Text

Request for Changes to Technical Specifications Refueling Operations - Relaxation of Requirements Applicable During Movement of Irradiated Fuel
	ML022200450
Person / Time
Site:	Salem
Issue date:	07/29/2002
From:	Garchow D; Public Service Enterprise Group
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LCR S02-010, LR-N02-0244
	Download: ML022200450 (52)
	v • d • e

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, New Jersey 08038-0236 JUL 2 9 2002 LR-N02-0244 LCR S02-01 0 0 sud U. S. Nuclear Regulatory Commission \uclear LI. C ATTN: Document Control Desk Washington, DC 20555-0001 REQUEST FOR CHANGES TO TECHNICAL SPECIFICATIONS REFUELING OPERATIONS - RELAXATION OF REQUIREMENTS APPLICABLE DURING MOVEMENT OF IRRADIATED FUEL SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 Gentlemen:

Pursuant to 10 CFR 50.90, PSEG Nuclear LLC (PSEG) hereby requests a revision to Appendix A of the Technical Specifications for the Salem Nuclear Generating Station, Units 1 and 2. In accordance with 10CFR50.91 (b)(1), a copy of this submittal has been sent to the State of New Jersey.

The purpose of this License Amendment Request is to provide flexibility in scheduling outage tasks and to modify unnecessarily restrictive containment closure and Fuel Handling Area Ventilation requirements. TSTF-51, Revision 2 is used as guidance in the preparation of this change request. This request revises the requirements for containment closure associated with the equipment hatch and personnel airlocks during Core Alterations and movement of irradiated fuel within the containment. This proposed change would allow the equipment hatch and the personnel airlocks to remain open during fuel movement in the containment provided administrative controls are developed and implemented, ensuring the closure of the equipment hatch and personnel airlock following a fuel handling accident within the containment building.

Appropriate Bases changes are included to reflect the proposed changes.

The basis for the proposed changes is the reanalysis of the limiting design basis Fuel Handling Accident (FHA) using the guidelines contained in 10CFR 50.67 and Regulatory Guide 1.183, Alternative Source Term. The proposed changes are based on NRC approval of selective implementation of Alternative Source Term methodology for the Salem Units 1 & 2 Fuel Handling Accident submitted by PSEG, letter LR-N02-0231 dated June 28, 2002. The Salem Units 1 & 2 UFSAR will be updated to reflect the enclosed amendment and analysis following NRC approval.

PSEG has evaluated the proposed changes in accordance with 10CFR50.91(a)(1), using the criteria in 10CFR50.92(c), and has determined this request involves no significant hazards considerations. An evaluation of the requested changes is provided in Attachment 1 to this letter. The marked up Technical Specification pages affected by the proposed changes are provided in Attachment 2.

PSEG requests approval of the proposed License Amendment by September 12, 2002 to be implemented within 30 days. This will support the start of Salem Nuclear Generating Station Unit 1, fifteenth (1R15) refueling outage that is scheduled for October 12, 2002.

ýoo1 95-2168 REV. 7/99

LR-N02-0244 LCR S02-010 JUL 2 9 20OZ Below is a summary of the commitments made in this submittal:

1. Develop and implement administrative controls to ensure that containment closure is accomplished within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following a Fuel Handling Accident within Containment.

Should you have any questions regarding this request, please contact Mr. Brian Thomas at 856 339-2022.

I declare under penalty of perjury that the foregoing is true and correct.

Sincerely, Executed on-i O'*

D. F.Prsrchow Vice Prsident - Operations Attachments (2)

C Mr. H. J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 U. S. Nuclear Regulatory Commission ATTN: Mr. R. Fretz, Project Manager - Salem Mail Stop 08B1 Washington, DC 20555-0001 USNRC Senior Resident Inspector - Salem (X24)

Mr. K. Tosch, Manager IV Bureau of Nuclear Engineering PO Box 415 Trenton, NJ 08625 2

Document Control Desk LR-N02-0244 LCR S02-010 SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 FACILITY OPERATING LICENSES DPR-70 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 EVALUATION OF REVISIONS TO THE TECHNICAL SPECIFICATIONS REFUELING OPERATIONS - RELAXATION OF REQUIREMENTS APPLICABLE DURING MOVEMENT OF IRRADIATED FUEL

Document Control Desk LR-N02-0244 LCR S02-010 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS RELAXATION OF REQUIREMENTS APPLICABLE DURING MOVEMENT OF IRRADIATED FUEL Table of Contents

1. D ES C R IPT IO N ................................................................................................................................. 1

2. P R O PO S E D C HA NG E ..................................................................................................................... 1

3. B A C KG R O UND ................................................................................................................................ 2

4. TECHNICAL ANALYSIS .......................................................................................................... 3

5. REGULATORY SAFETY ANALYSIS ......................................................................................... 6 5.1 No Significant Hazards Consideration ............................................................................ 6 5.2 Applicable Regulatory Requirements/Criteria ................................................................ 9

6. ENVIRONMENTAL CONSIDERATION .................................................................................... 10

7. R IS K S IG NIFICA NCE ..................................................................................................................... 10

8. R E F E R ENC ES ............................................................................................................................... 11

Document Control Desk LR-N02-0244 LCR S02-010 REQUEST FOR CHANGE TO TECHNICAL SPECIFICATIONS RELAXATION OF REQUIREMENTS APPLICABLE DURING MOVEMENT OF IRRADIATED FUEL

1. DESCRIPTION This letter is a request to amend Operating Licenses 50-272 and 50-311 for Salem Units 1 & 2 respectively.

The purpose of this License Amendment Request is to provide flexibility in scheduling outage tasks and to modify unnecessarily restrictive containment closure and Fuel Handling Area Ventilation System requirements. TSTF-51, Revision 2 is used as guidance.

The proposed amendment would allow movement of sufficiently decayed irradiated fuel within the containment building with the equipment hatch, personnel air locks and containment penetrations open. Operation of the Containment Purge Exhaust System (CPES) is not required during movement of sufficiently decayed fuel provided that the Auxiliary Building Ventilation System is in operation and taking suction from the containment atmosphere via the open containment airlocks.

The automatic isolation of the Containment Purge System, resulting from the containment radiation monitors, is being relaxed in order to maintain monitoring of post-Fuel Handling Accident (FHA) containment activity until the closure of the containment penetrations is accomplished.

Eliminating the automatic actuation allows the containment purge and exhaust system to remain in operation during refueling when the equipment hatch is open. It also assures that negative pressure in containment is maintained by either the containment purge system or the auxiliary building ventilation system with suction from the containment atmosphere via the open personnel airlocks. This action complies with the requirements listed in General Design Criteria (GDC 64),

which requires monitoring the reactor containment atmosphere following postulated accidents.

Manual capability for isolation of the containment purge system, if required, would remain unchanged. The amendment also would allow movement of irradiated fuel assemblies within the Fuel Handling Building with the Fuel Handling Area Ventilation System (FHAVS) in operation with no credit taken for filtration.

2. PROPOSED CHANGE The proposed changes to the Technical Specifications would revise the following sections:

a. Technical Specification Section Definition 1.8, previously not used, is being used for the revised definition of CORE ALTERATIONS, previously Definition 1.9. The proposed definition is identical to the Standard Technical Specifications, NUREG 1431 and better reflects the assumptions used for the (FHA) as described in UFSAR Section 15.4.6. The analyzed event is the drop of a fuel assembly and the subsequent rupture of the cladding of all the fuel rods in the assembly. Other off-normal events described during CORE ALTERATIONS are bounded by the FHA.

b. The current Definition 1.9a, Core Operating Limits Report, is relocated to Definition 1.9, previously used by the CORE ALTERATIONS definition. This is considered an administrative change and no relaxation of current requirements is proposed.

c. Deletion of the requirements for automatic isolation of the Containment Purge System during fuel movement within containment. TS Table 3.3-6, Item 2.a.l.a and 2.a.2.a are modified allowing the continuous monitoring of the containment atmosphere until containment closure is accomplished following a FHA. The containment ventilation provided by either the Containment Purge System or the Auxiliary Building Ventilation System (with personnel airlocks open) maintains outside airflow into containment and minimizes potential releases out of the open equipment hatch. The Auxiliary Building Ventilation System exhausts via the plant vent which, has installed radiation monitors providing for the continuous monitoring of Fuel Handling Building and containment atmosphere following a FHA with containment hatch and personnel airlocks open.

1

Document Control Desk LR-N02-0244 LCR S02-010 The applicable surveillances of Table 4.3-3, items 2.a.l.a and 2.a.2.a are modified to reflect the relaxations of automatic isolation of containment purge to allow for continuous monitoring of containment atmosphere following the FHA. In addition to the deletion of the Limited Condition for Operation (LCO) and Surveillance Requirement (SR) for the air particulate activity radiation monitor for containment purge & pressure vacuum relief isolation in Mode 6, this change also corrects an oversight during the submittal and approval of Amendment 79 (Unit 1) and Amendment 53 (Unit 2). Amendment 79/53 was issued on April 10, 1987 to revise accident monitoring instrumentation and radiation monitoring instrumentation requirements. As part of this change, the requirements to have the containment air particulate monitor (R1i1A) operable in Modes 1, 2, 3, and 4 was deleted from TS Table 3.3-6 Item 2.a.2.a requiring the R1i1A only to be operable in Mode 6 for purge & pressure vacuum relief isolation. Although the mode applicability was revised for LCO Table 3.3-6 to eliminate the Mode 1-4 requirement, an oversight was made in the request in that the corresponding surveillance requirement TS Table 4.3-3 Item 2.a.2.a was not revised to reflect the above change as well. Currently there is a mis-match between the surveillance table and the LCO table. This proposed TS change is correcting this oversight by deleting the requirement to perform the surveillance requirement for TS Table 4.3-3 item 2.a.2.a in modes 1-4.

d. The proposed change would revise Limiting Condition for Operation 3.9.4, Containment Building Penetrations, to allow the containment equipment hatch and the personnel airlocks to be open during movement of irradiated fuel assemblies within containment, provided they are capable of being closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months under administrative controls. A new surveillance Requirement (SR) would be added to verify the capability to install the containment equipment hatch within the required time, if the hatch is open, prior to the start of irradiated fuel movement within the containment building.

e. The Bases for this TS are being revised to define the capability to close containment following a FHA and describe the administrative controls required to comply with LCO implementation.

f. Deletion of Core Alterations from the Applicability and Actions in TS 3/4.7.6, Control Room Emergency Air Conditioning System. The analyzed event is the drop of a fuel assembly and the subsequent rupture of the cladding of all the fuel rods in the assembly.

Other off-normal events described during CORE ALTERATIONS are bounded by the FHA.

f. Relocating TS 3/4 3.9.9 Refueling Operations, Containment Purge and Pressure-Vacuum Relief Isolation System to be combined with TS 3.9.4, Containment Building Penetrations. The purpose of this relocation is two-fold; 1) Implementing consistency with the Improved Technical Specifications and, 2) expanding and clarifying the requirements for the Containment Purge System during movement of irradiated fuel assemblies within the containment.

g. Deletion of the SFP Filtration System and it's associated surveillances from TS 3/4 3.9.12 Fuel Handling Area Ventilation. In addition, Action a. is being revised to delete the requirement for suspending crane operation with loads over the storage pool. This action is not required since the FHA considers the drop of a fuel assembly to be the limiting scenario. The Control of Heavy Loads Program contains restrictions that do not allow movement of loads exceeding 2200 pounds (weight of a fuel assembly and associated handling device) over the spent fuel pool. Additionally, TS 3/4 9.7 prohibits travel of loads in excess 2200 pounds over fuel assemblies in the storage pool.

The marked up Technical Specification pages are included in Attachment 2.

2

Document Control Desk LR-N02-0244 LCR S02-010

3. BACKGROUND In December 1999, the NRC issued a new regulation, 10 CFR 50.67, which provides a means for power reactors to replace their existing accident source term with Alternative Source Term (AST).

