ML20148H535
| ML20148H535 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 03/21/1988 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | GPU Nuclear Corp, Jersey Central Power & Light Co |
| Shared Package | |
| ML20148H540 | List: |
| References | |
| DPR-16-A-120 NUDOCS 8803300019 | |
| Download: ML20148H535 (19) | |
Text
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o UNITED STATES o
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j NUCLEAR REGULATORY COMMISSION 3-f WASHINGTON. D. C. 20555
,e GPU NUCLEAR CODr0PATt0N AND JERSEY CENTRAL POWER & LIGHT COMPANY 1
DOCKET NO. 50-?19 OYSTER CREER MICLEAR GENERATING STATION AMENDMENT TO PROVISIONAL OPERATfNG LTCENSE Amendment No.120 License No. DPR-16 1.
The Nuclear Regulatory Comission (the Comission)'has found that:
A.
The application for amendment by GPU Nuclear Corporation and, et al.,
(the licensee), dated January 19, 1988 complies with the standards and requirements of the Atomic Eneroy Act of'1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter It B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (11 that the activities authorized by this amen &ent can be conducted without endangering the health and safety of the public, and (iii that such activities will be conducted in compliance with the Comission's regulations; 0.
The issuance of this amendrent will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this anendrent is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
8803300019 800321 PDR ADOCK 05000219 p
e E.
Accordingly, the licerse is anended by charges to the Techr.ical Specificaticns as indicated in the attachnent to this license atendment, and paragraph 2.C.(2) of Provisional Operating License No. CFR-16 is berehy arended to read as follows:
(?) Technical Spe,cifications The Technical Specifications contained in Appendices A and R. as revised through An.endment Fc.120, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical Specifications, i
3.
This license amendment is effective as of the date of issuance to be in;plenented within 30 days.
FOR THE NUCLEAR REGUL TORY OMMISSION s
I J hn '. Stolz, Directop j P
t Directorate I-V Division of Feactor Projects I/II Office of Nuclear Reactor Regulation Attachnent:
Charges to the Technical Specificatters Cate of Issuance:
March 21, 1988 4
5 i
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ATTACHMENT TO LICENSE AWENDMENT No.120 PROVISIONAL OPFRATING LICENSE Nu. DPP-16 DOCKET FO. 50-219 Peplace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated.
The revised pages are identified by amendnent number and contain vertical lines indicating the areas of change.
Remove Insert Page 1.0-2 Page 1.0-?
Pace 1.0 7 Page 3.1-18 Page 3.1-17 Dage 3.3-1 Page 3.3-1 Page 3.3-5 Page 3.3-5 Page 3.3-8 Page 3.3-8 Page 3.3-9a Page 3.3-9b Page 3.3-9i.
Page 3.4 2 Page 3.4-2 Page 3.4-4 Page 3.4-4 Page 3.a-5 Page 3.4-5 Page 3.5-?
Page 3.5-2 Page 3.8-1 Pace 3.A-1 Page 4.3-1 Page 4.3-1 Page 4.3-2 Page 4.3 2 i
I 4
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1.7 COLD SHUTOOWN The reactor is at cold shutdown when the mode switch is in the shutdown mode position, there is fuel in the reactor vessel, all operable control rods are fully inserted, and (except during reactor vessel presst.'e testing), the reactor l
coolant system maintained at less than 212'F and vented.
1.8 PLACE IN SHUTDOWN CONDITION Proceed with and maintain an uninterrupted normal plant shutdown operation until the shutdown condition is met.
1.9 PLACE IN COLD SHUT 00WN CONDITION Proceed with and maintain an uninterrupted normal plant shutdown operation until the cold shutdown condition is met.
1.10 PLACE IN ISOLATED CONDITION Proceed with and maintain an uninterrupted normal isolation of the reactor from the turbine condenser system including closure of the main steam isolation valves.
1.11 REFUEL H0DE Tne reactor is in the refuel mode when the reactor mode switch is ih the refuel moda position arid there is fuel in the reactor vessel.
In thit mode i
the refueling platform interlocks are in operation.
1.12 REFUELING OUTAGE For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of the end of the previous refueling outage, the test or surveillance need not be performed until the next regularly scheduled outage.
