ML19207B819
| ML19207B819 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 06/21/1979 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Brian Lee COMMONWEALTH EDISON CO. |
| References | |
| 780621, NUDOCS 7909050320 | |
| Download: ML19207B819 (22) | |
Text
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UNITED STATES y D c,%,( j NUCLEAR REGULATORY COMMISSION l/4,4
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. E WASHINGTON, D. C. 20555 A
%, ' u+....f Docket Nos. 50-373 and 50-374 JW 21979 Mr. Byron Lee, Jr.
Vice President Commonwealth Edison Company P. O. Box 767 Chicago, Illinois 60690
Dear Mr. Lee:
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION - LA SALLE COUNTY STATION, UNITS 1 & 2 As a result of our site visit on fire protection on April 2 to 5,1979, we find that we need additional information to complete our review of your fire protection systems. contains our request for additional information which supplements our February 26, 1979 request.
In addition, we are enclosing Enclosure 2 which requests additional information in the areas of instrument and control systems, materials engineering and radiological assessment. Our concerns in each of these areas were either discussed with or made known to your personnel.
Please inforn us af ter receipt of this letter of the date you can supoly the requested information 50 that we may factor that date into our review schedule.
Please contact us if you desire any discussio., or clarification of the inforration requested.
Sincerely, b
u t a n u. r'a rr'. M r-Light Water Reactors Branch No. 3 Division of Project '!anagement
Enclosures:
As Stated cc w/ enclosures:
See next page n,
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.a 790905 p 1
Mi oyron Lee, Jr. JUV 2 I 879 cc: Richard E. Powell, Esq.
Isham, Lincoln & Beale One First National Plaza 2400 Chicago, Illinois 60670 Dean Hansell, Esq.
Assistant Attorney General I'
State of Illinois
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188 West Randolph Street Suite 2315 Chicago, Illinois 60601 u
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ENCLOSURE 1 010.0 AUXILIARY SYSTEMS BRANCH Note: Questions which reference a specific fire zone in Unit I apply equally to the counterpart fire zone in Unit 2.
010.59 In your comparisc.; to Appendix A to BTP 9.5-1, you stated that LSCS is in (RSP) compliance with positionE.3.(b). However, at our site visit we were informed that only deluge control valves are electrically supervised, and other valves in the fire protection system are sealed with wire seals. Wire seals are not acceptc.ble. State your compliance with section E.3.(b) by providing electrical supervision with alarm and annunciation in the control room for all valves in the fire protection system, 010.60 In your comparison to Appendix A to BTP 9.5-1, you state that LSCS is not (RSP) in compliance with position E.3.(d) regarding standpipe and hose located inside containment. State your compliance with section E.3.(d) by providing standpipe and hose stations outside the containment access with sufficient hose so that all areas of containment can be reached.
010.61 Provide diagrams which indicate the routing of all flammable and combustible lia':id or gas lines in all plant areas containing or exposing safe shutdown systems, including those areas of the turbine building adjacent to, or inte-grated into the auxiliary building.
010.62 On page 9.5-4 of the FSAR, you state that " suspended ceilings are of negligible (RSP) combustibility." State your compliance with Section D.l.(f) of BTP 9.5-1, Appendix A, by providing sufficient data to demonstrate that such ceilings are of noncombustible construction as defined in Section B.4 of Regu-latory Guide 1.120, or replace the ceilings with materials which will comply.
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010.63 Fire Area 1 Refueling Floor On Page H.3-1 of the FSAR, you indicate that the walls and roof of this zone are metal on unprotected steel' supports, and that two open stairwells, as well as other unprotected floor openings, exist between this zone and lower elevatio..s of the reactor building. Analyze the effect on safe plant shut-down or on radiation releas-e if a fi.re were to cause the collapse of the
.. roof _and/or. walls of _this zone.
