ML20206T589
| ML20206T589 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 09/29/1986 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | Mangan C NIAGARA MOHAWK POWER CORP. |
| References | |
| NUDOCS 8610070068 | |
| Download: ML20206T589 (17) | |
Text
,
September 29, 1986 Docket No. 50-410 DISTRIBUTION:
- NM Attorney, 0GC NRC PDR JPartlow Mr. C. V. Mangan, Senior Vice President Local PDR EJordan Niagara Mohawk Power Corporation BWD-3 r/f BGrimes 300 Erie Boulevard West EAdensam ACRS (10)
Syracuse, New York 13202 MHaughey EHylton
Dear Mr. Mangan:
RBernero
Subject:
Corrections to the Final Draft Technical Specifications for Nine Mile Point, Unit 2 On September 10 and 11, 1986, we sent you the revised pages to the Final Draft Technical Specifications which were the result of our review of your requested changes to the Technical Specifications through August 22, 1986.
In subsequent discussions with your staff, errors in those pages were identified. The staff therefore reviewed the Final Draft Technical Specifications for additional errors.
The enclosed corrected pages for the Nine Mile Point, Unit 2 Technical Specifications contain changes resulting from either editorial or word processing errors and therefore do not alter the meaning of the specifications as provided in the final draft Technical Specifications provided on June 27, 1986 and September 10 and 11, 1986.
The enclosed corrected pages should replace the corresponding pages in your final draft Technical Specifications and should be included in your recertification of the revised final draft Technical Specifications as requested in our letter of September 10, 1986.
Sincerely,
/
Elinor G. Adensam. Director BWR Project Directorate No. 3 Division of BWR Licensing
Enclosure:
As stated cc: C. Schulten D. Vassallo 8610070068 060929 PDR ADOCK0">00gG nr&dw 0
y LAhD-3: DBL BWD-3:DBD D: W 3: DBL MHaughey/vag EHfiton EA e sam 09/j/86 09/g/86 09
/86 7
l S
fir. C. V. Mangan Nine Mile Faint Nuclear Station Niagara Mohawk Power Corporation Unit 2 cc:
Mr. Troy B. Conner, Jr., Esq.
Regional Administrator, Region I Conner & Wetterhahn II.S. Nuclear Regulatory Commission Suite 1050 631 Park Avenue 1747 Pennsylvania Avenue, N.W.
King of Prussia, Pennsylvania 19406 Washington, D.C.
20006 Mr. Paul D. Eddy Richard Goldsmith New York State Public Serice Syracuse University Cormission Collece of Law Nine File Point Nuclear Station -
E. I. White Hall Carnus Unit II Syracuse, New York I???3 P.O. Box 63 Lycoming, New York 13093 Ezra I. Bialik Assistant Attnrney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York. New York 10047 Resident inspectte Nirie Mile Point Nuclear Power Statior P. O. Box 99 Lycomina, New York 13093 Mr. John W. Keib, Esq.
Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13202 Mr. James Linville U. S. Nuclear Regulatory Commission Region I 631 Park Avenue Kina of Prussia, Pennsylvania 19406 Norman Rademacher, licensing Niacara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13?O2 00n Hill Niacara Mohawk Power Corooration Suite 550 4520 East West Highway Bethesda, Maryland 20814 7
- =
TABLE 4.11.1-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT MINIMUM OF DETECTION LIQUID RELEASE SAMPLING ANALYSIS TYPE OF ACTIVITY (LLD)(a)
TYPE FREQUENCY FREQUENCY ANALYSIS (pCi/ml) 1.
Batch Waste P
P Principal Gamma 5x10 7 Release Each Batch Each Batch Emitters (c)
Tanks (b)
- b. 2LWS-TK4B e
- c. 2LWS-TK5A
- d. 2LWS-TK5B P
One Batch /M Dissolved and 1x10 5 One Batch /M Entrained Gases (Gamma Emitters)
P M
H-3 1x10 5 Each Batch Composite (d)
Gross Alpha 1x10 7 P
Q Sr-89, St-90 5x10.s Each Batch Composite (d)
Fe-55 1x10 8 2.