Regulatory Guide 1.183 provides guidance for the implementation of alternate source terms. 10 CFR 50.67 requires licensees seeking to use AST to apply for a license amendment and include an evaluation of the consequences of the affected design-basis accidents. PSEG submitted a License Change Request via letter LR-N02-0231, dated June 28, 2002, which addresses these requirements by proposing the selective application of AST described in RG 1.183 in evaluating the radiological consequences of a FHA. As part of the implementation of the AST, the TEDE acceptance criterion of 10 CFR 50.67 (b)(2) replaces the previous whole body and thyroid dose guidelines of 10CFR 100.11 and 10 CFR 50, Appendix A, GDC 19 for the FHA. This LCR follows in part the guidance provided by TSTF-51, rev 2.

The proposed amendment is similar to amendments issued by the NRC (Reference 8m) allowing containment closure relaxations during fuel movement by applying the dose guidelines of 10 CFR 50.67 and Regulatory Guide 1.183, Alternative Source Term.

4. TECHNICAL ANALYSIS Each containment at Salem Nuclear Station is equipped with two personnel air locks, an equipment hatch, and a Containment Purge System. Technical Specification (TS) 3.9.4 requires that during Core Alterations or movement of irradiated fuel assemblies within containment, the associated equipment hatch be closed and secured with at least four bolts and at least one of the two doors in each personnel air lock be closed. TS 3.9.4 also requires that each penetration providing direct access from the containment atmosphere to the environment either be closed by an isolation valve, blind flange, or equivalent. The applicable design basis event is the Fuel Handling Accident inside containment. During Core Alterations or movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a fuel handling accident, involving dropping of a spent fuel assembly resulting in the rupture of the cladding of all the fuel rods in the assembly. In the re-analysis of this design basis event, airborne activity resulting from the initiating event (FHA) is assumed to be released to the environment over a 2-hour time period via the open equipment hatch and the plant vent not taking credit for filtration.

Following reactor shutdown, decay of the short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. Following sufficient decay time, the primary success path for mitigating the FHA no longer includes the functioning of the active containment systems.

Therefore, water level and decay time are the primary success paths for mitigating a FHA.

CORE ALTERATIONS, as presently defined in the Technical Specifications, is the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. As described in TSTF-51, Revision 2, accidents postulated to occur during core alterations include inadvertent criticality, fuel handling accident, and the loading of a fuel assembly or control component in an incorrect location. Generically, it was concluded that of these off normal occurrences, only the fuel handling accident results in cladding damage and potential radiological release. Consequently, it is being proposed that the definition of CORE ALTERATIONS be modified as shown in the TS mark-up. This change is consistent with the Improved Technical Specifications (ITS).

DOSE CALCULATIONS Calculations were performed to determine atmospheric dispersion factors (,/Qs) at the Salem Nuclear Generating Station (SNGS) control room (CR) air intake due to the FHA releases from the Containment Equipment Hatch and Plant Vent (PV).

Analyses were performed to determine the Exclusion Area Boundary (EAB), Low Population Zone (LPZ) and Control Room (CR) doses due to a FHA occurring in the containment building with containment equipment hatch and Personnel Locks opened. The FHA analysis was performed using the AST, the guidance in the Regulatory Guide 1.183, Appendix B, and TEDE 3

Document Control Desk LR-N02-0244 Attachment I LCR S02-010 dose criteria. The results of these calculations are within the regulatory acceptance criteria and are summarized below.

FUEL HANDLING ACCIDENT ANALYSIS DOSE RESULTS CONTROL ROOM EAB LPZ Results TEDE TEDE TEDE (rem) (rem) (rem)

Salem FHA Containment in 2.93E+00 4.15E+00 5.94E-02 Building Salem FHA in Fuel Handling 1.90E+00 4.15E+00 5.93E-02 Building Regulatory Acceptance 5.OOE+00 6.30E+00 6.30E+00 Criteria FUEL HANDLING BUILDING The Fuel Building of each nuclear unit at Salem Units 1 and 2 Nuclear Station is equipped with a Fuel Handling Area Ventilation System (FHAVS) with two exhaust and one supply fans. TS 3/4 9.12 presently requires that the FHAVS be operable and in operation while irradiated fuel is in the storage pool. There is one germane design basis event, the Fuel Handling Accident in the Fuel Building.

For movement of irradiated fuel, within the Fuel Handling Building, operation of the FHAVS ventilation is credited. TS 3/4 9.12 requires that the FHAVS be operable. If the FHAVS is inoperable, movement of irradiated fuel assemblies within the Fuel Building is to be immediately suspended. The LCO applicability is being modified to reflect the conditions under which a FHA could occur. The dose calculations performed, using AST, result in acceptable doses without crediting the FHAVS Filtration System thus; the Surveillances Requirements associated with the Filtration System are deleted. The Filtration System is not required for operability of the FHVAS.

Revised analyses of the FHA have been performed in support of this License Change Request.

The analyses were performed pursuant to Regulatory Guide 1.183 using the Alternative Source Term (AST) methodology. The source terms were calculated pursuant to R.G. 1.183. Releases from the reactor cavity or spent fuel pool were modeled pursuant to R.G. 1.183. The results were included as part of the selective application of Alternative Source Term applicable to the FHA, in the PSEG submittal LR-N02-0231, dated June 28, 2002.

ADMINISTRATIVE CONTROLS NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", provides guidance to reduce resulting doses and to avoid unmonitored releases.

This document recommends the development of contingency methods to promptly close the containment penetrations following Fuel Handling Accidents to enable ventilation systems to draw the release from a postulated Fuel Handling Accident in the proper direction such that it can be treated and monitored.

Even though containment closure is not credited in the dose calculations, PSEG proposes to develop administrative controls to provide reasonable assurance that containment closure, as a defense-in-depth measure, can be reestablished promptly following a Fuel Handling Accident to limit the releases much lower than assumed in the dose calculations. It is not necessary that the specific actions contained in these administrative controls be contained in the TS Bases. Plant procedures and changes thereto fall under the requirements of 10 CFR 50.59. The proposed administrative controls are shown below:

4

Document Control Desk LR-N02-0244 LCR S02-010 Containment Building Closure:

The following requirements shall be maintained to ensure defense-in-depth. Closure Controls are in effect whenever the affected Containment is open during operations within containment involving movement of irradiated fuel assemblies. The definition of an open containment penetration is a penetration that provides direct access from the containment atmosphere to the outside environment.

1. The equipment necessary to implement containment closure shall be appropriately staged prior to maintaining any containment penetration open including airlock doors and the containment equipment hatch.

2. Hoses and cables running through any open penetration, airlock, or equipment hatch should be configured to facilitate rapid removal in the event that containment closure is required.

3. The containment personnel airlock may be open provided the following conditions exist:

a. One door in each airlock is capable of being closed.

b. Hoses and cables running through the airlock shall employ a means to allow safe, quick disconnection or severance.

c. The airlock door is not blocked in such a way that it cannot be expeditiously closed.

Protective covers used to protect the seals/airlock doors or devices to keep the door open/supported do not violate this provision.

d. Personnel are designated and available with the responsibility for expeditious closure (within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) of at least one door on the containment airlocks following the FHA.

4. The containment equipment hatch may be open provided the following conditions exist:

a. The containment equipment hatch is capable of being closed or an equivalent closure device is available and can be closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

b. Hoses and cables running through the equipment hatch shall employ a means to allow safe, quick disconnection or severance.

c. The equipment hatch is not blocked in such a way that it cannot be expeditiously closed. Protective covers used to protect the seals/equipment hatch or devices to keep the hatch open supported do not violate this provision.

d. Necessary tools to install the equipment hatch and tighten at least four equipment hatch closure bolts are available or other methods to close the equipment hatch opening (i.e.,

restrict air flow out of the containment), such as a refueling hatch closure device, is staged at the work area along with the necessary installation tools.

e. A sufficient number of personnel are designated and available with the responsibility for expeditious closure (within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) of the containment equipment hatch opening following the FHA.

5. If containment closure would be hampered by an outage activity, compensatory actions will be developed.

6. Either the Containment Purge system or the Auxiliary Building Ventilation System with suction from the containment atmosphere, with associated radiation monitoring will be available whenever movement of irradiated fuel is in progress in the containment building and the equipment hatch is open. If for any reason, this ventilation requirement can not be met, movement of fuel assemblies within the containment building shall be discontinued until the flow path(s) can be reestablished, or close the equipment hatch (or an equivalent closure device is installed) and personnel airlocks. Periodic verification (once per shift) of this administrative control will ensure that air flow will be directed from containment to the Auxiliary Building or the Plant Vent where continuous monitoring will be in effect thus minimizing the potential for unmonitored releases out the open containment hatch following the FHA.

5

Document Control Desk LR-N02-0244 LCR S02-010

7. Personnel responsible for Containment Building Closure shall be trained and knowledgeable in using the procedure for executing containment closure. Walkdowns should be considered to demonstrate the closure capability including compensatory actions in the event of loss of electrical power.

Fuel Handling Building Closure:

The following requirements shall be maintained to ensure defense-in-depth. Closure Controls are in effect during operations within the Fuel Handling Building involving movement of irradiated fuel assemblies.

1. The Fuel Handling Building doors shall be maintained closed except for normal entry and exit unless a designated person is available to close the open door(s) should a FHA occur within the Fuel Handling Building.

2. The FHAVS, with associated radiation release monitoring will be available for the release flow path. If for any reason operation of the fuel handling area ventilation system flow path must be discontinued and the fuel building is open to the outside environment, fuel movement within the Fuel Handling Building shall be discontinued until the flow path can be reestablished, or until the openings to the outside environment are closed.

3. If the Fuel Handling Building closure would be hampered by an outage activity, compensatory actions will be developed.

Control Room Emergency Air Conditioning System (CREACS)

During movement of irradiated fuel assemblies, both CREACS normal outside air intakes should normally be open. If one intake is closed, movement of irradiated fuel assemblies will be suspended until the intake is reopened. These controls are governed by existing action requirements under Technical Specification 3.7.6.1. The actuation of CREACS during a FHA is performed by the radiation monitors located in the normal outside air intakes. Exceeding the setpoints of these radiation monitors will cause dampers to reposition to isolate the normal ventilation system from the Control Room Envelope, start the CREACS fans and open the appropriate outside emergency air intake. The radiation monitors in the Control Room normal outside intake are required to be OPERABLE during movement of irradiated fuel assemblies as governed by TS Table 3.3-6.

5. REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration PSEG Nuclear LLC (PSEG) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment" as discussed below:

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

Response: No.

An alternate source term calculation has been performed for Salem Nuclear Station that demonstrates that offsite and control room dose consequences of a postulated fuel handling accident remain within the limits provided sufficient decay has occurred prior to the movement of irradiated fuel without taking credit for certain mitigation features such as ventilation filter systems and containment closure. Fuel movement is allowed provided that irradiated fuel has undergone the required decay time.

The proposed amendment would allow movement of sufficiently decayed irradiated fuel within the containment building with the equipment hatch and personnel air locks open provided that administrative controls are implemented to promptly (within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) close the containment penetrations.

6

Document Control Desk LR-N02-0244 LCR S02-010 Either the Containment Purge system or the Auxiliary Building Ventilation System with suction from the containment atmosphere, with associated radiation monitoring will be available whenever movement of irradiated fuel is in progress in the containment building and the equipment hatch is open. If for any reason, this ventilation requirement can not be met, movement of fuel assemblies within the containment building shall be discontinued until the flow path(s) can be reestablished or close the equipment hatch and personnel airlocks. The amendment also would allow movement of irradiated fuel assemblies within the Fuel Handling Building with the Fuel Handling Area Ventilation System (FHAVS) in operation but no credit taken for filtration.

This amendment does not alter the methodology of the FHA or equipment used directly in fuel handling operations. Neither ventilation filter systems, the CPES nor the FHAVS, is used to actually handle fuel. Therefore neither of these systems is an "accident initiator".

Similarly, neither the equipment hatch, the personnel air locks, nor any other containment penetration, nor any component thereof is an accident initiator.

In the postulated Fuel Handling Accident, the revised dose calculations, performed using 10 CFR 50.67 and Regulatory Guide 1.183, Alternative Source Term, do not take credit for automatic containment purge isolation thus allowing for continuous monitoring of containment activity until containment closure is achieved. If required, containment purge isolation can be initiated manually from the control room.

