NLS2006017, Response to Request for Additional Information Fuel Handling Accident-Alternative Source Term Amendment
| ML061020078 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 04/07/2006 |
| From: | Edington R Nebraska Public Power District (NPPD) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NLS2006017, TAC MC8566 | |
| Download: ML061020078 (31) | |
Text
Nebraska Public Power District Always there when you need us NLS2C06017 April 7, 2006 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555-0001
Subject:
Response to Request for Additional Information Re: Fuel Handling Accident --
Alternative Source Term Amendment Cooper Nuclear Station, Docket No. 50-298, DPR-46
References:
- 1. Letter from U.S. Nuclear Regulatory Commission to R. Edington (Nebraska Public Power District), dated February 9, 2006, "Cooper Nuclear Station -
Request For Additional Information Re: Alternate Source Tern For Reevaluation of the Fuel Handling Accident Dose Consequences (TAC No.
MC8566)."
- 2. Letter from R. Edington (Nebraska Public Power District) to U. S. Nuclear Regulatory Commission, dated September 29, 2005, "License Amendment Request for Application of the Alternative Source Term for Reevaluation of the Fuel Handling Accident Dose Consequences" (NLS2005075).
- 3. Letter from R. Edington (Nebraska Public Power District) to U. S. Nuclear Regulatory Commission, dated January 16, 2006, "Revised and Supplemental Pages to License Amendment Request for Application of the Alternative Source Term for Reevaluation of the Fuel Handling Accident Dose Consequences" (NLS2006002)..
The purpose of this letter is for the Nebraska Public Power District to respond to the Request for Additional Information (RAI) provided in Reference 1 by the Nuclear Regulatory Commission regarding the previously submitted License Amendment Request of Reference 2, as supplemented by Reference 3. Please find the RAI responses in Attachment 1, as well as associated enclosures.
Should you have any questions concerning this matter, please contact Paul Fleming, Licensing Manager, at (402) 825-2774.
COOPERNUCLEARSTATION P.O. Box 98/ Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com
NLS2006017 Page 2 of 2 I declare under penalty of perjury that the foregoing is true and correct.
Executed on 41 70CD (Date)
Sincerely, i
( k. A Ran all K. Edington Vice President - Nuclear and Chief Nuclear Officer
/wv Attachment Enclosures cc: Regional Administrator W/attachment, enclosures US NRC - Region IV Cooper Project Manager w/attachment, enclosures US NRC - NRR Project Directorate IV-1 Senior Resident Inspector w/attachment, enclosures US'NRC - CNS Nebraska Health and Human Services w/attachment, enclosures Department of Regulation and Licensure NPG Distribution w/o attachment, enclosures CNS Records W/attachment, enclosures
NLS2006017 Page 1 of 4 Response to Request for Additional Information Re:
Fuel Handling Accident - Alternative Source Term Amendment Question 1:
The radiological dose analysis submitted in support of this license amendmen f request assumed the Regulatory Guide (RG) 1.183, Table 3, non-loss-of-coolant accident gap fractions. The release fractions listed in RG 1.183, Table 3, are acceptable for use with fiels having a peak burnup up to 62 gigawatt days per metric ton uranium (G WD/MTU) provided that the maximum linear heat generation rate does not exceed the 6.3 kilowatt perfoot peak rod average power for blurnups exceeding 54 GWD/MTU. Please verify that thefiuel used at Cooper Nuclear Station (CNS) meets these criteria.
Response
As discussed on Page 45 of Attachment I to Reference 2 regarding Footnote 1 1, the current core will remain within the constraints of the footnote. This has been confirmed with the fuel vendor. Core reload design control measures will ensure that future cores remain within the burnup limitation. The Nebraska Public Power District (NPPD) is monitoring the current ongoing dialogue between the Nuclear Regulatory Commission (NRC) and the Industry that seeks to modify or eliminate the footnote restriction.
Question 2:
The radiological dose analysis assumed that there is no means of isolating the radiological activity associated with afuiel handling accident (FHA) release within the reactor building (i.e., no credit is taken for maintaining the secondary containment boundary). Please identify all the possible release points for the FHA when the reactor building is open to the environment and the standby gas treatment system and secondary containment are inoperable. Also justify why the reactor building vent release pathway is assumed to bound all these other possible release pathwaysfor the control room dose assessment. Provide a figure, orfigures, showing stnrctures and assunmedpaths of airflow in your response.
Response
As stated on Page 49 of Attachment 1 to Reference 2, the Secondary Containment breach control strategy will not allow maintenance to be performed that compromises the structural integrity of Secondary Containment (e.g., removal of blowout panels, or core drills). Accordingly, NPPD understands this question to refer to airlocks, doors, and hatches that could be open to establish a path outside Secondary Containment. Enclosure 1 provides a revision to the FHA dose calculation that provides a more detailed justification for selecting the Reactor Building vent as the release path (Section 4.1). This calculation revision prov.ides a composite figure showing the configuration of the various openings relative to the Control Room intake (horizontal distance and angle), and a table listing th se openings with the associated dimensional characteristics. In summary, the
NLS2006017 Page 2 of 4 overriding considerations in selecting the Reactor Building vent as providing the bounding release point were: a) the comparatively short distance to the Control Room intake, b) the direct path to the outside environment, c) the relatively large cross-sectional area, and d) a driving force for the release.' However, it was noted that the Reactor Building Roof Hatch is closer to the Control Room intake than the vent, and similarly offers a direct path from the Refueling Floor to the environment. Therefore, as an additional conservatism, the Reactor Building Roof Hatch will be maintained closed during the movement of irradiated fuel in Secondary Containment.
