ML20206R616
| ML20206R616 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/12/1999 |
| From: | Hood D NRC (Affiliation Not Assigned) |
| To: | Mueller J NIAGARA MOHAWK POWER CORP. |
| References | |
| TAC-MA4335, NUDOCS 9901200041 | |
| Download: ML20206R616 (14) | |
Text
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y aug% UNITED STATES d
, s# j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 01 I
% +o January 12, 1999 Mr. John H. Mueller Chief Nuclear Officer Niagara Mohawk Power Corporation !
Nine Mile Point Nuclear Station Operations Building, Second Floor P.O. Box 63 Lycoming, NY 13093 I
SUBJECT:
REVISION TO TECHNICAL SPECIFICATION BASES FOR NINE MILE POINT NUCLEAR STATION, UNIT NO.1 (TAC NO. MA4335)
Dear Mr. Mueller:
By letter dated November 24,1998, Niagara Mohawk Power Corporation (NMPC) submitted revisions to the Technical Specifications (TSs) Bases for Nine Mile Point Nuclear Station, Unit No.1 (NMP1). The revision updates the " Bases for 2.1.1 Fuel Cladding - Safety Limit" and the
" Bases for 3.2.5 and 4.2.5 Reactor Coolant System Leakage Rate" to incorporate design changes affecting the reactor fuel, the reactor vessel water level instrumentation, and the drywell leak detection system. The reactor vessel water level instrumentation revision replaced the statement describing the lowest reactor pressure vessel level that could be monitored with a statement that describes the current capability to monitor water level throughout the entire fuel region. The core spray system flow parameters in the " Bases for 3.3.1 and 4.3.1 Oxygen Concentration" are updated to reflect the results of NMPC's most recent analyses. The Bases for TS 3.2.5,4.2.5,3.3.1, and 4.3.1 also include minor editorial changes, including typographical corrections.
By letter dated June 6,1997, NMPC submitted revisions to NMP1's TS " Bases for 3.4.4 and 4.4.4 Emergency Ventilation System," and " Bases for 3.4.5 and 4.4.5 Control Room Air Treatment System." The revisions involve the collection methods described in these TS Bases (i.e., use of a new canister sample collection method for the Control Room Air Treatment System and use of a slotted-tube collection method in sampling the Emergency Ventilation l
System filter trains). l
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The modified pages for the NMP1 TS Bases are enclosed, for insertion into the Technical i Specifications.
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oJ062 9901200041 990112 PDR ADOCK 05000220 P PDR 4 l "
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January 12, 1999 l
J. Mueller ;
if you have questions regarding this letter, contact me by phone at (301) 415-3049 or by electronic mail at dsh@nrc. gov.
Sincerely, Original signed by: !
l Darl S. Hood, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - l/II l Office of Nuclear Reactor Regulation i l
Docket No. 50-220
Enclosure:
Bases Pages cc w/ encl: See next page DISTRIBUTION: See next page 1
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, DOCUMENT NAME: G:\NMP1\LTRA4335.WPD To rec::ive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure
- "E" = Copy with attachment / enclosure ,"N" = No copy 0FFICE PM:PDI-1 lE LA:PDI4V l D:PDI-1 jff )5 l l l I NAME DHood/rst SLittlY SBajwa . G' DATE W '% 12/II/98 f/ /7/98f
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4 P J. Mueller If you have questions regarding this letter, contact me by phone at (301) 415-3049 or by electronic mail at dsh@nrc. gov.
Sincerely, DJDM Dari S. Hoad, Senior Project Manager Project Dire::torate I-1 Division of Reactor Projects - 1/II :
Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosure:
Bases Pages cc w/ encl: See next page l
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January 12, 1999 l J. Mueller l If you have questions regarding this letter, contact me by phone at (301) 415-3049 or by electronic mail at dsh@nrc. gov.
