ML20085B435
| ML20085B435 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/26/1995 |
| From: | Mckee P Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20085B442 | List: |
| References | |
| NUDOCS 9506150039 | |
| Download: ML20085B435 (12) | |
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4 UNITED STATES
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. NUCLEAR REGULATORY COMMISSION e
WASHINGTON, D.C. 20066 4001
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NORTHEAST NUCLEAR ENERGY COMPANY. ET AL.
DOCKET NO. 50-423 MILLSTONE NUCLEAR POWER STATION. UNIT NO. 3 l
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 115 License No. NPF-49 1.
The Nuclear Regulatory Commission (the Commission) has found that:-
A.
The application for amendment by Northeast Nuclear Energy Company, et al. (the licensee), dated January 23, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and-regulations set forth in 10 CFR Chapter I; j
B.
The facility will operate in conformity with the application, the f
provisions of the Act, and the rules and regulations of the Commission; i
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 I
conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will r.ot 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 Commission's regulations and all applicable requirements have been satisfied.
9506150039 950526 PDR ADOCK 05000423 P
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-49 is hereby amended to read as follows:
(2)
Technical Specifications The Technical Specificatjons contained in Appendix A, as revised through Amendment No.11
, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance, to be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Phillip F Kee, Directo Project D rectorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: May 26, 1995
4 ATTACHMENT TO LICENSE AMENDMENT NO. 115 FACILITY OPERATING LICENSE NO. NPF-(1 DOCKET NO. 50-423 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert viti viii ix ix xiv xiv 3/4 5-10 3/4 5-10 3/4 6-14 3/4 6-14 B 3/4 5-3 thru B 3/4 5-5 B 3/4 6-2 B 3/4 6-2 L
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LIMITINE C0ISITIONS FOR OPERATION Ale SURVEILLANCE REQUIREMENTS SECTION PAfig FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC l
ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >lpC1/ gram l
DOSE EQUIVALENT I-131..................
3/4 4-30 l
TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM.........................
3/4 4-31 l
3/4.4.9 PRESSURE / TEMPERATURE LIMITS...............
3/4 4-33 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS -
APPLICABLE UP TO 10 EFPY................
3/4 4-34 FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -
APPLICABLE UP TO 10 EFPY................
3/4 4-35 TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -
WITHDRAWAL SCHEDULE...................
3/4 4-36 Pressurizer.......................
3/4 4-37 Overpressure Protection Systems.............
3/4 4-38 FIGURE 3.4-4a NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (FOUR LOOP OPERATION) 3/4 4-40 FIGURE 3.4-4b NOMINAL MAXIMUM ALLOWABLE PORV SETPOINT FOR THE COLD OVERPRESSURE SYSTEM (THREE LOOP OPERATION).......
3/4 4-41 3/4.4.10 STRUCTURAL INTEGRITY 3/4 4-42 3/4.4.11 REACTOR COOLANT SYSTEM VENTS..............
3/4 4-43 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCLMULATORS......................
3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T GREATER THAN OR EQUAL TO 350*F 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,LESS THAN 350*F.........
3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK..............
3/4 5-9 3/4.5.5 pH TRIS 001UM PHOSPHATE STORAGE BASKETS.........
3/4 5-10 3/4.6 CONTAllNIENT SYSTEMS 3/4.6.1 PRIMARY CONTAIMENT Containment Integrity.................
3/4 6-1 Containment. Leakage 3/4 6-2 Containment Air Locks.................
3/4 6-5 Containment Pressure..................
3/4 6-7 NILLSTONE - LMIT 3 viii Amendment No. pp, 77, pp.115 oao.
1 IEEX j
LINITING COM ITIONS FOR OPERATION AW SURVEILLANCE REQUIREMENTS SECTION EAGE Air Temperature 3/4 6-9 Containment Structural Integrity...........
3/4 6-10 Containment Ventilation System............
3/4 6-11 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Cont'ainment Quench Spray System 3/4 6-12 Recirculation Spray System..............
