ML20214J354
| ML20214J354 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/22/1987 |
| From: | Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 86-796A, NUDOCS 8705280019 | |
| Download: ML20214J354 (30) | |
Text
10 CFR 50.55a VIHOINIA 13LECTHFC AND l'OWEH COMPANY R icar woxo, Vino ix A 20u6:
w.L.STEWANT Vice Passionvr woci.m*= oren.r m s.
gg United States Nuclear Regulatory Commission Serial No.
86-796A Attention: Document Control Desk N0/DJV:jmj Washington, D.C.
20555 Docket Nos.
50-338 License Nos.
NPF-4 Gentlemen:
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 ASME SECTION XI RELIEF REQUEST SUPPLEMENTAL INFORMATION Virginia Electric and Power Company letter dated February 6, 1987 requested relief from various examination requirements of ASME Section XI for the first inservice examination interval for North Anna Unit 1.
In a response to a request by your staff, this letter provides supplemental information regarding several of the relief requests.
The supplemental information is provided in the attachments which contain revised relief requests from nondestructive examinations (Attachment 1) and hydrostatic testing (Attachment 2) requirements. The complete attachments from our February 6, 1987 letter are included with revisions indicated by a vertical line in the right margin.
Very truly yours, W. L. Stewa Attachments Ok(i gR0528001, e,032, T $
ADOCK 05000338 p
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U. S. Nuclear Regulatory Commission
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101 Marietta Street, N.W.
Suite 2900 Atlanta, GA 30323 4
Mr. J. L. Caldwell NRC Senior Resident Inspector
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North Anna Power Station 1
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Attachmsnt 1 Page 1 of 12 RELIEF REQUESTS FOR NDE EXAMINATIONS Relief Request Description Page 1
B5.6, B-L-1 and B5.7, B-L-2
. Pump casing weld and 2
Pump Casing.
2 B6.7, B-M Valve Body 4
3 B2.9, B-I-2 and B3.8, B-I Cladding 5
4 B4.10, B5.5 and B6.5 of Category B-K-2 and C2.6, C3.4 and 7
C4.4 of Category C-E Support Componets 5
B5.1, B5.2, 5.3, B6.1, B6.2 and B6.3 of Category B-G 8 Pressure Retaining Bolting 6
B4.12 and B6.9 of Category B-G Pressure Retaining Bolting 9
7 C1.2, C-B - Nozzle to Vessel Welds 10 8
C1.4, C-D - Pressure Retaining Bolting 12 l
Page 2 of 12 ASME SECTION XI RELIEF REQUEST FROM CLASS 1 PUMP CASING EXAMINATIONS NORTH ANNA POWER STATION, UNIT 1 ASME SECTION XI EXAMINATION REQUIREMENT - 1974 Edition through Summer 1975 Addenda Item No. B5.6, Category B-L-1:
The examinations performed during each inspection interval shall include 100% of the pressure-retaining welds in at least one pump in each group of pumps performing similar functions in system (e.g., recirculating coolant pumps).
The examination method shall be volumetric.
Item No. B5.7, Category B-L-2:
One pump in each of the group of pumps performing similar functions in the system shall be examined during each inspection interval.- This examination may be performed on the same pump selected for the Category B-L-1 examination. The examination method shall be visual.
ALTERNATE EXAMINATION PROPOSED FOR ITEM NOS. B5.6 and B5.7 In lieu of Volumetric Examinations, Virginia Electric and Power Company proposes to perform:
1) 100% visual examination of the external surfaces only of one pump's casing welds to the extent and frequency of Examination Category B-L-2.
- 2) Surface examination to the extent practicable of the external casing weld (s) of one pump only, to the conditions and frequency of Examination Category B-L-2.
REASON FOR RELIEF REQUEST The North Anna Power Station Unit I reactor coolant pumps are Westinghouse Model 93 controlled leakage pumps. The Model 93 pumps casing is fabricated by welding four stainless steel castings together. Thus, there are three circumferential pressure boundary welds in the pumps that are to be examined in accordance with Category B-L-1 & B-L-2.
Since the installation of these pumps, it has been recognized that a volumetric examination of the casing welds is not practical with today's ultrasonic techniques.
Page 3 of 12 The physical properties of the stainless steel casting and weld material preclude a meaningful ultrasonic examination. Thus, the capability to examine these pump casing welds in the field did not exist until recently. In the spring of 1981 an examination was performed on one of the reactor coolant pumps at the R.E. Cinna plant using the miniature linear accelerator (MINAC), which was built under an EPRI sponsored program. This equipment has been made available to other utilities, and currently constitutes the only method available for the volumetric examination of reactor coolant pump casing welds.
The volumetric examination method is radiographic and is performed by placing the MINAC inside the pump casing and placing film on the outside of the pump. To perform the examination, the pump must be completely disassembled, including removal of the diffuser adapter and casing adapter. This amount of disassembly is far beyond the amount of disassembly performed for normal maintenance. Insulation must also be removed from the exterior of the pump casing.
The examination has been performed at four different sites, all of which have the Westinghouse Model 93 pump. The MINAC examination was performed at Ginna in the spring of 1981, at Point Beach Unit 1 in the fall of 1981, at Turkey Point Unit 3 early in 1982, and at H.B. Robinson Unit 2 later in 1982. No problems with the welds were found at any of the sites. A review of the original radiographs of the Point Beach Unit 1 pump was performed prior to the MINAC examination, and all the landmarks found were identified during the field examination with no apparent change.
