ML20149E640
| ML20149E640 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 01/05/1988 |
| From: | Perkins K Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20149E644 | List: |
| References | |
| NUDOCS 8801130422 | |
| Download: ML20149E640 (27) | |
Text
>Q Croo UNITED STATES
,og NUCLEAR REGULATORY COMMISSION g
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- E WASHINGTON, D. C. 20655 7;
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WISCONSIN ELECTRIC POWER COMPANY DOCVIT NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.110 License No. DPR-24 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated April 10, 1986 as revised by letter dated July 17, 1987 complies with the standards and requirements of theAtomicEnergyActof1954,asamended(theAct),andthe Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confonnity with tho application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assu.=nce (1) that the activities authorized by this amendment can be conJJ:ted without endangering the health 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 amendment will not be inimical to the common 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.
P 8801130422 880105 DR ADOCK O 6
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4 1 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No.
DPR-24 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.110, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective imediately upon issuance.
1 The Technical Specifications are to be implemented within 20 days from the date of issuance.
]
FOR THE NUCLEAR REGULATORY COMMISSION l
Kenneth E. Perkins, Director i
Project Directorate III-3 Division of Reactor Projects
Attachment:
Changes to the Technical Specifications i
Date of Issuance: January 5,1988
a asci, j UNITED STATES
'g NUCLEAR REGULATORY COMMISSION 3
g 7
- E W ASHINGTON, D. C. 20555 49.....,o WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.113 License No. DPR-27 1.
The Nuclear Regulatory Comission (the Comission) has found thec:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated April 10, 1986 as revised by letter dated July 17, 1987 complies with the standards and requirew,nts of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CER Chapter I; i
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) that the e.ctivities authorized by this arrendment can be conducted without endangering the health 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 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.
i i 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B cf Facility Operating License No.
DPR-27 is hereby amended to read as follows:
B.
_ Technical Specifications The Technical Specifi.:ations contained in Appendices A and B, as revised through Anendment No.113, are hereby incorporated in the license.
The licensee shall operate the facility in l
l accordance with the Technical Specifications.
1 I
3.
This license amendmeat is effective imediately upon issuance.
The Technical Speci'/ications are to be implemented within 20 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION a
Kenneth E. Perkins, Director Project Directorate III-3 Division of Reactor Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: January 5, 1988 I
i
a 1
ATTACHMENT TO LICENSE AMENDMENT NOS.110 AND 113 TO FACILITY OPERATING LICENSE N05. DPR-24 AND DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise,'opendix A and B Technical Specifications by removing the pages identifie3 below and inserting the enclosed pages.
The revised pages are identi.'ied by amendment number and contain marginal lines indicating the area of change.
APPENDIX A REMcVE INSERT 2
15.1 2 15.1-2 15.3.1-14a 15.3.1-14a 15.3.1-15 15.3.1-15 Table 15.3.5-5 (1 pg.)
Table 15.3.5-5 (1 pg.)
15.3.10-6 15.3.10-6 15.3.12-1 15.3.12-1 15.3.13-2 15.3.13-2 15.4.4-7 15.4.4-7 15.4.4-11 15.4.4-11 15.4.5-4 15.4.5-4 15.4.6-2 15.4.6-2 15.4.15-3 15.4.15-3 Figure 15.6.2-2 Figure 15.6.2-2 15.6.3-1 15.6.3-1 15.6.3-2 15.6.3-2 15.6.4/5.1 15.6.4/5.1 15.6.9-1 15.6.9-1 i
15.6.9-2 15.6.9-2 15.6.10-1 15.6.10-1 15.6.10-2
'5.6.10-2 15.6.12-1 15.6.12-1 l
15.7.8-1 15.7.8-1 l
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d.
Containment Integrity
- Containment integrity is defined to exist when:
- 1) All non-automatic containment isolation valves and blind flanges are closed as required.
- 2) The equipment hatch is properly closed.
- 3) At least one door in each personnel air lock is properly closed.
- 4) All automatic containment isolation valves are
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operable or are secured closed.
- 5) The uncontrolled containment leakage satisfies Specification 15.4.4.
e.
Protective Instrumentation Logic
- 1) Analog Channel An analog channel is an arrangement of components and modules as required to generate a single protective action signal when required by a plant condition. An analog channel loses its identity where single action signals are combined.
