Text

Response to Request for Additional Information, Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling
	ML093220821
Person / Time
Site:	Oyster Creek
Issue date:	11/13/2009
From:	Cowan P; Exelon Generation Co, Exelon Nuclear
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RA-09-083
	Download: ML093220821 (45)
	v • d • e

Exelon Nuclear www.exelonCOTP.COM Exelon.

200 Exelon Way Nuclear Kennett Square, PA 19348 10 CFR 50.90 RA-09-083 November 13, 2009 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Renewed Facility Operating License No. DPR-16 Docket No. 50-219

Subject:

Response to Request for Additional Information - Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling

References:

1) Letter from Pamela B. Cowan to U.S. Nuclear Regulatory Commission Technical Specification Change Request 338 - Secondary Containment Operability Requirements During Refueling, dated November 2, 2007

2) U.S. Nuclear Regulatory Commission facsimile dated March 25, 2008, Draft Request for Additional Information (RAI) Regarding Proposed License Amendment - Secondary Containment Operability Requirements During Refueling, Oyster Creek Nuclear Generating Station (Docket No. 50-219)

3) Letter from Pamela B. Cowan to U.S. Nuclear Regulatory Commission Response to Draft Request for Additional Information - AmerGen Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling, dated May 5, 2008

4) Letter from Pamela B. Cowan to U.S. Nuclear Regulatory Commission Supplemental Response - AmerGen Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling, dated July 3, 2008

5) U.S. Nuclear Regulatory Commission facsimile dated September 11, 2008, Draft Request for Additional Information (RAI) Regarding Proposed License Amendment - Secondary Containment Operability Requirements During Refueling, Oyster Creek Nuclear Generating Station (Docket No. 50-219)

6) Letter from Pamela B. Cowan to U.S. Nuclear Regulatory Commission Response to Request for Additional Information Supplemental Response -

AmerGen Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling, dated September 22, 2008 A-601 4ZK Attachment 8 transmittedherewith contains Sensitive Unclassified Non-Safeguards Information (SUNSI). When separatedfrom Attachment 8, this transmittaldocument is decontrolled.

U.S. Nuclear Regulatory Commission Secondary Containment Operability Response to NRC RAI Docket No. 50-219 November 13, 2009 Page 2

7) Letter from G. Edward Miller, U.S. Nuclear Regulatory Commission to Mr.

Charles G. Pardee, Exelon Generation Company, LLC, dated October 20, 2009, Oyster Creek Nuclear Generating Station - Request for Additional Information Regarding Secondary Containment Operability License Amendment Request (TAC No. MD7261)

By letter dated November 2, 2007 (Reference 1), Exelon Generation Company, LLC (Exelon)

(formerly AmerGen Energy Company, LLC) submitted a License Amendment Request (LAR) to revise the Oyster Creek Nuclear Generating Station (OCNGS) Technical Specifications (TS) to modify the requirements for Secondary Containment operability during-handling of irradiated fuel with sufficient decay.

Exelon provided additional information by letters dated May 5, 2008 (Reference 3) and September 22, 2008 (Reference 6), in response to U.S. Nuclear Regulatory Commission (NRC) requests for additional information (References 2 and 5) concerning this LAR. In addition, Exelon also provided supplemental information in a letter dated July 3, 2008 (Reference 4).

Subsequently, by letter dated October 20, 2009 (Reference 7), the NRC requested additional information in order to complete the review of the LAR. Attachment 1 to this letter restates the NRC's questions followed by Exelon's response. to this letter includes a revised Section 2.0, "ProposedChanges," which supersedes the information previously submitted in Section 2.0 of the November 2, 2007, submittal (Reference 1). Attachment 3 to this letter contains the revised TS mark-ups, and includes the revised TS Bases mark-ups. These mark-ups were modified to address the NRC concern with needing additional definition in the TS. Attachment 5 includes a copy of Calculation C-1 302-822-E310-081, Revision 1, "OysterCreek Onsite Atmospheric Dispersion (X/Q) for Fuel Handling Accident (FHA)." Attachment 6 includes a copy of Calculation C-1 302-822-E310-082, Revision 2, "OysterCreek Analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms (AST)." Attachment 7 includes a revised Section 5.0, "Regulatory Analysis," which supersedes the information previously submitted in Section 5.0 of the November 2, 2007, submittal (Reference 1). Attachment 8 includes annotated drawings showing the evaluated release locations. The information included in is considered "Sensitive Unclassified Non-Safeguards Information"(SUNSI) and should be withheld in accordance with 10 CFR 2.390.

Exelon has determined that the information provided in this response further supplements the information provided in the Technical Analysis in the original submittal (Reference 1) and supporting supplemental responses (References 3, 4, and 6), and it does not impact the Environmental Co'nsideration previously submitted. However, as indicated above, Exelon has revised Section 2.0, "Proposed Changes," and Section 5.0, "Regulatory Analysis," of the Reference 1 submittal to reflect the analysis provided in this response.

There are no regulatory commitments contained in this submittal.

U.S. Nuclear Regulatory Commission Secondary Containment Operability Response to NRC RAI Docket No. 50-219 November 13, 2009 Page 3 If any additional information is needed, please contact Mr. Richard Gropp at 610-765-5557.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 13th day of November 2009.

Respectfully, Pamela B. Cowan Director, Licensing and Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Response to NRC Request for Additional Information

2) Revised License Amendment Request Section 2.0, "Proposed Changes"

3) Revised Mark-ups of Technical Specifications Pages

4) Revised Mark-ups of Technical Specifications Bases Pages

5) Calculation C-1 302-822-E310-081, Revision 1, "OysterCreek Onsite Atmospheric Dispersion(X/Q) for Fuel HandlingAccident (FHA)"

6) Calculation C-1 302-822-E310-082, Revision 2, "OysterCreek Analysis of Fuel HandlingAccident (FHA) Using Alternative Source Terms (AST)"

7) Revised License Amendment Request Section 5.0, "Regulatory Analysis"

8) Annotated Drawings Showing Evaluated Release Locations (contains SUNSI) cc: Regional Administrator - NRC Region I w/o Attachment 8 NRC Senior Resident Inspector - OCNGS NRC Project Manager, NRR - OCNGS Director, Bureau of Nuclear Engineering, New Jersey Department of Environmental Protection

ATTACHMENT 1 Oyster Creek Nuclear Generating Station Response to NRC Request for Additional Information Secondary Containment Operability Requirements During Refueling Secondary Containment Operability Response to NRC RAI Page 1 of 16

Background

By letter dated November 2, 2007 (Reference 1), Exelon Generation Company, LLC (Exelon)

(formerly AmerGen Energy Company, LLC) submitted a License Amendment Request (LAR) to revise the Oyster Creek Nuclear Generating Station (OCNGS) Technical Specifications (TS) to modify the requirements for Secondary Containment operability during handling of irradiated fuel with sufficient decay.

Subsequently, by letter dated October 20, 2009 (Reference 2) the U.S. Nuclear Regulatory Commission (NRC) requested additional information in order to complete the review of the License Amendment Request (LAR). The specific questions are restated below followed by Exelon's response.

NRC Question 1 In the November 2, 2007, submittal, Attachment 2, Table 1, "Conformancewith Regulatory Guide (RG) 1.183 Main Sections," states that the submittal "conforms"to Regulatory Position 5.1.2 and that "[tlheanalysis takes no credit for safety related features." Later correspondence contained in calculation C-1302-822-E31 0-082, Revision 1, continues to state conformance to RG 1.183.1 Regulatory Position5.1.2 states:

Credit may be taken for accident mitigation features that are classified as safety-related, are required to be operable by technical specifications,are powered by emergency power sources, and are either automaticallyactuated or, in limited cases, have actuation requirements explicitly addressedin emergency operatingprocedures.

In the September 22, 2008, response to a Request for Information, Exelon stated:

AmerGen [Exelon] will update the UFSAR [UpdatedFinal Safety Analysis Report] to include the acceptable secondary containmentpenetrationsand openings that could be breached/opened while moving irradiatedfuel with sufficient decay ... Any additional penetrationsand openings not included in the UFSAR (as outlined in Table 1 below in response to NRC Question 2) must be analyzed in accordance with applicableregulatory requirements (i.e., 10CFR50.59) before relaxation of secondary containment requirements for movement of irradiatedfuel with sufficient decay. The method of evaluation used will demonstrate that radiologicalconsequences associatedwith the Fuel HandlingAccident (FHA) do not exceed applicable regulatorydose limits.

Based upon the above, the statement that, "[t]heanalysis takes no credit for safety related features," and the stated conformance to Regulatory Position 5.1.2 appearto be inconsistent.

By proposing a limited number of acceptablepenetrationsand openings that can be breached, Exelon credits the capability of any remainingsecondary containment accident mitigation features as being capable of performing their safety functions for the analyzed conditions for the duration of their mission times. 2 However, the licensee's proposedtechnical specification (TS) changes remove all requirements for all secondary containmentaccident mitigative features after24 hours. Instead, Exelon proposes that the secondary containment mitigative features are to be established in the UFSAR. Exelon's proposed deletion of TSs associatedwith secondary Secondary Containment Operability Response to NRC RAI Page 2 of 16 containmentoperability and incorporationof controls in the UFSAR is not consistent with Regulatory Position 5.1.2. In accordance with 10 CFR 50.36, "Technicalspecifications,"

Exelon's proposed continued reliance on some safety-related features of secondary containment to function or actuate to mitigate a design-basisaccident necessitates their inclusion in the TSs.

The TSs proposed in the originalLAR, which were not amended in the July 3, 2008, supplement or the September 22, 2008, RAI response, are insufficient for the NRC staff to find that the licensee has provided the lowest functional capabilityor performance level of equipment for safe operationof the facility that would provide reasonableassurancethat, in the event of an FHA when secondarycontainment is INOPERABLE, the dose consequences will meet NRC regulatoryrequirements.

Therefore, the NRC staff requests that the licensee provide revised TS changes, consistent with its proposed revised analysis of record, that ensure the lowest functional capabilityor performance level of equipment credited for functioning or actuating to mitigate the design basis fuel handling accident.

1According to calculation, C-1302-822-E31 0-082, Revision 1: 1) "This calculationdetermines the safety features required to assure that regulatorylimits in 10CFR50.67are met, and is performed in conformance with guidance for analysis of this event provided in Regulatory Guide (RG) 1.183, Appendix B. "2) "Dose models for both onsite and offsite are simplified and meet RG 1.183 requirements,"and 3)

"This analysis uses Alternative Source Term (AST) assumptionsper guidance in RG 1.183."

2 Per page 14 of calculation C-1302-822-E31 0-082, Revision 1 several structures and components are part of the primary success path and function to mitigate the Fuel Handling Accident. Specifically, Exelon states that the Commodities Penetrationon the RB North Wall, MAC Facility PersonnelAirlock, MAC FacilityEntrance, Trunion Room Door to Turbine Building are credited in the analysis.