The successful performance of this volumetric examination using the MINAC at four different sites demonstrates that the method is capable of satisfying ASME Section XI examination requirements; however, the performance of the examination has shown that there is a relatively high radiation exposure associated with it.
The total exposure associated with insulation removal, disassembly, examination, and reassembly of the pump has averaged about 40 man-rem.
There have been no defects identified by the four examinations performed on these pumps to date. A volumetric examination was attempted at North Anna in 1982. A radioactive source was placed within the pump casing and film around the outside. The developed film did not meet the density requirement for an acceptable examination. This examination was attempted twice at Surry. Both examinations yielded similar results.
The pumps casing examinations are also not justified from a cost / benefit perspective. The pump disassembly, examination and reassembly is estimated to cost $750,000.
Page 4 of 12 ASME SECTION XI RELIEF REQUEST FROM CLASS 1 VALVE BODY VT EXAMINATIONS NORTH ANNA POWER STATION, UNIT 1 ASME SECTION XI EXAMINATION REQUIREMENT - 1974 Edition through Summer 1975 Addenda Item No. B6.7, Category B-M-2:
Visual examination of the internal pressure boundary surfaces, on valves exceeding 4 in. nominal pipe size. One valve in each group of valves of the same constructional design, e.g., globe, gate, or check valve, manufacturing method and manufacturer that performs similar functions in the system shall be examined during each inspection interval. The examination may be performed at or near the end of the inspection interval.
ALTERNATE EXAMINATION PROPOSED In lieu of examination of each similar valve's interior on lines 4 inch nominal pipe size and larger during the interval, Virginia Electric and Power Company proposes to examine only those valves in this category which are disassembled during the remainder of the first interval for maintenance purposes.
REASON FOR RELIEF REQUEST Disassembly of a valve which has been functioning within acceptable parameters for the sole purpose of examination is contrary to good maintenance practices since the likelihood of failure may be increased. These components are subjected to an alternate form of performance and/or leakage monitoring, such as inservice valve testing, Appendix J leak rate testing, or primary coolant system leak detection.
Valves in this category are constructed of cast austenitic stainless or carbon steels, which have been identified as unlikely to experience failure by cracking. Finally, considering the uncertain benefit involved, it is difficult to justify the additional radiation exposure which would be incurred in the valve disassembly, examination, and valve reassembly.
We, therefore, believe that performing a visual examination of the interior of one valve in a group of similar valves within the Class 1 pressure boundary at North Anna Power Station Unit I during the first ten year inservice inspection interval does not provide an increase in safety above that provided by routine inservice valve testing and pressure testing required by ASME Section XI.
Therefore, the costs and radiation exposure associated with this examination are not justifiable.
Based en the preceding factors, we request relief from performing a visual examination of the interior surface of one valve in a group of similar valves on pipir.g 4 inch nominal pipe size and larger at North Anna Power Station Unit 1 during the first ten-year inservice inspection interval, unless the valve is disassembled for maintenance between now and the end of the interval.
Page 5 of 12 ASME SECTION XI RELIEF REQUEST FROM VT EXAMINATIONS OF CLASS 1 INTERIOR CLAD SURFACES OF VESSELS OTHER THAN REACTOR VESSELS NORTH ANNA POWER STATION, UNIT 1 ASME SECTION XI EXAMINATION REQUIREMENT - 1974 Edition through Summer 1975 Addenda Item No. B2.9, Category B-I-2 Pressurizer Cladding and Item No. B3.8, Category B-I-2, Heat Exchanger and Steam Generator Cladding "The examinations performed during each inspection interval shall cover 100% of the patch areas." The exam method shall be visual.
I ALTERNATE EXAMINATION PROPOSED FOR ITEM NOS. B2.9 and B3.8 Virginia Electric and Power Company proposes no additional examinations in these categories since later editions and addenda of ASME Section XI approved by the NRC and incorporated into 10 CFR 50.55a no longer require cladding examinations.
REASON FOR RELIEF REQUEST The Edition of the ASME Code,Section XI in effect for the first ten year inservice inspection interval at North Anna Power Station is the 1974 Edition with Addenda through Summer 1975. This edition of the Code requires reactor vessel, pressurizer and steam generator cladding examinations.
Subsequent edition and addenda to the Code which have been approved by the NRC for incorporation into 10 CFR 50.55a have deleted the cladding inspection.
Recognizing this deletion and the intent of the ASME Section XI examination to provide monitoring of component degradation over the plant's service interval, it is our position that the radiation exposure and cost associated with the cladding examinations are not commensurate with the increase in safety realized. The clad examination results obtained during the first inspection interval will not be directly comparable to examination results in later intervals.
Table 1 in the relief request presents a summary of the cladding examinations required by the first ten year inservice inspection plan and the results of those examinations performed to date.
Based on the preceding factors, we request relief from the remaining Pressurizer and Steam Generator cladding examinations at North Anna Power Station Unit 1 during the first ten year inservice inspection inte rval.
