RA-10-065, Response to Draft Request for Additional Information License Amendment Request Regarding Relocation of Selected Technical Specification Surveillance Frequencies to a Licensee Controlled Document

Response to Draft Request for Additional Information License Amendment Request Regarding Relocation of Selected Technical Specification Surveillance Frequencies to a Licensee Controlled Document

ML102430467
Person / Time
Site:	Oyster Creek
Issue date:	08/31/2010
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-10-065
Download: ML102430467 (43)

Text

Exekin.

x Ion Nm cii www exeloncorp orn 200 Exelon Way Nuciear Kennett Square, PA 19348 10 CFR 50.90 RA-10-065 August 31, 2010 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Response to Draft Request for Additional Information License Amendment Request Regarding Relocation of Selected Technical Specification Surveillance Frequencies to a Licensee Controlled Document

References:

1.

Letter from P. B. Cowan, Exelon Generation Company, LLC, to U.S. Nuclear Regulatory Commission, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3), dated October 30, 2009.

2.

Electronic transmission from G. Edward Miller, U.S. Nuclear Regulatory Commission, to Glenn Stewart, Exelon Generation Company, LLC, Oyster Creek Nuclear Generating Station

- Electronic Transmission, Draft Request for Additional Information Regarding License Amendment Request to Relocate Surveillance Frequencies to a Licensee Controlled Document (TAC No. ME2494), dated August 18, 2010.

In Reference 1, Exelon Generation Company, LLC (Exelon) submitted a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-1 6 for Oyster Creek Nuclear Generating Station (OCNGS). The proposed amendment would modify OCNGS TS by relocating selected Surveillance Requirement frequencies to a licensee-controlled document in accordance with TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control

- RITSTF [Risk-Informed Technical Specifications Task Force] Initiative 5b, dated March 18, 2009. The NRC reviewed the license amendment request and identified the need for additional information in order to complete their Exelon Nuclear 200 Exelon Way Kennett Square, PA 19348 RA-10-065 August 31, 2010 www.exeloncorp.com Nuclear 10 CFR 50.90 u.s. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 NRC Docket No. 50-219

References:

Response to Draft Request for Additional Information License Amendment Request Regarding Relocation of Selected Technical Specification Surveillance Frequencies to a Licensee Controlled Document 1.

Letter from P. B. Cowan, Exelon Generation Company, LLC, to U.S. Nuclear Regulatory Commission, "Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)," dated October 30, 2009.

2.

Electronic transmission from G. Edward Miller, U.S. Nuclear Regulatory Commission, to Glenn Stewart, Exelon Generation Company, LLC, "0yster Creek Nuclear Generating Station - Electronic Transmission, Draft Request for Additional Information Regarding License Amendment Request to Relocate Surveillance Frequencies to a Licensee Controlled Document (TAC No. ME2494)," dated August 18, 2010.

In Reference 1, Exelon Generation Company, LLC (Exelon) submitted a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-16 for Oyster Creek Nuclear Generating Station (OCNGS). The proposed amendment would modify OCNGS TS by relocating selected Surveillance Requirement frequencies to a licensee-controlled document in accordance with TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk-Informed Technical Specifications Task Force] Initiative 5b," dated March 18,2009. The NRC reviewed the license amendment request and identified the need for additional information in order to complete their

Response to Draft Request for Additional Information LAR Regarding Relocation of Selected TS Surveillance Frequencies to a Licensee Controlled Document Docket No. 50-219 August 31, 2010 Page 2 evaluation of the amendment request. A draft request for additional information (RAI) was electronically transmitted to Exelon on August 18, 2010 (Reference 2). Attachment 1 to this letter provides a restatement of the RAI along with Exelons response. Attachment 2 provides revised TS/Bases markups in response to the RAI.

In addition, TSTF-425, Revision 3, provided an optional insert to existing TS Bases to facilitate adoption of the TSTF traveler. The TSTF-425 TS Bases insert states as follows:

The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

Recently, several licensees submitting license amendment requests (LARs) for adoption of TSTF-425 have identified a need to deviate from this statement because it only applies to Surveillance Frequencies that have been changed in accordance with the Surveillance Frequency Control Program (SFCP) and does not apply to Surveillance Frequencies that are relocated to the SFCP but not changed. For Surveillance Frequencies relocated to the SFCP but not changed, the existing TS Bases description provides a valid description of the bases for the unchanged Surveillance Frequencies.

Therefore, upon implementation of the proposed change, where appropriate, the existing TS Bases information describing the bases for the Surveillance Frequencies will be relocated to the SFCP. This will ensure that the information describing the bases for unchanged Surveillance Frequencies is maintained.

Also, relative to the Bases insert, Exelon proposes to replace the TSTF-425 Bases insert specified above with a revised insert that reads The Surveillance Frequency(ies) is(are) controlled under the Surveillance Frequency Control Program, or otherwise modify the TS Bases, as appropriate, as indicated on revised proposed TS/Bases pages provided in.

Additionally, during the review of the TSTF-425 TS Bases issue, Exelon identified two additional proposed changes to the Bases for TS Sections 4.5 and 4.9, as indicated on TS pages 4.5-15 and 4.9-2, respectively, which are provided in Attachment 2 to this letter. These proposed changes are consistent with other proposed Bases changes in the original submittal (Reference

1) and as described above.

Also, the proposed notes specified on TS pages 4.1-10, 4.2-2, 4.4-2, 4.12-2, and 4.15-2 were revised for consistency with each other and with similar notes added to other TS pages included in the TS markups. The revised proposed TS pages are provided in Attachment 2 to this letter.

It should be noted that the revised proposed TS/Bases pages provided in Attachment 2 to this letter are specific to the issues identified in this RAI response, and supersede only the corresponding TS/Bases pages provided in the original submittal (Reference 1). As a result, the pages provided are a subset of the total TS/Bases pages provided in the original submittal and are not meant to be a complete replacement set of proposed TS/Bases pages. Therefore, the Response to Draft Request for Additional Information LAR Regarding Relocation of Selected TS Surveillance Frequencies to a Licensee Controlled Document Docket No. 50-219 August 31, 2010 Page 2 evaluation of the amendment request. A draft request for additional information (RAI) was electronically transmitted to Exelon on August 18, 2010 (Reference 2). Attachment 1 to this letter provides a restatement of the RAI along with Exelon's response. Attachment 2 provides revised TS/Bases markups in response to the RAI.

In addition, TSTF-425, Revision 3, provided an optional insert to existing TS Bases to facilitate adoption of the TSTF traveler. The TSTF-425 TS Bases insert states as follows:

liThe Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program. 1I Recently, several licensees submitting license amendment requests (LARs) for adoption of TSTF-425 have identified a need to deviate from this statement because it only applies to Surveillance Frequencies that have been changed in accordance with the Surveillance Frequency Control Program (SFCP) and does not apply to Surveillance Frequencies that are relocated to the SFCP but not changed. For Surveillance Frequencies relocated to the SFCP but not changed, the existing TS Bases description provides a valid description of the bases for the unchanged Surveillance Frequencies.

Therefore, upon implementation of the proposed change, where appropriate, the existing TS Bases information describing the bases for the Surveillance Frequencies will be relocated to the SFCP. This will ensure that the information describing the bases for unchanged Surveillance Frequencies is maintained.

Also, relative to the Bases insert, Exelon proposes to replace the TSTF-425 Bases insert specified above with a revised insert that reads 'The Surveillance Frequency(ies) is(are) controlled under the Surveillance Frequency Control Program,1I or otherwise modify the TS Bases, as appropriate, as indicated on revised proposed TS/Bases pages provided in.

Additionally, during the review of the TSTF-425 TS Bases issue, Exelon identified two additional proposed changes to the Bases for TS Sections 4.5 and 4.9, as indicated on TS pages 4.5-15 and 4.9-2, respectively, which are provided in Attachment 2 to this letter. These proposed changes are consistent with other proposed Bases changes in the original submittal (Reference

1) and as described above.

Also, the proposed notes specified on TS pages 4.1-10, 4.2-2, 4.4-2, 4.12-2, and 4.15-2 were revised for consistency with each other and with similar notes added to other TS pages included in the TS markups. The revised proposed TS pages are provided in Attachment 2 to this letter.

It should be noted that the revised proposed TS/Bases pages provided in Attachment 2 to this letter are specific to the issues identified in this RAI response, and supersede only the corresponding TS/Bases pages provided in the original submittal (Reference 1). As a result, the pages provided are a subset of the total TS/Bases pages provided in the original submittal and are not meant to be a complete replacement set of proposed TS/Bases pages. Therefore, the

Response to Draft Request for Additional Information LAR Regarding Relocation of Selected TS Surveillance Frequencies to a Licensee Controlled Document Docket No. 50-219 August 31, 2010 Page 3 TS/Bases pages provided in the Reference 1 submittal that are not included in Attachment 2 to this letter are still valid and remain part of the requested license amendment.

Exelon has concluded that the information provided in this response does not impact the conclusions provided in the original submittal (Reference 1).

This response to the request for additional information contains no regulatory commitments.

If you have any questions or require additional information, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 31 st day of August 2010.

Respectfully, Pamela B. Cowan Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC :

Response to Draft Request for Additional Information :

Revised Proposed Technical Specifications/Bases Pages cc:

Regional Administrator

- NRC Region I w/attachments NRC Senior Resident Inspector

- OCNGS NRC Project Manager, NRR

- OCNGS Director, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection Mayor of Lacey Township, Forked River, New Jersey Response to Draft Request for Additional Information LAR Regarding Relocation of Selected TS Surveillance Frequencies to a Licensee Controlled Document Docket No. 50-219 August 31, 2010 Page 3 TS/Bases pages provided in the Reference 1 submittal that are not included in Attachment 2 to this letter are still valid and remain part of the requested license amendment.

Exelon has concluded that the information provided in this response does not impact the conclusions provided in the original submittal (Reference 1).

This response to the request for additional information contains no regulatory commitments.

If you have any questions or require additional information, please contact Glenn Stewart at 610-765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 31 st day of August 2010.

Respectfully, d~aJ1I Pamela B. Cowan Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC : :

Response to Draft Request for Additional Information Revised Proposed Technical Specifications/Bases Pages cc:

Regional Administrator - NRC Region I NRC Senior Resident Inspector - OCNGS NRC Project Manager, NRR - OCNGS Director, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection Mayor of Lacey Township, Forked River, New Jersey w/attachments

ATTACHMENT 1 License Amendment Request Oyster Creek Nuclear Generating Station Docket No.. 50-219 License Amendment Request Regarding Relocation of Selected Technical Specification Surveillance Frequencies to a Licensee Controlled Document Response to Draft Request for Additional Information ATTACHMENT 1 License Amendment Request Oyster Creek Nuclear Generating Station Docket No. 50-219 License Amendment Request Regarding Relocation of Selected Technical Specification Surveillance Frequencies to a Licensee Controlled Document Response to Draft Request for Additional Information

Docket No. 50-219 Page 1 of 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST REGARDING RELOCATION OF SELECTED TECHNICAL SPECIFICATION SURVEILLANCE FREQUENCIES TO A LICENSEE CONTROLLED DOCUMENT In Reference 1, Exelon Generation Company, LLC (Exelon) submitted a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-16 for Oyster Creek Nuclear Generating Station (OCNGS). The proposed amendment would modify OCNGS TS by relocating selected Surveillance Requirement frequencies to a licensee-controlled document in accordance with TSTF-425, Revision 3, Relocate Surveillance Frequencies to Licensee Control

- RITSTF [Risk-Informed Technical Specifications Task Force] Initiative 5b, dated March 18, 2009. The NRC reviewed the license amendment request and identified the need for additional information in order to complete their evaluation of the amendment request. A draft request for additional information (RAI) was electronically transmitted to Exelon on August 18, 2010 (Reference 2). The question is restated below along with Exelons response.

RAI In a number of locations, the proposed wording would require that a SR [surveillance requirement] be performed in accordance with the Surveillance Frequency control Program.

Provide additional justification that this wording will not allow aspects of performing the SR other than the frequency of performance from being governed by the Surveillance Frequency Control Program. Alternately, propose revised wording that limits the control of the Surveillance Frequency Control Program to only the frequency of performing an SR.

RESPONSE

To make it clear that the Surveillance Frequency Control Program only controls changes to the frequency of surveillances that are included in the program, the proposed wording in accordance with the Surveillance Frequency Control Program inserted in the TS and Bases markups has been replaced with the wording at the frequency specified in the Surveillance Frequency Control Program as indicated in the revised proposed TS/Bases pages provided in.

REFERENCES:

1.

Letter from P. B. Cowan, Exelon Generation Company, LLC, to U.S. Nuclear Regulatory Commission, Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3), dated October 30, 2009.

2.

Electronic transmission from G. Edward Miller, U.S. Nuclear Regulatory Commission, to Glenn Stewart, Exelon Generation Company, LLC, Oyster Creek Nuclear Generating Station

- Electronic Transmission, Draft Request for Additional Information Regarding License Amendment Request to Relocate Surveillance Frequencies to a Licensee Controlled Document (TAC No. ME2494), dated August 18, 2010.

Docket No. 50-219 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST REGARDING RELOCATION OF SELECTED TECHNICAL SPECIFICATION SURVEILLANCE FREQUENCIES TO A LICENSEE CONTROLLED DOCUMENT Page 1 of 1 In Reference 1, Exelon Generation Company, LLC (Exelon) submitted a request for an amendment to the Technical Specifications (TS), Appendix A of Renewed Facility Operating License No. DPR-16 for Oyster Creek Nuclear Generating Station (OCNGS). The proposed amendment would modify OCNGS TS by relocating selected Surveillance Requirement frequencies to a licensee-controlled document in accordance with TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk-Informed Technical Specifications Task Force] Initiative 5b," dated March 18,2009. The NRC reviewed the license amendment request and identified the need for additional information in order to complete their evaluation of the amendment request. A draft request for additional information (RAI) was electronically transmitted to Exelon on August 18, 2010 (Reference 2). The question is restated below along with Exelon's response.

RAI In a number of locations, the proposed wording would require that a SR [surveillance requirement] be performed "in accordance with the Surveillance Frequency control Program."

Provide additional justification that this wording will not allow aspects of performing the SR other than the frequency of performance from being governed by the Surveillance Frequency Control Program. Alternately, propose revised wording that limits the control of the Surveillance Frequency Control Program to only the frequency of performing an SR.

RESPONSE

To make it clear that the Surveillance Frequency Control Program only controls changes to the frequency of surveillances that are included in the program, the proposed wording "in accordance with the Surveillance Frequency Control Program" inserted in the TS and Bases markups has been replaced with the wording "at the frequency specified in the Surveillance Frequency Control Program" as indicated in the revised proposed TS/Bases pages provided in.

REFERENCES:

1.

Letter from P. B. Cowan, Exelon Generation Company, LLC, to U.S. Nuclear Regulatory Commission, "Application for Technical Specifications Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3), II dated October 30, 2009.

2.

Electronic transmission from G. Edward Miller, U.S. Nuclear Regulatory Commission, to Glenn Stewart, Exelon Generation Company, LLC, "0yster Creek Nuclear Generating Station - Electronic Transmission, Draft Request for Additional Information Regarding License Amendment Request to Relocate Surveillance Frequencies to a Licensee Controlled Document (TAC No. ME2494)," dated August 18, 2010.

ATTACHMENT 2 License Amendment Request Oyster Creek Nuclear Generating Station Docket No. 50-219 Draft Request for Additional Information License Amendment Request Regarding Relocation of Selected Technical Specification Surveillance Frequencies to a Licensee Controlled Document Revised Proposed Technical Specifications/Bases Pages 41-2 4.3-1 4.5-6 4.7-1 4.9-1 4.1-3 4.3-2 4.5-9 4.7-2 4.92*

4.1-8 4.3-3 4.5-11 4.7-3 4.10-1 4.1-9 4.4-2 4.5-12 4.7-4 4.12-1 4.1-10 4.5-2 4.5-13 4.7-5 4.12-2 4.2-1 4.5-3 4.515*

4.8-1 4.15-2 4.2-2 4.5-4 4.6-1 4.8-2 4.17-1 4.2-4 4.5-5

• New proposed TS Bases page ATTACHMENT 2 License Amendment Request Oyster Creek Nuclear Generating Station Docket No. 50-219 Draft Request for Additional Information License Amendment Request Regarding Relocation of Selected Technical Specification Surveillance Frequencies to a Licensee Controlled Document Revised Proposed Technical Specifications/Bases Pages 4.1-2 4.1-3 4.1-8 4.1-9 4.1-10 4.2-1 4.2-2 4.2-4 4.3-1 4.3-2 4.3-3 4.4-2 4.5-2 4.5-3 4.5-4 4.5-5 4.5-6 4.5-9 4.5-11 4.5-12 4.5-13 4.5-15*

4.6-1 4.7-1 4.7-2 4.7-3 4.7-4 4.7-5 4.8-1 4.8-2 4.9-1 4.9-2*

4.10-1 4.12-1 4.12-2 4.15-2 4.17-1

• New proposed TS Bases page

4.1 PROTECTIVE INSTRUMENTATION Bases:

Surveillance intervals are based on reliability analyses and have been determined in accordance with General Electric Licensing Topical Reports given in References 1 through 5.Surveillance Frequencies are controlled under the Surveillance Frequency Control Program (SFCP).

The functions listed in Table 4.1.1 logically divide into three groups:

a.

On-off sensors that provide a scram function or some other equally important function.

b.

