ML12251A250
| ML12251A250 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/06/2012 |
| From: | Franke J Duke Energy Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC ME6527 | |
| Download: ML12251A250 (29) | |
Text
PDuke WEnergy Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 September 6, 2012 3F0912-02 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
Subject:
Crystal River Unit 3 - Response to Request for Additional Information to Support NRC Reactor Systems Branch (SRXB) Technical Review of the CR-3 Extended Power Uprate LAR (TAC No. ME6527)
References:
- 1. CR-3 to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate" (ADAMS Accession No. ML112070659)
- 2. Email from S. Lingam (NRC) to D. Westcott (CR-3) dated July 23, 2012, "Crystal River EPU LAR - Draft RAIs from SRXB Associated with Spent Fuel Storage (TAC No. ME6527)"
- 3. NRC to CR-3 letter dated August 3, 2012, "Crystal River Unit 3 Nuclear Generating Plant - Request For Additional Information For Extended Power Uprate License Amendment Request (TAC No. ME6527)"
(ADAMS Accession No. ML12213A303)
Dear Sir:
By letter dated June 15, 2011, Florida Power Corporation (FPC) requested a license amendment to increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt (Reference 1). On July 23, 2012, via electronic mail, the NRC provided a draft request for additional information (RAI) related to spent fuel storage needed to support the SRXB technical review of the CR-3 Extended Power Uprate (EPU) License Amendment Request (LAR) (Reference 2). By teleconference on July 26, 2012, FPC discussed the draft RAI with the NRC to confirm an understanding of the information being requested. On August 3, 2012, the NRC provided a formal RAI required to complete its evaluation of the CR-3 EPU LAR (Reference 3).
The attachment, "Response to Request for Additional Information - Reactor Systems Branch Technical Review of the CR-3 EPU LAR," provides the CR-3 formal response to the RAI.
This correspondence contains no new regulatory commitments.
A-ol Crystal River Nuclear Plant 15760 W. Powerline Street Crystal River, FL 34428
U.S. Nuclear Regulatory Commission 3F0912-02 Page 2 of 3 If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Superintendent, Licensing and Regulatory Programs at (352) 563-4796.
/,Yon A. Franke Vice President Crystal River Nuclear Plant JAF/krw
Attachment:
Response to Request for Additional Information - Reactor Systems Branch Technical Review of the CR-3 EPU LAR Enclosure CR-3 Spent Fuel Pool Boron Dilution Analysis xc:
NRR Project Manager Regional Administrator, Region II Senior Resident Inspector State Contact
U.S. Nuclear Regulatory Commission Page 3 of 3 3F0912-02 STATE OF FLORIDA COUNTY OF CITRUS Jon A. Franke states that he is the Vice President, Crystal River Nuclear Plant for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
Jo A. Franke ice President Crystal River Nuclear Plant The foregoing document was acknowledged before me this L04 day of SJJ
,LA_,,2012, by Jon A. Franke.
Signature of Notary Public State of Florida CARLYN E. PORTMANN Commission # DD 937553 resMarch 1, 2014 (Print, type, or stamp Commissioned Name of Notary Public)
Personally
/
Produced Known
-OR-Identification
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ATTACHMENT RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION
- REACTOR SYSTEMS BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR
U.S. Nuclear Regulatory Commission Attachment 3F0912-02 Page 1 of 7 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION -
REACTOR SYSTEMS BRANCH TECHNICAL REVIEW OF THE CR-3 EPU LAR By letter dated June 15, 2011, Florida Power Corporation (FPC) requested a license amendment to increase the rated thermal power level of Crystal River Unit 3 (CR-3) from 2609 megawatts (MWt) to 3014 MWt (Reference 1). On July 23, 2012, via electronic mail, the NRC provided a draft request for additional information (RAI) related to spent fuel storage needed to support the Reactor Systems Branch (SRXB) technical review of the CR-3 Extended Power Uprate (EPU)
License Amendment Request (LAR). By teleconference on July 26, 2012, FPC discussed the draft RAI with the NRC to confirm an understanding of the information being requested. The following provides the CR-3 formal response to the RAI needed to support the SRXB technical review of the CR-3 EPU LAR. For tracking purposes, each item related to this RAI is uniquely identified as SRXB X-Y, with X indicating the RAI set and Y indicating the sequential item number.
- 1)
(SRXB 1-1)
Section 2.8.6.2.2 of the technical report of the original EPU LAR dated June 15, 2011, states:
Studies performed in support of the evaluation discussed above demonstrate that continued use of a uniform axial burnup profile remains conservative for EPU conditions.
For bumup credit applications, uniform axial profile becomes non-limiting after accumulating some amount of depletion. Explain what is meant by this statement and provide the "studies" used to demonstrate that the uniform profile remains conservative for EPU.
Response
The statement regarding "Studies performed in support of the evaluation..." under the subheading, "Results," in Section 2.8.6.2, "Spent Fuel Storage," of the CR-3 EPU Technical Report (TR) (Reference 1, Attachments 5 and 7), refers to previous studies performed for the criticality analysis at pre-EPU conditions related to uniform (referred to herein as "flat") and distributed axial fuel burnup profiles. These studies were confirmed for applicability to fuel depleted at EPU conditions. The studies, as previously documented in the CR-3 Spent Fuel Pool Criticality Analysis Report (Reference 3), determined that the flat axial fuel burnup profile bounds the distributed axial fuel burnup profile over the enrichment and burnup ranges as shown in Table 1, "CR-3 Pool A Axial Burnup Distribution Effect," of this attachment. These studies were performed assuming fuel assemblies with axial blankets of 2.6 weight percent (wt%)
enrichment of Uranium 235 (U-235) at the top and the bottom six inches of the active fuel length, which is conservative with respect to the current CR-3 fuel assembly design containing 2.0 wt% U-235 enriched axial blankets. In addition, these studies were reviewed by the NRC staff during the review of CR-3 LAR #292 that requested modification of the fuel storage patterns in the CR-3 spent fuel pools. Based on this review, the NRC staff concluded that the flat axial fuel burnup profile is acceptable for evaluating fuel burnup greater than 30 gigawatt days per metric ton unit (GWd!MTU) (Reference 2).