Question 3:
Item 2 of Section 10.4.2 (Main Control Room Air Conditioning and Habitabil.ty Controls - Safety Design Basis) of the CNS Updated Safety Analysis Report (USAR) states that the main control room air-conditioning system and habitability controls isolate the main control room outside air intake on a Group 2 or Group 6primary containment isolation system isolation signal. This statement appears to conflict with CNS USAR Section 10.4.5.2 (Main Control Room Air Conditioning and IHabitability Controls - Control Room Emergency Filter
[CREF] System) which states that the control room envelope remains pressurized during a radiological emergency mode of operation by channeling all outside air through the CREFsystem. Please explain this apparent discrepancy.
Response
Section X-1 0.4.5.2 accurately describes the operation of the CREF System. When the CREF System initiates, the outside air intake damper for the normal unfiltered ventilation supply closes (HV-AOV-270), and the emergency bypass filter supply damper opens (HV-AOV-271). This description is consistent with the Background discussion of Technical Specification Bases B 3.7.4 (CREF System).
A USAR change has been made that provides clarification to USAR Section X,-
10.4.2 Item 2 that "the Main Control Room unfiltered outside air supply" isolates, rather than "the Main Control Room outside air intake" isolates.
Question 4:
The term "recently irradiatedfiuel assemblies" is being introduced in the applicability and action statementsfor several technical specifications (TSs). In some cases (i.e., TSs 3.3.6.2, 3.6.4.1, 3.6.4.2, and 3.6.4.3), this term is defined in the corresponding TS Bases as fuel that has occupied part of a critical reactor core within the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; in other cases (i.e., TSs 3.3.7.1 and 3.7.4), this term is defined in the corresponding TS bases as fiel that has occupied part of a critical reactor core within the previous 7 days. Please explain how using the same term (i.e., recently irradiatedfutel assemblies) with different definitions within the TSs will not cause confusion.
- 1.
As noted in Reference 2 (Attachment 1, page 52), physical interlocks and security concerns prevent both Reactor Building railroad airlock doors from being opened concurrently.
NLS2006017 Attachment I Page 3 of 4
Response
Question 5:
Resporse:
NPPD agrees that having different definitions for "recently irradiated fuel assemblies" could be a source of confusion. Accordingly, the term "lately irradiated fuel assemblies" will be used for the 7-day irradiation period. provides revised pages for TS 3.3.7.1 (CREF System Instrumentation) and TS 3.7.4 (CREF System), and associated TS Bases markups.
A commitment has been made (i.e., Commitment Number NLS2005075-02) to ensure that either a reactor building exhaust fan, a standby gas treatment (SGjT) system fan, or the CREF system is in operation prior to conductingfiuel handl ng operations when less than a 7-day decay time has elapsed. Since the CREF system initiates on high radiation detected in the reactor building exhaust plenum, either a reactor building exhaust fan or an SGTsysten fan must be operational to ensure ventilation flow in the exhaust plentun at the start of the FHA. Please describe the TSs or administrative controls that will be implemented to ensure that either a reactor building exhaustfan, an SGTsystem fan, or the CREF system will be in operationfor the entire 7-day decay time period afterfitel handling operations have begun.
The purpose of commitment NLS2005075-02 was to ensure there was a source of ventilation exhaust airflow at the start of the FHA that would cause CREF System initiation on high radiation in the Reactor Building Exhaust Plenum. If ventilation exhaust airflow were not available, then the CREF System would be started manually prior to the start of lately irradiated fuel movements. In this manner the CREF System safety function credited in the FHA would be met.
In responding to this RAI Question, it became clear that the CREF System Instrumentation Technical Specification (TS 3.3.7.1) already substantially provides the regulatory controls that implement this commitment. TS 3.3.7.1 requires an operable Reactor Building Ventilation Exhaust Plenum Radiation --
High function. The commitment establishes that adequate ventilation exhaust airflow is a support function to the radiation monitor during the movement of lately irradiated fuel. The TS 3.3.7.1 Bases will be revised as provided in to formalize this support feature. In this manner, ventilation exhaust airflow will be a TS Operability requirement for the radiation monitor. Revising the TS Bases will ensure that future changes are made under the provisions of 10 CFR 50.59, in accordance with TS 5.5.10, "Technical Specifications Bases Control Program." In the case of the election to operate with the CREF System manually initiated (in lieu of a running Reactor Building exhaust fan or SGT fan),
TS 3.3.7.1 would prevent the start of fuel movement until Required Action C. I had been completed ("Initiate CREF System"). Station procedures will ensure these Technical Specification-driven prerequisites are met prior to moving lately irradiated fuel.
NLS2006017 Page 4 of 4 With ventilation exhaust flow explicitly controlled within the TS Bases as provided in Enclosure 2, the NLS2005075-02 commitment is subsumed by the requirements of the Technical Specifications. Accordingly, NLS2005075-02 is no longer necessary and is withdrawn.
TS 3.7.4 requires CREF System operability during the movement of irradiated.
fuel assemblies. Under the existing CNS licensing basis, TS 3.7.4 allows continuation of fuel movements for up to a 7 day Completion Time if the CRE F System becomes inoperable after irradiated fuel movement has commenced. The TS Basis for this Completion Time is the low probability of a Design Basis Accident occurring during this time period. Since the probability of a Fuel Handling Accident is not increased by adoption of the Alternative Source Tenn, NPPD has not proposed TS changes that reduce this Completion Time.
Question 6:
The two comnmitments that have been made (i.e., Commitment Numnbers NLS2005075-01 and NLS2005075-02) are to be iniplemnented within 30 days of issuance of the requested license amendment. Could the movement of recently irradiatedfuel assemblies in secondary containment occur prior to the implementation of these two commitments?