Response - NRC Question 1 Exelon has determined it appropriate to revise the Technical Specifications (TS) and TS Bases pages to reflect that the locations (i.e., doors and hatches that do not involve disassembly of the Secondary Containment) as described in Table 1 of this attachment can be opened during movement of irradiated fuel provided that there is sufficient decay of the irradiated fuel. This decay period has been demonstrated in a design analysis to be 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months , the time at which Secondary Containment and Standby Gas Treatment System (SGTS) are no longer required to meet dose limitations for the Control Room (CR) and offsite locations. However, one SGTS circuit will be available to be used to draw any potential release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored. The only function required during this shutdown condition with fuel movement in progress with sufficient decay is for monitoring purposes and as a defense-in-depth measure, thereby further mitigating any possible dose to the public in the event of a fuel handling accident. Local grab samples at Secondary Containment openings may be employed as additional monitoring for radioactive airborne activity within the Secondary Containment. Provisions will be provided to ensure Secondary Containment openings as described in Table 1 of this attachment can be closed within one hour.

Secondary Containment Operability Response to NRC RAI Page 3 of 16 Exelon considers that these proposed changes are consistent with Regulatory Position 5.1.2.

The proposed TS and TS Bases changes are provided in Attachments 3 and 4 of this submittal.

NRC Question 2 In NRC Regulatory Issue Summary 2006-04, "Experience with Implementation of Alternative Source Terms," the NRC reiteratedits regulatoryposition that "Licenseesare responsible for identifying all releasepathways and for consideringthese pathways in their AST analyses, consistent with any proposed modification." During the course of the review, which includes its supplements and RAI responses, the licensee has provided three separatelists of analyzed or considered releasepoints and pathways. In reviewing these lists, the staff has identified variationsthat raise concerns regardingwhether the licensee has analyzed all potential release points and pathways to ensure that regulatorydose limits would be met in the event of an FHA.

Consistent with NRC's established regulatoryposition, the NRC staff requests that the licensee provide a comprehensive list of all analyzed and unanalyzed secondary containmentpotential release points and pathways to the environment and control room. These pathways should include those pathways to adjacentbuildings that could lead to the environment or to the control room (i.e. Secondary Containment HVAC ductwork, structuralopenings etc.). Additionally, the licensee's evaluation of the pathways should consider the effects of operabilityor inoperabilityof other safety systems such as the Secondary Containment Isolation Valves and Standby Gas Treatment System (SGTS). For each potential releasepoint or pathway, the licensee should provide the following:

a. The results of its dose analysis demonstratingthat 10 CFR 50.67 regulatorylimits are met;

b. If a dose analysis has not been performed, a technically sound basis for why this release point or pathway is bounded by other analyzed releasepoints; and

c. An explanation for how the existing proposed TS changes or, as necessary, new revised TSs will ensure that the dose limits are met.

Response - NRC Question 2 Exelon proposes to allow doors, penetrations, and hatch openings which do not require disassembly of the Secondary Containment to be opened during shutdown conditions with fuel movement in progress. These opening are described in Table 1 of this attachment, with one exception (i.e., Northwest RB Personnel Airlock to Office Building at Elev. 51'-3"). With respect to those doors and penetrations, these items are, and have always been, credited as being part of Secondary Containment. There are no new systems or components being credited and qualified as such in this analysis. One SGTS circuit will be available to be used to draw any potential release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored. The only function required during this shutdown condition with fuel movement in progress with sufficient decay is for monitoring purposes and as a defense-in-depth measure, thereby further mitigating any possible dose to the public in the event of a fuel handling accident. Local grab samples at Secondary Containment openings may be employed as additional monitoring for radioactive airborne activity within the Secondary Containment.

Provisions will be provided to ensure Secondary Containment openings can be closed within one hour.

Secondary Containment Operability Response to NRC RAI Page 4 of 16 The results of the dose analysis are dependent on many bounding analysis factors and assumptions as indicated in Exelon's July 3, 2008, supplemental response and included in Calculation C-1 302-822-E310-082, Revision 1, "OysterCreek Analysis of Fuel Handling Accident (FHA) Using Alternative Source Terms (AST)," and Revision 2 (Attachment 6). The varying factor for potential results is based on release pathway X/Q values, which is a function of location orientation in respect to the receptor points of the CR, Exclusion Area Boundary (EAB), and Low Population Zone (LPZ). The worst-case X/Q for any opening permitted to be open is used in the revised Calculation C-1302-822-E310-081, Revision 1, "OysterCreek Onsite Atmospheric Dispersion(X/Q) for Fuel HandlingAccident (FHA)" (Attachment 5). The calculation has determined that a 168-hour decay time is required for permitting openings, without the requirement of Secondary Containment or SGTS filtration. After this time, penetrations can be opened in accordance with Table 1 of this attachment with acceptable dose consequences.

The airlock on the northwest corner (51'-3" elevation) of the Reactor Building provides for a means of ingress and egress from the Reactor Building to the Office complex area (46'-6" elevation). Both doors are closed to provide the boundary except during transit. In addition, even during transit, the airlock is provided with two doors that have interlocks so that only one door will be opened at a time. This ensures that the Secondary Containment boundary is maintained, as required. The airlock enclosure is completely surrounded by concrete (walls, floors, and ceiling). The airlock is shown in more detail on one of the drawings (BR 4054) contained in Attachment 8. The CR is located on the 46'-6" elevation in the Turbine Building (northeast corner) adjacent to the Office complex. The CR has its own Heating, Ventilation and Air Conditioning (HVAC) system and is designed to maintain the CR at a pressure slightly higher than atmospheric pressure. This is a requirement specified in current OCNGS TS 4.17 B.

Those release locations considered in the analysis are described in Table 1 of this attachment.

NRC Question 3 In Table 4-3, "ParametersApplicable to AST Fuel HandlingAccident Dose Considerationsfor Oyster Creek Nuclear Generating Station," of the July 3, 2008, supplement, Exelon states that no credit is taken for filtrationby the SGTS. However, in the same supplement, Exelon provides a commitment to provide ". ...prompt methods... to enable ventilation systems to draw the release from a postulated fuel handlingaccident in the proper direction such that it can be treated and monitored." This appears to be a reference to the SGTS. Based on the proposed TS changes from the original LAR, it does not appearthat the SGTS will be required to be operable during non-recently irradiatedfuel handling operations (i.e., after the reactorhas been subcriticalfor 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). As such, it would not be available for performing the safety function described in the commitment. With the SGTS inoperable, the licensee would be unable to create the differential pressure inside the secondary containment necessary for the purposes of directing the radioactiverelease from a fuel handlingaccident through the SGTS filtration and the Main Stack. As such, otherpotential releasepoints or pathways, such as smallersecondary containmentpenetrations and pathways the licensee has determinedmust remain closed even after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of decay time, that have not been analyzed or considered may become relevant dose contributors. Therefore, the NRC staff requests the licensee provide the following information:

Secondary Containment Operability Response to NRC RAI Page 5 of 16

a. An analysis demonstrating that the SGTS can perform its safety function under all possible plant configurations relatedto secondary containment operabilityduring fuel handling operations. This analysis should consider the potential impacts of differentialpressures caused by local wind conditions.

b. Appropriate TS related to SGTS operabilityduring periods when it is credited for performing a safety function related to the mitigating the consequences of an FHA.

Response - NRC Question 3 The proposed TS require that one SGTS circuit be available within one hour after a fuel handling accident; therefore, the "prompt methods" as committed in Exelon's LAR submitted by letter dated November 2, 2007, can be satisfied. However, Exelon will not credit for the filtration or elevated release if any Secondary Containment penetration (e.g., door, hatch, or otherwise) is open during fuel movement.

Given the 168-hour delay prior to allowing penetrations to be open with the exception of the airlock as described (Northwest RB Personnel Airlock to Office Building at Elev. 51'-3"), SGTS is no longer required, as the exhaust is assumed to exit the building at the worst permitted release locations. This includes the diffuse area of the reactor building wall.

NRC Question 4 Based on the differences identified between the licensee's analyses and its statements regardingconformance with Regulatory Position5.1.2 discussed in question #1, the NRC staff requests that the licensee reevaluate its conformance with Regulatory Position 5.1.2 and provide additionaljustification that all credited accident mitigation features are classified as safety-related,are required to be operable by TSs, are powered by emergency power sources, and are either automatically actuatedor, in limited cases, have actuation requirementsexplicitly addressedin emergency operatingprocedures.

Response - NRC Question 4 There are no new accident mitigation features credited as a result of this LAR. The credited accident mitigation features are classified as safety-related, are required to be operable by TS, are powered by emergency power sources, and are automatically actuated. Secondary Containment openings as described in Table 1 of this attachment are permitted to be open during movement of irradiated fuel with a minimum of 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months decay. These Secondary Containment openings will be capable of being closed within one hour after a fuel handling accident. The exception identified in Table 1 of this attachment (i.e., Item 15 - Northwest RB Personnel Airlock to Office Building at Elev. 51'-3") is a safety-related Secondary Containment boundary and will remain closed.

NRC Question 5 Exelon has proposed the following commitment in the July 3, 2008, supplement:

Secondary Containment Operability Response to NRC RAI Page 6 of 16 Plantprocedures will continue to require that secondary containment integrity be maintained when handling heavy loads (greaterthan one fuel assembly), such as the reactor vessel head or dryer/separatorassembly, over the reactorcavity with fuel in the reactorvessel.

Currently, TS requirements 3.5.B. 1.c, 3.5.B. 1.d and 3.5.B. i.e (consistent with 10 CFR 50.36, Criterion 3) exist to require secondary containment integrity when heavy loads could cause a release of radioactive materials (i.e. reactorvessel head is on, operationsare not being performed in, above, or around the spent fuel pool that could cause release of radioactive materials, etc.). Exelon proposes to delete or modify these TS requirements. However, the licensee has not provided a technicaljustification for why these controls are no longer required to account for an FHA resulting from the potential drop of a heavy load. Such an accident has the potential to result in greaterfuel damage and radioactive release than that assumed in the license amendment. As such, the NRC staff request that the licensee provide a technical justificationfor why these limiting conditions for operation are not requiredto establish the lowest functional capability or performance levels for equipment required for safe operation of the facility (in accordancewith 10 CFR 50.36) for movement of heavy loads over the reactor cavity or spent fuel pool.

Response - NRC Question 5 Exelon modified TS Section 3.5.B, "SecondaryContainment," regarding the need to maintain Secondary Containment integrity during movement of heavy loads in, above, or around the reactor cavity and spent fuel pool. Refer to the discussion in Attachment 2 of this submittal regarding the proposed TS changes for controlling heavy loads over the reactor cavity and spent fuel pool. Attachment 3 contains the proposed TS mark-ups.