Analog devices coupled with a bi-stable trip that provides a scram function or some other vitally important function.

c.

Devices which only serve a useful function during some restricted mode of operation, such as startup or shutdown, or for which the only practical test is one that can be performed only at shutdown.

Group (b) devices utilize an analog sensor followed by an amplifier and bi-stable trip circuit. The sensor and amplifier are active components and a failure would generally result in an upscale signal, a downscale signal, or no signal. These conditions are alarmed so a failure would not go undetected. The bi-stable portion does need to be tested in order to prove that it will assume its tripped state when required.

Group (c) devices are active only during a given portion of the operational cycle. For example, the IRM is inactive during full-power operation and active during startup. Thus, the only test that is significant is the one performed just prior to shutdown and startup. The condenser Low Vacuum trip can only be tested during shutdown, and although it is connected into the reactor protection system, it is not required to protect the reactor. Testing at each REFUELING OUTAGEat the frequency specified in the Surveillance Frequency Control Program is adequate. The switches for the high temperature main steamline tunnel are not accessible during normal operation because of their location above the main steam lines. Therefore, after initial calibration and in-place OPERABILITY checks, they will not be tested between refueling shutdowns. Considering the physical arrangement of the piping which would allow a steam leak at any of the four sensing locations to affect the other locations, it is considered that the function is not jeopardized by limiting calibration and testing to refueling outages.

The CHANNEL FUNCTIONAL TEST verifies instrument channel operability. A successful test of the required contact(s) of a channel relay may be performed by the verification of a change in state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST or CHANNEL CALIBRATION of a relay. This is acceptable because all the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specification tests.

The logic of the instrument safety systems in Table 4.1.1 is such that testing the instrument channels trips the trip system to verify that it is OPERABLE. The testing may be performed by means of any series of sequential, overlapping, or total channel steps. However, certain systems require coincident instrument channel trips to completely test their trip systems. Therefore, SFCP Table 4.1.2 specifies the minimum trip system test frequency for these tripped systems. This assures that all trip systems for protective instrumentation are adequately tested, from sensors through the trip system.

OYSTER CREEK 4.1-2 Amendment No.: 171, 208, 263 4.1 PROTECTIVE INSTRUMENTATION Bases:

Surveillance intervals are based on reliability analyses and have been determined in accordance with General Electric Licensing Topical Reports given in References 1 through 5.Surveillance Frequencies are controlled under the Surveillance Frequency Control Program (SFCP).

The functions listed in Table 4.1.1 logically divide into three groups:

a.

On-off sensors that provide a scram function or some other equally important function.

b.

Analog devices coupled with a bi-stable trip that provides a scram function or some other vitally important function.

c.

Devices which only serve a useful function during some restricted mode of operation, such as startup or shutdown, or for which the only practical test is one that can be performed only at shutdown.

Group (b) devices utilize an analog sensor followed by an amplifier and bi-stable trip circuit. The sensor and amplifier are active components and a failure would generally result in an upscale signal, a downscale signal, or no signal. These conditions are alarmed so a failure would not go undetected. The bi-stable portion does need to be tested in order to prove that it will assume its tripped state when required.

Group (c) devices are active only during a given portion of the operational cycle. For example, the IRM is inactive during full-power operation and active during startup. Thus, the only test that is significant is the one performed just prior to shutdown and startup. The condenser Low Vacuum trip can only be tested during shutdown, and although it is connected into the reactor protection system, it is not required to protect the reactor. Testing at each REFUELING QUTAGEat the frequency specified in the Surveillance Frequency Control Program is adequate. The switches for the high temperature main steamline tunnel are not accessible during normal operation because of their location above the main steam lines. Therefore, after initial calibration and in-place OPERABILITY checks, they will not be tested between refueling shutdowns. Considering the physical arrangement of the piping which would allow a steam leak at any of the four sensing locations to affect the other locations, it is considered that the function is not jeopardized by limiting calibration and testing to refueling outages.

The CHANNEL FUNCTIONAL TEST verifies instrument channel operability. A successful test of the required contact(s) of a channel relay may be performed by the verification of a change in state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST or CHANNEL CALIBRATION of a relay. This is acceptable because all the other required contacts of the relay are verified by other Technical Specifications and non-Technical Specification tests.

The logic of the instrument safety systems in Table 4.1.1 is such that testing the instrument channels trips the trip system to verify that it is OPERABLE. The testing may be performed by means of any series of sequential, overlapping, or total channel steps. However, certain systems require coincident instrument channel trips to completely test their trip systems. Therefore, SFCP Table 4.1.2 specifies the minimum trip system test frequency for these tripped systems. This assures that all trip systems for protective instrumentation are adequately tested, from sensors through the trip system.

OYSTER CREEK Amendment No.: 171, 208, 26d 4.1-2

IRM calibration is to be performed during reactor startup. The calibration of the IRMs during startup will be significant since the IRMs will be relied on for neutron monitoring and reactor protection up to 38.4% of rated power during a reactor startup.

To ensure that the APRMs are accurately indicating the true core average power, the APRM5 are calibrated to the reactor power calculated from a heat balance. Limiting Safety System Settings (LSSS) 2.3.A.1 allows the APRMs to be reading greater than actual THERMAL POWER to compensate for localized power peaking. When this adjustment is made, the requirement for the absolute difference between the APRM channels and the calculated power to indicate within 2%

RTP is modified to include any gain adjustments required by LSSS 2.3.A.1.

LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 MWD/T Surveillance Frequency is controlled under the Surveillance Frequency Control Program is based on operating experience with LPRM sensitivity changes.

General Electric Licensing Topical Report NEDC-30851P-A (Reference 1), Section 5.7 indicates that the major contributor to reactor protection system unavailability is common cause failure of the automatic scram contactors. Analysis showed a weekly test interval to be optimum for Scram contactors are tested at the frequency specified in the Surveillance Frequency Control Program. The test of the automatic scram contactors can be performed as part of the CHANNEL CALIBRATION or CHANNEL FUNCTIONAL TEST of Scram Functions or by use of the subchannel test switches.

References:

(1)

NEDC-30851 P-A, Technical Specification Improvement Analyses for BWR Reactor Protection System.

(2)

NEDC-30936P-A, BWR Owners Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation), Parts 1 and 2.

(3)

NEDC-30851P-A, Supplement 1, Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation.

(4)

NEDC-30851P-A, Supplement 2, Technical Specification Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation.

(5)

NEDC-31 677P-A, Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation.

OYSTER CREEK 4.1-3 AMENDMENT NO.: 71,171, 208, 263, 266 IRM calibration is to be performed during reactor startup. The calibration of the IRMs during startup will be significant since the IRMs will be relied on for neutron monitoring and reactor protection up to 38.4%

of rated power during a reactor startup.

To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. Limiting Safety System Settings (LSSS) 2.3.A.1 allows the APRMs to be reading greater than actual THERMAL POWER to compensate for localized power peaking. When this adjustment is made, the requirement for the absolute difference between the APRM channels and the calculated power to indicate within 2°,fo RTP is modified to include any gain adjustments required by LSSS 2.3.A.1.

LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative input to the APRM System. The 1000 MVVD/T Surveillance Frequency is controlled under the Surveillance Frequency Control Program is based on operating experience \\\\'ith LPRM sensitivity changes.

General Electric Licensing Topical Report NEDC-30851 P-A (Reference 1), Section 5.7 indicates that the major contributor to reactor protection system unavailability is common cause failure of the automatic scram contactors. Analysis showed a weekly test interval to be optimum for Scram contactors are tested at the frequency specified in the Surveillance Frequency Control Program. The test of the automatic scram contactors can be performed as part of the CHANNEL CALIBRATION or CHANNEL FUNCTIONAL TEST of Scram Functions or by use of the subchannel test switches.

References:

(1)

NEDC-30851 P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System."

(2)

NEDC-30936P-A, "BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)," Parts 1 and 2.

(3)

NEDC-30851 P-A, Supplement 1, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation."

(4)

NEDC-30851 P-A, Supplement 2, "Technical Specification Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation."

(5)

NEDC-31677P-A, "Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation."

OYSTER CREEK 4.1-3 AMENDMENT NO.: 71,171,208,263,266

TABLE 4.1.1 Page 5 of 6 MINIMUM CHECK, CALIBRATION AND TEST FREQUENCY FOR PROTECTIVE INSTRUMENTATION Instrument Channel Check (Note 5)

Calibrate (Note 5)

Test (Note 5)

Remarks (Applies to Test & Calibration) 32.

LPRM Level a)

Electronics N/A 1/l2mo.

1/12mg.

b)

Detectors N/A Note 4 N/A 33.

RWCU HELB High Temperature N/A Each refuoling 1/3 mo.

Perform Channel Tests using the test outage switches.

Calibrate prior to startup and normal shutdown and thereafter check 4/s-and test 4/wic at the frequency specified in the Surveillance Frequency Control Program until no longer required.

Legend:

N/A = Not Applicable 1/s Once per shift 1/d Once per day 1/3d Once per 3 days I Iwk Once ner week II(IIU.

LIlIl..e pei IIIUIILII 1/3 mo.

Once every throo months 1/12 mo.

Once every 12 months 1/24 mo.

Once every 24 months OYSTER CREEK 4.1-8 Change: 7, Amendment No.: 171, 208, 259 TABLE 4.1.1 Page 50f6 MINIMUM CHECK, CALIBRATION AND TEST FREQUENCY FOR PROTECTIVE INSTRUMENTATION Instrument Channel 32.

LPRM Level Check (Note 5)

Calibrate (Note 5)

Test (Note 5)

Remarks (Applies to Test & Calibration) 33.

a)

Electronics b)

Detectors RWCU HELB High Temperature N/A N/A N/A

~

Note 4 Eash refl::Jeling

~

~

N/A

~

Perform Channel Tests using the test switches.

Legend:

Calibrate prior to startup and normal shutdown and thereafter check -+fs-and test 4fwk at the frequency specified in the Surveillance Frequency Control Program until no longer required.

N/A = Not Applicable 1/s Gnse per shift 1/d Gnse per day 1/dd Gnse per d days 1/wk Gnse per 'Neek 1/A10.

Gnse per A10nth 1/d A10.

Gnse every three A10nths 1/12 A10.

Gnse every 12 A10nths 1/24 A10.

Gnse every 24 A10nths OYSTER CREEK Change: 7, Amendment No.: 171,208,259 4.1-8

TABLE 4.1.1 Page 6 of 6 MINIMUM CHECK, CALIBRATION AND TEST FREQUENCY FOR PROTECTIVE INSTRUMENTATION NOTE 1:

Each automatic scram contactor is required to be tested at least once per week at the frequency specified in the Surveillance Frequency Control Program. When not tested by other means, the weekly test can be performed by using the subchannel test switches.

NOTE 2:

At least daily during reactor PO\\VER OPELTION At the frequency specified in the Surveillance Frequency Control Program, the reactor neutron flux peaking factor shall be estimated and flow-referenced APRM scram and rod block settings shall be adjusted, if necessary, as specified in Section 2.3 Specifications A.l and A.2.

NOTE 3:

Calibrate electronic bistable trips by injection of an external test current at the frequency specified in the Surveillance Frequency Control Programonee per 3 months. Calibrate transmitters by application of test pressure at the frequency specified in the Surveillance Frequency Control Programonce per 12 months.

NOTE 4:

Perform LPRM detectors calibration at the frequency specified in the Surveillance Frequency Control Program.every 1000 MWD/MT Merage Core Exposure NOTE 5:

Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

The following notes are only for Item 15 of Table 4.1.1:

A channel may be taken out of service for the purpose of a check, calibration, test or maintenance without declaring the channel to be inoperable.

a.

The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exists:

1)

Instrument indicates measured levels above the alarm setpoint.

2)

Instrument indicates a downscale failure.

3)

Instrument controls not set in operate mode.

4)

Instrument electrical power loss.

OYSTER CREEK 4.1-9 Change: 5, 7, Amendment No.: 71, 80, 95, 108, 171, 208, 263, 273 NOTE 1:

NOTE 2:

NOTE 3:

NOTE 4:

NOTES:

TABLE 4.1.1 Page 6 of6 MINIMUM CHECK, CALIBRATION AND TEST FREQUENCY FOR PROTECTIVE INSTRUMENTATION Each automatic scram contactor is required to be tested at least ease fler.. eel( at the frequency specified in the Surveillance Frequency Control Program. When not tested by other means, the weelHy-test can be performed by using the subchannel test switches.

At least daily dHriag reaster PO\\VER OPERATION At the frequency specified in the Surveillance Frequency Control Program, the reactor neutron flux peaking factor shall be estimated and flow-referenced APRM scram and rod block settings shall be adjusted, if necessary, as specified in Section 2.3 Specifications A.I and A.2.

Calibrate electronic bistable trips by injection of an external test current at the frequency specified in the Surveillance Frequency Control Progral11BRe-fJer 3 m8atfls. Calibrate transmitters by application of test pressure at the frequency specified in the Surveillance Frequency Control Program8ase Ber 12 m8atfls.

Perform LPRM detectors calibration at the frequency specified in the Surveillance Frequency Control Program.e.ef\\ lQQQ MWD/MT P"eraEe Cere EJ'fl8SHre Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

The following notes are only for Item 15 ofTable 4.1.1:

A channel may be taken out of service for the purpose of a check, calibration, test or maintenance without declaring the channel to be inoperable.

a.

The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs ifany of the following conditions exists:

1)

Instrument indicates measured levels above the alarm setpoint.

2)

Instrument indicates a downscale failure.

3)

Instrument controls not set in operate mode.

4)

Instrument electrical power loss.

OYSTER CREEK Change: 5,7, Amendment No.:

4.1-9 71,80,95,108,171,208,263,273

TABLE 4.1.2 MINIMUM TEST FREQUENCIES FOR TRIP SYSTEMS Trip System Minimum Test Frequency (Note 1) 1)

Dual Channel (Scram)

Same as for respective Instrumentation in Table 4.1.1 2)

Rod Block Same as for respective Instrumentation in Table 4.1.1 3)

DELETED DELETED 4)

Automatic Depressurization Each refueling outage each trip system, one at a time 5)

MSIV Closure Each refueling outage each closure logic circuit independently (1 valve at a time) 6)

Core Spray 1/3 mo and each refueling outago each trip system, one at a time 7)

Primary Containment Isolation Each refueling outage each trip circuit independently (1 valve at a time) 8)

Refueling Interlocks Prior to each refueling operation 9)

Isolation Condenser Actuation and Isolation Each refueling outage each trip circuit independently (1 valve at a time) 10)

Reactor Building Isolation and SGTS iui rncnctive Instrumentation Initiation in Table 1.1.1 11)

DELETED DELETED 12)

Air Eiector Offgass Line Isolation Each refueling outage 13)

Containment Vent and Purge Isolation 1/24 mo Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

OYSTER CREEK 4.1-10 Amendment No.: 108,116,144,160,171,193, 208, 273 TABLE 4.1.2 MINIMUM TEST FREQUENCIES FOR TRIP SYSTEMS Trip System 1)

Dual Channel (Scram) 2)

Rod Block 3)

DELETED 4)

Automatic Depressurization each trip system, one at a time 5)

MSIV Closure each closure logic circuit independently (1 valve at a time) 6)

Core Spray each trip system, one at a time 7)

Primary Containment Isolation each trip circuit independently (1 valve at a time) 8)

Refueling Interlocks 9)

Isolation Condenser Actuation and Isolation each trip circuit independently (1 valve at a time) 10)

Reactor Building Isolation and SGTS Initiation 11)

DELETED 12)

Air Ejector Offgass Line Isolation 13)

Containment Vent and Purge Isolation Minimum Test Frequency (Note 1)

Same as for respective Instrumentation in Table 4.1.1 Same as for respective Instrumentation in Table 4.1.1 DELETED Each refueling outage Each refueling outage 1/3 mo and each refueling outage Each refueling outage Prior to each refueling operation Each refueling outage Same as for respective Instrumentation in Table 4.1.1 DELETED Each refueling outage 1/24 mo Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

OYSTER CREEK 4.1-10 Amendment No.: 108,116,144,160,171,193,208,273

4.2 REACTIVITY CONTROL Applicability:

Applies to the surveillance requirements for reactivity control.

Qbiective:

To verify the capability for controlling reactivity.

Specification:

A.

SDM shall be verified:

1.

Prior to each CORE ALTERATION, and 2.

Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following the first criticality following any CORE ALTERATION.

B.

The control rod drive housing support system shall be inspected after reassembly.

C.

The maximum scram insertion time of the control rods shall be demonstrated through measurement and, during single control rod scram time tests, the control rod drive pumps shall be isolated from the accumulators:

1.

For all control rods prior to THERMAL POWER exceeding 40% power with reactor coolant pressure greater than 800 psig, following core alterations or after a reactor shutdown that is greater than 120 days.

2.

For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either a or b as follows:

a.1 Specifically affected individual control rods shall be scram time tested with the reactor depressurized and the scram insertion time from the fully withdrawn position to 90% insertion shall not exceed 2.2 seconds, and a.2 Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure prior to exceeding 40% power.

b.

Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure.

3.

On a frequency of less than or equal to once per 180 days of cumulative power operationAt the frequency specified in the Surveillance Frequency Control Program, for at least 20 control rods, on a rotating basis, with reactor coolant pressure greater than 800 psig.

D.

