION BY THE OFFICE OF NOCLEAR REACTOR R RELATED TO REQUESTS FOR RELIEF FROM INSERVICE INSPECTION REQUIREMENTS CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NOS. 'l AND 2 DOCKET NOS. 50-317 AND 50-318 ,

By applications dated November 6,1981 and December 21, 1982, Baltimore

- Gas and Electric Company (BG&E) requested relief from inservice inspection requirements of Section XI of the ASME Boiler and Pressure Vessel Code for Calvert Cliff.s Units 1 and 2. The proposed relief it, described herein. .

Discussion and Evaluation

1. Examination Reouirements for Reactor Vessel Closure Head Cladding The ASME Code Section XI,1974 Edition, with Addenda through Summer

. 1975, requires a visual and surface examination or a volumetric -

examination of the reactor vessel closure head cladding. By appli-cation dated November.6,1981, BG&E_, requested relief from this ex- .

amination requirement.

By adoption of the 1977 Edition of the ASME Code,Section XI, with Addenda through Summer 1978, BG&E is relieved of the requirement to .

ins'pect the reactor vessel head cladding. The commitment, contained in the November 6,1981 application, to perform a visual examination of the cladding is acceptable. Accordingly relief from the required l

. examination of the reactor vessel head cladding is, appropriate provided l that visual examination is continued.

2. Use of Code Case N-210. "Exemption to Hydrostatic Tests After Repairs't By appli, cation dated November 6,1981, BG&E requested relief from the ASME Code in that they desire to utilize Code Case N-210 in the course of performing the Inservice Inspection Program for Calvert Cliffs Units 1 and 2. The use of Code Case N-210 is endorsed by Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability-ASME Section XI Division 1." Revision 1 of Regulatory Guide 1.147, dated February 1982, indicates that Code Case N-210 has been annulled in that it has been l . included in the ASME Code by a subsequent revision. In such instances. -

! the continued use of the Code Case intent is sanctioned under the rules of the Code. Accordingly, we conclude that the use of Code Case N-210 for Calvert Cliffs Units 1 and 2 is acceptable, subject to the condition-i that, for repairs to piping, pumps and valves, the depth of the cavity not exceed 25 percent of the wall thickness.

3. Use of Code Case N-307, "Revised Ultrasonic Examination Volume O for Class 1 Bolting Examination Category B-G-1. Division 1, When the Examinations are Conducted from the Center-Drilled Hole" By application dated November 6,1981. BG&E requested relief from the 8529 y -

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ASME Code in that they desire to use Code Case N-307 in the course of O' perfonning the Inservice Inspection Program for Calvert Cliffs Units i I

1 and 2. Code Case N-307 allows relief from ultrasonic examination of studs in the volume that extends to within \" of the thn eade'd surface.

We.have previously reviewed the use of Code Case N-307 and have found

• it acceptable in that. indications are most likely to occur within %" of

- ,thethreadedsurface(thevolumetobeinspected)duetothehigherstress.

concentrations associated with the threads. Moreover, Code Case N-307 l was subsequently approved by Regulatory Guide 1.147 Revision 1, and is ..

therefore acceptable for use at Calvert Cliffs Units 1 and 2. ,

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4. Increased Inservice Leak Testino'in' Lieu of Hydrostatic Pressure Testina of Class 111 Component Coolino Water Systems By application dated November 6,1981. BG&E requested relief from the >

ASME Code as it applies to the inservice inspection of the Class III .

component cooling water system. Paragraph IWD-2410 requires hydrostatic Pressure Testing of Class III systems to 1.1. times design pressure .

during every ten year inspection interval. In their November 6, 1981 1 i

application, BG&E stated that "on the Component Cooling Water System main headers, where butterfly valves are installed, sufficient seal to maintain pressure on isolated portions of the system cannot be '

completed. BG&E proposed *that the Inservise Lea'll Test required every i 40-month period be performed on an annual basis to substitute for i

hydr'ostatic pressure testing of this system."

O In our letter and Safety Evaluation (SER) dated December 13, 1982 the NRC ap' proved relief from the hydrostatic test requirements of IWD-2410 for inservice inspection of the Class III service water. system. The service water system is similar to the component cooling water. system '

in that both systems rely on butterfly valves for pressure boundary -

1 solation. Our approval of relief in the December 13, 1982 SER was based upon (1) the inability of these butterfly valves to sustain ,

'the pressure required for a hydrostatic test (1.1 times the design )

pressure) and (2) the absence of a reasonable alternative other than annual leakage testing. From the above, we conclude that the relief requested in the November 6,1981 application, for hydrostatic testing of the component cooling water system, should be granted on the same basis as described in our SER dated December 13, 1982. This relief is based upon the commitment by BG&E to perform the.40-month inservice leak test on an annual basis for the component cooling water system.

5. Hydrostatic Testing of' Welds that Cannot be Isolated from the Steam

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senerator (Unit z only) , l

. In our letter and SER dated November 19, 1982, the NRC provided relief from the ASME Code requirement to perform hydrostatic tests on certain l' welds in lines that cannot. be isolated from the steam generator. These welds were associated with modifications to the auxiliary feedwater system. . '

• U This relief was based upon the desire not to perform a hydrostatic .

l test of the steam generator to test these welds since the steam ,

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generators are limited to a total of ten (10) hydrostatic tests during the lifetime of the plant. The next full hydrostatic test of the steam generators is scheduled during the 40-month inspection period which will begin in December 1983. ,

By application' dated ' December 21, 1982 BG&E identified additional welds  ;

associated with the auxiliary feedwater system modifications, for which relief had not previously been requested. These welds are also located such that they cannot be isolated from the steam generator in order to -

perform the required hydrostatic test. Accordingly, it is appropriatt to provide relief from the hydrostatic test requirement for these additional

- welds based upon the discussion presented in our SER dated November 19, 1982.

