ML19209A158
| ML19209A158 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 09/10/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19209A157 | List: |
| References | |
| NUDOCS 7910030029 | |
| Download: ML19209A158 (15) | |
Text
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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CONSUMERS POWER COMPANY DOCKET NO. 50-255 PALISADES PLANT AMENDMENT TO PROVISIONAL OPERATING LTCENSE Amendment No. 51 License No. DPR-20 1.
T.'.e Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for emendment by Consumers Power Company (the licensee) dated January 3,1978, as supported by information trantmitted by letters dated March 8,1977, June 24,1977, and November 28, 1977, complies with the standards and require-ments 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 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 Commission's regalations and all applicable requirements have been satisfied.
1084 217 7 91003 00M
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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 3.8 of Provisional Operating License No.
DPR-20 is hereby amended to read as follows:
B.
Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 51, are heret. incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION N,)
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i wn Dennis L. Ziema'nt, Chief Operating Reactors Branch #2 Division of Operating Reactors
Attachment:
Changes to the Technical Specifications Date of Issuance: September 10, 1979 1084 218
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ATTACHMENT TO LICENSE AMENDMENT NO. 51 PROVISIONAL OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Revise Appendix A by removing the pages described below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.
REMOVE INSERT i
2 3-la 3-la 3-2 3-2*
3-3 3-3 3-25a 3-29a 3-29a*
3-30 3-30 3-33 3-33 4-1 4-1 4-2 4-2 4-2a*
4-39 4-39
- There were no changes made to the provisions contained on these pages.
The Technical Specifications have merely been repositioned.
1084 219
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PALISADES PIANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Section Description Page No 1.0 DEFINITIONS 1-1 1.1 Reactor Operating Conditions 1-1 1.2 Protective Systems 1-2 13 Instrunentation Surveillance 1-3 1.h Miscellaneous Definitions 1-3 2.0 SAFET! LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 Safety Limits - Reactor Core 2-1 2.2 Safety Limits - Primary Coolant System Pressure 2-3 2.3 Limiting Safety System Settings - Reactor Protective System 2h 30 LIMITING CONDITIONS FOR OPERATION 3-1 3.1 Pri=ary Coolant System 3-1 3.1.1 operable Co=ponents 3-1 3.1.2 Heatup and Cooldown Rate 3k 3.1.3 Mini =us Conditions for criticality 3-15 3.1.4 Maximum Primary coolant Radioactivity 3-17 3.1.5 Pri=ary Coolant System Leakage Li=its 3-20 3.1.6 Mmm Pri=ary Coolant Oxygen and Halogens Concentrations 3-23 3.1.7 Pri=ary and Secondary Safety Valves
'-25 3.1.8 Overpressure Protection Systc=s 3-25a 3-26 3.2 Che=1 cal and volune Control System 3-29 33 Energency Core Cooling System 3-3h 3.h containment Cooling 3-38 3.5 Stea= and Feed *4ater Syste=s 3 ho 36 Containment Syste=
3-41 37 Electrical Syste=s 3 h6 3.6 Refueling operations 3-50 3.9 Effluent Release 1084 220 i
Amendment No. Jf, 51
31 PRD!ARY C00LANT SYSTEM (Contd) 3 1.1 coerable Ceepenents (Contd)
(2) Hydrostatic tests shall be conducted in accordance with applicable paragraphs of Section XI ASME Boiler & Pressure Vessel Code (1974).
Such tests shall be conducted with sufficient pressure on the seconda:7 side of the steam generators to restrict pri=ary to secondary pressure differential to a maximum of 1380 psi.
Maximum hydrostatic test pressure shall not exceed 1.1 Po plus 50 psi where Po is nominal operating pressure.
(3) Primary side leak tests shall be conducted at nor=al operating The te=perature sha11 be consistent with applicable pressure.
fracture toughness criteria for ferritic =aterials and shall be selected such that the differential pressure across the steam generator tubes is not greater than 1380 psi.
(b) Maxi==1 secondary hydrostatic test pressure shall not exceed 1250 psia. A minimum temperature of 100 F is required. Only ten cycles are per=itted.
(5) Maxi =us secondary leak test pressure shall not exceed 1000 psia.
A =inimum te=perature of 100 F is required.