NRC Question 6 In the November 2, 2007, submittal, Section 2 provides the proposedchanges. No justification is provided for the change labeled 2.4 and additionaljustificationis needed for changes 2.3, 2.5, 2.6 and 2.8. Many of the proposed changes cite a conformance with TSTF-51, Rev. 2 as the justification. However, the licensee'ssubsequent revisions to the originalLAR reduce its consistency with TSTF-51, Rev. 2. Therefore, the NRC staff requests that the licensee provide further detailedjustification for each proposed change.

Note: The NRC staff recognizes that the licensee's response to the RA/s above may result in significant changes to the TSs proposed in the November 2, 2007, submittal. Substantial changes could invalidate or rendermoot the originaljustification for the proposedchanges. In that case, the NRC staff encourages the licensee to completely revise Section 2.0, "Proposed Changes"and submit a new Section 2.0 which includes a clear technicaljustification for each proposed change.

Response - NRC Question 6 An updated Section 2.0 describing and justifying the proposed TS changes is provided in . The information contained in Attachment 2 of this submittal supersedes the information provided in Section 2.0 of Exelon's November 2, 2007, LAR submittal.

Secondary Containment Operability Response to NRC RAI Page 7 of 16 The justification for the proposed changes is based on the revised analysis submitted in Exelon's supplemental response dated July 3, 2008, and included in Calculation C-1302-822-E310-082, Revision 2, "OysterCreek Analysis of Fuel HandlingAccident (FHA) Using Alternative Source Terms (AST)," and Calculation C-1 302-822-E310-081, Revision 1, "Oyster Creek Onsite Atmospheric Dispersion (X/Q) for Fuel Handling Accident (FHA)." These revised calculations are included in Attachments 5 and 6 of this submittal.

NRC Question 7 In its amendment request, Exelon assumed that all radioactivitywill enter the control room through the HVAC intake ductwork. However, no technical basis for that assumption was provided. The NRC staff requests that the licensee provide a justification for this assumption. As part of its justification, the staff requests that the licensee reevaluate whether release pathways exist from the secondary containment into buildings connected to the control room. If such pathways exist, the staff requests that the licensee justify why the atmosphericdispersion factors used in its analysis are limiting. Finally, the staff requests that Exelon provide scale drawings showing the relationshipof the secondarycontainment to the control room.

Response - NRC Question 7 All radioactivity released during the FHA event is assumed to enter the CR through the CR HVAC intake louvers. This is conservative since the intake louvers are closer to the sources (than the CR itself), having higher X/Q values for any permitted opening. Therefore, it is conservative to use these X/Q values. There is one exception identified in Table 1 of this attachment (i.e., Item 15 - Northwest RB Personnel Airlock to Office Building at Elev. 51'-3").

This airlock must remain closed and this Secondary Containment boundary maintained during fuel movement.

As requested, Attachment 8 contains drawings that show the relationship of the Secondary Containment to the CR.

NRC Question 8 The NRC staff requests that the licensee explain how the monitoring of radioactivereleases resulting 'from an FHA or "inadvertentrelease of radioactivematerial" (GDC 63 and 64) will be accomplished with the secondarycontainment open. The currentlicensing basis for Oyster Creek assumes the secondary containment is operable during fuel handlingoperations and by extension would be operable during an FHA or an "inadvertentrelease of radioactive material" As such, any radiationmonitoring and filtering equipment inside secondary containment would have been designed, located, and calibratedbased on the current design and licensing basis.

The proposedchanges could impact the effectiveness of that monitoring equipment. For example, the timing of proceduralizedoperatoractions related to indicationsor alarms from this equipment could potentially be delayed or prevented by a reduced effectiveness of this equipment. The staff believes the ability to effectively monitor the radioactiverelease is critical to the protection of the public and plant personnel.

Secondary Containment Operability Response to NRC RAI Page 8 of 16 Response - NRC Question 8 Since one SGTS circuit will be available and running within one hour and all Secondary Containment openings will be capable of being closed within one hour, this will ensure that any release from a postulated fuel handling accident will be directed in such a manner that it can be treated and monitored. Local grab samples at Secondary Containment openings may be employed as additional monitoring for radioactive airborne activity within the Secondary Containment. Provisions will be provided to ensure Secondary Containment openings can be closed within one hour.

NRC Question 9 Regulatory Guide 1.194, "AtmosphericRelative Concentrationsfor Control Room Habitability Assessments at Nuclear Power Reactors," states that:

Diffuse source modeling should be used only for those situations in which the activity being releasedis homogenously distributedthroughout the buildingand when the assumed release rate from the buildingsurface would be reasonablyconstantover the surface of the building.

The release from the reactorcavity and spent fuel pool is to the area in the reactorbuilding that is above elevation (El.) 119'-3". The reactorbuilding is constructed entirely of reinforced concrete to the refueling floor level at El. 119'-3". Above the refueling floor, the structure is steel framework with insulated,corrosion-resistantmetal siding. Because of the differences in construction, leakage appearsto be more likely from the secondarycontainment above the refueling floor than from the secondarycontainmentbelow the refueling floor. Therefore, the NRC staff requests that the licensee justify the use of the entire exposed area of the reactorbuilding for calculation of the reactorbuilding diffuse source ratherthan using only the area of the building above the refueling floor where the materials of construction would be more likely to have releasepathways to the environment.

Additionally, the staff requests that the licensee provide a technical basis for its assumption that the activity being released will be homogenously distributedthroughout the building and that the release rate from the building surface will be reasonablyconstantover the surface of the building.

Response - NRC Question 9 A new diffuse area source has been determined using part of the metal siding surface area facing the CR HVAC intakes. Since a reasonable amount of mixing is anticipated, considering that temperature differences within the volume will cause turbulence, only 50% of this surface area is considered. Furthermore, this surface area is selected such that the resulting X/Q is maximized with respect to the worst-case CR intake location.

Due to the normal ventilation in effect during fuel movement the activity is conservatively distributed throughout the fuel handling area in determination of the diffuse area source X/Q.

However, for the purposes of this analysis, the released activity is only distributed into a volume of 100 ft3 .

The metal siding above the 119'-3" elevation is sealed, insulated, routinely tested via drawdown testing, and not expected to leak. The previous analysis included the entire surface area of the reactor building west wall in the consideration of a diffuse area source. A revised diffuse area Secondary Containment Operability Response to NRC RAI Page 9 of 16 NRC Question 10 Are any non-safety related systems and components credited in the alternatesource term analyses?If so:

a. Describe how this system will be electricallyseparatedfrom the safety-related system (provide a detailed discussion on how a fault on the non-Class 1E electrical circuit will not propagate to the Class 1E electrical circuit).

b. Describe the independence (e.g., electrical andphysical separation)and redundancy of these systems.

c. Describe how these systems meet the single failure criterion.

d. Describe how the operators will be notified in the event that these systems and components would become inoperable (e.g., control room annuciators).

e. Describe any impacts on seismic qualificationsof these systems and components.

Response - NRC Question 10 No non-safety related systems or components are credited as part of this LAR to revise the Technical Specifications regarding Secondary Containment operability requirements during refueling.

NRC Question 11 Are any loads being added to the Oyster Creek emergency diesel generators (EDGs)?If so, describe how the loads being added to the EDGs affect the capabilityand capacity of the EDGs (e.g., describe the impact of the proposedchange on the EDG ratings).

Response - NRC Question 11 No loads are required to be added to the OCNGS Emergency Diesel Generators (EDGs) as part of this LAR.

NRC Question 12 Provide the loading sequence for each EDG at Oyster Creek. In your response, describe the changes that have been made to the EDG loading sequence to support this LAR.

Response - NRC Question 12 See response to Question 11 above.

NRC Question 13 Secondary Containment Operability Response to NRC RAI Page 10 of 16 Provide a list and description of components being added to your 10 CFR 50.49 program due to this LAR. Confirm that these components are qualified for the environmental conditions they are expected to be exposed to.

Response - NRC Question 13 No additional equipment or components are being added to the 10 CFR 50.49 program as a result of this LAR.

NRC Question 14 Are there any changes in the chemical composition of the chemical spray solution as a result of this LAR? If so, provide the chemical composition and provide a detailed evaluation to show the components are qualified for the environmental conditions they are expected to be exposed to.

Also, describe, if any, changes in the operation of the chemical spray system and its impact on the environment.

Response - NRC Question 14 No chemical spray solution is credited as part of this LAR.

NRC Question 15 Confirm that Oyster Creek environmental qualification (EQ) analyses will continue to be based on Technical Information Document (TID)-14844, "Calculation of Distance Factors for Power and Test Reactor Sites" in the EQ program. Otherwise, provide the EQ analyses to support this LAR.

Response - NRC Question 15 Environmental Qualification (EQ) analyses will continue to be based on TID-14844 (Reference 3). No changes related to EQ are sought as part of this LAR.

References

1. Letter from Pamela B. Cowan to U.S. Nuclear Regulatory Commission Technical Specification Change Request 338 - Secondary Containment Operability Requirements During Refueling, dated November 2, 2007

2. Letter from G. Edward Miller, U.S. Nuclear Regulatory Commission to Mr. Charles G. Pardee, Exelon Generation Company, LLC, dated October 20, 2009, Oyster Creek Nuclear Generating Station - Request for Additional Information Regarding Secondary Containment Operability License Amendment Request (TAC No. MD7261)

3. TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," U.S. Atomic Energy Commission, March 23, 1962

Attachment 1 Secondary Containment Operability Response to NRC RAI Page 11 of 16 Table 1 Evaluated Release Locations Reactor Building West CR: <2.72 Since no specitic point-source leakage Wall (Modeled as a EAB: 1.16 is expected on this wall, the potential Diffuse Area) LPZ: 0.11 release source through this wall is best characterized as a diffuse area source.

As this is a modeling method used to identify a limiting release location, the TS does not allow any specific penetrations associated with this modeling.

Since only the metal siding portion of the east wall of the RB has the potential for leakage, the surface area is limited to this section of the wall. Furthermore, since complete mixing in the refueling area volume cannot be assumed, only 50% of the metal siding area is assumed in the determination of the diffuse area X/Q. This limited area is assumed to be at the worst case location with relationship to either CR intake location to maximize the calculated dose.

I + +

2 Drywell Access Facility CR: <1.41 West Wall: This is a temporary structure used as EAB: 1.16 1.61 E-03 the Drywell Support Center and Outage LPZ: 0.11 Command Center. It is connected to South Wall: the east Reactor Building personnel air 1.93E-03 lock through a series of temporary tunnels. These tunnels are not safety related; however, since they are connected to the RB personnel air lock, any air discharged through the air lock could be forced through the tunnel, being discharged at the doors to the temporary facility. These doors are closer to the CR intakes than the RB personnel air lock itself.

