NL-03-039, Reply to Request for Additional Information Proposed Amendment for Selective Adoption of Alternate Source Term
| ML030650319 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/04/2003 |
| From: | Dacimo F Entergy Nuclear Northeast |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NL-03-039 | |
| Download: ML030650319 (30) | |
Text
Entergy Nuclear Northeast Indian Point Energy Center 295 Broadway, Suite 1 P0 Box 249 Buchanan, NY 10511-0249 Tel 914 734 5340 Fax 914 734 5718 Fred Dacamo Vice President, Operations March 4, 2003 NL-03-039 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop O-P1 -17 Washington, DC 20555-0001
SUBJECT:
Indian Point Nuclear Generating Unit No.3 Docket No. 50-286 Reply to Request for Additional Information Regarding Proposed Amendment for Selective Adoption of Alternate Source Term Reference
- 1. Entergy letter IPN-02-044 to NRC; Proposed Changes to Technical Specifications for Selective Adoption of Alternate Source Term,"
dated June 5, 2002.
Dear Sir:
This letter provides additional information requested by the NRC regarding the proposed license amendment request submitted in Reference 1. The additional information was discussed with NRC staff during conference calls on February 6 and 27, 2003.
The information provided by this submittal is summarized below:
- 1. Commitment by Entergy Nuclear Operations, Inc (ENO) to perform tracer gas testing of the Indian Point 3 (IP3) control room envelope.
- 2. Explanation of the 1800 cfm control room ventilation inleakage assumption used in the dose analysis for the fuel handling accident.
- 3. Revise markup pages to reflect rewording of some of the technical specification changes originally proposed in Reference 1.
The requested information is provided in Attachment I and updated technical specification markup pages are provided in Attachment II. Attachment II also includes the proposed changes to the Technical Specification Bases, for information.
The revised wording of the proposed technical specifications remains consistent with the change description for the original request. The portion of the original request involving the relocation of certain requirements (involving fuel storage building ventilation) to a licensee controlled document is being withdrawn. A description of the requested rewording for all 0/
NL-03-039 Docket 50-286 Page 2 of 2 affected technical specifications is included in Attachment I. The additional supporting information provided in this submittal and the revised wording of the technical specifications do not alter the conclusions of the no significant hazards evaluation provided in Reference 1.
The changes provided in this submittal result in the addition of two new commitments and the withdrawal of a previous commitment. The commitments associated with this license amendment request are provided in Attachment Ill. If you have any questions or require additional information, please contact Mr. Kevin Kingsley at 914-734-5581.
1 declare under penalty of perjury that the foregQing is true and correct. Executed on
~Fred R. Dacimo :y Vice President, Operations Indian Point Energy Center cc:
Mr. Patrick D. Milano, Senior Project Manager Project Directorate I, Division of Reactor Projects 1/11 U.S. Nuclear Regulatory Commission Mail Stop 0 8 C2 Washington, DC 20555 Mr. Hubert J. Miller Regional Administrator Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector's Office Indian Point Unit 3 U.S. Nuclear Regulatory Commission P.O. Box 337 Buchanan, NY 10511 Mr. William M. Flynn New York State Energy, Research and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399 Mr. Paul Eddy New York State Dept. of Public Service 3 Empire Plaza Albany, NY 12223
ATTACHMENT I TO NL-03-039 RESPONSE TO NRC QUESTIONS REGARDING PROPOSED LICENSE AMENDMENT REQUEST FOR SELECTIVE ADOPTION OF ALTERNATE SOURCE TERM ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
NL-03-039 Attachment I Page 1 of 3 Question 1:
The revised fuel handling accident submitted with the license amendment request includes an assumption regarding the rate of unfiltered air inleakage into the control room. Establish a commitment to perform tracer gas testing to further support this assumption.
Response 1:
ENO has established an action plan regarding inleakage measurement and sealing of the control room envelope. ENO is currently evaluating the budget and schedule aspects associated with implementing this action plan. Based on a preliminary schedule for performing this testing to support a planned power uprate project at Indian Point 3, the initial testing would be completed prior to or during refueling outage 3R1 3, in the Spring of 2005. As a minimum, ENO commits to perform the tracer gas testing on a schedule that meets the requirements of the generic communication to be issued by the NRC on this subject.
Question 2:
The licensee dose analysis assumes an unfiltered air inleakage of 1800 cubic feet per minute (cfm) into the control room. Provide additional explanation of this assumption.
Response 2:
ENO considers the assumption of 1800 cfm to be reasonable because the dose analysis applies the inleakage value at the same time that the normal ventilation supply fans are providing 1500 cfm of unfiltered outside air for a total of 3300 cfm during the first 20 minutes following the initiating event. ENO verifies, in accordance with Technical Specification surveillance requirements that one ventilation train can maintain a slight positive pressure in the control room with a supply flow rate of < 400 cfm. Although positive pressure alone may not preclude localized inleakage, the 1500 cfm supply air should tend to offset the potential for inleakage from adjacent areas.
