ML20205L253

Boraflex Degradation Analysis

ML20205L253
Person / Time
Site:	Mcguire, McGuire
Issue date:	04/05/1999
From:	DUKE POWER CO.
To:
Shared Package
ML20205L214	List: ML20205L209 ML20205L225 ML20205L234 ML20205L243 ML20205L253
References
GL-96-04, GL-96-4, NUDOCS 9904140198
Download: ML20205L253 (44)

Text

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1 ATTACHMENT 8 i

MCGUIRE NUCLEAR STATION I BORAFLEX DEGRADATION ANALYSIS hA k0 b9 p P DR . '

Attachment 8 Page 1 of 9 MCGUIRE BORAFLEX DEGRADATION ASSESSMENT Table of Contents Section Page i

1.O BACKGROUND 2 1

2.0 PURPOSE 2 3.0 APPROACH 2 3.1 In-situ Testing 3 3.1.1 Method 3 3.1.2 In-situ Test Results for McGuire Unit 2 3 3.2 Computer Modeling 4 3.2.1 McGuire Unit 2 Model 5 3.2.2 McGuire Unit 1 Model 7 4.0 RACKLIFE ASSESSMENT 7 4.1 January 8, 1997 8 4.2 December 31, 1999 8 4.3 December 31, 2003 8

5.0 CONCLUSION

S 9 6.0 . REFERENCES 9 l l

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r2 Attachment 8 Page 2 of 9 1.O BACKGROUND The .McGuire spent fuel storage racks contain Boraflex neutron-absorbing panels. The Unit 1 racks were installed in January 1986, and the Unit 2 racks were installed in December 1984. The

! . function of the Boraflex panels is ensuring that reactivity of

( the stored fuel. assemblies is maintained within required limits.

l l Boraflex, as manufactured, is a silicon rubber material that l retains a powder of boron carbide neutron absorbing material.

l The Boraflex panels are enclosed in a formed stainless steel

wrapper sheet that is spot-welded to the storage tube. The l

wrapper sheet.is bent at each end to complete the enclosure of the Boraflex panel. The Boraflex panel is contained in the

plenum area between- the storage tube and the wrapper plate.

l Since the wrapper plate enclosure is not sealed, spent fuel pool water fills the enclosure.

It has been observed that af ter Boraflex receives a high gamma

' dose from the stored irradiated fuel (>10 2 rads) it can begin to j l . degrade and dissolve in the wet environment. Thus, the boron carbide poison material can be removed, thereby reducing the l poison worth of the Boraflex sheets. This phenomenon is documented in NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks" (Reference 6.1).

2.O PURPO3E One'of the key assumptions in the criticality calculation for the spent fuel storage racks is the boron-10 (B io) loading of the Boraflex. (Bi o is- the neutron-absorbing isotope of boron in boron carbide.) This attachment provides an assessment of the Boraflex in the McGuire spent fuel racks with consideration given to the degradation mechanism discussed above.

3.0 APPROACH Duke's approach to verifying McGuire's Boraflex is to periodically obtain results from quantitative in-situ l

measurements The first in-situ testing was performed in the McGuire Unit 2 spent fuel storage racks in January 1997 (see discussion below).

l l~ Additionally, Duke has used the RACKLIFE computer code, developed for the Electric Power Research Institute (EPRI), for estimating the condition of'the Boraflex through 2003. While Duke considers in-situ testing ~as its method of Boraflex verification, RACKLIFE is useful for an overall assessment of degradation. A RACKLIFE model can be-used to estimate degradation of each Boraflex panel for.some future date of interest. These results can be used for defining the fuel storage sub-regions of the pool and for

. determining which Boraflex panels should be in-situ tested.

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Attachment 8 i Page 3 of 9 I The following is a description of in-situ testing and RACKLIFE I modeling for the McGuire spent fuel storage racks.

3.1 IN-SITU TESTING 3.1.1 METHOD Northeast Technology Corporation (NETCO), under contract for the Electric Power Research Institute (EPRI), has developed the Boron-10 Areal Density Gage for Evaluating Racks (BADGER). This I system is used to measure the Bo i areal density (expressed as  !

2 grams of B i o per cm ) in spent fuel storage racks.

The BADGER system consists of a source head containing a Cf252 source and a detector head containing BF-3 detectors that are I lowered simultaneously into adjacent spent fuel storage cells. A I stepper motor and winch attached to the fuel bridge auxiliary hoist-allow the detector / source heads to be remotely located at desired elevations in the storage racks. The detector signals are fed into four pre-amplifiers and then to an electronics console that is positioned beside the pool. The signals are i recorded on a computer that also controls the stepper motor for

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positioning of the detector / source heads.

The principle of BADGER is measurement of thermal neutron attenuation by the Boraflex panel (s) between the source and detectors. The number of neutrons emitted by the Cfs2 2 source .

that reach the BF-3 detectors, is a function of the Bo areal i j l density in the Boraflex. The detector signal is low for Boraflex panels .with high Bo i areal density. Conversely, the detector signal is high for Boraflex panels with low B io areal density.

The BADGER equipment is calibrated by means of a calibration cell that is similar in construction to the spent fuel storage cells and that contains Boraflex panels of known Bo areal density.

i (Additional information regarding BADGER may be found in Reference 6.2) 3.1.2 In-Situ Test Results for McGuire Unit 2 I In a BADGER demonstration campaign in January 1997, 33 McGuire Unit 2 Boraflex panels were evaluated. Panels were selected to include those with the greatest gamma exposures. The results, excluding measurement uncertainties, are as follows:

l Boraflex Loss:

Reference 6.2 reports Boraflex loss for Region 1 and Region 2 in Tables 4.4 and 4.3, respectively. These findings are summarized below:

In Region 1, fifteen panels were evaluated. Boraflex loss ranged from zero ~ (for an unirradiated panel) to 33.33 percent. There j

1 Attachment 8 i Page 4 of 9 was a clear trend. for greater loss with increasing gamma exposure.

In Region 2, eighteen panels were evaluated. Foraflex loss ranged ~ from zero (for an unirradiated panel) to 15.85 percent.

There was no clear association between loss and gamma exposure. l l I l Generally, the loss of boron carbide from the panels was l relatively uniform. Gaps had formed in some of the panels (see

discussion below) and some limited thinning had occurred in some of Region 2 panels at the location of a 0.5 inch diameter

' inspection port.

Gaps in Boraflex Panels:

l While gap measurements are not the primary function of the BADGER l

test - equipment, an assessment .is made of gaps in the Boraflex panels based on the BADGER testing performed at McGuire in January 1997. The results of the gap measurements are presented in Tables 4-5 and 4-6 of Reference 6.2. Generally, the Region 1 panels were found to have one or two gaps per panel, and the Region 2 panels were found to have three or four gaps per panel.

None of the gaps exceeded four inches. The gaps appeared to be somewhat randomly distributed with no preferential elevation for gap formation.

3.2 Computer Modeling (RACKLIFE) i The McGuire Unit 2 Boraflex performance was modeled using the RACKLIFE computer code. RACKLIFE is a computer software package developed by NETCO under contract for EPRI. It is a stand-alone PC/ DOS executable program that computes the loss of boron carbide from Boraflex panels in fuel storage racks.

The RACKLIFE code is based on the following principles verified through extensive laboratory testing of irradiated Boraflex specimens as discussed in Reference 6.3:

a. Boraflex is. manufactured as a polydimethyl siloxane (silicon rubber) containing a powder of boron carbide, and a filler material of crystalline silica.

b. As Boraflex ages in the spent fuel pool environment, the polymer ?. atrix is gradually broken down and converted into amorphous silica. This is a function of gamma radiation and i exposure to the pool water,

c. ~ Amorphous silica is somewhat soluble in the spent fuel pool water at increasing rates with absorbed gamma dose, pool temperature, .and time. This solubilization is the physical mechanism that leads to removal of silica and boron carbide from l the storage racks.

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Attachment 8 Page 5 of 9

d. The amorphous silica and boron carbide from the Boraflex panels is transported into the spent fuel pool in a constant proportion. While the boron released from the spent fuel racks is indistinguishable from the boron in the boron acid added to PWR pools for criticality control (normally greater than 2000 ppm), silica concentrations in the pool are attributable almost I exclusively to the Boraflex since it is the only significant source of silica. Thus, the amount of boron carin ' de that is lost from the Boraflex can be calculated since the ratio of boron carbide to silica leaving the Boraflex is constant.

e. Silica concentration in the spent fuel pool water is a function of rack design, temperature, and operation of the pool clean-up system.