Continuous Grab Sample Grab Sample Principal Gamma 5x10 7 Releases M(e)
M(e)
Emitters (c)
I-131 1x10 8
- a. Service Water Dissolved and 1x10 s Effluent A Entrained Gases
~
(Gamma Emitters)
- b. Service Water H-3 1x10 5 Effluent B Gross Alpha 1x10 7
- c. Cooling Tower Grab Sample Grab Sample Sr-89, Sr-90 5x10.s Blowdown Q(e)
Q(e)
Fe-55 1x10 8
- d. Auxiliary Grab Sample Gram Sample Principal Gamma 5x10 7 Boiler M(f)
M(f)
Emitters (c)
Pump Seal and Sample l
Cooling Discharge H-3 1x10 6 (Service Grab Sample Grab Sample Water)
Q(f)
Q(f)
NINE MILE POINT - UNIT 2 3/4 11-2 SED * ^ nlE
RADI0 ACTIVE EFFLUENTS BASES LIQUID EFFLUENTS DOSE 3/4.11.1.2 (Continued)
Revision 1, October 1977 and R.G. 1.113 " Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Imple-menting Appendix I," April 1977.
This specification applies to the release of i
radioactive materials in liquid effluents from each unit at the site.
For units with shared radwaste treatment systems, the liquid effluents from the shared system are to be proportiorfed among the units sharing that system.
3/4.11.1.3 LIQUID RADWASTE TREATMENT SYSTEM The OPERABILITY of the liquid radwaste treatment system ensures that this sys-tem will be available for use whenever liquid effluents require treatment before release to the environment.
The requirement that the appropriate portions of this system be used when specified provides assurance that the releases of ra-dioactive materials in liquid effluents will be kept as low as is reasonably achievable.
This specification implements the requirements of 10 CFR 50.36a, GDC 60 of Appendix A to 10 CFR 50 and the design objective given in Section II.D of Appendix I to 10 CFR 50.
The specified limits governing the use of appro-priate portions of the liquid radwaste treatment system were specified as a suitable fraction of the dose design objectives set forth in Section II.A of Appendix I to 10 CFR 50 for liquid effluents.
This specification applies to -
the release of radioactive materials in liquid effluents from each unit at the site.
For units with shared radwaste treatment systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system.
3/4.11.1.4 LIQUID HOLOUP TANKS The tanks listed in this specification include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR 20, Appendix B. Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
3/4.11.2 GASE0US EFFLUENTS 3/4.11.2.1 DOSE RATE This specification is provided to ensure that the dose tate at any time at and beyond the SITE BOUNDARY from gaseous effluents from all units on the site will be within the annual dose limits of 10 CFR 20 to UNRESTRICTED AREAS.
NINE MILE POINT - UNIT 2 B3/4 11-2 SEP 2 41986
...c RADIOACTIVE EFFLUENTS a
BASES GASEOUS EFFLUENTS DOSE - 10 DINE-131, 10 DINE-133. TRITIUM, AND RADI0 ACTIVE MATERIAL IN j
PARTICULATE FORM i
3/4.11.2.3 (Continued) j milk and meat is assumed), and (4) deposition on the ground with subsequent exposure to man.
This specification applies to the release of radioactive materials in gaseous effluents from each unit at the site.
For units with shared radwaste treatment systems, the gaseous effluents from the shared sys-tem are proportioned among the units sharing that system.
3/4.11.2.4 & 3/4.11.2.5 GASEOUS RADWASTE TREATMENT SYSTEM AND VENTILATION EXHAUST TREATMENT SYSTEM
(
The OPERABILITY of the GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment before release to the environ-ment.
The requirement that the appropriate portions of these systems be used, l
when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept as low as is reasonably achievable.
This specification implements the requirements of 10 CFR 50.36a, GDC 60 of Appendix A to 10 CFR 50, and the design objectives given in Section II.D of Appendix I to 10 CFR 50.
Limits governing the use of appropriate portions of i
the system were specified as a suitable fraction of the dose design objectives set forth in Sections II.B and II.C of Appendix I to 10 CFR 50, for gaseous effluents.