The dose analysis is also conservative in the use of the 20-minute assumption for securing the normal unfiltered supply air and establishing the filtered supply air. Direct communication between the control room and the refueling cavity manipulator crane is established whenever changes in core geometry are taking place (FSAR Section 9.5.3). As a result control room personnel would be promptly notified of a fuel handling accident so that placing control room ventilation in the emergency mode could be accomplished well within the 20 minutes assumed in the analysis.
In addition, this assumption is consistent with other inleakage assumptions used in similar analyses for other plants (Docket 50-400 and 50-280 / 281) pending the resolution of generic issues related to control room habitability.
NL-03-039 Attachment I Page 2 of 3 Question 3:
Provide revised markup pages for the requested changes to the Technical Specifications based on the telecon of February 27, 2003 Response 3:
ENO proposes to modify the originally requested technical specification changes as follows:
- a. Withdraw proposed wording in LCO 3.9.3.a, regarding 'equipment hatch opening is capable of being closed.' Adopt TSTF-68, which is applicable to the personnel access door located in the equipment hatch closure plate.
The changes proposed in Reference 1 included the adoption of TSTF-68 for the personnel air locks as stated in item b, below. Based on the February 27, 2003 telecon with NRC staff, ENO is withdrawing the request to apply the 'capable of being closed' terminology to the equipment hatch opening. Instead, ENO will also apply the TSTF-68 change to the personnel access door located in the equipment hatch closure plate. As stated in Reference 1, ENO complies with the 'Reviewers Note' from TSTF-68 regarding the dose analysis and a commitment to implement administrative controls for the prompt closure of air lock doors. This commitment also applies to the personnel access door identified in LCO 3.9.3.a.
- b. No change to previous request to adopt TSTF-68 for the personnel air locks in LCO 3.9.3.b.
- c. No change to previous request to adopt TSTF-312 for use of administrative controls for unisolating containment penetrations in LCO 3.9.3.c.
- d. No change to previous request to delete Technical Specification 3.9.3.d and 3.9.3.e (regarding containment purge and containment pressure relief requirements) and associated surveillances.
As described in Reference 1, these requirements were established to support the NRC confirmatory dose analysis, associated with Amendment 175 (issued July 15, 1997), for the fuel handling accident. The new fuel handling accident analysis, based on alternate source term, replaces the licensing basis analysis established in Amendment 175.
Deleting these requirements is consistent with the new analysis and the Standard Technical Specifications (Reference 2). However, since this proposed change is not based on the adoption of a TSTF, ENO will agree to withdraw this change and resubmit at a later time, if review and approval of the change precludes the balance of this amendment from being approved in time to support the upcoming refueling outage scheduled to begin March 28, 2003.
NL-03-039 Attachment I Page 3 of 3
- e. Modify the previous request regarding the adoption of the TSTF-51 changes to the Applicability of LCO 3.9.3, by including the addition of the term 'recently' irradiated fuel.
ENO previously proposed, in Reference 1 to modify the Applicability of LCO 3.9.3 by deleting "During CORE ALTERATIONS", in accordance with TSTF-51. Since the new analysis of the fuel handling accident submitted for this license amendment request also supports adoption of the term 'recently irradiated fuel' in TSTF-51, ENO will include these changes to the Technical Specification and Bases.
- f.
Withdraw request for relocation to a licensee-controlled document of the requirements in Technical Specifications 3.3.8 and 3.7.13 regarding the fuel storage building emergency ventilation system and associated instrumentation. Instead, the proposed change will adopt TSTF-51, by adding the term 'recently irradiated fuel', as described in item e, to the Applicability of these Technical Specifications and the associated Bases.
REFERENCES:
- 1. Entergy Nuclear Operations, Inc letter to NRC, IPN-02-044, "Proposed Changes to Technical Specifications: Selective Adoption of Alternate Source Term and Incorporation of Generic Changes; TSTF-51, TSTF-68, and TSTF-312", dated June 5, 2002.
- 2. Standard Technical Specifications for Westinghouse Plants; NUREG-1 431, Revision 2.
- 3. TSFT-51, Revision 2, "Revise Containment Requirements During Handling Irradiated Fuel and Core Alterations," NRC approved November 1, 1999.
- 4. TSTF-68, Revision 2, "Containment Personnel Airlock Doors Open During Fuel Movement," NRC approved August 16,1999.