Response
Commitment NLS2005075-01 (as well as the commitments made in this response) have been designated to be in place concurrent with the requested 30-day implementation period of the license amendment. That was the intent of the wording in the "Committed Date" column in Reference 2. Thus, movement of irradiated fuel assemblies in Secondary Containment under the Alternative Source Term and associated TS changes will not occur prior to the implementation of these commitments. Commitment NLS2005075-02 has been withdrawn as discussed in the Question 5 response.
NLS2C0601 7 NEDC 05-031 Revision 22
- 2. The Radiological Dose Analysis provided as Attachment 1 of NEDC 05-031 (as submitted in Reference 2) is unchanged by this revision.
IIIrPTACHMENT 1 DESIGN CALCULATION COVER SHEET Page I of 12
Title:
Review of Alion Calculation ALION-CAL-Calculation Number: NEDC 05-031 NPPD-3236-001, Radiological Dose Analysis for a Fuel liandline Accident (FHA) at Cooper Nuclear Station CED/EE Number: EE 06-013 System/Structure: HV, SGT, SC Setpoint Change/Part Eval Number: N/A Component: N/A Discipline: Mechanical Design Classification: [X] Essential; [ ] Non-Essential SQAP Requirements Met? [ ] Yes; [X] N/A Propriet3ry Information Included? [ I Yes; [XI No PURPOSE:
fhis calculation incorporates by attachment Alion Science & Technology Calculation No. ALION-CAL-NPPI)-3236-001, ev. 0, ir. accordance with CNS Engineering Procedure 3.4.7. The calculation determines the dose to a Control Room ccupant and to a person at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) at the Cooper uclear Station site following a design basis Fuel Handling Accident (FHA). The analysis is performed using an klternate Source Term (AST) in accordance with the guidance provided by the NRC in Regulatory Guide 1.133 (July 2000) ant as authorized by 10 CFR 50.67. This calculation has been prepared as a Status 2 calculation for NRC review and will be as-built upon NRC approval.
Rev. 1 in orporates Rev. 1 of Alion's calculation, which was revised primarily to incorporate an additional Case 1 inalysis to determine total Control Room dose assuming 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> fuel decay time before fuel movement and without "REFS operation.
Rev. 2 adds Section 4.1 to evaluate additional potential Secondary Containment release points and upgrades. the
-alculation from Status 2 to Status 3. Alion calculation ALION-CAL-NPPD-323G-001, Rev. 1 is not changed by this evision. Revision 2 approvals are not applicable to the Alion calculation (Attachment 1).
ESULTS AND CONCLUSIONS:
he results are tabulated in Section 5 of Alion's calculation for each of the three (3) receptor locations:
- 1. Control Room,
- 2. Low lPopulation Zone (LPZ), and
- 3. Exclusion Area Boundary (EAB).
11 calculated doses were found to be below the stipulated limits. It is therefore concluded that the regulatory dose limits will not be exceeded following a postulated design basis FHA at Cooper Nuclear Station.
ATTACHMENTS:
- 1. Alion calculation ALION-CAL-NPPD-3236-001, Rev. 1 (including attachments thereto)
- 2. GE Litters REK:99-152 and REK:99-211 (References 6.10 and 6.9 of Alion's calculation) 2 3
Jim Drasler 6-A
? v-7 6
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3 - z Q L.
L 4j.
0 *!5 3)Wipt
%4 1
2 Alion Science & Technology Jim Drasler
- N/A Kyle 9/26/05 9/28/05 Hilgenfeld
- 9/25/05 umber Status Prepared By/Date Reviewed By/Date IDVed By/Date By/Date
- See microfilm for previous signatures Status Codes
- 1. Active
- 4. Superseded or Deleted
- 7. PRA/PSA
- 2. Information Only
- 5. OD/OE Support Only PROCEDURE 3.4.7 l
REVISION 26 PAGE 1 OF31
TTACHMENT 2 DESIGN CALCULATION CROSS-REFERENCE INDEX° 3 Page:
2 _ of 12 NEDC: 05-031 Rev. Number:
2 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM PENDING CHANG3ES NO.
DESIGN INPUTS REV. NO.
TO DESIGN INPIJTS 1
Refer to Attachment 1 listing, page 4 2
B&R Drawing 2060 N11 N/A 3
B&R Drawing 2209 3
N/A 4
B&R Drawing 4003 N31 N/A 5
B&R Drawing 4215 N02 N/A 6
B&R Drawing 4219 N05 N/A 7
B&R Drawing 4222 N01 N/A 8
B&R Drawing 4223 N02 N/A 9
B&R Drawing 4504 N15 DCNs 04-1494. 05-1210 10 B&R Drawing 4506 N08 DCN 06-0167 11 B&R Drawing 4507 N05 N/A 12 B&R Drawing 4535 N11 DCN 04-2654 13 B&R Drawing 4536 N25 DCNs 97-1523, 04-2655, 05-0846, 05-1211 14 6.SC.701 3
N/A 15 Technical Specification 3.4.6.1 Amendment Pending FHA Licensing 178 Amendment 16 NEDC 99-031 5
N/A PROCEDURE 3.4.7 l
REVISION 26 l
PAGE 2 OF(31 2
ZZTTACHMENT 2 DESIGN CALCULATION CROSS-REFERENCE IND'EX 3
Page:
3 of 12 NEDC:
05-031 Rev. Number:
2 Nebraska Public Power District DESIGN CALCULATION CROSS-REFERENCE INDEX ITEM NO.
AFFECTED DOCUMENTS REV. NUMBER 1
NEDC 99-032 3
2 USAR V-3.3 loep.xxl3 3
USAR VII-17 loep.xxl3 4
USAR X-4.5.2.1 loep.xxl3 5
USAR X-10.2.5.2 loep.xxl3 6_
USAR XIV-6 loep.xxl3 7
USAR XIV-7 loep.xxl3 8
USAR Appendix G loep.xxl3 9
TRM T 3.9 loep 07/07/05 10 TRM B 3.9 loep 06/16/05 11 Procedure 0.24 25 12 Procedure 0.50.5 2
13 Procedure 2.1.20 54, Admin. Hold 14 Procedure 2.1.20.1 12 15 Procedure 2.2.47 35 16 Procedure 2.3 A-1 17 17 Procedure 2.3 J-1 3
18 Procedure 2.3 R-2 10 19 Procedure 2.3 9-3-3 8
20 Procedure 6.LOG.602 39 21 _
Procedure 6.REFUEL.301 r5 22 Procedure 6.REFUEL.304 11 23 _
Procedure 6.REFUEL.305 10 24 Procedure 6.REFUEL.306 (3
25 Procedure 6.SC.501 14 26 Procedure 7.4.9 11 27 Procedure 7.4.10 12 PROCEDURE 3.4.7 REVISION 26 PAGE 3 OF 31 3
.TTACHMENT 2 DESIGN CALCULATION CROSS-REFERENCE INDEX3 Page:
4 _of 12 NEDC:
05-031 Rev. Number: _
2 ITEM NO.
AFFECTED DOCUMENTS REV. NUMBER 28 Procedure 7.4.13 12C1 29 Procedure 7.4.15 11 30 Procedure 7.4.17 1.2 31 Procedure 10.25 4,2 31 Procedure 10.25.1 1.6 33 Procedure 10.27 1.4 34 Breach Control Procedure New 35 DCD 3 06/1.3/03 36 DCD 4 04/1.1/05 37 DCD 5 06/24/04 38 DCD 6 02/07/03 39 DCD 7 03/07/05 40 DCD10 06/1.1/05 41 DCD 31 03/09/05 PROCEDURE 3.4.7 REVISION 26 PAGE 4 OF 31 4
I
ATTACHMENT 3 AFFECTED DOCUMENT SCREENING Page:
5
.of 12 NEDC:
05-031 Rev. Number:
2 I
The purpose of this form is to assist the Preparer in screening new and revised design calculations to determine potential impacts to procedures and plant operations.(i)
SCREENING QUESTIONS YES NO UNCERTAIN
- 1.
J)oes it involve the addition, deletion, or manipulation of a component or components which could impact a system lineup and/or checklist for valves, power supplies (breakers), process control switches, HVAC dampers, or instruments?
- 2.
Could it impact system operating parameters (e.g.,
temperatures, flowrates, pressures, voltage, or fluid chemistry)?
- 3.
I)oes it impact equipment operation or response such as valve closure time?
II
[XI II
[XI
[]
MX I I I I I I
- 4.
I)oes it involve assumptions or necessitate changes to the sequencing of operational steps?
- 5.
Does it transfer an electrical load to a different circuit, or impact when electrical loads are added to or removed from the system during an event?
[ ]
I I
[XI
[XI
[II
[II
- 6.
Does it influence fuse, breaker, or relay coordination?
- 7.
Does it have the potential to affect the analyzed conditions of the environment for any part of the Reactor Building, Containment, or Control Room?
- 8.
Does it affect TS/TS Bases, USAR, or other Licensing Basis documents?
II]
[XI
[XI II I ]
[ ]
[ ]
I
- 9.
Does it affect DCDs?
[XI]
II I I]
I
- 10.
Does it have the potential to affect procedures in any way not already mentioned (refer to review checklists in Procedure EDP-06)? If so, identify:
[]
[]
[Ix If all answers are NO, then additional review or assistance is not required.
If any answers are YES or UNCERTAIN, then the Preparer shall obtain assistance from the System Engineer and other departments, as appropriate, to determine impacts to procedures and plant operations. Affected documents shall be listed on Attachment 2.
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ATTACHMENT 4 DESIGN CALCULATION SHEET Page: 6 of 12 NEDC: 05-031 Rev. Number:
2 Nebraska Public Power District DESIGN CALCULATIONS SHEET PURF'OSE This calculation incorporates by attachment Alion Science & Technology Calculation No.
ALION-CAL-NPPD-3236-001, Rev. 0, in accordance with CNS Engineering Procedure 3.4.7. The calculation determines the dose to a Control Room operator and to a person at the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) at the Cooper Nuclear Station site following a design basis Fuel Handling Accident (FHA). The analysis is performed using an Alternate Source Term (AST) in accordance with the guidance provided by the NRC in Regulatory Guide 1.183 (July 2000) and as authorized by 10 CFR 50.67. This calculation has been prepared as a Status 2 calculation for NRC review and will be as-built upon NRC approval.
Rev. 1 incorporates Rev. 1 of Alion's calculation, which was revised primarily to incorporate an additional Case 1 analysis to determine total Control Room dose assuming 168 hour0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> fuel decay time before fuel movement and without CREFS operation. In addition, Rev. 1:
Added additional Assumption (9)
Ad ded Bromine source term table (Table 3)
Added un-decayed activity source term table (Table 4)
Added Bromine evaluation per RG1.183 conformance Nuclear inventory modified to 29 isotopes to accommodate bromines Shine analysis modified to provide a qualitative estimate of dose.
Incorporated various typos and rewording Rev. 2 adds Section 4.1, which provides an evaluation of additional Secondary Containment release points. This evaluation determines that the analysis of the FHA utilizing the Reactor Building Vent release path is bounding for all other release paths.
Additionally, Rev. 2 upgrades this calculation from Status 2 to Status 3.
