ML20086C036

Forwards New & Revised Relief Requests for CNS Second ten-yr ISI Program,In Response to Deficiencies Noted in Insp Rept 50-298/93-17

ML20086C036
Person / Time
Site:	Cooper
Issue date:	06/21/1995
From:	Mueller J NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS950119, NUDOCS 9507060245
Download: ML20086C036 (36)

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Nebraska Public' Power District u

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June 21,1995 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555

Subject:

Second Ten-Year Interval Inservice Inspection Relief Requests Cooper Nuclear Station, Docket No. 50-298, License No. DPR-46 I

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Reference:

NRC Inspection Report No. 50-298/93-17 issued to Nebraska Public Power District dated July 17,1993 Gentlemen:

. In accordance with 10CFR50.55a(g), the Nebraska Public Power District (District) is submitting to the Nuclear Regulatory Commission (NRC) new and revised relief requests for the Cooper Nuclear Station's (CNS) Second Ten-Year Inservice Inspection program (ISI). NRC Iuspection Report No. 50-298/93-17 identified certain deficiencies in the CNS ISI program. In response, the -

District has completed a thorough review of the program and has identified the need for additional

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relief from American Society of Mechanical Engineers (ASME) Code requirements that are i

considered impracticable. In addition, certain revisions and corrections to the existing approved relief requests were identified. The new and revised relief requests are included in the attachments j

to this letter. The District asks that these requests for relief be granted by September 1,1995 in order to support the refueling outage scheduled for October 1995.

The Fall 1995 refueling outage will be the last outage in the second ten-year inspection interval.

The current inspection interval has been extended in accordance with ASME Section XI, lWA-2400(c) as a result of the 1994 extended outage. The next inspection interval is scheduled to commence on January 1,1996, subject to the completion of the 1995 refueling outage.

I If you have any questions or require additional information regarding these relief requests, please i

call.

Sincerely, 1

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'J. II. Mueller te Manager F ;

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' U.S. Nuclear Regulatory Commission June 21,*1995 '

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Regional Administrator USNRC Region IV Arlington, TX NRC Resid_ent Inspector Cooper Nuclear Station J. R. Itall NRC NRR Project Manager Rockville, MD NPG Distribution i

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l COOPER NUCLEAR STATION SECOND TEN YEAR INTERVAL INSERVICE INSPECTION PROGRAM RELIEF REQUESTS l

The following is a list of Relief Requests:

Relief Request Description Number RI-01 Deleted, See Note 1 RI-02 Deleted, See Note 2 RI-03 Volumetric Examination of Nozzle Inner Radii (Reformatted, see Note 3)

RI-04 Deleted, See Note 1 RI-05 Inaccessible RHR Heat Exchanger Welds (Revised, see Note 4)

RI-06 Inaccessible Reactor Pressure Vessel Welds (Revised, see Note 5) r RI-07 Inspection of Reactor Vessel Support Skirt to Reactor Vessel Botton Head Wald (See Note 6)

RI-08 10-Year Hydrostatic Pressure Test Requirements for Class 1, 2 and 3 Systems RI-09 Pressure Testing of Non-safety System Process Piping In Containment Penetrations RI-10 Visual Examination (VT-2) In Redundant Systems for Buried Components RI-11 Visual Examination (VT-2) of Relief Valve Piping which Discharge into the Suppression Pool RI-12 Examination of Class 1 & 2 Longitudinal Seams RI-13 ISI of Snubbers included in CNS Technical Specifications RI-14 Alternative Rules for Selection and Examination of i

Class 1, 2 and 3 Integrally Welded Attachments RI-15 Inaccessible CRD Housing Welds RI-16 Examination of Pump Casing and Valve Body Internal Surfaces RI-17 ASNE Category B-K-1 and C-C, Integrally Welded Shear Lugs on Piping RI-18 Integrally Welded Attachments.

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l COOPER NUCLEAR' STATION SECOND TEN YEAR INTERVAL INSERVICE INSPECTION ~PROGRAN RELIEF REQUESTS Relief Request Descriptinn Number RI-19 Exemption From Pressure Testing Reactor'fessel Head Flange Seal Leak Detection System NOTE 1 Relief Requests RI-01 and RI-04 have been deleted.

Alternate welds have been selected in accordance with 10CFR50. 55a (b) (2) (ii) and Code Case N-408-2.

NOTE 2:

Relief Request RI-02, "ASNE Category C-F, RER Drywell Spray Internal to Drywell", was not granted.by the NRC E

and thus, removed per the Nay, 1987, addenda.

i NOTE 3:

Relief Request RI-03, " Volumetric Examination of Nossle-Inner Radii", was reformatted.

The basis for the relief and the alternate examination proposed are unchanged.

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NOTE 4:

Relief Request RI-05, " Inaccessible RER Heat Exchanger Welds", has been revised to correct an error in the weld number.

The bauis for the relief and the alternate examination proposed are unchanged.

NOTE 5:

Relief Request RI-06, " Inaccessible Reactor Pressure vessel Welds", has been revised to incorporate NRC Safety Evaluation dated January 27, 1986.