Following the first refueling outage, the time between successive tests or surveillance sns11 not exceed 20 months.*
1.13 PRIMARY CONTAINMENT INTEGRITY l
Primary containment integrity means that the drywell and adsorption chamber are closed and all of the following conditions are satisfied:
A.
All non-automatic primary containment isolation valves which are not required to be open for plant operation are closed.
B.
At least one door in the airlock is closed and sealed.
C.
All automatic containment isolation valves specified in Table 3.5.2 are operable or are secured in the closed position.
i D.
All blind flanges and manways are closed.
1
- The time between successive tests or surveillances shall not exceed 30 months prior to the cycle 10 refueling outage only.
OYSTER CREEK 1.C-2 Amendment No.: /, g,J/, /, 120 l
i
i i
i
- 1. 39 REACTOR VESSEL PRESSURE TESTING System pressure testing required by ASME Code Section XI, Article IWA-5000, including system leakage and hydrostatic tests, with reactor vessel completely water solid, core not critical and Section 3.2.A satisfied.
i I
l OYSTER CREEK 1.0-7
.n
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o TABLE 3.1.1 (CONTD) u.M k.
All four (4) drywell pressure instrument channels may be made inoperable during the integrated primary containment leakage rate test (See Specification 4.5), provided that the plant is in the cold shutdown 9
condition and that ne work is performed on the reactor or its connected systems which could result in Ni Iowering the reactor water level to less than 4'8" above the top of the active fuel.
1.
Bypass in IRM Ranges 8, 9, and 10.
L m.
There is one time delay relay associated with each of two pumps.
n.
Ore time delay relay per pump must be operable.
o.
There are two time delay relays associated with each of two pumps. One timer per puup is for sequence starting (SK1A, SK2A) and one timer per pump is for tripping the pump circuit breaker (SK7A, SK8A).
i p.
Two time delay relays per pump must be operable.
i q.
Manual initiation of affected component can be accomplished after the automatic load sequencing is completed.
r.
Time delay starts after closing of containment spray pump circuit breaker.
s.
These functions not required to be operable with the reactor temperature less than 212*F and the vessel i
head removed or vented or during reactor vessel pressure testing.
l w
t.
These functions may be operable or bypassed when corresponding portions in the same core spray system
- /.
logic train are inoperable per Specification 3.4.A.
i u.
These functions not required to be operable when primary containment integrity is not required to be i
maintained.
l k
These functions not required to be operable when the ADS is not required to be operable.
v.
w.
These functions must be operable only when irradiated fuel is in the fuel pool or reactor vessel and l
I secondary containment integrity is required per specification 3.5.B.
a y.
The number of operable channels may be reduced to 2 per Specification 3.9-E and F.
2 z.
The bypass function to permit scram reset in the shutdown or refuel mode with control rod block must be P
operable in this mode.
aa. Pump circuit breakers will be tripped in 10 seconds + 15% during a LOCA by relays SK7A and SK8A.
h cc Pump circuit breakers will trip instantaneously during a LOCA.
bb Only applicable during startup mode while operating in IRM range 10.
dd. If an isolation condenser inlet (steam side) isolation valve becomes or is made inoperable in the open y;.k
. position during the run mode comply with Specification 3.8.E.
If an AC motor-operated outlet (conden-sate return) isolation valve becomes or is made inoperable in the open position during the run mode comply o
with Specification 3.8.F.
ee. With the number of operable channels one less than the Min. No. of Operable Instrument Channels per l
Operable Trip Systems, operation may proceed until performance of the next required Channel Functional j
Test provided the inoperable channel is placed in the tripped condition within I hour.
j ff. This function is not required to be operable when the associated safety bus is noc required to be ener-gized or fully operable as per applicable sections of these technical specifications.
X
4
- 3. 3 REACTOR COOLANT Aoplicability: Applies to the operating status of the reactor coolant system.
Objective:
To assure tne structure integrity of the reactor coolant system.
Soecification:
A.
Pressure Temperature Relationships (i)
Reactor Vessel Pressure Tests - the minimum reactor vessel temperature at a given pressure shall be in excess of that indicated by the curve in Figure 3.3.l(a).
The maximum temperature for Reactor Vessel Pressure Testing is 250*F.
(ii)
Heatup and Cooldown Operations: Reactor noncritical--the minimum reactor vessel temperature for heatup and cooldown operations at a given pressure when the reactor is not critical shall be in excess of that in11cated by the curve in Figura 3.3.l(b).