- 5. TSTF-312, Revision 1, "Administratively Control Containment Penetrations," NRC approved August 16, 1999
ATTACHMENT IITO NL-03-039 TECHNICAL SPECIFICATION AND BASES MARKUP PAGES FOR PROPOSED LICENSE AMENDMENT REQUEST FOR SELECTIVE ADOPTION OF ALTERNATE SOURCE TERM ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
FSBEVS Actuation Instrumentation 3.3.8 3.3 INSTRUMENTATION 3.3.8 Fuel Storage Building Emergency Ventilation System (FSBEVS) Actuation Instrumentation LCO
3.3.8 APPLICABILITY
FSBEVS manual and automatic actuation instrumentation shall be OPERABLE.
17-czntt During movement of irradiated fuel in the fuel storage building.
A T5*TF "IT I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
Manual or automatic A.1 Place FSBEVS in Immediately FSBEVS actuation operation.
instrumentation inoperable.
DR A.2 Suspend movement of Immediately irradiated fuel in the fuel storage building.
"TSTF
- sI INDIAN POINT 3 Amendment 205 3.3.8-1
FSBEVS 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Fuel Storage Building Emergency Ventilation System (FSBEVS)
LCO 3.7.13 APPLICABILITY:
FSBEVS shall be OPERABLE.
During movement of irradiated fuel assemblies in the fuel storage building.
A
-TS-iF 51 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
FSBEVS inoperable.
A.1 Suspend movement of Immediately irradiated fuel assemblies in the fuel storage building.
-Tý'1 F SIT INDIAN POINT 3 Amendment 205 3.7.13-1
Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status:"
- a.
The equipment hatch closed and held in place by at least four bolts or the equipment hatch opening is closed using an
-*F-GI?
equipment hatch closure plate that may include a-e e
~
personnel access door:,
k0
- b.
One door in each air lock F-6B
- c.
Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
- 1. closed by a manual or automatic isolation valve, a blind flange, or equivalent, or
-Xt\\s-Qt A0 TC zq-t r r\\.X-kt VQ--%V ZAS r5T F APPLICABILITY:
- 2.
capable of being closed by OPERABLE Containment Purge and Pressure Relief Isolation System.
(Du*~ýCRE_
-AýLTERAýTIONS__,
During movement of irradiated fuel assemblies within containment.
A re-UAtii INDIAN POINT 3 Amendment 205 3.9.3-1
Containment Penetrations 3.9.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more containment A
spendE(ORE penetrations not in IOS required status.
ASuspend movement of Immediately irra di ated fuel assemblies within containment.
-rs-rF 3c?,.3-1 NOTE Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.
INDIAN POINT 3 3.9.3-2 Amendment 205
Containment Penetrations 3.9.3 SURVEILLANCE REOUIREHENTS SURVEILLANCE SR 3.9.3.1 Verify each required containment penetration is in the required status.
............ OT.......--...
Not required to m )t the reactor has b en t
.......!*...:.:.! !.!ou................
Verify tainment Purge System is ither:
- a.
closeýd a manual or automatic i Qlation valve, bli fla nge, or equivalent, r
- b.
aligned to disch rxge through the HEPA filters and charcoMl adsorbers.
FREQUENCY 7 days SR 3.9. 3.\\2-Verify each required containment purge system 92 days valve actuates to the isolation position on an actual or simulated actuation signal.
SR 3.9.3...
NOTE.........
Not required t be met if the re tor has been "subcritical for.*
50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.
INDIAN POINT 3 3.9.3-3 SURVEILLANCE REQUIREMENI I
i II Amendment 205
INSERTS FOR PROPOSED CHANGES TO TECHNICAL SPECIFICATION BASES Insert A:
involving handling recently irradiated fuel Insert B:
Due to radioactive decay, the FSBEVS instrumentation is only required to be OPERABLE during fuel handling involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />).
Insert C:
Due to radioactive decay, FSBEVS is only required to isolate during fuel handling accidents involving handling recently irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />).
Insert D:
c) the use of administrative controls to ensure prompt closure of any containment openings with direct access from the containment atmosphere to the outside atmosphere Insert E:
The containment personnel airlock doors and the personnel access door in the equipment hatch closure plate may be open during movement of irradiated fuel provided that at least one door in each opening is capable of being closed in the event of a fuel handling accident. In addition, the LCO is modified by a Note allowing penetration flow paths with direct access from the containment atmosphere to the outside atmosphere to be unisolated under administrative controls. Administrative controls, consistent with Appendix B of Regulatory Guide 1.183 (Reference 3), are required to assure that, in the event of a fuel handling accident inside containment, at least one door in each personnel access opening will be closed following an evacuation of containment, and penetration flow paths unisolated under administrative control will be promptly closed. The administrative controls assure that:
- 1. appropriate personnel are aware of the open status of the doors and penetration flow paths during movement of irradiated fuel assemblies within containment, and
- 2. specified individuals are designated and readily available to direct and perform isolation of affected openings in the event of a fuel handling accident, and
- 3. any obstructions (e.g., cables and hoses) that would prevent rapid closure of an open flow path can be quickly removed. Any cables or hoses to be disconnected should not be supplying services that support personnel safety (e.g., breathing air), and
- 4. during fuel handling operations and core alterations, ventilation system and radiation monitor availability should be assessed with the goal of minimizing the potential for radioactive releases, following a potential accident, even further below that provided by the natural decay that occurs following reactor shutdown.