Actual fuel handling operations are not affected by the proposed changes. Therefore, the probability of a Fuel Handling Accident is not affected with the proposed amendment. No other accident initiator is affected by the proposed changes.

The FHA in the Fuel Handling Building has been analyzed without credit for filtration by the FHAVS. The analyses of these design basis events were conducted with the Alternative Source Term Methodology in accordance with 10 CFR 50.67 and Regulatory Guide 1.183. These analyses show that the resultant radiation doses are within the limits specified in these documents.

The TEDE radiation doses from the analyses supporting this LCR have been compared to equivalent TEDE radiation doses estimated with the guidelines of R.G. 1.183. The new values are shown to be within the regulatory guidelines.

The revision to the definition of Core Alterations simply reflects the definition in the Standard Technical Specifications, NUREG 1431 for Westinghouse Plants and is supported by the bounding effects of the Fuel Handling Accident analysis.

The deletion of Core Alterations from the APPLICABILITY section of the affected LCO's is based on the fact that, during Core Alterations only the FHA results in cladding damage and potential radiological release. Consequently, the deletion of Core Alterations is consistent with industry approved practice and guidance documents (ex:

TSTF-51, revision 2).

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

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

Response: No.

The proposed change does not involve addition or modification to any plant system, structure, or component. The proposed amendment would permit the equipment hatch and personnel air locks to be open during movement of irradiated fuel. The proposed amendment does not involve any change in the operation of these containment 7

Document Control Desk LR-N02-0244 LCR S02-010 penetrations. Having these penetrations open does not create the possibility of a new accident.

The proposed amendment also would remove the requirements for operability of the FHAVS Filtration System during movement of sufficiently decayed irradiated fuel. It does not alter the operation of these systems. Therefore, the system is not an accident initiator. Modification of the requirements of operability for the system from the plant Technical Specifications does not create the possibility of a new accident.

The revision to the definition of Core Alterations simply reflects the industry position supported by the definition in the Standard Technical Specifications, NUREG 1431 for Westinghouse Plants and is supported by the bounding effects of the Fuel Handling Accident analysis.

The deletion of Core Alterations from the APPLICABILITY section of the affected LCO's is based on the fact that, during Core Alterations only the FHA results in cladding damage and potential radiological release. Consequently, the deletion of Core Alterations is consistent with industry approved practice and guidance documents (ex:

TSTF-51, revision 2).

The proposed amendment does not create the possibility of a new or different kind of accident than any previously evaluated.

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

Response: No.

The assumptions and input used in the analysis are conservative as noted below. The design basis Fuel Handling Accidents have been defined to identify conservative conditions. The source term and radioactivity releases have been calculated pursuant to Regulatory Guide 1.183 and with conservative assumptions concerning prior reactor operation. The control room atmospheric dispersion factors have been calculated with conservative assumptions associated with the release. The conservative assumptions and input noted above ensure that the radiation doses cited in this License Change Request are the upper bound to radiological consequences of a Fuel Handling Accident either in Containment or the Fuel Handling Building. The analyses show that there is a significant margin between the TEDE radiation doses calculated for the postulated Fuel Handling Accident using the Alternative Source Term and the acceptance limits of 10 CFR 50.67 and Regulatory Guide 1.183.

The revision to the definition of Core Alterations simply reflects the industry position supported by the definition in the Standard Technical Specifications, NUREG 1431 for Westinghouse Plants and is supported by the bounding effects of the Fuel Handling Accident analysis.

The deletion of Core Alterations from the APPLICABILITY section of the affected LCO's is based on the fact that, during Core Alterations only the FHA results in cladding damage and potential radiological release. Consequently, the deletion of Core Alterations is consistent with industry approved practice and guidance documents (ex:

TSTF-51, revision 2).

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PSEG concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

8

Document Control Desk LR-N02-0244 LCR S02-010 5.2 Applicable Regulatory Requirements/Criteria NRC Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors".

The NRC's traditional methods for calculating the radiological consequences of design basis accidents are described in a series of regulatory guides and Standard Review Plan (SRP) chapters. That guidance was developed to be consistent with the TID-14844 source term and the whole body and thyroid dose guidelines stated in 10 CFR 100.11.

Many of those analysis assumptions and methods are inconsistent with the ASTs and with the total effective dose equivalent (TEDE) criteria provided in 10 CFR 50.67. This guide provides assumptions and methods that are acceptable to the NRC staff for performing design basis radiological analyses using an AST. This guidance supersedes corresponding radiological analysis assumptions provided in other regulatory documents when used in conjunction with an approved AST and the TEDE criteria provided in 10 CFR 50.67.

PSEG used this regulatory guide extensively in the preparation of this "selective implementation". This application and the supporting analyses comply with this guidance as it applies to a Fuel Handling Accident.

Title 10, Code of Federal Regulations, Part 50 Section 67, "Accident Source Term".

10 CFR 50.67 permits licensees to voluntarily revise the accident source term used in design basis radiological consequences analyses. This document is part of a 10 CFR 50.90 license amendment application and evaluates the consequences of a design basis fuel handling accident as previously described in the Salem UFSAR.

Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors".

RG 1.183 supersedes corresponding radiological assumptions provided in other regulatory guides and standard review plan chapters when used in conjunction with an approved alternative source term and the TEDE provided in 10 CFR 50.67.

10 CFR 100, "Determination of Exclusion Area, Low Population Zone and Population Center Distance".

10 CFR 100.11 provides criteria for evaluating the radiological aspects of reactor sites. A footnote to 10 CFR 100.11 states that the fission product release assumed in these evaluations should be based on a major accident involving substantial meltdown of the core with subsequent release of appreciable quantities of fission products. A similar footnote appears in 10 CFR 50.67. In accordance with the provisions of 10 CFR 50.67(a), PSEG applied the dose reference values in 10 CFR 50.67 (b) (2) in the analyses in lieu of 10 CFR 100 for the Fuel Handling Accident.

NUREG-0800, Standard Review Plan, Section 15.7.4, "Radiological Consequences of Fuel Handling Accidents".

The SRP Section 15.7.4 describes the radiological effects of a postulated Fuel Handling Accident. The SRP does not directly refer to the guidance of RG 1.183 or 10 CFR 50.67.

Instead, it refers to regulatory documents, which are superseded by the selective application of the Alternative Source Term for the FHA.

10 CFR 50 Appendix A, General Design Criteria 19, Control Room PSEG has applied the guidelines provided by 10 CFR 50.67 and RG 1.183, which supersede the current requirements of GDC 19 for the Fuel Handling Accident.

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

9

Document Control Desk LR-N02-0244 LCR S02-010

6. ENVIRONMENTAL CONSIDERATION ENVIRONMENTAL ASSESSMENT/IMPACT STATEMENT Pursuant to 10 CFR 51.22(b), an evaluation of this license change request has been performed to determine whether or not it meets the criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) of the regulations.

Implementation of this amendment will have no adverse impact upon the Salem units; neither will it contribute to any significant additional quantity or type of effluent being available for adverse environmental impact or personnel exposure. The change does not introduce any new effluents or significantly increase the quantities of existing effluents. As such, the change cannot significantly affect the types or amounts of any effluents that may be released offsite. The new consequences of the revised Fuel Handling Accident analysis remain well below the acceptance criteria specified in 10 CFR 50.67 and Regulatory Guide 1.183.

It has been determined there is:

1. No significant hazards consideration,

2. No significant change in the types, or significant increase in the amounts, of any effluents that may be released offsite, and

3. No significant increase in individual or cumulative occupational radiation exposures involved.

Therefore, this amendment to the Salem TS meets the criteria of 10 CFR 51.22(c)(9) for categorical exclusion from an environmental impact statement.

7. RISK SIGNIFICANCE Based on the results of the conservative dose calculations provided in the previous submittal (LR N02-0231, dated June 28, 2002) and to support this submittal, the risk to the health and safety of the public as a result of a Fuel Handling Accident inside the containment with the equipment hatch open is minimal. Actual fuel handling accidents, which have occurred in the past, have resulted in minimal or no releases, which supports that the assumptions and methodology utilized in the radiological dose calculations, in accordance with Regulatory Guide 1.183, are very conservative. Other safety factors supporting the minimal risk significance involved with this request are: 1) Fuel Decay Time prior to start of fuel movement, 2) The water level (reactor cavity and spent fuel pool) that covers the fuel assemblies, 3) administrative controls for containment closure, and 4) ventilation and radiation monitoring availability.

In summary, based on the considerations discussed above:

a. There is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner,

b. Such activities will be conducted in compliance with the Commission's regulations, and

c. The issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

10

Document Control Desk LR-N02-0244 LCR S02-010 REFERENCES

a. PSEG Submittal to the NRC, LCR S02-03 (Decay Time), dated June 28, 2002.

b. UFSAR Section 9.4.3.1, Fuel Handling Area Ventilation.

c. UFSAR Section 15.4.6, Fuel Handling Accident.

d. PSEG Calculation S-C-ZZ-MDC-1920, Revision 1, Fuel Handling Accidents Occurring in Fuel Handling Building and Containment - AST Analysis for Relaxation of Containment Integrity.

e. PSEG Calculation S-C-ZZ-MDC-1912, Revision 0, CR x/Qs Using ARCON96 Code- Equipment Hatch, Plant Vent Releases.

f. NUMARC 93-01, Revision 3, " Industry Guidelines for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants", July 2000.

g. NUREG-0800, Standard Review Plan, Section 15.7.4, Rev.1, July 1981.

h. USNRC Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

i. 10 CFR 50.67, "Accident Source Term"

j. Salem Units I and 2 Technical Specifications.

k. USNRC Safety Guide 25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for BWRS and PWRS".

I. NUREG 1431, Standard Technical Specifications for Westinghouse PWRS, TSTF-51, Rev 2.

m. NRC Issued License Amendments involving Containment Closure Relaxations During Fuel Movement:

Turkey Point, SER Issued September 27, 2001

Brunswick, SER Issued March 14, 2002

Shearon Harris, SER issued July 30, 2001

Watts Bar, SER issued January 22, 2002

TMI Unit 1, SER issued October 2, 2001

Fort Calhoun, SER issued March 26,2002 11

Document Control Desk LR-N02-0244 LCR S02-010 SALEM NUCLEAR GENERATING STATION UNITS I AND 2 FACILITY OPERATING LICENSES DPR-76 AND DPR-75 DOCKET NOS. 50-272 AND 50-311 REFUELING OPERATIONS - RELAXATION OF REQUIREMENTS APPLICABLE DURING MOVEMENT OF IRRADIATED FUEL TECHNICAL SPECIFICATION PAGES WITH PROPOSED CHANGES The following Technical Specifications for Facility Operating License No. DPR-70 are affected by this change request:

Technical Specification Paqe 1.8 1-2 1.9 3/4.3.3.1 3/4 3-36 3/4 3-37 3/4 3-38 3/4.7.6.1 3/4 7-18 3/4 7-19 3/4.9.4 3/4 9-4 3/4.9.9 3/4 9-9 3/4.9.12 3/4 9-12 3/4 9-13 3/4 9-14 B 3/4.3.3.1 B 3/4 3-1a B 3/4 3-2 B 3/4.7.7 B 3/4 7-5c B 3/4.9.4 B 3/4 9-1 B 314 9-2 B 3/4 9-3 B 3/4.9.9 B 314 9-4 B 3/4.9.12 The following Technical Specifications for Facility Operating License No. DPR-75 are affected by this change request:

Technical Specification Page 1.8 1-2 1.9 3/4.3.3.1 3/4 3-39 3/4 3-40 3/4 3-41 3/4.7.6 3/4 7-15 3/4 7-16 3/4.9.4 3/4 9-4 3/4.9.9 3/4 9-10 3/4.9.12 3/49-13 3/4 9-14 3/49-15 B 3/4.3.3.1 B 3/4 3-1a B 3/4 3-2 B 3/4.7.7 B 3/4 7-5c B 3/4.9.4 B 3/4 9-1 B 3/4 9-2 B 3/4 9-3 B 3/4.9.9 B 3/4 9-4 B 3/4.912

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

1.7.1 All penetrations required to be closed during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

1.7.2 All equipment hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

CORE ALTERATION 1.8 NOT USCED ORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CORE A,^LTERATION 1.9 C RE ALTrERATION chall be the movem.ent or manipulation of any component within the reactor pessure ve.s.sel with the ve...l head reamove and fual in the vessel. Suspension of CORE AILTERITION shall not preclude completion of movement of a component to 2 safe "onsea'at"ve position.

