ML12139A198
| ML12139A198 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 05/16/2012 |
| From: | Schimmel M Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML121390261 | List: |
| References | |
| L-PI-12-038, TAC ME6984, TAC ME6985 | |
| Download: ML12139A198 (65) | |
Text
ENCLOSURE 4 CONTAINS PROPRIETARY INFORMATION -
WITHHOLD FROM PUBLIC DISCLOSURE IN ACCORDANCE WITH I 0 CFR 2.390 Xcel Energyg MAY 1 6 2012 L-PI-12-038 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant, Units 1 and 2 Dockets 50-282 and 50-306 Renewed License Nos. DPR-42 and DPR-60 Response to Requests for Additional lnformation (RAI) Associated with Spent Fuel Pool Criticalitv Chanqes (TAC Nos. ME6984 and ME69851 In a letter to the U.S. Nuclear Regulatory Commission (NRC) dated August 19, 201 1 (Agencywide Documents and Management System (ADAMS) Accession No. MLI 12360231), the Northern States Power Company, a Minnesota corporation doing business as Xcel Energy (hereafter "NSPM"), requested an amendment to the Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP). The proposed amendment requested TS changes and approval of the submitted criticality safety analysis methodology.
To complete their review, the NRC staff requested additional information by letter dated March 20,201 2 (ADAMS Accession No. ML120620389). to this letter provides the NSPM response to the March 20, 2012 request for additional information with reference to Enclosure 4 for certain information provided by Westinghouse. Enclosure 2 provides the non-proprietary version of the Westinghouse-prepared replies. Enclosure 3 provides the Westinghouse Application for Withholding Proprietary lnformation from Public Disclosure CAW-12-3471, accompanying Affidavit, Proprietary lnformation Notice, and copyright notice. Enclosure 4 contains the proprietary version of the Westinghouse-prepared replies. Enclosure 5 provides TS markups to address a clarification requested by NRC Staff.
As Enclosure 4 contains information proprietary to Westinghouse Electric Company LLC, it is supported by the enclosed affidavit (Enclosure 3) signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commission's regulations.
1717 Wakonade Drive East 0 Welch, Minnesota 55089-9642 Telephone: 651.388.1 121
Document Control Desk Page 2 Correspondence with respect to the copyright or proprietary aspects of the items provided in Enclosure 4 of this letter or the supporting Westinghouse affidavit should reference CAW-12-3471 and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
NSPM submits this supplement in accordance with the provisions of 10 CFR 50.90.
The supplemental information provided in this letter does not impact the conclusions of the Determination of No Significant Hazards Consideration and Environmental Assessment presented in the August 19, 201 1 submittal.
In accordance with 10 CFR 50.91, NSPM is notifying the State of Minnesota of this License Amendment Request (LAR) supplement by transmitting a copy of this letter to the designated State Official.
If there are any questions or if additional information is needed, please contact Glenn Adams at (61 2) 330-6777.
Summaw of Commitments This letter contains no new commitments:
I declare under penalty of perjury that the foregoing is true and correct.
MAY 1 6 2812 Mark A. Schimmel Site Vice President, Prairie Island Nuclear Generating Plant Northern States Power Company - Minnesota Enclosures (5) cc:
Regional Administrator, Region Ill, USNRC Project Manager, Prairie Island Nuclear Generating Plant, USNRC Resident Inspector, Prairie Island Nuclear Generating Plant, USNRC State of Minnesota (without enclosures)
ENCLOSURE I Spent Fuel Criticality Analysis Response to Requests for Addition Information (RAI)
In a letter to the U.S. Nuclear Regulatory Commission (NRC) dated August 19, 201 1 (Agencywide Documents and Management System (ADAMS) Accession No. MLI 12360231), the Northern States Power Company, a Minnesota corporation doing business as Xcel Energy (hereafter "NSPM"), requested an amendment to the Technical Specifications (TS) for Prairie Island Nuclear Generating Plant (PINGP). The proposed amendment requested TS changes and approval of the submitted criticality safety analysis methodology.
To complete their review, the NRC staff requested additional information by letter dated March 20,201 2 (ADAMS Accession No. ML120620389).
For clarity, the NRC RAI information is provided below in italics font and the NSPM response is provided in plain font.
Reactor Systems Branch (SRXB) RAls RAI SRXB-1:
Provide an analysis to show that modeling no gadolinia is conservative. The application references an analysis contained in NUREG/CR-6760 ("Study of the Effect of Integral Burnable Absorbers for PWR Burnup Creditly, which was based on a cask analysis with poison plates. These conclusions may or may not be applicable to the PINGP spent fuel pool analysis which assumed un-poisoned racks.
NSPM Response: Provided in Enclosure 4.
RAI SRXB-2:
Show that the 422V+ file1 assembly design is the appropriate reference fuel assembly design to use for all proposed storage ~~nfigurations.
NSPM Response: Provided in Enclosure 4.
RAI SRXB-3:
Availability of information related to criticality analysis of configurations such as the consolidated fuel rod storage canister is limited. Address the following to allow the NRC staff to evaluate these fuel storage configurations.
- a. Provide the details of the consolidated fuel rod storage canister analysis described on page 3-2 1 of WCAP-I 7400-P.