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Page 6 of 12 TABLE 1 - CLADDING EXAMS i
Pressurizer - Isometric Number - VRA-1-2100 Component Number Exam Date Results CP01 Scheduled for REF 87*
Steam Generator - Isometric Number - VRA-1-3100 Exam Date Results CP01 A REF 85 No Reportable Indications CP02 A REF 85 No Reportable Indications CP03 B Scheduled for REF 87*
CP04 B Scheduled for REF 87*
CP05 C REF 85 No Reportable Indications CP06 C REF 85 No Reportable Indications
- Relief Proposed REF 85 - Refueling Outage 1985 s
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Page 7 of 12 ASME SECTION XI RELIEF REQUEST FROM VT EXAMINATION OF CLASS 1, 2, & 3 HYDRAULIC SHOCK SUPPRESSORS NORTH ANNA POWER STATION UNIT 1 ASME SECTION XI EXAMINATION REQUIREMENTS - 1974 Edition through Summer 1975 Addenda Item Nos. B4.10, B5.5, and B6.5 Examination Category B-K-2 Support Components and Item NOS. C2.6, C3.4, and C4.4 Examination Category C-E-2, Support Components The area shall include the support components that extend from the piping, valve and pump attachment to and including the attachment to the supporting structure. The examination performed during each inspection l
interval shall cover all support components.
i The support settings of constant and variable spring type hangers, snubbers, and shock absorbers shall be verified.
Category IWD-2600(c) Support Components
" Supports (restraints), and hangers for components exceeding four-inch nominal pipe size whose structural integrity is relied upon to withstand i
design loads when the system function is required shall be visually examined to detect any loss of support capability and evidence of inadequate restraint."
ALTERNATE EXAMINATION PROPOSED FOR ITEM NOS. B4.10, B5.5, B6.5, C2.6, C3.4, C4.4, & CATEGORY IWD-2600(c)
Virginia Electric and Power Company requests permission to examine the i
hydraulic shock suppressors on Class 1,2, and 3 systems according to the requirements of the facility Technical Specifications in lieu of the Section XI requirements listed above.
REASON FOR RELIEF REQUEST The hydraulic shock suppressors on Class 1,2, and 3 systems are currently subjected to an ongoing examination and testing program detailed in the Technical Specifications which exceeds the requirements of ASME Section XI. The requirements of the Technical Specifications specify examinations to be performed at a minimum frequency of 18 months. The examination frequency is increased for test failures.
This l
program is designed to demonstrate continued operational readiness and structural integrity of shock suppressors. The visual examination boundary required by Technical Specifications covers the same areas as the ASME Section XI examination boundary. Additional visual examinations are performed beyond the ASME Section XI requirements to determine operability and freedom of movement. The Technical Specifications requirement for snubber examinations meet and exceed the ASME Section XI visual examination requirements for component supports.
Page 8 of 12 ASME SECTION XI RELIEF REQUEST FROM 100% NDE EXAMINATION ON PRESSURE RETAINING BOLTING 2 INCHES IN DIAMETER AND LARGER NORTH ANNA POWER STATION UNIT 1 ASME SECTION XI EXAMINATION REQUIREMENTS - 1974 Edition through Summer 1975 Addenda Examination Category B-G-1, Pressure Retaining Bolting, 2 Inches and Larger.
Item Nos. B5.1, B5.2, B5.3, B6.1, B6.2, and B6.3.
The areas shall include bolts, studs, nuts, bushings, washers, and threads in base material and flange ligaments between threaded stud holes.
The examinations performed during each inspection interval shall cover 100% of the bolts, studs, nuts, bushings, and threads in base material and flanges ligaments between threaded stud holes.
Bushings, threads, and ligaments in base material of flanges are required to be examined only when the connection is disassembled.
Bolting may be examined either in place under tension, when the connection is disassembled, or when the bolting is removed.
The examination shall be volumetric for item nos. B5.1 and B6.1.
Volumetric and surface examinations are required for item nos. B5.2 and B6.2.
A visual examination shall be performed on item nos. B5.3 and B6.3.
ALTERNATE EXAMINATION PROPOSED FOR ITEM NOS. B5.1, B5.2, B5.3, B6.1, B6.2, AND B6.3 Impracticality for performing the required examinations of the components in this category cannot be shown at this time. Virginia Electric and Power Company, therefore, withdraws this request.
Page 9 of 12 ASME SECTION XI RELIEF REQUEST FROM NDE EXAMINATION ON CLASS 1 PRESSURE RETAINING BOLTING LESS THAN 2 INCHES IN DIAMETER NORTH ANNA POWER STATION UNIT 1 ASME SECTION XI EXAMINATION REQUIREMENTS - 1974 Edition through Summer 1975 Addenda Examination Category B-G-2 Pressure Retaining Bolting, Less than 2 Inches.
1 Item No. B4.12 Piping Pressure Boundary and Item No. B6.9 Valve Pressure Boundary "The area shall include bolts, studs and nuts. The examinations performed during each inspection interval shall cover 100% of the bolts, studs, and nuts. Bolting may be examined in place under tension, when the connection is disassembled, or when the bolting is removed. The examination shall be visual."
ALTERNATE EXAMINATION PROPOSED FOR ITEM NOS. B4.12 and B6.9 Impracticality for performing a visual examination of the remaining 10 valves in the category cannot be shown. Virginia Electric and Power Company, therefore, withdraws this relief request.