(iii)
Power operations--The minimum reactor vessel temperature for power operations at a given pressure snall be in excess of that indicated by the curve in Fiqure 3.3.l(c).
Note: Figures 3.3.l(a), (b) and (c) apply when the closure heao is on the reactor vessel and studs are fully tensioned.
(iv)
Appropriate new pressure temperature limits must be approved as part of this Technical Specification when the reactor system has reached fif teen effective full power years of reactor operation.
B.
Reactor Vessel Closure Head Boltdown The reactor vessel closure head studs may be elongated
.020" (1/3 design preload) with no restrictions on reactor vessel terperature as long as the reactor vessel is at abnospheric pressure.
Full tensioning of the studs is not permitted unless the temperature of the reactor vessel flange and closure head flange is in excess of 100*F.
C.
Thermal Transients 1.
The average rate of reactor coolant temperature change during normal heatup and cooldown shall not exceed 100*F in any one hour period.
2.
The pump in an idle recirculation loop shall not be started unless the temperature of the coolant within the idle recirculation loop is within 50*F of the reactor coolant tenperature.
0YSTER CREEK 3.3.-l Amendment No.:
f(,120
e Transformation temperature.
The minimum temperature for pressurization at any time in life as to account for the toughness properties in the most limiting regions of the reactor vessel, as well as the effects of fast neutron i
embrittlement.
Figures 3.3.l(a), (b) and (c) are derived from an evaluation of the fracture toughness properties performed on the specimens contained in Reactor Vessel Materials Surveillance Program Capsule No. 2 (Reference 14).
The results of dosimeterwireanalyses(Reference 14)indicatedthattheneutronfluence{epl.0 MeV)2 at the 1/4T (T= vessel wall thickness) location at the end of 15 effective full power years of operation is 1.11 x 10 n/cm This value was used in the calculation of the adjusted reference nil-ductil".y temperature which, in turn, was used to generate the pressure-temperature curves in Figures 3.3.1(a),
1 (b) and (c).
The 2500F maximum pressure test temperature provides ample margin against violation of the minimum required temperature.
Secondary containment is not jeopardized by a steam leak during pressure testing, and the Standby Gas Treatment system is adequate to prevent unfiltered release to the stack.
j Stud tensioning is considered significant from the standpoint of brittle fracture only when the preload exceed approximately 1/3 of the final design value.
No vessel or closure stud minimum temperature requirements are considered necessary for preload values Delow 1/3 of the design preload with the vessel depressurized since preloads below 1/3 of the design preload result in vessel closure and average bolt stresses which are less than 20% of the yield strengths of the vessel and bolting materials.
Extensive service experience with these materials has i
confirmed that the probability of brittle fracture is extremely remote at these'
)
low stress levels, irrespective of the metal temperature.
The reactor vessel head flange and the vessel flange in combination with the double "0" ring type seal are designed to provide a leak tight seal when bolted together. When the vessel head is placed on the reactor vessel, only that portion of the head flange near the inside of the vessel rests on the vessel flange. As the head bolts are replaced and tensioned, the vessel head is flexed slightly to bring together the entire contact surfaces adjacent to the "0" rings of the head and vessel flange.
Both the head and the head flange have an NDT temperature of 40*F, and they are not subject to any appreciable neutron radiction exposure.
Therefore, the minimum vessel head and head flange temperature for bolting the head flange and vessel flange is established as 40'F + 60*F or 100*F.
Detailed stress analyses (4) were made on the reactor vessel for both steady state and transient conditions with respect to material fatigue.
The results of these analyses are presented and compared to allowable stress limits in Reference (4).
The specific conditions analyzed included 120 cycles of normal startup and shutdown with a heating and cooling rate of 100*F 9er hour applied continuously over a temperature range of 100*F to 546'F and for 10 cycles of emergency cooldown at a rate of 300'F per hour applied over the same range.
Thermal stresses from this analysis combined with the primary load stresses f all within ASME Code Section III allowable stress intensities.
Although the Oyster Creek Unit I reactor vessel was built in accordance with Section I of the ASME Code, the design criteria included in the reactor vessel specifications were in essential 9g eement with the criteria subsequently incorporated into Section III of the Code.l0(1 OYSTER CREEK 3.3-5 Amendment No.: )(,g,120
pH, enloride, and other chemical parameters are made to determine the cause of the unusual conductivity and instigate proper corrective i
action.