(6)
~n performin6 the tests identified in 3 1.1.e(h) and 3 1.1.er(5),
abcve, the secondary pressure shall not exceed the priz::ary pressure by = ore ths: 350 psi.
Nominal pr W 7 syste= operating pressure shall not exceed 2100 psia.
f.
The reactor inlet te=perature (indicated) shall not exceed the value g.
given by the following equation at steady state 100". power operation:
1 538.0 + 0.03938 (P-2060) + 0.0000h8k3 (P-2060)2 + 1 T
Where: T
= reactor inlet te=perature in F.
P = nominal operatin6 pressure in psia.
W = total recirculating mass flow in 10 lb/h corrected to the operating te=perature conditions.
Note: This equatien is shown in Figure 3-0 for a variety of = ass flev rates.
h.
A reactor coolant ptap shall not be started with one or more of the PCS cold leg temperatures 1 250*F unless 1) the pressurizer water volume is less than 700 cubic feet or 2) the secondary water temperature of each steam generator is less than 70*F above each of the PCS cold leg temperatures.
3-la Amendment g, 51
3.1 PRIMARY COOLANT SYSTEM (Contd)
Ba sis When primary c.colant boron concentrat'.on is being changed, the process must be uniform throughout the primary coolant system volume to prevent stratification of primary coolant at lower boron concentration which could result in a reactivity insertion.
Sufficient mixing of the pri-mary coolant is assured if one shutdown cooling or one primary coolant pump is in operation.( } The shutdown cooling pump will circulate the primary system volume in less than 60 minutes when operated at rated capacity.
The pressurtzer volume is relatively inactive, therefore will tend to have a boron concentration higher than rest of the pri-may coolant system during a dilution operation.
Administrative pro-cedures will provide for use of pressurizer sprays to maintain a nominal spread between the baron concentration in the pressurizer and the pri-mary system during the addition of boron. (
Both steam generators are required to be operable whenever the tempera-ture of the primary coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor.
Calculations have been performed to demonstrate that a pressure dif ferential of 1380 psi can be withstood by a tube uniformily thinned to 36% of its original nominal wall thickness (64% degrad-ation), while maintaining:
(1) A factor of safety of three between the acf.ual pressure dif-ferential and the pressure differential required to cause bursting.
(2)
Stresses within the yield stress for Inconel 600 at operating temperature.
(3) Acceptable stresses during accident conditions.
The maximum transient steam generntor dif ferential pressure is expected to occur during the loss of load accident.
The loss of load accident inititated from hot full power operating conditions and assuming a high pressurizer trip of 2277 psia is analyzed in Reference 3.
Results of this analysis indicate that the maximum steam ;enerator differential pressure is less than 1530 psi for the worst case assuming pressurizer spray and relief valves inoperable and assuming steam dump and turbine bypass operable.
The 1530 psi limit on transient pressure dif ferential is approximately 11% greater t that 3-2 Amendment No. 20, 51
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PRIMARY COOLANT SYSTEM (Contd) allowed during normal operation, so that substantial safety inargin exists between this pressure differential and the pressure differestial required for tube rupture.
Se;ondary side hydrostatic and leak testing requirements are consistent with AS:E BPV Section XI (19711 The differential maintains stresses in the steam generator tube valls within code allovable stresses.
The minimum te=perature of 100*F for pressurizing the steam generator second-ary side is set by the NDTT of the manvny cover of + ho F.
The transient analyses were perfor=ed assuming a vessel flov at hot zero power (532 F) of 126.9 x 10 lb/h minus 6% to account for flow measurement uncertainty and core flev bypass.(3) A steady state DNB analysis was also perfor=ed (assu=ing 115% overpower, 50 psi for pressure uncertainty, 3% for
- f. tov measurement uncertainty, and 3% for core flow bypass) in a parametric fashion to determine the core inlet te=perature as a function of pressure and flov for which the mini =us DN3R at 115% overpover is equal to 1.30.
The result of this steady state DN3 analysis was the following equation for limiting reactor inlet te=perature:
1 541.0 + 0.03938 (P-2060) + 0.0000k8h3 (P-2060)2 + 1.03h2 (W-120.2)
T A te=perature measurement uncertainty of 3 F was subtracted frem this limit in arriving at the LCO given in Section 3.1.1.g.
