ML19326C801
| ML19326C801 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 01/23/1973 |
| From: | Bernero R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8004280736 | |
| Download: ML19326C801 (11) | |
Text
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UNITED STAT ES f'
g ATOMIC ENERGY COMMISSION W ASHINGT ON D C.
2054's
,5 ham e JAN 2 3 G73 Docket No. 50-313 R. C. DeYoung, Assistant Director for Pressurized Water Reactor, L THRU:
A.
Schwencer, Chief, Pressurized Water Reactors Branch No. 4, Licensing REPORT OF SITE VISIT TO ARKANSAS NUCLEAR ONE - UNIT 1 Date:
November 29 - December 1, 1972 Location:
Arkansas Nuclear One Russellville, Arkansas and Arkansas Power and Light Company Little Rock, Arkansas
Purpose:
To review Arkansas Nuclear One - Unit 1 Electrical and Control Installation (See Attached Agenda)
Groups Participating:
(See Attached Attendance List)
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R. M. Bernero, Project Fbnager Pressurized Water Reactors Branch No. 4 Directorate of Licensing
Enclosures:
1.
Sununary Report 2
Agenda 3.
Attendance List cc:
S. H. Hanauer PWR Branch Chiefs R. C. DeYoung R. W. Klecker R. S. Boyd W. Morrison D. J. Skovholt G. Arlotto D. Muller R. F. Fraley R. R. Maccary RO (4)
D. Knuth Receptionist R. Tedesco V. Moore H. Denton REG Attendees R. Minogue AEC PDR Local PDR 736 8004280
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ENCLOSURE l_
1
SUMMARY
OF SITE VISIT AND FINAL ELECTRICAL DRAWING REVIEW NOVEMBER 29 - DECEMBER 1, 1972_
ARKANSAS NUCLEAR ONE - UNIT 1 DOCKET No. 50-313 The Arkansas Nuclear One pite was visited on November 29 and 30 for review of the electrical and control installations; the final session of the electrical and control drawing review was held at the Arkansas Power and Light Company Little Rock of fices on December 1,1972.
Drawing Review The items discussed in the drawing review included:
Air-Operated Valves It appears from electrical schematic review that some engineered safe-guards valves need air to operate on Engineered Safeguards Actuation System (ESAS) signal. The applicant agreed to review this matter because the Final Safety Analysis Report (FSAR) says that no air is required for accident response.
Dwg. E-196 We noted that the power < 22% permissive interlock for starting the reactor coolant pumps was not consistent with the < 15% power cited in the FSAR. The applicant agreed to check this.
Dwg. E-209 It was noted that reactor building outlet isolation valves CV-7401 and 7402 derive actuator power from the same breaker. The applicant agreed to examine this.
Dwn. E-216 We noted that the ESAS auxiliary relay contact (42X-5153). used to break
.the seal-in feature in the open circuit of the letdown cooler isolation valves appears to be wired incorrectly. The applicant agreed to examine this.
, SafeRuards Valves Torque Switches Electric-motor-operated valve design incorporates torque switches to cut motive power when the driving torque increases to a certain level.
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torque switch bypass is provided only for the first 5% of travel to We noted that, with enable the valve to overcome high starting torque.
the existing design, a momentary loss of power or a modest increase in The mechanical resistance might disable such a valve in mid stroke.-
applicant agreed to investigate this matter.
Dwn. E-232 We noted that Control Rod Drive / Reactor Coolant Pump cooling isolation valve CV-2221 was shown as a motor-operated valve here while Fig. 9-7 of the FSAR shows it as air-operated.
The applicant indicated that the valve is air-operated and will correct this drawing.
Core Flooding Tank (CFT) Isolatio:. Valves l
The indicator and alarm circuits associated with the motor-operated CFT isolation valves were discussed with respect to the applicant's FSAR response to AEC Question 6.7.
The applicant indicated that with his present design the position indication and position alarm functions are Based activated by separate contacts on the sama valve position switch.
on the applicant's indicated intention to have the power supply breakers to these valves locked open and tagged during normal operations, we stated the following design requirements.
Valve position visual indication should be available independent of a.
the actuator power supply.
b.
A valve-not-open alarm should be provided to signal if the valve is not fully open when reactor coolant pressure is above a preset value. 1Rie sensor for this alarm should be independent of the position indicator (item a.) and the pressure signals should be redundant and independent.
The Technical Specifications should require no criticality or prompt c.
shutdown unless both CFT isolation valves are open and the breakers are locked open and tagged.
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. Decay Heat Removal System (DilRS) Valve Interlocks The outstanding requirement of automatic closure and anti-opening inter-locks for the DHRS isolation valves (AEC Question 9.1) was discussed again. The applicant admitted that Fig. 9-12 in the FSAR incorrectly designates the two valves on either side of the reactor building pene-tration as the high pressure to low pressure transition and, therefore, the valves to be equipped with these special controls. A third valve, upstream of this pair, exists and this third valve (CV-1050) and the next valve in line (CV-1410) bound the high pressure section and will get the special controls. The third valve (CV-1424) in the low pressure line outside the reactor building will be treated as an ordinary isolation valve.
