ML20058N745
| ML20058N745 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/14/1993 |
| From: | Robert Stransky Office of Nuclear Reactor Regulation |
| To: | Stratman R CENTERIOR ENERGY |
| References | |
| TAC-M83726, NUDOCS 9312220239 | |
| Download: ML20058N745 (9) | |
Text
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i December 14, 1993 i
l Docket No. 50-440 l
Mr. Robert A. Stratman l
Vice President Nuclear - Perry l
Centerior Service Company P.O. Box 97, S270 Perry, Ohio 44081 l
Dear Mr. Stratman:
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SUBJECT:
CHANGES TO THE BASES FOR TECHNICAL SPECIFICATION SECTIONS 3/4.I.5, 3/4.6.3, 3/4.8.I, 3/4.8.2 AND 3/4.8.3 (TAC NO. M83726)
By letter dated March 30, 1992, the Cleveland Electric Illuminating Company, et al. (the licensee) proposed several changes to the Bases for the Technical Specifications for the Perry Nuclear Power Plant, Unit No. 1.
These changes modify Bases sections 3/4.I.5, " Standby Liquid Control System," and 3/4.8.I, 3/4.8.2 and 3/4.8.3, "A.C. Sources, D.C. Sources and Onsite Power Distribution Systems," to more accurately reflect plant conditions as described in the Updated Final Safety Analysis Report, as well as to correct several typographical errors contained in the Bases.
The staff has reviewed these proposed changes and finds them to be acceptable.
Enclosed are copies of revised Bases pages B 3/4 1-4, B 3/4 6-5, B 3/4 8-1 and B 3/4 8-2.
Where appropriate, overleaf pages have been provided for your convenience. This letter completes staff activities related to TAC No.
M83726.
Sincerely, Original signed by Robert J. Stransky Robert J. Sttansky, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation l
Enclosure:
TS Bases Pages DISTRIBUTJ14 Docket File Local & NRC PDRs GHill (2)
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UNITED STATES t
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[..Ikl. l. j NUCLEAR REGULATORY COMMISSION
- 3 ij WASHINGTON, D.C. 20566-0001 f
December 14, 1993 d
Docket No. 50-440 1
i I
Mr. Robert A. Stratman Vice President Nuclear - Perry j
Centerior Service Company P.O. Box 97, S270 Perry, Ohio 44081
Dear Mr. Stratman:
SUBJECT:
CHANGES TO THE BASES FOR TECHNICAL SPECIFICATION SECTIONS 3/4.1.5, 1
2 i
3/4.6.3, 3/4.8.1, 3/4.8.2 AND 3/4.8.3 (TAC NO. M83726)
By letter dated March 30, 1992, the Cleveland Electric Illuminating Company,.
et al. (the licensee) proposed several changes to the Bases for the. Technical Specifications for the Perry Nuclear Power Plant, Unit No.1.
These changes modify Bases sections 3/4.1.5, " Standby Liquid Control System," and 3/4.8.1, i
3/4.8.2 and 3/4.8.3, "A.C. Sources, D.C. Sources and Onsite Power Distribution i
Systems," to more accurately reflect plant conditions as described in the Updated Final Safety Analysis Report, as well as to correct everal i
typographical errors contained in the Bases. The staff has reviewed these proposed changes and finds them to be acceptable.
3 i
Enclosed are copies of revised Bases pages B 3/4 1-4, B 3/4 6-5, B 3/4 8-1 and l
B 3/4 8-2.
Where appropriate, overleaf pages have been provided for your J
convenience. This letter completes staff activities related to TAC No.
)
M83726.
,Si~'e
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/ 'btrt J. Stransky, Project Manager j
k Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosure:
2 TS Bases Pages cc w/ enclosure:
See next page I
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_..,.-_.__-..,...._,,.,--,,,_.m...-.
Mr. Robert A. Stratman Perry Nuclear Power Plant Centerior Service Company Unit Nos. I and 2 cc:
Jay E. Silberg, Esq, Mr. James W. Harris, Director Shaw, Pittman, Potts & Trowbridge Division of Power Generation 2300 N Street, N.W.
Ohio Department of Industrial Relations Washington, D.C.