The nominal full power inlet te=perature is 2*F less than the value given in Section 3 1.1.s to allow for drift within the te=perature control band. Thus, a total urcer-tainty of 5 F is applied to the limiting reactor inlet te=perature equation.
The limits of validity of this equation are:
1850 < Pressure < 2250 Psia 110.0 x 10 1 Vessel Flow < 130 x 10 Lb/h The restrictions on starting a Reactor Coolant Pump with one or more I'CS cold legs < 250*F are provided to prevent PCS pressure transients, caused by energy additions from the secondary system, which would exceed the limits of Appendix G to 10 CFR Part 50.
The PCS 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 rentricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 70*F above each of the PCS cold leg temperatures.(5)
References (1) FSAR, Sections 6.1.2.2 and 1L.3.2.
(2) FSAR, SEction L.3.7.
1084 (3) ri-NF-77-lS.
G)
Ei-NF-77-22.
" Palisades Plant Overpressurization Analysis," June, 1977, and " Palisades (5)
Plant Primary Coolant System Overpressurization Subsystem Description,"
October, 1977.
3-3 Amendment No. 71,51
3.1.8 Overpressure Protection Systems Specifications a.
When the temperature of one or more of the primary coolant system cold legs is < 250*F, ;wo power operated relief valves (PORVs) with a lif t setting of 5,400 psia, or a reactor coolant system vent of 3,1.3 square inches shall be operabic except as specified below:
(1) With one PORV innperable, either restore the inoperable PORV to operable status within 7 days or depressurize and vent the PCS through a > 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 PCS in a vented condition until both PORVs have been restored to operable status.
(2) With both PORVs inoperable, depressurize and vent the PCS thrcugh a 3,1.3 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 the PCS in a vented condition until both PORVs have been restored to operable status.
b.
In the event either the PORVs or the PCS vent (s) are used to mitigate a PCS pressure transient, a Special Report shall be prepared and submitted to the Commission within 30 days.
The repcrt shall describe the circumstances initiating the transient, the ef fect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.
Basis The OPERABILITY of two PORVs or an PCS vent opening of greater than 1.4 square inches ensures that the PCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or mcce of the PCS cold legs are 5,250*F.
Either PORV has adequate relieving capability to protect the PCS from overpressurization when the transient is limited to either (1) the start of an idle PCP with the secondary water temperature of the steam generator 5,70aF above the PCS cold leg temperatures (2) the start of a HPSI pump and its injection into a water solid PCS. (
or References
~
(1) " Palisades Plant Overpressurization Analysis," June, 1977, and " Palisades Plant Primary Coolant System Overpressurization Subsystem Description,"
October, 1977.
3-25a Amendment No. 51 1084 224
P00R 01 E l 3.3 DERGENCY CORE COOLING SYSTEM (Contd) 332 During power operstion, the require =ents of 3 31 c:ar be codified to allov one of the following conditions to be true at any one time. If the system ir not restored to meet the require =ents of 3 31 within the time period spec 1'fied below, the reactor shall be placed in a hot shutdovn condition sithin 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
If the require =ents of 3 3 1 are not met within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the reactor shall be placed in a cold shutdown ccndition within 2'. hours.
One safety injection tank may be inoperable for a period of no a.
more than one hour.
b.
One lov-pressure safety injection pu=p csy be inopemble pro-vided,the pu=p 'Is restored to operable status "vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The other low-pressure safety injection pump shall be tested
~
to demonstrate operability prior to initiating repair of the inoperable pump.
One high-pressure safety injection pump may be inoperable pro-c.
vided the pump is restored tg operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
The other high-pressure safety injection pump shall be tested to demonstrate 6perability prior to' initiating repair of the inoperable pump.
d.
One shutdown heat exchanger and one component cooling water heat exchanger may be inoperable for a period of no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Any valves, interlocks or piping directly associated with one e.
of the above components and required to function during acci-dent conditions shall be deemed to be part of that component and shall meet the same requirements as listed for that component.
f.