- b. In addition to the base case results shown on page 4-1 7 of WCAP-17400-P, provide a sensitivity analysis varying the enrichment to 3.0 weight percent 2 3 5 ~.
Page 1 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM
- c. Provide the details of the analysis described on page 4-21 of WCAP-I 7400-P to show the reactivity effects of the fuel assembly structural materials allowed to be placed in cells adjacent to the consolidated fuel rod storage canisters.
- d. Provide the details of the Array A and F interface analysis described on page 4-21 of WCA P-1 7400-P.
NSPM Response to items a through d: Provided in Enclosure 4.
- e. Provide the burnup ranges of the actual fuel rods stored in the consolidated fuel rod storage canisters and the failed fuel bin baskets.
NSPM Response to item e: As described in USAR 10.2.1 5.3, the fuel assemblies loaded into the consolidated rod storage canisters experienced assembly average exposures ranging from 28 to 40 GWDIMTU. Using the average assembly burnups from the original host assemblies, NSPM nuclear engineers have determined that pins currently residing in the failed fuel pin baskets range from approximately 15,788 MWDIMTU to 31,710 MWDIMTU. The values of actual burnup are immaterial because (as discussed in the LAR), the criticality analyses of the consolidated rod storage canisters and the failed fuel pin baskets were performed with no credit for burnup; fresh new fuel was assumed.
- f. Provide legible versions of Figure 3-2, "Failed Fuel Pin Basket, " Figure 3-3, "Fuel Rod Storage Canister, " and Figure 3-4, "Fuel Rod Storage Canister Cross Section, "
from WCA P-1 7400-P.
NSPM Response to item f: A legible version of Figure 3-2 is attached at the end of this Enclosure. Legible versions of Figures 3-3 and 3-4 are provided by Westinghouse in Enclosure 4.
RAI SRXB-4:
The application does not appear to discuss the use of axial blankets. Address the following related to axial blankets:
- a. If both blanketed and non-blanketed fuel assemblies are present in the pool, describe how the axial burnup profiles were modeled for both blanketed and non-blanketed assemblies at PINGP.
- b. Describe how the axial enrichment variation was modeled in the analysis for both blanketed and non-blanketed assemblies.
- c. Are annular pellets used for the blanketed regions? If so, describe and justify how the void was modeled.
NSPM Response: Provided in Enclosure 4.
Page 2 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM RAI SRXB-5:
Address the following related to the selection of design axial burnup profiles:
- a. Provide the analysis that determined how limiting shape assumptions were developed.
- b. Provide the actual PlNGP limiting relative burnup values at lower axial levels.
NSPM Response: Provided in Enclosure 4.
RAI SRXB-6:
Address the following related to rodded operation (Section 5.2 of WCAP-17400-P):
- a. Provide the analysis performed to determine that up to I gigawatt-day per metric ton uranium (GWd/MTU) of rodded operation would be "bounded by the design basis assemblies used to develop the burnup requirements."
NSPM Response to item a: Provided in Enclosure 4.
- b. Clarify the latter part of the following statement regarding how the evaluation will be performed:
"If an assembly experiences more than I GWd/MTU of core average rodded operation, the assembly shall either be treated as Fuel Category I or evaluated to determine which Fuel Category is appropriate for safe storage of the assembly."
NSPM Response to item b: The 1 GWdIMTU limit on rodded operation will be instituted in the appropriate procedure that characterizes fuel prior to creation of the Fuel Transfer Log. NSPM would expect this requirement to be included in the administrative procedure that directs the generation of Fuel Transfer Logs. As with any procedure, any change or exception to a particular step or particular limit (e.g., 1 GWdIMTU) would require review of the change pursuant to 10 CFR 50.59. That review may consider the fuel assembly's actual operating history to assess the effect of rodded operation. Further, any evaluation supporting that 50.59 review would necessarily be performed within the limits prescribed for "method of evaluation" (defined by NSPM's 50.59 Resource Manual).
RAI SRXB-7:
Regarding criticality code validation, the Interim Staff Guidance (ISG) DSS-ISG-20 10- I states:
"An acceptable means of including isotopes that are not explicitly represented in the critical experiments used in the validation would be to increase the bias and bias Page 3 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM uncertainty by an amount proportional to the reactivity worth of the isotopes not explicitly validated, "
WCAP-I 7400-P addresses the effect of fission products. Justify the treatment used.
NSPM Response: Provided in Enclosure 4.
RAI SRXB-8:
Confirm that PINGP has not used any reactivity control devices other than gadolinium and burnable poison rod assembly.
NSPM Response: As stated in the LAR, the Rod Cluster Control Assembly (RCCA) is another reactivity control device used at PINGP during core operation. Aside from gadolinium and burnable poison rod assemblies, NSPM confirms that no other reactivity control devices have been used at PINGP.
RAI SRXB-9:
Provide the analysis results for the misloaded configurations (i.e., Arrays B, C, and E).
NSPM Response: Provided in Enclosure 4.
RAI SRXB-10:
Provide a comparison of assemblies that use a range of expected burnable poison loading amounts (number of rods and weight percents) similar to Tables 3-3 and 3-5 of WCAP-17400-P.
NSPM Response: Provided in Enclosure 4.