010.64 Fire Area 2 Reactor Buildina (RSP)
(1) Figures 9.5-1, Sheet 5,and 9.5-2, Sheet 4 indicate for Fire Zone 231 that a single failure in the fire suppression system could cause the loss of the deluge system for the standby gas treatment system charcoal filter and the hose stations at columns ll A and 15C, leaving only 50 f t. of, hose at column 8.93C to combat a fire in the filter area.
In addition, Figure 9.5-la, Sheet 1, indicates the presence of redundant trains of safe shutdown cables in this area near the charcoal filters.
The deluge system should be converted to an automatic system, and that the length of hose at the column 8.9BC hose station be increased to 100 ft. to assure that fire protection water will be available to combat the fires in this area.
(RSP)
(2) The door from Fire Zone 2G to the equipment access air lock which leads to the outside is a large, non-rated, motor-driven swing door.
As a minimum, a 3 hr. rated fire door should be providea for this opening and an automatic sprinkler systen be provided in this area of the reactor building around the railroad tracks to protect against a fire in materials which may be present in this area.
The fire protection should i
also extend to areas in thi.; fire zone which contain redundant divisions of safe shutdnwn equipment and/or circuits.
See our position stated in Questions G10.54(4) and 010.56(7) concerning fire protection for redundant trains in close proximity to each other.
(RSP)
(3) Figure 9.5-1, Sheet 22, indicates that a 3 hr. rated fire door separates the off-gas building from the reactor building, Fire Zone 2G. However, on our site visit there were no labels on this door.
A 3 hr. rated fire I
door should be provided to close this opening.
(RSP)
(4) Figore 9.5-1, Sheet '9 indicates for Fire Zone 2H1, that the isolation dampers from this zone to the steam tunnel are 3 hr. fire rated dampers.
However, on our site visit we were informed that these dampers are not fira-ra ted.
'Three hr. rated fire door / dampers should be installed in these vent openings to provide a complete 3 hr. fire barrier between the reactor building and the steam tunnel.
010.6E Fire Area 4 Auxiliary Building (1) You indicate in the FSAR on Page H.3-57 that the roof over Fire Zone 4A, auxiliary building upper ventilation equipment floor, is mostly metal decking on 1 hr. protected structural steel. Analyze the effect on safe plant shutdown if a fire were to cause the collapse of the roof or the 1 hr.
rated walls of this zone.
Consider that such a structural collapse may affect the structural integrity of other floors or buildings, adjacent to the fire zone.
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(2) As 'in Question 010.65(1) above, analyze the effect on safe plant shutcown if a fire were to cause structural collapse of the ceiling or walls 6f Fire Zone 48, lower ventilation equipment floor.
(3) Your fire hazard analysis in the FSAR, Page H.3-64, states that the floor of Fire Zone 4C2, auxiliary building main floor, is a 2 hr. rated construction.
Figure 9.5-1 Sheet 13 indicates it is only 1 hr. rated.
For this area and any others with similar construction, analyze the effect on safe plant shutdown of a collapse of this floor slab. Consider, at least for this area, that the floor supports for the control room nay be tied in with floor supports of this area.
(RSP)
(4) During our site visit we observed a flammable liquid storage cabinet in the film room of Fire Zone 4C4, computer room area.
The flamable liquid storage cabinet should be removed from this area as stated in Section D.2a of Appendix A to BTP 9.5-1, and the storage of flammable liquids should be limited to areas located in non-safety related buildings.
(5) During out site visit we were infomed that the Halon suppression system (RSP) will not be installed in Fire Zone 4El, auxiliary equipment room.
As a minimum an automatic suppression systen should be installed in this area. Also the positions stated in Questions 010.54(4) and 010.56(7)
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should be implemented.
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(6) On page 9.5-12 of the FSAR you state that the manual deluge systems for the auxiliary electric equipment room, Fire Zone 4E2, supply air filters are operated from the same room. Indicate how the deluge systems will be 8
actuated assuming a fire in the filters which fills the auxiliary electric equipment room with dense smoke.
(RSP)
(7) During our site visit we oeserved that exhaust fans for the battery rooms are being installed to exhaust air from the floor of the rocms.
