ML20235A671
| ML20235A671 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/10/1987 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20235A638 | List: |
| References | |
| NUDOCS 8709230380 | |
| Download: ML20235A671 (170) | |
Text
, _ _ _ _
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to il w
SOUTHERN COMPANY SERVICES 1
INSPECTION, TESTING, AND ENGINEERING DEPARTMENT FOR GEORGIA POWER COMPANY INSERVICE INSPECTION PROGRAM SECOND 10-YEAR INTERVAL
)
E.I. HATCH NUCLEAR PLANT UNITS I AND 2
- O scs GPC REV.
DATE DESCRIPTION m-
,,,y, m,
m.
PMP'O MV'O M. ANT
~
SEE ATTATCHED APPROVAL SHEET
[
/
0 6/2i/n FOR REVISION O APPROVAL I
6/l$/87 GENERAL UPDATE N
n~a.
8709230330 070911 DR ADOCK 0500 1
EDWIN I. HATCH NUCLEAR PLANT - UNITS 1 AND 2 O
INSERVICE INSPECTION PROGRAM DISTRIBUTION LIST Copy Number Manual Holder 6
E.
C.
Sorrell GPC-Plant Hatch 7
E.
C.
Sorrell GPC-Plant Hatch 17 T. N.
Epps SCS-Birmingham 18 T. N.
Epps SCS-Birmingham 19 J.
P.
Kane GPC-Atlanta 20 L. T. Gucwa GPC-Atlanta 21 E.
C.
Sorrell GPC-Plant Hatch 22 E.
Burkett GPC-Plant Hatch 33 E.
Burkett GPC-Plant Hatch 24 D
S.
Read GPC-Plant Hatch 25 H.
L.
Sumner, Jr.
GPC-Plant Hatch 26 P.
E.
Fornell GPC-Plant Hatch 37 R.
T.
Oedamer SCS-Birmingham O
02221 1
TABLE OF CONTENTS n
'w)i i
1.0 I n t ro d u c t i o n................................................... 1 - 1
)
2.0 Cl as s 1 wi th Rel i e f Reque s ts................................... 2-1 3.0 Cl ass 2 wi th Rel i e f Reques t s................................... 3-1 4.0 Cl as s 3 wi th Rel i ef Reques ts.................................. 4-1
- 5. 0 Cl ass 1, 2, and 3 Supports wi th Relief Requests................ 5-1 6.0 Valve Testing wi th Relief Requests............................ 6-1 l
7.0 Pump Tes ting wi th Rel ief Requests............................. 7-1 8.0 Ger.stal Rel i e f Reque s ts........................................ 8-1 DC)
I
f.
1.0 INTRODUCTION
d 1.'1 GENERAL This docenent decribes the revised Inservice Inspection (ISI) and Inservice Testing (IST) Programs for Edwin I. Hatch Nuclear Plant, Units 1 and 2.
The previous program :;ubmitted for Hatch 1 and 2 by Georgia Power Company letter dated August 12, 1983, was prepared in accordance with the 1980 Edition of the American Society of Mechanical Engineers (ASME)Section XI with Addenda through Winter 1980.
This revised Program has been upgraded with the intent to meet, as much as l
practical, the requirements of the 1980 Edition of Section XI with Addenda through Winter 1981.
1.2 EFFECTIVE DAJE The revised Program shall go into effect on January 1, 1986.
l 1.3 SCOPE This document is a description of the ISI and IST Programs for Units 1 and 2 l
of Plant Hatch.
The programs for Class 1, 2, and 3 component examinations as well as for pump and valve surveillance testing are included.
1.4 COMPONENT UPGRADING
.O All plant' components have been reviewed to determine the appropriate classification for inservice inspection and testing.
was used for guidance in determining component classifications.
1 It must be noted that the classification of components as ASME Class 1, 2, or l
3 equivalent for this Program does not imply that the components were l
designed in accordance with ASME requirements.
The component design codes remain as stated in the FSAR.
1.5 SUBSEQUENT PROGRAM REVISIONS l
It is anticipated that this Program will be reviewed again near tha end of l
1 120 months of implementation.
At that time, the Program will be modified as 3
)
required to bring it into compliance with a later Nuclear Regulatory
)
Commission (NRC) aporoved edition and addenda of ASME Section XI.
I 1.6 RESPONSIBILITY 1
IJ Georgia Power Company, as owner, bears the overall responsibility for the
)
performance of the inservice inspection and testing activities in accordance with IWA-1400.
O 04071
p 1.7 RECORDS
! L]
Records and documentation of information and examination results, which provide the basis for evaluation and which facilitate comparison with results from previous and subsequent examir.ations, will be maintained and available for the active life of the plant in accordance with Section XI, IWA-6000.
l 1.8 METHODS OF EXAMINATION The method of examination planned for each area is delineated in subsequent sections.
Personnel performing nondestructive examinations will be trained in accordance with the American Society for Nondestructive Testing (ASNT)
" Recommended Practice SNT-TC-1A, Supplements and Appendices," as applicacle for technique and method used.
1.8.1 ULTRASONIC EXAMINATION (UT)
It is anticipated that most volumetric examinations will be performed ultrasonically.
Examinations will be conducted in accordance with the requirements of IWA-2232 of ASME Section XI except as requested by relief.
1.8.2 RADIOGRAPHIC EXAMINATION (RT)
Radiographic techniques will be used to supplement UT when necessary.
p 1.8.3 LIQUID PENETRANT EXAMINATION (PT)
V Liquid penetrant examinations in accordance with IWA-2222 of ASME Section XI will be performed whenever a surface examination is required on non magnetic components.
1.8.4 MAGNETIC PARTICLE EXAMINATION (MT)
Magnetic particle tests in accordance with IWA-2221 of ASME Section XI will be used when surface examination of carbon steel components is required.
1.8.5 VISUAL EXAMINATION (VT)
Visual examinations in accordar.ce with IWA-2210 of ASME Section XI are divided into four different categories:
I a.
VT-1 l
l This examination shall be conducted to determine the condition of the l
part, component, or surface examined.
b.
VT-2 This examination shall be conducted to locate evidence of leakage from pressure retaining components.
i 0407I 1
c.
VT-3 3
l This examination shall be conducted to determine the general mechanical and structural condition of components and their supports, d.
VT-4 This examination shall be conducted to determine conditions relating to j
the operability of components or devices.
1.9 STANDARDS FOR EXAMINATION EVALUATION The acceptance standards for Class 1 components will be either IWB-3000 of ASME Section XI or the Section III construction Code for the plant, as applicable.
For C16ss 2 and 3 components, Articles IWC-3000 and IWD-3000, respectively, are in the course of preparation.
Both Articles state that the rules of Article IWB-3000 may be used. Therefore, the acceptance standards for Class 2 and 3 components will be either Article IWB-3000 or the Section III construction Code for the plant.
The acceptance standards for Class 1, 2, and 3 component supports will be either IWF-3000 or the Section III construction Code for the plant, as applicable, n
( ")
Corrective measures for Class 1, 2, and 3 pressure testing will be determined per IWA-5000 of ASME Section XI.
Analysis of results will be performed and corrective measures will be determined per IWP-3000 for pump testing and IWV-3000 for valve testing.
1.10 REPAIR PROCEDURES Repairs to the pressure retaining boundary of ASME Class 1, 2, or 3 (equivalent) components will be performed in accordance with Article IWA-4000 utilizing Georgia Power company approved procedures which generally comply with the Code applicable to the construction of the component.
In addition, replacement of components will be performed per Article IWA-7000 of Section XI, to the extent practical.
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i EDWINI. HATCH IJNTT I BW70W HEAD MERIDIONAL +CIRCUMFERINTIAL WEl.0$
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EDWINI. HATCH UNIT 2 RPV BOTTOM HEAD MOCEDIONAL 4 CTRCUMFERENTIAL WELDS FIGURE 10 IO
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2.0 CLASS 1 WIT.H RELIEF REQUESTS G
General Table 1 provides a tabulation of the Class 1 pressure-retaining components subject to the inspection requirements of Subsection IWB of Section XI of the ASMi Boiler and Pressure Vessel Code, 1980 Edition with Addenda through Winter 1981.
These components will be inspected in accordance with the requirements of Subsection IWB to the extent practical.
This tabulation identifies the components to be inspected, the Section XI examination item and category, area to be examined, and the method of examination. Where relief from the inspection requirements of Subsection IWB is requested, information is provided which identifies the applicable Code requirements, justification for the relief requested, and the examination method to be used as an alternative.
Table IWB-2500-1 items not applicable to the Hatch Plant have also been listed and identified in the interest of completeness.
Hydrostatic Testing Hydrostatic testing will be conducted in a manner that will satisfy the requirements of IWA-5000 and IWB-5000. Where adjoining pipe sections have different test pressures, they will be separated whenever practicable and each section tested at its specified pressure. Where it is not practicable to separate adjoining sections of piping (e.g., the boundary is a check valve), the sections will be tested together at the lower of the specified (q
test pressures. No point in the piping shall be permitted to experience a "y
pressure greater than the specified test pressure for that piping.
I Weld Selection Criteria The extent of examination for Code Class 1 pipe welds (Category B-J) was determined by the requirements of Table IWB-2500 and Table IWB-2600 Category B-J of Section XI of the ASME Code in the 1974 Edition with Addenda through Summer 1975 as allowed by 10 CFR 50.55a.
In addition, terminal ends and high l
stress welds were chosen when practical to upgrade the overall selection criteria.
2.1 REQUESTS FOR RELIEF FROM ASME SECTION XI REQUIREMENTS 2.1.1 VOLUMETRIC EXAMINATION OF REACTOR PRESSURE VESSEL AND CLOSURE HEAD WELDS 2.1.1.1 Requirement From Which Relief is_ Requested Item No. Bl.11, Bl.12, Bl.21, and 81.22 of Table IWB-2500-1 of ASME Section XI require the volumetric. exs nination of reactor pressure vessel (RPV) and closure head circumferential, longitudinal, and meridional welds.
At Hatch, these examinations will be performed using ultrasonic techniques.
The applicable examination volumes are shown in Figures IWB-2500-1,
-2, and -3.
V i
O 04071 l
1 i
2
l The Code also requires that welds selected for examination are to be examined for essentially 100 percent of their length.
Relief from this requirement is requested.
2.1.1.2 Justification At Hatch, physical limitations prevent the examination of the entire length of these welds.
The 1974 Edition with Addenda through Summer 1975 and earlier editions of Section XI required that the examination cover at least 10 percent of the length of each longitudinal weld and 5 percent of the length of each circumferential weld.
