ML20210L864
| ML20210L864 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 09/25/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Nebraska Public Power District (NPPD) |
| Shared Package | |
| ML20210L867 | List: |
| References | |
| DPR-46-A-102, TAC 57610 NUDOCS 8610030320 | |
| Download: ML20210L864 (23) | |
Text
-
a 1.Eauq k,
UNITED STATES
/
E NUCLEAR REGULATORY COMMISSION g
L 4j WASHINGTON, D. C. 20555
\\...../
NEBRASKA PUBLIC POWER DISTRICT DOCKET NO. 50-298 COOPER NUCLEAR STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 102 License No. DPR-46 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Nebraska Public Power District dated April 26, 1985, as supplemented by submittals dated May 24, 1985, and June 14, 1985, and July 3, 1986 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the licensee is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-46 is hereby amended to read as follows:
8610030320 860'/25 PDR ADOCK 0000 8
P
e
-, (2) Technical Specification The Technical Specifications contained in Appendices A and B, as revised through Amendment No.102, are hereby incorporated in the q]
license. The licensee shall operate the facility in accordance l
with the Technical Specifications.
3.
This license amendment is effective.as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
.k w A{
W Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 25, 1986 1
r
' k I V t)e
C' i
e s
ATTACHMENT TO LICENSE AMENDMENT N0. 102 FACILITY OPERATING LICENSE N0. DPR-46 i
DOCKET NO. 50-298 i
1.
Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines Pages iv 50 52b 75 83 1
111
]
165 182 183 I
205 215 l
215d i
221 222 223 224 225 i
226 i
227 234 l
I l
1 i
{
.~.
. -.., -, _ _., - _ _ _, -._,,.,,,..~.---. ~,,-,_.
- i TABLE ~OF~ CONTENTS ~(Cont'd.)
Page No.
SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS 6.2 Review and Audit 220 6.2.1.A itation Operations Review Committee (SORC) 220 MA dembership 220 A..
Meeting Frequency 220 A.3 Quorum 220 A.4 Responsibilities 220 A.5 Authority 221 A.6 Records 221 A.7 Procedures 221 l
(
d l
6.2.1.B NPPD Safety Review and Audit Board (SRAB) 222 B.1 Membership 222 B.2 Meeting Frequency 222 B.3 Quorum 222 B.4 Review 222 i
B.5 Authority 223 j
B.6 Records 223 B.7 Audits 223 3
s 6.3 Procedures and Programs 225 l
l 6.3.1 Introduction 225 6.3.2 Procedures 225 6.3.3 Maintenance and Test Procedures 225
{
6.3.4 Radiation Control Procedures 225
.A High Radiation Areas 226 6.3.5 Temporary Changes 226 6.3.6 Exercise of Procedures 226 6.3.7 Programs 226
.A Systems Integrity Monitoring Program 226 l
.B Iodine Monitorihg Program 226
.C Environmental Qualification Program 226 5
.D Post-Accident Sampling System (PASS) 227 i
6.4 Record Retention 228 6.4.1 5 year retention 228 6.4.2 Life retention 228 6.4.3 2 year retention 229 6.5 Station Reporting Requirements 230 6.5.1 Routine Reports 230
.A Introduction 230
.B Startup Report 230
.C Annual Reports 230
.D Monthly Operating Report 231
.E Annual Radiological Environmental Report 231
.F Semiannual Radioactive Material Release Report 232 Amendment No. 191, 102
-iv *
,,,, g;.
i___
11, p, y y-, m COOPER NUCLEAR STATION r
g-TABLE 3.2.A (Page 1) g PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION INSTRUMENTATION 3
I n
Minimum Number Action Required of Operable WhenComponentl g
Instrument Components Per Operability is; Instrument I.D. No.
Setting Limit Trip System (1) Not Assured (2)t
?2 Main Steam Line High RMP-RM-251, A,B,C,6D
$ 3 Times Full Power 2
A or B ga Rad.
g Reactor Low Water Level NBI-LIS-101, A,B,C,6D #1 3+12.5" Indicated Level 2(4)
A or B
- g Reactor Low Low Low Water NBI-LIS-57 A & B #1
>-145.5" Indicated Level 2
A or B ya Level NBI-LIS-58 A & B #1 f5 Main Steam Line Leak MS-TS-121, A,B,C,6D
$ 200*F 2(6)
B Detection 122, 123, 124, 143, 144, g
145, 146, 147, 148, 149, w
150 v.