These can be done before limiting conditions, with respect to variables affecting the boundaries of the reactor coolant, are exceedeo.
Several technioues are available to correct off-standard re&ctor water auality conditions including removal of impurities from reactor water by the cleanup system, reducing input of impurities causing off-standard conditions by reducing power and reducing the reactor coolant temperature to less than 212*F.
The major benefit of reducing the reactor coolant temperature to less than 212*F is to reduce the temperature dependent corrosion rates and thereby provide time for the cleanup system to re-establish proper water cuality.
References (1) FOSAR, Volume I, Section IV-2 (2)
(Deleted)
(3)
(Deleted)
(4) Licensing Application Amendment 16, Design Requirements Section (5)
(Deleted)
(6) FOSAR, Volume I, Section IV-2.3.3 and Volume II, Appendix H (7) FOSAR, Volume I, Taole IV-2-1 (8) Licensing Application Anendment 34, Question 14 (9) Licensing Application Amendment 28. Item III-B-2 (10) Licensing Application Amendment 32, Question 15 (11) (Deleted)
(12) (Deleted)
(13) Licensing Application Amendment 16, Page 1 (14) GPUN TOR 725 Rev. 0: Testing and Evaluation of Irradiated Reactor l
Vessel Materials Surveillance Program Specimens.
l OYSTER CREEK 3.3-8 knendment No. : g,FI,120
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l a.
At least 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.
8.
If necessary to accomplish maintenance or modifications.to the core i
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 l
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 rated core spray to the reactor vessel, must remain operable to the extent that the pump and any necessary valves in each loop can be started or operated from the control room or from local control stations.
b.
The fire protection system is operable and, c.
Each core spray pump and all components in 3.4.A.8a are demonstrated to be operable every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
9.
If Specifications 3.4. A.7 and 3.4. A.8 cannot be met, the requirements of Specification 3.4.A.6 will be met and work will be initiated'to meet minimum operability requirements of 3.4.A 7 and 3.4.A.8.
10.
The core spray system is not required to be operable when the following conditions are met:
a.
The reactor mode switch is locked in the "refuel" or "shutdown" position.
b.
(1) There is an operable flow path capable of taking suction from the condensate storage tank and transferring water to the reactor vessel, and (2) The fire protection system is operable.
c.
The reactor coolant system is maintained at less than 212*F and vented (except during reactor vessel pressure testing).
l d.
At least one core spray pump, and system components necessary to deliver rated core spray flow to the reactor vessel, 0YSTER CREEK 3.4-2 Amendment No.: JfI,120
the reactor vessel, except as specified in Specifications 3.4.C.3, 3.4.C.4, 3.4.C.6 and 3.4.C.8.
2.
Tng absorption chamber water vo;ume shall not be less than 82,000 ftJ in order for the containment spray and emergency service j
water system to be considered operable.
3.
If one emergency service water system loop becomes inoperable, its associated containment spray system loop shall be considered inoperable.
If one containment spray system loop and/or its associatea emergency service water system loop becomes inoperable during the run moce, the reactor may remain in operation for a period not to exceed 7 days provided the remaining containment spray system loop and its associated emergency service water system loop each have no inoperable components and are demonstrated daily to be operable.
4.
If a pump in the containment spray system or emergency service water system becomes inoperable, the reactor may remain in operation for a period not to exceed 15 days provided the other similar pump is demonstrated daily to be operable. A maximum of two pumps may be inoperable provided the two pumps are not in the same loop.
If more than two pumps become inoperable, the limits of Specification 3.4.C.3 shall apply.
S.
During the period when one diesel is inoperable, the contair. ment spray loop and emergency service water system loop connected to the operable diesel shall have no inoperable components.
l l
6.
If primary containment integrity is not reauired (see i
Specification 3.5.A), the containment spray 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, tne reactor shall be placed in cold shutdown condition.
If j
the containment spray system or the emergency service water system l
becomes inoperable, the reacttr shall be placed in the cold 1
shutdown condition and 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, i
8.
The containment spray system may be made inoperable during the integrated primary containment leakage rate test recuired by Specification 4.5, provided that the reactor is maintained in the cold shutdown condition and that no work is performed on the 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.
