ML20138F668
| ML20138F668 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 12/10/1985 |
| From: | NEBRASKA PUBLIC POWER DISTRICT |
| To: | |
| Shared Package | |
| ML20138F625 | List: |
| References | |
| NUDOCS 8512160148 | |
| Download: ML20138F668 (22) | |
Text
4
-~'
1.
o'
' Revised Technical Specifications for Tables 3.2.A and 4.2.B. MCPR,'
Jet Pump Flow Mismatch Turbinc.Stop Valve Closure Scran Bypass Revised Pages:
11 73 41 102 50 137 71 151
.The attached proposed change contains clarifying changes to areas of the'
.present Technical Specifications as follows:
- 1.. Tables'on Pages 50.-71, and 73 of the present Technical Specifications do not accurately reflect the numbers of the instruments they refer to.
As
~
a result, the numbers have been changed to ensure their consistency as
~
. contained on the Equipment Data File on the District's database.
Additionally, one instrument originally contained on Table 4.2.3 (page 71) had been omitted by. error on another amendment request, and this change reinstates this item.
2.
A change has been made tio Section 3.5 cf the bases on page 102 of the Technical Specifications. 'This change is made to bring this section of the Technical Specifications into agreement ~with the wording used in.
another area of this same page. Additionally,. reference to the station' superintendent-title has been changed to " Division' Manager of Nuclear Operations" to ensure consistent use of titles.throughout the Technical Specifications.
-3..
.The present Technical Specifications in Section 3.6.F-(pages 137 and 151) discuss " Jet Pump Flow Mismatch", but the section is actually referring to recirculation - pump flow mismatch.
The present title'and contained text is not clear as to what this section discusses. As a result, the attached change-is proposed in order to ensure the clarity and purpose of
~
Section 3.6.F.
4.
A change to page 41 in the bases has been made in order to clarify the turbine stop valve closure scram bypass. No change.is made in setpoints or. operational procedures. -The change is made just to clarify this section.
Evaluation of this Revision with Respect to 10CFR50.92 A.
- The enclosed Te'chnical Specification change -is judged to involve no significant hazards based on the following:
1.=
Does the proposed. license amendment involve a significant increase in. the probability or consequences of an accident previously, evaluated?
Evaluation:
These proposed changes do not change existing equipment, surveillances,- or procedures.
The changes add clarity to the present Technical Specifications by rewording existing sections that are not exactly clear, or make changes to component numbering that e
8512160148 851210 PDR ADOCK 05000298:
p -.
=_
t are not, correct.
The intent and purpose of the sections.do not change.. As :a result. it ' is the District's assessment that this change doeslnot-increase the probability or consequences of an accident previously evaluated.
-2..
Does.the proposed license amendment create the possibility for_a new or different kind of accident from any accident previously evaluated?
~
Evaluation:
~ This proposed change.does not make any changes to the present mode of ' operation, but adds clarity to ambiguous Technical Specifications. Therefore, the proposed license amendment does not s create the possibility of,a new or different kind of accident from.
any accident previously evaluated.
- 3. Does the proposed amendment involve. a' significant reduction in a margin of safety?
Evaluation:
The proposed change does not affect any equipment or procedures that would affect a margin of safety, but provide more clarity in the operating Technical Specifications. This factor makes the Technical Specifications clearer and while not-providing a clear increase in a margin of. ' safety,. does not result -in a reduction in a margin of safety.
B.
Additional-basis for proposed no significant k hazards consid'eration determination:
The commission has provided. guidance concerning the application of the-standards for determining whether a significant~ hazards consideration exists by providing certain examples (48CFR14870). The examples include:
"(i) A _ purely administrative change to Technical Specifications.
For example, a change to achieve consistency throughout the, Technical
. Specifications, correction of an error, or a change in nomenclature."
It.
is the District's belief the proposed change is encompassed by the above example.
J h
c TABLE OF CONTENTS (cont'd)
Page No.
SURVEILLANCE LIMITING CONDITIONS FOR OPERATION REQUIREMENTS
~ 3.5 CORE AND CONTAINMENT COOLING SYSTEMS 4.5 114 - 131 A.
Core Spray and LPCI Subsystems A
114 B.
Containment Cooling Subsystem (RHR Service Water)
B 116 lC.
HPCI Subsystem C
117 D.
RCIC Subsysten D
118 E.
Automatic Depressurization System E
119 F.
Minimum Low Pressure Copling System Diesel
' Generator Availability F
120 G.
Maintenance of Filled Discharge Pipe G
122 H.
Engineered Safeguards Compartments Cooling H
123
~
3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A.
Thermal and Pressurization Limitations A
132 B.
Coolant Chemistry B
133a C.
Coolant Leakage C
135 D.
Safety and Relief Valves D
136 E.
Jet Pumps E
137 F.
Recirculation Pump Flow Mismatch F
137 l
G.
Inservice Inspection G
137 H.
Shock Suppressors (Snubbers)
H 137a 3.7 CONTAINMENT SYSTEMS 4.7 159 - 192 A.
159 B.
Standby Gas Treatment System B
165 C.
165a D.