There are four (4) doors associated with the D/W Access Facility. All 4 doors are located in roughly the same direction relative to the CR intakes. The X/Q for the door on the northern wall would be

Attachment 1 Secondary Containment Operability Response to NRC RAI Page 12 of 16 Table 1 Evaluated Release Locations Dounaea Dy tne door on tne west wail, which is closer to CR intakes A and B.

Similarly, the two doors on the southern wall facing the RB, would bound the door on the west wall. The X/Q for the south eastern-most door would be bounded by the south western-most door, which is closer to CR intakes A and B.

3 Commodities CR: <1.29 (1) This is a flanged penetration, for which penetration on RB EAB: 1.16 1.77E-03 procedures are currently in place for South Wall Elevation LPZ: 0.11 routing commodities through the 23'-6" penetration without violating the secondary containment boundary.

However, for the purpose of determining the maximum dose to CR personnel, this penetration is considered to be open during shutdown conditions with movement of fuel with sufficient decay in progress.

4 Commodities CR: <3.81 *1) This is a flanged penetration, for which penetration on RB EAB: 1.16 5.21 E-03 procedures are currently in place for North Wall Elevation LPZ: 0.11 routing commodities through the 23'-6" penetration without violating the secondary containment boundary.

However, for the purpose of determining the maximum dose to CR personnel, this penetration is considered to be open during shutdown conditions with movement of fuel with sufficient decay in progress.

5 Commodities CR: <1.02 t This is a flanged penetration, for which penetration on RB EAB: 1.16 [1.40E-03] procedures are currently in place for East Wall, Elevation LPZ: 0.11 routing commodities through the 23'-6" penetration without violating the secondary containment boundary.

However, for the purpose of determining the maximum dose to CR personnel, this penetration is considered to be open during shutdown conditions with movement of fuel with sufficient decay in progress. This penetration is adjacent to the RB east personnel airlock and is considered bounded by

Attachment 1 Secondary Containment Operability Response to NRC RAI Page 13 of 16 Table 1 Evaluated Release Locations me airiOCK ýwnicn is signimicanviy iess than the bounding X/Q used in the dose analysis. Therefore, a specific X/Q was not calculated..

6 RB Roof Hatch CR: <1.33 "I The roof hatch is a personnel access EAB: 1.16 1.82E-03 way to the RB roof. Although not LPZ: 0.11 opened routinely, it is considered to be a potential release location in the event that it is open during shutdown conditions with movement of fuel with sufficient decay in progress.

7 Stack Tunnel Door CR: <0.62 (1) This is the access location to certain EAB: 1.16 8.55E-04 SGTS dampers and equipment to the LPZ: 0.11 east of the Reactor Building. It is modeled as a potential release location if opened during shutdown conditions with movement of fuel with sufficient decay in progress.

8 East RB Airlock Door CR: <1.02 (1) The east RB Airlock door is connected EAB: 1.16 1.40E-03 to the temporary tunnel leading to the LPZ: 0.11 Drywell Access Facility. However, since this temporary facility is not safety related, leakage from this location is postulated during movement of fuel with sufficient decay in progress.

9 South East RB Airlock CR: <1.29 t Since the south east RB Airlock door is Door EAB: 1.16 [1.77E-03] in the same relative direction as the LPZ: 0.11 south RB commodities penetration and farther away, it is considered bounded by the south RB commodities penetration. Therefore, a specific X/Q was not calculated.

10 Reactor Building Truck CR: <1.33 t1) Since the RB Truck Airlock door is in Airlock EAB: 1.16 [1.82E-03] the same relative direction and farther LPZ: 0.11 away from other calculated openings, this penetration is considered bounded by the south RB commodities penetration and the RB roof hatch.

Therefore, a specific X/Q was not calculated.

11 Isolation Condenser CR: <1.33 (1 Since these vents (on the RB east wall)

Vents EAB: 1.16 [1.82E-03] are in the same general direction as the LPZ: 0.11 RB roof hatch (and farther away) with respect to the CR intakes, a release

Attachment 1 Secondary Containment Operability Response to NRC RAI Page 14 of 16 Table 1 Evaluated Release Locations trom tnis location is considered bounded by the RB roof Hatch.

Therefore, a specific X/Q was not calculated.

12 MAC Facility Entrance CR: <4.84 (1) These double doors are modeled as a EAB: 1.16 6.62E-03 single penetration.

LPZ: 0.11 13 (MAC Facility CR: 4.93 The MAC Facility Personnel Airlock Personnel Airlock) EAB: 1.16 6.75E-03 exits out of tornado/missile Northwest RB LPZ: 0.11 protection area located on the north Personnel Airlock at RB wall (23'-6" elev.). This is the Elevation 23'-6" location with the maximum X/O for the dose analysis and results in the maximum dose for all penetrations allowed to be open during shutdown conditions with movement of fuel with sufficient decay.

14 Trunnion Room Door CR: <2.72 (1) The Trunnion Room is a subset of to Turbine Building EAB: 1.16 3.73E-03 secondary containment and houses the LPZ: 0.11 outboard MSIVs. The single access door is not an airlock. Access to this room is permitted (via TS) during operation through intermittent opening of the door under administrative controls.

15 Northwest RB CR: >5.0 The Northwest RB Personnel Airlock Personnel Airlock to EAB: 1.16 X/Q Not leads from the RB to the Office Building Office Building at LPZ: 0.11 Calculated on the 51' 3" elevation (Columns RF Elevation 51'-3" Based on and R6). Its closure is credited in the Inspection of analysis of the FHA. The upper Closeness to containment personnel air lock performs CR Access no active function in response to the Door postulated accident; however, its leak tightness is required to ensure that the NOT Permitted release of radioactive materials from to Be Open primary containment is restricted to During Fuel those leakage paths assumed in the Movement accident analysis, and the fission products released by the FHA will be treated by the SGT System. It was not originally on the list of penetrations allowed to be open. Since this airlock opens into the Office Building (close to the CR entrance), it is NOT permitted to

Attachment 1 Secondary Containment Operability Response to NRC RAI Page 15 of 16 Table 1 Evaluated Release Locations oe open aue to its proximity to tne UH HVAC intakes and the CR access door.

Based on this closeness, a specific X/Q 4 4 was not calculated.

16 Southwest RB CR: <2.72k, This airlock leads from the Personnel Airlock to EAB: 1.16 [3.73E-03] southwest RB to the Turbine TB at Elevation 6'-5" LPZ: 0.11 Building on the 3'-6"elevation (Columns J and R6). It was not originally on the list of penetrations allowed to be open.

However, it is assumed to be open during shutdown conditions with movement of fuel with sufficient decay in progress.

Since this location is in the same general direction as the Trunnion Room door and is farther away, the X/Q is considered bounded by the Trunnion Room door.

Therefore, a specific X/Q was not 4

calculated.

+

17 Northwest RB CR: <4.93W' This airlock leads from the northwest Personnel Airlock to EAB: 1.16 [6. 75E-03] RB to the Turbine Building on the 3'-6" TB at Elevation -1'-11" LPZ: 0.11 elevation (Columns RG and R6). It was not originally on the list of penetrations allowed to be open. However, it is assumed to be open during shutdown conditions with movement of fuel with sufficient decay in progress. Since this location is in the same general direction as the MAC Facility personnel airlock and is farther away, the X/Q is considered bounded by the MAC Facility personnel airlock. Therefore, a specific X/Q was not calculated.

18 Floor Plug to SW RB CR: <2.72 {" Since this location is in the same Corner Room EAB: 1.16 [3.73E-03] general direction as the Trunnion Room Elevation 23'-6" LPZ: 0.11 door and is farther away, the X/Q is considered bounded by the Trunnion Room door. Therefore, a specific X/Q was not calculated.

19 Floor Plug to NW RB CR: <4.93 t1) Since this location is in the same Corner Room EAB: 1.16 [6.75E-03] general direction as the MAC Facility Elevation 23'-6" LPZ: 0.11 personnel airlock and is farther away,

Attachment 1 Secondary Containment Operability Response to NRC RAI Page 16 of 16 Table 1 Evaluated Release Locations trie x/u is considered bounded by the MAC Facility personnel airlock.

Therefore, a specific X/Q was not calculated.

20 Service Water Pipe CR: <1.02 1.40E-03 At elevation 41'-6" and located Penetration EAB: 1.16 approximately 64' south of the North LPZ: 0.11 face of the Reactor Building (1) CR Dose estimated using a ratio to the maximum X/Q of 6.75E-03 sec/m 3 allowed.

A'TACHMENT 2 Revised License Amendment Request Section 2.0, "Proposed Changes" Page 1 of 4

Background

By letter dated November 2, 2007, Exelon (formerly AmerGen) submitted Technical Specification Change Request 338, "SecondaryContainment Operability Requirements During Refueling," requesting changes to the Oyster Creek TS that would modify Secondary Containment requirements during refueling operations. Exelon provided additional information concerning this proposed LAR by letters dated May 5, 2008, July 3, 2008, and September 22, 2008.

Subsquently, in the NRC's Request for Additional Information (RAI) dated October 20, 2009, the NRC indicated that based on the information provided, significant changes to the TS may have resulted that could invalidate or render moot the original justification for the proposed changes.

Therefore, the NRC is requesting that Exelon completely revise Section 2.0, "Proposed Changes,"and submit a new Section 2.0 which includes a clear technical justification for each proposed change.

Accordingly, the information provided below contains a new Section 2.0, "ProposedChanges,"

for the proposed LAR. This Section 2.0 supersedes the information provided in Section 2.0 of Exelon's November 2, 2007, submittal. The new Section 2.0 describes each of the proposed changes along with the supporting justification.

2.0 Proposed Changes The proposed changes to the Oyster Creek TS are being requested to allow certain Secondary Containment operability requirements to be modified during fuel handling operations when handling fuel that has not occupied part of a critical reactor core within the previous 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months .

The proposed changes have been evaluated and are supported by reanalysis of the radiological consequences of a Fuel Handling Accident (FHA) utilizing AST methodology previously reviewed and approved by NRC for use at Oyster Creek in TS Amendment No. 262, dated April 26, 2007. The movement of sufficiently decayed irradiated fuel is consistent with TSTF-51, Revision 2 to NUREG-1433, Volume 1, Revision 2, "StandardTechnical Specifications -

General Electric Plants."

Existing Secondary Containment integrity requirements specified in TS 3.5.B will remain applicable when fuel handling operations involve fuel that has occupied part of a critical reactor core within the previous 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months .

2.1 TS Definition 1.52, RECENTLY IRRADIATED FUEL, is being added to specify when the existing TS provisions for Secondary Containment integrity remain applicable. The proposed TS definition specifies a minimum decay time of 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months before Secondary Containment requirements can be altered.

