ML20028B678
| ML20028B678 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 11/23/1982 |
| From: | Toner K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| References | |
| TASK-03-01, TASK-3-1, TASK-RR NUDOCS 8212030137 | |
| Download: ML20028B678 (58) | |
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Consumers Power Company General offices: 1945 West Pernali Road, Jackson, MI 49201 + ($17) 788 0550 November 23, 1982 Dennis M Crutchfield, Chief Operating Reactors Branch No 5 Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT - SEP TOPIC III-1, CLASSIFICATIONS OF STRUCTURES, COMPONENTS AND SYSTEMS (SEISMIC AND QUALITY)
The enclosed report provides Consumers Power Company evaluation of SEP Topic III-1, Classification of Structures, Components and Systems, for the Big Rock Point Plant. The report provides information on 8 areas of concern and it consists of the following appendices:
Appendix I
- Fracture Toughness Appendix II
- Full Radiography Requirements Appendix III
- Valves Appendix IV
- Pumps Appendix V
- Tanks Appendix VI
- Pipe Cyclic Loads on Piping / Fatigue Analysis Appendix VII
- Unidentified Codes Based upon the results of the evaluations of the syste.s and components considered, it is concluded that the Big Rock Point Plant meets current licensing criteria.
A W
Kerry A Toner Senior Licensing Engineer CC Administrator, Region III, USNRC NRC Resident Inspector-Big Rock Point Attachment oc1182-008%142 8212030137 821123 PDR ADOCK 05000155 P
Q s
INTRODUCTION SEP Topic III-1, Quality Group Classification of Components and Systems, was developed to ensure systems and components in the Big Rock Point Nuclear Plant were designed, fabricated, installed and tested to quality standards that reflect the importance of their safety functions. To assess this, the NRC contracted the Franklin Research Center to prepare a report (Reference 1) which addresses safety margins of systems and components in light of the changes that have taken place in the design and licensing criteria.
Upon NRC review of Reference 1, several items were noted which could not be dispositioned due to a lack of information. These items were then submitted (Reference 2) to Consumers Power for disposition.
This report contains those dispositions. In some cases, information was not available to disposition each item per the NRC guidelines. However, in each of these cases an evaluation was completed, or it was concluded, that the item did not pose a significant hazard to plant safety. Therefore this SEP Topic is considered adequately addressed and complete. The disposition and eval-uations are contained in each of the following appendices. Appendices cor-respond to the eight(8) items of concern listed in Reference 2, page 3 and 4 Appendix I - Fracture Toughness Appendix II - Full Radicgraphy Requirements Appendix III - Valves A'ppendix IV - Pumps Appendix V - Tanks Appendix VI - Pipe Cyclic Loads on Piping / Fatigue Analysis Appendix VII - Unidentified Codes REFERENCES 1 - Technical Evaluation Report; Quality Group Classifications of Systems and Components for Consumers Power Company Big Rock Point Plant dated April 9, 1982 by Franklin Research Center.
2 - Letter from DMCrutchfield, NRC, to DJVandeWalle, CPCo, dated April 16, 1982.
o RP1082-0019B-NS03
B 1
FRACTURE TOUGHNESS APPENDIX I CONSUMERS POWER COMPANY BIG ROCK POINT i
SEP TOPIC III-1, QUALITY GROUP CLASSIFICATIONS OF STRUCTURES, COMPONENTS AND SYSTEMS (SEISMIC AND QUALITY) i
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i RP1082-0015A-NS03 28 pays w
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ITEM Fracture Toughness Requirements CONCERN Current code requires that the fracture toughness of certain Class 1, 2 and 3 components be evaluated to ensure adequate margin between the lowest service temperature and the nil ductility transition temperature. For 66 of the 71 components reviewed by the Franklin Institute, there was insufficient infor-mation to determine if fracture toughness was an issue.
RESPONSE
Table 4-1 of the Franklin Report was reviewed to determine the 66 items of This list is included in Table I-1.
concern.
Design and construction information was gathered on the 66 items and compared to the established criteria exempting components from fracture toughness These criteria, detailed in Section 4.1.1 of Appendix A of the requirements.
Franklin Report are listed below:
Class 1
-Materials whose nominal thickness is 5/8 in. or less
-Bolts 1 in, or less
-Pipes, fittings, pumps and valves with nominal pipe size six (6) in. or less
-Austenitic stainless steels
-Non-ferrous materials
-Drop weight tests are not required for martensitic high alloy chromium (series 4xx) and precipitation hardened s sels a
Class 2
-Exemptions listed for Class 1 components
-Materials whose lowest service temperature exceeds 150 F are exempt from impact testing
-Commonly used plate, forging, and casting materials list in Table A4-2 of Reference I when used in components whose lowest service temperature (LST) exceeds the tabulated nil ductility transition i
temperature (TNDT) by at least the thickness-dependent value A, deter-mined from Figure A4-1 of Reference 1.
Class 3
_ Exemption listed for Class I components
-Materials listed in Table A4-3 in the thickness shown when the LST for the componant is at or above the tabulated temperature.
RP1082-0015A-NS03
e 3
-Materials for components for which the LST exceeds 100 F 4
are exempt from impact testing.
Table I-1 summarizes the results of the above comparison. As can be seen, many items were exempted, because of materials (18 itema), components thick-ness (26 items), components diameter (33 items) or other reasons (9 items).
This left 8 items remaining to be dispositioned.
- 1) Steam Drum
- 2) Steam Drum Nozzles (GT or ET 4 in. dia.)
- 3) Poison Tank
- 4) Reactor Coolant Pressure Boundary Isolation Valves
- 5) Containment Penetration Valves
- 6) Condensate Feedwater System
- 7) Main Steam System
- 8) Reactor Water Claan-Up System Evaluations were completed on these items as noted in Table I-1.
These evaluations are included in this appendix.
RP1082-0015A-NS03
4 TABLE I-l FRACTURE TOUGHNESS REQUIREMENTS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES I
2 3
4 RECIRCULATION SYSTEM
- 1) Steam Drum Acceptable per Evaluation FHSR 5.2.2 Shell Evaluation 1(attached) llead i
- 2) Steam Drum Acceptable per Evaluation FilSR 5.2.2 d
' Nozzles ( 4 in.)
Evaluation 2(attached)
- 3) Pumps ASTM A 351 Cr CF8M Technical Evaluation Report Body x
Science Applications Inc.
Cover x
FHSR 5.3.4 Impeller x
- 4) Valves I
24" Cate x
ASTM A351 Gr CF8M (Cast)
FHSR 5.4.2.2 20" Cate x
ASTM A351 Gr CF8M (Cast)
FHSR 5.4.2.2 j
20" Butterfly x
ASTM A351 Cr CF8 (Cast)
FHSR 5.4.2.4
- 1) Exempt due to austenitic stainless steel NOTE:
- 2) Exempt due to thickness less than or equal to.625 in.
LT = Less Than
- 3) Exempt due to diameter less than or equal to 6 in.
GT = Greater Than
- 4) Exempt due to other reasons, i.e. nonferrous material LT or ET = Less Than or Equal To i
RP1082-0015B-NS03
TABLE I-l FRACTURE TOUGHNESS REQUIREMENTS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4 LIQUID POISON SYSTEM.
- 5) Piping and Valve Beyond x
All Piping 3 in, diameter P&lD M-107, 121, Bechtel Spec M-53 Isolation Valves
- 6) Poison Tank Acceptable per Evaluation M-27-1-2 CPCo L-6022-BC-1 Nat'l Tank & Boiler Co Evaluation 3(attached)
- 7) Nitrogen Bottles x
Piping to Bottles =.5 in. dia.
