ML20198J154

Forwards Amend 100 to License DPR-46 & Safety Evaluation. Amend Changes Tech Specs to Correct Error in Bases Which Defines Limiting Control Rod Pattern,To Reflect Organizational Changes

ML20198J154
Person / Time
Site:	Cooper
Issue date:	05/20/1986
From:	Long W Office of Nuclear Reactor Regulation
To:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
Shared Package
ML20198J158	List: ML20198J154
References
TAC-60331, NUDOCS 8606020185
Download: ML20198J154 (2)

Text

p UNITED STATES NUCLEAR REGULATORY COMMISSION a

wassiNoToN. o. c. 20sss g,

%...../

NEBRASKA PUBLIC POWER DISTRICT DOCKET N0. 50-298 COOPER NUCLEAR STATION AMEN 0 MENT TO FACILITY OPERATING LICENSE f

Amendment No.100 License No. DPR-46 t

i 1.

The Nuclear Regulatory Comission (the Comission) has found that:

j A.

The application for amendment by Nebraska Public Power District dated December 10, 1985, as supplemented by submittal dated January 13, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the

.i Comission's rules and regulations set forth in 10 CFR Chapter I; M

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the g

Comission; 1

C.

There is reasonable assurance (i) that the activities authorized t

by this amendment can be conducted without endangering the health

'l and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; 1

D.

The issuance of this amendment will not be inimical to the comon 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 Comission'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:

l 8606020185 860520 PDR ADOCK 05000298 i

P PDR l

. (2) Technical Specification The Technical Specifications contained in Appendices A and B, as revised through Amendment No.100, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION s'm sf h/A Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing

Attachment:

Changes to the Technical Soecifications Date of Issuance: May 20, 1986 9

ATTACHMENT TO LICENSE AMENDMENT NO. 100 1

FACILITY OPERATING LICENSE NO. DPR-46 DOCKET NO. 50-298 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines, i

Pages ii 10 1

41 50 71 73 102 137 151 171 173 174 220 221 222 i

224 l

225 231 236 237 1

1 l

l 1

TABLE OF COSTENTS (cont'd)

Pace No.

SURVEILLASCE LIMITING CONDITIONS FOR OPERATION REOUIREMENTS 3.5 CCRE AND CONTAINMENT COOLING SYSTDis 4.5 114 - 131 A.

Core Spray and LPCI Subsystems A

114 B.

Containment Cooling Subsystem (RNR Service Water)

B 116 C.

HPCI Subsystem C

117 D.

RCIC Subsystem D

118 E.

Autocatic Depressurization System E

119 T.

Minimum Low Pressure Cooling System Diesel Generator Availability T

120 G.

Maintenance of Tilled Discharge Pipe G

122 H.

Engineered Safeguards Compartments Cooling H

123 3.6 PRIMARY SYSTEM BOUNDARY 4.6 132 - 158 A.

Thersal and Pressurization Limitations A

132 3.

Coolant Chemistry B

133a C.

Ceolant Leakage C

135 D.

Safety and Relief Valves D

136 E.

Jet Pumps E

137 T.

Recirculation Pump Flow Mismatch T

137 l

C.

Inservice Inspection G

137 H.

Shock Suppressors (Snubbers)

H 137a 3.7 CONTAINMENT SYSTEMS 4.7 159 - 192 A.

Primary Containment A

159 B.

Standby Cas Treatment System 3

165 C.

Secondary Containment C

165a D.

Primary Containment Isolation Valves D

166 3.8 MISCELLANEOUS RADIOACTIVE MATERIAL SOURCES 4.8 185 - 186 3.9 AUXILIARY ELECTRICAL SYSTEMS 4.9 193 - 202 A.

Auxiliary Electrical Equipment A

193 B.

Operation with Inoperable Equipment B

195

3. C CCEI ALTERATIONS 4.10

'203 - 209 A.

Refueling Interlocks A

203 3.

Core Monitoring B

205 C.

Spent Tuel Pool Water Level C

205a D.

Time Limitation D

205a E.