This specification applies to the release of radioactive materials.
in gaseous effluents from each unit at the site.
For units with shared rad-waste treatment systems, the gaseous effluents from the shared system are l
proportional among the units sharing that system.
3/4.11.2.6 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the GASEOUS RADWASTE TREATMENT SYSTEM is maintained below the flamability limits of hydrogen and oxygen.
Automatic control features are included in the system to prevent the hydrogen concen-trations from reaching these flammability limits.
These automatic control features include injection of dilutants to reduce concentrations below flamma-bility limits.
Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of GDC 60 of Appendix A to 10 CFR 50.
3/4.11.2.7 MAIN CONDENSER - 0FFGAS Restricting the gross radioactivity rate of noble gases from the mai.n condenser offgas provides reasonable assurance that the total body exposure to an indi-vidual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR 100 in the event this effluent is inadvertently discharged directly to the environment without treatment. This specification implements the requirements of GDC 60 and 64 of Appendix A to 10 CFR 50.
NINE MILE POINT - UNIT 2 B3/4 11-5
i 4
~
6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Superintendent - Nuclear Generation shall be responsible i
for overall unit operation and shall delegate in writing the succession to-this responsibility during the Superintendent's absence.
l 6.1.2 The Station Shift Supervisor - Nuclear (or during the Supervisor's absence from the control room, a designated individual) shall be responsible i
for the control room command function.
A management directive to this effect, signed by the Vice President - Nuclear Generation shall be reissued to all j
station personnel annually.
- 6. 2 ORGANIZATION 0FFSITE 6.2.1 The offsite organization for unit management and technical support j
shall be as shown on Figure 6.2.1-1.
UNIT STAFF I
6.2.2 The unit organization shall be as shown on Figure 6.2.2-1 and:
a.
Each on-duty shift shall be composed of at least the minimum shift crew shown in Table 6.2.2-1; i
b.
At least one Licensed Operator shall be in the control room when fuel is in the reactor.
In OPERATIONAL CONDITIONS 1, 2, or 3, at least one 1
Licensed Senior Operator or Licensed Operator shall be at the controls of j
the unit.
i c.
A Radiatian Protection Technician
- shall be on site when fuel is in the l
- reactor,
}
d.
At le.ast two Licensed Operators shall be present in the control room during j
reactor startup, scheduled reactor shutdown, and during recovery from reactor j
trips.
I e.
A Licensed Senior Operator shall be required in the control room during OPERATIONAL CONDITIONS 1, 2, and 3 and when the emergency plan is acti-l vated.
This may be the Station Shift Supervisor - Nuclear, the Assistant i
Station Shift Supervisor - Nuclear or other individuals with a valid senior operator license.
When the emergency plan is activated in OPERA-TIONAL CONDITIONS 1, 2, or 3 the Assistant Station Shift Supervisor -
Nuclear becomes the Shift Technical Advisor and the Station Shift i
- The Radiation Protection Technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.
This provision does not permit any shift j
crew position to be unmanned upon shift change due to an oncoming crewman l
being late or absent.
(
NINE MILE POINT - UNIT 2 6-1 SEP 2 4 W L---
ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS
_ ROUTINE REPORTS ANNUAL REPORTS 6.9.1.5 (Continued) whole-body dose received from external sources should be assigned to specific major work functions.
~
b.
The results of specific activity analysis in which the primary coolant exceeded the limits of Specification 3.4.5.
The following information shall be included:
(1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before the first sample in which the limit was exceeded; (2) Results of the last iso-topic analysis for radiofodine performed before exceeding the limit, results of analysis while the limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit.
Each result should include datt and time of sampling and the radioiodine concentrations; (3)
Cleanup system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> before the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a func-tion of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.
c.
Documentation of all challenges to safety / relief valves; and i
d.
Any other unit unique reports required on an annual basis.
MONTHLY OPERATING REPORTS 6.9.1.6 Routine reports'of operating statistics and shutdown experience, in-cluding documentation of all challenges to the main steam system safety / relief valves, shall be submitted monthly to the Director, Office of Resource Manage-ment, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555, with a copy to the Regional Administrator of the Regional Office of the NRC no later than the 15th of each month following the calendar month covered by the report.