P&ID M-107 CORE SPRAY SYSTEM
- 8) Pumps Core Spray Pump x
4 in. Lines In & Out of Pumps P&ID M-123 Spec M-14
- 9) Piping, Valves & Fittings in Containment Only Carbon Steel Pipe 6" in this P&ID M-123 Core Spray Test Loop x
test loop of 8 in, nominal size Class E-11 Sch 40 pipe. This pipe has a Spec M-53 322 in, wall thickness RP1082-0015B-NS03
TABLE I-l FRACTURE TOUGHNESS REQUIREMENTS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4
- 10) Piping, Valves & Fittings Outside Containment 4
Return Pipe to x
Only pipe GT 6 in. is this return P&ID M-123 i
Core Spray Pump line of 8 in. OD Sch 40. This Class E-Il pipe has a.322 in, wall thickness Spec M-53
- 11) Core Spray Sparger x
Austenitic Stainless Steel G.E. Report NEDC 21974, Appendix II, Rev 1
- 12) Suction Strainers x
Attached to 6 in. OD Pipe P&ID M-123 Heat Exchangers 13)
Primary Side x
14 CA Tube Material has minimum P.O. M-31 Southwestern wall thickness less than.625 in.
Engineering /M31-1-2 DM 65904 14)
Secondary Side x
Tube Sheet & Shell = A212B CM3182 Wall thickness =.375 in.
AM-6a5903 BACKUP CORE SPRAY SYSTEM
- 15) Piping, Fittings & Valves x
Piping is LT or ET 6 in. Dia. OD P&ID M-123 Inside Containment Class E-Il & C-6 Spec M-53 RP1082-0015B-NS03
i TABLE I-l FRACTURE TOUGHNESS REQUIREMENTS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4
- 16) Piping, Fittings &
x Piping is 6 in. Dia. OD P&ID M-123 1
Valves Outside Contaiment 8 in. Piping is Sch 40, Class E-Il & F-7 Nom. Wall =.322 in.
Spec M-53
- 17) Core Spray Nozzle x
ASTM A479 Cr 316 SS (Small Nozzles) x Elbows ASTM A182 Gr 0740-G30940 F304SS G.E. Dwg 761E293 x
Center Manifold ASTM A479 Gr 304SS ENCLOSURE SPRAY &
BACKUP SYSTEM Piping is 11 or ET 6 in. Dia. OD P&ID M-123
- 18) Piping, Fittings & Valves x
x Piping to Valves is 8 in. Sch 40 ASME Section II'i, NC2311(b)(5)
Nom. Wall =.322 in.
P.O. M-108 Sh 2 i
P.O. M-108 1
Spec M-53 & ANSI B36.10-1979
- 19) Spray Rings x
Seamless Carbon Steel Pipe.
P&ID M-123
)
P&ID M-123 Pipe. ASTM A-106 Gr B Spec M-53 4 In. OD Piping i
RP1082-0015B-NS03
-_=
1 TABLE I-l FRACTURE TOUGHNESS REQUIREMENTS i
i ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4 EMERGENCY COOLING SYSTEM Emergency Condenser
- 20) Shell Side x
A212 Gr A, 1/2 in. Plate P&ID M-107, M-107
- 21) Tube Side x
ASTM A213 PT 304 M7-1-7 Spec, M7-2-3 Spec, M7-5-10-1 Spec
- 22) Piping Fittings & Valves x
C.S. Pipe LT or ET 61n. Dia P&ID M-107 Class A-5 (Tube Side)
The only C.S. pipe is the 1
condenser 4
)
- 23) Piping, Fittings & Valves x
The only C.S. Pipe is the condenser P&ID M-107. Class A-5, Spec M-53 (Shell Side) vent which is 14 in. schedule 40 Nom Wall is.438 in.
REACTOR DEPRESSURIZATION SYSTEM
- 24) Relief Valves i
RV 5002 x
RV 5003 Piping to Valves 6 in. Diameter C.E. Dwg E230-104-5 RV 5000 j
RV 5046 RV 5045 RV 5001 I
- 25) Pipe Fittings & Valves x
All Piping LT 6 in. 0.D.
P&ID M-121 C.E. Dwg 230-104-5 to Relief Valves Spec 34490-1300-503 1
RP1082-0015B-NS03
TABLE I-l FRACTURE TOUGHNESS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4 FIRE PROTECTION SYSTEM
- 26) Discharge Lines to Steam x
10 in. Sch. 80 Pipe 740-B41021 Sh 2 Drum F.. closure Nom. Wall =.594 in.
M-53 Spec
- 27) Pumps - Electric x
Piping to Pumps is 8 in. Sch 40, P&ID M-123, Spec M-14 -16 & 53 Diesel x
Nom. Wall =.322 in.
ASME Section III, NC2311(b)(5)
Jockey x
Piping is 2.5 in. Diameter ANSI B16.10-1979 28) Piping,
- 28) Piping Fittings & Valves x
Carbon Steel, A53 Gr A P&ID M-123, Class F-7 Spec M-53 Nominal Wall =.322 in.
8 in. Sch 40 SAFETY RELIEF VALVES j
- 29) RV 5000 x
RV 5001 RV 5002 Maximum Pipe Size P&ID M-121. P.O. M-87 i
RV 5003 is 3 in. Diameter RV 5045 RV 5046 REACTOR COOLANT PRESSURIZED BOUNDARY
- 30) Piping from Reactor up to &
x Piping = A351, A358, A376, Spec M-53 P&ID M-107 & M-121 including First Isolation Type 316 SS Valve RP1082-0015B-NS03
TABLE I-l FRACTURE TOUGHNESS REQUIREMENTS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4 l
I REACTOR PRESSURE COOLANT BOUNDARY
- 31) Isolation Valves MO N001A x
ASTM A351 Gr CF8M FHSR 5.4.2 MO N001B ASTM A351 Gr CF8M FHSR 5.4.2 MO N003A ASTM A351 Gr CF8M FHSR 5.4.2 MO N003B ASTM A351 Gr CF8M FHSR 5.4.2 VFW 9 Acceptable per Evaluation Evaluation 4A(attached)
MO 7050 Acceptable per Evaluation Evaluation 4B(attached)
SV 4984 x
10 in. Pipe Sch 80 P&ID A-203 Nom. Wall =.594 in.
SV 4985 x
10 in. Pipe Sch 80 Spec 34490-1300-503 Nom. Wall =.594 in.
SV 4986 x
10 in. Pipe Sch 80 ANSI B36.10-1979 Nom. Wall =.594 in.
3V 4987 x
10 in. Pipe Sch 80 MO N006A ASTM A351 Gr CF8 FHSR 5.4.2 MO N006B ASTM A351 Gr CF8 FHSR 5.4.2 MO 7056 & MO 7058 Acceptable per Evaluation Evaluation 4C(attached)
Balance of Valves x
Piping LT or ET 6 in. Diameter
- 32) Containment Penetration SEP Topic VI-4 Valves CV 4094, 4095, 4096, 4097 Acceptable per Evaluation Evaluation SA(attached) l CV 4031, 4102 & 4025 x
LT 6 in. Pipe SEP Topic VI-4 RP1082-0015B-NS03
l TABLE I-1 FRACTURE TOUCHNESS REQUIREMENTS ITEM DESCRIPTION EXEMPTION COMMENTS /DISPOSTION REFERENCES 1
2 3
4
- 32) Containment Penetration Valves (continued)
CV4103 VEC-301 x
LT 6 in. Pipe SV 9155 & SV 9156 x
LT or ET A in. Pipe SEP Tr.pic VI-4 110 7050 Acceptable per Evaluation Refer to Evaluation 4B MO 7065 x
LT or ET 6 in. Pipe VFW 9 & VFW 304 Acceptable per Evaluation M-53 Spec, ANSI B36.10-1979 Evaluation 5B(attached)
VRD 310 & 311, CV 4117 x
LT or ET 6 in. Pipe P&ID M-106 CV 4091, 4092, 4093, 4027 x
LT or ET 6 in. Pipe
- 33) Control Rod Drive x
Nozzles = SA213TP304 G.E. Drawing i E-201-798-9 Housings Flange = SA182F304 M2-30-9 Backing Ring = Carbon Steel Wall Thick, ness
.5 in.
- 34) Control Rod Drive System x
All Piping LT or ET 6 in. Diameter P&ID M-122, M-121 M-106, M-107
- 35) Spent Fuel Pit Cooling x
Piping from Fuel Pit to Heat P&ID M-111, Class M-1, F-3, Systems Exchanger is 4 in. Dia. C.S.
Spec M-53 Piping LT 6 in. Dia.
