ML071360339

Technical Specification Change Request No. 327 - Modify Technical Specifications for Primary Containment Oxygen Concentration

ML071360339
Person / Time
Site:	Oyster Creek
Issue date:	05/16/2007
From:	Cowan P AmerGen Energy Co
To:	Document Control Desk, NRC/NRR/ADRO
References
2130-07-20497
Download: ML071360339 (29)

Text

c 2 130-07-20497 May 16,2007 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Technical Specification Change Request No. 327 - Modify Technical Specifications for Primary Containment Oxygen Concentration Pursuant to 10 CFR 50.90, AmerGen Energy Company, LLC hereby requests changes to the Technical Specifications included in Oyster Creek Nuclear Generating Station (OCNGS)

Operating License No. DPR-16. The proposed changes modify Technical Specifications (TS) 3.5.A.6 to adopt the inerting/de-inerting requirements of the Standard Technical Specifications, which require inerting the primary containment to less than 4 percent oxygen concentration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding 15 percent of rated thermal power, and allow de-inerting the containment 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing thermal power to less than 15 percent of rated during a plant shutdown. Also, a new TS condition will be added to identify required actions if the primary containment oxygen concentration increases to greater than or equal to four volume percent while in the RUN MODE. In addition, definitions of Thermal Power and Rated Thermal Power (RTP) will be added to the TS. provides an evaluation supporting the proposed TS changes. Attachment 2 contains the marked-up TS pages for the proposed changes. The TS Bases pages are provided for information only, and do not require NRC approval.

AmerGen Energy Company, LLC requests approval of these changes by May 11,2008. This license amendment will be implemented within 60 days of approval.

The proposed changes to the Technical Specifications have undergone a safety review in accordance with Section 6.5 of the OCNGS TS. No new regulatory commitments are established by this submittal.

2130-07-20497 May 16,2007 Page 2 We are notifying the State of New Jersey of this application for changes to the Technical Specifications by transmitting a copy of this letter and its attachments to the designated State Official.

If any additional information is needed, please contact Dave Robillard at (610) 765-5952.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 16th day of May 2007.

Respectfully, Director - Licensing & Regulatory Affairs AmerGen Energy Company, LLC : Oyster Creek Technical Specification Change Request No. 327, Evaluation of Proposed Changes : Oyster Creek Technical Specification Change Request No. 327, Proposed Technical Specification Changes (Mark-up) cc: S. J. Collins, Administrator, USNRC Region I G. E. Miller, USNRC Project Manager, Oyster Creek M. S. Ferdas, USNRC Senior Resident Inspector, Oyster Creek P. Baldauf, Director, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection File No. 04034

ATTACHMENT 1 Oyster Creek Technical Specification Change Request No. 327 Evaluation of Proposed Changes 2130-07-20497 Page 1 of 6 1.O DESCRIPTION This letter proposes to amend Operating License No. DPR-16 for Oyster Creek Nuclear Generating Station (OCNGS).

The proposed changes would revise the OCNGS Operating License to modify Technical Specifications (TS) 3.5.A.6 to adopt the inertinglde-inertingrequirements of the improved Standard Technical Specifications (STS) (Reference l ) , which require inerting the containment to less than four (4.0) volume percent (Yo)oxygen concentration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding fifteen (15) o/o of rated thermal power (RTP), and allow de-inerting the containment 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing thermal power to less than 15% of RTP during a plant shutdown. Definitions of Thermal Power and Rated Thermal Power (RTP) will also be added to the TS. Additionally, various TS and TS Bases pages are revised to capitalize THERMAL POWER and RATED THERMAL POWER, to maintain consistency with typical TS format.

2.0 PROPOSED CHANGE

S The proposed changes modify Specification 3.5.A.6 by adopting inerting and de-inerting requirements that are consistent with the guidance of the STS (Reference 1). The STS uses 15% (RTP) as the reference condition for inerting and de-inerting the containment, rather than the mode selector switch being in the RUN MODE. The proposed changes will require inerting containment to less than 4.0 volume percent oxygen concentration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding 15% RTP during startup and allow de-inerting the containment 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing thermal power to less than 15% of RTP during a plant shutdown. The changes will also establish consistency with the STS required action in the event primary containment oxygen concentration is not within the specified limit.

The proposed changes would add the following definition of rated thermal power and replace current TS 3.5.A.6 as follows:

1.49 RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 1930 MWt.

1.50 THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

3.5.A.6. The primary containment oxygen concentration shall be less than 4.0 volume percent while in the RUN MODE, except as specified in 3.5.A.6.a and 3.5.A.6.b below.

a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP following startup, to1 2130-07-20497 Page 2 of 6

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to c 15% RTP prior to the next scheduled reactor shutdown.

c. If the primary containment oxygen concentration is greater than or equal to 4.0 volume percent while in the RUN MODE, except as specified in 3.5.A.6.a or 3.5.A.6.b above, restore oxygen concentration to c 4.0 volume percent within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise reduce THERMAL POWER to 5 15% RTP within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Definitions of Thermal Power and Rated Thermal Power (RTP) will also be added to the TS. Additionally, various TS and TS Bases pages are revised to capitalize THERMAL POWER and RATED THERMAL POWER, to maintain consistency with typical TS format.

Further, the proposed changes would incorporate the STS Bases for the primary containment oxygen concentration into the OC TS Bases on page 3.5-10.

The introductory phrase of TS 3.5.A.6, after completion of the startup test program and demonstration of the plant electrical output, is deleted since these activities were completed at the time of the initial plant startup; therefore this is an obsolete statement.

No changes are proposed to the surveillance requirement since both the current OC TS and the STS are consistent in requiring weekly checks of the primary containment oxygen concentration.

3.0 BACKGROUND

During normal operation and following a design basis accident (i.e., Loss of Coolant Accident (LOCA)) the primary containment atmosphere must be inerted to ensure that hydrogen combustion cannot occur. lnerting is achieved by purging the primary containment with nitrogen until oxygen concentration is less than four percent by volume.

When the primary containment atmosphere is inerted, operators cannot access the containment to perform required surveillances and leak inspections unless they wear self contained breathing apparatus. TS 3.5.A.6 provides for a 24-hour time period following startup or before shutdown in which the primary containment atmosphere does not have to be inerted. The 24-hour time period is a reasonable amount of time for plant personnel to perform required inspections and then to complete inerting or shutdown evolutions. The current reference condition for this 24-hour time period is when the reactor mode switch is placed in the run position, and for shutdown the reference condition for the 24-hour time period is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reaching the shutdown condition.

