ML20129C487
| ML20129C487 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 10/15/1996 |
| From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20129C466 | List: |
| References | |
| NUDOCS 9610240039 | |
| Download: ML20129C487 (8) | |
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1 ATTACHNENT 2 CONSUNERS POWER COMPANY BIG ROCK P0 INT PLANT DOCKET 50-155 CURRENT TECHNICAL SPECIFICATION PAGES MARKED UP TO ILLUSTRATE CHANGES Submitted October 15, 1996 i
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7 Pages 9610240039 961015 PDR ADOCK 05000155 P
3.5.2 Operatina Reouirements Water addition to the containment sphere must be manually stopped before the accumulated water level reaches an elevation of 596 feet.
3.6 CONTAINMENT RE0VIREMENTS l
j7 Containment sphere integrity shall be maintained during power operation, refueling operation, shutdown and cold shutdown conditions
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except as specified by a system of procedures and controls to be established for occasions containment must be breached during cold J
shutdown.
If containent integrity cannot be maintained, the plant k
shall be brought to the snatdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and to the cold shutdown condition 4*.hin the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
8 d.3.7 CONTAINMENT SNlC;1C ((AXACC TESTI"C LE AK. KATE TESTN PRDM2 Am M
For the purpose of this specification, leakage rate is defined as the d
M percent of the contained atmosphere (weight basis) which escapes per 6
7 day (24 hrs) under the defined pressure conditions through any leaks d
in the containment boundary and all isolation valves and their x
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associated piping.
A The maximum allowable integrated leakage rate shall not exceed 0.5%/ day bq b;
of the containment atmosphere (weight basis) at the design pressure of gW g
27 psig. The procedure for containment sphere leakage testing shall be:
(a) At least once every 6 months, the personnel lock, the equipment
- 3 b d lock and the sphere supply-and-exhaust ventilation valves shall be o
jf-pressurized, dth air to 23 psig, to test their leak tightnens.
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The sum of leakage rates from these valves and locks shall be less E
than 0.25%/ day of the containment atoosphere (weight basis) at 23
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psig.
h (b) Each reactor shutdown for refueling, but in no case at intervals kb o
greater than two years, the following valves shall be tested for i5 I'
operability from both the manual and automatic modes of operation d
d and, at the same time, shall be tested for leak tightness by means g
of a pressure test utilizing air or the normal working fluid at a pressure not less than 23 psig:
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$23.
d[
Main Steam Isolation (MO-7050) 4h y
MainsteamDrain(No-7065) pIs Clean-Up System Resin Sluice (CV-4091, CV-4092, CV-4093) h a
Reactor.and Fuel Pit Drain Isolation (CV-4027, CV-4117) d Reactor Enclosure ' Clean Sump Isolation (CV-4031, CV-4102) te d
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Reactor Enclosure Dirty Susp Isolation (CV-4025, CV-4103)
E Reactor Enclosure Treated Waste Valve (CV-4049) a
- 0perability, automatic controls and instrumentation tests required only if valve is opened for use during operation.
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. b't to Amendment M7,117 January 16, 1996
4.1.2 (Centd) is above 300 psig. The shutdown cooling system shall be OPERABLE and ready for service during REFUELING OPERATIONS and the breakers for MO-7070 and NO-7071 shall be tagged "open".
The primary coolant shall be sampled and analysed daily during periods of POWEE OPERATION. The following are absolute limits which if exceeded shall necessitate reactor SHUTDOWN. Corrective action will necessarily be taken at more stringent limits to minimize the possibility of these absolute limits ever being reached.
j Conduct ivity (Micronho/cm) h inum 5
l Maximum Trawient*
10 pH (Loicr and Upper Limits) 4.0 and 10.0 Chlorid,eIonQpe) 1.0 100 Boron ((pm) r i
f Isotopic analysis of the primary coolant to determine the DOSE EQUIVALENT I-131 concentration shall be performed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during periods of operation.
1.
If the DOSE EQUIVALENT I-131 concentration exceeds 0.2 pC1/mi and is less than or equal to 4.0 pCi/st, isotopic analysis to determine DOSE EQUIVALENT I-131 shall be performed every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> until the activity is less than 0.2 pci/al.
2.