The exhaust air systems'should be modified to remove air at the ceiling level in each of the battery rccms to prevent pocketing of hydrogen gas at the ceiling.
(RSP)
(8) The battery rooms should be provided with automatic suppression systems unless they are separated from other areas of the clant by 3 hr. fire rated barriers.
(RSP)
(9) Fire Zone 4F1, Page H.3-86.
During our site visit we observed an enclosed bus duct which passed vertically through Fire Zone 4F1, Division 1 essential switchgear room at the south wall. 'de were advised that tha bus duct penetra-tions of the floor and ceiling slabs will be sealed around the outside of the duct only - that there would not be a seal inside the cable duct. This is not acceptable.
Indicate your compliance with Section D.l(j) of Aopendix A to BTP 9.5-1 which requires that all penetrations of rated fire barriers be sealed to provide a fire resistance equivalent to that of the fire barrier.
This includes seals in fully enclosed electrical raceways where the raceway itself is not enclosed in a 3 hr. rated fire barrier.
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(10) Fire Zone 4F3, Page H.3-89.
Durino our site visit we were informed tnat the Halon suppression systea will not be installed to protect the cables above the suspended ceiling in Fire Zone 4F3 auxiliary building ground floor.
Indicate your compliance with BTP 9.5-1 Appendix A Section D.l(f) and provide adequate fire detection and suppression systems for the cable in the concealed space.
In addition, the area below the ceilings should be provided with an automatic fire detection system.
010.66 Fire Area 5 Turbine Buildino (1) Fire Zone 5A4, cable area elevation 749 feet, contains ESF Division 2 cables from both units.
Describe the procedure used to shut do.m both units ith available onsite power if a fire were to cause loss of these Division 2 cables from both units.
Consider that in case of loss of offsite power, there is only one Division i diesel generator to handle the neces-sary loads of both units.
Show that this diesel has adequate capacity for shutting down both units.
(RSP)
(2) Because of the relatively high concentration of cables in Fire Zones 5A4 and 5B13, balance-of-plant cable area including so.1e Division 2 cables, it is our position that an automatic water suppression system be installed in these areas in accordance with Section D.3.(c) of BTP 9.5-1 Appendix A.
(RSP)
(3) The wall which separates Fire Zone 539, motor driven reactor feed pump room, from Fire Zone 8Al, HPCS diesel ventilation equipment room, is not a rated fire barrier.
This wall should be upgraded to provide a 3 hr. rated fire barrier between these two zones to prevent a fire in Zone 539 from spreading into Zone 8A1.
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(4) The'125-V battery and cables for ESF Division 1 of Unit 2 are located in Fire Zone SCll, turbine building ground floor. The 2 hr. fire rated walls which separate the battery room from the remainder of this zone, as well as the floor and ceiling of the battery room, should be upgraded to provide a 3 hr. fire barrier to separate the battery room from Fire Zone SCll, and that all ESF cable associated with these bc.teries be separated from the turbine building by 3 hr. fire rated barriers.
010.67 Fire Area 7 Diesel Generator Buildina (RSP)
(1) Smoke or other products of combustion entering the diesel generator room air intakes could render the diesel generators inoperable. Such a scenario which could render all diesel generators for one unit inoperable is possible given fires in any of the outside transformers adjacent to the diesel genera-tor building, a fire in any of the diesel fuel oil tank rooms, day tank rooms, or diesel generator rooms, all of which vent to the roof of the diesel generator building.- This is also true for a fire in the bus duct cables immediately outside the air intakes for the diesel generators.
This situa-tion is not acceptable. Modify your design such that a fire will not render all the diesel generators inoperable.or describe how you will safely shut down either one, or both units, if such a fire occurs concurrent with the loss of offsite power.