Each withdrawn control rod shall be exercised at Ioast once per 31 daysat the frequency specified in the Surveillance Frequency Control Program. This test shall be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power operation is continuing with two or more inoperable control rods or in the event power operation is continuing with one fully or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out. The surveillance need not be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable rods has been reduced to less than two and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause of an immovable control rod.

OYSTER CREEK 4.2-1 Amendment No: 178, 198, 249, 266, 275 4.2 REACTIVITY CONTROL Applicability:

Applies to the surveillance requirements for reactivity control.

Objective:

To verify the capability for controlling reactivity.

Specification:

A.

80M shall be verified:

1.

Prior to each CORE ALTERATION, and 2.

Once within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following the first criticality following any CORE ALTERATION.

B.

The control rod drive housing support system shall be inspected after reassembly.

C.

The maximum scram insertion time of the control rods shall be demonstrated through measurement and, during single control rod scram time tests, the control rod drive pumps shall be isolated from the accumulators:

1.

For all control rods prior to THERMAL POWER exceeding 40%

power with reactor coolant pressure greater than 800 psig, following core alterations or after a reactor shutdown that is greater than 120 days.

2.

For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "a" or "b" as follows:

a.1 Specifically affected individual control rods shall be scram time tested with the reactor depressurized and the scram insertion time from the fully withdrawn position to 90%

insertion shall not exceed 2.2 seconds, and a.2 Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure prior to exceeding 400/0 power.

b.

Specifically affected individual control rods shall be scram time tested at greater than 800 psig reactor coolant pressure.

3.

On a frequency of less than or equal to once per 180 days of cumulative power operationAt the frequency specified in the Surveillance Frequency Control Program, for at least 20 control rods, on a rotating basis, with reactor coolant pressure greater than 800 psig.

Amendment No: 178, 198, 249,266,275 4.2-1 o.

Each withdrawn control rod shall be exercised at least once per 31 daysat the frequency specified in the Surveillance Frequency Control Program. This test shall be performed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the event power operation is continuing with two or more inoperable control rods or in the event power operation is continuing with one fully or partially withdrawn rod which cannot be moved and for which control rod drive mechanism damage has not been ruled out. The surveillance need not be completed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if the number of inoperable rods has been reduced to less than two and if it has been demonstrated that control rod drive mechanism collet housing failure is not the cause of an immovable control rod.

OYSTER CREEK

E.

Surveillance of the standby liquid control system shall be as follows:

1. Pump operability Note lOnce/3 months

2. Boron concentration Note lOnce/month determination

3. Functional test Note lOnce every 24 months

4. Solution volume and Note lOnce/day temperature check

5. Solution Boron-lO Note lOnce every 24 months.

Enrichment Enrichment analyses shall be received no later than 30 days after sampling. If not received within 30 days, notify NRC (within 7 days) of plans to obtain test results.

Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted above.

F.

At specific power operation conditions, the actual control rod configuration will be compared with the expected configuration based upon appropriately corrected past data. This comparison shall be made every equivalent full power month. The initial rod inventory measurement performed with equilibrium conditions are established after a refueling or major core alteration will be used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle.

G.

The scram discharge volume drain and vent valves shall be verified open at least once per 31 daysat the frequency specified in the Surveillance Frequency Control Program, except in shutdown mode*, and shall be cycled at least one complete cycle of full travel at the frequency specified in the Surveillance Frequency Control Programat least quarterly.

H.

All withdrawn control rods shall be determined OPERABLE by demonstrating the scram discharge volume drain and vent valves OPERABLE. This will be done at the frequency specified in the Surveillance Frequency Control Programat least once per refueling cycle by placing the mode switch in shutdown and by verifying that:

a.

The drain and vent valves close within 30 seconds after receipt of a signal for control rods to scram, and b.

The scram signal can be reset and the drain and vent valves open when the scram discharge volume trip is bypassed.

• These valves may be closed intermittently for testing under administrative control.

Corrected: 12/24/84 OYSTER CREEK 4.2-2 Amendment No.: 64, 74, 75,124,141,159, 172, 178 Change: 25 E.

Surveillance of the standby liquid control system shall be as follows:

1. Pump operability

2. Boron concentration determination

3. Functional test

4. Solution volume and temperature check

5. Solution Boron-10 Enrichment Note 1Once/3 menths Note 1Once/month Note 1Once every 24 months Note 1Once/day Note 1Once every 24 months.

Enrichment analyses shall be received no later than 30 days after sampling. If not received within 30 days, notify NRC (within 7 days) of plans to obtain test results.

Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted above.

F.

At specific power operation conditions, the actual control rod configuration will be compared with the expected configuration based upon appropriately corrected past data. This comparison shall be made every equivalent full power month. The initial rod inventory measurement performed with equilibrium conditions are established after a refueling or major core alteration will be used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle.

G.

The scram discharge volume drain and vent valves shall be verified open at least once per 31 daysat the frequency specified in the Surveillance Frequency Control Program, except in shutdown mode*, and shall be cycled at least one complete cycle of full travel at the frequency specified in the Surveillance Frequency Control Programat least quarterly.

H.

All withdrawn control rods shall be determined OPERABLE by demonstrating the scram discharge volume drain and vent valves OPERABLE. This will be done at the frequency specified in the Surveillance Frequency Control Programat least once per refueling cycle by placing the mode switch in shutdown and by verifying that:

a.

The drain and vent valves close within 30 seconds after receipt of a signal for control rods to scram, and b.

The scram signal can be reset and the drain and vent valves open when the scram discharge volume trip is bypassed.

• These valves may be closed intermittently for testing under administrative control.

OYSTER CREEK Corrected: 12/24/84 4.2-2 Amendment No.: 64, 74, 75,124,141,159,172,178 Change:~

The monthly control rod exercise test at the frequency specified in the Surveillance Frequency Control Program serves as a periodic check against deterioration of the control rod system. Experience with this control rod system and analysis performed for an industry-wide BWR initiative that was approved by NRC have indicated that monthly tests are adequate, and that rods which move by drive pressure will scram when required as the pressure applied is much higher.The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The requirement to exercise the control rods within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a condition with two or more control rods which are valved out of service or one fully or partially withdrawn control rod which can not be moved provides assurance of the reliability of the remaining control rods.

Pump operability, boron concentration, solution temperature and volume of standby liquid control system 4 are checked on a frequency consistent with instrumentation checks described in Specification 4.1 at the frequency specified in the Surveillance Frequency Control Program.

Experience with similar systems has indicated that the test frequencies are adequate. The only practical time to functionally test the liquid control system is during a refueling outage. The functional test includes the firing of explosive charges to open the shear plug valves and the pumping of demineralized water into the reactor to assure operability of the system downstream of the pumps. The test also includes recirculation of liquid control solution to and from the solution tanks.

Pump operability is demonstrated on a more frequent basis. This test consists of recirculation of demineralized water to a test tank. A continuity check of the firing circuit on the shear plug valves is provided by pilot lights in the control room. Tank level and temperature alarms are provided to alert the operator to off-normal conditions.

Figure 3.2.1 was revised to reflect the minimum and maximum weight percent of sodium pentaborate solution, and the minimum atom percent of B-i 0 to meet 10 CFR 50.62(c)(4). Since the weight percent of sodium pentaborate can change with water makeup or water evaporation, frequent surveillances are performed on the solution concentration, volume and temperature. The sodium pentaborate is enriched with B-b at the chemical vendors facility to meet the minimum atom percent.

Preshipment samples of batches are analyzed for B-b enrichment and verified by an independent laboratory prior to shipment to Oyster Creek. Since the B-b enrichment will not change while in storage or in the SLCS tank, the surveillance for B-b enrichment is performed or a 24 month intervalat the frequency specified in the Surveillance Frequency Control Program. An additional requirement has been added to evaluate the solutions capability to meet the original design shutdown criteria whenever the Boron-i 0 enrichment requirement is not met.

The functional test and other surveillance on components, along with the monitoring instrumentation, gives a high reliability for standby liquid control system operability.

OYSTER CREEK 4.2-4 Amendment No.: 75, 124, 159, 172, 178, 275 The monthly control rod exercise test at the frequency specified in the Surveillance Frequency Control Program serves as a periodic check against deterioration of the control rod system. Experience with this control rod system and analysis performed for an industry wide BVVR initiative that was approved by NRCf&1 have indicated that monthly tests are adequate, and that rods 'Nhich move by drive pressure will scram when required as the pressure applied is much higher.The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. The requirement to exercise the control rods within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a condition with two or more control rods which are valved out of service or one fully or partially withdrawn control rod which can not be moved provides assurance of the reliability of the remaining control rods.

Pump operability, boron concentration, solution temperature and volume of standby liquid control system(4) are checked on a frequency consistent with instrumentation checks described in Specification 4.1 at the frequency specified in the Surveillance Frequency Control Program.

Experience with similar systems has indicated that the test frequencies are adequate. The only practical time to functionally test the liquid control system is during a refueling outage. The functional test includes the firing of explosive charges to open the shear plug valves and the pumping of demineralized water into the reactor to assure operability of the system downstream of the pumps. The test also includes recirculation of liquid control solution to and from the solution tanks.

Pump operability is demonstrated on a more frequent basis. This test consists of recirculation of demineralized water to a test tank. A continuity check of the firing circuit on the shear plug valves is provided by pilot lights in the control room. Tank level and temperature alarms are provided to alert the operator to off-normal conditions.

Figure 3.2.1 was revised to reflect the minimum and maximum weight percent of sodium pentaborate solution, and the minimum atom percent of 8-10 to meet 10 CFR 50.62(c)(4). Since the weight percent of sodium pentaborate can change with water makeup or water evaporation, frequent surveillances are performed on the solution concentration, volume and temperature. The sodium pentaborate is enriched with 8-10 at the chemical vendor's facility to meet the minimum atom percent.

Preshipment samples of batches are analyzed for B-1 0 enrichment and verified by an independent laboratory prior to shipment to Oyster Creek. Since the 8-10 enrichment will not change while in storage or in the SLCS tank, the surveillance for 8-10 enrichment is performed 00 a 24 month intervalat the frequency specified in the Surveillance Frequency Control Program. An additional requirement has been added to evaluate the solution's capability to meet the original design shutdown criteria whenever the 8oron-10 enrichment requirement is not met.

The functional test and other surveillance on components, along with the monitoring instrumentation, gives a high reliability for standby liquid control system operability.

OYSTER CREEK 4.2-4 Amendment No.: 75, 124, 159, 172, 178, 275

4.3 REACTOR COOLANT Applicability:

Applies to the surveillance requirements for the reactor coolant system.

Oblective:

To determine the condition of the reactor coolant system and the operation of the 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 shall be periodically removed, tested, and evaluated to determine the effects of neutron fluence on the fracture toughness of the vessel shell materials. Pressure and temperature curves are contained in the Pressure and Temperature Limits Report (PTLR).

B.

lnservice inspection of ASME Code Class 1, Class 2 and Class 3 systems and components shall be performed in accordance with Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a, except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a.

C.

Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR, Section 50.55a, except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a.

D.

A visual examination for leaks shall be made with the reactor coolant system at pressure during each scheduled refueling outage or after major repairs have been made to the reactor coolant system in accordance with Article 5000, Section Xl. 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 tested in accordance with Specification C above. Setpoints shall be as follows:

Number of Valves Set Points (psig) 4 1212+/-36 5

1221+/-36 F.

A sample of reactor coolant shall be analyzed at least every 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sat the frequency specified in the Surveillance Frequency Control Program for the purpose of determining the content of chloride ion and to check the conductivity.

OYSTER CREEK 4.3-1 Amendment No.: 82, 90, 120, 150, 151, 164, 188, 195, 261, 268, 269 4.3 REACTOR COOLANT Applicability:

Applies to the surveillance requirements for the reactor coolant system.

Objective:

To determine the condition of the reactor coolant system and the operation of the 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 shall be periodically removed, tested, and evaluated to determine the effects of neutron fluence on the fracture toughness of the vessel shell materials. Pressure and temperature curves are contained in the Pressure and Temperature Limits Report (PTLR).

B.

Inservice inspection of ASME Code Class 1, Class 2 and Class 3 systems and components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR, Section 50.55a, except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a.

C.

Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR, Section 50.55a, except where specific written relief has been granted by the NRC pursuant to 10 CFR, Section 50.55a.

D.

A visual examination for leaks shall be made with the reactor coolant system at pressure during each scheduled refueling outage or after major repairs have been made to the reactor coolant system in accordance with Article 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 tested in accordance with Specification C above. Setpoints shall be as follows:

Number of Valves 4

5 Set Points (psig) 1212 +/- 36 1221 +/- 36 F.

A sample of reactor coolant shall be analyzed at least every 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sat the frequency specified in the Surveillance Frequency Control Program for the purpose of determining the content of chloride ion and to check the conductivity.

OYSTER CREEK 4.3-1 Amendment No.: 82, 90, 120, 150, 151,164,188,195, 261,268,269

Primary Coolant System Pressure Isolation Valves Srecification:

1.

Periodic leakage testing(a)on each valve listed in Table 4.3.1 shall be accomplished prior to exceeding 600 psig reactor pressure every time the plant is placed in the cold shutdown 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 preceding 9 months, whenever the valve is moved whether by manual actuation or due to flow conditions, and after returning the valve to service after maintenance, repair or replacement work is performed.

H.

Reactor Coolant System Leakage 1.

Unidentified leakage rate shall be calculated at the frequency specified in the Surveillance Frequency Control Programat least once every Ihours.

2.

Total leakage rate (identified and unidentified) shall be calculated at the frequency specified in the Surveillance Frequency Control Programat 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 equipment drain tank flow integrator shall be conducted at the frequency specified in the Surveillance Frequency Control Programat least once per 24 months.

I.

An inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC staff positions on schedule, methods, personnel, and sample expansion included in the generic letter or in accordance with alternate measures approved by the NRC staff.

Bases:

Data is available relating neutron fluence (E>1.OMeV) and the change in the Reference Nil-Ductility Transition Temperature (RTNDT). Pressure and temperature curves are contained in the Pressure and Temperature Limits Report (PTLR).

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 (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.

• NRC Order dated April 20, 1981.

OYSTER CREEK 4.3-2 Amendment No.: 97,118,120,151,154, 188, 193, 263, 269 G.

Primary Coolant System Pressure Isolation Valves Specification:

1.

Periodic leakage testing(a) on each valve listed in Table 4.3.1 shall be accomplished prior to exceeding 600 psig reactor pressure every time the plant is placed in the cold shutdown 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 preceding 9 months, whenever the valve is moved whether by manual actuation or due to flow conditions, and after returning the valve to service after maintenance, repair or replacement work is performed.

H.

Reactor Coolant System Leakage 1.

Unidentified leakage rate shall be calculated at the frequency specified in the Surveillance Frequency Control Programat-least once every 4hours.

2.

Total leakage rate (identified and unidentified) shall be calculated at the frequency specified in the Surveillance Frequency Control Programat 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 equipment drain tank flow integrator shall be conducted at the frequency specified in the Surveillance Frequency Control Programat least once per 24 months.

I.

An inservice inspection program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC staff positions on schedule, methods, personnel, and sample expansion included in the generic letter or in accordance with alternate measures approved by the NRC staff.

Bases:

Data is available relating neutron fluence (E>1.0MeV) and the change in the Reference Nil-Ductility Transition Temperature (RTNOT)' Pressure and temperature curves are contained in the Pressure and Temperature Limits Report (PTLR).

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 (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.

• NRC Order dated April 20, 1981.

OYSTER CREEK 4.3-2 Amendment No.: 97,118,120,151,154, 188, 19d,26d,269

the existence of ASME Section Xl. For this reason, the degree of access required by ASME Section Xl is not generally available and will be addressed as requests for relief in accordance with 10 CFR 50.55a(g).

Experience in safety valve operation shows testing in accordance with the ASME Code is adequate to detect failures or deterioration. The as-found setpoint tolerance is specified in the ASME CM Code as the owner-defined tolerance or +/- 3% of valve nameplate set pressure. An analysis has been performed which shows that with all safety valves set 36 psig higher, the safety limit of 1375 psig is not exceeded.

Conductivity instruments continuously monitor the reactor coolant. Experience indicates that a periodic check of the conductivity instrumentation at least every 72 houreat the frequency specified in the Surveillance Frequency Control Program 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 frequency is adequate.

OYSTER CREEK 4.3-3 Amendment No.: 82, 90, 261, 268 the existence of ASME Section XI. For this reason, the degree of 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).

Experience in safety valve operation shows testing in accordance with the ASME Code is adequate to detect failures or deterioration. The as-found setpoint tolerance is specified in the ASME OM Code as the owner-defined tolerance or +/- 3%

of valve nameplate set pressure. An analysis has been performed which shows that with all safety valves set 36 psig higher, the safety limit of 1375 psig is not exceeded.

Conductivity instruments continuously monitor the reactor coolant. Experience indicates that a periodic check of the conductivity instrumentation at least every 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sat the frequency specified in the Surveillance Frequency Control Program 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 frequency is adequate.

OYSTER CREEK 4.3-3 Amendment No.: 82,90,261,268

Item Frequency C.

Containment Cooling System

2. Motor-operated valve operability Note lEvery 3 months

3. Pump compartment water-Note lOnce/week and after each entry tight doors closed D. Emergency Service Water System

1. Pump Operability Note lOnce/3 months. Also after major maintenance and prior to startup following a refueling outage.

E. Control Rod Drive Hydraulic System

1. Pump Operability Note lOnce/month. Also after major maintenance and prior to startup following a refueling outage.

F. Fire Protection System

1. Pump Operability Note I Once/month. Also after major maintenance and prior to startup following a refueling outage.

2. Isolation valve operability Note IOnce/3 months. Also after major maintenance and prior to startup following a refueling outage.

Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted.

Bases:

It is during major maintenance or repair that a systems design intent may be violated accidentally.

Therefore, a functional test is required after every major maintenance operation. During an extended outage, such as a refueling outage, major repair and maintenance may be performed on many systems. To be sure that these repairs on other systems do not encroach unintentionally on critical standby cooling systems, they should be given a functional test prior to startup.

Motor operated pumps, valves and other active devices that are normally on standby should be exercised periodically to make sure that they are free to operate. Motors on pumps should operate long enough to approach equilibrium temperature to ensure there is no overheat problem. Whenever practical, valves should be stroked full length to ensure that nothing impedes their motion. Testing of components per OC Inservice Testing Program in accordance with the ASME Code once every 3 months provides assurances of the availability of the system. The Control Rod Hydraulic pumps and Fire Protection System pumps are not part of the lnservice Test Program per the ASME Code and will continue to be tested for operability once per monthat the frequency specified in the Surveillance Frequency Control Program. Engineering judgment based on experience and availability analyses of the type presented in Appendix L of the FDSAR indicates that testing these components more ofter than once a month over a long period of time does not significantly improve the system reliability. Also, at this frequency of testing wearout should not be a problem through the life of the plant.

OYSTER CREEK 4.4-2 Amendment No.: 109,160, 210, 268 Freguency C.

Containment Cooling System

2. Motor-operated valve operability

3. Pump compartment water-tight doors closed D. Emergency Service Water System

1. Pump Operability E. Control Rod Drive Hydraulic System

1. Pump Operability F. Fire Protection System

1. Pump Operability

2. Isolation valve operability Note 1Evory 3 months Note 1Onco/wook and after each entry Note 1Onco/3 months. Also after major maintenance and prior to startup following a refueling outage.

Note 1Onco/month. Also after major maintenance and prior to startup following a refueling outage.

Note 1Onco/month. Also after major maintenance and prior to startup following a refueling outage.

Note 1Once/3 months. Also after major maintenance and prior to startup following a refueling outage.

Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted.

Bases:

It is during major maintenance or repair that a system's design intent may be violated accidentally.

Therefore, a functional test is required after every major maintenance operation. During an extended outage, such as a refueling outage, major repair and maintenance may be performed on many systems. To be sure that these repairs on other systems do not encroach unintentionally on critical standby cooling systems, they should be given a functional test prior to startup.

Motor operated pumps, valves and other active devices that are normally on standby should be exercised periodically to make sure that they are free to operate. Motors on pumps should operate long enough to approach equilibrium temperature to ensure there is no overheat problem. Whenever practical, valves should be stroked full length to ensure that nothing impedes their motion. Testing of components per OC lnservice Testing Program in accordance with the ASME Code once ovory 3 months provides assurances of the availability of the system. The Control Rod Hydraulic pumps and Fire Protection System pumps are not part of the Inservice Test Program per the ASME Code and will continue to be tested for operability once per monthat the frequency specified in the Surveillance Frequency Control Program. Engineering judgmont based on m<perience and availability analyses of the type presented in Appendix L of the FDSAR indicates that testing these components more often than once a month ovor a long period of time does not significantly improve the system reliability. Also, at this frequoncy of testing '..learout should not be a problem through the life of the plant.

OYSTER CREEK 4.4-2 Amendment No.: 109,160,210,268

b.

If the airlock is opened during a period when Primary Containment is not required, it need not be tested while Primary Containment is not required, but must be tested at a prior to returning the reactor to an operating mode requiring PRIMARY CONTAINMENT INTEGRITY.

D.

Primary Containment Leakage Rates shall be limited to:

1.

The maximum allowable Primary Containment leakage rate is 1.0 La. The maximum allowable Primary Containment leakage rate to allow for plant startup following a type A test is 0.75 La. The leakage rate acceptance criteria for the Primary Containment Leakage Rate Testing Program for Type B and Type C tests is 0.60 La at a, except as stated in Specification 4.5.D.2.

2.

Verify leakage rate through each MSIV is 11.9 scfh when tested at 20 psig.

3.

The leakage rate acceptance criteria for the drywell airlock shall be 0.05 La when measured or adjusted to Pa.

E.

Continuous Leak Rate Monitor 1.

When the primary containment is inerted, the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements.

2.

This monitoring system may be taken out of service for the purpose of maintenance or testing but shall be returned to service as soon as practical.

F.

Functional Test of Valves 1.

All automatic primary containment isolation valves shall be tested for automatic closure by an isolation signal at the frequency specified in the Surveillance Frequency Control Programduring each REFUELING OUTAGE and the isolation time determined to be within its limit. The following valves are required to close in the time specified below:

Main steam line isolation valves:

3 seconds and 10 seconds 2.

Each automatic primary containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on OYSTER CREEK 4.5-2 Amendment No.: 132, 186, 196, 250 b.

If the airlock is opened during a period when Primary Containment is not required, it need not be tested while Primary Containment is not required, but must be tested at Pa prior to returning the reactor to an operating mode requiring PRIMARY CONTAINMENT INTEGRITY.

O.

Primary Containment Leakage Rates shall be limited to:

1.

The maximum allowable Primary Containment leakage rate is 1.0 La. The maximum allowable Primary Containment leakage rate to allow for plant startup following a type A test is 0.75 La. The leakage rate acceptance criteria for the Primary Containment Leakage Rate Testing Program for Type B and Type C tests is

~0.60 La at Pa, except as stated in Specification 4.5.0.2.

2.

Verify leakage rate through each MSIV is ~ 11.9 scfh when tested at ~ 20 psig.

3.

The leakage rate acceptance criteria for the drywell airlock shall be

~ 0.05 La when measured or adjusted to Pa.

E.

Continuous Leak Rate Monitor 1.

When the primary containment is inerted, the containment shall be continuously monitored for gross leakage by review of the inerting system makeup requirements.

2.

This monitoring system may be taken out of service for the purpose of maintenance or testing but shall be returned to service as soon as practical.

F.

Functional Test of Valves 1.

All automatic primary containment isolation valves shall be tested for automatic closure by an isolation signal at the frequency specified in the Surveillance Frequency Control Programduring each REFUELI~JG OUTAGE and the isolation time determined to be within its limit. The following valves are required to close in the time specified below:

Main steam line isolation valves:

~ 3 seconds and ~ 10 seconds 2.

Each automatic primary containment isolation valve shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work is performed on OYSTER CREEK 4.5-2 Amendment No.: 132, 186, 196,250

the valve or its associated actuator by cycling the valve through at least one complete cycle of full travel and verifying the isolation time limit is met.

Following maintenance, repair or replacement work on the control or power circuit for the valves, the affected component shall be tested to assure it will perform its intended function in the circuit.

3.

During each COLD SHUTDOWN, each main steam isolation valve shall be closed and its closure time verified to be within the limits of Specification 4.5.F.1 above unless this test has been performed within the last 92 days.

4.

Reactor Building to Suppression Chamber Vacuum Breakers a.

The reactor building to suppression chamber vacuum breakers and associated instrumentation, including setpoint, shall be checked for proper operation at the frequency specified in the Surveillance Frequency Control Programevery three months.

b.

At the frequency specified in the Surveillance Frequency Control ProgramDuring each REFUELING OUTAGE, each vacuum breaker shall be tested to determine that the force required to open the vacuum breaker from closed to fully open does not exceed the force specified in Specification 3.5.A.4.a. The air-operated vacuum breaker instrumentation shall be calibrated at the frequency specified in the Surveillance Frequency Control Programduring each REFUELING OUTAGE.

5.

Pressure Suppression Chamber - Drywell Vacuum Breakers a.

Periodic OPERABILITY Tests At the frequency specified in the Surveillance Frequency Control ProgramOnce every 3 months and following any release of energy which would tend to increase pressure to the suppression chamber, each OPERABLE suppression chamber - drywell vacuum breaker shall be exercised.

Operation of position switches, indicators and alarms shall be verified at the frequency specified in the Surveillance Frequency Control Programevery 3 months by operation of each OPERABLE vacuum breaker.

b.

REFUELING OUTAGE TeetsThe following tests, with the exception of b(4),

are performed at the frequency specified in the Surveillance Frequency Control Program.

(1)

All suppression chamber - drywell vacuum breakers shall be tested to determine the force required to open each valve from fully closed to fully open.

(2)

The suppression chamber

- drywell vacuum breaker position indication and alarm systems shall be calibrated and functionally tested.

OYSTER CREEK 4.5-3 Amendment No.: 144, 186, 196, 210, 221 the valve or its associated actuator by cycling the valve through at least one complete cycle of full travel and verifying the isolation time limit is met.

Following maintenance, repair or replacement work on the control or power circuit for the valves, the affected component shall be tested to assure it will perform its intended function in the circuit.

3.

During each COLD SHUTDOWN, each main steam isolation valve shall be closed and its closure time verified to be within the limits of Specification 4.5.F.1 above unless this test has been performed within the last 92 days.

4.

Reactor Building to Suppression Chamber Vacuum Breakers a.

The reactor building to suppression chamber vacuum breakers and associated instrumentation, including setpoint, shall be checked for proper operation at the frequency specified in the Surveillance Frequency Control Programevery three months.

b.

At the frequency specified in the Surveillance Frequency Control ProgramDuring each REFUELING OUTAGE, each vacuum breaker shall be tested to determine that the force required to open the vacuum breaker from closed to fully open does not exceed the force specified in Specification 3.5.AA.a. The air-operated vacuum breaker instrumentation shall be calibrated at the frequency specified in the Surveillance Frequency Control Programduring each REFUELI~JG OUTAGE.

5.

Pressure Suppression Chamber - Drywell Vacuum Breakers a.

Periodic OPERABILITY Tests At the frequency specified in the Surveillance Frequency Control ProgramOnce every d months and following any release of energy which would tend to increase pressure to the suppression chamber, each OPERABLE suppression chamber - drywell vacuum breaker shall be exercised.

Operation of position switches, indicators and alarms shall be verified at the frequency specified in the Surveillance Frequency Control Programe¥efY d months by operation of each OPERABLE vacuum breaker.

b.

REFUELI~JG OUTAGE TestsThe following tests, with the exception of b(4),

are performed at the frequency specified in the Surveillance Frequency Control Program.

(1)

All suppression chamber - drywell vacuum breakers shall be tested to determine the force required to open each valve from fully closed to fully open.

(2)

The suppression chamber - drywell vacuum breaker position indication and alarm systems shall be calibrated and functionally tested.

OYSTER CREEK 4.5-3 Amendment No.: 144, 18(3, 19(3,210,221

(3)

At least four of the suppression chamber - drywell vacuum breakers shall be inspected, If deficiencies are found, all vacuum breakers shall be inspected and deficiencies corrected such that Specification 3.5.A.5.a can be met.

(4)

A drywell to suppression chamber leak rate test shall be performed once every 24 months to demonstrate that with an initial differential pressure of not less than 1.0 psi, the differential pressure decay rate shall not exceed the equivalent of air flow through a 2-inch orifice.

G.

Reactor Building 1.

Secondary containment capability tests shall be conducted after isolating the reactor building and placing either Standby Gas Treatment System filter train in operation.

2.

The tests shall be performed at the frequency specified in the Surveillance Frequency Control Programat least once per operating cycle (interval not to exceed 20 months) and shall demonstrate the capability to maintain a 1/4 inch of water vacuum under calm wind conditions with a Standby Gas Treatment System Filter train flow rate of not more than 4000cfm.

3.

A secondary containment capability test shall be conducted at each refueling outage prior to refueling.

4.

The results of the secondary containment capability tests shall be in the subject of a summary technical report which can be included in the reports specified in Section 6.

H.

Standby Gas Treatment System 1.

The capability of each Standby Gas Treatment System circuit shall be demonstrated by:

a.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 18 months, after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation, and following significant painting, fire, or chemical release in the reactor building during operation of the Standby Gas Treatment System by verifying that:

(1) The charcoal absorbers remove 99% of a halogenated hydrocarbon refrigerant test gas and the HEPA filters remove 99% of the DOP in a cold DOP test when tested in accordance with ANSI N510-1 975.

OYSTER CREEK 4.5-4 Amendment No.: 144,186,193,219 (3)

At least four of the suppression chamber - drywell vacuum breakers shall be inspected. If deficiencies are found, all vacuum breakers shall be inspected and deficiencies corrected such that Specification 3.5.A.5.a can be met.

(4)

A drywell to suppression chamber leak rate test shall be performed once every 24 months to demonstrate that with an initial differential pressure of not less than 1.0 psi, the differential pressure decay rate shall not exceed the equivalent of air flow through a 2-inch orifice.

G.

Reactor Building 1.

Secondary containment capability tests shall be conducted after isolating the reactor building and placing either Standby Gas Treatment System filter train in operation.

2.

The tests shall be performed at the frequency specified in the Surveillance Frequency Control Programat least once per operating cycle (interval not to exceed 20 months) and shall demonstrate the capability to maintain a 1/4 inch of water vacuum under calm wind conditions with a Standby Gas Treatment System Filter train flow rate of not more than 4000cfm.

3.

A secondary containment capability test shall be conducted at each refueling outage prior to refueling.

4.

The results of the secondary containment capability tests shall be in the subject of a summary technical report which can be included in the reports specified in Section 6.

H.

Standby Gas Treatment System 1.

The capability of each Standby Gas Treatment System circuit shall be demonstrated by:

a.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 18 months, after every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation, and following significant painting, fire, or chemical release in the reactor building during operation of the Standby Gas Treatment System by verifying that:

(1) The charcoal absorbers remove ~99% of a halogenated hydrocarbon refrigerant test gas and the HEPA filters remove

~99% of the DOP in a cold DOP test when tested in accordance with ANSI N510-1975.

OYSTER CREEK 4.5-4 Amendment No.: 144,186,193,219

(2)

Results of laboratory carbon sample analysis show 95%

radioactive methyl iodide removal efficiency when tested in accordance with ASTM D 3803-1989 (30°C, 95% relative humidity, at least 45.72 feet per minute charcoal bed face velocity).

b.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 18 months by demonstrating:

(1)

That the pressure drop across a HEPA filter is equal to or less than the maximum allowable pressure drop indicated in Figure 4.5.1.

(2)

The inlet heater is capable of at least 10.9 KW input.

(3)

Operation with a total flow within 10% of design flow.

c.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 30 days on a STAGGERED TEST BASIS by operating each circuit for a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

d.

Anytime the HEPA filter bank or the charcoal absorbers have been partially or completely replaced, the test per 4.5.H.1.a(as applicable) will be performed prior to returning the system to OPERABLE STATUS.

e.

Automatic initiation of each circuit at the frequency specified in the Surveillance Frequency Control Programevery 18 months.

Inerting Surveillance When an inert atmosphere is required in the primary containment, the oxygen concentration in the primary containment shall be checked at the frequency specified in the Surveillance Frequency Control Programat least weekly.

J.

Drywell Coating Surveillance Carbon steel test panels coated with Firebar D shall be placed inside the drywell near the reactor core midplane level. They shall be removed for visual observation and weight loss measurements during the first, second, fourth and eighth refueling outages.

K.

Instrument Line Flow Check Valves Surveillance The capability of a representative sample of instrument line flow check valves to isolate shall be tested at the frequency specified in the Surveillance Frequency Control Programat least once per 24 months. In addition, each time an instrument line is returned to service after any condition which could have produced a pressure flow disturbance in that line, the open position of the flow check valve in that line shall be verified. Such conditions include:

OYSTER CREEK 4.5-5 Amendment No.: 132,186,216,219 Corrected by letter of 3/18/02 (2)

Results of laboratory carbon sample analysis show ;:::95%

radioactive methyl iodide removal efficiency when tested in accordance with ASTM 03803-1989 (30°C, 95% relative humidity, at least 45.72 feet per minute charcoal bed face velocity).

b.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 18 rnonths by demonstrating:

(1)

That the pressure drop across a HEPA filter is equal to or less than the maximum allowable pressure drop indicated in Figure 4.5.1.

(2)

The inlet heater is capable of at least 10.9 KW input.

(3)

Operation with a total flow within 10% of design flow.

c.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 30 days on a STAGGERED TEST BASIS by operating each circuit for a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.

d.

Anytime the HEPA filter bank or the charcoal absorbers have been partially or completely replaced, the test per 4.5.H.1.a(as applicable) will be performed prior to returning the system to OPERABLE STATUS.

e.

Automatic initiation of each circuit at the frequency specified in the Surveillance Frequency Control Programevery 18 rnonths.

I.

Inerting Surveillance When an inert atmosphere is required in the primary containment, the oxygen concentration in the primary containment shall be checked at the frequency specified in the Surveillance Frequency Control Programat least weekly.

J.

Drywell Coating Surveillance Carbon steel test panels coated with Firebar 0 shall be placed inside the drywell near the reactor core midplane level. They shall be removed for visual observation and weight loss measurements during the first, second, fourth and eighth refueling outages.

K.

Instrument Line Flow Check Valves Surveillance The capability of a representative sample of instrument line flow check valves to isolate shall be tested at the frequency specified in the Surveillance Frequency Control Programat least once per 24 rnonths. In addition, each time an instrument line is returned to service after any condition which could have produced a pressure flow disturbance in that line, the open position of the flow check valve in that line shall be verified. Such conditions include:

OYSTER CREEK 4.5-5 Amendment No.: 132,186,216,21Q Corrected by letter of 3/18/02

Leakage at instrument fittings and valves Venting an unisolated instrument or instrument line Flushing or draining an instrument Installation of a new instrument or instrument line L.

