ML20113D541
| ML20113D541 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/10/1985 |
| From: | Mcdonald R ALABAMA POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 TAC-51088, TAC-51089, NUDOCS 8504150353 | |
| Download: ML20113D541 (13) | |
Text
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Mailing Address i
Alibimt Power CompIny
'g 600 North 18th Streat Post Office Box 2641 o
Birmingham. Alabama 35291 Telephone 205 783-6090 R. P. McDenald Senior Vice President Flintridge Building
/\\labama POWCf April 10,1985 m
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Docket Nos. 50-3%
50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Connission Washington, D. C.
20555 Attention:
Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Units 1 and 2 Regulatory Guide 1.97 Compliance Gentlemen:
Alabama Power Company provided to the NRC on March 30, 1984 for Unit 2 and June 29, 1984 for Unit 1 Compliance Reports regarding the provisions of Regulatory Guide 1.97.
On February 7,1985, the NRC issued the " Interim Regulatory Guide 1.97 Report, Joseph M. Farley Nuclear Plant Units 1 and 2" in which additional information was requested. provides the requested information.
In the March 30,1984 Unit 2 Compliance Report, Variables 19/32 and 19/33 (Power Supply) were stated to be in compliance with the provisions of Regulatory Guide 1.97. A subsequent review of the compliance of these variables determined a more appropriate listing of these variables as not in compliance but justified. Attachment 2 provides revised pages for Variables 19/32 and 19/33 for Unit 2.
These revised pages have already been incorporated into the June 29, 1984 Unit 1 Compliance Report.
If you have any questions, please adviso.
Yours truly,
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1 7
R. P. Mcdonald '
RPM / JAR:gri-D42 Attachments cc:
See page 2 415 53 850410 I
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A K 05000348 F
Mr. S. A. Varga April 10, 1985 U. S. h. clear Regulatory Comunission Page 2 cc: Mr. L. B. Long Dr. J. N. Grace Mr. E. A. Reeves Mr. W. H. Bradford 1
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r-Response to NRC Request for Additional Information -
NRC Section 3.3.2:
Seismic Qualification Requirement Deviation (Unit No. 2)--Results of the ongoing qualification program to verify the structural adequacy of the main control board and the main control board termination cabinets, is needed from the licensee. The licensee should commit to making any changes that are identified as necessary, by the results of this evaluation, to meet Regulatory Guide 1.97 requirements.
I APCo Response:
A seismic qualification program to verify the seismic structural adequacy of the main control board (MCB) and the Regulatory Guide (R.G.) 1.97 display devices mounted on the MCB has been completed.
This seismic qualification program concluded that the structure of the MCB is seismically qualified for use in Farley Nuclear Plant Unit 2 in accordance with the provisions of IEEE 344-1971 as described in FSAR Chapter 3.10.
The seismic qualification program, however, did identify some devices on the MCB whose mounting does not meet seismic mounting requirements and some devices which must be replaced.
A R.G.1.97 variable which has a display on the MCB is not considered to be seismically qualified unless the display device is seismically mounted and seismically qualified and all other display devices on the referenced section of the board are seismically mounted. Display devices other than the R.G.1.97 variable displays are considered to require seismic mounting because the mounting of these devices could fail during a seismic event and disable a R.G.1.97 display.
Identified below are the devices on each section of the MCB which will be modified or replaced to bring that section of the MCB into compliance with IEEE 344-1971 as described in Farley Nuclear Plant FSAR Chapter 3.10.
Devices which are R.G.1.97 displays are so identified by their respective variable number (s).
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Response to NRC Request for Ariditional Information -
Regulatory Guide 1.97 Page 2 APCo Response (continued):
MCB MCB R.G. 1.97 Section Item No.
Description Action Variable No.
