ML19323C764
| ML19323C764 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/04/1976 |
| From: | Goller K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19323C736 | List: |
| References | |
| FOIA-80-162 NUDOCS 8005190151 | |
| Download: ML19323C764 (30) | |
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NUCLE AR REGUL.ATORY CC'AMISSICN
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CONSUMERS POWER COM2ANY DOCKET NO. 50-155 BIG ROCK POINT PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 10 License No. DPR-6 1.
The Nuclear Regulatory Commission (the Co==ission) has found that:
A.
The applications for amendment by the Consumers Power Company (the licensee) dated July 25, 1975 and supplements thereto dated August 22, 1975, September 8,1975, Noveder 26, 1975, February 4,1976, February 27,1976, March 26,1976, April 30, 19 76, May 10,19 76 and May 11, 1976; October 13,19 75, as modified by letters dated April 28,1976 and May 11 and 25,1976; August 15, 1974, as modified by letters dated Noveder 14, 1974, Dececher 17,1974, March 10,1975, April 29,1975 and October 9, 1975; and Deced er 5, 1975, comply with the standards and requirements of the Atomic Energy Ac cf 1954, as amended (the Act), and, that in view of the uemption granted by the Commission on May 26, 1976, comply with the Co= mission'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 Cocaission; C.
There is reasonable assurance (i) 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 Co::missibn's re8ulations; 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.
After weighing the environmental aspects involved, the issuance of this amendment is in accordance with 10 CFR Part 51 of the Com=ission's regulations and all applicable requirements have been satisfied.
3 2.
On l'ay 26, 19 76, the Cotnission by Memorandu= and Order granted to the Consumers Power Co=pany certain exe=ptions from the requirements of 10 CFR 50.46 for the captioned facility.
3.
Accordingly, Facility Operating License No. DPR-6 is hereby a= ended as follows:
A.
Change the Technical Specifications as indicated in the attach =ent to this license amend =ent.
B.
Revise item 2.C(2) of the license to read:
(2)
Technical Specifications
~
The Technical Specifications contained in Appendix A as issued May 1,1964, as revised, are hereby incorporated in this license.
The licensee shall operate the facility in accordance with the Technical Specifications as revised.
C.
Add the following as item 2.C(3) of the license:
(3)
The licensee is subject to the conditions set forth in Section III.d of the Commission's Memorandum and Order dated May 26, 1976.
4.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C0141SSION Karl R. Goller, Assistant Director for Operating Reactors Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 4,1976 b
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'F ATTACHMENT TO' LICENSE AT;DME';T NO.10 FACILITY OPERATING LICENSE NO. DPR-6 DOCKET NO. 50-155 The Technical Specifications attached to Facility Operating License No. DPR-6 are hereby changed as follows:
1.
P.eplace Contents page iv vith the attached new page.
Delete Section 3 5 2(a) in its entirety.
3.
Delete Section 3.5.2(c) in its entirety.
4.
Replace Section 3.7(e) with the following:
"An integrated leakage rate test shall be conducted on the containment sphere at approximately three equal intervals during each 10-year service period."
Delete the following phrases in chronological order from Section 5.
4.1.2(b):
", core spray and backup core spray systems" and
" core spray system,"
and "and the fire water makeup system to the condenser hotwell" and I
" core spray system and",
- 6. Delete Section h.2.l(a) in its ent.cety.
- 7. Delete Section 4.2.l(b) in its entirety.
- 8. Delete the lant sentence in Section h.2.6.
l lv N,
9 Delete the colu::ns er. titled Reload B & C and Reload E and S
add the following colu=n to Table 5.1.
General Reload G-10 Gac::etry, Fuel Rod Array 11 x 11 t, ~
Rod Pitch, Inches 0.577 UO R ds 2
109 j
Cobalt - Bearing Corner Rode k
Gadolinium - Bearing UO Rods b
2 Inert Spacer Capture Rod (Zr-2) 1 Zircaloy Rods 3
p Spacers per Bundle 3
Fuel Rod Cladding Material Zr-2 Wall Thickness, Inches 0.03h Fuel Rods Outside Rod Dia=gter, Inches 0.449 Fuel Stacked Density, Percent Theoretical 91.6 f
Active Fuel Length. Inches Standard Rod 5
TO Fill Gas Helium 3,95%
10 Delete the references to Reloads B, C sad E in Note 1 of Table 5.1.
IL Delete the reference to Reload E in Note 3 of Table 5.1.
h __. _
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I2 Change Section 5.2.l(b) to read as follows:
5.2.l(b)
Reactor Operation
'Ihe reactor operation shall be so limited as to be consistent with the most conservative of the following:
TABLE 1 Reload E-G and Modified E-G Reload F & J-2 Reload G G-lU Minimum Core Burnout Ratio at Overpower 1.5' 1.5**
- 1. 5" Transient Minimum Burnout Ratio in Event of kas of Recirculation Pumps From Rated Power 1.5 1.5 1.5 2
Maximum Heat Flux at overpower, Btu /h-ft 500,000 395,000 h0T 000 2
Maximum Steady State Heat Flux, Btu /h-ft h10,000 32h,000 333,600 Maximum Average Planar Linear Heat Generation Rate, Steady State, kW/Ft Stability Criterion: Maximum Measured Zero-to-Peak Flux Amplitude, Percent of Average operating Flux 20 20 20 Maximum Steady State Power Level, MW 2k0 240 2h0 g
Maximum Value of Average Core Power Density 8 240 W, kW/L h6 h6 46~
t Ilominal Reactor Pressure During, Steady State Power Operation, psig 1335 1335 1335 Minimum Recirculation Flow Rate Lb/h (Except During Pu=p Trip Tests or Natural 6
6 6
Circulation. Tests as Outlined in Section 8) 6 x 10 6 x 10 6 x 10 Maximum mwd /T of Contained Uranium for an Individual Bundle 23,500 23,500 23,500 Rate-of-Change-Of-Reactor Power During Power Operation:
Control rod withdrawal during power operation shall be auch that the average rate-of-change-of-reactor power is less than 50 MW, per minute when power is less than 120 MW less than 20 MW per minute whed power is between 120 and 200 MW, and 10 b,t Per Mnute when power is Mveen 200 and 240 W
- g t
t
- Based on correlation given in " Design Basis for Critical Heat Flux Condition in Boiling Water Reactors," by J. M. Healzer, J. E. Hench, E. Janssen and
.S. Levy, September 1966 (APED 5286 and APED 5286, Part 2).
