ML20035C472
| ML20035C472 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/02/1993 |
| From: | Fenech R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9304080010 | |
| Download: ML20035C472 (8) | |
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Temest.ee vaney Autnarny Pest offee Bom 2000. SaddrDaisy. Tennessee 373772000 f
1 Robert A Fenech vce Pres oent. Seouoyah Nuclear Plant l
April 2, 1993 i
i U.S. Nuclear Regulatory Commission
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ATIN: Document Control Desk Washington, D.C. 20555 i
Gentlemen:
In the Matter of
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Docket Nos. 50-327-Tennessee Valley Authority
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50-328 SEQUOYAH NUCLEAR PLANT (SQN) - CLARIFICATION FOR IN-SERVICE PRESSURE TEST (ISPT) RELIEF REQUESTS ISPT-2 AND ISPT-3
Reference:
TVA letter to NRC dated November 17, 1992, "Sequoyah Nuclear j
Plant (SQN) - Units 1 and 2 - Revision to In-Service Pressure Test (ISPT) Program in Support of Cycle 6 Refueling. Outages"
.t In the referenced letter, TVA submitted three relief requests to NRC I
(entitled ISPT-2, -3, and -4) that were associated with American Society of Mechanical Engineers (ASME) pressure test activities scheduled for l
SQN's Cycle 6 refueling outages. Two relief requests' contained in TVA's referenced letter (ISPT-2 and -3) listed Code Case N-498, " Alternate Rules for 10-Year Hydrostatic Pressure Testing for Class 1 and 2' i
Systems," as the impractical requirement.. During subsequent discussions between SQN Site Licensing staff'and NRC, NRC recommended that TVA.-
clarify the impractical requirement section of ISPT-2 and ~3 to reference j
the ASME code requirement in. lieu of the Code Case N-498 alternative requirement. The recommended clarifications to ISPT-2 and -3 are.
j provided in the enclosure. Revision bars are also provided to indicate l
changes.
The_ revised relief requests,(ISPT-2 and--3) will supersede i
those previously provided to you in TVA's' referenced letter.
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U.S. Nuclear Regulatory Commission Page 2 April 2, 1993 Please direct questions concerning this' issue to D. V. Goodin at (615) 843-7734.
Sincerely,
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- r. d Robert A. Fenech Enclosure cc (Enclosure):
Mr. D. E. LaBarge, Project Manager U.S. Nuclear Regulatory Commission One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 l
NRC Resident Inspector Sequoyah Nuclear-Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite.2900 Atlanta, Georgia 30323-0199 i
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ENCLOSURE
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SEQUOYAH NUCLEAR PLANT IN-SERVICE PRESSURE TEST (ISPT) 6 REVISED RELIEF REQUESTS
.i ISPT-2 AND ISPT-3
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REQUEST FOR RELIEF ISPT-2 System:.
Safety injection (63)-
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~I Drawings:
Final Safety Analysis Report.(FSAR) Figure 6.3.2-1 Components:
Pressure boundary piping between:
(1). Hot-leg' injection lines:
l Loop 1 - from check valve 63-641_to check valve 63-640 (8 inches), check valve 63-543 (2 inches), and valve._
j FCV-63-163 (3/4 inch)
I Loop 2 - from check valve 63-559 to check valve 63-547-(2 inches), and valve FCV-63-165 (3/4 inch) 4 Loop 3 - from check valve 63-644 to check valve 63-643 (8 inches), check valve 63-545 (2 inches), and valve FCV-63-164 (3/4 inch)
Loop 4 - from check valve 63-558 to check valve-63-549 (2 inches), and valve FCV-63-166 (3/4 inch)
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i (2) Cold-leg injection lines:
Loop 1_- from check valve 63-560 to check valve 63-622 i
(10 inches), check valve 63-633 (6 inches), check valve 63-551 (2 inches), and valve FCV-63-117 (3/4 l
inch)
Loop 2 - from check valve 63-561 to check valve 63-623-(10 inches), check valve 63-632 (6 inches), check.
valve 63-553 (2 inches), and valve FCV-63-97 (3/4 inch)
Loop 3 - from check valve 63-562 to check valve 63-624' (10 inches), check valve 63-634 (6 inches), check valve 63-555 (2 inches), and valve FCV-63-79 (3/4 inch)
Loop 4 - from check valve 63-563 to check valve 63-625 lI (10 inches), check valve 63-635 (6 inches), check valve 63-557 (2 inches), and valve FCV-63-69 (3/4 inch)
(3) Cold-leg injection lines (1 1/2 inches) from' check valves63-586 (loop 1),63-587 (loop'2),63-588 (loop 3), and 63-589 (loop 4), to check valve 63-581 (3 i
inches) and isolation valve FCV-63-24 (1 inch)
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Class: 1
'!I Function: Reactor coolant system (RCS)' pressure boundary l
Impractical Requirement: American Society of Mechanical Engineers,Section XI, Subsection IWB-2500, Table IWB-2500-1, Category B-P,
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-l Footnote 1 states that the " Entire pressure. retaining' boundary of-the reactor coolant system is subject to system pressure test conducted in accordance with IWA-5000 with
-l the exceptions specified in IWA-5214 when system pressure l
tests are conducted for repaired, replaced or altered-i components."
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Basis For i
Relief: The subject injection line segments are located between the primary and secondary safety-injection check valves. The r
hot-leg injection line segments are not pressurized during normal operation or during cold shutdown. The cold-leg injection line' segments are pressurized to_the pressure of:
the safety-injection accumulators (650 pounds per square inch gauge [psig]) during normal operation.