NGMCB2500A-AB Al 124 MLB-4 Modify
- A2 23 NR-46 Modify
- A2 24 NR-47 Modify
- A2 128 CPLB Modify
- 19/20,19/21 A2 135 PR-950 Replace 16 A2 137 LR-35948 Replace 10 A3 165 TSLB-1 Modify
- A3 65 TSLB-2 Modify
- A3 66 TSLB-3 Modify
- NGMCB2500B-AB B1 55 FR-478 Modify
- B1 56 FR-488 Modtfy*
81 SB LR-476 Modify
- 81 59 LR-477 Replace 4
NGMCB2500C-AB C
11 LI-459A, LI-460 Modify
- C 40 TI-451. TI-452 Modify
- C 50 TR-413 Pep 1 ace 2
C 53 LR-459 Modify
- C 54 PR-444 Modify
- C 55 TR-410 Replace 3
C 61 RPI Modify
- 1009 C
62 YN-4056A Modify *
- Mounting Modifications I
Response to NRC Request for Additional Information -
l Regulatory Gaide 1.97 l
Page 3 i
APCo Response (continued):
A seismic qualification program to verify the seismic structural adequacy of the main control board termination cabinets has also been completed. The seismic qualification program concluded that additional fasteners for the internal device mounting panel are required to bring the cabinets into compliance with the seismic requirements of Farley Nuclear Plant Unit 2.
The modifications necessary to bring the main control board termination cabinets into compliance with tha seismic qualification provisions of IEEE 344-1971 as described in FSAR Chapter 3.10 will be performed.
2.
NRC Section 3.3.3:
Neutron Flux (intermediate range)--The licensee should commit to providing redundant power sources for this variable.
APCo Response:
The means available to Alabama Power Company for providing redundant power sources for this variable have been evaluated and it has been determined that two primary options exist. The first option involves changing the power supply for one of the neutron flux instrumentation loops. This option would require a complete rework of the loop i
including cabling and raceways and components which would comprise the instrument loop in order to match the train orientation of the power supply.
In addition, realignment of the neutron flux instrumentation loop NE35 or NE36 would result in a major change to the original design of the nuclear instrumentation and reactor protection systems.
The new neutron flux monitoring loop was designed to serve not only as a means of resolving an ambiguity between the two existing loops but also as a response to the requirements of 10CFR50 Appendix R.
Realignment of this loop would have a significant impact on the fire protection compliance program. Furthermore, custom fabricated cables, containment penetrations and raceway would have to be replaced and relocated at considerable expense. The other option would be the installation of a new, completely redundant instrument loop aligned to Train B.
The estimated cost for such a new loop would be approximately $140,000 per unit.
Response to NRC Request for Additional Information -
Regulatory Guide 1.97 Page 4 APCo Response (continued):
While a complete electrical ( AC and DC) Train A failure as currently configured would disable all intermediate range neutron flux instrumentation, the probability of such a fault is very low and does not justify the considerable expense involved in pursuing either of the above two options.
It should be emphasized that the only identified failure that would disable the neutron flux instrument is a complete loss of Train A AC and DC power.
Single failures such as inverter failures, battery failures, battery charger failures, 4160 volt bus fault or DC bus fault would not disable the neutron flux instrument. Also, as stated in the previous Compliance Report submittals of March 30 and June 29, 1984, it is the opinion of Alabama Power Company that the imposition of instrumentation design criteria for accident monitoring which is more stringent than the instrumentation design criteria for accident mitigation is not justified.
3.
NRC Section 3.3.16:
Pressurizer Heater Status--The licensee should install electric current instrumentation in accordance with the regulatcry guide recommendations.
APCo Response:
Alabama Power Company participated in the Westinghouse Owners Group (WOG) effort which developed predefined symptom-based event-related recovery strategies for responding to emergency transients. This effort was part of a multi-utility response to the provisions of NUREG-0737, Item I.C.1, Short-term Accident and Procedures Review, in which various transients and accidents were reviewed and response procedures developed. The predefined recovery strategies provide guidance to recover the plant to a normal operational state or a known safe state from which repair, if required, can be accomplished. The recovery strategies for each transient or accident were integrated with the other transient and accident strategies to ensure that the overall response plan was appropriate. Response strategies for the Revision 1 of the Emergency Response Guidelines were developed by the WOG and have been validated and are approved for implementation by the NRC.