Note that this relief request is still applicable per 10CFR50.55 a (g) 6 ( A) (3) (iv) because the augmented examinations of the RPV have been scheduled for the first period of the third inspection interval.

NOTE 6:

Relief Request RI-07, " Inspection of Reactor Vessel Support Skirt to Reactor Vessel Bottom Head Weld", was submitted under separate cover on November 1, 1994.

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o COOPER NUCLEAR STATION INSERVICE IN8PECTION PROGRAM RELIEF REQUEST RI-03 REVISION 1 code Class:

1

Subject:

Volumetric Examination of Nozzle Inner Radii Components:

Category B-D Reactor Head Nozzles N6A, N6B, and N7 Q9de Requirements i

Reference; ASME Section XI 1980 Edition with the Winter 1981 Addenda.

Table IWB-2500-1 requires a volumetric examination of the nozzle inner radius section on the reactor vessel head nozzles.

Basis For Relief:

The reactor vessel closure head is removed during refueling outages which allows access to the inner surface of the head nozzles.

'A dye penetrant test of the inner radii would be more sensitive in detecting surface defects than an ultrasonic test from the outside diameter.

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Erggesed Alternative In lieu of performing the Code required examinations, CNS proposes to perform a surface examination of the reactor head nozzles inner radii.

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J COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-05 REVISION 1 Code. Class:

2 Subjects.

Inaccessible RHR Heat Exchanger Welds.

Components:

' Category C-A Circumferential welds in vessels.

Code Requirement 31

Reference:

ASME Section XI, 1980 ' Edition with the Winter 1981 Addenda Table IWC-2500-1 requires volumetric examination of circumferential welds RHR-CA-3A and RHR-CA-3B, in one of the RHR heat exchangers.

Basis For Relieff The weld joint configuration is not accessible _ for volumetric examination.

The configuration is shown in Figure RI-05-1 on page 2.

Limited access also precludes a surface examination.

Proposed Alternativet In lieu of performing the Code required examinations, CNS proposes to perform a visual inspection of the area during a system leakage test each inspection interval.

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l COOPER NUCLEAR STATION i.18ERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-05 REVISION 1 Figure RI-05-1 I

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.l-i COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAN RELIEF REQUEST NUNBER RI-06

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REVISION 1 Code Class:

1 Subjects Inaccessible Reactor Pressure Vessel Welds.

Components:

Category B-A circumferential welds in vessels.

Code. Requirements:

Reference; ASME Section XI, 1980 Edition with the Winter 1981 4

Addenda Table IWB-2500-1 requires a volumetric examination of the following circumferential welds:

HMB-BB-1 HMB-BB-2 HMB-BB-3 HMB-BB-4 HMB-BB-5 HMB-BB-6 VLA-BA-1 VLA-BA-2 VLA-BA-3 VLB-BA-1 VLB-BA-2 VLB-BA-3 VCB-BA-2 Basis For Relief:

The Cooper Nuclear Station construction permit was issued before the effective date of implementation for ASME Section XI and thus the plant was not designed to meet the requirements of inservice inspection; therefore, 100%

compliance is not feasible or practical.

Access to the reactor vessel beltline region is not possible.

The reactor vessel is insulated with permanent reflective insulation and surrounded by a concrete biological shield.

The annular space between the inside diameter of ' the insulation and the outside diameter of the reactor vessel is a nominal 2 inches.

There is no working space to remove the insulation panels from the vessel, which precludes both direct and remote examination of the outside surface.

The interior surface is clad and the arrangement of the vessel internals, shroud and jet pumps make an internal volumetric examination of these welds impractical for a meaningful examination using the technology available at the beginning of the interval.

Parts of longitudinal seams VLA-BA-1, 2,

and 3, however, appear to be accessible from openings around the recirculation riser nozzles N2A, N2E, and N2H respectively.

Again these seams are not 100%

accessible.

The scanning surface would require a minimum of 17 inches from the weld.

This surface area is only available for a few inches - closest to a nozzle.

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COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAN RELIEP REQUEST NUNBER RI-06 REVISION 1 Proposed 11ternativat In lieu of performing the Code required examinations, CNS proposes to perform:

(a)

When the recirculation riser nozzle to vessel welds (Category B-D) are examined, a best effort examination of longitudinal shell welds in the beltline region,. VLA-BA-1, shall be performed; 2,

and 3,'

(b)

The accessible portions of the bottom head meridional welds, HMB-BB-1 through 6, will be examined on a best effort basis when the vessel support skirt weld'is examined; (c) ~The beltline region weld areas shall be visually examined from the vessel-interior during the ' Category. B-N-1 and B-N-2 examinations; (d)

The areas of the lower head and the shield annulus below the vessel shall be visually inspected during a system hydrostatic test; and (e)

One circumferential shell weld and 50 inches of longitudinal seam shell weld above the shield wall shall be examined.

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COOPER NUCLEAR STATION i

U INSERVICE INSPECTION PROGRAN RELIEF REQUEST NUMBER RI-07 REVISION E Code class:-

1.

Subjects-Inspection of. Reactor Vessel Support Skirt.to l

Reactor Vessel Bottom-Head Weld.