E Main Steam Line High MS-dPIS-116,A,B,C,6D 2 150% of Rated Steam 2(3)
B Flow 117, 118, 119 Flow Main Steam Line Low MS-PS-134, A,B,C,6D
> 825 psig 2(5)
B Pressure High Drywell Pressure PC-PS-12, A,B C,6D
$ 2 psig 2(4)
A or B High Reactor Pressure RR-PS-128 A & B
$ 75 psig i
D Main Condenser Low MS-PS-103, A,B,C,6D
> 7" Hg (7) 2 A or B Vacuum Reactor Water Cleanup RWCU-dPIS-170 A & B f 200% of System Flow 1
C System High Flow e
e
NOTES FOR TABLE 3.2.A (cont'd.)
Isolations 1.
Secondary Containment Isolation 2.
Start Standby Gas Treatment System Group 7 Isolation Signals:
1.
Reactor Low Low Low Water Level (3-145.5 in) l 2.
Main Steam Line High Radiation (<3 times full power background)
Isolations:
1.
Reactor Water Sample Valves t
I i
i E
i Amendments Nos.
55, 58, 88, 101, 102
-52b--
..,.,7,g., -.
-f
.m
__m i
y I
4 COOPER NUCLEAR STATION o
l g
TABLE 4.2.B (Page 6) g RCIC TEST & CALIBRATION FREQUENCIES E'
sg Instrument-Item Item I.D. No.
Functional Test Freq.
Calibration Freq.
Check!
2 O
Instrument Channels D
1.
Reactor High Water Level NBI-LIS-101 A & C, #2 Once/ Month (1)
Once/3 Months Once/ Day
-g 2.
Reactor Low Water Level 10A - K79 A & B 10A-Once/ Month (1)
Once/3 Months once/ Day K80 A & B g-3.
RCIC High Turbine Exhaust RCIC-PS-72, A & B Once/ Month (1)
Once/3 Months None y
Press.
4.
RCIC Low Pump Suction Press.
RCIC-PS-67-1 Once/ Month (1)
Once/3 Months None 3
5.
RCIC Steam Line Space Excess RCIC-TS-79, A,B,C, & D Once/ Month (1)
Once/Oper. Cycle None Temp.
RCIC-TS-80, A,B,C, & D Once/ Month (1)
Once/Oper. Cycle None RCIC-TS-81, A,B,C, & D Once/ Month-(1)
Once/Oper. Cycle None RCIC-TS-82, A,B,C, & D Once/ Month (1)
Once/Oper. Cycle None 6.
RCIC Steam Line High AP RCIC-dPIS-83 Once/ Month (1)
Once/3 Months None RCIC-dPIS-84 Once/ Month (1)
Once/3 Months None d
7.
RCIC Steam Supply Press.. Low RCIC-PS-87, A,B,C, & D Once/ Month (1)
Once/3 Months
.None
'8 8.
RCIC Low Pump Disch. Flow RCIC-FIS-57 Once/ Month (1)
Once/3 Months None 9.
Pump Disch. Line Low Pressure CM-PS-269 Once/3 Months Once/3 Months None 10.
RCIC Turbine Conditional RCIC-TDR - K9 Once/ Month (1)
Once/Oper. Cycle None I
Supv. Alarm Timer 11.
RCIC Steam Line High AP RCIC-TDR-K-12 Once/ Month Once/Oper. Cycle None Actuation Timer RCIC-TDR-K-32 Once/ Month once/Oper. Cycle None Logic Systems (4)(6) j 1
1.
Logic Bus Power Monitor Once/6 Months N.A.
2.
RCIC Initiation once/6 Months N.A.
3.
Turbine Trip Once/6 Months N.A.
4.
RCIC Automatic Isolation Once/6 Months N.A.
J A
i 4
l J )
~
3.23 BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious con-sequences. This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems. The objectives of the specifications are (1) to assure the effectiveness of the protective instrumentation when required even during periods when portions of such systems are out of service for maintenance, and (2) to prescribe the trip settings required to assure adequate performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from i
the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
A.
Primary Containment Isolation Functions Actuation of primary containment valves is initiated by protective instru-mentation shown in Table 3.2 A which senses the conditions for which isola-i I
tion is required. Such instrumentation must be available uhenever primary containment integrity is required, t
j The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
g The low water level instrumentation, set to trip at 176.5" (+12.5") above the top of the active fuel, closes all isolation valves except those in Groups 1, 4, 5, j
and 7.
Details of valve grouping and required closing times are given in Specification 3.7.