EXTENT OF REVIEW Alion's calculation was performed under their own QA program, which included an independent technical review. Therefore, the NPPD review does not include in-depth checks of mathematical calculations, but rather focuses on general acceptability (if design inputs, assumptions, methodology, and conclusions. Any significant comments or concerns identified during the review have been resolved with Alion and incorporated.
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ATTACHMENT 4 DESIGN CALCULATION SHEET Page: 7 of 12 NEDC: 05-031 Rev. Number:
2 REVIEW
SUMMARY
Alion's calculation is organized into a single main portion along with Attachments A through H, which include the computer files as well as Alion's Design Review Checklist.
- 1.
Purpose - The purpose of the calculation is as given above and as stated in Section 1 of Alion's calculation. This section was reviewed and found to be acceptable.
- 2.
Design Inputs - Design Inputs are contained in the Cross Reference Index given on page 4 of Alion's calculation and are discussed in Section 2 of Alion's calculation. The design inputs were reviewed and found to be acceptable. Non-status 1 inputs were verified using additional information and were found to lie acceptable for use in this calculation.
The revised Cross Reference Index, starting on page 2 of this calculation, identifies Design Inputs and Affected Documents that reflect the change from Status 2 to Status 3.
- 3. Assumptions - Major assumptions are identified in Section 3 of Alion's calculation.
Additional assumptions are inferred in the input documents used and identified throughout Alion's calculation by inference according to context and use. The assumptions were reviewed and found to be acceptable.
- 4.
Methodology - The methodology is described in Section 4 of Alion's calculation. The methodology was reviewed and found to be acceptable.
4.1 Release Point Evaluation -Alion calculation ALION-CAL-NPPD-3236-001,, evaluates the Fuel Handling Accident assuming a release point at the Reactor Building Vent. Additional ingress/egress points that could be open to establish a release path for radioactive materials outside Secondary Containment are identified in to Surveillance Procedure 6.SC.701, "Technical Specification Verification of Secondary Containment Access Doors and Hatches." The expected release of radioactive material at these additional points is evaluated to demonstrate a radioactive material release at the Reactor Building Vent is bounding.
Alion's calculation is performed assuming no credit for reduction of radioactive material released from the reactor building as a result of operation of engineered safety features (i.e., SGTS). (Attachment 1, Assumption 4) The Fuel Handling Accident is analyzed assum.tng fuel decay times of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Base Case) and 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> not crediting the operation of the Control Room Emergency Filtration System (CREFS) (Case 1).
(Attachment 1, Assumption 5) Atmospheric dispersion factors (X/Q) assuming a ground PROCEDURE 3.4.7 1
REVISION 26 l
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ATTACHMENT 4 DESIGN CALCULATION SHEET Page: 8 of 12 NEDC: 05-031 Rev. Number:
2 level release from the reactor building vent were used in the analysis. (Attachment 1, Section 4.1) The reactor building vent analysis assumes a release rate of 4.576 X 104 (ft3/min.). (Attachment 1, Section 2.3) 4.1.1 Control Room Location - The location and cross-sectional area of the Reactor Building Vent and the additional ingress/egress points with respect to the CREFS intake are summarized in Table 1, and graphically represented in Figure 1.
To assure operability of the Reactor Building Exhaust Plenum Radiation Monitor, a Reactor Building exhaust fan, or a Standby Gas Treatment System fan will be operating.
Forced outflow with a Reactor Building exhaust fan egresses from the Reactor Building vent. A release with an SGT fan operating is via the Elevated Release Point. An elevated release is less bounding than the vent release based on X/Q values calculated in NEDC 99-031. With one of these fans in service, a potential inflow into Secondary Containment through the additional ingress/egress points is created with the potential to reduce or eliminate the release of radioactive material. As a minimum, for a case where CREFS is running in lieu of a Reactor Building exhaust fan or SGT fan, the outflow through the additional points is expected to occur over a longer period of time due to natural air diffusion as opposed to the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> forced outflow assumed in the FHA analysis.
The following paragraphs summarize the expected release of radioactive material at the additional ingress/egress points assuming a FHA on the refuel floor (elev. 1001).
Release points 3, 4, 6 and 7 provide a direct release from Secondary Containment. These points are located at least three times further away from the CREFS intake than the Reactor Building Vent. A release at these points would require transport of the radioactive material through numerous compartments to a lower level in Secondary Containment which will reduce the amount of radioactive material available for release to the environment and any potential release would be counteracted by an expected air inflow or as a worst case, via a non-forced outflow.
Release points 8, 9, 10 and 11 provide a release point from Secondary Containment that is located in a compartment within permanent site structures. These points are located a minimum of 20 percent further away from the CREFS intake than the Reactor Building Vent. A release from Secondary Containment at these points would require transport of the radioactive material through numerous compartments to a lower level in Secondary Containment which will reduce the amount of radioactive material available for release to the environment, require transport of the radioactive material through at least one additional compartment outside Secondary Containment further reducing the amount of radioactive material available for release and would be counteracted by an expected air inflow or as a worst case, via a non-forced outflow.
PROCEDURE 3.4.7 l
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ATTACHMENT 4 DESIGN CALCULATION SHEET Page: 9 of 12 NEDC: 05-031 Rev. Number:
2 I
I Release point 2 provides a release point from Secondary Containment that is located in a compartment within a permanent site structure. Although this point is located marginally closer to the CREFS intake, a release at this point would require transport of the radioactive material through numerous compartments to a lower level in Secondary Containment which will reduce the amount of radioactive material available for release to the environment, require transport of the radioactive material through at least one additional compartment outside Secondary Containment further reducing the amount of radioactive material available for release and would be counteracted by an expected air inflow or as a worst case, via a non-forced outflow.