- Based on Exxon Nuclear Corporation Synthesized Hench Levy.
"'To be determined by linear extrapolation from Tabla 2 attached.
l
i 3 e
- 13. Following the table in Section 5.2.l(b), add the attached Table 2 -
MAPLHGR (kw/ft).
- 14. Delete Figures 5.2 and 5.3 and renumber existing Figures 5.' through 5.8 as 5.2 through 5.6, respectively. Add new Figure 5.7 attached.
- 15. Delete Section 6.1.h(a) in its entirety.
16.
Delete Section 6.1.4(b) in its entirety.
17.
Delete Section 6.1.4(c) in its entirety.
18.
Change Section 6.1.5(b) to read as follows:
"The emergency condenser system control initiation sensors shall be functionally tested not less frequently than once every 12 months."
19.
Delete Section 6.1.6 in its entirety.
20.
Delete the entire entry of " Post-Incident Spray System -
Automatic Control Operation" in the table contained in Section 7.6.
21.
Change the words " Reactor emergency cooling systems trip circuits",,
in the table in Section 7.6 to " Emergency Condenser Trip Circuits.
22.
Replace Table 8.2 with the attached new Table 8.2.
- 23. Incorporate additional pages 11-1 through 11-20.
These pages are the new format of the Specifications to be issued in the future and are identified as Section 11.0 but contain the numbering sequence of the new format, i.e., 3.1.5/4.1.5 etc.
Amendment No. 10 is identified for each new page in the lower Corner.
/
i I
CONTENTS (Contd)
Page No.
6.5.1 Plant Review Committee (PRC) 10-5 6.5.2 Safety and Audit Review Board (SARB)................
10-7 6.6 Reportable Occurrence Action.
10-11 6.7 Safety Limit Violation.
10-11 6.8 Procedures...
10-11 6.9 Reporting Requirements.
10-12 6.10 Record Retention.
10-20 6.11 Radiation Protection Program.
10-22 6.12 Respiratory Protection Program.
10-22 11.0 (Section 3.1.4/4.1.4) Emergency Core Cooling System 11-1 (Section 3.1.5/4.1.5) Reactor Depressurization System 11-8 (Section 3.3.4/4.3.4) Containment Spray System.
11-15 (Section 3. 5. 3/4. 5. 3) Emergency Power Sources 11-18 w.
i i
i iv Amendment No. 10
o
.e
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This table should follow Table 1 cf S-etien 5 2.1(b))
TABLE 2 MAPLHGR (kW/Ft)
Planar Average Exposure (Wd/STU)
_ Modified F E-G, F, J-2
& NFS-DA G-1U Reload G Reload 0
6.38 6.40 200 9.5 9.4 907 6.86 1,814 6.87 2,041 6.79 4,536 6.76 6.90 5,000 9.9 9.7 9,072 7.05 9,979 6.86 10,000 9.9 9,7 13,608 6.97 14,515 7.25 15,000 9,8 9.6 18,144 7.25 19,051 6.95 20,000 8.7 8.6 25,000 8.7 8.3 25,401 7.05 27,216 7.28 Amendment No. 10 N
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TABLE 8.2 EEI UO -
Center elt 2
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PuO2 mediate Advanced
- FS-DA Minimum Core Burnout Ratio at Overpower 1.5" 1 5*
1.5*
15 Transic..; Mininun Burnout Ratio in Event of Loss of Recirculation From Rated Power 1.5 1.5 15 1.5
. !hxiatra Heat Flux at Overpower, 402,000 Btu /h-Ft2 500,000 Maxinu.;}teadyStateHeatFlux, Etu/h-F.
h10,000 500,000 500,000 329,000 Maxitte Average Planar Linear Heat (See Note 1)
Generation Rate, Cteady State, kW/Ft Stability Criterion:
M1ximum Measured 7.ero-to-Feak Flux A.plitude, Percent of Averace Operating Flux 20 20 Maximu-Steady State Power Level,1&
2h0 2h0 t
1;cninal Reactor Prcosure Durin;;
Steady State Power Operation, Psig 1,335 1,335 Minin.u= Recirculation Flow Hate, Lb/h (Exce;.t During Punp Trip Tests or Uatural 6
6 Circulation Tests as Cutlined in Sec 8) 6 x 10 6 x 10 Maximum !Nd/T of Contained Uranium for 23,500 nn Individual Bundle 23,500 Nunber of Bundles:
Pellet UO 1
3 2
Povder UO 1
2 2
Rate-of-Change-of-Reactor Power During Power Operation:
Control rod withdrawal during power operation shall be such that the averaSe rate-of-change-of-reacter power is less than 50 IGt per minute when power is less than 120 IN less than 20 int per minute when power is between 120 and 200 IW, and 10 h.,i per minute when power is between 200 and 2hD ;&g.
g g
- Based upon critical heat flux correlation, APED 5286.
- Ro longer used in reactor.
