ST-HL-AE-5015, Errata to Application for Amends to Licenses NPF-76 & NPF-80 Incorporating Changes to Updated FSAR to Reflect Increase in Spent Fuel Pool Heat Loads

Errata to Application for Amends to Licenses NPF-76 & NPF-80 Incorporating Changes to Updated FSAR to Reflect Increase in Spent Fuel Pool Heat Loads

ML20084U453
Person / Time
Site:	South Texas
Issue date:	05/30/1995
From:	Groth J HOUSTON LIGHTING & POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20084U395	List: ST-HL-AE-5015, Errata to Application for Amends to Licenses NPF-76 & NPF-80 Incorporating Changes to Updated FSAR to Reflect Increase in Spent Fuel Pool Heat Loads ST-HL-AE-5102, Forwards Errata to Proposed License Amend Re Increase in Spent Fuel Pool Heat Loads,Including Resubmittal of Proposed License Amend,Annotated W/Date on Which Original Submittal Signed
References
ST-HL-AE-5015, NUDOCS 9506130190
Download: ML20084U453 (80)

Text

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o The Light companySouth llouston Lighting & Power exas Project ectric ,crerating Station P. O. Bom m Whorth, Texas m83 May 30,1995 ST-HL-AE-5015 File No.: G20.01, G21.01 10CFR50.90 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499 Proposed License Amendment Concerning an Increase in Soent Fuel Pool Heat Loads Pursuant to 10CFR50.90, Houston Lighting & Power Company proposes to amend its Operating Licenses NPF-76 and NPF-80 by incorporating the attached changes to the Updated Final Safety Analysis Report for the South Texas Project Units I and 2. The purpose of this license change is to revise the Spent Fuel Pool heat load licensing basis to provide greater flexibility for normal refueling practices.

Current South Texas Project licensing basis calculations for heatload to the Spent Fuel Pool are based on the assumption that the entire core is offloaded to the Spent Fuel Pool Juring refueling, and all but typically 65 assemblies are returned to the reactor. This is based on a twelve-month refueling cycle. Under the present 18-month refueling cycle, refueling results in typically 88 fuel assemblies being left in the Spent Fuel Pool. As a result, precautions have been taken to ensure that the heat balance limits are met so that the heat load remains within the licensing basis. To ensure South Texas Project remained within the design basis, fuel offload to the Spent Fuel Pool was not permitted until the decay heat was less than that assumed in the licensing basis.

The licensing basis calculations have been reperformed using the methodology given in Section 9.l.3 of the Standard Review Plan, assuming more fuel assemblies being offloaded as a basis for the heat load. In this case the calculated maximum temperatures may exceed the limits identified in the Safety Evaluation Report.

Pursuant to 10CFR50.59, the calculated increase in the Spent Fuel Pool temperature 4 decreases the margin of safety and therefore requires review as an unreviewed safety question by the Nuclear Regulatory Commission. However, the attached safety evaluation shows that the increase does not constitute a significant hazard.

rroject Manager on Behalf of the Participants in the south Texas Project

Houston Lighting & Power Company South Texas Project Electric Generating Station ST-HL-AE-5015 File No.: G20.01, G21.01 Page 2 Houston Lighting &_ Power has reviewed the attached proposed amendment pursuant to 10CFR50.92 and determined that it does not involve a significant hazards consideration. In addition, Houston Lighting & Power has determined that the proposed amendment satisfies the criteria of 10CFR51.22(c)(9) for categorical exclusion from the requirement for an environmental assessment.

The South Texas Project Nuclear Safety Review Board and the Plant Operations Review Committee have reviewed and approved the proposed changes.

Houston Lighting & Power requests that the effective date of this amendment be 30 days after the date of Nuclear Regulatory Commission approval. Although this request is neither exigent nor an emergency, issuance of this amendment by the Nuclear Regulatory Commission by August 31,1995, is requested.

In accordance with 10CFR50.91(b), Houston Lighting & Power is providing the State of Texas with a copy of this proposed amendment.

If you should have any questions concerning this matter, please contact Mr. K. J. Taplett at (512) 972-8416 or me at (512) 972-8664.

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, J. F. Groth Vice President, Nuclear Generation PLW/lf Attachments: 1) Summary and Description of the Proposed Changes

2) No Significant Hazards Consideration Determination

3) Marked-Up Updated Final Safety Analysis Report Pages

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-V l Houston Lighting & Power Company 'I ST-HL-AE-5015 s South Texas Project Electric Generating Station File No.: G21.01 '

Page 3 -!

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Leonard J. Callan Rufus S. Scott.

Regional Administrator, Region IV- Associate General Counsel i

.U. S. Nuclear Regulatory Commission Houston Lighting & Power Company - -l 611 Ryan Plaza Drive, Suite 400 ' P. O. Box 61067 Arlington, TX 76011-8064 Houston, TX 77208  !

Thomas W.' Alexion Institute of Nuclear Power Project Manager Operations - Records Center '

U. S. Nuclear Regulatory Commission - 700 Galleria Parkway .  ;

Washington, DC 20555-0001 13H15 Atlanta, GA 30339-5957

David P.' Lovelesy ,

Dr. Joseph M. Hendrie .

Sr; Residsnt In'spector 50 Bellport Lane .

c/o U. S. Nuclear Regulatory Comm. Bellport, NY ' 11713  :

P. O. Box 910 Bay City, TX 77404-0910 Richard A. Ratliff Bureau of Radiation Control .

J. R. Newman, Esquire Texas Department of Health  !

Morgan, Lewis & Bockius 1100 West 49th Street 1800 M Street, N.W. Austin, TX 78756-3189 Washington, DC 20036-5869 U. S. Nuclear Regulatory Comm.

K. J. Fiedler/M. T. Hardt Attn: Document Control Desk  !

City Public Service Washington, D. C. 20555-0001 P. O. Box 1771  ;

San Antonio, TX 78296  ;

J. C. Lanier/M. B. Lee J. R. Egan, Esquire  ;

City of Austin Egan & Associates, P.C.  ?