2.2 TS Section 3.5.B.1 is being revised to include an exception for maintaining Secondary Containment integrity at all times. This exception is intended to permit movement of irradiated fuel during refueling operations without Secondary Containment integrity when the reactor has been subcritical for greater than 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months provided certain provisions are met. These provisions include: 1) only certain evaluated release locations can be opened, 2) one Standby Gas Treatment System (SGTS) circuit shall be available, and 3) in the event of fuel handling accident the available SGTS circuit shall be running within one hour and all Secondary Containment openings must be closed within one hour.

This will ensure that, in the event of a radiological release from a fuel handling accident, Page 2 of 4 air flow is directed such that it can be treated and monitored. This proposed change has been evaluated and is supported by reanalysis of the radiological consequences of a FHA utilizing AST methodology.

2.3 TS Section 3.5.B.1 .c and d are revised to allow the handling of irradiated fuel without Secondary Containment integrity when the reactor has been subcritical for greater than 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months . Secondary Containment integrity is required when handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months ), or when work is being performed on the reactor or its connected systems in the reactor building which has the potential to drain the reactor vessel.

Secondary Containment is not required with the reactor vessel head or drywell head in place, as allowed by current TS.

TS Section 3.5.B.1 .c is being revised to read: "No work is being performed on the reactoror its connected systems in the reactorbuilding which has the potential to drain the reactorvesseL" This replaces the existing requirement that: "The reactorvessel head or the drywell head are in place," which is being relocated to Section 3.5.B.1 .d.

TS Section 3.5.B.1 .d is being revised to stipulate that one of the following must be met:

1) "The reactorvessel head or the drywell head are in place," or 2) The reactorhas been subcriticalfor greaterthan 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months ." This replaces the existing requirement that: "No work is being performed on the reactoror its connected systems in the reactorbuilding which could result in inadvertent releasesof radioactivemateriaL."

These proposed changes are considered conservative and consistent with TSTF-51, Revision 2.

2.4 TS Section 3.5.B.1 .e is being revised. The existing wording: "No operationsare being performed in, above, or around the spent fuel pool or reactorcavity that could cause release of radioactivematerials,"will be revised to restrict movement of heavy loads over the reactor cavity or spent fuel pool when Secondary Containment is not available.

The proposed changes will still limit the type of activities that can be performed in, above, or around the spent fuel pool or reactor cavity that could cause release of radioactive materials.

2.5 TS Section 3.5.B.4.a action statement for loss of SECONDARY CONTAINMENT INTEGRITY or inoperable secondary containment isolation valves (SCIVs) during power operation is being revised to include a restriction on the movement of heavy loads in, above, or around the spent fuel pool.

2.6 TS Section 3.5.B.4.b.(1) action statement for loss of SECONDARY CONTAINMENT INTEGRITY or inoperable secondary containment isolation valves (SCIVs) during refueling is being revised to include a restriction on the movement of RECENTLY IRRADIATED FUEL. The original request to delete the restriction on activities that could reduce the shutdown margin has been re-evaluated and is no longer being considered.

This specific requirement will be retained in the TS.

2.7 TS Section 3.5.B.4.b.(2) action statement for loss of SECONDARY CONTAINMENT INTEGRITY or inoperable secondary containment isolation valves (SCIVs) during refueling is being revised to read as follows: "Ceaseall work on the reactoror its connected systems in the reactorbuilding which could result in potentialto drain the Page 3 of 4 reactor vessel." The existing phrase "...inadvertentreleases of radioactivematerials" was replaced by the phrase ". ..potential to drain the reactorvessel." This proposed change is considered conservative and consistent with TSTF-51, Revision 2.

2.8 TS Section 3.5.B.4.b.(3) action statement for loss of SECONDARY CONTAINMENT INTEGRITY or inoperable secondary containment isolation valves (SCIVs) during refueling is being revised to include a restriction on the movement of heavy loads in, above, or around reactor cavity or the spent fuel pool.

2.9 TS Section 3.5.B.5 is a new action statement for loss of SECONDARY CONTAINMENT INTEGRITY or inoperable secondary containment isolation valves (SCIVs) during refueling that has been added to control the movement of irradiated fuel without Secondary Containment integrity if the reactor has been subcritical for greater than 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months under certain conditions. These conditions include: 1) only certain evaluated release locations are permitted to be open, 2) one SGTS circuit shall be available, and

3) in the event of a fuel handling accident the available SGTS circuit shall be running within one hour and all Secondary Containment openings must be closed within one hour. This will ensure that, in the event of a radiological release from a fuel handling accident, air flow is directed such that it can be treated and monitored. This proposed change has been evaluated and is supported by reanalysis of the radiological consequences of a FHA utilizing AST methodology. These new requirements were also discussed in 2.2 above.

2.10 TS Section 3.5.B.6.b.(3), which has been renumbered as 3.5.B.7.b.(3), is revised to specify that during refueling the availability requirements for the SGTS apply when: 1) fuel handling operations involve RECENTLY IRRADIATED FUEL, 2) when performing operations with the potential to drain the reactor vessel, or 3) for movement of heavy loads in, above, or around the reactor cavity or spent fuel pool. The original request to delete the restriction on activities that could reduce the shutdown margin has been re-evaluated and is no longer being considered. This specific requirement will be retained in the TS.

2.11 TS Section 3.5.B.7.c, which is being renumbered from 3.5.B.6 to 3.5.B.7, is a new section that was added related to SGTS operability requirements during refueling operations. This new section would permit SGTS operability requirements to be altered when moving irradiated fuel provided the reactor has been subcritical for greater than 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months and all release locations from the Secondary Containment can be closed within one hour.

2.12 TS Section 3.5.B.9 is a new section that was added related to Secondary Containment requirements specifying that in the event of a fuel handling accident, SGTS must be running within one hour and all Secondary Containment openings must be closed within one hour following the fuel handling accident.

2.13 Table 3.5-1 is a new table that has been added for TS Section 3.5.B.5 describing release locations that are acceptable to be opened once the reactor has been subcritical for greater than 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months . There is one exception identified in the table (i.e., Item 15 -

Northwest RB Personnel Airlock to Office Building at Elev. 51' 3") and this penetration must remain closed. The release locations identified in the table are the only release pathways evaluated and acceptable to be opened during movement of irradiated fuel.

Other potential release pathways may be opened provided that any potential release Page 4 of 4 from the pathway will be bounded by a pathway previously evaluated as identified in the table.

2.14 TS Section 3.17 C and D, are being revised to specify that operability requirements for the Control Room Heating, Ventilating, and Air-Conditioning (HVAC) System during refueling apply when handling RECENTLY IRRADIATED FUEL, and to substitute the wording "...the potential to drain the reactor vessel ...." in lieu of "...inadvertent release of radioactive material...," consistent with TSTF-51, Revision 2. In addition, these sections are being revised to include restrictions on the movement of heavy loads in, above, or around the reactor cavity or spent fuel pool.

2.15 TS Bases Sections 3.5 and 4.5 are revised to incorporate the basis for the proposed changes, including the basis for the term RECENTLY IRRADIATED FUEL assuming a minimum decay time of 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months , the need for maintaining SGTS operability or availability, the permitted release locations, and to incorporate the defense-in-depth guidelines contained in TSTF-51, Revision 2.

ATTACHMENT 3 Revised Technical Specifications Page Mark-ups Affected Pages 1.0-9 3.5-5 3.5-6 3.5-7 3.5-7a (new page) 3.5-7b (new page) 3.5-7c (new page) 3.17-1

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.

1.51 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heatup and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.23.

1.52~ RECENTLY IRRADIATED FUEL RECENTLY IRRADIATED FUEL is fuel that has~occupied part of a critical reactor core~

within the previous 168 hours0.00194 days 0.0467 hours 2.777778e-4 weeks 6.3924e-5 months . This 168-hour perio6d may be reduced to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if all secondar ontaihnmentopenings are closed and t~he Standby Gas Treatmen~t System is OPERABLE.

OYSTER CREEK 1.0-9 Amendment No. 266, 2-69, XXX

8. Shock Suppressors (Snubbers)

a. All safety related snubbers are required to be operable whenever the systems they protect are required to be operable except as noted in 3.5.A.8.b and c below.

b. With one or more snubbers inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber(s) to operable status.

c. If the requirements of 3.5.A.8.a and 3.5.A.8.b cannot be met, declare the protected system inoperable and follow the appropriate action statement for that system.

d. An engineering evaluation shall be performed to determine if the components protected by the snubber(s) were adversely affected by the inopera-bility of the snubber prior to returning the system to operable status.

B. Secondary Containment

1. Secondary containment integrity shall be maintained at all times unless all of the following conditions are met except as specified in Specification 3`5.bi5:

a. The reactor is subcritical and Specification 3.2.A is met.

b. The reactor is in the cold shutdown condition.

c. N*'*Er iTbik per d bh its +connected-systems in the reactor buildifh n1 which*nasi teotenalto dr reacto

d. Nq:~jbjgp(rfoiifid_6R thc rdqctr~r or) if

&6e,`-octgd &y-;stems 4tnhc Feactor build ng w-hich-NO u FntiS o4 ibR .,t ods'Ft4 ldabc of radior ntact o 6644i4Qnrieof~the followihng-arle met 1)Th~e reactor ve~ssel head ý',r the dir-yweled are ini place.,

2) The reactor has beenssbcriticdallfr rete th an 1.6 8 hou rs.

e. No movement of heavy loads in, above, or.aro.und th.e

.. .rea..cto,.r-....

cavity or spent fubi, pool.

e7, Nbpemtie are being performed in,abo'&_'oG*r a&eupd the SpcRit fueltrg oe htol cause,,Feefasc ef ad'acti've rhaterialc OYSTER CREEK 3.5-5 Amendment No.: 32,74,86, 87,100, XXX

2. Upon the accidental loss of SECONDARY CONTAINMENT INTEGRITY, restore, SECONDARY CONTAINMENT INTEGRITY within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , except as provided in specification 3.5.B.3.

3. With one or more of the automatic secondary containment isolation valves inoperable:

a. Maintain at least one automatic secondary containment isolation valve in each affected penetration OPERABLE.

b. Within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months restore the inoperable automatic secondary containment isolation valve(s) to OPERABLE status or isolate each affected penetration with at least one valve secured in the closed position.

4. If Specifications 3.5.B.2 or 3.5.B.3 cannot be met:

a. During Power Operation:

(1) Have the reactor mode switch in the shutdown mode position within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

(2) Cease all work on the reactor or its connected systems in the reactor building which could result in inadvertent releases of radioactive materials.

(3) Cease all operations in, above or around the Spent Fuel Storage Pool that could cause release of radioactive materials.