Note 1:
MAPLHGR (kW/Ft)
Planar Avg Exposure (mwd /STU)
NFS-DA 0
6.38 2041 6.79 4536 6.76 9979 6.86 13608 6.97 l
19051 6.95 Amendment No. 10 25401
- 7 05
4
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Limiting Condi tions for Operation Surveillance Requirement i
ll.3.1.h EMERGENCY CORE COOLING SYSTD4 11.h.l.h EMERGENCY CORE COOLING SYSTEM Applicability:
Applicability:
{
Applies to the operating status of the emer-Applies to periodic testing requirements for gency core cooling system.
the emergency core cooling systems.
l Objective:
Objective:
f.
To assure the capability of the emergency To verify operability of the emergency core core cooling system to cool reactor fuel in cooling systems.
l the event of a Loss of Coolant Accident.
Specification:
l Specification:
A.
Each month the following shall be performed:
A.
The two core spray systems (original and j
redundant) shall be operable whenever Verify the operability of MO-7051,
-7061, -7066, -7073 and -707h by remote
.the plant is in a power operation con-i g,g g dition.
The original core spray system shall also be operable during refueling, Leak testing of the core spray heat j
operations.
exchanger.
B.
The core spray recirculation system shall be operable whenever the plant Verify that the check valve between MO-7051 and -7061 is not stuck shut.
is in a power operation condition.
C.
The core spray recirculation heat ex-B.
At each shutdown the following shall be changer shall not be taken out of ser-perfonned.
vice during power operation for periods exceeding four (4) hours. The heat Verify the operability of MO-7070 and exchanger shall be considered inoperable
-7071 by remote manual actuation.
and out of service if tube bundle leak-age exceeds 0.2 gpm.
11-1 Amendment No. 10
Limiting Conditions for Operation Surveillance Requirement ll.3.1.h EMERGENCY CORE COOLING SYSTEM (Contd)
~
ll.h.l.h ' EMERGENCY CORE COOLING SYSTD4 (Contd )
D.
Both fire pumps (electric and diesel)
C.
At least once every six (6) months, and the piping system to the core spray except for periods of continuous shutdown. when the following shall system tie-ins shall be operable when-ever the plant is in a power operation be performed prior to startup:
condition and refueling.
Automatic actuation of the core spray system valves with water flow manually blocked (M0-7051, -7061, -7070 and E.
If Specifications A, B, C, and D are not
-7071).
met, a normal orderly shutdown shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the reactor shall be shut down as described in Section Operability check of the Core Spray Recirculation System.
1.2.5(a) within twelve (12) hours and shut down as described in Section 1.2.5(a) and (b) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No work D.
At each major refueling outage, the fol-shall be performed on the reactor or its loving shall be performed:
connected systems when irradiated fuel is in the reactor vessel which could Calibration of core spray system actua-result in lowering the reactor water level tion and pressure and flow instrumentation.
below elevation 610'5".
Verify that the two core spray system containment isolation check valves are not F.
Until such time as the effectiveness of-stuck shut.
redundant core spray nozzle has been Operability of the check valves between proven, the fire water makeup system t MO-7051 and MO-7061 and MO-7070 and MO-7071.
the condenser het well shall be operable and ready for service during power oper-Calibration of fire system basket strainer ation.
If the fire water makeup system differential pressure switches.
becomes inoperable and not corrected, a normal orderly shutdown shall be initi-Operability check of the core spray j
ated within one (1) hour nnd the reactor recirculation system.
shall be shut down as described in Sec-tion 1.2 5(a) within twelve (12) hours E.
Instruments shall be checked, tested and and shut down as described in Section calibrated at least as frequently as 1.2.5(a) and (b) within the following listed in Table ll.h l.h(a).
2h hours.
C.
Instrument set points shall be au speci-fled in Table 11.3 1.h(a).
11-2 Amendment No. 10
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TABLES 11 3.1.ha AND ll.h.l.ka Instrumentation That Initiates Cores Spray ll.h.1.ha Surveillance Requirement
. 11.3.1.ha Limiting Conditions for Operation Instrument Trip Trip System Limiting Conditions for Test Including Instrument Parameter Logie Set Point Operability Valve Actuation Calibration Open Core Spray Valves n
Low Reactor Water One of Two for 2610'5" Elev Power Operation Once Every Six Months Each Major Level (b)
Each of Two (22'8" Above and Refueling of Operation Refueling Valves in Series Core)
Operations (a)
Other Than Cold Shut-down f
Steam Drum Pressure One of Two for
>200 Psig Power Operation Once Every Six Months Each Major I
I4v (b)
Each of Two and Refueling of Operation Refueling Valves in Series Operations (a)
Other Than Cold Shut-down J
Notes for Tables 11.3.1.ha and 11.h.l.ha (a) Initiation of valve operation requires both low reactor water level coincident with lov steam drum pressure.
(b) The primary core spray system shall be'available for use during refueling operations and the backup system shall be closed and operation of the backup core spray valves shall be blocked or otherwise defeated while the piping section from the valves to the reactor head is dismantled.
~
11-3 Amendment L 10 u &,.am_L-e
_,,, _.un.
a---ww-~2----
w Bases:
The core spray system consists of two automatically actuated independent double capacity piping headers capable of cooling reactor fuel for a range of Loss of Coolant Accidents.
Either system by itself is capable of pro-viding adequate cooling for postulated large breaks in all locations. When adequate depressurization rates are achieved in the postulated small-break situation, either core spray system provides adequate cooling.
For the largest possible pipe break, a flow rate of approximately 400 gpm is required after about 20 seconds.
Each core spray system has 100% cooling capacity from each spray header and each pump set.