- 36) Heat Exchangers x
Tube Wall =.049 in.
M-10-2-2 Shell Wall =.3125 M-10-5-3 Largest Pipe to or from Heat Exch.
P&lD M-Ill is 4 in.
RP1082-0015B-NS03
4 TABLE I-l FRACTURE TOUGHNESS REQUIREMENTS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4 CONDENSATE /FEEDWATER SYSTEM
- 37) Piping from Outer-Most Pipe in Containment = 10 in. Dia.
P&ID M-106 & M-121 Containment Isolation Valve Acceptable per Evaluation Evaluation 6.(attached) up to and including the Shut-Off Valve & Connecting x
Connecting Piping LT or ET 6 in.
P&iD H-106 & M-121 Piping up to and including Dia. Piping at Shutoff Valves the first Shutoff Valve (VFW2 & VFW6) is 6 in. Dia.
MAIN STEAM SYSTEM
- 38) Steam Line from Drum to M07050 Acceptable per Evaluation Evaluation 7A(attached)
- 39) Drain Line from Steam Line x
Pipe Size = 1.5 in.
P&ID M-121 l
to MOV 7065
- 40) Steam Line from MOV 7050 Acceptable per Evaluation P&ID M-121, M-106 to Stop Valve and Evaluation 7B(attached)
Connecting Piping to x
Bal, of pipe LT 6 in. Dia, first Shutoff Valve
- 41) Drain Line from Valve x
Pipe Size = 1.5 in.
P&ID M-106 110V 7065 to CV 4107 M-121
- 42) REACTOR WATER CLEANUP x
x A376 TP304 SS-from RV to P&ID M-107, 121 SYSTEM Regen. Ht. Ex.
Class C-3, A-2 All Carbon Steel (CS)
Spec M-53 Pipe 3 in. Dia. OD RP1082-0015B-NS03
TABLE I-1 FRACTURE TOUCHNESS REQUIREMENTS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4 REACTOR WATER CLEANUP SYSTEM (Continued) i j
- 43) Cleanup Demineralizer Acceptable per Evaluation P&ID M-107, M-28-4-2, M-28 Evaluation 8A(attached)
- 44) Non Regen Heat Exchanger (Tube Side) x Tube Material = Non Ferrous P&ID M-107 Southwestern Engineering 45)
(Shell Side) x 12 in. Pipe Sch 30 EM 65161 1
Min Wall.33 in.
EM 86365
- 46) Regenerative Heat Exchanger (Tube Side) x Nonferrous Tubes P&ID M-107 47)
(Shell Side)
Acceptable per Evaluation Southwestern Engineering i
~
EM 65149, EM-36924, DM-86366 Evaluation 8B(attached)
REACTOR SHUTDOWN s
COOLING SYSTEM J
- 48) Pumps x
Piping in and out is P&ID M-107, P.O. M-14 LE & ET 6 in. Dia. OD x
RP1082-0015B-NS03
's
TABLE I-l FRACTURE TOUCHNESS REQUIREMENTS l
ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4 REACTOR SHUTDOWN i
COOLING SYSTEM (continued) 49)
Heat Exchangers (Tube Side) x 18 Ga. Tubes 5/8 OD Spec 3159-H8 50)
(Shell Side) x A-53 Cr B P&ID M-107 16 in. OD Sch 20,.312 Nom. Wall P.O. M-8, Dwg. G40107 51)
Gland Coolers (Tube Side) x SA 304 SS P&ID M-107, M-4-16-2 52)
(She11' Side) x Carbon Steel Dwg. JS1259 3/16 in. thick shell*
H-14-16-2 53)
Pipings, Valves &
x A-106 Gr B. Pipe 2 in. Dia.
P&ID M-107 Fittings REACTOR COOLING WATER SYSTEM 54)
Pumps Piping to pumps is 8 in. Sch 40 P&ID M-Ill, P.O. M-14, M14-5-3 Nom. Wall =.322 in.
Heat Exchangers (Tube Sije) x SB 111 Tubes, #18 BWG 3/4 in. OD M6-1-5, Criscom Russel AA3994, M6-2-1
- Per phone conversation with Sales Engineer, Engine Cooler Co.
RP1082-0015B-NS03
TABLE I-l FRACTURE TOUCHNESS ITEM DESCRIPTION EXEMPTION COMMENTS / DISPOSITION REFERENCES 1
2 3
4 L
REACTOR COOL _ING WATER
. SYSTDi = (cortinued)
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- 56) Heat Exchangers
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x-SA 285 @.C P.O. M-6, M6-4-1
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Nom. Jhici!= 5/16 in....
A n
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8 in. Dia. :rsh',40,' Pipe.
P&ID M-ll!
- 57) Piping, Valver & Fittings x
x
~
A-53 Cr A Class F-3 a
Nom. Wall =.322 in.
Spec M-53:
All othfr. pipe,6 in.
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- 58) Reactor Cooling Water Tank
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- 59) Pip 9:g & Valves from Hotwell x
x x
to' Condensate Storage Tank Other~ pipe'LTfor ET 6 in.
Spec M-53, Cinys M-J. ; 4
,. E-8 in, pipe, Sch 30"=.322,19., wall n
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- 60) Service & Instrument Air,'f X' g 31 Piping LT 6 in. Dia..,
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ITEM DESC:tIPTION EXEMPTION COMMEhTS/DISPOSTION REFERENCES I
I 2
3 4
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HAKEUP_AND C,0NDENSATE DEMINERALIZER SYSTEM (continued)
- 63) Service Water System x
10 in. Pipe Sch 40 P&ID M-lll Nom. Wall =.365 in.
Spac M-53, Class F-10 F-2 i
REACTOR RECIRCULATION PUMP SEAL WATER SYSTEM
- 64) lleat Exchanger (Tube Side) x Wall =.065 in, P&ID M-Ill & M-237 65)
(Shell Side) x Wall Thickness =.375 in.
Byron Jackson IC-2172, Bechtel M2-420-2, Spec M-53
- 66) Piping & Valves x
x All Piping LT or ET 1.5 in.
P&ID M-237 Some Piping A376TP304 Spec M-53 1
ii l
RP1082-0015B-NS03 i
EVALUATION 1 STEAM DRUM The Steam Drum Shell and Heads are preoperationally hydrostatically tested with the Nuclear Steam Supply System at a pressure of (1.1)(Operating Pres-sure). Since the operation pressure of the primary system is 1350 psig, the hydrostatic test pressure is 1500 psia. The lowest service temperature (LST) of the steam drum is the calculated minimum metal temperature expected during normal operation whenever the pressure within the component exceeds 20% of the preoperational system hydrostatic test pressure. This pressure is calculated to be 300 psia at which the temperature would be 417F (Reference A).
The materal for both the steam drum shell and heads is SA212B which is identical to ASTM A212B. ASTM A212B has since been replaced by ASTM 516 Gr 70 for which the T is 0 F (Reference B; Table A4-2).
The temperature margin over TNDT' A,forIhe4.375inthickvesselisdeterminedtobe52F(ReferenceB; Figure ND A4-1).
By the method of evaluation for Class 1 components in Reference B:
LST - T
>A 417F-0JDJ52F 417 F>52 F Due to the LST exceeding T by 100 F or more, the fracture mechanics ap-proachofAppendixGinRefDT erence C is not required. This evaluation confirms the adequacy of the steam drum shell and head with respect to fracture tough-ness.
I AP1082-0019A-NS03
EVAIUATION 2 STEAM DRUM N0ZZLES ( 4 in.)
The applicable steam drum nozzles are made from SA105 Gr II and the thickest wall section is 3.875 inches. The nozzles receive hydrostatic testing with the Nuclear Steam Supply System and therefore possess a lowest service temp-is erature of 417F (Note Evaluation 1).
The temperature margin above TNDT calculated to be 48F (Figure A4-1, Reference B).
T is not readly available NDT for thi,s material but if the LST exceeds the reference nil ductility tran-by 100F then fracture mechanics is not required.
sition temperature (RTIn this case, the LST b )417F.
LST - T
= 100F or ET
=
n ca es that the T ias to be 317F or 417F - TNDT NDT f r this greater before fracture mechanics is required. This magnitude of TTherefore,themater rum material is highly unlikely.
nozzles is considered adequate.