20037 P. O. Box 825 Columbus, Ohio 43216 Mary E. O'Reilly Centerior Energy Corporation The Honorable Lawrence Logan 300 Madison Avenue Mayor, Village of Perry Toledo, Ohio 43652 4203 Harper Street Perry, Ohio 44081 Resident Inspector's Office The Honorable Robert V. Orosz U.S. Nuclear Regulatory Commission Mayor, Village of North Perry Parmly at Center Road North Perry Village Hall Perry, Ohio 44081 4778 Lockwood Road North Perry Village, Ohio 44081 Regional Administrator, Region III U.S. Nuclear Regulatory Commission Attorney General 799 Roosevelt Road Department of Attorney General Glen Ellyn, Illinois 60137 30 East Broad Street Columbus, Ohio 43216 Lake County Prosecutor Lake County Administration Bldg.
Radiological Health Program j
105 Main Street Ohio Department of Health Painesville, Ohio 44077 Post Office Box 118 Columbus, Ohio 43266-0118 Ms. Sue Hiatt OCRE Interim Representative Ohio Environmental Protection Agency 8275 Munson DERR--Compliance Unit Memtor, Ohio 44060 ATTN: Zack A. Clayton P. O. Box 1049 Terry J. Lodge, Esq.
Columbus, Ohio 43266-0149 618 N. Michigan Street, Suite 105 Toledo, Ohio 43624 Mr. Thomas Haas, Chairman Perry Township Board of Trustees Ashtabula County Prosecutor 3750 Center Rd., Box 65 25 West Jefferson Street Perry, Ohio 44081 Jefferson, Ohio 44047 State of Ohio Mr. Kevin P. Donovan Public Utilities Commission Cleveland Electric Illuminating Company East Broad Street Perry Nuclear Power Plant Columbus, Ohio 43266-0573 P. O. Box 97, E-210 Perry, Ohio 44081 David P. Igyarto, General Manager Cleveland Electric Illuminating Company James R. Williams, Chief of Staff Perry Nuclear Power Plant Ohio Emergency Management Agency P. O. Box 97, SB306 2825 West Granville Road Perry, Ohio 44081 Worthington, Ohio 43085
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REACTIVITY CONTROL SYSTEMS BASES CONTROL RODS (Continued)
Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR.
The overtravel position feature provides the only positive mechanical means of determining that a rod is properly coupled and therefore this check must be performed prior to achieving criticality after completing CORE ALTERATIONS that could have affected the control rod coupling integrity. The subsequent check is performed as a backup to the initial demonstration.
In order to ensure that the control rod patterns can be followed and therefore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure.
The amount of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system. The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL ROD PROGRAM CONTROLS The rod withdrawal limiter system input power signal orginates from the first stage turbine pressure.
When operating with the steam bypass valves open, this signal indicates'a core power level which is less than the true core power.
Consequently, near the low power setpoint and high power setpoint of the rod pattern control system, the potential exists for non-conservative control rod withdrawals. Therefore, when operating at a sufficiently high power level, there is a small probability of violating fuel Safety Limits during a licensing basis rod withdrawal error transient.
To ensure that fuel Safety Limits are not violated, this specification prohibits control rod withdrawal when a biased power signal exists and core power exceeds the specified level.
Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to i
result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident. The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal. When THERMAL POWER is greater than 20% of RATED THERMAL POWER, there is no possible rod worth which, if dropped at the design rate of the velocity limiter, could result in a peak i
enthalpy of 280 cal /gm. Thus requiring the RPCS to be OPERABLE when THERMAL POWER is less than or equal to 20% of RATED THERMAL POWER provides adequate control.
PERRY - UNIT 1 B 3/4 1-3
i REACTIVITY CONTROL SYSTEMS BASES CONTROL ROD PROGRAM CONTROLS (Continued) will not be withdrawn or inserted.pervision to assure that out-of-sequence rods The RPCS provide automatic su The analysis of the rod drop accident is presented in Section 15.4 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.
The RPCS is also designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during higher power operation.
A dual channel system is provided that, above the low power setpoint, restricts the withdrawal distances of all control rods. This restriction is greatest at highest power levels.
3/4.1.5 STANDBY LIOUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern.
To meet this objective it is necessary to inject a quantity of boron which produces a concentration of 660 ppm in the reactor core. To allow for potential leakage and imperfect mixing this concentration is increased by 25%. The required concentration is achieved by having a minimum available guantity of 4260 gallons of sodium pentaborate solution; 13.8% by weight, containing a minimum of 5236 lbs. of sodium pentaborate. This quantity of solution is the net amount above the pump suction, thus allowing for the portion that cannot be injected. The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permissible sodium pentaborate solution volume range which adequately compensates for the positive reactivity effects due to temperature and xenon during shutdown. The temperature requirement for the sodium pentaborate solution is necessary to ensure that the sodium pentaborate remains in solution.