Standby Cas Treatment System E

205a F.

Core Standby Cooling Systems T

205a C.

Control Room Air Treatment C

206 H.

Spent Tual Cask Handling H

206 3.11 FUIL RODS 4.11 210 - 214e k

A.

Average Flanar Linear Heat Generation Rate (APLHCR)

A 210 B.

Linear Heat Generation Rate (LHCR)

B 210 C.

Minimum Critical Power Ratio (MCPR)

C 212 D.

Thermal-hydraulic Stability D

212a Amendment flo. H,97.100 l

GEMAC 1 4.75" 400*

IN.

SSO.'

848,75"-

332*

INST N T TAP ocd-750"-

R E ACTOR l

722.25* VESSEL w M iAkGE FLANCE 700"-

\\

[

650"- 640.O* 3yNg LINE g =-

YARWAY GEMAC

)

BARTON YARWAY J

ROSEWCUNT

• 60" " - - -

- HPCt 8 :- +60"- -

+ 60* 576.7U +60"

-575E (8)

('

8h 42.5 %HI W

~

S52.56"

+20tf m" -544.2S(4) 27.5"(4)- LD ALARM iS.

_S29.2S (3)

REACTCR 544.25" (4) n e..

.01______ j!ca^". ::eS*(3) 12 s'6 7S'

. o :. _ Si6.75.e SOT-II*8)

Lis.101 LI.94 g,4

(-FEED WATMd479.p(2)

. 37"( 2 )

(9 5)

}

INrTIATE RCIC, HPCl,

,$Q,0 4 SCI *-

TRP REClRC. PWIPS 400"-

-371.25 (1)

-145.S*(l)

.lSI 7-INITIATE RHR, C. S.,

.. 150" 366.75" O'

350".- - 352.56 START OIE'SEL 8 TAF

• 352.56 --

CONTRIBUTE Td A.D.S.

d Q (931'.6")

LI-8 5 313 5"-.39" tj I9

• 0)

IN STRUMENT 2/3 /

3Og.

3 RACK CORE HEIGHT

( 25 5 8 25 6 )

PERMESSIW W

252.56".. 800 2S0"- h

'I*

l WATER LEVEL PCMENCLATURE (9 3) 201-208.56" LEVEL HEIGE ABCVE INSTRUME?U HT A8OV

(_.$EN2 ir81 NO.

VESSEL ZERO READifG TAF S

NCZZLE 16 1. S RC

[ jn,)

(;g)

ISO -

NOZZLE r (8) 575.25 M.S 222.69 (7) 559.25 42.5 206.69 (4) 544.25 27.5 191.69 10 0*-

(3) 529.25 12.5 176 69 (2) 479.75 37 127.19 (1) 371.25 445 5 18.69 S&-

d.

E LE VATION 917*.O*

VESSEL aOTTOu FIGURE 2.1.1 REACTOR WATER LEVEL INDICATION CORRELATION Amendment No.,J f, 100 -

?

'P.G CONDITIONS FOR CPERATION SURVEILLANCE RE0UTREMENTS 3.1 BASES (Cont.d) 4.1 BASES (Cont.d) ence paragraph VII.5.7 FSAR).

Thus

ero 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 cever scram and rod block condition. Thus, only the power range. The IRM's if the calibration were perfor=ed dur-and APRM's provide adequate coverage ing operation, flux shaping weuld not 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 Siasing Network, is not signifi-operable in the Refuel mode assures cant and therefore, to avoid spurious that shifting to the Refuel mode scrams, a calibration frequency of during reactor power operation does each refueling outage is established, not diminish the protection provided i

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 protect the reactor. The actual ins t ru=e nt.

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) to 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 i

stage pressure is reduced due to units on a continuous basis, flow through the bypass valves while reactor power is unchanged.

2.