RP1082-0019A-NS03
EVALUATION 3 POISON TANK The Poison Tank contains borated water for use in the shutting down of the reactor vessel. To maintain good solubility of the boric acid in the water, the tank is maintained at a temperature of 150F by 2 immersion heaters.
In the unlikely event that electrical power to the heaters if terminated, power is drawn from the emergency diesel generator. A back-up emergency diesel generator is also available for further support. It is reasonable to conclude therefore, that the LST of the poison tank is 150F. The tank is made from SA212B and has a all thickness of 2.5 inches. The T f this material, as NDT determined in Evaluatian 1, is 0F and the temperature margin over TTheevaluationofkke,A,is determined to be 30F (Reference B; Figure A4-1).
poison tank is as follows:
LST - TNDT> ^
140F - oF > 30F 140 F>30-F Since LST - T is greater than A, the poison tank satisfactorily meets the fracturetougbATuess criteria for Class 2 Components. This evaluation confirms the adequacy of the poison tank with respect to fracture toughness.
RP1082-0019A-NS03
EVALUATION 4A & SB FEEDWATER SYSTEM: VALVES VFW9 & VFW304 Both valves are constructed from A216 GR WCB material with a wall thickness 1.28 inches. VFW304 is closed during Nuclear Steam Supply System hydrostatic testing, while VFW 9 is left open. Feedwater supplied from the high pressure heater when the plant is operating at a 50MW electric is at a temperature of 333F. Since the system typically operates above this, it can be reasonably concluded that the LST is 333F. From Table A4-2 T is +30F (Reference B).
From Figure A4-1 (Reference B), the temperature margk above T is 30F. The NDT evaluation of these valves is as follows:
LST - TNDT >^
333F - 30F > 30F 303F > 30F Since LST exceeds T by 100F or more, the fracture mechanics of Appendix G.
NDT Reference C is not required.
It can be concluded that the material from which these valves are made is satisfactory with respect to fracture toughness.
t i
RP1082-0019A-NS03
EVALUATION 4B REACTOR PRESSURE COOLANT BOUNDARY - VALVE M07050 Valve M07050 is closed during preoperational hydrostatic testing of the Nuclear Steam Supply System. This valve therefore experiences a test pressure of 1500 psia as determined in Evaluation 1.
The lowest service temperature as defined in Reference B will be the lowest calculated metal temperature ex-pected during normal operation whenever the pressure within the component exceeds 20% of the preoperational system hydrostatic test pressure. This pressure is calculated to be 300 psia at which the temperature will be 417F (Reference A).
The valve housing is made of A216 Gr WCB for which the TNDT s
+30F (Reference B; Table A4-2) and the nominal wall thickness is 2.3125 inches. The corresponding temperature margin above TNDT, A, is 30F (Reference B. Figure A4-1).
By the method of evaluation for Class I components in Reference B:
LST - T
- ^
NDT 417F - 30F > 30F 387F> 30F Due to the LST being greater than T by 100F or more,- the fracture mechanics approachinAppendixG.ReferenceCkTs not required. This evaluation confirms the adequacy of Valve M07050 with respect to fracture toughness.
RP1082-0019A-NS03
,_e
-~-.n-,-
-,w,
,.7,.- -,-
4-m--
r----e,w
EVALUATION 4C REACTOR COOLANT PRESSURE BOUNDARY - VALVES M07045 & M07058 These two valves are identical and experience similar conditions of operation.
Both are made from A216 Gr WCB material with a wall thickness of 1.41 inches.
The valves are automatically closed when the system pressure reaches 300 psig which corresponds to a temperature greater than 420F. Since the valves are pressurized to 1500 psia during Nuclear Steam Supply System hydrostatic testing, a LST of 417F will be used (refer to Evaluation 1).
The temperature is 30F per Figure A4-1 of Reference B.
T is +30F per margin above T TableA4-2ofk![erence1. Theevaluationofthesevalvesk!asfollows:
LST - TNDT> ^
417F - 30F > 30F 387F> 30F Since LST exceeds T by 100F or more, the fracture mechanic's approach of NDT Appendix G. Reference C is not required. It can be concluded that the ma-terial of the valves M07056 & M07058 is satisfactory with respect to fracture toughness.
l RP1082-0019A-NS03
EVALUATION SA CONTAINMENT PENETRATION VALVES: CV4094, CV4096, CV4095 & CV4097 Valves CV4094 & CV4096 are made of cast iron ar.d valves CV4095 & CV4097 are made from cast steel. All four valves have a wall thickness of 1.05 inches.
The LST of these valves is 50F and from Figure A4-1 of Reference B, the temperature margin above T is 30F T
r ese materials is not readily NDT NDT available but would have tc exceed 20F to violate the fracture toughness for these Class 2 components. These valves are on containment ventilation lines that delivers and exits air in the containment. In addition, these valves serve to isolate the containment by checking air flow. Valves CV4096 & CV4097 are on a supply line while valves CV4094 5 CV4095 are on an exhaust line.
Both valves on each line are actuated sicultaneously by the sensing devices and/or systems. All valves are set to close at slightly less than atmospheric pressure to maintain containment integrity.
In addition all valves close in the event of any SCRAM regardless of pressure conditions. The conditions of pressure these valves are subject to is less than or equal to approximately 62 psia. Above'this pressure, the prime concern becomes the containment struc-i ture itself. This stress range is considerably lower than the rated working pressure of 260 psi @ 200F for valves CV4094 i CV4096 (Material Group 1.1; Reference D).
Some conservatism is built into this rating as it is unlikely that these valves will ever experience 200r since they are outside of con-tainment. At lower temperatures, the rating is higher,than that listed.
Valves CV4095 & CV4097 are subject to no more than *900 f t-lbs of operating torque which is well below the 2900 f t-lb maximum operating torque (Reference E).
It is concluded that these valves satisfactorily meet the fracture toughness requirements the ventilation system impcses upon them.
- Per telecon with manufacturer.
RP1082-0019A-NS03 w--.
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4-,&i-,
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+-, -
EVALUATION 6 FEEDWATER SYSTEM PIPING The piping of concern is 10 inches in diameter, schedule 120, A106 Gr B Carbon Steel with a wall thickness of.844 inches. From Figures A4-1 of Reference B, the temperature margin over T is 30F When the plan is operating at 50 MW electric,thefeedwatertemperaIurepass.ing through this piping is 333F.
ND Since the plant normally operates above this wattage, it is reasonable to conclude that 333F is the LST.
The T of the above material is not readily available. However, the T wouldkIvetobegreaterthan303Fbeforeimpacttestingwouldberequir!b LST - INDT "
F,TNDT "
333F - I
=
NDT As T
's in the magnitude of 303F or greater are highly improbable, Consumers Powerbompanyconcludesthat the feedwater piping material used in the con-ND struction of the Big Rock Nuclear Power Plant is adequate.
I i
RP1082-0019A-NS03 l
EVALUATION 7A PIPING FROM STEAM DRUM TO VALVE M07050 The piping of concern is 12 inch nominal diameter, A106 GR B with a nominal wall thickness of 1.0 inches. This piping is pressurized during preoperation hydrostatic testing of the Nuclear Steam Supply System (Refer to Evaluation 1).
Using the same conditions as in Evaluation 1, the lowest service temp-THe temperature margin above T is 30F erature of this piping is 417F.
frA106GrBisnoNDTreadily (Reference B; Figure A4-1).
While the TNDT available, to fail the fracture toughness criteria, this temperature would have to be 387F by the following analysis:
LST - TNDT> A T
-^
NDT NDT> 417 F - 30F > 387F T
w uld have to be In the same way, for Appendix G. Reference C analysis, TNDT As TNDT's in the magnitude of 317F or higher are highly unlikely, Consumers Power Company concludes that the main ~ aim piping material used in the construction of Big Rock Point Nuclear Plant is adequate.
i RP1082-0019A-NS03 i
EVALUATION 7B MAIN STEAM SYSTEM: PIPING AFTER M07050 TO FIRST SHUT 0FF VALVE The piping under evaluation is 12 inch diameter, Schedule 120, A106 Gr B carbon steel with a nominal wall thickness of 1.0 inches. Motor operated valve M07050 remains closed until a pressure of 500 psig is obtained, after-which it opens to supply the steam turbine. The temperature of the steam at this pressure is greater than 465F (Reference A).