RAI SRXB-11:
Describe the sun/eillance program on the rod cluster control assemblies (RCCAs) credited to ensure the required subcritical margin.
NSPM Response: NSPM is not crediting any current or future surveillance program for an RCCA discharged to the Spent Fuel Pool (SFP). This position is based on the proven integrity of RCCAs in reactor service and the general observation that the Technical Specifications do not require any such surveillance while the RCCA is in service during reactor power operations. This position is also manifest in accident analyses, wherein the neutron poison integrity of RCCAs is not assumed to be depleted.
Once an RCCA is discharged to the spent fuel pool, it is not subjected to any depletive flux and no further poison depletion is assumed to occur. Thus, if the integrity of the RCCA is sufficient during its operational life and it is exposed to no depletive flux while in the SFP, no surveillance program is necessary for an RCCA discharged to the SFP.
Page 4 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM Notwithstanding the inservice integrity of the RCCA neutron poison described above, the analysis performed to support Array G conservatively assumed a 25% reduction in the poison content of the credited RCCA. This is conservative based on the following discussion in which inservice RCCA neutron poison integrity is inferred from reactor startup testing surveillance:
The design lifetime of RCCAs is approximately 12 effective full power years and is based on mechanical degradation in the RCCA tips exposed to the core neutron flux and not based on poison depletion. During each refueling, the reactivity worth of all control rod banks is specifically measured during startup testing. The acceptance criteria for any measured control rod worth are < 15% from predicted worth for any individual bank or < 8% of predicted total worth. Therefore, the assumed value of a 25% reduction from the design reactivity is conservative.
RAI SRXB-12:
For several fuel categories (i.e., 3, 5, and 6), the same fuel category number is proposed for storing both "Cycles 1-4" and "> Cycle 4" fuel assemblies. What specific controls will be in place to prevent using the wrong loading curve?
NSPM Response: In accordance with fleet procedure, NSPM will make all procedure changes necessary to safely implement the proposed amendment, including establishment of human performance controls (as necessary) to prevent use of the wrong loading curve.
As discussed in the LAR (Enclosure 1 page 10 of 28), NSPM does not perceive the use of the subject fuel categories as a particular human performance challenge:
"With respect to fuel characterization, the only significant parameter change in the proposed TS is the use of Core Operating Cycle to distinguish which TS table applies to a given irradiated fuel assembly. This change causes no net increase for a human error because any risk associated with identifying the wrong operating cycle is comparable to the current risk of mistakenly identifying a non-gadolinium assembly as a gadolinium assembly. Neither of these parameters is visually evident on a fuel assembly, but is derived from fuel or operating records and administratively tied to the particular fuel assembly serial number identification (ID). In that regard, these two different parameters (Cycle designation vs, gadolinium content) are substantially equivalent in risk from a human factors standpoint."
Accordingly, NSPM would not expect to develop any specific controls on these loading curves beyond the programmatic controls placed on fuel characterization. As described in the LAR (Enclosure 1, page 13 of 28), the Fuel Transfer Log is prepared and verified by qualified individuals.
Page 5 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM RAI SRXB-13:
Propose the Technical Specifications (TS) language for the interface requirements, or provide justification/explanation for the removal of the interface requirements from the PlNGP TSs.
NSPM Response: As described in the LAR, the proposed TS adequately describe all possible fuel placement and all possible interfaces, as follows:
- 1. Proposed TS 3.7.17 requires that each fuel assembly stored in the SFP shall satisfy the loading restrictions of TS 4.3.1 (emphasis added).
- 2. Proposed TS 4.3.1.I
.e requires that the storage racks be maintained with fuel assemblies loaded in accordance with Figure 4.3.1-1.
- 3. Figure 4.3.1-1 provides only seven (7) allowable arrays, with a caveat to allow Array F to interface with Array A, and no other.
Thus, the TS require that each fuel assembly meets one of the seven arrays at all times, whether it is residing in the middle of an array or residing in the interface between two different arrays. This is explained further in proposed TS Bases 3.7.1 7 (insert A):
Array interface requirements: Technical Specifications provide only one special interface requirement between different arrays. This specific interface is described in Figure 4.3.1-1 Note 7 (Array F shall interface only with Array A) and was specifically analyzed. Otherwise, the Technical Specifications do not provide any unique rules for the interface between arrays. Rather, the Technical Specifications require that all fuel in the spent fuel pool satisfy one of the required arrays, even in transitions between two major arrays.
In a telephone conference on April 11, 2012, NRC Staff requested further clarification of the specification based on precedent established by another licensee. Thus, NSPM proposes a TS revision to add a clarifying note to TS Figure 4.3.1-1 similar to that provided in Florida Power & Light letter to NRC dated March 31, 2012 (ADAMS Accession No. ML12094A317). The precedent note could not be adopted verbatim because of geometrical differences between PlNGP and the precedent configuration; most notably, PlNGP includes a 3x3 array whereas the precedent was based only on 2x2 array interfaces. The proposed TS note is provided in Enclosure 5.
Health Physics and Human Performance Branch (AHPB) RAls RAI AHPB-1:
Describe the proposed procedure changes, additions, and deletions, including Procedure Title and Number.