Electric Utility Department 2300 N Street, N.W.

721 Barton Springs Road Washington, D.C. 20037 Austin, TX 78704 i i

C. A. Johnson ,

Central Power and Light Company ,

P. O. Box 289, Mail Code: N5012 ,

Wadsworth, TX 77483 I

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I UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION In the Matter of - )

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Houston Lighting & Power ) Docket Nos. 50-498 Company, et al., -) 499- l

) r South Texas Project )  :

Units I and 2 )

AFFIDAVIT J. F. Groth, being duly sworn, hereby deposes and says that he is Vice President, Nuclear Generation of Houston Lighting & Power Company; that he is duly authorized to sign and file  :

with the Nuclear Regulatory Conunission the attached proposed amendment to the South Texas Project Units 1 and 2 concerning an increase in spent fuel pool heat loads; is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of his  ;

knowledge and belief.

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J. F. Groth ice President, Nuclear Generation STATE OF TEXAS )  ;

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Subscribed and sworn to before me, a Notary Public in and for The State of Texas this 8C "' day of Ray ,1995.

No .b]Lkb Notar Public in and for The State of Texas 1

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ST-HL-AE-5015 '

File No.: G20.01 G21.01 1

ATTACHMENT 1 ,

SUMMARY

AND DESCRIPTION OF THE PROPOSED CHANGES l

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a Attachment 1 Page 1 of 14 SPENT FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTH CYCLES 1.0

SUMMARY

Currently, the South Texas Project spent fuel pool heatup calculations assume the entire core is discharged to the Spent Fuel Pool, and all but typically one-third of the core (65 assemblies) is reinserted in the reactor during each refueling outage. However, this is based on use of a twelve-month refueling cycle. Houston Lighting & Power has performed an analysis based on an eighteen-month refueling cycle under which typically eighty-eight assemblies are left in the Spent Fuel Pool during each refueling outage. Increasing the assumed number of fuel assemblies being discharged results in an increased heat load to the Spent Fuel Pool. Houston Lighting & Power proposes to revise the Updated Final Safety Analysis Report Section 9.1 and Tables 9.1-1 and 9.1-5 to reflect the recalculated Spent Fuel Pool heatup temperatures. ,

Using methods more severe than prescribed in Section 9.1.3 of the Standard Review Plan, Houston Lighting & Power has determined that the calculated maximum temperature in the Spent Fuel Pool, with the 18-month cycle Vantage 5H fuel, may exceed the limits specified in the Safety Evaluation Report for the South Texas Project. Pursuant to 10CFR50.59, this increase in temperature is a decrease in the margin of safety, and therefore requires review ,

as an unreviewed safety question. However, the following safety evaluation shows thet the increase does not constitute a significant hazard.  ;

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SPENT FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTN CYCLES  !

2.0 BACKGROUND

I The current refueling practice at the South Texas Project is as follows:

1. Remove the entire core (193 fuel assemblies) and place it in the Spent Fuel Pool.

This is a full-core offload and is defined as the Abnormal Maximum condition in Section 9.1.3 of the Standard Review Plan.

2. Transfer a fraction of the 193 fuel assemblies from the Spent Fuel Pool back to the reactor for use in the next cycle. Tne fuel assemblies not transferred remain in the Spent Fuel Pool. This is the Normal Maximum condition as defined in Section 9.1.3 of the Standard Review Plan.

Currently, the South Texas Project licensing basis assumes one-third of the core (65 assemblies) is discharged into the Spent Fuel Pool for each normal refueling. This assumption was based on 12-month refueling cycles. In April 1991 and December 1991, respectively, South Texas Project Units 1 and 2 began the first 18-month fuel cycles. For an 18-month fuel cycle, typically 88 fuel assemblies can be discharged to the Spent Fuel Pool for a given fuel cycle. To ensure the South Texas Project remains within the licensing basis for heat load to the Spent Fuel Pool, fuel offload to the Spent Fuel Pool is not permitted until the decay heat is less than that assumed in the licensing basis.

Houston Lighting & Power has revised the Spent Fuel Pool heatup analysis to incorporate the routine refueling practice of full-core offload and 18-month cycles. In the revised Spent Fuel Pool heatup analysis, Heuston Lighting & Power has also conservatively accounted for the higher peaking factor and enthalpy rise factor due to the Vantage 5H Fuel Upgrade program. The revision changes the licensing basis of the plant. -

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Page 3 of 14 l SPEur FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTH CYCLES l 1

3.0 PROPOSED CHANGE

S Houston Lighting & Power proposes to modify the Updated Final Safety Analysis Report Section 9.1 and Tables 9.1-1. The proposed change would revise the Spent Fuel Pool heatup temperatures in the Updated Final Safety Analysis Report Table 9.1-1. Particularly, the peak Spent Fuel Pool temperatures for the following cases would exceed the licensing limits given in Safety Evaluation Report, Supplement 6, Appendix BB:

= Normal Maximum Case:

Current 145.7 F (65 assemblies discharged 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> after shutdown, one cooling train in operation)

Proposed 155.0 F (88 assemblies discharged 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown, one cooling train in operation) -

= Rapid Refueling Case:

Current 129.2 F (65 assemblies discharged 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> after shutdown, two cooling trains)

Proposed 135.0 F (88 assemblies discharged 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown, two cooling trains)

= Peak Spent Fuel Pool Temperature:

Rapid Refueling Case:

Current 150.7 F (65 assemblies discharged 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> after shutdown, one cooling train)

Proposed 155.0 F (193 assemblies discharged 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown, two cooling trains) 202.0 F (193 assemblies discharged 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown, one cooling train)

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Attachment 1 Page 4 of 14 SPENT FUEL POOL HEATUP FOR Futt-CORE OFFLOAD AND 18-MONTH CYCLES Abnormal Maximum Case:

Current 155.4 F (193 assemblies discharged 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown, two cooling trains)

Proposed 205.0 F (193 assemblies discharged 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown, one cooling train) 156.0 F (193 assemblies discharged 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown, two cooling trains)

Temperatures for the other cases presented in the revised Updated Final Safety Analysis Report Table 9.1-1 would remain below the Standard Review Plan criteria of 140 F for the Normal Maximum cases, and no boiling condition for the Abnormal Maximum cases.