For both Hatch units, the minimum lengths described above were met during previous examinations.
The following evaluation (listed by code item No.) gives the percentage of weld accessible for examination and the additional welds to be examined so that the total weld length examined equals 100 percent of the required examination length.
Item Bl.11 - Hatch Unit 1 - There are two circumferential welds (C-3 and C-4) in the beltline region of the RPV, as shown in Figure 1.
Weld C-4 has three access doors through the concrete shield, and removable RPV insulation in these areas was provided during the design.
These three access ports allow the manual examination of approximately 15 percent of the weld. Weld C-3 has two usable access doors allowing approximately 10 percent coverage; therefore, a total of only 25 percent of the beltline area welds can be examined during the second 10 year interval.
During the second 10-year interval, portions of welds C-2 and C-5 will also be examined in order that the total equivalent length being examined p
equals the length of one beitline circumferential weld.
V Item 81.11 - Hatch Unit 2 - There are two circumferential welds (2C-3 and 2C-4) in the beltline region of the RPV, as shown in Figure 2.
These welds have permanent tracks which were installed during con >truction prior to the preservice examinations.
Projected examination coverage indicates that approximately 348 inches (47 percent) of weld 2C-3 can be examined using a mechanized system and 234 inches of weld 2C-4 (32*4) can be examined.
Therefore, a total of approximately 79 percent of the circumferential beltline area welds can be examined during the second 10 year interval.
During the second 10 year interval, portions of welds 2C-2 and 2C-5 will also be examined in order that the total equivalent length being examined equals the length of one beltline circumferential weld.
In addition, the required examinations for the third 40-month period of the first 10 year interval will be completed.
I_ tem 81.12 - Hatch Unit 1 - Using the same access doors as described above, approximately 20 percent to 30 percent of C-3-A, 20 percent to 30 percent of C-3-B, and 10 percent to 15 percent of weld C-3-C can be manually examined.
Therefore, a total of 50 percent to 75 percent of the beltline longitudinal weld, can be examined during the second 10 year 04071 k
R i
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interval.
Also, during the second 10 year interval, sufficient weld U
lengths will be selected from welds C-2-A, C-2-B, C-2-C and/or C-4-A, C-4-B, and C-4-C to ensure that the equivalent length of a beltline longitudinal weld is examined.
)
i Item B1.12 - Hatch Unit 2 - From the preservice data it is apparent that longitudinal welds 2C-3-A, 2C-3-B, and 2C-3-C can be 100 percent examined using a mechanized system with pole tracks installed during construction.
One of these three welds will be 100 percent examined during the second i
10 year interval.
In addition, the required examinations for the third j
40-month period of the first 10 year interval will be completed.
Item B1.21 - Hatch Unit 1 - Circumferential bottom head weld C-7 (Figure 9) will be 100 percent examined to the extent practical during the second 10 year interval.
If it is found during examinations that 100 percent coverage cannot be obtained, specific relief will be requested at that time. One circumferential closure head weld will also be 100 percent examined during the interval.
Item 81.21 - Hatch Unit 2 - Circumferential bottom head weld 2C-7 (Figure 10) had 73 percent of its weld length examined during preservice.
This weld will be 100 percent eramined to the extent practical during the second 10 year interval.
If it is found during examinations that 100 percent coverage cannot be obtained, specific relief will be requested at that time.
One circumferential closure head weld will also be 100 percent examined during the interval.
In addition, fq g
the required examinations for the third 40-month period of the first 10-year interval will be completed.
Item B1.22 - Hatch Unit 1 - One of the bottom head meridional welds extending from circumferential weld C-5 to C-7 (Figure 9) will be 100 percent examined to the extent practical during the interval.
If it is found during examinations that 100 percent coverage cannot be obtained, specific relief will be requested at that time.
One meridional closure head weld will also be 100 percent examined during the interval.
Item Bl.22 - Hatch Unit 2 - One of the bottom head meridional welds extending from circumferential weld 2C-5 to 2C-7 (Figure 10) will be 100 percent examined to the extent practical during the interval.
If it is found during examinations that 100 percent coverage cannot be obtained, specific relief will be requested at that time.
One meridional closure head weld will also be 100 percent examined during the interval.
In addition, the required examinations for the third 40-month period of the first 10 year interval will be completed.
Testing in Lieu of Section XI Requirements 2.1.1.3 1
1 For future examinations of RPV circumferential, longitudinal, and meridional welds, the examinations will be performed to the extent possible as described in the previous paragraphs.
2.1.2 Georgia Power has withdrawn Relief Request 2.1.2.
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0407I L___ __ ___
t'l 2.1.3 VOLUMETRIC EXAMINATION OF REACTOR PRESSURE VESSEL AND CLOSURE HEAD V
N0ZZLE-TO-VESSEL WELOS AND N0ZZLE INSIDE RADIUS SECTIONS 2.1.3.1 Requirement From Which Relief is Requested Item No. B3.90 and B3.100 of Table IWB-2500-1 of ASME Section XI require the volumetric examination of RPV and closure head nozzle-to-vessel welds and nozzle inside radius sections.
The applicable examination volume; are shown in Figures IWB-2500-7(a) through (d).
At Hatch, these examinations will be performed using ultrasonic techniques.
2.1.3.2 Justification Examination limitations exist for the nozzle examinations at Hatch Units 1 and 2 due to a combination of permanent physical obstructions. At Hatch Unit 1, an insulation support ring (Figure 1) is welded just above the N2A through K recirculation system inlet nozzles and the N1A and B recirculation system outlet nozzles. Welded thermocouple are near nozzles N48 and N4D which partially limit coverage. The limitations exist regardless of transducer size.
Hatch Unit 2 has limitations for the examination of the 2N4A and C feedwater nozzles due to interference from adjacent nozzles (Figure 2) and the transition area of the nozzles where they are welded to the shell.
As before, transducer size has very little impact on the coverage of these feedwater nozzle welds.
Showing the nozzle-to-vessel weld as N to V and the nozzle inside radius section as IRS, the table below shows the affected nozzle, minimum coverage, and reason for limitation.
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Hatch Unit 1 Nozzle Limited Examinations Minimum Coverage Reason N2A N to V; IRS 85%
Ins. Support Ring N2B N to V; IRS 85%
Ins. Support Ring N2C N to V; IRS 85%
Ins. Support Ring N2D N to V; IRS 85%
Ins. Support Ring N2E N to V; IRS 85%
Ins. Support Ring N2F N to V; IRS 85%
Ins. Support Ring N2G N to V; IRS 85%
Ins. Support Ring N2H N to V; IRS 85%
Ins. Support Ring N2J N to V; IRS 85%
Ins. Support Ring N2K N to V; IRS 85%
Ins. Support Ring N1A N to V; IRS 85%
Ins. Support Ring N1B N to V; IRS 85%
Ins. Support Ring N4B N to V; IRS 95%
Thermocouple N40 N to V; IRS 95%
Thermocouple Hatch Unit 2 Nozzle Limited Examinations Minimum Coverage Reason 2N4A N to V; IRS 85%
Aojacent Nozzle p
2N4C N to V; IRS 85%
Adjacent Nozzle
.d l
04071 i
Il 2.1.3.3 Testing in Lieu of Section XI Requirements w/
Not applicable to this relief request.
2.1.4 VOLUMETRIC EXAMINATION OF AUSTENITIC AND DISSIMILAR METAL PIPING WELDS 2.1.4.1 Requirement From Which Relief is Requested Item No. 85.130, B5.50, B9.11, B9.12, and B9.31 of Table IWB-2500-1 of ASME Section XI require a volumetric and surface examination of austenitic and dissimilar metal piping welds.
In addition, Item No. M4.10 of Table IWB-2500-1 requires either a volumetric or a surface examination of the pressure retaining welds in control rod drive (CRD) housings.
These volumetric examinations are to be performed using ultrasonic techniques in accordance with Paragraph IWA-2232 of Section XI.
This paragraph specifies that austenitic and dissimilar metal piping welds are to be examined in accordance with Article 5 of ASME Section V.
2.1.4.2 Justification Article 5 of ASME Section V does not provide the detailed guidance necessary to examine austenitic and dissimilar metal piping welds with the exception of austenitic piping welds which have been repaired by weld overlay.
These clad overlaid piping welds will be examined in accordance with Article 5 of Section V and Appendix III of Section XI.
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2.1.4.3 Testing in Lieu of Section XI Requirements Ferritic piping welds will be examined per Appendix III of Section XI. To provide consistency, austenitic and dissimilar metal piping welds will also be examined in accordance with Appendix III.
2.1.5 SURFACE EXAMINATION OF REACTOR PRESSURE VESSEL SUPPORT SKIRT WELD 2.1.5.1 Requirement From Which Relief is Requested Item No. 88.10 of Table IWB-2500-1 of ASME Section XI requires volumetric or surface (as applicable) examination of integrally welded attachments (support skirt), in accordance with the examination requirements shown on Fig. No.
At both Hatch units these examinations will be performed using either liquid penetrant or magnetic particle techniques.
2.1.5.2 Justification At both Hatch units, examination area C-D of Fig. No. IWB-2500-13 is not accessible for meaningful examination because of location and geometric configuration of welded areas.
Physical access by the examiner is also restricted because of high radiation and obstruction due to CRD housings and its support system.
The combination of these factors prevents these welds from being examined from inside the support skirt.
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Hatch Unit I has no access through the support skirt; therefore, the inside b
surface of the reactor vessel support skirt is totally inaccessible.
As 04071
_.___m._
O shown in: Figure 6 for Hatch Unit 2, the support skirt near the skirt-to-U-
vessel weld is very limited.
Health Physics indicates that the dose rate in this area during a recent Hatch Unit 2 outage was approximately 180 mr/hr; L
however, it is very difficult to quantify the total exposure for an j
examination.
Magnetic particle techniques cannot be used due to the space restrictions. The use of dye penetrant would require a very thorough cleaning of the weld and adjacent base material to remove rust and scale.
The preparation of the weld would potentially have to be performed using techniques such as wire brushes since power tools may not fit into the limited area.
2.1.5.3 Testing-in Lieu of Section XI Requirements As an alternate, both units will have a surface' examination performed on the OD of 100 percent of the weld during the second 10 year interval.
Also, a limited ultrasonic examination will be performed to the extent practical to provide'as much coverage as possible'of the weld.