Primary Containment Isolation Valves D
166 3.8 MISCELLANEOUS RADI0 ACTIVE MATERIAL SOURCES 4.8 185 - 186 3.9
' AUXILIARY ELECTRICAL SYSTEMS 4.9 193 - 202 A.
Auxiliary Flactrical Equipment A
193 B.
Operation with Incperable Equipment B
195 3.10 CORE ALTERATIONS 4.10 203 - 209 A.
Refueling Interlocks A
203 B.
Core Monitoring B
205 C.
Spent Fuel Pool Water Level C
205 D.
Time Limitation D
206 E.
Spent Fuel' Cask Handling E
206 3.11 FUEL RODS 4.11 210 - 214e A.. Average Planar Linear Heat Generation Rate (APLHGR)
A 210 B.
Linear Heat Generation Rate (LHGR)
B 210 C.
Minimum Critical Power Ratio (MCPR)
C 212 t
LIMITING CONDITIONS FOR OPERATION SURVEILLANCE RE0UIREMENTS r
3.1 BASES (Cont.d) 4.1 BASES (Cont.d) ence paragraph VII.5.7 FSAR). Thus zero flow signal will be sent to half the IRM System is not required in of the APRM's resulting in a half the "Run" mode. The APRM's cover scram and rod block condi 5 n.
- Thus,
(
only the power range. The IRM's if the calibration were perrormed dur-
)
and APRM's provide adequate coverage ing operation, flux shaping would not I
in the startup and intermediate range, be possible. Based on experience at other generating stations, drift of The requirement to have the scram instruments, such as those in the functions indicated in Table 3.1.1 Flow Biasing Network, is not signifi-operable in the Refuel mode assures cant and therefore, to avoid spurioca that shifting to the Refuel mode scrams, a calibration frequency of i
during reactor power operation does each refueling outage is established.
not diminish'the protection provided by the reactor protection system.
Group (C) devices are active only dur-ing a given portion of the operational Turbine stop valve scram occurs at cycle. For example, the IRM is active 10% of valve closure. Below 30% of during startup and inactive during the rated turbine first stage full-power operation. Thus, the only pressure, the scram signal due to test that is meaningful is the one the turbine stop valve closure may performed just prior to shutdown or be bypassed because the flux and startup; i.e.,
the tests that are pressure scrams are adequate to performed just prior to use of the i
protect the reactor. The actual instrument.
scram bypass setpoint. however, is implemented at <25% of rated Calibration frequency of the instru-turbine first stage pressure (or ment channel is divided into two 179 psig).co compensate for groups. These are as follows:
possible turbine trips during bypass valve testing. During 1.
Passive type indicating devices bypass valve testing, the first that can be compared with like j
stage pressure is reduced due to units on a continuous basis.
flow through the bypass valves i
while reactor power is unchanged.
2.
Vacuum tube or semi-conductor dev,1ces and detectors that Turbine control valves fast' closure drift or lose sensitivity.
initiates a scram based on pressure switches sensing Electro-Hydraulic Experience with passive type instru-l Control (EHC) system oil pressure.-
ments in generating stations and sub-The switches are located on the stations indicates that the specified Control Valve Emergency Trip oil calibrations'are adequate. For those header, and detects the loss of devices which employ amplifiers, etc.,
oil to hold the valves open.
drift specifications call for drift to be less that 0.4%/ month; i.e.,
in This scram signal is also bypassed the period of a month a maxinum drift when turbine first stage pressure of 0.4% could occur, thus providing is less than 179 psig.
for adequate margin.
The requirements that the IRM's be in-serted in the core when the APKM's read 2.5 indicated on the scale in the Startup and Refuel modes assures that -
COOPER NUCLEAR STATION TABLE 3.2.A (Page 1)
PRIMARY CONTAINMENT AND REACTOR VESSEL ISOLATION INSTRUMENTATION Minimum Number Action Required of Operable
'When Component Instrument Components Per.
Operability is Instrument I.D. No.
Setting Limit Trip System (1) Not Assured (2)
Main Steam Line High RMP-RM-251, A,B,C,6D
$ 3 Times Full Power 2
A'or B Rad.
Reactor Low Water Level NBI-LIS-101, A,B,C,&D #1 1+12.5" Indicated Level 2(4)
A or~B l
Reactor Low Low Water NBI-LIS-57 A & B #2 1-37" Indicated Level 2
A or B Level NBI-LIS-58 A & B #2 Reactor Low Low Low Water NBI-LIS-57 A & B #1 3-145.5" Indicated Level 2
A or B Level NBI-LIS-58 A & B #1 Main Steam Line Leak MS-TS-121, A B.C,6D
< 200*F 2(6) 3 Detection 122, 123,.124, 143, 144, 145, 146, 147, 148, 149, 150 Main Steam Line High MS-dPIS-Il6 A,B,C,6D
< 140% of Rated Steam 2(3)
B Flow 117, 118, 119 Flow Main Steam Line Low MS-PS-134, A,B,C,6D 3 825 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
< 200% of System Flow 1
C System High Flow
COOPER NUCLEAR STATION
' TABLE 4.2.B (Page 2)
RHR SYSTEM TEST & CALIBRATION FREQUENCIES
~
Functional Instrument Item Item I.D. No.
