ML20081K812
| ML20081K812 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/28/1983 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20081K795 | List: |
| References | |
| NUDOCS 8311100272 | |
| Download: ML20081K812 (18) | |
Text
{{#Wiki_filter:- _ _ _ _ _ - _ O GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION (DOCKET NO. 50-219) PROVISIONAL OPERATING LICENSE DPR-16 Applicant hereby requests the commission to change Appendix A of the above-captioned license as follows: 1.- Sections to be changed: a..Section 2.1
- b. Section 2.3 -
- c. Section 3.2
- d. Section 3.4
- e. Section 3.10
- 2. Extent of changes:
- a. Section 2.1 Safety Limit - Fuel Cladding Integrity Supplement No.- 1 to TSCR No. 96 changes the water level safety limit back to 4'8" from the previous limit of 10".
- b. Section 2.3 Limiting Safety System Settings.
Supplement No. 1 to TSCR No. 96 incorporates a new scram setting for Recirculation Flow at 117% of rated flow.
- c. Section 3.2 Reactivity Control Supplement No. 1 to TSCR No. 96 revises the control rod withdrawal sequences and establishes the maximum in sequence rod worth to be 1.0% AK.
- d. Section 3.4 Emergency Cooling Supplement No. 1 to TSCR No. 96 changes the water level limit back to 4'8" from the previous revision which had set a limit of 10".
- e. Section 3.10 Core Limits Supplement No.1 to TSCR No. 96 changes the 11CPR Limits to 1.4 from a previous revised limit of 1.3 and furnishes the maximum allowable average planar LHGR curves for Five Loop operation and for Four Loop operation (Figures 3.10-4 and 3.10-5, respectively).
- 3. Change requested:
The requested change is shown on the attached Technical Specification pages 2.1-1, 2.1-4, 2.3-3, 2.3-8, 3 2-2, 3.2-5, 3.2-6, 3.2-9, 3.4-2, 3.4-4, 3.4-6, 3.4-7, 3.4-8, 3.10-3, 3.10-12, and 3.10-13. 8311100272 831028 PDR ADOCK 05000219 p PDR _. _ _ _ _ _ - - _ _ )
4
- 4. Discussion:
Supplement No. 1 to TSCR No. 96. incorporates Amendments 5 and 6 to the General Electric NEDO 24195 Oyster Creek Reload Fuel Application into the final Cycle 10 Reload Core Design. This~ change request supplement also changes the water level safety, limit back to the original setting of 4'8" from the previously submitted setting of 10".. General Electric NEDO 24195 shall. comprise Amendment No. 80 to the Facility Description and Safety Analysis Report. t 4
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o g_ - 2.1-1 SECTION 2 1 SAFETY' LIMITS,AND LIMITING SAFETY SYSTEM SETTINGS 2.1 -SAFETY LIMIT - F UEL CLADDING INTEGRITY Applicability: Applies to the interrelated variables associated with fuel thermal behavior. Objective: To establish limits on the important thermal-hydraulic variables to assure the integrity of the fuel cladding. Specifications: A. When the reactor pressure is greater than or equal to 800 psia and the core flow is greater than or equal to 10% of rated, the existence of a minimum critical i power ratio (MCPR) less than 1.07 shall constitute violation of the fuel cladding integrity safety limit. B. When the reactor pressure is less than 800 psia or the core flow is less than 10% of rated, the core thermal power shall not exceed 25% of rated thermal power. ,~ C. In the event that reactor parameters exceed the - limiting safety. system settings in specification 2.3 1 and a reactor scram is not initiated by the associated i protective instrumentation, the reactor shall be . brought to, and remain in, the cold' shutdown condition until an analysis is performed to determine whether the safety limit established in specification 2.1.A and 2.1.B was exceeded. D. During all modes of reactor operation with irradiated
- fuel in the reactor vessel, the water level shall not be less-than-4'8" above the top of active fuel.
