ML20076A968
| ML20076A968 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/23/1979 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20076A975 | List: |
| References | |
| NUDOCS 7905070337 | |
| Download: ML20076A968 (20) | |
Text
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S, UNITED STATES
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- NUCLEAR REGULATORY COMMISSION g {..P *(). j WASHINGTON, D. C. 20665
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THE CONNECTICUT LIGHT AND POWER COMPANY, THE HARTFORD ELECTRIC LIGHT COMPANY, WESTERN MASSACHUSETTS ELECTRIC COMPANY, AND NORTHEAST NUCLEAR ENERGY COMPANY DOCKET NO. 50-336 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 l
AMENDMENT TO FACILITY OPERATING LICENSE
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Amendment No. 50 License No. OPR-65 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Connecticut Light and Power Company. The Hartford Electric Light Company, Western Massachusetts Electric Company and Northeast Nuclear Energy Company (the licensees) dated December 8,1977, and supplemental information dated December 3,1976, March 8 and 22 and June 9,1977, 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, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission'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 Commission's regulations and all applicable requirements have been satisfied.
790507033'i4
. 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. DPR-65 is hereby amended to read as follows:
(2 ) Technical Specifications The Technical Specifications contained in Appendices
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A and B, as revised through Amendment No.50, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the i
Technical Specifications.
3.
This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY C0tiMISSION O
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. [c Robert W. Reid, Chief 3
Operating Reactors Cranch M
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Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: March 23,1979 0
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ATTACHMENT TO LICENSE AMENDMENT NO. 50 FACILITY OPERATING LICENSE NO. DPR-65 DOCKET NO. 50-336 Replace the following pages of the Appendix "A" Technical Specifications The revised pages are identified by Amendment with the enclosed pages.
The numoer and contain vertical lines indicating the area of change.
corresponding overleaf pages are also provided to maintain document l
completeness.
Pages V
i 3/4 4-1 3/4 4-21a (added) l 3/4 4-21b (added) i, 3/4 5-7 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2a i
B 3/4 4-11 B 3/4 4-12 6-21 f
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s INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS PAGE SECTION 3/4.4.4 PRESSURIZER.................... 7.......;........... 3/4 4-4 4
3/4.4.5. STEAM GENERATORS.....................................
3/4 4-5 3/4.4.6 REACTOR COOLANT SYSTEM. LEAKAGE..l.y.'..................
3/4 4-8 i
Lea kag e Detection Sys tems........................'....
3/4 4-8 Re, actor Coolant System $ Leakage.......................
3/4 4-9 3/4.4.7 CHEMISTRY..............q.............................. 3/4 4-10 3/4.4.8 SPECIFIC ACTIVITY....................................
3/4 4-13 3/4.4.9 PRESSURE / TEMPERATURE LIMITS..........................
3/4 4-17 Reactor Coolant System.:.............................
3/4 4-17 Pressurizer..........................................
3/4 4 '
Overpressure Protection Sy tems......................
3/4 4-21a l
4 3/4.4.10 STRUCTUR5C INTEGRITY................................. 3/4 4-22' 3/4.4.11 cbRE BARREL M0VEMENT.................................
3/4'4-32 J
3/4.5 EMERGENC'/ SCORE COOLING S_YSTEMS (ECCS) 1 3/4.5.1 SAFETY INJECTION TANKS...............................
3/4 5-1
,'h avg 300'F.......................
3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T e
3/4.5.3 ECCS SUBSYSTEMS fi
< 3 0 0
- F......................'.
3 (' 4 5 - 7 l
avg l
3/4.5.4 \\ REFUELING WATER STORA6,C TANK.........................
3/4 5-8 s
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4 s.*
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l MILLSTONE - UNIT 2 V
'tw Amendment No. 50 k
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT..................................
3/4 6-1 Containment Integrity................................
3/4 6-1 Containment Leakage..................................