CORE OPERATING LIMITS REPORT 1.9a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these operating limits is addressed in individual specifications.

DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The SALEM - UNIT 1 1-2 Amendment No. 20 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT IN5RTRIIMFNT C)PFRARI F M('f1F5R RETPOINT RANGE ACTION INSTRUMENT OPERABLE MODES SETPOINT

1. AREA MONITORS

a. Fuel Storage Area 1 _515 mR/hr 10-1104 mR/hr 19

b. Containment Area 2 1,2,3&4 _0103 R/hr 1-10z R/hr 23

2. PROCESS MONITORS

a. Containment Y) 20 112

1) Gaseous Activity 6 Set at lass than or equal 40--1-Gpm a) Purge & Pressure and to 50% of the 10CFR20 Vacuum Relief concentratio-n limits Isolation for gaseonus effluents relea~rd to unreatrintad aea&

1,2,3,4&5 per ODCM Control 3.3.3.9 b) RCS Leakage 1 1,2,3&4 N/A 101-106 cpm 20 Detection

2) Air Particulate Activity A A*

a) Purge & Pressure 1 0 ý2x oaG~gruna IQ 44.--- cpm 22 Va-cuum Relief

-I-soatoe (NOT USED) b) RCS Leakage 1 1,2,3&4 N/A 101_106 cpm 20 Detection

With fuel in the storage pool or building.

The plant vent noble gas monitor may also function in this capacity when the purge/pressure-vacuum relief isolation valves are open.

SALEM - UNIT 1 3/4 3-36 Amendment No. 236

TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 19 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 20 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

ACTION 22 (Not Used) W.th the number of channels OPERABLE lass than required by the Minimum Channels OPER:ABLE requ .... iement, comply with the -ACTION requirements of Specifi-ation ".9.9.

ACTION 23 - With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s),

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 24 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel(s) to OPERABLE status within 7 days or initiate and maintain operation of the Control Room Emergency Air Conditioning System (CREACS) in the pressurization or recirculation mode of operation. CORE ALTERATIONS and movement of irradiated fuel assemblies will be suspended during operation in the recirculation mode.

ACTION 25 - With no channels OPERABLE in a Control Room air intake, immediately initiate and maintain operation of the CREACS in the pressurization or recirculation mode of operation. CORE ALTERATIONS and movement of irradiated fuel assemblies will be suspended during operation in the recirculation mode.

SALEM - UNIT 1 3/4 3-37 Amendment No.-22-5

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNELS SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECKS CHECKS CALIBRATION TEST REQUIRED

1. AREA MONITORS

a. Fuel Storage Area S M R Q

b. Containment Area S M R Q 1,2, 3&4

2. PROCESS MONITORS

a. Containment Monitors

1) Gaseous Activity a) Purge & Pressure S M R Q 1, 2, 3, 4T&-5&

Vacuum Relief Isolation b) RCS Leakage S M R Q 1,2,3 &4 Detection

2) Air Particulate Activity Purge & Pres'ure I A M S.. . .

A a) SM R t 12 3 4. &S I VaumRelie Iratioaton (NOT USED) b) RCS Leakage S M R Q 1,2,3&4 Detection

With fuel in the storage pool or building.

SALEM - UNIT 1 3/4 3-38 Amendment No. 157

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY AIR CONDITIONING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6.1 The common control room emergency air conditioning system (CREACS) shall be OPERABLE with:

a. Two independent air conditioning filtration trains (one from each unit) consisting of

1. Two fans and associated outlet dampers,

2. One cooling coil,

3. One charcoal adsorber and HEPA filter array,

4. Return air isolation damper.

b. All other automatic dampers required for operation in the pressurization or recirculation modes.

c. The control room envelope intact.

APPLICABILITY: ALL MODES and during movement of irradiated fuel assemblies and during CORE ALTERATIONS.

ACTION: MODES 1, 2, 3, and 4

a. With one filtration train inoperable, align CREACS for single filtration train operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and restore the inoperable filtration train to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With CREACS aligned for single filtration train operation and with one of the two remaining fans or associated outlet damper inoperable, restore the inoperable fan or damper to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With the Control Room Envelope inoperable, restore the Control Room Envelope to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With one or both series isolation damper(s) on a normal Control Area Air Conditioning System (CAACS) outside air intake or exhaust duct inoperable, close the affected duct within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one isolation damper secured in the closed position or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . (Refer to ACTION 24 of Table 3.3-6.)

The CREACS is a shared system with Salem Unit 2 SALEM - UNIT 1 3/4 7-18 Amendment No.490

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

e. With one or both isolation damper(s) on an outside emergency air conditioning air intake duct inoperable, close the affected duct within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one isolation damper secured in the closed position and restore the damper(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

f. With any isolation damper between the normal CAACS and the CREACS inoperable, secure the damper in the closed position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6 or during movement of irradiated fuel assemblies and during CORE ALTERPATIONS.

a. With one filtration train inoperable, align CREACS for single filtration train operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or suspend CORE .A.LTERATIONS and movement of irradiated fuel assemblies.

b. With CREACS aligned for single filtration train operation with one of the two remaining fans or associated outlet damper inoperable, restore the fan or damper to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or suspend CORE A.LTElA.TIONS a.-Nd a ovement of irradiated fuel assemblies.

c. With two filtration trains inoperable, immediately suspend CORE ALTEPRATIONS and movement of irradiated fuel assemblies.

d. With the Control Room Envelope inoperable, immediately suspend CORE.

.A.LTERPATIONS N .,d imovement of irradiated fuel assemblies.

e. With one or both series isolation damper(s) on a normal CAACS outside air intake or exhaust duct inoperable, immediately suspend CORE A.LTERATIONS and-movement of irradiated fuel assemblies until the affected duct is closed by use of at least one isolation damper secured in the closed position. (Refer to ACTION 24 of Table 3.3-6.)

f. With one or both series isolation damper(s) on an outside emergency air conditioning air intake duct inoperable, immediately suspend CORE .A.LTEPRA.TIONS And movement of irradiated fuel assemblies until the affected duct is closed by use of at least one isolation damper secured in the closed position. To resume CORE A'LTERATIONS or movement of irradiated fuel assemblies, at least one emergency air intake duct must be operable on each unit.

g. With any isolation damper between the CAACS and the CREACS inoperable, immediately suspend CORE A'LTERATIONS 2nd movement of irradiated fuel assemblies until the damper is closed and secured in the closed position.

SALEM - UNIT 1 3/4 7-19 Amendment No.-190

REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment hatch inside door is capable of being closed and held in place by a minimum of four bolts, or an equivalent closure device installed and capable of being closed,

b. A minimum of one door in each airlock is capable of being closed

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by the Containment Purge and Pressure-Vacuum Relief Isolation System.

Note: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During CORE ALTER.ATIONS or movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE A.LTERA.TIOhS or movement of irradiated fuel in the containment building. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by a manual or automatic containment isolation valve at least once per 7 days.

4.9.4.2 Once per refueling prior to the start of movement of irradiated fuel assemblies within the containment building, verify the capability to install, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the equipment hatch. Applicable only when the equipment hatch is open during movement of irradiated fuel in the containment building.

4.9.4.3 Verify, once per 18 months, each required containment purge isolation valve actuates to the isolation position on a manual actuation signal.

SALEM - UNIT 1 3/4 9-4 Amendment No. 21 I

a

-I n

I. '3 0

0 p n n

m a -I ii r

7 2

z 7 4 )7 0** )

a I:

TI I 2 n a

ii 7 j

.3: n 3 I ) 7 a a 1*

3 a

T 4 )

. *1 a n

3 T a 3 4

-I 3:

3 r ( 7 3 4 p

3 I I 3 !

3 3 .5 3.

o *i 9 I 3

I S 1 1 3! 3 4

- I 3 3 1 3 I 3 4 CL - 4.

I) 3D 3' )

.3 U

3 I 3 4 9 1 3 I 3 4 3

3 1 I) I - 4 3 I 3 4 3 I 3 3 3 .4 3

3 3 4 3 I r

3 II 3 I 3

REFUELING OPERATIONS FUEL HANDLING AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 The Fuel Handling Area Ventilation System shall be OPERABLE with:

a. Two exhaust fans and one supply fan OPERABLE and operating, and

b. Capable of maintaining slightly negative pressure in the Fuel Handling Building.

APPLICABILITY: Dhngemovermdatent ofeirradiin the storage pool.

During movement of irradiated fuel within the Fuel Handling Building ACTION:

a. With no Fuel Handling Area Ventilation System OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with lonad over the storage pool ntilthe Fuel Handling Area Ventilation System is restored to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required ventilation system shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that, the Fuel Handling Building is maintained at a slightly negative pressure with respect to atmospheric pressure.

b. At least once per 31 days by verifying both exhaust fans and one supply fan start and operate for at least 15 minutes, if not operating already.

c. At least once per 18 months by verifying a system flowrate of 19,490 cfm +/- 10% during system operation.

.1 At least once per 31 days by initiating flow through the HEPA. tuto an cfarca dofo ri and verifying that the train operates for at least 15 minuter, a.2 Prior to an during mo~vement of irra-diated he-a' seamblies Or crane operation over the storFage pool in the Fuel Handling Building:

I.BRoth e~xhausrtfans and one supply fan must beOPIERABLE and operating with flow being directed through the HEPA and charcoal filters.

2.AII damnpers required to divert the entire ai4flW through the W9EPA'chraroal filter train are OPER-ABLE and inthe position reequid to divert ful exhau st 1loM through the HE.Alch arcoal 3.IDUctWork, dampers and housings whic-h Will ensure all post-accident exhausted air isprccessed through the HEPAtcharcoal filter train are nat 41The- fuelI handling area ismaintained at a negative pressure equal to or more negative than I1I8 inch water gauge relative to- the-oultsid-e atmosphere, and 5-.At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the-reafter verify both exhaust fans and one supply fan operating wt the entir flo1-w being directed through the HEP2A and c-har-c-oal fles

h. At least once per 18- months- -or(1)after any str-actural maintenance on the HEPA filter or charcoal

-ad-old-e-r Housings, or t2) fTowuing painingri, We Or Gir PIIIILa, nIUIea-s inI an -ni .

commnicaingwith the systemn, by; SALEM - UNIT 1 3/4 9-12 Amendment No.2 REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 1.Verifying that with the. ventilation system. operating at a, flow ra-te of 19,190 cfm + 10% and exhausting through the HEPA filters-and charcol .'*t.adeorbers, the total bypass flowu of the ventilation sys-te-m to the- fac*iliy vent, including leakage through the ventilation systemn dive~fing-valves, is!5 i YAc wnie n tne- ventilation system is,reetec by aumining coic uur at tne storage pc ventilation system intake 2.Verifying that the charcoal adsorber. remove > 99% of a halogenated hydrcarbon refrigern t test gas when they are tested in,-pace while operating the vYetilation system at a flow..,, rate of 19,490 cfm +/- 10%.

3.Ver4'ing that the HEP Afitte r bhanks remora-ve >Ž A99 %of the -DO-PWhen they are tested in place while operating the ventilation system at a-flows rate of 19,190 cfm 1n 4.Veedfying within 31 days after re.mo.val, from the FHV unit, that a laboratory test of a sample of the cha*rcoal2 adso.-rer. w.,hen o-btained* in a*ccorda.nce with R~egula2torsP Position- C.6.b of Regulatory Guide 1.52, Revision 2,March 1978, showsA9 the methyl ioide penetration less, tha 5.0% when tested ina~ccdanace with ASTMA D3803-1989 at a temperature of 300C and a rel~ative hu midit' of 95%.