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Page 10 of 12 ASME SECTION XI RELIEF REQUEST FROM UT EXAMINATION ON CLASS 2 N0ZZLE TO VESSEL WELDS NORTH ANNA POWER STATION UNIT 1 l
3 ASME SECTION XI EXAMINATION REQUIREMENT - 1974 Edition through Summer 1975 Addenda Item No. C1.2, Category C-B: The examination shall include 100% of the nozzle to vessel attachment welds. The examination method shall be volumetric.
4 ALTERNATE EXAMINATION PROPOSED Due to the configuration of the RHR heat exchanger nozzle (reinforcing plate), Virginia Electric and Power Company proposes to perform:
1) 100% visual examinations of the external surfaces of the weld and reinforcing plate during hydrostatic pressure test.
2)
Insulation will be removed to allow for inspection of the tell-tale hole for leakage during hydrostatic testing.
REASON FOR RELIEF REQUEST The nozzle to vessel welds of the RHR heat exchangers are completely
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covered by a reinforcing plate and are not accessible for examination as required by IWC-2500. See attached sketch. The reinforcing plate 4
covering the RHR heat exchanger nozzle-to-vessel welds contains tell-tale holes such that visual examinations can be performed for evidence of leakage.
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I Page 12 of 12 ASME SECTION XI RELIEF REQUEST FROM NDE EXAMINATION ON CLASS 2 BOLTING LESS THAN 2 INCHES IN DIAMETER NORTH ANNA POWER STATION UNIT 1 ASME SECTION XI EXAMINATION REQUIREMENT - 1974 Edition through 1975 Addenda Item No. C1.4, Category C-D: " Visual examinations performed during each inspection interval shall cover 100% of the bolts, studs, nuts, bushings, washers and threads in base material and flange ligaments between threaded studs holes.
Nondestructive examination shall be performed on 10% of the bolting in each joint, but not less than two bolts or studs per joint.
Bushings, threads, and ligaments in base material of flanges are required to be examined only when the connection is disassembled.
Bolting may be examined either in place under tension, when the connection is disassembled or when the bolting is removed.
The examination shall be visual and either surf ace or volumetric."
ALTERNATE EXAMINATION PROPOSED The original intent of this relief request was to take advantage of later editions of the code.
Impracticality of examination for the remaining 4 valves in this category cannot be shown. Virginia Electric and Power Company withdraws this relief request.
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Paga 1 of 15 RELIEF REQUESTS FOR 10-YEAR HYDE 0 STATIC TESTS Relief Request Description Page 1
Chemical and Volume Control - Over 2
pressurize Reactor Coolant Pump Seal 2
Chemical and Volume Control - Check valves prevent lines from being pressurized 3
3 Chemical and Volume Control - Check 4
valves prevent lines from being pressurized 4
Feed Water - Check Valves allow over-5 pressurizing Steam Generator 5
Feedwater and Chemical and Volume Control -
6 Overpressurizing Pump Suction Piping 7
6 Reactor Coolant - Reactor Vessel inspection 7
7 Residual lleat Removal - Class I Components 8
8 Safety Injection - Check Valves allow 9
pressuring Reactor Coolant System 9
Safety Injection - Overpressurizing Reactor 10 Coolant System 10 Safety Injection - Check Valves allow over-11 pressurizing Reactor Coolant System 11 Various - Request testing per Steam 12 Generator Manufacturer 12 Various - Test Pressure Limit 15 i
Page 2 of 15 RELIEF REQUEST 1 SYSTEM - Chemical and Volume Control COMPONENTS - Piping on drawing 11715-FM-95C between the pumps and first flange from pumps.
PUMPS LINE 1-RC-P-1A 2"-CH-14-1502 1-RC-P-1B 2"-CH-15-1502 1-RC-P-1C 2"-CH-16-1502 CODE REQUIREMENTS - System Hydrostatic Test per IWB-5222 PROPOSED ALTERNATIVE EXAMINATION - The normal system leakage test after each refueling is an adequate examination.
REASON FOR RELIEF - Pressurizing the piping listed above will also pressurize the number one seal of the reactor coolant pumps. This could potentially damage the number one seal. Relief was granted to Surry Power Station Unit 2 per Safety Evaluation Report dated 01/24/86 for the same situation as described above.
Page 3 of 15 RELIEF REQUEST 2 SYSTEM - Chemical and Volume Control COMPONENTS Piping located on drawing 11715-FM-95C between the valves listing below Valves Line HVC-1311 and 1-CH-328 2"-CH-68-1502 1-CH-325 and 1-CH-496 3"-CH-1-1502 CODE REQUIREMENTS - Class 1 System Hydrostatic Test per IWB-5222.
Po=2500 psig, To=496*F, Test Pressure is 2550 psig per IWB-5222.
PROPOSED ALTERNATIVE EXAMINATION - As an alternative, the Reactor Coolant System will be pressurized to a pressure as close as practical to 2335 psig but not less than 2300 psig while the reactor is in a shutdown condition to create a pressure boundary at Check Valves 1-CH-328 and 1-CH-325.
The components listed above will then be tested to a pressure (2300 psig < test pressure < 2335 psig) as close as practical to the Reactor Coolant System pressure using a charging pump.
REASON FOR RELIEF - Check valves 1-CH-328 and 1-CH-325 prevent the components listed above from being pressurized without pressurizing the Reactor Coolant System. The code required test pressure of 2550 psig will overpressurize the Reactor Coolant System.