Additional calculations were performed to evaluate the impact of EPU operation on the axial burnup profile of a U-235 blanketed Mark-B-HTP fuel assembly. The fuel burnup results are
U.S. Nuclear Regulatory Commission Attachment 3F0912-02 Page 2 of 7 provided in Table 2, "Axial Burnup Profile for Blanketed Assemblies - EPU," of this attachment and the comparison of the pre-EPU and EPU axial burnup profiles indicate minor deviations as shown in Figure 2.8.6.2-1, "Comparison of Pre-EPU and EPU Axial Burnup Distributions," of Section 2.8.6.2 of the CR-3 EPU TR. A bounding distributed axial profile was calculated by selecting the minimum relative burnup of each node of the various assembly burnups from the planned EPU core design. The results also indicate that the uniform axial profile continues to bound the distributed axial profile. The conclusion of these calculational studies is consistent with the results and conclusion of the non-EPU spent fuel pool criticality analysis. As such, the flat (i.e., uniform) axial burnup profile continued to be used when evaluating the impact of EPU operation on the CR-3 spent fuel pool criticality analysis.
Table 1: CR-3 Pool A Axial Burnup Distribution Effect Enrichment 2.0 2.5 3.0 3.5 4.0 4.5 5.0 (wt% U-235)
Burnup (GWd/MTU) 6.41 13.26 19.56 25.47 31.94 37.46 42.73 Bumup Profile flat flat flat flat flat flat flat k-calc 0.7801 0.7822 0.7832 0.7856 0.7810 0.7797 0.7805 stan dev 0.0006 0.0005 0.0005 0.0006 0.0006 0.0006 0.0006 Bumup Profile 40 40 40 40 40 40 40 k-calc 0.7748 0.7770 0.7770 0.7771 0.7722 0.7700 0.7695 stan dev 0.0006 0.0006 0.0005 0.0006 0.0005 0.0006 0.0005 Burnup Profile 50 50 50 50 50 50 50 k-calc 0.7770 0.7789 0.7771 0.7761 0.7717 0.7710 0.7695 stan dev 0.0005 0.0006 0.0006 0.0005 0.0005 0.0006 0.0006 Max 0.7801 0.7822 0.7832 0.7856 0.7810 0.7797 0.7805 Max Profile flat flat flat flat flat flat flat
U.S. Nuclear Regulatory Commission 3F0912-02 Attachment Page 3 of 7 Table 2: Axial Bumup Profile for Blanketed Assemblies - EPU Axial Segment Relative Burnup (cm)
Bounding Profile 0 to 15.2 0.2847
.... 0 to.!........
L5.2..
0.84.......
15.2 to i 25.06 0.7152 25.06 to 38.73 0.9099 2 5 0 _ ! t... 8 7 0 -9..9 38.73 to 58.73 1.0309 58.73 to 78.73 1.0814 78.73 to 98.73 1.0954 98.73 to 118.73 1.0985 118.73 to 138.73 1.099 138.73 to 158.73 1.0992 158.73 to 178.73 1.0997 178.73
_ to_ 198.73 1.1008 198.73 to 218.73 1.1023 218.73 to 238.73 1.1039 238.73 to 258.73 1.0998 258.73 to 278.73 1.0885 278.73 to 298.73 1.0598 298.73 to 318.73 0.9807 _
318.73 to 332.4 0.8141
_ *3,3,,2
-4,,. i 4....2..................... 0....
2_
_7.........
332.4 to 3422.6 0.6267 342.26 to 357.46 0.2515
- 2)
(SRXB 1-2)
Provide the spent fuel pool boron dilution analysis Specification 4.3.1.
supporting the revised CR-3 Technical Response:, "CR-3 Spent Fuel Pool Boron Dilution Analysis," to this attachment is provided to show the time required to reach a spent fuel storage rack multiplication factor (ker) limit of 0.95 in the CR-3 spent fuel pools during a boron dilution event. This analysis supports a CR-3 EPU licensing basis change request to credit the use of soluble boron in the CR-3 spent fuel pools to preclude spent fuel pool criticality accidents as allowed by 10 CFR 50.68(b)(4). The analysis determined that the worst case credible spent fuel pool boron dilution event is the shear of a Fire Service System line and assumes a maximum fire water pump run-out flowrate of 3100 gpm.
The analysis also assumes the initial spent fuel pool soluble boron concentration is at the minimum CR-3 Technical Specification limit of 1925 ppm boron and dilutes to a final soluble boron concentration of 571 ppm, which assures that the maximum kff is less than or equal to 0.945 under accident conditions.
The analysis concludes that the time to reach a minimum allowable boron concentration of 571 ppm is 77 minutes for Pool A, 60 minutes for Pool B, and 137 minutes when Pools A and B are connected. Assuming a conservative time of 5 minutes for the spent fuel pool water level to reach the high level alarm, plant personnel have at least 55 minutes following receipt of the high
U.S. Nuclear Regulatory Commission Attachment 3F0912-02 Page 4 of 7 level alarm to terminate the event. Thus, it is concluded that, for credible dilution sources, there is sufficient time for plant personnel to identify and terminate a boron dilution event prior to reaching the criticality limits required by 10 CFR 50.68(b)(4).
- 3)
(SRXB 1-3)
Pool A racks contain Carborundum neutron absorber panels. Provide a detailed assessment of the current and future conditions of the Carborundum neutron absorber panel in terms of the following:
- a.
uniform Boron-10 loss from the absorber panels, and
- b.
local degradation such as gapping, cracking and/or scalloping.
Response
Boron-10 Loss from Carborundum Absorber Panels Since the installation of the Carborundum high density spent fuel racks in Pool A in early 1982, the spent fuel rack boron carbide (B4C) sample surveillances have been performed on six occasions. Various sample packets have been removed and B4C poison samples examined. In addition, during each refueling outage, the gamma sample holder is relocated, as directed by the reactor engineer, to the highest dose rate area of spent fuel Pool A to ensure the samples experience the highest dose to accelerate any degradation of the Carborundum samples.
Table 3, "CR-3 Spent Fuel Pool A Carborundum Sample Data," of this attachment provides the results of the previous six B4C sample surveillances performed from 1984 to 2004.