- Containment isolation valves are discussed in FSAR Section 5.2.
Unit 1 - Amenchent No. 43.50.85.110 Unit 2 - Amendment No. 48 E6,89,113 15.1-2
. ~..
If leakage is to another system, it will be detected by the plant radiation monitors and/or water inventory control.
Continuous monitoring of steam generator tube leakage is accomplished by either the individual unit Air Ejector Radiation Monitor, the combined Air Ejector Radiation Monitor, or the Steam Generator Blowdown Radiation Monitor in combination with periodic surveillance of the primary coolant activity. Backup monitoring can be accomplished by sampling secondary coolant gross activity.
References FSAR Section 6.5, 11.2.3 l
Unit 1 - Amendment No.10,110 Unit 2 - Amendment No. J2,113 15.3.1-14a
E.
Maximum Reactor Coolant Oxygen and Chloride and Fluoride Concentration For Power Operation Specification:
1.
The concentration of oxygen in the reactor coolant shall not exceed 0.1 ppm.
2.
The concentration of chloride and of fluoride in the reactor coolant shall not exceed 0.15 ppm each.
3.
If the oxygen concentration and the chloride or fluoride concentration of the reactor coolant simultaneously exceed the limits given in 1) and 2) respectively, corrective action is to be taken immediately to return the system to within nonnal operation specifications.
If the nonnal operational limits are not achieved within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor is to be brought to a hot shutdown condition.
If the system is not brought to within specifications within an additional 24-hour period, I
the system is to be brought to a cold shutdown condition, and the cause of the out-of-specification operation ascertained and corrected.
Basis:
By maintaining the oxygen, chloride and fluoride concentration in the reactor coolant within the limits as specified by E 1), 2) and 3), the functional integrity of the materials of the Reactor Coolant System is assured under all operating conditions.(I)
If these limits are exceeded, measures can be taken to correct the condition, e.g., replacement of ion exchange resin or adjustment of the hydrogen concentration in the volume control tank (2), and further because of the time-I dependent nature of any adverse effects arising from oxygen, chloride and fluoride concentration in excess of the limits, it is unnecessary to shut down l immediately since the condition can be corrected. Thus the period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for corrective action to restore the concentrations within the limits has been established.
If the corrective action has not been effective at the end of the 24-hour period, then the reactor will be brought to the hot shutdown l
Unit 1 - Amendment No.110 Unit 2 - Amendment No.113 15.3.1-15
4 TABLE 15.3.5-5 (Cont'inued)
MINIMUM NO. OF OPERABLE OPERATOR ACTION IF CONDITIONS OF COLUMN 2 4
NO.
FUNCTIONAL. UNIT CHANNELS CHANNELS CANNOT BE MET 7.
Containment High Range 3
2 If operability cannot be restored within Radiation Monitor seven days after failure, prepare a specia!
report to be submitted within thirty days in accordance with 15.6.9.2.D.
l 1
8.
Containment High Range 2
1 If operability cannot be restored within 48 Pressure Monitor hours, be in hot shutdown within twelve hours.
I 9.
- a. C'entainment Water Level 2
1 Operation may continue up to thirty days.
If
}
Keyway operability cannot be restored, be in hot j
shutdown within the next twelve hours.
t l
- b. Containment Water Level 2
1 If the operability cannot be restored within 1
Sump B Continuous Indication 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, be in hot shutdown within twelve l
hours.
I j
10.
C6ntainment Hydrogen Monitors 4
1 If operability cannot be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, be in hot shutdown within the next~
i six hours.
j 11.
Reactor Vessel Fluid Level 4
1 If operability cannot be restored within 48 System hours, be in hot shutdown within the next twelve hours.
1 12.
In-Core Thennocouples 4/ core 2/ core If operability of at least two themocouples quadrant quadrant per core quadrant cannot be restored within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, be in hot shutdown within the next twelve hours.
~
13.
Main Steam Line Radiation 1/ steam 1/ steam If operability cannot be restored within seven Monitors (SA-11) line line days, prepare a special report to be submitted within thirty days in accordance with 15.6.9.2.E.
l Unit 1 - Amendment No. 92,110 Uni't 2 - Amendment No. 96,113
a.
The RCCA does not drop upon removal of stationary gripper coil voltage.
b.