The following Updated Final Safety Analysis Report changes clarify the refueling practices and design of the Spent Fuel Pool and are included in this submittal for completeness:

(a) Updated Final Safety Analysis Report Section 9.1.2.3, Safety Evaluation for Spent Fuel Storage, will be revised to reflect that a complete loss of Spent Fuel Pool water inventory is not a credible event.

(b) Updated Final Safety Analysis Report Section 9.1.3.1.2, describing dewatering protection for the Spent Fuel Pool Cooling and Cleanup System, will be revised to describe that draindown is not possible for credible design basis breaks.

(c) A new Updated Final Safety Analysis Report Section 9.1.3.2.2, " Spent Fuel Pool Cooling During Refueling Operation," will be added to describe the current Spent Fuel Pool cooling method during refueling operation.

(d) Updated Final Safety Analysis Report Section 9.1.3.3.1, describing availability and reliability of the Spent Fuel Pool Cooling and Cleanup System, will be revised to discuss 18-month refueling cycles.

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Attachment 1 Page 5 c114 SPENT FUEL. POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTH C.YCLES (e) Updated Final Safety Analysis Report Table 9.1-1, " Spent Fuel Pool Cooling and Cleanup System Design Parameters," will be replaced to provide the revised Spent Fuel Pool temperatures and heat loads for various scenarios and the 18 month fuel cycles. The title of the table will also be changed to read " Spent Fuel Pool Heatup Analysis Results."

(f) Updated Final Safety Analysis Report Table 9.1-5, " Spent Fuel Poo! Cooling and Cleanup System Failure Modes and Effects Analysis," will be revised to include effects of failure that may result in a complete loss of Spent Fuel Pool cooling.

(g) Editorial changes are included in various other sections.

Updated Final Safety Analysis Report markups of the proposed changes are provided as Attachment 3.

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6 Attachment 1 Page 6 of 14 SPENT FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTH CYCLES 4.0 SAFETY EVALUATION This safety evaluation addresses the following topics:

1. Spent Fuel Pool time to boil

2. Spent Fuel Pool boil-off rate i

3. Normal Maximum case temperature

4. Rapid Refueling case temperature

5. Abnormal Maximum case temperature

6. Peak clad temperature l

7. Peak Boraflex (poison) temperature l

8. Stainless steel fuel box temperature l

9. Spent Fuel Pool liner temperature l

10. Spent Fuel Pool concrete temperature

11. Boiling dose consequences

12. Spent Fuel Pool calculated heatup rate comparison with plant data The safety evaluation includes the effects of the following:

(a) Full-core offload for routine refueling operation, (b) Fuel reload cycles of 18 months, (c) Higher peaking factor (Fq) to incorporate the Vantage 5H fuel upgrade effects, (d) Higher enthalpy rise factor (FAh) to incorporate the Vantage 5H fuel upgrade effects.

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4 Attachment 1 Page 7 of 14 SPENT FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTH CYCLES 4.1 SJPENT FUEL POOL TIME TO HOIL The impact of the proposed changes on the Spent Fuel Pool boiling conditions have been evaluated. In the event of a postulated complete loss of Spent Fuel Pool cooling capability and Abnormal Maximum case conditions, the time-to-boil is calculated to be 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, compared to the 2.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br /> under the current licensing basis. The analysis conservatively assumes that both cooling trains simultaneously fail just when the Spent Fuel Pool reaches its maximum temperature of 156 F. The 0.06 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> difference in time-to-boil is well within the error bounds of the calculation, and is partially due to the added conservatism of 2%

higher heat load for power measurement uncertainty. The South Texas Project Emergency Operating Procedures conservatively require 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> to restore cooling to the Spent Fuel Pool following a loss of cooling incident.

4.2 SPENT FUEL POOL HOIL-OFF RATE Houston Lighting & Power has determined that the maximum revised Spent Fuel Pool boil-off rate would be 131 gpm for the Abnormal Maximum case, while the current boil-off rate given in the Safety Evaluation Report is 135 gpm. Therefore, there is no adverse impact on the Spent Fuel Pool boil-off rate as a result of the proposed changes. The current means of makeup, described in Updated Final Safety Analysis Report Section 9.1.3.3.2, would not be affected.

4.3 NORMAL MAXIMUM CASE TEMPERATURE Houston Lighting & Power has evaluated the effects of the proposed changes on the Spent Fuel Pool bulk water temperature for the Normal Maximum Case. The peak bulk water temperature for this case (150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> after shutdown) would be 155 F. In the current safety analysis, the maximum pool temperature is 145.7 F, and is given in Supplement 6 to the Safety Evaluation Report, Appendix BB, Section 5.1. In the current safety analysis, the Spent Fuel Pool temperature remains above the Standard Review Plan limit of 140 F for 11.5 days. For the proposed changes, the maximum bulk temperature would be greater than the 140 F Standard Review Plan criteria for approximately 710 hours0.00822 days <br />0.197 hours <br />0.00117 weeks <br />2.70155e-4 months <br /> (30 days).

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The 30-day period in which the Spent Fuel Pool bulk water temperature would remain i above 140 F would not have an adverse affect on the structuralintegrity of the Spent Fuel Pool (see Sections 4.8 and 4.9.) The design temperatures for all Spent Fuel Pool i components (e.g., demineralizers, heat exchangers, etc.) are greater than the proposed Spent  ;

Fuel Pool temperatures presented in the revised Updated Final Safety Analysis Report Table 9.1-1. The peak temperature calculated for the Normal Maximum Case would be below the 225 F used in the original structural steel analysis. >

t 4.4 RAPID REFUELING CASE TEMPERATURE .

r Houston Lighting & Power has evaluated the effects of the proposed changes on the Spent Fuel Pool bulk water temperature for the Rapid Refueling Case (full-core offload 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown). The peak bulk water temperature for this case, with two Spent Fuel Pool cooling trains in operation, would be 155 F. The current safety analysis assumes one-third of the core is offloaded to the Spent Fuel Pool 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> after shutdown, resulting in a maximum pool temperatuce of 129.2 F.