The table shows results for gamma exposed and water spent fuel rack samples. Each gamma and water sample packet contains ten individual B4C samples positioned from the top (Sample 1) to bottom (Sample 10). Concentrating on the more limiting gamma exposed samples, nine of the ten individual samples in the table show consistent results with Sample 2 being the outlier.
Sample 2 has a vent hole, which causes damage to the individual sample directly adjacent to the hole. In 1998, Sample 2 in Packet 4 had the B4C and backing material missing, and in 2004 Sample 2 in Packet 9 had a hole completely through the sample. It is unclear if some of the damage occurred while the samples were in the gamma sample holder via flow induced erosion, or whether the damage was caused by the decontaminating process of spraying down the samples upon removal from the sample holder. Regardless of the erosion mechanism, the damage has been limited to the surface area directly adjacent to the sample packet vent holes, and has not been observed in the other samples above or below Sample 2. The actual rack vent holes are located approximately 8 inches above the active fuel height. As a result, Sample 2 is excluded from the total average weight loss column in Table 3 and a separate column is provided showing the average weight loss associated with Sample 2.
Table 3 also shows a large percent loss over the first two years from 1982 to 1984, but greatly reduced and consistent percent losses from 1984 to 2004 excluding Sample 2. This apparent large reduction in weight during this two year period was due to improperly establishing the initial baseline density of the pre-exposed samples. During the first surveillance period, the vendor did not properly dry the samples to remove excess moisture prior to weighing the pre-exposed samples to establish an initial baseline density. Therefore, the actual pre-exposed densities are unknown, and the measured weight loss over each subsequent surveillance period is
U.S. Nuclear Regulatory Commission Attachment 3F0912-02 Page 5 of 7 conservatively high compared to their pre-exposed weights.
As a result, the more accurate measure of B 4C weight loss is shown from 1984 forward. Including the Sample 2 data and the conservative pre-exposed densities, the Carborundum neutron absorber panel sampling history indicates approximately a 6% average B4 C weight loss over the 22 year period.
Local Degradation of Carborundum Absorber Panels During each high density spent fuel rack B4C sample surveillance, each individual sample surface is also examined for the appearance of texture, discoloration, cracking, scalloping, spalling (chipping), blisters, voids, and separation of the B4C granular surface from the fiberglass backing. Excluding Sample 2, the other individual samples have shown consistent results with some incidences of discoloration, but no major flaws.
As for swelling, the B4C sample surveillances do not include a thickness measurement; however, in reviewing the 2004 samples against the single CR-3 unexposed B4C sample, there is no visual indication of swelling through 2004. In addition, CR-3 has not experienced stuck assemblies or the inability to insert assemblies in any cells of the high density spent fuel racks due to swelling as experienced at other facilities.
As for gapping, the Carborundum sheets used at CR-3 consist of one continuous piece versus the Carborundum panels that are stacked atop one another to form the full length panel. As such, individual panel shrinkage makes the panel type material more susceptible to gaps and not the sheet type material. For gapping to occur in the sheet type material requires cracking completely separating the individual sheet. To date, CR-3 has not experienced any cracking in the spent fuel rack sample.
In-situ Boron-10 Areal Density Gauge for Evaluating Racks (BADGER) testing can reveal if any of the Carborundum neutron absorber sheets have gaps or separations from cracking. Per the CR-3 license renewal application and associated RAI response letters, FPC has committed to BADGER testing as part of the enhancement of the Fuel Pool Rack Neutron Absorber Monitoring Program, which will be implemented prior to CR-3 extended operation.
As described in the FPC to NRC letter dated January 27, 2010, regarding an RAI associated with the CR-3 license renewal application (Reference 4), the BADGER surveillance test interval will be performed every 10 years after the first BADGER test performance prior to CR-3 extended operation. The BADGER test will also be staggered with a 10 year interval for the B4C sample surveillances, staggering them at 5 year intervals to ensure either B4C sample surveillances or BADGER testing is performed every 5 years.
U.S. Nuclear Regulatory Commission 3F0912-02 Attachment Page 6 of 7 Table 3: CR-3 Spent Fuel Pool A Carborundum Sample Data Gamma Samples Sample Sample Sample Total Individual Sample Weight Losses Average Average Loss -
Removal Age Packet Gamma 1
2 3
4 5
6 7
8 9
10 Loss Sample 2 Date (Years)
Number Dose (Rad)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
04/15/1982 0.0 02/17/1984 1.8 1
6.74E+09 2.670 2.870 2.520 2.260 2.290 2.350 2.250 2.230 2.290 2.370 2.410 2.359 11/04/1985 3.6 10 l.OOE+10 2.201 2.669 2.367 2.109 2.214 2.118 2.220 2.419 2.195 2.098 2.261 2.216 02/19/1988 5.8 2
1.43E+10 2.807 4.362 2.744 2.897 2.525 2.623 2.729 2.506 2.539 2.709 2.844 2.675 05/14/1993 11.1 3
2.02E+10 4.040 10.250 3.530 6.230 3.720 3.590 3.650 3.640 3.600 3.740 4.599 3.971 06/18/1998 16.2 4
3.32E+10 4.912 17.860 4.420 4.560 4.496 4.450 4.190 4.096 4.190 4.005 5.718 4.369 05/20/2004 22.1 9
4.02E+10 5.537 21.010 5.066 4.263 4.259 4.329 4.365 4.331 4.740 4.491 6.239 4.598 Totals from 1988 to 1998 1.89E+10 2.105 13.498 1.676 1.663 1.971 1.827 1.461 1.590 1.651 1.296 2.874 1.693 Totals Projected to 2036 (Extended Life) 1.01E+I 1 12.273 64.204 10.429 9.585 10.566 10.175 9.040 9.419 10.023 8.638 15.435 10.017 Water Samples Sample Sample Sample Total Individual Sample Weight Losses Average Average Loss -
Removal Age Packet Gamma 1
2 3
4 5
6 7
8 9
10 Loss Sample 2 Date (Years)
Number Dose (Rad)
(
M)
(%)
M%)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
(%)
04/02/1982 0.0 02/17/1984 1.8 13 82.31 2.200 2.270 2.310 2.150 2.100 2.150 2.270 2.280 2.090 2.120 2.194 2.186 11/04/1985 3.6 21 157.45 1.787 1.989 1.815 1.756 1.728 1.619 1.661 1.699 1.620 2.105 1.778 1.754 02/19/1988 5.8 15 445.04 2.627 2.839 2.464 2.376 2.265 2.296 2.562 2.294 2.169 2.262 2.415 2.368 05/14/1993 11.1 16 1590.63 3.850 3.940 3.510 3.150 2.710 2.930 3.290 2.860 2.680 3.010 3.193 3.110 06/18/1998 16.2 18 1858.6 5.000 4.450 3.590 3.000 2.990 3.037 2.790 2.660 3.120 4.350 3.499 3.393 05/20/2004 22.1 19 2040.8 5.414 4.970 2.708 2.221 2.235 2.394 2.637 2.908 3.168 4.578 3.323 3.140
U.S. Nuclear Regulatory Commission Attachment 3F0912-02 Page 7 of 7
- 4)
(SRXB 1-4)
Provide a quantitative assessment of the impact of panel degradation mechanisms noted in RAI 3 on the criticality analysis of record. Estimate the date on which the worst panel will exceed the assumed condition in the criticality analysis or record.