Thus, specifying both systems to be fully operational vill assure to a high degree core cooling if the core spray system is required.
In addition, the original core spray is required to be operable during refueling operations to provide fuel cooling in the unlikely event of an inadvertent draining of the reactor vessel.
The core spray systems receive their water supply from the plant fire system. The plant fire system nupply in from Lake Michigan via two redundant 1,000 gpm fire pumps, one electric and one diesel driven. These pumpu nLart automatically on decaying fire system pressure.
The core sp.my recirculation system is provided to prevent excessive water buildup in the containment sphere and to provide for long-tenn, post-accident cooling. The system consists of two pumps (h00 gpm each) and a heat ex-changer. The pumps take a suction from the lower levels of containment and discharge to the core spray headers.
The system is actuated manually when the water level in the containment rises to elevation 587 feet. The 587-foot elevation vill be achieved between 6 to 2In hourn opernt. ion of one core sprny and one contninment. spray nyntem.
11-4 Amendment No. 10
Bases:
(Contd)
A test tank and appropriate valving is provided in the core spray recirculation system so the pump suction c:nditions and the flow characteristics of the system can be periodically tested.
One core spray recirculation pump has adequate capacity to provide fuel cooling at anytime following a Loss of Coolant Accident. Continuous containment spray operation is not required during the post-accident recirculation phase if only one recirculation pump is available.
The operable status of the various systems and components is to be demonstrated by periodic tests. Some of these t sts vill be performed while the reactor is operating in the power range.
If a component is found to be inoper-tble, it will be possible in most cases to effect repairs and restore the system to full operability within a er ralatively short time. For a single component to be inoperable does not negate the ability of the system to per-farm its function, but it reduces the redundancy provided in the reactor design and thereby limits the ability to tolerate additional equipment failures.
If it develops that (a) the inoperable component is not repaired within the specified allowable time period; or (b) a second component in the same or related system is found to be inoperable, the reactor will initially be removed from service which will provide for a reduction of the decay hrat from the fuel and consequential reduction of cooling requirements after a postulated Loss of Coolant Accident.
If the malfunction cannot be rapidly corrected, the reactor will be cooled to the shutdown condition using normal cooldown procedures.
In this condition, release of fission products or damage of the fuel elements is not c:nsidered possible.
The plant operating procedures will require immediate action to effect repairs of an inoperable component and, therefore, in most cases, repairs will be completed in less than the specified allowable repair times. The limiting times to repair are intended to:
(1) Assure that operability of the component will be rentored promptly and yet. (2) allow sufficient time to effect repairs using safe and proper procedures.
The leakage rate limit for the core spray recirculation system heat exchanger has been established y,
to assure detection of any degradation of the integrity of the heat exchanger.
11-5 Amendment No. 10
Bases:
(Contd)
As a result of an evaluation of the effect testing frequency on emergency core coo. ling system reliabilityb and because f a lack of test data to prove the effectiveness of the redundant (nozzle) core spray system spray 10 distribution the surveillance requirements for the original core spray system have been increased.
In addition, time periods allowed for operation with the original (ring) core spray system out of service have been reduced significantly.
Further changes in surveillance and operability requirements vill be requested prior the refueling outage presently scheduled for Spring 1977 based on modifications to make the core spray cystems more testable and following proof of nozzle spray effectiveness.
The fire water makeup system to the condenser hot well was provided as a temporary means of reducing peak fuel clad temperature under postulated small and intermediate sized pipe breaks until the Reactor Depressurization System could be completed.
It is still required until nozzle spray distribution patterns are demonstrated.
References:
1.
Consumers Power Company letter to Directorate of Licensing, USAEC, dated May 18, 1972.
2.
Technical Specifications Change No 26 dated July 27, 1971.
3.
FHSR, Section 12.
k.
FHSR, Section 3 5.
FHSR, Section 5 6.
Consumers Power Company letter to Directorate of Licensing, USAEC, dated September 22, 1972.
7 FHSR, Section 13 t:s S.
" Big Rock Point Plant Hydrological Survey," Great Lakes Research Division, Special Report. No 9, Ayer, J. C., et al, Nov 1961.
9.
Consumers Power Company letter to the Secretary of the Commissien, USNRC dnt.ed March 26, 1976.
10.
Comments by the Director, Nuclear Reactor Regulation Relating to the Request for Exemption of the Big Rock Point Nuclear Power Plant From the Requirements of 10 CFR 50.h6 dated April 19, 1976.
I 11-6 Amendment Ho, 10
(NOTE:
This is the new format of the Specifications to be issued in the future.
".h e re fo re, the numbering system may conflict with existing sections.
Both are still applicable.)
Limiting Conditions for uperation Survcillance Requirement 3.1.5 REACIT)R DEPRESSURIZATION SYSTEM 4.1.5 REACTOR DEPRESSURIZATION SYSTEM Applicability:
Applicability:
Applies to the operating status of the Reactor Applies to periodic testing requirements for Depressurization System (RDS).
the RDS.
Objective:
Objective:
n~
To assure the operability of the RDS and when To verify operability of the RDS.
working in conjunction with the emergency core cooling system to allow cooling of the reactor Specification:
fuel in the event of a Loss of Coolant Accident.
A.
The isolation valves shall be test-operated Specification:
at least once every three months per Sec-tion IWV-3410 Summer 1973 Addenda of the A.
The RDS shall be operable at all power ASME BGPV Code Section XI.
levels and when the reactor is critical with the head on or when in hot shutdown B.
The depressurizing valves shall be test-conditions.
operated during each cold shutdown; how, ever, in the case of frequent cold shut-downs, these valves need not be exercised B.