Vacuum tube or semi-conductor devices and detectors that Turbine control valves fast closure drift or lose sensitivity.

initiates a scram based on pressure j

switches sensing Electro-Hydraulic Experience with passive type instru-Control (EHC) system oil pressure, ments in generating stations and sub-The switches are located on the stations indicates that the specified Centrol 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 drif t to be less that 0.4"/ month; i.e.,

in This scram signal is also bypassed the period of a month a maximum 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 APRM's read 2.5 indicated on the scale in the Startup and Refuel modes assures that 4

Amendment No.g100 41

COOPER NUCLEAR STATION g

TABLE 3.2.A (Page 1)

E PRIMARY CONTAINMENT AND REACTOR VESSEL IS01.ATION INSTRl! MENTATION a

z 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 liigh EMP-RM-251, A.B.C,&D

$ 3 Times Full Power 2

A or B Rad.

Reactor Low Water f.evel NBI-LIS-101 A.B.C,&D #1 3+12.5" Indicated I.evel 2(4)

A or B l

Reactor Low Low Water NBI-l.IS-57 A & B #2 3-37" Indicated Level 2

A or B Level NBI-LIS-58 A & B #2 Reactor I.ow I.ow 1.ow Water NBI-LIS-57 A & B #1 3-145.5" Indicated Level 2

A or B I.evel NBI-I.IS-58 A & B #1 Main Steam I.iue 1.eak MS-TS-121, A.B.C,&D

$ 200*F 2(6)

B

,g Detection 122, 123, 124, 143, 144, 145, 146, 147, 148, 149, 150 Main Steam Line High MS-dPIS-116 A.B.C,6D

$ 140% of Rated Steam 2(3)

B Flow 117, 118, 119 Flow Main Steam I.ine 1.ow MS-PS-134, A.B.C,6D 3 825 psig 2(5)

H Pressure e

liigh Drywell Pressure PC-PS-12 A,B,C,&D

$ 2 psig 2(4)

A or B liigh Reactor Pressure RR-PS-128 A & B 3 75 psig I

p Main Condenser 1.ow MS-PS-103 A,B.C,&D 3 7" lig (7) 2 A or B Vacuum Reactor Water Cleanup RNCU-dPIS-170 A & B 3 200% of System Flou I

C System liigh Flow I

COOPER NUCLFi.: STATION TABI.E 4.2.B :Page 2) jf H11R SYSTI'H TEST & CAI.1 LHATION FREQlfENCIES g

ce a

~

Inst reI$cnt Functional d

Item item 1.D. No.

Test Freq.

Calibration Freq.

Check Instrumentation o

I.

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 i

NBI-PIS-52 B & D 5.

Drywell Press.-Containment PC-PS-Il9, A.B.C 6.D once/ Month (1) Once/3 Months Mone Spray 6.

RIIR Pump Discharge Press.

RHR-PS-120, A B.C & D Once/ Month (1) Once/3 Months None 7.

RilR Pump Discharge Press.

RHR-PS-105 A,B,C & D Once/ Month (I) Once/3 Honths None 8.

RilR Pump Low Flow Switch RHR-dPIS-125 A & B Once/ Month (1) Once 3 Honths None 9.

RHR Pump Start Time Delay RHR-TDR-K70 A & B Once/ Month (1) Once/Oper. Cycle None 10.

RHR Injection Valve Close T.D.

RilR-TDR-K45 IA & IB Once/ Month (I) Once/Oper. Cycle None 8

RHR Pump Start Time Delay RHR-TDR-K75, A & B Once/ Month (!) Once/Oper. Cycle None 12.

RilR lleat Exchanger Bypass T.D.

RIIR-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.

Lov Voltage Relays 27 X 3/lA (7)

None 15.

Low Voltage Relays 27 X 3/IB (7)

None 16.

I.ow Voltage Relays 27 x 2/lF, 27 X 2/lc (7)

None

~

17.

Low Voltage Relays 27 X 1/lF, 27 X 1/lG (7)

None 18.

Pump Disch. Line Press. I.ow CH-PS-266, CM-PS-270.

Once/3 Months once/3 Honths None 19.

Emergency buses tindervoltage,

2 7/ l F-2, 2 7/ l FA-2, 2 7 / IC-2, once/Honth once/18 Honths once/12 hrs.