The RCCA does not step in properly when the proper voltage sequences are applied to the control rod drive mechanism coils.
It shall then be assumed inoperable until it has been tested to verify that it does drop.
c.
If the bank demand position is greater than or equal to 215 steps, or, less than or equal to 30 steps, and the rod position indicator channel shows a misalignment from the bank demand position of 15 inches, the RCCA shall be assumed inoperable I
until it has been tested to verify that it does step properly.
d.
If the bank demand position is between 215 steps and 30 steps, and the rod position indicator channel shows a misalignment from the bank demand position of 7.5 inches, the RCCA shall be l
assumed inoperable until it has been tested to verify that it does step properly.
2.
Specification 15.3.10.C.1.b can be modified by the following:
a.
If an RCCA does not step in upon demand, up to six hours is j
allowed to determine whether the problem with stepping is an electrical problem.
If the problem cannot be resolved within
]
six hours, the RCCA shall be assumed. inoperable until it has been verified that it will step in or would drop upon demand.
b.
If more than one RCCA does not step in, apparently due to electrical problems, the situation shall be rectified or clearly defined that it is an electrical problem and the RCCAs are capable of dropping upon demand or an orderly shutdown shall commence within six hours.
3.
No more than one inoperable RCCA shall be permitted during sustained power operation.
4.
When it has been detemined that an RCCA does not drop on removal of stationary gripper coil voltage, the shutdown margin shall be main-tained by boration as necessary to compensate fo the withdrawn worth of the inoperable RCCA.
If sustained power operation is anticipated, the insertion limit shall be adjusted to reflect the worth of the inoperable RCCA.
Unit 1 - Amendment No. A9,76, 110 15.3.10-6 Unit 2 - Amendment No. E5,80, 113 7
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15.3.12 CONTROL ROOM EMERGENCY FILTRATION Applicability:
Applies to the operability of the control room emergency filtration.
Objective:
To specify functional requirements of the control room emergency filtration during power operation and refueling operation.
Specification:
1.
Except as specified in 15.3.12.3 below, the control room emergency filtra-tion system shall be operable at all times during power operation and refueling operation of either unit.
a.
The results of in-place cold D0P and halogenated hydrocarbon tests, I
conducted in accordance with Specification 15.4.11, on HEPA filter and charcoal adsorber banks shall show a minimum of 99% DOP removal and 99% halogenated hydrocarbon removal.
b.
The results of laboratory charcoal adsorbent tests, conducted in accordance with Specification 15.4.11, shall show a minimum of 90% removal of methyl iodide.
If laboratory analysis results for in-place charcoal indicate less than 90% nethyl iodide removal, this specification may be met by replacement with charcoal adsorbent which has been verified to achieve 90% minimum removal and which has been stored in sealed containers, and retesting the charcoal adsorber bank for halogenated hydrocarbon removal.
c.
The results of fan testing, conducted in accordance with specifica-tion 15.4.11, shall show operation within + 10% of design flow.
i Unit 1 - Amendment No.110 15.3.12-1 Unit 2 - Amendment No.113
_ _ _ _ _ - _ - _ - - - _ _ _ _ _ _ _ _ _ _ = - _ - _ -
c.
Basis Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during plant startup and shutdown. The consequence of an inoperable snubber is an increase in the prooability of structural damage to piping as a result of seismic or other events ini-tiating dynamic loads.
It is therefore required that all snubbers required to protect the p;'imary coolant system, and other safety related systems or componer.ts, be operable during reactor operation.
Because the snubber protee. tion is required only during relatively low probability events, a period of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is allowed for repairs or replace-ment.
In case a shutdown is required, the allowance 6f 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to reach a Cold Shutdown condition will permit an orderly shutdown consistent with standard operating procedures.
Since plant cower operation should not connence with known defective safety relateo equipment, Specification l
15.3.13.4 prohibits reactor startup wich inoperable snucbers.
1 l
Unit 1 - Amendment No. J/,110 15'3.13-2 Unit 2 - Amendment No. 22, 113
2.
Visual inspection shall be made for excessive leakage from components of the system. Any significant leakage shall be measured by collection and weighing or by another equivalent method.
B.
Ac:eptar.ce Criterion The maximum allowable leakage from the fiesidual Heat Removal System compenents (which includes valve ster..s. flanges and pump seals) shall not exceed two gallons per hour.
