ML20002C669
| ML20002C669 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/04/1972 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20002C667 | List: |
| References | |
| NUDOCS 8101100688 | |
| Download: ML20002C669 (12) | |
Text
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ATTACHMENT A CHANGE NO. 30 TO THE TECHhiCAL SPECIFICATIONS FACILITY LICENSE NO. DPR-6
_ CONSUMERS P0k'ER COMPANY DOCKET NO._50-155 IN_SERVIC_E INSPECTION OF REACTOR COOLANT SYS_T_ EMS 1.
Add a new section, "Section 9",
to the Technical Specifications to read as follows:
"9 0 PRIMARY SYSTEM BURVEILLANCE r,.1 APPLICABILITY Applies to in-service structural surveillance of primary system co=ponents.
92 OMECTIVE To assure the ir.tegrity of the reactor pressure boundary.
93 spec:FICATIONS (a) Post-operational inspections shall be made according to the methods and intervals indicated in Table 9 3(a) and IS-2h2 of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 edition, with su==er 1971 addenda.
(b) The structural integrity of the primary system boundary shall be maintained at the level required by the original acceptance standards throughout the life of the plant.
Any evidence as a result of the tests outlined in Table 9 3(a), that defects have developed or grown shall be investigated, including evaluation of comparable areas l
of the primary system.
(c) Sufficient records of each inspection shall be maintained to allow co=parison and evaluation of future tests.
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$i (d) A surveillance program to monitor radiation-induced changes in the mechanical and impact properties of the reactor vessel material shall be maintained as required by Section L.l.l(h) of these Technical Specifications.
9.4 BASIS This inspection program implementsSection XI of the 1971 A5xI Boiler and Pressure Vessel Code to the maximum extent practical.
It is recognized that plant design and construction were com-pleted approximately seven years prior to the development of Section XI and it is therefore not possible to comply fully with the code. Areas of exception in the Big Rock Point In-Service Inspection Program are based primarily on experience gained during the three previous in-service inspections con-ducted prior to formulation of this program. Based on this previous experience, a detailed review of piping and equipment fabrication drawings, plant physical constraints (1 and Sec-tion XI, Table 9 3(a) was developed to summarize the plant in-service inspection requirements. Significant exceptions to Section XI are as follows:
Paragraph IS-211.2 Remote visual examination will be used for visual inspections.
Depending on the physical arrangement of the area to be in-spected, it may or may not be possible to obtain the resolution rpecified in IS-211.1. In any circumstance, the best resolation possible shall be obtained for the inspection.
Paragraph IS-232 No preoperational examination can be made because the plant has been in operation since 1962. It would.also be impractical to attempt an examination of this nature at this time due to high radiationfieldsand/ordifficultiesingainingaccesstosome areas of the reactor coolant pressure boundary.'
l (1) September 29, 1971 letter from R. L. Haueter to i
Dr. Peter A. Morris
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I Table IS-261 Item 1.1 - Longitudinal and circumferential welds in the core region are not accessible because the outside of the reactor vessel is closely surrounded by concreta and access from the inside is not feasible due to the proximity of the thermal shield.
Item 1.2 - Longitudinal and circumferential welds in the reactor vessel shell are not accessible with existing equip-ment. Meridional welds in the lower head are not accessible because of the core support plate and poor geometry due to 40 penetrations.
Item 1 3 - We shall attempt to inspect reactor vessel to flange weld from flange surface. The weld is 22" below the surface of the flange and attenuation may limit the validity of this inspection.
Ite: 1.h - Primary nozzle to reactor vessel welds are inacces-sible due to configuration of the reactor vessel in-ternals and/or lack of equipment te perform an in-spection of this nature nor does y,lant decign permit inspection of these nozzle welds from the reactor vessel exterior.
Item 1.5 - This item has been deleted with the exception of the control rod drive "J" welds since plant design and present technology do not permit a volumetric exam-inatinn of this type of penetration.
Item 1.6 - None of the reactor vessel penetrations, except clo-sure head penetrations, are visually accessible because of plant design.
Item 1.7 - A device is being developed to attempt to incpect volumetrically the six reactor vessel steam outlet nozzles primary nozzle to safe-end welds. These welds are not accessible either visually or by sur-face examination. The two 3" vent nozzles in the reactor vessel head are accessible for visual and surface inspections. Previous attempts to inspect
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these welds volumetrically were unsuccessful due to the physical configuration of the nozzles. All other reactor vessel primary nozzle to safe-end welds are inaccessible because plant design will not allow an inspection of this type.
Item 1.12 - Th$s item has been deleted since plant design does not allow access to reactor vessel support pads.
