ML20197H038
| ML20197H038 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/08/1984 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20197H028 | List: |
| References | |
| NUDOCS 8406180254 | |
| Download: ML20197H038 (7) | |
Text
o 4.3-1
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4.3 REACTOR COOLANT l
Applicability: Applies to the surveillance requirements for the reactor coolant system l
Objective:
To determine the condition of the reactor plant coolant f
system and the oe< *ation of the safety devices related to it.
I Specification: A. Neutron flux monitors shall be installed in the reactor vessel adjacent to the vessel wall at the core midplane level.
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 NOTT from Figure 3.3.1.
B. Inservice inspection of ASME Code Class 1, Class 2 and Class 3 systems and components shall be perfonned in accordance with 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 the NRC pursuant to 10 CFR, Section 50.55a(g)(6)(1).
1 C. Inservice testing of ASME Code Class 1. Class 2 and Class 3 pumps and valves shall be perfonned in aCCordance with Section XI of the ASME Boiler and l
Pressure Vessel Code and applicable Addenda as required 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)(1).
l D. A visual examination for leaks shall be made with the reactor coolant system at pressure during each scheduled l
refueling outage or after major repairs have been made t
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to the reactor coolant system in accordance with Article j
5000,Section XI.
The requirements of specification 3.3. A shall be met during the test.
E. Each replacement safety valve or valve'that has been repaired shall be bench checked for proper set point. A i
minimum of 5 of the valves shall be bench checked or replaced with a bench checked valve each refueling outage such that all valves are checked in three successive refueling outages, to ensure set points are as follows:
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Number of Valves Set Point (psig) 4 1212 + 12 i
4 1221 7 12 4
1230 I 12 4
1239 T 12 F. A sample 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 purpose of determining the t
i content of the chloride ion and to check the conductivity.
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8406180254 840608 DR ADOCK 05000219
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4.3-2
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Bases:
Numerous data are available relating integrated flux and the change in Nil-Ductility Transition Temperature (NDTT) in various steels.
The base j
metal has been demonstrated to be relatively insensitive to neutron i
irradiation (see Expected NDT changes in FDSAR Table IV-1-1, and Figures i
IV-2-9 and IV-2-10).
The most conservative data has been used in i
Specification 3.3.
The integrated flux at the vessel wall is calculated l
from core physics data and will be measured using flux monitors installed inside the vessel. The measurements of the neutron flux at the vessel l
wall will be used to check and if necessary correct, the calculated data to determine an accurate flux. From this a conservative NDT temperature determined. Since no shift will occur until an integrated flux of can pnyt is reached, the confimation can be made long before an NOTT 10 l shift would occur.
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The inspection program will reveal problem areas should they occur, before a leak develops.
In addition, extensive visual inspection for leakr, will be made on critical systems. Oyster Creek was designed and constructed prior to the existence of ASME Section XI. For this reason, the degree of i
access required by ASME Section XI is not generally available and will be addressed as " requests for relief" in accordance with 10 CFR 50.55a(g).
i Experience in safety valve operation shows that a check of approximately
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1/3 of the safety valves per year is adequate to detect failures or l
l deterioration.
The tolerance value is specified in Section I of the ASME Code at +1% of design pressure. An analysis has been performed which shows that with all safety. valves set 12 psic higher the safety limit of i
1375 psig is not exceeded.
i Conductivity instruments continuously monitor the reactor coolant.
l Experience indicates that check of conductivity instrumentation at least i
every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is adequate to ensure accurate readings.
The reactor water sample will also be uSed to determine the chloride ion content to assure that the limits of 3.3.E are not exceeded.
The chloride ion content will not change rapidly over a period of several days; therefore, the sampling i
frequency is adequate.
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4.3-3 l
F.
Primary Coolant System Pressure Isolation Valves Specification:
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1.
Periodic leakage testing (a) on eacn valve listed in table 4.3.1 shall be accomplished l
prior to exceeding 600 psig reactor pressure every time the plant is placed on the cold shutdown condition for refueling, each time the plant is placed in a cold shutdown j
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 i
replacemant work is performed.