These strategies utilize pressurizer breaker status indication to provide the status of the pressurizer heaters.
Response to NRC Request for Additional Information -
Regulatory Guide 1.97 Page 5 APCo Response (continued):
The Emergency Response Procedures in use at Farley Nuclear Plant were developed from the previously discussed WOG effort and do not utilize pressurizer heater current for accident mitigation. The status of the pressurizer heaters can be adequately determined using a combination of pressurizer heater breaker position and pressurizer pressure. The addition of current instrumentation is therefore not necessary at the Farley Nuclear Plant to meet the provisions of Regulatory Guide 1.97.
Alabama Power Company responded to NUREG-0737, Item II.E.3.1 by letter of January 14, 1981 in which the capability for emergency power supply for pressurizer heaters was discussed.
4.
NRC Section 3.3.19:
Steam Generator Pressure--The licensee should clarify the range and the extent of the deviation, identify the lowest safety valve actuation setpoint and provide any additional justification deemed necessary.
APCo Response:
R.G.1.97 states that the range guideline for Steam Generator Pressure is from atmospheric pressure to 20% above the lowest safety valve setting.
The lowest Steam Generator safety valve setting at Farley Nuclear Plant is 1075 psig; therefore, the R.G.1.97 guideline for Farley Nuclear Plant is 0-1290 psig.
The range of the existing instrumentation is 0-1200 psig which is 90 psig below that of the R.G. 1.97 guideline. Alabama Power Company feels that the 0-1200 psig range is acceptable since the highest actuation setpoint of the Main Steam Safety valves is 1129 psig.
Allowing for 3% accumulation above this actuation point, the maximum credible steam line pressure is 1163 psig. This pressure is within the indicated range of the existing instrumentation.
The confusion generated by the Alabama Power Company original response stems from FSAR Table 7.5-1 which states that Steamline Pressure Range is 0-1300 psig.
This FSAR Table is currently being revised to correctly specify Steamline Pressure Range as 0-1200 psig.
5.
N,RC Section 3.3.23:
Radiation Exposure Rate--The licenspe shou}d install Category 2 instrumentation with a range of 10-' to 10 R/hr in areas where access is required to service equipment important to safety; the licensee should identify the range of the rate monitor in the control room.
Response to NRC Request for Additional Information -
Regulatory Guide 1.97 Page 6 APCo Respense:
R.G.1.97 states that the instrumentation for the radiation exposure rate inside buildings or areas where access is required to service equipment important to safety must be able to detect significant releases, provide for long-term surveillance and provide for release assessment. The instrumentation currently installed at Farley Nuclear Plant provides for this.
In response to NUREG-0737. Item II.B.2, Alabama Power Co,npany conducted a shielding design review to ensure adequate access to vital areas and protection of safety-related equipment by design changes, increased permanent or temporary shielding, or post-accident procedural controls. All modifications resulting from this review have been completed. Therefore, the portable monitoring instrumentation described is gufficignt. The range of the Control Room Radiation Monitor is 10- to 10 R/hr.
JAR /gri-D42 l
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Revised Pages, Unit 2 Compliance Report 1
Variable Pages 19/32.0-1 19/32.3-1 19/33.0-1 19/33.3-1
REGULATORY GUIDE 1.97 CATEGORY 1 COMPLIANCE CPORT VARIABLE 19/32: CONTAINENT ISOLATION VALVE STATUS PENETRATION No. 71 - LEAK RATE TEST TPNS No(s) - INSIDE: Blind Flange 0UTSIDE: Q2P23ZS3238-N
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EETS RESOLUTIONS TO GUIDELINES R.G. 1.97 NONCOMPLIANCES 1.
QUALIFICATION a)
ENVIR0841 ENTAL NO JUSTIFY b)
SEISMIC NO JUSTIFY 2.