2.1.6 REACTOR PRESSURE VESSEL N0ZZLE TO SAFE-END WELDS (NOMINAL PIPE SIZE
<4 INCHES) 2.1.6.1 Requirement From Which Relief is Requested item No. 85.20 of Table IWB-2500-1 of ASME Section XI requires a surface examination of the RPV nozzle-to-safe-end welds with nominal pipe size
<4 inches.
Relief from this requirement is requested.
2.1.6.2 Justification The nozzle-to-safe-end welds for the nozzles listed Flow are physically inaccessible for surface examination.
The affected nozzles are:
N10 2N10 N11A 2N11A N11B 2N11B N12A 2N12A N12B 2N12B N16A 2N16A N168 2N16B These 2-inch instrument nozzles have very limited access due to the design of the concrete shield.
Each nozzle has small doors that can be opened allowing 12 to 18 inches of access.
However, due to the distance the RPV wall is recessed from the outside of the shield wall (e.g., insulation thickness, air gap,'and shield thickness) the weld cannot be physically reached.
The 2-inch RPV bottom head-drain-nozzle-to-safe-end weld is exempted by IWB-1220(c) of ASME Section XI from examination.
2.1.6.3 Testing in Lieu of Section XI Requirements The nozzle-to-safe-end welds listed above will receive a remote visual examination with the exception of the 2-inch drain nozzle welds.
In
(
addition, these nozzles will be pressure tested per IWB-5000 of ASME Section XI since they are located within the hydrostatic test boundary of the nuclear steam supply system.
0407I
i l
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2.1.7 VOLUMETRIC AND SURFACE EXAMINATIONS OF PRESSURE RETAINING WELDS IN l /T
' \\J PIPING WITH NOMINAL PIPE SIZE 24 INCHES s
2.1.7.1 Requirement From Which Relief is Requested Item No. 89.11 of Table IWB-2500-1 of ASME Section XI requires the examination of pressure retaining welds in piping that are located within flued head containment penetration assemblies.
These welds and their penetration assembly number for Hatch 1 are listed below.
Hatch 2 has no pressure retaining welds in the penetration assemblies.
Weld Identification No.
Pen. No.
Weld Identification No.
Pen. No.
1821-1 FW-18A-7A X-9A IB21-1MS-24A-17A X-7A 1821-1FW-188-6A X-9B 1821-1MS-248-16A X-78 1E51-1RCIC-4-D-20A X-10 1821-1MS-24C-16A X-7C 1E41-1HPCI-10-D-15A X-11 1321-1MS-24D-17A X-7D 1E11-1RHR-20B-D-13A X-12 1E11-1RHR-24A-R-3A X-13A 1E11-1RHR-24B-R-3B X-138 1G31-1RWCU-6-D-15B X-14 1G31-1RWCV-6-D-15C X-14 1E21-1CS-10A-3A X-16A 1E21-1CS-108-4A X-16B 1E11-1RHR-4-HS-6A X-17
/-'s 2.1.7.2 Justification (v)
These welds are inaccessible for examination due to the design of the flued head.
All twelve circumferential butt welds, except the two located in the reactor water cleanup (RWCU) penetration, are carbon steel.
The two stainless steel welds that are located in the RWCU penetration were made to replace a Type 304 55 pipe that had undergone IGSCC.
The welds involved are a flued head with a Type 308L corrosion resistant clad on the inside surface to a Type 304L solution annealed pipe (<0.035 percent carbon), and a Type 304L pipe-to pipe weld.
These welds were made in accordance with the guidelines of NUREG-0313 to minimize susceptibility to IGSCC.
i 2.1.7.3 Testing in Lieu of Section XI Requirements A UT baseline was performed on the two new welds in the RWCU system while they were accessible during repair to ensure a high quality weld.
In accordance with IWB-5221 of ASME Section XI, a system leakage test is to be performed on all 12 welds prior to startup following each reactor refueling outage.
All pipe-to penetration (flued head) welds outside containment will be examined volumetrically.
In addition, a surface examination will be g-)
performed on the accessible weld (s) of the flued head penetration assembly.
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04071
2'.1.8 ASME CLASS 1 (EQUIVALENT) VALVES EXCEEDING 4-INCHES NOMINAL PIPE SIZE AND ASME CLASS.1 (EQUIVALENT) PUMP CASINGS 2.1.8.1 Requirement From Which Relief is Requested Table IWB-2500-1, Item No. B12.50 of ASME Section XI requires a visual
[
examination of the internal pressure boundary surfaces of one valve in each group of valves that are of the same constructional design, such as globe, gate, or check valve, and manufacturing method and that are performing similar functions in the system.
Table IWB-2500-1, Item No. B12.20 of ASME Section XI requires a visual examination of the internal pressure boundary surface of one of the two reactor recirculation system pumps.
2.1.8.2 Justification Disassembly of these valves and pumps for the visual examination during the inspection interval, in the absence of other required maintenance, represents an unnecessary exposure to radiation and contamination.
Valves on the reactor recirculation (RC) system and the residual heat removal (RHR) system suction lines would involve off-loading'the fuel elements and draining the l
RPV prior to disassembly as a precautionary measure. Work on the RC system l
. pump discharge valves and the RHR system injection valves would require the installation of plugs in the jet pump risers.
Preparatory work of this scope is considered impractical for the sole purpose of conducting a visual O.
examination. Contamination levels in the valves and pumps associated with the RC system loops are particularly high due to the physical location at the bottom of the system. During routine maintenance, the valve body and the pump casing internal surfaces are visually examined.
Many of the valves, particularly the containment isolation valves are disassembled for maintenance of leak-tightness.
Disassembly of other Class 1 valves and the pumps solely for internal examination is counter to the ALARA guidelines to keep the occupational dose rates as low as reasonably achievable.
In view of the cost in man-rem and in view of the minimal benefits obtained, we conclude that this Code requirement does not provide sufficient benefits to justify the exposure.
2.1.8.3 Testing in Lieu of Section XI Requirements Class 1 pumps and Class 1 valves exceeding 4 inches nominal pipe size are subject to visual examination of the internal surfaces when disassembled for maintenance. The coverage provided by examinations during routine l
maintenance coupled with periodic leak tests and hydrostatic tests will provide adequate assurance of the structural integrity of the Class 1 pumps and valves, while keeping exposure to radiation and contamination as low as reasonably achievable.
2.1.9 PRESSURE RETAINING WELDS IN CONTROL R00 DRIVE HOUSINGS 2.1.9.1 Requirement From Which Relief is Requested Table IWB-2500-1, Item No. B14.10 of ASME Section XI requires a volumetric or surface examination of the pressure retaining welds in 10 percent of the 0407I
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peripheral control rod drive housings.
Each housing has a pipe-to pipe weld located near the RPV and a pipe-to-flange weld.
.2.1.9.2 Justification The examination of these welds are limited because of the location and design
'of the housings.
Physical accessibility by an examiner is extremely limited by the close proximity of'the housings to each other and by the support arrangement. Also, the insert and withdraw lines to the control rod drive system are connected at the top of the housing flange and limit access to much of the lower weld. The combination of these factors limit the examination of these welds.
-Figure 4.2-8 of the Hatch' Unit 2 FSAR shows that there are 28 peripheral CRD housings. Each housing has an attachment weld to the reactor vessel and a weld joining the housing to the flange.
Section 4.2 of the FSAR shows that the failure of a CRD housing weld will produce a maximum leakage rate of 840 gal / min. The available makeup systems are RCIC-400 gal / min, CRD-160 gal / min., and the transfer system to feedwater-1000 gal / min _.
Therefore, the reactor can be shut down and ccoled in an orderly manner using makeup systems supplied by onsite power, as required by IWB-1220.
Since loss of coolant would occur during normal operation, it is _ our interpretation that the service transformer is-the source of onsite power. -(Note:
Hatch Unit 1 should have essentially the same leakage rates and makeup capabilities).
2.1.9.3 Testing in Lieu of Section XI Requirements O
These welds will be pressure tested per IWB-5000 of ASME Section XI since they are located within the hydrostatic test boundary of the nuclear steam
' supply system, t
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0407I
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I f.)N NOTES FOR TABLE 1
~
(1) During each refueling outage, the reactor pressure vessel (RPV) closure studs are normally left in place; therefore, only a volumetric examination will be performed (ASME Section XI Table IWB-2500-1, Item No. B6.20).
If the studs are removed, both a surface and volumetric examination will be performed (ASME Section XI Table IWB-2500-1, Item No. 86.30).
(2) The Class 1 socket welds (Item No. 89.40 of Table IWB-2500-1 of ASME Section XI) for both units are exempted from surface examination by paragraph IWB-1220(a).
However, selected austenitic piping welds may receive supplemental surface (PT) examinations.
(3) Prior to January 1, 1984, ultrasonic thickness measurements were performed each refueling outage on the RPV closure head at both units (Hatch 1-six outages and Hatch 2-three outages) to determine if thinning has occurred.
These measurements found no apparent thinning and therefore, these examinations will be performed once every 40 months at each unit.
[s v
0407I t
I 3.0. CLASS 2 WITH RELIEF REQUESTS General l
Table 2 provides a tabulation of the Class-2 pressure-retaining components subject to the inspection requirements of Subsection IWC of Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition with Addenda through Winter 1981. These components will be inspected in accordance with the requirements of Subsection IWC to the extent practical.
This tabulation identifies the componants to be inspected, the Section XI examination item and category, area to be examined, and the method of examination. Where relief from the inspection requirements of Subsection IWC is requested, information is provided which identifies the applicable Code requirements, justification for the relief request, and the examination method to be used as an alternative.
Table IWC-2500-1 items not applicable to Plant Hatch have also been listed and identified in the interest of completeness.
Hydrostatic Testing l
-Hydrostatic testing will be conducted in a manner that will satisfy the requirements of IWA-5000 and IWC-5000. Where adjoining pipe sections have different test pressures, they will be separated whenever practicable and each section tested at its specified pressure. Where it is not practicable to separate adjoining sections cf piping (e.g., the boundary is a check valve), the sections will be tested together at the lower of the specified
' O test pressures.
No point in the piping will be permitted to experience a pressure greater than the specified test pressure.
Weld Selection Criteria As required by 10 CFR 50.55a (Code of Federal Regulations), the extent of examinations for all Class 2 ECCS piping welds was determined by the requirements of Paragraph IWC-1220, Table IWC-2520 (Category C-F and C-G welds), and Paragraph IWC-2411 of Section XI, 1974 Edition with Addenda through Summer 1975. To make this plan even more comprehensive, those welds with high S values were selected for examination to the extent practical.
m However, some exemptions such as pressure / temperature were not used for particular systems.