The administrative controls must also be consistent with any pertinent assumptions in the dose analysis for the fuel handling accident. Note that the Indian Point 3 Final Safety Analysis Report (Reference 2) specifies that "No movement of irradiated fuel in the reactor is made until the reactor has been subcritical for at least 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />." Therefore, the FSAR prohibits movement of any fuel that can be classified as "recently irradiated."
Insert F:
However, if personnel access doors or containment penetration flow paths are unisolated during any movement of irradiated fuel assemblies in containment, administrative controls are established to ensure prompt closure of these openings in the event of a fuel handling accident.
FSBEVS Actuation Instrumentation B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Fuel Storage Building Emergency Ventilation System (FSBEVS) Actuation Instrumentation BASES BACKGROUND The FSBEVS ensures that radioactive materialsfin the fuel building atmosphere following a fuel handling accident are filtered and adsorbed prior to exhausting to the environment.
The system is described in the Bases for LCO 3.7.13, Fuel Storage Building Emergency Ventilation System (FSBEVS).
The system initiates filtered ventilation of the fuel storage building automatically following receipt of a high radiation signal from fuel storage building area radiation monitor, R-5.
High radiation levels detected by the fuel storage building area radiation monitor, R-5, initiates fuel storage building isolation and starts the FSBEVS.
These actions function to prevent exfiltration of contaminated air by initiating filtered ventilation, which imposes a negative pressure on the fuel storage building.
Following an Area Radiation Monitor (R-5) signal or local manual actuation to the emergency mode of operation, the FSBEVS ventilation supply fans stop automatically and the associated ventilation supply dampers close automatically.
The charcoal filter face dampers (inlet and outlet dampers) open automatically, if not already open.
Additionally, the rolling door closes, if open, and the inflatable seals on the man doors and rolling door are actuated.
The FSB exhaust fan continues to operate.
APPLICABLE SAFETY ANALYSES A
The FSBEVS ensures that radioactive materials in thenf el storage building atmosphere following a fuel handling accident are filtered and adsorbed prior to being exhausted to the environment when the FSBEVS is aligned and operates as described in the Bases for LCO 3.7.13, Fuel Storage Building Emergency Ventilation (continued)-
INDIAN POINT 3 Revision 0 B 3.3.8 - 1
FSBEVS Actuation Instrumentation B 3.3.8 BASES APPLICABLE SAFETY ANALYSES (continued)
System (FSBEVS).
This action reduces the radioactive content in the fuel building exhaust following a LOCA or fuel handling accident so that offsite doses remain within the limits specified in 10 CFR 100 (Ref. 1).
The FSBEVS actuation instrumentation satisfies Criterion 3 of 10 CFR 50.36.
The LCO requirements ensure that instrumentation necessary for local manual and automatic actuation of the FSBEVS is OPERABLE.
Manual and automatic FSBEVS actuation instrumentation consists of one channel of Fuel Storage Building Area Radiation Monitor (R-5) and one channel of manual actuation.
Manual actuation from the fan house and automatic FSBEVS actuation instrumentation are Operable when both the Fuel Storage Building Area Radiation Monitor (R-5) signal and manual initiation will cause the realignment of the FSBEVS to the accident mode of operation as described in the Bases for LCO 3.7.13, Fuel Storage Building Emergency Ventilation System (FSBEVS).
The setpoint for Fuel Storage Building Area Radiation Monitor (R-5) is established in accordance with the FSAR (Ref. 2).
APPLICABILITY The manual FSBEVS initiation must be OPERABLE when moving~irradiated fuel assemblies in the fuel storage building, to ensure the FSBEVS operates to remove fission products associated with leakage after a fuel handling acciden High radiation itiation of the FSBEVS must be OPERABLE in any MODE during movement o irradiated fuel assemblies in the fuel storage building to ensure automatic initiation of the FSBEVS when the potential forafuel handling accident exists.
(continued)-
INDIAN POINT 3 LCO B 3.3.8 - 2 Revision 0
FSBEVS Actuation Instrumentation B 3.3.8 BASES ACTIONS The most common cause of channel inoperability is outright failure or drift of the bistable or process module sufficient to exceed the tolerance allowed by Reference 2.
Typically, the drift is found to be small and results in a delay of actuation rather than a total loss of function.
This determination is generally made during the performance of a COT, when the instrumentation is set up for adjustment to bring it within specification.