NSPM Response: Below is a list of the procedure changes that NSPM expects to make in support of the proposed amendment. In accordance with 10 CFR 50 Appendix Page 6 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM B, NSPM will make all procedure changes necessary to safely implement the proposed amendment. Please note that this list is tentative, pending further changes that may evolve during the amendment review process. Further changes may be required when formal implementation reviews are initiated per site Engineering Change processes initiated after LAR approval. Also, this list does not include all administrative changes to align new Technical Specification table and figure numbering. These changes will be identified as part of the formal implementation processes.
Expected procedure changes include:
D5.3, "Handlins of Non-Fuel Components in the Spent Fuel Pool ". (Procedure number and title are tentative). NSPM expects to issue a new procedure to manage movement of non-fuel components within the spent fuel pool. The primary purpose of the new procedure will be to provide assurance that the empty cells credited in Arrays B, C, D and E remain empty when manipulating non-fuel material. This new procedure will contain the following attributes:
0 The procedure will contain reference to the new Technical Specification TableIFigure numbering.
The procedure will define non-fuel components to distinguish non-fuel from fuel movement managed in existing procedure D5.1.
0 The procedure will require use of component transfer logs for non-fuel components being placed in spent fuel storage racks. These transfer logs will meet the same preparation, review and performance requirements as the existing logs used to control movement of fuel (Reference existing D5.1 and SWl-NE-8).
The new procedure will require the presence of an individual qualified and trained to understand the limitations of placement of non-fuel within the spent fuel pit storage racks.
0 The procedure will contain provisions to assure compliance with the analysis completed to support the new criticality analysis.
- 2. D5.1, "Spent Fuel Pit Fuel Handlins Operations" is expected to be revised to implement the requirements for handling fuel in the spent fuel pool area. This procedure will be revised to include the following attributes:
align with the new Technical Specification TablelFigure numbering 0
distinguish fuel from non-fuel movement managed in the proposed new procedure D5.3.
add the new analyzed spent fuel pool temperature for normal conditions ( 4 5 0 degrees F)
- 3. D5.2, "Reactor Refuelins Operations" provides the requirements for fuel handling during normal refueling activities in both the containment structure and spent fuel pool. This procedure is expected to be revised to include the following attributes:
0 align with the new Technical Specification TableIFigure numbering.
Spent Fuel Criticality Analysis - Response to RAls NSPM 0
distinguish fuel from non-fuel movement managed in the proposed new procedure D5.3.
0 add the new analyzed spent fuel pool temperature for normal conditions ( 4 5 0 degrees F)
- 4. Surveillance Procedure SP 1951, "Spent Fuel Pool lnventorv Verification" is presently completed before fuel movement in the pool to further verify Technical Specification compliance for all proposed moves. (Meets the requirements of TS SR 3.7.17.1). This procedure will be revised to include the following attributes:
align with the new Technical Specification TableIFigure numbering.
0 assure all preplanned fuel moves will comply with all the fuel category and array requirements delineated in the new Technical Specifications.
0 eliminate any reference to Gadolinia loading (It is no longer credited in analyses)
Surveillance Procedure SP 1350, "Spent Fuel Pool lnventorv Verification" is presently completed after fuel movement in the pool to further verify Technical Specification compliance for all new moves. (Meets the requirements of TS SR 3.7.17.2). This procedure will be revised to include the following attributes:
0 align with the new Technical Specification TableIFigure numbering.
eliminate reference to old wording (i.e, checkerboarding) and add new wording to assure the verification addresses all new arrays and associated restrictions.
0 include the necessary steps to address LAR Enclosure 7 Commitment 1 (i.e.,
"The procedure revisions will include a post campaign validation of all affected Spent Fuel Pool (SFP) locations, whether fresh fuel is involved in the campaign or not").
- 6. Administrative Work Instruction SWI-NE-22, "SHUFFLEWORKS Fuel Movement Planning Svstem" is managed by the Nuclear Engineering Staff and provides detailed instructions for generating Fuel Transfer Logs using the software package SHUFFLE WORKS^^.
0 This procedure will be revised to align with the new Technical Specifications sections. The SHUFFLE WORKS^^ program will need to be reprogrammed to accommodate the new Fuel Categories and Arrays. Upon completion of programming, this procedure will require a major revision to adopt the new program operating requirements and software operating instructions.
- 7. Administrative Procedure SWI-NE-8, "Generation of Fuel Transfer Logs" is managed by the Nuclear Engineering Staff and provides instructions for creating the Fuel Transfer Logs used to control fuel placement. This procedure will be revised to include the following attributes:
align with the new Technical Specification TableIFigure numbering.
eliminate any reference to gadolinia loading (It is no longer credited in analyses) assure compliance with new Technical Specification requirements (including the appropriate steps to address LAR Enclosure 7 Commitments 1 and 2, excluding Page 8 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM the post campaign surveillance commitment addressed in the changes to SP 1 350).
RAI AHPB-2:
Do any of the proposed procedure changes require changes, additions, or deletions to personnel actions, including cognitive "actions" such as monitoring, calculating, or interpreting?
NSPM Response: Based on review of the proposed amendments and associated requirements, there are no significant changes to the types of cognitive actions taken by personnel. The types of cognitive actions required to perform fuel characterization and safe placement are described in considerable detail in the Human Performance review provided in the LAR (Enclosure 1 starting on page 10 of 28). As noted therein, the only new cognitive action with consequence deals with placement of hardware in new Arrays B through E. As discussed in the response to AHPB-1 and AHPB-3, NSPM expects that plant procedures and associated training may have to be revised to address this new cognitive action.