Component Category B-H Integral Attachment Weld Cada maauir - nts:

Reference; ASME Section XI,.1980 Edition with the ' Winter 1981

i Addenda Table IWB-2500-1 requires a surface examination to be performed on areas A-D and B-C of Figure IWB-2500-13.

Basis for Relief f

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The reactor vessel bottom head was constructed with a weld build-up around the circumference of the head that was designed as the attachment point for the reactor vessel support. skirt (Figure RI-07-1).

_This weld build-up was machined and heat treated along with l

the reactor vessel.

The support skirt is attached to the weld build-up by means-of a full penetration butt weld..As can be seen in Figure RI-07-2, the.

design of this weld is such that the surface examination requirements of Figure IWB-2500-1 (Area'A-D)-can not be met'due to lack of accessibility.

The interior surface area is not accessible due to the configuration of the vessel skirt, lower vessel head insulation, close proximity of CRD housings, and, the lack of j

adequate skirt access manways and manueuverability spacc.

There is a severe angle between the vessel bottom head and the vessel skirt I

which physically prohibits access to the root area of the weld..

However, a partial coverage volumetric examination of this volume.

can be performed from one side (B-C) of the weld.-

Based upon a j

review of the fabrication drawings, it is estimated that - an i

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ultrasonic examination.can.be performed on volume A-B-C-D, as shown on Figure RI-07-2, using a 0. degree longitudinal' wave search unit,.

45 degree shear wave search unit, and a 60 degree shear wave search j

unit.

Due to the configuration, the scanning will be performed.in one direction only.

Performance of this partial l coverage volumetric examination, coupled with a surface examination on the l

B-C side of the weld, will provide adequate assurance of the

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structural integrity of the weld.

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COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-07 REVISION 1 Basis for Relief (continued):

A fracture mechanics evaluation has been performed for the support skirt to lower RPV head weld.

The results of the evaluation indicate that, for a surface flaw, growth rate is extremely small, and such growth would provide negligible adverse impact upon the integrity of the weld for the remainder of plant life.

Proposed Alternative As an alternate examination, CNS will perform a surface examination on the accessible side of the subject weld (area B-C),

and an ultrasonic examination of area A-B-C-D, as shown on Figure RI-07-2, each inspection interval.

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i COOPER NUCLEAR STATION INSERVICE' INSPECTION PROGRAN i

RELIEF REQUEST NUNBER RI-07 REVISION 1 FIGURE RI-07-1 REACTOR VESSEL BOTTOM HEAD WELD BUILD-UP

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a COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAN RELIEF REQUEST NUNBER RI-07 REVISION 1 ZIGURE RI-07-2 REACTOR VESSEL SUPPORT SKIRT ATTACHMENT' WELD 45' HMC-80-1 UPPER 8 E SEGMENT 33 BR B N_

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COOPER NUCLFAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-08 REVISION O j

Code Class:

1, 2,

and 3 Subjects 10-Year hydroctatic pressure test requirements for Class 1, 2,

and 3 systems Components:

All Class 1,

2, and 3 piping systems which are subject to the pressure test requirements of ASME Section XI.

Coda Requirements:

Reference; ASME Section XI, 1980 Edition with the Winter 1981 Addenda Tcble IWA-5210-1, Categories B-P (Class 1), C-H (Class 2),

D-A, D-B, and D-C (Class 3) requires a system hydrostatic pressure test (IWA-5211d) and accompanying VT-2 at least once each inspection interval.

Basis For Relief 2 Approved Code Case N-498 currently allows Class 1 and 2 System Hydrostatic testing at a reduced pressure equal to system nominal operating pressure. The recent ASME approved Code Case N-498-1, i

while repeating these requirements for Class 1 and 2,

also clarifies the intent of using installed plant instrumentation without the need for test gauging or the imposed requirements of

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IWA-5260 when performing these nominal operating pressure tests.

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Performing system pressure tests on Class 1

and 2

systems

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consistent with the requirements of N-498-1, together with the I

applicable volumetric examinations in accordance with the ISI Program, provides a level of quality and safety equivalent to, or l

greater than, that provided by the Code hydrostatic test pressure and instrumentation requirements.

Code Case N-498-1 also permits the reduced pressure testing in lieu of Hydrostatic Tests for Class 3 Systems.

CNS employs a very active erosion / corrosion monitoring and control program which periodically measures wall thickness in selected C). ass 3 piping and components.

This program primarily focuses on j

those portions of piping which are most susceptible to erosion, microbiologically influenced corrosion (MIC) and other identified corrosion mechanisms which are inherent to the service water and Page 1 of 2 l

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COOPER NUCLEAR STATION l

INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-08 REVISION 0-nazia For Relief (continued):

like systems.

The screening criteria for selection of piping and i

components to be chosen for " Thickness Examination" includes:

(1) sections susceptible to wall thinning by erosion, (2) low flow sections and, (3) intermittent or no flow sections.

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l It is CNS's intention to select those portions of piping and l

l components for examination most susceptible to erosion and corrosion thereby giving a conservative representation of overall pressure boundary integrity.