I Also, the power operated relief valves (PCV-1456 and PCV-1455C) of the Reactor Coolant System are designed to limit the pressurizer pressure to a value below the fixed high-pressure reactor trip setpoint (2385 psig). The relief valve setpoints are 2335 psig.
It is not desirable to take the Reactor Coolant System above the power operated relief valve setpoint.
Similar relief was granted to Surry Power Station Unit 2 per Safety Evaluation Report dated 01/24/86 (Docket 50-281) for components with the same design configuration, i
Page 4 of 15 RELIEF REQUEST 3 SYSTEH - Chemical and Volume Control COMPONENTS - Piping between the valves listed below located on drawings 11715-FM-95C and'11715-FM-95B.
Valves Line 1-CH-496, 1-CH-HCV-1311, and 1-MOV-1289A 3/4"-CH-240-1502 2"-CH-68-1502 3"-CH-1-1502 3"-CH-79-1502 CODE REQUIREMENT - Class 2 System Hydrostatic Test per IWC-5222.
Since there are no relief valves for the above components, test pressure per IWC-5222 is 3419 1
psig.
PROPOSED ALTERNATE EXAMINATION - As an alternative, the Reactor Coolant System will be pressurized to a pressure as close as practical to 2335 psig but not less than 2300 psig while the reactor is in a shutdown condition to create a pressure boundary at Check Valves 1-CH-328 and 1-CH-496.
The components listed above will then be tested to a pressure (2300 psig < test pressure <
2335 psig) as close as practical to the Reactor Coolant System pressure using a charging pump.
1 REASON FOR RELIEF - Check Valves 1-CH-328, 1-CH-325 and 1-CH-496 prevent the components listed above from being pressurized without pressurizing the Reactor Coolant System.
The Code required test pressure of 3419 psig will overpressurize the Reactor Coolant System.
Also, the power operated relief valves (PCV-1456 and PCV-1455C) of the Reactor Coolant System are designed to limit the pressurizer pressure to a value be.aw the fixed high-pressure reactor trip setpoint (2385 psig). The relief valve setpoints are 2335 psig.
It is not desirable to take the Reactor Coolant System above the power operated relief valve setpoint.
Page 5 of 15 RELIEF REQUEST 4 SYSTEM - Feedwater COMPONENTS - Piping between the valves listed below located on drawing 11715-FM-74A.
4 Valve Connecting Line Valve 1-FW-62 3"-WAPD-10-601 to 3"-WAPD-9-601 1-FW-68 1-FW-64 3"-WAPD-9-601 1-FW-68 1-FW-93 3"-WAPD-12-601 to 3"-UAPD-11-601 1-FW-100 1-FW-96 3"-WAPD-11-601 1-FW-100 1-FW-126 3"-WAPD-14-601 to 3"-WAPD-13-601 1-FW-132 1-FW-128 3"-WAPD-13-601 1-FW-132 1-FW-278 4"-WAPD-39-601 to 3"-WAPD-10-601 1-FW-68 CODE REQUIREMENTS - Class 2 System Hydrostatic Test per IWC-5222 and IWA 5213(d). P =1400 psig, T <200*F, Test Pressure d
d is 1540 psig per IWC-5222.
PROPOSED ALTERNATIVE EXAMINATIONS - Since the components listed above cannot be pressurized without pressurizing the steam generators, they must be tested per the required manufacturer's hydrostatic test method for the steam generators. Therefore, the proposed alternative examination is the examination described in the Westinghouse Technical Manual for the secondary side of the steam generators. The examination is to pressurize the secondary side of the steam generators to 1356 psig, hold for 30 minutes, and then reduce to the design pressure (1085 psig) for 3 1/2 hours. A VT-2 examination will then be performed.
REASON FOR RELIEF - Due to check valves 1-FW-132, 1-FW-100, and 1-FW-68 the piping listed above cannot be pressurized without pressurizing the steam gecorators. The code required test pressure of 1540 psig would over-pressurize the steam generators.
6 Page 6 of 15 RELIEF REQUEST 5 SYSTEM - Feedwater, Chemical and Volume Control and Safety Injection COMPONENTS - Centrifugal pumps and discharge piping to first isolation valve on drawings 11715-FM-74A, 11715-FM-95B, 11715-FM-96A Pumps 1-FW-P-2, 1-FW-P-3A, 1-FW-P-3B 1-CH-P-1A, 1-CH-P-1B, 1-CH-P-1C 1-SI-P-1A,1-SI-P-1B CODE REQUIREMENT - System Hydrostatic Tests per IWC-5222 and IWD-5223.
PROPOSED ALTERNATIVE EXAMINATION - As an alternative, the test pressure for the pump discharge and associated piping extending to the first shutoff valve on the discharge side of.he pump shall be the same as that required for the piping and components on the suction side of the pump.
In support of this alternative, the 1980 Edition, winter of 1981 Addenda of ASME Section XI (approved by the NRC) paragraph IWA-5224(d) allows the system test boundary interface to be the first shutoff valve on the discharge side of the centrifugal pump when the primary system pressure ratings on the suction and discharge sides differ, REASON FOR RELIEF - Centrifugal pumps and the portions of the pump discharge lines up to the first isolation valve cannot be isolated from the pump suction. If the discharge piping were pressurized to the required test pressure, then the suction piping and/or pump would be subjected to a pressure far in excess of its design with the potential for permanent damage to piping and components.