If the Trip Setpoint is less conservative than the tolerance specified by Reference 2, the channel must be declared inoperable immediately and the appropriate Condition entered.
A.1 and A.2 This condition applies when the manual or automatic FSBEVS initiation capability is inoperable.
The Required Action is to immediately place the system in operation as described in the Bases for LCO 3.7.13, FSBEVS.
This accomplishes the actuation instrumentation function that may have been lost and places the unit K in a accident mode of operation.
Alternatively, movement of
ýirradiated fuel assemblies in the fuel storage building must be suspended immediately to eliminate the potential for events that could require FSBEVS actuation.
The Completion Time of immediately requires that the Required Action be pursued without delay and in a controlled manner.
SURVEILLANCE REQUIREMENTS SR 3.3.8.1 Performance of the CHANNEL CHECK once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ensures that a gross failure of instrumentation has not occurred.
A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.
A CHANNEL CHECK for a single channel instrument is satisfied by verification that the sensor or the signal processing equipment has not drifted outside its limit.
(continued)
INDIAN POINT 3 Revision 0 B 3.3.8 - 3
FSBEVS B 3.7.13 BASES BACKGROUND (continued) safe into the required emergency mode position.
Note that the inflatable seals on man doors and truck door are not required for maintaining the FSB at these required post accident conditions.
A push button switch adjacent to the 95' elevation door leading to the Fan House allows the Fuel Storage Building Exhaust Fan to be momentarily shut down and air removed from the man door seal to allow the door to be opened for FSB ingress or egress when in the emergency mode of operation. The fan will automatically restart and the door is resealed after a preset time has elapsed (approximately 30 seconds).
The FSBEVS is discussed in the FSAR, Sections 9.5, and 14.2 (Refs. 1 and 2, respectively).
APPLICABLE SAFETY ANALYSES The FSBEVCS design basis is established by the consequences of the limitina Design Basis Accident (DBA),
which is a fuel R
fLT handling accident The analysis for a fuel handling accident assumes that the FSB exhaust fan can maintain the FSB at a slight negative pressure (i.e., < -0.125 inches water gauge) with respect to atmospheric pressure with the exhaust flow rate < 20,000 cfm.
Under these conditions, all FSB ventilation exhaust is assumed to be directed through the roughing filters, HEPA filters, and charcoal filters and is released to the environment via the plant vent.
This ensures that offsite post accident dose rates are wlin requi red XC limitsh AltsoL 1
requires the ABILITY ofhe FSBEVS wh i d sr i ated fue femblies are belhmoved wihi the FSB, Sanalys'is indica ts that offs*e post accident l*se rates wi l be#
hinre d
swihout tw peration of FSBEVS if the
".'It lat 1 has ha'i.
continuou'-45 day decay period.* This analysis is decied in Reference Z The FSBEVS satisfies Criterion 3 of 10 CFR 50.36.
(continued)_
INDIAN POINT 3 B 3.7.13 - 3 Revi si on 0
FSBEVS B 3.7.13 BASES LCO This LCO requires that the Fuel Storage Building Emergency Ventilation System is OPERABLE and the FSB boundary is intact.
This ensures that the required negative pressure is maintained in the FSB and FSB ventilation exhaust is directed through the roughing filters, HEPA filters, and charcoal filters and is released to the environment via the plant vent.
Failure of the FSBEVS or the FSB boundary could result in the atmospheric release from the fuel storage building exceeding the 10 CFR 100 (Ref. 3) limits in the event of a fuel handling acciden
_A The FSBEVS is considered OPERABLE when the individual components necessary to control exposure in the fuel storage building are OPERABLE.
FSBEVS is considered OPERABLE when its associated:
- a.
Exhaust fan is OPERABLE;
- b.
Roughing filter, HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration function;
- c.
Ductwork and dampers are OPERABLE as needed to ensure air circulation can be maintained through the filter;
- d.
Ventilation supply fan trip function and ventilation supply isolation dampers closure function are OPERABLE or secured in incident position; and
- e.
FSBEVS charcoal filter bypass dampers are closed and leak tested.
The inflatable seals on man doors and truck door are not required for maintaining the FSB at these required post accident conditions.
Additionally, the FSBEVS is not rendered inoperable when the FSBEVS exhaust fan is momentarily shut down and air removed from the door seal to allow the door to be opened for FSB ingress or egress when in the emergency mode of operation.
Requirements for the OPERABILITY of the Area Radiation Monitor (R-5) and associated instrumentation that initiates the FSBEVS are addressed in LCO 3.3.8, "Fuel Storage Building Emergency Ventilation System Actuation Instrumentation."
(continued)
INDIAN POINT 3 B 3. 7. 13 - 4 Revision 0
FSBEVS B 3.7.13 BASES LCO Requirements for leak testing the FSBEVS charcoal filter bypass (continued) dampers following closure are governed by the IP3 FSAR.
