ML050890117
| ML050890117 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/27/2005 |
| From: | Jamil D Duke Power Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 04-CN-004 | |
| Download: ML050890117 (17) | |
Text
Ma Duke FifPower.
A Duke Energy Company D.M. JAMIL Vice President Duke Power Catawba Nuclear Station 4800 Concord Rd. / CNOI VP York, SC 29745-9635 803 831 4251 803 831 3221 tax January 27, 2005 U.S. Nuclear Regulatory Commission Attention:
Document Control Desk Washington, D.C. 20555
Subject:
Duke Energy Corporation Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 Request for Relief Number 04-CN-004 Reply to Request for Additional Information Pursuant to 10 CFR 50.4, please find attached the subject reply.
The format of the reply is to restate the NRC question, followed by Catawba's response.
The NRC questions were transmitted via fax dated November 10, 2004.
There are no regulatory commitments contained in this letter or its attachment.
If you have any questions concerning this material, please call L.J. Rudy at (803) 831-3084.
Very truly yours, LJR/s Attachment www. dukepower. corn
Document Control Desk Page 2 January 27, 2005 xc (with attachment):
W.D. Travers, Regional Administrator U.S. Nuclear Regulatory Commission, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 E.F. Guthrie, Senior Resident Inspector U.S. Nuclear Regulatory Commission Catawba Nuclear Station S.E. Peters, Project Manager (addressee only)
U.S. Nuclear Regulatory Commission Mail Stop 0-8 G9 Washington, D.C. 20555-0001
REQUEST FOR ADDITIONAL INFORMATION CATAWBA NUCLEAR STATION UNITS 1 AND 2 SECOND 10-YEAR INSERVICE INSPECTION INTERVAL REQUEST FOR RELIEF NUMBER 04-CN-004 DOCKET NUMBERS 50-413 AND 50-414 In its letter dated May 18, 2004 Duke Energy Corporation (the licensee) submitted Request for Relief Number 04-CN-004 proposing an alternative pursuant to 10 CFR 50.55a(a)(3)(ii) for portions of ASME Code Class 1 piping and components connected to the Reactor Coolant System (RCS) that are normally isolated from direct RCS pressure (2235 psig) during their normal operation. They are isolated from the reactor coolant loop by their location, either upstream of a check valve, between 2 check valves or between 2 closed valves that must remain closed during the unit's operation (or Startup) in Modes 3, 2 or 1.
The NRC staff has reviewed the licensee's request for relief and has determined that additional information is necessary to complete the review of the relief request. The request for additional information (RAI) is requested below:
- 1.
State the 10-year inservice inspection (ISI) interval in which Request for Relief 04-CN-004 applies and provide the start and end dates of the applicable 10-year ISI interval.
Response to Question 1:
This relief is requested for Catawba Units 1 and 2, second 10-year inspection intervals.
The Unit 1 interval began June 29, 1995 and ends June 29, 2005.
Unit 1 has one more outage prior to the end of the interval.
The Unit 2 interval began August 19, 1996 and ends August 19, 2006.
Unit 2 has one more outage prior to the end of the interval.
- 2.
In the request for relief Code Case N-498-1 was listed as an alternative to rules for system leakage testing.
Was Code Case N-498-1 invoked under Regulatory Guide 1.147, Revision 12 for the applicable 10-year ISI interval program? Since Code Case N-498-4 is now approved for general use in Regulatory Guide 1.147, Revision 13 was it considered for this request for relief?
Attachment Page 1
Response to Question 2:
Catawba is using Code Case N-498-1, "Alternative Rules For Ten Year System Hydrostatic Testing For Class 1, 2, and 3 Systems." This Code Case is approved for use by Request for Relief Serial #94-GO-001. Regulatory Guide 1.147, Revision 12 endorsed Code Case N-498-1 for use with no limitations.
Duke Energy Corporation evaluated Code Case N-498-4 and determined that there was no value added to that from currently used Code Case N-498-1.
- 3.
In the request for relief the licensee referenced the Crane Technical Paper #410 published by the Crane Valve Group of the Crane Valve Manufacturing Company.
Provide a background regarding this paper as to how it applies to the Catawba Nuclear Station, Units 1 and 2.
Has this paper been approved by the NRC for the use at Catawba Nuclear Station, Units 1 and 2?
Response to Question 3:
Crane Technical Paper No. 410, titled, "Flow of Fluids Through Valves, Fitting, and Pipe," 1988, is published by the Engineering Department of Crane Company.
This technical paper provides common mathematical relationships, through various formulas and nomographs, which can be applied to forced fluid flow.
As such, this document presents mathematical equations that are commonly accepted.