Helays (Degraded Voltage) 27/ICB-2 20.

Emergency Buses 1.oss of 27/lF-1, 27/lFA-1, 27/IC-1, once/ Month Once/18 Honths Once/12 hrs.

Voltage Relays 27/lGB-1, 27/ET-1, 27/FT-2 21.

Emergency Buses IIndervoltage 27X7/lF, 27K7/lc Once/Honth Unce/la Months None Relays Timers l

w COOPER NUCLEAR STATION g

e TABLE 4.2.B (Page 4) 5.

IIPCI TEST & CALIBRATION FREQUENCIES 2

5 Functional Instrument Item Item I.D. No.

Test Freq.

Calibration Freq.

Check 1.

Reactor Low Water Level NBI-LIS-72. A.B.C. & D, #3 Once/Honth (1) Once/3 Honths once/ Day 2.

Reactor liigh Water I.evel NBI-LIS-101. (B & D #2)

Once/Honth (1) Once/3 Honths once/ Day l

3.

liigh Drywell Pressure 14A - K5 A & B (7)

(7) None 14A - K6 A & B (7)

(7) None g

4.

HPCI Turbine liigh Exhaust itPCI-PS-97 A & B Once/Honth (1) Once/3 Honths None o

Press.

5.

IIPCI Pump Low Suction Press.

HPCI-PS-84-1 Once/Honth (1) Once/3 Honths None 6.

HPCI Pump Low Discharge Flow IIPCI-FS-78 Once/Honth (!) Once/3 Honths None 7.

IIPCI Low Steam Supply Press.

HPCI-PS-68, A,B C. & D Once/Honth (1) Once/3 Honths None 8.

HPCI Steam Line liigh AP HPCI-dPIS-76 Once/Honth (1) Once/3 Honths None HPCI-dPIS-77 Once/Honth (1) Once/3 Honths None 9.

IIPCI Steam Line Space liigh IIPCI-TS-101 A,B,C, & D Once/Honth (1) Once/Oper, Cycle None Temp.

102, 103, 104, HPCI-TS-125, 126, 127, 128 RHR-TS-150,151,I52,153,154 O

155,156,157,158,159,160,161 8

10.

Emergency Cond. Stg. Tk, Low HPCI-LS-74 A & B Once/Honth (1) Once/3 Honths None Level I!PCI-LS-75 A & B Once/Honth (1) Once/3 Honths None II.

Suppression Chamber High HPCI-LS-91 A & B Once/Honth (1) Once/3 Honths None Water Level 12.

HPCI Cland Seal Cond,llotwell HPCI-LS-356 B Once/Honth (1) Once/3 Honths None 4

Level llPCI-LS-356 A Once/Honth (1) Once/3 Hanths None 13.

HPCI Control Oil Pressure Low HPCI-PS-2787-H Once/Honth (1) Once/3 Honths None llPCI-PS-2787-L Once/Honth (1) Once/3 Honths None 14 Turbine Condition Supr. Alarm IIPCI-TDR-Kl4 Once/Honth (1) Once/Oper. Cycle None Actuation Timer 15.

Pump Disch. Line Low Press.

CH-PS-268 Once/3 Honths Once/3 Honths None 16.

HPCI Turbine Stop Valve Hon.

IIPCI-LMS-4 Once/Honth N.A.

None 17.

Sup. Chamber HPCI Suction Viv.

HPCI-LMS-2 Once/Honth fl. A.

None Once/0 er. Cycle None 18.

IIPCI Steam Line liigh AP HPCI-TDR-K33, once/Honth P

Actuation Timer

!!PCI-TDR-K43 Once/Honth once/Oper. Cycle None Logic (4)(6) 1.

Logic Bus Power Monitor Once/6 Honths N.A.

2.

IIPCI Initiation Once/6 Honths U.A.

3.

IIPCI Turbine Trip once/6 Honths H.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 red withdrawal from locations of high poet.r density during high power level operation.

Two channels are pro-vided, and one of thest may be bypassed from ths. console for maintenance and/or testing.