Sincerely, Original signed by:
Darl S. Hood, Senior Project Manager Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket No. 50-220
Enclosure:
Bases Pages cc w/ encl: See next page 1 l
DISTRIBUTION: See next page l
l DOCUMENT NAME: G:\NMP1\LTRA4335.WPD Tc r:ceive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure ."N" = No copy 0FFICE PM:PDI 1 lE LA:PDI-&V l D:PDI 1 ef/ h l l l NAME DNood/rst SLitt Y SBajwa . G' DATE LV 12/ I'/98 t 1E/ P/90
/g/ff Official Record dopy i
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. DATED: J a n u a ry ~ 12, 1999
' REVISION TO TECHNICAL SPECIFICATION BASES FOR NINE MILE POINT NUCLEAR STATION, UNIT NO.1 M @l37 - I PUBLIC i PDI-1 Reading '
J. Zwolinski
. S. Bajwa l S. Little -
D. Hood
.OGC G. Hill (2), T-5 C3 '
W. Beckner '
'ACRS
' C. Hehl, Region i e
cc: Plant Service list t
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DATED: January 12, 1999 REVISION TO TECHNICAL SPECIFICATION BASES FOR NINE MILE POINT NUCLEAR STATION, UNIT NO.1 Docket File -
PUBLIC PDl-1 Reading J.Zv 4!nski S. Bajwa S. Little D. Hood OGC G. Hill (2), T-5 C3
- w. Beckner ACRS C. Hehl, Region I cc: Plant Service list
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John H. Mueller Nine Mile Point Nuclear Station ,
Niagara Mohawk Power Corporation Unit No.1 cc:
l Regiona! Administrctor, Region I
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U.S. Nuclear Regulatory Commission 475 Allendale Road l King of Prussia, PA 19406 i
Resident inspector U.S. Nuclear Regulatory Commission ,
l P.O. Box 126 l Lycoming, NY 13093 Charles Donaldson, Esquire L Assistant Attorney General New York Department of Law 120 Broadway New York, NY 10271 Mr. Paul D. Eddy State of New York Departmei 'af Public Service Power Division, System Operations l
3 Empire State Plaza '
Albany, NY 12223 i
i Mr. F. William Valentino, President i
~ ~ ~ '
New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 Mark J. Wetterhahn, Esquire Winston & Strawn 1400 L Street, NW Washington, DC 20005-3502 Gary D. Wilson, Esquire -
l Niagara Mohawk Power Corporation
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300 Erie Boulevard West Syracuse, NY 13202 i Supervisor
- Town of Scriba Route 8, Box 382 Oswego, NY 13126 I
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l REVISED TECHNICAL SPECIFICATION BASES PAGE LIST l
Nine Mile Point Nuclear Station. Unit No.1 i
Replace the following pages of the Technical Specifications Bases with the attached pages. j The revised page contains vertical lines in the page margin indicating the areas of change. i l
Remove 1019.t1 l 16 16 i 103 103 104 104 126 126 i
176 176 180 180 l
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Enclosure i
BASES FOR 2.1.1 FUEL CLADDING - SAFETY LIMIT During periods when the reactor is shut down, consideration must also be given to water level requirements, due to the effect of decay :
helt. If reactor water level should drop below the top of the active fuel during this time, the ability to cool the core is reduced. This reduction in core cooling capability could lead to elevated cladding temperatures and clad perforation. The core will be cooled sufficiently to [
prsvent clad melting should the water level be reduced to two-thirds of the core height.