For valves which isola,te at this level this trip setting 35_
i is adequate to prevent core uncovery in the case of a break in the largest (1
line assuming a 60 second valve closing time. Required closing times are less than this.
The low low low reactor water level instrumentation is set to trip when the water level is 19" (-145.5") above the top of the active fuel. This trip closes Groups 1 and 7 Isolation Valves (Reference 1), activates the remainder of the CSCS subsystems, and starts the emergency diesel generators. These trip level settings were chosen to be high enough to prevent spurious actuation but low enough to initiate CSCS operation and primary system isolation so that post accident cooling can be accomplished, 2
Amendments Nos. 75, $3, 58, 101, 102
.,zygy,
,v 2
3.4 BASES (cont'd.)
The volume versus concentration requirement of the solution is such that, should evaporation occur from any point within the curve, a low level alarm will annunciate before the temperature versus concentration requirements are exceeded.
The quantity of stored boron includes an additional margin (25 percent) beyond the amount needed to shutdown the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water.
A minimum quantity of 2650 gallons of solution having a 16 percent sodium pentaborate concentration, or the equivalent as shown in Figure 3.4.1, is required to meet this shutdown requirement. For the minimum required pumping rate of 38.2 gpm, the maximum net storage volume of the boron I
solution is established as 4780 gallons.
l 4.4 BASES i
l STANDBY LIQUID CONTROL SYSTEM i
Experience with pump operability indicates that the monthly test, in combination with the tests during each operating cycle, is sufficient to maintain pump performance. The only practical time to fully test the liquid control system is during a refueling outage. Various components of the system are individually tested periodically, thus making unnecessary more frequent testing of the entire system.
l The bases for the. surveillance requirements are given in subsection III.9.6 of 8
the Final Safety Analysis Report, and the details of the various tests are discussed in subsection III.9.5.
The solution temperature and volume are l
l checked at a frequency to assure a high reliability of operation of the system f
should it ever be required.
4s l
8L-at i
u
\\
Amendments Nos. JOI, 102
-111-
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.7. (cont'd.)
4.7 (cont'd.)
B.
Standby Gas Treatment System B.
Standby Gas Treatment System 1.
Except as specified in 3.7.B.3 below, 1.
At least once per operating cycle both standby gas treatment systems the following conditions shall be shall be operable at all times when demonstrated.
secondary containment integrity is required.
a.
Pressure drop across the combined HEPA filters and charcoal adsorber 2.a. The results of the in-place cold DOP banks is less than 6 inches of and halogenated hydrocarbon leak tests water at the system design flow at < design flow (1780 CD1) and at a rate.
reactor buil'ing pressure 3.25" Wg on HEPA filters and charcoal adsorber b.
Inlet heater input is capable of banks respectively shall show 399%
reducing R.H. from 100 to 70% R.H.
DOP removal and 199% halogenated hydrocarbon removal.
2.a. The tests and sample analysis of Specification 3.7.B.2 shall be
- b. The results of laboratory carbon performed at least once per year l
sample analysis shall show 399%
for standby service or after every radioactive eethyl iodide removal 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation and with inlet conditio3s of: velocity following significant painting, 342 FPM, 31.75 mg/m inlet methyl fire or chemical release in any iodide concentration, >70% R.H.
and $30 C.
ventilation zone communicating with the system.
- c. Each fan shall be shown to provide
after each complete or partial l
replacement of the HEPA filter f
f 3.
From and after the date that one bank or after'any structural F
standby gas treatment system is made maintenance'on the system housing.
i l
or found to be inoperable for any i
reason, reactor operation is permis-
- c. Halogenated hydrocarbon testing sible only during the succeeding shall be performed after each l
seven days unless such system is
' complete or partial replacement sooner made operable, provided that of the charcoal adsorber bank during such seven days all active or after any structural main-components of the other standby tenance on the system housing.,
gas treatment system, and its associated diesel generator, shall
- d. Each system shall be operated be operable.
with the heaters on at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> every month.
Fuel handling requirements are speci-fled in Specification 3.10.E.
- e. Test sealing of gaskets for housing doors downstream of the HEPA filters and charcoal adsorbers shall be performed at, and in conformance with, each test performed for compliance with
)
Specification 4.7.B.2.a and i
Specification 3.7.B.2_.a.
3.
System drains where present shall i
be inspected quarterly for adequate water level in loop-se.21s.