E. During all modes of operation except when the reactor head is off and the reactor is flooded to a level above the main steam nozzles, at least two [2] recirculation loop suction valves and their associated discharge valves will be in the full open position. k e 4 -e. - --e.s..rw+ y- ,.__y -ww_,_ ..e..,# , , . ...,,,m- ,y ,,, . , . . . .,o,,m, . ,,-,y-,,,.,,-,-+ .-~+,,e ,,,.,..w ,%--- 9..,,--,,,, g- .- - .,-- ..-
2.1-4 within the actiie fuel region and extends up through the moisture separators. For the purpose of this specification water level is defined to include mix-ture level during power operations. The lowest point at which the water level can present-ly be monitored is 4'8" below the-top of active fuel. Although the lowest reactor water level limit which ensures adequate core cooling is the top of the active fuel, the safety limit has been conservatively established at 4'8" above the top of active fuel. Specification 2.1.E assures that an adequate flow path exists from the annular space, between the pressure vessel wall and the core shroud, to the core region. This provides for good communication between these areas, thus assuring that reactor water level instru-ment readings are indicative of the water level in the core region. REFERENCES (1) NED0-24195, General Electric Reload Fuel Application for Oyster Creek.
3.4-2 I system becomes inoperable, the reactor shall be placed in the cold shutdown condition and no work shall be performed on the reactor or its con-nected systems which could result in lowering the reactor water level ~to less than 4'8" above the top of.the active fuel.
- 7. If necessary to accomplish maintenance or modifi-
- cations to the core spray systems, their power supplies or water supplies, reduced. system avail- .
ability.is permitted when the reactor is: (a) ! maintained in the cold shutdown condition or (b) in the refue1~ mode with the reactor coolant system maintained at less than 212*F.and vented, and (c)-no work is performed on the reactor vessel and connected systems that could result in lowering the reactor water level to less than 4'8" above the top of the active fuel. Reduced Core Spray System Availability is minimally defined as follows:
- a. At leart one core spray pump, and system components necessary to' deliver rated core spray to the reactor vessel, must remain operable to the-extent that the pump and any necessary valves can be started or operated from the control room or from local control
- stations.
- b. The fire protection. system is operable, and
- c. These systems are demonstrated to be operable on a weekly basis.
e'
- 8. If necessary to accomplish maintenance or modifi-cations to the core spray systems, their power supplies or water supplies, reduced system availability is permitted when the reactor is in the refuel mode with the reactor coolant system maintained at less than 212*F or in the startup mode for the purposes of low power physics testing. Reduced core spray system availability is defined as follows:
- a. At least one core spray pump in each loop, and system components necessary to deliver 4
. Amendment No.
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3.4-4
- e. (1) No work shall be performed on the reactor or its connected systems which could result in lowering the reactor water level to less than 4'8" above the top of the active fuel and the conden-sate storage tank level is greater than thirty (30) feet (360,000 gallons). At least two redundant systems including core spray pumps and system components must remain operable as defined in d.
above. OR (2) The reactor vessel head, fuel pool gate, and separator-dryer pool gates are removed and the water level is above elevation 117 feet. NOTE: When filling the reactor cavity
- from the condensate storage tank and draining the reactor cavity to the con-
. densate storage tank, the 30 foot limit does not apply provided there is suf- . ficient amount of water to complete the flooding operation.
- B. Automatic Depressurization System
- 1. Five electromatic relief valves of the automatic depressurization' system shall be operable when-
. the reactor-water temperature is greater than l 212*F and pressurized above 110 psig, except as specified in 3.4.B.2. The automatic pressure relief function of these valves (but not the automatic depressurization function) may be
- inoperable or bypassed during the system hydro-
! static pressure test required by ASME Code Section XI, IS-500 at or near the end of each ten 4 year inspection interval.
- 2. If at any time there are only four operable elec-tromatic relief valves, the reactor may remain in operation for a period not to exceed 3 days pro-vided the motor operated isolation and condensate Amendment No.
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3.4-6 provided the two pumps are not in the same loop. If more than two pumps become inoserable, the limits of Specification 3.4.C.3.sia11 apply..
- 5. During the period when one diesel is inoperable, the containment' spray loop and. emergency service water. system loop connected to the operable diesel shall have no inoperable components..
