ML20235Y460
| ML20235Y460 | |
| Person / Time | |
|---|---|
| Site: | River Bend  |
| Issue date: | 03/03/1989 |
| From: | Calvo J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20235Y463 | List: |
| References | |
| NUDOCS 8903140170 | |
| Download: ML20235Y460 (11) | |
Text
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GULF STATES UTILITIES COMPANY DOCKET NO. 50-458 R1VEREENDSTATION, UNIT 1 AFO R ENT TO FACILITY __ OPERATING LICENSE Amendment No. 35 License No. NPF-47 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Gulf States Utilities Company (the licensee) dated September 28, 1988, as supplemented Novenber 30, 1988, and modified February 6,1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as arnended, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable kssurance:
(1) that the activities authorized by this amendmer.t can be conducted without endangering the he61th and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this license amendment wi not be inimical to the comen 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.
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Accordir. gly, the license is arended by changes to the Technical j
l Spccificaticris as indicated in the attachnent to this license amendrent and Paragraph 2.C.(2) of Facility Operating License No. NPF-47 is hereby an, ended to read as follows:
(2) Technical Specifications and Environmental Protection Plan l
The Technical Specifications contained in Appendix A, as revised through l
Amendment No. 35 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. GSU shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance.
l FOR THE NUCLEAR REGULATORY COMMISSION d
G.
G A-O Jose A. Calvo, Director Project Directorate - IV Division cf Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reacter Regulatier, Attachrrent:
Changes to the Technical Specifications Date of Issuance: March 3, 1989
ATTACHMENT TO LICENSE AVENDMENT NO. 35 FACILITY OPEFATING LICENSE NO. NPF-47 ECCKET NO. 50-458
' Replace the fc110 wing page of theLAppendix "A" Technical Specifications with the enclosed page. The revised P45E is identified by Anendnent number and contains a vertical line ir.dicating the area of changt. Overiccf page prcvided to meintain document completer.tss, PEF 0VE PAGES INSERT PAGES 1-6 1-6 3/4 6-2 3/4 6-2 3/4 9-6 3/4 9-6 B 3/4 6-1 s 3/4 6-1 l
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i 5a A
DEFINITIONS l
of a logic circuit, from sensor through and including the actuated device, to verify OPERABILITY.
The LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.
MEMBER (5) 0F THE PUBLIC 1.24 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupa-tionally associated with the plant.
This category does not include employees of the utility, its contractors or vendors.
Also excluded from this category are persons who enter the site to service equipment or to make deliveries.
This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.
MINIMUM CRITICAL POWER RATIO 1.25 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.
OFFSITE DOSE CALCULATION MANUAL (ODCM)
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1.26 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints.
It shall also contain a table and figure defining current radiological environmental monitoring sample locations.
OPERABLE - OPERABILITY 1.27 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of )erforming its specified function (s) and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are e.lso capable of performing their related support function (s).
OPERATIONAL CONDITION - CONDITION 1.28 An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.
PHYSICS TESTS 1.29 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.
RIVER BEND - UNIT 1 1-5 I
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DEFINITIONS PRESSURE BOUNDARY LEAKAGE l
1.30 PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a reactor coolant system component body, pipe wall or vessel wall.
PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING 1.31 PRINARY CONTAINMENT INTEGRITY - FUEL HANDLING shall exist when:
All containment penetrations required to be closed during accident I
a.
conditions are closed by at least one manual valve, blind flange,
{
or deactivated automatic valve secured in its closed position.
Up to twelve vent and drain line pathways may be opened under administrative control for the purposes of surveillance testing l
proviced the total calculated flow rate through the open vent and drain line pathways is less than or equal to 70.2 cfm.
b.
All containment hatches are closed.
c.
Each containment air lock is in compliance with the requirements of Specification 3.6.1.4.
PRIMARY CONTAINMENT INTEGRITY - OPERATING 1.32 PRIMARY CONTAINMENT INTEGRITY - OPERATING shall exist when:
a.
All containment penetrations required to be closed during accident conditions are either:
1.
Capable of being closed by an OPERABLE containment automatic isolation system, or 2.
Closed by at least one manual valve, blind flange, or deacti-vated automatic valve secured in its closed position, except as provided in Specification 3.6.4.
b.
All containment equipment hatches are closed and sealed.
c.
Each containment air lock is in compliance with the requirements of Specification 3.6.1.4.
d.
The containment leakage rates are within the limits of Specification 3.6.1.3.
e.
The suppression pool is in compliance with the requirements of Speci-fication 3.6.3.1.
f.
The sealing mechanism associated with each primary containment penetra-tion; e.g., welds, bellows or 0-rings, is OPERABLE.
