ML20065G950
| ML20065G950 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/28/1982 |
| From: | Bayne J POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | Vassallo D Office of Nuclear Reactor Regulation |
| References | |
| JPN-82-75, NUDOCS 8210040317 | |
| Download: ML20065G950 (10) | |
Text
POWER AUTHORITY OF THE STATE OF NEW YORK 10 CoLUMous CincLE NEW YORK. N. Y.10019 (2128 397 6200
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t GEORGE L. ING ALLg e r r e...
wtCE CMal asAN JOSEPH R. SCHMIEoER K6 CHARD M. FLYNN e.so..,
c..P CosERT 1. MILLONZI September 28, 1982 J O H N W. B.OST. O N..
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FREatmen R.eL4Rx 1""' l."J.l.".*.*!*"""
l JPN-82-75 THOM AS R. FREY 1":'.*".:'ffl."JJ'::"'
Director of Nuclear Reactor Regulation U.S.
Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Mr. Domenic B.
Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Inservice Inspection Program Request for Additional Information
References:
1)
P.J Early (PASNY) to T.A.
Ippolito (USNRC) dated September 10, 1979 regarding Inservice Inspection Program - Inservice Examination of Welds and Supports (JPN-79-57) 2)
D.B.
Vassallo (USNRC) to L.W. Sinclair (PASNY) dated April 14, 1982 - same subject 3)
J.P.
Bayne (PASNY) to D.B. Vassallo (USNRC) dated June 21, 1982 - same subject (JPN-82-52)
Dear Sir:
The Power Authority submitted, via Reference 1, inservice inspection program relief requests for our FitzPatrick nuclear power facility.
As a result of your initial review, you requested additional information on these relief requests via Reference 2.
We found it necessary, for the j
reasons detailed in Reference 3, to reschedule the submittal for that l
additional information.
Through discussion with members of your staff, we clarified some of your questions.
As a result, attached is a partial response to Reference 2.
I I
l Ob 9210040317 820928 PDR ADOCK 05000333 O
PDR l
I
Items 8 and 10 require further research or study by the Authority and therefore are not addressed in the attachment to this letter.
Specifically, some of this information can only be obtained when the plant is shut down.
Other infor-mation requires the completion of an industry survey.
A response to the remaining two questions will be provided by December 8, 1982 or approximately 30 days after the start of the next scheduled outage.
Also discussed with your staff was the schedule for the formal granting of these relief requests.
We are proceeding on the basis that these relief requests will be granted upon identification of those specific components not inspected (per our response to items II.1-II.6), at the end of the ten year inspection interval.
If you have any questions, please contact Mr.
J.A.
Gray, Jr.
of my staff.
Very truly yours,
\\
J.. P. Ba? 'hY'
' Executive Vice President Nuclear Generation cc:
Mr. Ron Barton United Engineers & Constructors, Inc.
30 S.
17th Street Philadelphia, PA 19101 Mr.
J.
Linville Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 136
- Lycoming, N.Y.
13093 Mr.
G.
Freund c/o S.A.I.
P.O. Box 696 Idaho Falls, Idaho 83402
Attachment to JPN-32-75 Pcwer Authority of the State of New York James A. Fit: Patrick Nuclear Pcwer Plant RESPCNSE TO APRIL 14, 1982 REQUEST FOR ADDITICNAL INFORMATICN - INSERVICE INSPECTICN PRCGRAM These responses are keyed to the enclosure of the April 14, 1982
- letter, D.3. Vassallo (USNRC) to L.W.
Sinclair regarding the James A.
Fit: Patrick Inservice Inspection Program.
I.
General 1.
A relief request will be submitted prior to May 1, 1983
- cr relief from the volumetric examination of the vessel to support skirt weld (Category 3-H).
Surface examination methods will be used in lieu of the volumetric methods required by the 1974 Edition Summer 1975 Addenda ASME Section XI.
Relief will be requested because of the difficulty obtaining full volumetric coverage due to the gecmetry of the weld.
New editions of the ASME code recognize this and permit surface examination.
II.
Relief Recuests Frem the Sectember 10, 1979 ISI Program Relief is requested from the requirement tha t all bolts, studs and nuts be inspected each inspection interval.
(la)
In lieu of this, bolted components which are disassembled for other maintenance or repair will have their bolts in-spected.
This relief would apply to the Incore Monitoring Housing Assemblies and the Control Red Drive (CRD) assemblies.