5.Ver\Ifing a system flow rate of 19,490 cm, +/-r10% duning system operation.

c.After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber Dperation by verifying wil thin 31 days after re-moval forom the FHV unit, that a laborato.y analysis of a representative .rbr.. , sample, when Obtained in accordance with Regulatory Position C.6.b of Regulatory Guic ae1.2 Revision 2, March 197-8, shws a meathyl iodide penetration less than 5.0% when taetem- IA 2r-r-01-02ArA a : 4..db Artan'on Maim A-4-11.01 boas loon annrf +, Ira f r A

-4~lf mddh f 0rO/.

^aai,

vvv* a*c v, . u , mp %v, v . , . ,. ., ,, .- v. . ,

SALEM - UNIT 1 3/4 9-13 Amendment No. 245

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

d. At least once per 18 months by:

1 that the pressure drop across the co.mb.ined

.Verifying . HE.PA filters :and charcoal.ad.rber ba Si Water Gauge wniie operatng the ventilation system at a nowMrate or wi WW +/-T, 2.Verifi*n* that the air flow distribuitions u*niform Within 20% across HEPA filters and charco-l

.Verifying that the ventation systemn maintains the spent fuel storage pool area at a negative pressure of >_1.12 inc-heas Water Gauge relative to the ouwtsides a2tmosphere during system oprtin e- Afterf each complete or partial replacemeAnt of a-HEPA filtr ba2nk by verifying that the HEPA filter bhankse remove > 99%A of the DOP when they are tested in place While operating the filter train at a flow rate of 19,190 crf. +/- 10%.

f. After each complete or partial replacemen~t of a charcoal absorber bank by verifying that the charcoa absorbers remove > 99% of a2aoeae hydrocarbon refrigerant test gas when the" are_ tested- inrplace while operating the filter train at a flowN ra-tA Of 19,190 *cf 10/.

SALEM - UNIT 1 3/4 9-14 Amendment No.-245

INSTRUMENTATION BASES field sensors and signal processing equipment for these channels are assumed to operate within the allowances of these uncertainty magnitudes.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," and Supplements to that report. Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

Th iodnalar rRip SetpgiSt fo -h.Continment and P2resure R*li*f system du'-ng MOD*E 6 i

.Pure etabi;fshed to enwure that nthe event of 2 handling

-Iel accide;nt insie cotainment, progmpt otion "ill occu-r to ensure ca~lul2aed oft its doses reain1 below I OCF=RI QO limits. Prompt ioaonWill also ensu1reha Contrhl Roomsdoae Hing a fuel handing accident willdremA bnlolits follo The aiar1Crip se5.oi7nd lueula of Tale3.13 for theaRi2A Whilei Mode 6 Wil bke auponisoltaingenthpeConatinm csrablisedtbae Purge and P~ressure Relief System When Gona~inment gaseous actiVity, levels roach 50% of the more connerawtive IOCFR20 concentration liisfrrlaato unrartricted aras. These concentration i~mits re specified in !0CFR2O, Appendix B, Table 11,Column I- A setpoint base-d on 50% of the I 0CFR2 coGncentF2ation limit_ Will be low enough to ensure that prompt CnimntP-urge and Pressrer Relief system isoltionn occu~ during a fuel handling accident and high enough to prevent unnecesr. Coaimn Pug and Pressure Relief systemn isltinaused by routine outage activities.

In the postulated Fuel Handling Accident, the revised dose calculations. performed using 10 CFR 50.67 and Regulatory Guide 1.183. Alternative Source Term, do not take credit for automatic containment purge isolation thus allowing for continuous monitoring of containment activity until containment closure is achieved. If required, containment purge isolation can be initiated manually from the control room.

SALEM - UNIT 1 B 3/4 3-1 a Amendment No. 2.36 1

INSTRUMENTATION BASES 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION (Continued)

CROSS REFERENCE - TABLES 3.3-6 AND 4.3-3 TIS Table Instrument Description Acceptable RMs Item No. Channels la Fuel Storage Area 1R5 or I R9 lb Containment Area 1R44A and B 2ala Containment Gaseous Activity Purge & PressureNacuum 1R12A or 1R41A, and Relief Isolation D(1) (2) 2alb Containment Gaseous Activity RCS Leakage Detection 1R12A 2a2a Containment Air Pa^ rtoicu-,,late Activity Purge & I4R141A Pressure.a!Acr--m Relief Isolation (NOT USED) 2a2b Containment Air Particulate Activity RCS Leakage Detection 1R11A 2b1 Noble Gas Effluent Medium Range Auxiliary Building Exhaust 1R41 B System (Plant Vent) & D (1X3X5) 2b2 Noble Gas Effluent High Range Auxiliary Building Exhaust IR41C &D "".J'o System (Plant Vent) 2b3 Noble Gas Effluent Main Steamline Discharge - Safety Valves I R46 and Atmospheric Steam Dumps 2b4 Noble Gas Effluent Condenser Exhaust System 1R15 3a Unit I Control Room Intake Channel 1 (to Unit I Monitor) 1R1 B-1 Unit 1 Control Room Intake Channel 2 (to Unit 2 Monitor) 2R1 B-2 Unit 2 Control Room Intake Channel 1 (to Unit 2 Monitor)

Unit 2 Control Room Intake Channel 2 (to Unit I Monitor) 2R1 B-1 IRI B-2 Immediate action(s), in accordance with the LCO Action Statements, means that the required action should be pursued without delay and in a controlled manner.

(1) The channels listed are required to be operable to meet a single operable channel for the Technical Specificiation's "Minimum Channels Operable" requirement.

(2) The setpoint applies to 1R41 D. The measurement range applies to 1 R41A and B which display in uCi/cc using the appropriate channel conversion factor form cpm to uCi/cc.

SALEM - UNIT 1 B 3/4 3-2 Amendment No. January 28, 1998

PLANT SYSTEMS BASES The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix "A", 10 CFR 50 except for the Fuel Handling Accident, where the allowable doses to Control Roomi personnel are provided by 10CFR 50.67 and RG 1.183, Alternative Source Term.

3/4.7.7 AUXILIARY BUILDING EXHAUST AIR FILTRATION SYSTEM The Auxiliary Building Ventilation System (ABVS) consists of two major subsystems. They are designed to control Auxiliary Building temperature during normal and emergency modes of operation, and to contain Auxiliary Building airborne contamination during Loss of Coolant Accidents (LOCA). The two subsystems are:

1. A once through filtration exhaust system, designed to contain particulate and gaseous contamination and prevent it from being released from the building in accordance with 10CFR20, and

2. A once through air supply system, designed to deliver outside air into the building to maintain building temperatures within acceptable limits. For the purposes of satisfying the Technical Specification LCO, one supply fan must be administratively removed from service such that the fan will not auto-start on an actuation signal; however, the supply fan must be OPERABLE with the exception of this administrative control.

These systems operate during normal and emergency plant modes. Additionally, the system provides a flow path for containment purge supply and exhaust during Modes 5 and 6. Either the Containment Purge system or the Auxiliary Building Ventilation System with suction from the containment atmosphere, with associated radiation monitoring will be available whenever movement of irradiated fuel is in progress in the containment building and the equipment hatch is open. If for any reason, this ventilation requirement can not be met, movement of fuel assemblies within the containment building shall be discontinued until the flow path(s) can be reestablished or close the equipment hatch and personnel airlocks.

The exhaust system consists of three 50% capacity fans that are powered from vital buses. These fans exhaust from a common plenum downstream from three High Efficiency Particulate Air (HEPA) filter banks, two of which, 11 & 12 can be interchangeably aligned to discharge to a single carbon adsorber bed. Filter unit 11 is limited in capacity and can only be aligned to the ECCS areas of the Auxiliary Building for HEPA only or HEPA + Carbon modes of filtration. Filter unit 12 can be used to ventilate the normal areas of the Auxiliary Building in HEPA only, or when used in conjunction with 13, may be used to ventilate the ECCS areas of the Auxiliary Building in HEPA + Carbon. Filter unit 13 does not communicate with the carbon adsorber housing and is used for exhausting air from the normal areas of the Auxiliary Building during any plant Mode or purging the Containment Building during Modes 5&6. The fans are designed for continuous operation, to control the Auxiliary Building pressure at -0.10" Water Gauge with respect to atmosphere.

The supply system consists of two 100% capacity fans that are powered from vital buses, and distribute outdoor air to the general areas and corridors of the building through associated ductwork.

AUXILIARY BUILDING VENTILATION ALIGNMENT MATRIX Unit 11 from ECCS HEPA only, with Unit 12 from Aux. Normal HEPA only; or SALEM - UNIT 1 B 3/4 7-5c Amendment No.22-8

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2000 ppm) ensure that: 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on K*ff of no greater than 0.95 which includes a conservative allowance for uncertainties, is sufficient to prevent reactor criticality during refueling operations.

The sampling and analysis required by surveillance requirement 4.9.1.2 ensures the boron concentration required by Limiting Condition of Operation 3.9.1 is met. Sampling and analysis of the refueling canal is required if water exists in the refueling canal, regardless of the amount.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS During CORE A'LTE.RATIONS or movement of irradiated fuel assemblies within containment the requirements for containment building penetration closure capability and OPERABILITY ensure that a release of fission product radioactivity within containment will not exceed the guidelines and dose calculations described in Reg. Guide 1.183, Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Reactors. In MODE 6, the potential for containment pressurization as a result of an accident is not likely. Therefore, the requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements during CORE A.LTEGRATIONS or movement of irradiated fuel assemblies within containment are referred to as "containment closure" rather than containment OPERABILITY. For the containment to be OPERABLE, CONTAINMENT INTEGRITY must be maintained. Containment closure means that all potential containment atmosphere release paths are closed or capable of being closed. Closure restrictions include the administrative controls to allow the opening of both airlock doors and the equipment hatch during fuel movement provided that: 1) the equipment inside door or an equivalent closure device installed is capable of being closed with four bolts within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by a designated personnel: 2) the airlock door is capable of being closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by a designated personnel, 3) either the Containment Purge System or the Auxiliary Building Ventilation System taking suction from the containment atmosphere are operating and 4) the plant is in Mode 6 with at least 23 feet of water above the reactor pressure vessel flange.

Administrative requirements are established for the responsibilities and appropriate actions of the designated personnel in the event of a Fuel Handling Accident inside containment. These requirements include the responsibility to be able to communicate with the control room, to ensure that the equipment hatch is capable of being closed, and to close the equipment hatch and personnel airlocks within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the event of a fuel handling accident inside containment. These administrative controls ensure containment closure will be established in accordance with and not to exceed the dose calculations performed using guidelines of Regulatory Guide 1.183.

SALEM - UNIT 1 B 3/4 9-1 Amendment No. 2-1-7

3/4.9 REFUELING OPERATIONS BASES The containment serves to limit the fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10CFR100 and Reg. Guide 1.183, Alternative Source Term, as applicable. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The Containment Equipment Hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into or out of containment. During GORE ALTERATIONRS or movement of irradiated fuel assemblies within containment, the Containment Equipment Hatch inside door can be open provided that: 1) It is capable of being closed with four bolts within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by designated personnel, 2) either the Containment Purge System or the Auxiliary Building Ventilation System taking suction from the containment atmosphere are operating and 3) The plant is in Mode 6 with at least 23 feet of water above the reactor pressure vessel flange. Good engineering practice dictates that the bolts required by the LCO are approximately equally spaced.

An equivalent closure device may be installed as an alternative to installing the Containment Equipment Hatch inside door with a minimum of four bolts. Such a closure device may provide penetrations for temporary services used to support maintenance activities inside containment at times when containment closure is required; and may be installed in place of the Containment Equipment Hatch inside door or outside door. Penetrations incorporated into the design of an equivalent closure device will be considered a part of the containment boundary and as such will be subject to the requirements of Technical Specification 3/4.9.4. Any equivalent closure device used to satisfy the requirements of Technical Specification 3/4.9.4.a will be designed, fabricated, installed, tested, and utilized in accordance with established procedures to ensure that the design requirements for the mitigation of a fuel handling accident during refueling operations are met. In case that this equivalent closure device is installed in lieu of the equipment hatch inside door, the same restrictions and administrative controls apply to ensure closure will take place within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following a FHA inside containment.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during operation in MODES 1, 2, 3, and 4 as specified in LCO 3.6.1.3, "Containment Air Locks". Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown, when containment closure is not required and frequent containment entry is necessary, the air lock interlock mechanism may be disabled. This allows both doors of an airlock to remain open for extended periods.