Release point 5 provides a direct release from Secondary Containment. This point is located marginally closer to the CREFS intake. However, the cross-sectional area of this point is less than 30 % of the Reactor Building Vent cross-sectional area and any potential release would be counteracted by an expected air inflow or as a worst case, via a non-forced outflow.
Based on the above discussion, it is concluded that the analysis of the FHA utilizing the Reactor Building Vent release path is bounding for all other release paths with respect to Control Room Dose.
4.1.2 ELAB and LPZ Locations - The EAB and LPZ receptor points are analyzed using (X/Q) values established in calculation NEDC 99-036. (Attachment 1, Section 2.5) The (X/Q) values were calculated utilizing a minimum distance from the Elevated Release Point (ERP) to the site boundary in the East-South-East sector of 920 meters for the EAB, and a distance of 1 mile from the center of the reactor for the LPZ.
The additional ingress/egress points that could be open to establish a path outside Secondary Containment are all located north and west of the ERP to provide a greater distance to the site boundary than that used in the EAB analysis. The railroad airlock (Point 3), which is farthest from the center of the reactor, is within 200 feet of the center of the :eactor. A change of this distance in relation to the 1 mile distance used for the LPZ dose calculation is considered to have a minimal potential impact on results because of the small difference with respect to the overall distance.
Based on the previous discussion on the amount of radioactive material available for release at the additional ingress/egress points, the minor change in distance used to calculate (X/Q) values and the relatively low Accumulated Dose at these locations, it is concluded that the analysis of the FHA utilizing the Reactor Building Vent release path is representative of all other release paths with respect to the EAB and LPZ dose analysis.
Any impact on accumulated dose at the EAB or LPZ due to a release at any of the additional ingress/egress points would be minimal.
Z PROCEDURE 3.4.7 l
REVISION 26 l
PAGE 9 OF 31 9
TRUE NORTH Page 10 of 12 NEDC 05-031 Rev. Number 2 I
I I
CONTROL BUILDING a_
CONTROL ROOM A
INTAKE I
5 R
- 1 BUILDING 8
\\
7 LEEND
- 1. REACTOR BUILDING VENT
- 2. PERSONNEL AIRLOCK
- 3. RAILROAD AIRLOCK
- 4. ALTERNATE SHUTDOWN ROOM DOOR
- 5. REACTOR BUILDING ROOF HATCH
- 6. HP(,J ROOM HATCH
- 7. SOJTHWEST VESTIBULE
- 8. REACTOR BUILDING HVAC AIRLOCK
- 9. RRIAG HVAC EXHAUST AIRLOCK 1O.RRhMG HVAC FILTER DOORS ll.RHF' A HEAT EXCHANGER ROOM CEILING PLUG 3.
8 REACTOR BUILDING POSSIBLE RELEASE POINTS FIGURE 1
A.TTACHMENT 4 DESIGN CALCULATION SHEET Page: 11 of 12 NEDC: 05-031 Rev. Number:
2 TABLE 1 Possible Release Points Comparison Table Release Description/
N-S E-W Distance Angle From Cross-se(tion Point Location (Ft.)
(Ft.)
(Ft-)
True North (Ft.2)
Reactor Building I
Vent 53.33 S 20.33 E 57 125 32 R-1001 Overhead Personnel Airlock 40.33 S 29.58 E 50 42 R-903 1
Railroad Airlock 145.67 S IG0.42 W 217 194 433 R-903 Alternate 4
Shutdown Hatch 168.83 S 59.08 E 179 127 8
R-903 Reactor Building 5
Roof Hatch 45.33 S 17.58 E 49 125 9
R-1001 Overhead G
l HPCI Hatch 175.67 S 92.42W 198 174 165 R-859 Overhead Southwest 7
Vestibule 183.33 S 53.92 W 191 1G3 21 R-903 Reactor Building 8
HVAC Airlock 70.83 S 0.42 WV 71 146 35 R-958 RRNIG Exhaust 9
Airlock G5.42 S 27.17 W 71 1G9 21 R-97G RRMG Filter 10 Airlock 46.67 S 54.83 W 72 196 14 R-976 RHR A Heat 11 Exchanger Hatch 49.83 S 48.83 W 70 190 59 R-958 I
PROCEDURE 3.4.7 I
REVISION 26 I
PAGE 11 OF 31 l
11
ATTACHMENT 4 DESIGN CALCULATION SHEET Page:
'12 of 12 NEDC: 05-031 PRev. Number:
2
- 5.
Results / Conclusions - Results and conclusions are given in Section 5 of Alion's calculation. The results and conclusions were reviewed and found to be acceptable.
All calculated doses are below the corresponding regulatory limits.
Calculation Section 4.1.1 demonstrates that the Reactor Building Vent release point bounds, other identified Secondary Containment release points for calculating the dose to a control room occupant following a design basis Fuel Handling Accident at Cooper Nuclear Station.
Calculation Section 4.1.2 demonstrates that the analysis of the FHA utilizing the Reactor Building Vent release path is representative of all other identified Secondary Containment release paths with respect to the EAB and LPZ dose analysis. Any impact on accumulated dose at the EAB or LPZ due to a release at any of the additional ingress/egress points would be minimal.
- 6. References -References are listed in Section 6 of Alion's calculation. The references were reviewed and found to be acceptable.