, 6. If primary' containment integrity is not required (see Specification 3.5.A), the containment spray 4
system may be made inoperable.
- 7. .If' Specifications 3.4.C.3, 3.4.C.4, 3.4.C.5 or 3.4.C.6 are not met, the reactor shall be placed in the cold shutdown condition. If the contain-ment spray system or the emergency service water system.becomes inoperable, the reactor shall be placed in the cold shutdown condition and no work shall be performed on the reactor or its connec-
-ted systems which could result in lowering the reactor water level to less than 4'8" above the top of the active fuel.
- . :8. The containment spray system may be made inoper-able during the integrated primary containment leakage rate test required by Specification-4.5, provided'that the reactor is maintained in the cold shutdown condition and that no work is
-performed on1the reactor or its connected. systems which could result in lowering the reactor level to less than 4'8" above.the top of the active fuel.
D. Control Rod Drive Hydraulic System
- 1. The control rod drive (CRD) hydraulic system shall tne operable when the reactor water -
temperature is above 212*F except as specified in 3.4.D.2 below.
- 2. If one CRD hydraulic pump becomes inoperable when the reactor water temperature is above 212 F, the reactor may remain in operation for a period not to exceed 7 days provided the second CRD hydrau-lic pump is operating and is checked at least
.once every 8 hours. If this condition cannot be met,'the reactor water temperature shall be reduced to(212*F.
L l Amendment No. i
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3.4-7 E. Core Spray and Containment Spray Pump Compartments
~ Doors The core spray and containment spray pump compartments doors shall be closed at all times except during passage.in order to conside.- the core spray system and the containment spray system operable.
F. Fire Protection System
- 1. The fire protection system shall be operable at all times with fuel in the reactor vessel except as specified in Specification 3.4.F.2.
- 2. If the fire protection system becomes inoperable during the run mode, the reactor may remain in operation provided both core spray system loops are operable with no inoperable components.
Bases: This specification assures that adequate emergency su core ecclir.c capability is availabic when the cere E spray system is required. Based on the loss-of-coolant analysis for the worst line break, a core spray of at least 3400 gpm is required within 35 seconds to assure effective core cooling.*l 1 Thus, if one loop becomes' inoperable, the operable loop is capable of providing cooling to the core and the reactor may remain in operation for a period of 7 days provided repairs can be completed within that time. The 7 days is based upon the consideration discussed in the bases of Specification 3.2 a d the pump operability tests of Specification 4.4. If repairs cannot be made, the reactor is depressurized and vented to prevent pressure buildup and no work is allowed to be performed on the reactor which could result in lowering the water level below 4'8" above the top of active fuel. Each core. spray loop contains redundant active compo-nents. Therefore, with the loss of one of these components the-system ~is still capable of supplying
-* Core Spray' System 2 is required to deliver 3640 gpm.
Amendment No.
3.4-8 rated flow and the system as a whole (both~1 oops) can
. tolerate an additional single failure of one of its Lactive. components and still perform the intended func-tion and' prevent clad melt. Therefore, if a redundant
. active component fails, a longer repair period is jus-
'tified basedRon-the consideration given in the bases of Specification 3.2. The consideration indicates
.that for a one out of 4 requirement the time out of service would be 1F 30 days
= 17.5 days TT7T
- 1.71 LSpecification 3.4.A.5 ensures that if one diesel is out ofLservice for repair, the core spray system loop on the other' diesel must be operable with no compo-nents out of service. This ensures that the loop can perform its intended function, even assuming one of
~itstactive components fails. If this condition is not met, the reactor is'placed in a condition where core spray is no longer required..