RACKLIFE performs a mass balance of SiO 2 in the pool and within the wrapper plate plenum that encapsulates the Boraflex panels.

A simple explanation of the mass balance is that the total SiO2 released by the Boraflex panels, in aggregate, is affected by the amount of SiO 2 in solution in the pool water and the amount removed over time by the clean-up system. The contribution of 3 each panel to the bulk SiO2 quantity is determined, based on the irradiation-time history of the panel. All other factors being equal, panels with higher gamma exposures have higher SiO 2 '

releases, and for those with equal gamma exposures, the ones that received the dose early in life have SiO2 releases. Having calculated the SiO 2 released by each panel, RACKLIFE then calculates the boron carbide released, based on the fixed ratio of boron carbide to SiO 2. A detailed discussion of the RACKLIFE code may be found in Reference 6.3.

It is important to note that Duke will not use the RACKLIFE code for Boraflex verification. RACKLIFE will be used to identify lead panels for in-situ testing and to provide an estimate of future condition. In-situ testing will be used to verify the Boraflex.

3.2.1 MCGUIRE UNIT 2 MODEL A RACKLIFE model was developed for the Unit 2 spent fuel pool with the following input (Note: a detailed description of the RACKLIFE inputs may be found in Reference 6.3.):

1. Dimensional data for the pool, storage racks, and Boraflex.

2. Spent fuel pool water data including temperature and silica concentration.

3. Data for the irradiated fuel assemblies stored in the racks, including enrichment, burnup, discharge date, end of cycle power fraction, reactor cycles in which the assembly operated.

4. Dates and locations where each of the irradiated fuel assemblies was stored in the spent fuel racks.

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Attachment 8 Page 6 of 9 One of the key RACKLIFE inputs is the escape coefficient assumed for the storage racks. This coefficient is associated with the rate at which the spent fuel pool water moves through the stainless steel wrapper that encapsulates the Boraflex panel.

The more "open" the wrapper is to the pool, the greater the resulting degradation rate. A higher escape coefficient is used for a more open wrapper.

i The approach ured for McGuire Unit 2 was to vary the Region 1 and Region 2 escape coefficients to obtain the best match between the RACKLIFE results and the BADGER results for the tested panels.

The escape coefficients thus determined were 1.25 for Region 1 and 0.05 for Region 2.  !

The RACKLIFE results are adjusted based on a 95/95 worst case statistical evaluation of RACKLIFE error for the panels tested by '

BADGER.

A comparison of the BADGER results and worst-case RACKLIFE results, expressed as a percentage of the minimum as-built Bo i areal density for each region (0.0216 g/cm2 and 0.0075 g/cm 2 for Region 1 and Region 2, respectively) is shown below:

Unit 2 Percent January 1997 Boraflex Loss BADGER RACKLIFE Region 1 Panels (nominal) (worst case)

A 23 South 0% 4%

C 13 East 19% 28%

C 13 North 30% 29%

D 13 East 29% 26%

E 2 West 11', 18%

E 13 East 25% 38%

E 13 North 37% 38%

E 13 West 34% 41%

F 2 East 17% 21%

F 12 West 22% 36%

F 13 East 29% 41%

F 13 North 31% 40%

F 14 East 29% 40%

G 12 East 28% 34%

H 13 West 25% 30%

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1 Attachment 8 Page 7 of 9 Unit 2 Percent January 1997 Boraflex Loss BADGER RACKLIFE Region 2 Panels (nominal) (worst case)

BB 78 South 0% 19%

DD 78 East 17% 19%

DD 78 North 11% 19%

.DD 78 South -3% 19%

FF 78 East 5% 19%

FF 78 North 19% 19%

HH 78 East -3% 19%

HH 78 North 15% 19%

KK 3 South 0% 18%

KK 78 North 0% 19%

KK 78 South 3% 19%

KK 78 West 13% 19%

MM 78 East 0% 19%

MM 78 North 4% 19%

MM 78 South 4%- 19%

MM 78 West 19% 19%

PP 78 East 1% 19%

PP 78 South 3% 19%

3.2.2 McGuire Unit 1 Model BADGER testing was only performed in the Unit 2 spent fuel pool.

Thus, no Unit 1 in-situ test results are available for comparison to .RACKLIFE. However, the design and construction of the Unit 1 storage racks are identical to Unit 2. Therefore, the escape coefficients and adjustments determined for the Unit ? model are applicable to the Unit 1 model.

4.0 3LACKLIFE ASSESSMENT To provide flexibility in fuel storage, the criticality analysis subdivides Region 1 and Region 2, as follows:

Region lA is assumed to have'Boraflex degraded 75 % frcm the original design minimum (25% remaining) and Region 1B is assumed to have Boraflex degraded 100% (0% remaining).

. Region 2A is assumed to have Boraflex degraded 50 % from the  ;

original-design minimum (50% remaining) and Region 2B is assumed l to have Boraflex degraded 100% (0% remaining).

In the Unit-1 spent fuel racks, only the Region lA and Region 2A designations are assigned. In the Unit.2 spent fuel racks, Region lA, Region lB and Region 2A designations are assigned.

Worst-case RACKLIFE assessments for the sub-Regions are presented l below for various in-service dates.

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l-Attachment 8 Page 8 of 9 i

4.1 January 8, 1997 Unit 2 i The worst panel losses, expressed as percentage of the minimum as-built' Bi o areal density for the sub-Regions, are as follows:

Region 1A- 34%

Region'1B- 41%

Region 2A- 19%

Unit 1 l

I No specific RACKLIFE comput%tions were performed for Unit 1 for January 8, 1997. Calculations for - later in-service dates demonstrate that degradation in Unit 1 is enveloped by Unit 2.

4.2 December 31, 1999 Unit 2 l

l The worst-case RACKLIFE results were computed for December 31, 1999. The worst panel losses, expressed as percentage of the minimum as-built Bi o' areal density for the sub-Regions, are as

! follows:

Region 1A- 50%

Region 1B- 60%

Region 2A- 21%'

Unit 1

-No specific'RACKLIFE computations were performed for Unit 1 for December 31, 1999. Calculations for later end dates demonstrate-that. degradation in Unit 1 is enveloped by Unit 2.

4.3 December 31, 2003 l Unit 2~

The worst-case RACKLIFE results were computed for December -31, 2003. The -worst panel . losses, expressed as percentage of the minimum as-built Bo i areal density for the sub-Regions, are as follows:

Region 1A 81%

Region 1B- 97%

Region 2A- 22%-

Unit l'

'A ~ worst-case - RACKLIFE model was developed . for Unit 1 with the l

l Attachment 8 l Page 9 of 9 l escape - coefficients and error corrections used for the Unit 2 model. This approach is justified since the storage racks are of identical design.

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The worst. panel losses, expressed as percentage of the minimum as-built B i o areal density for the sub-Regions, are as follows:

Region lA- 64%

Region 2A- 21%

5.0 CONCLUSION

S The initial in-situ verification for the McGuire Unit 2 spent fuel racks in January .1997 showed the Boraflex has degraded in both Region 1 and Region 2. Additional in-situ testing will be performed at. a frequency of three years, starting in 2000, to confirm the Boraflex levels assumed in the revised criticality analysis.

The Unit 2 RACKLIFE model produced results consistent with the January 1997 in-situ test results. Using the rack escape coefficients determined for the Unit 2 model, a Unit 1 RACKLIFE model was developed, and it shows Unit 1 Boraflex is less degraded than Unit 2 Boraflex. RACKLIFE assessments for the Unit i

1 and Unit 2 pools for December 31, 2003 show the Boraflex is not expected to degrade to less than the values assumed in the criticality calculation. In-situ testing will be employed to verify the actual Boraflex condition.

In the near term, Duke will continue investigations into options to address degrading Boraflex at McGuire. These options include replacement of the storage racks, insertion of additional neutron poison (rack or fuel assembly inserts), more stringent controls on fuel reactivity and storage patterns, and chemical inhibitors j currently under investigation by EPRI.

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6.O REFERENCES 6.1 "Boraflex Degradation in Spent Fuel Pool Storage Racks," NRC i Generic Letter 96-04, June 26, 1996.  ;

6.2 " BADGER, a Probe for Nondestructive Testing of Residual l Boron-10 Absorber Density in Spent-Fuel Storage Racks:

Development and Demonstration, EPRI TR-107335, October 1997".