Suppression Chamber Surveillance 1.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per day the suppression chamber water level and temperature and pressure suppression system pressure shall be checked.

2.

A visual inspection of the suppression chamber interior, including water line regions, shall be made at the frequency specified in the Surveillance Frequency Control Programat each major refueling outage.

3.

Whenever heat from relief valve operation is being added to the suppression pool, the pool temperature shall be continually monitored and also observed until the heat addition is terminated.

4.

Whenever operation of a relief valve is indicated and the suppression pool temperature reaches 160°F or above while the reactor primary coolant system pressure is greater than 180 psig, an external visual examination of the suppression chamber shall be made before resuming normal power operation.

M.

Shock Suppressors (Snubbers)

As used in this specification, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.

1.

Each snubber shall be demonstrated OPERABLE by performance of the following inspection program:

a.

Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of the categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 4.5-1. The visual inspection interval for each type of snubber shall be determined based upon the criteria provided in Table 4.5-1.

OYSTER CREEK 4.5-6 Amendment No.: 182, 186, 216 Leakage at instrument fittings and valves Venting an unisolated instrument or instrument line Flushing or draining an instrument Installation of a new instrument or instrument line L.

Suppression Chamber Surveillance 1.

At the frequency specified in the Surveillance Frequency Control ProgramAt-least once per day the suppression chamber water level and temperature and pressure suppression system pressure shall be checked.

2.

A visual inspection of the suppression chamber interior, including water line regions, shall be made at the frequency specified in the Surveillance Frequency Control Programat each major refueling outage.

3.

Whenever heat from relief valve operation is being added to the suppression pool, the pool temperature shall be continually monitored and also observed until the heat addition is terminated.

4.

Whenever operation of a relief valve is indicated and the suppression pool temperature reaches 160°F or above while the reactor primary coolant system pressure is greater than 180 psig, an external visual examination of the suppression chamber shall be made before resuming normal power operation.

M.

Shock Suppressors (Snubbers)

As used in this specification, "type of snubber" shall mean snubbers of the same design and manufacturer, irrespective of capacity.

1.

Each snubber shall be demonstrated OPERABLE by performance of the following inspection program:

a.

Visual Inspections Snubbers are categorized as inaccessible or accessible during reactor operation. Each of the categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 4.5-1. The visual inspection interval for each type of snubber shall be determined based upon the criteria provided in Table 4.5-1.

OYSTER CREEK 4.5-6 Amendment No.: 182, 186,216

f.

Snubber Service Life Monitoring A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by Specification 6.10.2.1.

Concurrent with the first inservice visual inspection and at least once per 24 months thereafter, the installation and maintenance records for each snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review.

If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be re-evaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall be indicated in the records. Service life shall not at any time affect reactor operations.

N.

Secondary Containment Isolation Valves 1.

Each secondary containment isolation valve shall be demonstrated operable prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator by cycling the valve through at least one complete cycle of full travel. Following maintenance, repair or replacement work on the control or power circuit for the valves, the affected component shall be tested to assure it will perform its intended function in the circuit.

2.

At the frequency specified in the Surveillance Frequency Control ProgramAt lea6t once per refueling outage, all valves shall be tested for automatic closure by an isolation signal.

OYSTER CREEK 4.5-9 Amendment No.: 168,186,219 f.

Snubber Service Life Monitoring A record of the service life of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained as required by Specification 6.10.2.1.

Concurrent with the first inservice visual inspection and at least once per 24 months thereafter, the installation and maintenance records for each snubber shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be re-evaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall be indicated in the records. Service life shall not at any time affect reactor operations.

N.

Secondary Containment Isolation Valves 1.

Each secondary containment isolation valve shall be demonstrated operable prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator by cycling the valve through at least one complete cycle of full travel. Following maintenance, repair or replacement work on the control or power circuit for the valves, the affected component shall be tested to assure it will perform its intended function in the circuit.

2.

At the frequency specified in the Surveillance Frequency Control Programl\\t least onGe per refueling outage, all valves shall be tested for automatic closure by an isolation signal.

OYSTER CREEK 4.5-9 Amendment No.: 168,186,219

A Primary Containment Leakage Rate Testing Program has been established to implement the requirements of 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

Guidance for implementation of Option B is contained in NRC Regulatory Guide 1.163, Performance Based Containment Leak Test Program, Revision 0, dated September 1995.

Additional guidance for NRC Regulatory Guide 1.163 is contained in Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance Based Option of 10 CFR 50, Appendix J, Revision 0, dated July 26, 1995, and ANSI/ANS 56.8-1 994, Containment System Leakage Testing Requirements. The Primary Containment Leakage Rate Testing Program conforms with this guidance as modified by approved exemptions.

The maximum allowable leakage rate for the primary containment (La) is 1.0 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design basis LOCA maximum peak containment pressure (Pa). As discussed below, Pa for the purpose of containment leak rate testing is 35 psig.

The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends. Whenever a double gasketed penetration (primary containment head equipment hatches and the absorption chamber access hatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of 35 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure.

Monitoring the nitrogen makeup requirements of the inerting system provides a method of observing leak rate trends and would detect gross leaks in a very short time. This equipment must be periodically removed from service for test and maintenance, but this out-of-service time be kept to a practical minimum.

Automatic primary containment isolation valves are provided to maintain PRIMARY CONTAINMENT INTEGRITY following the design basis loss-of-coolant accident. Closure times for the automatic primary containment isolation valves are not critical because it is on the order of minutes before significant fission product release to the containment atmosphere for the design basis loss of coolant accident. These valves are highly reliable, see infrequent service and most of them are normally in the closed position. Therefore, a-testing at the frequency specified in the Surveillance Frequency Control Programduring each REFUELING OUTAGE is sufficient.

Large lines connecting to the reactor coolant system, whose failure could result in uncovering the reactor core, are supplied with automatic isolation valves (except containment cooling).

Closure times restrict coolant loss from the circumferential rupture of any of these lines outside primary containment to less than that for a main steam line break (the design basis accident for outside containment line breaks). The minimum time for main steam isolation valve (MS IV) closure of 3 seconds is based on the transient analysis that shows the pressure peak 76 psig below the lowest safety valve setting. The maximum time for MSIV closure of 10 seconds is based on the value assumed for the main steam line break dose calculations and restricts coolant loss to prevent uncovering the reactor core. Per the ASME Code, the full closure test of the MSIVs during COLD SHUTDOWNs will ensure OPERABILITY and provide assurance that the valves maintain the required closing time. The provision for a minimum of 92 days between the tests ensures that full closure testing is not too frequent. The MSIVs are partially stroked quarterly periodically as part of reactor protection system instrument surveillance testing.

OYSTER CREEK 4.5-11 Amendment No.: 132,186,196,219,221,250, A Primary Containment Leakage Rate Testing Program has been established to implement the requirements of 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

Guidance for implementation of Option B is contained in NRC Regulatory Guide 1.163, "Performance Based Containment Leak Test Program", Revision 0, dated September 1995.

Additional guidance for NRC Regulatory Guide 1.163 is contained in Nuclear Energy Institute (NEI) 94-01, "Industry Guideline for Implementing Performance Based Option of 10 CFR 50, Appendix J," Revision 0, dated July 26, 1995, and ANSI/ANS 56.8-1994, "Containment System Leakage Testing Requirements." The Primary Containment Leakage Rate Testing Program conforms with this guidance as modified by approved exemptions.

The maximum allowable leakage rate for the primary containment (La) is 1.0 percent by weight of the containment air per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design basis LOCA maximum peak containment pressure (Pa). As discussed below, Pafor the purpose of containment leak rate testing is 35 psig.

The penetration and air purge piping leakage test frequency, along with the containment leak rate tests, is adequate to allow detection of leakage trends. Whenever a double gasketed penetration (primary containment head equipment hatches and the absorption chamber access hatch) is broken and remade, the space between the gaskets is pressurized to determine that the seals are performing properly. The test pressure of 35 psig is consistent with the accident analyses and the maximum preoperational leak rate test pressure.

Monitoring the nitrogen makeup requirements of the inerting system provides a method of observing leak rate trends and would detect gross leaks in a very short time. This equipment must be periodically removed from service for test and maintenance, but this out-of-service time be kept to a practical minimum.

Automatic primary containment isolation valves are provided to maintain PRIMARY CONTAINMENT INTEGRITY following the design basis loss-of-coolant accident. Closure times for the automatic primary containment isolation valves are not critical because it is on the order of minutes before significant fission product release to the containment atmosphere for the design basis loss of coolant accident. These valves are highly reliable, see infrequent service and most of them are normally in the closed position. Therefore, a-testing at the frequency specified in the Surveillance Frequency Control Programduring each REFUELING OUTAGE is sufficient.

Large lines connecting to the reactor coolant system, whose failure could result in uncovering the reactor core, are supplied with automatic isolation valves (except containment cooling).

Closure times restrict coolant loss from the circumferential rupture of any of these lines outside primary containment to less than that for a main steam line break (the design basis accident for outside containment line breaks). The minimum time for main steam isolation valve (MSIV) closure of 3 seconds is based on the transient analysis that shows the pressure peak 76 psig below the lowest safety valve setting. The maximum time for MSIV closure of 10 seconds is based on the value assumed for the main steam line break dose calculations and restricts coolant loss to prevent uncovering the reactor core. Per the ASME Code, the full closure test of the MSIVs during COLD SHUTDOWNs will ensure OPERABILITY and provide assurance that the valves maintain the required closing time. The provision for a minimum of 92 days between the tests ensures that full closure testing is not too frequent. The MSIVs are partially stroked quarterly periodically as part of reactor protection system instrument surveillance testing.

OYSTER CREEK 4.5-11 Amendment No.: 1J2,186,1Q6,21Q,221,250,

~

Surveillance of the suppression chamber-reactor building vacuum breaker consists of OPERABILITY checks and leakage tests (conducted as part of the containment leak-tightness tests). These vacuum breakers are normally in the closed position and open only during tests or an accident condition. As a result, a testing frequency of three months for OPERABILITY is considered justified for this equipment. Inspections and calibrations are performed during the REFUELING OUTAGEs, this frequency being based on equipment quality, experience, and engineering judgement.The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.

The 14 suppression chamber-drywell vacuum relief valves are designed to open to the full open position (the position that curtain area is equivalent to valve bore) with a force equivalent to a 0.5 psi differential acting on the suppression chamber face of the valve disk. This opening specification assures that the design limit of 2.0 psid between the drywell and external environment is not exceeded. At the frequency specified in the Surveillance Frequency Control ProgramOnce each REFUELING OUTAGE, each valve is tested to assure that it will open in response to a force less than that specified. Also, it is inspected to assure that it closes freely and operates properly.

The containment design has been examined to establish the allowable bypass area between the drywell and suppression chamber as 10.5 in2 (expressed as vacuum breaker open area). This is equivalent to one vacuum breaker disk off its seat 0.371 inch; this length corresponds to an angular displacement of 1.25°. A conservative allowance of 0.10 inch has been selected as the maximum permissible valve opening. Valve closure within this limit may be determined by light indication from two independent position detection and indication systems. Either system provides a control room alarm for a non-seated valve.

At the end of each refueling cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by at least 1 psi with respect to the suppression chamber pressure. The pressure transient (if any) will be monitored with a sensitive pressure gauge.

If the drywell pressure cannot be increased by I psi over the suppression chamber pressure it would be because a significant leakage path exists: in this event, the leakage source will be identified and eliminated before POWER OPERATION is resumed.

If the drywell pressure can be increased by 1 psi over the suppression chamber, the rate of change of the suppression chamber pressure must not exceed a rate equivalent to the rate of air flow from the drywell to the suppression chamber through a 2-inch orifice.

In the event the rate of change of pressure exceeds this value, then the source of leakage will be identified and eliminated before POWER OPERATION is resumed.

The drywell suppression chamber vacuum breakers are exercised at the frequency specified in the Surveillance Frequency Control Programevery 3 months and immediately following termination of discharge of steam into the suppression chamber. This monitoring of valve operability is intended to assure that valve operability and position indication system performance does not degrade between refueling inspections. When a vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights are designed to function as follows:

Full Closed 2 Green

- On (Closed to 0.10 open) 1 Red

- Off Open 0.10 2 Green

- Off (0.10 open to full open) 2 Red

- On OYSTER CREEK 4.5-12 Amendment No. 128,186,196,210,211,219 Corrected by letters of 3/18/02 and 4/5/02 Surveillance of the suppression chamber-reactor building vacuum breaker consists of OPERABILITY checks and leakage tests (conducted as part of the containment leak-tightness tests). These vacuum breakers are normally in the closed position and open only during tests or an accident condition. As a result, a testing frequency of three months for OPERABILITY is considered justified for this equipment. Inspections and calibrations are performed during the REFUELING OUTAGEs, this frequency being based on equipment quality, experience, and engineering judgement.The Surveillance Frequencies are controlled under the Surveillance Frequency Control Program.

The 14 suppression chamber-drywell vacuum relief valves are designed to open to the full open position (the position that curtain area is equivalent to valve bore) with a force equivalent to a 0.5 psi differential acting on the suppression chamber face of the valve disk. This opening specification assures that the design limit of 2.0 psid between the drywell and external environment is not exceeded. At the frequency specified in the Surveillance Frequency Control ProgramOnce each REFUELING OUTAGE, each valve is tested to assure that it will open in response to a force less than that specified. Also, it is inspected to assure that it closes freely and operates properly.

The containment design has been examined to establish the allowable bypass area between the drywell and suppression chamber as 10.5 in2 (expressed as vacuum breaker open area). This is equivalent to one vacuum breaker disk off its seat 0.371 inch; this length corresponds to an angular displacement of 1.25°. A conservative allowance of 0.10 inch has been selected as the maximum permissible valve opening. Valve closure within this limit may be determined by light indication from two independent position detection and indication systems. Either system provides a control room alarm for a non-seated valve.

At the end of each refueling cycle, a leak rate test shall be performed to verify that significant leakage flow paths do not exist between the drywell and suppression chamber. The drywell pressure will be increased by at least 1 psi with respect to the suppression chamber pressure. The pressure transient (if any) will be monitored with a sensitive pressure gauge. If the drywell pressure cannot be increased by 1 psi over the suppression chamber pressure it would be because a significant leakage path exists: in this event, the leakage source will be identified and eliminated before POWER OPERATION is resumed. If the drywell pressure can be increased by 1 psi over the suppression chamber, the rate of change of the suppression chamber pressure must not exceed a rate equivalent to the rate of air flow from the drywell to the suppression chamber through a 2-inch orifice. In the event the rate of change of pressure exceeds this value, then the source of leakage will be identified and eliminated before POWER OPERATION is resumed.

The drywell suppression chamber vacuum breakers are exercised at the frequency specified in I the Surveillance Frequency Control Programevery 3 months and immediately following termination of discharge of steam into the suppression chamber. This monitoring of valve operability is intended to assure that valve operability and position indication system performance does not degrade between refueling inspections. When a vacuum breaker valve is exercised through an opening-closing cycle, the position indicating lights are designed to function as follows:

OYSTER CREEK Full Closed (Closed to 0.10" open)

Open 0.10" (0.10" open to full open) 4.5-12 2 Green - On 1 Red - Off 2 Green - Off 2 Red - On Amendment No. 128,186,196,210,211,219 Corrected by letters of 3/18/02 and 4/5/02

At the frequency specified in the Surveillance Frequency Control ProgramDuring each refueling outage, four suppression chamber-drywell vacuum breakers will be inspected to assure components have not deteriorated. Since valve internals are designed for a 40-year lifetime, an inspection program which cycles through all valves in about 1/10th of the design lifetime is extremely conservative. The alarm systems for the vacuum breakers will be calibrated at the frequency specified in the Surveillance Frequency Control Programduring each refueling outage. This frequency is based on experience and engineering judgement.

Initiating reactor building isolation and operation of the standby gas treatment system to maintain a 1/4 inch of water vacuum, tests the operation of the reactor building isolation valves, leakage tightness of the reactor building and performance of the standby gas treatment system. Checking the initiating sensors and associated trip channels demonstrates the capability for automatic actuation. Performing the reactor building in leakage test prior to refueling demonstrates secondary containment capability prior to extensive fuel handling operations associated with the outage.

Verifying the efficiency and operation of charcoal filters at the frequency specified in the Surveillance Frequency Control Programonce per 18 months gives sufficient confidence of standby gas treatment system performance capability. A charcoal filter efficiency of 99% for halogen removal is adequate.

The in-place testing of charcoal filters is performed using halogenated hydrocarbon refrigerant which is injected into the system upstream of the charcoal filters. Measurement of the refrigerant concentration upstream and downstream of the charcoal filters is made using a gas chromatograph. The ratio of the inlet and outlet concentrations gives an overall indication of the leak tightness of the system. Although this is basically a leak test, since the filters have charcoal of known efficiency and holding capacity for elemental iodine and/or methyl iodide, the test also gives an indication of the relative efficiency of the installed system. The test procedure is an adaptation of test procedures developed at the Savannah River Laboratory which were described in the Ninth AEC Cleaning Conference.*

High efficiency particulate filters are installed before and after the charcoal filters to minimize potential releases of particulates to the environment and to prevent clogging of the iodine filters.