APPLICABILITY During movement of irradiated fuel in the fuel storage building, the FSBEVS is requiredto be OPERABLE to mitigate the consequences ofX fuel handling accident.
ACTIONS A.1 MCn t When the FSBEVS is inoperable during movement offirradiated fuel assemblies in the fuel storage building, action must be taken to place the unit in a condition in which the LCO does not apply.
Action must be taken inmnediately to suspend movement of irradiated fuel assemblies in the fuel storage building.
This does~not preclude the movement of fuel to a safe position.
're-tA SURVEILLANCE REQUIREMENTS SR 3.7.13.1 This SR requires periodic verification that the FSBEVS charcoal filter bypass dampers are installed and leak tested.
This SR is performed by a visual verification that the bypass dampers are installed and an administrative verification that required leak testing was performed following the last installation of the dampers. Requirements for leak testing the FSBEVS charcoal filter bypass dampers following closure are governed by the IP3 FSAR.
This SR is performed prior to movement ofAirradiated fuel assemblies in the fuel storage building, and once per 92 days thereafter.
The 92 day Frequency is appropriate because the bypass dampers are operated under administrative controls which provides a high degree of assurance that the dampers will remain in the required position.
This Frequency has been shown to be acceptable through operating experience.
(continued)
INDIAN POINT 3 B 3.7.13 - 5 Revision 0
Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations BASES BACKGROUND During lRE ýALE ERATIONS or movement ofjrradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be restricted from escaping to the environment when the LCO requirements are met.
In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, "Containment."
In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent.
The LCO requirements are referred to as "containment closure" rather than "containment OPERABILITY."
Containment closure means that all potential escape paths are closed, except for the OPERABLE Purge System Penetration.
Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.
The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation ýexsures are maintained well within the Io(.FZSo,,
requirements o 1
O(Ff-Additionally, the containment provides LAD
-TV 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 and out of containment.
During CO E ALTERATIONS or CQ
+Ayd movemento-ofglirradiated fuel assemblies within containment, the equipment hatch must be held in place by at least four bolts.
Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced.
In lieu of maintaining the equipment hatch in place for containment closure, a temporary closure device may be used to maintain containment closure during core (continued)
INDIAN POINT 3
,,i[,
Revision 0 B 3.9.3 - 1
Containment Penetrations B 3.9.3 BASES BACKGROUND (continued) movement ofirradiated fuel assemblies within containment.
The temporary closure device may provide-penetrations for temporary services or personnel access.
The temporary closure device will be designed to withstand a seismic event and designed to withstand a pressure which ensures containment closure during refueling operations.
The closure device will provide the same level of protection as that of the equipment hatch for the fuel handling accident by restricting direct air flow from the containment to the environment.'--*
'!-,,t R
The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "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, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.
During CORE A movement of recA.I..5 irradiated fuel assemblies within containment, containment closure is required; therefore, the door interlock mechanism may remain disabled, but one air lock o must always remain closed.
The requirements for cont a n' e
trion closure ensure that a release of f_ sion product radioactivity within containment will be trictedN m escapi to-e nvironmnt.
Ti osur:e restricions ar sufficiNt to re'Sbkrict fi ion pruct r*doacti
- ty rele e fromesntainme due to fuel handl'ng
~cci~
,hit durg refu ing.
The Containment Purge System consists of the 36-inch containment purge supply and exhaust ducts.
The supply system includes roughing filters, heating coils, fan and a containment penetration with two butterfly valves for isolation.
The exhaust system includes a containment penetration with two butterfly valves for isolation and can be aligned to discharge to the (continued)
INDIAN POINT 3 Revision 0 B 3.9.3 - 2
Containment Penetrations B 3.9.3 BASES BACKGROUND atmosphere through the plant vent either directly or through the (continued)
Containment Purge Filter System (i.e., a filter bank with roughing, HEPA and charcoal filters).
The Containment Purge System must be isolated when in Modes 1, 2, 3 or 4 in accordance with requirements established in LCO 3.6.3, Containment Isolation Valves.
In Modes 5 and 6, the Containment Purge System may be used for containment ventilation.
When open, the Containment Purge System isolation valves are capable of closing in response to the detection of high radiation levels in accordance with requirements established in LCO 3.3.6, Containment Purge and PressureRelief Isoato Isrmnain(e
.J'Dei io~ati on cApbl hC
- nanetP g ytmmust be gned hag-brg t'mnt~u iter S'y em durin CORE mini nu shtdr fo *sp The Containment Pressure Relief Line (i.e., Containment Vent) consists of a single 10-inch containment vent line that is used to handle normal pressure changes in the Containment when in Modes 1, 2, 3 and 4 (Ref.
1).