C.
Corrective Action Repairs shall be made as required to maintain leakage within the cceeptance criterion of IV-B.
D.
Test Frequency Tests of the Residual Heat Removal System shall be conducted at shutdown for major refueling.
V.
Annual Inspection A detailed visual examination of the accessible interior and exterior surfaces of the containment structure and its components shall be perfcor.ed annually and prior to any integrated leak test, to uncover any evidence of deterioration which may affect either the containment's structural integrity or leak-tightness. The discovery of any significant deterioration shall be accompanied by corrective actions in accordance with acceptable procedures, l
' nondestructive tests and inspections, and local testing where practical, prior to the conduct of any integrated leak test. Such repairs shall be reported as part of the test results.
Unit 1 - Amer.dment No.110 15.4.4-7 Unit 2 - Amendment No.113
IX. Liner Plate A.
The liner plate will be examined before the initial pressure test to determine the following:
(1) Locate areas which have inward deformations. The magnitude of the inward deformations will be measured and recorded.
The arei, will be permanently marked for future reference.
The inward deformations will be measured between the angle stiffeners which are on 15-inch centers. The measurements will be accurate to +.01 inch.
(2) Try to locate areas having strain concentrations by visual examination paying particular ettention to the condition of the liner surface.
Record the locatior, of any areas having strain concentrations.
B.
Shortly after the initial pressure test and at about one year after initial start-up, reexamine the areas located in section (A).
Measure and record inward deformations.
Record observations per-taining to strain concentrations.
C.
If the difference in the measured inward deformations exceeds 0.25 l
inch (for a particular location) and/or changes in strain concentra-tion exist, then an investigation will be made. The investigation l
will determine the cause and any necessary corrective actior,.
D.
The surveillance program will only be continued beyond the one year l
after initial start-up inspection if some correctivo action was needed.
If requiret, the frequency of inspection for a continued surveillance program will be dec.. mined shortly after the "one year after initial start-up inspection".
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1 l
Lan c 1 - Amendment No.110 15.4.4 11 Unit 2 - Amendment No.113
operability of these systems is therefore to combine systems tests to be performed during refueling shutdowns, with more frequent component tests, which can be performed during reactor operation.
The systems tests demonstrate proper automatic operation of the Safety Injection and Containment Spray Systems.
With the pumps blocked from starting, a test signal is applied to initiate automatic action, and verification is made that the components receive the safety injection signal in the proper sequence. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry.(I)
During reactor operation, the instrumentation which is depended on to ini-tiate safety injection and containment spray is generally checked weekly and the initiating circuits are tested monthly (in accordance with Specifi-cation 15.4.1).
In addition, the active components (pumps and valves) are to be tested monthly to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The test interval,
of one month is based on the judgement that more frequent testing would not significantly increase the reliability (i.e. the probability that the compo-nent would operate when required), yet more frequent testing wculd result in increased wcar over a long period of time.
Other systems that are also important to the emergency cooling function are the accumulators, the Component Cooling System, the Service Water System and the containment fan coolers. The accumulators are a passive safeguard.
In accordance with Specification 15.4.1 the water volume and pressure in the accumulators are checked periodically. The other systems mentioned operate when the reactor is in operation and by these means are continuously monitored for satisfactory perforTnance.
References (1) FSAR Section 6.2.
Unit 1 - Amendment No.110 15.4.5-4 Unit 2 - Amendment No.113 1
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3.
Each diesel generator shall be given an inspection, at least annually, following the manufacturer's recommendations for this class of stand-by service.
4.
Each fuel oil transfer pump shall be run monthly.
The above tests will be considered satisfactory if all applicable equipment operates as designed.
B.
Station Batteries 1.
Every month the voltage of each cell (to the nearest 0.05 volt),
the specific gravity and temperature of a pilot cell in each battery and each battery-voltage shall be measured and recorded.
2.
Every 3 months the specific gravity, the height of electrolyte, I
and the amount of water added, for each cell, and the temperature of every fifth cell, shall be measured and recorded.
3.
At each time data is recorded, new data shall be compared with old to detect signs of abuse or deterioration.
4.
Each battery shall be subjected to a load test at intervals reconmended by the manufacturer but not exceeding five years.
The battery voltage as a function of time shall be monitored to establish that the capacity is sufficient to carry the loads as delineated in FSAR Table 8.2-3 for the specified length of l
time.