1
Page 7 of 15 RELIEF REQUEST 6 l
SYSTEM - Reactor Coolant COMPONENTS - Bottom of reactor vessel CODE REQUIREMENTS - Visual Examination per IWA-5240 during System Hydrostatic Test.
PROPOSED ALTERNATIVE EXAMINATION - The proposed alternative examination is to examine the bottom of the reactor vessel for evidence of leakage during the 10 year vessel inspection.
REASON FOR RELIEF - The system hydrostatic test for the Reactor Coolant System is performed at Hot Standby, MODE 3.
The bottom of the reactor vessel is inaccessible due to temperature and radiological concerns.
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j RELIEF REQUEST 7 f
SYSTEH - Residual Heat Removal l
COMPONENTS - Piping located on drawing 11715-FM-94A between valves l
listed below.
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Valves Line MOV-1701 and MOV-1700 14"-RH-1-1502 CODE REQUIREMENTS - Class 1 System Hydrostatic Test IWB-5222.
Po=2235 psig, To=650 degrees F, Test pressure per IWB-5222 is 2280 psig.
PROPOSED ALTERNATIVE EXAMINATION - As an alternative, the components listed above will be tested in accordance with 4
IWC-5222. The test pressure will be 584 psig as determined by the setpoint of relief valves RV-1721A and RV-1721B (467 psig).
This i
alternative is considered sufficient since the 1
relief valves are set at 467 psig. As a result, line 14-RH-1-1502-Q1 should not see a pressure significantly higher than 467 psig.
In addition, MOV-1700 and MOV-1701 will not open if the Reactor j
Coolant pressure is >660 psig.
t REASON FOR RELIEF - During the system hydrostatic test of the primary system, MOV-1700 is closed in addition to MOV-1701 in order to prevent possible overpressurization of the Residual Heat i
Removal System. Thus, the portion of the i
RHR system identified above cannot be pressurized with the primary system and due to design, it cannot be pressurized without opening one of the MOV's.
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Page 9 of 15 RELIEF REQUEST 8 SYSTEM - Safety Injection COMPONENTS - Piping between the following sets of valves located on station print 11715-FM-96B Valves Lines 1-SI-83, 1-SI-190, and 1-SI-195 6"-SI-131-1502 1-SI-86, 1-SI-192, and 1-SI-197 6"-SI-133-1502 1-SI-89, 1-SI-194, and 1-SI-199 6"-SI-132-1502 1-SI-95, 1-SI-211, and 1-SI-204 6"-SI-19-1502 2"-SI-59-1502 1-SI-99, 1-SI-209, and 1-SI-203 6"-SI-21-1502 2"-SI-61-1502 1-SI-103, 1-SI-213, and 1-SI-205 6"-SI-16-1502 2"-SI-63-1502 CODE REQUIREMENTS - Class 1 System Hydrostatic Test per IWB-5222.
Po=2235 psig, To=160*F, Test Pressure per IWB-5222 is 2432 psig.
PROPOSED ALTERNATIVE EXAMINATION - As an alternative, the Reactor Coolant System will be pressurized to a pressure as close as practical to 2335 psig but not less than 2300 psig while the reactor is in a shutdown condition to create a pressure boundary at the first valve of each set listed above. These components will then be tested to a pressure (2300 psig < test pressure < 2335 psig) as close as practical to the Reactor Coolant System pressure using a charging pump. The Reactor Coolant System will'be borated equal to or greater than cold shutdown boron concentration.
REASON FOR RELIEF - The first valve in each set prevent the components listed above from being pressurized without pressurizing the Reactor Coolant System. The power operated relief valves (PCV-1456 and PCV-1455C) of the Reactor Coolant System are designed to limit the pressurizer pressure to a value below the fixed high-pressure reactor trip setpoint (2385 psig). The relief valve setpoints are 2335 psig which is below the test pressure of 2432 psig.
It is not desirable to take the Reactor Coolant System above the power operated relief valve setpoint.
Similar relief was granted to Surry Power Station Unit 2 per Safety Evaluation Report dated 01/24/86 (Docket 50-281) for components with the same design configuration.
Page 10 of 15 RELIEF REQUEST 9 SYSTEM - Safety Injection COMPONENTS - Piping and valves listed below and located on drawings 11715-FM-96A and 11715-FM-96B.
Valve Connecting Line Valve 4
MOV-1890C and MOV-1890D 10"SI-18-1502/10"-SI-238-1502 to 6"-SI-133-1502 1-SI-197 I
to 6"-SI-132-1502 1-SI-199 to 6"-SI-131-1502 1-SI-195 MOV-1890A 10"-SI-15-1502 to 6"-SI-16-1502 1-SI-213 to 6"-SI-130-1502 1-SI-211 to 6"-SI-19-1502 i
MOV-1890B 10"-SI-140-1502 to 6"-SI-21-1502 1-SI-209 1-SI-193 2"-SI-55-1502 1-SI-194 1-SI-191 2"-SI-53-1502 1-SI-192 1-SI-188 2"-SI-51-1502 1-SI-190 CODE REQUIREMENTS - Class 2 System Hydrostatic Test per IWC-5222.