The following summarizes the general weld selection criteria for Class 2 systems:
RHR, Core Spray, and HPCI 1.
Examine all required welds within the 10-year interval using the 1974 Code with Addenda through Summer 1975 for selection and the 1980 Code with Addenda through winter 1981 for technique.
2.
Apply high stress and terminal ends when practiczl.
i 0408I
/'~'s 3.
Do not use pressure / temperature exemption.
V 4.
In addition to Code requirements examine (UT or surface as applicable) welds on branch connection lines greater than one inch in diameter that could impact safety-related function of system out to the first closed manual valve, reverse check valve, or power operated valve.
Otherwise,
~
component connections, piping and associated valves, and vessels (and their supports), that are 4 inches nominal pipe size and smaller are exempt.
5.
Examine 100% of attachment welds within 10 years where the base material of the attachment is greater than or equal to 3/4 inches thick.
Other Class 2 Systems As permitted in 10 CFR 50.55a, the remaining Class 2 piping welds were selected using the 1974 Edition of the Code with Addenda through Summer 1975. The following components were exempted per IWC-1220:
1.
Components in systems where both the design pressure and temperature are equal to or less than 275 psig and 200 F., respectively.
2.
Components in systems or portions of systems, other than emergency core cooling systems, which do not function during normal reactor operation.
3.
Component connections, piping and associated valves, and vessels (and their supports), that are four inches nominal pipe size and smaller.
3.1 REQUESTS FOR RELIEF FROM ASME SECTION XI REQUIREMENTS 3.1.1 VOLUMETRIC EXAMINATION OF PRESSURE RETAINING WELOS IN CLASS 2 VESSELS 3.1.1.1 Requirement From Which Relief is Requested Item No. C1.10, C1.20, and C1.30 of Table IWC-2500-1 of ASME Section XI require the volumetric examination of Class 2 vessel shell circumferential, head circumferential, and tubesheet-to-shell circumferential welds, respectively.
The volumetric examination of the residual heat removal (RHR)
)
system heat exchanger circumferential welds will be performed using ultrasonic techniques.
The required examination volumes are shown in ASME Section XI Figures IWC-2500-1 and -2.
Relief from this requirement is requested.
3.1.1.2 Justification The shell and head circumferential weld examinations are limited by vessel supports adjacent to these welds.
In addition, the ultrasonic examination of the head circumferential weld from the head side cannot be perforiaed due to configuration.
The examination volume as required by Figure IWC-2500-2 for the tubesheet-to-shell weld cannot fully be met due to configuration also.
l The evaluation below lists the percentage of examination coverage and the l
physical limitations for each unit.
O 04081
1 I
l l
I (V
i Hatch Unit i i
There are three Category C-A circumferential welds in each of the two RHR heat exchangers.
These welds and their UT limitations are given below.
(See Figure 4).
1E11 - 2Hx-A(B)-1 Shell Head to Upper Shell Ring - These welds cannot be i
examined from the shell head side due to the curvature of the j
head.
Only about 65 inches of a total circumference of 179 inches l
(approximately 36 percent) can be examined from the upper shell ring side due to support interference.
1E11 - 2Hx-A(B)-2 Upper Shell Ring *o Lower Shell Ring - Complete coverage is obtained from the upper sheil-ring side and 0 percent i
coverage from the lower shell-ring side due to support interference.
1E11 - 2Hx-A(B)-3 Lower Shell-Ring to Flange - Complete coverage is obtained from the lower-shell-ring side.
Examination from the l
flange side cannot be performed due to the geometry.
Hatch Unit 2 l
1 There are three Category C-A circumferential welds in each of the two RHR heat exchangers.
These welds and their UT limitations are given below.
(See i
n Figure 3).
l C
j 2E11 - 2Hx-A(B)-1 Shell Head to Upper Shell Ring - These welds cannot be i
examined from the shell-head side due to the curvature of the head.
Only about 65 inches of a total circumference of 179 inches (approximately 36 percent) can be examined from the upper shell-ring side due to support interference.
2E11 - 2Hx-A(B)-2 Upper Shell-Ring to Lower Shell Ring - Complete coverage is obtained from the upper shell-ring side and approximately 36 percent coverage from the lower shell-ring side due to support interference.
2E11 - 2Hx-A(B)-3 Lower Shell-Ring to Flange - Complete coverage is obtained from the lower shell-ring side.
Examination from the flange side cannot be performed due to the geometry.
3.1.1.3 Testing in Lieu of Section XI Requirements The ultrasonic examination of the shell and head circumferential welds will be supplemented by a surface examination.
Tne tubesheet-to-shell weld cannot be properly prepared for surface examination nor can the examination be performed due to the tubesheet studs and nuts adjacent to the weld.
In j
I addition to the examinations described above, system pressure tests per Article IWC-5000 of ASME Section XI will be performed on these welds.
AU 04081 i
3
(~')
3.1.2 SURFACE EXAMINATION OF WELDED ATTACHMENTS ON RHR, CORE SPRAY, HPCI,
\\~ '
AND RCIC SUCTION LINES FROM TORUS 3.1.2.1 Requirement From Which Relief is Requested Item No. C3.20 of Table IWC-2500-1 of ASME Section XI requires 100 percent l
surface examination of integrally welded attachments on piping.
Figure IWC-2500-5 determines the surface area of this examination.
Suction lines for RHR, core spray, HPCI, and RCIC systems penetrating the torus are seal welded to the outside surface of the torus wall.
Relief is requested from the surface examination of these welded attachments.
3.1.2.2 Justification l
At both Hatch units, these integrally welded attachments on lines penetrating the torus are obstructed by reinforcement plates added after plant construction.
The evaluation below lists the percentage of examination l
coverage and describes the physical limitations for each unit.
Hatch Unit 1 As shown in Figure 7, the suction piping is surrounded by reinforcing ribs which may limit access on one or more sides of the pipe, in particular, when using magnetic particle techniques.
This method is 7s preferred since the torus has a heavy coating of paint, and removing the paint and cleaning the surface to perform penetrant examinations would i'~
be extremely difficult with the space limitations. As shown in Figure 7 I
approximately 80 percent - 100 percent of the RCIC (1E51) and HPCI I
(1E41) welds can be examined, approximately 50 percent - 75 percent of the core spray (1E21) welds, and approximately 25 percent or less of the RHR (IEll) welds.
Hatch Unit 2 As shown in Figure 8, the welds are totally inaccessible to perform surface examinations; therefore, a visual examination will need to be performed in lieu of the Code requirements.
3.1.2.3 Testing in lieu of Section XI Requirements Hatch Unit 1 The examinations will be performed to the extent possible as described in the previous paragraph, concerning Hatch Unit 1.
i l
Hatch Unit 2 Visual examination (VT-1) in accordance with IWA-2211 will be performed I
to insure the integrity of these attachments.
V(3 0408I
l
's 3.1.3 SURFACE EXAMINATION OF PRESSURE RETAINING WELDS IN CLASS 2 PUMPS l
^
CJ l
3.1.3.1 Requirement From Which Relief is Requested Item No. C6.10 cf Table IWC-2500-1 of ASME Section XI requires a surface j
examination of the pump casing welds from one Class 2 pump in each group of pumps that are of similar design, size, function, and service in a system.
{
Relief from this requirement is requested.
3.1.3.2 Justification This relief request applies to the core spray pumps and RHR pumps on Hatch Unit 2 only.
The Hatch Unit 1 pumps have a different design and do not contain pressure retaining welds (Item No. C6.10).
As shown in Figure 5, the j
pressure retaining welds 2E11-1RHR-PLP-A-1 through -6 are completely encased I
in the suction casing and can be accessed only when the pump is completely disassembled.
These welds are not the welds considered to be pressure retaining pump casing welds; therefore, it is impractical to disassemble the pump (s) solely to examine these welds.
(Note: At least one of the six pumps j
has been disassembled for maintenance and the welds examined).
)
3.1.3.3 Testing in Lieu of Section XI Requirements Class 2 pump casing welds are subject to surface examination when disassembled for maintenance.
The coverage provided by examinations during
,y routine maintenance coupled with hydrostatic tests will provide adequate
j assurance of the structural integrity of these pumps, while keeping exposure t
to radiation and contamination as low as reasonably achievable.
k 0408I
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4.0 CLASS 3 COMPONENTS WITH RELIEF REQUESTS Table 3 provides a tabulation of the Class 3 pressure-retaining components subject to the inspection requirements of Subsection IWD of Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition with Addenda through Winter 1981.
These components will be inspected in accordance with Subsection IWD to the extent practical with the exception of the relief requests outlined in Paragraph 4.1 below.
Table IWD-2500-1 items not applicable to Plant Hatch have also been listed and identified in the interest of completeness.
Where adjoining pipe sections have different test pressures, they will be separated whenever practicable and each section tested at its specified pressure. Where it is not practicable to sep& rate adjoining sections of piping'(e.g., the boundary is a check valve), the sections will~be tested together at the lower of the specified pressures.
No point in the piping will be pressurized above the specified test pressure.
4.1 REQUESTS FOR RELIEF FROM ASME SECTION XI REQUIREMENTS 4.1.1 SYSTEM PRESSURE TESTS ON CLASS 3 SMALL DIAMETER PIPING I
Relief Request 4.1.1 has been withdrawn.
4.1.2 SYSTEM PRESSURE TESTS ON CLASS 3 BURIED PIPING 4.1 ~. 2.1 Requirement From Which Relief is Reauested Table IWD-2500-1 of ASME Section XI requires a system pressure test for Class 3 components including buried piping.
4.1.2.2 Justification The following portions of systems were designed without including provisions for testing buried piping as required by Paragraph IWA-5244.
Hatch Unit No. 1 Plant service water system from the valve pit at the intake structure to the diesel generator building and from the valve nit to the reactor building wall.
RHR service water system from the valve pit at the intake structure to the reactor building wall..
Reactor core isolation cooling system from valve 1E51-F009 to valve 1E51-F010.
High pressure coolant injection system from valve IE41-F004 to valve IE41-F010.
Hatch Unit No. 2 Plant service water system from the valve pit at the intake structure to the O
diesel generator building and from the valve pit to the reactor building wall.
0409I
l l
O RHR service water system from the valve pit at the intake structure to the i
ej reactor building wall.
In addition, the visual examination for leakage at the ground level is not feasible since a majority of the piping is buried under asphalt.
4.1.2.3 Testing in Lieu of Section XI Requirements Normal system functional testing demonstrates leaktight integrity of all safety-related buried piping.