To date, approximately 7 of 42 (approximately 16%) In-core Monitor assemblies (3 bolts per assembly) have been dis-assembled for maintenance and inspected.
Approximately 79 of 137 (58%) Centrol Rod Drive Assemblies have been disassembled and inspected.
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The fisassenhly area rac:ation cse rate is approximate.,y 150-200 millarem/hcur whcie bcdy.
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s Final Safety Analysis Repcrt, Section 3.5 (pgs. 3.5-17, 35-13) each Centrol Red Drive (CRD) housing bolt has an allcwable lead
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the eight bcits would have to fail :
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Complete failure of all eight CRD hcusing belts result in a maximum leakage of 340 gym which :.s belcw -he make-up capacity cf
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n Leakage Detection System.
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ccclant pressure bcundary would not occur.
Leakage :. rom a complete ral.,ure of the in-ccre
=cnitoring belt:.ng, (assum'ng ccmplete separation) would result in leakage less than the ncrma' make-up rate prov:.ded by the EPCI systems.
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2a)
Figure 3.3-1 of the recently revised Fit: Patrick FSAR is a sketch which indicates the relative locations Of both the housing to vessel "D" weld and the stub tube to housing weld.
Note also that these component penetrations are belcw the core support plate in the vessel bottom head.
The difficulty of inspecting c =penents below the' core suppcrt plate prevents an adequate examination of the housing-to-vessel "J weld.
Any Leakage due to a ccmplete or partial failure of these welds would be detected by the Drywell Leakage Detecnicn System.
In addition, the area beneath the vessel bot:cm head will be inspected for leakage during the ten year hydrostatic test.
2b)
As discussed in the Fi :Patrzek F5AR, (Section 3.5.6.1; weld a ::=plete circumferential failure of attachment would resui: in the separation Of the hcusing from the vessel.
The C n=r:1 Red Drive and hcusing wculd be driven into the CRD Suppcru S tructure (Fit: Patrick F5AR Figure
- 3. 5-3 ).
In the wcrs: case, this wculd resul: in a 0. 06 inch diame:rt al clearance be: ween he hcusing and reac r
vessel.
Reacter water wculd leak at a race of appr:ximately 440 gym.
Normal make-up would be provided by the Reactor Core Isciation Cccling System (RCIC) at 400 gpm, and the CRC 4
pumps at 60 gpm.
Any leakage due :
a c mplete :r partial f ailure of the:e bolts would be detected by the Dryvell Leakage Detection 5ystem.
3)
We anticipate that all Main Steam Relief Valve bolting will be inspected within each ten year interval provided current valve testing requirements do not change.
There are four flanges which are part of the head spray portion of the RHR system.
The same two flanges are disassembled during every refueling outage as part of the reactor disassembly.
The bolting associated with these flanges will be inspected at least once per ten year inter-val.
The bolting of the other flanges will be inspected only upon disassembly of these components for other maintenance or repair.
Two flanges serve as decontamination connections for each Recirculation System loop.
Inspection of the "A"
- loop four inch flange has already been performed.
- However, there is no scheduled maintenance for these components, bolt inspection will be performed only when disassenbly Is required 4a)
Visual examination of the internal surfaces of the recircu-lation pumps is impractical because of the cost, difficulty in disassembly of the pump, and associated cersonnel radiation exposure.
Disassembly, strictly for inspection has several detrimental effects:
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5a) 5ee response to item la above.
5b)
Disassembly increases the bolt failure rate. Any leakage frem the reactor recirculation system will be detected by the Drywell Leakage Detection System.
Increased personnel radiation exposure will result frcm more frecuent inspection.
The Authority considers that disassembly, for bolt inspection alene, does not result in an increased level of safety.
The "A" recirculation pump mechanical seal has been replaced.
The mechanical seal bolts were examined at that time.
This represents an inspection of half the bolts *rithin this categerJ.
6a)
See response to item la above.
6b)
See response to items. Ib and 5b above.
6c)
A :ero-degree ultrasenic test of a valve bcdy would not be useful because:
1)
It'dces not provide a reliable method for detecting shallow surface indications.
ii)
It does not reliably detect material cracks, especially if these cracks are not nearly perpendicular to the sound beam.
iii)
These valves are either cast or forged with a grain structurc that causes severe ultrasenic sound beam attenuation.
This will further reduce the ability to detect surface flaws.
iv)
Many of these valves are made of stain-less steel, which further attenuates the ultrasonic sound beam.
v)
The Drywell Leakage Detection System provides a means for detecting valve leakage.