Test Freq.
Calibration Freq.
Check Instrumentation 1.
Drywell liigh Pressure PC-PS-101, A, B, C & D Once/ Month (1) Once/3 Months.
None 2.
Reactor Vessel Shroud Level NBI-LITS-73, A & B #1 Once/ Month (1) Once/3 Months once/ Day 3.
Reactor Low Pressure RR-PS-128 A & B Once/ Month (1) Once/3 Months None 4.
Reactor Low Pressure NBI-PS-52 A & C Once/ Month (1) Once/3 Months None NBI-PIS-52 B & D 5.
Drywell Press.-Containment PC-PS-Il9, A,B,C.& D Once/ Month.(1) Once/3 Months None Spray 6.
RHR Pump Discharge. Press.
RHR-PS-120, A,B,C & D-Once/ Month (1) Once/3 Months None 7.
RHR Pump Discharge Press.
RHR-PS-105, A,B,C & D Once/ Month (1) Once/3 Months None 8.
RHR Pump Low Flow Switch RHR-dPIS-125 A & B Once/ Month (1) Once 3 Months None 9.
RIIR Pump Start Time Delay RHR-TDR-K70, A & B Once/ Month '(1) Once/Oper. Cycle None
- f 10.
RilR Injection Valve Close T.D.
RHR-TDR-K45 1A & IB Once/ Month (1) Once/Oper. Cycle None l
8 11.
RilR Pump Start Time Delay RHR-TDR-K75, A & B Once/ Month (1) Once/Oper. Cycle None 12.
RIIR Heat Exchanger Bypass T.D.
RHR-TDR-K93, A & B Once/ Month (1) Once/Oper. Cycle None 13.
RilR Cross Tie Valve Position RHR-LMS-2 Once/ Month (1)
N.A.
14.
Low Voltage Relays 27 X 3/lA (7)
None 15.
Low Voltage Relays 27 X 3/lB (7)
None 16.
Low Voltage Relays 27 x 2/lF, 27 X 2/lG (7)
None 17.
Low Voltage Relays 27 X 1/lF, 27 X 1/lG (7)
None 18.
Pump Disch. Line Press. Low CM-PS-266, CM-PS-270 Once/3 Months once/3 Months None 19.
Emergency buses Undervoltage 27/lF-2, 27/lFA-2, 27/1C-2, Once/ Month once/18 Months once/12 hrs.
Relays (Degraded Voltage) 27/lGB-2 20.
Emergency Buses Loss of 27/lF-1, 27/lFA-1, 27/lG-1, once/ Month once/18 Months once/12 hrs.
Voltage Relays 27/lGB-1, 27/ET-1, 27/ET-2 21.
Emergency Buses Undervoltage 27X7/lF, 27X7/lG Once/ Month Once/18 Months None Relays Timers
a
- COOPER' NUCLEAR STATION TABLE 4.2.B'(Page 4)
HPCI-TEST & CALIBRATION FREQUENCIES Functional.
Instrument Item
-Item I.D. No.
Test Freq.
Calibration Freq.
Check 1.
Reactor Low Water Level NBI-LIS-72, A,B,C, & D, f3 Once/ Month (1) Once/3 Months Once/ Day 2.
Reactor High Water Level NBI-LIS-101, (B & D #2)
Once/ Month (1) Once/3 Months Once/ Day l
3.
High Drywell Pressure' 14A -- K5 A & B (7)
(7) None 14A - K6 A & B
. (7)
(7) None 4.
HPCI Turbine High Exhaust HPCI-PS-97 A & B Once/ Month (1) Once/3 Months None
. Press.
- 5..HPCI Pump Low Suction Press.
HPCI-PS-84-1 Once/ Month (1) Once/3 Months None 6.
HPCI Pump Low Discharge Flow HPCI-FS-78.
Once/ Month -(l) Once/3 Months None 7.
HPCI Low Steam Supply Press.
HPCI-PS-68, A,B,C, &.D Once/ Month (1) Once/3 Months None 8.
HPCI Steam Line High AP HPCI-dPIS-76 Once/ Month (1) Once/3 Months None HPCI-dPIS-77 Once/ Month (1) Once/3 Months None 9.
HPCI Steam Line Space High HPCI-TS-101, A,B,C, & D Once/ Month (1) Once/Oper. Cycle None Temp.
102, 103, 104, HPCI-TS-125, 126, 127, 128 RHR-TS-150,151,152,153,154, d
155,156,157,158,159,160,161 8
~10.
Emergency Cond. Stg. Tk. Low HPCI-LS-74 A & B Once/ Month (1) Once/3 Mor.. hs None Level HPCI-LS-75 A & B Once/ Month (1) Once/3 Months None 11.
Suppression Chamber High HPCI-LS-91 A & B Once/ Month (1) Once/3 Months None Water Level 12.
IIPCI Gland Seal Cond. Hotwell HPCI-LS-356 B Once/ Month (1) Once/3 Months None Level HPCI-LS-356 A Once/ Month (1) Once/3 Months None 13.