4.1.3 PRESSURE TEST ON CLASS 2 AND 3 WATER SYSTEMS I
4.1.3.1 Requirement From Which Relief is Requested IWC-5222 and IWD-5223 of ASME Section XI requires that Class 2 and 3 systems l
be tested at a pressure of 1.10 or 1.25 times the design pressure.
Relief is 3
l requested from testing those portions where it is necessary to use a butterfly valve six inches in diameter or greater as a hydrostatic test boundary valve.
4.1.3.2 Justification Butterfly valves are basically flow control valves and are not intended to be block valves.
The normal leakage thrcugh these large valves makes it impractical to attain and maintain the hydrostatic test pressure.
1 V
4.1.3.3 Testing in Lieu of Section XI 11 requirements j
i Since there are no viable means to test those portions of the system j
described above at a higher pressure, a hydrostatic test at normal operating pressure (i.e., system functional test or system inservice test) will be performed on those portions of these systems which have a butterfly valve 6 inches in diameter or greater.
4.1.4 SYSTEM PRESSURE TESTS ON CLASS 3 TUBING <1 INCH IN DIAMETER 4.1.4.1 Requirement _From Whicn Relief is Requested Table IWD-2500-1 of ASME Section XI requires thet a system functional (or inservice) test and a system hydrostatic test be acc.omplished for the pressure retaining boundary.
Relief is requested from visually examining small instrument tubing <1 inch in diameter in the plant service water (PSW) system and the RHR service water (RHRSW) System which does not provide a cooling function for a safety-related component.
4.1.4.2 Justification These lines do not provide cooling water to safety-related equipment; therefore, loss of the line would not cause loss of a safety-related component.
In addition, icakage through the tubing would not degrade the overall capability of these service water systems, since such flow would be
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negligible when compred to the rated flow of the pumps (PSW 3500 gpm each pump and RHRSW 4000 gpm each pump).
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1
,- ~3 4.1.4.3 Testing in Lieu of Section XI Requirements 4
All piping 21 inch providing a safety-related function will be pressure tested.
In addition, any tubing providing cooling water to a safety-related component will also be pressure tested.
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g 5.0 CLASS 1, 2, AND 3 SUPPORTS WITH RELIEF REQUESTS Table 4 provides a tabulation of the Class 1, 2, and 3 component supports subject to the inspection requirements of Subsection IWF,Section XI of the ASME Boiler and Pressure Vessel Code, 1980 Edition wi;h Addenda through Winter 1981.
These components will be inspected in accordance with the requirements of Subsection IWF to the extent practical.
This tabulation identifies the components to be inspected, the Section XI examination item and category, area to be examined, and the method of examination. Where relief from the inspection requirements of Subsection IWF is requested, l
information is provided which identifies the applicable Code requirements, J
justification for the relief request, and the examination method to be used as an alternative. Table IWF-2500-1 items not applicable to Plant Hatch have also been listed and identified in the interest of completeness.
Paragraph IWF-1230, Supports Exempt from Examination and Test, is in the l
course of preparation.
Later Addenda through Winter 1984 of Section XI do a
not provide guidance for the exemption criteria for these components.
J Georgia Power company has therefore decided to incorporate the exemption criteria for Class 1, Class 2, and Class 3 component supports found in Paragraphs IWB-1220, IWC-1220, and IWD-1220 of ASME Section XI, respectively.
5.1 REQUESTS FOR RELIEF FROM ASME SECTION XI REQUIREMENTS 5.1.1 REQUIREMENT TO VERIFY HOT OR COLD SETTINGS ON SPRING CANS AND SNUBBERS h
5.1.1.1 Requirement From Which Relief is Requested Subparagraph IWF-3410(a)(5) of ASME Section XI states that a component support condition which is unacceptable for continued service is improper hot or cold positions for spring supports or snubbers.
Relief from this requirement is requested.
5.1.1.2 Justification There are no exact design positions on the scales for spring supports or snubbers but an operational range where the indicator should be located.
Georgia Power Company will verify that the spring support is in the operable range plus acceptable tolerances; i.e., within analyzed hot and cold load settings.
The support may not exactly show the precise hot or cold setting because the analysis may have used a conservative temperature; i.e., the plant may not see the temperature analyzed of that specific system on the specific day when the inservice inspection was performed.
The intent of the inspection is to verify that the spring can is not outside the range specified in the analysis and that the can is not bottomed out.
5.1.1.3 Testing in Lieu of Section XI Requirements The visual examination will verify that the indicator falls within the operational limits.
O 04091 l
N(D 5.1.2 TESTING OF HYDRAULIC AND MECHANICAL TYPE SNUBBERS
(
5.1.2.1 Requirement From Which Relief is Requested Article IWF-5000 of ASME Section XI outlines the inservice test requirements for hydraulic and mechanical type snubbers.
Relief from this requirement is requested.
5.1.2.2 Justification Instead of using IWF-5000, the ongoing testing program per the Plant Technical Specifications will be performed.
This testing program is designed to demonstrate the functional integrity of the snubbers and exceeds the requirements of Article IWF-5000/
5.1.2.3 Testing in Lieu of Section XI Requirements As noted in 5.1.2.2, the testing program per the Plant Technical Specifications will be performed.
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g 6.0 INSERVICE INSPECTION OF VALVES The valve testing program is based on the 1980 Edition of ASME Section XI with Addenda through Winter 1981.
Valves in the program are listed in the valve test tables and will be tested in accordance with the code unless otherwise specified.
Several features of the program are discussed below.
Containment Isolation Valves Containment isolation valves (CIVs) will have a type C leakage test performed I
per the requirements of 10 CFR 50, Appendix J.
In addition, each valve or penetration will have a specific leakage limit assigned to it by Georgia Power Company.
Limits will be based on type of valve, size of valve, the number of valves tested in parallel paths, and historical leakage data.
These limits are available for review in the Long-Term Inservice Inspection Pump and Valve Test Plans stored in the Plant Hatch document control room.
Pressure Isolation Valves Pressure isolation valves, as a rule, will be tested in the same manner as the CIVs and the observed leakage escalated per IWV-3423 to the function maximum pressure differential value.
The maximum leakage for each valve will be 0.5 gpm per inch of nominal valve size or a maximum of 5 gpm (water).
If air is used as the test medium, the leakage will be conservatively assumed to be water leakage. Testing experience at Plant Hatch has shown that when air is used as the test medium the resulting leakage is higher than when water is used !n the same test volume; therefore, it is conservative to test with air instead of water.
l Valves to be Tested During Cold Shutdown and Refueling l
Valve testing will commence as soon as possible into the cold shutdown but no later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the shutdown.
Valve testing will continue during the shutdown until complete or until plant startup and return to power.
Any testing not completed at one cold shutdown will be performed during subse uent cold shutdowns before the next refueling.
During refueling, any valve scheduled for a refueling test will be tested.
l Also, any valve identified to be tested at ccid shutdown that has not been tested during the previous three months will be tested during the refueling.
Remote Indicating Lights Valves with remote indicating lights will be observed at least once every 2 j
years to verify that the valve operation is accurately indicated.
l Fail-Safe Valves Unless otherwise specified in the program tables, the only fail-safe valves are either air operated or solenoid valves.
Stroking the valve full cycle j
during normal testing causes less of power to the actuator as required by IWV-3415.
Therefore, additional testing to prove the fail-safe capability O-will not be performed.
1 0410I i
b
. Control' Rod Drive System Valves
-f See Relief Request 6.1.22 Valves Testing During Cold Shutdown See Relief Requests 6.1.9 and 6.1'.19.
In addition to the Relief Requests addressed above, relief is specifically requested for each valve in the Valve Test Project Tables as required. All relief requests are located following these tables.
Compressed Air Systems Valve Testing Section 9.3.1.4, Safety Evaluation of Compressed Air systems, of the Final Safety Analysis Report (FSAR) states:
"Because compressed air is not essential for the safe shutdown of the plant, the compressed air systems do not switch automatically to operation from diesel generator. power following a loss of normal power."
.Section 15.1.28.2, Analysis of Effects and Consequences of Loss of Instrument
. Air, also states:
"All equipment using instrument air is designed to fail to a position 9O.
that is consistent with the safe shutdown of the plant."
From these statements made in the FSAR, we have concluded that no valves in the instrument air and service air systems should be tested under the rules of ASME Section XI.
Thermal Relief Valves The RHR heat exchangers were provided with small relief valves during their i
fabrication which provide no safety function. Much larger re' lief valves which provide the required relief capacity are inciuded in this program.
l However, potential leakage through the thermai relief valves is tested for during the Appendix J local leak rate tests.
Passive Power Operated !!alv_es_
Power operated valves wisich are nnt required to change position to accomplish a specific safety-related functicii will have their positions verifieo quarterly and each time the valves are cycled.
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Class 1, 2, and 3 Valve Testing Program Notes:
- 1.. This valve will.not be stroked because it is a passive containment isolation valve.
It is normally closed and does not have to open to perform any. safety-related function.
2.
Note 2 is left blank intentionally.
3.
The test frequency is as required'by IWV-3511. The opening pressure will be determined per the requirements of ASME'PTC 25.3-1976.
4.
The testing of the standby liquid control system explosive-actuated valves will be performed per the Technical Specifications. requirements which exceed the. requirements of IWV.
5.
The opening function of this normally closed check valve is proven every 3 months.by pump operability testing.
Design flow will be passed through the valve.
.6.
.This valve is exempted from testing per IWV-1200. The function-of this flow control' valve is system control only.
7.
This normally closed fail open valve opens when the diesel generator starts. Operability of the valve is proven by pump testing.
8.
This normally closed valve opens and closes at set pressures. The safety-related function of this valve is in the closed position as a containment isolation valve. Operability of the valve is proven during pump testing.
9.
This valve is a fail-open A0 pressure control valve that opens and closes at set pressures.
The primary function of the valve is to regulate system pressure as'the pump is started up.
Operability of the valve is proven by-pump testing.
- 10. This valve is a self-modulating pressure control valve that is exempt from testing per IWV-1200(a).
The fail-open position will be verified during pump testing by observing that the valve is open when there is no flow through the valve.
11.
This check valve is normally full open; therefore, it does not have to change position to perform any safety-related function.
12.
This valve is used for testing valves T48-F323A-L (or 2T48-F323A-L) and has no safety related function.
1 l
i Lo I
04111
O closure and position indication lights.
13.
This normally-open feedwater check valve has an air assist for tight Closure will be proven each cold shutdown, but not more frequent than once per 3 months, by observing the indicating lights.