Dwg. E-280 We noted inconsistencies between designations of the reactor building coolers on this drawing and those shown on Figure 9-6 in the FSAR. The applicant acknowledged these and stated that installed equipment connec-tion reversals have been discovered in the service water system and must be corrected.
Emergency Feedwater System (EFW)
The emergency feedwater system design was discussed in relatica to the AEC request for information presently outstanding (Questions 14.31 and 14.12). It was noted that the present design does not meet single failure criteria in such areas as equipment location, power sources, and actuator circuits. We also confirmed that the motor-driven EFW pump is connected to non-essential bus Al; therefore, in the event of loss of off-site power, manual connection to the emergency buses is necessary.
In addition, bus Al is not seismic Category I.
We also noted that FSAR Fig. 7-22 showed the steam supply connection for the EFW pump upstream of the relief valves for steam generator 24A and downstream from the relief valves for steam generator 24B. The applicant agreed to check whether this difference could affect the performance of the EFW pump turbine.
Dwg. E-241 The applicant confirmed that no time delay is built into the control circuits of the reactor ouilding spray pumps other than that required for diesel loading sequence.
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It was noted that the engineered safety feature trip output relay contact designation for the reactor building cooler fans IB and 1D were The applicant agreed to investigate this.
in error.
Dwg. E-16 We noted that both pairs of the engineered safety features switchgear room coolers are fed from the same bus (B55). Thus, a failure of this We indicated bus would cause loss of cooling capability in both rooms.
that if loss of cooling impaired the proper function of the switchgear the design should be modified to provide independent power sources.
Dwg. E-19 We noted that the filter recirculation bleeder valves (cf. FSAR Fig. 6-
- 10) were powered from the same bus as the fan whose failure they are to We indicated that the power source for each valve protect against.
should be shared with the opposite fan.
FSAR Table 8-1 This table shows that the instrument air compressors, all main turbine generator lube oil pumps, and the turning gear motor are prevented from We noted that shutting off running when an ESAS signal is present.
these pumps could lead to severe mechanical damage of the main unit andWe asked the applicant to confirm his intent to operate in this manner.
stated that our concern is not for mechanical damage of their generator, but only to identify all the loads that will be borne by the diesel generator so that we can evaluate its capacity.
SITE REVIEW The plant installation was reviewed following the attached agenda.
The following matters were noted:
Reactor Building Pressure Sensors The installed reactor building high trip pressure sensors provide an analog output signal rather than a digital signal as indicated by the
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- FSAR and the as-built Reactor trotection System (RPS) logic drawings.
The applicant was informed of thiv inconsistency and agreed to inform us 1
of his resolution.
Separation of Equipment Two of the three redundant ESAS reactor coolant pressure sensors were found to be installed on the same instrument rack. Concern was expressed to the applicant about an event such as pipe whip which might lead to the simultaneous f ailure of both these units. The applicant agreed to consider this.
The doors separating adjacent redundant equipment spaces such as the diesel generator rooms and the 4160V switchgear rooms are not watertight The applicant agreed to reevaluate the potential inleakage designs.
rate from piping f ailures against the door leakage and floor irainage capacities.
We found that the two emergency battery rooms are directly connected by a large ventilation duct since the exhaust fan for one battery room is located within the other battery room.
The applicant agreed to recon-sider this design.
Raised Floors The RPS cabinets are located in the rear area of the control room on a raised floor. The cables entering these cabinets run beneath this i
raised floor with no separation barriers except flexible plastic conduit j
sleeving.
We' expressed our concern to the applicant that this cable arrangement is highly vulnerable to common mode failures and appears to violate the applicants own separation criteria as documented in the The applicant claimed that safety is not compromised because of FSAR.
the f ail safe (upon loss of power) characteristics of the RPS design.
A similar arrangement of cabling under a raised floor was found for the rod drive control cabinets in the computer room above the control room.
Here too the separation criteria may be violated.
Overhead Cables in Control Room We noted that there were many heavy cables in open raceways in the over-
. head space of the Control Room.
These cables were identified as the 48V We rod drive control cables each carrying a current of 40-50 amps.
expressed concern that these cables constitute a potential fire source-1 l-t
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which can cause a loss of the combined control room for both plants.
i-The applicant claimed that the cables are derated and only a limited number of these cables carry current at any one time.
Emergency Coolers in Control Room The presence of the package emergency cooler in the Control Room was We discussed the potential for damage to cabinets nearby.
noted.