The Fuel Zone Water Level Monitoring System (FZWLMS) instrumentation has an indicated range which allows cont:nuous indication of r: actor water level from below the bottom of the active fuel to above the maximum normal water level. The reactor vessel tap for the low- -
1:w-law water level instrumentation is located approximately 7 feet 11 inches below the minimum normal water level or spproximately 4 feet 6 inches above the top of the active fuel. The low-low-low water level trip point is 6 feet 3 inches (-10 inches indicator scale) below ;
the minimum normal water level (Elevation 302'-9"). The 20 inch difference between the reactor vessel tap and the trip point resulted from 4 gn ev luation of the recommendations contained in General Electric Service information Letter 299 "High Drywell Temperature Effect on i R: actor Vessel Water Level Instrumentation." The low-low-low water level trip point was raised 20 inches to conservatively account for ,
po:sible differences in actual to indicated water level due to potentially high drywell temperatures. The safety limit has been established hers to provide a point which can be monitored and also can provide adequate margin. However, for performing major maintenance as cpecified in Specification 2.1.1.e redundant instrumentation will be provided for monitoring reactor water level below the low-low-low wstir level set point. (For example, by installing temporary instrument lines and reference points to redundant level transmitters so that the rc:ctor water level may be monitored over the required range). In addition, written procedules, which identify all the valves which have the pot:ntial of lowering the water level inadvertently, are established to prevent their operation during the major maintenance which requires the water level to be below the low-low level set point.
The thermal power transient resulting when a scram is accomplished other than by the expected scram signal (e.g., scram from neutron flux following closure of the main turbine stop valves) does not necessarily cause fuel damage. However, for this specification a safety limit vioIItion will be assumed when a scram is only accomplished by means of a backup feature of the plant design. The concept of not cpproaching a safety limit provided scram signals are operable is supported by the extensive plant safety onelysis.
AMENDMENT NO.142 16
4 BASES FOR 3.2.5 AND 4.2.5 REACTOR COOLANT SYSTEM LEAKAGE RATE Allowable leakage rates of coolant from the reactor coolant system have been based on the predicted and experimentally observed behavior ;
cf cracks in pipes and on the ability to makeup coolant system leakage in the event of loss of offsite a-c power. The normally expected background leakage dua to equipment design and the detection capability for determining coolant system leakage were also considered in cstablishing the limits. The behavior of cracks in piping sntems has been experimentally and analytically investigated as part of the USAEC sponsored Reactor Primary Coolant System Rupture Study (the Pipe Rupture Study). Work utilizing the data obtained in this study indicates l that leakage from a crack can be detected before the crack grows to a dangerous or critical size by mechanically or thermally 'mduced cyclic !
loading, or stress corrosion cracking or some other mechanism characterized by gradual crack growth. This evidence suggests that for :
1::akage somewhat greater then the limit specified for unidentified leakage, the probability is small that imperfections or cracks associated [
r with such leakage would grow rapidly. However, the establishment of allowable unidentified leakage greater than that given in 3.2.5 on the basis of the data presently available would be premature because of uncertainties associated with the data. For leakage of the order of 5 .
gpm as specified in 3.2.5, the experimental and analyticat data suggest a reasonable margin of safety that such leakage magnitude would nst result from a crack approaching the critical size for rapid propagation. Leakage of the magnitude specified can be detected reasonably in a matter of a few hours utilizing the available leakage detection schemes, and if the origin cannot be determined in a reasonably short time, the plant should De shut down to allow further investigation and corrective action.
inspection and corrective action is initiated when unidentified leekage increases at a rate in excess of 2 gpm, within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period or ,
1:ss. This minimizes the possibility of excessive propagation of intergranular stress corrosion cracking. ;
A totalleakage of 25 gpm is well within the capacity of the control rod drive system makeup capability (page ill-7 of the First Supplement)*.
As discussed in 3.1.6 above, for leakages within this makeup capability, the core wili remain covered and automatic pressure blowdown wilf ;
nrt be actuated.
4 Leakage is detected by having all unidentified leakage routed to the drywell floor drain tank and identified leakage routed directly to the drywell equipment drain tanks. Identified leakage includes such items as recirculation pump seal leakage.
