ML19343D250
| ML19343D250 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/20/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19343D244 | List: |
| References | |
| NUDOCS 8105040043 | |
| Download: ML19343D250 (9) | |
Text
Attachment 3 e
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G ORDER FOR MODIFICATION OF LICENSE (EVENT V) 0YSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 Insert the following pages in the Appendix A Technical Specifications Pages 3.3-2*
3.3-2a 3.3-2b*
3.3 4.3-l*
4.3-la 4.3-2*
' '4. 3 - 9
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- ~There are no che.iges to the provisions contained on this page; it is merely included for pagination purposes.
810504Oet
3.3-2 i
D.
Reactor Coolant System Leakage Reactor coolant leakage into the primary containment from unidentified sources shall not exceed 5 gpm.
In addition, the total leakage in the containment, identified and unidentified, shall not exceed 25 gpm.
If'these conditions cannot be met, the reactor will be placed in the cold shutdown condition.
E.
Reactor Coolant Quality 1.
The reactor coolant quality shall not exceed the following limits during power operation with steaming rates to the turbine-condenser of less than 100,000 pounds per hour, conductivity 2 u cho/cm chloride ion 0.1 ppm 2.
The reactor coolant quality shall not exceed the fol.*owing limits during power operation with steaming rates to the o
turbine-condenser of at least 100,000 pounds per hour.
j conductivity 10 sho/cm chloride ion 1.0 ppm IfSpecifihation3.'3.E.1and3.3.E.2cannotbemet, the 3.
reactor shall be placed in the cold shutdown conditien.
F.
Recirculation Loop Operability 1.
The reactor shall not be operated with one or more recirculation loops out of service except as specified in Specification 3.3.F.2.
2.
Reactor operation with one idle recirculation loop is permitted provided that the idle loop is not isolated I
f rom the reactor vessel.
t 3.
If Specifications 3.3.F.1 and 3.3.F.2 cre not met the reactor shall be placed in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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- 3. 3-2a G.
Primary Coolant System Pressure Isolation Valves Applicability:
Operational Conditions - S;artup and Run Modes; applies to j
the operational status of the primary coolant system pres-sure isolation valves.
Objective:
To increase the reliability of primary coolant system pres-r sure isolation valves thereby reducing the potential of an intersystem loss of coolant accident.
Speci ficatiori:
1 During reactor power operating conditions, the integrity of all pressure isolation valves listed in Table 3.3.1 i*
shall be demonstrated. Valve leakage shall not exceed the amounts indicated.
2.
If Specification 1 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in the cold shutdown conditidn within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
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NRC Order dated April 20, 1981 i
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3.3-2b The reactor coolant system (
is's primary barrier against the Bases:
release of fission products to the environs.
In order to provide
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assurance that this berrier is maintained at a high degree of integrity, restrictions have been placed on the operating conditions to which it can be subjected.
The Oyster Creek reactor vessel was designed and manufactured in accordance with General Electric Specification 21A1105 and AsiE Section I as discussed in Reference 13.
The original operating 10aitations were based upon the require-ment that the minimum temperature for pressurization be at least 60'F greater than the nil ductility transformation temper ature. The minimum temperature for pressurization at any time in life had to account for the toughness properties in the most 1Laiting regions of the reactor vessel, as well as the ef fects of f ast neutron embrittlement.
Figures 3.3.1 is derived from an evaluation of the fracture toughness properties performed for Oyster Creek.
(Reference l
- 12) in an effort to establish new operating limits. The results of neutron flux dosimeter analyses in Reference 12 indicate that the total f ast neutron fluence (>l Mev) expected for Oyster Creek at the end of teS8' ""
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- full power years of operation is 1.22 x 10 nyt on the inside surface of the reactor vessel core region shell.
A conservative fast neutron fluence of 75% of this value is assumed it the 1/4 T (one quarter of wall thickness) location for the preparation of the pressure / temperature curves in Firure 3.3.1.
t Stud tensioning is considered significant from the standpoint of brittle fracture only when the preload exce.ed approximately 1/3 of the final design value.
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No vessel or clorure stud minimum temperature requirements are considered necessary for preload values below 1/3 of the design preload with the vessel depressurized since preloads below 1/3 of the design preload result in ver sel closure and average bolt stresses which are *.ess than 20% of.the yield strengths of the vessel and bolting naterials.
Extensive service experience with these materials I
has confirmed that the probability of brittle fracture is estremely remote at these low stress levels, irrespective of the metal temperature.
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Amencnent i;o.15, C,
3.3-8 TABLE 3.3.1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum (a)
System Valve No.
Allowable Leelage Core Spray System 1 NZO2A 5.0 GPM NZO2C 5.0 GN1 Core Spray System 2 NZO2B 5.0 GPM NZO2D 5.0 GPM i
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Footnote:
(al.
Leakage rates less than or equal to 1.0 gpm are considered acceptable.
l 2.
Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not l
exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
3.
Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm re considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible _ rate of 5.0 gpm by 50% or greater.
4.
Leakage rates greater than 5.0 gpm are considered unacceptable.
5.
Test differential pressure shall not be less th:n 150 psid.