Amendment No. 58, 79, 50, 82, 88, 101, 102
~
-165-4.:xy n
a 3.7.B & 3.7.C BASES (cont'd) i High efficiency particulate absolute (HEPA) filters are installed before and
\\
after the charcoal adsorbers to minimize potential release af 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 leakage for the charcoal adsorbers and HEPA filters. The laboratory carbon sample test results should indicate a radio-active methyl iodide removal efficiency of at lease 99 percent for expected accident conditions.
If the performance of the HEPA filters and charcoal l
adsorbers are as specified, the resulting doses will be less than the 10 CFR 100 guidelines for the accidents analyzed.
l Only one of the two standby gas treatment systems is needed to cleanup the reactor building atmosphere upon containment isolation.
If one system is j
found to be inoperable, there is no immediate threat to the containment t
system performance and reactor operation or refueling operation may continue i
while repairs are being made.
If neither system is operable, the plant is brought to a condition where the standby gas treatment system is not required.
k 4.7.B & 4.7.C BASES i
3 l
Standby Gas Treatment System and Secondary Containment 2
l I
l Initiating reactor building isolation and operation.of the standby gas treatment system to maintain at least a 1/4 inch of water vacuum within the secondary I
containment provides an adequate test of the operation of the reactor building I
isolation valves, leak tightness of the reactor building and performance of the st.ndby gas treatment system. Functionally testing the initiating sensors and associated trip channels demonstrates the capability for automatic actuation.
j Performing these tests prior to: refueling will demonstrate secondary containment l
capability prior to the time the primary containment is opened for refueling.
Periodic testing gives sufficient confidence of reactor building integrity and
!g l
standby gas treatment system performance capability.
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
'E filters and adsorbers are not clogged by excessive amounts of foreign matter.
E 4
~
A 7.8 kw heater is capable of maintaining relative humidity below 70%. Heater i
capacity and pressure drop should be determined at least once per operating 4
g
{
cycle to show system performance capability.
t The frequency of tests and sample analysis are necessary to show that the HEPA.
filters and charcoal adsorbers can perform as evaluated. Tests of the charcoal adsorbers with halogenated hydrocarbon refrigerant shall be performed in accor-dance with ANSI N510-1980. The test cannisters that are installed with the
~
adsorber trays should be used for the charcoal adsorber efficiency test. Each j
sample should be at least two inches in diameter and a length equal to the thickness of the bed.
If test results are unacceptable, all adsorbent in the system shall be replaced 4
i i
j Amendment No. $@, 82, 88, 191, 102
-182-
<7
4./.B & 4.7.C BASES with an adsorbent qualified according to Table 1 of Regulatory Guide 1.52, Revision 2, March, 1978. The replacement tray for the adsorber tray removed for the test should meet the same adsorbent quality. Tests of the HEPA fil-ters with DOP aerosol shall be performed in accordance to ANSI N310-1980.
Any filters found defective shall be replaced with filters qualified pursuant to Regulatory Position C.3.d. of Regulatory Guide 1.52, Revision 2, March, 1978.
l All elements of the heater should be demonstrated to be functional end operable during the test of heater capacity. Operation of the heaters will prevent moisture buildup in the filters and adsorber syttem.
With doors closed and fan in operation, DOP aerosol shall be sprayed externally along the full linear periphery of each respective door to check the gasket i
seal. Any detection of DOP in the fan exhaust shall be considered an unacceptable test result and the gaskets repaired and test repeated.
If system drains are present in the filter /adsorber banks, loop-seals must be used with adequate water level to prevent by-pass leakage from i
the banks.
If significant painting, fire or chemical release occurs such that the HEPA filter or charcoal adsorber could become contaminated from the fumes, chemicals or foreign material, the same tests and sample analysis shall be performed as required for operational use.
The determination of significance shall be made by the operator on duty at the time of the incident. Knowledgeable staff members should be consulted prior to making this determination.
Demonstration of the automatic initiation capability and operability of filter cooling is necessary to assure system performance capability.
If one standby gas treatment system is inoperable, the other system must be tested gr l
daily.
This substantiates the availability of the operable system and thus e
i reactor operation or refueling operation can continue for a limited period
}
of time.
17 t
3.7.D & 4.7.D BASES 3.
Primary Containment Isolation Valves I
Double isolation valvta are provided on lines penetrating the primary con-,'
tainment and open to the free space of the containment.
Closure.of one of the valves in each line would be sufficient to maintain the integrity of the pressure suppression system. Automatic initiation is required to minimize the potential leakage paths from the containment in the event of a loss-of-coolant accident.