6.3' "The RACKLIFE Boraflex Rack Life Extension Computer Code: j Theory and Numerics", DRAFT, NETCO, May 1997.

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ATTACHMENT 9 MCGUIRE NUCLEAR STATION PROPOSED REVISION TO UFSAR CHAPTER 16, " SELECTED LICENSEE COMMITMENTS" .

i 16.9 AUXILIARY SYSTEMS l

l 1

16.9-9 SPENT FUEL POOL STORAGE RACK POISON MATERIAL COMMITMENT

a. The Region 1 panel average storage rack poison material Boron 10 areal density shall be greater than or equal to:

0.005 gm Bidcm' for Region 1 A 0 gm Bi dem" for Region 1B

b. The Region 2 panel average storage rack poison material Boron 10 areal density shall be greater than or equal to:

0.003 gm Bi dema for Region 2A 0 gm Bidem for Region 2B APPLICABILITY: l When a fuel assembly is stored in a spent fuel rack cell location.

l REMEDIAL ACTION: For Units 1 and 2

a. With a panel average spent fuel pool storage rack cell poison material not within limits:

1. Perform SR 3.7.14.1. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter until the l affected fuel assembly is moved, and;

2. Verify that the fuel assembly in the affected location meets LCO 3.7.15(b) for Region 1 or LCO 3.7.15(d) for Region 2 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

b. If Remedial Action a. 2 is not met, immediately initiate action to move the affected fuel assembly to an acceptable location.

TESTING REQUIREMENTS:

a. Verify that the panel average spent fuel pool storage rack poison material is within limits every three years.

16.9-23

I BASES:

The McGuire spent fuel storage racks contain Boraflex neutron-absorbing panels that surround each storage cell on all four sides (except for peripheral sides). The function of these Boraflex l panels is to ensure that reactivity of the stored fuel assemblies is maintained within required l limits. Boraflex, as manufactured, is a silicon rubber material that retains a powder of boron l carbide (84C) neutron absorbing material. The Boraflex panels are erclosed in a formed l stainless steel wrapper sheet that is spot-welded to the storage tube. The wrapper sheet is bent at each end to complete the enclosure of the Boraflex panel. The Boraflex panel is contained in the plenum area between the storage tube and the wrapper plate. Since the wrapper plate enclosure is not sealed, spent fuel pool water is free to circulate through the i plenum.

It has been observed that after Boraflex receives a high gamma dose from the stored irradiated fuel (>10' rads) it can begin to degrade and dissolve in the wet environment. The potential degradation mechanisms with respect to boraflex in s,3ent fuel storage racks include:

(1) gamma radiation-induced shrinkage of boraflex and the potential for developing tears or gaps in the material, and (2) gradual long-term boraflex degradation over the intended service life of the racks as a result of gamma irradiation and exposure to the spent fuel pool environment.

Thus, the B4C poison material can be removed, thereby reducing the poison worth of the Boraflex sheets. This phenomenon is documented in NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks". To address this degradation, the spent fuel racks have been analyzed taking credit for soluble boron as allowed in WCAP-14416-NP-A,

" Westinghouse Spent Fuel Rack Criticality Analysis Methodology," Revision 1, November 1996.

This methodology ensures that the spent fuel rack multiplication factor, kon is less than or equal to 0.95. Codes, methods and techniques used in the McGuire criticality analysis are used to satisfy this kon criterion. The spent fuel storage racks are analyzed to allow storage of fuel assembiies with enrichments up to a maximum of 4.75 weight percent Uranium-235 while i maintaining k n <_0.95 including uncertainties, tolerances, bias, and credit for soluble boron.

Soluble boron credit is used to offset uncertainties, tolerances, and off-normal conditions and to provide subcritical margin such that the spent fuel pool kon is maintained less than or equal to I 0.95. The soluble boron concentration required to maintain k n less than or equal to 0.95 under l normal conditions is 440 ppm. In addition, sub-criticality of the pool (k.n < 1.0) is assured on a  !

95/95 basis without the presence of the soluble boron in the pool. Credit is taken for reactivity depletion due to fuel burnup and reduced credit for the Boraflex neutron absorber panels.

The limits specified for the panel average storage rack poison material Boron 10 areal density ensures the kon of the spent fuel pool will always remain < 1.00, assuming the pool to be j flooded with unborated water. The specified limit of Boron 10 areal density in boraflex i preserves the assumptions used in the analyses of the potential criticality accident scenarios. l These limits are the minimum required concentration for fuel assembly storage. The criticality l analysis performed shows that the acceptance criteria for criticality is met for the storage of fuel I assemblies with soluble boron credit, reduced credit for the Boraflex panels and the storage configurations and enrichment limits Specified by LCO 3.7.15. The storage configuration requirements specified by LCO 3.7.15 establish four regions within the spent fuet pool storage racks. Figure 16.9-1 illustrates the four regions for the Unit 1 spent fuel pool and Figure 16.9-2 illustrates the four regions for the Unit 2 pool. The limits specified are not applicable if a storage cell location does not contain a fuel assembly. l 16.9-24

REMEDIAL ACTIONS The remedial actions associated with this SLC are designed to ensure that an unplanned criticality event cannot occur as a result of degraded boraflex conditions. R3 medial Action a.1.

verifies the Spent Fuel Pool boron concentration to be within Technical Specification 3.7.14 limits. These limits are based on the cycle-specific Core Operating Limits Requirements ,

(COLR) document. The COLR Spent Fuel Pool boron concentration cannot be less than 2475 l ppm soluble boron for any specific cycle. This is the initial boron concentration used in the

]

Spent Fuel Pool boron dilution analysis. If SR 3.7.14.1 indicates boron concentrations less than the acceptable level, the associated remedial actions are to immediately suspend movement of fuel assemblies in the pool area and to immediately initiate boron additions to raise the boron j concentration to acceptable levels. Remedial Action a.2. deterniines if the assembly can be qualified for storage in Region 1B or Region 2B. If the assembly can be stored in one of these regions, then it will not have to be moved and the Remedial Actions can be immediately exited.

If Remedial Action a.2. cannot be met, then action is to be initiated immediately to move the affected assembly to an acceptable location. There may be circumstances that will prevent the 1 movement of the affected assembly in a reasonable time period. For example, if the pool is nearly full, there may not be enough spaces available to meet the required storage configurations of LCO 3.7.15. In this case, it is acceptable to continue Remedial Action a.1.

until the affected fuel assembly can be moved to an acceptable location. The daily verification of boron concentration per SR 3.7.14.1 ensures the assumptions used in the associated i criticality analyses are maintained. There is a large amount of margin between the COLR I boron concentration and the boron concentration needed maintain subcritical conditions in the Spent Fuek Pool. Daily verifications are considered to be adequate to ensure that no dilution evolution could go undetected for an extended period resulting in boron concentrations less than the minimum amounts newsary for maintaining subcritical conditions.

TESTING REQUIREMENTS 1

The testing requirements will verify that the Boron 10 areal density is within acceptable limits.

The preferred method for verifying the Boron 10 areal density would be in-situ testing at least every three years. Testing may be performed more frequently based on engineering judgment, spent fuel pool water chemistry, and modeling projections of boraflex degradation.

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REFERENCEG:

1. UFSAR, Section 9.1.2.

2. Issuance of Amendments, McGuire Nuclear Station,1Jnits 1 and 2 (TAC NOS. M89744 and M89745), November 6,1995.

3. Double contingency principle of ANSI N16.1-1975, as specified in ine April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

4. UFSAR, Section 15.7.4.

5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

6. NRC Generic Letter 96-04: Boraflex Degradation in Spent Fuel Pool Storage Racks, June 26,1996

7. WCAP-14416-NP-A, Westinghouse Spent Fuel Rack Criticality Analysis Methodology, Revision 1, November 1996 1

l I

I-16.9-28

ATTACHMENT 10 MCGUIRE NUCLEAR STATION PROPOSED REVISIONS TO TECHNICAL SPECIFICATION BASES

l Spent Fuel Pool Boron Concentration B 3.7.14 ENTIRE 3.7.14 BASES TO BE E ED e NNNG B 3.7 PLANT SYSTEMS PAGES.