During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, containment closure may be required; therefore, the door interlock mechanism may remain disabled, and both doors of each containment airlock may be open if 1) At least one door of each airlock is capable of being closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by a dedicated individual, 2) either the Containment Purge System or the Auxiliary Building Ventilation System taking suction from the containment atmosphere are operating and 3) the plant is in Mode 6 with at least 23 feet of water above the reactor pressure vessel flange.

In the postulated Fuel Handling Accident, the revised dose calculations, performed using 10 CFR 50.67 and Regulatory Guide 1.183, Alternative Source Term, do not take credit for automatic containment purge isolation thus allowing for continuous monitoring of containment activity until containment closure is achieved. If required, containment purge isolation can be initiated manually from the control room.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent. Equivalent isolation methods may include the use of a material that can provide a temporary atmospheric pressure, ventilation barrier. Any equivalent method used to satisfy the requirements of Technical Specification 3/4.9.4.c. 1 will be designed, fabricated, installed, tested, and utilized in accordance with established procedures to ensure that the design requirements for the mitigation of a fuel handling accident during refueling operations are met.

SALEM - UNIT 1 B 3/4 9-2 Amendment No. 2-4 3/4.9 REFUELING OPERATIONS BASES The surveillance requirement 4.9.4.2 demonstrates that the necessary hardware, tools, and equipment are available to close the equipment hatch. The surveillance is performed prior to movement of irradiated fuel assemblies within the containment. This surveillance is only required to be met when the equipment hatch is to be open during fuel movement.

The OPEFRABILITY of the containment purge system ensumres that the containment vent and purge penetrations Will be automatically isolated upon detection of high radiation levels within the contairnment The OPERABIL ITY of this system is required to rastrict the rlea~se of radionactiv material from the co-nt2inment atmorphere to the environment.

3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that: 1) manipulator cranes will be used for movement of control rods and fuel assemblies, 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140°F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification. A minimum flow rate of 1000 gpm is required. Additional flow limitations are specified in plant procedures, with the design basis documented in the Salem UFSAR. These flow limitations address the concerns related to vortexing and air entrapment in the Residual Heat Removal system, and provide operational flexibility by adjusting the flow limitations based on time after shutdown. The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

SALEM - UNIT 1 B 3/4 9-3 Amendment No. 2-1-7

REFUELING OPERATIONS BASES For support systems: Service Water (SW) and Component Cooling (CC), component redundancy is necessary to ensure no single active component failure will cause the loss of Decay Heat Removal. One piping path of SW and CC is adequate when it supports both RHR loops. The support systems needed before entering into the desired configuration (e.g., one service water loop out for maintenance in Modes 5 and 6) are controlled by procedures, and include the following:

A requirement that the two RHR, two CC and two SW pumps, powered from two different vital buses be kept operable A listing of the active (air/motor operated) valves in the affected flow path to be locked open or disabled.

Note that four filled reactor coolant loops, with at least two steam generators with at least their secondary side water level greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop. This ensures that a single failure does not cause a loss of decay heat removal.

With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 (NOT USED) COGINTAINMENT PUIRGE AND PRIESSUR IRCUUMI r D4IRELI EF *21 ATION SYSTEM Thea OPERABIL ITY of this system enasuresr that the containment vent and purgje penetrations ;w-ill be, u.toma..cally isolated upon dete..tion of high radiat.i.on l .within the cont.inment. The O.RABLITY of this system is required to reastrict the releaste of radioarctive material frm the containment atmosphere to the environment.

3/4.9.10 and 3/4/9/11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM The lietations on the Fuel handling area Ventilation system ensure that all ramdnoactie material releasd from a drpped oirdie fuelpassemtbly will be filtered through the HiEPA f and-witelr harcal adsorber prior to discharge to the atmosphere. The OPER-AB-ILITY of this syste-m iG consiStent with the.

assumptions of the accident analyses. I abnratory testing of thAe arbon Radolriarci perm in 2accordance With ASTh 102802-1989 with an accreptance criteria that isdetermined by applying a minium sfe* factor of 2 to the charcoal filter reoarapl efficiency Gredited in the design basis dose analysis as specified in Generic- Letter 99-02.

The operability of the Fuel Handling Area Ventilation System during movement of irradiated fuel ensures that a release of fission product radioactivity within the Fuel Handling Building will not exceed the guidelines and dose calculations described in Reg. Guide 1.183. Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Reactors.ensures all building exhaus*t fio;w is pro..ssod through the HEPAlcharc-al filter train whenever. a Fuel HandlinqGAccident is possible. This.ill minimize .. ite doses folo;i-ng the Dost-lated Fuel Handl*n a.ccident.

SALEM - UNIT 1 B 3/4 9-4 Amendment No. 245

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

1.7.1 All penetrations required to be closed during accident conditions are either:

a. Capable of being closed by an OPERABLE containment automatic isolation valve system, or

b. Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except for valves that are opened under administrative control as permitted by Specification 3.6.3.1.

1.7.2 All equipment hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

CORE ALTERATION 1.8 NOT USED CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control I components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position CORE ALTERA.TION 1 "CORE ALTIERATION ANall ha the mo..ve.ment or manipulation of any component within the reac2or pressure vessel with the ve*sel head removed 2nd f.elin the vessel. Suspension of CORE

.AI-TEIRATION shall not prFeclude com~pletion of moevement Of a component to a rafe annseriafive Position.

CORE OPERATING LIMITS REPORT 1.9a- The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.9. Unit operation within these operating limits is addressed in individual specifications.

DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries per gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The SALEM - UNIT 2 1-2 Amendment No. 1-97 I

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM/TRIP MEASUREMENT INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION INSTRUMENT OPERABLE

1. AREA MONITORS

a. Fuel Storage Area 1 _515 mR/hr 101_104 mR/hr 23

b. Containment Area 2 1,2,3&4 _0103 R/hr 1-10 7 R/hr 26

2. PROCESS MONITORS

a. Containment

1) Gaseous Activity 6- Set at less than or equal 401.06 cpm 26

-1 a) Purge & Pressure Vacuum Relief concentrFtion liFmAts Isolation frcggaSaau* r, fflueantrt realeased to unrstricted 1,2,3,4&5 per ODCM Control 3.3.3.9 b) RCS Leakage 1 1,2,3&4 N/A 101-106 cpm 24 Detection

2) Air Particulate Activity c-a)Purge &Pressure !Sm x9aGrgronAQ 404406 cpm 25 Vacuum-Relie l6Glati~f (NOT USED) b) RCS Leakage 1 1,2,3&4 N/A 101-106 cpm 24 Detection

With fuel in the storage pool or building.

The plant vent noble gas monitor may also function in this capacity when the purge/pressure-vacuum relief isolation valves are open.

SALEM - UNIT 2 3/4 3-39 Amendment No. 24 TABLE 3.3-6 (Continued)

TABLE NOTATION ACTION 23 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 24 With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.7.1.

ACTION 25 (Not Used) W^th the number of ch.nneal.,OPRABI En185 than required by the MinimAum Channelre OPERABLE re rquir-ement, comply with the ACTION requirements of Specification 3.9.9 ACTION 26 - With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 27 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel(s) to OPERABLE status within 7 days or initiate and maintain operation of the Control Room Emergency Air Conditioning System (CREACS) in the pressurization or recirculation mode of operation. CORE ALTERATIONS and movement of irradiated fuel assemblies will be suspended during operation in the recirculation mode.

ACTION 28 - With no channels OPERABLE in a Control Room air intake, immediately initiate and maintain operation of the CREACS in the pressurization or recirculation mode of operation. CORE ALTERATIONS and movement of irradiated fuel assemblies will be suspended during operation in the recirculation mode.

SALEM - UNIT 2 3/4 3-40 Amendment No.-206

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNELS SOURCE CHANNEL FUNCTIONAL SURVEILLANCE INSTRUMENT CHECKS CHECKS CALIBRATION TEST REQUIRED

1. AREA MONITORS

a. Fuel Storage Area S M R Q

b. Containment Area S M R Q 1,2, 3&4

2. PROCESS MONITORS

a. Containment Monitors

1) Gaseous Activity a) Purge & Pressure S M R Q 1, 2, 3, 4,5-&6 Vacuum Relief Isolation b) RCS Leakage S M R Q 1,2, 3&4 Detection

2) Air Particulate Activity a) Pu2'rg & Pressure R ha n' 1, 2, 3,& 4, &6 Vacuum ReliRe 1=12fian (NOT USED) b) RCS Leakage S M R Q 1,2, 3&4 Detection

With fuel in the storage pool or building.

SALEM - UNIT 2 3/4 3-41 Amendment No. 4-38

PLANT SYSTEMS 3/4.7.6 CONTROL ROOM EMERGENCY AIR CONDITIONING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.6 The common control room emergency air conditioning system (CREACS)* shall be OPERABLE with:

a. Two independent air conditioning filtration trains (one from each unit) consisting of:

1. Two fans and associated outlet dampers,

2. One cooling coil,

3. One charcoal adsorber and HEPA filter array,

4. Return air isolation damper.

b. All other automatic dampers required for operation in the pressurization or recirculation modes.

c. The control room envelope intact APPLICABILITY: ALL MODES and during movement of irradiated fuel assemblies and duri-ngGOR A LTE.RATIONS.

ACTION: MODES 1, 2, 3, and 4

a. With one filtration train inoperable, align CREACS for single filtration train operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and restore the inoperable filtration train to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With CREACS aligned for single filtration train operation and with one of the two remaining fans or associated outlet damper inoperable, restore the inoperable fan or damper to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With the Control Room Envelope inoperable, restore the Control Room Envelope to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With one or both series isolation damper(s) on a normal Control Area Air Conditioning System (CAACS) outside air intake or exhaust duct inoperable, close the affected duct within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one isolation damper secured in the closed position or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(Refer to ACTION 27 of Table 3.3-6.)

The CREACS is a shared system with Salem Unit 1 SALEM - UNIT 2 3/4 7-15 Amendment No.1-73 I

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)

e. With one or both isolation damper(s) on an outside emergency air conditioning air intake duct inoperable, close the affected duct within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one isolation damper secured in the closed position and restore the damper(s) to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

f. With any isolation damper between the normal CAACS and the CREACS inoperable, secure the damper in the closed position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5 and 6 or during movement of irradiated fuel assemblies and during CORE ALTERATIONS.

a. With one filtration train inoperable, align CREACS for single filtration train operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or suspend CORE .A.LTERAXTIONS and ovement of irradiated fuel assemblies.

b. With CREACS aligned for single filtration train operation with one of the two remaining fans or associated outlet damper inoperable, restore the fan or damper to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or suspend CORE A.LTEPRA.TIONS and movement of irradiated fuel assemblies.

c. With two filtration trains inoperable, immediately suspend CORE ALTERPATIONS .d movement of irradiated fuel assemblies.

d. With the Control Room Envelope inoperable, immediately suspend CORE QLTERATIONS _And movement of irradiated fuel assemblies.

e. With one or both series isolation damper(s) on a normal CAACS outside air intake or exhaust duct inoperable, immediately suspend CORE A.LTERA ONS and movement of irradiated fuel assemblies until the affected duct is closed by use of at least one isolation damper secured in the closed position. (Refer to ACTION 27 of Table 3.3-6.)

f. With one or both series isolation damper(s) on an outside emergency air conditioning air intake duct inoperable, immediately suspend CORE A'LTERATIONS and :ovement of irradiated fuel assemblies until the affected duct is closed by use of at least one isolation damper secured in the closed position. To resume CORE ALTERATITO.NS or movement of irradiated fuel assemblies, at least one emergency air intake duct must be operable on each unit.

g. With any isolation damper between the CAACS and the CREACS inoperable, immediately suspend CORE ALTEPRATIONS and movement of irradiated fuel assemblies until the damper is closed and secured in the closed position.