Tripping of one of the channale vill block erroneous rod withdrawal soon enough to prevent fuel damage.

Tnis system backs i

up the operator who withdraws control rods according to written se-l quences.

The spatified restrictions with one channel out of service conservatively assure that fuel damage vill not occur due to rod with drawal errors when this condition exists.

A limiting control roi pattern is a pattern which results in the core being on a thermal hydraulic limit (i.e.. MCPR equals the operating limit as defined on Figure 3.11 and I.HGR = as defined in 1.0.A.4).

During use of such petterns it is judged that testing of the RBM system prior to withdradal of such rods to assure its operability vill assure that inproper withdrawal does not occur.

It is the responsibility of the Reactor Engineer to identify these limiting patterns and the desig-nated code 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 persor.nel qualified to perform this function may be designated by the Division Manager of Nuclear Operations.

C.

Seram Insertion Times Aha 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 great'er than the safety lir.it.

The surveillance requirement for scram testing of all the control rods after each refueling outage and 10% of the control rods at 16-veek 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 BVR's with cc.ntrol rod drives the same as those on Cocper 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 higti z.eutron 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 vben compared to the typical time delay of about 210 milliseconds estimated from scram test results.

Approximately the first 90 nil 11 seconds 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.

Amendment No [. M M 100

-102--

A LIMITING CCNDITIONS FOR OP2 RATION SURVEZLLANCE REQUIREMENTS 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 pu=p flow in-following conditions do not occur sic-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 shall be initiated a.

The recirculation pu=p flow differs and the reactor shall be in a Cold by more than 15 frem 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 flow characteristics.

b.

The indicated value of core flow rate varies from the value derived from loot flow measurements by more than 10%.

c.

The diffuser to lover plenu= differen-tial pressure reading on an individual jet pu=p varies from the cean of all jet pu=p differential pressures by more than 10%.

]

F.

Recirculation Pump Flev Miscatch F.

Recirculation Pu=p Flow Mismatch l

1.

Deleted.

1.

Deleted.

I 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 C.

Inservice Inspection To be considered operable, com-Inservice inspection shall be per-ponents shall satisfy the require-formed in accordance with the ments contained in Section XI of requirements for ASMI Code Class 1 the ASME Boiler and Pressure Vessel 2, and 3 couponents contained in Code and applicable Addenda for Section XI of the ASME Boiler and continued service of ASMI Code Pressure Vessel Code and applicable Class 1, 2, and 3 components except Addenda as required by 10 CFR 50, where relief has been granted by the Section 50.55a(g), except where Co= mission pursuant to 10 CFR 50, relief has been granted by the Section 50.55a(g)(6)(i).

Co= mission pursuant to 10 CFR 50, Section 50.55a(g)(6)(1).

Amendment flo.)(,)(,100

_t37

~ 3 i

.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.

T.

Recirculation Pump Flow Mismatch l

l l

1 1

Requiring the discharge valve of the lower speed loop to re=ain closed until the speed of faster 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

{

~*bration of the jet pump risers will not occur.

G.

Inservice Inspection The inservice inspection program conforms to the requirements of 10 CFR 50, Section 50.55a(s). 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 ASHI 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 vill be reported to the Commission prior to the beginning of each 10-year inspection period if they are known to be required at that time. Deviations which are identified during the course of inspection vill be reported quarterly throughout the inspection period.

Amendment No.g)N.100

-151-

TABLE 3.7.2 TESTABLE PENETRATIONS WITH DOUBLE 0-RING SEALS PEN. NO DESCRIPTION X-1A Dryvell equipment batch X-1B Dryvell equipment hatch X-2 Drywell airlock door l

X-4 Drywell head access hatch X-6 CRD removal hatch X-35A TIP "D" Penetration X-35B TIP "A" Penetration X-35C TIP "C" Penetration X-35D TIP "B" Penetration i

X-35E TIP N Purge Connection 3

X-200A Suppression chamber access hatch X-2003 Suppression chamber access hatch X-213B Suppression chamber drain flange Dryvell head Stabilizer Assembly Inspection Ports (8)

Amendment No.