It is CNS's position that performing systen pressure tests on Class 3 systems consistent with the requirements of N-498-1, l

together with augmented test programs (e.g.

erosion / corrosion l

monitoring for piping determined to be most susceptible to erosion I

and corrosion), provides a level of quality and safety equivalent to, or greater than, that provided by the Code hydrostatic test pressure and' instrumentation requirements.

Proposed Alternative As a supplement to existing Section XI requirements, CNS will adopt the provisions of Code case N-498-1.

In lieu of performing a hydrostatic pressure test at a pressure above nominal operating pressure or system pressure for which overpressure protection is

required, as required by Table IWA-5210-1, Examination Categories B-P, C-H, D-A, and D-B, a system pressure test at nominal operating pressure and temperature shall be performed.

In lieu of instrumentation requirements specified in IWA-5260, existing plant instrumentation will be used per IWA-5212(b).

Where instrumentation may be required and does not exist, the rules of IWA-5260 shall be used.

For Class 3 Systems, CNS shall also continue to maintain and implement an erosion / corrosion monitoring program for piping determined to be most susceptible to erosion and corrosion, as previously described.

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COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-09 REVISION O Code Class:

2

Subject:

Pressure testing of non-safety system process piping in a containment penetration.

Components:

Category C-H Class 2 penetration piping and valves, where the inboard and outboard piping outside the containment isolation valves is classified as nonsafety-related, code Requirements:

Reference; ASME Section XI, 1980 Edition with the Winter 1981 Addenda Table IWC-2500-1 requires a system pressure test, inservice or functional (IWC-5221), and a system hydrostatic test (IWC-5222) in conjunction with a VT-2 visual examination to be performed on pressure retaining components, piping and valves.

Basis For Relief The portion of piping that penetrates containment and the associated inboard and outboard containment isolation valves are required to be constructed in accordance with Class 1 or Class 2 design requirements.

In the instance where the piping penetration is for a nonsafety-related system, the sole safety function of the penetration piping and associated valves is to provide containment isolation and maintain containment integrity in the event of a failure of the attached nonsafety-related piping.

In all cases during normal plant operation, the isolation valves associated with these penetrations are maintained in the locked closed position, are administratively closed (controlled procedurally), or they close automatically upon receipt of a containment isolation signal or on loss of flow.

The integrity of these penetrations is verified by 10CFR50, Appendix J, leakage testing.

Additionally, per Code Case N-522, " Pressure Testing of Containment Piping Section XI, Division 1," the ASME Section XI Code Committee has determined that pressure testing in accordance with 10CFR50, Appendix J, is an acceptabl6 alternative to the pressure testing requirements of Table IWC-2500-1, Category C-H, for piping that penetrates the containment vessel and is attached to non-Code Class piping.

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COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM f

RELIEF REQUEST NUMBER RI-09 REVISION 0 i

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Basis For Relief fContinued)1 Performing system pressure tests each inspection period and a i

hydrostatic test each inspection interval as required by Section XI would _be redundant to Appendix J testing.

Additional pressure J

testing per the requirements of Table IWC-2500-1, Category C-H, would provide no significant increase in quality or safety.

Pressure testing of piping in nonsafety-related systems penetrating

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containment pursuant to the requirements of 10CFR50, Appendix J, in lieu of Section XI pressure testing provides an acceptable level of quality and safety.

Proposed Alternative As an alternative to existing Section XI requirements, CNS will i

adopt the provisions of Code Case N-522.

Pressure testing shall be performed in accordance with the requirements of 10CFR50, Appendix J,

in lieu of the additional requirements specified in Table IWC-2500-1, Category C-H.

This alternate testing shall be applicable to class 2 containment penetration piping and associated valves attached to non-safety system (non-Class) piping.

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A COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-10 REVISION 0 Code Class:

3 Subjects visual examination (VT-2) in redundant systems for buried components.

Components:

Category C-H, pressure test of buried service water critical supply headers leading from the service water building to the control building.

Code Requirements:

Reference; ASME Section XI, 1980 Edition, Winter 1981 Addenda In redundant systems where the buried components are nonisolable, i

the visual examination VT-2 shall consist of a

test which determines the change in flow between the ends of the buried pipe (IWA-5244 (b)).

Basis For Relief 2 Isolation valves are installed in the redundant buried portion of the service water critical supply headers.

Buried components in redundant systems that are isolable are not addressed in IWA-5244.

However, leakage testing of the buried piping is impractical because the isolation valves located in the service water building and the control building are large butterfly valves which are extremely unreliable for performing a pressure isolation function.

Each critical header supplies two RHRSW booster pumps, one REC heat 1

exchanger and one diesel generator.

A butterfly isolation valve is j

installed in the main header in the service water building and in each of these branch supply lines in the control building.

Since the valves are not designed to be leak tight, these five butterfly valves would provide multiple leakage paths._

Leak testing this buried piping to determine the rate of pressure loss would require extensive valve seat maintenance and would not provide conclusive test results.

1 Current Code rules allow determining a change in flow between the ends of the buried components (IWA-5244(a) and -5244(b)).

Flow instruments are installed in the service water lines in the control building.

However there are no flow instruments installed in the system upstream of the buried piping.

Accurate flow measurements using temporary flow instrumentation (e.g., ultrasonic flow meters) are not possible due to insufficient runs of straight pipe between the pump discharge and the buried piping.