The maximum closure times for the automatic isolation valves of the primary containment and reactor vessel isolation control system have been selected in consideration of the design intent to prevent core uncovering following pipe breaks outside the primary containment and the need to contain released fission products following pipe breaks inside the primary containment.
These valves are highly reliable, have a ' low service requirement, a,nd most are normally closed. The initiating sensors and associated trip channels are also checked to demonstrate the capability for automatic isolation.
The test interval of once per operating cycle for automatic initiation Amendment No. 82, 101, 102
-183-
,~.,
-se==-e
bIMITINGCONDITIONSFOROPERATION SURVEILLANCE REQUIREMENTS 3.10.A (Cont'd) 4.10 (Cont'd) 6.
Any number of control rods may be withdrawn or removed from the reactor core providing the following conditions are satisfied:
a.
The reactor mode switch is locked in the " refuel" position. The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed on a withdrawn con-trol rod after the fuel assemblies in the cell containing (controlled by) that control rod have been removed from the reactor core. When fuel is present in the reactor vessel, all other refueling interlocks shall be operable.
B.
Core Monitoring B.
Core Monitoring l
During core alterations two SRM's Prior to making any alterations to shall be operable, one in the core the core, the SRM's shall be quadrant where fuel or control rods functionally tested and checked for are being moved and one in an ad-neutron response. Thereafter, while jacent quadrant. For an SRM to be required to be operable, the SRM's I
considered operable, the following will be checked daily for response l
conditions shall be satisfied:
(or every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> until 3 cps is attained if the spiral reload 1.
The SRM shall be inserted to the normal technique is being used).
l Perating level.
(Use of special move-Og able, dunking type detectors during J
initial fuel loading and major core
{
alterations in place of normal detec-i tors is permissible as long as the F
detector is connected to the normal 0-SRM circuit.)
2.
Operable SRM's shall have c minimum of g
3 cps except as specified in 3 and 4 below.
3.
Prior to spiral unloading, the SRM's shall have an initial count rate of 3 cps. During spiral unloading, the count rate on the SRM's may drop below 3 cps.
Amendment No. (T, 191, 102 i
-205-
,y
~. _. _ _
SURVEILLANCE REQUIREMENTS LIMITING' CONDITIONS FOR OPERATION -
.12 Additional Safety Related Plant 4
3.12 Additional Safety Related Plant Capabilities Capabilities Applicability _:
Applies to the surveillance require-Applies to the operating status of the ments for the main control room venti-main control room ventilation system, lation system, the reactor building the reactor building closed cooling closed cooling water system and the water system and the service water service water system which are required by the corresponding Limiting Conditions system.
for Operation.
Objective:
l Objective:
To verify that operability or availa-e To assure the availability of the main bility under conditions for which these control room ventilation system, the capabilit:es are an essential response reactor building closed cooling water to station abnormalities.
system and the service water system upon the conditions for which the capability is an essential response to station abnormalities.
Main Control Room Ventilation A.
Main Control Room Ventilation A.
At least once per operating cycle, the 1.
Except as specified in Specification pressure drop across the combined HEPA 1.
3.12.A.3 below, the control room air filters and charcoal absorber banks treatment system, the diesel shall be demonstrated to be less than l'
generators required for operatiran of 6 inches of water at system design flow this system and the main control room rate.
air radiation monitor shall be oper-fi able at all times when containment 4
integrity is required.
7 j
2.a. The tests and sample analysis of 2.a. The results of the in-place cold DOP Specification 3.12.A;2 shall be per-T i t
l 1
and halogenated hydrocarbon tests formed at least once per year for at 3 design flow (341 CFM) and at standby service or after every 720
[
, control room pressure on HEPA fil-hours of system operation and fol-ters and charcoal adsorber banks lowing significant painting, fire respectively shall show > 99% DOP or chemical release in any ventila-removal and > 99% halogenated tion zone communicating with the '
hydrocarbon removal, system.
Cold DOP testing shall be performed b.
The results of laboratory carbon after each complete or r cial replace-b.
sample analysis shall show > 99%
ment of the HEPA filter bank or after radioactive methyl iodide removal any structural maintenance on the velocity with inlet conditiogs of:
system housing.
>22 FPM, >1.75 mg/m inlet iodide concentration, > 95% R.H. and $30*C.
Halogenated hydrocarbon testing shall c.
Each fan shall be shown to provide be performed after cach complete or c.
341 CFM t10%.
partial replacement of,the charcoal
~
absorber bank or after any structural maintenance on the system housing.