SALEM - UNIT 2 3/4 7-16 Amendment No. 1-73 I

REFUELING OPERATIONS CONTAINMENT BUILDING PENETRATIONS LIMITING CONDITION FOR OPERATION 3.9.4 The containment building penetrations shall be in the following status:

a. The equipment hatch inside door is capable of being closed and held in place by a minimum of four bolts, or an equivalent closure device installed and capable of being closed,

b. A minimum of one door in each airlock is capable of being closed

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by the Containment Purge and Pressure-Vacuum Relief Isolation System.

Note: Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During CORE ALTERPATIONS or movement of irradiated fuel within the containment.

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment building penetrations shall be determined to be either in its required condition or capable of being closed by a manual or automatic containment isolation valve at least once per 7 days.

4.9.4.2 Once per refueling prior to the start of movement of irradiated fuel assemblies within the containment building, verify the capability to install, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , the equipment hatch. Applicable only when the equipment hatch is open during movement of irradiated fuel in the containment building.

4.9.4.3 Verify, once per 18 months, each required containment purge isolation valve actuates to the isolation position on a manual actuation signal.

SALEM - UNIT 2 3/4 9-4 Amendment No.1-9Q

REFUELING OPERATIONS CONTAiNMENT PURGE AND PRESSURE VACUU'JM RELIEFISOL..1ATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 Tha Containment Purge and Pressure.... Vac Relief isolation system shall bha OPE21=-RABLE.

APPLICABILITY: MODE 6.

AGT40W W'th the Containment Purge and Pressure*-AVcum Reliaef isolation system inop*erble, clce ea2ch of the Purge and Preu.re. Vacuum Relief penetrFaons providing direct arccessom kon the containmn atmosphere to the outside atmosphere. The provisions Of SpeGification 2.03 are not applicable.*

SURVElI I ANCE R=QUl I1REMENTS 4.9.9 The Containment Purge and Pressurre Va-cu-um Relief isolation system shall be demonstrated

,-irf oce pe 7- days during CORE OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months p.or to the sto. - -- and at least ALTIER-ATIONS by verifying that containment Purge and Prerssure Vacuu Relief isolatio ocuson manua.R initiatn tl and on a high radiation test signal from each of the onta-inment r-diation monitoring instrumentatinn channels.

SALEM - UNIT 2 3/4 9-10 Amendment:

REFUELING OPERATIONS 3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.12 The Fuel Handling Area Ventilation System shall be OPERABLE with:

a. Two exhaust fans and one supply fan OPERABLE and operating, and

b. Capable of maintaining slightly negative pressure in the Fuel Handling Building.

APPLICABILITY: Whene..er irra-diated- fuel, is in the st-rage pool During movement of irradiated fuel within the Fuel Handling Building ACTION:

a. With no Fuel Handling Area Ventilation System OPERABLE, suspend all operations involving movement of fuel within the storage pool or ,rae operatio.. n with .ads aver the storage pe*o until the Fuel Handling Area Ventilation System is restored to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required ventilation system shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the Fuel Handling Building is maintained at a slightly negative pressure with respect to atmospheric pressure.

b. At least once per 31 days by verifying both exhaust fans and one supply fan start and operate for at least 15 minutes, if not operating already.

c. At least once per 18 months by verifying a system flowrate of 19,490 cfm +/- 10% during system operation.

1. At least once per 31 days by initiating flow through the HEP2A filter and charcoal

. ,adrber train and verifying that the train operates for at least 15 mninutes.

2.2 Prior to 2nd du ring movement of iraitdfuel asse-8mblieS Or crane operation over the storage pool in the Fuel Handling .Building:

I .Bot-h exhauset fansm and one supply fan must be O-PIER-ABL-E and operating with flowA b-eing directed through the HEPIA and charc-al filters.

2.All dampers required to di-ert the entire ai*..eow through the HEpaPtcharcoal filter train are OPER.ABLE and inthe position requir4d to die.rt.i-ll exhaust flo, through the HEP.charcoa filter tr-ain-.

3.Du-t

,,- k, dampers and housings- Whih Il'l esure all poet-_accident exh usted air is processed through the ERI-r charcoal filter trai;n re inta*t 4The fel handling area is maintained at a negative pressure equal to or more negative than I Q inch water gauge rela2tive to the auide datmosphere, and 5.At leGAastone per 21 hour2.430556e-4 days 0.00583 hours 3.472222e-5 weeks 7.9905e-6 months s- the-re-after verify bo~th exhausatfans and onesspply fan operating with the entire flowA being directed through the HEP1A and charcoal filters

h. At IA*st once per 18 mnGths or (1) after any s&tntrl maintQeAnarnA the HEPA filter or charcoal adsorber housings, or (2) following painting, fire Or chemic*,,val release in any Ventilation zone co-mmunIcating with the systeme,by:

SALEM -UNIT 2 3/4 9-13 Amendment No. 244

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 1.Veroifing that With the ventilation system operating at a flow rate of 19,190 cfan+ 10% and exhausting through the HEPA filters and charcoal a*derbers, the total bypass filo of the venteilation system to the facility vent, including leakage through the ventilation systenm diverting valves, 6is! 1 % When the ventilation system- is teested by admi*ing cold DOP at the storage pool ventilation system rntake.

2. Verifying that the c-.harcoal adsIrbers remove > 99% of a halogenated hydrocarbon test gas and that thea HERA filter hbanks remonave Q99% of tha DO*P when they refrigerant a2re tes-ted in place using the test procedure guidance of Regulator; Positions C.5.a, C.5.cEr and- C.5.d of Regulatoy Guide 1.52, Revision 2, March 1978- (except forthe prvsosof ANSI N5IO Sections 8 and 9), and the system flow rate is 19,190 cfm +/-

3. Verifying within 31 days after remo-val from the F=HV unit, that aa laboratory' test of a sample of thea charcoal adsorber, when obtained ;in .... accord-ance. with Regulatory Position C.6.b. of Regulatory Guide 1.52, Revision 2, March 1978, shows the methyl iodide penetra*tion less than 5.0% 'whenA tested in accordance with ASTM D3803- 1989 at a 0

tamrperatre a of 30 C and :a ralativa humidity of 95%.*

S.... . . ... ... . .. A*

4. Nlenrinyg a system flowl rate of 1W,4WU cfF +/- 1U0% aunng system operatin v -

C. Aftwr every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal ad.e rber operation by verifying within 31 days after removal that a laboratory analysis of a representa-tive- carbon sample, when obtained in fmrom T-he FHV unit, aaccordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, show-s a methyl iodide penetration less than 5.0% when tested in accordance With A*STM 3803 0

1989 at a temperature of 30 C and a relative humidity of 95%.

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banksis less than or equal to 1 inche-s Water Gauge while operating the system at a flow rate of 19,1490 c.fm. 10%.

2. Deleted.

Verifying that the system maintans the spent fuel storage pool :area- at a negative 3.

pressure of greater than or equal to 118 inc-hesc WAater Gauge rela.tive to the- outsid-e wi

atmosne~re -r-aulnna system oprin.n SALEM - UNIT 2 3/4 9-14 Amendment No. 226

REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

A. After each complete or partial replacement Of a HE=PA filter bank by verifying that the HEPA filter bnsre-move greater than or equal to 9%Of the DOP when119 they are tested in place While opewr*ing the system at a flo-A ra*te Of 19,190 cfmo

10%.

After each complete or partial replacement of .2acharco absorber bank by verifying that the c-harcoal absorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are in-placea tested w.hile operating the systemn at a flow- rateo 19,1490 cfm +/-k10-%-.

SALEM - UNIT 2 3/4 9-15 Amendment No.

INSTRUMENTATION BASES these uncertainties are factored into the determination of each Trip Setpoint. All field sensors and signal processing equipment for these channels are assumed to operate within the allowances of these uncertainty magnitudes.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," and Supplements to that report. Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

Th isolatin alarrtlrip setpoint for thes Containimen.t Purge and Pressure Relief system during MODE 6 ii ectablished to ennure th~at inthe event of 2 fuel handling accident iniecontainment, pr~ompt ioainwl occu_'r to ens-re calc'-lated .ffaite doses remain b"el'; I.0CFR CIOimits. Prompt isolationwill 210a eAnnsre that Cntr*ol Room dosers following a fuel handing accidentw ill ramain belol GDC 19 limitr. ThelaFrmltrip setpoint v~alu oaf Table 33-6 for thia R12.A wNhile i~n de 6 will be established based upon isolating the Containment Purge and Pre..ure Relief System when containment gaseous ativity lieels reach 509 of the more c.on*e.ative 1*CFR20 concentration imits. for raelase to unrestricted areas. These co.ncentratinn lim-its are SPecified in1OCF20, Appendix B, Table 11, Colu-min 1. A. setpoint based- on 5-0% of the 10OC-FR concentration imits Will beSlow enough to ensure that. ,pro. CnInment Purge and Pressure Relief system ilation occurs... during a fuel handling accit an high enough to prevent unne.essa

.r,ontainm.e..nt Purge and Presau-re Relief system isola*tions caused by routine outage activities.

In the postulated Fuel Handling Accident, the revised dose calculations, performed using 10 CFR 50.67 and Regulatory Guide 1.183, Alternative Source Term, do not take credit for automatic containment purge isolation thus allowing for continuous monitoring of containment activity until containment closure is achieved. If required, containment purge isolation can be initiated manually from the control room.

SALEM - UNIT 2 B 3/4 3-1 a Amendment No. 24-7U I

INSTRUMENTATION BASES 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION (Continued)

CROSS REFERENCE - TABLES 3.3-6 and 4.3-3 T/S Table Instrument Description Acceptable RMs Item No. Channels la Fuel Storage Area 2R5 or 2R9 lb Containment Area 2R44A and B 2ala Containment Gaseous Activity Purge & Pressure/Vacuum 2R12A or 2R41A, B Relief Isolation and D(1) (2) 2alb Containment Gaseous Activity RCS Leakage Detection 2R12A 2a2a Containment Air P1rticulte Actii-t Purge & Press'-re-ac*u'-'m 2R4--A Relief Isolatio (NOT USED) 2a2b Containment Air Particulate Activity RCS Leakage Detection 2R11A Noble Gas Effluent Medium Range Auxiliary Building Exhaust 2R45B(3) 2bl System (Plant Vent)

Noble Gas Effluent High Range Auxiliary Building Exhaust 2R45C(3) 2b2 System (Plant Vent) 2b3 Noble Gas Effluent Main Steamline Discharge - Safety Valves 2R46 and Atmospheric Steam Dumps 2b4 Noble Gas Effluent Condenser Exhaust System 2R15 3a Unit 2 Control Room Intake Channel 1 (to Unit 2 Monitor) 2R1 B-1 Unit 2 Control Room Intake Channel 2 (to Unit I Monitor) 1R1 B-2 Unit 1 Control Room Intake Channel 1 (to Unit 1 Monitor)

Unit 1 Control Room Intake Channel 2 (to Unit 2 Monitor) 1RI B-1 2R1B-2 (1) The channels listed are required to be operable to meet a single operable channel for the Technical Specification's "Minimum Channels Operable" requirement.

(2) For Mode 6, the setpoint applies to 2R41D using 2 x Background from 2R4*.A For Modes 1, 2, 3, 4 & 5, the setpoint applies to 2R41 D per Specification 3.3.3.9. The measurement range applies to 2R41A and B which display in uCi/cc using the appropriate channel conversion factor form cpm to uCi/cc.

(3) If 2R45 is out of service 2R41 may be used to meet the technical specification action requirement.

SALEM UNIT 2 B 3/4 3-2 Amendment No. 1-73 I

PLANT SYSTEMS BASES The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion (GDC) 19 of Appendix "A", 10 CFR Part 50. The re-analysis of the Fuel Handling Accident is based on the criteria of 10 CFR 50.67 and Reg. Guide 1.183, Alternative Source Term, which replace GDC 19 for Control Room doses.