P = 2485 psig, Design Temperature is less than 280*F,Testpressureis2733.5psig.
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i PROPOSED ALTERNATE EXAMINATION - As an alternative, the Reactor Coolant System will be pressurized to a pressure as close as practical to 2335 psig but not less than 2300 psig while the reactor is in a shutdown condition to create a pressure boundary at Check Valves 1-SI-83, 1-SI-86, 1-SI-89, 1-SI-95, 1-SI-99 and 1-SI-103.
These components will then be tested to a pressure (2300 psig < test pressure < 2335 l
psig) as close as practical to the Reactor Coolant System pressure using a test pump.
i REASON FOR RELIEF - Check valves 1-SI-83, 1-SI-86, 1-SI-89, 1-SI-95, 1-SI-99 and 1-SI-103 prevent the components listed i
above from being pressurized without pressurizing-the Reactor Coolant System. The Code required test i
pressure of 2733.5 psig will overpressurize the Reactor Coolant System.
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Page 10a of 15 l
RELIEF REQUEST 9 l
REASON FOR RELIEF (continued)
The power operated relief valves (PCV-1456 and PCV-1455C) f l
of the Reactor Coolant System are designed to limit the l
pressurizer pressure to a value below the fixed high-pressure reactor trip setpoint (2385 psig). The relief valve setpoints are 2335 psig which is below the l
test pressure of 2733.5 psig.
It is not desirable to l
take the Feuctor Coolant System above the power operated i
relief valve setpoint.
l Similar relief was granted to Surry Power Station Unit 2 per Safety Evaluation Report dated 01/24/86 (Docket 50-281) for components with the same design configuration.
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Page 11 of 15 F
RELIEF REQUEST 10 SYSTEM - Safety Injection COMPONENTS - Piping between the sets of valves listed below located on drawing 11715-FM-96B.
Valves Line MOV-1865A and 1-SI-125 12"-SI-123-1502 1-SI-123 and 1-SI-125 3/4'-SI-78-1502 MOV-1865B and 1-SI-142 12"-SI-124-1502 1-SI-140 and 1-SI-142 3/4"-SI-84-1502 MOV-1865C and 1-SI-159 12"-SI-125-1502 1-SI-157 and 1-SI-159 3/4"-SI-80-1502 1
CODE REQUIREMENTS - Class 2 System Hydrostatic Test per IWC-5222.
P =2485 psig, T O200*F Test pressure per the d
d code is 2733.5 psig since there is no over pressure protection for the above components.
PROPOSED ALTERNATIVE EXAMINATION - As an alternative it is requested that the Class 2 components listed above be tested per IWB-5222. The nominal operating pressure is 660 psig and temperature is 120*F.
Thus, testing per IWB-5222 would require a test pressure of 724 psig. This should be adequate considering the nominal operating conditions.
i REASON FOR RELIEF - Check valves 1-SI-125, 1-SI-142, and 1-SI-159
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at the Class 1 and 2 system boundaries prevent i
the pressurization of the above components without pressurizing the primary system. The required test pressure is 2733.5 psig as stated above, which would over pressurize the primary system.
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Page 12 of 15 RELIEF REQUEST 11 SYSTEM - Main Steam (Steam) Decay Heat Release Feedwater Chemical Feed Blowdown COMPONENTS - Steam generator and piping located on drawings 11715-FM-70B, 11715-EM-74A, 11715-FM-89D, 11715-FM-98A, and 11715-FM-102A.
Component Connected Pipping Component 1-RC-E-1A 32"-SHP-1-601 SV-MS-101A to 32"-SHP-22-601 SV-MS-102A SV-MS-103A c
SV-MS-104A SV-MS-105A l
to 6"-SHP-37-601/1"-SHP-84-601 PCV-MS-101A j
to 3"-SHP-64-601 1-MS-327 1-MS-18 to 1 1/2"-SHPD-6-601 1-MS-22 to 1/2"-SHPD-71-601 1-MS-26 1-RC-E-1A 32"-SHP-1-601 1-MS-35 NRV-MS-101A 3"-SilP-60-601 1-MS-344 1-RC-E-1A 32"-SHP-1-601 to 3"-SHP-45-601 1-MS-344 to 3"-SHP-531-601 NRV-MS-103A to 1"-SHP-518-601 1-MS-346 1-MS-348 i
1-RC-E-IA 2"-SS-302-601 1-SS-576 1-RC-E-1A 32"-SHP-1-601 to 32"-SHP-22-601 to 3"-SDHV-1-601 to 4"-SDilV-4-601 HCV-MS-104 1-RC-E-1A 16"-WFPD-24-601 1-FW-68 1-FW-47 to 3/4"-CFPD-1-1CN9 1-WT-39 1-RC-E-1A 2"-WGCB-4-601 1-BD-1 2"-WGCB-5-601 1-BD-4 1"-WCCB-6-601 1-BD-2