Since a full-core ofDoad is the normal refueling practice at the South Texas Project, Houston Lighting & Power has also performed an analysis for this case considering a single active failure (e.g., loss of one Spent Fuel Pool cooling train). The peak bulk temperature -

was calculated to be 202 F. This temperature is lower than the "no boiling" criterion for the full-core ofuoad case specified in Standard Review Plan 9.1.3. This case was not previously considered in the Safety Analysis Report but is included here for completeness.

This case is bounded by the Abnormal Maximum case (see Section 4.5).

4.5 AllNORMAL MAXIMUM CASE TEMPERATURE For the Abnormal Maximum Case with full-core offload, Standard Review Plan Section 9.1.3 requires that the temperature of the pool remain below the boiling point. Also, a single active failure (e.g. loss of one Spent Fuel Pool cooling train) need not be considered for the Abnormal Maximum case. In the current safety analysis, the maximum Spent Fuel Pool temperature for this case is 155.4 F. The peak Spent Fuel Pool bulk water msc.~ om wws m3i

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SPENT FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTH CYCLES ,

e temperature for the proposed change would be 156 F. This is much lower than the "no  :

boiling" criterion specified for this case in Standard Review Plan 9.1.3. Therefore, the j revised Spent Fuel Pool water temperatures would be acceptable, given the conservative j assumptions and calculational models used to determine the heat loads and temperatures.  !

Since a full-core offload is the normal refueling practice at the South Texas Project, Houston Lighting & Power has also performed an analysis for this case considenng a single active failure involving loss of one Spent Fuel Pool cooling train. The peak bulk l temperature was calculated to be 205 F. This temperature is also lower than the "no l boiling" criterion specified in Standard Review Plan 9.1.3. This case was not previously l considered in the Safety Analysis Report but is included here for completeness. .

In the event of a postulated complete loss of Spent Fuel Pool cooling capability, the time-to-boil was calculated to be 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The time-to-boil analysis has been performed ,

for the Abnormal Maximum case with complete loss of cooling. The analysis .

conservatively assumes that both cooling trains fail just when the Spent Fuel Pool reaches its maximum temperature of 156 F. The South Texas Project design basis time-to-boil, given in Supplement 6 to the Safety Evaluation Report, Appendix BB, is 2.86 hours9.953704e-4 days <br />0.0239 hours <br />1.421958e-4 weeks <br />3.2723e-5 months <br />.

4.6 PEAK CLAD TEMPERATURE  ;

Houston Lighting & Power's analysis shows that the maximum fuel cladding temperature would be 223 F for the South Texas Project Rapid Refueling case (full core offload,100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> decay), which is higher than the current value of 202 F. To conservatively account for the Vantage 5H Fuel effects, the calculation for the peak clad temperature uses a higher peaking factor (Fq) of 2.7 versus 1.2 used previously and an enthalpy rise factor (FAh) of ,

1.62 versus 1.0 used previously. At the top of the fuel racks (about 23 ft below the Technical Specification minimum water level), the saturation temperature is approximately 238 F. This gives a subcooling margin of over 15 F for the peak clad temperature.

Therefore, no boiling would occur at the location of the peak clad temperature and the fuel cladding integrity would be maintained. Note that for the Abnormal Maximum case, the ,

peak clad temperature was calculated to be 216 F.

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SPENT FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTH CYCLES ,

l 4.7 PEAK BORAFLEX (POISON) TEMPERATURE Houston Lighting & Power's analysis shows that the maximum Boraflex poison temperature i due to gamma heating will not exceed the previously calculated temperatures. Input from consultants currently investigating this phenomenon indicates that the short duration of the maximum temperature will not substantially affect the integrity of the Boraflex. Therefore, the change does not have an adverse effect on the integrity of the Boraflex.

i 4.8 STAINLESS STEEL FUEL BOX TEMPERATURE Thermal analysis for the stainless steel rack walls shows the temperature drop through the stainless steel is less than 0.5 F. The thermal stresses in the rack walls resulting from this temperature difference would be negligible. Therefore, the stainless steel structural integrity will be maintained.

4.9 SPENT FUEL POOL LINER TEMPERATURE The Spent Fuel Pool liner plate and gates are designed to be exposed to water containing i boric acid solution at a pool temperature of 212 F. A Spent Fuel Pool temperature of 225 F was used for evaluating material properties for determining the allowable stresses.

The calculated Spent Fuel Pool temperatures are less than 212 F; therefore, the design conditions have been met.

4.10 SPENT FUEL POOL CONCRETE TEMPERATURE The concrete structure is designed for a pool temperature of 212 F. A maximum pool water temperature of 156 F has been calculated for the Abnormal Maximum case of full-core offload (the current licensing basis temperature for this case is 155.4 F). Therefore, the existing concrete design conditions have been met.

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h Attachment 1 Page 11 of 14 SPENT FUEL POOL HEAT 1JP FOR FULL-CORE OFFLOAD AND 18-MONTH CYCLES 4.11 HOILING DOSE CONSEOUENCES The Spent Fuel Pool boiling dose consequences following a complete loss of Spent Fuel l l Pool cooling for the full-core offload and 18-month reload cycles are described in the l

Updated Final Safety Analysis Report Section 9.1.3.3.4. This calculation assumes a full core discharge at 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown and instantaneous boiling in the Spent Fuel i Pool. The Spent Fuel Pool heatup analysis for the bounding case was performed for a full l core discharge at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown. The heatup calculation shows the Spent Fuel Pool reaches 202*F in 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br />. Therefore, the boiling dose assumption of Spent Fuel l Pool boiling in 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> is conservative and bounding. The results of the evaluation l show that the dose consequences of iodine release due to Spent Fuel Pool boiling are significantly below the allowable dose requirements of 10 CFR 100. Therefore, there will be no significant increase in hazards to the health and safety of the public.