Response
As shown in Table 3, "CR-3 Spent Fuel Pool A Carborundum Sample Data," of this attachment, B4C sample surveillance periods from 1988 to 1998 represent the steepest rate of B4C weight loss over the 22 years.
Table 3 shows the total accumulated dose along with the overall individual and average B4C weight loss for the ten year period. These totals exclude Sample 4 from 1993, which, along with Sample 2, is an outlier compared to the other nine samples from 1993 and the other Sample 4 surveillances throughout the 22 year period. Projecting this ten year weight loss rate from 2004 to the end of the CR-3 period of extended operation, including the Sample 2 data and conservative pre-exposed densities, the approximate average weight loss through 2036 is expected to be approximately 15.5% average B4C sample weight loss with a total accumulated gamma dose of approximately 1.OOE+l 1 rads. Based on previous vendor testing of this type of poison samples, a 20% weight loss of a B4C sample is equivalent to a 15%
Boron-10 material degradation.
Therefore the CR-3 spent fuel rack Carborundum absorber panels are expected to remain within the 15% Boron-10 material degradation assumption of the spent fuel pool criticality analysis through the extended period of operation at the EPU power level.
Based on the B4C individual sample surface examinations during the previous surveillances and over 22 years of exposure to CR-3 spent fuel pool conditions, the samples, with the exception of Sample 2, did not experience loss of B4C grains leaving appreciable voids, scalloping, or spalling. Also, the B4 C samples did not contain appreciable cracks, blisters, or separation of the B4C granular surface from the fiberglass backing.
To date the Sample 2 damage has been limited to the surface area directly adjacent to the sample packet vent holes, and no such degradation has been observed in any of the other samples. The actual rack vent holes are located approximately 8 inches above the active fuel height, and if such deterioration were occurring in the racks, this Sample 2 degradation mechanism would have to be global and reach down through the lower samples (3 through 10) to reach into the top of the active fuel region.
Therefore, it is reasonable to conclude that these local degradation mechanisms will not adversely impact the capability of the Carborundum neutron absorber panels from performing their function and that the spent fuel pool criticality analysis will remain valid through the extended period of operation at the EPU power level.
In addition, the response to RAI B.2.37-2 associated with the CR-3 license renewal application, FPC to NRC letter dated January 27, 2010 (Reference 4), provides a description of how the material condition of the Carborundum neutron absorber panels will be monitored during the period of extended operation. To avoid duplication of NRC staff reviews, FPC proposes further questions regarding the estimated schedule on when the degradation of the Carborundum neutron absorber panels will exceed the assumed condition in the criticality analysis of record be included as part of the CR-3 license renewal application review.
U.S. Nuclear Regulatory Commission Attachment 3F0912-02 Page 8 of 7 References
- 1.
FPC to NRC letter dated June 15, 2011, "Crystal River Unit 3 - License Amendment Request #309, Revision 0, Extended Power Uprate."
(ADAMS Accession No. MLI 12070659)
- 2.
NRC to FPC letter dated October 25, 2007, "Crystal River Unit 3 -
Issuance of Amendment Regarding Fuel Storage Patterns in the Spent Fuel Pool (TAC No. MD3308)."
(ADAMS Accession No. ML072910317)
- 3.
Holtec Report HI-2063559, Revision 1, "Criticality Analysis of Additional Patterns for Crystal River 3 Pools A & B," dated September 19, 2006. (Proprietary)
- 4.
FPC to NRC letter dated January 27, 2010, "Crystal River Unit 3 - Response to Request for Additional Information for the Review of the Crystal River Unit 3, Nuclear Generating Plant, License Renewal Application (TAC NO. ME0274) and Amendment #9." (ADAMS Accession No. ML100290366)
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 ENCLOSURE 1 CR-3 SPENT FUEL POOL BORON DILUTION ANALYSIS
SYSTEM#
CALC. SUB-TYPE PRIORITY CODE QUALITY CLASS SF N/A 3
S NUCLEAR GENERATION GROUP Fll-0001 (Calculation #)
CR-3 SDent Fuel Pool Dilution Analysis (Title including structures, systems, components)
FII BNP UNIT NCR3 [I] HNP FI]RNP I] NCP W-1ALL APPROVAL M Electronically Approved REV PREPARED BY REVIEWED BY SUPERVISOR Signature Signature Signature 0
Name Name Name Tyson Huntsman/
Lewis Wells Michael T. Floyd Ryan A. Stephens Date Date Date 09/05/2012 09/05/2012 09/05/2012 (For Vendor Calculations)
Vendor Vendor Document No.
Owner's Review By Date
Calculation No.
Fll-0001 Page ii of iv Revision 0
TABLE OF CONTENTS Table of Contents ii Revision Summary iii Document Indexing Table iv Purpose 1
Assumptions 1
Design Inputs 1
Scenarios 4
Analysis 5
Results / Conclusions 6
References 9 - Record of Lead Review 3 pages
Calculation No.
F1 1-0001 Page iii of iv Revision 0
Revision Summary Rev. #
Revision Summary (list ECs incorporated) 0 Initial issuance of this calculation in support of the Spent Fuel Pool Criticality Analysis for EPU conditions.
Calculation No.