The limiting conditions for operation of
~
m re often than once every three months per
==
the instrumentation and actuating circuitry Section IWV-3410 Summer 1973 Addenda of the which initiates and controls the RDS are ASME BGPV Code Section XI.
given in Table 3.5.2.h.
C.
The instrumentation shall be functionally tested, calibrated and checked as indi-cated in Tabic 4.5.2.h.
11-7 Amendment No. 10
Limitin2 Conditions for Operation Surveillance Requi rement i
3.1.5 REACTOR DEPRESSURIZATION SYSTEM lContd) 4.1.5 REACTOR DEPRESSURIZATION SYSTDI (Contd)
Action:
D.
System Logic shall also be functionally tes ted as indicated in Table 4.5.2.h.
1.
Should one depressurizing valve or isola-tion valve become inoperable in the closed E.
Should one input or output channel fait.
position, the reactor may remain in opera-the remaining three channels shall be i
tion for a ceriod not to exceed seven (7) tested within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least nce days.
The remaining valves and actuating each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> until the system is testored circuitry shall be demonstrated to be to normal, operable within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least once each 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> until the system is restored F.
The UPS battery surveillance is as to operable status, described in Section 11.4.5.3.
f f
2 Should one isolation valve or depressuri-zing valve become inoperable in the open G.
The RDS containment genetration assem-position, during power operation, the blies seal pressure shall be examined plant will be brought to the cold shutdown at six-month intervals.
condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.
Only one RDS valve train, one input channel, one output channel and one UPS power supply may be out of service at any one time.
If these components are not returned to operable status within seven (7) days, a normal orderly shutdown shall be initiated within one (1) hour and the reactor shall be shut-
,3 down as described in Section 1.2.5(a) within twelve (12) hours and shutdown as described in Section 1.2.5(a) and (b) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
11-8 Amenilmen t No. IO
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Tables 3.5.2.h and b.5.2.h Instrumentation That Initiates RDS Operation 3 5 2.H - Limiting Conditions for Operation 4.5 2.11 - Surveillance Requirecient Minimum Protective Operable Conditions for Instrument Instrunent Channel Partsneter Channels Limiting Set Point Operability Trip Test Calibration Trip Low Steam Drum 3
Above or Equal to At Power Levels Monthly Each Major Level 25" Below Center Whenever the Refueling Line Reactor Is Critical With the Head On or when in llot Shutdown Fire Pump (s) 3 2,100 Paig Ditto Monthly Each Major Discharge Pressure Refueling Low Reactor 3
> 2'8" Above Top Monthly Each Major Water Level of Active Fuel Refueling 120-Second 3
< 120 Seconds Fol-Monthly liach Major Time Delay loving Low Steam Refueling Drum Level Signal Monthly Input Channels 3
A Through D Monthly Output Channels 3
I Through IV Monthly Monthly cFire Pump Start 1
cReference Specifications 3.1.4 and 4.1.h for Bases.
11-9 Ama:ndment No. 10 9
Limiting Conditions for Operation Surveillance Requirement 3.1.5 REAC'IDR DEPRESSURIZATION SYSTEM (Contd) 4.1.5 REACTOR del'RiiSSURIZATION SYST11M (Contd) 1 Bases:
The RDS provides for both manual and automatic depressurization of the primary system to allow injection of the core spray following a small-to-intermediate size break in the primary system. This will allow core cooling with the objective of preventing excessive fuel clad temperatures. The design of this system is based on the specified initiation set points described in Table 3 5 2.h.
Transient analyses reported in Section 6 of the RDS Descrip-tion, Operation and Performance Analysis submitted August 15, 19714 to the Directorate of Licensing USAEC, to demonstrate that these conditions result in adequate safety margins for both the fuel and the system pressure.
Performance analysis of the RDS is considered only with respect to its depressurizing effect in conjunction vit.h Therefore, no credit is taken for steem cooling of the core which provides further conservatism core spray.
to the emergency core cooling system.
These specifications ensure the operability of the RDS under all conditions for which the automat.ic or manual depressurization of the system is an essential response to the transient described above.
11-10 Amendment No. 10 t
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i Bases: (Contd)
'One RDS valve can remain out of service in the closed position for seven days because of redunda c the remaining RDS valves are test-operated immediately as described in Specification h l 5 C n y, provided for the actuation on the depressurizing system are reached, all the valves in the four blovdown paths are When conditions opened.
Each blovdown path is designed to pass lhh lb/second of steam at 1350 psig which is a third of the the required rate of depressurization. required total flow rate. ' Therefore, failure of one flow path to open u has been provided for the RDS which initiates action to mitigate the conse dent.
This set of specifications provides the limiting conditions of operation for the HDS.
the specifications are (1) to assure the effectiveness of the protective instrumentation when required even The objectives of during periods when portions of such systems are out of service for maintenance and (ii) to prescribe th settings required to assure adequate performance.
tional tests and calibrations, any one channel may be bypassed.Tb conduct the required input channel maintenance or If the input channel is not bypassed when functional tests and calibrations are performed, actual trip signals supersede test and calibration condition s.
The minimum functioncl testing frequency used in this specification is based on a frequency that has proven acceptable and conforms to that of the existing reactor protection system.
Four plant variablec are monitored and used as inputs to the actuation system.
level, (2) reactor water level, (3) motor-driven fire pump discharge pressure and (h) diesel engine driven fire These are (1) steam drum vater pump discharge pressure.
channels which are physically and electrically isolated from one another.These variables are jointly proce in each of the four variables is associated with each of the four input channels.Each gate into another channel.
One sensor enabled when two of the four input channels are in a tripped state.