Item 1.15 - The core support plate brackets are not accessible for either direct or remote visual examination due to the physical configuration of the reactor vessel.
Item 3.1 - The steam drum longitudinal welds will not be inspec-ted. They are located near the horizontal plane through the center line of the steam drum and are, therefore, inaccessible due to the high radiation field and the inaccessibility of the sides of the steam drum. The circu:ferential steam drum welds will be inspected only within 30 of are from the top of the steam drum.
Item 3.2 - The steam drum prima a nozzle to reactor vessel velds and inside radius dections will not be inspected from the drum exterior due to the high radiation field that exists beneath the steam drum and the extreme dif-ficulty in removing insulation from the steam drum.
The high radiation field that exists inside the dru=
precludes inspection from the inside.
Set on nozzle; on clean-up system heat exchanges will r;t be inspected volumetrically due to geometry.
Item 3 3 - The se.fe-end to nozzle weld on the steam drum vent line will not be inspected volumetrically because j
its geometry is such that it does not allow proper j
placement of an ultrasonic probe.
Item 3.6 - The entire inspection of steam drum integrally velded supports will be concentrated on the upper supports due to the extremely high radiation area under the steam drum.
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Item 4.1 - The following safe-end to pipe and carbon steel to stainless steel transition welds will not be in-spectedduetotheplantphysicallayoutand/or high radiation areas:
Transition welds:
6" shutdown system outlet line.
4" core spray line.
Safe-end to pipe veld:
Reactor vessel recirculating inlet nozzles.
Reactor vessel instrument nozzles.
Reactor vessel emergency coolant inlet nozzle.
Reactor vessel unloading outlet nozzle.
Reactor vessel poison inlet nozzle.
Steam drum vent nozzle - will inspect visually and surface, not volumetric (refer to Item 3.3).
Reactor vessel steam outlet nozzles - will inspect volue-trically only.
Item 4.2 - The longitudinal welds on the main recirculating system steam risers are inaccessible due to plant design and high radiation areas.
Some circumferential welds in portions of the var-ious systems that comprise the reactor coolant pres-sure boundary are inaccessible due to plant physical design and/or high radiation fields.
Item 4.3 - Not applicable.
Item 4.5 - Approximately 50% of the integreally welded pipe supports on the main recirculating system are in-accessible due to physical location and high radia-tion areas and will not be inspected.
Item 4.6 - The piping restraints on the downcom'ers will not be inspected due to physical plant layout.
Item 5.1 - Not applicable.
Item 5 3 - Not applicable.
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Item 5. 4 - Not applicable.
Item 5.6 - Not applicable.
Item 6.1 - Not applicable.
Item 6.2 - Main recirculating pun:p discharge and butterfly valves are not isolable from the reactor; therefore, they will not be inspected.
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Table 9 3(a)
PRIMARY COOIANT SYSTEM SURVEILIA'CE I
Item No.
(Corresponds to Examination i
Numbers in Category g
i Table IS-261)
Table IS-251 Components and Parts To Be Examined Method h
Reactor Vessel and Closure Head 1.2 B
Closure Head Welds Volumetric 4
I 13 C
Reactor Vessel to Flange Weld Volumetric f
Closure Head to Flange Weld Ditto b
1.5 E-1 Control Rod Drive Penetration "J" Welds Volumetric Closure Head Nozzle Welds Ditto 1.6 E-2 Closure Head Nozzle Welds Visual 5
1.7 F
Primary Nozzle to Safe-End Welds:
3 Reactor Vessel Steam Outlet Nozzles Volumetric 3
Access Ports in Closure Head Volumetric, Visual O
and Surface j
s 4
1.8 G-1 Closure Studs and N2ts Volumetric and Visual or Surface
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19 G-1 Ligaments Between Threaded Stud Holes Volueetric
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1.10 G-1 Closure Washers Visual G-2 Pressure Retaining Bolting Visual I
1.11
Item Fo.