- 14. This normally. closed. valve is opened during testing only; therefore, it is passive'and does not require testing.
- 15..This valve is locked open and has no safety-related function requiring stroking the valve.
- 16. This valve functions only as a pressure / flow control valve; therefore, it is exempt per IWV-1200.
- 17. These thermal relief valves on the RHR heat exchangers were supplied by the manufacturer to meet code requirements.
The operability of these valves is not required since relief valves with much larger capacities were added to the system during design.
The safety-related closed position of these valves is proven each refueling outage during leak rate testing.
- 18. The jockey pump system is a continuously operating system that ensures that RHR and core spray remain water filled.
Excess flow from the jockey pumps is returned either through valves 1(2) E11-F044A or B, or both, depending upon alignment.
The safety-related closed position of each valve i
is proven each refueling outage during leak rate testing.
)
I J
lO 04111
6.1 RELIEF REQUESTS FOR CLASS 1, 2, AND 3 VALVES 6.1.1 TEST REQUIREMENT IWV-3521 requires that check valves be exercised at least once per 3 months.
6.1.1.1 Basis For Relief This containment isolation valve is normally open with design flow passing through it during normal operation.
The only F c.ctical method of verifying
{
closure is to introduce reverse flow througn the valve cnd measure leakage.
6.1.1.2 Alternate Testing This verification is performed each refueling outage during local leak rate testing.
6.1.2 TEST REQU:REMENT IWV-3423(d) requires that gate valves with the functional differential pressure greatar than 15 psi be tested for leaksge in the same direction as when the valve is performing its function.
6.1.2.1 Basis for Relief The correct direction is to pressurize from the inboard side of the valve; however, the piping on the inboard side runs directly from the valve to the torus (or the reactor vessel) and cannot be pressurized for testing.
6.1.2.2 Alternate Testing This containment isolation valve will be leak rate tested in a nonconservative reverse direction as addressed in the containment leak rate test program for the Type C leakage tests.
6.1.3 TEST REQUIREMENT IWV-3521 requires that check valves be exercised at least once per 3 months.
6.1.3.1 Basis for Relief This valve is designed to prevent a water leg from forming in the relief line discharge line under vacuum conditions.
Since there is no flow through the valve and it is a simple check valve, it cannot be tested as required.
6.1.3.2 Alternate Testing Each refueling outage (when the containment is de-inerted) the movement of the disc will be observed to see that it moves freely.
O 04111
6.1.4. TEST REQUIREMENT.
IWV-3411 requires that valves. be exercised at least once per 3 months.
6.1.4.1. Basis for Relief Th"s valve cannot be full-stroked during power operation because flow to the
{
core would be reduced by one-half.
The valve control circuitry does not allow partial stroking of the valve.
1 6.1.4.2. Alternate Test _in_g This valve will be tested each cold shutdown but.not more' frequently than once per 3 months.
-1 6.1.5 TEST REQUIREMENT IWV-3521 requires that check valves be exercised at least once per 3 months.
6.1.5.1 Basis for Relief i
This normally closed standby liquid control check valve cannot be opened l
without introducing flow through it with a pressure greater than reactor i
pressure.
6.1.5.2 Alternate Testing Design flow through this valve is achieved once per 18 months due to the requirements of the Technical Specifications.
6.1.6 TEST REQUIREMENT-IWV-3521 requires that check valves be exercised at least once per 3 months and IWV-3421 requires that Category A valves be leak tested.
6.1.6.1 Basis for Relief Since there is no valve between the check valve and the torus the line cannot be pressurized to ensure closure of the valve.
This valve is the inboard containment isolation valve for the core spray test line.
The outboard isolation is a closed system as defined in the Appendix J program. This valve -is sealed from the primary containment atmosphere because the test line terminates below the water level of the torus and the leakage is not included in the Type C local leak rate testing.
6.1.6.2 Alternate Testing The integrated leak rate test will prove that leakage through the valve and closed system does not occur.
There are no other practical means of testing 3
this check valve.
i m
04111
r_____
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6.1.7 TEST REQUIREMENTS IWV-3521. requires that check valves be exercised at least once per 3 months.
6.1.7.1 Basis for Relief These normally closed check valves do not receive " full-open" flow during normal operation'or shutdown conditions.
There has been no practical method demonstrated to fully stroke these valves.
6.1.7.2 Alternate Testing i
At least one of the RHR. system check valves will receive shutdown cooling flow through it during cold shutdown conditions; however, the' flow is not sufficient to fully open the valve.
~
Valve integrity will be proven during leak rate tests each refueling outage because this valve is a pressure isolation valve and must pass the leakage tests.
One of.the two valves in the system will be disassembled on a rotating basis every other refueling outage.
If the valve is determined to be nonfunctional, so that design flow would not pass through the valve, the other valve would then be disassembled and inspected, t
'6.1.0 TEST REQUIREMENT IWV-3521 requires that check valves be exercised at least once per cycle.
6.1.8.1 Basis for Relief Testing of this normally closed check valve during normal operation requires removing the associated RHR train from. operational status in order to relieve the differential pressure across the valve, thereby decreasing the level of plant reliability.
6.1.8.2 Alternate Testing The valve will be tested each refueling outage but not more frequently than once-per 3 months.
6.1.9 TEST REQUIREMENT IWV-3417(a) states that if an increase in stroke time of 25 percent or more from the previous test for valves with stroke times greater than 10 seconds or '50 percent or more for valves with stroke times less than or equal to 10 seconds is observed, test frequency shall be increased to once each month until corrective action is taken.
Relief is requested for valves normally tested during cold shutdown.
6.1.9.1 Basis for Relief Valves th are normally tested during cold shutdown or refueling cannot be tested once each month.
Stroking these valves during power operation may place the plant in an unsafe condition.
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6.1.9.2 ' Alternate Testing These valves will be stroked every cold shutdown but not more frequently than once per 3 months.
6.1.10 -TEST REQUIREMENT IWV-3411 requires that valves be exercised at least.once per 3 months.
6.1.10.1 Basis for Relief Valve E41-F007 cannot be closed during normal. operation because its failure in the closed position would result in the loss of the HPCI system.
Valve E41-F006 cannot be opened during normal power operation without first closing-E41-F007.
.6.1.10.2 -Alternate Testing The operability of 1(2)E41-F006 and 1(2)E41-F007 will be demonstrated during l
each cold shutdown but not more frequently than once per 3 months.
6.1.11. TEST REQUIREMENT IWV-3423 requires that valves be leak tested with the pressure differential in the same direction as when the valve is performing its function.
6.1.11.1 Basis for Relief This turbine exhaust containment isolation valve is a normally closed stop check valve with the closure mechanism in the " locked open" position.
The.
valve then functions as a simple check valve.
The piping on the inboard side 1
of.the valve runs directly from the valve to the torus and cannot be pressurized-for testing.
6.1.11.2 Alternate Testing As an alternate, the valve is closed with the closure mechanism and leak tested from the reverse side. The leak test, as defined in the containment leak rate test program, is conservative since the test pressure tends to lift the disc from the seat much in the same manner as reverse testing a globe valve.
6.1.12 TEST REQUIREMENT IWV-3521 requires that check valves be exercised every 3 months.
6.1.12.1 Basis for Relief
('
Check valves 1(2)E51-F001,1(2)E51-F002,1(2)E51-F028,1(2)E51-F040,.
l 1(2)E41-F021, 1(2)E41-F022, 1(2)E41-F040, and 1(2)E41-F049 are located on the 3
RCIC or HPCI turbine steam exhaust lines.
During quarterly pump testing j
these valves are partially stroked during system operation; however, test conditions do not provide sufficient flow to prove that the valves are fully open.
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L__________-__-_________-
i i
6.1.12.2 Alternate Testing One of these valves will be disassembled on a rotating basis each refueling i
outage.
If the valve is determined to be.non-functional, so that design flow l
would not pass through the' valve, an additional valve will be disassembled.
Failure of this valve would require disassembly of the remaining valves.
6.1.13 TEST REQUIREMENT IWV-3521 requires that check valves be exercised at least once per 3 months.
6.1.13.1 Basis for Relief Valve 1(2)E41-F045 is a normally closed check valve located on the HPCI pump l
suction line and cannot be stroked.
This valve does not see flow during any i
normal mode of' reactor operations or shutdown conditions.
Testing of this valve would require pumping water from the torus to the condensate storage tank, thereby lowering the water quality in the tank.
6.1.13.2 Alternate Testing Degradation of this valve is n,t expected because wear does not occur in a closed check valve.
To ensure that the valve will open if needed, it will be disassembled every other refueling outage to prove that the disc is free to move.
h 6.1.14 TEST REQUIREMENT IWV-3521 requires that check valves be exercised at least once per 3 months.
6.1.14.1 Basis for Relief Normally closed check valves P41-F024A & B, P41-F025A & B, P41-F026A & B, P41-F028A & B, and P41-F522A & C, (or Unit 2 valves 2P41-F024A & B, 2P41-F025A & B, 2P41-F026A & B, 2P41-F028A & B, and 2P41-F321) supply service water to RHR and core spray, HPCI, and RCIC pump and pump room coolers.
During quarterly testing of the pumps, the coolers are placed in operation,
-thereby stroking these valves.
However, the design of the system does not provide for positive verification of the flow rate through each valve.
6.1.14.2 Alternate Testing One valve will be disassembled each refueling outage on a rotating basis to ensure that the design function of the valve can be achieved.
If the valve
.is determined to be nonfunctional, an additional valve will be disassembled.
Failure of this valve will require the disassembly of the remaining valves for that unit.
6.1.15 TEST REQUIREMENT IWV-3413(b) requires stroke. times shall be measured to the nearest second, I
for stroke times 10 seconds or less, or 10 percent of the specified limiting stroke time for full-stroke times longer than 10 seconds.
0411I
-IWV-3417 requires that if the stroke time increases by 25 percent from the previous test'for valves with full-stroke times greater than 10 seconds or 50 percent for valves will full-stroke times less than 10 seconds, the test frequency shall be increased to.once each month until corrective action is taken.
6.1.15.1 Basis For Relief These valves are air operated valves without indicating lights or control switches. Measurement of stroke times can be performed only by observation of the stem movement when the associated room cooler is placed into operation. This type of testing does not provide the accuracy required by IWV-3413(b) and IWV-3417.
6.1.15.2 Alternate Testing A maximum stroke time which will be as short as practical will be assigned to each valve.
If'the measured stroke tima exceeds this value, the valve will be declared inoperable.