PROCESS CONTROL PROGRAM (PCP) 1.33 The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Part 20, 10 CFR Part 61, 10 CFR Part 71 and i
RIVER BEND - UNIT 1 1-6 Amendment No.35
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT PRIMARY CONTAINMENT INTEGRITY - OPERATING LIMITING CONDITION FOR OPERATION 3.6.1.1 PRIMARY CONTAINMENT INTEGRITY - OPERATING shall be maintained.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2* and 3.
ACTION:
Without PRIMARY CONTAINMENT INTEGRITY - OPERATING, restore PRIMARY CONTAINMENT INTEGRITY - OPERATING within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.6.1.1 PRIMARY CONTAINMENT INTEGRITY - OPERATING shall be demonstrated:
a.
After each closing of each penetration subject to Type B testing, except the primary containment air locks, if opened following Type A or B test, by leak rate testing the semis with gas at Pa, 7.6 psig, and verifying that, when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Surveillance Requirement 4.6.1.3.d for all other Type B and C penetrations, the combined leakage rate is less than 0.60 La.
b.
At least once per 31 days by verifying that all primary containment
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penetrations ** not capable of being closed by OPERABLE primary con-tainment automatic isolation valves abd required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in position, except as provided in Specification 3.6.4.
By verifying each primary containment air lock is in compliance with c.
the requirements of Specification 3.6.1.4.
d.
By verifying the suppression pool is in compliance with the requirements of Specification 3.6.3.1.
"See Special Test Exception 3.10.1
- Except valves, blind flanges, and deactivated automatic valves which are located inside the primary containment, steam tunnel or drywell, and are locked, sealed or ctherwise sscured in the closed position. These penetrations shall be vartfied clo:ed during each COLD SHUTDOWN except such verification need not be performed more often than once per 92 days.
t RIVER BEND - UNIT 1 3/4 6-1
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4 CONTAINMENT SYSTEMS PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING LIMITING CONDITION FOR OPERATION 3.6.1.2 PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING shall be maintained.
APPLICABILITY:
Operational Condition *#
ACTION:
Without PRIMARY CONTAINMENT INTEGRITY - FUEL' HANDLING, suspend handling of irradiated fuel in the primary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
SURVEILLANCE REQUIREMENTS 4.6.1.2 PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING shall be demonstrated:
a.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to entering and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during Operational Condition
- by verifying that all primary con-tainment penetrations required to be closed during accident condi-tions are closed by hatches, valves, blind flanges, or deactivated automatic valves secured in position.**
l b.
By verifying each containment air lock is in compliance with the requirements of Specification 3.6.1.4.
"When handling irradiated fuel in the primary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
- Up to twelve vent and drain line pathways may be opened under administrative control for the purposes of surveillance testing provided the total calculated flow rate through the open vent and drain line pathways is less than or equal to 70.2 cfa.
- See Special Test Exception 3.10.1.
RIVER BEND - UNIT 1 3/4 6-2 Amendment No.35
REFUELING OPERATIONS 3/4.9.3 CONTROL ROD POSITION LIMITING CONDITION FOR OPERATION 3.9.3 All control rods shall be inserted.*
APPLICABILITY: OPERATIONAL CONDITION 5, during CORE ALTERATIONS.**
ACTION:
With all control rods not inserted, suspend all other CORE ALTERATIONS.
SURVEILLANCE REQUIREMENTS 4.9.3 All control rods shall be verified to be inserted, except as above specified:
a.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to:
1.
The start of CORE ALTERATIONS.
2.
The withdrawal of one control rod under the control of the reactor mode switch Refuel position one-rod-out interlock, b.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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"Except (1) control rods removed per Specification 3.9.10.1 or 3.9.10.2 or (2) one control rod may be withdrawn under control of the reactor mode switch Refuel position one-rod-out interlock.
- See Special Test Exception 3.10.3.
RIVER BEND - UNIT 1 3/4 9-5
S REFUELING OPERATIONS 4
3/4.9.4 DECAY TIME i
LIMITING CONDITION FOR OPERATION j
i 3.9.4 The reactor shall be subcritical for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />."
APPLICABILITY:
OPERATIONAL CONDITION 5, during movement of irradiated fuel in the reactor pressure vessel.
ACTION:
With the reactor subtritical for less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel.
l SURVEILLANCE REQUIREMENTS
'4.9.4 The reactor shall be determined to have been subtritical for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verification, prior to movement of irradiated fuel in the reactor pressure vessel, of the date and time of subcriticality.
I "The reactor shall be suberitical for at least 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> prior to opening vent and drain line pathways under the provisions of Specification 3.6.1.2.