HPCI Control Oil Pressure Low HPCI-PS-2787-H Once/ Month (1): Once/3 Months None HPCI-PS-2787-L once/ Month (1) Once/3 Months None 14.
Turbine Condition Supr. Alarm HPCI-TDR-K14 Once/ Month (1) Once/Oper. Cycle None Actuation ?'mer 15.
Pump Disch. Line Low Press.
CM-PS-268 Once/3 Months once/3 Months None 16.
HPCI Turbine Stop Valve Mon.
HPCI-LMS-4 Once/ Month N.A.
None 17.
Sup. Chamber HPCI Suction Viv.-
HPCI-LMS-2 Once/ Month N.A.
None 18.
HPCI Steam Line High AP HPCI-TDR-K33, Once/ Month once/Oper. Cycle. None Actuation Timer HPCI-TDR-K43 Once/ Month once/Oper.-Cycle. None Logic (4)(6) 1.
Logic Bus Power Monitor-Once/6' Months N.A.
. 2.
HPCI Initiation Once/6 Months N.A.
3.
HPCI Turbine Trip Once/6 Months N.A.
?
3.3 and-4.3 BASES:
(Cont'd) 5.
.The Rod Block Monitor-(RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high power level operation. Two channels are pro-vided, and one of these may be bypassed from the console for maintenance and/or testing. Tripping of one of the channels will block erroneous rod withdrawal soon enough to prevent fuel damage. This system backs up the operator who withdraws control rods according to written se-quences. The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod with-drawal errors when this condition exists.
A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit (i.e., MCPR equals the safety limit, l
and LHGR = as defined in 1.0.A.4).
During use of such patterns, it is judged.that testing of the RBM system prior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur. It is the responsibility of the Reactor Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable control rods in other than limiting patterns. Other person-nel qualified to perform this function may be designated by the Division 1 Manager of Nuclear Operations.
C.
Scram Insertion Times The control rod system is designed to bring the reactor suberitical at a rate fast enough to prevent fuel damage; i.e.,
to prevent the MCPR from becoming less than the safety limit. The limiting power transient is defined in Reference 3.
Analysis of this transient shows that the negative reactivity rates resulting from the scram provide the required protection, and MCPR remains greater than the safety limit.
The surveillance requirement for scram testing of all the control rods after each refueling outage and 10% of the control rods at 16-week intervals _is adequate for determining the operability of the control rod system yet is not so frequent as to cause excessive wear on the
. control rod system components.
The numerical values assigned to the predicted scram performance are based on the analysis of data from other BWR's with control rod drives the same as those on Cooper Nuclear Station.
The occurrence of scram times within the limits, but significantly longer than the average, should be viewed as an indication of a systematic problem with control rod drives.
In the analytical treatment of the transients which are assumed to scram on high neutron flux, 290 milliseconds are allowed between a neutron sensor reaching the scram point and start of motion of the control rods. This is adequate and conservative when compared to the typical time delay of about 210 milliseconds estimated from scram test results. Approximately the first 90 milliseconds of each of these time intervals result from the sensor and circuit delays; at this point, the pilot scram solenoid deenergizes. Approximately 120 milliseconds later,
-102-
- h LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS i
'3.6.E Jet Pumps 4.6.E.
Jet Pumps
'1.
Whenever the reactor is in the start-1.
Whenever there is recirculation flow up or run modes, all jet pumps shall with the reactor in the startup or be operable. If it is determined run modes, jet pump operability shall that a jet pump is inoperable, or be checked daily by verifying that the if two or more jet pump flow in-following conditions do not occur sim-struments failures occur and cannot
.ultaneously:
be corrected within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an orderly shutdown shallibe initiated a.
The recirculation pump flow differs and the reactor shall be in a Cold by more than 15% from the established
- Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, speed ficw characteristics.
~
b.
The' indicated value of core flow rate varies from the value derived from loop flow measurements by more than 10%.
c.
The diffuser to lower plenum differen-tial pressure reading on an individual jet pump varies from the mean of all jet pump differential pressures by more than 10%.
. l F.-
Recirculation Pump Flow Mismatch F.
Recirculation Pump Flow Mismatch l
.1.
-Deleted.
1.
Deleted.
l ' 2.
Following one recirculation pump operation, the discharge valve of l
the low speed recirculation pump may not be opened unless the speed of the faster pump is equal to or less than 50% of its rated speed.
G.
Inservice Inspection G.
~ Inservice Inspection
. Inservice inspection shall be per-To be considered operable, com.
formed in accordance with the ponents shall satisfy,the require-requirements for ASME Code Class 1, ments contained in Section XI of.
2, and 3 components contained in
~
the ASME Boiler and Pressure Vessel Section XI of the ASME Boiler and Code and' applicable Addenda for.
-Pressure Vessel Code and applicable continued service of ASNE Code Addenda as required by 10 CFR 50,
. Class 1, 2, 'and 3 components except Section 50.55a(g), except where where relief has been granted by the relief has been granted by the Commission pursuant to 10 CFR 50 Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
Section 50.55a(g)(6)(1).
-137-
,n.-
L
.3.6.E'& 4.6.E BASES- (Cont'd) jet pump body; however, the converse is not true. The lack of any substantial stress in the jet pump body makes failure impossible without an initial nozzle riser system failure.
