ML20008D830
| ML20008D830 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/29/1980 |
| From: | Kintner L Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8010230086 | |
| Download: ML20008D830 (19) | |
Text
SY no y
?g UNITED STATES y') -
g NUCLEAR REGULATORY COMMISSION 5
C W ASHING TON, D. C. 20555 s
/
SEP 2 91980 Docket No. 50-364 APPLICANT: Alabama Power Company FACILITY:
Joseph M. Farley Nuclear Plant, Unit 2
SUBJECT:
SUMMARY
OF SEPTEMBER 18, 1980 MEETING REGARDING REVIEW 0F OPERATING LICENSE APPLICATION The purpose of the meeting was to hear and discuss applicant's proposed response to several staff positions in the review for a full power operating license. is a list of attendeas.
1.
Containment Purge Position Applicant proposed a long-term solution to staff's position on containment venting and purging - to keep purging to less than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year per reactor.
For Farley 2, this position cannot be met with presently-installed equipment because (1) pressure buildup inside containment is about one pound per square inch (psi) per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> due to venting of air-operated valves inside containment, ano (2) purging for access to perform periodic surveillance and main-tenance requires at least 270 hours0.00313 days <br />0.075 hours <br />4.464286e-4 weeks <br />1.02735e-4 months <br /> per year to achieve ALARA conditions.
The present vent and purge system has 48-inch butterfly valves and 18-inch butterfly bypass valves, with ducts and filters designed for a few inches of water differ-ential pressure.
The present system cannot be operated by allowing containment pressure building to 1 psi and then venting becase the greater-than-design pressure would damage ducts and filters.
The applicant has proposed to add a 3-inch bypass line to allow plant operation with the containment closed part of the time. Containment pressure would be allowed to build up to about 1 psi in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and then it would be vented to atmospheric pressure in about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> - thus having venting for 1/3 of the time ( 2920 hours0.0338 days <br />0.811 hours <br />0.00483 weeks <br />0.00111 months <br /> per year). Three containment isolation valves (globe seats) wouH be included in eacn bypass line so that in the event of an accident, the line would be isolated.
Space between the valves would be vented to the penetration room so that any leakage during the accident would be filtered.
The 3-inch, hard-seat, valves are more reliable and leak-tight than the present 18-inch butterfly valves.
Applicant is providing information to the staff to demonstrate operability of the 48-inch and 18-inch purge valves, which are used to reduce radioactivity due to noble gases inside containment prior to entry for periodic surveillance and maintenance. The minimum purge time, using 48-inch valves is 270 hours0.00313 days <br />0.075 hours <br />4.464286e-4 weeks <br />1.02735e-4 months <br /> per year. Use of the 18-inch valves would require a much longer purge time. The Farley Plant has separate charcoal filters to remove iodine by recirculation of containment air prior to entry.
&O10 @
. ~
2-i The Staff in Containment Systems Branch said that applicant's proposal did not meet staff's position for purging or venting no more than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year. Applicant' requested an appeal to higher levels of management.
In the afternoon, applicant appealed to Mr. W. Butler, Chief. Containment Systems Branch and Mr. L. Rubenstein, AD for Core and Containment Systems.
4 After hearing applicant's proposal, Mr. Butler asked if alternative systems had been considered, for example:
(1) Use of containment air to supply air for the instrument air system for valves inside containment thus eliminating containment air pressure buildup from venting valves.
(2) Use of throttling valves in the present 48-inch purge system to protect filters from overpressurization.
l Use of a storage tank and air corpresson inside containment to stora (3) excess air from valves at a high pressure and then release it to a tank outside containment in a short time interval.
Applicant said it had considered the first alternative and that space for the 4
system near or in containment was a main deterrant. Use of the second alter-native was also precluded because of space requirements for the throttling l
valves (25 feet long and 10 feet high). The third alternative is similar to 2
those used in smaller reactors and had not been sized for Farley Plant.
Applicant and staff agreed that the applicant's proposal and an evaluation of the above three alternatives for a long term solution should be docketed for review by the staff.