Therefore, direct measurement of the change in flow between the ends of the buried Page 1 of 2

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COOPER NUCLEAR STATION 1

INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-10 REVISION 0 Basis For Reliefs (Continued).

piping is not practical.

The installation of permanent flow instruments would require significant system modifications which would be burdensome.

The cost of these modifications, when weighed against the benefits, are not justifiable.

Proposed Alternativer In lieu of performing a visual examination VT-2 in accordance with IWA-5244, CNS shall use existing plant instrumentation for determining the integrity of buried pipe.

Discharge pressure is indicated by pressure gauges provided at each individual pump (SW-PI-360 A,B,C,&D).

Service water pumps A & C discharge to a common header, as do pumps B & D.

Each header is provided with pressure indication prior to exiting the intake structure (SW-PI-383 A&B).

When these headers resurface in the control building, pressure indication (SW-PI-384 A&B) and flow indication (SW-FI-385 A&B and SW-FI-364 A&B) are provided.

The integrity of the buried piping is verified during quarterly pump testing.

Using the downstream flow instruments, flow rate is set at the fixed test reference value and documented in the test record.

The pump discharge pressure is then measured and used to determine the head produced by the pump.

Head and flow rate are interdependent variables which, together, define pump hydraulic performance.

As the pump degrades, the developed head will decrease at the reference flow rate.

Due to the location of the flow rate instruments (downstream of the buried piping), a decrease in pump head during testing may also indicate side-stream leakage into the isolated non-critical header or through-wall leakage in the buried portion of the service water system piping.

This is because the head developed by the pump decreases as flow rate increases.

Significant through-wall leakage would be evident because the total flow rate would increase even though the downstream indicated flow rate is set at the reference value.

Therefore, a satisfactory quarterly service water pump test also verifies the integrity of the buried system supply piping.

Should the pump test results fall in the required action range, additional tests and evaluations will be performed to determine whether the unsatisfactory test results are due to side-stream leakage past butterfly isolation valves, degraded pump performance, or through-wall leakage.

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COOPER NUCLEAR STATION IW8ERVICE INSPECTION PROGRAM-RELIEF REQUEST NUMBER RI-11 REVISION 0

Code Class:

3

Subject:

Visual examination (VT-2) of relief valve piping which discharge into the suppression pool'.

Components:

Relief Valve (s) MS-RV-71ARV, MS-RV-71BRV, MS-RV-71CRV,_MS-RV-71DRV, MS-RV-71ERV, MS-RV-71FRV, MS-RV-71GRV, and MS-RV-71HRV. discharge piping to the suppression pool.

code Requirementat Reference; ASME Section XI, 1980 Edition, Winter 1981 Addenda IWD-5223 (f).

For safety or relief valve piping which discharges into the containment pressure suppression pool, a pneumatic test (at a pressure of 90% of the pipe submergence head of water) that demonstrates leakage integrity shall be performed in lieu of a system hydrostatic test.

HERis For Relief These relief valves are currently actuated once each operating cycle commensurate with Reactor Vessel pressure 2 100 psig.

Suppression Pool temperature and levels monitored during this test substantiate the integrity of the discharge piping by its ability to direct flow from the relief valve to the suppression pool.

The code required 10 year pressure test of the discharge piping

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with a pneumatic test at a pressure of 90% of the pipe submergence head of water equates to an applied pressure of approximately 1.17 psig equivalent to the 3 feet of submerged piping.

This Code requirement has been removed from the 1994 Addenda of ASME Section XI 1992 Edition.

Current test parameters significantly exceed Code requirements in piping pressurization and frequency.

Performance of the current Code required testing would not increase the margin of assurance for safety beyond current test parameters, and would only serve as i

a redundant inferior test requirement.

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s COOPER. NUCLEAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-11 REVISION O

Proposed Alternativet In lieu of performing a visual examination VT-2 in accordance with the requirements specified in IWD-5223 (f), CNS shall use existing plant surveillance tests of the operability of each Main Steam Safety Relief Valve to demonstrate the integrity of the discharge piping.

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COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-12 REVISION 0 Code Class:

1 and 2 j

Subjects Examination of Class 1 and 2 Longitudinal Piping Welds Components:

All designated Code Class 1 and 2 longitudinal piping welds that are identified and included in the CNS Inservice Inspection Program.

Code Recuiremntat Reference; ASME Section XI, 1980 Edition with the Winter 1981 Addenda class 1:

Table IWB-2500-1 requires a surface and volumetric examination be performed on longitudinal pipe welds on piping greater than or equal to NPS 4 and a surface examination only on longitudinal pipe welds less than NPS 4.

Class 2:

Table IWC-2500-1 requires a surface and volumetric examination be performed on longitudinal pipe welds having a nominal pipe wall thickness greater than 1/2 inch. It also specifies and requires a surface examination only for those longitudinal pipe welds having a nominal pipe wall thickness of 1/2 inch and less and for those contained in pipe branch connections having a pipe diameter greater than NPS 4.