F.
Recirculation Pump Flow Mismatch g
Requiring the discharge valve of the lower speed loop'to remain closed until
-the speed offfaster pump is equal to or less than 50% of its rated. speed provides assurance when' going from one to two recirculation pump operation that excessive l
' vibration of the j et pump risers will not occur.
G.
Inservice Inspection The inservice inspection. program conforms to the requirements of 10 CFR 50, Section 50.55a(g). 'Where practical, the inspection of components conforms to the requirements of ASME Code Class 1, 2, and-3 components contained in
.Section XI of the ASME Boiler and Pressure Vessel Code. If a Code required inspection is impractical, a request for a deviation from that requirement is submitted to the Commission in accordance with 10 CFR 50, Section 50.55a(g)(6)(1).
Deviations which are needed from the procedures prescribed in Section XI of the ASME Code and applicable Addenda will be reported to the Commission prior to the beginning of each 10-year 'nspection period if they are known to be required at that time. Deviations which are identified during the. course of inspection will be reported quarterly throughout the inspection period.
-151-
..n.,
i
- t l
. Revised Technical Specifications for Testable Penetrat' ions, Primary Containment Testable Isolation Valves Revised Pages:
171 173 174
~
Tables 3.7.2 and 3.7.4 list penetrations and primary containment isolation valves, respectively, which are testable.
It was noted that there was equipment which was being tested (by LLRT's), but that these items did not appear on Tables 3.7.2 and 3.7.4.
As a result, the tables have been revised to reflect the actual equipment that is tested.
Evaluation of this Revision with Respect to 10CFR50.92 A.
The enclosed Technical Specification change is judged.to involve no significant hazards based on the following:
1.
Does the proposed license. amendment involve a significant increase in the probability or consequences of an ac :ident previously evaluated?
Evaluation:
The proposed change does not change existing equipment, surveillances, or procedures, but updates t.hese tables to reflect the actual items tested. ~ It is therefore the District's assessment that these changes do not involve an increase in the probability or consequences of an accident previously evaluated.
2.
Does the proposed license amendment create the possibility for a new or different kind of ' accident from any ace'ident previously evaluated?
Evaluation:
The proposed change does not introduce any new mode of operation, but specifies more limiting testing requirements for components which further enforces the design objectives given in the USAR. On this basis the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3.
Does the proposed amendment involve a significant reduction in a margin of safety?
Evaluation:
This change adds to these tables testing requirements which, although they are presently subject to testing anyway, if not added to these tables, could reduce a margin of safety. Due to this fact, this change does not involve a reduction in a margin of safety but because it results in more stringent Technical Specifications will result in an improvement in a margin of safety.
L-B.
Additional basis for_ proposed no significant hazards consideration determination:
The commission has provided' guidance concerning the application of the
. standards for determining whether a significant hazards consideration exists by providing certain examples (48CFR14870). The examples include:
".(ii) A change that constitutes an additional limitation, restriction, or control,not presently included in the Technical Specifications." It is the District's. belief that the proposed change is encompassed by the above example.
O
f TABLE 3.7.2 TESTABLE PENETRATIONS ~WITH DOUBLE 0-RING SEALS
. PEN. NO-DESCRIPTION X-1A Drywell equ'ipmen't hatch X-1B Drywell equipment hatch X-2
- Drywell airlock door l
l X-4 Drywell. head access hatch X-6 CRD removal hatch X-35A TIP "D" Penetration X'-35B TIP "A" Panetration X-35C TIP "C" Penetration X-35D' TIP "B" Penetration X-35E' TIP N Purge Connection 2
-X-200A Suppression chamber access hatch X-200B.
Suppression chamber access hatch X-213B Suppression chamber drain flange Drywell head Stabilizer Assembly Inspection Ports (8) 4 4
-171-e s
+
r_., -., _, -.
f-..,.,._, _.,,..,...,
e.,_. -e.-..
.. ~.
- - ~ -,. - -
r i
y.wr e
+-
t TABLE 3.7.4 PRIMARY CONTAINMENT TESTABLE ISOLATION VALVES TEST PEN. NO.
VALVE NUMBERS MEDIA X-7A MS-AO-805 and MS-AO-86A, Main Steam Isolation Valves Air X-7B MS-A0-80B and MS-AO-86B, Main Steam Isolation Valves Air X-7C MS-AO-80C and MS-AO-86C, Main Steam Isolation Valves Air
~X-7D MS-AO-80D and MS-AO-86D, Main Steam Isolation valves Air.
X-8 MS-MO-74 and MS-M0-77, Main Steam Line Drain Air X-9A RF-15CV and RF-16CV, Feedwater Check Valves t
Air X-9A RCIC-A0-22, RCIC-MO-17, and RWCU-15CV, RCIC/RWCU Connection to Feedwater Air X-9B RF-13CV and RF-14CV, Feedwater Check Valves Air X-9B HPCI-A0-18 and HPCI-MO-57, HPCI Connection to Feedwater Air X-10 RCIC-MO-15 and RCIC-MO-16, RCIC Steam Line Air X-Il HPCI-M0-15 and HPCI-MO-16, HPCI Steam Line Air X-12 RHR-MO-17 and RHR-MO-18, RRR Suction Cooling' Air X-13A-RHR-MO-25A and RHR-MO-27A, RRR Supply to RPV Air X-13B RHR-M0-25B and RHR-MO-27B, RHR Supply to RPV Air
~
X-14 RWCU-MO-15 and RWCU-MO-18, Inlet to RWCU System Air X-16A.