B 3.7.14 Spent Fuel Pool Boron Concentration --

BASES BACKGROUND in the two region poison fuel storage rack (Refs.1 and 2) design, the spent fuel pool is divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate pools. Region 1, with 286 storage positions, is designed to accommodate new fuel with a maximum nominal enrichment of 4.75 wt%

U-235 (maximum tolerance of 0.05 wt%), which have accumulated minimum burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1. Fuel assemblies not meeting the criteria of Table 3.7.15-1 shall be stored in accordance with Figures 3.7.15-1 through 3.7.15-3. Region 2, with 1177 storage positions, is designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups in accordance with the accompanying LCO.

The water in the spent fuel pool normally contains soluble boron, which results in large suberiticality margins under actual operating conditions.

However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting ka of 0.95 be evaiuated in the absence of soluble boron. Hence, the design of the spent fuel storage racks is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the spent fuel pool fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 3) allows credit for soluble boron under other abnormal or accident conditions, since only a single accident need be considered at one time. For example, the most severe accident scenario is associated f with the movement of fuel from Region 1 to Region 2, and accidental misloading of a fuel assembly in Region 1 or Region 2. This could potentially increase the reactivity of the spent fuel pool. To mitigate these postulated criticahty related accidents, boron is dissolved in the pool water. Safe operation of the two region poison fuel storage rack with no j movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with LCO 3.7.15, " Spent Fuel Assembly Storage." Prior to movement of an assembly, it is necessary to perform SR 3.7.14.1.

l APPLICABLE Most accident conditions do not result in an increase in the reactivity of SAFETY ANALYSES either of the two regions. Examples of these accident conditions are the loss of cooling (reactivity increase with decreasing water density) and the McGuire Units 1 and 2 B 3.7.14-1 Revision No.4-

)

1 ENTIRE 3.7.14 BASES TO BE REPLAC b O " " " "

l FOLLOWING PAGES. B3 4 BASES \

\

l APPLICABLE SAFETY ANALYSES (continued) dropping of a fuel assembly on the top of the rack. However, accidents can be postulated that could increase the reactivity. This increase in

! reactivity is unacceptable with unborated water in the storage pool. Thus, i

for these accident occurrences, the presence of soluble boron in the storage pool prevents criticality in both regions. The postulated accidents are basically of two types. A fuel assembly could be incorrectly transferred from Region 1 to Region 2 (e.g., an unirradiated fuel assembly or an hsufficiently depleted fuel assembly). The second type of postulated accidents is associated with a fuel assembly which is dropped adjacent to the fully loaded Region 2 storage rack. This could have a small positive reactivity effect on Region 2. However, the negative reactivity effect of the soluble boron compensates for the increased reactivity caused by either one of the two postulated accident scenarios. The accident analyses is provided in the UFSAR, Section 15.7.4 (Ref. 4).

The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 5).

i i

I LCO The spent fuel pool boron concentration is required to be within the limits '

specified in the COLR. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference 4. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement wiain the spent fuel pool. l A!'PLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

\

ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

i When the concentration of boron in the fuel storage pool is less than

! required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies.

I CO is not met while moving irradiated fuel assemblies in MODE 5 J N J McGuire Units 1 and 2 B 3.7.14 2 Revision No. t

Spent Furi Pool Boron Conctntration B 3.7.14 BASES i

[ ACTIONS (continued) . _ _ _ I i

or 6, LCO 3.0.3 would not be applicable. ing irradiated fuel assemblies while in MODE 1,2,3, or 4, th .m movement is independent of reactor operation. Thereforu, inability to suspend movement of fuel assemblies is not sufficient reason to require e m clo shutdown.

SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time.

REFERENCES 1. UFSAR, Section 9.1.2.

2. Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOS. M89744 and M89745), November 6,1995.

J

3. Double contingency principle of ANSI N16.1-1975, as specified in )

the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A). l

4. UFSAR, Section 15.7.4.

5. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

4 ENTIRE 3.7.14 BASES TO BE REPLACED WITH FOLLOWING PAGES.

I McGuire Units 1 and 2 B 3.7.14-3 Revision No. e- l l

Spent Fuel Pool Boron Concentration B 3.7.14 NEW 3.7.14 BASES B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Boron Concentration BASES BACKGROUND in the two region poison fuel storage rack (Refs.1 and 2) design, th spent fuel pool is divided into two separate and distinct regions.

Region 1, with 286 storage positions, is designed and generally reserved for temporary storage of new or partially irradiated fuel. Region 2, with 1177 storage positions, is designed and generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying burnup levels.

The McGuire spent fuel storage racks contain Boraflex neutron-absorbing panels that surround each storage cell on all four sides (except for peripheral sides). The function of these Boraflex panels is to ensure that the reactivity of the stored fuel assemblies is maintained within required limits. Boraflex, as manufactured, is a silicon rubber material that retains a powder of boron carbide (B4C) neutron absorbing material. The Boraflex panels are enclosed in a formed stainless steel wrapper shee that is spot-welded to the storage tube. The wrapper sheet is bent each end to complete the enclosure of the Boraflex panel. The Borafle panel is contained in the plenum area between the storage tube and t wrapper plate. Since the wrapper plate enclosure is not sealed, spe t fuel pool water is free to circulate through the plenum. It has been observed that after Boraflex receives a high gamma dose from the stored irradiated fuel (>10' rads) it can begin to degrade and dissolve in the w et l environment. Thus, the B4C poison material can be removed, thereay reducing the poison worth of the Boraflex sheets. This phenomenon is documented in NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks". A To address this degradation, each region of the spent fuel pool has be divided into two sub-regions; with and without credit for Boraflex. For th  !

regions taking credit for Boraflex, a minimum amount of Boraflex wa '

assumed that is less than the original design minimum Bio areal density.

The McGuire spent fuel storage racks have been analyzed taking credit for soluble boron as allowed in Reference 3. The methodology ensures that the spent fuel rack multiplication factor, kg, is less than or equal t 0.95 as recommended in ANSI /ANS-57.2-1983 (Ref. 4) and N guidance (Ref. 5). The spent fuel storage racks are analyzed to all w storage of fuel assemblies with enrichments up to a maximum no nal enrichment of 4.75 weight percent Uranium-235 while maintaining k 5 McGuire Units 1 and 2 B 3.7.14-1 Revision No.

Spent Fu::I Pool Boron Conc:ntration NEW 3.7.14 BASES B 3.7.14 BASES s x_

BAOKGROUND (continued) 0.95 including uncertainties, tolerances, bias, and credit for soluble boron. Soluble boron credit is used to offset uncertainties, tolerances, and off-normal conditions and to provide subcritical margin such that the spent fuel pool kon is maintained less than or equal to 0.95. The soluble boron concentration required to maintain kon less than or equal to 0.95 under normal conditions is 440 ppm. In addition, sub-criticality of the pool (k.n < 1.0) is assured on a 95/95 basis, without the presence of the soluble boron in the pool. The criticality analysis performed shows that the acceptance criteria for criticality is met for the storage of fuel assemblies when credit is taken for reactivity depletion due to fuel burnup, the presence of integrated Fuel Burnable Absorber (IFBA) rods, reduced credit for the Boraflex neutron absorber panels and storage con'igurations and enrichment limits Specified by LCO 3.7.15.

APPLICABLE Most accident conditions do not result in an increase in reactivity of the f SAFETY ANALYSES racks in the spent fuel pool. Examples of these accident conditions are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly between rack modules (rack design precludes this condition), and the drop of a fuel assembly between rack modules and the pool wall.

However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage pools. The first is a drop or placement of a fuel assembly into the cask loading area. The second is a significant change in the spent fuel pool water temperature (either the loss of normal cooling to the spent fuel pool water which causes an increase in the pool water temperature or a large makeup to the pool with cold water which causes a decrease in the pool water temperature) and the third is the misloading of a fuel assembly into a location which the restrictions on location, enrichment, burnup and rmmber of IFBA rods is not satisfied. I For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 6) can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against l a criticality accident. Thus, for these postulated accident conditions, the {

presence of additional soluble boron in the spent fuel pool water (above the 440 ppm required to maintain kon less than or equal to 0.95 under normal conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

! Calculations were performed to determine the amount of soluble boron required to offset the highest reactivity increase caused by either of  :

1 McGuire Units 1 and 2 B 3.7.14-2 Revision No.

NEW 3.7.14 BASES Sp:nt Furl Pool Boron Conc:ntration y B 3.7.14 l

D .

APPLICABLE SAFETY ANALYSES (continued)

, these postulated accidents and to maintain ken less than or equal to 0.95.

l It was found that a spent fuel pool boron concentration of 1170 ppm was adequate to mitigate these postulated criticality related accidents and to maintain ken less than or equal to 0.95. Specification 3.7.14 ensures the I spent fuel pool contains adequate dissolved boron to compensate for the j increased reactivity caused by these postulated accidents.