With redundant pumps and explosive injection valves and with a highly reliable control rod scram system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
Surveillance requirements are established on a frequency that assures a high reliability of the system. Once the solution is established, boron concentration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.
1.
C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NED0-10527, March 1972 2.
C. J. Paone, R. C. Stirn and R. M. Young, Supplement I to NED0-10527, July 1972 3.
J. M. Haun, C. J. Paone and R. C. Stirn, Addendum 2, " Exposed Cores,"
Supplement 2 to NED0-10527, January 1973 PERRY - UNIT 1 B 3/4 1-4 REVISED BY NRC LETTER DATED 12/14/93
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i CONTAINMENT SYSTEMS I
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BASES t
DEPRESSURIZATION SYSTEMS (Continued) l In addition to the limits on temperature of the suppression pool water, operating procedures define the action to be taken in the event a safety-relief
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valve inadvertently opens or sticks open. As a minimum this action shall l
include:
(1) use of all available means to close the valve, (2) initiate suppression pool water cooling, and (3) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety relief valve, where possible, to assure mixing and uniformity i
of energy insertion to the pool.
The containment spray system consists of two 100% capacity loops, each with three spray rings located at different elevations about the inside circumference of the containment. RHR pump A supplies one loop and RHR pump B supplies the other.
RHR pump C cannot supply the spray system. Dispersion of the flow of water is effected by 346 nozzles in loop A and 344 nozzles in' loop B, enhancing the condensation of water vapor in the containment volume and preventing overpressurization.
Heat rejection is through the RHR heat exchangers. The turbulence caused by the spray system aids in mixing the containment air volume to maintain a homogeneous mixture for H control.
2 The suppression pool cooling function is a mode of the RHR system and functions as part of the containment heat removal system. The purpose of the system is to ensure containment integrity following a LOCA by preventing excessive containment pressures and temperatures.
The suppression pool cooling mode is designed to limit the long term bulk temperature of the pool to 185*F considering all of the post-LOCA energy additions. -The suppression pool cooling trains, being an integral part of the RHR system, are redundant, safety-related component systems that are initiated following the recovery of the reactor vessel water level by ECCS flows from the RHR system. ~ Heat rejection to the i
emergency service water is accomplished in the RHR heat exchangers.
The suppression pool make-up system provides water from the upper contain-I ment pool to the suppression pool by gravity flow through two 100% capacity dump lines following a LOCA. The quantity of water provided is sufficient to account for all conceivable post-accident entrapment volumes, ensuring the long term energy sink capabilities of the suppression pool and maintaining the water coverage over the uppermost drywell vents.
During refueling, there will be administrative control to ensure the make-up dump valves will not be opened.
3/4.6,4 CONTAINMENT ISOLATION VAtVES i
I The OPERABillTY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or GDC 54 through 57 of Appendix A to 10 CFR 50. pressurization of the co Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA.
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PERRY - UNIT 1 B 3/4 6-5 REVISED BY NRC LETTER DATED 12/14/93
CONTAINMENT SYSTEMS BASES All required Containment Isolation Valves are listed in the PNPP Unit 1 Plant Data Book. The opening of normally locked or sealed closed containment isolation valves under administrative controls in accordance with footnote # includes the (1) stationing an operator, who is in constant following considerations:
corrnunicatien with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.
The above considerations do not apply to the nonnally locked closed (LC) Fire Protection system manual hose reel containment isolation valves IP54-F726 and
-F727 when opened as necessary to supply fire mains when handling irradiated fuel in the primary containment, during CORE ALTERATIONS and operations with a potential for draining the reactor vessel.
3/4.6.5 VACUUM RELIEF 3/4.6.5.1 CONTAINMENT VACUUM RELIEF AND 3/4.6.5.2 CONTAINMENT HUMIDITY CONTROL L
Vacuum breakers are provided on the containment to prevent an excessive vacuum from developing inside containment during an inadvertent or improper operation of the containment spray.
Four vacuum breakers and their associated isolation valves are provided. Any two vacuum breakers provide 100% vacuum relief.
The containment vacuum relief system is designed to prevent an excessive vacuum from being created inside the containment following inadvertent initia-tion of the containment spray system. By maintaining temperature / relative humidity within the limits for acceptable operation shown on Figure 3.6.5.2-1, the maximum containment vacuum created by actuation of both containment spray loops will be limited to approximately -0.7 psig.