An efficiency of 99% is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by testing with DOP at testing medium.

The 95% methyl iodide removal efficiency is based on the formula in GL 99-02 for allowable penetration [(100%

- 90% credited in DBA analysis) divided by a safety factor of 2].

If the allowable penetration is 5%, the required removal efficiency is 95%.

If laboratory tests for the adsorber material in one circuit of the Standby Gas Treatment System are unacceptable, all adsorber material in that circuit shall be replaced with adsorbent qualified according to Regulatory Guide 1.52. Any HEPA filters found defective shall be replaced with those qualified with Regulatory Position C.3.d of Regulatory Guide 1.52.

• D.R. Muhabier. In Place Nondestructive Leak Test for Iodine Adsorbers. Proceedings of the Ninth AEC Air Cleaning Conference. USAEC Report CONF-660904, 1966 OYSTER CREEK 4.5-13 Amendment No.: 186, 195, 219 At the frequency specified in the Surveillance Frequency Control ProgramDuring eaoh refueling outage, four suppression chamber-drywell vacuum breakers will be inspected to assure components have not deteriorated. Since valve internals are designed for a 40-year lifetime, an inspection program which cycles through all valves in about 1/1 Oth of the design lifetime is extremely conservative. The alarm systems for the vacuum breakers will be calibrated at the frequency specified in the Surveillance Frequency Control Programduring eaoh refueling outage. This frequenoy is based on experienoe and engineering judgement.

Initiating reactor building isolation and operation of the standby gas treatment system to maintain a 1/4 inch of water vacuum, tests the operation of the reactor building isolation valves, leakage tightness of the reactor building and performance of the standby gas treatment system. Checking the initiating sensors and associated trip channels demonstrates the capability for automatic actuation. Performing the reactor building in leakage test prior to refueling demonstrates secondary containment capability prior to extensive fuel handling operations associated with the outage.

Verifying the efficiency and operation of charcoal filters at the frequency specified in the Surveillance Frequency Control Programonoe per 18 months gives sufficient confidence of standby gas treatment system performance capability. A charcoal filter efficiency of 99%

for halogen removal is adequate.

The in-place testing of charcoal filters is performed using halogenated hydrocarbon refrigerant which is injected into the system upstream of the charcoal filters. Measurement of the refrigerant concentration upstream and downstream of the charcoal filters is made using a gas chromatograph. The ratio of the inlet and outlet concentrations gives an overall indication of the leak tightness of the system. Although this is basically a leak test, since the filters have charcoal of known efficiency and holding capacity for elemental iodine and/or methyl iodide, the test also gives an indication of the relative efficiency of the installed system. The test procedure is an adaptation of test procedures developed at the Savannah River Laboratory which were described in the Ninth AEC Cleaning Conference.*

High efficiency particulate filters are installed before and after the charcoal filters to minimize potential releases of particulates to the environment and to prevent clogging of the iodine filters.

An efficiency of 99%

is adequate to retain particulates that may be released to the reactor building following an accident. This will be demonstrated by testing with DOP at testing medium.

The 950/0 methyl iodide removal efficiency is based on the formula in GL 99-02 for allowable penetration [(100%

- 900/0 credited in DBA analysis) divided by a safety factor of 2]. If the allowable penetration is ~5%, the required removal efficiency is ~95%. If laboratory tests for the adsorber material in one circuit of the Standby Gas Treatment System are unacceptable, all adsorber material in that circuit shall be replaced with adsorbent qualified according to Regulatory Guide 1.52. Any HEPA filters found defective shall be replaced with those qualified with Regulatory Position C.3.d of Regulatory Guide 1.52.

• D.R. Muhabier. "In Place Nondestructive Leak Test for Iodine Adsorbers." Proceedings of the Ninth AEC Air Cleaning Conference. USAEC Report CONF-660904, 1966 OYSTER CREEK 4.5-13 Amendment No.: 186, 195,219

The operability of the instrument line flow check valves are demonstrated to assure isolation capability for excess flow and to assure the operability of the instrument sensor when required.

The representative sample consists of an approximately equal number of EFCVs such that each EFCV is tested at least every 10 years (nominal). The nominal 10 year interval is based on other performance-based testing programs, such as Inservice Testing (snubbers) and Option B to 10 CFR 50, Appendix J, EFCV test failures will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability is maintained. Operating experience has Demonstrated that these components are highly reliable and that failures to isolate are very Infrequent. Therefore, testing of a representative sample was concluded to be acceptable from a relaibility standpoint.

9 Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily periodically is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and also observed during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

References (1)

Licensing Application, Amendment 32, Question 3 (2)

FDSAR, Volume I, Section V-1.1 (3)

GE-NE 770-07-1 090, Oyster Creek LOCA Drywell Pressure Response, February 1991 (4)

Deleted (5)

FDSAR, Volume I, Sections v-I.5 and V-i.6 (6)

FDSAR, Volume I, Sections V-i.6 and Xlll-3.4 (7)

FDSAR, Volume I, Section XIll-2 (8)

Licensing Application, Amendment ii, Question 111-18 (9)

GE BWROG B2i-00658-0l, Excess Flow Check Valve Testing Relaxation, dated November 1998 OYSTER CREEK 4.5-15 Amendment No.: 165,186, 216, 219 Corrected by letter 3/1 8/02 The operability of the instrument line flow check valves are demonstrated to assure isolation capability for excess flow and to assure the operability of the instrument sensor when required.

The representative sample consists of an approximately equal number of EFCV's such that each EFCV is tested at least every 10 years (nominal). The nominal 10 year interval is based on other performance-based testing programs, such as Inservice Testing (snubbers) and Option B to 10 CFR 50, Appendix J, EFCV test failures will be evaluated to determine if additional testing in that test interval is warranted to ensure overall reliability is maintained. Operating experience has Demonstrated that these components are highly reliable and that failures to isolate are very Infrequent. Therefore, testing of a representative sample was concluded to be acceptable from a relaibility standpoint. (9)

Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters GaUy-periodically is I sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and also observed during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance that no significant damage was encountered. Particular attention should be focused on structural discontinuities in the vicinity of the relief valve discharge since these are expected to be the points of highest stress.

References (1)

Licensing Application, Amendment 32, Question 3 (2)

FDSAR, Volume I, Section V-1.1 (3)

GE-NE 770-07-1090, "Oyster Creek LOCA Drywell Pressure Response," February 1991 (4)

Deleted (5)

FDSAR, Volume I, Sections V-1.5 and V-1.6 (6)

FDSAR, Volume I, Sections V-1.6 and XIII-3.4 (7)

FDSAR, Volume I, Section XIII-2 (8)

Licensing Application, Amendment 11, Question 111-18 (9)

GE BWROG B21-00658-01, "Excess Flow Check Valve Testing Relaxation," dated November 1998 OYSTER CREEK 4.5-15 Amendment No.: 165,186,216,219 Corrected by letter 3/18/02

4.6 RADIOACTIVE EFFLUENT Applicability:

Applies to monitoring of gaseous and liquid radioactive effluents of the Station during release of effluents via the monitored pathway(s). Each Surveillance Requirement applies whenever the corresponding Specification is applicable unless otherwise stated in an individual Surveillance Requirement. Surveillance Requirements do not have to be performed on inoperable equipment.

Obiective:

To measure radioactive effluents adequately to verify that radioactive effluents are as low as is reasonable achievable and within the limit of 10 CFR Part 20.

Specification:

A.

Reactor Coolant Reactor coolant shall be sampled and analyzed at the frequency specified in the Surveillance Frequency Control Programat least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for DOSE EQUIVALENT 1-131 during RUN MODE, STARTUP MODE and SHUTDOWN CONDITION.

B.

NOT USED.

C.

Radioactive Liguid Storage 1.

Liquids contained in the following tanks shall be sampled and analyzed for radioactivity at the frequency specified in the Surveillance Frequency Control Programat least once per 7 days when radioactive liquid is being added to the tank:

a.

Waste Surge Tank, HP-T-3; b.

Condensate Storage Tank.

D.

Main Condenser Offcias Treatment RELOCATED TO THE ODCM.

E.

Main Condenser Offcias Radioactivity 1.

The gross radioactivity in fission gases discharged from the main condenser air ejector shall be measured by sampling and analyzing the gases.

a.

At the frequency specified in the Surveillance Frequency Control Programat least once per month, and b.

When the reactor is operating at more than 40 percent of rated power, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after an increase in the fission gas release via the air ejector of more than 50 percent, as indicated by the Condenser Air Ejector Offgas Radioactivity Monitor after factoring out increase(s) due to change(s) in the THERMAL POWER level.

F.

Condenser Offqas Hydrogen Concentration The concentration of hydrogen in offgases downstream of the recombiner in the Offgas System shall be monitored with hydrogen instrumentation as described in Table 3.15.2.

G.

NOT USED.

H.

NOT USED.

OYSTER CREEK 4.6-1 Amendment No.: 108,126,166,191, 266 4.6 RADIOACTIVE EFFLUENT Applicability:

Applies to monitoring of gaseous and liquid radioactive effluents of the Station during release of effluents via the monitored pathway(s). Each Surveillance Requirement applies whenever the corresponding Specification is applicable unless otherwise stated in an individual Surveillance Requirement. Surveillance Requirements do not have to be performed on inoperable equipment.

Objective:

To measure radioactive effluents adequately to verify that radioactive effluents are as low as is reasonable achievable and within the limit of 10 CFR Part 20.

Specification:

A.

Reactor Coolant Reactor coolant shall be sampled and analyzed at the frequency specified in the Surveillance Frequency Control Programat least once every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for DOSE EQUIVALENT 1-131 during RUN MODE, STARTUP MODE and SHUTDOWN CONDITION.

B.

NOT USED.

C.

Radioactive liqUid Storage 1.

Liquids contained in the following tanks shall be sampled and analyzed for radioactivity at the frequency specified in the Surveillance Frequency Control Programat least once per 7 days when radioactive liquid is being added to the tank:

a.

Waste Surge Tank, HP-T-3; b.

Condensate Storage Tank.

D.

Main Condenser Offgas Treatment RELOCATED TO THE ODCM.

E.

Main Condenser Offgas Radioactivity 1.

The gross radioactivity in fission gases discharged from the main condenser air ejector shall be measured by sampling and analyzing the gases.

a.

At the frequency specified in the Surveillance Frequency Control Programat least once per month, and b.

When the reactor is operating at more than 40 percent of rated power, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after an increase in the fission gas release via the air ejector of more than 50 percent, as indicated by the Condenser Air Ejector Offgas Radioactivity Monitor after factoring out increase(s) due to change(s) in the THERMAL POWER level.

F.

Condenser Offgas Hydrogen Concentration The concentration of hydrogen in offgases downstream of the recombiner in the Offgas System shall be monitored with hydrogen instrumentation as described in Table 3.15.2.

G.

NOT USED.

H.

NOT USED.

OYSTER CREEK 4.6-1 Amendment No.: 108,12i,1ii,191, 2ii

4.7 AUXILIARY ELECTRICAL POWER Applicability:

Applies to surveillance requirements of the auxiliary electrical supply.

Objective:

To verify the availability of the auxiliary electrical supply.

Specification:

A.

Diesel Generator 1.

Each diesel generator shall be started and loaded to not less than 80% rated load at the frequency specified in the Surveillance Frequency Control Programevery two weeks.

2.

The two diesel generators shall be automatically actuated and functionally tested at the frequency specified in the Surveillance Frequency Control Programduring oach refueling outage. This shall include testing of the diesel generator load sequence timers listed in Table 3.1.1.

3.

Deleted.

4.

The diesel generators fuel supply shall be checked following the above tests.

5.

The diesel generators starting batteries shall be tested and monitored per Specification 4.7.B.

B.

Diesel Generator Starting Batteries 1.

Weekly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the following:

a.

The active metallic surface of the plates shall be fully covered with electrolyte in all batteries.

b.

The designated pilot cell voltage is greater than or equal to 2.0 volts.

c.

The overall battery voltage is greater than or equal to 112 volts while the battery is on a float charge.

d.

The pilot cell specific gravity, corrected to 77° F, is greater than or equal to 1.190.

2.

Quarterly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the specific gravity for each fourth cell is greater than or equal to 1.190 when corrected to 77°F. The specific gravity and electrolyte temperature of every fourth cell shall be recorded for surveillance review.

3.

Annual Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the specific gravity for each cell is greater than or equal to 1.190 when corrected to 77° F. The electrolyte temperature and specific gravity for every cell shall be recorded for surveillance review.

OYSTER CREEK 4.7-1 Amendment No.: 144,189,197, 227, 236,245 Corrected by letter of 10/15/2004 4.7 AUXILIARY ELECTRICAL POWER Applicability:

Applies to surveillance requirements of the auxiliary electrical supply.

Objective:

Specification:

To verify the availability of the auxiliary electrical supply.

A.

Diesel Generator 1.

Each diesel generator shall be started and loaded to not less than 80%

rated load at the frequency specified in the Surveillance Frequency Control Programevery two weeks.

2.

The two diesel generators shall be automatically actuated and functionally tested at the frequency specified in the Surveillance Frequency Control Programduring each refueling eutage. This shall include testing of the diesel generator load sequence timers listed in Table 3.1.1.

3.

Deleted.

4.

The diesel generators' fuel supply shall be checked following the above tests.

5.

The diesel generators' starting batteries shall be tested and monitored per Specification 4.7.B.

B.

Diesel Generator Starting Batteries 1.

'Neekly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the following:

a.

The active metallic surface of the plates shall be fully covered with electrolyte in all batteries.

b.

The designated pilot cell voltage is greater than or equal to 2.0 volts.

c.

The overall battery voltage is greater than or equal to 112 volts while the battery is on a float charge.

d.

The pilot cell specific gravity, corrected to 77°F, is greater than or equal to 1.190.

2.

Quarterly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the specific gravity for each fourth cell is greater than or equal to 1.190 when corrected to 77°F. The specific gravity and electrolyte temperature of every fourth cell shall be recorded for surveillance review.

3.

Annual Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the specific gravity for each cell is greater than or equal to 1.190 when corrected to 77°F. The electrolyte temperature and specific gravity for every cell shall be recorded for surveillance review.

OYSTER CREEK 4.7-1 Amendment No.: 144,189,197,227,236,245 Cerrected by letter of 10/15/2004

4.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 12 months, the diesel generator battery capacity shall be demonstrated to be able to supply the design duty loads (diesel start) during a battery service test.

5.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 24 months, the following tests will be performed-(perform during plant shutdowns or during 2l-month Diesel Generator inspections):

a.

Battery capacity shall be demonstrated to be at least 80% of the manufacturers rating when subjected to a battery capacity discharge test.

b.

If a Diesel Generator Starting Battery is demonstrated to have less than 85% of manufacturers ratings during a capacity discharge test, it shall be replaced within 2 years.

C.

Station Batteries Weekly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the following:

a.

The overall battery voltage is greater than or equal to the minimum established float voltage.

b.

Each station battery float current is 2 amps when battery terminal voltage is greater than or equal to the minimum established float voltage of 4.7.C.1.a.

2.

Monthly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the following:

a.

The electrolyte level in each station battery is greater than or equal to minimum established design limits.

b.

The voltage of each pilot cell is greater than or equal to 2.07 volts while the respective battery is on a float charge.

c.

The electrolyte temperature of each station battery pilot cell is greater than or equal to minimum established design limits.

3.

Quarterly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the voltage of each connected cell is greater than or equal to 2.07 volts while the respective battery is on a float charge.

Oyster Creek 4.7-2 Amendment No. 142, 189, 197, 227, 245 Corrected by lotter of 10/15/2004 4.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 12 months, the diesel generator battery capacity shall be demonstrated to be able to supply the design duty loads (diesel start) during a battery service test.

5.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 24 months, the following tests will be performed-(perform during plant shutdowns or during 24 month Diesel Generator inspections):

a.

Battery capacity shall be demonstrated to be at least 80%

of the manufacturers' rating when subjected to a battery capacity discharge test.

b.

If a Diesel Generator Starting Battery is demonstrated to have less than 85%

of manufacturers ratings during a capacity discharge test, it shall be replaced within 2 years.

C.

Station Batteries 1.

VVeekly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the following:

a.

The overall battery voltage is greater than or equal to the minimum established float voltage.

b.

Each station battery float current is s 2 amps when battery terminal voltage is greater than or equal to the minimum established float voltage of 4.7.C.1.a.

2.

Monthly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the following:

a.

The electrolyte level in each station battery is greater than or equal to minimum established design limits.

b.

The voltage of each pilot cell is greater than or equal to 2.07 volts while the respective battery is on a float charge.

c.

The electrolyte temperature of each station battery pilot cell is greater than or equal to minimum established design limits.

3.

Quarterly Surveillance will be performed at the frequency specified in the Surveillance Frequency Control Program to verify the voltage of each connected cell is greater than or equal to 2.07 volts while the respective battery is on a float charge.

Oyster Creek 4.7-2 Amendment No. 142, 189, 197,227,245 Corrected by letter of 10/15/2004

4.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 24 months:

a.

The station battery capacity shall be demonstrated to be able to supply the design duty cycle loads during a battery service test. The modified performance discharge test may be substituted for the service test.

b.

(i)

Verify required station battery charger supplies 429 amps for the B MG Set charger, 600 amps for the NB static charger, and 500 amps for the C charger, for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at greater than or equal to the minimum established float voltage, or (ii)

Verify each required battery charger can recharge the battery to the fully charged state while supplying the normal steady state DC loads during station operation, after a battery discharge to the bounding design basis event discharge state.