(RSP)
(2) During our site visit we observed that the doors to and between each of the diesel generator rooms are not pmvided with curbs to contain a fuel oil spill. All entrances to the diesel generator rooms, including the doors between the rooms, should be provided with curbs to contain a fuel uil spill,
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(3) During our site visit we observed that the discharge nozzles for the CO 2
systems in the diesel generator rooms were limited to a' single line of nozzles directly )ver the diesel generators.
Indicate the design parameters for this total flooding system and pmvide the results of tests which verify that the design concentration can be attained using this single line of nozzles throughout the rooms for the required soak time.
(4) Describe the provisions utilized to prevent a rupture at any point in the fuel oil lines from draining the entire contents of one fuel oil tank.
(RSP)
(5) Your design of the diesel generator building, with the fuel oil storage tanks located below the diesel generators, is not in compliance with Section F.10 of BTP 9~.5-1 Appendix A.
Provide a liquid tir,ht, 3 hr. fire rated barrier, including all penetration seals, between the fuel oil storage tank rooms and the diesel generator rooms, and all other areas.
In addition, the fuel oil storage tank area and any other area which is susceptible to a large diesel fuel oil spill (if conditions of (4) above are possible) shoulo be provided with a backup automatic or manual fixed suppression system in addition to the primary automatic suppression system.
OlC.68 Verify that the safety-related cable trays in Fire Zones 10Al,1031 and 10C3, Off-Gas Building,do not contain circuits required for safe plant shutdown.
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010.69 Class IE and associated circuit cables at the La Salle County Station (RSP) are marked at each end and in junction boxes.
Cables, located in cable trays, are not marked and the markings, located in junction boxes, lacked sufficient durability to be readily recognized as color coded markings.
Visual verification that the cable instullation.
is in conformance with separation criteria was not possible, In accordance with the specific identification criteria of IEEE Standard 384-1974 and Regulatory Guide 1.75 (Revision 1), we recuire that both Class lE and associated cables be marked in a manner of sufficient durability and at a suffcient. number of points to facilitate visual verification that the installation is in conformance with separation criteria. Cables must be marked either before or during installations.
Since this has not been done at the La Salle station, we require a detailed review of cable identification and routing methods.
In this regard, we requir 3 a positive physical check of the installed cable be performed usine high-frequency tracing (or other method) and sample testing techniques.
We will request that I&E review your positive physical cneck of the installed cable.
J10.70 In accordance with Section 8.3.1.3.2 of the FSAR, exposed conduits are supposed to be marked by color codes at the beginning and the end of the run, on both sides of a wall through which the conduit passes, and at both sides of junction boxes.
Class lE conduits at the La Salle County Station are not marked by color code in accordance with the FSAR.
Correct this
- ficiency.
We will request that I&E inspect this deficiency.
11 0. 71 In regard to separation between Class IE and non-Class IE cable (RSP) trays and cables, the separation is less than 3 ft. between trays separated vertically, the separation is non-existent between cables rising from both Class IE and non-Class IE trays in cable spreading areas (cables are bundled).
It is our concern that fa "ures or faults in non-Class IE cables will degrade Class IE c caits belcw an acceptable level.
For the non-Class IE cables rising from trays with no separation from Class IE cables, we consider that they are associated circuits and should meet the guidelines for associated circuits of IEEE standard 384-1974 and Regulatory Guide 1.75.
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For vertical separation of less than 3 f t. between Class IE and non-Class IE trays, the results of an analysis should be provided in accordance with section 5.1.1.2 cf IEEE standard 334-19)4, that demonstrates that failure or faults in non-Class IE circuits will not degrade Class IE circuits below an acceptable level or these non-Class IE circuits must be considered associated circuits and meet the applicable requirements of IEEE standard 384-1974 and Regulatory Guide 1.75 for associated circuits.
010.72 To assure that redundant safety related cable systems are separated from each other so that both are not subject to damage from a single fire hazard, the following information for each Class IE system required to bring the plant to safe cold shutdown should be. provjded:
(1)' Provide a Table listing electrical eouioment recuired nr
. essential for safe shutdown.
(2) Define each equipments location by fire area, (3) Define each equipments redundant counterpart with a description of its locations with respect to its redundant counterpart.