REDUNDANCY YES 3.
POWER SUPPLY NO 4.
CHANNEL AVAILABILITY N/A 5.
QUALITY ASSURANCE YES 6.
DISPLAY AND RECORDING a)
DISPLAY YES b)
RECORDING YES 7.
RANGE YES 8.
EQUIPMENT IDENTIFICATION NO JUSTIFY 9.
INTERFACES (isolation)
YES
- 10. SERVICING, TESTING, CALIBRATION YES
- 11. HUMAN FACTORS YES
- 12. DIRECT MEASUREMENT YES 19/32.0-1 Rev. 1
l 3.
P0ER SUPPLY YARI ABLE 19/32: CONTAINENT ISOLATION VALVE STATUS PENETRATION No. 71
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EXISTING CONDITION
'The position indicating circuit for valve MOV3238-N is powered from an MCC which is not tavided with onsite standby power from a diesel generator and is not backed-up by battery.
JUSTIFICATION
- i.,
Power for the valve position indicating circuit is derived from the same A
. source as the operating power for the valve.
During a loss of power, the
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valve position indication would be disabled, but at the same time, the valve would not be functional.
The motor operated isolation valve for this penetration is nomally closed and would remain closed in an accident condition.
The penetration is used only during integrated leak-rate testing and this valve would not be opened during normal power operation. The penetration is also flanged-off inside the containment by a blind flange.
If the power to the valve were to fail, the resulting loss of position indication would be recognized, and therefore, the operator would not be led to defeat or fail to accomplish a required safety function.
In addition, the blind flange inside containment would maintain containment penetration isolation in the event of valve failure.
19/32.3-1 Rev. 1
r REGULATORY GUIDE 1.97 CATEGORY 1 COMPLIANCE REPORT VARIABLE 19/33: CONTAINENT ISOLATION VALVE STATUS PENETRATION No. 72 - LEAK RATE TEST TPNS No(s) - INSIDE: BLIND FLANGE OUTSIDE: Q2P23ZS3239-N EETS RESOLUTIONS TO GUIDELINES R.G. 1.97 NONCOMPLIANCES 1.
QUALIFICATION a)
ENVIROM1 ENTAL NO JUSTIFY b)
SEISMIC NO JUSTIFY 2.
REDUNDANCY YES k
3.
POWER SUPPLY NO 4.
CHANNEL AVAILABILITY N/A S.
QUALITY ASSURANCE YES 6.
DISPLAY AND RECORDING a)
DISPLAY YES b)
RECORDING YES 7.
RANGE YES 8.
EQUIPMENT IDENTIFICATION NO JUSTIFY 9.
INTERFACES (isolation)
YES 10.
SERVICING, TESTING, CALIBRATION YES 11.
HUMAN FACTORS YES 12.
DIRECT MEASUREMENT YES 19/33.0-1 Rev.1
r 3.
POWER SUPPLY VARIABLE 19/33: CONTAINENT ISOLATION VALVE STATUS PENETRATION No. 72 EXISTING CONDITION
'The position indicatipg circuit for valve MOV3239-N is powered from an MCC which is not provided with onsite standby power from a diesel generator and is not backed-up by battery.
JUSTIFICATION
- i.,
Power for the valve position indicating circuit is derived from the same
. source as the operating power for the valve.
During a loss of power, the
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' valve position indication would be disabled, but at the same time, the valve would not be functional.
The motor operated iso 16 tion valve for this penetration is normally closed and would remain closed in an accident condition.
The penetration is used only during integrated leak-rate testing and this valve would not be opened during normal power operatiun.
i the containment by a blind flange.The penetration is also flanged-off inside 1
the resulting loss of position indication would be recognized, andIf the p therefore, the operator would not' be led to defeat or fail to accomplish a required safety function.
In addition, the blind flange inside containment would maintain containment penetration isolation in the event of valve i
failure.
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19/33.3-1 Rev. 1 L._
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