The Containment Pressure Relief Line is equipped with three quick-closing butterfly type isolation valves, one inside and two outside the containment which isolate automatically in accordance with requirements established in LCO 3.3.2, "Engineered Safety Feature Actuation System (ESFAS)
Instrumentation",
and LCO 3.3.6, "Containment Purge System and Pressure Relief Line Isolation Instrumentation." -h Containment Pressure Relief Line discharges to the atmosphere via the Containment Auxiliary Charcoal Filter System i.e., a filter bakhhrugig EAandl charcal pirtssu)e cagsithe oontainment we nMds1 P, 3ad4(e.1.TeCnanetPessure Relief Lie usean*oaerOEAEAiNs o
aqusppifed mini umre numbe-cofin hoursefl hye Contatinont Pressu, eoefne mustd rn awo otisoed beauehe Containment Auxich ryoat rco al te ayst e it red to q
eme tested insccordance witssure S eiefi LineIsolation
- 5.
V nst rntFitr Tt Prog m ed Continmnt Pessre Rlie Lin dichares o th (ontinued)i INDIAN POINT 3 B 3.9.3 - 3 Revision 0
Containment Penetrations B 3.9.3 or r'yqj bQ-wýsalAO U"rAQY o~riitmcb"k QLol BASES BACKGROUND The other contai m.t penetrations that provide direct access (continued) from containment atm sphere 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 must be approved - ir-eee 4d4-n*i h
0 CFR 50.5&9 and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during fuel movements.
APPLICABLE SAFETY ANALYSES During CRE ALTERAT,ON,
movement ofAirradiated fuel assemblies within containment, the most severe radiological consequences result w rz,-
- fl rom a fuel handling accidentý The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 2).
Fuel handling accidents, analyzed in Reference 2, include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The release of radioactivity from the containment following a fuel handling accident is limited by the following:
a)
The requirements of LCO 3.9.6, "Refueling Cavity Water Level;"
b)
The minimum decay time of 44-hours prior to CORE ALTERATIONS;
- and, c)
The requirements of this LCO to either isolate the C tainment Pi.r, ge System or lign the sysmt discharge hrough e HEPA hours llowing the actor shutdo SThis c
bination f requiremen s ensures tha
.the release o xfission
- product r iatvi *, subsequen *to a fuel ha dling accident,*
esults in
- ses that xre well wit in the guidel ine values specified i10 CFR 100.* StandarRiew Plan,mSection 15.7.4 Rev. 1 (Ref.ye 3),efines "wea within"0 CFR 100 tb be 25% or le of the 10 CFR (conti nued)
INDIAN POINT 3 B 3.9.3-4 Revision 0
Containment Penetrations B 3.9.3 BASES 1
APPLICABLE SAFETY ANALYSES (continued)
Containment penetrations satisfy Criterion 3 of 10 CFR 50.36.
LCO
_rrNSUVT - E APPLICABILITY This LCO limits the consequences of a fuel handling accidentýin EU containment by limiting the potential escape paths for fission product radioactivity released within containment.
The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge system penetrations.
For the OPERABLE containment purge system penetrations, this LCO ensures that these penetrations are isolable by the Containment Purge isolation i srmnai
'1:
1
,t equre ment 1
iso ate t e ontainment Purge S ste align the system to disc rge through the HE filters an arcoal absorbe for a minimum o the first 550hu olwn he atrsutwn r~equired to 11 "t ofst a
'nepoueowti euired limits.
The
\\
`Pr-Rean 1*solated because theS
ýcontainment P*ssure Relief L m
eut reman otebcus th Coainment Aux)i~ir Charcoal ter System i rbrqie ob Co tm is n~o *required to be tes in accordanice ith Specifica*on 5.5.10, Ven lation Filter Test Pro4 am.
The OPNEAILITY requir ents for this L ensure that the autom e syste valves meet assumptions us in the afety analysito ensure thhtreleases th ugh the valves are fi red and can terminated, uch that radiological doses are within he acceptanc' limit.
The containment penetratii requirements are applicable during-GGRE ALTERATIONS-or movement of irradiated fuel assemblies within
\\i-,'
containment because this is when there is a potential for fuel handling accident.
In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1.
In MODES 5 r
and 6, whenCORE ALTERATIONS at movement ofirradiated fuel" assemblies within containment not being conducted, the (continued)
INDIAN POINT 3 B 3.9.3 - 5 Revision 0
Containment Penetrations B 3.9.3 BASES APPLICABILITY potential forNYfuel handling accident does not exist.
(continued)
Therefore, under these conditions no requirements are placed on containment penetration status.
'niL-*
- ?
- CUZ, ACTIONS A.1 If the containment equipment hatch, air locks, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Purge system isolation instrumentation not capable of automatic actuation when the purge and exhaust valves are open, the unit must be placed in a condition where the isolation function is not needed.