(4) eeasei movemrentof heavy2 loads in, abo,**'

or around the&slpent fuel ~p~ol.

b. During refueling:

(1) Cea~se fuel handling opýerations involving RECENTLY IRRADIATED FUEL or. activities Which~ co~uld reduce the shuitdown margin.

(excluding react~or coolant temp~erature changes)..

(2) Cease all work on the reactor or its connected systems in the reactor build ng which could result in in5d.&ld"tr.la*c f ?adi*ati?6

ý -,-.vessel.

(3) Ceas fe loads in, abov.e, o.ra"r, -und the'rieciorjcavity-br spent fueli Q aI;IqeplcrpqAtib in,abovcI r areund&

001c~c theSpnt uc Stra ool thatýý couldj qau4pe

",777diaticmaciae OYSTER CREEK 3.5-6 Amendment No.: 14,18,32,74,103, 168, XXX

5 ,D~urnn rfuLel 6fuel-is: permitted nihgibvdmeit of'jirraddatd without 'secondary con~tainmeint integrity if th~e reactor has~

been, IubcriCt#icalV for..qI reter~than-468 hours wh-en:

I" 'I"'atinslised" in Tble3"'

pro , 09".1W0.si ustMecosed excetimon' ftWrunniigagwaitI neihoureleanealaseco r contaiment

6. Two separate and independent standby gas treatment I system circuits shall be operable when secondary containment is required except as specified by Specification I

3.5.1B.67.

The release locations'listed in Table 3.5-1 areIthe only releaseolocations that have been evaluated and -are,acceptable, to be opened during movement of irradiated fuel, with~ one noted exception (item14 - Northwest RB `6rso6nnelAi;rlock .to OtOCfefBuIldinig at-.elev.,51 '-3". Other potential release pathways may be opeped',pr9,v,,_ _n,,oehaii~ea0,rj th aha will be bounded by a pathway previousl vlae sietfe nt~tbe (2,One standby: g asýtreatmentcircuit shall býe.runnngwthJ* one h~ur to direct flow in.the roper-direction in the event of a.rad ioogical release. from fuel4 handling accident such that, it can: be treatieid*ndm* onit-re"id...

OYSTER CREEK 3.5-7 Amendment No.: 167, 168, 211,233, XXX

67. With one standby gas treatment system circuit inoperable:

a. During Power Operation:

(1) Verify the operability of the other standby gas treatment system circuit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . If testing is required to demonstrate operability and significant painting, fire, or chemical release has taken place in the reactor building within the previous 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , then demonstration by testing shall take place within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the expiration of the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, and (2) Continue to verify the operability of the standby gas treatment system circuit once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months until the inoperable standby gas treatment circuit is returned to operable status.

(3) Restore the inoperable standbygas treatment circuit to operable status within 7 days.

b. During Refueling:

(1) Verify the operability. of the other standby gas treatment system within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . If testing is required to demonstrate operability and significant painting, fire, or chemical release has taken place in the reactor building within the previous 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , then demonstration by testing shall take place within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of the expiration of the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, and (2) Continue to verify the operability of the redundant standby gas treatment system once per 7 days until the inoperable system is returned to operable status.

(3) Restore the inoperable standby gas treatment system to operable status within 30 days or cease all~~~~~~~

fuol'i- a!ri hadlng cocatrtocoperati641 6ol hdu6tn h~d'v argin (oXcuin rthOP G99t 44mpratwifrcchanges.of the following:-

rcqiGtorcol`ii (a)','Activ~ities that could reduce shutdown6 mar'okgih '(excLidinrgreator coplaiit tembperat -urechanges).

R ECENTLY'IRR DIATED FUEL.'

(c): Opera.tion IiiI§w~ith ýthe .potentiafl 'to drjain theýý reactoo.-vesse!*

NY AMoemet hevIy in, aov, or around~hjatrcvt or spent-fuel, pool.v (0e:' Open-in-g'ppnetr~a~tions in accordance With,,

oopn.ý hal beccdsed OYSTER,'CREEK OYSTER CREEK 3.5-la

C. Standby gas treatment system availability.c.an-,eiodifi, d durn , movementot =r,radiated* fuel' whihen (1)>. Thne reactor has been,,' a I- r-e

.1.8houiand ,; u~tclorrae hn Al *release locatiohnslrom tesecondarW containment cans be d!;osqded 78 If Specifications 3.5.B.&6 and 3.5.1B3. are not met, reactor shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the condition of Specification 3.5.B.1 shall be met.

Sin t he e'v'ent of "afuelh idi ing accide-n't; one -st-a6d&y`-g~

treatment circuit *must be*running within one*hour, and all sec*ndary containment openings shall be *losed within onehourý following the fuieihandling accident.(1, One standby gas treatment circuit shall be capable of running within one hour to direct flow in the proper.ddirection' inthe event of 4radiological releasefrom a fuel handling accident'suc ithat itcan,

,kd rnon-itoired.

be treated an OYSTRCEK> 3.5-7b

Table 3.5-1 n,; !p'Re aeý "cai KmThe*,releaselocationsJ 6iementofirradiate are"the only release6 1sted'iTable'13.65i loaton-s th'at have be-enevaluatedand are"acceptable*to fue6,'withonenoýted exception(ltemn14'N- t1hw estRB6Pers nnen *I~cktoffi~eBudingtEev5V3)Other' pened1durd g*

e uated-asldentifiedinithe~tableo yltitheDrwel qiAcbtc'ý ac~t(tm? aiov qpnP idea- e:Eait -Rp ilc

'nbe Afer theA 6-our-deayp-idL-jgreui 6eatbi'hereqfuireentohvte ailck cIe canberlxd

( At any*i*me t he-access pointS'to the c itern 1tabore) canbe pen proyidedthe ,. sothweRB.t RBIersoneb* e"l oksed-(MACFacili!ty, Pers~o nnel. Airlock)iscclosed. $;After~thei68.*h ourdecay~period dunng* ref ueh ng!*operatIns,* the* req uiir, e en~t::t'avetht dairlock*

relaxed. 0,6loseied N-~b __~~'W~,-6!

(M kC-4ilitý R.flo7L

3.17 Control Room Heating, Ventilating, and Air-Conditioning System Applicability: Applies to the operability of the control room heating, ventilating, and air conditioning (HVAC) system and Control Room Envelope (CRE) boundary.

NOTE -----------------------------------------------------

The CRE boundary may be opened intermittently under administrative control.

Obiective: To assure the capability of the control room HVAC system and CRE boundary to minimize the amount of radioactivity, hazardous chemicals, or smoke from entering the control room in the event of an accident.

Specifications:

A. The control room HVAC system shall be operable during all modes of plant operation.

B. With one control room HVAC system determined inoperable for reasons other than specification D:

1. Verify once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the partial recirculation mode of operation for the operable system, or place the operable system in the partial recirculation mode; and

2. Restore the inoperable system within 7 days, or prepare and submit a special report to the Commission in lieu of any other report required by Section 6.9, within the next 14 days, outlining the action taken, the cause of the inoperability and the plans/schedule for restoring the HVAC system to operable status.

C. With both control room HVAC systems determined inoperable for reasons other than specification D:

1. During Power Operation: place the reactor in the cold shutdown condition within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months

2. During Refueling:

(a) Cease RIECENTLY *I4RADIEDATE UEb. L e e'w handling operations; and S1/2 teall wor;k;on hnereactororfs-connecteosystems n*tne. reactor.

bujilding which could resulWin' iddna pe.44440ease- of ~raalio-a-tiuo, atrilth e, potential to drain the reactor vessqell.; and (s)*:>Cease movement of heay0adn ocat spent fuzel pool.

D. When one or both control room HVAC systems are determined inoperable due to an inoperable CRE boundary:

1. During Power Operation: actions to implement mitigating actions shall be performed immediately, verification that the mitigating actions are in place shall be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and the CRE boundary shall be restored to operable status within 90 days.

2. D"rn 'Refu 6e 1i-hig'ivmn

~ o difd ~igiil&inttimn~~irr (a) Immediately suspend movement of ... 4,.,RECENTLY IRRADIATED).FUEL assemblies in the containment; and (b)K. . 4rImimediateyinitiateiaction to suspend; operat6ons with .th potential to'drainý th.e reator vessel;;'-and (6)K Immiediately suspend move'me'nt of heavy l'oýa'd-s in,aboveor aroUndJtIe reactoir, cavity..ndsent fuel, pool.

OYSTER CREEK 3.17-1 Amendment No.: 115, 139, 167, 211, 225, 246, XXX

ATTACHMENT 4 Revised Technical Specifications Bases Page Mark-ups 3.5-8 3.5-11 3.5-11 a (new page) 3.5-11 b (new page) 3.5-11 c (new page) 3.5-1 id (new page) 3.5-12 3.5-12a 3.5-12b (new page) 4.5-13

Bases:

Specifications are placed on the operating status of the containment systems to assure their availability to control the release of any radioactive materials from irradiated fuel in the event of an accident condition. The primary containment system(1 ) provides a barrier against uncontrolled release of fission products to the environs in the event of a break in the reactor coolant systems.

Whenever the reactor coolant water temperature is above 212'F, failure of the reactor coolant system would cause rapid expulsion of the coolant from the reactor with an associated pressure rise in the primary containment. Primary containment is required, therefore, to contain the thermal energy of the expelled coolant and fission products which could be released from any fuel failures resulting from the accident. If the reactor coolant is not above 212 0F, there would be no pressure rise in the containment. In addition, the coolant cannot be expelled at a rate which could cause fuel failure to occur before the core spray system restores cooling to the core. Primary containment is not needed while performing low power physics tests since procedures and the Rod Worth Minimizer would limit rod worth such that a rod drop would not result in any fuel damage. In addition, in the unlikely event that an excursion did occur, the reactor building and standby gas treatment system, which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below 40 R'-4 100 app lic0*able 1FR limits.

The absorption chamber water volume provides the heat sink for the reactor coolant system energy released following the loss-of-coolant accident. The core spray pumps and containment spray pumps are located in the comer rooms and due to their proximity to the torus, the ambient temperature in those rooms could rise during the design basis accident. Calculations(7) made, assuming an initial torus water temperature of 100°F and a minimum water volume of 82,000 ft3 , indicate that the comer room ambient temperature would not exceed the core spray and containment spray pump motor operating temperature limits and, therefore, would not adversely affect the long-term core cooling capability. The maximum water volume limit allows for an operating range without significantly affecting accident analyses with respect to free air volume in the absorption chamber. For example, the containment capability(8) with a maximum water volume of 92,000 ft. 3 is reduced by not more than 5.5% metal-water reaction below the capability with 82,000 ft. 3 .

Experimental data indicate that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160°F during any period of relief valve operation with sonic conditions at the discharge exit. Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.

OYSTER CREEK 3.5-8 Amendment No.: 76, 1-96, XXX

Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads. It is, therefore, required that all snubbers required to protect the primary coolant system or any other safety system or component be OPERABLE whenever the systems they protect are required to be OPERABLE.