Fll-0001 Page iv of iv Revision 0
Document Indexing Table Document ID Number Function Relationship to Calc.
Action Type (e.g., Calc No., Dwg.
(i.e. IN for design (e.g. design input, assumption basis, (specify if Doc. Services No., Equip. Tag No.,
inputs or references; reference, document affected by (e.g. CALC, DWG, Procedure No.,
OUT for affected results) or Config. Mgt. to Add,
- TAG, Sotaenm n
ouet)Deleted or Retain) (e.g.,
PROCEDURE, Software name and documents)
CM Add, DS Delete)
SOFTWARE) version)
TAG FSP-1 IN DESIGN INPUT ADD TAG FSP-2A IN DESIGN INPUT ADD TAG FSP-2B IN DESIGN INPUT ADD PROCEDURE OP-406 IN REFERENCE ADD PROCEDURE OP-418 IN REFERENCE ADD PROCEDURE OP-880 IN REFERENCE ADD DWG 302-182 IN DESIGN INPUT ADD DWG 302-231 IN DESIGN INPUT ADD DWG 302-621 IN DESIGN INPUT ADD DWG 521-110 IN DESIGN INPUT ADD DWG 521-111 IN DESIGN INPUT ADD CALC 191-0006 IN DESIGN INPUT ADD CALC M98-0033 IN DESIGN INPUT ADD CALC M99-0008 IN DESIGN INPUT ADD CALC F97-0014 IN DESIGN INPUT ADD CALC HNP-M/MECH-IN DESIGN INPUT ADD 1099 CALC M98-0055 IN DESIGN INPUT ADD OTHER ITS IN DESIGN INPUT ADD OTHER 00031-000 IN DESIGN INPUT ADD I
F 4
F F
F 4
F I
F (For the purpose of creating cross references to documents in the Document Management System and equipment in the Equipment Data Bas
Calculation No.
Fl1-0001 Page 1 of 9 Revision 0
1 Purpose The purpose of this calculation is to identify and calculate the bounding dilution flow rates for the Crystal River 3 (CR3) Nuclear Plant spent fuel pools (SFPs), and perform a parametric study of times required to reach specified critical spent fuel pool boron concentrations. This analysis supports a request for a Technical Specification change which will credit boron concentration as a method to maintain a sub-critical configuration in the A and B spent fuel pools as a result of Extended Power Up-rate (EPU) at CR3. The analysis will be performed as directed by the Kopp Memo (REF 7.4.4), and will consider all possible dilution initiating events (including operator error).
It will also justify the surveillance interval for verifying the Technical Specification minimum pool boron concentration is maintained. The analysis will be performed such that it will apply to any spent fuel pool configuration; 'A' and 'B' pools connected, or 'A' and 'B' pools separated by a physical barrier.
2 Assumptions 2.1 SFP floor drains do not exist To be conservative, it is assumed that none of the water from a pipe break is entering the floor drains and that the entire volume of water from the break is entering the SFP volume. This assumption will reduce the time required to reach the minimum critical boron concentration.
2.2 The boron concentration is homogenous throughout the spent fuel pool To be conservative, it is assumed that all water mixes instantly, thereby neglecting additional time required for the pool volume to reach boron equilibrium.
2.3 The initial SFP level is at the low level alarm setpoint To conservatively reduce the time necessary to reach minimum critical boron, it is assumed that the spent fuel pool level is at the low level alarm setpoint. Plant procedures set a minimum spent fuel pool level 156' 6" ( REF 7.1.1) at which operator action is required.
3 Design Inputs 3.1 Nuclear Services Closed Cycle Cooling Water System (SW)
The Nuclear Services Closed Cycle Cooling Water System (SW) is an intermediate cooling system that removes heat from safety and non-safety related components during all plant operation. The SW system is a closed loop system in order to help prevent direct leakage of radiation from nuclear support systems in the plant to the environment. Though the SW system is not used directly to mitigate any accidents, it is used to provide cooling to components in systems that are necessary to mitigate an accident.
Calculation No.
Fll-0001 Page 2 of 9 Revision 0
Water from both the Spent Fuel Cooling System and the SW System flows through the Spent Fuel Heat Exchangers (SFHE 1A & 1B). Table 3.1-1 shows the heat exchanger characteristics for both the tube side (SF) and the shell side (SW) (REF 7.3.2, 7.3.3, and 7.4.2).
Table 3.1-1 SFHE 1A/1B Operating Characteristics Tube Shell 7.5 x 10A5 7.5 x 10A5 Mass Flow Rate Ibm/hr Ibm/hr Pressure 77 Psig 189 Psig T in 127.2 Deg-F 95 Deg-F TOut 115.5 Deg-F 106.7 Deg-F No. of Tubes 148 Tube OD 3/4in Tube Thickness 20 BWG. Avg.
- 0.035 in Tube Length 13.58 ft
- From Cameron Hydraulic Data 3.2 Fire Service System (FS)
The Fire Service System will be used as the bounding analysis for this calculation. The analysis will be based on operating the bounding 2000 GPM Fire Service pump at run-out conditions.
Pump run-out of the 2000 GPM FSP is 3100 GPM based on inspection of the pump curve located in REF 7.3.6. The pump discharge pressure of 220 ft at the flow rate of 3100 GPM is estimated to be equivalent to piping losses and elevation change for a pipe shear on the SFP building floor, and is the maximum flow rate capability of the system.
3.3 Demineralized Water System (DW)
The DW system is used to provide demineralized water to the plant. It supplies demineralized water to the plant through the use of two identical 480V electric motor driven pumps. Each of these pumps is rated at 150 GPM and 300' total developed head (REF 7.3.7). For the purpose of this analysis, the Demineralized Water System will be bounded by the 2000 GPM Fire Service System pumps, and therefore will not be evaluated.
3.4 Spent Fuel Pool Design The Fuel Handling Building consists of two spent fuel pools (Pools 'A' & 'B') that can be used to store spent fuel. The two pools are connected by a canal that can also have a gate inserted to separate the two pools. Additionally, the B pool is connected to the Cask Loading Pit by another canal with a removable gate. These gates are used to allow draining of an individual area of the pool without lowering the water level in the remainder of the system.
Typical operation at Crystal River 3 has both gates removed so that the A and B pools and the
Calculation No.