The primary means of determining the reactor coolant leakage rate is by monitoring the rate of rise in the levels of the drywell floor and equipment drain tanks. Checks will be made every four hours to verify that no alarms have been actuated due to high leakage. For sump l )
inflows of one gpm, changes on the order of 0.2 gpm can be detected within 40 minutes. At inflows between one and five gpm, changes on the order of 0.5 gpm can be detected in eight minutes. ,.
'FSAR AMENDMENT NO.142 103
L BASES FOR 3.2.5 AND 4.2.5 REACTOR COOLANT SYSTEM LEAKAGE RATE i
Another method of determining reactor coolant leakage rate is by monitoring for excess leakage in the drywell floor and equipment drain ,
tanka. This system monitors the change in tank volume over accurate time periods for the full range of tank instrumentation. If the leakage is high enough, an alarm is actuated indicating a leak rate above the predetermined limit (Section V.B)*.
Additional information is available to the operator which can be used for the shift leakage check if the drywell sumps level alarms are out of servL~a. The integrated flow pumped from the sumps to the waste disposal system can be checked. ,
t QualitItive infctmation is also available to the operator in the form of indication of drywell atmospheric conditions. Continuous leakage from i the primary coolant system would cause an increase in drywell temperature. Any leakage in excess of 15 gpm of steam would cause a i continuing increase in drywell pressure with resulting scram (First Supplement)*.
Either the rate of rise leak detection system, the excess leakage detection system or the integrated flow can be utilized to satisfy l Specification 3.2.5.b.
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t AMENDMENT NO.142 104 ]
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6 BASES FOR 3.3.1 AND 4e3.1 OXYGEN CONCENTRATION The four percent by volume oxygan concentration eliminates the possibility of hydrogen combustion following a loss-of-coolant accident (Section Vll-G.2.0 and Appendix E-II.5.2)*. The only way that significant quantities of hydrogen could be generated by metal-water r: action would be if the core spray system failed to sufficiently cool the core. As discussed in Section Vil-A.2.O*, each core spray system I i
will deliver, as a minimum, core spray sparger flow as shown on Figure Vil-2*. In addition to hydrogen generated by metal-water reactiori, significant quantities can be generated by radiolysis. (Technical Supplement to Petition for Conversion from Provisional Operating License to Full Tcrm Operating License). ;
t At r: actor pressures of 110 psig or less, the reactor will have been shutdown for more than an hour and the decay heat will be at sufficiently low values so that fuel rods will be completely wetted by core spray. The fuel clad temperatures would not exceed the core ,
spray water saturation temperature of about 344*F. ,
I The occurrence of primary system leakage following a major refueling outage or other scheduled shutdown is much more probable than the occurrence of the loss-of-coolant accident upon which the specified oxygen concentration limit is based. Permitting access to the drywell for 1::ck inspections during a startup is judged prudent in terms of the added plant safety offered without significantly reducing the margin of saft.ty. Thus to preclude the possibility of starting the reactor and operating for extended periods of t'une with significant leaks in the primtry system, leak inspections are scheduled during startup periods when the primary system is at or near rated operating temperature ,
cnd pressure. The 24-hour period to provide inerting is judged to be reasonable to perform the leak inspection and establish the required exygen concentration.
The primary containment is normally slightly pressurized during periods of reactor operation. Nitrogen used for inerting could leak out of the contrinment but air could not leak in to increase the oxygen concentration. Once the containment is filled with nitrogen to the required j conc ntration, no monitoring of oxygen concentration is necessary. However, at least once a week, the oxygen concentration will be ;
determined as added assurance that Specification 3.3.1 is being met.
s "FSAR ,
AMENDMENT NO.142 126 l i
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.,1 BASES FOR 3.4.4 AND 4.4.4 EMERGENCY VENTILATION SYSTEM The emergency ventilation system is designed to filter and exhaust the reactor building atmosphere to the stack during secondary containment isolation conditions. Both emergency ventilation system fans are designed to automatically start upon high radiation in the rsactor building ventilation duct or at the refueling platform and to maintain the reactor building pressure to the design negative pressure so es to minimize in-leakage. Should one system fail to start, the redundant system is designed to start automatically. Each of the two fans his 100 percent capacity. .