2.
Overcurrent Protection for Electric Cable Penetrations Through Containment 3
Applicant has agreed to provide a preliminary design and schedule for installa-j tion of any modifications needed to meet staff's position by October 20, 1980.
4 The NRC Licensing Project Manager.for Farley-2 stated that this date was not
~
compatible with our prcspective decision date for full power license (December 1),
i for which a response from applicant was scheduled for the week of September 22.
Mr. Oliver ringsley, Manager of Nuclear Engineering Technical Support for APCo, advised that the fuel loading date and power escalation date have slipped two weeks.
Fuel load is now scheduled for October 15 instead of September 29 and power escalation is December 15 instead of December 1.
On the basis of appli-4 cant's slipped dated for startup, the LPM said the October 20 date was accept-able for a response, if it is substantially complete. The scheduled date for SE input for this item will be' the week of October 27. While applicant's response has slipped four weeks, its startup schedule has slipped only two weeks. The LPM said that.he could accommodate a few items coming in after scheduled response due date but that too many items slipping could cause the prospective decision date to be later than applicant's scheduled power escalation date.
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_2
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l 3.
Alternate Shutdown Panel The staff in Power Systems Branch gave the status and tentative position for the alternate shutdown panel to be used to shutdown the plant in the event fire in the cable spreading room or control room damages controls and instruments in the control room.
A key element in staff's position is that there should be the capability to maintain hot shutdown from the panel sending people into the plant to control systems locally.
The applicant has studied seven systems it considers necessary to bring the plant to a cold shutdown, and has provided the results of these studies to the staff.
Applicant studies are based on the criteria that operator actions (locally in the plant) are pennissible, if there is a reasonable amount of time to perform them.
Further applicant's studies are based on proceeding through hot shutdown to cold shutdown, without a requirement to maintain hot shutdown for a specified time.
The staff discussed two apparent problems in the present arrangement for l
alternate shutdown panel:
(1) The fire may short power supplies.o many of the required system pumps and valves, blowing fuses or opening circuit breakers. The replacement of the fuses or closing circuit breakers cannot reasonably be expected
~
to be done on all required systems in the required time.
The use of a second set of fuses or circuit breakers in required systems would elimi-nate this problem.
s (2) The valves and pumps in required systems cannot be controlled from the alternate shutdown panel but must be monitored and controlled locally.
The addition of instruments and controls to the panel sufficient to main-tain hot shutdown with proper separation from control room panel circuits could mitigate this problem.
The staff agreed that local operation of equipment for bringing the plant to a cold shutdown is a reasonable design criteria.
The staff in Power Systems Branch and applicant's representati /es discussed the alternate shutdown panel in a separate meeting. Additional information required for this review will be sent by the app!. cant during the week of September 22.
4.
Completion of Construction Mr. R. Wessman, Specialist in Field Coordination for the Office of Inspection and Enforcement, brought to applicant's attention several matters that would facili-tate 18E's verification that safety systems have been properly installed and tested and that procedures are ready to be used. He called particular attention to equipment and procedures required to be used by the Physical Security Plan, waste processing system, fire protection system, and various administrative and plant procedures required for safey.
He asked'that the Resident Inspector, W. Bradford, be kept informed on all matters relevent to I&E inspections and
< verifications that must be completed prior to issuing a fuel loading and low power testing license.
5.
Intersystem Leakage of Valves that Separate High Pressure Systems from -
Low Pressure Systems Staff in Reactor Systems Branch summarized its position in SRP 3.9.6 that requires a leakage test of each valve separating a high pressure system from a low pressure system after each disturbance of -the valve.
By disturbance is meant, for example, the opening of check valves where two check valves in series are used to contain high pressure.
Applicant said its approach to meeting staff's position is to use valve reliability t_o determine frequency of tests, consider the use of extra valves in the system, and thorough inspection of high pressure piping. is a copy of a meeting handout describing its approach to valve tests.