CS-MO-11A and CS-MO-17A, Core Spray to RPV Air X-16B CS-MO-llB.and CS-MO-12B, Core Spray to RPV Air
- X-17 RHR-MO-32 and RHR-MO-33, RPV Head Spray Air X-18 RW-732AV and RW-733AV, Drywell Equipment Sump Discharge Air X-19 RW-765AV and RW-766AV, Drywell Floor Drain Sump Discharge Air
~X-25 PC-232MV and PC-238AV, Purge and Vent Supply to Drywell Air
.X-25 ACAD-1305MV and ACAD-1306MV, Supply to Drywell Air X-26 PC-231MV, PC-246AV, and PC-306MV Purge and Vent Exhaust l
from Drywell Air X-26 ACAD-1310MV, Bleed from Drywell Air
-173-
E t.
' '.h 7
E TABLE 3.7.4 (page 2)
PRIMARY CONTAINMENT TESTABLE-ISOLATION VALVES TEST PEN.'NO.
VALVE NUMBERS MEDIA LX-39A.
RHR-MO-26A and RHR-MO-31A, Drywell Spray Header Supply Air
.X-39B RHR-MO-26B and RHR-MO-31B, Drywell Spray Header Supply Air X-39B
- ACAD-1311MV and ACAD-1212MV, Supply to Drywell Air
-X-41 RRV-740AV and RRV-741AV, Reactor Water Sample Line Air X-42 SLC-12CV and SLC-13CV, Standb'y Liquid Control Air X-205 PC-233MV and PC-237AV, Purge and Vent Supply to Torus Air X-205 PC-13CV and PC-243AV, Torus Vacuum Relief Air X-205 PC-14CV and PC-244AV, Torus Vacuum Relief Air X-205 ACAD-1303MV and ACAD-1304MV, Supply to Torus Air X-210A RCIC-MO-27'and RCIC-13CV, RCIC Minimum Flow Line Air X-210A~
RHR-MO-21A, RHR to Torus Air X-210A RHR-MO-16A, RHR-10CV, and RHR-12CV, RHR Minimum Flow Line Air
.X-210B RHR-M0-21B, RHR to Torus Air-X-210B.
HPCI-17CV and HPCI-MO-25, HPCI Minimum Flow Line' Air X-210B RHR-MO-16B, RHR-llc'V, and RHR-13CV, RHR Minimum Flow Line Air X-210A and'211A RHR-MO-34A, RHR-M0-38A, and RHR-MO-39A, RHR to Torus Air X-210B and 211B1 RHR-MO-34B, RHR-MO-38B, and RHR-MO-39B, RHR to Torus' Air X-211B
'ACAD-1301MV and ACAD-1302MV, Supply to Torus Air X-212 RCIC-15CV and RCIC-37, RCIC Turbine Exhaust Air X-214 HPCI-15CV and HPCI-44, HPCI Turbine Exhaust Air X-214 HPCI-AO-70 and HPCI-AO-71, HPCI Turbine Exhaust. Drain Air X-214~
RHR-MO-166A ana RHR-MO-167A RHR Heat Exch. Vent Air
.X-214 RHR-MO-166B and RHR-MO-167B RHR Heat Exch. Vent Air-
-X-220 PC-230MV, PC-245AV, and PC-305MV Purge and Vent Exhaust Air l from Torus X-220 ACAD-1308MV, Bleed from Torus Air X-221 RCIC-12CV and RCIC-42, RCIC Vacuum Line Air X-222 HPCI-50 and HPCI-16CV, HPCI Turbine Drain Air
-174-
r-t 1.
e Attachmsnt 3
~
Revised Technical Specifications-for Section 6 Administrative Controls Editorial Changes
. Revised Pages: 220 237
-The District-~recently added two new management positions to its organization.
These. positions have been identified as the Maintenance Manager and the Outage & Modifications Manager..The. Maintenance Manager will be in charge of the Maintenance Group. (Electrical, Mechanical, Maintenance Scheduling).
The Operations Manager used to fulfill these' functions but his maintenance duties have been delegated to allow him more time in operations management.
The Outage.-& Modifications Manager will be responsible for outage-related and design modification activities. The attached change shows the revision to the Organization Chart as a result of these new positions, and the Maintenance Managers addition to' SORC.
Additionally, the Reactor Engineer Supervisor's title has been changed to Operations Engineering Supervisor. This c'..
e is a change in. title only and does ~ not alter this Supervisor's duties. ur responsibilities, and does not change CNS's organizational structure.
! Evaluation of this Revision with Respect to 10CFR50.92 A.
The. enclosed Technical Specification change is judged to involve no significant hazards based on the following:
1.