Fll-0001 Page 3 of 9 Revision 0
Cask Loading Pit are all connected. To be conservative for this calculation, analysis will be preformed assuming that the cask pit gate is in place and it is isolated from the A and B pools.
The spent fuel pools and transfer canals are all built as a concrete structure with a stainless steel liner. The transfer canals are designed such that their bottom depth is above the top of the fuel racks. This is to prevent a loss of water in one section of the pool from uncovering the fuel in the other section of the pool if the removable gate is not in place. Table 3.4-1 was generated by input data taken from CR3 calculation F97-0014, CR-3 Spent Fuel Pool Temperature Rise from Fuel in the Pool, and used to determine the total volumes of the spent fuel pool.
Table 3.4-1 Total Raw Spent Fuel Pool Volumes SSC L (N-S)
Wume Volume W)
Det Voue Vlm ft Ft ft Cu-ft gal Pool'A' 23.917 32.125 38.166 29324.2 219360.4 Pool 'B' 23.917 32.500 38.166 29666.5 221921.0 Cask Pit 13.000 13.000 38.166 6450.1 48249.8 Cutout Pool 'B' 23216.5 173671.3 Total I
I Total A+B 52540.7 1 393031.7]
Table 3.4-1 gives the overall dimensions and volumes of the spent fuel pools (REF 7.3.4).
The area included in Pool 'B' must have the area occupied by the 13' by 13' cask pit subtracted to get the actual area of Pool 'B'. The Cask Loading Pit was not included in the total volume because it is assumed that the gate is in place, separating it from the rest of the SFP. Table 3.4-2 gives the partial Spent Fuel Pool volumes. These values are adjusted for spent fuel racks, spent fuel assemblies, and transfer tube platforms. For conservatism in Tables 3.4-1and 3.4-2, it was assumed that the water level in the SFP is on the low level alarm setpoint, resulting in reduced operator response times. Input data and detailed calculations of the adjusted 'A' and 'B' spent fuel pool volumes can be found in calculation F97-0014, Revision 8. The adjusted volumes given in Table 3.4-2 will be used in the dilution analysis when determining boron dilution times.
Table 3.4-2 Spent Fuel Pool Volumes Total Adjusted SSC Volume Volume I Cu-ft gal Pool 'A' 26602.0 198996.8 Pool 'B' 20612.2 154190.0 Total 47214.2 353186.7
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4 Scenarios 4.1 Break in the Spent Fuel Cooling System in heat exchangers The Spent Fuel (SF) Cooling System consists of two independent and redundant cooling loops. The SF system interacts with the Service Water (SW) system through the Spent Fuel Heat Exchangers (SFHEs) (REF 7.2.1). The heat exchangers have a tube and shell design in which a break in the tubes could result in leakage of unborated water into the spent fuel pools.
The shell side has a design pressure of 200 psig. The tube side design pressure of 125 psig (REF 7.4.2). A break in a tube within the heat exchanger could result in an unborated water leak into the spent fuel system due to the higher pressure on the shell side of the tubes. This scenario is bounded by the FSW line shear of 3100 GPM because the estimated max flow for this scenario is 100 GPM.
4.2 Break in the Fire Service piping The Fire Service System (FS), consists of three 2.5" standpipes and the associated hose reels (REF 7.2.3). One is located along the north wall, another along the east wall, and the last along the southern most west wall. A break in one of these pipes is bounded by the 3100 GPM Fire Service pump run-out analysis.
4.3 Break in Demineralized Water piping Another possible source is the Demineralized Water System (DW). This system includes a 2" header running along the north side of the spent fuel pool. This header consists of one 1.5" valve and four 3/4" valves. Additionally, there are also another five individual 3/4" standpipes, each containing a single valve (REF 7.2.2). Four of these are along the east wall of the FHB and the last is along the southern most west wall. A break in one of these pipes is bounded by the 3100 GPM Fire Service pump run-out analysis.
4.4 Misalignment of valve There are a total of ten demineralized water valves that could be possible sources of unborated water to the SFP if misaligned. The valve configuration of this system is as described in scenario 3. There is one 1.5" valve that is a drain valve for the DW system. The nine remaining 3/4" valves are all hose connection isolation valves. These hose connections are typically used to wash down the areas in the FHB. All ten of these DW valves are normally administratively closed by REF 7.1.2. The Fire Service System consists of three separate 2.5" standpipes and their associated hose reels. Each of the three standpipes is reduced to an associated 1.5" hose reel isolation valve that is administratively sealed closed by REF 7.1.3. Misalignment of any of these valves is bounded by the 3100 GPM Fire Service pump run-out analysis.
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F1 1-0001 Page 5 of 9 Revision 0
4.5 Break in FHB roof storm drain The FHB roof storm drains are an additional possible source of dilution. Because these storm drains are only gravity fed, a break in these lines in considered highly unlikely. As a result, a break in a storm drain will not be considered as a credible source of SFP boron dilution.
4.6 Tank ruptures in the vicinity of the pool All large capacity tanks at the Crystal River 3 Nuclear Plant are located outside of the area that could affect the SFP. Therefore, a tank rupture is not considered to be a credible dilution source to the SFP.
4.7 Dilution events initiated in the Reactor Coolant System There are two (2) transfer tubes that separate the Reactor Building (RB) Fuel Transfer Canal (FTC) from the spent fuel pool. Each of these tubes is designed with a gate valve and a blind flange to separate the FTC from the SFP when not in refueling operations. During the course of a refueling outage these valves may be opened for 3-4 weeks to permit fuel transfer between the spent fuel pool and the reactor vessel. During this time, there is the possibility of a dilution of the Reactor Coolant System (RCS) that would directly affect the SFP. Technical Specification 3.9.1 requires RCS boron concentration to be maintained above the Core Operating Limits Report (COLR) limit during Mode 6. Procedure FP-203, Offloading and Refueling Operations, administratively requires caution tags to be issued for applicable valves that may cause a dilution of the RCS during Mode 6. The possibility of dilution of the SFP due to a dilution of the RCS is considered to be minimal and is therefore not analyzed.
5 Analysis All scenarios described in Section 4 are either bounded by the pump run-out analysis for the Fire Service System given in Section 3.2, or they are not considered to be credible dilution sources. The bounding analysis is estimated to be conservative by approximately a factor of 10 to the estimated worst case realistic SFP dilution flow rate of approximately 300 GPM.