High efficiency particulate absolute (HEPA) filters are installed before and after the charcoal adsorbers to minimize potential release of particulates to the environment and to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radioiodine to the environment. The in-place test results should indicate a system leak tightness of less than 1 percent bypass Isakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal efficiency of at least 90 percent for expected accident conditions. If the cfficiencies of the HEPA filters and charcoal adsorbers are as specified, the resulting doses will be less than the 10CFR100 guidelines for the eccidents analyzed. Operation of the fans significantly different from the design flow will change the removel efficiency of the HEPA filters cnd charcoal adsorbers.
Only one of the two emergency ventilation systems is needed to cleanup the reactor building atmosphere upon containment isolation. If one '
system is found to be inoperable, there is no immediate threat to the containment system performance and reactor operation or refueling operation may continue while repairs are being made. If neither circuit is operable, the plant is brought to a condition where the emergency ventilation system is not required.
Pressure drop across the combined HEPA filters and charcoal adsorbers of less than 6 inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Heater capability and pressure drop should be determined at least once per operating cycle to show system performance capability.
The frequency of tests ami sample analysis are necessary to show that the HEPA filters and charcoal adsorbers can perform as evaluated.
The charcoal adsorber efficiency test should allow for charcoal. sampling to be conducted using an ANSI /ASME N510-1980 approved method. If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified in Table 5-1 of ANSI 509-1980.
AMENDMENT NO.142 176
t BASES FOR 3.4.5 AND 4.4.5 CONTROL ROOM AIR TREATMENT SYSTEM The control room air treatment system is designed to filter the control room atmosphere for intake air. A roughing filter is used for recirculation flow during normal control room air treatment operation. The control room air treatment system is designcd to maintain the control room pressure to the design positive pressure (one-sixteenth inch water) so that allleakage should be out leakage. The control room air treatment system starts automatically upon receipt of a LOCA (high cywell pressure or low-low reactor water level) or Main St:am Line Break (MSLB) (high steam flow main-steam line or high temperature main-steam line tunnel) signal. The system can also be minually initiated.
High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorber.
The charcoal adsorbers are installed to reduce the potential intake of radioiodine to the control room. The in-place test results should indicate a system leak tightness of less than 1 percent bypass leakage for the charcoal adsorbers and a HEPA efficiency of at least 99 percent removal of DOP particulates. The laboratory carbon sample test results should indicate a radioactive methyl iodide removal officiency of at least 90 percent for expected accident conditions. If the efficiencies of the HEPA filter and charcoal adsorbers are as specified, adequate radiation protection will be provided such that resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria for Nuclear Power Plants, Appendix A to 10CFR Part 50. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers.
If the system is found to be inoperable, there is no immediate threat to the control room and reactor operation or refueling operation may continue for a limited period of time while repairs are being made. If the makeup system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> or refueling operations are terminated.
Prassure drop across the combined HEPA filters and charcoal adsorbers of less than six inches of water at the system design flow rate will indicate that the filters and adsorbers are not clogged by excessive amounts of foreign matter. Pressure drop should be determined at least once per operating cycle to show system performance capability.
The frequency of tests and sample analysis are necessary to show the HEPA filters and charcoal adsorbers can perform as evaluated.
The charcoal adsorber efficiency test should allow for charcoal sampling to be conducted using an ANSI /ASME N510-1980 approved msthod. If test results are unacceptable, all adsorbent in the system shall be replaced with an adsorbent qualified according to Table 5-1 of ANSI 509-1980. The replacement charcoal for the adsorber tray removed for the test should meet the same adsorbent quality. Any HEPA filters found defective shall be replaced with filters qualified pursuant to ANSI 509-1980.
AMENDMENT NO. Md,161 180