3/4 6-13 3/4.6.3 CONTAIMENT ISOLATION VALVES.............
3/4 6-15 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors 3/4 6-16 Electric Hydrogen Recombiners 3/4 6-17 3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM Steam Jet Air Ejector 3/4 6-18 3/4.6.6 SECONDARY CONTAINMENT Supplementary Leak Collection and Release System...
3/4 6-19 Secondary Containment Boundary............
3/4 6-22 Secondary Containment Boundary Structural Integrity.................
3/4 6-23 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Valves 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING FOUR LOOP OPERATION 3/4 7-2 TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING THREE LOOP OPERATION 3/4 7-2 t
l NILLSTONE - UNIT 3 ix Amendment No. JJ JJ, 77, pf, J77115 0308 l
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DGEX BASES SECTION EME TABLE B 3/4.4-1 REACTOR VESSEL FRACTURE TOUGHNESS PROPERTIES.. B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE................. B 3/4 4-10 3/4.4.10 STRUCTURAL INTEGRITY.................. B 3/4 4-15
'3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............. B 3/4 4-15 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS...................... B 3/4 5-1 3/4.3.2 and 3/4.5.3 ECCS SUBSYSTEMS............... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK.............. B 3/4 5-2 3/4.5.5 pH TRIS 00IUM PHOSPHATE STORAGE BASKETS......... B 3/4 5-3 3/4.6 CONTAIMENT SYSTEMS 3/4.6.1 PRIMARY CONTAIMENT................... B 3/4 6-1 j
3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS
.......... B 3/4 6-2 3/4.6.3 CONTAINMENT ISOLATION VALVES.............. B 3/4 6-3 t
3/4.6.4 COMBUSTIBLE GAS CONTROL................. B 3/4 6-3 i
3/4.6.5 SUBATMOSPHERIC PRESSURE CONTROL SYSTEM......... B 3/4 6-3b
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3/4.6.6 SECONDARY CONTAl m ENT.................. B 3/4 6-4 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE...................... B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION..... B 3/4 7-3 3/4.7.3 REACTOR PLANT COMPONENT COOLING WATER SYSTEM...... B 3/4 7-3 3/4.7.4 SERVICE WATER SYSTEM.................. B 3/4 7-3 3/4.7.5 ULTIMATE HEAT SINK................... B 3/4 7-3 3/4.7.6 FLOOD PROTECTION.................... B 3/4 7-4 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM........ B 3/4 7-4 3/4.7.8 CONTROL ROOM ENVELOPE PRESSURIZATION SYSTEM....... B 3/4 7-4 3/4.7.9 AUXILIARY BUILDING FILTER SYSTEM............ B 3/4 7-4 3/4.7.10 SNUBBERS
........................ B 3/4 7-5 MILLSTONE - UNIT 3 xiv Amendment No. f), pp.115 ons
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EllEMENCY CORE C00 LIM SYSTEll5 3/4.5.5 nH TRIS 00llBI PHOSPHATE STORME RASKETS LIMITING COMITION FOR OPERATICII 3.5.5 The trisodium phosphate (TSP) dodecahydrate Storage Baskets shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3 and 4 AGIIM:
With the TSP Storage Baskets inoperable, restore the system TSP Storage Baskets to OPERABLE status within 7 days or be in at least HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOW the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REQUIRDIENTS 4.5.5 The TSP Storage Baskets shall be demonstrated OPERABLE at least once each REFUELING INTERVAL by verifying that a minimum total of 974 cubic feet of TSP is contained in the TSP Storage Baskets.
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NILLSTONE LSIIT N0. 3 3/4 5-10 Amendment No.115 am
f Intentionally Left Blank i
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i MILLSTONE - UNIT 3 3/4 6-14 Amendment No. JJ, pp, 115
kgip Wit ildif ul L II I I EHEK alCY C0RE C00 LING SYSTEMS 3/4.5.5 TRIS 0DItal PH0SPHATE STORAGE BASKETS BASES l
BACKGROUND 1
1 i
Trisodium phosphate (TSP) dodecahydrate is stored in porous wire mesh l
baskets on the floor or in the sump of the containment building to ensure that lodine, which may be dissolved in the recirculated reactor cooling water i
i following a loss o'f coolant accident (LOCA), remains in solution. TSP also helps inhibit stress corrosion cracking (SCC) of austenttic stainless steel components in containment during the recirculation phase following an accident.