In order to provide sufficient time to accommodate surveillances and leak inspections during startup, transfer to the RUN MODE is delayed as long as possible to delay the start of the 24-hour inerting window.

Further, the current TS requirement does not provide sufficient time to perform in-containment equipment maintenance without a complete shutdown of the reactor.

2130-07-20497 Page 3 of 6

4.0 TECHNICAL ANALYSIS

The proposed changes add a definition of R I P to TS and modify Specification 3.5.A.6 by adopting inerting and de-inerting requirements which are consistent with the guidance of the STS (Reference 1). The STS uses the reference condition of 15% RTP for inerting and de-inerting rather than the reactor mode switch being in the RUN position during startup, and prior to a scheduled shutdown.

The addition of the definitions of THERMAL POWER and RTP to TS are necessary to define RTP as used in proposed TS 3.5.A.6. The value of RTP is currently defined as 1930 MWt in OCNGS Operating License Condition C.1.

The proposed changes to TS 3.5.A.6 will require inerting the containment to less than 4.0 volume percent oxygen concentration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding 15% RTP during startup and allow de-inerting 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing thermal power to less than 15%

RTP during a plant shutdown. The basis for this change, as stated in STS, is that at less than 15% RTP the potential for an event that generates significant hydrogen is low and the primary containment need not be inert. The STS bases further states that the probability of an event that generates hydrogen occurring within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a startup, or within the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before a shutdown, is low enough that these windows, when the primary containment is not inerted, are also justified.

Making the 24-hour time period contingent upon core thermal power will allow placing the reactor mode switch in the RUN position earlier during the startup sequence. In addition, the proposed changes will provide additional time to perform surveillances and maintenance inside the primary containment before the containment is inerted.

Therefore, implementing these changes will improve overall plant reliability.

Further, basing the 24-hour period to de-inert the containment on when the reactor power is less than 15% RTP, rather than prior to a scheduled shutdown, provides a more explicit requirement as to when the 24-hour period starts. Also, the proposed changes avoid references to the ambiguous term scheduled shutdown since this activity is not defined in either the STS or the Oyster Creek TS.

Finally, the proposed changes adopt the Limiting Condition for Operation (LCO) actions requirement in the STS. The current TS 3.5.A.6 does not identify a required action in the event the containment oxygen concentration is not in compliance with this LCO. Non-compliance with TS 3.5.A.6 results in entering TS LCO 3.O.A, which requires placing the plant in the COLD SHUTDOWN condition within 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> of entering LCO 3.0.A. The STS action statement for the containment oxygen concentration LCO limits the required action to reducing reactor power to less than 15% RTP within the following 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, if primary containment oxygen concentration can not be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The 24-hour timeframe to restore primary containment oxygen concentration to the TS value avoids unnecessary plant transients.

2130-07-20497 Page 4 of 6

5.0 REGULATORY ANALYSIS

5.1 NO SIGNIFICANT HAZARDS CONSIDERATION AmerGen has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed changes modify the Technical Specifications (TS) by adding definitions of Rated Thermal Power (RTP) and Thermal Power to TS and adopting containment inerting and de-inerting requirements that are consistent with the guidance of NUREG-1433, Standard Technical Specifications - General Electric Plants, BWR/4 (STS), Revision 3.1. Additionally, various TS and TS Bases pages are being revised to capitalize THERMAL POWER and RATED THERMAL POWER, to maintain consistency with typical TS format. The proposed changes will allow inerting of the primary containment within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding 15%

(RTP) during a plant startup, and de-inerting 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing thermal power to less than 15% RTP during a plant shutdown. Also, a new TS condition will be added to identify required actions if the primary containment oxygen concentration increases to greater than or equal to four volume percent while in the RUN MODE. The proposed changes do not alter the physical configuration of the plant, nor do they affect any previously analyzed accident initiators. The accident analysis assumes that a Loss of Coolant Accident (LOCA) occurs at 1OOoh power.

The consequences of a LOCA at less than 15% RTP would be much less severe, and produce less hydrogen than a LOCA at 100% power.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed changes add definitions of Rated Thermal Power (RTP) and Thermal Power to TS and adopt the STS guidance regarding containment inerting/de-inerting requirements. Additionally, various TS and TS Bases pages are being revised to capitalize THERMAL POWER and RATED THERMAL POWER, to maintain consistency with typical TS format. The proposed changes introduce no new mode of plant operation and they do not involve any physical modification to the plant. The proposed changes are consistent with the current safety analysis assumptions. No setpoints are being changed which would alter the dynamic response of plant equipment. Accordingly, no new failure modes are introduced.

2130-07-20497 Page 5 of 6 Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed changes add definitions of Rated Thermal Power (RTP) and Thermal Power to TS and adopt the STS guidance regarding containment inerting/de-inerting requirements. Additionally, various TS and TS Bases pages are being revised to capitalize THERMAL POWER and RATED THERMAL POWER, to maintain consistency with typical TS format. Adoption of the STS reference point operating condition of 15% RTP adds operational flexibility related to the performance of inspections and maintenance inside primary containment during plant startup and shutdown. Making the 24-hour time period contingent upon core thermal power, rather than reactor mode switch position during a plant startup, will enable placing the mode switch in the RUN position sooner. The proposed changes do not affect any assumptions or conclusions contained in the plant safety analyses, which assume that a LOCA occurs at 100% power. The current Limiting Condition for Operation action requirement and shutdown reference condition for de-inerting involve a complete reactor shutdown. Changing this requirement to 15% RTP, avoids the potential for an unnecessary plant transient.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based upon the above, AmerGen concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

5.2 APPLICABLE REGULATORY REQUIREMENTS/CRITERIA 10 CFR 50.36, "Technical Specifications," provides the regulatory requirements for the content required in a licensee's TS. Criterion 3 of 10 CFR 50.36(~)(2)(ii)requires a limiting condition for operation to be established for a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. While the LOCA analysis results in a hydrogen concentration level significantly below the concentration at which hydrogen can ignite in air, the containment inerting design feature was provided to preclude a hydrogen-oxygen reaction under conditions more severe than can currently be foreseen (Reference 2).