The actuation of the RDS is The input channel is in a tripped state upon coincidence and subsequent pro motor-driven) and (3) low reactor water level.
level sensor associated with the input channel indicates a level of 25" below steam drum center linea This lov steam drum level signal initiates a two-minute delay which allows a containment evacuation interval prior to system blowdown and also pennits the incorporation of operator input to the system initiation logic specified in the design basis (Reference Section 3 3.D of the August 15, 1974 RDG Description, Operation and 9
11-11 Amendment No. 10 L
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Bases: (Contd)
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Performance Analysis).
For the latter, the operator is provided with manual timer reset capability for each of the four input channels at the control panel.
The lov steam drum level signal is also used to generate a fire pump start signal.
Verification of a fire pump start and thus verification that a source of core spray water is available at,the core spray valves is obtained when the pressure switch associated with the input channel at either fire pump discharge has tripped, corresponding to a pressure equal, to or exceeding 100 psig.
This variable is used as an enabling input to the actuation system to prevent depressurizing the reactor coolant aystem when the source of coolant' required to cool the core is not available.
A low reactor water level signal is generated when the input channel reactor water level sensor indicates a level 12'8" above the top of active fuel.
Lov reactor water level is confimation of the IDCA and with the other two inputs present (time delayed lov drun level and core spray water availability) causes the automatic trip of the input channel.
Thene trip level settings were chosen to be low enough to prevent spurious actuation but high enough to initiate RIC operation so that post-accident cooling can be accomplished.
Upon failure of an uninterruptible power supply (UPS) or a channel. power supply, the affected channel fault con-dition is alarmed as " channel 'X' unavailable."
Power failures associated with input channels cause the coin-cidence trip conditions for the input channels to change from 2-out-of-b to 2-out-of-3 The output channel actuation coincidence reverts to 3-of-3 upon failure of an output channel power supply.
Input channel bypass capability is provided to perinit bypassing any one input channel at a time.
The bypass feature is used to bypass a channel when the channel has failed to the " trip" state and/or when channel mainte-nance is required.
Bypassing of an input channel in the " trip" state or for maintenance causes the coincidence trip condition of the input channel to be. changed from 1-out-of-3 or 2-out-of-h, respectively, to 2-out-of-3 The input channel bypassed condition is alamed as " channel
'X' unavailable" and " bypassed."
Should an output channel require maintenance or should a single fault cause an output channel subchannel trip (two independent subchannels operate in 2 of 2 coincidence), the output chaanel actuation capability can be disabled by removing the associated 125 V DC supply.
The 125 V. DC supply to an output channel is disabled via
{
a circuit breaker in its respective UPS. 'ihe disabling of an output channel is alarmed as " channelX' unavailable."
Since 3-out-of-b output channels are required to assure design requirements are met (one output channel operaten one depressurizing valve and one isolation valve), the failure of one output channel vill not preclude achieving the required rate of depressurization. This redundancy also enables maintenance to be performed on one output channel while the plant is in operation.
l I
Amendment No. 10 h.---~% m.
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I Bases: (Contd)
' Once the RDS actuation system output channels are enabled (at least two input channels are in a tripped state or a manual trip is initiated) and tripped, they remain in that condition until they are manually reset. This reset can be accomplished.only after the initiating signals (ie, input channel trips or manual trip) have been restored to levels at which RDS operation is not required.
Separate, independent and one-hour sources of electrical power are provided, through four divisions, to accom-plish the detection of the LOCA and the completion of the depressurization.
Each of the divisions (1, 2, 3 and b) is supplied with electrical power from one of four independent uninterruptible power supplies (UPS) consistinc of a battery charger, a battery and an inverter.
Each UPS has output of 120 V AC, 60 Hz and 125 V DC.
Divisions 1 and 2 normally receive power from the existing 480 V AC Bus lA.
Divisions 3 and h are supplied by h80 V AC Bus 2A.
Normal station power to Busses lA and 2A can be provided by one of three sources:
(1) The station turbine generator, (2) the 138 kV transmission line or (3) the 46 kV transmission line.
Should none of these sources be available, provision is included for sup-Plying input power from the h80 V AC Bus 2B which is tied to the emergency diesel.
If all 480 V AC power is lost, the UPS is capable of sustaining its output for one hour.
Since only 3-out-of-h blowdown paths are required to assure adequate depressurization, the single system failure of one UPS division will not preclude achieving the required rate of depressurization.
This redundancy also enables maintenance to be performed on the UPS while the plant is in operation.
Four new containment penetration assemblie's are used in transmitting electrical power, control and instrumenta-tion signals between equipment located inside the containment building and facilities located external to the containment building.
These electrical penetrations are velded into spare containment penetration sleeves. The penetration assemblies are designed in accordance 'riin IEEE 31r and are seismically and environmentally quall-fled to the RDS design conditions.
The pressure retaining portion of the assemblics is designed and fabricated to the requirements of Subsection NE, Class MC vessels, of Section III of the ASME Code.
The penetration assemblies include a single aperture seal and a double electrical conductor seal and are designed to operate with the internal cavity pressurized with nitrogen at approximately 27 psig.
The relatively maintenance-free seal assemblics dictate a minimum inspection frequency of twice annually.
j Amendment No. 10 11-13
Limiting Conditions for Operation Surveillance Requirement ll.3 3.h CONTAINMENT SPRAY SYSTD4 11.h.3.h CONTAINMENT SPRAY SYSTD4 Applicability:
Applicability:
Applies to the operating status of the con-Applies to the testing of the containment tainment spray system.
spray system.
Objective:
Objective:
w.
'Do assure the capability of the containment To verify the operability of the containment spray system to reduce containment pressure spray system.
in the event of a Loss of Coolant Accident.
Speci fication :
Specification:
A.
Once each operating cycle, the following A.
During power operation each of the two shall be performed:
containment spray systems shall be operable.