6.
Consideration of LOCA During Plant Startup and Shutdown During plant startup and shutdown, ECCS systems are necessarily prevented from operating. By letter dated August 25, 1980, staff has requested applicant to describe precredures for mitigating -the consequences of a LOCA, is it were to occur during plant startup or shutdown.
Applicant provided a draft response during the meeting.
It is expected to be l
similar to those provided for other Westinghouse plants (e.g., North Anna 2) l 7.
Containment Sump Performance Staff in Reactor Systems Branch gave the status of its review of applicant's response that provides a description of instrumentation to detect excessively low ECCS flow from the containment sump following LOCA, and means for ccrrecting the condition. Additional information is needed to assure that the emergency l
operations director will be made aware of the condition.
l Three alternatives were discussed:
(1) Procedures requiring that flow be monitored periodically and recorded in a 1og (hourly).
(2) An alarm on low flow.
(3) A person in the control -room who is dedicated to monitor the flow.
j 8.
Modification of Power Supplies to Solenoids for Auxiliary Feedwater Flow Control Valves Staff in Power Systems Branch discussed applicant's proposed modifications (letter dated September 8,1980).
Staff expressed concern that the modified design would not permit the operator to modulate the flow control valves -
v
. therefore feedwater would be full flow or off.
The resulting operation of feedwater may not be desirable. Staff requested (and applicant provided September 19) the schematic diagrams for the modifications.
k Le ter L.
intner,-Project Manager Licensing Branch No. 2 Division of Licensing
Enclosure:
As stated cc w/ enclosure:
See next page I
Mr..F. L. Clayton, Jr., Senior Vice President Alabama Power Company Post Office Box 2641 Birmingham, Alabama 35291 cc:
Mr. Alan R. Barton Executive Vice President Alabama Power Company-Post Office Box 2641-Birmingham, Alabama 35291 Mr. Ruble A. Thomas Vice President Southern Company-Services, Inc.
Post Office Box 2625 Birmingham, Alabama 35202 Mr. George F. Trowbridge Shaw,.Pittman, Potts and Trowbridge 1800 f1 Street, N. W.
Washington, D. C.
20036 Mr. W. Bradford NRC Resident Inspector P. O. Box 1814 Dothan, Alabama 36302 l
l' E
I l-I
MEETING
SUMMARY
DISTRIBUTION Docket File D. Muller NRC PDR
- R. Ballard Local PDR W. Regan NSIC D. Ross TIC P. Check TERA R. Satterfield NRR Reading
- 0. Parr-LB #2 File F. Rosa H. Danton W. Butler E. Case W. Kreger D. Eisenhut R. Houston R. Purple T. Murphy B. J. Youngblood L.'Rubenstein A. Schwencer T. Speis F. Miraglia W. Johnston J. Miller J. Stolz G. Lainas S. Hanauer R. Vollmer W. Gammill J. P. Knight F. Schroeder R. Bosnak D. Skovholt F. Schauer M. Ernst R. E. Jackson R. Baer LKintner Project Manager C. Berlinger Licensing Assistant-MService K. Kniel Attorney, OELD G. Knighton I&E (3)
A. Thadani ACRS (16)
D. Tondi R. Tedesco G. Lear V. Noonan-S. Pawlicki V. Benaroya Z. Rosztoczy W. Haass 5
NRC
Participants:
Others:
'BCC:
Applicant & Service List i
I p...
...e.'..-w....-
.. - +
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- + - * - - - *..
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ENCLOSURE 1-ATTENDANCE LIST SEPTEMBER 18, 1980 MEETING WITH NRC - APCo
-NAME ORGANIZATION
NRC/DSI/CSB 1
l
-*W. J. Shapaker NRC/DSI/CSB B. D. McKinney, Jr.
. Alabama Power Company l
- 0. D. Kingsley, Jr.
Alabama Power Company Ozen Batum Southern Company Services, Inc.