This is accomplished by immediately
- l, suspending -6.R.-A
.E. NS..
an-A movement offlirra laRed fuel assemblies within containment.
Performance of these actions shall not preclude completion of movement of a component to a safe position.
SURVEILLANCE REQ UIREMENTS 1.3 (s.IA6 C-10SA~ of, Mq~
ý
-lt SR 3.9.3. 1 This Surveil ance demonstrates that each of the containment penetrations "cgui' b
s i
iticn i-in t1ht
-post-t-on, The Surveillance on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing.
Also the Surveillance will demonstrate that each valve operator has motive power, which will ensure that each valve is capable of being closed by an OPERABLE automatic containment purge and exhaust isolation signal.
A%
A
) The Surveillance is performed
-L4y-7 daysf[
T
-e.~movement of irradiated fuel assemblies within containment.
The Surveillance interval is selected to be commensurate with te-nerma!
"-a*r-on-o A
surveillance before the start of refueling operations will-pr-cvid Lwa L~i~e srvci~anc r~6
~no (continued)
INDIAN POINT 3 B 3.9.3 - 6 Revision 0
Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS SR 3.9.3.1 (continued) vci-if i during the applicable period for this LCO.
As such, this Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the containment will not result in a release of fission product radioactivity to the environment.
SR 3.9.3.2 requsiret obe puerormed or ts isf t rer holat e ben sualbcrtic for > 550 h rs.
These restri ions ensure that t offsite dose limit for a 1e handling accidestof 75 rem to the yroid at the erusion area undary (i.e., 25 pC cent of the 10 CFR\\Cart 100 1yi t of 300 rem)ins ment by either fitring any releaseK'from the contu ment or by da greater denau time before fuelhnhandling duriin arefuln opermitios.vey2.onhTHANLCLIRTO S R 3.9.3\\2 This Surveillance demonstrates that each containment purge and exhaust valve actuates to its isolation position on an actual or simulated high radiation signal.
The 92 day Frequency ensures that this SR is performed prior to this function being required and periodically thereafter. In LCO 3.3.6, the Containment Purge system isolation instrumentation requires a CHANNEL CHECK every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and a COT every 92 days to ensure the channel OPERABILITY during refueling operations. Every 24 months a CHANNEL CALIBRATION is performed. SR 3.6.3.5 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements.
(conti nued)
INDIAN POINT 3 Revision 0 B 3.9.3 - 7
Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS SR 3.9.3.\\ (continued)
These Surveillances performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment.
,SR 3.9.3.4 iettedo Thi SR verifies t the requiredCo tainment Buildin\\ Purge System testin is performed i accordance with pecification 5S5 10, Ventilation Filter Test rogram (VFTP).
e VFTP includes esting HEPA filte>NPerformance, arcoal adsorber ficiency, minim system flow rkte, and the p sical properties f the activated\\
"arcoal (gener'al use and fol1 wing specific op* ations).
Specific te frequencies\\ ad additional 'formation are di cussed in detail in th VFTP.
REFERENCES
- 1.
FSAR, Section 5.3.
- 2.
FSAR, Section 14.2. --NRE6-O*8O, Section 15.7.1, Rcy. 1, July 1981.
- 4.
10 CFR 50 Appendix A, "General Design Criteria," Criterion 19, Control Room.
LitkcA-O fMrr*r X-,k INDIAN POINT 3 B 3.9.3 - 8 Revision 0
ATTACHMENT III TO NL-03-039 COMMITMENTS REGARDING PROPOSED LICENSE AMENDMENT REQUEST FOR SELECTIVE ADOPTION OF ALTERNATE SOURCE TERM ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
NL-03-039 Attachment III Page 1 of 1 COMMITMENTS REGARDING PROPOSED LICENSE AMENDMENT REQUEST FOR SELECTIVE ADOPTION OF ALTERNATE SOURCE TERM - FUEL HANDLING ACCIDENT Number Commitment Comment / Schedule IPN-02-044-1 ENO will establish administrative controls to Commitment from ensure prompt closure of containment openings original amendment in the event of a fuel handling accident in the request maintained.
containment building.
NL-03-039-1 ENO commits to the completion of tracer gas On a schedule that testing of the control room envelope to meets the requirements determine the flow rate of unfiltered air of the generic inleakage.
communication to be issued by the NRC on this subject NL-03-039-2 ENO commits to implement the guidelines of Prior to implementation NUMARC 93-01, Revision 3, Section 11.2.6, of the requested consistent with the 'Reviewer's Note' contained amendment.
in TSTF-51, Revision 2.
WITHDRAWN ------
IPN-02-044-2 ENO will relocate the requirements of Technical Commitment no longer Specifications 3.3.8 and 3.7.13 to the Technical applies. Proposed Requirements Manual.
amendment no longer requests relocation of these requirements.