3/4.6.5.3 DRYWELL VACUUM BREAKERS Drywell vacuum breakers are provided on the drywell to prevent drywell flooding due to differential pressure across the drywell and to equalize pressure between the drywell and containment.
Two drywell vacuum breakers and their associated isolation valves are provided. Any one vacuum breaker can provide full vacuum relief capability.
3/4.6.6 SECONDARY CONTAINMENT Secondary containment is designed to minimize any ground level release of radioactive material which may result from an accident. The Shield Building provides secondary containment during normal operation when the containment is sealed and in service. At other times, the containment may be open and, when required, secondary containment integrity is specified.
Establishing and maintaining a vacuum in the annulus with the annulus exhaust gas treatment system, along with the surveillance of the doors, hatches, and valves, is adequate to ensure that there are no violations of the integrity of the secondary containment.
The OPERABILITY of the annulus exhaust gas treatment systems ensures that sufficient iodine removal capability will be available in the event of a LOCA.
The reduction in containment iodine inventory reduces the resulting site PERRY - UNIT 1 B 3/4 6-6 Amendment No. 44
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ELECTRICAL POWER SYSTEMS BASES 3/4.8.1. 3/4.8.2 and 3/4.8.3 A.C. SOURCES. D.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS The OPERABILITY of the A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will bc available to supply the safety related equipment required for (1) the safe shutdown of the facility and (2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A" to 10 CFR 50.
The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least Division 1 or 2 of the onsite A.C. and D.C.
power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. or D.C. source. Division ~; supplies the high pressure core spray (HPCS) system only.
The A.C. and D. C. source allowable out-of-servi:e times are based on Regulatory Guide 1.93, " Availability of Electrical rower Sources," December 1974 as modified by plant specific analysis and diesel generator manufacturer recommendations. When diesel generator Division 1 or Division 2 is inoperable, J
there is an additional ACTION requirement to verify that all required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diesel generator Division i or Division 2 as a source of emergency power, are al;o OPERABLE. This requ'rement is intended to provide assurance that a loss of offsite power event will not result in a complete loss of safety function of critical systems during the period diesel generator Division 1 or l
Division 2 is inoperable. The term verify as used in this context means to administrative 1y check by examining logs or other information to determine if certain components are out-of-service for maintenance or other reasons.
It does not mean to perform the surveillance requirements needed to demonstrate the OPERABILITY of the component.
The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that (1) the facility can be maintained in the shutdown or refueling condition for extended time periods and (2) sufficient instrumentation and control capability 1s available for monitoring and maintaining the unit status.
The surveillance requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies,"
March 10, 1971, and Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants,"
Revision 1, August 1977 as modified by plant specific analyses and diesel generator manufacturer recommendations.
PERRY - UNIT I B 3/4 8-1 REVISED BY NRC LETTER DATED 12/14/93
l ELECTRICAL POWER SYSTEMS BASES 1
A.C. SOURCES. D.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS (Continued)
The surveillance requirements for demonstrating the OPERABILITY of the unit batteries meet the intent of the recommendations of Regulator Guide 1.129 l
" Maintenance Testing and Replacement of Large Lead Storage ua.. aries for Nuclear Power Plants", February 1978, and IEEE Std 450-1980, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations." Some surveillance intervals differ from those recommended in IEEE Std 450-1980, as identified in USAR Table 1.8-1.
Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on float charge, connection resistance values and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates and compares the battery capacity at that time with the rated capacity.
Table 4.8.2.1-1 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage and specific gravity. The limits for the designated pilot cells float voltage and specific gravity, greater than 2.13 volts and not more than.015 below the manufacturer's full charge specific gravity or battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity. The normal limits for each connected cell for float voltage and specific gravity,
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greater than 2.13 volts and not more than.020 below the manufacturer's full I
charge specific gravity with an average specific gravity of all the connected I
cells not more than.010 below the manufacturer's full charge specific gravity, ensures the OPERABILITY and capability of the battery.
Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8.2.1-1 is permitted for up to 7 days. During this 7 day period:
(1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity of all the cells, not more than.020 below the manufacturer's recommended full charge specific gravity ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual l
cell's specific gravity ensures that an individual cell's specific gravity will not be more than.040 below the manufacturer's full charge specific gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than 2.07 volts, ensures the battery's capability to perform its design function.
PERRY - UNIT 1 B 3/4 8-2 REVISED BY NRC LETTER DATED 12/14/93