3/4 6-2 C o n ta i nme n t A i r L o c k s................................ 3/4 6-6 I n te rna l P re s su re.................................... 3/4 6-8 A i r Te mpe ra t u re...................................... 3/4 6-9 Containment Structural Integri ty..................... 3/4 6-10 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS.................
3/4 6-12 3
Containment Spray System.............................
3/4 6-12 4
Containment Ai r Reci rculation System................. 3/4 6-14 3/4.6.3 CONTAINMENT ISOLATION VALVES.........................
3/4 6-15 3/4.6.4 COMBUSTIBLE GAS CONTR0L..............................
3/4 6-20 Hydrogen Analyzers...................................
3/4 6-20 Electric Hydrogen Recombiners - W................... 3/4 6-21 Hydrogen Purge System................................
3/4 6-23 Post-Incident Recirculation Systems..................
3/4 6-24 3/4.6.5 SECONDARY CONTAINMENT................................
3/4 6-25 Enclosure Building Filtration System................. 3/4 6-25 lh Enclosure Building Integri y.........................
3/4 6-28 l
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MILLSTONE - UNIT 2 VI s.
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3/4.4 REACTOR COOLANT SYSTEM REACTORCOOLANTLOOPj, LIMITING CONDITION FOR OPERATION I
3.4.1 Four reactor coolant pumps shall be in operation.
i ApPLICABIL!iY: As noted below, but excluding MODE 6.*
l I
ACTION:
MODES 1 and 2:
With less than four reactor coolant pumps in operation, be in HOT STANOBY within 4 hqurs.
MODES 3, 4** and S**:
Operation may proceed provided at least one reactor coolant loop is in operation with an associated reactor coolant pump or shutdown cooling pump.# The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.1 The Flow Dependent Selector Switch shall be detemined to be in the 4 pump position within 15 minutes prior to making the reactor critic'al and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.
O
- SeeSpecialTestExce3 tion 3.lc.4.
- A reactor coolant pump shal.1 r.ct be : tarted with one er 'more of the RCS cold leg temperatures < 275'F unless 1) the pressurizer water volume is less than 600 cuEic feet or 2) the secondary (water temperature of each steam generator is less than 43'F 31*F when measured by a surface contact instrument) above the coolant temperature in the reactor vessel.
dAll reactor coolant pumps and shutdown cooling pumps may be de-energized for up to I hour, provided no operations are pemitted which could cause dilution of the reactor coolant system boron concentration.
M'LLSTONE - UNIT 2 3/4 4-1 Amendment No, 50
REACTOR COOLANT SYSTEM SAFETY YALVES - SHUTDOWN _
LIMITING CONDITION FOR OPERATION
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I A minimum of one pressurizer code safety valve shall be OPERABLE l
3.4.2 with a lift setting of 2500 PSIA + 1%.
APPLICABILITY: N0 DES 4 and 5, q
i ACTION:
l With no pressurizer code safety valve OPERABLE, imediately suspend all operations involving positive reactivity changes and place an OPERABLE i
shutdown cooling loop into operation.
O SURVEILLANCE REQUIREMENTS The pressurizer code safety valve shall be demonstrated OPERABLE 4.4.2 per Surveillance Reguirement 4.4.3.
O MILLSTONE - UNIT 2 3/44-2
PRESSURIZER 1
LIMITING CONDITION FOR OPERATION l
l 3.4.9.2 The pressurizer temperature shall be limited to:
a.
A maximum heatup of 100*F in any one hour period, b.
A maximum cooldown of 200*F in any one hour period, and c.
A maximum spray water temperature differential of 350*F.
APPLICABILITY: MODES 1, 2, 3, 4 and 5.
ACTION:
With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evalu3 tion to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 1
4.4.9.2 The pressurizer temperature and spray water temperature differen-tial shall be determined to be within the limits at ledt once per hour during system heatup or cooldown.