When1the reactor is in the shutdown ~or refueling mode and'the reactor coolant system-is less than 212*F and vented-and no work is being performed that could result in lowering the water level to less than 4'8" above the core, the likelihood of a leak or rupture leading to uncovering of the core is very low. The only source of energy that must be removed is decay heat'and one day after shutdown this heat generation rate is conservatively calculated to be not more than 0.6% of rated power. Sufficient core spray flow to cool the core can be supplied by one core spray pump or one of the two fire protection system pumps under these conditions. - When itL1s necessary to perform repairs on-the core spray system components, power supplies or water sources, Specification 3.4.A.7 per-mits reduced cooling system capability to that which could provide sufficient core spray flow from two independent sources. Manual initiation of these
- systems is adequate since it can be easily accomplish-ed within 15 minutes during which time the temperature rise in the reactor will not reach 2200*F.
i l L Amendment No. 1
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2.3-3
~ FUNCTION LIMITING SAFETY SYSTEM SETTINGS C. . Reactor High, d 1060 psig Pressure, Scram D. Reactor High Pressure, 2@ 1070 psig Relief Valves Initiation 3@ 1090 psig E. Reactor High Pressure, d 1060 psig with time delay Isolation Condenser n3 seconds l Initiation F. Reactor High Pressure, 4 @ 1212 psig Safety Valve Initiation 4 @ 1221 psig + 12 psi 4 @ 1230 psig 4 @ 1239 psig G. Low Pressure Main Steam 2 825 psig Line, MSIV Closure H. Main Steam Line Isolation d 10% Valve Closure f rom Valve Closure, Scram full open I. Reactor Low Water Level, d 11'5" above the top of the Scram active fuel as indicated under normal operating conditions J. Reactor Low-Low Water 6 7'2" above the top of the Level, Main Steam Line active fuel as indicated Isolation Valve Closure under normal operating conditions K. Reactor Low-Low Water n 7'2" above the top of the Level, Core Spray active fuel .
l Initiation L. Reactor Low-Low Water 6 7'2" above the top of the Level, Isolation Con- active fuel with time denser Initiation delay n3 seconds M. Turbine Trip, 10 percent turbine stop Scram valve (s) closure from full open N. Generator Load Rejection, Initiate upon loss of oil Scram pressure from turbine acceleration relay O. Recirculation Flow, Scram 6 71.4 Mlb/hr (117% of rated flow) 1
2.3-8 s During periods when the reactor is. shut down, decay heat is present and adequate water level must be maintained to provide core cooling. Thus, the low-low level trip. point of 7'2" above the core is provided to actuate the core spray system to provide cooling water should the-level drop to this-point. In addition, the normal reactor feedwater system and control rod-drive hydraulic system provide protection for the water level safety limit both > when the reactor is operating at power and in the shutdown condition. The turbine stop valve (s) scram is provided to anticipate
. the pressure, neutron flux, and heat flux increase caused by the rapid closure of the turbine stop valve (s) and failure of the turbine bypass system. l y The generator load rejection scram is provided to anticipate the rapid increase in pressure'and neutron flux ! resulting from fast. closure of the turbine control valves to-a load rejection and failure of the turbine bypass 4
system. This' scram is initiated by the loss of turbine acceleration relay oil pressure. The timing for this scram is almost identical to the turbine trip. l The total recirculation flow scram is provided to terminate a flow increase transient. Flow transient are normally protected against by employing the kg factor and use of mechanical stops on the recirculation pumps. Oyster Creek does not have mechanical stops on its recirculation pumps and maximum flow is beyond the limit for which the kg factor provides protection. The recirculation flow scram is set to the maximum flow level corresponding to the kg curve to be used (Section 3.10), i References (1) FDSAR, Volume I, Section VII-4.2.4.2 (2) FDSAR, Amendment 28, Item III.A-12 (3) FDSAR, Amendment 32, Question 13 (4) Letters, Peter A. Morris, Director, Division of Reactor Licensing, USAEC to John E. Logan, Vice President, Jersey "entral Power and Light Company, dated November 22, 1967'and January 9, 1968 (5) FDSAR, Amendment 65, Section B.XI. (6) FDSAR, Amendment 65, Section B.IX. r e l
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3.2-2 the rod program is being followed. A startup without the RWM as described in this subsection shall be reported in a special report to the Nuclear Regulatory Commission (NRC) within 30 days of the startup stating the reason for the failure of the RWM, the action taken to. repair it and the schedule for completion of the repairs. Control rod withdrawal sequences shall be established with a banked position withdrawal sequence so that the rod drop accident design limit of 280 cal /gm is not exceeded. For control rod withdrawal sequences not in strict compliance to BPWS, the maximum in sequence rod worth shall be 61.0% AK.