RAI AHPB-3:
Will changes, additions, or deletions to training be required?
NSPM Response: Prior to implementation of the proposed amendment, formal review for revised training will be performed in accordance with fleet guidance and the training program. This formal review would not be completed until the amendment approval is imminent and NSPM has confidence that the requirements and the Technical Specifications are in their final form. However, to be responsive to this question, NSPM offers below some description of training that might be expected prior to implementation.
NSPM1s initial assessment indicates that no significant training for plant operators or nuclear engineers will be necessary because the fundamental procedures and responsibilities for safely placing fuel assemblies per TS will not be significantly changed. Fuel handling will continue to be governed by existing procedures using the fuel transfer logs generated per SWI-NE-08 to place fuel within the storage racks.
These fuel transfer logs will continue to be generated by qualified nuclear engineering staff to assure all fuel is placed in acceptable locations. Therefore, the changes associated with the new allowable spent fuel arrays will be relatively transparent to fuel handling operators.
However, to address minor differences between the proposed requirements and the current requirements, NSPM expects that training will be enhanced as follows:
- 1.
Licensed Senior Reactor Operators (SROs) are responsible for Technical Specification compliance. Because Technical Specifications will change, one-time Spent Fuel Criticality Analysis - Response to RAls NSPM training will be provided to all licensed SROs to assure familiarity with the new requirements prior to implementation. To assure all future licensed operators are adequately trained, the training lesson plans associated with Fuel Handling Operations may be modified to address these new requirements.
The Nuclear Engineering staff presently generates the fuel transfer logs per plant procedure SWI-NE-8. Nuclear Engineers are qualified to perform this activity as described in qualification procedure FL-ESP-RXP-002M. Therefore, in conjunction with the proposed changes to SWI-NE-8 and the new procedure D5.3, one time training may be provided to all Reactor Engineers qualified to FL-ESP-RXP-002M prior to implementation. Future personnel qualifications may have to be conducted on the revised SWI-NE-08 and associated procedures which will address the new analysis limitations. This qualification may also qualify the Nuclear Engineering Staff to perform the new oversight function of monitoring non-fuel movement within the pool as required by the new procedure D5.3.
- 3. The requirement to maintain empty cells for several of the arrays is new (i.e., non-fuel material cannot be place in the cells credited as empty). Historically, movement of non-fuel components in the pool or placement of non-fuel components in storage racks has been controlled via the work control process. A new procedure has been proposed to manage movement of non-fuel components within the spent fuel pool (D5.3). Therefore, one time general familiarization training may be provided to all work groups who are typically involved with work in the spent fuel pool. This includes the following work groups:
Licensed Reactor Operators 0
Plant Operations 0
Plant Services 0
Nuclear Engineers Radiation Protection 0
Work Planning 0
Construction Laborers Other than that described above, no additional training is anticipated. The engineering change process will identify and document completion of this training and changes to lesson plans prior to implementation.
RAI AHPB-4:
What are the criteriahequirements for qualifying a Nuclear Engineer to be able to do the tasks involved in fuel movement?
NSPM Response: The following tasks related to fuel movement may be performed by a Nuclear Engineer:
Page 10 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM Preparation and review of the Fuel Transfer Logs. These logs provide detailed instructions dictating all fuel movement within the spent fuel pit area or reactor cavitylvessel.
Selection of fuel assemblies for dry fuel storage.
0 Perform the duties of a Fuel Accountability Engineer. This position provides independent verification of fuel movement in the spent fuel pit area when the control room is not involved in fuel movement.
The qualification requirements for each of these tasks are discussed below:
Preparation and review of Fuel Transfer Logs:
Nuclear Engineers generate the fuel move sequences per plant procedure SWI-NE-8, Generation of Fuel Transfer Logs. Nuclear Engineers are qualified to perform this activity as described in the associated qualification guide.
Selection of Fuel for Drv Fuel Storage:
Nuclear Engineers may evaluate fuel assemblies for insertion into dry fuel storage casks per plant procedure. Personnel are qualified to perform this activity as described in the associated qualification guide.
Performance of Fuel Accountabilitv Engineer Duties:
Nuclear Engineers may perform the function of the Fuel Accountability Engineer per plant instructions. Personnel are qualified to perform this activity as described in the associated qualification guide.
RAI AHPB-5:
Will the proposed LAR require changes to any personnel qualifications in support of this NSPM Response: The control of fuel placement within the storage racks will continue to be governed by existing procedures and qualifications. Placement of fuel within the storage racks is controlled by use of Fuel Transfer Logs generated by Nuclear Engineers or Senior Reactor Operators (SROs) who are knowledgeable and trained about the storage requirements. Reactor Engineers must qualify to the requirements of the qualification guide FL-ESP-RXP-002M and SROs via required knowledge of Technical Specifications. Therefore, compliance with the new arrays will be relatively transparent for personnel involved in fuel handling operations. For this reason, no changes to personnel qualifications are necessary with the possible exception of the following:
Procedure D5.3 has been proposed to govern the movement of non-fuel material in the spent fuel pool. The primary purpose is to assure spent fuel rack cells credited as empty for several arrays remain empty. For this reason, a person familiar with the storage requirements will be required to monitor non-fuel movement in the spent fuel pool and to sign off on transfer logs when placing non-fuel components in the storage Page 11 of 15 Spent Fuel Criticality Analysis - Response to RAls racks. This task is nearly identical to the existing requirements for the fuel transfer logs as used in D5.1. However, since fuel is not being moved in D5.3, a nuclear engineer knowledgeable in storage requirements via qualification to FL-ESP-RXP-002M will qualify to monitor non-fuel movement and to sign off the transfer log when necessary.