5.

The following tests will be performed to verify battery capacity (perform during nIgt shutdowns for Station Batteries B and (

a.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 60 months, battery capacity shall be demonstrated to be at least 80% of the manufacturers rating when subjected to a performance discharge test or a modified performance discharge test.

b.

Performance discharge tests or modified performance discharge tests of station battery capacity shall be given at least once per 12 months when:

(i)

The station battery shows degradation, or (ii)

The station battery has reached 85% of expected life with battery capacity < 100% of manufacturers rating.

c.

Performance discharge tests or modified performance discharge tests of station battery capacity shall be given at least once per 24 months when the battery has reached 85% of expected life with battery capacity 100% of manufacturers rating.

Oyster Creek 4.7-3 Amendment No. 245 4.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 24 months:

a.

The station battery capacity shall be demonstrated to be able to supply the design duty cycle loads during a battery service test. The modified performance discharge test may be substituted for the service test.

b.

(i)

Verify required station battery charger supplies ~ 429 amps for the B MG Set charger, ~ 600 amps for the AlB static charger, and ~ 500 amps for the C charger, for ~ 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at greater than or equal to the minimum established float voltage, or (ii)

Verify each required battery charger can recharge the battery to the fully charged state while supplying the normal steady state DC loads during station operation, after a battery discharge to the bounding design basis event discharge state.

5.

The following tests will be performed to verify battery capacity (perform during plant shutdowns for Station Batteries B and C):

a.

At the frequency specified in the Surveillance Frequency Control ProgramAt least once per 60 months, battery capacity shall be demonstrated to be at least 80%

of the manufacturers' rating when subjected to a performance discharge test or a modified performance discharge test.

b.

Performance discharge tests or modified performance discharge tests of station battery capacity shall be given at least once per 12 months when:

(i)

The station battery shows degradation, or (ii)

The station battery has reached 850/0 of expected life with battery capacity < 100% of manufacturer's rating.

c.

Performance discharge tests or modified performance discharge tests of station battery capacity shall be given at least once per 24 months when the battery has reached 85%

of expected life with battery capacity

~ 100% of manufacturer's rating.

Oyster Creek 4.7-3 Amendment No. ~

Basis: The biweekly tests of the diesel generator& tests are primarily to check for failures and deterioration in the system since last use. The manufacturer has recommended the two week test interval, based on experience with many of their engines. One factor in determining this test interval (besides checking whether or not the engine starts and runs) is that the lubricating oil should be circulated through the engine approximately every two wooksperiodically. The diesels should be loaded to at least 80% of rated load until engine and generator temperatures have stabilized (about one hour). The minimum 80% load will prevent soot formation in the cylinders and injection nozzles.

Operation up to an equilibrium temperature ensures that there is no over-heat problem.

The tests also provide an engine and generator operating history to be compared with subsequent engine-generator test data to identify and correct any mechanical or electrical deficiency before it can result in a system failure.

The test during refueling outages is more comprehensive tests, including procedures that are most effectively conducted at that time. Theseperformed at the frequency specified in the Surveillance Frequency Control Program, include automatic actuation and functional capability tests, to verify that the generators can start and assume load in less than 20 seconds and testing of the diesel generator load sequence timers which provide protection from a possible diesel generator overload during LOCA conditions.

The diesel generator batteries are challenged every two weeksat the frequency specified in the Surveillance Frequency Control Program to perform the 80% load test. This effectively performs an uninstrumented battery service test. The biweekly diesel start, when combined with the annual battery service test, provides an extensive amount of data on battery performance characteristics. This test data negates the need to lower the battery performance test interval from biennial to annually.

The diesel batteries shall be tested and monitored in accordance with the requirements of Specification 4.7.B to ensure their viability. The requirement to replace any battery in-the next refueling outage or within 2 years which demonstrates less than 85% of manufacturers capacity during a capacity discharge test provides additional assurance of continued battery operability.

Verifying, per 4.7.C.1.a, battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the battery chargers, which support the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state while supplying the continuous steady state loads of the associated DC subsystem. On float charge, battery cells will receive adequate current to optimally charge the battery. The voltage requirements are based on the minimum float voltage established by the battery manufacturer (2.17 V per cell average, or 130.2 V at the battery terminals). This voltage maintains the battery plates in a condition that supports maintaining the grid life (expected to be approximately 40 years for B station battery; 20 years for C station battery). The weekly frequency is consistent with manufacturer recommendations and IEEE Standard 450-1 995The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

OYSTER CREEK 4.7-4 AMENDMENT NO: 142,189,197,227, 236, 245 Corrected by letter of 10/15/2004 Basis: The biweekly tests of the diesel generators tests are primarily to check for failures and deterioration in the system since last use. The manufacturer has recommended the two week test interval, based on experience with many of their engines. One factor in determining this test interval (besides checking whether or not the engine starts and runs) is that the lubricating oil should be circulated through the engine approximately e'/eF)' two weeksperiodically. The diesels should be loaded to at least 80%

of rated load until engine and generator temperatures have stabilized (about one hour). The minimum 80%

load will prevent soot formation in the cylinders and injection nozzles.

Operation up to an equilibrium temperature ensures that there is no over-heat problem.

The tests also provide an engine and generator operating history to be compared with subsequent engine-generator test data to identify and correct any mechanical or electrical deficiency before it can result in a system failure.

The test during refueling outages is more comprehensive tests, including procedures that are most effectively conducted at that time. Theseperformed at the frequency specified in the Surveillance Frequency Control Program, include automatic actuation and functional capability tests, to verify that the generators can start and assume load in less than 20 seconds and testing of the diesel generator load sequence timers which provide protection from a possible diesel generator overload during LOCA conditions.

The diesel generator batteries are challenged every two weeksat the frequency specified in the Surveillance Frequency Control Program to perform the 80%

load test. This effectively performs an uninstrumented battery service test. The biweekly diesel start, when combined with the annual battery service test, provides an extensive amount of data on battery performance characteristics. This test data negates the need to lower the battery performance test interval from biennial to annually.

The diesel batteries shall be tested and monitored in accordance with the requirements of Specification 4.7.B to ensure their viability. The requirement to replace any battery ffi..-

the next refueling outage or within 2 years which demonstrates less than 85°,fo of manufacturers capacity during a capacity discharge test provides additional assurance of continued battery operability.

Verifying, per 4.7.C.1.a, battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the battery chargers, which support the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state while supplying the continuous steady state loads of the associated DC subsystem. On float charge, battery cells will receive adequate current to optimally charge the battery. The voltage requirements are based on the minimum float voltage established by the battery manufacturer (2.17 V per cell average, or 130.2 V at the battery terminals). This voltage maintains the battery plates in a condition that supports maintaining the grid life (expected to be approximately 40 years for B station battery; 20 years for C station battery). The 'Neekly frequency is consistent with manufacturer recommendations and IEEE Standard 450

+9QaThe Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

OYSTER CREEK 4.7-4 AMENDMENT NO: 142,189,197,227,236,245 Corrected by letter of 10/1512004

Verifying battery float current while on float charge (4.7.C. 1.b) is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state. The float current requirements are based on the float current indicative of a charged battery. Use of float current to determine the state of charge of the battery is consistent with IEEE Standard 450-1995. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe weekly frequency is consistent with IEEE Standard 150-1995.

This Surveillance Requirement (4.7.C.1.b) provides that the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of 4.7.C.1.a. When this float voltage is not maintained the Actions of 3.7.D.1 are being taken, which provide the necessary and appropriate verifications of the battery condition.

Furthermore, the float current limits are established based on the float voltage range and is not directly applicable when this voltage is not maintained.

The 4.7.C.2.a minimum established design limit for electrolyte level ensures that the plates suffer no physical damage and maintains adequate electron transfer capability. For the station batteries, this is the minimum level mark on the side of the battery cell. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is consistent with IEEE 450-1995.

Surveillance Requirements 4.7.C.2.b and 4.7.C.3 require verification that the cell float voltages are equal to or greater than 2.07 V. The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramThe frequencies for cell voltage verification (monthly for pilot cell, and quartorly for each connected cell) are consistent with IEEE Standard 450-1995.

Surveillance Requirement 4.7.C.2.c verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (i.e., 60 degrees Fahrenheit for station battery B; 50 degrees Fahrenheit for station battery C). Cell electrolyte temperature is maintained above these temperatures to assure the battery can provide the required current and voltage to meet the design requirements. Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is consistent with IEEE Standard 450-1995.

A battery service test, per 4.7.C.4.a, is a special test of the station battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC auxiliary electrical power system.

The discharge rate and test length corresponds to the design duty cycle requirements.

Surveillance Requirement 4.7.C.4.b verifies the design capacity of the station battery chargers.

The battery charger supply is based on normal steady state DC loads during station operation and the charging capacity to restore the battery from the design minimum charge state to the fully charged state. The minimum required amperes and duration ensures that these requirements can be satisfied. The battery is recharged when the measured charging current is 2 amps.

Surveillance Requirement 4.7.C.4.b(i) requires that each required station battery charger (i.e.,

only one charger per station battery required for compliance with 3.7.A.4) be capable of supplying the amps listed for the specified charger at the minimum established float voltage for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The ampere requirements are based on the output OYSTER CREEK 4.7-5 Amendment No. 24 Verifying battery float current while on float charge (4.7.C.1.b) is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state. The float current requirements are based on the float current indicative of a charged battery. Use of float current to determine the state of charge of the battery is consistent with IEEE Standard 450-1995. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe weekly frequency is consistent with IEEE Standard 450 1995.

This Surveillance Requirement (4.7.C.1.b) provides that the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of 4.7.C.1.a. When this float voltage is not maintained the Actions of 3.7.D.1 are being taken, which provide the necessary and appropriate verifications of the battery condition.

Furthermore, the float current limits are established based on the float voltage range and is not directly applicable when this voltage is not maintained.

The 4.7.C.2.a minimum established design limit for electrolyte level ensures that the plates suffer no physical damage and maintains adequate electron transfer capability. For the station batteries, this is the minimum level mark on the side of the battery cell. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is consistent with IEEE 450 1995.

Surveillance Requirements 4.7.C.2.b and 4.7.C.3 require verification that the cell float voltages are equal to or greater than 2.07 V. The Surveillance Frequencies are controlled under the Surveillance Frequency Control ProgramThe frequencies for cell voltage verification (monthly for pilot cell, and quarterly for each connected cell) are consistent with IEEE Standard 450 1995.

Surveillance Requirement 4.7.C.2.c verifies that the pilot cell temperature is greater than or equal to the minimum established design limit (Le., 60 degrees Fahrenheit for station battery B; 50 degrees Fahrenheit for station battery C). Cell electrolyte temperature is maintained above these temperatures to assure the battery can provide the required current and voltage to meet the design requirements. Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity. The Surveillance Frequency is controlled under the Surveillance Frequency Control ProgramThe Frequency is consistent with IEEE Standard 450

+Wa.

A battery service test, per 4.7.C.4.a, is a special test of the station battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC auxiliary electrical power system.

The discharge rate and test length corresponds to the design duty cycle requirements.

Surveillance Requirement 4.7.C.4.b verifies the design capacity of the station battery chargers.

The battery charger supply is based on normal steady state DC loads during station operation and the charging capacity to restore the battery from the design minimum charge state to the fully charged state. The minimum required amperes and duration ensures that these requirements can be satisfied. The battery is recharged when the measured charging current is S 2 amps.

Surveillance Requirement 4.7.C.4.b(i) requires that each required station battery charger (Le.,

only one charger per station battery "required" for compliance with 3.7.A.4) be capable of supplying the amps listed for the specified charger at the minimum established float voltage for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The ampere requirements are based on the output OYSTER CREEK 4.7-5 Amendment No. 24a

4.8 ISOLATION CONDENSER Arplicability:

Applies to periodic testing requirements for the isolation condenser system.

Obective:

To verify the operability of the isolation condenser system.

Specification:

A.

Surveillance of each isolation condenser loop shall be as follows:

[tem Freciuency 1.

Operability of motor-Note lOnce/3 months operated isolation valves and condensate makeup valves.

2.

Automatic actuation and Note lEach refueling outage functional test.

(interval not to exceed 20 months) or following major repair.

3.

Shell side water volume Note lOnce/day check 4.

Isolation valve (steam side) a.

Visual inspection Note lEach refueling outage b.

External leakage check Each primary system Leak test c.

Area temperature check Note lOnco/shift Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted above.

Basis:

Motor-operated valves on the isolation condenser steam and condensate lines and on the condensate makeup line that are normally on standby should be exercised periodically to make sure that they are free to operate. The valves will be stroked full length every time they are tested to verify proper functional performance. This frequency of testing is consistent with instrumentation tests discussed in Specification 4.1. Testing of these components per the ASME Code once every 3 monthsat the frequency specified in the Surveillance Frequency Control Program provides assurance of availability of the system. Also, at this frequency of testing, wearout should not be a problem throughout the life of the plant.

The automatic actuation and functional test will demonstrate the automatic opening of the condensate return line valves and the automatic closing of the isolation valves on the vent lines to the main steam lines. Automatic closure of the isolation condenser steam and condensate lines on actuation of the condenser pipe break detectors will also be verified by the test.

It is during a major maintenance or repair that a systems design intent may be violated accidentally. This makes the functional test necessary after every major repair operation.

By virtue of normal plant operation the operators daily observe the water level in the isolation condensers.

In addition, isolation condenser shell side water level sensors provide control room annunciation of condenser high or low water level.

OYSTER CREEK 4.8-1 Amendment No. 209, 268 4.8 ISOLATION CONDENSER Applicability:

Applies to periodic testing requirements for the isolation condenser system.

Objective:

To verify the operability of the isolation condenser system.

Specification:

A.

Surveillance of each isolation condenser loop shall be as follows:

Frequency 1.

2.

3.

4.

Operability of motor-operated isolation valves and condensate makeup valves.

Automatic actuation and functional test.

Shell side water volume check Isolation valve (steam side) a.

Visual inspection b.

External leakage check c.

Area temperature check Note 1Oncela months Note 1Each refueling outage (interval not to exceed 20 months) or following major repair.

Note 1Once/day Note 1Each refueling outage Each primary system Leak test Note 1Once/shift Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted above.

Basis:

Motor-operated valves on the isolation condenser steam and condensate lines and on the condensate makeup line that are normally on standby should be exercised periodically to make sure that they are free to operate. The valves will be stroked full length every time they are tested to verify proper functional performance. This frequency of testing is consistent with instrumentation tests discussed in Specification 4.1. Testing of these components per the ASME Code once e'lePj a monthsat the frequency specified in the Surveillance Frequency Control Program provides assurance of availability of the system. Also, at this frequency of testing, wearout should not be a problem throughout the life of the plant.

The automatic actuation and functional test will demonstrate the automatic opening of the condensate return line valves and the automatic closing of the isolation valves on the vent lines to the main steam lines. Automatic closure of the isolation condenser steam and condensate lines on actuation of the condenser pipe break detectors will also be verified by the test. It is during a major maintenance or repair that a system's design intent may be violated accidentally. This makes the functional test necessary after every major repair operation.

By virtue of normal plant operation the operators daily observe the water level in the isolation condensers. In addition, isolation condenser shell side water level sensors provide control room annunciation of condenser high or low water level.

OYSTER CREEK 4.8-1 Amendment No. 209, 268

At the frequency specified in the Surveillance Frequency Control Program Each refueling outage the insulation will be periodically removed from the steam side isolation valve and the external valve bodies will be inspected for signs of deterioration. Additionally, special attention is specified for these valves during primary system leakage tests and the temperature in the area of these valves is checked at the frequency specified in the Surveillance Frequency Control Programonce each shift for temperature increases that would indicate valve leakage. The special attention given these valves in the design and during their construction along with the indicated surveillance is judged to be adequate to assure that these valves will maintain their integrity when they are required for isolation of the primary containment.

Reference (1)

Licensing Application, Amendment 32, Question 5 OYSTER CREEK 4.8-2 At the frequency specified in the Surveillance Frequency Control Program Each refueling outage the insulation will be periodically removed from the steam side isolation valve and the external valve bodies will be inspected for signs of deterioration. Additionally, special attention is specified for these valves during primary system leakage tests and the temperature in the area of these valves is checked at the frequency specified in the Surveillance Frequency Control Programonce each shift for temperature increases that would indicate valve leakage. The special attention given these valves in the design and during their construction(1) along with the indicated surveillance is judged to be adequate to assure that these valves will maintain their integrity when they are required for isolation of the primary containment.

Reference (1)

Licensing Application, Amendment 32, Question 5 OYSTER CREEK 4.8-2

4.9 REFUELING Applicability:

Applies to the periodic testing of those interlocks and instruments used during refueling.

Objective:

To verify the operability of instrumentation and interlocks in use during refueling.

Specification: A.

The refueling interlocks shall be tested prior to any fuel handling with the head off the reactor vessel, at woekly intervalsat the frequency specified in the Surveillance Frequency Control Program thereafter until no longer required and following any repair work associated with the interlocks.

B.

Prior to beginning any core alterations, the source range monitors (SRMs) shall be calibrated. Thereafter, the SRMs will be checked daily, tested monthly, and calibrated at the frequencies specified in the Surveillance Frequency Control Programevory 3 months until no longer required.

C.

Within four (4) hours prior to the start of control rod removal pursuant to Specification 3.9.E verify:

1.