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4.3-1 4.3 REACTOR C00LAh7
. Applicability:
Applies to the surveillance requirements for the reactor j
coolant system.
' Objective:
To determine the condition of the reactor coolant system and the operation of the safety devices related to it.
Specification:
A.
Neutron flux monitors shall be installed in the reactor vessel adjacent to the vessel vall at the core midplane icvel.
The monitors shall be removed and tested at the first refueling outage to experimentally verify the calculated values of integrated neutron flux that are used to determine the NDIT from Figure 3.3.1.
B.
Non-des.ructive' examinations shall be made on the con-ponents as spe:ified in Table 4.3.1.
Any indication of a defect shall be investigated and evaluated.
C.
A visual examination for les.*es shall be made with the reactor co'olant sys. tem at pressu-e during each scheduled refueling 'o'Otage or af ter major repairs have been mace to the reactor coolant system.
The require =ents of specification 3.3.A shall be met during the test.
D.
Each replacement safety valve or valve that has been repaired shall be bench checked for the proper set point. A minimum of 5 of the valves shall be bench checked or replaced with a bench checked valve each re-fueling outage such that all valves are checked in three successive refueling outages, to insure set points are as follows:
Number of Val'es Set Point (psig) v l
4 1212 2 12 4
1221 12
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4 1230 = 12 4
1239
- 12 E.
A sa=ple of reactor coolant shall be analyzed at least every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for the purpcse of determining the cen-tent of chloride ion and to check the cenductivity.
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o.3-la F.
Primary Coolant System Pressure Isolation Valves Speci fi cation:-
1.
Per odic leakagt testing (a) on each valve listed in table
4.3.2.shall be accomplished prior to exceeding 600 psig reactor pressure every time the plant is placed on the cold shutdewn condition for refueling, each time the plant is placed in a cold 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, and prior to returning the valve to service after maintenance, repair or replacehent work is performed.
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/2 (a)To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indi-cators) 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.
.C 3rcer dcted April 20. 1931 t
4.3-2 o
Basis Numerous data are available relating integrated flux and the change in Nil-Ductility Transition Temperature (NDTT) in various steals.
The base metal has been demonstrated to be relatively insensitive to neutron 1 :adiation (see expected NDT changes in FDSAK Table IV-1-1, and Figures IV-2-9 and IV-2-10).
The most conservative data has been used in i
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Specification 3.3.
The integrated flux at the vessel wall is' calculated from core physics data and will be measured using flux monitors installed inside the vessel.
The measurements of the neutron flux at the vessel vall will i
be used to check and if necessary correct, the calculated data to determine an accurate flux.
From this a conservative j
NDT temperature can be determined.
Since no shift will occur until an integrated flux of 1017 nvc is reached, the confirmation can be made loag before an NDIT shift would occur.
Prior to operation the' reactor coolant system will be free '
of gross defects and the facility has been designed such that gross defects should not occur throughout life; however,
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to determine the status of the coolant system to ensure that gross defects are not developing this surveillance program was developed.. This insp.ection vill reveal problem areas should they occ'ur before a leak develops.
In addition, extensive visual inspection for leaks will be made on criti-cal systems.
The inspection period is based on the observed rate of growth of defects from fatigue studies sponsored by the AEC.
These. studies show that it requires thousands of stress cycles, at stresses beyond any conceived in a i
reactor system to propagate a crack and it is thus concluded that the frequency is adequate.
The access provisions for in-service inspection has been compared v3,th the access require-
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ments.of the proposed N-45 Code for In-Service Inspection of Nuclear Reactor Coolant Systems.. The degree of access required by N-45 is not generally available, however, solu-j I
l metric inspection of accessable areas has been proposed.
It 1s considered appropriate to evaluate the results obtained from compliance with this Technical Specification and the state of the art before establishing a long term inspection program.
j Experience in safety valve operation shows that a check of cpproximately 1/3 of the safety valves per year is adequate i
to detect failures or deterioration.
The tolerance value is specified in Section 1 of the ASME Code at 21% of design pressure.
An analysis has been performed which shows that with all safety valves set 12 psig higher the safety limit of 1375 psig is not exceeded.
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4.3-9 TABLE 4.3.2 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES
- s MaxTmum(*)
Systen Valve No.
Allowable Leal: age Core Spray System 1 NZ02A 5.0 GPM NZO2C 5.0 GPM Coca Spray System 2 NZ02B 5.0 GPM NZO2D 5.0 GPM i
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I Footnote:
(a j Leakage rates less than or equal to 1.0 gpm are considered acceptuble.
l.
2.
' Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm.are considered acceptable if the latest reasured rate has not exceeded the rate determined by 'the previous test by an amount that reduces the margin between neasured leakage rate and.the maximum permissible rate of 5.0 gpm by 50% 'or greater.
I Leakage rates greater than 1.0 gpm but less than or equal.to 5.0 gpm 3.
are considered unacceptable if the latest measured rate exceeded the i
rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
4.
Leakage rates greater than 5.Q gpm 'are considered unacceptable.
5.
Test differential. pressure shall not be less than 150 psid.
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..: Order dated April 20, 1981 f
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