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT
- 6.9.1.7 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The initial report shall be submitted before May 1 of the year after the plant achieves initial criticality.
The Annual Radiological Environmental Operating Report shall include sum-maries, interpretations, and an analysis of trends of the results of the radio-logical environmental surveillance activities for the report period, including a comparison, as appropriate, with preoperational studies, operational controls,
- A single submittal may be made for a multiple unit site. The submitte' should combine those sections that are common to all units at the site.
NINE MILE POINT - UNIT 2 6-19 SEP 2 4 W
ADMINISTRATIVE CONTROLS REPORTING REQUIREMENTS ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORTS 6.9.1.7 (Continued) previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.
The reports shall also include the results of the land use census required by Specification 3.12.2.
The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the OFFSITE DOSE CALCULATION MANUAL, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.
In the event that some indivi-dual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.
The missing data shall be submitted as soon as possible in a supplemental report.
The reports shall also include the following:
a summary description of the Radiological Environmental Monitoring Program; at least two legible' maps
- cover-l ing all sampling locations keyed to a table giving distances and directions from the centerline of one reactor; the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 3.12.3; dis-cussion of all deviations from the Sampling Schedule of Table 3.12.1-1; and dis-cussion of all analyses in which the LLD required by Table 4.12.1-1 was not achievable.
SEMIANNUAL RADI0 ACTIVE EFFLUENT RELEASE REPORT **
6.9.1.8 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be sub-mitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date the plant achieves initial criticality.
The Semiannual Radioactive Effluent Release Reports shall include a sumary of the quantities of radioactive liquid and gaseous effluents and solid waste re-leased from the unit as outlined in Regulatory Guide 1.21, " Measuring, Evalua-ting, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power
- One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.
- A single submittal may be made for a multiple unit site. The submittal should combine those sections that are common to all units at the site; however,' for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.
NINE MILE POINT - UNIT 2 6-20 SEP 2 4 man
TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ag APPLICABLE CHANNEL CHANNEL OPERATIONAL o
INSTRUMENT CHECK CALIBRATION CONDITIONS
',i 1.
Reactor Vessel Pressure M
R 1, 2 2.
Reactor Vessel Water Level e3 a.
Fuel Zone M
R 1, 2, 3
-4 b.
Wide Range M
R 1, 2, 3 m
3.
Suppression Pool Water Level a.
Narrow Range M
R 1,2,3 b.
Wide Range M
R 1,2,3 4.
Suppression Pool Water Temperature M
R*
1, 2 5.
Suppression Chamber Pressure M
R 1, 2 6.
Suppression Chamber Air Temperature M
R$
1, 2 y
7.
Drywell Pressure a.
Narrow Range M
R 1, 2 (p
b.
Wide Range M
R 1, 2 g
8.
Drywell Air Temperature M
R*
1, 2 9.
Drywell Oxygen Concentration M
R 1, 2 10.
Drywell Hydrogen Concentration Analyzer and Monitor M
Q**
1, 2 11.
Safety / Relief Valve Position Indicators M
R 1, 2 12.
Drywell High Range Radiation Monitors M
Rt 1, 2, 3 13.
RHR Heat Exchanger Service Water Radiation Monitor M
R 1, 2, 3 14.
Refuel Platform Area Radiation Monitor M
R tt 15.
Neutron Flux a.
APRM M
R 1, 2 b.
IRM M
R 1, 2 c.
SRM M
R 1
l 16.
Primary Containment Isolation Valve Position Indication Mtti R***
1, 2 E=
b3
)
I i
INSTRUMENTATION I
MONITORING INSTRUMENTATION I
FIRE DETECTION INSTRUMENTATION LIMITING CONDITIONS FOR OPERATION 3.3.7.8 As a minimum, the fire detection instrumentation for each fire detec-tion zone shown in Table 3.3.7.8-1 shall be OPERABLE.
i APPLICABILITY:
Whenever equipment protected by the fire detection instrument l
is required to be OPERABLE.