All electrical connections will be checked for tightness.
Unit 1 - Amendment No. 2,110 15.4.6-2 Unit 2 - Amendment flo. 2,113 e-94=ww..-
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G.
Fire Pump Diesel Battery and Charger Test Frequency 1.
a.
Verify electrolyte level above Weekly l
the plates b.
Verify that the overall battery Weekly voltage is > 24 volts 2.
Verify the specific gravity is Quarterly appropriate for continued service of the battery 3.
a.
Verify that the battery, cell 18 months plates and battery racks show no visual indication of. physical damage or abnormal deterioration b.
Verify that the battery to battery 18 months and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material Basis ilormally, the fire protection system is not in use. :llowever, the system components are required to perform as designed in the event of a fire emergency.
The National Fire Protection Association and the plant insurance carrier have specified periodic tests and inspections to demonstrate fire protection equipment operability. The listed tests and inspections include and exceed the requirements of these organi:ations. Testing more frequently than that listed is not considered necessary to insure operability and perfo rmance.
Unit 1 - Amendment No. 32,110 15.4.15-3 Unit 2 - Amendment No. 36,113 i
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15.6.3 Facility Staff Qualifications 15.6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions or as clarified in 15.6.3.2 through 15.6.3.4.
15.6.3.2 Except as provided in 15.6.3.3, the Health Physicist shall meet j
the following requirements:
a.
The individual shall have a bachelor's degree or the equivalent in a science or engineering subject, including some formal training in radiation protection. For purposes of this para-graph, "equivalent" is as follows:
(1) Four years of formal schooling in science or engineering; or (2) Four years of applied radiation protection experience at a nuclear facility; or (3) Four years of operational or technical experience or training in nuclear power; or (4) Any combination of the above totalling four years.
b.
Except as provided in d., below, the individual shall have at least five years of professional experience in applied radia-tion protection. A master's degree in a related field is equivalent to o'le year of experience and a doctor's degree in a related field is equivalent to two years of experience, c.
Except as provided in d.,
below, at least three of the five i
years of experience shall be in applied radiation protection work in a nuclear facility dealing with radiological problems similar to these encountered in nuclear power plants, d.
If the individual has a bachelor's degree specifically in health physics, radiological health, or radiation protection, at least three years of professional experience is required; if the individual has a master's or a doctor's degree specif-ically in health physics, radiological health, or radiation protection, at least two years of professional experience is required. This experience shall be in applied radiation i
protection in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power plants, l
15.6.3-1 Unit 1 - Anendment No. A3,52, 110 Unit 2 - Amendment No. AB,58, 113 i
15.6.3.3 In the event the position of Health Physicist is vacated and the' proposed replacement does not meet all the qualifications of 15.6.3.2, but is determined to be otherwise well qualified, then concurrence of HRC shall be sought in approving the qualification of that individual.
15.6.3.4 -The Duty Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering disci line with specific training p
in plant design and response and analysis of the plant for transients and accidents. The Duty Technical Advisor shall also receive training in plant design and layout including the capabilities of instrumentation and controls in the control room, i
i 1
Unit 1 - Amendment No. 52,110 15.6.3-2 Unit 2 - Amendment No. 58,113 i
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15.6.4 TRAllill1G 15.6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Superintendent - Training and shall meet or exceed the requirements and recommendations of Section 5.5 of AtiS!
til8.1-1971 and Appendix "A" of 10 CFR Part 55.
15.6.4.2 A training program for the Fire Brigade shall meet or exceed the requirements of Section 27 of the flFPA Code-1976, except that the meeting frequency may be quarterly, 15.6.5 REVIEW Atl0 AUDIT 15.6.5.1 Hanager's Supervisory Staff 15.6.5.1.1 The fianager's Supervisory Staff (tiSS) shall function to advise the Manager on all matters related to nuclear safety.
15.6.5.1.2 The Manager's Supervisory Staff shall be selected from the following:
Chairman: Manager - Point Beach fluclear Plant Member:
General Superintendent Member:
Superintendent - Operations Member:
Superintendent - tiaintenance &
Construction Member:
Superintendent - Engineering, Quality
& Regulattry Services Member:
Superintendeit - Training Member:
Superintenden - Technical Services Member:
Superintendent - Reactor Engineering Member:
Radiochemist Member:
Health Physicisc Member:
Superintendent - Instrumentation &
Control 15.6.5.1.3 Alternate members may be appointed by the MSS Chaiman to serve on a temporary basis; however, no more than two alternates shall vote in MSS at any one time. Such appointment shall be in writint.