1.
Autom' tic actuation of the contain-ment spray valve MO-706h (with water flow manually blocked).
B.
If Specification A is not met, a normal orderly shutd)wn shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the 2.
Calibration of flow instrumentation.
reactor shall be shut down as des-B.
At least once every six (6) months, cribed in Section 1.2.5(a) within except for periods of continuous 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and shutdown as described shutdown when the following shall
=
in Section 1.2.5(a) 6 (b) within the be. performed prior to startup:
.following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
V ri fy p rability of power-operated C.
Operability of the fire water supply va ves required for proper system and recirculation systems is governed actuation.
by Specification 11.3 1.h.
C.
Surveillance of fire water supply and recirculation systems is governed by Specification b.l.h.
1 D.
Instrument channels shall be tested and calibrated as listed in Table ll.h.3.h(a) 11-14 l
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TABLE 11.h.3.h Instrumentation That Initiates Enclosure Spray 11.h.3.h Gurveillance Requirement
_11.3.3.5 Limiting Conditions for Operation Instrument Trip Trip System Conditions for Test Including Instrument Parameter Logic Set Point Operability Valve Actuation Calibration Enclosure High 1 of 2
$2.2 Psig (a)
Power Operation Once Every Six Months Each Major Pressure and Refueling of Operations Other Refueling Operation Than Shutdown Time Delay (b) 1 of 1 213 Min, Power Operation Once Every Gix Months Each Major sl5 Min (a) and Refueling of Cperations Other Refueling Operation Than Shutdown (a) Primary enclosure spray setting.
(b) The time delay device requires power to perform the tripping function.
by the valve control circuit.
This supply is provided
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11-15 Amendment No. 10
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Bases:
The containment spray systems are provided to reduce pressure in the containment following a Loss of Coolant Accident. They are initially supplied from the fire water system and later by the core spray recirculation system.
They are not required to be in service at reactor coolant temperatures of 212 F or below because the resultant Loss of Coolant Accident pressure is not sufficient to pressurize the containment.
Operation of only one system is sufficient to provide the required containment spray flow.
The specified flow of apprcxLaately h00 gpm is sufficient to remove post-accident core energy releases including a substan-tial chemical reaction involving hydrogen generation to below design vals's.
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The operable status of these systems and components is demonstrated by periodic tests.
If a component is found to be inoperable, it vill be possible in most cases to effect repairs and restore the system to full operability within a relatively short time.
If a single system becomes inoperable, a redundant system has been provided with the ability to perform the spray function, but it reduces the redundancy provided by plant design and limits the ability to tolerate additional equipment failures.
Initiation of the containment spray system assures that containment design pressure vill not be exceeded due to hycrogen generation assuming the core spray systems do not function.
It has been conservatively calcu-lated that the energy release following a complete core meltdown (assuming no containment spray systems or core spray systems operate) would bring the containment pressure to approximately the design value (27 psig) about 15 minutes after the postulated accidant had occurred. Subsequent LOCA analysis system modifications and regulations have limited H2 generation such that it is no longer significant and calculations show that containment sprays are not required to prevent containment design pressures from being exceeded. Thus, the automatic actuation time of the primary containment spray system has been established at 15 minutes so as to allow the operator adequate time to evaluate and block actuation, if system operation is not required.
References:
1.
FHSR, Section 3 2.
Additional information in support of Proposed Technical Specification Change No 8 dated March 17, 1966.
3 Safety Evaluation by the Research and Power Reactor Safety Branch, Division of Reactor Licensing, Conaumers Power Company, Proposed Change No 8 dated April lb, 1966.
11-16 Amendment W.
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Limiting Conditi ns for Operation Surveillance Hequirement 11 3 5 3 DfERGENCY POWER SCUSCES 11.. 5 3 DERGENCY POWER SOURCES Applicability:
Applicability:
Applies to the operational status of the Applies to the periodic testing requirements emergency power sources.
for the emergency power sources.
Obj ective:
Objective:
w To assure the capability of the emergency To assure the operability of the emergency power sources to provide power required for power sources to provide emergency power in emergency equipment in the event of a Loss the event of a Loss of Coolant Accident.
of Coolant Accident.
Specification:
Specification:
A.
The emergency power system surveillance A.
For all reactor operating conditions ex-will be performed as indicated below.
In cept' cold shutdown, there shall normally addition, components on which maintenance be available one 138 kV line, one 46 kV has been performed will be tested.
line, one diesel generator system, one station battery system, and four RDS un-1.
During each operating cycie -
interruptible power supplies including batteries, except as specified below:
(a) Test of automatic initiation sensors and load test the emer-1.
Refueling operations and related gency diesel,to 180-200 kW testing may be conducted with the generator output for at least 138 kV line de-energized.
20 minutes.
P.
The h6 kV line or the diesel gener-(b) Test and calibrate the following ator may be out of service for repair instruments and controls nusoel-for periods up to three (3) days during ated with diesel generator:
. reactor operation and for extended periods during refueling or shutdown (1)
Fuel oil level, operationss (2) Oil Pressure tripping.
(3) Water temperature tripping.
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11-17 Amendment 'to. 10 L
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t Lir.itir.g Conditions for Operatior.
Surveillance Requirement 11.3 5.3 EMERGENCY POWER SOURCES (Contd) 11.h 5.3 EMERGENCY POWER SOURCES (Contd) 3 The diesel generator fuel supply shall (h) Overspeed tripping.
be adequate for~one-day operation.
(5) Battery undervoltage alarm.
4.