.m
-215-Amendment No. S/, SS, 191, 102
- :* P - -: (-;:, m... w;,
~
- =.*.=,?rqe,;p p ol
.9
~
- _~
l 3.1'2 BASES 6
l A.
Main Control Room Ventilation System l
The control room ventilation system is designed to filter the control room j
atmosphere for intake air and/or for recirculation during control room t
isolation conditions. The system is designed to automatically start upon control room isolation and to maintain the control room pressure to the design positive pressure so that all leakage.should be out leakage.
1
+
High efficiency particulate absolute (HEPA) filters are installed before the
)
charcoal adsorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential intake of radiciodine to j
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 1
and HEPA filters. The laboratory carbon sample test results should indicate l
1 a radioactive methyl iodide removal efficiency of at least 99 percent for expected accident conditions. If the performance of the HEPA filters and
{
charcoal adsorbers are as specified, the resulting doses will be less than the allowable levels stated in Criterion 19 of the General Design Criteria l
for Nuclear Power Plants, Appendix A to 10 CFR Part 50.
j If the system is found to be inoperable, there is no immediate threat to the control room and reactor operation may continue for a limited period i
of time while repairs are being made.
If the system cannot be repaired within seven days, the reactor is shutdown and brought to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Reactor Building Closed Cooling Water System i
l l
The reactor building closed cooling water system has two pumps and one heat exchanger in each of two loops. Each loop is capable of supplying g
the cooling requirements of the essential services following design j-(
accident conditions with only one pump in either loop.
J
- l The system has additional flexibility provided by the capability of inter-lp-connection of the two loops and the backup water supply to the critical l
loop by the service water system. This flexibility and the need for only i
one pump in one loop to meet the design accident requirements justifies i,
the 30 day repair time during normal operation and the reduced requirements j
l during head-off operations requiring the availability of LP,CI or the core
\\
j spray systems.
4
)
C.
Service Water System
}
j The service water system consists of four vertical service water pumps 3
located in the intake structure, and essociated strainers, piping, valving 1
and instrumentation. The pumps discharge to a common header from which I
independent piping supplies two Seismic Class I cooling water loops and one turbine building loop. Automatic valving is provided to shutoff all supply to the turbine building loop on drop in header pressure thus assuring supply to the Seismic Class I loops each of which feeds one diesel generator, two RHR i
service water booster pumps, one control rocu basement fan coil unit and one RBCCW I
1 I
Amendment No. 82, pp, 191, 102
-215d-
...y m y
--w,.,
.v.-
-e---.,e e
-w
,,w, gr r-w m w
,w--,-.w-,
y---- --s--
.-ww.---r-
.e
,r-g,- - - - - - <ee-*-*-m--w,,w
,,v+
,Wy=-
- v e w--
y e-g qv-
-1.-
a 6.2 (cont'd) f.
Investigate all violations of Technical Specifications, including reporting evaluation and recommendations to prevent recurrence, to the Vice President - Nuclear and to the Chairman of the NPPD Safety Review and Audit Board.
g.
Perform special reviews and investigations and render reports thereon as requested by the Chairman of the Safety Review and Audit Board.
h.
Review all reportable events specified in Section 50.73 to 10CFR I
Part 50.
i.
Review drills on emergency procedures (including plant evacuation) and adequacy of communication with off site groups.
l j.
Periodically review procedures required by Specifications 6.3.1, 6.3.2, 6.3.3, and 6.3.4 as set forth in administrative procedures.
5.
Authority j
a.
The Station Operations Review Committee shall be advisory.
b.
The Station Operations Review Committee shall recommend to tha l
l Division Manager of Nuclear Operations approval or disapproval of proposals under items 4, a through e and j above.
In case l
of disagreement between the recommendations of the Station r
Operations Review Committee and the Division Manager of Nuclear Operations, the course determined by the Division Manager of Nuclear Operations to be the more conservative will be followed.
A written summary of the disagreement will be sent to the Vice President - Nuclear and to the NPPD Safety Review and Audit Board.
The Station Operations Review Committee shall' report to the c.
/I Chairman of the NPPD Safety Review and Audit Board on all re-5 views and investigations conducted under items 4.f. 4.g. 4.h,
[_
and 4.1.
l d.
The Station Operations Review Committee shall make determinations i
regarding whether or not proposals considered by the Committee
[
involve unreviewed safety questions. This determination shall be subject to review by the NPPD Safety Review and Audit Board.
t 6.