1 Unit 1 - Amendment tio. H.110 15.6.4/5-1 Unit 2 - Amendnent tio. 95.113 e
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15.6.9 PLAllT REPORTING REQUIRE!!Ef1TS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following program for reporting of operating information shall be followed.
Reports should be addressed to the Regional Administrator, Region III, unless other-l wise noted.
15.6.9.1 Routine Reports A.
Startup Report 1.
A sumary report of plant startup and power escalation testing which addresses each of the tests identified in the FSAR and includes a l
general description of the measured values obtained during the test program and a compari-son of these values with design predictions and specifications must be submitted under the following conditions:
a.
Receipt of an operating license.
b.
Amendment to the license involving a planned increase in power level.
c.
Installation of fuel that has a different design I
or has been manufactured by a different fuel
- supplier, d.
Modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
Any corrective actions that were required to obtain satisfactory operations shall also be described.
2.
This report shall be submitted within the earliest time frame of the following:
1 Unit 1 - Amendnent No. A3,110 15.6.9-1 Unit 2 - Amendment No. AS, 113 p
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a.
90 days following completion of the startup tests.
b.
90 days following resumption or comencement of commercial, power operation, c.
9 months following initial criticality.
B.
Annual Results and Data Report 1.
A results and data report covering the period of the previous calendar year shall be submitted prior to March 1 of each year.
2.
This report shall include:
a.
Complete results of steam generator tube inservice inspection completed during the calendar year as required by specification 15.4.2.A.7 b.
A tabulation on an annual basis of the number of station, utility, and other personnel receiving exposures greater than 100 mrem / year and their associated man-rem exposure according to work and job functions. The dose assignments to various duty functions may be estimates based on pocket dosimeter, T!.D or film badge measurements. Small exposures totalling less than 20 percent of the individual total dose need not be accounted for.
In the aggregate, at least 807, of the total whole body dose received from external sources shall be assigned to specific major work functions, c.
A description of facility changes, tests or experiments as required pursuant to 10 CFR 50.59(b).
d.
A tabulation of all challenges to the pressurizer po'wer operated relief valves or pressurizer safety valves:
Unit 1 - Miendment 79,110 15.6.9-2 Unit 2 - Amendment EA, 113
15.6.10 PLANT OPERATING RECORDS Specification Records and logs relative to the following items shall be retained for 5 years I
unless a longer period is required by applicable regulations.
A.
Records of normal plant operation, including power levels and periods of op.ution St each power level shall be retained for 5 years except those records of transient or operational m cles for reactor coolant system (RCS) components having a limited number <
design transients, shall be retained for the duration of the operating : 'ense.
B.
Records of p'.incipal maintenance activities, including inspection, repair, substitution, or replacement of items of equipment pertaining to nuclear safety shall be retained for a period of 5 years where these requirements do not conflict with requirements of 10 CFR 50.49(j),10 CFR 50.59, and surveillance requirements of these Technical Specifications. The quality assurance, environmental qualification, installation, and service life ricords of. components covered by these requirements shall-be retained for the duration of the Operating License.
C.
Records of Licensee Event Reports.
D.
Records of installation, environmental qualification, periodic checks, inspections, and calibrations of equipment pertaining to nuclear safety to verify that surveillance requirements tre being met will be retained for the duration of the Operating License. All other records of this type will be retained for 5 years.
E.
Records of new and spent fuel inventory and assembly histories (5 years fc11owing transfer).
F.*
Records of dusign modifications made to systems and equipment, including drawings, as described in the FSAR.
G.*
Rec'ords of plant radiation and contamination surveys.
l H.*
Records of off-site environmental surveys, l
\\
1.*
Records of radiation exposure of all individuals entering radiation controlled areas of the plant, including records for preparation of i
NRC-4 forms, bioassay and whole body counting results; and records of l
- Items will be retained for the duration of the Operating License.
l Unit 1 - Amendment No. PE,110 15.6.10-1 Unit 2 - Amendment No. 97,113
individual exposures exceeding 40-MPC hour limits, including evaluations and actions taken.