Any valve, interlock or pipe associated with the safety injec-tion and shutdown cooling system and which is not covered under 3 3 2e above but, which is required to function during accident conditions, may be inoperable for a period of no more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Prior to initiating Icpairs, all valves and interlocks in the system that provide the duplicate function shall be tested to demonstrate operability, jgj Amendment No. y/I, 51
3.3 EMERGENCY CORE COOLING SYSTEM (Contd) g. A maximum of one high-pressure safety injection pump shall be OPERABLE whenever the temperature of one or more of the PCS cold legs is < 250*F. Basis The normal procedure for starting the reactor is, first, to heat the primary coolant to near operating temperature by running the primary coolant pumps. The reactor is then made critical by withdrawing control rods and diluting boron in the primary coolant.( With this mode of start-up, the enerdy stored *n the primary coolant dur-ing the approach to criticality is subetantially equal to that during power operation and, therefore, e.11 engineered safety features and auxiliary cooling systems are required to be fully operable. Ih ing low-temperature physics tests, there is a negligible amount of stored energy in t,he primary coolant; therefore, an accident comparable in 1084 226 3-30 Amendment No. 51
P00R~0RIGINR 33 susrumac1 coaz cootrso srsTr< (contd) that 25% of their conbined disebarge rate is lost from the prinary coolant system out the break. The transient bot spot fuel clad temperatures for the break sizes considered are shovu on FSAR Figures 14.17 9 to 14.17 13 These de::custrate that the ~4- "uel clad tenperatures that could occur over the break sice spec-trum are well below the nelting tenperature of zirconiu= (3300 F). Malfunction of the Lov Pressure Safety injection Flow control valve could defeat the Lov Pressure Injection feature of the ECCS; there-fore, it is disabled in the 'epen', mode (by isolating the air supply) during plant operation. This action assures that it vill not block flow during Safety I=jection. The inadvertent closing of any one of the Safety Injection bottle isolation valves in conjunction with a LOCA has not been analyced. To provide assurance that this vill not occur, these valves ate electrically locked open by a key switch in the control rec =. In addition, prior to critical the valves are checked open, and then the h80 volt breakers e.t MCC 9 are opened. Thus, a failure of a breaker and a svitch are required for any of the valves to close. The limitation for a maximum of one high pressure safety injection pump to be operable, and the Surveillance Requirement to verify all high pressure safety injection pumps except the required operable pump to be inoperable below 250*F, provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. References (1) FSAR, Section 9.10. 3. (2) FSAR, Section 6.l. 3-33 Amendment No. [, 51
I e P00R~0RIGl01 k.0 SURVEILLANCE REQUIRDENTS k. 0.1 Surveillance require =ents shall be applicable during the reactor operating conditions associated with individual Limiting Conditions for Operation unless otherwise stated in an individual surveillance require =ent. L.O.2 Unless otherwise specified, each surveillance requirement shall be perfor=ed within the specified ti=e interval with: a. A =axi=u= allovable extension not to exceed 25% of the surveil-lance interval, and b. A total maximu co=bined interval time for any three consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval. k.1 INSTRUMD*f ATION AND CON'"ROL Atelicability Applies to the reactor protective syste= and ether critical instru=enta-tion and controls. Objective To specify the minimu frequency and type of surveillance to be appited to critical plant instru=entr. tion and controls. Specifications Calibration, testing, and checking of instrument channels, "eactor pro-tective syste= and engineered safeguards syste= logic channels and =iscellaneous instrument syste=s and controls shall be perfor=ed as specified in 4.1.1 and in Tables 4.1.1 to 4.1.3. 4.1.1 Overpressure Protection Systems a. Each PORV shall be demonstrated operable by: 1. 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 PORV is required operable and at least once per 31 days thereafter when the PORV is required operable. 2. Perfc mance of a channel calibration on the PORV actuation channel at least once per 18 months. 3. Verifying the PORV isolation valve is open at least once per 72 hours when the PORV is being used for overpressure protection. 4. Testing in accordance with the inservice inspection requirements for ASME Section XI, Section IWV Category C valves. 1084 228 4-1 Amendment No. Jd, 51