1 Fuel that is damaged during a LOCA will release iodine in several l
chemical forms to the reactor coolant and to the containment atmosphere. A I
portion of the iodine in the containment atmosphere is washed to the sump by 1
containment sprays (i.e., Quench Spray and/or Containment Recirculation Spray). The emergency core cooling water is borated for reactivity control.
This borated water causes the sump solution to be acidic.
In a low pH (acidic) solution, dissolved iodine will be converted to a volatile form. The volatile lodine will evolve out of solution into the containment atmosphere, significantly increasing the levels of airborne iodine. The increased levels of airborne iodine in containment contribute to the radiological releases and increase the consequences from the accident due to containment atmosphere leakage.
After a LOCA, the components of the core cooling and containment spray systems will be exposed to high temperature borated water.
Prolonged exposure to the core cooling water combined with stresses imposed on the components can cause SCC. The SCC is a function of stress, oxygen and chloride concentrations, pH, temperature, and alloy composition of the components.
High temperatures and low pH, which would ba present after a LOCA, tend to promote SCC. This can lead to the failure of necessary safety systems or components.
Adjusting the pH of the recirculation solution to levels above 7.0 prevents a significant fraction of the dissolved iodine from converting to a volatile form. The higher pH thus decreases the level of airborne iodine in containment and reduces the radiological consequences from containment atmosphere leakage following a LOCA. Maintaining the solution pH 2 7.0 also reduces the occurrence of SCC of austenitic stainless steel components in containment. Reducing SCC reduces the probability of failure of components.
Granular TSP dodecahydrate is employed as a passive form of pH control for post LOCA containment spray and core cooling water. Baskets of TSP are placed on the floor or in the sump of the containment building to dissolve NILLSTONE UNIT N0. 3 8 3/4 5-3 Amendment No.115 0303
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n ENERSENCY C0RE COOLING SYSTEMS BASES (continued)
BACKGROUND (continued) from released reactor coolant water and containment sprays after a LOCA.
Recirculation of the water for core cooling and containment sprays then provides mixing to achieve a uniform solution pH. The dodecahydrate form of TSP is used because of the high humidity in the containment building during normal operation.
Since the TSP is hydrated, it is less likely to absorb large amounts of water from the humid atmosphere and will undergo less physical and chemical change than the anhydrous form of TSP.
APPLICABLE SAFETY ANALYSES The LOCA radiological consequences analysis takes credit for iodine retention in the sump solution based on the recirculation water pH being 2 7.0.
The radionuclide releases from the containment atmosphere and the consequences of a LOCA would be increased if the pH of the recirculation water were not adjusted to 7.0 or above.
LIMITING CONDITION FOR OPERATION The TSP is required to adjust the pH of the recirculation water to 2 7.0 after a LOCA. A pH 2 7.0 after a LOCA is necessary to prevent significant amounts of iodine released from fuel failures and dissolved in the recirculation water from converting to a volatile form and evolving into the containment atmosphere. Higher levels of airborne iodine in containment may increase the release of radionuclides and the consequences of the accident. A pH 2 7.0 is also necessary to prevent SCC of austenitic stainless steel components in containment. SCC increases the probability of failure of components.
i The required amount of TSP is based upon the extreme cases of water volume and pH possible in the containment sump after a large break LOCA. The minimum required volume is the volume of TSP that will achiava a sump solution pH of 2 7.0 when taking into consideration the maximum possible sump water volume and the minimum possible pH. The amount of TSP needed in the 1
containment building is based on the mass of TSP required to achieve the desired pH. However, a required volume is specified, rather than mass, since it is not feasible to weigh the entire amount of TSP in containment.