The proposed changes do not affect the assumptions or conclusions contained in the plant safety analyses. The proposed changes do not involve any physical changes to the plant design, nor do they impact any accident initiators. The accident analysis in Chapter 15 of the Updated Final Safety Analysis Report (UFSAR) (Reference 3) 2130-07-20497 Page 6 of 6 assumes that a LOCA occurs at 100% RTP. The consequences of a LOCA below 15%

RTP would be less severe and would produce less hydrogen.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation," or would change an inspection or surveillance requirement. However, the proposed amendment does not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51-22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review, Paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, Paragraph (b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 PRECEDENT The NRC approved similar license amendments for the James A. FitzPatrick Nuclear Power Plant on December 28, 1994 (Amendment 221) (TAC No. M90050), and the Pilgrim Nuclear Power Station on April 10, 2006 (Amendment 218) (TAC No. MC7056).

8.0 REFERENCES

1) NUREG-1433, "Standard Technical Specifications - General Electric Plants, BWW4" (STS), Revision 3.1 , December 1, 2005.

2) Oyster Creek Generating Station - Updated Final Safety Analysis Report, Section 6.2.5.3, Design Evaluation.

3) Oyster Creek Generating Station - Updated Final Safety Analysis Report, Section 15.6.5, Loss of Coolant Accident.

ATTACHMENT 2 Oyster Creek Technical Specification Change Request No. 327 Proposed Technical Specification Changes (Mark-up)

Paaes Insert Page 1.O-6 1.O-8 2.1 -1 2.1-2 2.3-1 2.3-4 2.3-6 3.1 -5 3.1-6 3.1-17 3.3-3a 3.5-4 3.5-10 3.6-4 3.10-4 4.1-3 4.2-1 4.6-1

Insert 1 1.49 RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 1930 MWt, 1.50 THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

Insert 2 The primary containment oxygen concentration shall be less than 4.0 volume percent while in the RUN MODE, except as specified in 3.5.A.6.a and 3.5.A.6.b below.

a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP following startup, to,

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

C. If the primary containment oxygen concentration is greater than or equal to 4.0 volume percent while in the RUN MODE, except as specified in 3.5.A.6.a or 3.5.A.6.b above, restore oxygen concentration to < 4.0 volume percent within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, otherwise reduce THERMAL POWER to 5 15% RTP within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Insert 3 All nuclear reactors must be designed to withstand events that generate hydrogen either due to the zirconium metal water reaction in the core or due to radiolysis. The primary method to control hydrogen is to inert the primary containment. With the primary containment inert, that is, oxygen concentration < 4.0 volume percent (v/o), a combustible mixture cannot be present in the primary containment for any hydrogen concentration. An event that rapidly generates hydrogen from zirconium metal water reaction will result in excessive hydrogen in primary containment, but oxygen concentration will remain < 4.0 v/o and no combustion can occur. The primary containment oxygen concentration must be within the specified limit when the primary containment is inerted, except as allowed by Specifications 3.5.A.6.a or 3.5.A.6.b during startups and shutdowns. The primary containment must be inert when reactor power is greater than or equal to 15% RTP, since this is the condition with the highest probability of an event that could produce hydrogen. Specification 3.5.A.6.a requires that primary containment be inerted within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding 15 percent RTP during startup. Specification 3.5.A.6.b allows de-inerting to commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reaching 15 percent RTP during the shutdown sequence.

lnerting the primary containment is an operational problem because it prevents containment access without an appropriate breathing apparatus. Therefore, the primary containment is inerted as late as possible in the plant startup and de-inerted as soon as possible in the plant shutdown. As long as power is less than 15 percent RTP, the potential for an event that generates significant hydrogen is low and the primary containment need not be inert.

Furthermore, the probability of an event that generates hydrogen occurring within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a startup or within the last 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before a shutdown, is low enough that these windows, when the primary containment is not inerted, are also justified. The 24-hour time period is a reasonable amount of time to allow plant personnel to perform inerting or de-inerting.

1.28 -FRACTIONOF RATED POWER (FRP)

The F&ICTION OF RATED POWER is the ratio of core 1.29 TOP OF ACTIVE FUEL (TAF)- 353.3 inches above vessel zero.

1.30 REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

1.31 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE is that leakage which is collected in the primary containment equipment drain tank and eventually transferred to radwaste for processing. ,

1.32 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE is all measured leakage that is other than identified leakage.

1.33 PROCESS CONTROL PLAN The PROCESS CONTROL PLAN shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20,61 and 71, State regulations, burial ground requirements, and other requirements goveming the disposal of solid radioactive waste.

1.34 AUGMENTED OFFGAS SYSTEM (Am)

The AUGMENTED OFFGAS SYSTEM is a system designed and installed to holdup andor process radioactive gases from the main condenser offgas system for the purpose of reducing the radioactive material content of the gases before release to the environs.

1.35 MEMBER OF THE PUBLIC A MEMBER OF THE PUBLIC is a person who is not occupationally associated with AmerGen Energy Company, LLC and who does not normally fnxluent the Oyster Creek Nuclear Generating Station site. The category does not include contractors, contractor employees, I

vendors, or persons who enter the site to make deliveries, to service equiprnenk work on the site, or for other purposes associated with plant functions.

1.36 OFFSITE DOSE CALCLJLATION MANUAL IODCW The OFFSITE DOSE CALCULATIONuL\MJAL shall contain the methodology and OYSTER CREEK

1.42 AVERAGE PLANAR LINEAR HEAT GENERATION R A E c The AVERAGE PLANAR LWEAR NEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the heat generation rate per unit length of fuel rod for all the fuel rods in the specified bundle at the specified height divided by the number of Fuel rods in the fuel bundle at that height.

1.43 CORE OPERATING LIh?lITS REPORT The Oyster Creek CORE OPERATING LIMITS REPORT (COLR) is the document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9. ? .f. Plant operation within these operating limits is addressed in individual specifications.

1.44 LOCAL LINEAR HEAT GENERATIOX RATE The LOCAL, LrmEAR HEAT GENERATION RATE (LLHGR) shall be applicable to a specific planar height and is equal to the AVERAGE PLANAR LINEAR GENERATION RATE (APLHCR) at the specified height multiplied by the local peaking factor at that height.

1.45 SHUTDOWN MARGIN (SDM)

SHUTDOT" MARGIN is the amount of reactivity by which the reactor would be subcritical when the control rod with the highest reactivity worth is fully withdrawn, all other operable control rods are fully inserted, all inoperable control rods are at their current position, reactor water temperature is 689F, and the reactor fuel is xenon free. Determination of the control rod with the highest reactivity worth includes consideration of any inoperable control rods which are not fully inserted.

1.46 IDLE RECIRCULATION LOOP A recirculation loop is idle when its discharge valve is in the closed position and its discharge bypass valve and suction valve are in the open position.

I .47 ISOLATED RECIRCULATION LOOP A recirculation loop is fully isolated when the suction valve, discharge valve and discharge bypass valve are in the closed position.