The pipe in consideration therefore, is exposed to steam temperatures between 465F and 580F (operating temperature). The LST is therefore set at 465F, the temperature margin, A, over T is 30F as per Figures A4-1, Reference B.
T is not readily availabeforthismaterialbutiftheLSTexceedstheheferencenilductility transition temperature (RT
) by 100F then fracture mechanics is not re-quired.
Inthiscase,thek[ Tis 465F,therefore:
LST - T
> 100F or ET 465F - T
>100F ET indicates that the T has to be 365F or greater before fracture mechanics is required. ThismagnbudeofT f r this material is highly unlikely.
r ET Therefore, the material of the main steam system piping is considered ade-quate.
RP1082-0019A-NS03 l
o EVALUATION BA DEMINERALIZER TANK The demineralizer tank is 1.875 inches thick and is made of SA212B material.
It has been established that T for this material is 0F (Evaluation 1) and the temperature margin above T
, A, is 30F (Reference C; Figure 4A-1).
The N
demineralizer is fed from the nonregenerative heat exchanger by the clean-up system pump. Water exiting the nonregenerative heat exchanger is at 100F during normal operation. A malfunction on the part of the nonregenerative It is heat exchanger would only serve to produce higher temperature water.
therefore reasonable to conclude the LST of the demineralizer tank to be 100F.
Using the method of evaluation for Class 1 components given in Reference B, the demineralizer tank evaluation is as follows:
LST - TNDT > A 110F - 0F >30F 110F > 30F r m re, the fracture mechanics approach Due to the LST exceeding T y
ofAppendixG,ReferenceCkTs not required. From this evaluation, it can be N
concluded that the fracture toughness of the demineralizer tank material is adequate.
EVALUATION 8B REGENERATIVE HEAT EXCHANGER SHELL is The shell of this heat exchanger is made from SA-106 Gr B for which T Thewallthicknessif1.0inchesforwhichthklemp-not readily available.
The LST of the shell is 100F. The eraturemarginaboveT{DJ,,A,is30F.
evaluation of the shel
,, g,77,,,
LST - T
>A 110F - T NDT T
0F
(
NDT >
While T f greater than or equal to 80F for this material is unlikely, NDT i
fracture mechanics of Appendix G, Reference C is required for this Class 1 However, insufficient data is available to use the analysis of component.
Since this T is unlikely and due to the satis-Appendix G, Reference C.
factoryperformanceofthiscomponentkkmorethan20yearsofoperation, Consumers Power Company feels this material is adequate with respect to fracture toughness.
RP1082-0019A-NS03 1
APPENDIX I REFERENCE TABLE Reference A:
1967 A.S.M.E. Steam Tables Reference B:
Quality Group Classification of Components and Systems (SEP III-1), Consumers Power Company, prepared by Franklin Research Center, Philadelphia, PA.
Reference C:
A.S.M.E.Section III, Subsection NA, " Appendices",
1980 Reference D:
ANSI B16.34 - 1981, " Valves-Flanged and Buttvelding End" Reference E:
ANSI /AWWA C504-80, "AWWA Standard for Rubber Seated Butterfly Valves" l
RP1082-0019A-NS03
1 RADIOGRAPHY REQUIREMENTS APPENDIX II CONSUMERS POWER COMPANY BIG ROCK POINT SEP TOPIC III-1, QUALITY GROUP CLASSIFICATIONS OF COMPONENTS AND SYSTEMS
)
RP1082-0007A-NS03 8 pages
2
-ITEM Radiography requirements of:
Class 1, 2 and 3 Vessels Class 1, 2 and 3 Piping and Valves Class 1 and 2 Pumps Safety Relief Valves CONCERN Radiography Requirements - The licensee should provide the following infor-mation:
a)
Radiography requirements imposed on Class 2 and 3 vessels for which Code Case 1273N was not invoked and with welded joint thicknesses less than 1-1/2 inches.
b)
Code edition and Class for components designed to ASME Section III.
c)
Radiography requirements for piping and valves designed only to ASA B31.1 (1955).
d)
Radiography requirements for all Class 1 and 2 pumps.
e)
Confirm that radiography requirements were inforced for safety relief valves designed to ASME Section I.
f)
Radiography requirements for Class 1 vessels for which Code Case 1270N or 1273N were not invoked.
RESPONSE
Consumers Power Company conducted an intensive research program to locate information on radiography requirements for the above mentioned items. Item b) above will be addressed in Appendix VIII of this report.
Sources of information included:
Bechtel Piping Specifications The Big Rock Point Final Hazards Summary Report The Big Rock Point Technical Specification The Big Rock Point Equipment Manual The Bechtel Correspondence Files The Big Rock Point Vendor Drawing Files The Big Rock Point Purchase Requisition Files The Big Rock Point Equipment Database The Big Rock Point 40-Year ISI Master Plan From these sources of information the following items were concluded:
RP1082-0007A-NS03
3 VESSELS No radiography information was located on the Shutdown Gland Coolers, RCP Seal Water System Heat Exchanger, LPS Nitrogen Bottles, Spent Fuel Pit Cooling System Heat Exchanger.
The Gland Coolers, RCP Seal Water Heat Exchangers and Nitrogen Bottles are addressed in the Pressure Test Program in Section XI.
The SFP Cooling System Heat Exchanger is non-classed and therefore not con-sidered safety related.
The following radiography information was located on the remainder of the vessels:
Regenerative Heat Exchanger - This vessel was constructed and subse-quently installed as a replacement in 1972 and at that time 100%
radiography was employed on the new vessel.
Non-Regenerative Heat Exchanger - This vessel was also constructed and subsequently installed as a replacement in 1972 and 100% radiography was performed on the vessel at that time.
Emergency Condenser - Radiography was performed on the shell and nozzle welds. Inspection reports indicate that 125 radiographs were reviewed.
Code Cases 1270N and 1272N were employed.
Cooling Water Heat Exchanger - Inspection reports indicate that four radiographs were performed.
Core Spray Heat Exchanger - Radiography was performed on the channel weld. Code Case 1272N was implemented on the tube side of the vessel.
Peactor Shutdown Cooling Heat Exchanger - This vessel implemented Code Case 1273N.
PIPING Radiography information was not located for all piping systems built to B31.1.
However, the Inservice Inspection Program for Big Rock Point covers Class 1, 2 and 3 piping systems.
Primary Coolant Loop - Construction was to ASME Section I and the following radiography was performed:
1.
100% radiography was performed on the seam welds of the risers.
2.
All shop and field circumferential welds were fully radio-graphed and penetrant tested.
RP1082-0007A-NS03
4 Reactor Depressurization System - This system was designed and con-structed to ASME Section III 1971 Edition. Radiography requirements are listed below:
1.
Longitudinal Welds - 100% Radiography 2.
Butt Welds - 100% Radiography 3.
Branch Connections More Than 4" - 100% Radiography Bechtel Piping Specification Requirements - (Relavent to B31.1 Piping Only)
Spec #3159-M-113 Shop Fabricated Piping: All butt welds fabricated in accordance with ASME Section I will be 100% radiographed.
Spec #3159-M-119, Field Fabricated Piping: Welds fabricated in accor-dance with ASME Section I will be 100% radiographed.
VALVES No information was locatad for valves that were built to B31.1 or the safety relief valves. However, valve bodies are covered by the ISI program. Also, safety related valves are included in the Big Rock Point Pump and Valve Program.
The following radiography information was found for valves:
Primary Coolant System - All valve castings were fully radiographed after heat treatment to comply with ASTM Standards for Class 2 cast-ings.
Redundant Core Spray - The valves were built and inspected to ASME Section I, 1970 Edition.
Reactor Depressurization System - The RDS valves were designed, fabri-cated and inspected to ASME Section III Class 1, 1971 Edition.