Code Case N-408-2, " Alternative Rules for Examination of Class 2 Piping;Section XI, Division 1",

Section (c) states as an alternative to Table IWC-2500-1 the requirements of Table 1 of the Code Case may be used for welds in austenitic stainless steel or high alloy piping, Code Category C-F-1.

Table 1 requires a surface and volumetric examination be performed on Class 2 longitudinal pipe welds where the nominal pipe wall thickness is greater than or equal to 3/8 inch for piping greater than NPS 4.

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Code Case N-408-2, " Alternative Rules for Examination of Class 2 Piping;Section XI, Division 1",

Section (d) states as an alternative to Table IWC-2500-1 the requirements of Table 2 of the Code Case may be used for welds in carbon steel or low alloy j

piping, Code Category C-F-2.

Table 2 requires a surface and j

volumetric examination be performed on Class 2 longitudinal pipe welds where the nominal pipe wall thickness is greater than or Page 1 of 2

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t COOPER NUCLEAR STATION INSERVICE INSP ET' ION PROGRAM RELIEF REQUEST NGHBER RI-12 REVISION O equal to 3/8 inch for piping greater than NPS 4.

Basis for Relief 2 The area of the longitudinal seam weld which is most susceptible to failure is that portion immediately adjacent to the circumferential weld.

During the circumferential welding procese, this area is most likely to undergo material

changes, resulting in flaw development and potential failure.

This critical area is included in the required volume of material examined during the volumetric scanning of the circumferential weld.

PImposed Alternative:

CNS proposes as an alternative to the Code required volumetric examination and/or surface examination of Class 1

and 2

longitudinal pipe welds, to perform the examinations in accordance with ASME Section XI Code Case N-524,

" Alternative Examination Requirements for Longitudinal-Pipe Welds in Class 1 and 2 Piping;Section XI, Division 1".

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COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM' RELIEF REQUEST NUMBER RI-13 REVISION 0 code Class:

1, 2,

and 3 Subjects ISI of Snubbers included in CNS Technical Specifications components:

All designated Code Class 1, 2,

and 3 snubbers that are identified and included in the Cooper Nuclear Station Inservice Inspection Program Code Requirements:

Reference; ASME Section XI, 1980 Edition with the Winter 1981 Addenda Preservice examination per IWF-2200 and Inservice examination per IWF-2400.

Preservice testing per IWF-5200 and Inservice testing per IWF-1 5300/5400.

Basis for ReliefI Currently, the Cooper Nuclear Station (CNS)

Technical Specifications (TS) include a comprehensive program for visual examination and functional testing of all safety related hydraulic and mechanical snubbers, including all ASME Code Class designated 1,

2, and 3 snubbers.

A significant portion of the safety related snubbers at CNS are also Code Class. The overlap of the visual examination and testing programs per ASME Section XI and Technical Specifications for the Code Class snubbers presents an unnecessary redundancy.

The Technical Specification snubber visual examination program and the visual examination program required by subsection IWF of ASME Section XI are similar in content.

Both programs include parallel criteria for operability, schedule, and sample size.

The Technical Specification snubber testing program and the snubber testing program required by Subsection IWF of ASME Section XI are also very similar in content.

Both programs include parallel requirements for operability testing.

Similar requirements for

' testing frequency, sample

size, and additional sampling for failures are also included in both programs.

Regarding test frequency and sample

size, CNS's Technical Specification Program calls for testing a representative sample of i

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I COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAN RELIEF REQUEST NUMBER RI-13 REVISION 0 10% of the total snubber population every 18 months during a shutdown. Due to the representative sampling, 10% of the Code Class-snubbers will be tested, as required by Subsection IWF of Section XI.

Over a ten outage cycle, 100% of the total snubber population will be tested.

Regarding sample expansion for

failures, CNS's Technical Specificat3on Snubber Program is similar to the requirements of Subsection IWF of Section XI in that for each snubber which does not meet the functional test criteria, an additional 10% of that type of snubber shall be functionally tested.

The snubber visual examination and testing programs of Technical Specification 3.6.H/4.6.H meet the intent of the program required by ASME Section XI, Subsection IWF.

No additional safety benefits will be realized by imposing both programs on the Code class snubbers at CNS.

Proposed Alternatiy_g1 CNS will perform visual examinations of Code Class snubbers in accordance with the latest approved revision of the Station Technical Specifications in lieu of the requirements of IWF-2000 and IWF-2400.

CNS will perform functional testing of Code Class snubbers in accordance with the latest approved revision of the Plant Technical Specifications in lieu of the requirements of IWF-5200 and IWF-5300/5400.

The testing results will be tracked and reported per Station Technical Specifications.

As required by IWF-1300 the examination of Code Class snubber integral and non-integral attachments will be performed in accordance with Subsection IWF.

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COOPER NUCLEAR STATION.

INSERVICE IN8PECTION PROGRAM RELIEF RFQUEST NUMBER RI-14 REVISION 0 code Class:

1, 2 and 3

Subject:

Alternative rules for the selection and examination 1 of Class 1, 2 and 3 integrally welded attachments.

Components:

All integrally welded attachments in examination categories B-K-1, C-C, D-A, and D-B.

Code Requirementat Reference; ASME Section XI, 1980 Edition with the Winter 1981 Addenda B-K-1 Volumetric or surface examination as applicable of integrally welded attachments exceeding 5/8" design thickness.