4.12 SPENT FUEL POOL CALCULATED HEATUP RATE, CO5fPARISON WITH PLANT DATA )

Spent Fuel Pool heatup rate data was obtained during loss of Spent Fuel Pool cooling while motor-operated valves were being tested. The measured heatup rate was 0.3 F/hr. Based on the same Spent Fuel Pool conditions, the calculated heatup rate was determined using the Standard Review Plan Branch Technical Position ASB 9-2 assumptions and decay heat formulations. The heatup rate was calculated to be 0.7 F/hr. Since the calculated heatup rate over-estimates the measured heatup rate, the results presented in Updated Final Safety i Analysis Report Table 9.1-1 are conservative.

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5.0 CONCLUSION

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i The safety evaluation includes the effects of full-core offload,18-month reload cycles, higher peaking factors (Fq), and higher enthalpy rise factors (FAh) which incorporate the '

effects of Vantage V5H fuel upgrade. The evaluation shows that the small reduction in margin of safety does not significantly increase the hazards and is not a safety concern because the following conditions are acceptable:

1. The time-to-boil due to loss of all Spent Fuel Pool cooling is 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, which is ,

higher than the 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> used in the South Texas Project's Emergency Operating Procedures.

2. In the event of Spent Fuel Pool boiling, the reactor makeup water pump can provide i sufficient makeup water to meet the boil-off rate of 131 gpm. The assured pool '

makeup source is the Seismic Category I Low Head Safety Injection system. The Low Head Safety Injection pump has the capability to pump 1900 gpm, much in excess of the 131 gpm Spent Fuel Pool boil-off rate. The Low Head Safety Injection -

system mode of recovery is described in Supplement 6 to the Safety Evaluation Report, Appendix BB.

3. During a full-core offload refueling operation, various means of cooling, described in the Updated Final Safety Analysis Report Section 9.1.3.3.2, are available which ensure that the safety analysis described in Updated Final Safety Analysis Report Section 9.1.3 will remain valid.

4. For the design basis case of Rapid Refueling full-core offload, the fuel clad integrity is not compromised.

5. For the design basis case, the integrity of the Spent Fuel Pool Boraflex is not adversely impacted.

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f Attachment 1 Page 13 of 14 SPENT FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 1840 NTH CYCLES

6. Thermal stresses in the Spent Fuel Pool stainless steel rack walls are negligible and the stainless steel structural integrity is maintained. .

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7. The Spent Fuel Pool liner plate and gates are designed to be exposed to water containing boric acid solution at a pool temperature of 212 F. The calculated Spent Fuel Pool temperatures are less than 212 F; therefore, the design conditions have been met.

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8. The Spent Fuel Pool concrete structure is designed to be exposed to water at a pool temperature of 212 F. The calculated Spent Fuel Pool temperatures are less than 212 F; therefore, the design conditions have been met.

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9. The calculated Spent Fuel Pool boiling doses are conservative and well below the l limits of 10 CFR 100. l

10. The Spent Fuel Pool calculation overestimates the heatup rate as compared to plant measured data. Therefore, the calculated results presented in Updated Final Safety Analysis Report Table 9.1-1 are conservative and bounding.

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6.0 REFERENCES

1. NUREG-0800, US NRC Standard Review Plan, Section 9.1.3 " Spent Fuel Pool Cooling and Cleanup System," Revision 1, July 1981.

2. NUREG-0781, Safety Evaluation Report Related to the Operation of South Texas Project, Units 1 and 2, including Supplements 3, 6, and 7, Section 9.1 and Appendix BB " Safety Evaluation by the Office of the Nuclear Reactor Regulation Related to the Increase in the Spent Fuel Capacity Through the Use of High Density Storage Racks."

3. NUREG-1346, Technical Specifications, South Texas Project, Unit Nos.1 and 2, l Docket Nos. 50-498 and 50-499, Appndix "A" to License Nos. NPF-76 and NPF-80, 1 Section 3/4.9.3 " Decay Time," Mar.h 1989. l

4. South Texas Project Updated Final Safety Analysis Report, Sections 9.1.2,9.1.3, &

9.2.2, Revision 3.

5. EPRI TR-103300, " Guidelines for Boraflex Use in Spent-Fuel Storage Racks," ,

December 1993.

6. ST-HL-AE-2417, " Expansion of the Spent Fuel Pool Storage Capacity Using High Density Spent Fuel Racks;" letter from G. E. Vaughn to U. S. Nuclear Regulatory Commission, dated March 8,1988.  ;

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~, 4 ST-HL-AE-5015 File No.: G20.01, G21.01 l

ATTACHMENT 2 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION l

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.: i ST-HL-AE-5015 File No.: G20.01, G21.01 Attachment 2 ,

Page 1 of 2 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION i

i!ouston Lighting & Power has determined that the proposed changes do not involve a i signirk/.nt hazards consideration as defined in 10 CFR 50.92:

1. 'he proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated because:

(a) The Spent Fuel Pool conditions are not indicative of accident initiators. ,

(b) Design and operability requirements of equipment important to safety are 7

~

not affected.

(c) If only one Spent Fuel Pool cooling train is available, boiling would not occur and the Spent Fuel Pool components would remain within their  !

. design bases. ,

(d) The complete loss of Spent Fuel Pool cooling event has previously been analyzed and described in Supplement 6 to the Safety Evaluation Report, j Appendix DB. The dose consequences for this event have been evaluated and the safety evaluation is described in Updated Final Safety Analysis Report Section 9.1.3.3.4. The results of the evaluation show that the Spent Fuel Pool components would remain within their design bases. Also, the dose consequences of iodine release as a result of Spent Fuel Pool boiling are significantly below the allowable dose limits of 10 CFR 100. l

2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated because:

(a) The operability of safety-related equipment is not impacted.