The l
minimum required volume is based on the manufactured density of TSP dodecahydrats. Since TSP can have a tendency to agglomerate from high humidity in the containment building, the density may increase and the volume decrease during normal plant operation. Due to possible agglomeration and increase in density, estimating the minimum volume of TSP in containment is conservative with respect to achieving a minimum required pH.
MILLSTONE UNIT N0. 3 B 3/4 5-4 Amendment No.115 0303
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EMERGENCY CORE C0OLING SYSTEMS BASES (continued)
APPLICABILITY In MODES I, 2, 3, and 4, a design basis accident (DBA) could lead to a fission product release to contatnment that leaks to the secondary containment boundary. The large break LOCA, on which this system's design is based, is a
. full-power event. Less severe LOCAs and leakage still require the system to be OPERABLE throughout these MODES. The probability and severity of a LOCA decrease as core power and reactor coolant system pressure decrease. With the reactor shut down, the probability of release of radioactivity resulting from such an accident is low.
In MODES 5 and 6, the probability and consequence of a DBA are low due to the pressure and temperature limitations in these MODES. Under these conditions, the SLCRS is not required to be OPERABLE.
h ACTIONS If it is discovered that the TSP in the containment building sump is not within limits, action must be taken to restore the TSP to within limits.
During plant operation, the containment sump is not accessible and corrections may not be possible.
The 7-day Completion Time is based on the low probability of a DBA occurring during this period. The Completion Time is adequate to restore the volume of TSP to within the technical specification limits.
If the TSP cannot be restored within limits within the 7-day Completion Time, the plant must be brought to a MODE in which the LCO does not apply.
The specified Completion Times for reaching MODES 3 and 4 are those used throughout the technical specifications; they were chosen to allow reaching the specified conditions from full power in an orderly manner and without challenging plant systems.
SURVEILLANCE REQUIREMENTS Surveillance Reauirement 4.5.5 Periodic determination of the volume of TSP in containment must be performed due to the possibility of leaking valves and components in the containment building that could cause dissolution of the TSP during no' mal r
operation. A Frequency of once each REFUELING INTERVAL is required to determine visually that a minimum of 974 cubic feet is contained in the TSP Storage Baskets.
This requirement ensures that there is an adequate volume of TSP to adjust the pH of the post LOCA sump solution to a value 2 7.0.
The periodic verification is required every refueling outage, since access to the TSP baskets is only feasible during outages. Operating experience has shown this Surveillance Frequency acceptable due to the margin in the volume of TSP placed in the containment building.
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i NILLSTONE UNIT N0. 3 8 3/4 5-5 AmeI1dmentNo.115 0303
CONTAHOIENT SYST21 na m 3/4.6.1.6 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 facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 60 psia in the event of a LOCA.
A visual inspection in conjunction with the Type A leakage tests is sufficient to demonstrate this capability.
l 3/4.6.1.7 CONTAINMENT 'fENTILATION SYSTEM The 42-inch containment purge supply and exhaust isolation valves are required to be locked closed during plant operation since these valves have not been demonstrated capable of closing during a LOCA or steam line break accident. Maintaining these valves closed during plant operations ensures that excessive quantities of radioactive materials will not be released via the Containment Purge System. To provide assurance that these containment valves cannot be inadvertently opened, the valves are locked closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or prevents power from being supplied to the valve operator.
The Type C testing frequency required by 4.6.1.2d is acceptable, provided that the resilient seats of these valves are replaced every other refueling outage.
3/4.6.2 DEP 3/4.6.2.1 and 3/4.6.2.2 CONTAINMENT OVENCH SPRAY SYSTEM and RECIRCULATION SPRAY SYSTEM The OPERABILITY of the Containment Spray Systems ensures that containment depressurization and iodine removal will occur in the event of a LOCA.
The pressure reduction, iodine removal capabilities and resultant containment leakage are consistent with the assumptions used in the safety analyses.
MILLSTONE - UNIT 3 B 3/4 6-2 Amendment No. JJ 115 0306 l
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