O.
Control Rod Drive Hydraulic System 1.
Tne control rod drive (CRO) hydraulic system shall be operable
)
when the reactor water temperature is above 212*F except as specified in 3.4.0.2 and 3.4.0.3 below.
OYSTER CREEK 3.4-4 Amendment No.:
15, W, 16, 120 Correction:
12/24/84
2.
If one CR0 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 hydraulic pump is operating and is checked at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
If this condition cannot be met, the reactor water temperature shall be reduced to <( 212*F.
3.
During reactor vessel pressure testing, at least one CR0 pump snall be operable.
E.
Core Spray and Containment Spray Pump Compartments'0cors The core spray and containment spray pump compartments-doors shall be closed at all times except during passage in order to consider the core spray system and the containment spray system operable.
F.
Fire Protection System j
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 fir 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 adeouate emergency core cooling capability is available when the core spray system is required. Based 1
on the loss-of-coolant analysis for the worst line break, a core spray of at least 34Q core cooling.*lg)gpm is required with 35 seconds to assure effective 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 and 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 components.
Therefore, with the loss of one of these components the system is still capable of supplying rated flow and the system as a whole (both loops) can tolerate an additional single failure of one of its active components and still perform the intended function and prevent clad melt.
Therefore, if a redundant active component f ails, a longer repair period is justified based on the consideration given in the bases of Specification 3.2.
The consideration indicates that for a one out of 4 recuirement the time out of service would be
)
30 days W" 1.71 " 17.5 days
' core Spray System 2 is reauired to deliver 3640 gpm.
0YSTER CREEK 3.4-5 Amendment No.: Jf',120 t
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4 b.
(1) Tnere is an operable flow path capable of taking su: tion from the condensate storage tank and transferring water to the reactor vessel, and (2) The fire protection system is operable, c.
The reactor coolant system is maintained at less than 212'F arn vented.
d.
At least one core spray pump, and system compontats necessary to deliver rated core spray flow to the reactor vessel, must remain operable to tne extent tnat the pump and any necessary valves can be started or operated from the control room or from local control stations, and the torus is mechanically intact.
e.
(1) No work shall be performed on the reactor or its connected systems whien could result.in lowering the reactor water level to less than 4'8" above the top of the active fuel and the condensate 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 f eet.
NOTE: When filling the reactor cavity from the condensate storage tank and draining the reactor cavity to the condensate storage tank, the 30 foot limit does not apply provided there is a sufficient amount of water to complete the flooding operation.
3.
Primary containment integrity shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212*F and fuel is in the reactor vessel except while performing low power pnysics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 Mwt or during reactor vessel pressure l
testing.
a.
With one or more of the containment isolation valves shown in Table 3.5.2 inoperable:
0YSTER CREEK 3.5-2 Amendment No.: j#, g, N, 120
3.8
!SCt.ATION CONDENSES Appiirabi'ity: ApAlies to operating status of the isolation condenser.
Objectid:
To assure heat removal capability under conditions of reactor vessel isolation from its normal heat sink.
Spec i f i c.at i M :
A.
The two isolation condenser loops shall be, operable during
~
power operation and whenever the reactor coolant temperature is greater than 212*F except as specified in C, below or during reactor vessel pressure testing.
B.
The shell side of each condenser shall contain a minimum water volume of 22, 730 gallons.
If the minimum volume cannot be maintained or if a source of makeup water is not available to the condenser, the condenser snail be considered inoperable.
C.
If one isolation condenser becomes inoperable during the run mode the reactor may remain in operation for a period not to exceed 7 days provided the motor operated isolation and condensate makeup valves in the operable isolation condenser are demonstrated daily to be operable.
D.
If Specification 3.8. A and 3.8.B are not met, or if an inoperable isolation condenser cannot be repaired witnin 7 days, the reactor shall be placed in the cold shutdown condition.
E.
If an isolation condenser inlet (steam side) isolation valve (V-14-30, 31, 32 or 33) becomes or is made inoperable, in the open oc:,ition during the run mode, the redundant inlet isolation valve shall be demonstrated operable.
If the inoperable valve is not returned to service within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> oeclare the affected isolation condense.e inoperable, isolate it and comply with Specification 3.8.C.
F.