J.*
Records of gaseous and liquid radioactive material released to the environ-ment and dilution of these wastes.
K.*
Records of any special reactor tests or experiments.
L.
Records of changes made in the Operating Procedures.
M.
Records of sealed source and fission detector leak tests and results performed pursuant to Specification 15.4.12, including annual physical inventory results verifying accountability of sources.
N.
Records of training, qualification and requalification for NRC-licensed personnel shall be retained for the duration of the operator's license per 10 CFR 55 requirements.
Records of fire brigade member training, including drill critiques shall be maintained."or 3 years in accordance with 10 CFR 50, Appendix R Section I.4 requirements.
0.*
Records of in-service inspections performed pursuant to these technical specifications.
P.*
Records of Quality Assurance activities required by the QA Manual shall be maintained for the duration of the Operating License except those QA records relating to radioactive materials shipping packages, which shall be main-tained for the lifetime of the packaging per '#0 CFR 71.91(c) requirements.
Q.*
Records of reviews performed for changes made to procedures or equipment, or reviews of tests and experiments pursuan. '.o 10 CFR 50.59 and as required per Specification 15.6.5.1.6.
R.*
Records of meetings of the Manager's boervisory Staff and the Off-Site Review Comittee.
S.*
Records of avlyses for radiological environmental monitoring.
T.
Records of radioactive material shipments having a specific activi? of greater than 0.002 microcurie / gram shall be retained for a period of 2 years in accordance with 10 CFR 71.91(a).
U.
Records concerning the Security Plan, procedures, testing, maintenance, and j
audit shall be maintained in accordance with the Comission-approved PBNP fiodified Amended Security Plan.
- Items will be retained for the duration of the Operating License.
15.6.10-2 Unit 1 - Amendment No.97,110 Unit 2 - Amen 6nent No. JN,113
15.6.12 ENVIRONMENTAL QUALIFICATION A.
By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualificd in accordance with the provisions of:
Division of Operating Reactors "Guidelines for Evaluating Environmental Qualification of Class lE Electrical Equipment in Operating Reactors" l
(D0R Guidelines); or NUREG-0588, "Interin Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," December 1979.
e Copies of these documents are attached to Order for Modification of Licenses DPR-24 and DPR-27 dated October 24, 1980.
B.
By no later than December 1,1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to docume.it the degree of compliance with the D0R Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.
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I Ofttt 66166 10/16/B0, Unit 1 - Amendnent No, 110 15.6.12-1 Unit 2 - Amendment No.
113
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15.7.8 ADMINISTRATIVE CONTR01.S 15.7.8.1 Duties of the Manager's Supervisory Staff The duties of the Manager's Supervisory Staff with respect to these radiological effluent technical specifications are listed in specification 15.6.5.1.6 at items J. and k.
15.7.8.2 Audii.s A.
An audit of the activities encompassed by the Offsite Dose Calculation Manual and the Process Control Program and its implementing pracedures shall be performed at least once every 24 months utilizing either offsite licensee personnel or a consulting fim.
B.
An audit of the radiological environmental monitoring program and the results thereof shall be perfomed at least once every li' months utilizing either offsite licensee personnel or a qualified consulting fim.
i C.
The results of the aujits in A and B above shall be trans-mitted to the Vice-President - Nuclear Power and.the Chairman of the Offsite Review Comittee.
15.7.8.3 Plant Operating Procedures The ODCM and the PCP shall be established and maintained in accordance with 'he provisions of specification 15.6.8.
Effluent t
and environmental monitoring shall be addressed in the Quality Assurance Program.
15.7.8.4 RETS Reporting Requirements The following written reports shall be submltted to the Administra-i tor, U.S. Nuclear Regulatory Comission Region III with a copy to the Director, Office of Inspection and Enforcement, USNRC, Washington, D.C. 20555 within the time periods specified.
A, t emiannual Mcnitoring Report A report within 60 days after January 1 and July 1 ear.h year l
for the six month period or fraction thereof, ending June 30 and December 31 containing:
1.
Infomation relative to the quantities of liquid, gaseous and solid radioactive effluents released from the facility, and effluent volumes used in maintaining the releases Unit 1 - Amendment No. 97,110 15.7.8-1 Unit 2 - Amendment No. J01 113
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