- R. L. George Alabama Power Company
- A. A. Vizzi Bechtel J. H. Bell Bechtel
- B. J. George Southern Company Services, Inc.
l
- E. R. Burns Southern Company Services, Inc.
l
- L. E. Conway Westinghouse l
- D. B. Pierce Westinghouse
- Bob Hollands Alabama Power Company
- Jessie Love Bechtel Kenneth W. McCracken Alabama Power Company i
l-L. L. Kintner LB#2/ DOL /NRC
- R.
Fitzpatrick PSB/NRC
- 0.
Chopra
-PSB/NRC l
- A.
Ungara PSB/NRC
- R.
Wessman I&E/NRC
- J.
Knox PSB/NRC
- A. Capucci MEB/NRC-
- M. Mendonca RSB/NRC
'*F. Orr RSB/NRC
- T.
Dunning ICSB/NRC
- Part Time
ENCLOSURE 2 HAND 0UT - SEPTEMBER 18,1980 f1EETING INTERSYSTEM LEAKAGE COMMENTS I.
Valves which protect low-pressure systems from Reactor Coolant System pressure should be considered pressure-isolation valves.
The leak-tightness of these valves should be ensured by periodic leak testing to prevent exceeding the design pressure of the low-pressure systems thus preventing an intersystem LOCA.
The periodic leak testing of pressure-isolation valves should be performed during each refueling outage as the plant is coming up in power.
II.
Pressure isolation valves should be included in the Pump and Valve Program as Category A or AC per IWV-2000 of Section XI of the ASME Code, 1974 Edition with Addenda through Summer 1975. These valves should meet the leak rate test requirements of IWV-3420 of Section XI with the exception that the maximum permissible leakage rate shall be 1 GPM per valve.
III.
The NRC presently has a proposed change to the plant Technical Specifi-cations regarding pressure-isolation valves.
This change is satisfac-tory except for Surveillance Pequirement 4.4.7.2(d) which requires that the valves be leak tested each time they are disturbed.
This requirement should be deleted.
IV.
The Clessi/ Class 2 boundary should not be considered the isolation point to be protected by redundant isolation valves and by periodic leak testing.
The Class 2 high-pressure piping does not need protec-tion.
The proper isolation point is the high pressure / low pressure boundary. This is the underlying assumption in both the WASII-1400 Report and the EPRI Interfacing-Systems LOCA Report. See paragraph VIII.
l l
V.
In cases where pressure isolation is provided by two valves, both l
will be independently leak tested.
When three or more valves provide l
isolation, three of the valves need to be leak tested.
By testing three valves in series, the probabilistic f ailure rate decreases by two orders of magnitude.
A-
I S
_ - 1 _ _:-'
SimWEF VI.
The pressure-isolation valves to be leak tested during each refueling outage with their estimated probability of' failure are listed below:
Valve Number NUREG 0677 Probability Q2E11V051A Q2E11V051B Q2E11V051C Q2E11V021A 51"e hv
///f Q2E11V021B Q2E11V021C Q2E11V042A Q2E11V042B (continued)
Valve Number NUREG-0677 Probability Q2E21V077A Q2E21V077B e '"
Q2E217076A 2
/8#YY '
Q2E21V076B Q2E11V044*
6 Q2E11V016A**
Q2E11V001A**
M Q2E11V016B**
Q2E11V001B**
g=g Total Probability I.b
- See ISI Relief Request 3.1.17 for more information.
- See ISI Relief Request 3.1.8 for more information.
VII.
The following valves are Class 1 and form a Class 1/ Class 2 boundary.
However, they do not perform a pressure-isolation function and will not be leak tested:
Q2E21V062A Q2E21V062B Q2E21V062C These valves are the second valves off the RCS loops and are part of l
a series of five (5) normally-closed high-pressure valves.
The prob-ability of five valscc'failing simultaneously is so small as to be disregarded. Leak testing these va'.ves will not significantly increase the icve.1 of plant safety.
Q2E21V066A Q2E21V066B Q2E21V066C l
These valves are the second valves off the RCS loops and are part of a series of four (4) normally-closed high-pressure valves.