This relief includes the following additional inspections:

1. Surfa'ce Examination after the final weld , pass. ,

2. An Inservice Examination of the components at a pressure corresponding to 100% reactor power. An Inservice Examination of the components in the HOT STANDBY mode (which is approximately 50 psi greater than 100% nor, mal operating pressure). ,

3. A 100% Volumetric , Examination utilizing ultrasonic and/or radio-graphy methods. .

A f'inal issue raised by BG&R in~ their December 21, 1982 application relates to the surface examination of welds after removing half of O

the first layer by grinding. While this technique was endorsed by '

the NRt in our November 19, 1982 SER, for welds'for which relief from hydrostatic testing was granted, we concur with the licensee that this is not a requirement of the applicable repair code (USAS 8 31.7).

. Accordingly, the removal of weld material and subsequent surface exami- -

nation is not a required procedure for welds associated with the auxiliary feedwater modification as described in our SER of November 19, 1982 and the BG&E application dated December 21, 1982. .

Conclusion - .

The relief from the Code is based upon our review of the information sub-mitted by BGLE to support the determination that compliance with the ASME Code inservice inspection requirements would be impractical for the facility.

We have determined that the inspection from which this relief is sought is impractical and pursu. ant to 10 CFR 550.55a(g)(6)(i), that the granting of

-l this relief is authorized by law and will not endanger life or property, or the epmen defense and security, and is otherwise in the public interest.

In making this determination, we have given due consideration to the burden that could result if these requirements were imposed on the facility. We have' determined that the granting of this relief does not involve a signi- i ficant increase in the probability or consequences of an accident previously 1 evaluated, does not create the possibility of an accident of a type different' i from any evaluated previously, and does not involve a significant reduction in a margin of safety; and thus, does no't involve a significant hazards O consideration. We have concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the' health'and safety of the public will not be endangered by operation in the proposed manner. <

and (2) such. activities will be conducted in compliance with the Conrais- .

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O ion , resuiations and the 1 uance or.t81, re uef iii. net.se iai 4eai. ...

to the comen defensp and. security dr"to the health'and safety of the public. Furtheremore, we have determined that the granting of t!}is ,

relief from ASME Code requirements does not authorize a change in affluent types or total amounts nor an increase in power level and will not result in any significant invironmental impact. We have concluded that the granting

, of this relief i's insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that neither an environmental impact

• statement nor a negative declaration and environmental impact appraisal need to be prepared in connection with this action. -

Dated: JAN 2 4 E .

Principal Contributors:

D. Jaffe

• G. Johnson Y O

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. 7590-01 l

UNITED STATES NUCLEAR REGULATORY COMMISSION ,

DOCKET NOS. 50-317 AND 50-318 BALTIMORE GAS AND ELECTRIC COMPANY

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NOTICE OF GRANTING OF RELIEF FROM ASME SECTION XI

, INSERVICE INSPECTION REQUIREMENTS , ,

The U.-S. Nudlear Regulatory Comission (the Comission) has granted i a relief from certain requirements of the ASME Code,Section XI, "Rules e for Inservice Inspection of Nuclear Powe'r Plant Components" to Baltimore GasandElectricCompany(thelicensee),whichrevisedtheinserviceinspec-tion program for Calvert Cliffs Nucitar Power Plant, Unit Nos. I and 2. The

. ASME Code requirements are'incorporatedly veference into the Comission's

. i rules and regulations in 10 CFR Part 50. The relief is effective as of O

.its da.te of issuance. -

' The code relief, granted i'n accordance with 10 CFR Part 5 , Section

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50.55'a(g)(6)(1). relates to (1) Examination of reactor vessel closure head

' cladding, (2) Code Case N-210. "Exemption to Hydrostatic Tests after Repairs,"

(3) Code Case,N-307 for Centerdrilled Hole Ultrasonic Examination of Studs, (4) Increased inservice leak testing in lieu of hydrostatic pressure testing of the Class.III component cooling water system, and (5) Hydrostatic testing of welds that cannot be isolated from the steam generators (Unit 2 only). -

The Comission has' determined that the granting of this reifef will  ;

not result in any significant environmental impact and that pursuant to -

10 CFR g51.5(d)(4) an environmental impact statement or negative declaration i and environmental, impact appraisal need not be prepared in connection with this action.,

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1 7590-01f l O .

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. r For further. details with respect to this actjon, see (1) the licensees requests for relief from code requirements dated November 6, 1981 and December'21,1982 and (2) the Comission's related Safety Evaluation. All of these items are available for public inspection at the Commission's

• Public Document Room,1717 H Street, N. W. Washington, D. C. 20555, and at the Calvert County Library, Prince Frederick, Maryland. A copy of item (2) may be obtained upon request addressed to the U. S. Nuclear Regulatory ,

.. Comission, Washi.ngton, .D. C. 20555, Attention: Director Division of Licensi.ng. - -

' Dated at Bethesda, Maryland this 24th day of January,1983.

, FOR THE NUtt. EAR REGUi.ATORY COMMISS10?i J .

Robert A. Clark, Chief -

Operating Reactors Branch #3 Division of Licenstng l l

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