If Specifications A.2 or A.3 are not met, a normal orderly shutdown shall (c)
Verify tha automatic transfer of be initiated within one (1) hour and station psuer from the 138 kV the reactor shall be shuc down as line to the 46 kV line.
described in Section 1.2.5(a) within twelve (12) hours and shut down as (d)
Verify the automatic transfer of described in Seccion 1.2.5(a) and (b) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
During power sources for the lY and 2Y instrument and control panels, refueling operations cease all changes which could affect reactivity.
(e)' Verify the cells cell plates, and ba :tery ra'.4s show no visual 5.
The station battery system shall be indicacion of thysical damage or operable under all condicinas except abnornal deterioration.
during cold shutdown.
If the station battery is inoperable, no actions shall (f) Verify the cell-to-cell and ter-be taken which result in a reactivity minr1 connections are clean, tight, addition, except cooldown, or which frr.e of corrosion and coated with might result in the primary coolant anti-corrosion material.
system being drained.
(g) Verify that the battery charger 6.
If Specification A.5 is not met a g
will supply at least 30 amnares normal orderly shutdown of the reactor at a minimum of 135 volts for shall be initiated within one (1) hour at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. See note below, and the reactor shall be shut down as described in Section 1.2.5(a) within (h)
Verify that the battery capacity twelve (12) hours and shut down as is adequate to supp1v and ~~intain described in Section 1.2.5(a) and (b) in OPERABLE status all of the actual within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />..
emergency loads for the design time interval when the batterv is subjected 7,
one RDS unintorruptible power supply te a battery service test.
The design including battery may be out of service tiae interval for the RDS batteries as described in Section 3.1.5 Action 3.
is one hour and elgl'c hours for the station battery. Fee note below.
11-18 53 ""' I"* "
No. 10
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Limiting Conditions for Operation Surveillance Requirement-11.4.5.3 EMERGENCY POWER SOURCES (Contd)
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2.
Monthly -
(a)
Test start diesel generator and operate at 1 cast the fire pump as a load to 480 V Bus 2B for at least 20 minutes.
(b)
Verify that the cell voltage is A 2.0 volts and specific gravity is 2 1.2 of each cell of the station battery and the RDS batteries.
(c)
Test operate the rod position motor generator set.
3.
Weekly - The electrolyte level of each pilot cell is between the minimum and maximum level indication marks.
The pilot cell specific gravity, corrected to (77)oF, is ): 1.2.
The pilot cell voltage is pt 2.0 volts.
s The overall battery voltage is > 125 volts.
Test start the diesel generator and run for warn-up period.
4.
Sixty Months - At least once per 60 months during shutdown verify that the battery capacity is at least 80% of the manufacturers rating when subjected to a performance discharge test.
'lh is p e r-formance discharge test shall be performed nuhnequent to the satisfactory completion of the required battery service test.
Sm-Note below.
11-19 Amendment No. 10
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wr..m av >ta.u w.w.-n-v a+
n=- w w
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Note:
'Ihe:a curveillance items shall become ef fective prior to startup f olluwing the 19/7 refueling outage.
Bares:
Normal station power can Le provided by the station turbine generator, the 133 k'l tra::.: mission line or the h6 kV line.
These sour:es are adequate to pro tide emergency a-c power.
When " ne cble, a sin 61e emergency diesel generator rated at 200 kW starts sutaraticed;;.
V *.her,e so'ir :es is sysi 1-to 480 V Bus 23.
The weekly star
- i:.g tes' p: c;de emerge:.:y a-e Isrer tection of moving parts.
is base 1 t.a Manufacturer's Bulletin ?37h3-1 f?r relubrication pro-voltage sensors; one to detect 4 css of norr.11 power on Bus lA and the other to provid output prior to automatic closure of the generator output breaker.
to assure that the normal Buses lA and 2A are isolated prior to closing the generator output breakerAdditional breake prevents overloading of the ge:.erator at the uwltching ;:eriod.
This The diesel fuel oil tank is sized for two-day full load operation.
provide fuel makeup.
One-day supply is considered adequate to A single station battery supplies power for normal station services and is sized for emergency uses including valves and controls for Loss of Coolant Accidents.
g;nerator output if normal station power sources are not available.The battery can be charged from the emergency diesel The primary core spray valves and the primary containment spray valv'e are operated and controlled by power from the station battery.
from normal station power sources or the emergency diese1 generator.The backup core spray valves and ba RDS uninterruptible power supplies supply each division (except division 5).with electrical power.(UPS) A, B, C, and D, each consisting of a batte Each UPS has outputs of 120 VAC, 60 llz, and 125 VDC.
One of these batteries supplies control power for the emergency diesel generator.
Divisions 1 and 2 and 3 and 4 normally receive power from 480 VAC busses lA and 2A, respectively.
or both busses, provision is included for supplying input power from 480 VAC bus 2B which isIn the event of loss of power to eithe tied to the emergency diesel generator.
if all 480 VAC power is lost, the RDS UPS is capable of sustaining LLs outputs for one hour.
The
,2 a tation battery has adequate capacity to carry normal loads plus an assumed fatture (locked rotor current) of the DC lube oil pump for 54 minutes without the battery charger and still provide suf ficient power for equipment required to operate during a LOCA.
If steps are taken to reduce nonessential loads during a loss of of f-site power (such as ptrt of the emergency lights) additional time (up to five hours) can be gained from the time of the loss of the charger until the battery would no longer have sufficient power for equipment required to operate during a IUCA.
a tetion battery and the four (4) RDS batteries will be considered operable 1he if they are essenttally futly charged and the battery charger is in service.
Additionally prior to the startup following t he 1977 ref ueling outage, succe.stui cowpiction of nervice testing and performance discharge testing w' chin each operating cycle and each sixty months, rccpectively, will further establish battery reliability.
IL-20 Amendment No.,10 m,, m. _ _. - - - -- -
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