, MILLSTONE - UNIT 2 3/4 4-21 Amendment No. 45
F REACTOR COOLANT SYSTEM OVERPRESSURE PROT _ECTION SY_ STEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:
Two power operated relief valves (PORVs) with a lift setting a.
of 1 450 psig, or b.
A reactor coolant system vent of 1 1.3 square inches.
1 APPLICABILITY: When the temperature of one or more of the RCS cold legs is 1275'F, except when the reactor vessel head is removed.
ACTION:
With one PORV inoperable, either restore the inoperable PORY a.
to OPERABLE status within 7 days or depressurize and vent the i
RCS through a 1 1.3 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status, i
b.
With both PORVs inoperable, depressurize and vent the RCS through a > 1.3 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain tee RCS in a vented condition until both PORVs have been restored to OPERABLE status.
In the event either the PORVs or the RCS vent (s) are used to c.
mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.
d.
The provisions of Specification 3.0.4 are not applicable.
MILLSTONE - UNIT 2 3/4 4-21a Amendment No. 50 i
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REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS
a.
Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to entering a condition in which the FORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required OPERABLE.
b.
Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.
c.
Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.
d.
Testing in accordance with the inservice test requirements for ASME Category C valves pursuant to Subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code 1974 Edition, and Addenda through Summer 1975.
4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
- when the vent (s) is being used for overpressure protection.
l O
- Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
MILLSTONE - UNIT 2 3/4 4-21b Amendment No. 50
REACTOR COOLANT SYSTEM STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of components (except steam generator tubes) identified in Section 1.2.14 of the FSAR as Safety Class 1 com-ponents and of the steam generator secondary side circumferential shell welds shall be maintained at a level consistent with the acceptance criteria in Specification 4.4.10, APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
a.
With the structural integrity of any of the above components not conforming to the above requirements and T
> 200*F, either imediately isolate the affected compon8M or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, b.
With the structural integrity of any of the above components not conforming to the above requirements and the unit in COLD SHUTDOWN, restore the structural integrity of the affected component to within its limits prior to increasing the Reactor Coolant System temperature above the minimum temperature required by NDT considerations.
SURVEILLANCE REQUIREMENTS 4.4.10 The following inspection program shall be perfomed:
a.
Inservice Inspections The structural integrity of the Safety Class 1 components shall be demonstrated by verifying their acceptability per the requirements of Articles15-200 and 1S-500 of Section XI of the ASME Boiler and Pressure Vessel Code, dated July 1971, including the Sunner 1971 Addendum, as j
outlined by the inspection program shown in Table 4.4-4.-
The structural integrity of the steam generator secondary side circumferential shell welds shall be demonstrated by verifying their acceptability per the requirements of Tables ISC-261. ISC-251 l
and Section ISC-240 of Section XI of the ASME Boiler and Pressure Vessel Code, Winter 1972 Addendum, as outlined by the inspection program shown in Table 4.4-4.
i i!LLSTONE - UNIT 2 3/4 4-22
EMERGENCY CORE COOLING SYSTEMS ECCS SUBSYSTEMS - T.yg < 300*F LIMITING CONDITION FOR OPERATION 1
3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:
One# OPERABLE high-pressure safety injection pump, and l
a.
An OPERABLE flow path capable of taking suction from the refuel-b.
ing water storage tank on a safety injection actuation signal and automatically transferring suction to the containment sump 2
on a sump recirculation actuation signal.
APPLICABILITY: MODES 3* and 4.
ACTION:
With no ECCS subsystem OPERABLE, restore at least one ECCS a.
subsystem to OPERABLE status within one hour or be in COLD SHUTOOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
In the event the ECCS is actuated and injects water into the b.
Reactor Coolant System, a Special Report shall be prepared and i
submitted to the Consnission pursuant to Specification 6,9.2 within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to date, 4
h SURVEILLANCE REQUIREMENTS The ECCS subsystem shall be demonstrated OPERABLE per the applicable i 4.5.3.1 Surveillance Requirements of 4.5.2.