Since existing nuclear engineering and SRO qualifications will suffice, an adequate pool of personnel should be available for spent fuel evolutions with non-fuel components and there are no changes to personnel qualifications necessary. However, since the required qualifications are a subset of those specified in FL-ESP-RXP-002M, NSPM may develop new qualification guides to establish a specific pool of workers to perform the non-fuel monitoring functions of D5.3. If so, the new qualification guide will contain qualifications for all the criteria related to fuel placement in the spent fuel racks presently in qualification guide FL-ESP-RXP-002M.
RAI AHPB-6:
Will any changes fo human/sysfem inferfaces be required to support fhe proposed LAR, such as changes to boron concen frafion alarm sefpoinfs?
NSPM Response: No. As described in the LAR (Enclosure 1, page 25 of 28), man-machine interfaces are not changed. Fuel and non-fuel components will be handled as currently prescribed. As described in the LAR (Enclosure 1, page 6 of 28), no change to the minimum SFP boron concentration limit is required, and therefore, no setpoint requirements are needed. The existing boron dilution analysis remains bounding for the new loading configurations.
RAI AHPB-7:
Are there any surveillance requiremenfs to visually verify on a periodic basis fhaf a sample of ID Numbers/Confents and Locafions are consisfenf wifh the ID Numbers/Confenfs and Locafions in Shuffleworks?
NSPM Response: No. As stated in the LAR Enclosure 1 (page 12 of 28), no such surveillances are performed, which is consistent with other benchmarks in the industry.
As discussed in the LAR Enclosure 1 (starting on page 12 of 28), NSPM has reasonable assurance of fuel assembly identification by the other surveillances and procedures described therein.
RAI AHPB-8:
Provide a defailed descripfion of how the non-conservafisms in the current analysis were identified and describe the corrective and preventive actions being taken to correcf and prevent non-conservafisms in fhe proposed and future analyses. Iden fify completion dafes in the response.
NSPM Response: In August 201 0, the non-conservatisms in the spent fuel criticality analysis-of-record were discovered by Westinghouse while developing new criticality Page 12 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM analyses to accommodate the expected effects of the PlNGP Extended Power Uprate (EPU) project. At the time, consideration was being given to using the methodology of the analysis-of-record (WCAP-16517-NP) for the EPU. To validate the integrity of that analysis for supporting the additional EPU effects and the new regulatory thresholds of the Interim Staff Guidance (ISG) DSS-ISG-2010-001, Westinghouse performed an internal review and discovered that certain plant parameters used in the calculation of nuclear fuel depletion were nonconservative in relation to actual PlNGP conditions and those listed in the WCAP, Initial nonconservatisms included the values for operating core fuel temperature, core boron concentration, and the axial burnup profile. The discovery was documented in the Westinghouse corrective action program and a subsequent extent-of-condition review revealed that the value for theoretical density of the fuel was also non-conservative.
Corrective Actions. The discovered non-conservatisms have been identified in the Westinghouse and PlNGP corrective action programs. To restore SFP operability for the existing inventory of spent fuel, Westinghouse performed an interim criticality analysis with corrected, conservative values for the given parameters and produced revised burnup vs. enrichment curves for qualifying fuel to meet the required subcriticality limits in the TS. These revised curves were implemented at PlNGP in accordance with NRC Administrative Letter 98-10, as described in the LAR. One fuel assembly had to be relocated to demonstrate compliance with the interim restrictions.
Preventive Actions. To prevent non-conservatisms in the proposed future analyses, NSPM and Westinghouse carefully selected parameters that bounded actual fuel design and operating conditions for the analysis supporting the LAR. Besides the corrective action program, this careful selection of parameters was also prompted by the ISG, particularly items 2.b (Depletion Analysis - Reactor Parameters) and 3.a (Criticality Analysis - Axial Burnup Profile). Justification of these parameters is provided in to the LAR. Furthermore, to ensure that the values summarized in the licensing report were actually used in the analysis, NSPM performed an audit of the Westinghouse analyses in May of 201 I. Also, as the preparer of the analysis, Westinghouse investigated other causal factors to the condition and has implemented actions in accordance with their corrective action program.
Completion Dates. The NSPM corrective action program (CAP) tracked these non-conservatisms in three separate CAP items. One of these was completed January 19, 201 1, and another completed March 9, 201 1. The Westinghouse CAP item was closed on October 12, 201 1. The third NSPM CAP item remains open pending approval and implementation of the proposed license amendment.
RAI AHPB-9:
Provide detailed descriptions of any other corrective action items over the past five years that involve mis-positioning of items in the reactor core or the spent fuel pool.