The pressurization of these line segments to a test pressure equivalent to nominal RCS pressure (2235 psig) i during Modes 4, 5, or 6 is not possible because of
-l insufficient RCS pressure to keep the primary check valve closed against test pressure. Pressurization of these line segments to full RCS pressure during Modes 1, 2, or 3 would risk injection of cold water into the RCS.
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Full compliance with the code would require'either removal l
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of the primary check valve disks or installation of
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temporary piping to provide a flow path around the primary check valve. -This option requires a modification to SQN's RCS, which would place an unusual hardship on the plant staff and would require several days of critical path outage time for inctallation and removal.
I Alternative '
i Testing: The cold-leg injection line segments will be visually examined (VT-2) during the RCS leakage test conducted during start-up following each refueling outage. This leakage test is performed at safety-injection accumulator pressure (nominally 650 psig).
The hot-leg injection line segments will be visually examined (VT-2) once every ten years with the unit in Mode 3.
The pressure during'this test will be the discharge pressure of the safety-injection pump, which is
<q oroximately 1500 psig.
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REQUEST FOR. RELIEF ISPT-3 System: Reactor coolant (68)
Chemical and volume control (62)
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Drawings: FSAR Figures 5.1-1 and 9.3.4-1 Component: Pressure boundary piping between:
(1) Drain lines from:
l Loop 1 - valve 68-549 to 68-550 (2' inches) and 68-551 (3/4 inch) 3 Lcop 2 - valve 68-553 to 68-554 (2' inches) and 68-593 to blind flange (3/4 inch)
Loop 3 - valve 68-581 to 68-582 (2 inches)
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- i Loop 4 - valve 68-557 to 558 (2 inches)
(2) Reactor vessel head vent (3/4 inch) from:
i Valve 68-597 to flange (3/4 inch), valve 68-602 to flange (3/4 inch), valves FSV-68-394 and FSV-68-395 to valves FSV-68-396 and FSV-68-397 l
=i (3) Pressurizer spray vents (3/4 inch) from:
Valve 68-594 to flange, and valve 68-577 to flange (4) Excess letdown drain (3/4 inch) from valve 62-701 to I
(5) Reactor coolant pump seal drain and vent lines (3/4 inch) from:
Loop 1 - valve 62-572 to flange, valve 62-580 to flange r
Loop 2 - valve 62-573 to flange, valve 62-581 to flange Loop 3 - valve 62-575 to flange, valve 62-582 to flange l
Loop 4 - valve 62-574 to flange, valve 62-583 to flange
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Class:
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Function: Reactor coolant pressure boundary
.j 1mpractical l
Requirement: American Society of Mechanical Engineers,Section XI, Subsection IWB-2500, Table IWB-2500-1, Category B-P, Footnote 1 states that the " Entire pressure retaining boundary of the reactor coolant system is subject to system pressure test conducted in accordance with IWA-5000 with the exceptions specified in IWA-5214 when system pressure tests are conducted for repaired, replaced or altered-
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components."
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Basis For t'
Relief: Various piping segments are located in open-end tailpipes that serve as vent, drain, test, or fill lines. Manual j
valves and flanges bound these piping segments to provide the design-required double isolation at the reactor coolant pressure boundary. These piping segments are not normally pressurized.
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Pressure testing of these piping segments at nominal 7
operating pressure in Mode 3 would require that the inboard isolation valve be opened when the reactor coolant system j
(RCS) is at full temperature and pressure (547 degrees j
Fahrenheit and 2235 psig). This action would violate the design requirement for double isolation. valve protection.
The potential for spills when opening the. system presents a significant risk of personnel contamination. Pressure testing in Mode 6 would require that a hydrostatic pump be connected at each segment location. Ilowever, for some segments there is no connection available and would require l
a modification for installation of a pump connection. These piping segments are located in high-radiation areas, and testing would result in high-personnel radiation exposures.
A breakdown of the dose estimates for each radiation area in the plant is provided below:
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RCS Loop Drains 6 items at 10 person-hours per item 300 millirem (mrem)/ hour 18.000 person-roentgen equivalent man (person-rem)
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Reactor Vessel Head Vents 2 items at 10 person-hours per item 1
150 mrem / hour.
t 2 items at 8 person-hours per item 20 mrem / hour 3.320 person-rem t
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Pressurizer Spray Vents
'f 2 items at 10 person-hours per item 200 mrem / hour 4.000 person-rem I
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Excess Letdown Drain 1 item at 8 person-hours per item 50 mrem / hour 0.400 person-rem l
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RCS Seal Drains and Vents j
4 items at 8 person-hours per item
.20 mrem / hour a
4 items at 8 person-hours per item 50 mrem / hour 2.240 person-rem-Based on estimated durations and actual survey data from SQN's Cycle 5 outages, a total dose estimate of 27.960 l
person-rem is predicted for the subject pressure test.
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These piping segments are visually inspected each refueling outage as the unit returns to operation. These segments are I
not specifically pressurized past the first isolation valve i
for this inspection.
It is possible that the piping is pressurized because of leakage at the first isolation valve. With these inspections being performed'approximately six times in each inspection interval, the increase in 1
I safety achieved from the required nominal operating pressure test is not commensurate with the hardship of performing such testing.
Alternative Testing: These piping segments will continue to be visually inspected following each refueling outage for leakage and evidence of past leakage during the RCS leakage test.
This test is conducted with the RCS at full operating temperature and l
pressure.
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