Specification 4.3.1.1 c. requires that the spent fuel rack ken be less than or equal to 0.95 when flooded with water borated to 440 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent l fuel pool before the 0.95 ken design basis is exceeded. The spent fuel l pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron ,

l concentration to 440 ppm is not a credible event.

The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 5).

LCO The spent fuel pool boron concentration is required to be within the limits specified in the COLR. The specified concentration of dissolved boron in l the spent fuel pool preserves the assumptions used in the analyses of the potential criticality accident scenarios as described in Reference 4. This j concentration of dissolved boron is the minimum required concentration f for fuel assembly storage and movement within the spent fuel pool.

l APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

I ACTIONS A.1 and A.2 1

1 The Required Actions are modified by a Note indicating that LCO 3.0.3  !

does not apply.

When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of i

, an accident or to mitigate the consequences of an accident in progress. l This is most efficiently achieved by immediately suspending the l movement of fuel assemblies. The concentration of boron is restored ,

simultaneously with suspending movement of fuel assemblie .  !

l 1 i

McGuire Units 1 and 2 B 3.7.14-3 Revision No.

l _

r

]

NEW 3.7.14 BASES Spent Fu:1 Pool Boron Conc 3ntration

\\ B 3.7.14 BASES ACTIONS (continued)

If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time.

REFERENCES 1. UFSAR, Section 9.1.2.

2. Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOF f61744 and M89745), November 6,1995.

3. WCAP-14416-NP-A, Westinghouse Spent Fuel Rack Criticality Analysis Methodology, Revision 1, November 1996.

, 4. American Nuclear Society, "American National Standard Design l l k Requirements for Lig'n t Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSl/ANS-57.2-1983, October 7,1983.

5. Nuclear Reguistory Commission, Memorandum to Timothy Collins 1 from Laurence Kopp, " Guidance on the Regulatory Requirements i for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19,1998.

6. Double contingency principle of ANSI N16.1-1975, as specified i the April 14,1978 NRC letter (Section 1.2) and implied in t proposed revision to Regulatory Guide 1.13 (Section 1 ,

l Appendix A).

l 7. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

8. UFSAR, Section 15.7.4. l 1

McGuire Units 1 and 2 B 3.7.14-4 Revision No.

i j

I Sp:nt Fu:1 Ass:mbly Storags ENTIRE 3.7.15 BASES TO BE l REPLACED WITH FOLLOWING B 3.7~ PLANT SYSTEMS AGES.

B 3.7.15 Spent Fuel Assembly Storage BASES BACKGROUND in the two region poison fuel storage rack (Refs.1 and 2) design, the spent fuel pool is divided into two separate and distinct regions which, for the purpose of criticality considerations, are considered as separate f

pools. Region 1, with 286 storage positions, is designed to {

accommodate new fuel with a maximum nominal enrichment of 4.75 wt%

U-235 (maximum tolerance of 0.05 wt%), which have accumulated 1 minimum burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1. Fuel assemblies not meeting the criteria of Table 5.7.15-1 shall be stored in accordance with Figures 3.7.15-1 through 3.7.15-3. Region 2, with 1177 storage positions, is designed to accommodate fuel of various initial enrichments which have accumulated minimum burnups in accordance with the accompanying LCO.

The water in the spent fuel pool normally contains soluble boron, which results in large subcriticality margins under actual operating conditions.

However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting km of 0.95 be evaluated in the absence of soluble boron. Hence, the design of the spent fuel storage racks is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the spent fuel pool fully loaded. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 3) allows credit for soluble boron under other abnormal or accioent condQns, since only a single accident need be considered at one time. For e .ople, the most severe accident scenario is associated with the movement of fuel from Region 1 to Region 2, and accidental mistoading of a fuel assembly in Region 1 or Region 2. This could potentially increase the reactivity of the spent fuel pool. To mitigate these postulated criticality related accidents, boron is dissolved in the pool water. Safe operation of the two region poison fuel storage rack with no movement of assemblies may therefore be achieved by controlling the location of each assembly in accordance with the accompanying LCO.

Prior to movement of an assembly, it is necessary to perform SR 3.7.14.1.

APPLICABLE The hypothetical accidents can only take place during or as a result of the

, SAFETY ANALYSES movement of an assembly (Ref. 4). For these accident occurrences, the l presence of soluble boron in the spent fuel pool (controlled by LCO 3.7.14, " Spent Fuel Pool Boron Concentration") prevents criticality in s --

t McGuire Units 1 and 2 B 3.7.15-1 Revision No.4-

l Spent Fusl Ass!mbly Storags B 3.7.15

! BASES ENTIRE 3.7.15 BASES TO BE

-REFLACEL WITH FOLLOWING PAGES.

APPLICABLE SAFETY ANALYSES (continued) both regions. By close'y controlling the movement of each assembly by checking the location of each assembly after movement, the time period for potential accidents may be limited to a small fraction of the total operating time. During the remaining time period with no potential for accidents, the operation may be under the auspices of the accompanying LCO.

The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 5).

LCO The restrictions on the placement of fuel assemblies within the spent fuelk pool, in accordance with Tables 3.7.15-1 and 3.7.15-3, in the accompanying LCO, ensures the k n of the spent fuel pool will always remain < 0.95, assuming the pool to be flooded with unborated water.

Fuel assemblies not meeting the criteria of Tables 3.7.15-1 and 3.7.15-3 f shall be stored in accordance with Figures 3.7.15-1,3.7.15-2 and 3.7.15-3, and Tables 3.7.15-2 and 3.7.15-4.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pool.

ACTIONS M Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the LCO, the immediate action is to initiate actio to make the necessary fuel assembly movement (s) to bring the configuration into compliance.

j If unable to move irradiated fuel assemblies while in MODE 5 or 6, i LCO 3.0.3 would not be applicable, if unable to move irradiated fuel assemblies while in MODE 1,2,3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not f sufficient reason to require a reactor shutdown.

t McGuire Units 1 and 2 8 3.7.15-2 Revision No.G-

i l

Sp:nt Futi Ass:mbly Storage l

B 3.7.15 BASES ENTIRE 3.7.15 BASES TO BE met % ACED WITH FULLOW1NG PAGES.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies by adrninistrative means that the initial enrichment an burnup of the fuel assembly is in accordance with the configurations specified in the accompanying LCO.

REFERENCES 1. UFSAR, Section 9.1.2.

l 2. Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOS. M89744 and M89745), November 6,1995.

/

3. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the

, proposed revision to Regulatory Guide 1.13 (Section 1.4, l Appendix A).

l

4. UFSAR, Section 15.7.4.

5. 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

l l

l l

1 l

l 1

McGuire Units 1 and 2 B 3.7.15-3 Revision Note- ,

I l

r i

Spent Fuel Assembly Storage B 3.7.15 NEW 3.7.15 BASES B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage BASES BACKGROUND in the two region poison fuel storage rack (Refs.1 and 2) design, the spent fuel pool is divided into two separate and distinct regions.

Region 1, with 286 storage positions, is designed and generally reserved for temporary storage of new or partially irradiated fuel. Region 2, with 1177 storage positions, is designed and generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying burnup levels.

The McGuire spent fuel storage racks contain Boraflex neutron-absorbing panels that surround each storage cell on all four sides (except for peripheral sides). The function of these Boraflex paneic is to ensure that the reactivity of the stored fuel assemblies is maintained within required limits. Boraflex, as manufactured, is a silicon rubber material that retains a powder of boron carbide (B4C) neutron absorbing material. The Boraflex panels are enclosed in a formed stainless steel wrapper sheet that is spot-welded to the storage tube. The wrapper sheet is bent at each end to complete the enclosure nf the Boraflex panel. The Boraflex panel is contained in the plenum area between the storage tube and the wrapper plate. Since the wrapper plate enclosure is not sealed, spent fuel pool water is free to circulate through the plenum. It has been observed that after Boraflex receives a high gamma dose from the stored irradiated fuel (>10' rads) it can begin to degrade and dissolve in the wet environment. Thus, the B4C poison material can be removed, thereby reducing the poison worth of the Boraflex sheets. This phenomenon is documented in NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks".