Request for Relief No. 04-CN-004, with reference to Crane Technical Paper No.
410, applied the following formula to express the fluid pressure-flow relationship for fluid flow through a fixed orifice.
-xL
=.33L 2235 The above formula is derived from Equation 2-23, on page 2-14, of Crane Technical Paper No. 410, 1988.
L(gpm) = CA 2g(144)dP(psi) 2235 For the example presented in the Request for Relief, if a leak (Li) were projected to be present at 2235 psig (dPi), that same leak would be present at 250 psig Attachment Page 2
(dP2), but at only 33% of the original leak rate.
By maintaining other variables constant and simplifying for like terms, Crane Equation 2-23 becomes:
LeakRate (L2)=
xL1 x L
.33L1 dP 1 2235 As identified in the original Request for Relief, inspections that reveal no leakage at 250 psig (during VT-2 examination) give high confidence that no leakage would be present at 2235 psig.
- 4.
In Section IV "Bases for Requesting Relief" of the licensee's relief the licensee noted in general that there would be physical exposure to significant safety hazards such as high temperatures, and pressures during the Code leakage testing.
Provide additional detail for each "Portion" of the relief regarding physical safety hazards to workers.
Response to Question 4:
Primary system pressure and temperature for Catawba Class A piping is 2235 psig and 557 degrees F. These valves are the first and second valves at the Class A boundary extremity.
Manipulation of these components to reactor coolant system pressure and temperature would subject personnel in these areas to a substantial safety hazard.
- 5.
In Section IV "Basis for Requesting Relief" of the licensee's relief the licensee stated that there would be additional radiation exposure if the Code system leakage tests were performed, but dose rates were not provided.
Provide the dose rates that workers would be exposed to during the Code leakage tests for each "Portion" of the relief.
Response to Question 5:
Note: The following two-letter Duke Energy Corporation system designators are utilized in this and some subsequent responses:
NC -
Residual Heat Removal/Low Head Injection System NI -
Intermediate Head Injection System NV -
Chemical and Volume Control/High Head Injection System Attachment Page 3
The following sketches give locations and dose rates for valves that are in Portions 1 through 7.
02203'.
Dose rates In his area are 500 rnrevhrcontact and 200 mremihr General Area Attachment Page 4
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Attachment Page 5
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- 6.
In Portion 3 the length of the segment of the piping was identified to be 3 inches.
Provide the segment length of the piping in Portions 1, 2, 4, 5, 6, and 7 in which relief is requested.
Response to Question 6:
The following are the approximate segment lengths for portions 1, 2, 3, 4, 5, 6, and 7.
Portion #1 Portion #2 Portion #3 Portion #4 Portion #5 Portion #6 Portion #7
= 97 feet
= 302 feet
= 32 feet
= 765 feet
= 383 feet
= 77 feet
= 20 feet
- 7.
In the provided P&IDs attached to the licensee's request for relief the material of the piping was Attachment Page 11
listed as stainless steel, but was not specific regarding the type of stainless steel nor were there any discussions in the relief on corrosion mechanisms.
Provide the material type for each "Portion" or piping segment of the relief and if any corrosion mechanisms have been identified during the service life of the subject piping segments.
Response to Question 7:
Throughout the Question 7 response, the following template is utilized for the information transmitted (where applicable):
Catawba Pipe Specification (PS), Design Pressure, Design Temperature, Pipe Diameter and Schedule, ASME Piping Materials Designation, Components in Portion or Piping Segment (e.g., fittings, socket welds (SW), butt welds (BW)), Component Rating, ASME Component Materials Designation Portion 1:
Auxiliary spray line, boundary valves INV37A, 1NV861, 1NV38 PS 2501.1, 2500 psia, 650 degrees F, 2" SA376 Grade 304, fittings, SW, SA182, F304 Portion 2:
Residual heat removal suction off reactor coolant loop, boundary valves 1ND1B,
- 1ND2A, 1ND36B, 1ND37A PS 2501.1, 2500 psia, 650 degrees F, 12" SA376 Grade 316, fittings, BW, SA403, 316 Portion 3:
NV to NI safety injection to cold legs, boundary valves 1NI351, 1NI15 (loop A), 1NI352, 1NI17 (loop B), 1NI353, 1NI19 (loop C), INI354, INI21 (loop D)