6.1.16 TEST REQUIREMENT IWV-3411 requires that valves be exercised at least once per 3 months.
6.1.16.1 Basis for Relief Closure of this normally open valve would totally interrupt flow to the drywell coolers.
This interruption may cause an increase in drywell temperature which would require removing the unit from operation.
The valve cannot be partially stroked due to control circuitry.
{
i 6.1.16.2 Alternate Testing
)
This valve will be stroked every cold shutdown but not more frequently than once per 3 months.
6.1.17 TEST REQUIREMENT IWV-3411 requires that valves be exercised at least once per 3 months.
6.1.17.1 Basis for Relief Closure of this valve during normal power operation would interrupt flow to the turbine building equipment normally cooled by service water.
This valve cannot be partially stroked due to control circuitry.
6.1.17.2 Alternate Testing i
This valve will be stroked every cold shutdown but not more frequently than
_once per 3 months.
J 6.1.18 TEST REQUIREMENT O
i IWV-3411 requires that valves be exercised at least once per 3 months.
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0411I j
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(a~'l 6.1.18.1 Basis for Relief Closure of this valve during normal operation vould shut off the cooling water to a reactor recirculation system pump.
6.1.10.2 Alternate Testing i
This valve will be stroked every cold shutdown but not more frequently than once per 3 months.
6.1.19 TEST REQUIREMENT IWV-3417(b) and IWV-3523 state that when corrective action is required as a result of tests made during cold shutdown, the condition shall be corrected before startup.
6.1.19.1 Basis for Relief Startup of the plant is governed by the Technical Specifications.
6.1.19.2 Alternate Testing Under such conditions startup shall be permitted as provided in the Technical Specifications.
f-3 6.1.20 TEST REQUIREMENT
,J IWV-3411 requires that valves be exercised at least once per 3 months.
6.1.20.1 Basis for Relief If this valve is opened during normal operation, a single failure of the second isolation valve in the line or operator error may result in damage to the lower pressure piping and equipment.
Due to control circuitry this valve cannot be partially stroked.
6.1.20.1 Alternate Testing This valve will be stroked every cold shutdown but not more frequently than once per 3 months.
6.1.21 (This Relief Request Number is Not Used) 6.1.22 TEST REQUIREMENT IWV-1100 provides the rules and requirements for inservice testing of certain Class 1, 2, and 3 valves which are required to perform a specific function in l
shutting down a reactor to the cold shutdown condition or in mitigating the j
consequences of an accident.
f The control rod drive (CRD) system valves will not be individually tested per
,r-this requirement.
\\_-
04111
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6.1.22.1 Basis for Relief V
The Plant Technical Specifications require all operable withdrawn control rods to be exercised at least once per week when above a designated power level.
After each refueling outage, all control rods capable of normal insertion shall be scram timed from the fully withdrawn position. Also, 10 percent of i
the rods shall be scram timed from the fully withdrawn position at least once per 120 days of operation.
The Technical Specifications give the allowable insertion times for these tests.
6.1.22.2 Alternate Testing The Technical Specifications adequately demonstrate the operability of the control rod drive system and the additional requirements of ASME Section XI would not increase the level of safety.
Therefore, Technical Specifications testing will be used in lieu of the ASME Section XI requirements.
6.1.23 TEST REQUIREMENT Pelief Request 6.1.23 has been withdrawn.
6.1.24 TEST REQUIREMENT Relief is requested from applying the requirements of IWV-3427(b) to the main h
steam isolation valves (MPLs:
1821-F022A through D, 1821-F028A through D, d
2821-F022A through D, and 2B21-F028A through D).
I 6.1.24.1 Basis for Relief ASME Section XI requires that the valves be tested once per 24 months unless a test shows that the margin between the measured leakage rate and the j
maximum permissible rate has been reduced by 50 percent or greater.
In this
]
case the frequency of testing should be doubled (or approximately every 12 1
months) to coincide with a cold shutdown. Also, trending of the leakage rates is required.
1 The Technical Specifications require that these valves be tested on a much
{
more frequent basis or once per refueling outage.
These 24-inch globe valves are tested by applying 28 psig between the inboard and outboard valves with a J
maximum allowable leakage of 11.5 SCFH.
This allowable leakage is small when compared to other Category A valves.
Some leakage may occur through mechanically sound valves because the inboard valve is reverse tested, which tends to lift the globe from the seat, thereby allowing leakage that would not occur during a normal pressurization.
Since the normal testing frequency applied by the Technical Specifications are very similar to those applied by IWV-3427(b), the provisions of IWV-3427(b) need not be applied.
Using the conservative test, with low allowable leakages at a frequency of once per refueling outage, any significant degradation of the valve will be detected.
D 04111
[d 6.1.24.2 Alternate Testing These valves will oe conservatively tested each refueling outage using the parameters as defined in the Technical Specifications.
6.1.25 TEST REQUIREMENT IWV-3413(b) requires stroke times to be measured to the nearest second, for stroke times 10 seconds or less, or 10 percent of the specified limiting stroke time for full-stroke times longer than 10 seconds.
IWV-3417 requires that if the stroke time increases by 25 percent from the previous test for valves with full-stroke times greater than 10 seconds or 50 percent for valves will full-stroke times less than 10 seconds, the test frequency shall be increased to once each month until corrective action is taken.
6.1.25.1 Basis for Relief The normally open transversing incore probe containment isolation valves are quick closure valves with a closure time of approximately 1 or 2 seconds.
l Application of IWV-3413(b) and IWV-3417 to these valves would provide no j
useful function.
I l
6.1.25.2 Alternate Testing l
(O)
A maximum stroke time which will be as short as practical will be assigned to
'~'
each valve.
If the measured stroke time exceeds this value, the valve will be declared inoperable.
6.1.26 TEST REQUIREMENT IWV-3410(g) and IWV-3520(c) state that when corrective action is required as a result of tests made during cold shutdown, the condition shall be corrected l
before startup.
6.1.26.1 Justification Under such conditions startup shall be permitted as provided in the Technical Specifications.
6.1.27 TEST REQUIREMENT IWV-3413(b) requires stroke times shall be measured to the nearest second, i
for stroke time 10 seconds or less, or 10 percent of the specified limiting l
stroke time for full-stroke times longer than 10 seconds.
I IWV-3417 requires that if the stroke time increases by 25 percent from the previous test for valves with full-stroke times greater than 10 seconds or 50 percent for valves will full-stroke times greater than 10 seconds, the test frequecy shall be increased to once each month until corrective action q
is taken.
V 0411I L_ _ _-_ _
O 6.1.27.1 Basis for Relief Solenoid valves are quick actuating valves and the application of IWV-3413(b) and IWV-3417 to these would provide no useful function.
6.1.27.2 Alternate Testing A maximum stroke time which will be as short as practical will be assigned to each valve.
If the measured stroke time exceeds this value, the valve will g
be declared inoperable.
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O 0411I
7.0 INSERVICE TESTING OF PUMPS 4
The 1980 ASME Section XI Code with Addenda through Winter 1981 requires inserv'ce testing of pumps in accordance with Section IWP.
The inservice testin program for Class 1, 2, and 3 pumps is described in the following tables Where full compliance with the Code is not practical, relief has been requested.
Per IWP-3210, Georgic Power Company has set its own allowable ranges of test parameters.
Vibration Limits For the RHR service water pumps, the plant service water pumps, and the standby diesel generator service water pumps, the following ranges are different than those found in Table IWP-3100-2.
Reference Quantity Acceptable Alert
, Required Action l
0<V s2.0 0<Vs2V 2V <Vs3V V>3V r
r r
r r
2<V s5 0<Vs(2+V )
(2+V )<Vs(4+V )
V>(4+V )
I r
r r
r r
V >5 0<Vs(1.4V )
1.4V <Vs1.8V Y>1.8V r
r r
r r
V is procedurally set to be a minimum of 1.5 mils.
Per manufacturers r
recommendations, these service water pumps will operate satisfactorily with l
e vibrations in excess of 4 mils.
Flow and Differential Pressure Limits For the RHR pumps, the core spray pumps, the standby liquid control pumps, the RHR service water pumps, the standby diesel generator service water pumps, the jockey pumps (AP only, Relief Request 7.1.7), the HPCI pump, and I
the RCIC pump, the following ranges are different than those found in Table IWP-3100-2.
Test Quantity Acceptable Alert Required Action AP 0.93 to 1.05 AP 20.9 and <0.93 AP
<0.9 or >1.05 AP R
R g
Q 0.94 to 1.05 Q 2.9 and <.94 Q
<0.9 or >1.05 Q R
R R
These values were determined through a review of historical data and the use of such will not impact the ability to detect degradation of the pumps.
5 04111
O For the plant service water pumps, the following ranges are different than those found in Table IWP-3100-2.
Reference Quantity Acceptable Alert Required Action l
AP 0.93 to 1.1 AP 20.9 and <0.93 AP
<0.9 or >1.1 or AP R
R R
Q 0.94 to 1.1 Q
20.9 and <0.94 Q
<0.9 or >1.1 Q R
R R
These values were determined through a review of historical data and the use of such will not impact the ability to detect degradation of the pumps.
The variance for these pumps is due primarily to using a manual butterfly valve for throttling and the subsequent inability to match the flowrate of the previous test.
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t 7.1 RELIEF REQUESTS.FOR CLASS 1, 2, and 3 PUMPS 7.1.1 TEST REQUIREMENT IWP-4310 requires that.the temperature of bearings outside the main flow path
.be measured.
7.1.1.1 Basis for Relief The bearings (or bearing surfaces) are cooled by'the process fluid that is j
being pumped.
7.1.1.2 Alternate Testing
'The bearings will be inspected for wear whenever the pump is disassembled for maintenance.
l 7.1.2 TEST REQUIREMENT Table IWP-4110-1 requires that flow be measured within 12 percent of full scale.
7.1.2.1 Basis for Relief Flowrate is measured by the change in the standby liquid control' test tank i
level during a two minute test period.
O.
7.1.2.2 Alternate Testing The standby ' liquid control (SBLC) is aligned so that each pump takes suction from a demineralized water source and discharges through a threttle valve to a test tank. The pump is placed into operation and the throttle valve adjusted to obtain a reference discharge pressure.
The level of the test tank is then measured and the pump is run for two minutes. After the two minute run, the tank level is again measured.
Flowrate is then determined by the following equation:
Flow (gpm) = A Tank level (in.) x 4.81 gal /in.
2 min For a situation in which the flowrate is measured by instrument, a 0-100 gpm instrument would normally be used for the SBLC pump flowrate of approximately 43 gpm.