That the reactor mode switch is locked in the refuel position and that the one rod out refueling interlock is operable.

2.

That two (2) SRM channels, one in the core quadrant where the control rod is being removed and one in an adjacent quadrant, are operable and inserted to the normal operation level.

D.

Verify within four (4) hours prior to the start of control rod removal pursuant to Specification 3.9.F and at the frequency specified in the Surveillance Frequency Control Programat least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, until replacement of all control rods or rod drive mechanisms and all control rods are fully inserted that:

1.

the reactor mode switch is locked in the refuel position and the one rod out refueling interlock is operable.

2.

Two (2) SRM channels, one in the core quadrant where a control rod is being removed and one in an adjacent quadrant, are operable and fully inserted.

3.

All control rods not removed are fully inserted with the exception of one rod which may be partially withdrawn not more than two notches to perform refueling interlock surveillance.

4.

The four fuel assemblies surrounding each control rod or rod drive mechanism being removed or maintained at the same time are removed from the core cell.

OYSTER CREEK 4.9-1 Amendment No.: 23,43 4.9 REFUELING Applicability:

Applies to the periodic testing of those interlocks and instruments used during refueling.

Objective:

To verify the operability of instrumentation and interlocks in use during refueling.

Specification: A.

B.

C.

The refueling interlocks shall be tested prior to any fuel handling with the head off the reactor vessel, at weekly intervalsat the frequency specified in the Surveillance Frequency Control Program thereafter until no longer required and following any repair work associated with the interlocks.

Prior to beginning any core alterations, the source range monitors (SRMs) shall be calibrated. Thereafter, the SRM's will be checked-GaHy, tested monthly, and calibrated at the frequencies specified in the Surveillance Frequency Control Programevery 3 months until no longer required.

Within four (4) hours prior to the start of control rod removal pursuant to Specification 3.9.E verify:

1.

That the reactor mode switch is locked in the refuel position and that the one rod out refueling interlock is operable.

2.

That two (2) SRM channels, one in the core quadrant where the control rod is being removed and one in an adjacent quadrant, are operable and inserted to the normal operation level.

D.

Verify within four (4) hours prior to the start of control rod removal pursuant to Specification 3.9.F and at the frequency specified in the Surveillance Frequency Control Programat least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter, until replacement of all control rods or rod drive mechanisms and all control rods are fUlly inserted that:

1.

the reactor mode switch is locked in the refuel position and the one rod out refueling interlock is operable.

2.

Two (2) SRM channels, one in the core quadrant where a control rod is being removed and one in an adjacent quadrant, are operable and fully inserted.

3.

All control rods not removed are fully inserted with the exception of one rod which may be partially withdrawn not more than two notches to perform refueling interlock surveillance.

OYSTER CREEK 4.

The four fuel assemblies surrounding each control rod or rod drive mechanism being removed or maintained at the same time are removed from the core cell.

4.9-1 Amendment No.: ~

E.

Verify prior to the start of removal of control rods pursuant to Specification 3.9.F that Specification 3.9.F.5 will be met.

F.

Following replacement of a control rod or rod drive mechanism removed in accordance with Specification 3.9.F, prior to inserting fuel in the control cell, verify that the bypassed refueling interlocks associated with that rod have been restored and that the control rod is fully inserted.

Basis:

The refueling interlock&

1 are required only when fuel is being handled and the head is off the reactor vessel. A test of these interlocks prior to the time when they are needed is sufficient to ensure that the interlocks are operable. The testing frequency for the refueling interlocks is based upon engineering judgment and the fact that the refueling interlocks are a backup for refueling procedures.The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The SRMs 2 provide neutron monitoring capability during core alterations. A calibration using external testing equipment to calibrate the signal conditioning equipment prior to use is sufficient to ensure operability. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.The frequencies of testing, using internally generated test signals, and recalibration, if the SRMs are required for an extended period of time, are in agreement with other instruments of this type which are presented in Specification 4.1.

The surveillance requirements for control rod removal assure that the requirements of Specification 3.9 are met prior to initiating control rod removal and at appropriate intervals thereafter.

References:

(1)

FDSAR, Volume I, Section Vll-7-2.5 (2)

FDSAR, Volume I, Sections Vll-4.2.2 and Vll-4-5.1 OYSTER CREEK 4.9-2 Amendment No.: 23,13 Basis:

E.

Verify prior to the start of removal of control rods pursuant to Specification 3.9.F that Specification 3.9.F.5 will be met.

F.

Following replacement of a control rod or rod drive mechanism removed in accordance with Specification 3.9.F, prior to inserting fuel in the control cell, verify that the bypassed refueling interlocks associated with that rod have been restored and that the control rod is fully inserted.

The refueling interlocks(1) are required only when fuel is being handled and the head is off the reactor vessel. A test of these interlocks prior to the time when they are needed is sufficient to ensure that the interlocks are operable. The testing frequency for the refueling interlocks is based upon engineering judgment and the fact that the refueling interlocks are a backup for refueling procedures.The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.

The SRM's(2) provide neutron monitoring capability during core alterations. A calibration using external testing equipment to calibrate the signal conditioning equipment prior to use is sufficient to ensure operability. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program.The frequencies of testing, using internally generated test signals, and recalibration, if the SRM's are required for an extended period of time, are in agreement 'Nith other instruments of this type which are presented in Specification 4.1.

The surveillance requirements for control rod removal assure that the requirements of Specification 3.9 are met prior to initiating control rod removal and at appropriate intervals thereafter.

References:

(1)

(2)

FDSAR, Volume I, Section VII-7-2.5 FDSAR, Volume I, Sections VII-4.2.2 and VII-4-5.1 OYSTER CREEK 4.9-2 Amendment No.: ~

4.10 ECCS RELATED CORE LIMITS Applicability:

Applies to the periodic measurement during power operation of core parameters related to ECCS performance.

Obiective:

To assure that the limits of Section 3.10 are not being violated.

Specification:

A.

Average Planar LHGR.

The APLHGR for each type of fuel as a function of average planar exposure shall be checked daily at the frequency specified in the Surveillance Frequency Control Program during reactor operation at greater than or equal to 25% rated thermal power.

B.

Local LHGR.

The LHGR as a function of core height shall be checked at the frequency specified in the Surveillance Frequency Control Program4a during reactor operation at greater than or equal to 25% rated thermal power.

C.

Minimum Critical Power Ratio (MCPR).

1.

MCPR shall be checked at the frequency specified in the Surveillance Frequency Control Program8a during reactor operation at greater than or equal to 25% rated thermal power.

2.

The MCPR operating limit shall be determined within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of completing scram time testing as required in Specification 4.2.C.

Bases:

The term daily in Technical Specification 4.10 shall be conservatively interpreted as once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (with normal grace allowance). This applies to Technical Specification 4.10 surveillance requirements only.

The LHGR shall be periodically checked daily at the frequency specified in the Surveillance Frequency Control Program to determine whether fuel burnup or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are moved daily, a daily periodic check of power distribution is adequate.

The minimum critical power ratio (MCPR) is unlikely to change significantly during steady state power operation so that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an acceptable frequency for surveillance.

In the event of a single pump trip, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sthe surveillance intervalfrequency specified in the Surveillance Frequency Control Program remains acceptable because the accompanying power reduction is much larger than the change in MAPLHGR limits for four loop operation at the corresponding lower steady state power level as compared to five loop operation. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency specified in the Surveillance Frequency Control Program is also acceptable for the APRM status check since neutron monitoring system failures are infrequent and a downscale failure of an APRM initiates a control rod withdrawal block, thus precluding the possibility of a control rod withdrawal error.

OYSTER CREEK 4.10-1 Amendment No.: 75, 249 ECR OC 04-00575 Corrected by letter of 10/20/04 4.10 ECCS RELATED CORE LIMITS Applicability:

Applies to the periodic measurement during power operation of core parameters related to ECCS performance.

Objective:

Specification:

To assure that the limits of Section 3.10 are not being violated.

A.

Average Planar LHGR.

The APLHGR for each type of fuel as a function of average planar exposure shall be checked GaUy-at the frequency specified in the Surveillance Frequency Control Program during reactor operation at greater than or equal to 25%

rated thermal power.

B.

Local LHGR.

The LHGR as a function of core height shall be checked at the frequency specified in the Surveillance Frequency Control ProgramGaHy during reactor operation at greater than or equal to 25%

rated thermal power.

C.

Minimum Critical Power Ratio (MCPR).

1.

MCPR shall be checked at the frequency specified in the Surveillance Frequency Control ProgramGaHy during reactor operation at greater than or equal to 25%

rated thermal power.

2.

The MCPR operating limit shall be determined within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of completing scram time testing as required in Specification 4.2.C.

Bases:

The term "daily" in Technical Specification 4.10 shall be conservatively interpreted as once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (with normal grace allowance). This applies to Technical Specification 4.10 surveillance requirements only.

The LHGR shall be periodically checked GaUy-at the frequency specified in the Surveillance Frequency Control Program to determine whether fuel burnup or control rod movement has caused changes in power distribution. Since changes due to burnup are slow, and only a few control rods are moved daily, a GaUy-periodic check of power distribution is adequate.

The minimum critical power ratio (MCPR) is unlikely to change significantly during steady state power operation so that 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is an acceptable frequency for surveillance. In the event of a single pump trip, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />sthe surveillance intervalfrequency specified in the Surveillance Frequency Control Program remains acceptable because the accompanying power reduction is much larger than the change in MAPLHGR limits for four loop operation at the corresponding lower steady state power level as compared to five loop operation. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency specified in the Surveillance Frequency Control Program is also acceptable for the APRM status check since neutron monitoring system failures are infrequent and a downscale failure of an APRM initiates a control rod withdrawal block, thus precluding the possibility of a control rod withdrawal error.

OYSTER CREEK ECR OC 04 00575 4.10-1 Amendment No.: 75, 249 Corrected by letter of 10/20/04

4.12 Alternate Shutdown Monitoring Instrumentation Applicability:

Applies to the surveillance requirements of the alternate shutdown monitoring instrumentation.

Obiective:

To specify the minimum frequency and type of surveillance to be applied to the alternate shutdown monitoring instrumentation.

Specification:

Each of the alternate shutdown monitoring channels shown in Table 4.12-1 shall be demonstrated operable by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise notedshown in Table 4.12-1.

Basis:

The operability of the alternate shutdown monitoring instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of hot shutdown of the plant from locations outside of the control room. The type and frequency of surveillances frequencies required in Table 4.12-1 are consistent with or more conservative than the BWR Standard Technical Specificationsare controlled under the Surveillance Frequency Control Program.

OYSTER CREEK 4.12-1 Amendment No.: 161, 263 4.12 Alternate Shutdown Monitoring Instrumentation Applicability:

Applies to the surveillance requirements of the alternate shutdown monitoring instrumentation.

Objective:

Specification:

To specify the minimum frequency and type of surveillance to be applied to the alternate shutdown monitoring instrumentation.

Each of the alternate shutdown monitoring channels shown in Table 4.12-1 shall be demonstrated operable by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies specified in the Surveillance Frequency Control Program unless otherwise notedshovln in Table 4.12-1.

The operability of the alternate shutdown monitoring instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of hot shutdown of the plant from locations outside of the control room. The type and frequency of surveillances frequencies required in Table 4.12 1 are consistent with or more conservative than the BVVR Standard Technical Specificationsare controlled under the Surveillance Frequency Control Program.

OYSTER CREEK 4.12-1 Amendment No.: 161,263

TABLE 4.12-1 ALTERNATE SHUTDOWN MONITORING INSTRUMENTATION CHANNEL CHANNEL Functional Limit CHECK (Note 1)

CALIBRATION (Note 1)

I Reactor Pressure M

Q I

Reactor Water Level (fuel zone) n/a Q

I Condensate Storage Tank Level M

R I

Service Water Pump Discharge Pressure M

R I

Control Rod Drive System Flowmeter M

R I

Shutdown Cooling System Flowmeter n/a R

I Isolation Condenser B Shell Water Level M

R I

Reactor Building Closed Cooling Water Pump Discharge Pressure M

R Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

M Monthly Q

Quarterly I

R Refueling Outage I

OYSTER CREEK 4.12-2 Amendment No.: 11l, 161, 263 TABLE 4.12-1 ALTERNATE SHUTDOWN MONITORING INSTRUMENTATION CHANNEL CHANNEL Functional Limit CHECK (Note 1)

CALIBRATION (Note 1)

Reactor Pressure M

Q Reactor Water Level (fuel zone) n/a Q

Condensate Storage Tank Level M

R Service Water Pump Discharge Pressure M

R Control Rod Drive System Flowmeter M

R Shutdown Cooling System Flowmeter n/a R

Isolation Condenser "B" Shell Water Level M

R Reactor Building Closed Cooling Water Pump Discharge Pressure M

R Note 1: Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

M Monthly Q

Ql:Jartorly R

Rofl:Joling Ol:Jtago OYSTER CREEK 4.12-2 Amendment No.: 114, 161, 263

TABLE 4.15.2 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL CHANNEL SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECK (h)

CHECK CALIBRATION(f)(h)

TEST (h)

REQUIRED (a)

1. Main Condenser D

N/A Q(g)

M (c)

Offgas Treatment System Hydrogen Monitor Leçend: D once per 21 hrs; M once per 31 days; Q once per 92 days; N/A = Not Applicable.

TABLE 4.15.2 NOTATIONS (a)

Instrumentation shall be OPERABLE and in service except that a channel may be taken out of services for the purpose of a check, calibration, test or maintenance without declaring it to be inoperable.

(c)

During main condenser offgas treatment system operation.

(f)

The CHANNEL CALIBRATION shall be performed according to established station calibration procedures.

(g)

A CHANNEL CALIBRATION shall include the use of at least two standard gas samples, each containing a known volume percent hydrogen in the range of the instrument, balance nitrogen.

(h)

Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

OYSTER CREEK 4.15-2 Amendment No.: 137, 145, 155, 166, 263 TABLE 4.15.2 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS INSTRUMENT

1. Main Condenser Offgas Treatment System Hydrogen Monitor CHANNEL CHECK (h)

SOURCE CHECK N/A CHANNEL CALIBRATION(f){h)

Q(g)

CHANNEL FUNCTIONAL TEST (h)

M CHANNEL SURVEILLANCE REQUIRED (a)

(c)

Legend: D onco per 24 hrs; M once per 31 days; Q once per Q2 days;

---N/A =Not Applicable.

TABLE 4.15.2 NOTATIONS (a)

Instrumentation shall be OPERABLE and in service except that a channel may be taken out of services for the purpose of a check, calibration, test or maintenance without declaring it to be inoperable.

(c)

During main condenser offgas treatment system operation.

(f)

The CHANNEL CALIBRATION shall be performed according to established station calibration procedures.

(g)

A CHANNEL CALIBRATION shall include the use of at least two standard gas samples, each containing a known volume percent hydrogen in the range of the instrument, balance nitrogen.

(h)

Surveillance Frequencies are specified in the Surveillance Frequency Control Program unless otherwise noted in the table.

OYSTER CREEK 4.15-2 Amendment No.: 137,145,155, 166,263

4.17 Control Room Heating, Ventilating, and Air-Conditioning System Alicability:

Applies to surveillance requirements for the control room heating, ventilating, and air conditioning (HVAC) systems.

Obiective:

To verify the capability of each control room HVAC system to minimize the amount of radioactivity from entering the control room in the event of an accident.

Specification: Surveillance of each control room HVAC system shall be as follows:

A.

At least once monthlyAt the frequency specified in the Surveillance Frequency Control Program: by initiating, from the control room, the partial recirculation mode of operation, and by verifying that the system components are aligned such that the system is operating in this mode.

B.

At least once every refueling outageAt the frequency specified in the Surveillance Frequency Control Program: by verifying that in the partial recirculation mode of operation, the control room and lower cable spreading room are maintained at a positive pressure of 1/8 in. WG relative to the outside atmosphere.

Basis:

Periodic surveillance of each control room HVAC system is required to ensure the operability of the system. The operability of the system in conjunction with control room design provisions is based upon limiting the radiation exposure to personnel occupying the control room to less than a 30-day integrated dose of 5 rem TEDE for the most limiting design basis accident.

OYSTER CREEK 4.17-1 Amendment No.: 115, 139, 262 4.17 Control Room Heating. Ventilating. and Air-Conditioning System Applicability:

Applies to surveillance requirements for the control room heating, ventilating, and air conditioning (HVAC) systems.

Objective:

To verify the capability of each control room HVAC system to minimize the amount of radioactivity from entering the control room in the event of an accident.

Specification: Surveillance of each control room HVAC system shall be as follows:

A.

At least once monthlyAt the frequency specified in the Surveillance Frequency Control Program: by initiating, from the control room, the partial recirculation mode of operation, and by verifying that the system components are aligned such that the system is operating in this mode.

B.

At least once every refueling outageAt the frequency specified in the Surveillance Frequency Control Program: by verifying that in the partial recirculation mode of operation, the control room and lower cable spreading room are maintained at a positive pressure of ~ 1/8 in. WG relative to the outside atmosphere.

Basis:

Periodic surveillance of each control room HVAC system is required to ensure the operability of the system. The operability of the system in conjunction with control room design provisions is based upon limiting the radiation exposure to personnel occupying the control room to less than a 3D-day integrated dose of 5 rem TEDE for the most limiting design basis accident.

OYSTER CREEK 4.17-1 Amendment No.: 115, 139, 262