1.48 OPERATIONAL CONDITION The reactor plant operational status as to criticality, reactor mode switch position, reactor coolant temperature, andor specific system status. These conditions consist of POWER OPERATION, STARTUP MODE, SHUTDOWN CONDITION, COLD SHUTDOWN CONDITION, and REFUEL MODE. A change or entry into an operating condition is signified by movement of the reactor mode switch or a change in reactor coolant temperature from <2 12°F to 22 12°F.

OYSTER CREEK I .o-8 Amendment No.: w -* ,

SECTION 2 SAFETY LIMITS AND LIMIT IN^ SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMIT - FUEL CLADDING INTEGRITY Applicability: Applies to the interrelated variables associated with fuel thermal behavior.

Obiective: To establish limits on the important thermal hydraulic variables to assure the integrity of the fuel cladding.

Specifications:

A. When the reactor pressure is greater than or equal to 800 psia and the core flow is greater than or equal to 10% of rated, the existence of a minimum CRITICAL POWER RATIO (MCPR) less than 1.10 for both four or five loop operation and 1.12 for three loop operation shall constitute violation of the fuel cladding integrity safety limit.

I s than 800 psia or the core flow is less than 10 shall not exceed 25% of ers exceed the limiting safety system settings in Specification 2.3 and a reactor scram is not initiated by the associated protective instrumentation, the reactor shall be brought to, and remain in, the COLD SHUTDOWN CONDITION until an analysis is performed to determine whether the safety limit established in Specification 2.1 .A and 2.1.B was exceeded.

D. During all modes of reactor operation with irradiated fuel in the reactor vessel, the wat level shall not be less than 4'8" above the TOP OF ACTIVE FUEL.

OYSTER CREEK 2.1-1 Amendment No.:

Bases:

The fuel cladding integrity safety limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedure used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity safety limit is defined as the CRITICAL POWER RATIO in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transit~onconsidering the power distribution within the core and all uncertainties.

The Safety Limit MCPR") is determined using the General Electric Thermal Analysis Basis, GETAB(*),which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. The power distribution uncertainty is treated in accordance with a NRC approved . The revised analysis results in lower SLMCPR values primarily due to an improved treatment of the power distribution uncertainty that reduces the conservatism of the GETAB method of power allocation. All other uncertainties are consistent with the GETAB basis. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) - Boiling Length (L), GEXL, correlation.

The use of the GEXL correlation is not valid for the critical power calculations at pressures below 800 psia or core flows less than 10% of rated. Therefore, the fuel cladding integrity safety limit is protected by limiting the cor At pressures below 800 psia, the core elevation pressure drop (0 power, 0 r than 4.56 psi. At low power and all flows this pressure differential is maintained in the bypass region of the core. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and all flows will always be greater than 4.56 psi. Analyses show that with a flow of 28 x 1O3 Ibs/hr bundle flow, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, bundle flow with a 4.56 psi driving head will be greater than 28 x 1O3 lbs/hr irrespective of total core flow and independent of bundle power for the range of bundle powers of concern. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at t 3.35 MWt. With the des tors this corresponds to a c than 50%. Thus, a core limit of 25% for reactor pressur flow less than 10% is co

+ --

Plant safety analyses have shown that the scrams caused by excee assure that the Safety Limit of Specification 2.1 .A or 2.1.8 will not are checked periodically to assure the insertion times are adequate. Th transient resulting when a scram is accomplished other than by the expected scram signal (e.g.,

OYSTER CREEK 2.1-2 Amendment No,: -,e

2.3 L I M ~ T I NSAFETY

~ SYSTEM SETTlNGS

~ i t v : to trip settings on automatic protective devices relatsd to varis.3;es cn

~ ~ ~ p ~ i c ~ b iApplies which safety limits have been placed.

0biective: To provide automatic corrective action to prevent the safety limits from t s i ~ g exceeded.

Specification: Limiting safety system settings shall be as follows:

FUNCTION LI~ITINGSAFETY SYSTEM SETTINGS A. Neutron Flux, Scram A.l APRM When the reactor mode switch is in ths Run p&iinil, the APRM flux scram setting shall be the minimum c;i:

For W 20.0x1061b/hr.

The applicable stability protection settings, as d3iimd in the COLR, with a rgaxirnum setpoint of 120.0% for core flew ecjzal to I

61 x 10 Ib/hr and greater, where: /----------- ----.

hFLPD = maximum fraction of limiting power density where the limiting power density for each bundle is the design linear heat generation rate for thai bundle.

The ratio of FRP/MFLPD shall be set equal to 1.O unless the actual operating value is less than 1.13 in which case the actual operating value will be used.

This adjustment may be ~ c c o ~ p ~ i by she~

increasing the APRM gain and thus reducing the flow reference APRM High F l u Scram Curve Sy the reciprocal of the APRM gain change.

A.2 IRM A.3 APRM Downscale coincident with IRM Upscale OYSTER CREEK 2.3-1 Amendment No.: 7?,?El !l+-?Q81 235,

2.3 LIMITING SAFETY SYSTEM SETTINGS

-Bases:

Safety limits have been established in Specifications 2.1 and 2.2 to protect the integiitj af p , ~

IC A&,/:-

fuel cladding and raactor coolant system barriers, respectively. Automatic protectit 0 Liz. rue§ have been provided in the plant design for corrective actions to prevent the safety limits f j - ~ m being exceeded in normal operation or operational transients caused by reasonably e : i p ~ % d single operator error or equipment malfunction. This Specification establishes the trip sezings r__A---^.- --.-*----,-*--.-.& -,

for these automatic protection devices. c-

,ri----- +-------d

-c. -. ,.-+.-A .-- - _....---_x..i I

/

1 The Average Power Range Monitor, APRM, $rip setting has been established io assuis GSJS:

reaching the fuel cladding integri neutron flux. However, near the balance, so that the neutron flux For slow maneuvers, su er transferred to the water follow the neutron flu For fast transients, the ter due to the effect of the fuel time constant. Therefore, when the neutron ftux increases to the scram setting, the percent increase in heat flux and power transferred water will be less than the percent increase in neutron flux.

The APRM trip setting will be varied automatically with recirculationflow, with the trip ssi1ing at the rated flow of 61.O x 1O8 lblhr or greater being 120.0% of rated neutron flux. Based on a complete evaluation of the reactor dynamic performance during normal operation as wglt BS expected maneuvers and the various mechanical failures, it was concluded that sufficient protection is provided by the simple fixed scram setting (2,3).However, in response to expressed beliefs (4) that variation of APRM flux scram with recirculation flow is a prudent measure to ensure safe plant operation, the scram setting will be varied with recirculation flaw.