,100

-171-

TABLE 3.7.4 PRII'ARY CONTAINMENT TESTABLE ISOLATION VALVES TEST

?dS. M..

VALVE NL'MBERS MEDIA l

a-la MS-AO-80A and MS-AO-86A, Main Steam Isolation Valves Air X-7B MS-AO-808 and MS-AO-865, 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-MO-77, Main Steam Line Drain Air X-9A RF-15CV and RF-16CV, Feedwater Check Valves Air X-9A RCIC-AO-22. RCIC-MO-17, and RWCU-15CV, RCIC/RWCU Connection to Feedwater Air X-9B RF-13CV and RF-14CV, Feedwater Check Valves Air A-9b HPCI-AO-18 and HPCI-MO-57. HPCI Connection to Feedwater Air X-10 RCIC-MO-15 and RCIC-MO-16 RCIC Steam Line Air X-11 HPCI-MO-15 and HPCI-MO-16. HPCI Steam Line Air X-12 RHR-MO-17 and RHR-MO-18, RHR Suction Cooling Air X-13A RHR-MO-25A and RHR-MO-27A, RHR Supply to RPV Air X-13B RHR-MO-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-12A, Core Spray to RPV Air X-163 CS-MO-11B and CS-MO-128, Core Spray to RPV Air X-17 RRR-MO-32 and RRR-MO-33 RPV Head Spray Air X-18 RW-732AV and RW-733AV, Dryvell Equipment Sump Discharge Air X-19 RW-765AV and RW-766AV, Dryvell 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 Dryvell Air X-26 ACAD-1310MV, Bleed from Drywell Air Amendment No. A6. g g, 100

_173-

TABLE 3.7.4 (pego..

PRIMARY CONTAINMENT TESTABLE ISOLATION VALVES TEST PEN. NO.

VALVE STMBERS MEDIA X-39A RHR-MO-26A and RHR-MO-31A, Dryvell Spray Header Supply Air X-395 RHR-MO-268 and RHR-MO-31B, Drywell Spray Header Supply Air X-39B ACAD-1311MV and ACAD-1312MV, Supply to Drywell Air X-41 RRV-740AV and RRV-741AV, Reactor Water Sample Line Air X-42 SLC-12CV and SLC-13CV, Standby 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-2103 RHR-MO-21B, RHR to Torus Air X-210B HPCI-17CV and HPCI-MO-25. HPCI Minimum Flow Line Air X-210B RHR-MO-16B, RHR-11CV, and RHR-13CV, RHR Minimum Flow Line Air X-210A and 211A RNR-MO-34A, RHR-MO-38A, and RHR-MO-39A, RHR to Torus Air X-2108 and 211B RER-MO-34B, RHR-M0-38B, and RHR-MO-39B, RHR to Torus Air X-211B ACAD-1301MV and ACAD-1302MV, Supply to Torus Air i

j X-212 RCIC-15CV and RCIC-37 RCIC Turbine Exhaust Air X-214 HPCI-15CV and HPCI-44. HPCI Turbine Exhaust Air i

X-214 HPCI-AO-70 and HPCI-AO-71, HPCI Turbine Exhaust Drain Air i

X-214 RHR-MO-166A and 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 [

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 4

Amendment No,j#, A % 100

-174-

1 REVIEk' AND AUDIT 6.2.1 The organization and duties of committees for the review and audit of station operation shall be as outlined below:

A.

Station Operations Review Committee (SORC) 1.

Membership:

a.

Chairman: Division Manager of Nuclear Operations b.

Technical Staff Manager c.

Operations Manager d.

Technical Manager e.

Operations Supervisor f.

Maintenance Supervisor g.

Instrument and Control Supervisor h.

Chemistry and Health Physics Supervisor 1.

Plant Engineering Supervisor j.

Operations Engineering Supervisor l

k.

Computer Applications Supervisor 1.

Maintenance Manager

[

m.

Quality Assurance Manager - non-voting member.