PS 2501.1, 2750 psia, 650 degrees F, 1-1/2" schedule 160, SA376 Grade 304, fittings, SW, 6000 lb, SA182, F304 Portion 4:
NI cold leg accumulator to cold legs, boundary valves 1NI54A, 1NI60, INI181, INI165 (loop A), 1NI65B, INI71, 1NI180, 1NI167 (loop B), 1NI76A, 1NI82, 1NI175, INI171 (loop C), 1NI88B, lNI94, lNI176, lN1169 (loop D)
PS 2501.1, 2500 psia, 650 degrees F, 10" schedule 140, SA376 Grade 316, fittings, BW, SA403, 316 PS 2501.1, 2500 psia, 650 degrees F, 6" schedule 160, SA376 Grade 304, fittings, BW, SA403, 304 Attachment Page 12
PS 2501.1, 2500 psia, 650 degrees F, 2" schedule 160, SA376 Grade 304, fittings, SW, 6000 lb, SA182, F304 Portion 5A:
NI safety injection to hot legs, boundary valves 1NI157, 1NI156 (loop A), INI160, 1NI159 (loop D)
PS 2501.1, 2500 psia, 650 degrees F, 6" schedule 160, SA376 Grade 304, fittings, BW, SA403, 304 PS 2501.1, 2500 psia, 650 degrees F, 4" schedule 160, SA376 Grade 304, fittings, BW, SA403, 304 PS 2501.1, 2500 psia, 650 degrees F, 2" schedule 160, SA376 Grade 304, fittings, SW, 6000 lb, SA182, F304 Portion 5B:
NI safety injection to hot legs, boundary valves INI129, 1N1128, INI134 (loop B), 1NI125, 1NI124, 1NI126 (loop D)
PS 2501.1, 2500 psia, 650 degrees F, 8" schedule 160, SA376 Grade 304, fittings, BW, SA403, 304 PS 2501.1, 2500 psia, 650 degrees F, 6" schedule 160, SA376 Grade 304, fittings, BW, SA403, 304 PS 2501.1, 2500 psia, 650 degrees F, 2" schedule 160, SA376 Grade 304, fittings, SW, 6000 lb, SA182, F304 Portion 6:
Reactor coolant loop drain lines, boundary valves lNC4, lNC5 (loop A), 1NC94, 1NC95 (loop B), lNC13, iNC115, 1NC106 (loop C), lNC19, 1NC20 (loop D)
PS 2501.1, 2500 psia, 650 degrees F, 2" schedule 160, SA376 Grade 304, fittings, SW, 6000 lb, SA182, F304 Portion 7:
Reactor coolant loop vent lines, boundary valves 1NC298, lNC299, bypass valves lNC311, 1NC312 PS 2501.1, 2500 psia, 650 degrees F, 3" schedule 160, SA376 Grade 304, fittings, BW, SA403, 304 The above listed materials are all austenitic stainless steel.
These stainless steels have performed without evidence of general corrosive attack in pressurized water reactor coolant service.
There are no known concerns with flow accelerated corrosion, erosion, or boric acid corrosion.
Stress Corrosion Cracking (SCC) has been observed in isolated cases with Type 304 and 316 stainless steels based on oxygenated reactor coolant environments that experience long periods of stagnation.
However, there are restrictive controls on the reactor coolant system chemistry, including oxygen and chlorides at Catawba during startup and operation.
These controls, along with periodic flushes during Attachment Page 13
refueling outages, limit the exposure of this piping to this detrimental environment.
Furthermore, Catawba has no history of corrosion degradation of these lines and volumetric examinations of the welds within this scope of piping to date have not indicated any reportable indications.
- 8.
Provide a discussion on the ASME Code,Section XI nondestructive examinations, if any that the subject piping segments received or will receive during the plant's current 10-year inservice inspection program plan.
Response to Question 8:
Portion 1:
Segments of this piping receive a surface examination in accordance with Examination Category B-J, Pressure Retaining Welds in Piping (B9.40).
Portion 2:
Segments of this piping receive a volumetric and surface examination in accordance with Examination Category B-J, Pressure Retaining Welds in Piping (B9.11).
Portion 3:
Segments of this piping receive a surface examination in accordance with Examination Category B-J, Pressure Retaining Welds in Piping (B9.40).
Portion 4:
Segments of this piping receive a volumetric and surface examination in accordance with Examination Category B-J, Pressure Retaining Welds in Piping (B9.11, B9.12, B9.32, B9.40).
Portion 5:
Segments of this piping receive a volumetric and surface examination in accordance with Examination Category B-J, Pressure Retaining Welds in Piping (B9.11, B9.21, B9.40).
Portion 6:
Segments of this piping receive a surface examination in accordance with Examination Category B-J, Pressure Retaining Welds in Piping (B9.21, B9.40).
Portion 7:
Attachment Page 14
Segments of this piping receive a surface examination in accordance with Examination Category B-J, Pressure Retaining Welds in Piping (B9.21).
Attachment Page 15