~(Corresponds to Exanination Numbers in Category Table IS-261)
Table IS-251 Components and Parts To Be Examined Method I-Reactor Vessel and Closure Head (Contd) 1.13 I-1 Closure Head Cladding Visual and Surface i
or Voluuhetric
,A 1.14 I-1 Reactor Vessel Cladding Visual f
1.15 N
Reactor Vessel Thermal Shield Support Brackets and
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Interior Surfaces VisualJy Accessible Visual ij Heat Exchangers and Steam Drum 31 B
Longitudinal and/or Circumrerential Welds for:
Visual and Volumetric Clean-Up Regenerative Heat Exchangers Ditto i
Clean-Up Nonregenerative Heat Exchangers j
Clean-Up Demineralizer Tank 6
j Steam Drum (60 Arc at Top of Circumferential Welds)
Emergency Condenser i
t 32 D
Primary Nozzle to Vessel Head Welds for:
Visual and Volumetric O
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Set-In Nozzles on Clean-Up Regenerative Heat Exchanger Ditto
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Clean-Up Demineralizer i
Emergency Condenser 3
Steam Outlet and Relief Valve Nozzles on Steam Drum 33 F
Primary Nozzle to Safe-End Welds Visual and Surface and Volumetric a
2 J
3.4 G-1 Pressure Retaining Bolting Visual and Volumetric f
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Item No.
(Corresponds to Examination Numbers in Category Table IS-261)
Table IS-251 Components and Parts To Be Examined
!!ethod Heat Exchangers_and Steam Drum (Contd) 4 i
35 G-2 Prescure Retaining Bolting Visual A
3.6 H
Integrally Welded Vessel Supports Near Top of Steam Visual and Volumetric Drum 37 I-2 Steam Drum Cladding Visual Piping Pressure Boundary 1.1 F
Safe-Ends or Dissimilar Metal Welds:
4 Transitions at Clean-Up System Heat Exchangers Visual and Surface and Volumetric Transition at Shutdown System Inlet Ditto Transition at HCV-30 in Control Rod Drive System Control Rod Drive System Inlet to Clean-Up System Bypass on Control Rod Drive Pump to cooling Ileader Dypass r
Reactor Vessel Steam Outlet Nozzles Volumetric Steam Drum Downcomer Nozzles Visual and Surface and Volumetric Steam Drum Riser Nozzles Ditto Steam Drum Vent Nozzles Visual and Surface
Item No.
(Corresponds to Examination Numbers in Category Table IS-261) -Table IS-251 Components and Parts To Be Examined Method Piping Pressure Boundary (Contd) h.2 J
Circumferential Pipe Welds:
Visual and Volumetric No of Welds To Be Inspected Estimated No of During Inspec-Accessible System tion Interval Welds Main Steam 13 32 Feed-Water lh 45 Emergency Condenser 13 45 Clean-Up 40 122 Shutdown 11 34 Maist Recirculating 25 68 Core Spray 4
14 Redundant Core Spray 9
33 Poison >2" 12 22 h.4 G-2 Pressure Retaining Bolting <2" Visual 8
h.5 K-1 Integrally Welded Pipe Supports (Except Those Above Visual and Volumetric 585' Level in Recirculating Pump Room)
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4.6 K-2 Piping Support and Hanger (Except Those Between 595' Level and 636' Level in Recirculating Purep Room)
Vistn1 h.7 J-2 Circumferential and Longitudinal Pipe Welds and Branch Pipe Connection Welds Visual h.8 J-1 Socket Welds 2" and Grcntar Visus 1 and Surrace k
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- )l Item No.
(Corresponds to - Examination Numbers in Category Table 18-261)
Table IS-251 Components and Parts To Be Examined Method jh.
Pump Pressure Boundary a
Visual 52 L-2 Pump Casings r
L 5.5 G-2 Pressure Retaining Dolting <2" Visual y
Visual
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57 K-2 Supports and Hangers
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Valve Pressure Boundary
^d!t AO 6.2 M-2 Valve Bodies Visual IA r7 63 F
Valve-to-Safe-End Welds Visual and Volumetric
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6.4 G-1 Pressure Retaining Bolting >2" Visual and Volumetric M.
6.5 G-2 Pressure Retaining Bolting <2" Visual h
6.6 K-1 Integrally Welded Supports Visual and Volumetric O
Visual 6.7 K-2 Supports and Hangers L
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2.
Change Section 1.1.2 to read as follows:
"1.1.2 Sections 1 through 7 present the plant operating limits (in-cluding original fuel) through the period of the full-ters license. In the order of presentation, Section 1.0 concerr.:
the introduction, Section 2.0 concerns the site, Sections 3 0 through 6.0 cover the plant and its systems, and Section 7.0 presents procedures for plant start-up and procedures for nor-tal and emergency operation of the plant. This section alsc includes administrative and procedural safeguards to the extent that these have a potential effect on nuclear safety. Section 8.0 describes Phase II of the Research and Development Progran, including parametric variations and operating limits on R&D fuel during Phase II.
Section 9 0 presents the pri=ary systen surveillance requirements."
- 3. Add the following at the end of the Table of Contents:
"9 0 Primary Syste: Surveillance................
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