Alternate members shall be appointed in writing by the Division Manager of Nucleat Operations to serve on a temporary 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 Chair =an.

3.

Quorum: Division Manager of Nuclear Operations or his designated alternate plus four other members including alternates.

4 Responsibilities:

Review all proposed normal, abnormal, maintenance and emergency a.

operating procedures specified in 6.3.1, 6.3.2, 6.3.3, and 6.3.4 and proposed changes thereto: and any other proposed procedures or changes thereto determined by any member to ef fect nuclear

safety, b.

Review all proposed tests and experiments and their results, which involve nuclear hazards not previously reviewed for conformance with technical specifications.

Submit tests which may constitute an unreviewed safety question to the NPPD Safety Review and Audit Board for review.

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 safety significance to the NPPD Safety Review and Audit Board for review, Review station operation to detect potential nuclear safety e.

hazards.

Amendment No. JW,' gg,100

-220-

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

,I 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 10CTR Part 50.

1.

Review drills on emergency procedures (including plant evacuation) and adequacy of communication with off site groups.

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 The Station Operations Review Committee shall be advisory.

a.

b.

The Station Operations Review Committee shall recommend to the Division Manager of Nuclear Operations approval or disapproval of proposals under items 4, a through e and j above. In case of disagreement between the recommendations of the Station 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 vill be followed. A written su==ary of the disagreement will be sent to the Vice Ptesident - Nuclear and to the l NPPD Safety Review a6d Audit Board, The Station Operations Review Committee shall report to the c.

Chairman of the NPPD Safety Review and Audit Board on all re-views and investigations conducted under items 4.f. 4.g. 4.h, and 4.1.

d.

The Station Operations Review Committee shall make determinations 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.

6.

Records:

Minutes shall be kept for all meetings of the Station Operations Review Committee and shall include identification of all documen-Amendment No. J0', Jf',)W,M 100

-221-1

6.2 (Cont'd) 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.

l 7.

Procedures:

l Written administrative procedures for Committee operation shall be prepared and maintained describing the method for submission j

and content of presentations to the committee, provisions for use of subcommittees, review and approval by members of written Committee evaluations and recommendations, lissemination of minutes, and such other matters as may be appropriate.

B.

NPPD Safety Review and Audit Board (SRAB) l Function: Ths Board shall function to provide independent review and audit of designated activities.

1

\\

i 4

\\

?

I L

Amendment No.M JM JV,' 100

-222 -

4 I

6.2 (cont'd) c.

Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

d.

Proposed changes to Appendix A Technical Specifications or 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 safety.

l g.

All reportable events specified in Section 50.73 to 10CFR Part 50.

h.

Any indication of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.

i. Minutes of meetings of the Station Operations Review Co=mittee.

j. Disagreement between the reco=mendations 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 en the results of investiga-tions and recommendations to prevent or reduce the probability of recurrence.

5.

Authorit': The NPPD Safety Review and Audit Board shall report to and be advisory to the Vice President - Nuclear on those areas of responsi-

{

bility specified in Specifications 6.2.1.B.4 and 6.2.1.B.7.

Amendment No. g M,M, 100 624-

L.

(icnt'd) 6.

Recordst Minutes shall be recorded for all meetings of the b' PPD Safety Review and Audit Board and shall identify all documentary material reviewed. Copies of the minutes shall be forwarded to the Vice President - Nuclear and the Division Hanager of Nuclear l

Operations, and such others as the Chairman may designate within one month of the meeting.

7.

Audits:

Audits of selected aspects of plant operation shall be perfor=ed 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 are reviewed by SRAB. A representative portion of procedures and records of the activities performed during the audit period shall be audited and, in addition, observations of performance of operating and maintenance activities shall be included. These audits shall encompass:

Amendment flo. MM 100

-225-

1.

A tabulation on an annual casts of the number of statica, utility and other personnel (including contractors) re-ceiving exposures greater than 100 mrem /yr and their associated man rem exposure according to work and job functions, 1/

e.g.,

reactor operations and surveillance.

inservice inspection, routine maintenance, special main-tenance (describe maintenance), vaste processing, and refueling. The dose assignment to various duty functions may be estimates based on pocket dosimeter TLD, or film badge measurements.