The purpose of an engineering evaluation is to determine if the components protected by the snubber were adversely affected by the inoperability of the snubber. This ensures that the protected component remains capable of meeting the designed service. A documented visual inspection will usually be sufficient to determine system OPERABILITY.

Because snubber protection is required only during low probability events, a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed for repairs or replacements.

Secondary containment 5 ) is designed to minimize any ground level release of radioactive materials which might result from a serious accident. The reactor building provides secondary containment during reactor operation when the drywell is sealed and in service and provides primary containment when the reactor is shutdown and the drywell is open, as during refueling.

Because the secondary containment is an integral part of the overall containment system, it is required at all times that primary containment is required. Moreover, secondary containment is required during fuel handling operations inJivii'n~igRE"CENIR.ADIand"'

wtheneyer pserfcr"n ed.on rectedtwosbeing systemtsh,'t he reactorJ uinl6 i~l~ciui.

r~u tjii h~potential W., a,~IJ n el.liraet&

~ ~ tm buiedi i 68asoesae5ove pentotir B!a'ses .Taibl,bI- 5-1 open ý(v Airl-ock,toý Office.BJ uidi ing, prIo:vided, one sit'and~by ga When secondary containment is not maintained, the additional restrictions on operation and maintenance give assurance that the probability of inadvertent releases of radioactive material will be minimized. Maintenance will not be performed on systems which connect to the reactor vessel lower than the top of the active fuel unless the system is isolated by at least one locked closed isolation valve.

OYSTER CREEK 3.5-11 Amendment No.: 18,74,75,100,196,21 1, XXX ECR OC 04-00842 Corrected: 12/24/84

Bases Table 3.5-1*'-

EV a,luated Release Location's This is a tem porarys'tuctu re ;used§as the. Drywell SupporCentear nd.

Outage om'nand s nte Itconnectedt the east Reactor, Building personn elaaj?1ck through a Iser ies oiem porarytunnelsý These tunnels arenot~safety'related, Thoweverq'sirnce they are:*

connected o theR B"ersonne air lock' any arm.dischargedthrough theair""l*ockcouldb.fedd through the tunnelbng-discharged at-,.,,

th e~doo rs to the temp' a-,,faclity' Tese doors:are closer"to theCR' intakestha eroneair

.hpBIockitsef There are four,.(-4)ido 66o1ia'ted with'th6D.s a I

4ýdorsaýare medirectionhrelative to theCR

`ocatedmirouqtgh

'itakes:, The' kdifo[ the oa

.tte n noeth'wnalI would be bounded, by the door:onthe'west wall whichsq osertCR' intakes A,and B.'

Sieilarlyl..thle tWo'doors~on the southe6ifwallW facing the; RBB6would bound th 'Zhdta~f&the south eastern-'mb'st`

door would beboun'deýbyl he§south jwesterpi-.mostý door*,wlich isŽ closer to CRintakes-. A* andai .B.

2 CG ommodities"ppenetrationron, Yes This is aIflan g'ed; '. 6t ti"'ý4.fbr;*Nhich :*pr currently in' RB SouthtWall Elev23'-6" P!a efor routing born oditiesthroughlthe penetration without violatinga the,....o pf.. co ry.tair ent .oburndaay.However. for1the purposeoof" determining fu thelmaxim~um s flCth ... dose~to *C* personnel,*thispenetrationis

- I* I , ,, s considered to~be durnggnconditions with-movemen of>

fuel withfffs uffiCnie ,deoa'yr inprogress7 3 Com'm"oditiespe6netration ~on Ys Thsi flndpnertnowicpoeuesa urnlyn RB Nrth Wapll dlv236 r- eniainwtotvoai the seonaIry:Icntain~met bunary.,n 'However, Itor the~urpose -of determnining the'maxim uim dose to CR: personnel, this~pene~trationis2 considered to be ope'n during shutdown .conditiqqs witmovemqy engfc fue[,with'Ssufticint decay inprogress.,

35-1* a

Bases- .Table'&3.5-1, Release Locations 6vlutd 4oditispenetration onh Y is isaflangedieetratino.foowhich'.procedureae cur*entl ir'!

RB:East'Wall Elev 23'-6, placelfor routing com m oditi*s thrgh h1e'petration without violatig; the econdaryco.ntainmrnert bo'undaiy. H~'eerfotep ~ . O oeo determining theremaximum dose to6C_-R p.ersonhne ,-this-penetra:ionJ._is considered to beop.en during shýUtdown "cbondtionswithim ovement of6 fuel!with *su fficient cia in pos f estir;* tion~is-a'6jact*to4 the RB'eastpe*sosnnel[airock and-is considere'boundedd bytl the, ailrlock (which is signihcantly s tused: inthe dose__anlySS_.n 'Terefoe__a__s 3eclftTere.fors h1i otalculatte"..

5 3Boof atch ~Y ~Th-'roof hatch is'a-ýpersonne accss,,;t~6`the not olperneclroutinelyjt it* *nsiderelt-rd 6 R rof Alithough"'

otent alI"release locaton nos op`n"'d8"t' in the event that itsen, eng uteown*con itons*w movener of fuelwithsufficient decay in progress&*

6 Stac'k Tunnel[D6o Y& This isteacs o.cto'ocran ST'capes'adýeup'n t*6 the'eastFf theIeacor6 tBuildir . ]gsitsmade eda's*aipoite*ntial release loainioeedrn udw:6hiins ih oeet of fuel with suffiicent ca'y'nprogr, s.*

EsThast D' RB Airlo~kDp

~~Ts~'flck door is ~connectetd',toth idyiary~tu~nnp leadih to~the DrywellrAcce~ss ,Facility. Hjweer, sic hstmbrr aiiyi not safety related,: leakage from this loaini otltd uig movement of fuel with sufficient decay inppordess.

8 South East RB AilocK Poor Y

es Since *e south.east RB Aiock door. s in thesam. relative'direction as thesouith RB com modities penetratio *and "farther -away,*it ist considereld bounaded by tIrhe south RBIcom ooaIteIspe Theirefor&ea specific X/Qwas not calculated 9 Ratr; BUildihg Tru~ck~ e Sinc ~th~e RB'.Truck'Airlock do si h sa ae-eaiddrcind 4rokfairther awayfromn other 'calculated ope ins' ki en aini considered bounded byjthe south 4R Brrýcom ties,,teS e re ion* andA the;'R* oof. aatch. ,therefore,, aspecif ic X);was, notcalculated.

10 Iso6ltion Cd&niideriients, Yes Since the;se vents (on the RB east. wall))ari h'-the ,'§a.negeneral /

directi astli eRBBroof hatch (and farpneraway) wýithrespect to thbe, CR intakes" a>elease from tlhis iocation istconsidrede',boUndeeoby the RB'roof Hatch. Thereforeaseii iIQ titcal 8uliated,,

1,T MVAC FaciliftyEhtre esThesedk uble, door aremoedas iglrt ti 3.5-r lb

Bases Table 3:5-.11 Evalfuate dhReea's-e,.

i Lo'c`atio0n's (MVACt'acility'-Prsonne, Y&§ The MAC Fa~cility 'Pers~onnel

ýirlockjý ~~~~~~protection-area locate'dn't'he n~orth'B "'ý73-%,#ý) TfiJthe,.

FersonneLAilockatd Elev locationwith tlhe'maximum X/Q forTthedosýeoanalysis'and`resuts iin th ainim, dosefor"all'peneýýtrati'on~s all'owed to'bepedriý,

s'htdown d conditions with 'movement;ofifuebwith sufficentdecay,,

i ooo ose e..rn un........innR o mm i......

sas se ..fs o.. a c.. n . nm ta . d ouse..

. .entan ss.......

Tuib'ine Buildin eý utý Access__ _totlis__omis intemrm ite~opemnii u

tof the door unIderadm.

droe permitted_(via__S)_dungtoperaheontr miiIstratve"'contr'ols.ý-i ugnfi Nort '§et RBPe~s 666 1 No9 The6Northwes't, RBPersonnel Airlock leads from theR&Bto the*!Offi&e Airlock to Offi6p'iqil'ding ,atý Bq~li n e51V" '3 el~e~'vIatI ion (Colu mns RF and .IR 16).1 ,,Is, losure is' Ellem. 51'-3" credited in'the analysis~of thee FHA. The iupep&containment'personnel[

air loc6kperforms no active function *,,'responseto'the postulated-accident;hhwever, its leak*tightnness is required to ensure that fit release' of radioactive materials from primary containm ent is. restricted to those leakage. pathssassumed in the accident-analysis,ýand the fission products released bythe FHA Awill bet*t eated by the SGT System.I It was not originallyonIthe Ist of penetrations 'allowed to be, oIpen. Since t i ilc pn noteOffice. Building ,(closeJto `thee R entrance) it sNOpermtted* topbe opn doo a6 its< proximity.oý

, cdue to onOX this o the,ý;CRkV.% .*OntakessandihefCX.access.ad clseness, a,,s becifc-X/Qwas not :calcula-te6d,-.

15 Southwiest RB.11Psonhel Yes Thisi 'airl66k~leads from thesou** stRBBtdi'theTdrbine.

AirlocktoiBat..Elev +-~.*675" ildin the'3'-6' elevation (ColunimnsJ.andclR6)'..týasnot-,

onginallyontthe"list of penetrations allowed to beopen.'

Flowever'&rit is:assum ed to be open du r*sn hutdown ncondiios witl) movement of fuewith sufficient decayin progbressiiiSince!

th~is.dlocationýis in the'.sam e, general'directtoin as the Trunnrioný',

Room .dofartheawa*the X/Q is considered boundledd yVte Tru'nnion Roo idoor. Therefore, a pecific X/Q~was not,

__________calculatedý.

16 Northwest` RB Pe'rsonrnel YeO Thiý airlock-leads fd rmthe northwest RB to theTurlbine. B3uildih`g dh Airlocks to TI3 at Elev. -1'41" the 3 -6"elevatrion(ColumnsRRand R6). It wa's 'not originally on the, list of ppenetrations'allowed to be*olen'. Howevert isPas.'sumed to be open during shutdownconrditionS ,with movementoffuel With sufficient, decay in progress. Since this location is',in the~same general, direction' I asihe MAC Facility persomnnlairlo c and.Pk isfarthe ,raway, ',the, X/Q is considered bou~ndedj by th M~AC7 Fa ility personneI'airlock. Therefore,

_____________________________ isecific, X,/Q :was. -notcalculated.

3.5-lic

Basis Table 315-11 E ui~ated Rel seLocations 17 Fld6r?* g'-t6, RBCom *rn, es Since thisSloatisnn iittih samegen ,,eral--as" t ihhib

,pqrm 4ne>2~6 frilither a(yj~he X/Q is oýrdb somd&&d udd.bt 7 Ti'urid;5km i2 dii dll1r Therefore,, a ispeci [/iX/O was;;not~calculatea ontomroil,.ý 18 Floor PlUgto4N RB Corner Yss Since thislocat'iohis in the same generaldýcjection ashie, MAT.acility.