- 3. The average of the scram insertion times of all operable control rods shall be no greater than:
Rod Length Insertion Time Inserted (Percent) (Seconds) 5 0.375 20 0.900 50 2.00 90 5.00 The average of the scram insertion times for the three fastest control rods of all groups of four
. control rods in a two by two array shall be no greater than:
Rod Length Insertion Time Inserted (Percent) (Seconds) 5 0.398 20 0.954 50 2.120 90 5.300 Any four rod group may contain a control rod which is valved out of service provided the above requirements and Specification 3.2.A are met. Time zero shall be taken as the de-energization of the pilot scram valve solenoids.
- 4. Control rods which cannot be moved with control rod drive pressure shall be considered inoper-able. If a partially or fully withdrawn control rod drive cannot be moved with drive or scram pressure the reactor shall be brought to a
n_ , 3.2-5
'for-the entire subsequent fuel cycle. The criterion will be satisfied by demonstrating Specification 4.2.A at the beginning of each fuel' cycle with the core in the cold, xenon-free condition. This demonstration will include consideration for the calculated reactivity characteristic during'the following operating cycle and the uncertainty in this calculation.
The control rod drive housing support restricts the outward movement of-a: control rod to less than 3 inches in the extremely remote event of a housing failure (2). The amount of. reactivity which could be added by this small 1 amount of .rcx1 ' withdrawal, -which is less than a normal single withdrawal increment, will not contribute to any damage to the reactor coolant-system. The support is not required when no fuel is in the core since no nuclear consequences could occur in the absence of fuel. The support is not required if the reactor coolant system is
'at atmospheric pressure since there would then be no. -
driving force to rapidly eject:a drive housing. The support 11s not required if all control rods are fully inserted since the reactor would remain subcritical even in the event of complete ejection of the strongest control
- rodt51, i
The Rod Worth Minimizer (4) provides automatic
-supervision of conformance-to the specified control rod patterns. It serves as a back-up to procedural control of 1
control rod worth. In the' event that the RWM is out of f service when required, a licensed operator can manually fulfill the control rod pattern conformance functions of the RWM in-which case the normal procedural controls are backed up by~ independent procedural controls to assure conformance during control rod withdrawal. This allowance to perform ~a startup without the RWM is limited to once each calendar year to assure a high operability of the RWM
- which is preferred over procedural controls.
n Control rod drop accident (RDA) results for plants using banked position withdrawal sequences (BPWS) show that in all casesJthe peak fuel enthalpy in a RDA. would be much less than the 280 cal /gm design limit even with the
- maximumn; incremental rod worth. The BPWS is developed prior to initial operation of the unit following any refueling outage and the requirement that the operator e follow the BPWS-is supervised by the RWM or a second licensed operator. If it is necessary to deviate l slightly from the BPWS sequence (i.e., due to an l inoperable control rod) no further analysis is'needed if i -the maximum incremental rod worth in the modified sequence
- is.dl.0%-4K. An incremental control rod worth of $ 1.0% AK l
will not result in a peak fuel enthalpy above the design l- limit of 280 cal /gm as documented in reference 10.
r 3.2-6 The BPWS Limits the reactivity worths of control rods and together with the integral. rod velocity limiters and the
. action of the control rod drive system limits potential reactivity insertion such~ that the results of a control rod drop accident will not exceed a maximum fuel energy content of 280 cal /gm. Method'and basis for the rod drop accident analyses are documented in Reference 5.