To address this degradation, each region of the spent fuel pool has been divided into two sub-regions; with and without credit for Boraflex. For the regions taking credit for Boraflex, a minimum amount of Boraflex was assumed that is less than the original design minimum B10 areal density. l To address this degradation, each region of the spent fuel pool has been '

divided into two sub-regions; with and without credit for Boraflex. For the regions taking credit for Boraflex, a minimum amount of Boraflex was 1 assumed that is less than the original design minimum B10 areal density. l' l

Two storage configurations are defined for each region; Unrestricted and Restricted storage. Unrestricted storage allows storage in all cells without restriction on the storage configuration. Restricted storage a"ows j storage of higher reactivity fuel when restricted to a certain storage t k ~

I

~ - l McGuire Units 1 and 2 B 3.7.15-1 Revision No.

l l

Sp:nt Fu:1 Ass:mbly Storage NEW 3.7.15 BASES BASES _ /

BACKGROUND (continu

/

configuration with lower reactivity fuel. A third loading p rn, Checkerboard storage, was defined for Regions 1B, 2A and Checkerboard storage allows storage of the highest reactivity fuel in ea region when checkerboarded with empty storage cells.

The McGuire spent fuel storage racks have been analyzed taking credn  !

for soluble boron as allowed in Reference 3. The methodology ensures j that the spent fuel rack multiplication factor, kon, is less than or equal tc 0.95 as recommended in ANSI /ANS-57.2-1983 (Ref. 4) and NRC guidance (Ref. 5). The spent fuel storage racks are analyzed to allow storage of fuel assemblies with enrichments up to a maximum nomina enrichment of 4.75 weight percent Uranium-235 while maintaining k n s 0.95 including uncertainties, tolerances, bias, and credit for soluble boron. Soluble boron credit is used to offset uncertainties, tolerances, and off-normal conditions and to provide suberitical margin such that the spent fuel pool k.n is maintained less than or equal to 0.95. The soluble 4 boron concentration required to maintain k n less than or equal to 0.95 under normal conditions is 440 ppm. In addition, sub-criticality of the pool (k.n < 1.0) is assured on a 95/95 basis, without the presence of the i soluble boron in the pool. The criticality analysis performed shows that the acceptance criteria for criticality is met for the storage of fuel ,

assemblies when credit is taken for reactivity depletion due to fuel  !

burnup, the presence of Integrated Fuel Burnable Absorber (IFBA) rods,  !

reduced credit for the Boraflex neutron absorber panels and storage configurations and enrichment limits Specified by LCO 3.7.15.

l APPLICABLE Most accident conditions do not result in an increase in reactivity of the SAFETY ANALYSES racks in the spent fuel pool. Examples of these accident conditions are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly between rack modules (rack design precludes this condition), and the drop of a fuel assembly between rack modules and the pool wall.

However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage pools. The first is a drop or I placement of a fuel assembly into the cask loading area. The second is a significant change in the spent fuel pool water temperature (either the loss of normal cooling to the spent fuel pool water which causes an increase in the pool water temperature or a large makeup to the pool with cold water which causes a decrease in the pool water temperature) and the third is the mistoading of a fuel assembly into a location which the restrictions on location, enrichment, burnup and number of IFBA rods is

! not satisfied.

For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter McGuire Units 1 and 2 B 3.7.15-2 Revision No.

E

f.

l Sp:nt Furl Assembly Storage l B 3.7.15 NEW 3.7.15 BASES

! BASES /

w

' APPLICABLE SAFETY ANALYSES (continued) -

(Ref. 6) can be applied. This states that one is not required to two unlikely, independent, concurrent events to ensure protection againsf a criticality accident. Thus, for these oostulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 440 ppm required to maintain kon less than or equal to 0.95 under j normal conditions) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikelv event.

Calculations were performed to determine the amount of soluble boro required to offset the highest reactivity increase caused by either of thest postulated accidents and to maintain k n less than or equal to 0.95. It was found that a spent fuel pool boron concentration of 1170 ppm was adequate to mitigate these postulated criticality related accidents and t3 l maintain k n less than or equal to 0.95. Specification 3.7.14 ensures th B l spent fuel pool contains adequate dissolved boron to compensate for the '

increased reactivity caused by these postulated accidents.

Specification 4.3.1.1 c. requires that the spent fuel rack kon be less than or equal to 0.95 when flooded with water borated to 440 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that ,

sufficient time is available to detect and mitigate a dilution of the sper l l fuel pool before the 0.95 k.n design basis is exceeded. The spent f '

pool boron dilution ana!ysis concluded that an unplanned or inadverten event which could result in the dilution of the spent fuel pool boron concentration to 440 ppm is not a credible event.

The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7).

LCO a The restrictions on the placement of fuel assemblies within the Region 1 A of the spent fuel pool, which have accumulated burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1 or number of IFBA rods greater than or equal to the minimum qualifying number of IFBA rods in Table 3.7.15-12 in the accompanying LCO, ensures the k n of the spent fuci pool will always remain s 0.95, assuming the pool to be flooded with water borated to 440 ppm. Fuel assemblies not meeting the criteria of Tables 3.7.15-1 or 3.7.15-12 shall be stored in accordance with  ;

Fiqure 3.7.15-1.

N- )

4 1

l f I i  !

McGuire Units 1 and 2 B 3.7.15-3 Revision No.

c l

I

t. . 1

Sp:nt Fu 1 Ass:mbly Storage B 3.7.15 BASES EW 3.7.15 BASES LCO (continued) h /

The restrictions on the placement of fuel assemblies within the Region 1B of the spent fuel pool, which have accumulated burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-3 in the accompanying LCO, ensures the kon of the spent fuel pool will always remain s 0.95, assuming the pool to be flooded with water borated to 440 ppm. Fuel assemblies not meeting the criteria of Table 3.7.15-3 shall be stored in accordance with either Figure 3.7.15-2 and Table 3.7.15-2 for Restricted storage, or Figure 3.7.15-3 for Checkerboard storage.

9 The restrictions on the placement of fuel assemblies within the Region 2A of the spent fuel pool, which have accumulated burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-6 in the accompanying LCO, ensures the k n of the spent fuel pool will always remain s 0.95, assuming the pool to be flooded with water borated to 440 ppm. Fuel assemblies not meeting the criteria of Table 3.7.15-6 shall be stored in accordance with either Figure 3.7.15-4 and Table 3.7.15-7 for Restricted storage, or Figure 3.7.15-5 for Checkerboard storage.

l 1

d j The restrictions on the placement of fuel assemblies within the Region 2B of the spent fuel pool, which have accumulated burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-9 in the accompanying LCO, ensures the ken of the spent fuel pool will always remain s 0.95, assuming the pool to be flooded with water borated to 440 ppm. Fuel assemblies not meeting the criteria of Table 3.7.15-9 shall be '

stored in accordance with either Figure 3.7.15-6 and Table 3.7.15-10 for Restricted storage, or Figura 3.7.15-7 for Checkerboard storage.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pool.  !

ACTIONS M  ;

I Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the configuration of fuel assemblies stored in the spent fuel pool is

-in nennrdance with the LCO, the immediate action is to initiate action i McGuire Units 1 and 2 B 3.7.15-4 Revision No.

l l

Sp:nt Fa::1 Ass:mbly Storage B 3.7.15 NEW 3.7.15 BASES BASES /

LCO (continued) to make the necessary fuel assembly movement (s) to bring t e configuration into compliance.

If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies by administrative means that the fuel assembly is in accordance with the configurations specified in the accompanying LCO.

l-REFERENCES 1. UFSAR, Section 9.1.2.

2. Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOS. M89744 and M89745), November 6,1995.

3. WCAP-14416-NP-A, Westinghouse Spent Fuel Rack Criticality Analysis Methodology, Revision 1, November 1996.

4. American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI /ANS-57.2-1983, October 7,1983.

5. Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, " Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19,1998.

6. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, l Appendix A).

l 7, 10 CFR 50.36, Technical Specifications, (c)(2)(ii).

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% /

McGuire Units 1 and 2 B 3.7.15-5 Revision No.

ATTACHMENT 11 REVISED MCGUIRE NUCLEAR STATION TECHNICAL SPECIFICATION BASES

o Sp:nt Fu:1 Pool Boron Conc:ntration B 3.7.14 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Boron Concentration BASES BACKGROUND in the two region poison fuel storage rack (Refs.1 and 2) design, the

( spent fuel pool is divided into two separate and distinct regions. 1 Region 1, with 286 storage positions, is designed and generally reserved J for temporary storage of new or partially irradiated fuel. Region 2, with 1 1177 storage positions, is designed and generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying burnup levels.