Records:
\\
Minutes shall be kept for all meetings of the Station Operations Review Committee and shall include identification of all documen-tary material reviewed; copies of the minutes shall be forwarded to the Chairman of the NPPD Safety Review and Audit Board and the Vice President - Nuclear within one month.
7.
Procedures:
Written administrative: procedures for Committee operation shall be prepared and maintained describing the method for submission and content of presentations to the committee, provisions.for use of subcommittees, review and approval by members of 'ritten w
Committee evaluations and recommendations, dissemination of minutes, and such other matters as may be appropriate.
Amendment No. 50, $2, $J, $$, 100, 101, 102
-221-
- f t.g 7,
6.2 (Cont'd)
B.
NPPD Safety Review and Audit Board (SRAB)
Function: The Board shall function to provide independent review and audit of designated activities.
1.
Membership:
a.
Chairman b.
Vice-Chairman c.
Five Members d.
Consultants (as required) l The Board members shall collectively have the capability required to review problems in the following areas: nuclear power plant operations, nuclear engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering, quality assurance practices, and other appropriate fields associated with the l
unique characteristics of the nuclear power plant involved.
I When the nature of a particular problem dictates, special i
consultants will be utilized.
l Alternate members shall be appointed in writing by the Board Chairman to serve on a temporary basis; however, no more than two alternates shall serve on the Board at any one time.
l l
2.
Meeting frequency: Semiannually, and as required on call of the l
Chairman.
3.
Quoram: Chairman or Vice Chairman, plus four members including alternates. No more than a minority of the quorum shall be from groups holding line responsibility for the operation of the plant.
[
f 4.
Review: The following subjects shall be reported to and reviewed by the NPPD Safety Review and Audit Board.
e
\\
The safety evaluations for 1) changes to procedures, equipment a.
or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
b.
Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
Amendment No. 50, 82, 85', 100, 101, 102 l
-222-2xx< x
d
- 5. 2* (cont'd) c.
Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
Proposed changes to Appendix A Technical Specifications or d.
r the CNS Operating License.
e.
Violations of applicable codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.
f.
Significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear j
safety.
l g.
All reportable events specified in Section 50.73 to 10CFR j
Part 50.
i h.
Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.
I 1.
Minutes of meetings of the Station Operations Review Committee.
J
- j. Disagreement between the recommendations of the Station Operations Review Committee and the Division Manager of Nuclear Operations.
k.
Review of events covered under e,f,g, and h above include reporting to appropriate members of management on the results of investiga-i tions and recommendations to prevent or reduce the probability 1
of recurrence.
l f" 5.
Authority: The NPPD Safety Review and Audit Board shall report to and j
be advisory to the Vice President'- Nuclear on'those areas of j
responsibility specified in Specifications 6.2.1.B.4 and 6.2.1.B.7.
i~~
6.
Records:
l<
Minutes shall be recorded for all meetings of the NPPD Safety l
Review and Audit Board and shall identify all documentary i:
material reviewed. Copies of the minutes shall be forwarded to j
the Vice President - Nuclear and.the Division Manager of Nuclear \\
Operations, and such others as the Chairman may designate within
[
one month of the meeting.
7.
Audits:
i l
Audits of selected aspects of plant operation shall be performed under the cognizance of SRAB with a frequency commensurate with their safety significance. Audits performed by the Quality Assurance Department which meet this specification shall be considered to meet the SRAB audit requirements if the audit results r
i are reviewed by SRAB. :A representative portion of procedures and records of the activities performed during the audit period shall be audited and, in a'ddition, observations of performance of. operating and maintenance activities shall be included. These audits shall encompass:
Amendment No. 89, 82, 85, 191, 102
,,3x7g,l
-223-n-
-,.-,y
.m
-,-n
.-,,r-
- 6. 2*
(cont'd) a.
Verification of compliance with internal rules, procedures (for example: normal, off-normal, emergency, operating, maintenance, surveillance, test, and radiation control procedures) and applicable license conditions at least once per 24 months.
b.
The training, qualification, and performance of the operating staff at least once per 24 months.
c.
The Emergency Plan and implementing procedures at least once per 12 months.
d.
The Security Plan and implementing procedures at least once per 12 months.
1 I
e.
The facility fire protection and its implementing procedures at j
least once per 24 months.
f.
A fire protection and loss prevention inspection will be performed j
utilizing either qualified off-site licensee personnel or an out-i side fire protection consultant at least once per 12 months.
i I
l l
g.
An inspection and audit by an outside qualified fire protection j
consultant shall be performed at least once per 36 months.
i j
h.