RIVER BEND - UNIT 1 3/4 9-6 Amendment No.35
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3.4.6 CONTAINMENT SYSTEMS BASES-3/4.6.1 CONTAINMENT 3/4.6.1.1 PRIMARY CONTAINMENT INTEGRITY - OPERATING PRIMARY CONTAINMENT INTEGRITY OPERATING ensures that the rel' ease of radio-active materials from the primary containment atmosphere will bc restricted to those leakage paths and associated leak rates assumed in the a;cident analyses.
This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.
3/4.6.1.2 PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING PRIMARY CONTAINMENT INTEGRITY - FUEL HANDLING ensures that the release of radio-active materials from the primary containment atmosphere will be restricted to those leakage pathways and associated leak rates assumed in the accident anal-ysis. This restriction will limit the site boundary radiation doses to less than 25% of the 10 CFR 100 limits during a postulated fuel handling accident within the primary containment.
The footnote allows the opening of vent and drain line pathways for the purposes of performing leak rate surveillance testing.
Offsite doses as a result of a postulated fuel handling accident inside primary containment were calculated based upon the flow rate which would be produced through twenty open 3/4 inch vent and drain line pathways with a 0.367 inch water gauge differential pressure between the containment and auxiliary buildings. This would result in a total calculated flow rate out of these open vent and drain lines of 70.2 cubic feet per minute. Accordingly, this footnote allows the opening of up to twelve vent and drain line pathways for the purposes of performing leak rate surveillance testing provided the total calculated flow rate, considering the specific con-tainment and auxiliary building differential pressure, does not exceed 70.2 cubic feet per minute as used in the above analyses.
3/4.6.1.3 PRIMARY CONTAINMENT LEAKAGE The limitations on primary containment leakage rates ensure that the total primary containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure of 7.6 psig Pa.
As an added conservatism, the measured overall integrated leakage rate is further limited to less than or equal to 0.75 La, during performance of the periodic tests, to account for possible degradation of the prie.1ry containment leakage barriers q
between leakage tests.
Operating experience with the main steam line isolation valves has indi-l cated that degradation has occasionally occurred in the leak tightness of the q
valves; therefore, the special requirement for testing these valves.
The surveillance testing for measuring leakage rates is consistent with
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the requirements of Appendix J to 10 CFR 50.
I RIVER BEND - UNIT 1 B 3/4 6-1 Amendment No.35
CONTAINMENT SYSTEMS BASES I
3/4.6.1.4 PRIMARY CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the primary containment air locks are required to meet the restrictions on PRIMARY CONTAINMENT INTEGRITY-OPERATING and the primary containment leakage rate given in Specifica-tions 3.6.1,1 and 3.6.1.3.
The' specification makes allowances for the fact
'that there may be long periods of time when the air locks will be in a closed and secured pcsition during reactor operation.
Only one closed door in each air lock is required to maintain the integrity of the primary containment.
3/4.6.1.5 MAIN STEAM-POSITIVE LEAKAGE CONTROL SYSTEM Calculated doses resulting from the maximum leakage allowance for the isolation valves in the postulated LOCA situations would be a small fraction of the 10 CFR 100 guidelines. Operating experience has indicated that degra-dation has occasionally occurred in the leak tightness of the MSIVs.
Inclu-sion of the specified leakage control system will prevent untreated leakage from the MSIVs when isolation of the primary system and containment is required.
This system includes the Main Steam Shutoff Valves (MSSV).
3/5.6.1.6 PRIMARY CONTAINMENT STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the unit.
Structural integrity is required to ensure that the containment will withstand the maximum pressure of 7.6 psig in the event of a LOCA.
A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.
3/4.6.1.7 PRIMARY CONTAINMENT INTERNAL PRESSURE The limitations on primary containment internal pressure ensure that the containment peak pressure of 7.6 psig does not exceed the design pressure of 15.0 psig during LOCA conditions or that the external pressure differential does not exceed the design maximum external pressure differential of + 0.6 psid or the differential at which water would overflow the weir wall into the dry-well of 0.58 psid. The limit of -0.3 to + 0.3 psig for initial primary con-tainment internal pressure will limit the peak primary containment pressure to 7.6 psig which is less than the design pressure and is consistent with the safety analysis.
3/4.6.1.8 PRIMARY CONTAINMENT AVERAGE AIR TEMPERATURE The limitation on primary containment average air temperature ensures that the containment peak air temperature does not exceed the design temperature of 185'F during LOCA conditions and is consistent with the safety analysis.
RIVER BEND - UNIT 1 B 3/4 6-2 Amendment No.35
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