The control rod system is designed to bring the reactor subcritical from a scram signal at a rate fast enough to prevent fuel damage. Scram reactivity curve for the transient analyses is calculated and evaluated with each reload core. In the analytical treatment of the tran-sients, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods. This is adequate and conservative when compared to the typical time delay of about 210 millisec-onds estimated from scram test results. Approximately the first 90 milliseconds of each of these time intervals result from the . sensor and circuit delays when the pilot scram solenoid de-energizes. Approximately 120 millisec-onds later, the control rod motion is estimated to actually begin. However, 200 milliseconds is conserva-tively assumed for this time interval in the trar.sient analyses and this is also included in the allowable scram insertion times of Specification 3.2.B.3. The specified
. limits provide sufficient scram capability to accommodate failure to scram of any one operable rod. This failure is in addition to any inoperable rods that exist in the core, provided that those inoperable rods met the core reacti-
-vity Specification 3.2.A.
Control rods (6) which cannot be moved with control rod drive pressure are clearly indicative of an abnormal operating condition on the affected rods and are, there-fore, considered to be inoperable. Inoperabe rods are i valved out of service to fix their position in the core and assure predictable behavior. If the rod is fully inserted and then valved out of service, it is in a safe position of maximum contribution to shutdown reactivity. If it is valved out of service in a non-fully inserted position, that position is required to be consistent with the shutdown reactivity limitation stated in Specification 3.2.A, which assures the core can be shutdown at all times
, with control rods. Before rod is valved out of service in a non-fully inserted position an analysis is performed to insure specification 3.2.A is met.
Amendment No.
. 3.2-9 During each. fuel cycle excess operating reactivity varies as fuel depletes and as any burnable poison in supplementary control is burned. The magnitude of this -
excess reactivity is indicated by the integrated worth of control rods inserted into the core, referred to as the control rod inventory in the core. As fuel burnup progresses,-anomalous behavior in the excess reactivity may be detected by comparison of actual rod inventory with expected inventory based on appropriately corrected past datc. Experience at Oyster Creek and other operating BWR's indicates that the control rod inventory should be predictable to the equivalent of one percent in reactivity. Deviations beyond this magnitude would not be expected and would require thorough evaluation. One percent reactivity limit is considered safe since an insertion of this reactivity into the core would not lead to transients exceeding design conditions of the reactor system. I
References:
l (1) FDSAR, Volume I, Section III - 5.3.1 (2) FDSAR, Volume I, Section VI-3 (3) FDSAR, Volume I, Section III - 5.2.1 (4) FDSAR, Volume I, Section VII-9 (5) NEDO-24195, General Electric Reload Fuel Application for Oyster Creek. (6) FDSAR, Volume I, Section III-5 and Volume II, Appendix B (7) FDSAR, Volume I, Sections VII - 4.2.2 and VII - 4.3.1 (8) FDSAR, Volume I, Section VI-4
'(9) FDSAR, Amendment No. 55, Section 2 (10) C. J. Paone, Banked Position Withdrawal Sequence, January 1977 (NEDO-21231) i' l
l l
3.10-3 C. - Minimum Critical Power Ratio .(MCPR) , During steady state power operation, MCPR shall be greater than or equal to the following: APRM Status MCPR Limit
- 1. If any-two (2)- LPRM assemblies which 1.40 l are input to the APRM system and are separated in distance by less than three '(3) times the control rod pitch contain a combination of (3) out of four (4)-detectors located in either the A and B or C.and D levels which g are failed or bypassed i.e., APRM-channel or LPRM input bypassed or inoperable.
- 2. If any LPRM input to the APRM system 1.40 l at the B, C, or D level is failed or bypassed or any APRM channel is in-operable.(or bypassed).
- 3. All B, C, and D LPRM inputs to the 1.40 l APRM system are operating and no APRM channels are inoperable or bypassed.
When APRM status changes due to instrument failure (APRM or LPRM input failure), the MCPR requirement for the degraded condition shall be met within a time interval of eight (8) hours, provided that the control rod block is placed in operation during this interval. For core flows other than rated, the nominal value for MCPRlshall be increased by a factor of kg, where kg is-as shown in Figure 3.10-6. If at any time during power operation it is determined by normal-surveillance that the limiting value for MCPR is being exceeded for reasons other than instru-ment failure, action shall be initiated to restore operation to within the prescribed limits. If the steady state-MCPR is not returned to within the prescribed limits within two [2] hours, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours. During this period, surveillance and corresponding action shall continue Amendment No.
-3 4
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