An increase in the APRM scram trip setting would decrease the margin present before the fusi cladding integrity safety limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide a reasonable range for maneuvering during operE.ticn.

Reducing this operating margin would increase the frequency of spurious scrams, which could have an adverse effect on reactor safety because of the resulting thermal stresses and the unnecessaty challenge to the operators, Thus, the APRM scram trip setting was sslecied because it provides adequate margin for the fuel cladding integrity safety limit and yet allows operating margin that reduces the possibility of unnecessary scrams, The scram trip setting must be adjusted to ensure that the LHGR transient peak is not on of maximum fraction of limiting power density (MFLPD) and The scram setting is adjusted in accordance with the formula in e MFLPD is greater than the fraction of the rated power (FRP). Th? 1 i adjustment may be accomplished by increasing the APRM gain and thus reducing the flow 1 referenced APRM High Flux Scram Curve by the reciprocal of the APRM gain change.

L-.-.*.. -dk OYSTER CREEK 2.3-4 Amendment No. 7!,?5, Z W ? ? , 23.5, w

The Rod Worth Minimizer is not required beyond 10% of rated power. The ability of the IRMs to terminate a rod withdrawal transient is limited due to the number and location of IRM /-+

., -\

i detectors. An evaluation was performedthat showed by maintaining a minimum recirculation  !,

flow of 39.65~10Ib/hr in range 10 a complete rod withdrawal initiated at 35% of rated power or I 1 $-* i less would not result in violating the fuel cladding safety limit. Therefore, a rod block on the c~

IRMs at less than 35% of rated power would be adequate protection against a rod withdrawal i -4 \

transient. i ri i

Reactor power level may be varied by moving control rods or by varying the r~irculation flow \

rate. The APRM system provides a control rod block to prevent gross rod withdrawal at i 1 \

constant recirculation flow rate to protect against grossly exceeding the MCPR Fuel Cladding \

Integrity Safety Limit. This rod block trip setting, which is automatically varied with recirculation loop flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow P range. The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationship. Therefo steady-state operation, is at 115% of the trip setting. The actual power distribut~onin the core is established by specified control rod sequences and is monitored continuously by the incore LPRM system. As with APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction of limiting power density exceeds the fraction of the rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gains. I The settings on the reactor high pressure scram, anticipatory scrams, reactor coolant system relief valves and isolation condenser have been established to assure never reaching the reactor coolant system pressure safety limit as well as assuring the system pressure does not exceed the range of the fuel cladding integrity safety limit. In addition, the APRM neutron flux scram and the turbine bypass system also provide protection for these safety limits, e.g.,

turbine trip and loss of electrical load transients (5). In addition to preventing power operation above 1060 psig, the pressure scram backs up the other scrams for these transients and other steam line isolation type transients. Actuation of the isolation condenser during these transients removes the reactor decay heat without further loss of reactor coolant thus protecting the reactor water level safety limit.

The reactor coolant system safety valves offer yet another protective feature for the reactor coolant system pressure safety limit since these valves are sized assuming no credit for other pressure relieving devices. In compliance with Section I of the ASME Boiler and Pressure Vessel Code, the safety valve must be set to open at a pressure no higher than 103% of design pressure, and they must limit the reactor pressure to no more than 110% of design pressure.

The safety valves are sized according to the Code for a condition of main steam isolation valve closure while operating at 1930 MWt, followed by (1) a reactor scram on high neutron flux, (2) failure of the recirculation pump trip on high pressure, (3) failure of the turbine bypass valves to open, and (4) failure of the isolation condensers and relief valves to operate. Under these conditions, a total of 9 safety valves are required to turn the pressure transient. The ASME B&PV Code allows an as-found f 3% of setpoint pressure variation in the lift point of the valves. The as-left Safety Valve setpoint tolerance requirement will remain f 1% per GE NEDC-31753P (approval letter dated March 8, 1993) recommendation, This variation is recognized in Specification 4.3.

OYSTER CREEK 2.3-6 Amendment No.: , I * ,-2e

High drywell pressure probides a secoiid means of initiating the core spray to mitigate the consequences of loss-of-coolant accident. Its trip setting of 5 3 . 5 psig initiates the core spray in time to provide adequate core cooling. The break size coverage of high drywell pressure was discussed above. Low-low water level and high drywell pressure in addition to initiating core spray also causes isolation valve closure. These settings are adequate to cause isolation to minimize the offsire dose within required limits;

• It is permissible to make the drywell pressure instrument channels inoperable during performance of the integrated primary containment leakage rate test provided the reactor IS iii the COLD SWLTDOWN condition. The reason for this is that the Engineered Safety Features. which are effective i n case of a LOCA under these conditions, will still be effective because they will be activated (when the Engineered Safety Features system is required as identified in the technical specification of the system) by iow-low reactor water level.'

The scram discharge volume has two separate instrument volumes utilized to detect water accumulation. The high water level is based on the design that the water in the SDIV's. as detected by either set of level instruments, shall not be allowed to exceed 29.0 gallons: thereby.

permitting 137 control rods to scram. To provide further margin. an accumulation of not more rhan 14.0 gallons of water, as detected by either instrument volume. will resuit in a rod block and an alarm. The accumulation of not more than 7.0 gallons of water. as detected in either insrrurnent volume will result in an alarm.

Detailed aiialyses of transients have shown that sufficientprotection is provided by other scrams below 4 5 % power to permit bypassing of the turbine trip an However. for operational convenience. 40% of rated reactor the setpoint below which these trips are bypassed. This setpoint is c capacI t y .

A low condenser vacuum scram trip of 20 inches Hg has been provided t o protect the main condenser in the event that vacuum is lost. A loss of condenser vacuum would cause the turbine stop valves to close. resulting in a turbine trip transient.

The low condenser vacuum trip provides a reliable backup to the turbine trip. Thus, if then is a failure of the turbine trip on low vacuum, the reactor would automatically Scram at 20 inches Hg. The condenser is capable of receiving bypass steam until 7 inches H g vacuum thereby mitigating the transient and providing a margin. .

Tne settings to isolate the isolation condenser in the event of a break in the steam or condensare lines are based on the predicted maximum flows that these systems would experience during operation, thus permitting operation while affording protection in the event of a break. The settings correspond to a flow rate of less than three times the normal flow rate of 3.2X10' Ib/hr.