The probabil-ity of four valves failing simultaneously is so small as to be disregarded.
Leak testing these valves will not significantly increase the level of plant safety. In addition, V066A, B, C, and V062A, B, C, discharge into V051A, B, C, which will be leak tested each refueling outage.
Q2E21V077C This valve is the first valve off the RCS loop but is a part of a series of four (4) normally-closed high-pressur.e valves.
Q2E21V078A Q2E21V079A Q2E21V078B Q2E21V079B Q2E21V078C Q2E21V079C These valves are the second valves off the RCS loops and are a part of a series of four (4) normally-closed high-pressure valves.
17 In addition, all of these thirteen (13). valves are limited in the amount of reverse leakage that they can pass.
In line with these valves are locked, throttled globe valves whose maximun flow rate (193 CPM) is controlled by the plant Technical Specifications Surveillance Require-ment 4.5.2(1) on page 3/4 5-6.
Also, NUREG-0677 states on page 21 that at Sequoyah Nuclear Plant the 9
charging and boron injection systems were excluded because they were rated at reactor design pressure.
Farley Unit 2 has a similar design.
NUREG-0677 also states, "The accumulators were excluded because a break in those systems would not esult in coolant being lost outside of con ta inment ".
This statement applies to Farley Unit 2.
Therefore, the following accumulator discharge check valves are not considered pressure-isolation valves:
Q2E21V032A Q2E21V032B Q2E21V032C Q2E21V037A Q2E21V037B Q2E21V037C However, these valves are leak tested to verify closure at each refueling outage as the plant cones up in power.
See ISI Relief Requests 3.1.15 and 3.1.34.
VIII.
In the e= 2nt that the Class 1 valves should fail and the high-pressure Class 2 piping is subjected to RCS pressure, a LOCA would not result.
The high-pressure Class 2 piping is rated at the same pressure as reactor coolant piping.
The ASME Code Section III requires that Class 2 piping welds receive a radiograph and dye penetrant examination to ensure structural integrity.
In addition, the piping receives a hydrostatic pressure test, a visual examination for leakage, and a visual inspection of support capability of all hangers and supports.
The preservice inspection of Class 2 piping was originally planned to be in accordance with the 1971 Edition of Section XI with Addenda through Winter 1972.
The PSI program was voluntarily upgraded to the 1974 Edition with Addenda through Summer 1975.
The NRC has approved the Farley Unit 2 preservice inspection program and has stated that the PSI is in compliance with the ASME Code and with 10 CFR 50.55a(g)(2).
The Unit 2 inservice inspection program is essentially identical to the Unit 1 ISI program which has also been approved by the NRC.
The PSI and ISI programs exist for'the purpose of ensuring that the systems function safely over the 40-year life of the plant.
4
J. M. FARLEY NUCLEAR PLANT UNIT 2 RCS PRESSURE-ISOLATION VALVES i
Intersystem LOCA Probability Flow Path:
Flow Path:
Total. System:
Number of Type of Original -
Revised -
- Revised System Flow Pathn.,
Interface No Testing Leak Testing Probability Residual 2
Two 1.8 x 10
- 4.2 x 10'
- 8.4 x 10 '
-4
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. gate valves
-6
-9
-8 Low-Head 6
Three check 3.0 x 10
- 7.4 x 10
- 4.4 x 10 Safety Injection, valves Cold Leg
-4
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Two check valves 2.8 x 10
- 4.8 x 10'
- 9.6 x 10 Safety Injection, and one normally-r closed
~g Hot Leg motor-operated
++ l. S x io
- 3 0 x 10 gate valve 4
- Lenk testing will be performed every refueling on all valves in the " Type of Interface" column.
~ ~ ~
- Leak testing will be performed every refueling and whenever flow occurs through the flow path on all valves in the LHSI Hot Leg.
-6 1.02 x 10 Total Intersystem LOCA Probability or 5.2 x 10-i
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