All high-pressure safety injection pumps, except*the above required 4.5.3.2 OPERABLE pump, shall be demonstrated inoperable at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is < 275'F by verifying that the motor circuit breakers have been removed from their electrical power supply circuits.
With pressurizer pressure < 1750 psia.
'A maximum of one high-pressure safety injection pump shall be OPERABLE whenever the temperature of one or more of the RCS cold legs is 5,275'F.
MILLSTONE - UNIT 2 3/4 5-7 Amendment No. #, 50 1
EMERGENCY CORE COOLING SYSTEMS REFUELING WATER STORAGE TANK
!JMITINGCONDITIONF0_ROPERATION 3.5.4 The refueling water storage tank shall be OPERABLE with:
a.
A minimum contained volume of 370,000 gallons of borated
- water, b.
A minimum boron concentration of 1720 ppm, A minimum water temperature of 50*F when in MODES 1 and 2, and c.
d.
A minimum water temperature of 35'F when in MODES 3 and 4.
APPL.ICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the refueling water storage tank ' inoperable, restore tank to OPERABLE status 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 COLD SHUTDOWN within the.next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:
a.
At least once per 7 days by:
1.
Verifying the water level in the tank, and 2.
Verifying the boron concentration of the water.
l b.
When in MODES 3 and 4, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature is > 35'F when the RWST ambient air I
temperature is < 35'F.
When in MODES 1 and 2, at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying c.
the RWST temperature is > 50*F when the RWST ambient air temperature is < 50'F.
MILLSTONE - UNIT 2 3/4 5-8
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.30 during all nonnal operations and anticipated transients.
STARTUP and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for the Power Level-High, Reactor Coolant Flow-Low, and Thermal Margin / Low Pressure trips have been reduced to their specified values.
Reducing these trip setpoints ensures that the ONBR will be maintained above 1.30 during j
three pump operation and that during two pump operation the core void fraction will be limited to ensure parallel channel flow stability within the core and thereby prevent premature DNB.
A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure consi-derations require pl. ant cooldown if component repairs and/or corrective actions cannot be made within the allowable out-of-service time.
The restrictions on starting a Reactor Coolant Pump during MODES 4 and 5 with one or more RCS cold legs < 275'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 43*F S
(31*F when measured by a surface contact instrumentl above the coolant
)
temperature in the reactor vessel.
3/4.4.2 and 3/4.4.3 SAFETY VAL'iES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 296,000 lbs Der hour of saturated steam at the valve ~
setpoint.
The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capa-bility and will prevent RCS overpressurization.
MILLSTONE - UNIT 2 B 3/4 4-1 Amendment No. 50
REACTOR COOLANT SYSTEM BASES During operation, all pressurizer code safety valves must be OPERASLE to prevent the RCS from being pressurized above its safety limit of 2750 The combined relief capacity of these valves is sufficient to psia.
limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reacto is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.
3/4.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure The steam bubble also protects the pressurizer surges during operation.
code safety valves and power operated relief valve against water relief.
The power operated relief valve and steam bubble function to relieve RCS Operation of the power oper'ated 4
pressure during all design transients.
relief valve in conjunction with a reactor trip on a Pressurizer--
Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS The program for inservice inspection of steam will be maintained.
generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1, in combination with a Supplementary Inservice Inspection Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the Program.
event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice condittor Inservice inspection of steam generator tubing that lead to corrosion.
also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found If the to result in negligible corrosion of the steam generator tubes.
secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking.
y MILLSTONE - UNIT 2 B 3/4 4-2 Anendment No. 27, 77, 50
REACTOR COOLANT SYSTEM BASES Stress corrosion cracking couli also be initiated from the primary side, if sufficiently large tute strains were introduced as a result of the primary and secondary effects of denting. The Supplementary Inspection Program assures that tubes that have developed excessive strains will be identified and removed from service, on a preventive basis, before cracking would develop, in accordance with the Tube Plugging Criteria described in Specification 4.4.5.2.