If the DOSE EQUIVALENT I-131 exceeds 4.0 pC1/m1, the plant shall be placed in a SHUTDOWN condition with the main steam
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isolation valve closed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
(c) Leakane Limits 1.
If the primary coolant system leakage exceeds 1 spa and the pource of leakage is not identified, the reactor shall be SHUTDOWN as described in Section 1.2.5(s) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cooldown to a COLD SHUTDOWN condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
2.
If leakage from the primary coolant system exceeds 10 gym, the reactor shall be SMUTDOWN as described in Section 1.2.5(a) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cooldown to a COLD SHUTDOWN condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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- Conductivity is expected to increase temporarily after startups from cold shutdown. The maximum transient value here stated is the maximum permissible and applies only to the period subsequent to a cold shutdown between criticality and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 201 rated power.
23 Amendment 107 February 19, 1992 TS SECTION 4
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4.1.3 ' Racetor Safety System Bypass The following tabulation gives the permissive functional cor.ditions during which certain reactor safety system sensors are bypassed by the reactor safety system mode selector switch. A keylock reactor mode switch shall be provided having " Shutdown," " Refuel," "8ypass Dump Tank" and "Run" positions.
These positions shall'have the following functions:
Mode nelector Switch Position Trip Functio' massed I
Run Mone *)
i Bypass Dump Tank *)
Low Steam Drum Water Level I
Recirculation Waterline Valves Closed a
Steam Line Backup Isolation Valve 3
Closed High Water Level in Scram Dump Tank (b)
High Condenser Pressure RefuelId) i Low Steam Drum Water Level Recirculation Waterline Valves closed I
l Steam Line Backup Isolation Valve Closed
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High condenser Pressure d.LDLU STEAM DR>fA LUATEf?.
i Shutdown 57 RAM LIUE.DOCKUP I6DLATDU VRLVE j
C LDSE.D
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H16H C0f0DEMfEC. PEECCultC RL.CW.CVLA T)Dhl lNRTLtLIN E VALOLS Cl:DCL D (a) Control rod withdrawal !s prevented by interlock while switch is in 4
this mode posi' tion.
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(b) Bypass of this trip function is necessary to enable emptying the
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dump tank af ter a scram.
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i (c)With the mode switch in the " shutdown" position, both the scram j
circuit and the control rod withdrawal circuit are open. The ventilating duct circuit power supply is transferred to a point which provides penetration closure protection through signals from "high containment sphere pressure" and " low water level in reactor vessel."
s This permits normal ventilation in the containment sphere during shutdown when the control rods are held in the full-in position.
Me..f 0.;.. ;;;. ;;f;;y ey;;;; ;i;;;;i; ;;; typ;;;d ;ir.;; ;te.e i; ;,
- e :: Athf cr re-*-^1
-ad (d)With the mode switch in the refuel position and the crane positioned over the reactor vessel, crane operation is prevented if any one rod is withdrawn from full-in position.
(*)Righ condenser pressure reactor trip is automatically bypassed any time steam drum pressure is below a set point maximum of 500 psis.
54 Amendment 107 February 19, 1992 TS SECTION 6
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i ADMINISTRATIVE CONTROLS l
6.9.2.2.A (Contd) i 3.
Solid Waste i
The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined i
by 10 CFi Part 61) shipped offsite during the report period a.
Container burial volume, b.
Total curie quantity (specify whether determined by measurement or estimate),
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- Principal radionuclides (specify whether determined by c.
measurement or estimate),
d.
Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),
Type of container (e.g., LSA, Type A, Type 8, Large e.
Quantity), and f.
Solidification agent or absorbent (e.g., cement, asphalt).
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4.
Radiological Impact on Man g
i The Radioactive Effluent Release Report " '
'O d;y; cft:: ?::=:; 1 ef ::d ;::: d:11 include potential doses to individuals and populations calculated using measured effluent and averaged meteorological data in accordance with the methodologies in the Offsite Dose Calculation Manual.
a.
Total body and significant organ doses (greater than 1 m1111 Rem) to individuals in unrestricted areas from receiving water-related exposure pathways.
b.
ne maximum offsite air doses (greater than 1 millitad) due to beta and gamma radiation at locations near ground level from gaseous effluents.
c.