(4) Identify the essential cabling (instrumentation, control, and power) for each equipment.
(5) Describe the routing of each essential cable identifted in
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item (4) (by fire area) from source to terminations.
(6) Identify each location where essential cables are located in the same fire area with their redundant counterpart, (7) For each location identified in item (6), describe the effects on safe shutdown f f both redundant cables are lost from an exposure fire.
010.73 Section H.3.4.3 (page H.3 63) of Appendix H to the FSAR fndicates that for the design basis fire in the control room:
(1) the auxiliary equipment room contains the remote shutdown panels, (2) all circuits in the remote shutdown panels are electrically isolated from the main control room, and (3) remote shutdown circuits are unaffected by loss o,f the control room circutts, To assure that the electrical isolation between the control room and the remote shutdown systems is sufficient to preclude a design basis fire in the control room, in the auxiliary equipment room, or at remote shutdown control locations from reducing the safe cold shutdown capability below an acceptable level. provide the following information-(a) Identify each circuit located on the hot shutdown panel required for shutdown with a description of how it is isolated from the control room circuitry.
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(b) For each circuit identified in item (a), provide detailed electrical schematic drawings which clearly describe the electrical isolation between the hot shutdown panel and control room.
(c) Identify each circutt required for safe cold shutdown located in the control rcom but r.ot en the safe shutdown panels located in the auxiliary equfpment room.
(d) For each circutt identified in item (c), provide the results of a,
analysts tuat demonstrates that failurt (open, short, or hot short) of these circuits due to a design basis fire in the control room will not affect their remote safe shutdown capabtitty.
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- I EtiCLOSURE 2 030.0 IriSTU.]1TATI Yi Afl0 COTITROL SYSTEMS BRA ;CH 031.263 We do not understand your response to Part 2 of Question (G.2.1.2)
(G.3.3.3) 031.248.
Therefore, please clarify the following:
(QO31.248)
(1) Part 2 of Quastion 031.248 deals with a discrepancy between [~
a stated 20 cercent and ?5 Dercent reactor cumo soeed trio.
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response refers to FSAR Section G.2.1.2.6.2e which discusses a 25 percent flow valve position interlock.
(2) The response to' Questions 031.248 Part 2 also references FSAR Section G.3.3.3.10.2.2 which states that the recircula-tion pumps can be transferred to high speed from either tne icw speed motor generator or a dead start if the speed is
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less than 20 cercent.
Figure G. A-3, which is also referenced in
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the response, deals with a 18' percent valve position interlock.
(3)
It is our unde-standing that the actual design in-cludes a valve pdsition interlock which is set at 25 nercent of stroke and a speed interlock which is set at 20 percent.
Are m.
L, these understandings correct?
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031.264 (RSP) Your response to Question 031.35 states that no nuclear steam supply
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[QO31.35) shutoff system isolation valves are provided with the r.anual r-override feature.
The response then states that the reactor
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coolant sample lines and valves in the sample lines for post
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ranual override of the isolation signal. These manual overrides p.
were rot found during our review of the nuclear steam supply shutoff system final design drawings.
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Existing manual overrides must be shown in the final design drawings. Further, you shc ild describe how the manual override complies with IEEE 2/9-1971, Sections 4.11 through 4.14.
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Therefore, please amend the FSAR accordingly, 031.260 Discrepancies have been found in various final design drawings (RSP)
(F807E157TD) during our review.
Exac.ples of the problem areas are:
(F807E166TD)
(F7.3-13)
(1) The reference table of General Electric (GE) Figure 307E152TD (Nuclear Steam Supply Snutoff System, Sheet 1) shcws contacts 1-2 of coil K30 as being a spare. On sheet 6 of the same figure, this contact pair is shown as being used in indicating C.
light circuitry.