Does the proposed license amendment involve a.significant increase in the probability 'or consequences of an ' accident previously evaluated?
Evaluation:
The proposed. change does not change existing equipment, surveillances, or. operating procedures.
It is the District's opinion based on this fact, that this change does not affect the
. probability or consequences of an accident previously evaluated.
2.
Does the proposed license amendment create the possibility for a new or different kind of accident from any accident previously' evaluated?
Evaluation:
The proposed change does not alter our present modes of operation or create a new mode'of operation.
Therefore, the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.
3..
Does the proposed amendment involve a significant reduction in a margin of safety?
Evaluation:
This ~ proposed license amendment does change our administrative procedures by adding these Managers positions. - By dividing the k
I o
previous responsibilities of the Operations Manager, the District feels that both maintenance and operations can benefit from the increased attention.
Also, the addition of an Outage &
Modifications Manager will enhance the activities that take place in these E areas.
For these reasons it is felt that this increased attention will result in an improvement in the margin of safety.
B.
Additional basis for proposed no significant - hazards consideration determination:
The commission' has provided guidance concerning the application of the standards for determining whether.a significant hazards consideration exists by. providing certain examples (48CFR14870). The examples include:
"(ii) A change that constitutes an additional limitation, restriction, or control not presently included in the Technical Specifications." It is
~
the District's belief that the proposed change is encompassed by the above example.
r
- s
-4 3
). s '
is y'.
H
/
F 16.2
. REVIEW AND AUDIT a
a
-6.2.~1
-The organization and'dutie's of' committees for the review and audit of station operation sha'll be as outlined below:
- A.'
Station Operations Review Committee (SORC) 1.
' Membership:
Chairman:' Division, Manager of Nuclear Operations a.
b.
_ Technical Staff.' Manager
/
/
c.
Operations Manager d.
Technical Manager' e.
,0perations Supervisor
~
f.
.MaintenanceSupervisir g.
Instrument and Control Supervisor g
~h.
Chemistri and Health Physics' Supervisor, i.
Plant Engineering Supervisor j.
Operations Engineering Supervisor l
k.
Computer Applicatio'as Supervisor 1.-
Maintenance Managei' p
l m.
Quality Assurance Manager - non-voting member.
Alternate taembers shall be appointed in writing 'by the Division 7
-Manage'r'of Nuclear Operctlons to serve on a teaporary basis; however, no more than;two alternates shall. serve on the Committee at any one.
time.-
- 2..
Meeting Frequency: Monthly, and as required on call of the Chairman.'
3.
Quorum: Division Manager of Nuclear Operations or his designated alternate plus four other members including alternates.
4.
Responsibilities:,
a.
Review all. proposed normal, abnoru:al, maintenance and emergency-Joperating procedures specified in 6.3.1 f6.3.2, 6.3.3, and 6.3.4
.and proposed changes thereto: and any othe iproposed procedures.
. or changes thereto determined by any membei to'eflect nuclear' satfety.
b.
Review all' proposed tests and experiments and their results, which involve nuclear hazards not previcusly reviewed for conformance with technical specifications., Subhit tests which may constitute an unrev'iewed safety question to the NPPD Safety Review and Audit Board for review, t
c.
Review proposed changes to Technical Specifications.
d.
Review proposed changes or modifications to station systems or equipment-as discussed;in the SAR or'which-involves an unre -
viewed safety question as defined in '10CfR50.59(c). Submit
- changes to equipment or systems having safecy significance to the NPPD Safety Review and Audit Board for reviev;
}
e.
Review station operation to detect potential nuclear safaty hazards.
e
-220-)
k I
-~
..,,,.k
NUCLEAR OPERATIONS CIVI SION MANAGER s
STATION OPERATIONS FIRE PROTECTION REVIEW COMMITTEE ENGINEER M (SORC)
I TECH NIC A L QUALITY ASSURANCE STAFF l
MANAGER - CNS MANAGER l
l l
l l
l 1
I 1
T RAINING TECHNICAL MAWTE NANCE OPERATIONS OUTAGE &
ADMINISTRATIVE i
MOOlFCATIONS SERVICES MANAGER MANAGER MANAGER MANAGER MANAGER MANAGER a
d l
l i
OPERATIONS SPEC 1AL MAINTENANCE SUPERVISOR SECURITY AOMINIS TR ATIVE M ATE RIAL ERVISO R E
ER I
SUPERVISOR SUPERVISOR SUPERVISOR i
tu i
W J
N CHEM & HP COMPUTER PLANT OPERATlDNS APPLICATIONS ENGINEER ENGINEERING SUPERVISOR SUPER VISOh SUPERVISOR SUPERVISOR i
l I
I8C OPERATIONS 1/8 ONE/ SHIFT "a$0N SUPER VISOR 2/S TWO/ SHIFT l
t 3/S THREE/ SHIFT l
l RO. NRC REACTOR OPERATORS LICENSE SHIFT SRO-NRC SENIOR REACTOR OPERATORS l
l SUPERVISORS (SRCl i
i UCENSE I
I t/s W-FUNCTIONAL POSITION ONLY ELECTRICAL MAINTENANCE MECHANICAL
.'l
{
PHYSICALLY LOCATED IN GENERAL OFFICE, p
2/Sl STATION OPER.