(b) The probability of safety-related equipment malfunctioning is not increased. ,

(c) The scope of the change does not establish a potential new accident precursor.

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ST-HL-AE-5015 File No.: G20.01, G21.01 (d) The Spent Fuel Pool design considers design basis heat loads for the modified refueling procedure which includes a full-core offload.

Attachment 2 Page 2 of 2 NO SIGNIFICANT HA7ARDS CONSIDERATION DETERMINATION - (Continued) i (e) For the design basis case, the integrity of the Spent Fuel Pool Boraflex is not adversely impacted.

3. The proposed changes do not involve a significant reduction in a margin of safety because:

(a) No fuel damage would occur as a result of the proposed change. ,

(b) Technical Specification operability and surveillance requirements are not reduced.

(c) The Spent Fuel Pool boiling doses would be significantly below the allowable dose limits of 10 CFR 100.

(d) The modified refueling procedure (full-core offload) continues to have acceptable  !

margins of safety. l (e) For the design basis case, the integrity of the Spent Fuel Pool Boraflex is not adversely impacted.

Based on the safety evaluation presented above for the proposed changes, Houston Lighting &

Power has determined that the health and safety of the public will not be jeopardized. Therefore, the proposed changes do not involve a significant hazards consideration.

IMPLEMENTATION PLAN Houston Lighting & Power requests an implementation time of 30 days from the effective date of the approved license amendment to facilitate distribution and to make appropriate changes to plant documents. [

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ST-HL-AE-5015 l File No.: G20.01, G21.01 l

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1 ATTACHMENT 3 J

Marked-up Updated Final Safety Analysis Report Pages SPENT FUEL POOL HEATUP FOR FULL-CORE OFFLOAD AND 18-MONTH CYCLES i l

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- ST-HL-AE-5015 File No.: G20.01, G21.01 c

The following Updated Final Safety Analysis Report pages are provided in support of this amendment. Proposed revisions are indicated as appropriate.

TC 9-1 (Table of Contents)

TC 9-12 (List of Tables) 9.1-1* 9.2-10*

9.1-2* 9.2-11

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• Pages with no changes shown are provided to support review of proposed License Amendment.

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l TABLE OF CONTENTS '

CHAPTER 9 AUXILIARY SYSTEMS 1

Section Title Pace l 9.1 FUEL STORAGE AND HANDLING 9.1-1 9.1.1 New Fuel Storage 9.1-1 t 9.1.1.1 Design Bases 9.1-1 l 9.1.1.2 Facilities Description 9.1-1 9.1.1.3 Safety Evaluation 9.1-2 9.1.2 Spent Fuel Storage 9.1-2 9.1.2.1 Design Bases 9.1-2 9.1.2.2 Facilities Description 9.1-4 9.1.2.3 Safety Evaluation 9.1-5 9.1.3 Spent Fuel Pool Cooling and Cleanup System 9.1-5  ;

9.1.3.1 Design Bases 9.1-5 ,

9.1.3.1.1 Spent Fuel Cooling 9.1-6 ,

9.1.3.1.2 Dewatering Protection 9.1-6 9.1.3.1.3 Water Purification 9.1-6 9.1.3.2 System Description 9.1-6 9.1.3.2.1 Component Description 9.1-8 9.1.3.2.2 Spent Fuel Pool Cooling During Refueling Operations 9.1-9 9.1.3.3 Safety Evaluation 9.1-10 9.1.3.3.1 Availability and Reliability 9.1-10 9.1.3.3.2 Spent Fuel Storage Area Dewatering 9.1-10 i 9.1.3.3.3 Water Quality 9.1-10 9.1.3.3.4 Spent Fuel Pool Boiling Dose Analysis 9.1-11 9.1.3.4 Instrumentation Application 9.1-11 9.1.3.4.1 Temperature 9.1-11 9.1.3.4.2 Pressure 9.1-11  ;

9.1.3.4.3 Flow 9.1-11 9.1.3.4.4 Level 9.1-11 +

9.1.3.5 Tests and Inspections 9.1-12 9.1.4 Fuel Handling System (FHS) 9.1-12 9.1.4.1 Design Bases 9.1-12 9.1.4.2 System Description 9.1-12 9.1.4.2.1 New Fuel Handling 9.1-12 9.1.4.2.2 Refueling Procedure 9.1-13 9.1.4.2.2.1 Phase I - Preparation 9.1-14 9.1.4.2.2.2 Phase II - Reactor Disassembly 9.1-14 9.1.4.2.2.3 Phase III - Fuel Handling 9.1-15 9.1.4.2.2.4 Phase IV - Reactor Assembly 9.1-16 9.1.4.2.3 Spent Fuel Shipment 9.1-16 9.1.4.2.4 Component Description 9.1-17 9.1.4.2.4.1 Refueling Machine 9.1-17 9.1.4.2.4.2 Fuel Handling Machine 9.1-18 9.1.4.2.4.3 New Fuel Elevator 9.1-19 9.1.4.2.4.4 Fuel Transfer System 9.1-19 9.1.4.2.4.5 Rod Cluster Control Change Fixture 9.1-19 9.1.4.2 4.6 Spent Fuel Assembly Handling Tool 9.1-20 9.1.4.2.4.7 New Fuel Assembly Handling Tool 9.1-20 u musmsmo 7C 9-1 Revision 3