b. The PCS vent (s) shall be verified to be open at least once per 12 hours when the vent (s) is being used for overpressure protection except when tha 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. Basis Failures such as blown instrument fuses, defective indicators, and f aulted amplifiers which result in " upscale" or "downscale" indication can be easily recognized by simple observation of the functioning of an instrument or system. Furthermore, such failures are, in many cases, revealed by alarm or annunciator action and a check supplements this type of built-in surveil-lance. Based on experience in operation of both conventional and nuclear plant systcos when the plant is in operation, a checking frequency of once-per-shift is deemed adequate for reactor and steam system instrumentation. Calibrations are performed to insure the presentation and acquisition of accurate information. The power range safety channels are calibrated daily against a heat balance standard to account for errors induced by changing rod patterns and core physics parameters. Other channels are subject only to the "drif t" errors induced within the instrumentation itself and, consequently, can tolerate longer intervals between calibration. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recali-bration is performed at each refueling shutdown interval. Substantial calibration shifts within a channel (essentially a channel failure) will be revealed during routine checking and testing procedures. Thus, minimum calibration frequencies of one-per-day for the power range safety channels, and once each refueling shutdown for the process system channels, are considered adequate. The minimum testing f requency for tho e instrument channels connected to the reactor protective system is based on an estimated average unsafe failure rate of 1.14 x 10~ failure / hour per channel. This estimation is based on limited operating experience at conventional and nuclear plants. An " unsafe failure" is defined as one which negates channel operability and which, due to its nature, is revealed only when the channel is tested or attempts to respond to a bona fide signal. 1084 229 4-2 Amendment No. d, 51
For the specified one-month test interval, the average unprotected time is 360 hours in case of a failure occurring between test intervals, thus the probability of failure of one channel between test intervals is -5 360 x 1.14 x 10 or 4.1 x 10~ Since two channels must fail in order to negate the safety function, the probability of simultaneous failure -8 of two-out-of-four channels is (4.1 x 10- ) = 6.9 x 10 This repre-sents the fraction of time in which each four-channel system would have one operable and three inoperable channels and equals 6.9 x 10~ x 8769 hours per year, or 2.16 seconds / year. These estimates are conservative and may be considered upper limits. Testing intervals will be adjusted as appropriate based on the accumu-lation of specific operating history. The testing frequency of the process instrumentation is considered adequate (based on experience at other conventional and nuclear plants on Consumers Power Company's system) to maintain the status of the instruments so as to assure safe operation. As the reactor protec-tion system is not required when the plant is in a refueling shutdown cond it ion, routine testing is not required. Those instruments which are similar to the reactor protective system instruments are tested at a similar frequency and on the same basis. 1084 230 4-2a Amendment No. 51
4.6 SAFETY INJECTION AND CONTJINMENT SPRAY SYSTD4S TESTS Applicability Applies to the safety injection system, the containment spray system, chemical injection system and the containment cooling system tests. Objective To verify that the subject systems will respond promptly and perform their intended functions, if required. Specifications 4.6.1 Safety Injection System System tests shall be performed at each reactor refueling interval. a. A test safety injection signal vill be applied to initiate opera-tion of the system. The safety injection and shutdown cooling system pump motors may be de-energized for this test, b. The system test vill be considered satisfactory if control board indication and visual observations indicate that all camponents have received the safety injection signal in the proper sequence and timing (ie, the appropriate pump breakers shall have opened and closed, and all valves shall have completed their travel). All high pressure safety injection pumps except those otherwise c. required to be operable shall be demonstrated inoperable at least once per 12 hours whenever the temperature of one or more of the PCS cold legs is < 250 F by verifying that the control system fuses and their fuse holders for the HPSI pumps (P66A, P66B and P66C) have been removed from the circuit. 4.6.2 Containment Spray System System tests shall be performed at each reactor refueling interval. a. The test shall be performed with the isolation valves in the spray supply lines at the containment blocked closed. Operation of the system is initiated by tripping the normal actuation instrumenta-tion. b. At least every five years the spray nozzles shall be verified to be open, The test will be considered satisfactory if visual observations c. indicate all components have operated satisfactorily. 4.6.3 Pumps The safety injection pumps, shutdown cooling pumps, and contain-a. ment spray pumps shall be started at intervals not to exceed three months. Alternate manual starting between control room con-sole and the C-33 panel shall be practiced in the test program. 1084 231 ( Amendment No.51 h-39}}