ACTION:
?
a.
With any, but not more than one-half the total in any fire zone, Function N*
fire detection instruments shown in Table 3.3.7.8-1 inoperable, restore the inoperable Function N* instrument (s) to OPERABLE status within 14 days or within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour.
b.
With more than one-half the Function N* fire detection instruments in any I
fire zone shown in Table 3.3.7.8-1 inoperable or with any Functions S* or l
X* instruments shown in Table 3.3.7.8-1 inoperable, or with any two or more adjacent instruments shown in Table 3.3.7.8-1 inoperable, within I hour establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour.
i c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.8.1 Each of the above required fire detection instruments which are accessible during unit operation shall be demonstrated OPERABLE.at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST.' Fire detectors which are not accessible during unit operation shall be demonstrated OPERABLE i
by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless performed in the previous 6 months.
i 4.3.7.8.2 The NFPA Standard 72D supervised circuits supervision associated with l
the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months.
l 4.3.7.8.3 The non-supervised circuits associated with detector alarms between the instruments and the control room shall be demonstrated OPERABLE at least j
once per 31 days.
{
j
- These letters are found in the alpha-numeric fire zone designation and are explained in the foct. note to Table 3.3.7.8-1 l
NINE MILE POINT - UNIT 2 3/4 3-91
-SEP 2 41986- - -. -
TABLE 4.3.7.11-1 (Continued)
RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS a
E CHANNEL MODES IN WilICll CHANNEL SOURCE CHANNEL FUNCTIONAL S'1VEILLANCE m
o INSTRUMENT CHECK CHECK CALIBRATION TEST REQUIRED Y
4.
Main Stack Effluent c5
- a. Noble Gas Activity Monitor t D
M R(a)
Q(c) l
-4
- b. Iodine Sampler W
NA NA NA N
- c. Particulate Sampler W
NA NA NA
- d. Flow-Rate Monitor D
NA R
Q R
- e. Sampler Flow-Rate Monitor D
NA R
Q a
M
=
4 O
L4 e
i
TABLE 3.3.9-2 PLANT SYSTEMS ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE 1.
Feedwater System / Main Turbine Trip System a.
Reactor Vessel Water Level - High <202.3 in.*
-<209.3 in.
l Level 8 2.
Service Water System a.
Discharge Bay Level 1275' Elev.
1275' 2-3/4" Elev.
b.
Intake Tunnel 1 & 2
->39*F
->38 F Water Temperature c.
Service Water Bay
>234' Elev.
>233' 1-1/4" Elev.
d.
Service Water Pumps Discharge 110 psid 114.5 psid l
Strainer Differential Pressure -
Train "A" e.
Service Water Pumps Discharge 110 psid 114.5 psid l
Strainer Differential Pressure-Train "B" f.
Service Water Su; ply Header NA NA Discharge Water Temperature g.
Service Water Inlet Pressure for EDG*2 (HPCS, Division III)
- 1) Division I Supply Header
>25 psig
>17.5 psig
- 2) Division II Supply Header
>22 psig
>17.5 psig
- See Bases Figure B3/4 3-1.
4 NINE MILE POINT - UNIT 2 3/4 3-114 SEP 2 4 HE6
g e
TA8LE 3.6.3-1 (Centinued)
PRIMARY CONTAINMENT ISOLATION VALVES a
5 ISOLATION VALVE ISOLATION MAXIMUM CLOSING g
VALVE NO.
VALVE FUNCTION GROUP SIGNAL (a)
TIME (SECONDS) 2RHS*MOV142(j)(a) RHS Drain to Radwaste Outside IV 4
A,Z,F,RM 30 2RHS*MOV149(j)(m) RHS Drain to Radwaste Inside IV 4
A,Z,F,RM 25 e
2RHS*SOV35 A/B (j)(m)
A,Z,F,RM 5
2RHS*SOV36 A/B (j)(a)
A Z,F,RM 5
2RDS*A0V124(k)
SCRAM Discharge volume vent NA NA 2RDS*A0V132(k)
SCRAM Discharge volume vent NA NA 2RDS*A0V123(k)
SCRAM Discharge volume drain NA NA 2RDS*A0V130(k)
SCRAM Discharge volume drain NA NA
,s B.