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I (a)
To 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, j
l NRC Order dated April 20, 1981 b
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4.3-4 l
TABLE 4.3.1 l
PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum (a)
System Valve No.
Allowable Leakage Core Spray System 1 NZO2A 5.0 GPM NZ02C 5.0 GPM Core Spray System 2 NZ028 5.0 GPM NZ020 5.0 GPM i
Footnote:
(a) 1.
Leakage rates less than or equal to 1.0 gpm are considered acceptable.
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 exceeded the rate determined by the previous test by an amount that reduces t.he margin between measured leakage rate and
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the maximum permissible rate of 5.0 gpm by 50% or greater.
3.
Leakage rates greater than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate i
detennined by the previous test by an amount that reduces the j
margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
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4.
Leakage rates greater than 5.0 gpm are considered unacceptable.
5.
Test differential pressure shall not be less than 150 psid.
NRC Order dated April 20, 1981 i
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'T OYSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE N0. DPR-16 DOCKET NO. 50-219 TECHNICAL S.'ECIFICATION CHANGE REQUEST N0.123 Pursuant to 10 CFR 50.91 an analysis concerning significant hazards considerations is provided below:
1)
Section to be changed:
Administrative cnanges indicated below are applicable to Section 4.3 of Appendix A.
2)
Extent of cham:
Paragraphs 4.3.B and 4.3.C have been changed to reflect the Commission's recommended wording for the conduct of inservice inspection and inservice testing. The bases have also been changed accordingly. Table 4.3.1 has been removed since it has been superceded by the Oyster Creek Inservice Inspection Program for the second 10-year interval.
The existing table 4.3.2 has been renumbered 4.3.1 along with the references to it.
3)
Discussion:
Examples of amendments that are considered not likely to involve significant hazards considerations were provided in the Federal Register on April 6,1983 (48FR 14870). Technical Specification Change Request No.123 meets the provisions of example. (i) (as referenced above) in that it is an administrative change.
Copies of modified pages 4.3-1, 4.3-2, 4.3-3, and 4.3-4 are being provided for clarification purposes and the convenience of the Conmiission.
The Commission has recommended the wording employed in modified paragraphs 4.3.8 and 4.3.C and the elimination of program details elsewhere in the text.
This change will make clearer the inservice testing and inspection requirements as described in 10 CFR 50.55a and Section XI of the ASME Boiler and Pressure Yessel Code, the details of which are implemented by the inservice inspection and testing programs for Oyster Creek. The proposed text will eliminate the need for frequent changes to the Technical Specifications whenever l
program changes or requests for relief are made.
4)
Determination:
We have detennined that the subject change request involves no significant hazards in that operation of the Oyster Creek' Nuclear Generating Station in accordance with Technical l
Specification Change Request No.123 would not:
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1)
Involve a significant increase in the probability or consequences of an accident previously evaluated; or 2)
Create the possibility of a new or different kind of accident from any previously evaluated; or 3)
Involve a significant reduction in a margin of safety.
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OYSTER CREEK NUCLEAR GENERATING STATION (00CKET NO. 50-219)
PROVISIONAL OPERATING LICENSE NO. OPR-16 i
L Applicant hereby requests the Commission to reissue Appendix A to the above captioned license:
t 1)
Section to be changed:
Administrative changes indicated below are applicable to Appendix A.
2)
Extent of change:
Remove detailed ISI and IST requirements of Section 4.3 and replace with Commission's recommended wording.
3)
Changes requested:
As indicated on modified pages 4.3-1, 4.3-2, 4.3-3 and 4.3-4.
Delete existing table 4.3-1, renumber existing table 4.3.2, "4.3.1" and appropriately change / remove references to the two tables within the text.
4)
Discussion:
This Technical Specification Change Request modifies Section 4.3 by deleting detailed ISI and IST requirements and replacing them with general references to 10 CFR 50.55a and ASME Boiler and Pressure Vessel Code,Section XI.
This is in response to a Commission recommendation and will reduce the complexity of l
administering the programs.
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