Upon initiation o f the alternate shutdown panel, this function is bypassed to prevent spurious isolation due to fire induced circuit faults.

OYSTER CREEK 3.1-5 Amendment No.: 13.49.-112,349;-GI-,-I?&

• Correction 1 I /30/87

The s e t t i n g o f ten times t h e stack release l i m i t f o r i s o l a t i o n of the a i r - e j e c t o r offgas l i n e i s t o permit the operator t o perform normal, immediate remedial action i f the stack l i m i t i s exceeded. The time necessary f o r t h i s action would be extremely short when considering the annual averaging which i s allowedunder 10 CFR 20.106, and, therefore, would produce i n s i g n i f i c a n t e f f e c t s on doses t o the public. .

Four r a d i a t i o n monitors are provided which i n i t i a t e i s o l a t i o n o f t h e reactor b u i l d i n g and operation o f the standby gas treatment system. Two monitors a r e located i n the v e n t i l a t i o n ducts, one i s located i n t h e area o f the r e f u e l i n g pool and one i s located i n the reactor vessel head storage area. The t r i p l o g i c i s b a s i c a l l y a 1 out o f 4 system. Any upscale t r i p w i l l cause the desired action. T r i p s e t t i n g s o f 17 mr/hr i n the duct and 100 mr/hr on the r e f u e l i n g f l o o r are based upon i n i t i a t i n g standby gas treatment system so as not t o exceed allowed dose rates o f 10 CFR 20 a t the nearest s i t e boundary.

The SRM upscale o f 5 x lo5 CPS i n i t i a t e s a rod block so t h a t the chamber can be relocated t o a lower f l u x area t o maintain SRM c a p a b i l i t y a t power i s increased t o the IRM range. Full scale reading i s 1 x 10 CPS.

This5rod block i s bypassed i n IRM Ranges 8 and higher since a l e v e l o f 5 x 10 CPS i s reached and the SRM chamber i s a t i t s f u l l y withdrawn p o s i t ion.

The SRM downscale rod block. o f 100 CPS prevents the instrument chamber from being withdrawn too f a r from the core during t h e period t h a t i t i s required t o monitor the neutron f l u x . This downscale rod block i s also bypassed i n IRM Ranges 8 and higher. I t i s not required a t t h i s power l e v e l since good i n d i c a t i o n e x i s t s 19 the Intermediate Range and t h e SRM w i l l be reading approximately 5 x 10 CPS when using IRM Ranges 8 and higher .

The IRM downscale rod block i n conjunction with t h e chamber f u l l - i n p o s i t i o n and range switch setting, provides a r o d block t o assure t h a t the I R M i s i n i t s most s e n s i t i v e condition before startup. I f the two l a t t e r conditions are s a t i s f i e d , control rod withdrawal may commence even i f the IRM i s not reading a t l e a s t 5%. However, a f t e r a substantial neutron f l u x i s obtained, the rod block s e t t i n g prevents the chamber from being withdrawn t o an i n s e n s i t i v e area o f the core.

The APRM downscale s e t t i n g o f 1 2 / 1 5 0 f u l l scale i s provided i n t h e RUN MODE t o prevent control rod withdrawal without adequate neutron monitoring.

High flow i n the main steamline i s set a t 120% o f r a t e d flow. A t t h i s s e t t i n g the i s o l a t i o n valves close and i n the event o f a steam l i n e break l i m i t the loss o f inventory so t h a t f u e l occur. The 120% flow would correspond t o the would e i t h e r i n d i c a t e a l i n e break or too h i g Temperature sensors are provided i n the steam l i n e tunnel t o provide ; f closure of the main steamline i s o l a t i o n valves should a break o r l e a k occur i n t h i s area o f the plant. The t r i p i s set a t 50°F above ambient temperature a t rated power. This s e t t i n g w i l l cause i s o l a t i o n t o occur for main steamline breaks which r e s u l t i n a f l o w o f a few pounds p e r minute o r greater. I s o l a t i o n occurs soon enough t o meet the c r i t e r i o n o f no c l a d perforation.

OYSTER CREEK 3.1-6 Amendment No. : ;I-pFp-;b@+?--&?&

w?+%--wwR-

TABLE 3 ,CONTD)

Sheet Y of 13

~ d i v i d uelectromatic

~i relief valve control switches shall not be placed in the Off position for more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (total time for all control switches) in any 30-day period and only one relief valve control switch may be placed in the Off position at a time. I

1. With two core spray svstems OPERABLE:

1. A maximum of two core spray booster pump differential pressure (CUP) switches may be inoperable provided that the switches are in opposing ADS trip system [i.e., only: either RV-40 A&D or RV-40 B&rC]. Place the relay contacts associated with the inoperable d/p switch(es) in the de-energized position, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Restore the inoperable (Vp switch(es) within 8 days, or declare ADS inoperable and take the action required by Specification 3.4.B.3;

2. If two inoperable dfp switches are in the same ADS trip system [ i s . , RV-40 A&B or RV-40 C&D], place the reiay contacts associated with the inoperable d p switchtes) in the de-energized position, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Restore the inoperable cVp switches within 4 days, or declare ADS inoperable and take the action required by Specification 3.4.B.3.

With only one core spray system OPERABLE:

If one or more d/p switches become inoperable in the OPERABLE core spray system, declare ADS inoperable and t&e the action required by Specification 3.4.B.3.

-9

j. Not required betow 40% of rated reactor .-Jd*

k. All four (4)drywell pressure instrument channels may be made inoperable during the integrated primary containment leakage rate test (See Specification 4S), provided that the plant is in the COLD SHUTDOWN condition and that no work is performed on the reactor or its connected systems which could result in lowering the reactor water level to less than 48 above the TOP OF THE ACTIVE FUEL.

1. Bypass in IRM Ranges 8,9, and 10.

m, There is one time delay relay associated with each of two pumps.

n. One time delay relay per puny must be OPERABLE.

OYSlXR CREEK 3.1.17 Amendment No. m;

2. The circuit breaker of the recirculation pump motor generator set associated with an ISOLATED RECIRCUL,ATION LOOP shall be open and defeated from operation.

3. An ISOLATED RECIRCULATION LOOP shall not be returned to service unless the reactor is in the COLD SHUTDOWN condition.

b. When there are two inoperable recirculation loops (either two IDLE RECIRCULATION LOOPS or one IDLE RECIRCULATION LOOP

• ~~*-.--..%---- -- --,

RECIRCULATION LOOP) the reactor core not exceed 90% of rated power.