3/4.7.7 AUXILIARY BUILDING EXHAUST AIR FILTRATION SYSTEM The Auxiliary Building Ventilation System (ABVS) consists of two major subsystems. They are designed to control Auxiliary Building temperature during normal and emergency modes of operation, and to contain Auxiliary Building airborne contamination during Loss of Coolant Accidents (LOCA). The two subsystems are:

1. A once through filtration exhaust system, designed to contain particulate and gaseous contamination and prevent it from being released from the building in accordance with 10CFR20, and

2. A once through air supply system, designed to deliver outside air into the building to maintain building temperatures within acceptable limits. For the purposes of satisfying the Technical Specification LCO, one supply fan must be administratively removed from service such that the fan will not auto-start on an actuation signal; however, the supply fan must be OPERABLE with the exception of this administrative control.

These systems operate during normal and emergency plant modes. Additionally, the system provides a flow path for containment purge supply and exhaust during Modes 5 and 6. Either the Containment Purge system or the Auxiliary Building Ventilation System with suction from the containment atmosphere, with associated radiation monitoring will be available whenever movement of irradiated fuel is in progress in the containment building and the equipment hatch is open. If for any reason, this ventilation requirement can not be met, movement of fuel assemblies within the containment building shall be discontinued until the flow path(s) can be reestablished or close the equipment hatch and personnel airlocks.

The exhaust system consists of three 50% capacity fans that are powered from vital buses. These fans exhaust from a common plenum downstream from three High Efficiency Particulate Air (HEPA) filter banks, two of which, 21 & 22 can be interchangeably aligned to discharge to a single carbon adsorber bed. Filter unit 21 is limited in capacity and can only be aligned to the ECCS areas of the Auxiliary Building for HEPA only or HEPA + Carbon modes of filtration. Filter unit 22 can be used to ventilate the normal areas of the Auxiliary Building in HEPA only, or when used in conjunction with 23, may be used to ventilate the ECCS areas of the Auxiliary Building in HEPA + Carbon. Filter unit 23 does not communicate with the carbon adsorber housing and is used for exhausting air from the normal areas of the Auxiliary Building during any plant Mode or purging the Containment Building during Modes 5&6. The fans are designed for continuous operation, to control the Auxiliary Building pressu re at -0.10" Water Gauge with respect to atmosphere.

The supply system consists of two 100% capacity fans that are powered from vital buses, and distribute outdoor air to the general areas and corridors of the building through associated ductwork.

SALEM - UNIT 2 B 3/4 7-5c Amendment No.209

3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on minimum boron concentration (2000 ppm) ensure that: 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel. The limitation on Kff of no greater than 0.95 which includes a conservative allowance for uncertainties, is sufficient to prevent reactor criticality during refueling operations.

The sampling and analysis required by surveillance requirement 4.9.1.2 ensures the boron concentration required by Limiting Condition of Operation 3.9.1 is met. Sampling and analysis of the refueling canal is required if water exists in the refueling canal, regardless of the amount.

3/4.9.2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

3/4.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS During CORE. ALTER.ATIONS or movement of irradiated fuel assemblies within containment the requirements for containment building penetration closure capability and OPERABILITY ensure that a release of fission product radioactivity within containment will be restricted from leaking to the enveir-ment not exceed the guidelines and dose calculations described in Reg Guide 1.183, Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Plants. In MODE 6, the potential for containment pressurization as a result of an accident is not likely. Therefore, the requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements during CORE .A.LTERATIONS or movement of irradiated fuel assemblies within containment are referred to as "containment closure" rather than containment OPERABILITY. For the containment to be OPERABLE, CONTAINMENT INTEGRITY must be maintained. Containment closure means that all potential release paths are closed or capable of being closed. Closure restrictions must be r.ufficient to pro.id an atmospherc .. etilation barrier to r..t.rirt r2idoni aci m.atrial ral. .a.d from :a fuel e*lement rupture during refueling operatio-n. include the administrative controls to allow the opening of both airlock doors and the equipment hatch during fuel movement provided that: 1) the equipment inside door or an equivalent closure device installed is capable of being closed with four bolts within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by a designated personnel: 2) the airlock doors are capable of being closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by designated personnel .3) either the Containment Purge System or the Auxiliary Building Ventilation System taking suction from the containment atmosphere are operating and 4) the plant is in Mode 6 with at least 23 feet of water above the reactor pressure vessel flange.

Administrative requirements are established for the responsibilities and appropriate actions of the designated personnel in the event of a Fuel Handling Accident inside containment. These requirements include the responsibility to be able to communicate with the control room, to ensure that the equipment hatch is capable of being closed, and to close the equipment hatch and personnel airlocks within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the event of a fuel handling accident inside containment. These administrative controls ensure containment closure will be established in accordance with and not to exceed the dose calculations performed using guidelines of Regulatory Guide 1.183.

SALEM - UNIT 2 B 3/4 9-1 Amendment No. 449

REFUELING OPERATIONS BASES The containment serves to limit the fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10CFR100 and Reg Guide 1.183, Alternative Source Term, as applicable. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The Containment Equipment Hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into or out of containment. During GORE ALTERATIONS or movement of irradiated fuel assemblies within containment can be open provided that:

1) it is capable of being closed with four bolts within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by designated personnel, 2) either the Containment Purge System or the Auxiliary Building Ventilation System taking suction from the containment atmosphere are operating and 3) the plant is in Mode 6 with at least 23 feet of water above the reactor pressure vessel flange.the Containment Equipment Hatch inid door must be held in pla*c, by at least four bolts. Good engineering practice dictates that the bolts required by the LCO are approximately equally spaced.

An equivalent closure device may be installed as an alternative to installing the Containment Equipment Hatch inside door with a minimum of four bolts. Such a closure device may provide penetrations for temporary services used to support maintenance activities inside containment at times when containment closure is required; and may be installed in place of the Containment Equipment Hatch inside door or outside door. Penetrations incorporated into the design of an equivalent closure device will be considered a part of the containment boundary and as such will be subject to the requirements of Technical Specification 3/4.9.4. Any equivalent closure device used to satisfy the requirements of Technical Specification 3/4.9.4.a will be designed, fabricated, installed, tested, and utilized in accordance with established procedures to ensure that the design requirements for the mitigation of a fuel handling accident during refueling operations are met. In case that this equivalent closure device is installed in lieu of the equipment hatch inside door, the same restrictions and administrative controls apply to ensure closure will take place within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following a Fuel Handling Accident inside containment.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during operation in MODES 1, 2, 3, and 4 as specified in LCO 3.6.1.3, "Containment Air Locks". Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown, when containment closure is not required and frequent containment entry is necessary, the air lock interlock mechanism may be disabled. This allows both doors of an airlock to remain open for extended periods. During CORE A.LTERATIONS or movement of irradiated fuel assemblies within containment, containment closure may be required; therefore, the door interlock mechanism may remain disabled,-bu4 on. air lock door mu-st 2lhay, r.main closed and both doors of each containment airlock may be open if:

1) At least one door of each airlock is capable of being closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by dedicated personnel,) 2) either the Containment Purge System or the Auxiliary Building Ventilation System taking suction from the containment atmosphere are operating and 3) The plant is in Mode 6 with at least 23 feet of water above the reactor pressure vessel flange.

In the postulated FHA, the revised dose calculations performed using RG 1.183 criteria, do not assume automatic containment purge isolation thus allowing for continous monitoring of containment acitivity until the release pathways are isolated. If required, manual isolation of containment purge can be initiated from the control room.

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by an OPERABLE automatic isolation valve, or by a manual isolation valve, blind flange, or equivalent. Equivalent isolation methods may include the use of a material that can provide a temporary atmospheric pressure, ventilation barrier. Any equivalent method used to satisfy the requirements of Technical Specification 3/4.9.4.c. 1 will be designed, fabricated, installed, tested, and utilized in accordance with established procedures to ensure that the design requirements for the mitigation of a fuel handling accident during refueling operations are met.

SALEM - UNIT 2 B 3/4 9-2 Amendment No. -

REFUELING OPERATIONS BASES The surveillance requirement 4.9.4.2 demonstrates that the necessary hardware, tools, and equipment are available to close the equipment hatch. The surveillance is performed once per refueling prior to the start of movement of irradiated fuel assemblies within the containment. This surveillance is only required to be met when the equipment hatch is open.

The OPERABILITY of the conAinment purge sys-tam a-enuresg that the- containment vent and purge Penetrations Will bD automglatically dupon detection of high rad*iation lWWlS Within the contaffinment.

The OPELRABIITY of this system is requir9d t reatrict the re'l2se of rad*oactive mrateril from the contanment =atmosphere to the environment 3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

3/4.9.6 MANIPULATOR CRANE The OPERABILITY requirements for the manipulator cranes ensure that: 1) manipulator cranes will be used for movement of control rods and fuel assemblies, 2) each crane has sufficient load capacity to lift a control rod or fuel assembly, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analyses.

3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirements that at least one residual heat removal loop be in operation ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140°F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification. A minimum flow rate of 1000 gpm is required. Additional flow limitations are specified in plant procedures, with the design basis documented in the Salem UFSAR. These flow limitations address the concerns related to vortexing and air entrapment in the Residual Heat Removal system, and provide operational flexibility by adjusting the flow limitations based on time after shutdown. The requirement to have two RHR loops OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

For support systems: Service Water (SW) and Component Cooling (CC), component redundancy is necessary to ensure no single active component failure will cause the loss of Decay Heat Removal. One piping path of SW and CC is adequate when it supports both RHR loops. The support systems needed before entering into the desired configuration (e.g., one service water loop out for maintenance in Modes 5 and 6) are controlled by procedures, and include the following:

A requirement that the two RHR, two CC and two SW pumps, powered from two different vital buses be kept operable A listing of the active (air/motor operated) valves in the affected flow path to be locked open or disable.

SALEM - UNIT 2 B 3/4 9-3 Amendment No. 499

REFUELING OPERATIONS BASES Note that four filled reactor coolant loops, with at least two steam generators with at least their secondary side water level greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop. This ensures that single failure does not cause a loss of decay heat removal.

With the reactor vessel head removed and 23 feet of water above the reactor pressure vessel flange, a large heat sink is available for core cooling. Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

3/4.9.9 (Not Used)CONTA.INMENT PURGE A.ND PRESSURE-VACU UM RELIEF ISO'.ATION SYSTEM The OPERABILITY of thios system ensu1res that theq containment vent and purge penetrations will be.

automaticall'y isolated upon detection of high radiation le40el6 Within the containment. The OPERABILITY4 of this system is required. to restrict the release of radioac.tive material fom the containment atmosphere to the environment.

3/4.9.10 and 3/4/9/11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

3/4.9.12 FUEL HANDLING AREA VENTILATION SYSTEM The irmitationi on the Fuel handling areaVentilation system enurge that all iradioactie fmateril released fo a dopped i r it faue assembly will be filtered through the HElA filters and ecehardo adsorbor prior to discr-harge to the atmosphere. The 0OPERABIL ITY of this, system isconsistaent with the assumptionsz of the accident analyses. Laboratory testing of the carbon adsorber is peuformed i 2accordnance With ASTM 133803-1989 With a-n acce-ptance criteria that isdetermined by applying a miimm aftyfactor Of 2-to the c-har-coal filter remova~l efficiencGy credited in the design b~asis doasa analysis as, specified in Generic Leatter 99-0:2.

The operability of the Fuel Handling Area Ventilation System during movement of irradiated fuel ensures I that a release of fission product radioactivity within the Fuel Handling Building will not exceed the guidelines and dose calculations described in Reg. Guide 1.183, Alternative Radiological Source Term for Evaluating Design Basis Accidents at Nuclear Power Reactors.!- building eX.-haus.t flA is pr-cessed through the HEPA*-charo 1.filter train W..hnever a. Fuel Handling Accident is possible. This will minimiZe oftife do-ses following the postulated Fuel Handling Accaid-ent.

SALEM - UNIT 2 B 314 9-4 Amendment No. 226

LR-N02-0244, Request for Changes to Technical Specifications Refueling Operations - Relaxation of Requirements Applicable During Movement of Irradiated Fuel

Text

Navigation menu

LR-N02-0244, Request for Changes to Technical Specifications Refueling Operations - Relaxation of Requirements Applicable During Movement of Irradiated Fuel

Text

Navigation menu

Search