Remote Manual D
2RHS*MOV15 A,8 Containment Spray to Drywell Outside IV's 12 RM NA 2RHS*MOV 1 A,B,C RHS Pump Suction Outside IVs 12 RM NA l
2RHS*MOV30 A,8 RHS Test Line to SP Outside IVs 12 RM NA 2RHS*MOV25 A,8(n) Containment Spray to Drywell Outside IVs 12 RM NA 2RHS*MOV24 A,B,C RHS/LPCI to RPV Outside IVs 12 RM NA 2CSH*MOV118(n)
CSH Suction from SP Outside IV 12 RM NA 2CSH*MOV105 HPCS Min Flow Bypass Outside IV 12 RM NA 2CSH*MOV107 CSH to RPV Dutside IV 12 RM NA l
2CSL*MOV112 CSL Suction from SP Outside IV 12 RM NA l
2CSL*MOV104 CSL to RPV Outside IV 12 RM NA 2ICS*MOV136(n)
ICS Suction from SP Outside IV 12 RM NA 2ICS*MOV143(n)
ICS Min flow to SP Outside IV 12 RM MA to B
TABLE 3.6.3-1 (Centinued)
PRIMARY CONTAINMENT ISOLATION VALVES a
E ISOLATION VALVE ISOLATION MAXIMUM CLOSING g
VALVE NO.
VALVE FUNCTION GROUP SIGNAL (a)
TIME (SECON05) l E
H D.
Other e
g Safety Relief i
[
2RHS*RV20 A,B,C(d)
RHS RV disch. to SP Outside IVs i
2RHS*RV61 A,B,C(d)
RHS RV disch. to SP Outside IVs 2RHS*RV108(d)
RHS RV disch. to SP Outside IVs i
2RHS*RV110(d)
2RHS*RV139(d)
RHR Hdr. Flush to Radwaste RV i
2RHS*RV152(n)
SDC Supply from RCS RV Inside IV i
2RHS*RVS6 A,B(d)
2RHS*SV34 A,B(d)
RHS HX steam supply Safety valves 2RHS*SV62 A,8(d)
RHS HX steam supply Safety valves j
4 2RHS*RVV35 A,B(d)
RHS Vacuum Breakers e
I 2CSL*RV10S(d)
CSL RV Disch. to SP Outside IV 2CSL*RV123(d)
CSL RV Disch. to SP Outside IV l
2RHS*RVV36 A,B(d)
RHS Vacuum Breakers 4
i 2CCP*RV170(n)
CCP RV Discharge Inside IV l
2CCP*RV171(n)
CCP RV Discharge Inside IV 2CSH*RV113(d)
CSH RV Disch.' to SP Outside IV' 2CSH*RV114(d)
CSH RV Disch. to SP Outside IV 4
0 N
b m-B l
.l TABLE 3.6.3-1 (Centinued)
PRIMARY CONTAINMENT ISOLATION VALVES a
E ISOLATION VALVE ISOLATION MAXIMUM CLOSING g
VALVE NO.
VALVE FUNCTION GROUP SIGNAL (a)
TIME (SECONOS)
'i!
Excess Flow Check (el Reactor Instrumenta-
{
tion Lines Z
2ISC*EFV1 Inst. Line from MSS m
2ISC*EFV2 Inst. Line from N14,200*
2ISC*EFV3 Inst. Line from N14,160*
2ISC*EFV4 Inst. Line from N13,190" 2ISC*EFV5 Inst. Line from N14,20*
2ISC*EFV6 Inst. Line from N14,340*
w}
2ISC*EFV7 Inst. Line from N13,10 2ISC*EFV8 Inst. Line from N12,160*
J, 2ISC*EFV10 Inst. Line from N12,200*
H 2ISC*EFV11 To 2?SC*FT47K,FT48B 2ISC*EFV13 To 2ISC*FT47H 2ISC*EFV14 Vessel Bottom Tap, loop A Jet Pump l
2ISC*EFV15 Inst. Line from N12,340*
2ISC*EFV17 Inst. Line from N12,20*
21SC*EFV18 To 2ISC*FT47J,FT48A 21SC*EFV20 To 2ISC*FT47E 2ISC*EFV21 Vessel Bottom Tap for CSH, RDS 2ISC*EFV22 Vessel Bottom Tap for WCS and Loop B J.P.