Roon*Ele'vb.23ý61 personne lai rock{aindis farther away, the *Q codieie dbou ded2.

66si by thet91MAC, Faacilite personnel airlock-T:he ef ores tefciwa 1noci Yes A lle d.ti' and located apprx..........

41,-6..

______Penetration, __________he___Reactor___ Building The release locat"1nshisted in Bases:Table 3.5-1 are theonii'release 6ocations'that'have been evaluated anid are accidptabie'to! be.opnened during mov~ement~of irradiiatilu'e~iwth the noted exceptnio(Item1.4 -.-Norhwest RB Personnel Airlock toOff ice, Buildirg atElev.' 517-3 theepotent e.enedipli d[that anyhPotentiIalIrelease fromihe, pathway will be u*ed'b ,y pathway previlousyi evaluated ',de ntified.inthe.table.

¶,h* At any time, theaccesspoints tohe DryweWAccess Facdlityi(tem l above)can be open providedd the East RB A0rlock, (Itemin7 above)'s closed. After thel 68-"hourd aperiod, d ring refuein eratis, hrq~irfet to hav~e th~ilckcoe can be.11add'4' (3) At any ti me, the..accesspoints~td~t hthe" "Facility EntranceA'(It~rii11iaboVe)`an` b eop, nrovided thefeqthvN iSti RB ' tlil 23"- "MA acity .ersionrelF R oc i)s4closedq-After te16 1h, o upst he eqte airloc clse the can _p Jeaxed With?5,tgar~ tDitsArrea in ,_.oefWing,.v-l yatin ispovded ,

D6ifftsreia) ac.!r.nt!rc'cl4akaQg~eie S no. speeu* -`et--! q-eq i s'9 the- t ....

DifuseAre).best chrceizda_ diffuse area surc As --this is a model igrmethod, used to identfy-:a'l" a lIn e 66, toit T d p~e.etrationsas.sociated.. withthis modlelini Since only thgemetalh4 sidiigportion'of the east wall *fthe,*B hasthe potentialfor, lalagetH'e~surface`.areais mltedin 1thi~ssection of thewall. ,Furthermmore,.*since complete.m, imxng n'et refueligara'vblumeqcannot be.

assumed, only -50% ffthe`Jmetal siding areaiAs ,assumed i `thede*terminationt Iof ttlediffuse.area',X/Q-This limited area is assumed to be at the-wprst case lcatiion witi re! ationship to ejtIherjQR,'ntake locatolnto maximize the calculated dose.

35id

Plant procedures'idi rect, intecondary conta mnt,Jtegity beibmataned~wlhen.

handing~heavyA adsq(greaterp.than one, fueassemby)such as *threactdr vessel head6r dryer/separhtý?r)'lh,, above orar undtereactor cavity w'ithzfuel'inth&e react"or vessel; top pwdeadditinial~protectionn.'Plantprpcedures als dfirect*hthat seconary ontaii ieit~ ntegrity .bemaintained when' handling heavy loads in, above or around th~e spent f iej~pk, The trunnion room door is not an access opening for the passage of personnel and equipment into the reactor building. During all modes of operation, the trunnion room is a low traffic area and momentar openings of the door would be limited and administratively controlled and have little effect on SGTS and HVAC.

The standby gas treatment system(6) filters and exhausts the reactor building atmosphere to the stack during secondary containment isolation conditions, with a minimum release of radioactive materials from the reactor building to the environs. Due to radioactive.decay, dunng fueI>.

handi ng 90eratins Icontrolssaer these e .lyorequIred-iwhen handingq(RECENTLYI IRRADIATED FUEL; Eor;during operations"with"thepotentialto the rea tor vese "w

pooWs heh~ri.... g~ .l i;V!orar~un&

= the reactorcvitr s 4e InSection 3.5.B.5 and 3.5.B.6 of the Technical Specification, the use of the word "Circuits" actually means 'Trains" as the word trains is used in the following paragraph.

Two separate filter trains are provided, each having 100% capacity( . There is a section of ductwork upstream and downstream that is common to both filter trains. If one filter train becomes inoperable, there is no immediate threat to secondary containment and reactor operation may continue while repairs are being made. Since the test interval for this system is one month (Specification 4.5), the time out-of-service allowance of 7 days is based on considerations presented in the Bases in Specification 3.2 for a one-out-of-two system.

There is also only one vital power supply to the SGTS automatic initiation controls and for the operation of the heating coils for both filter trains.

Therefore, the SGTS is not mechanically nor electrically single failure proof. However, manual actuation of the SGTS is not vulnerable to single failures and is an acceptable backup to automatic initiation.

Two automatic secondary containment isolation valves are installed in each reactor building ventilation system supply and exhaust duct penetration. Both isolation valves for each supply duct penetration are located inside the secondary containment boundary, and the two exhaust duct penetration isolation valves are located outside of the secondary containment boundary. Removal of an inboard supply or exhaust valve (closest to the boundary) is permitted only when secondary containment is not required. The outboard isolation supply or exhaust valve can be removed when secondary containment is required as long as the inboard valve is secured in the closed position. Due to radioactive decay, durin64-fuel ~h'anjdling operations the sedo"nrdar'donitafinment isolation valves are only requred' to be OPERABLE when, handling REENTLY IRRADIATED FUEL; during o teraions'with the6potentill to drain the reactor vessel;orwhen handlinheavy loadsji above, or !ound the reactor 6it.rspentfuel poiol.

OYSTER CREEK 3.5-12 Amendment No.: 14,1,,79,86,97,168,196,230, XXX ECR OC 04-00842 Corrected: 12/24/84

The a~dditio'n'ofhtl'term'RFitCEN'1TLY~ IRRADIATED F:UEL associated ~ihhnln raitdfe inl'econdaryjcpntainmert function. Technica Spcfction requireenss 40lcal sice hademon6rstrated- that. after 'sufficient radioactivedea has'il 6cUIrre .d;,ofsien a halsis dontrlro eraiatoroses.'resultinig from afe a~i~acdnrmi eovteii~o T~olb~igd~lnes,'arkeih'clu ed inte assessmendtbloVytm eoe fro svice durFing move nto'fi'R'adiated fuel-'

661in,ilhAndlinglco re altera~tions, ventilaio'n system andrdainmntr ava iI bIit4~ep W6NMARIC 91 -06) shbuld'b6 _6siessedd,'itfi_ r~esp toi filttlonand onitring4 of relieases fromffuel F6Io ýrLsht'down;'adioactivity in'i

.thifje i'd"dea'y"s',aEway-fair rapidly. The basis of the Techn(cal S~ec-Ific"tlom, operbiltwa~identis the reduction in doses rdue to suclidca 'Te~jalof ni~it~liin veitiatin-' sstem and radiation 'moh itor availabiliyi ordc oe

'even, further below, that provided by the natural decay.

Asingle nofirmal or contingency method to promptly close secondary containment penetratlbns s houI d be, develo ped for tho'se t imes when sec'oi dary conta inime!n't igý noqired.,Such prompt nmetho~ds need not-completely blrck the 6netrt~ rb

.cpble of "esistinigpressure. The puJrpose of~th'e "prompt methods'!'m~rtioned:e above are to enable ventilation systems to draw the release fro'm a postulI ated fuel' handling accident in the proper direction such that it cari-be treaIted and mntrd

.Inh addition'to the above, ln the event of a fuel handling accident, one standby gas treatment circuit shall'be running within one hour and 'all secondary containment openings must, be closed 'within-one hour., These actions will enable ventilation' systems to draw the release, fromri, a otlted'fuel handling accident in the proper direction suh that it canbbetreated ,and monitored as discussed above.

Ai'fel hiandling aci@de~nt is'an event that could result in the release ofsinictqutteso fission. produc rts ri'thl acidental dropping ofaq mentý (incluingý fuel budlsnto irradiatefue in&i the reactorqcore or spentful

References:

(1) FDSAR, Volume I, Section V-1 (2) FDSAR, Volume I, Section V-1.4.1 (3) FDSAR, Volume I, Section V-1.7 (4) Licensing Application, Amendment 11, Question 111-25 (5) FDSAR, Volume I, Section V-2 (6) FDSAR, Volume I, Section V-2.4 (7) Licensing Application, Amendment 42 (8) Licensing Application, Amendment 32, Question 3 (9) Robbins, C. H., "Tests on a Full Scale 1/48 Segment of the Humboldt Bay Pressure Suppression Containment," GEAP-3596, November 17, 1960.

(10) Bodega Bay Preliminary Hazards Summary Report, Appendix I, Docket 50-205, December 28, 1962.

(11) Report H. R. Erickson, Bergen-Paterson To K. R. Goller, NRC, October 7, 1974.

Subject:

Hydraulic Shock Sway Arrestors.

(12) General Electric NEDO-22155 "Generation and Mitigation of Combustible Gas Mixtures in Inerted BWR Mark I Containment" June 1982.

(13) Oyster Creek Nuclear Generating Station, Mark I Containment Long-Term Program, Plant Unique Analysis Report, Suppression Chamber and Vent System, MPR-733; August, 1982.

OYSTER CREEK 3.5-12a Amendment No.: !4,19,79,86,97,168,196,230, XXX ECR OC 04-00842 Corrected: 12/24/84

(14) Oyster Creek Nuclear Generating Station, Mark I Containment Long-Term Program, Plant Unique Analysis Report, Torus Attached Piping, MPR-734; August, 1982.

(15) AmerGen Calculation C-1302-243-E170-087, "Wetwell-to-Drywell Vacuum Breaker Sizing."

(16) General Electric NEDE-24802, "Mark I Containment Program Mark I Wetwell-to-Drywell Vacuum Breaker Functional Requirements, Task 9.4.3," April, 1980.

(14), Technic~lSpecifati'o11hTask Foice (T!STF) Iorloved'Stan dard~

Techinical Specif iations* 'TravemlerTSTF-!515-ARev. 2 (NUMAR'91-, Gdidelihnesr lndust:Actinsto Assess Shutdow n Mangement" 6*StERkCREk 1-5712b

During 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 during 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 eftesll ,"-fu

,and*ing,11i6aininvolvingn REC1*ENTLY IRAITED FUEL associated with the outage. Verifying the efficiency and operation of charcoal filters once 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]. Ifthe allowable penetration is <5%, the required removal efficiency is t95%. 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, XXX

RA-09-083, Response to Request for Additional Information, Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling

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RA-09-083, Response to Request for Additional Information, Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling

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