1 Flexible Sheathing on Cables We noted.an apparent-lack of conformity to separation criteria in cabinets and panels due to the disposition of cables run in flexible the cables in question were sheathing. The applicant pointed out that not safety related and that the sheathing is added as a conservative design option.
Cable Spreading Room We noted that the Cable Spreading Room is very densely stacked although We the safety related cables were installed in solid steel conduit.
expressed our concern about fire hazards.
Diesel Fuel Supply We questioned the very limited fuel capacity of the diesel generator day The appli-tank (375 gallons, app oximately 1-1/2 hours at full power).
cant noted that the diesel fuel transfer pumps will automatically replenish the day tank from the emergency vault.
t Fire Protection System in Diesel Generator Rooms
'Each
-The sprinkler system for the diesel generator rooms was discussed.
room has a sep'arate. sprinkler manifold supplied by a remotely (Control Room) operated water supply valve. The supply valve is located outside
'the room and the sprinkler manifold inside the room is equipped with conventional fusible sprinkler heads to prevent inadvertent flooding of
-the space if the supply valve is opened by mistake.
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. Steam and Feedwater Lines A walk-through review was made of the arrangement of the ste.a and feed-water lines outside containment considering the ef fects of failure of The main steam lines leave containment end. pass through the such pipes.
Auxiliary Building high above the new fuel and epent fuel storage areas.
The horizontal runs are mounted on shelf-like supports with the line f rom the south generator (24A) running low and closer to the Reactor Building and the line f rom the north generator (24B) above and set The offset of the two f arther away from the Reactor Building surf ace.
lines permits-installation of all the steam relief valves on the top of the steam lines with all of the relief valve exhaust stacks venting up Where the 36-inch main steam lines run out through through the roof.
the side of the Auxiliary Building, they both turn downward outside; the north line takes a slight turn so that it runs directly above the south The large main steam block valves are line before it turns downward.
located in the vertical runs of the steam lines just below the downward turn; the steam lines are seismic Category I up to and including these valves.
The 24-inch main feedwater lines both enter the Reactor Building through the south penetration room; the lines are about 40 feet apart in that room and the maximum length of the run in the room appears to be no more than 40 feet. The emergency feedwater line for the south generator is in this room, about 10 f eet below the main feedwater line for the same The emergency feedwater line for the north generator is in generator.
the other penetration room.
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AGENDA FOR SITE VISIT _
ARKANSAS NUCLEAR ONE - UNIT 1 i
November 29-30, 1972 1.
Control Room General layout Nuclear & Reactor Protection instrument arrangement & layout a.
b.
Rod position indication Protection System initiation & bypass switch arrangements c.
d.
Diesel control board Cabling in control room (separation, loading, etc.)
e.
f.
Radiation monitoring g.
2.
Cable runs & cable spreading area a.
General layout b.
Degree of separation Diverse wiring c.
Tray or wireway density (pa: king) d.
Fire detection & protection e.
3.
Switchgear Rooms General layout Physical & electrical separation of redundant units a.
b.
Potential for damage due to fire, missiles, etc.
c.
d.
Cable installation Fire detection & protection e.
4.
Battery Installations a.
General layout Physical & electrical separation b.
Potential for damage due to fire, missiles, etc.
c.
d.
Fire detection & protection 5.
Diesel Generators a.
General layout b.
Physical & electrical separation c.
Fuel supply system d.
Fire detection & protection
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Reactor _ Building & Auxiliary Building Protection system instrument arrangement & layout a.
Potential for instrument damage due to_ fire, missiles, etc.
b.
instruments Separation of piping & wiring to redundant c.
d.
Provision for testing protection instruments 4
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f ATTENDANCE l
AP&L/BECHTEL/B&W/AEC MEETINGS NOVEMBER 29. 30 & DECEMBER 1,1972_
DATES PRESENT_
TITLE _
COMPANY _
NAME All Project Manager All AP&L Asst. Engineer All W. Cavanaugh D. Ructer Asst. Engineer 29-30 J. Grisham QA Electrical 29 E, Quattlebaum Production Engineer 29-30 J. McAlister Reactor Technician 29-30
- P. Almond Chief QA 30 Asst. Chief Engineer
- N. Moore Manager, Substation Design 30 i
- R. Toler 30
- S. Gritmett Project Supervisor 29-30
- W. Houston Project Engineer All Bechtel Licensing Engineer All C. Katanics J. Oszewski Elect. Group Supv.
All J. Haidinger Senior Elect. Engr.
29-30 l
W. Kunz instrumentation Leader All F. Silberman Assoc. Project Manager 29-30 B&W H. Baker System Engineer 29-30
- A. Lloyd System Engineer A1.
- C. Strempke Project Manager 29-30 AEC R.~Bernero Elect. & Control Sys.
All T. Ippolito Elect. & Control Sys.
29-30
- J. Calvo Principal Reactor Insp.
- V. Brownlee
- Part-time on dates indicated
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