The McGuire spent fuel storage racks contain Boraflex neutron-absorbing panels that surround each storage cell on all four sides (except for peripheral sides). The function of these Boraflex panels is to ensure that the reactivity of the stored fuel assemblies is maintained within required l limits. Boraflex, as manufactured, is a silicon rubber material that retains '

a powder of boron carbide (B4C) neutron absorbing material. The

, Boraflex panels are enclosed in a formed stainless steel wrapper sheet that is spot-welded to the storage tube. The wrapper sheet is bent at each end to complete the enclosure of the Boraflex panel. The Boraflex panel is contained in the plenum area between the storage tube and the  !

wrapper plate. Since the wrapper plate enclosure is not sealed, spent  !

fuel pool water is free to circulate through the plenum. It has been i observed that after Boraflex receives a high gamma dose from the stored  !

irradiated fuel (>10' rads) it can begin to degrade and dissolve in the wet environment. Thus, the B4C poison material can be removed, thereby reducing the poison worth of the Boraflex sheets. This phenomenon is i documented in NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks".

To address this degradation, each region of the spent fuel pool has been divided into two sub-regions; with and without credit for Boraflex. For the regions taking credit for Boraflex, a minimum amount of Boraflex was assumed that is less than the original design minimum Bio areal density.

The McGuire spent fuel storage racks have been analyzed taking credit for soluble boron as allowed in Reference 3. The methodology ensures that the spent fuel rack multiplication factor, kon, is less than or equal to 0.95 as recommended in ANSl/ANS-57.2-1983 (Ref. 4) and NRC guidance (Ref. 5). The spent fuel storage racks are analyzed to allow storage of fuel assemblies with enrichments up to a maximum nominal '

enrichment of 4.75 weight percent Uranium-235 while maintaining kons l l

McGuire Units 1 and 2 B 3.7.141 Revision No.

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Sp:nt Fu:1 Pool Boron Conc:ntration B 3.7.14 l

l BASES l BACKGROUND (continued) 0.95 including uncertainties, tolerances, bias, and credit for soluble boron. Soluble boron credit is used to offset uncertainties, tolerances, and off-normal conditions and to provide subcritical margin such that the spent fuel pool kon is maintained less than or equal to 0.95. The soluble boron concentration required to maintain kon less than or equal to 0.95 under normal conditions is 440 ppm. In addition, sub-criticality of the pool (k.n < 1.0) is assured on a 95/95 basis, without the presence of the soluble boron in the pool. The criticality analysis performed shows that the acceptance criteria for criticality is met for the storage of fuel assemblies when credit is taken for reactivity depletion due to fuel l burnup, the presence of Integrated Fuel Burnable Absorber (IFBA) rods, reduced credit for the Boraflex neutron absorber panels and storage configurations and enrichment limits Specified by LCO 3.7.15.

APPLICABLE Most accident conditions do not result in an increase in reactivity of the SAFETY ANALYSES racks in the spent fuel pool. Examples of these accident conditions are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly between rack modules (rack design precludes this condition), and the I drop of a fuel assembly between rack modules and the pool wall.

However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage poow. The first is a drop or .

placement of a fuel assembly into the cask loading area. The second is a l significant change in the spent fuel pool water temperature (either the loss of normal cooling to the spent fuel pool water which causes an increase in the pool water temperature or a large makeup to the pool with cold water which causes a decrease in the pool water temperature) and the third is the misloading of a fuel assembly into a location which the restrictions on location, enrichment, burnup and number of IFBA rods is not satisfied.

i For an occurrence of these postulated accidents, the double contingency l principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter  !

(Ref. 6) can be applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for these postulated accident conditions, the presence of additional soluble boron in the spent fuel pool water (above the 440 ppm required to maintain k n less than or equal to 0.95 under i normal conditions) can be assumed as a realistic initial condition since l not assuming its presence would be a second unlikely event.

Calculations were performed to determine the amount of soluble boron l required to offset the highest reactivity increase caused by either of I

McGuire Units 1 and 2 B 3.7.14-2 Revision No.

1 i

I

i l Sp:nt Fu:1 Pool Boron Concentration B 3.7.14 i

BASES APPLICABLE SAFETY ANALYSES (continued) these postulated accidents and to maintain k.n less than or equal to 0.95.

It was found that a spent fuel pool boron concentration of 1170 ppm was adequate to mitigate these postulated criticality related accidents and to maintain k.n less than or equal to 0.95. Specification 3.7.14 ensures the spent fuel pool contains adequate dissolved boron to compensate for the increased reactivity caused by these postulated accidents.

Specification 4.3.1.1 c. requires that the spent fuel rack k n be less than or equal to 0.95 when flooded with water borated to 440 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent fuel pool before the 0.95 k n design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron concentration to 440 ppm is not a credible event.

The concentration of dissolved boron in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 5).

LCO The spent fuel pool boron concentration is required to be within the limits specified in the COLR. The specified concentration of dissolved boron in the spent fuel pool preserves the assumptions used in the analyses of the potential criticality accident scenarios as described in Reference 4. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pool.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pool.

ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the fuel storage pool is less than requi. red, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the l

movement of fuel assemblies. The concentration of boron is restored j simultaneously with suspending movement of fuel assemblies.

McGuire Units 1 and 2 B 3.7.14-3 Revision No.

Spent Futi Pool Boron Concantration B 3.7.14 1

BASES i ACTIONS (continued) I If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation. Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies that the concentration of boron in the spent fuel pool is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of pool water is expected to take place over such a short period of time.

REFERENCES 1. UFSAR, Section S.1.2.

2. ' Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOS M89744 and M89745), November 6,1995.

l

3. WCAP-14416-NP-A, Westinghouse Spent Fuel Rack Criticality Analysis Methodology, Revision 1, November 1996.

4. American Nuclear Society, "American National Standard Design  ;

Requirements for Light Water Reactor Fuel Storage Facilities at l Nuclear Power Plants," ANSI /ANS-57.2-1983, October 7,1983.

i

5. Nuclear Regulatory Commission, Memorandum to Timothy Collins i from Laurence Kopp, " Guidance on the Regulatory Requirements l for Criticality Ana!ysis of Fuel Storage at Light Water Reactor Power Plants," August 19,1998.

6. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the  ;

proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

7. 10 CFR 50.36 Technical Specifications, (c)(2)(ii).

8. UFSAR, Section 15.7.4. ,

McGuire Units 1 and 2 B 3.7.14-4 Revision No.

l Sp:nt ;-'usl Ass:mbly Storage B 3.7.15 BASES B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage ,

BASES BACKGROUND in the two region poison fuel storage rack (Refs.1 and 2) design, the i spent fuel pool is divided into two separate and distinct regions.

Region 1, with 286 storage positions, is designed and generally reserved for temporary storage of new or partially irradiated fuel. Region 2, with 1177 storage positions, is designed and generally used for normal, long term storage of permanently discharged fuel that has achieved qualifying burnup levels.

The McGuire spent fuel storage racks contain Boraflex neutron-absorbing panels that surround each storage cell on all four sides (except for  ;

peripheral sides). The function of those Boraflex panels is to ensure that the reactivity of the stored fuel assemblies is maintained within required limits. Boraflex, as manufactured, is a silicon rubber material that retains i

a powder of boron carbide (B4C) neutron absorbing material. The Boraflex panels are enclosed in a formed stainless steel wrapper sheet that is spot-welded to the storage tube. The wrapper sheet is bent at each end to complete the enclosure of the Boraflex panel. The Boraflex panel is contained in the plenum area between the storage tube and the wrapper plate. Since the wrapper plate enclosure is not sealed, spent fuel pool water is free to circulate through the plenum. It has been observed that after Boraflex receives a high gamma dose from the stored irradiated fuel (>10' rads) it can begin to degrade and dissolve in the wet environment. Thus, the B4C poison material can be removed, thereby reducing the poison worth of the Boraflex sheets. This phenomenon is

-documented in NRC Generic Letter 96-04, "Boraflex Degradation in Spent Fuel Pool Storage Racks".

To address this degradation, each region of the spent fuel pool has been  ;

divided into two sub-regions; with and without credit for Boraflex. For the j regioris taking credit for Boraflex, a minimum amount of Boraflex was assumed that is less than the original design minimum B10 areal density.

To address this degradation, each region of the spent fuel pool has been divided into two sub-regions; with and without credit for Boraflex. For the regions taking credit for Boraflex, a minimum amount of Boraflex was assumed that is less than the original design minimum B10 areal density.