It appears that the valves were also inspected using Specification
- 34490-1300-502.
l RP1082-0007A-NS03
5 Specification #34490-1300-502 for Class 1, 2 and 3 nuclear valves requirements are listed below:
NON-DESTRUCTIVE EKAMINATION REQUIRMENTS - Specification f34490-1300-502 RADIOGRAPHIC EXAMINATION CAST STAINLESS STEEL Class 1 Class 2 Class 3 Yes Yes Not (2)
(2)
Req'd FORGED STAINLESS STEEL OR BAR STOCK Class 1 Class 2 Class 3 Not Not Not Req'd Req'd Req'd CAST CARBON STEEL Class 1 Class 2 Class 3 Yes Yes Not (2)
(2)
Req'd PRESSURE RETAINING WELDS Class 1 Class 2 Class 3 Yes Yes Yes l
(6)
(6)
(9) l i
NOTES:
1 - Ultrasonic Examinatien may be used in place of or in combination with Radiographic Exacination except where the grain or configuration is such that meaningful results cannot be obtained by Ultrasonic Methods.
2 - Not required for fillet and socket welds attachment welds, seal wolds and hard surfacing welds.
+
RP1082-0007A-NS03
6 PUMPS No information was found for the Core Spray System, Fire Protection or the Shutdown Cooling System pumps.
These pumps'are not included in the Inservice Inspection Program but all pumps that are considered safety-related are included in the Pump Inservice Inspection Program.
Reactor Circulating Pumps - The pump case is 100% radiographed.
QUALITY CONTROL The following paragraphs were taken from a Bechtel quality control statement issued by Bechtel prior to construction:
1.
Equipment and material (except some standard, off-the-shelf items) are inspected in the shop by Bechtel inspection. They examine mill test reports, inspect items during manufacture for compliance with specifications, workmanship, defects, dimensional tolerances, etc.
2.
Piping and equipment falling under the ASME Code are inspected during shop fabrication by a National Board inspector. The vendor stamps completed vessels and furnishes data reports to Bechtel. For shop fabricated items which will be assembled in the field, partial data reports are furnished.
3.
Field assembly of ASME Code equipment and piping is inspected by a National Board inspector. Partial data reports are completed. Either hydro or pneumatic tests are made as appropriate.
4.
Interconnecting piping between the reactor and steam drum for this plant has been ruled to be non-ASME Code. It will be produced and f abricated to the ASA Code for pressure piping, but inspection will be performed by a National Board inspector and data sheets furnished comparable to those required for ASME. Field erection, welding, radiographing, and hydro-testing will be carried out just as for ASME piping, and will be Natior.al Board inspected. We expect the latter to be done by Hartford, but arrangements have not yet been made.
BIG ROCK POINT ISI REQUIREMENTS Big Rock Point Inservice Inspection Program, Table 11-1 outlines the inservice inspection program implemented at Big Rock Point. This program meets the requirements set forth in Section XI of the ASME Code 77S78. The table is set up to show the number of volumetric and surface examintions scheduled to be performed over a 40-Year period.
Big Rock is presently at the end of the first 10-Year interval of the 40-Year Master Plan. By the middle of 1983, approximately 25% of planned inspections will be complete.
RP1082-0007A-NS03
f TABLE 11-1 Total Total Welds Scheduled for Total Welds Scheduled for Number Volumetric Testing I.iquid Penetrant Testing of Welds During 40-Year Plan During the 40-Year Plan Class 1 Reactor Pressure Vessel 126 47 9
Reactor Pressure Vessel Closure Head 31 12 1
Emergency Condenser 16 12 0
Clean-Up Regenerative Heat Exchanger 56 31 2
Clean-Up Nonregenerative Heat Exchanger 10 6
4 Steam Drum 69 37 11 Clean-Up Demineralizer Tank 11 10 0
Emergency Condenser System 72 51 16 Liquid Poison System 93 2
30 i
Shutdown System 55 20 20 Reactor Clean-Up System 294 3
164 Main Steam System 441 46 152 i
Main Recirculation System 195 84 72 Core Spray System 13 5
5 Redundant Core Spray 38 29 29 Control Rod Drive System 99 0
40 Feedwater System 21 6
10 Reactor Depressurization 118 8
35 Reactor Clean-Up Pump 15 11 11 Main Recirculation Pump 6
0 0
Valve Bodies 52 2
0 Valve Bolting 76 6
0 NOTE: Above totals include branch connectors and exclude lugs and supports.
RP1082-0007A-NS03 j
TABLE 11-1 (continued)
Total Total Welds Scheduled for Total Welds Scheduled for Number Volumetric Testing Liquid Penetrant Testing of Welds During 40-Year Plan During 4d-Year Plan Class 2 Control Rod Drive 19 18 18 Core Spray System 53 0
45 Feedwater System 88 65 68 Main Steam System 27 21 0
Post-Incident Piping Cooling 36 0
32 Shutdown Cooling System 83 0
73 Valve Bodies 31 0
30 Scram Dump Tank 4
4 1
Class 3 Service Water System 19 0
0 Reactor Cooling Water 13 0
0 Reactor Shutdown System 35 0
0 Service Water System 5
0 0
NOTES:
1.
Welds that rquire both volumetric and surface examination have been counted only once under the volu-metric category.
2.
These numbers represent close approximations.
3.
Total welds = total welds in the system.
4.
Volumetric = exams perform using UT.
5.
Surface = exams performed using PT.
6.
Several welds in each system have been excluded or exempted from examination due to inaccessability, size, etc.
RP1082-0007A-IIS03
r
{
1 l
l 4
i 1
VALVES APPENDIX III CONSUMERS POWER COMPANY BIG ROCK POINT PLANT SEP III-1, QUALITY GROUP CLASSIFICATIONS OF COMPONENTS AND SYSTEMS f
i I
l i
l RP1082-0002A-NS03 7 pages
1 APPENDIX III SYSTEMATIC EVALUATION SAFETY TOPIC 111-1 QUALITY GROUP CLASSIFICATION OF COMPONENTS AND SYSTEMS ITEM Valves CONCERN Provide on a sample basis for Class 1, 2 and 3 valves, information regarding the design of the valve in order to evaluate if they meet current body shape and pressure-temperature rating requirements.
RESPONSE
Of approximately 3939 valves in the plant, 111 (46 Class 1, 46 Class 2, 1 Class 3, 3 Class 4, 15 National Fire Protection Code-Class F) are listed in Procedure TV-30, "ASME Boiler and Pressure Vessel Code Section XI, IWV and IWP Testing Program." These valves were selected in accordance with the testing requirements of NRC correspondence dated January 16, 1978. All other valves are censidered for operating convenience or maintenance and are therefore not included in the Inservice Inspection (ISI) Program or the scope of this item.
Consumers Power Company intends to perform the following evaluation on a sample basis as requested by the NRC:
Class 1 Valves - Compare actual body shape with body shape rules of Section NB-3544, ASME Code Section III, Division 1, 1977 through 1978 Summer Addenda.
If the shapes are not significantly dif-ferent, they are considered adequate.
Class 2 Valves - Identify pressure-temperature rating requirements of Class 2 and 3 valves and compare them to the pressure-temperature ratings of ANSI B 16.34, 1977, " Steel Valves."
To develop the sample for evaluation, the valves were grouped according to valve type, class (either 1 or 2/3), and manufacturer (See Table III-1).
Data sheets and drawings of the valves were obtained and reviewed with the following results:
1 Class 1 Valves - There is insufficient information to perform the body shape j
evaluation for Class 1 valves. Nevertheless Consumers Power RP1082-0002A-NS03
- ~ -... -,
r 2
Company considers that the Class 1 valves should be adequate for the following reasons
- 1) The plant has been in service since 1962 and in this 20-year period, corrective actions have been taken to resolve valve problems which have occurred.
- 2) The valves are included in the ISI program which requires periodic operability testing and examination of the pressure boundary integrity.
Class 2 and 3 The results of the evaluation of these valves are Valves summarized in Table III-2(a). Table III-2(b) is a supplemental table showing results for a similar evaluation of Class 1 valves.