__/

7

_/"

3. if Specifications 3.3.F.1and 3.3.F.2are not met, an orderly shutdown shall be initiated immediately until all operable control rods are fully inserted and the reactor is in either the REFUEL MODE or SHUTDOWN CONDITION within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4. With reactor cooiant temperature greater than 212°F and irradiated fuel in the reactor vessel, at least one recirculation loop discharge valve and its associated suction valve shall be in the full open position.

5. If Specification 3.3.F.4 is not met, immediately open one recirculation loop discharge valve and its associated suction valve.

6. With reactor coolant temperature less than 212°F and irradiated fuel in the reactor vessel, at least one recirculation loop discharge valve and its associated suction valve shall be in the full open position unless the reactor vessel is flooded to a level above 185 inches TAF or unless the steam separator and dryer are removed.

OYSTER CREEK 3.3-3a Amendment No :

5. Pressure Suppression Chamber - Drywell Vacuum Breakers

a. When primary containment is required, all suppression chamber -

drywell vacuum breakers shall be OPERABLE except during testing and as stated in Specification 3.5.A.5.b and c, below.

Suppression chamber - drywell vacuum breakers shall be considered OPERABLE if:

(1) The valve is demonstrated to open from closed to fully open with the applied force at all valve positions not exceeding that equivalent to 0.5 psi acting on the suppression chamber face of the valve disk.

(2) The valve disk will close by gravity to within not greater than 0.10 inch of any point on the seal surface of the disk when released after being opened by remote or manual means.

(3) The position alarm system will annunciate in the control room if the valve is open more than 0.10 inch at any point along the seal surface of the disk.

b. Five of the fourteen suppression chamber - drywell vacuum breakers may be inoperable provided that they are secured in the I

closed position. With one of the nine required suppression chamber

- drywell vacuum breakers inoperable, restore one vacuum breaker to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

c. One position alarm circuit for each OPERABLE vacuum breaker may be inoperable, provided that each OPERABLE suppression chamber - drywell vacuum breaker with one defective alarm circuit, and associated remaining position alarm circuit are verified to be OPERABLE immediately, and monthly in accordance with 4.5.FS.a.

Additionally, a daily verification using the OPERABLE position alarm circuit that the affected vacuum breaker is closed shall be performed.

d. If Specifications 3.5 .AS (a), (b) or (c) can not be met, the reactor shall be PLACED IN the COLD SHUTDOWN CONDITION within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

7. Deleted.

OYSTER CREEK 3.5-4 Amendment No.: 2!, 32, W , $7, !% 24-Q

The capacity of the I4 suppression chamber to drywell vacuum relief valves is sized to limit the external pressure of the drywell during post-accident drywell cooling operations to the design limit of 2 psi. They are sized on the basis of the Bodega Bay pressure suppression A calculation (I5' was performed in accordance with NEDE-24802('6) to determine the required number of vacuum breakers by using a mass and energy balance to determine vacuum breaker flow area. The results of the calculation indicate that 8 vacuum breakers are required to provide vacuum relief capability. An additional vacuum breaker is included for single failure criteria, bringing the total required to 9.

Each suppression chamber drywell vacuum breaker is fitted with a redundant pair of limit switches to provide fail safe signals to panel mounted indicators in the reactor building and alarms in the control room when the disks are open more than 0.1" at any point along the seal surface of the disk. These switches are capable of transmitting the disk closed-to-open signal with 0.01 movement of the switch plunger. Continued reactor operation with failed components is justified because of the redundancy of components and circuits and, most importantly, the accessibility of the valve lever arm and position reference external to the valve. The fail-safe feature of the alarm circuits assures operator attention if a line fault occurs.

Conservative estimates of the hydrogen produced, consistent with the core cooling system provided, show that the hydrogen air mixture resulting from a loss-of-coolant accident is considerably below the flammability limit and hence it cannot bum, and inerting would not be needed. However, inerting of the primary containment was included in the proposed design and operation. The 5% oxygen limit is the oxygen concentration limit stated by the American Gas Association for hydrogen-oxygen mixtures below which combustion will not occur.(4) The 4% oxygen limit was established by analysis of the Generation and Mitigation of Combustible Gas Mixtures in Inerted BWR Mark I Containments."*'

OYSTER CREEK 3.5-10 Amendment No.: 43 3.6.A 10 CFR 100, am implemented by sRP Section 15.6.4, require. t h a t t h e r a d i o l o g i c a l conaequences of f a i l u r e of a main steam l i n e outmi.de containment be limited t o small fraction. of t h e expoaure guideline. of 10 CFR 100. During Syatomatic EValU8tiOn Program (SIP) t o r Oyater Creek, an independent asmrssmnt of t h e radiological conmrquencem of a main steam line f a i l u r e o u t r i d e containment (SEP Topic XV-18) w a 8 performed by the NRC 5 t a f f .

The amaemmnmnt determined t h a t i f t h e eximting Oymter Creek Technical Specification l i m i t for primuy coolant iodine a c t i v i t y ( 8 . 0 u c i t o t a l iodine par gram) i m used, t h e p o t e n t i a l o f f s i t e dome. would excrclcl t h e applicable dome l i m i t . The ataff recommended t h a t Oymter C r e e k maintain t h e primary coolant radioiodine a c t i v i t y within t h e General Slectric Standard Technical Spacification (NUREG-0123) l i m i t ( 0 . 2 uCf/grua DOSZ EQUIVALENT 1-131), which would m e e t t h e acceptance criteria.

However, t h e s t a f f . uralyaem f o r o y s t e r Creek showed t h a t small-line f a i l u r e 8 are more limiting than t h e main ateam l i n e f a i l u r e . 10 CFR ZOO, a. impluam+ed by SRP Section 13.6.2, requirem t h a t t h e r8diological conm.quencea of f a i l u r e O f amall l i n e s carrying primary coolant outmid. containment be limited t o s l m r i l l fractionm of t h e expoaura quidelines of 10 CFR 100.

During t h e evaluation o f SER Topic XV-16 wRadiologlical Conmequencem of F a i l u r e o t Saul1 Lines Carrying primary c M > L a n t Outaide C o n t a i m n t

• t h o Staff determined t h a t Oyster Cr-k doem not comply w i t h current acceptance c r i t e r i a . The S t a f f recommended t h a t t h e General Electric Standard TeChnkC8l Spocification (NUREG-0123) l i m i t (0.2 uCi/grua DOSE EQWIVIUEST 1-131) f o r raactor coolant radioiodine .r+ivity br adoptad in order t o enmure t h a t t h e radiological conaequences t o t h e environment from a f a i l u r e o f small l i n e s a r e acceptably low.