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LIST OF TABLES CHAPTER 9 Table Title Page 9.1-1 Spent Fuel Pool Cooling and Clcantg Sys w Occign Parsmctcr Heatup Analysis for 18-henth Reload Cycles 9.1-30 9.1-2 Spent Fuel Pool Cooling and Cleanup System, Component Design Parameters 9.1-31 9.1-3 FHB 15/2-Ton Crane - Compliance with Pegulatory Guide 1.104 9.1-35 9.1-4 NSSS Vendor Recommended Specifications and Guidelines for Spent Fuel Pool Water Purity 9.1-37 9.1-5 Spent Fuel Pool Cooling and Clean-Up System Failure Modes and Effects Analysis 9.1-38 9.1-6 Spent Fuel Pool Boiling 9.1-39 9.2.1-1 Essential Cooling Water System Equipment Data 9.2-36 9.2.1-2 Essential Cooling Water System, Failure Modes and Effects Analysis 9.2-38 9.2.1-3 Evaluation of Corrosive / Scale-Forming Tendencies of the Essential Cooling Pond 9.2-43 9.2.1-4 ECW Relief Valves to which ASME B&PV Code Case N-242-1 Applies 9.2-44 9.2.2-1 Component Cooling Water System Design Parameters 9.2-45 9.2.2-2 CCW Water Chemistry, Nomal Plant Operation 9.2-48 9.2.2-3 Component Cooling Water System, Failure Modes and Effects Analysis 9.2-49 9.2.2-4 Component Cooling Water Requirements 9.2-63 9.2.2-5 CCWS Relief Valves to Which ASME B&PV Code Case N-242-1 Applies 9.2-65 9.2.5-1 Maximum Heat Rejected to the Essential Cooling Water System for the Safe Shutdown Unit 9.2-67 9.2.5-2 Maximum Heat Rejected to the Essential Cooling Water System for the LOCA Unit 9.2-68 1 9.2.5-3 Maximum Consumptive Use of Water in the l Essential Cooling Pond for a Safe Shutdown in One Unit and LOCA in th". Other Unit 9.2-69 9.2.5-3.1 Maximum Consumptive Use of Water in the Essential Cooling Pond for the Safe Shutdown of Two Units 9.2-70 9.2.5-4 Combinations of Phenomena Considered in the Design of the Ultimate Heat Sink 9.2-71 9.2.5-5 Summary of ECP Performance 1 During Minimum Cooling 9.2-72 j 9.2.7-1 Reactor Makeup Water System Component Design -

Parameters 9.2-73 9.2.7.2 Reactor Makeup Water System Failure Modes l and Effects Analysis 9.2-74 l 9.3-1 Gas Vessels - Design Data 9.3-64 i

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. i 9.1 FUEL' STORAGE AND HANDLING r

i Facilities for the receipt and storage of new fuel and the storage'and transfer of spent fuel are housed in the Fuel Handlin5 Building (FHB). A >

separate and independent FHB is provided for each unit of the South Texas  !

Project Electric Generating Station (STPEGS). Each FHB is designed as a {

controlled-leakage seismic Category I structure. The design of the FHB [

Heating, Ventilating and Air-Conditioning (HVAC) System is discussed in ,

Section 9.4.2. The structural design considerations are described in Section 3.8.4.

9.1.1 New Fuel Storage 9.1.1.1 Desien Bases. The new fuel storage pit is a reinforced '

concrete pit and an integral part of each seismic Category I FHB. This pit provides temporary dry storage for approximately one-third of a core (66 fuel assemblies) of new fuel. The fuel is stored in racks (Figure 9.1.1-1) .

composed of individual vertical cells fastened together to form three 2 x 11 modules which may be bolted to anchors in the floor and walls of the new fuel  !

storage pit. The new fuel racks are classified as seismic Category I ,

components, as defined by Re5ulatory Guide (RG) 1.29, and American Nuclear Society (ANS) safety class (SC) 3 (Section 3.2).  !

The new fuel racks are designed with a center-to-center spacing of 21 inches.

This spacing provides a minimum of 12 in, between adjacent fuel ~ assemblies.

This separation is sufficient to maintain a suberitical array assuming optimum moderation. Space between storage positions is blocked to prevent insertion j of fuel. All rack surfaces that come into contact with the fuel assemblies  ;

are made of annealed authentic stainless steel, and the support structure is ,

painted carbon steel. {

The racks are designed to withstand normal operating loads, as well as to .

I remain functional with the occurrence of a Safe Shutdown Earthquake (SSE).

The new fuel racks are designed to withstand a maximum uplift force of 5,000  !

pounds and to meet the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, Appendix XVII.

The new fuel storage pit access hatch is a three-section cover. This cover will minimize the introduction of dust and debris into the pit. The cover is designed to withstand the impact force of a new fuel assembly dropped from the maximum elevation allowed by the 2-ton hoist of the FHB overhead crane.

In addition, space is provided for the storage of fuel during refueling inside the Reactor Containment Building (RCB). See Section 9.1.2.1 for a description of the racks.

9.1.1.2 Facilities Descriotion. The FHB abuts the south side of the RCB and is adjacent to the west side of the Mechanical-Electrical Auxiliaries Building (MEAB) of each unit. The locations of the two FHBs are shown in the l station plot plan on Figure 1.2-3. For general arran5ement of the new fuel storage facilities, refer to Figures 1.2-39 through 1.2 48.

New fuel assemblies are received in ths receiving area of each FHB and temporarily stored in the shipping containers in the new fuel handling area.

9.1-1 Revision 0

. STPEGS UFSAR 5

In the new fuel handling area,-each new fuel assembly is removed from its ,

shipping container and inspected visually to confirm the assembly has not been damaged during' shipment. The new fuel assemblies are transported from the ,

inspection area to the new fuel storage pit or to the new fuel elevator by the L 15/2 ton, . dual-service FHB crane. The 2-ton hoist of this crane is designed to handle new fue1' assemblies. New fuel handling is discussed in detail-in

.Section 9.1.4. Use of the 2-ton hoist of the 15/2-ton crane or of ' the fuel- ,

handling machine to handle new fuel ensures that the design uplift of the  ;

racks will not be exceeded.

The new fuel storage pit is situated in the approximate center-of each FHB.

- The floor of the new fuel storage pit is at El. 50 ft-3 inches. The new fuel storage pit access hatch is provided with a three section protective cover at El. 68 ft. The fuel assemblies are loaded into the new fuel storage racks through the top and stored vertically.  !

9.1.1.3 Safety Evaluatiou. Units 1 and 2 of the STPEGS are each provided with separate and independent fuel handling facilities.