Small exposures totaling less than 20%

of the individual total dose need not be accounted for.

In

~

the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

2.

A su'mmary description of facility changes, tests or experi-ments in accordance with the requirements of 10CTR50.59(b).

3.

Pursuant to 3.8. A. a report of radioactive source leak testing. This report is required only if the tests r. veal the presence of 0.005 microcuries or more of removable contamination.

4.

Documentation of all challenges to relief valves or safety valves.

D.

Monthly Operating Report Routine reports of operating statistics, shutdown experience, and a narrative summary of operating experience relating to safe operation of the facility, shall be submitted on a monthly basis to the individual designated in the current revision of Reg.

Guide 10.1 no later than the tenth of each month following the calendar month covered by the report.

6.5.2 Reportable Events A Reportable Event shall be any of those conditions specified in Section 50.73 to 10CTR 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/

This tabulation supplements the requirements of 520.407 of 10CFR Part 20.

Amendment No.,Sf$ 100

-231-m

N a

1 ar+

• o PRrwNT VIE PRESIDENT

- NUCLEAR

-o TECHNICAL STAFF VIEW D

AUDIT BOARD MANAGER, NPG (S R AB) g NUCLEAR OPERATIONS NUCLEAR SERvCES QUALITY ASSURANCE DIVISION. MANAGER DIVISION. MANAGER DIVISION. MANAGER l

l FIRE PROTECTION OUAILTY ASSURANCE QUAILTY ASSURANCE PROG R AM MAN AGER - CNS MANAGER - G.O.

NUCLEAR FUELS ADMIR SITE NUCLEAR NUCLEA R LICENSING NUCLEAR ENGINEERING NUCLEAR PHCMECF~

SUP POR T S SAFETY DEPARTMENT DE P A R T MENT DE PA R TMENT MANAGER OFFICE DEf%RTMENT-MANAGE R DEPARTMENT-MANAGER MANAGER Figur e 6.1.1 NPPD Nuclear Power Group Organi olion Chart i

W to NUCLEAR OPER A TIONS ro DIVISION

=

MANAGER o

STATION OPERATIONS FIRE PROTECTION REVIEW COMMITTEE

- - - - ~ ~ ~

ENGINEER N (SO R C)

QUALITY ASSURANCE

$7 7 8

MANAGER - C NS MANAGER I

I I

TRAINING TECHNICAL MAINTE NANCE OP ERATIONS OUTAGE &

ADMINISTRATIVE MOOWEATIONS SERVICES MANAGER MANAGER MANAGER MANAGER MANAGER MANAGER OPER A T RO N S SPECIAL MANTENANCE TRAINING PRCLIECTS SUPERVISOR N

SUPER VISO R E NGIN E E R SECURITY AOMINIS TR ATIVE MATERIAL

)

d I

SUPPORT SUPERVISOR SUPE RVISOR SUPERVISOR i

CHEM S HP COMPUTER PLANT OPERATDNS APPLIC ATIONS ENGINEER ENGINEERNG SUPERVISOR SUPE R VISOR SUPERVISOR SUPERVtSOR 18C OPE R ATION S SUPE RVISOR I/S ONE/ SHIFT S UPE R VISOR IS R O) 2/S TWO/ SHIFT l

3/5 THREE/ SHIFT I

RO NRC REACTOR OPERATORS LICENSE SHIFT SRO.NRC SE NIOR REACTOR OPERATORS SUPERVISORS (S RG UCENSE 8/ S

q. FUNCTIONAL POSITION ONLY l

ELECTRICAL MAINT E MNCE MECHANICAL 1

PHYSICALLY LOCATED IN GENERAL OFFICE PLANNER /

SUPERVISOR SCHEDULE R SUPERVISOR 2/Sl UNIT OPER IROl giSTATON OPER I (UNLICENSED)

Figure 6.I.2 NPPD Cooper Nuclear S tation Organization Chart