Attachmet 2 Page 13 of 15 COMPONENTS CONNECTED PIPING COMPONENTS 1-RC-E-1B 32"-SHP-2-601 to 32"-SHP-23-601 SV-MS-101B SV-MS-102B SV-MS-103B SV-MS-104B SV-MS-105B to 6"-SHP-38-601/1"-SHP-85-601 PCV-MS-101B to 3"-SHP-65-601 1-MS-325 1-MS-57 to 1 1/2"-SHPD-8-601 1-MS-60 to 1/2"-SHPD-73-601 1-MS-64 1-RC-E-1B 32"-SHP-2-601 1-MS-74 NRV-MS-101B 3"-SHP-61-601 1-MS-353 1-RC-E-1B 32"-SHP-2-601 to 3"-SHP-61-601 to 3"-SHP-46-601 1-MS-353 to 3"-SHP-533-601 NRV-MS-103B to 1"-SHP-520-601 1-MS-356 1-MS-357 1-RC-E-1B 2"-SS-225-601 1-SS-218 1-RC-E-1B 32"-SHP-2-601 to 32"-SHP-23-601 to 3"-SDHV-2-601 to 4"-SDHV-4-601 HCV-MS-104 1-RC-E-1B 16"-WFPD-23-601 1-FW-79 1-FW-100 to 3/4"-CFPD-2-1CN9 1-WT-51 1-RC-E-1B 2"-WGCB-7-601 1-BD-10 2"-WGCB-8-601 1-BD-13 1"-WCCB-9-601 1-BD-11 1-RC-E-1C 32"-SHP-3-601 to 32'-SHP-24-601 SV-MS-101C SV-MS-102C SV-MS-103C SV-MS-104C SV-MS-105C to 6"-SHP-39-601/1"-SHP-86-601 PCV-MS-101C to 3"-SHP-66-601 1-MS-95 1-MS-323 to 1 1/2"-SHPD-7-601 1-MS-98 to 1/2"-SHPD-75-601 1-MS-412
Page 14 of 15 COMPONENTS CONNECTED PIPING COMPONENTS 1-RC-E-1C 32"-SHP-3-601 1-MS-112 NRV-MS-101C 3"-SHP-62-601 1-MS-362 1-RC-E-1C 32"-SHP-3-601 to 3"-SHP-62-601 to 3"-SHP-47-601 1-MS-362 to 3"-SHP-532-601 NRV-MS-103C to 1"-SHP-519-601 1-MS-365 1-MS-336 1-RC-E-1C 2"-SS-227-601 1-SS-217 1-RC-E-1C 32"-SHP-3-601 to 32"-SHP-24-601 to 3"-SDHV-3-601 to 4"-SDHV-4-601 HCV-MS-104 1-RC-E-1C 16"-WFPD-22-601 1-FW-111 1-FN-132 to 3/4"-CFPD-3-1CN9 1-9f-67 1-RC-E-1C 2"-WGCB-10-601 1-BD-19 2"-WGCB-11-601 1-BD-22 1"-WCCB-12-601 1-BD-20 CODE REQUIREMENT - Class 2 System Hydrostatic Test per IWC-5222 and IWA-5213(d). For Feedwater components P =1100 psig, g
T >200'F, test pressure per IWC-5222 wouId be 1375 psig.
d For the Chemical Feed Components P =1775 psig, T<200*F,testpressureperIWC-5232wouldbe1952.5 d
psig. The remaining components have P =1 85 psig, d
T
, test pressure per IWC-5222 would be 1356 psig.
d PROPOSED ALTERNATIVE EXAMINATION - The Westinghouse Technical Manual for the Steam Generator requires the secondary side to be pressurized to 1356 psig, held for 30 minutes and then reduced to the design pressure (1085 psig) for a sufficient time to permit proper examination of welds, closures and surfaces for leakage or weeping.
The secondary side will be held at 1356 psig for 30 minutes and then at 1085 psig for a minimum of 3 1/2 hours in accordance with the Code. A VT-2 examination will then be performed.
RELIEF FOR REQUEST - Westinghouse, the manufacturer of the steam generators, gives specific testing requirements for the steam generator which must also be applied to the components listed above because these components cannot be isolated from the stenm generators.
Page 15 of 15 RELIEF REQUEST 12 SYSTEM - Component Cooling, Chemical and Volume Control, Fuel Pit Cooling, Safety Injection, Quench Spray, Recire Spray, Service Water and Sampling.
COMPONENTS - Piping and components included in the system hydrostatic test boundary.
CODE REQUIREMENT - Per IWA-5265(b)..."the imposed pressure on any component, including static head, will not exceed 106% of the specified test pressure for the system."
PROPOSED ALTERNATIVE EXAMINATION - Hydrostatic testing of systems that cannot be isolated to meet the system test pressure at the test boundary high point and the 106% system test pressure maximum at the test boundary low point shall be conducted by pressurizing to the system test pressure at the low point in the test boundary.
REASON FOR RELIEF - Unisolable portions of the above systems within the system hydrostatic test boundary are located throughout the plant such that there are variations in elevation within the boundaries that would result in imposed pressures in excess of six percent of the specified test pressure.
It is Virginia Electric and Power Company's desire to limit the test pressure imposed on system components to 106% of the specified test pressure (as required by paragraph IWA-5265(b)). Thus, due to the effects of static head, portions of the piping at higher elevations will be subjected to a test pressure lower than that specified. There is no practical method for isolating the piping segments to achieve the required test pressure at all elevations.
In a Safety Evaluation Report on Duane Arnold Energy Center (Docket No. 50-331) dated 03/31/86, relief was granted from IWA-5265(b) for situations as described above.