_5_
7)
Under the postulated conditions of loss of coolant, from the three inch (I.D. of 2.9") reactor core isolation cooling (RCIC) steam line and from the three inch (I.D. of 2.63")
Control Rod Drive return line, the reactor can be shv.tdown and cooled down in an orderly manner.
In this event, make-up would be provided by the HPCI system using on site power as described in the James A.
FitzPatrick FSAR Sections 6.3 and 6.4.
3)
Addi:icnal time is recuired :: research and evaluate
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. s inch and 3.500 inch wall = ice.
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., - _4 2.,3 2
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4 a..:.n.. s t. _,.o-t with wall :hicknesses between 3. 275 and 0. 5 00 inch.
9a)
Welds lccated Outside and immediatelf ad acent :=
centainment penetrations will be examined in accer-dance w th =cde recuiremenzs.
Ei:her surface or volumetric examinations will be performed depending u=cn the catec.erv. Or class of the weld.
These examinations wil'1 include the full length of the veld L
and will be perf ormed a: the required frequency.
9b)
All centainment penetrations are constructed :: direct leakage in:0 primary containment where 1: will be detected bv. the Drv.well Leakage Detecnicn 5-tem.
The Drywell Leakage Detection System ( DLDS ) is described in the recently updated Fit: Patrick FSAP.,
Secticn 4.10.
Surveillance requirements and limiting conditions for operatien Of the DLOS are in the Fit: Patrick Technical Specifications.
9c)
A ccmplete set of Fit: Patrick Inservice Insrection
)
Isemeteric drawings were provided 20 Mr. Gecrge Freund of 5.A.I.
(P. D.
Scx 696, Idano Falls, Idaho 33402) by letter dated August 20, 1982.
These drawings allistrate containment penetrations and nearny welds.
l 9d:
L sted belcw 15 the Code Categcry f or welds where relief is requested:
System Weld Code Cateccry Control Red Drive 3-J
.sC.n
- _ s.
~ce-4 I
Main 5 team 3-J Feecwater 3-1 Core Spray 3-:
RWC 3-;
(.
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z-a. a n d u -.-
n I-' Refer :: Inservice Inspection Iscmeter c drawings..
9e)
There are ac dissimilar metal welds included in this relief request.
10)
Additional time is required to verify the gecmeerv.
of these welds and to prepare sketches of their location.
_7_
Table 1 RDKHOR VESSEL PRESSURE RETAINING BOLTING SMALLER THAN 2 INCHES IN DIAMETER In-Core Monitoring Penetrations Control Rod Drive Penetrations (8 bolts per assembly)
(S bolts per assembly) 12-45 36-21 22-19 22-39 30-51 30-27 20-45 44-21 26-19 26-39 34-51 34-27 28-45 28-17 30-19 30-39 10-47 38-27 36-45 12-13 34-19 34-39 14-47 42-27 12-41 20-13 38-10 38-39 18-47 46-27 20-41 28-13 42-10 42-39 22-47 50-27 36-41 36-13 46-10 46-39 26-47 02-23 04-37 44-13 50-19 02-35 30-47 06-23 12-37 12-09 06-15 06-35 34-47 10-23 20-37 36-09 10-15 10-35 38-47 14-23 28-37 20-05 14-15 14-35 42-47 18-23 36-37 28-05 18-15 18-35 06-43 22-23 44-37 36-05 22-13 22-35 10-43 26-23 20-33 26-15 16-35 14-43 30-23 28-33 30-15 30-35 18-43 34-23 36-33 34-13 34-35 22-43 38-23 04-29 38-15 38-35 26-43 42-23 12-29 42-1.-
42-35 30-43 46-23 20-29 46-15 46-35 34-43 50-23 28-29 06-11 50-35 38-43 36-29 10-11 02-31 42-43 44-29 14-11 06-31 46-43 12-25 18-11 10-31 06-39 20-25 22-11 14-31 10-39 28-25 26-11 18-31 14-39 34-21 30-11 22-31 18-39 12-21 34-11 96-31 02-19 20-21 38-11 30-31 06-19 28-21 42-11 38-31 10-19 46-11 12-31 14-19 10-07 46-31 18-19 14-07 18-03 50-31 18-07 12-03 02-27 22-07 26-03 06-27 26
^7 30-03 10-27 30-07 34-03 14-27 34-07.
18-51 18-27 38-07 22-51 22-27 42-07 26-51 26-27