The required accuracy of this instrunent would be 12 percent or +2 gpm.
This corresponds to a 10.83-inch water level in the test tank.
1
-Therefore, the accuracy of the measured flowrate should be well within Code allowance.
7.1.3 TEST REQUIREMENT Table IWP-4110-1 requires that pressure be measured within 12 percent of full scale.
I r, d
1 0411I
i 7.1.3.1 Basis-for Relief Inlet pressure is detertained for this group by measuring the river level at
.the intake structure.
The differential pressure is then:
AP = P, + (114.5 ft - River Water Level) x 0.433 where AP is.the differential pressure and P the outlet pressure.
l g
7.1.3.2 Alternate Testing This method of measurement is well within the code requirement for the-determination of the differential pressure.
7.1.4 TEST REQUIREMENT Table IWP-3100-1 requires that the proper lubricant level or pressure be observed.
7.1.4.1 Basis for Relief This pump is lubricated by the process fluid that is being pumped.
7.1.4.2 Alternate Testing-The applicable parameters in Table IWP-3100-1 will be measured.
7.1.5 TEST REQUIREMENT Table IWP-3100-1 requires that inlet pressure, differential pressure, and flowrate be measured.
7.1.5.1. Basis for Relief The standby liquid control pumps are positive displacement pumps for which the outlet pressure is constant regardless of inlet pressure.
No AP or inlet pressure measurements will be made.
7.1.5.2 Alternate Testing The outlet pressure and the flowrate is measured.
7.1.6 (Relief Request Number Not used) 7.1.7 TEST REQUIREMENT Table IWP-3100-1 requires that both the differential pressure and the flowrate be measured.
7.1.7.1 Basis for Relief The only function of this pump is to maintain a water inventory in the RHR and core spray piping system and flow rate instrumentation is not required.
0411I
)
l l
i I
l 1
(V)
However, dischar'ge is into a fixed resistance system; therefore, measurement l
of both parameters is not required to determine degradation of the pump.
]
l 7.1.7.2 Alternate Testing i
The differential pressure will be measured during each pump test.
7.1.8 TEST REQUIREMENT
~
Table IWP-4110-1 requires that pump monitoring instruments be within the following accuracies:
1 Pressure
+ 2% of full scale l
Differential Pressure 7 2% of full male Flowrate 7 2% of full scale 7
Speed
_ 2% of full scale Temperature
+ 5% of full scale Vibration Amplitude
+ 5% of full scale j
7.1.8.1 Basis for Relief Instrumentation is used which meets the acceptable instrument accuracies defined in Table IWP-4110-1.
However, the total loop accuracy is not easily determined (there are several calculational techniques) and the total loop accuracy may fall outside the Code limits.
m
(
7.1.8.2 Alternate Testing The instrumentation used by Georgia Power Company to measure pump operability parameters should provide data that is sufficiently accurate to allow assessment of pump condition and to detect pump degradation.
t
[h 04111
g 8.0 GENERAL RELIEF REQUESTS Georgia Power Company also has several relief requests which do not apply directly to any individual section but are general in nature.
8.1 REQUESTS FOR RELIEF FROM ASME SECTION XI REQUIREMENTS 8.1.1 BASIC CALIBRATION BLOCKS FOR PIPE WELD EXAMINATIONS 8.1.1.1 Requirement From Which Relief is Requested Appendix III of Section XI delineates the requirements for the design and fabrication of basic calibration blocks for pipe weld examinations.
It specifies that the basic calibration block shall be fabricated with notches and that the basic calibration block nominal diameter and thickness be equivalent to the component to be examined.
Relief is requested so that existing basic calibration blocks may be used for pipe weld examinations.
s 8.1.1.2 Justification The majority of existing Hatch basic calibration blocks used for pipe weld examinations were fabricated with diameters, thicknesses, and side-drilled holes in accordance with the 1974 Edition of ASME Section V.
For the two primary reasons listed below, these same basic calibration blocks will be used to provide the most meaningful and thorough examinations possible:
1.
Side-drilled holes as calibration reflectors result in a more sensitive ultrasonic examination than one using notches.
2.
Correlation of ultrasonic data with previous examinations as required by Subarticle IWA-1400 of Section XI makes it necessary that these basic calibration blocks be used so future examination results can be correlated with past results.
8.1.1.3 Testing in Lieu of Section XI Requirements The basic calibration blocks using side-drilled holes as calibration reflector' will be used for the majority of the pipe weld examinations.
O 0411I
1
-8.1.2 ALLOW HATCH 2 TO START ITS SECOND 10-YEAR INSPECTION INTERVAL ON s_
JANUARY 1, 1986
'8.1.2.1 Requirement from Which Relief is Requested According to 10 CFR Part 50.55a(g)4(ii), inservice examination of components during successive 120-month inspection intervals shall comply with the requirements of the latest edition and addenda of the approved ASME Section XI Code 12 months prior to the start of the 120-month inspection interval.
Permission is requested to start the Hatch 2 second 10-Year inspection interval ahead of schedule such that the two Hatch units would be under the i
same edition and addenda of the Code. The second 40-month inspection period for Hatch 2 ended on'May 5, 1986.
Hatch 2 would normally update for the second 10 year inspection interval in September 1989.
8.1.2.2 Justification The Inservice Inspection requirements for Hatch 1 are presently in compliance with ASME Section XI, Winter 1981 Addenda, effective January 1, 1986.
A move l
to start the second 10 year interval for Hatch 2 on January 1, 1986, would require compliance with ASME Code Section XI, Winter 1981 Addenda, which is more stringent. A more comprehensive inspection would be achieved.
In l
addition, compliance with the same Code for both units will enhance the possibility of detecting a generic problem and would also reduce cost l
involved in maintaining two separate programs. The RPV welds will also be examined at the same point in time as was planned in the earlier program (i.e., those welds that were scheduled during the third 40-month period will
~'
be examined during the first 40-month period of the new 10-year interval).
In discussions with the NRC HNP licensing project manager and appropriate NRC staff personnel, they have indicated that this is a reasonable approach. The proposed change will not endanger public health and safety.
8.1.2.3 Testing in Lieu of Section XI Requirements None.
8.1.3 INCORPORATION OF CHANGES IN DESIGN, TESTING, AND PROCEDURES WITHOUT ANY UNREVIEWED SAFETY QUESTIONS INTO THE INSERVICE INSPECTION PROGRAM / PLAN 8.1.3.1 Requirement From Which Relief is Requested 10 CFR Part 50.59 allows changes to a nuclear facility and procedural changes in accordance with Plant Technical Specifications and the safety analysis report, without prior approval from the NRC, provided there are no unreviewed l
safety questions.
The facility's Inservice Inspection Program / Plan should be revised to include these changes, L
Changes such as valve operator, stroke time, pump performance, welds, etc.,
made in compliance with 10 CFR 50.59, which affects the ASME Section XI requirements may not be shown on the existing revision of the ISI program
- ()
plan document.
A relief is requested from this inconsistency.
0411I i
i 8.1.3.2 Justification Records of these changes will be maintained and changes in ASME Section XI requirements pertaining to these changes will be complied with.
However, a revision to the ISI Program Plan (s) for every minor change is an unrealistic, costly, and time consuming task.
Any delay in revising the ISI program
(
plan (s) will not endanger public health and safety.
8.13.3 Testing in Lieu of Section XI Requirements Records of all changes will be maintained.
Changes affecting ASME Section XI
-requirements shall be incorporated into the ISI Program Plan (s) whenever a need for their update is warranted.
8.1.4 EXEMPTING SUBSECTION IWE, AND THE REFERENCES TO IWE IN OTHER SUBSECTIONS 8.1.4.1 Requirement from Which Relief is Requested Table IWE-2500-1, Code categories E-A, E-A-1, E-B, E-C, E-D, E-E, E-F, E-G, and E-P of ASME Section XI, Winter 1981 Addenda delineates examination requirements of containment.
8.1.4.2 Justification Federal Register, Vol. 48, No. 28/ Monday, February 7,1983, Page 5532 relative to 10 CFR Part 50 Codes and Standards for Nuclear Power Plants; Item No. 4, indicates that Subsection IWE " Requirements for Class MC Components of Light Water Cooled Power Plants" was added to Section XI by'the Winter 1981 Addenda.
However, 10 CFR Paragraph 50.55a presently only incorporates those portions of Section XI that address the ISI requirements for Class 1, 2, and 3 components and their supports.
The regulations.do not currently address the ISI of containments.
Since this amendment is only intended to update current regulatory requirements to include the latest Code addenda, the requirements of Subsection IWE are not imposed upon Licensees by the NRC as a result of this amendment.
8.1.4.3 Testing in Lieu of Section XI Requirements None.
8.1.5 DUE DATE FOR OWNER'S DATA REPORT FOR INSERVICE INSPECTION, FORM NIS-1 (Relief Request 8.1.5 has been withdrawn) l-8.1.6 REFERENCE SYSTEM FOR ALL WELDS AND AREAS SUBJECT TO SURFACE OR VOLUMETRIC EXAMINATION 8.1.6.1 Requirement From Which Relief is Requested Section IWA-2600 of ASME Section XI, with Addenda through Winter 1981, requires the establishment of a reference system for all welds and areas V
1 04111 E_-_---.--------__
subject to surface or volumetric examination.
This system shall permit identification of each weld, the location of each weld centerline, and marking at regular intervals along the length of each weld.
Relief from this l
requirement is requested.
8.1.6.2 Justification At Hatch Units 1 and 2, physical and radiological limitations prevent the actual marking of the majority of the RPV welds, nozzle welds, and piping welds.
In many instances, these limitations actually prohibit complete examination of the welds.
Relief requests have already been submitted herein for the welds which cannot be completely examined and thus will be examined to the extent possible.
Many of the welds which could be accessed are covered by insulation.
It would be inappropriate to require the manhours of exposure involved to remove and replace the insulation to facilitate marking.
Each weld that has been or will be examined is referenced to permanently located fixtures in the immediate area of the subject weld.
These references include such items as nozzles, plant azimuth and floor identifications, welded attachments, and concrete embedments.
tach weld is described in the inspection plan and on the examination reports in such a manner as to make it unique.
For each weld that is examined, the starting point is identified and the direction of travel is noted on the examination report in order that any indications can be accurately located.
The configuration of the welds at Hatch permits the inspectors to locate the weld edges which are used to locate the weld centerline.
8.1.6.3 Testing in Lieu of Section XI Requirements None.
1 (m) 04111 l
_ - _ _ _ _ _ _ _ _ _