If an AC motor-operated isolation condenser outlet (condensate return) isolation valve (V-14-36 or 37) becomes or is made inoperable in the open position in the run mode, return the valve to service within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or declare the affected isolation condenser inoperable, isolate it and comply with Specification 3.8.C.
Basis:
The purpose of the isolation condenser is to depressurize the reactor and to remove reactor decay heat in the event that the turbing) generator and main condenser is unavailable as a heat sink.li Since the shell side of the isolation condensers operate at atmospheric pressure, they can accomplish their purpose when the reactor temperature is sufficiently above 212*F to provide for the heat transfer corresponding to reactor decay he at. Tne tube side of the isolation condensers form a closed loop with the reactor vessel and can operate without reducing the reactor coolant water inventory.
0YSTER CREEK 3.8-1 knendment No. : X,120
4.3 REACTOR COOLANT Applicability: Applies to the surveillance reouirements for the reactor coolant system.
Objective:
To determine the condition of the reactor coolant system and the operation of tne safety devices related to it.
Specification:
A.
Materials surveillance specimens and neutron flux monitors shall be installed in the reactor vessel adjacent to the wall at the midplane of the active core.
Specimens and monitors snall be periodically removed, tested, and evaluated to detemine the effects of neutron fluence on the f racture toughness of the vessel shell materials.
The results of these evaluations shall be used to assess the adecuacy of tne P-T curves of Figures 3.3,1(a), (b) and (c). New curves shall be generated as reouired.
B.
Inservice inspection of ASME Code Class 1, Class 2 and Class 3 systems and components shall be performed in accordance witn Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a(g),
except where specific written relief has been granted by *,ne NRC pursuant to 10 CFR, Section 50.55a(g)(6)(i).
C.
Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves snall be performed in accordance with Section XI of tne ASME Boiler and Pressure Vessel Code and applicablo Addenda as reauired by 10 CFR, Section 50.55a(g), except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(i).
D.
A visual examination for leaks shall be made with the reactor coolant syste:n at pressure during each scheduled refueling outage or af ter major repairs have been made to the reactor coolant system in accordance with Article 5000,Section XI.
Tne reouirements of specification 3.3.A shall be met during the test.
E.
Each replacement safety valve or valve that has been repaired shall be tested in accordance with subsection IWV-3510 of Section XI of the ASME Boiler and Pressure Vessel Code.
Setpoints shall be as follows Number of Valves Set Points (psig) 4 1212 + 12 4
1221 7 12 4
1230 7 12 4
1239}12 F.
A sample of reactor coolant shall be analyzed at least every 72 nours for the purpose of determining the content of chloride ion and to check the conducti'rity.
0YSTER CREEK 4.3-1 knendment No.:
$2',90,120
..t
- G.
Primary Coolant System Pressure Isolation Valves Specif) cation:
1.
Periodic leakage testing (a) on each valve listea in table 4.3.1 shall be accomplishec prior to exceeaing 600 psig reactor pressure every time the plant is placeo in tne colo snutdown concition for refueling, each time tne plant is placed in a colo shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in the ' preceeding 9 months, whenever the valve is moved whether 6y manual actuation or due to flow concitions, and after returning the valve to service af ter maintenance, repair or replacement work is performea.
H.
Reactor Coolant System Leakage 1.
Unidentified leakage rate shall be calculated at least once every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
2.
Total leakage rate (identified anc unidentifiec) shall be calculatea at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
3.
A channel calibration of the primary containment sump flow integrator and the primary containment eouipment drain tank flow integrator shall be conducted at least once per 18 months.
Bases:
Data is available relating neutron fluence (E>1.0 MeV) and the change in the Reference Nil-Ductility Transition Temperature (RTNOT).
The pressure-temperature (P-T) operating curves of Figures 3.3.l(a), (b) and (c) were developed based on the results of testing and evaluation of specimens removed from the vessel af ter 8.38 EFPY of operation.
Similar testing and analysis will be performea throughout vessel life to monitor the effects of neutron irradiation on tne reactor vessel shell materials.
The inspection program will reveal problem areas should they occur, before a leak develops.
In addition, extensive visual inspection for leaks will be made on critical systems. Oyster Creek was designed and constructed prior to idTo satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.
- NRC Order dated April 20, 1981.
Corrected:
1/28/86 OYSTER CREEK 4.3-2 Amendment No.: K, #/,p11,120
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