2ISC*EFV23 To 2ISC*FT48C and Postaccident Sampling 2ISC*EFV24 To 21SC*FT480 and Postaccident Sampling 21SC*EFV25 To 2ISC*FT47L 2ISC*EFV26 To 2ISC*FT47C 21SC*EFV27 To 2ISC*FT47A 2ISC*EFV28 To 2ISC*FT47R u
2ISC*EFV29 To 2ISC*FT47G 2ISC*EFV30 To 2ISC*FT47N m
2ISC*EFV31 To 2ISC*FT43A A
2ISC*EFV32 To 2ISC*FT47T g
2ISC*EFV33 To 2ISC*FT47V,FT48C
ELECTRICAL POWER SYSTEMS ELECTRICAL EQUIPMENT PROTECTIVE DEVICES PRIMARY CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES LIMITING CONDITIONS FOR OPERATION 3.8.4.2 All primary containment penetration conductor overcurrent protective devices
- shall be OPERABLE.
AiPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
a.
With one or more of the primary containment penetration conductor overcurrent protective devices
- inoperable, declare the affected system or component inoperable and apply the appropriate ACTION statement for the affected system and:
1.
For 13.8-kV circuit breakers, deenergize the 13.8-kV circuits by tripping the associated redundant circuit breaker (s) within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify the redundant circuit breaker (s) to be tripped at least once every 7 days thereafter.
2.
For 600 volt MCC circuit breakers, remove the inoperable circuit breaker (s) from service by opening the breaker within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and verify the inoperable breaker (s) to be in the open position at least once every 7 days thereafter.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
The provisions of Specification 3.0.4 are not applicable to overcurrent devices in 13.8-kV circuits which have their redundant circuit breakers tripped or to 600-volt circuits which have the inoperable circuit breaker disconnected.
SURVEILLANCE REQUIREMENTS 4.8.4.2 Each of the primary containment penetration conductor overcurrent pro-tective devices
- shall be demonstrated OPERABLE:
a.
At least once per 18 months:
1.
By verifying that the medium voltage 13.8-kV circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers of each voltage level and performing:
- Excluded-from this specification are those penetration assemblies that are capable of withstanding the maximum current available because of an electrical fault inside containment.
NINE MILE POINT - UNIT 2 3/4 8-28 SEP 2 4 586 a
~, a SPECIAL TEST EXCEPTIONS 3/4.10.7 SPECI_AL INSTRUMENTATION -_ INITIAL CORE LOADING LIMITING CONDITIONS FOR OPERATION 3.10.7 During initial core loading within the Startup Test Program the pro-visions of Specification 3/4.9.2 may be suspended provided that at least two source range monitor (SRM) channels with detectors inserted to the normal operating level are OPERABLE with:
a.
One of the required SRM channels continuously indicating
- in the control room, b.
One of the required SRM 4etectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant,**
c.
The RPS " shorting links" shall be removed prior to and during fuel
- loading, d.
The reactor mode switch is OPERABLE and locked in the Refuel position.
APPLICABILITY:
OPERATIONAL CONDITION 5 ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving initial core loading.
SURVEILLANCE REQUIREMENTS 4.10.7.1 Within one hour prior to and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during the initial core loading verify that:
a.
The above required SRM channels are OPERABLE by:
1.
Performance of a CHANNEL CHECK ***
2.
Confirming that the above required SRM detectors are at the normal operating level and located in the quadrants required by Specification 3.10.7.
"Up to 16 fuel bundles may be loaded without a visual indication of count rate.
- The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detec-tors are connected to the normal SRM circuits.
- May be performed by use of movable neutron source.
NINE MILE POINT - UNIT 2 3/4 10-7 SEP 2 41986