C-C Surface examination of all integrally welded attachments exceeding 3/4" design thickness.

D-A,B,C Surface examination of -11 integrally welded attachments.

corresponding to those component supports selected by IWF-2510(b).

BARis For Relief Code Case N-509,

" Alternative Rules for the Selection and Examination of Integrally Welded Attachments,Section XI, Division 1,

provides an alternative to the tables of IWB/C/D-2500-1 for integrally welded attachments.

The alternative requires a surface examination of 10% of the integrally welded attachments associated with the component supports selected for examination under IWF-2510.

In addition an examination is required whenever component support member deformation is identified.

This Code Case recognizes the results of over 20 years of inservice inspections and the considerable attention that component supports have j

received through JRC bulletins.

Proposed Alternativer

-l In lieu of performing the Code required examinations, CNS proposes to examine integrally welded attachments in accordance with Code Case N-509 requirements.

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COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAN RELIEF REQUEST NUMBER RI-15 REVISION 0 Code Class:

1 Subjects Inaccessible CRD housing welds.

Components:

Category B-O CRD housing welds.

Code Requirements:

Reference; ASME Section XI, 1980 Edition with the 1981 Addenda Table IWB-2500-1, Category B-O requires a surface examination to be performed on 10% of the peripheral CRD housing welds.

t, Rasis For Relieft There are thirty six CRD housings on the periphery.

Each housing has an upper and lower weld.

A surface examination of 10% of these welds would require the welds in four housings be examined.

The upper CRD housing welds are located inside the reactor vessel skirt.

The twelve inch diameter hole in the reactor vessel support skirt is too small to permit access for a surface examination.

The lower CRD housing welds are partially accessible, however the adjacent CRD housings prevent surface examination of approximately 50% of the weld.

Proposed Alternativ,11 In lieu of performing the Code required examinations, CNS proposes to examine 50% of eight peripheral CRD lower housing welds during the inspection interval and visually examine (VT-2) the remaining CRD housing welds (upper and lower) in conjunction with the class 1 system leakage test after each refueling outage.

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COOPER NUCLEAR STATION INSERVICE IN3PECTION PROGRAM RELIEF REQUEST NUMBER RI-16 REVISION 0 l

Code Class:

1 i

Subjects Examination of Pump Casing and Valve Body Internal Surfaces components:

All designated Category B-L-2 and B-M-2 pump and valves that are identified and included in the CNS Inservice Inspection Program Code Requirements Reference; ASMP 3ection XI, 1980 Edition with the Winter 1981 Addenda Table IWB-2500-1, Code Category B-L-2 and B-M-2 requires a visual (VT-3) examination, to the extent practical, of the component internal pressure boundary.

Basis For nelisi Later editions of ASME Section XI (1989 and later) clarify tnat the VT-3 examination is required only when a pump or valve is disassembled for maintenance, repair or volumetric examination.

This is not stated in the 1980 Editioa through Winter 1981 Addenda of ASME Section XI, but it was not the intent of the Code to require disassembly of pumps and valves solely for the purpose of conducting a visual examination of the internal pressure retaining boundary surfaces. This would require unnecessary maintenance of components without offering a commensurate increase in safety.

PIpposed Alternative In lieu of performing the Code required examinations,.CNS proposes to perform visual (VT-3) examination only when a pump or valve is disassembled for maintenance, repair or volumetric examination.

Examination of the internal pressure boundary shall be performed to the~ extent practical.

Examination is required only once during the inspection interval.

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l }c COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-17 REVISION O Code Class:

1 and 2

Subject:

Integrally Welded Shear Lugs on Piping.

Components:

Category B-K-1 and C-C Integrally Wo3ded Attachments.

code Requirements:

Reference; ASME Section XI, 1980 Edition with the Winter 1981 Addenda For piping, Table IWB-2500-1, Code Category B-K-1, and Table IWC-2500-1, Code Category C-C, require a surface examination of integrally welded attachments meeting the requirements of the Table footnotes. The extent of the examination essentially includes 100%

of the length of the attachment weld at each attachment subject to examination.

Basis For Relief Certain integrally welded attachments on class 1 and 2 pipe supports are shear lugs adjacent to a pipe clamp or restraint.

Component support shear lugs on horizontal piping runs prevent movement of the support along the axis of the pipe.

Shear lugs on vertical piping runs transfer load from the pipe to the support in the downward direction.

Shear lugs are typically welded on the two sides orthogonal to the support by a groove or a fillet weld as shown in Figures IWB-2500-15 or IWC-2500-5(a).

Sometimes the shear lug is attached by a fillet weld all around as shown in Figure IWC-i 2500-5(b).

In order to examine 100% of the surface for 1/2" on l

either side of the weld, the pipe clamp or restraint must be disassembled.

The Code does not usually require a component to be i

disassembled solely for examination.

Disassembly of the component l

support may require considerable time, the erection of an alternate i

support and, depending on location, may result in significant radiation exposure.

Performing a surface examination of the accessible portions of the lug without removal of the component support will cover approximately 80% or more of the required surface and is sufficient to detect service induced flaws in the attachment welds.