Futhermore, the potential causative factors for developing) excessive tube strain have been eliminated, or greatly reduced by (a the steam generator repairs that were implemented during the November 1977 Outage, (b) condenser integrity resulting from the condenser retubing implemented ~
in the May 1977 Outage, and (c) the phasing in of the Full Flow Condensate Polishing System during Cycle 2.
[
The extent of cracking during plant operation would be limited by
(_/
the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 0.5 GPM, per steam generator). Cracks having a primary-to-4 secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 0.5 gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this Itmit will require plant shutdown and an unscheduled inspection, during which the leaking tubes, and certain deformed tubes, will be located and plugged.
Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations. Plugging will be required for all tubes with imperfec-h tions exceeding the plugging limit of 40% of the tube nominal wall V
thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Mechanical tube defects caused by loose parts are unlikely, based on experimental data addressing loose parts effects. To prov'ide conclusive assurance of the validity of this statement, and to demonstrate that.
hypothesized degradation does not occur, suspect tubes are to be inspected during the " Scheduled Inspection" as defined in Specification 4.4.5.2.4.
Whenever the results of any steam generator tubing inservice inspection fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption of plant operation. Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratcry examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.
MILLSTONE - UNIT 2 B 3/4 4-2a Amendment No. 22, 37, 50
1 REACTOR COOLANT SYSTEM BASES i
for piping, pumps and valves. Below this temperature, the system pressure i
must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure c:mpliance with the requirements of Appendix H to 10 CFR Part 50.
The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fati-O gue analysis performed in accordance with the ASME Code requirements.
The OPERABILITY of two PORVs or an RCS vent opening of greater than 1.3 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are 1 275"F. Either PORV has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator 1 43*F (31*F when measured by a surface contact instrument) above the coolant temperature in the reactor vessel or (2) the start of a HPSI pump and its injection into a water solid RCS.
3/4.4.10 STRUCTURAL INTEGRITY The required inspection programs for the Reactor Coolant System components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.
To the extent practicable, the inspection program for the Reactor Coolant System components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code " Inservice Inspection of Nuclear Reactor Coolant Systems" dated July 1, 1971.
All areas scheduled for volumetric examination have been pre-service examined using equipment, techniques and procedures anticipated for use during post-operation examinations. To assure that consideration is given to the use of new or improved inspection equipment, techniques and procedures, the Inservice Inspection Program will be periodically reviewed on a 5 year basis.
The use of conventional nondestructive, direct visual'and remote visual test techniques can be applied to the inspection of most reactor coolant loop components except the reactor vessel. The reactor vessel requires special consideration because of the radiation levels.
g MILLSTONE - UNIT 2 B 3/4 4-11 Amendment No. 50
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REACTOR COOLANT SYSTEM BASES The techniques anticipated for inservice inspection include visual inspections, ultrasonic, radiographic, magnetic particle and dye pene-trant testing of selected parts.
The nondestructive testing for repairs on components greater than 4 inches diameter gives a high degree of confidence in the integrity of.
the system, and will detect any significant defects in and near the new welds. Repairs on components 4 inches in diameter or smaller receive a surface examination which assures a similar standard of integrity. In each case, the leak test will ensure leak tightness during nomal operation.
For nomal opening and reclosing, the structural integrity of the Reactor Coolant System is unchanged. Therefore, satisfactory performance of a system leak test at 2250 psia following each opening and subsequent reclosing is acceptable demonstration of the system's structural integrity.
These leak tests will be conducted within the pressure-temperature limitations for Inservice Leak and Hydrostatic Testing of Specification 3.4.9.1 and Figure 3.4-2.
Inspection of the pipe hangers and supports provides assurance that 1
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these devices are operated within permissible travel and/or loading limits.