It is used for illustrative purposes only and is not a requirement for this report. There is sufficient information to evaluate the pressure-temperature ratings of 4 of the 12 groups of Class 2/3 valves. Three groups of Class 2/3 valves lended themselves to comparison with Class 1 valves that passed an evaluation as shown in Table III-2(b). The final result is that 7 of the 12 groups representing 22 of the 47 Class 2/3 valves were evaluated and determined acceptable. The tcaciaing valves could not be evaluated but are considered adequate for the same reasons given for Class 1 valves above.
RP1082-0002A-NS03
TABLE III-1 i
VALVE CROUPINGS i
1 2
3 4
5 6
7 8
9 10 11 CIDBE BALL CHECK CHECK Btr1TERFLY EXPIDSIVE CATE Black, Sivalls in Bryson Worcester Allis Chalmers Edward Allis Chalmers Conax Powell Class 1 Class 2 Class 1 Class 2 Class 1 Class 1 Class 1 Class 1 Class 2 Class 1 Class 2 CV4020*
CV4025 CV4092*
CV4091*
CV4094*
VW304*
CV4095*
CV4121*
CV4118* M07050*
M07065*
1 CV4050*
CV4027*
CV4093 CV4096 VW9(1)
CV4097 CV4122 CV4119 M07051 M07068 CV4117 CV4031 CV4120 M07052 i
CV4123 M07053 CV4049 CV4124 M07056 CV4102 M07057 CV4103 M07058 CV4105 M07059 M07061*
M07062 M07063
- Valves chosen for evaluation RP1082-0002A-NS03
TABLE III-1 VALVE CROUPINGS 1
2 3
4 5
6 7
8 9
10 11 GIDBE BALL CHECK CHECK BUITERFLY EXPIDSIVE GATE Black Sivalls & Bryson Worcester Allis Chalmers Edward Allis Chalmers Conax Powell Class 1 Class 2 Class 1 Class 2 Class 1 Class 1 Class 1 Class 1 Class 2 Class 1 Class 2 CV4020*
CV4025 CV4092*
CV4091*
CV4094*
Vfw304*
CV4095*
CV4121*
CV4118* M07050*
M07065*
CV4050*
CV4027*
CV4093 CV4096 VFW9(1)
CV4097 CV4122 CV4119 M07051 M07068 CV4117 CV4031 CV4120 M07052 CV4123 M07053 CV4049 CV4124 M07056 CV4102 M07057 CV4103 M07058 CV4105 M07059 MO7061*
M07062 MO7063
- Valves chosen for evaluation RP1082-0002A-NS03
e e'
TABl.E III-1 (continued)
VALVE CROUPINGS 12 13 14 15 16 17 18 19 20 21 22 RELIEF RELIEF CATE CATE CATE CATE GATE Crosby Edward Velan Crane Anchor Class 1 Class 2 Class 4(3) Class 2 Class 1 Class 2 Class 3 Class 1 Class 2 Class 1 Class 2 RV5000* RV5018* RV5043 RV5083* M07072* M07064* N07066* CV7071* M07070* CV4180 VPIl RV5001 RV5019 RV5062 RV5084*
CV4181 VPI2 RV5002 RV5049 RV5063 CV4182 VPI3 RV5003 RV5050 CV4183 VPI4 RV5045 RV5051 CV4184 VPI6 RV5046 VPI7 VPI8 VPI9 RP1082-0002A-NS03
TABLE 111-2a RESULTS OF PRESSURE-TEMPERATURE RATING EVALUATION OF CLASS 2/3 VALVES (2)
Category Valve ASTM Material Material Max. Temp Max. Pressure Rating Allowable Remarks Specification Group No.
F psig Pressure 2
CV4027 216 WCB 1.1 60 24 600#
1480 4
CV4091 A 105 1.1 140 1450 1500#
3590
-^
2200 5000#
-999J' /'" See Note 1 9
CV4118 351(18-8 S.S.)
2.1 145
/
11 M07065 650 1700 See Note 4
, 13 RV5018 Carbon Steel 1.1 425 300,,
e Insufficient Information
+-
.s 15,
RV5083 Cd Gen Steel 1.1 Insufficient Iaiornation
.s 17 M07064 216,WCB 1.1 150#
e w See Note S '
., /
18 M07066 A 1h
- ' ~
See Note 5
/
p -
s m.
/
~
20 M07070.216 WCB 1.1 582 1350
'500#
27M -
See Note 3
,. A
+
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-b'
/w m*
v,r '
w,
~ * -
_ y-
/_
V y'
s m
~
e n
r.
a e
~
w
\\~s n
w
?
x
^'
eu
/
,^
j's. -
s N 'Ns
~
s-x e
m
_, o y
-y.,
%. ^ ^
" N
^
RP1082-0002A-NS03
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a
f as
/
I
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a s-TABLE Ill-2b
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SUPPLEMENTAL INFORMATION
'/
PRESSURE-TEMPERATURE RATING EVALUATION
's-
/
~..
1 0F CLASS 1 VALVES f
i e
./
Category Valve ASTM Material Material Max. Temp Max. Pressure Rating Allowable Remarks Specification Group No.
F psig Pressure psig 1
CV4920 216 WCB 1.1 160 2200 15008 3507 3
CV4092 A 105 1.1 140 1450 15008 3590 6
VFW304 216 WCB 1.1 375 2000 1500*
3198 i
8 CV4121 351 (18-8 S.S.)
2.1 600 1700 5000#
6230 i
10 M07050 216 WCB 1.1 650 1700 1500#
2685 See Notes 3,4 M07061 216 WC3 1.1 582 1350 15004 2782 See Notes 3,5 19 M07071 216 WCB 1.1 582 1350 1500#
2782 See Note 3 NOTES:
1.
This valve has a rating of 5000#, however, 4500# was used in the evaluation because it is the highest rating given in ANSI B16.34-1977.
The resulting allowable pressure is therefore a conservative value and acceptable for the evaluation.
2.
Unless otherwise specified, all data was obtained from valve data sheets and drawings. Material Groups and Allowable Pressures obtained from ANSI B16.34-1977.
3Property "ANSI code" (as page type) with input value "ANSI B16.34-1977.</br></br>3" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process..
Data obtained from valve specification.
4 M07065 has the same operating conditions, manufacturer, and is in the same system as M07050 - a Class 1 valvee Since M07050 passed the evaluation it is believed that M07065 would also if sufficient information was available.
5.
M07064 and M07066 are in the same line as M07061 - a Class 1 valve. Since these 3 valves would have similar design criteria and M07061 passed evaluation, it is assumed that M07064 and M07066 would pass also if sufficient information were available.
E RP1082-0002A-NS03
1 PUMPS APPENDIX IV CONSUMERS POWER COMPANY BIG ROCK POINT SEP TOPIC III-1, QUALITY GROUP CLASSIFICATIONS OF COMPONENTS AND SYSTEMS I
l t
I i
l i
L RP1082-0008A-NS03 3 pages i
r o
2 ITEM Pumps involved in the following systems: Core Spray, Fire Protection, Reactor Shutdown Cooling and Reactor Cooling Water.
CONCERN Provide the codes, standards, or manufacturers specifications used for de-signing these pumps.
EVALUATION An extensive search of the Big Rock Point FHSR, Pump and Valve Program, engineering drawings, vendor drawings and correspondence records was initiated in an attempt to provide the necessary information. Table IV-1 provides the codes of standards or manufacturers specifications used in designing these Further pumps, as well as the materials of construction of the bowl or case.
information on the pumps in each system follows.
CORE SPRAY SYSTEM PUMPS These two pumps were furnished in accordance with Bechtel Specification M-14 and were built in strict accordance with ASME Code Case 1272. They are regularly inspected using approved procedures of the Big Rock Point Pump and Valve Program per Section IWP of the ASME Section XI Code (1977 Edition thru Summer 1978 Addenda).
MRE PROTECTION SYSTEM PUMPS This system utilizes three pumps. The jockey fire pump and the electric motor driven fire pump were furnished in accordance with Bechtel Specification M-14.
The diesel engine driven fire pump was furnished in accordance with Bechtel Specification M-16, which references the National Fire Protection Agency for design approval. Both the electric motor and diesel engine-driven fire pumps The are inspected regularly per the Big Rock Point Pump and Valve Program.
jockey fire pump is not safety-related and is not included in the program.