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COOPER NUCLEAR STATION INSERVICE INSPECCION PROGRAN RELIEF REQUEST NUMBER RI-17 REVISION O Proposed Alternativer In lieu of performing the Code required examinations, CNS proposes to examine the accessible portions of piping integrally welded attachments contained in Code Categories B-K-1 and C-C without removal of the piping component support.

For those piping component support shear lugs where 100% of the required surface examination area is not obtained the extent of examination will be documented in the record of inspection.

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i COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAN RELIEF REQUEST NUNBER RI-18 REVISION 0 Code Class:

2 Subjects Integrally Welded Attachments Components:

Category C-C, Integrally Welded Attachments to the Residual Heat Removal (RHR) Pump Casing.

Code Requirements:

Reference:

ASME Section XI, 1980 Edition with the Winter 1981 Addenda Table IWC-2500-1 requires a surface examination to be performed on pump integral attachment welds defined by the areas in Figure IWC-2500-5.

Basis for RolleLL Each RHR pump has an integrally welded attachment connecting the pump to the pump baseplate located on the underside of the pump as shown on Figure RR-18-1.

This weld is completely inaccessible and examination is not possible.

Proposed Alternatives As an alternate examination, CNS will perform a VT-2 visual examination of the applicable pump and baseplate in conjunction with the Class 2 system pressure test required by Category C-H.

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COOPER NUCLEAR' STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUNBER RI-18 REVISION 0 FIGURE RI-18-1 RESIDUAL HEAT REMOVAL PUMP INTEGRAL ATTACHMENT W

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.e COOPER NUCLEAR STATION INSERVICE IN8PECTION PROGRAM RELIEF REQUEST NUMBER RI-19 REVISION 0 Code Class:

1 subjects Exemption From Pressure Testing Reactor Vessel Head Flange Seal Leak Detection System.

Components:

Category B-P Pressure Test of Line No. 1-MS-152-1" Code Requirements:

Reference; ASME Section XI, 1980 Edition with the Winter 1981 Addenda IWB-5210 (a) (1) requires that pressure retaining components following opening and closing within each system boundary be subjected to a system leakage test after pressurization to nominal operating pressure.

IWB-5210 (a) (2) requires the pressure retaining components within each system boundary to be subjected to a system hydrostatic pressure test.

Basis for Relief:

The Reactor Vessel Head Flange Leak Detection Line is separated from the reactor pressure boundary by one passive membrane, a silver plated 0-ring located on the vessel flange.

A second 0-ring is located on the opposite side of the tap in the vessel flange (See Figure RR-19-1).

This line is required during plant operation in order to indicate failure of the inner flange seal 0-ring.

Failure of the 0-ring would result in the annunciation of a High Level Alarm in the control room.

On this annunciation, control room operators would quantify the leakage rate from the 0-ring and tnen isolate the leak detection line from the drywell sump by closing the valves NBI-AOV-736AV and NBI-AOV-737AV (see Figure RI-19-1).

This action is taken in order to prevent steam cutting of the 0-ring and the vessel flange.

Failure of the inner 0-ring is the only condition under which this line is pressurized.

The configuration of this system precludes hydrostatic testing while the vessel head is removed bt:ause the odd configuration of the vessel tap coupled with the high test pressure requirement (1000 psig minimum), prevents the tap in the flange from being temporarily plugged.

Adequate testing cannot be performed when the head is installed because the seal prevents complete filling of the line, which has no vent available.

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COOPER NUCLEAR STATION INSERVICE IN8PECTION PROGRAM RELIEF AEQUEST NUMBER RI-19 j

REVISION 0 Aggia for Relief fCentinued)!

j operational testing of this line is precluded because the line will only be pressurized in the event of a failure of the inner 0-ring.

It is extremely impractical to purposely fail the inner 0-ring in order to perform a test.

Based on the above, CNS requests relief from the ASME Section XI requirements for static and operational pressure testing of the Reactor Vessel Head Flange Seal Leak Detection System.

PI.gposed Alternative A VT-2 visual examination will be performed on the line during vessel flood-up in a refueling outage.

The hydrostatic head developed due to the water above the vessel flange during flood-up will allow for the detection of any gross defects in the line.

This examination will be performed once each refueling outage.

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COOPER NUCLEAR STATION INSERVICE INSPECTION PROGRAM RELIEF REQUEST NUMBER RI-19 REVISION O i

FIGURE RI-19-1 HEAD FLANGE SEAL LEAK DETECTION DOUBLE 'O' RING SURFACE N

i N

'N RPV WALL

/

1 -MS-157-1" (NOT TO SCALE)

CLADDING N01-A0V-7.36h x

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NBI-A0V-737k Page 3 of 3 i

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LIST OF NRC COMMITMENTS l

l ATTACHMENT 3 Correspondence Not NLS950119 The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments.

Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGF The proposed alternatives to the individual relief Contingent upon NRC request.s contained in the attachments to NLS950119 will approval requested by be adopted into the Second Ten-Year ISI program September 1, 1995 contingent upon Imc approval.

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l PROCEDURE NUMBER 0.42 REVISION NUMBER 0 PAGE 10 OF 16

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