3/4.4.11 CORE BARREL MOVEMENT This specification is provided to ensure early detection of excessive core barrel movement if it should occur. Neutron noise levels are used to continually monitor core support barrel (CSB) motion. Change in motion is manifested as changes in the four excore neutron detector signals. Base-line core barrel movement Alert Levels and Action Levels at nominal THERMAL POWER levels of 20%, 50%, 80% and 100% of RATED THERMAL POWER will be detemined during the reactor startup test program, Data from these detectors is to be reduced in two foms. RMS values are computed from the Amplitude Probability Density (APD) of the signal amplitude. These RMS magnitudes include variations due both to various neutronic effects and internals motion. Consequently, these signals alone can only provide a gross measure of CSB motion. A more accurate assessment of CSB motion is obtained from the Auto and Cross Power Spectral Densities These data (PSD, XPSD), phase (d) and coherence (COH) of these signals.
result from a Spectral Analysis (SA) of the excore detector signals.
A modification to the required monitoring program may be justified by an analysis of the data obtained and by an examination of the affected parts during the Dlant shutdown at the end of the first fuel cycle.
MILLSTONE - UNIT 2 8 3/4 4-12 Amendment No. U, D, 50
i ADMINISTRATIVE CONTROLS THIRTY-DAY WRITTEN REPORTS (Continued) completed copy of a licensee event report form.
Information provided on the licensee event report form shall be supplemented, as needed, by addi-tional narrative material to provide complete explanation of the circum-stances surrounding the event.
Reactor protection system or engineered safety features instrument a.
settings which are found to be less conservative than those estab-lished by the technical specifications but which do not prevent the fulfillment of the functional requirements of affected systems, b.
Conditions leading to operation in a degraded mode permitted by a limiting condition for operation or plant shutdown required by a limiting condition for operation.
Observed inadequacies ia the implementation of administrative or c.
V procedural concrols which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety features systems.
d.
Abnormal degradation of systen.3 other than those specified in 6.9.1.8.c. above, designed to contain radioactive material resulting from the fission process.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of t
Inspection and Enforcement Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification:
a.
Inoperable Seismic Monitoring Instrumentation,
- ks Specification 3.3.3.3.
b.
Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4 Safety Class 1 Inservice Inspection Program Review, c.
Specification 4.4.10.1.
l d.
Core Barrel Movement, Specifications 3.4.11 and 4.4.11.
i e.
ECCS Actuation, Specificatinns 3.5.2 and 3.5.3.
f.
Fire Detection Instrumentation, Specification 3.3.3.7.
Fire Suppression Systems, Specifications 3.7.9.1 and 3.7.9.2.
h.
RCS Overp.ressure Mitigation, 5pecification 3.4.9.3.
l g.
MILLSTONE - UNIT 2 6-21 Amendment No. 36, 50
r ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:
Records and logs of facility operation covering time interval at a.
each power level.
b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c.
All REPORTABLE OCCURRENCES submitted to the Commission.
d.
Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
e.
Records of reactor tests and experiments.
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f.
Records of changes made to operating procedures.
l g.
Records of radioactive shipments.
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h.
Records of sealed source leak tests and results.
1 1.
Records of annual physical inventory of all sealed source material of record.
6.10.2 The following records shall be retained for the duration of the facility operating license:
a.
Records and drawing changes reflecting facility design modifica-tions made to systems and equipment described in the Final Safety Analysis Report.
f b.
Records of new an,i irradiated fuel inventory, fuel transfers and assembly burnup histories.
c.
Records of facility radiation and contamination surveys.
d.
Records of radiation exposure for all individuals entering radiation control areas.
e.
Records of gaseous and liquid radioactive material released to the environs, f.
Recnrds of transient or operational cycles for those facility components designed for a limited number of transients or cycles.
MILLSTONE - UNIT 2 6-22 Amendment No. ),
36
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