REACTOR SHUTDOWN COOLING AND REACTOR COOLING WATER SYSTEMS These four pumps (two in each system) were furnished in accordance with Bechtel Specification M-14.
These pumps are not considered safety related and are not included in the Big Rock Point Pump and Valve Program. The Reactor Shutdown Cooling System cools the NSSS after the NSSS temperature cools below 425 F and the pressure is less than 3000 psig after shutdown. The Reactor Cooling Water System cools components in the Containment Building during plant operation.
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TABLE IV-1 PUMPS CODE. STANDARD PUMP OR MANUFACTURERS SPEC MATERIAL Core Spray System fASMECodeCase1272N-5 Cast Iron Diesel Fire Protection Design Approved by NFPA, Cast Iron Characteristics Specified In NFPA Pamphlet #20 Jockey Fire Protection Bechtel Spec M-14 Cast Iron Electric Fire Protection Bechtel Spec M-14 Cast Iron RSCW System Bechtel Spec M-14 4-6% Cr Alloy Steel RCW System Bechtel Spec M-14 Cast Iron Correspondence of 12/21/61 from Bechtel to Consumers Power Bechtel Spec M-16 f
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1 TANKS APPENDIX V CONSUMERS POWER COMPANY BIG ROCK POINT-SEP TOPIC III-1, QUALITY GROUP CLASSIFICATIONS OF COMPONENTS AND SYSTEMS
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2 ITEM Tanks involved in the following systems: Liquid Poison, Service and Instru-ment Air, and Reactor Cooling Water.
CONCERN a)
Confirm that the atmospheric storage tanks meet current compressive stress requirements, b)
Confirm that the 0 to 15 PSIG storage tanks meet current tensile allow-ables for biaxial stress field conditions, c)
Specifications for tanks built to codes other than ASME Section VIII (1959).
d)
Information on the standards used in the design of the reactor cooling water tank and nitrogen bottles.
EVALUATION An extensive search of the Big Rock Point FHSR, engineering drawings, tech-nical specifications, standard operating procedures, vendor drawings and correspondence records was initiated in an attempt to provide the necessary information. Table V-1 summarizes the information found. Further information on each of these systems follows.
LIQUID POISON SYSTEM The liquid poison tank was designed and built in accordance with the ASME Section VIII code. It is normally maintained at a pressure of 0 to 50 psig and a temperature of 150 F.
However, the opt sting conditions (which this tank was designed for) are 2000 psig and 650 1 The nitrogen bottles are standard commercial gas bottles. Their pressur, is checked three times daily (once per shift) and there is also a low pressure alarm to the control room set at 1945 psig.
SERVICE AND INSTRUMENT AIR SYSTEM The tanks in question were the piping compressor tanks, of which the Big Rock Point Plant has none. The system operates at 105 psig and does have air receivers and air dryers which were designed and built to the ASME Code for Unfired Pressure Vessels.
REACTOR COOLING WATER SYSTEM The reactor cooling water tank is a cement structure attached to the bottom part of the containment structure. The walls are sealed with a waterproof sealer.
It stores water for the Reactor Cooling Water System that cools certain components in the Containment Building during plant operation.
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TABLE V-1 TANKS CODE, STANDARD OR MANUFACTURERS PUMP SPEC MATERIAL THICKNESS Liquid Poison ASME B&PV Code A-212B 2 1/4" latest Ed. as of 4/6/61 RCW N/A Cement Sides & Bottom N/A Boiler Plate Top Piping Compressor N/A N/A N/A Nitrogen Bottles N/A N/A N/A IBechtel Spec M-27 NOTE: N/A Not Available RP1082-0012A-NS03
i CYCLIC LOADS ON PIPING FATIGUE ANALYSIS ON CLASS 1 VESSELS APPENDIX VI CONSUMERS POWER COMPANY BIG ROCK POINT SEP TOPIC III-1, QUALITY GROUP CLASSIFICATIONS OF STRUCTURE, COMPONENTS AND SYSTEMS l
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ITEM Cyclic loading on Class 1 piping and fatigue analysis requirements for Class 1 vessels.
CONCERN Calculations should be provided to assess the impact on the usage factor of gross discontinuities in Class 1 Piping Systems for a medium and large number of cyclic loads.
Demonstrate compliance with current fatigue analysis requirements for all Class 1 Vessels.
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EVALUATION Consumers Power Company cannot provide the evaluations which the staff wants in the near future. With regard to both piping and vessels, CPCo does not have the kind of good load information that the Franklin Research Center believes we do.
However, the following points are offered.
CLASS 1 PIPING CPCo understands the code re'quirements. The Franklin Report notes in great detail the code sections of original and present use for Big Rock Point and new plants respectively. This is a fine condensing of code requirements but does not provide the information CPCo needs, namely loads. Any General Electric or Bechtel stress analysis of any piping is out of date or non-existent, with the exception of the RDS.
In the past few years some pipe stress work has been performed on subsystems at Big Rock Point. This work has been performed to assist in modifications.
Results to date have demonstrated that although certain design and analysis documentation is not available. The piping consistently meets code require-ments for thermal expansion and deadweight. The cyclic loads of concern are unclear, the only loads that can be reasonably specified are to thermal expansion.
Class 1 Piping is within the scope of the ISI Program. The program is nearing I
l the end of the first 10-Year interval with very few defects noted. The plant has been operating for about 20-Year with good operating history.
Despite the idea that SEP Topic III-6 is supposed to contain the seismic issue and that the seismic issue is excluded from III-1, it is the SEP Topic III-6 which will generate the analysis which are required to generate the loads required to conduct any fatigue analysis. CPCo is, in fact, doing a great deal of pipe work for SEP Topic III-6, which is the seismic issue.
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.. a required to produce the stresses by which fatigue resistance is determined.
Also, the mechanical loads from piping are not well defined and will not be until the seismic program is complete. As Big Rock Point does have a good operating history temperaures and stresses on vessel nozzles will not be generated.
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UNIDENTIFIED CODES APPENDIX VII CONSUMERS POWER COMPANY BIG ROCK POINT SEP TOPIC III-1, QUALITY GROUP CLASSIFICATIONS OF COMPONENTS AND SYSTEMS T
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_ ITEM Unidentified codes for certain components.
CONCERN Some components listed in Tables 4-1 and 4-2 of the Franklin Report lacked information on the design codes of the components.
Those components are as follows:
Core Spray Sparger Suction Strainers Emergency Condenser Valves (Tube Side)
Emergency Cooling System Piping (ECS - Tube Side)
Reactor Depressurization System Relief Valves Reactor Depressurization System Piping & Valves Reactor Uater Clean-up Non-Regen Hx (Shell Side)
Reactor Water Clean-up Regen Hx (Tube & Shell Side)
Reactor Shutdown Cooling System (Gland Coolers)
Reactor Cooling Water Heat Exchanger Reactor Cooling Water Tank Standby Diesel Generator System Piping Standby Diesel Generator System Valves
RESPONSE
Tables 4-1 and 4-2 of the Franklin Report list design codes for Class 1,
2, 3 components. That information was lacking for the above components. A docu-ment search was conducted in an attempt to define the missing code informa-tion. Purchase specifications, drawing files, correspondence files and the FHSR were consulted with the following results:
Core Spray Sparger - The Core Spray Sparger was replaced in 1979. The a.
replacement core spray sparger was designed and fabricated in accordance with ASA B31.1-1955. However, it is stated in Appendix II to General Electric Report NEDO-21974 that ASA B31.1-1955 does not distinguish between types of loads (primary / secondary) or types of stress (bending /memb rane). Therefore, ASME Section III Subsection NG (Core Support Structure) will be used to catagorize load and stress. Year and addenda were not specified for ASME Section III in that report.
b.
Suction Strainers - No design codes were identified for the suction strainers. This item is not considered to be a safety hazard and there-fore it is felt that no further evaluation is necessary.
c.
Emergency Condenser Valves (Tube Side) - It was determined that the Emergency Condenser itself was designed to ASME Section VIII. No design codes were identified however for the ECS valves or piping.
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