5.1 Pump Run-out Bounding analysis will be performed by determining the flow rate of water from a single 2000 GPM FS pump at run-out conditions introduced into the Spent Fuel Pool. Run-out conditions are determined using the pump curve for the FS pumps. Fire Service pump curves are found in Attachment A of calculation M98-0055, Fire Service Design Pressure and Temperature, and show a flow rate of 3100 GPM.
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F1 1-0001 Page 6 of 9 Revision 0
5.2 Boron Dilution The boron dilution rate for a constant volume is based on the following differential equation (REF 7.3.5):
Equation 5.1 Vdc
= -QC + QCl, dt Where:
C= boron concentration of mixing volume (ppm)
Q= flowrate of water into mixing volume (GPM)
V= finite mixing volume (gallons)
Cin= boron concentration of water entering mixing volume (ppm)
This equation assumes the boron concentration remains homogenous throughout the volume.
It can also conservatively be assumed that the boron concentration of the water entering the mixing volume is equal to zero, thereby removing the QCin term from the equation. Solving Equation 5.1 above for time (t) yields Equation 5.2 below.
Equation 5.2 Where C, is the initial boron concentration of the mixing volume at t=0. The final boron concentration, Cf, is the final concentration desired in the mixed volume. When determining time for operator response, ratios of the final to initial boron concentration can be used within the formula. This will allow the calculation results to be used for a variety of initial and final boron concentrations.
6 ResultslConclusions 6.1 Pump Run-out Pump run-out for the Fire Service pump was determined from calculation M98-0055, Fire Service Design Pressure and Temperature. Attachment A of M98-0055 contains pump performance curves for the three 2000 GPM pumps. Run-out conditions correspond to the last point on the pump curves of 3100 GPM.
6.2 Boron Dilution Boron dilution analysis is completed using Equation 5.2 and the known spent fuel pool volumes and flow rate for the FS pump run-out. Table 6.2-1 was generated based on ratios of initial (C, ) and final (Cf) boron concentrations.
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F1l1-0001 Page 7 of 9 Revision 0
Table 6.2-1 Boron Dilution Analysis Pool A Pool B Pool A &B Cf/Co connected min min min 0.1 148 115 262 0.2 103 80 183 0.3 77 60 137 0.4 59 46 104 0.5 44 34 79 0.6 33 25 58 0.7 23 18 41 0.8 14 11 25 0.9 7
5 12 1.0 0
0 0
The Technical Specification boron concentration limit is 1925 ppmB. The minimum allowable boron concentration to maintain sub-criticality in the SFP is identified as 571 ppmB by REF 7.4.3. These boron concentrations yield a limiting Cf/Co ratio of 571/1925 or 0.296.
6.3 Spent Fuel Pool Fill Times The time required to reach the high level alarm setpoint of the SFP can be determined based on the flow rates of water into the SFP. Per REF 7.3.1, the Hi-Level Alarm setpoint is 159' 0".
The required volume change in the pool to reach the Hi-Level Alarm point can be determined by multiplying the difference in elevations, 2.5 feet, by the area of the pool in the given configuration. Given this required volume change, necessary operator response time due to actuation of a Hi-Level Alarm can be determined for given flow rates into the SFP. All times to reach the Hi-Level Alarm have been rounded up to the nearest minute to be conservative.
The area of Pool 'A' is Length (23.92 ft) x Width (32.125 ft), or 768.3 Sq-ft. This results in a volume addition of 1920.8 Cu-ft or 14,369 gallons in order to reach the Hi-Level Alarm. For this volume and a flow rate of 3100 GPM, it would take 5 minutes before the alarm is received by the operators.
To create a limiting scenario for the area of Pool 'B', the required fill volume is determined by including the area of the cask pit and cask pit gate as well. The area of Pool 'B' is Length (23.92 ft) x Width (32.5 ft) - Cask Pit cutout (13' x 13') + Cask Pit (10' x 10') + Cask Pit Gate (3' x 3'), or 717.2 Sq-ft (REF 7.2.4 and 7.2.5). This results in a volume addition of 1793.2 Cu-ft or 13,415 gallons in order to reach the Hi-Level Alarm. For this volume and a flow rate of 3100 GPM, it would take 5 minutes before the alarm is received by the operators.
The area with Pool 'A' & 'B' connected will be made conservative by also adding the area of the transfer canal between the two pools as well as the area of the cask loading pit and it's transfer canal. This will result in the maximum time to reach the Hi-Level Alarm. The 'A' to 'B' canal is 4' by 3', the cask pit to 'B' canal is 3' by 3', and the cask loading pit is 10' by 10' (REF
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7.2.4 and 7.2.5). Adding these areas to the already determined areas of the 'A' and 'B' pools yields a total area of 1,497.6 Sq-ft. This results in a required volume change of 3,744 Cu-ft, or 28,007 gallons. For this volume and a flow rate of 3100 GPM, it would take 10 minutes before the alarm is received by the operators.
The bounding scenario to reach the Hi-Level alarm is 10 minutes when Pools 'A' and 'B' are connected and 5 minutes when they are isolated. This is the maximum time it would take for operators to realize they have a leakage source into the SFP. At this time, operators would take action to determine the source of in-leakage and prevent further dilution of the SFP. The minimum boron concentration required to maintain sub-criticality is identified in REF 7.4.3 as 571 ppmB. Using this limiting boron concentration, and a starting boron concentration at the Technical Specification limit of 1925 ppmB (REF 7.4.1) yields Table 6.3-1 below.
Table 6.3-1 Comparison of Alarm Time to Time to Reach Minimum Boron Time to Reach Time to Minimum Pool Alarm Allowable Configuration Boron min min Pool 'A' 5
77 Pool'B' 5
60 Pool 'A + B' 10 137 Table 6.3-1 shows that for the worst case, plant personnel will have 55 minutes following notification of a dilution event to mitigate the effects of that event.
As a result of the bounding analysis performed in this calculation, it can be concluded that for all possible sources of spent fuel pool boron dilution, there will be sufficient warning to plant personnel of a dilution event before the minimum allowable boron concentration of 571 ppmB (REF 7.4.3) is reached. The Technical Specification surveillance interval of seven days for verifying the minimum SFP boron of 1925 ppmB is maintained is sufficient because a dilution event cannot occur without generating a Hi-Level Alarm and notifying operators.