The LCO mtatemont parmitting power oporation t o continue f o r limited t i m e period@w i t h t h e primary c o o l a n t s a p r c i f i c a c t i v i t y greater than 0.2 microcuriem par gram DOSE EQUXVALENT 1-131, b u t lemr than o r equal to 4.0 microcurie EQUIVALENT 1-131, aCcOrrrmOdatea poaaible iodine phenomenon which may occur following change# i n The r y p r t i n q of cumulative operating t i m o w i t h microcuriem pot gram DOSE EQUIVALENT 1-131 will allow m u f f i c i e n t i tFrrm for Commim8ion t o evaluate t h e circumrtance8. 1I Info~a~io obtained n on iodine spiking will be umed t o aamomm t h e parameters ammociated w i t h spiking phanoiiena. A reduction i n frequency o f i s o t o p i c analyaim following power chang8m may be /

p.m!~i@.fbl@i f j u m t i f i e d by thm data obtained.  !

The surveillance requirement8 provido adequate assurance t h a t exCe8mive specific a c t i v i t y level# i n t h o reactor coolant w i l l b 8 detect& i n s u f f i c i e n t time t o taka c o r r e c t i v e action. i i

3.6.8 RELOCATES TO THE ODCX OYSTER CREEA 3.604 Amendment No. : 4Pp--5-01Cyt36 I

The APRM response is used to predict when the rod block occurs in the analysis of the rod withdrawal error transient. The transient rod ;ositis: at the rod block and corresponding MCPR can be determined. The MCPR has been evaluated for different APRM responses which would result from changes in the APRM status as a consequence of bypassed APRM channel and/or failed/bypassed LPRM inputs. The steady state MCPR required to protect the minimum transient CPR for the worst case APRM status condition (APRM Status 1) is determined in the rod withdrawal error transient analysis. The steady state MCPR values for APRM status conditions 1, 2, and 3 will be evaluated each cycle. For those cycles where the rod withdrawal error transient is not the most severe transient the MCPR value for APRM status conditions 1, 2, and 3 will be the same and be equal to the limiting transient MCPR value.

The time interval of Eight (8) hours to adjust the steady state of MCPR to account for a degradation in the APRM status is justified on the basis of instituting a control rod block which precludes the possibility of experiencing a rod withdrawal error transient since rod withdrawal is physically prevented. This time interval is adequate to allow the operator to either increase the MCPR to the appropriate value or to upgrade the status of the APRM system while in a condition which prevents the possibility of this transient occurring.

Transients analyzed each fuel cycle will be evaluated with respect to the operational MCPR limit specified in the COLR.

The purpose of the kr factor is to define operating limits at other than rated flow conditions. At less than 100% flow the required MCPR is the product of the operating limit MCPR and the k, factor. Specifically, the kf factor provides th I margin to protect against a flow

~ - - - - - - - - - - - % . -

increase transient.

The kf factor curve R, here developed generically using the flow control line corresponding t g r a t e d core flow. For the manual flow control mode, the kffactors were calculated such that at the maximum flow state (as limited by the pump scoop tube set point) and the corresponding core power (along the rated flow control line), the limiting bundles relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs were calculated at different points along the rated flow control line corresponding to different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR determines the value of kf.

The kf factor also provides the required thermal margin to protect against reactor thermal hydraulic instability. The kt factor establishes the required MCPR at low flow conditions such that if a reactor thermal hydraulic instability were to occur, the MCPR Safety Limit would not be exceeded.

OYSTER CREEK 3.10-4 Amendment No.: 7 2 3 5 -

IRfl calibration I S to be perfornied durin$ reactor startup The calibration of the IRMs dunng startup will be significant since the IRhls will be relied on for neutron monitoring and reactor protection u p 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)I! 3 A 1 allows the APffiVs to be reading greater than actual compensate for localized power peakiny When this adjustment is m absolute difference between the .APffiMchannels and the calculated power to indicate within 2% I RTP ismodified to-mchdeany yam 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 I000 IM\NDIT Frequency is based on operating experience with LPRM sensitivity changes.

General Electric Licensing Topical Report NEDC-3085 1 P-A (Reference I), 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. The test of the automatic scram contactors can be performed as part of the Channel Calibration or Test of Scram Functions or by use of the subchannel test switches.

References:

(I) NEDC-3085 1 P-A, "Technical Specification Improvement Analyses for BWR Reactor Protection System."

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

NEDC-3085 1 P-A, Supplement I , "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation I' MDC-3085 1 P-A, Supplement.2, "Technical Specification' Improvement Analysis for BWR lsolation Instrumentation Common to RPS and ECCS Instrumentation."

NEDC-3 1677P-A. "Technical Specification lmprovement Analysis for BWR Isolation Actuation Instrumentation."

OYSTER CREEK 4.1-3 AMENDMENT NO.: 7i;irr ,*

4.2 REACTIVITY CONTROL Amlicability: 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 a c c u m u l a t o r s k - ~ ~ - - ~ - - ~

1. For all control rods prior to xceeding 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 operation, for at least 20 control rods, on a rotating basis, with reactor coolant pressure greater than 800 psig.

D. Each partially or fully withdrawn control rod shall be exercised E . least once each week. 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 I~ l l ory 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: #&-S$,

-a!&

4.6 IOACTIVE EFFLUENT A ~ ~ I i c a b i l iApplies

~: to monitoring of gaseous and liquid radioactive effluents of the Station during release c effluentsvia 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 equipmer Objective: To measure radioactive effluents adquately to verify that radioactive effluents are as low 11s is reasonably achievable and within the h i t of 10 CFR Part 20. I A.

Reactor cooiant shall be sampled and d y z e d at 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 SHUTDOVVN CONDXTION.

B.. NOT USED.

c.

1. Liquids contained in the following tanks shall be sampled and analyzed for radioactivity at 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.

I). Main Condenser 0 - Treatmen4 RELOCATED TO THE ODCM.

E.

1. The gross radioactivity in fission gases discharged @omthe main condenser air ycctor shall be measured by sampiing and analyzing the gases.

a. at least once per month, and

b. When the reactor is operating at mom than 40 percent of rated p o w , within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> aftrx 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 increa!m(s) due to change(s) in the IWd. /,"--------- ------.-

The concentration of hydrogen in offgases downstream of the recombiner in the Wgkf System ShaIl be monitored with hydrogen instrumentation as described in Table 3.15.2.

G. NOTUSED.

H. NOTUSED.

OYSTER CREEK 4.6-1 Amendment No.:tBrs;-)