Include status and preventive actions taken, and proposed completion dates.
Page 13 of 15 Spent Fuel Criticality Analysis - Response to RAls NSPM NSPM Response: Over the past five years, there have been no corrective action items that involve mis-positioning of items in the PlNGP reactor core or SFP. A search of the Prairie Island corrective action database and interviews with the Nuclear Engineering staff revealed no instances of mis-positioning of items in the reactor core or SFP over the aforementioned time period.
Page 14 of 15
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Westinghouse-Prepared RAI Responses Non-Proprietary Note: Enclosure 2 pages are identified in the header as "Attachment 2, Revision 1" (which relates to Westinghouse transmittal annotations) and is paginated in the footer as "[ ] of 45".
45 pages follow Revision 1 Provide an analysis to show that modeling no gadolinia is conservative. The application references an analysis contained in NUREGICR-6760 ("Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit"), which was based on a cask analysis with poison plates. These conclusions may or inay not be applicable to the PINGP spent file1 pool analysis which assumed un-poisoned racks.
Response
NUREG-6760 reviews the use of Integral Burnable Absorbers (IBAs) for their impact on reactivity and concludes:
"The analyses described in this report conclusively demonstrate that, with the exception of the Westinghouse IFBA rods, the neutron multiplication factor for an assembly without IBAs is always greater (tl~oughout burnup) than the neutron multiplication factor for an assembly with IBAs, including U02-Gd203, U02-Er203, and AI2O3-B4C rods. Therefore, for those IBAs other than IFBAs, burn~~p credit criticality safety analyses inay simply and conservatively neglect the presence of the IBAs by assuming nonpoisoned equivalent enrichment fi~el. Considering the variations in IBA assembly designs, neglecting the presence of the IBAs is an important simplifying assumption that does not add significant unnecessary conservatism."
The analyses performed in developing the Reference 1 concli~sion involved cask criticality calculations which included fixed poison panels between assemblies. In the PINGP SFP, there are no fixed neutron absorbers and so the neutron spectrum is different from that in Reference 1 and therefore may or may not be covered by the study performed in Reference 1. Westinghouse has confirmed the validity of the Reference 1 conclusions for the PINGP SFP based on the analyses described below.
Revision I Revision 1 Figure 1-1 [
Revision 1 Figure 1-2 [
Revision 1 Figure 1-3 [
Revision 1 a,c Figure 1-4 [
Revision 1 a,c Figure 1-5 [
Revision 1 a,c Figure 1-6 [
Revision 1 a,c Revision 1 Figure 1-7 [
Revision 1 Figure 1-8 [
Revision 1 a,c
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Revision 1 References
- 1. C. E. Sanders, et al., "Study of the effect of Integral Burnable Absorbers for PWR Burnup Credit",
NUREGICR-6760, Oak Ridge National Laboratory, Oak Ridge, TN, March 2002.
- 2. WCAP-17400-P, "Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", July 2011.
Revision 1 Show that the 422V+ fuel assembly design is the appropriate reference fuel assembly design to use for all proposed storage configurations.
Response
For the purposes of SFP criticality safety analyses, the Westinghouse fuel designs fall into two general categories: Optimized Fuel Assembly (OFA), and Standard (422V+). The he1 designs and their differences are discussed in Section 3.1 of Reference 1. Note that while ENC fuel is present in the pool it is significantly less reactive than 422V+ as discussed in Section 3.1 and is not considered in the response.
References
- 1. WCAP-17400-P, "Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", July 2011.
Revision 1
- 3) Availability of information related to criticality analysis of configurations such as the consolidated fuel rod storage canister is limited. Address the following to allow the NRC staff to evaluate these fuel storage configurations.
a) Provide the details of the consolidated fuel rod storage canister analysis described on page 3-21 of WCAP-17400-P.
b) In addition to the base case results shown on page 4-17 of WCAP-17400-P, provide a sensitivity analysis varying the ensichment to 3.0 weight percent 2 3 5 ~.
c) Provide the details of the analysis described on page 4-21 of WCAP-17400-P to show the reactivity effects of the fuel assembly structural materials allowed to be placed in cells adjacent to the consolidated fuel rod storage canisters.
d) Provide the details of the Array A and F interface analysis described on page 4-21 of WCAP-17400-P.
e) Provide the burnup ranges of the actual fuel rods stored in the consolidated fuel rod storage canisters and the failed fuel bin baskets.
f ) Provide legible versions of Figure 3-2, "Failed Fuel Pin Basket," Figure 3-3, "Fuel Rod Storage Canister," and Figure 3-4, "Fuel Rod Storage Canister Cross Section," from WCAP-17400-P.
Response
Consolidated fuel rod storage canisters were analyzed in a bounding manner by conservatively modeling parameters important to reactivity such as enrichment, fuel burnup, and fuel rod pitch. Additionally, the composition of the material in the non-fuel cells was selected to bound the reactivity impact of the as-loaded non-fuel cells. The maximum enrichment stored in a consolidation canister is 4.0 wt% 2 3 5 ~.
Furthermore, all of the fuel that has been consolidated has been irradiated. [
Revision 1 Figure 3-1 [
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Revision 1 References
- 1. DSS-ISG-2010-1, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools." Accession Number ML110620086, Nuclear Regulatory Colmnission, Rockville, MD, October 201 1.