- 7. Attachments
- 1. Alion calculation ALION-CAL-NPPD-3236-001, Rev. 1 (including attachments thereto)
- 2. GE Letters REK:99-152 and REK:99-211 (References 6.10 and 6.9 of Alion's calculation)
PROCEDURE 3.4.7 l
REVISION 26 PAGE 12 OF31l 12
NLS200601 7 Revised TS and TS Bases Changes Revised TS Markup Pages
- 1) Page 3.3-63
- 2) Page 3.7-8
- 3) Page 3.7-9 Revised TS Clean Pages
- 1) Page 3.3-63
- 2) Page 3.7-8
- 3) Page 3.7-9 Revised TS Bases Markup Pages
- 1) Page B 3.3-189
- 2) Page B 3.7-18
- 3) Page B 3.7-19
- 4) Page B 3.7-20
CREF System Instrumentation 3.3;.7.1 Table 3.3.7.1-1 (page 1 of 1)
Control Room Emergency Filter System Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE
- 1. Reactor Vessel Water 1,2,3, 2
>-42 inches Level -Low Low (Level 2)
(a)
SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4
- 2. Drywell Pressure - High 1,2,3 2
< 1.84 psig SR 3.3.7.1.3 SR 3.3.7.1.4
- 3. Reactor Building Ventilation 1,2,3, 2
< 49 mR/hr Exhaust Plenum (a),(b)
SR 3.3.7.1.2 Radiation - High SR 3.3.7.1.3 SR 3.3.7.1.4 (a)
During operations with a potential for draining the reactor vessel.
(b)
During eORE ALTtrtf+k-Ht-P*'-ftand Iduring movement of lately irradiated fuel assemblies in the secondary containment.
Amendment 24-2 3.3-63 4212ffi
CREF System
.. 7.4 3.7 PLANT SYSTEMS 3.7.4 C:ontrol Room Emergency Filter (CREF) System LCO 3.7.4 APPLICABILITY:
The CREF System shall be OPERABLE.
MODES 1, 2, and 3, During movement of lately irradiated fuel assemblies in the secondary containment, During CORE ALTcERAFIoN-,
During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION TICME A.
CREF System inoperable.
A.1 Restore CREF System to 7 days OPERABLE status.
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Cooper 3.7-8 Amendment No. +7-8
CREF System
- 3.7.4 ACTICNS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and
NOTE-------------
associated Completion LCO 3.0.3 is not applicable.
Time of Condition A not rnet during movement of lately irradiated fuel C.1 Suspend movement of lately Immediately assemblies in the irradiated fuel assemblies in secondary containment; the secondary containment.
ALTER ATIONS, or AND during OPDRVs.
C.2 Suspend CORE Immediately ALTERATIONS.
AND C.32 Initiate action to Immediately suspend OPDRVs.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREF System for > 15 minutes.
31 days (continued)
Cooper 3.7-9 Amendment No. -78
CREF System Instrumentation 3.:3.7.1 Table 3.3.7.1-1 (page 1 of 1)
Control Room Emergency Filter System Instrumentation APPLICABLE MODES OR REQUIRED OTHER CHANNELS SPECIFIED PER TRIP SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM REQUIREMENTS VALUE
- 1. ReactorVessel Water 1,2,3, 2
>-42Inches L~vel - Low Low (Level 2)
(a)
SR 3.3.7.1.2 SR 3.3.7.1.3 SR 3.3.7.1.4
- 2. Drywell Pressure - High 1,2,3 2
< 1.84 psig SR 3.3.7.1.3 SR 3.3.7.1.4
- 3. Reactor Building Ventilation 1,2,3, 2
<49 mR/hr Exhaust Plenum (a),(b)
SR 3.3.7.1.2 Radiation - High SR 3.3.7.1.3 SR 3.3.7.1.4 (a)
Duiing operations with a potential for draining the reactor vessel.
(b)
Duding movement of lately irradiated fuel assemblies in the secondary containment.
Cooper 3.3-63 Amendment
CREF System 3.7.4 3.7 PLNNT SYSTEMS 3.7.4 C:ontrol Room Emergency Filter (CREF) System LCO 3.7.4 APPLICABILITY:
The CREF System shall be OPERABLE.
MODES 1, 2, and 3, During movement of lately irradiated fuel assemblies in the secondary containment, During operations with a potential for draining the reactor vessel (OPDRVs).
ACTIONS CONDITION REQUIRED ACTION TIME A.
CREF System inoperable.
A.1 Restore CREF System to 7 days OPERABLE status.
B.
Required Action and B.1 Be in MODE 3.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition A not AND met in MODE 1, 2, or 3.
B.2 Be in MODE 4.
36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (continued)
Cooper 3.7-8 Amendment No.
CREF System 3.7.4 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and NOTE-------------
associated Completion LCO 3.0.3 is not applicable.
Time of Condition A not rnet during movement of lately irradiated fuel C.1 Suspend movement of lately Immediately assemblies in the irradiated fuel assemblies in secondary containment or the secondary containment.
during OPDRVs.
AND C.2 Initiate action to Immediately suspend OPDRVs.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Operate the CREF System for > 15 minutes.
31 days (continued)
I I
I I
Cooper 3.7-9 Amendment No. 4-8
CREF System Instrumentation B 3.3.7.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY
- 3. Reactor Building Ventilation Exhaust Plenum Radiation -
High (continued)
The Reactor Building Ventilation Exhaust Plenum Radiation -
High Function is required to be OPERABLE in MODES 1, 2, and 3 and during movement of lately irradiated fuel assemblies in the secondary containment, GGRE ALTERATIONS, and operations with a potential for draining the reactor vessel (OPDRVs), to ensure control room personnBl are protected during a pipe break resulting in significant releases of radioactive steam and gas, fuel handling event, or vessel draindown event. During MODES 4 and 5, when these specified conditions are not in progress (e.g., ECRE AUTERATIGNSOPDRVs), the probability of a pipe break resulting in significant releases of radioactive steam and gas or fuel damage is low; thus, the Function is not required. Due to radioactive decay, this Function is only required to initiate the CREF System during fuel handling accidents involving handling lately irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 7 days). During the movement of lately irradiated fuel. Reactor Building ventilation exhaust flow (provided by either a Reactor Building ventilation exhaust fan or SGT fan) is a required support function.