E (2) FSAR Figure 7.3-13 (Nuclear Steam Supply Shu;off System, sheet
- 1) shows thE drywell high pressure signal as isolating various valves (E12-F023, E12-F009, E12-F008, E12-F049, E12-F000A,
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E12-F0408, E12-F053A and E12-F053B) in the RHR System.
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the logic in GE Figure 807E152TD, sheet 6, shows valves
@.i E12-F009, E12-F023, E12-F008, E12-F053A and E12-F05313
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,w (3) FSAR Figure 807E152TD, sheet 6 (FSAR Figure 807E152TD, sheet 8), the isolation logic shows the drywell high pressure
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and reactor low water level 2 isolation signals operate contacts of Kl A, B, C, D and C71A - K4A, B, C, D respectively. However, on the same page the isolation logic for high drywell pressure shows that the high dryvell pressure signal operates contacts of C71A-K4A, E C, D While sheets 5'snd 5A of the GE drawing (FSAR Sheets 6 and
- 7) show contacts of KlA. B, C, D as being controlled by reactor icw water level 't level 2.
Also, GE Figure 807E16eTD, e
sheets 10 and 13, shows drywell high pressure to be associated with contacts of C71 A-K4A, B, C, D.
(Hence tne logic for K1 and K4 is reversed on more t~ n one sheet.)
(4) FSAR Figure 7.3-11 refers to FSAR Figure 7.3-15, sheet 2, zone B-ll for the details of valve steam leakoff detection for valve E12-F009.
Zone B-ll does not exist on the referenced IED (as determined by the response to Question nt.
IT-031.15).
ff Revise the final system design drawings to resolve ali such discrepancies and submit the revised drawings in accordance with RG 1.70, Chapter 7.0.
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The responses to Parts 2 and 5 of Question 031.137 and Part 2 (Q031.137)
(Q031.240) of Question 031.240 do not demonstrate that the devices which are used to protect the Class lE nuclear instrumentation from the non-Class lE control systems are suitably qualified. Therefore, the above cited responses are unacceptable.
Provide amended responses to the cited questions which demonstrate that suitable (Class lE) isolation is provided to protect the outputs from the nuclear
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instrumentation.
031.267(RSP)
(QO31.240)
The response to Question 031.240 Part 3 is incomplete and, therefore, unacceptable.
Revise the response to include a description of the effect of Test 1 on the inputs and the bases for accepting such voltage transients in the nuclear instrumentation circuits which w
4 feed their signals to this type of isolator.
031.263 The meaning of your response to Question 031.137 Part 7 and the F,031.137)
(QO31.240) follow up Question 031.240 Part a is not clear.
State that the isolation of the ESF status signals occurs within the ESF equipment via relay contacts and that the subject ESF status input cabinets contain no Class lE wiring or provide a clear description and i,
justification for any alternative design.
The response to Question 031.256 Part 3 is unclear.
Describe the r.
031.269 (QO31.256) signal source for K62413 and K624C when the power supply which is common to K611, K612, K613, and K624A fs.
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031.270 Clarify the discrepancy between FSAR Figure 9.4-1 Sheet 6 (F9.4-1)
(IE-0-4432AD) and FSAR Drawing IE-0-4432AD with regard to the control of OVC16YA.
t 031.271 Justify your assertion that a failure of relay CREY-VCC80XA, (IEO-4432AD)
OXY-VC088X1 or 0AE-VC090XI does not constitute a single failure which will leave the control room unisolated when isolation is m
required.
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120.0 MATERIALS ENGINEERING BRANCH 121.11 In Section 5.2.3.3.1.4, of the FSAR, " Operating Limits Based on Fracture Toughness," a factor of 2 F per ft-lb is used to convert longitudinal Charpy V-notch impact data obtained at the 30 ft-lb level to estimates of the data at the 50 f t-lb level.
It is stated that these estimates plus a 30 F adjustment for specimen orientation are based on information tabulcted in WRC Bulletin 217, " Properties of Heavy Section Nuclear Reactor Steels," and other fracture toughness tests.
Explicitly state the procedures used to verify these factors, including a sample calculation and any data, other than that in WRC Bulletin 217, used as a basis for the estimates.