Revision 1
- 4) The application does not appear to discuss the use of axial blankets. Address the following related to axial blankets:
a) If both blanketed and non-blanketed fuel assemblies are present in the pool, describe how the axial burnup profiles were modeled for both blanketed and non-blanketed assemblies at PINGP.
b) Describe how the axial enrichment variation was inodeled in the analysis for both blanketed and non-blanketed assemblies.
c) Are annular pellets used for the blanketed regions? If so, describe and justify how the void was modeled.
Response
The analysis in Reference 1 modeled the fuel in the blanket region conservatively as detailed below.
Both blanketed and non-blanketed fuel assemblies are stored in the SFP at PINGP. [
Revision 1 References
- 1. WCAP-17400-P, "Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", July 2011.
Address the following related to the selection of design axial burnup profiles:
- a. Provide the analysis that determined how limiting shape assumptions were developed.
Revision 1 Revision I Address the following related to the selection of design axial burnup profiles:
- b. Provide the actual PINGP limiting relative burnup values at lower axial levels.
Table 5-1: Axial Burnup Profiles Revision I References
- 1. WCAP-17400-P, "Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", July 2011.
- 2. J. C. Wagner, et al., "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses", NUREGICR 6801, Oak Ridge National Laboratory, Oak Ridge, TN, March 2003 Revision 1 Address the following related to rodded operation (Section 5.2 of WCAP-17400-P):
- a. Provide the analysis perfor ed to deter ine that up to 1 gigawatt-day per etric ton uraniu
( Wd
) of rodded operation would e ounded y the design asis asse lies used to develop the urnup re uire ents.
Response
[
Revision 1 Figure 6-1 [
Revision 1 arc Figure 6-2 [
Revision 1 arc Figure 6-3 [
33 of 45 Revision 1 a,c Figure 6-4 [
Revision 1 Figure 6-5 [
Revision 1 a,c L
Figure 6-6 [
Revision 1 a,c Figure 6-7 [
37 of 45 Revision 1 arc 7
Figure 6-8 [
Revision 1 a,c 7
Figure 6-9 [
Revision 1 Figure 6-10 [
lqC 40 of 45 Revision 1 arc Revision 1 Regarding criticality code validation, the Interim Staff Guidance (ISG) DSS-ISG-2010-1 states:
"An acceptable means of including isotopes that are not explicitly represented in the critical experiments used in the validation would be to increase the bias and bias uncertainty by an amount proportional to the reactivity worth of the isotopes not explicitly validated."
WCAP-17400-P addresses the effect of fission products. Justify the treatment used.
Response
[
r Revision 1 Revision I References
- 1. WCAP-17400-P, "Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", July 2011.
- 2.
WCAP-17094-P, "Turkey Point Units 3 and 4 New Fuel Storage Rack and Spent Fuel Pool Criticality Analysis", February 201 1
- 3. S. M. Bowman, et al., "Experience With the SCALE Criticality Safety Cross-Section Libraries",
NUREGICR-6686, October 2000.
- 4. J. C. Dean and R.W. Tayloe "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," NUREGICR-6698, NUREGICR-6698, January 2001 Revision I Provide the analysis results for the mis-loaded configurations (i.e., Arrays B, C, and E).
Res~onse:
[
Table 9-1: The Unborated Results of the Misloaded Configuration Calculations Revision 1 Provide a comparison of assemblies that use a range of expected burnable poison loading amounts (number of rods and weight percents) similar to Tables 3-3 and 3-5 of WCAP-17400-P.
Ressonse References
- 1. WCAP-17400-P, "Prairie Island Units 1 and 2 Spent Fuel Pool Criticality Safety Analysis", July 2011.
Marked Up Technical Specification Page Note: Changes are shown as the new note (number 8) indicated by double-underlined text. These changes are incremental to the version proposed in the original license amendment request.
1 page follows
Design Features 4.0 Figure 4.3.1-1 (continued)
Allowable Storage Arrays See notes 1 - F&below for use of Figure 4.3.1-1 Notes:
- 1. In all arrays, an assembly of higher Fuel Category number can replace an assembly designated with a lower Fuel Category number.
- 2. Category 1 is fuel up to 5.0 weight percent U-235 enrichment and does not credit burnup.
- 3. Fuel Categories 2 through 6 are determined from Tables 4.3.1-2 or 4.3.1-3.
- 4. An "R" designates a location that requires insertion of an RCCA in the fuel assembly.
- 5. An "X" designates a location that requires an empty cell, except that the empty cells in Array F may store assembly structural materials including nozzles, guide tubes, and grids.
- 6. An empty (water-filled) cell may be substituted for any fuel-containing cell in all storage arrays.
- 7. Array F shall only interface with Array A, and no other.
- 8. Except for the center rodded assembly of the 3x3 Array G and the special interface defined between Array A and Array F. each assembly location is part of up to four 2x2 arrays 4assemby in the lower right, lower left. upper right, upper left) and each assembly must simultaneously meet the r e a - u i r e m e n - m ~ a r t.
Prairie Island Units 1 and 2 Unit 1 -Amendment No.
Unit 2 -Amendment No.