Organ doses (greater than 1 millites) to individuals in unrestricted areas from radioactive iodine and radioactive material in particulate form from the major pathways of esposure.
d.
Total body doses (greater than 1 mantes) to 'the population and aveease doses (greater than 1 milliten) to individuals in the population from receiving water-related pathways to a distance of 50 alles from the site.
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e6 Amendment 107 February 19, 1992 TS SECTION 10
ADMINISTRATIVE CONTROLS
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- 6.12.1 (Contd) c.
A radiation protection qualified individual (e.g., Health Physics Technician) with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area, and shall perform periodic radiation surveillance as specified by the O.st;tg xd Radiation Protection Supervisor in the l
RWP.
g mec 6.12.2 In addition to the requirements of Spe fication 6.12.1, areas accessible to personnel with radiati levels greater than 1000 mrem /h
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at 30 cm (12 inches) but less than 00 rad /h at I meter from the
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radiation source or from any ;;ai e from which the radiation
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penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control l
of the Shift Supervisor on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area.
In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.
For individual high radiation areas accessible to personnel with radiation levels of greater than 1000 mR/h that are located within large areas where no enclosure exJsts for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activated as a warning device.
6.13 ENVIRONMENTAL QUALIFICATION (Deleted) 6.14 PROCESS CONTROL PROGRAM (PCP) 6.14.1 Changes to the PCP shall be submitted to the Commission in the Radioactive Effluent Release Report for the period in which
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the change (s) was made effective. This submittal shall contain:
a.
Sufficiently detailed information to support the rationale for the change; b.
A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and c.
Documentation of the fact that the change has been reviewed and approved by the responsible Nuclear Operations Department per CPC-2A (Quality Program).
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90 Amendment H 4, 116 August 2, 1995
Limitina Conditions for Ooeration Surveillance Requirements 1
l 11.3.5.3 ENERC DCY POWER SOURCES (Contd) 11.4.5.3 ENERC DCY POWER SOURCES (Contd) initiated within one (1) hour and (h) Verify that the capacity of the station battery, the RDS batteries
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the reactor shall be shutdown as described in Section 1.2.5(a) within and the alternate shutdown battery l
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twelve (12) hours and shutdous as is adequate to supply and described in Section 1.2.5(a) and maintain in OPERABLE status 2
(b).within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
all of the actual emergency loads for the design time 7.
One RDS uninterruptible power supply interval when the battery is including battery may be out of service subjected to a battery service
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test. The design time interval i
as described in section 3.1.5 Action Ka..
for the RDS batteries is one l
S.
During reactor power operaties, the hour, two hours for the 138 kV line may be out of service for station battery and seventy-j two hours for the alternate repair for periods up to three (3) days.
shutdown battery.
9.
If Specification A.8 is not met, a j
normal orderly shutdown shall be (i) Test and calibrate the 2400 volt j
initiated within one (1) hour and bus undervoltage trip control
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the reacter shall be shutdown as components as follows:
described in section 1.2.5(a) within
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twelve (12) hours and shutdown as (1) The undervoltage relays described in Section 1.2.5(a) and 127-10ZY, ZZ and YZ will (b) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
drop out on decreasing voltage of no lower than 107.1 volts, after a delay j
S.
During power and refueling operat.cos of <.6 seconds.
I the 2400 volt bus undervoltage ermponents j
shall be operable or placed in.the tripped (2) The auxiliary timing relay l
condities, except during the monthly 162-104 will be actuated channel functiemal testing period.
after a 10 2 0.5 second time delay upon receiving a signal from all three (3) j l
undervoltage relays.
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l 110 A W at 107 f
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l February 19, 1992 i
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TB SECTION 11
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BASES FOR 13.1.3.1 - GASEOUS EFFlVENT DOSE RATE (Contd) if or to less than or equal to 3000 nrems/he to the skin. These release rate
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limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1
1500 arems/yr.
The required detection capabilities for radioactive materials in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs).
Detailed discussion of the LLD and other detection limits can be found in HASL Procedures Manual, RASL-300 (revised annually), Currie, L A, " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal Chen 40, 586-93 (1968), and Hartwell, J K " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).
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135 Amendment M7,116 August 2, 1995