121.12 To demonstrate compliance with Section IV.A.3 of Appendix G, 10 CFR Part 50, provide the results from the C impact tests y
for materials for piping, pumps and valves.
121.13 It is stated in the FSAR that at 10 F the reactor vessel closure stud materials for Unit No. I have a minimum Charpy impact energy and lateral expansion of 43 ft-lb and 23 mils a
respectively. The reactor vessel stud materials ir. Unit No. 2 c~
have a reported minimum Charpy impact energy of and a lateral po--
expansion of 40 ft-lbs and 24 mils respectively. To demonstrate L-l comf iance with Section IV.A.4 of Appendix G,10 CFR Part 50, provide the Charpy impact test results for bolting and other fastener materials in Unit Nos. 1 and 2.
121.14 Referring to Table 5.2-11 of the FSAR, the statement is made that for the nozzles, flange and shell regions near geometric discontiruities a RT of 40 F is used in lieu of a dis-continuity analyses S0' demonstrate compliance with Section IV.A.2.b of Appendix G.
Provide a calculation to demonstrate theadequacyofRT,9brthedeterminationofthereactorvessel of 40 F in those regions and show how
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f this value is used' pressure temperature operation limits.
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121.15 Section IV.B of Appendix G,10 CFR Pert 50, requires that the reactor vessel beltline material poss?ss a minimun upper-shelf t"
energy of 75 ft-lbs as determined fron' Charpy impact test data
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on unirradiated specimens conducted in accordance with g'.
paragraph NB-2322.2(a) of the ASME Code. To demonstrate compliance with Section IV.B, provide the upper-shelf Charpy h',
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impact test data for the reactor vessel beltline materials.
If upper-shelf energies of less than 75 ft-lbs were obtained, c
analyses and data also must be submitted to demonstrate 7"
adequate margins of safety from deterioration by neutron irradiation.
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121. E Section II.B of Appendix H of 10 CFR Part 50 requires that materials from the reactor beltline region be monitored by a surveillance program complying with ASTM E 185-73, " Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels."
The FSAR indicates that the specimens selected for the surveil-
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lance program are from representative materials frca orientation and location not in conformance to ASTM E 185-73.
Provide the following information for the specimens in the surveillance program:
(1)
Identify the surveillance specimens taken from the base and weld materials in the beltline region of the reactor vessel, including the plate and weld identification number, specimen orientation, and the location from which the specimens were taken. Also identify the base and weld metal surveillance specimens that were not taken from the actual base and weld materials in the reactor vessel beltline region. The identification should include:
(a) Materials specification, heat number and material identification number C"
(b) Weld wire e
(c) Weld flux (d) Weld process (e) Heat treatment (f) C impact energy test results C
y (g) RTNDT r-(h) Copper content h[Ysh W
(2)
Provide technical justification for deviation from the hy requirement of Section II.C.1 of Appendix H that the G'
test specimens in the surveillance program be taken from
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alongside the fracture toughness test specimens which were N
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required by Section III.A of Appendix G of 10 CFR Part 50.
The information should demonstrate that the test specimens i
are fully representative of the materiais and processes
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used for the fabrication of the beltline region of the reactor vessel.
(3)
In response to Question 121.3, the statement is made that based on experience at the General Electric Company the amount of shift measured on irradiated longitudinal test pr==-
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specimens will be essentially the same as the shif t in equivalent transverse scecimens.
Provide the Charpy impact test data to demonstrate that the RT,g shif t and the decrease in the upper-shelf energy leves due to neutron irradiation are equivalent regardless of the specimen orientation.
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3 ?,0. 0 PADIOLOGICAL ASSESSME'lT 331.23 Describe permanent shielding provided to assure (12.3) acceptable radiation levels in potentially occupied areas in the vicinity of the spent fuel transfer process.
If very high radiation areas are projected, describe precautions taken to prevent inadvertent pertsnnel access during fuel transfer.
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