The Radiological Environmental Monitoring Program and the Offsite Dose Assessment Manual with their implementing procedures at least once every 24 months.
t -
o
\\
Amendment No. 52, $5, $$, 19@, 191, 102
-224-f.l
.,xgy
6.3 PROCEDURES AND PROGRAMS 6.3.1 Introduction Station personnel shall be provided detailed written procedures to be used for operation acd maintenance of system components and systems that could have an effect on nuclear safety.
6.3.2 Procedures Written procedures and instructions including applicable check off lists shall be established, taplemented, and maintained for the following:
A.
Normal startup, operation, shutdown and fuel handling operations of the station including all systems and components involving nuclear safety.
B.
Actions to be taken to correct specific and foreseen potential or actual malfunctions of safety related systens or components including responses to alarms, primary system leaks and abnormal reactivity changes.
C.
Emergency conditions involving possible or actual releases of radio-active materials.
D.
Implementing procedures of the Security Plan and the Emergency Plan.
E.
Implementing procedures for the fire protection program.
F.
Administrative procedures for shift overtime.
G.
Implementing procedures for the Offsite Dose Assessment Manual.
6.3.3 Maintenance and Test Procedures The following maintenance and test procedures will be provided to satisfy routine inspection, preventive maintenance programs, and operating license requirements.
1 A.
Routine testing of Engineered Safeguards and equipment as required by the facility License and the Technical Specifications.
B.
Routine testing of standby and redundant equipment.
C.
Preventive or corrective maintenance of plant equipment and systems j
that could have an effect on nuclear safety.
D.
Calibration and preventive maintenance of instrumentation that could affect the nuclear safety of the plant.
E.
Special testing of equipment for proposed changes to operational procedures or proposed system design changes.
6.3.4 Radiation Control Procedures Radiation control procedures shall be maintained and made available to all station personnel. These procedures shall show permissible radiation exposure, and shall be consistent with the requirements of 10 CFR 20.
Amendment No. 82, 85, 89, 100, 101, 102 l
-225-
.' 6.3 (cont'd)
A.
High Radiation Areas In lieu of the " control device" or " alarm signal" required by Paragraph 20.203 (c) (2) of 10 CFR 20 each High Radiation Area (100 mrem /hr or greater) shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by requiring notification and permission of the shift supervisor. Any individual or group of indi-viduals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.
i 6.3.5 Temporary Changes Temporary changes to procedures which do not change the intent of the f
original procedure may be made, provided such changes are approveo by two members of the operating staff holding SR0 licenses. Such changes shall be documented and subsequently reviewed by the Division Manager l
of Nuclear Operations within one month.
i i
6.3.6 Exercise of Procedures I
Drills of the Emergency Plan procedures shall be conducted annually, including a check of communications with offsite support groups. Drills l
on the procedures specified in 6.3.2.A, B, and C above shall be con-ducted as part of the retraining program.
6.3.7 Programs The following-programs shall be established:
A.
Systems Integrity Monitoring Program A program shall be established to reduce leakage to as low as practical levels from systems outside the primary containment during a serious accident that would or could contain highly radioactive fluids. This program shall include provisions establishing preventive maintenance and periodic visual inspection requi,rements, and leak testing requirements for l
each system at a frequency not to exceed refueling cycle intervals.
j B.
Iodine Monitoring Program 4
A program shall be established to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include training of personnel, procedures for monitoring and provisions for maintenance of sampling and analysis i
equipment.
C.
Environmental Qualification Program A.
By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the, degree of compliance with the D0R Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.
Amendment No. 82, $3, 701, 102 yml
-226-I
.=
'j $6.T (cont'd)
D.
Post-Accident Sampling System (PASS)
A program shall be established to. ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. This program shall include training of personnel, procedures j
for sampling and analysis and provisions for operability of sampling and analysis equipment.
4 i
l
?
I l
i i
I i
1 1
l l
i I
I I
2 I
Amendment No. 90, 101, 102
-227*
,, n. ;.
.: a6.5.2 Rsportable Evsnts A Reportable Event shall be any of those conditions specified in 2
Section 50.73 to 10CFR Part 50.
The NRC shall be notified and a report submitted pursuant to the requirements of Section 50.73.
Each Reportable Event shall be reviewed by SORC and the results of this review shall be submitted to SRAB and the Vice President -
Nuclear.
1 1
b l
l l
I i
I i
e i
u I
I Amendment No. 86, 89, 191, 102
-234-
..,3 y.,7, y
.___.,