( Two storage configurations are defined for each region; Unrestricted and l Restricted storage. Unrestricted storage allows storage in all cells without restriction on the storage configuration. Restricted storage allows storage of higher reactivity fuel when restricted to a certain storage l

McGuire Units 1 and 2 B 3.7.15-1 Revision No.

Sp nt Fu:1 Ass:mbly Storage B 3.7.15 BASES i

BACKGROUND (continued) configuration with lower reactivity fuel. A third loading pattern, Checkerboard storage, was defined for Regions 1B, 2A and 2B.

Checkerocard ctorage allows storage of the highest reactivity fuelin each region when checkerboarded with empty storage cells.

l The McGuire spent fuel storage racks have been analyzed taking credit for soluble boron as allowed in Reference 3. The methodology ensures that the spent fuel rack multiplication factor, k n, is less than or equal to 0.95 as recommended in ANSI /ANS-57.2-1983 (Ref. 4) and NRC guidance (Ref. 5). The spent fuel storage racks are analyzed to allow storage of fuel assemblies with enrichments up to a maximum nominal enrichment of 4.75 weight percent Uranium-235 while maintaining kon s 0.95 including uncertainties, tolerances, bias, and credit for soluble boron. Soluble boron credit is used to offset uncertainties, tolerances, and off-normal conditions and to provide subcritical margin such that the spent fuel pool k.n is maintained less than or equal to 0.95. The soluble boron concentration required to maintain k.n less than or equal to 0.95 under normal conditions is 440 ppm. In addition, sub-criticality of the pool (k.n < 1.0) is assured on a 95/95 basis, without the presence of the soluble boron in the pool. The criticality analysis performed shows that the acceptance criteria for criticality is met for the storage of fuel assemblies when credit is taken for reactivity depletion due to fuel burnup, the presence of Integrated Fue! Burnable Absorber (IFBA) rods, reduced credit for the Boraflex neutron absorber panels and storage configurations and enrichment limits Specified by LCO 3.7.15.

APPLICABLE Most accident conditions do not result in an increase in reactivity of the SAFETY ANALYSES racks in the spent fuel pool. Examples of these accident conditions are the drop of a fuel assembly on top of a rack, the drop of a fuel assembly i between rack modules (rack design precludes this condition), and the drop of a fuel assembly between rack modules and the pool wall.

However, three accidents can be postulated which could result in an increase in reactivity in the spent fuel storage pools. The first is a drop or placement of a fuel assembly into the cask loading area. The second is a significant chanpe in the spent fuel pool water temperature (either the loss of normal cooling to the spent fuel pool water which causes an increase in the pool water temperature or a large makeup to the pool with cold water which causes a decrease in the pool water temperature) and the third is the misloading of a fuel assembly into a location which the restrictions on location, enrichment, burnup and number of IFBA rods is not satisfied.

For an occurrence of these postulated accidents, the double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter McGuire Units 1 and 2 8 3.7.15-2 Revision No.

l Sp:nt Fu:1 Ass:mbly Storage B 3.7.15

! BASES APPLICABLE SAFETY ANALYSES (continued)

(Ref. 6) can be applied. This states that one is not required to assume l two unlikely, independent, concurrent events to ensure protection against l a criticality accident. Thus, for these postulated accident conditions, the '

presence of additional soluble boron in the spent fuel pool water (above the 440 ppm required to maintain k.n less than or equal to 0.95 under normal conditions) can be assumed as a realistic initial condition since  ;

not assuming its presence would be a second unlikely event.

Calculations were performed to determine the amount of soluble boron required to offset the highest reactivity increase caused by either of these postulated accidents and to maintain kon less than or equal to 0.95. It was found that a spent fuel pool boron concentration of 1170 ppm was adequate to mitigate these postulated criticality related accidents and to maintain ken less than or equal to 0.95. .o7ecification 3.7.14 ensures the spent fuel pool contains adequate dissoh d boron to compensate for the increased reactivity caused by these postulated accidents.

Specification 4.3.1.1 c. requires that the spent fuel rack k n be less than or equal to 0.95 when flooded with water borated 13 440 ppm. A spent fuel pool boron dilution analysis was performed which confirmed that sufficient time is available to detect and mitigate a dilution of the spent I fuel pool before the 0.95 k n design basis is exceeded. The spent fuel pool boron dilution analysis concluded that an unplanned or inadvertent event which could result in the dilution of the spent fuel pool boron l concentration to 440 ppm is not a credible event.

The configuration of fuel assemblies in the spent fuel pool satisfies Criterion 2 of 10 CFR 50.36 (Ref. 7).

LCO a The restrictions on the placement of fuel assemblies within the Region 1 A of the spent fuel pool, which have accumulated burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-1 or number of IFBA rods greater than or equal to the minimum qualifying number of IFBA rods in Table 3.7.15-12 in the accompanying LCO, ensures the kon of the spent fuel pool will always remain s 0.95, assuming the pool to be flooded with water borated to 440 ppm. Fuel assemblies not meeting the criteria of Tables 3.7.15-1 or 3.7.15-12 shall be stored in accordance with Figure 3.7.15-1.

l l

l l

McGuire Units 1 and 2 8 3.7.15-3 Revision No.

r Sp nt Funi Asssmbly Storage i B 3.7.15 BASES LCO (continued) 12 The restrictions on the placement of fuel assemblies within the Region 1B of the spent fuel pool, which have accumulated burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-3 in the accompanying LCO, ensures the k n of the spent fuel pool will always remain s 0.95, assuming the pool to be flooded with water borated to 440 ppm. Fuel assemblies not meeting the criteria of Taule 3.7.15-3 shall be stored in accordance with either Figure 3.7.15-2 and Table 3.7.15-2 for Restricted storage, or Figure 3.7.15-3 for Checkerboard storage.

c The restrictions on the placement of fuel assemblies within the Region 2A of the spent fuel pool, which have accumulated burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-6 in the accompanying LCO, ensures the k n of the spent fuel pool will always remain s 0.95, assuming the pool to be flooded with water borated to 440 ppm. Fuel assemblies not meeting the criteria of Table 3.7.15-6 shall be stored in accordance with either Figure 3.7.15-4 and Table 3.7.15-7 for Restricted storage, or Figure 3.7.15-5 for Checkerboard storage.

d The restrictions on the placement of fuel assemblies within the Region 2B of the spent fuel pool, which hav3 accumulated burnup greater than or equal to the minimum qualified burnups in Table 3.7.15-9 in the accompanying LCO, ensures the kon of the spent fuel pool will always remain s 0.95, assuming the pool to be flooded with water borated to 440 ppm. Fuel assemblies not meeting the criteria of Table 3.7.15-9 shall be stored in accordance with either Figure 3.7.15-6 and Table 3.7.15-10 for Restricted storage, or Figure 3.7.15-7 for Checkerboard storage.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel

^

pool.

ACTIONS .A_a Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with the LCO, the immediate action is to initiate action McGuire Units 1 and 2 8 3.7.15-4 Revision No.

Sp:nt Fut! Ass mbly Storaga B 3.7.15 BASES LCO (continued) to make the necessary fuel assembly movement (s) to bring the configuration into compliance.

If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel aseemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation. Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS This SR verifies by administrative means that the fuel assembly is in accordance with the configurations specified in the accompanying LCO.

REFERENCES 1. UFSAR, Section 9.1.2.

2. Issuance of Amendments, McGuire Nuclear Station, Units 1 and 2 (TAC NOS. M89744 and M89745), November 6,1995.

1

3. WCAP-14416-NP-A, Westinghouse Spent Fuel Rack Criticality j Analysis Methodology, Revision 1, November 1996. t

4. American Nuclear Society, "American National Standard Design Requirements for Light Water Reactor Fuel Storage Facilities at Nuclear Power Plants," ANSI /ANS-57.2-1983, October 7,1983.

5. Nuclear Regulatory Commission, Memorandum to Timothy Collins from Laurence Kopp, " Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light Water Reactor Power Plants," August 19,1998.

6. Double contingency principle of ANSI N16.1-1975, as specified in the April 14,1978 NRC letter (Section 1.2) and implied in the ,

! proposed revision to Regulatory Guide 1.13 (Section 1.4, I Appendix A).

7. 10 CFR 50.36. Technical Specifications, (c)(2)(ii).

(

i f McGuire Units 1 and 2 B 3.7.15-5 Revision No.

l

!