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7 References 7.1 Procedures 7.1.1 OP-406 Rev. 89, Spent Fuel Cooling System 7.1.2 OP-418 Rev. 48, Demineralized Water System 7.1.3 OP-880 Rev. 66, Fire Service System 7.2 Drawings 7.2.1 302-621 Rev. 54, Spent Fuel Cooling 7.2.2 302-182 Rev. 64, Condensate and Demineralized Water Supply for Nuclear Plant 7.2.3 302-231 Rev. 91, Fire Service Water 7.2.4 521-110 Rev. 10, Spent Fuel Pit - Floor Details 7.2.5 521-111 Rev. 05, Spent Fuel Pit - Liner Details 7.3 Calculations 7.3.1 191-0006 Rev. 1, Spent Fuel Storage Pool A&B Level Loop Accuracy 7.3.2 M98-0033 Rev. 2, Spent Fuel Cooling (SF) Design Pressure and Temperature 7.3.3 M99-0008 Rev. 0, Nuclear Services Closed Cycle System (SW) Design Pressure and Temperature 7.3.4 F97-0014 Rev. 8, CR-3 Spent Fuel Pool Temperature Rise from Fuel in the Pool 7.3.5 HNP-M/MECH-1099 Rev. 0, Spent Fuel Pool Boron Dilution Analysis 7.3.6 M98-0055 Rev. 0, Fire Service Design Pressure and Temperature 7.3.7 M93-0021 Rev. 2, Auxiliary Building Demineralized Water Storage Tank ( DWT-1)
Volume 7.4 Other Documents 7.4.1 ITS, Improved Technical Specifications 7.4.2 00031-000 Rev. 2, Installation, Operation & Maintenance of Shell and Tube Heat Exchangers 7.4.3 Holtec Report, HI-2063559 Rev. 2, Criticality Analysis of Additional Patterns for Crystal River 3 Pools A & B.
7.4.4 NRC Memo, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, Acc. # ML072710248.
Calculation No.
F1l1-0001 Attachment 1
Page 1 of 3 Revision 0
Record of Lead Review Document F11-0001 Revision 000 The signature below of the Lead Reviewer records that:
the review indicated below has been performed by the Lead Reviewer; appropriate reviews were performed and errors/deficiencies (for all reviews performed) have been resolved and these records are included in the design package; the review was performed in accordance with EGR-NGGC-0003.
Design Verification Review L1 Design Review F-1 Alternate Calculation Mi Qualification Testing E--
Engineering Review El Owner's Review Ej Special Engineering Review_______________________
L-I YES L-I N/A Other Records are attached.
I,Anwic WaIlkII Safety Analysis 09/04/2012 Lead Reviewer (print/sign)
Discipline Date Item No.
Deficiency Resolution
-q1 Need to reference the Kopp memo (probably under "Purpose" and address every aspect of what is says should be considered:
"The analysis should consider all possible dilution initiating events (including operator error), dilution sources, dilution flow rates, boration sources, instrumentation, administrative procedures, and piping. This analysis should justify the surveillance interval for verifying the technical specification minimum pool boron concentration."
How is the surveillance interval addressed and justified?
Reference was added to the purpose.
Surveillance interval was justified because a dilution event cannot occur without reaching the high level alarm and notify operators.
Calculation No.
Fll-0001 Attachment 1
Page 2 of 3 Revision 0
2 There should be an acceptance criteria section which explains why 571 ppm was used as the final concentration (Cf).
Suggestion for wording:
"The worse case misloaded assembly required boron concentration is greater than 571 ppmB (Pool B) to maintain a keff less than 0. 95. Since it is a higher boron concentration than that required for Pool A, use of 571 ppmB will result in the least amount of time from event initiation to the lowest acceptable final boron concentration (Cf). Per IOCFR 50.68(b) (4), The kef for a fully unborated condition must still be less than 1.0. "
I don't think there is an acceptance criteria for this calc. The 571 ppm is an input value to the calculation and produces our results. Making 571 an acceptance criteria makes it seem to me like that is the number we are trying to calculate.
I would think if anything it would be an assumption, however I believe the calc is sufficient as is. If you disagree we can discuss this further.
3 Under "Purpose", add in that the licensing Done basis for ITS 4.3.1 is changing for the EPU to credit borated water.
4 Section 2.3. I do not believe that the low Removed reference to ITS.
level of 156' 6" bounds the ITS limit (156') in that use of the ITS value would result in less time for operator actions. Suggest removing the sentence about the ITS.
5 Need to have the Holtec report as a Done reference.
6 Is there any criterion on timing to identify I have discussed time requirements with Ops and terminate a dilution event? The final and we will be implementing the necessary paragraph states "...there will be sufficient actions in the site APs. The necessary changes time for plant personnel to indentify and will be made to plant procedures as directed by, terminate..." How do we know this? Without EC 71193. I will change it to say "... plant this information, one can postulate that it will personnel will have 55 minutes to identify and take the plant longer than 55 minutes to terminate..." so that it does not appear we are react to a SFP level alarm.
saying there is sufficient time without a basis.
7 Please add references for 10x10, 3x3 and References 7.2.4 and 7.2.5 were added.
4x3 volume adders.
8 Section 5. Please provide details as to how Changed wording to "... 10 to the estimated the bounding analysis is conservative by a worst case realistic SFP dilution flow rate of factor of 10 to the worst case realistic approximately 300 GPM.
condition.
Calculation No.
F1l1-0001 Attachment I
Page 3 of 3 Revision 0
9 Not a deficiency. Independently performed a volume calculation (alternate calculation) and determined that the volumes are as follows:
Pool A - 23558 ft3 Pool B - 19003 ft3 Total - 42561 ft3 The values calculated by in this calculation are smaller and are therefore conservative and appropriate. The differences are attributable to conservatisms introduced ignoring different volumes of water. Note that the raw volume calculations have already been design verified and my check was to confirm their conservatism.
None required.
10 Not a deficiency. Performed an alternate None required.
calculation for the time to reach minimum allowable boron concentrations. Confirmed calculation of time to alarm.