ACTIONS A Note has been provided to modify the ACTIONS related to CREF System instrumentation channels. Section 1.3, Completion Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each additional failure, with Completion Times based on initial entry into the Condition. However, the Required Actions for inoperable CREF System instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable CREF System instrumentation channel.
A.1 Because of the diversity of sensors available to provide isolation signals and the common interface with the Secondary Containment isolation Instrumentation, allowable out of service time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Functions 1 Cooper B 3.3-189 i i i( 104
CREF System B 3.7.4 BASES APPLICABLE SAFETY ANALYSES The ability of the CREF System to maintain the habitability of the control room is an explicit assumption for the safety analyses presented in the USAR, Chapters X and XIV (Refs. 1 and 2, respectively). The CREF System is assumed to operate following a loss of coolant accident and a fuel handling accident involving handling lately irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 7 days).
The CREF System satisfies Criterion 3 of 10 CFR 50.36(c)(2)(ii) (Ref. 4).
LCO The CREF System is required to be OPERABLE, since total system failure could result in exceeding a dose of 5 rem to the control room operators in the event of a DBA.
The CREF System is considered OPERABLE when the individual components necessary to control operator exposure are OPERABLE.
The system is considered OPERABLE when its associated:
- a.
Fans are OPERABLE (one supply fan, the emergency booster fan and the exhaust booster fan);
- b.
HEPA filter and charcoal adsorber are not excessively restricting flow and are capable of performing their filtration functions; and
- c.
Ductwork, valves, and dampers are OPERABLE, and air circulation can be maintained.
In addition, the control room boundary must be maintained, including the integrity of the walls, floors, ceilings, ductwork, and access doors, such that the pressurization limit of SR 3.7.4.4 can be met. However, it is acceptable for access doors to be open for normal control room entry and exit, and not consider it to be a failure to meet the LCO.
Cooper B 3.7-18 1 VE) 1&+
CREF System B 3.7.4 BASES APPLICABILITY In MODES 1, 2, and 3, the CREF System must be OPERABLE to control operator exposure during and following a DBA, since the DBA could lead to a fission product release.
In MODES 4 and 5, the probability and consequences of a DBA are reduced because of the pressure and temperature limitations in these MODES. Therefore, maintaining the CREF System OPERABLE is not required in MODE 4 or 5, except for the following situations under which significant radioactive releases can be postulated:
- a.
During operations with potential for draining the reactor vessel (OPDRVs); and
- b.
During GeRE ALTERATIONS and eb.
During movement of lately irradiated fuel assemblies in the secondary containment. Due to radioactive decay, the CREF System is only required to be OPERABLE during fuel handling involving handling lately irradiated fuel (i.e., fuel that has occupied part of a critical reactor core within the previous 7 days).
ACTIONS A.1 The inoperable CREF System must be restored to OPERABLE status within 7 days. With the unit in this condition, there is no other system to perform control room radiation protection. The 7 day Completion Time is based on the low probability of a DBA occurring during this time period.
B.1 and B.2 In MODE 1, 2, or 3, if the inoperable CREF System cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE that minimizes risk. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.
Cooper B 3.7-19 Coopr B3.7-9 Revisic t-
CREF System B 3.7.4 BASES ACTIONS CA, IC2 and C.3 The Required Actions of Condition C are modified by a Note indicating that LCO 3.0.3 does not apply. If moving lately irradiated fuel assemblies while in MODE 1, 2, or 3, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of lately irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.
During movement of lately irradiated fuel assemblies in the secondary containment, during GORE AL-TRATFqs, or during OPDRVs, if the inoperable CREF System cannot be restored to OPERABLE status within the required Completion Time, activities that present a potential for releasing radioactivity that might require isolation of the control room must be immediately suspended. This places the unit in a condition that minimizes risk.
If applicable, eeRE ALTERATIONS and movement of lately irradiated fuel assemblies in the secondary containment must be suspended immediately. Suspension of these activities shall not preclude completion of movement of a component to a safe position. Also, if applicable, actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and the subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.
SURVEILLANCE REQUIREMENTS SR 3.7.4.1 This SR verifies that the CREF System in a standby mode starts on demand and continues to operate. The system should be checked periodically to ensure that it starts and functions properly. As the environmental and normal operating conditions of this system are not severe, testing the system once every month provides an adequate check on this system. Since the CREF System does not contain heaters, the system need only be operated for > 15 minutes to demonstrate the function of the system. The 31 day Frequency is based on the known reliability of the equipment.
Cooper B 3.7-20
ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS© Correspondence Number: NLS2006017 The following table identifies those actions committed to by Nebraska Public Power District (NPPD) in this document. Any other actions discussed in the submittal represent intended or planned actions by NPPD. They are described for information only and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.
COMMITMENT COMMITTED DATE COMMITMENT NUMBER OR OUTAGE The Reactor Building Roof Hatch will be Concurrent with 30-maintained closed during the movement of NLS2006017-01 day implementation of issued license irradiated fuel in Secondary Containment.
amedment.
amendment.
The TS 3.3.7.1 Bases will be revised as Concurrent aith 30-provided in Enclosure 2 to formalize this day implementation support feature [airflow in the Reactor amedment.
Building exhaust plenum].
I 4
.4 4
4 4
+
4
- 4.
4 PROCEDURE 0.42 REVISION 19 PAGE 21 OF 27