UNIT OPER (RO)
SUPERVISOR SCHEDULER SUPERVISOR y3 1 (UNLICENSED)
Figure 6.l.2 NPPD Cooper Nuclear S tation Organization Chort i
[
t j
i i
l l
l
i Attachmznt.4 s
Revised Technical Specifications for' I
- Reactor Water Level Indication
/
.g.
N Revised Pages:
10 r
Figure 2.1.1 '(page 10) o f.- the Technical Specifications shows that ? level indication' (remote and -local) and RHR -interlock functions are provided by Yarway transmitters.
Recently a new. Rosemount pressure transmitter was -
installed that used the-same taps as the Yarway transmitter.
The new Rosemount; pressure transmitter will provide remote : level--indication on -
N, '
Pane 12 9-3 -in the control room (which the Yarway used to provide), while the Yarway.still provides local indication and.RHR interlock functions. This was done '. because the Yarway was susceptible to electrical " noise", would not -
' i' calibrate within the system accuracy.of 13.5% of span, and because of low AC
~
. signal level:the. signal was not repeatable.
The new transmitter meets the 3
requirements of IEEE 323-1974 and is classified IE Essential, which results in F -
a more reliable' transmitter.
' Evaluation of this Revision with Respect to 10CFR50.92
'A. -
The. enclosed Technical Specification change is -judged to involve no' significant hazards based on the following:
1.
Does the proposed license amendment involve a significant increase in the probability or Econsequences of an accident previously evaluated?
e Evaluation:
The change of transmitters to;a 1E qualified component will increase the reliability of-the, remote 'indicar ina.
Because1of this the proposed change does ' n'ot 4nvolve a sisnificant increase in the probability or consequences of ee. e ela it previously evaluated.
\\.
is.
3
'2.
Does the proposed license amena(_at w si:e the possibility for a new or different kind ' of ' a'ecident - from any accident previously evaluated?
,1
/
Evaluation:
The instrumentation being changed provides indication only, and
.because of increased reliability will not create the possibilitrifor a new or different kind of accident from any accident previously evaluated.
3.
Does the proposed amendment involve a signifhant reduction in a margin of. safety?'
Evaluation:
-Indication reliability will be improved by a factor of 3 which will enhance the margin of safety.
i
cf _; ; f o' ' '(,.
t s.
s
- ,t._ l B.
Additional ~ basis for proposed no. significant hazards consideration determination:
The commission has provided guidance concerning the application of the standards for-determining whether a significant hazards consideration
-,/h i.
exists by providing certain examples (48CFR14870). The examples include:
1
"(1) A purely administrative change to Technical Specifications.
For example, a change to achieve consistency throughout the Technical Specifications, correction of an error, or a change in nomenclature."
It-
.is_the District's' belief the proposed change is encompassed by the above example.
s.
p.:3 -
.k s
a s
e e ')
e GEMAC 916.75"
-400" IN.
848.75" - 332" TAP 80 6-750"-
REACTOR 722.25" VESSEL
T LINE 600" YARWAY GEM C 5
BARTON YARWAY ROSEMOUNT
+ 60" ' - - - - - -
- HPCI S :- +60"- - - +60" 576.75" - + 60"
-575.25(8) 158 818) 42.5 %HIALARM 552.56"- +200" 55 6--544.25(4) 27.5'T4)- 1.0 ALARM IS.
- 9,5 (3)
.-12.5"(3) 544.25" (4)
INSTRUMDg-3 0,,-- E 86
--0"-----0"- - 516.75
-O',
500"-
ZERO LIS - 101 tI-94 g9,4)
(-FEED WATM
_479.75(2)
-37"( 2 )
(9 5)
_455,O" INITIATE RCIC, HPCI, 450"-
TRIP RECIRC. PUMPS 400"-
-371.25 (I)
-L45.5" ( l )
-150" 7--tNITIATE RHR, C. S.,
. - 150" 366.75" TAF+ 352.56',-- 0" 350"-
352.56,,
START DIESE 8 CONTRIBUTE A. a S.
Bht Q (93t'-6")
LI-85 (9-5)
IN STRU MENT 31 5"- -39" d
2/
3Og 2
RACK CORE HEIGHT
( 25 -5 8 25-6)
PERMISSIVE W
252.56"- -10 6 25 1 5
LI-9i WATER LEVEL NOMENCLATURE
- (9 3) '
200"_
208.56" LEVEL _ HEIGHT ABCNE INSTRUMENT HT ABOVE
$E syag-NO.
VESSEL ZERO READING TAF (NOZZLE 161.5"{E
( in.)
(in.)
C 150 -
NOZZLE p --
(8) 575.25 58.5 222.69 (7) 559.25 42.5 206.69 (4) 544.25 27.5 191.69 100"-
(3) 529.25 12.5 176.69 (2) 479.75
-37 127.19 (1) 371.25
-14 5.5
- 18. G9 50"-
d.-
E LE VATION 917'-O" V ESSEL BOTTO M FIGURE 2.1.1 REACTOR WATER LEVEL INDICATION CORRELATION