Flood protection of each FHB is diicussed in Section 3.4.1. Flooding of the new fuel storage pit from fluid sources inside either FHB is not considered credible since all fluid systems components are located well below the elevation of the new fuel storage pit access hatch. A floor drain is provided -

in the new fuel storage pit to minimize collection of water.

The applicable design codes and the ability of the FHB to withstand various external loads and forces are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7. Missile protection of the FHBs is discussed in Section 3.5. Failure of nonseismic systems or structures will not decrease the degree of suberiticality provided in the new fuel storage pit.

In accordance with American National Standards Institute (ANSI) N18.2, the design of the normally dry new fuel storage racks is such that the effective multiplication factor will not exceed 0.98 with fuel of the highest anticipated enrichment in place, assuming optimum moderation (under dry or

~ fogged conditions). For the unborated flooded condition, assuming new fuel of the highest anticipated enrichment in place, the effective multiplication  !

factor does not exceed 0.95. Credit may be taken for the inherent neutron- <

a'usorbing effect of the materials of construction. I l

The new fuel assemblies are stored dry, the 21-in spacing ensuring a safe I geometric array. Under these conditions, a criticality accident during I refueling and storage is not considered credible. Consideration of criticali-ty safety analysis is discussed in Section 4.3.

Design of the facility in accordance with RG 1.13 ensures adequate safety under both normal and postulated accident conditions. The new fuel storage racks also meet the requirements of General Design Criterion (GDC) 62.

9.1.2 Spent Fuel Storage 9.1.2.1 Desien Bases. The spent fuel pool (SFP) is a stainless steel-lined reinforced concrete pool and is an integral part of each FHB. All spent l

9.1-2 Revision 0

o STPECS UFSAR fuel racks are classified as seismic Category I, as defined by RG 1.29, and as ANS SC 3. 3 The spent fuel storage facility provides storage capacity for 1,969 high density poison spent fuel racks in a honeycomb array in each unit. Two storage regions are provided in the SFP. There are 288 storage cellsFigure in 3 Region 1 type racks and 1,681 storage cells in Region 2Thetypesix racks.

Region 1 rack 9.1.2 2 shows the pool layout for both Units 1 and 2.

modules are located in the northwest corner of the spent fuel pool.

The Region 1 racks have 10.95-in. nominal center-to-center spacing with removable poison assemblies between the cells. This region is conservatively designed to accommodate unirradiated fuel at enrichments to 4.0 weight l3 percent. Region 1 storage cells are each bounded on four sides by a water box except on the periphery of the pool. The neutron poison material (Boraflex")

is located in these water boxes. This is accomplished by capturing two sheets of the poison material on two outside opposite faces of a thin-walled rectan-gular box. The poison sheets are captured under thin stainless steel sheets which are intermittently velded all around to the thin-walled rectangular box.

This configuration allows any irradiation gas formation in the poison to escape. A locking device engages the structure under the lead-in guide to hold the assembly in place. Special tools are provided for unlocking, removing, . reinstalling, and locking this poison assembly. The axial location of the poison with respect to the active fuel region is provided and main-tained by this welded assembly structure. Figure 9.1.2-3 shows a typical Region 1 spent fuel rack.

The reactivity characteristics of fuel assemblies which are to be placed in the spent fuel storage racks are determined and the assemblies are categorized by reactivity. Alternately, if necessary, all assemblies may be treated as if each assembly is of the highest reactivity class until the actual assembly reactivity classification is determined. Section 5.6 of the Technical Specifications provides the definitions of the reactivity classifications and the allowed storage patterns. Fuel assemblies are loaded into the racks in a geometrically safe configuration to ensure rack suberiticality.

3 Fuel assembly reactivity requirements for close packed storage and checker-board storage are specified in the Technical Specifications. The boron concentration of the water in the spent fuel pool is maintained at or above the minimum value needed to ensu'.e that the rack K,g is less than or equal to 0.95 in the event of misplaced assemblies in the close packed storage areas or in checkerboard storage areas. Consideration of criticality safety is discussed in Section 4.3.

The Region 2 racks have a 9.15-in. nominal center-to-center spacing with fixed poison material surrounding each cell. A sheet of neutron poison material is j3 captured between the side walls of all adjacent boxes. To provide space for the poison sheet between boxes, a double row of matching flat round raised areas are coined into the side walls of all boxes. The raised dimension of these locally formed areas on each box wall is half the thickness of the poison sheet. The boxes are fusion welded together at all these local areas.

The poison sheets are scalloped along their edges to clear these areas.

Figure 9.1.2-4 shows a typical Region 2 spent fuel rack.

9.1-3 Revision 3

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STPEGS UFSAR The axial location of the poison with respect to the active fuel region is provided and maintained by the structure of each box. At the outside periphery of each rack, a sheet of poison material is captured under thin stainless sheets which are intermittently welded all around to the box.

All rack surfaces that come into contact with fuel assemblies are made of annealed austenitic stainless steel. These materials are resistant to corrosion during normal and emergency water quality conditions. The racks are designed to withstand normal operating loads as well as to remain functional with the occurrence of an SSE. The racks are designed with adequate energy absorption capabilities to withstand the impact of a dropped spent fuel assembly from the maximum lift height of the spent fuel pit bridge hoist. The racks are designed to withstand a maximum uplift force equal to the uplift force of the bridge hoist. The 14-in, and 16-ir. racks also meet the require-ments of ASME Code,Section III, Appendix XVII. The high-density spent fuel racks meet the criteria of Appendix D to Standard Review Plan (SRP) 3.8.4.

Shielding for the SFP is adequate to protect plant personnel from exposure to radiation in excess of published guideline values as stated in Section 12.1.

A depth of approximately 10 f t of water over the top of the spent fuel assemblies will limit direct radiation to 2.5 mR/hr (surface dose rate).

The FHB Ventilation Exhaust System is designed to limit the offsite dose in the event of a significant release of radioactivity from the fuel, as discussed in Sections 12.3.3, 15.7.4, and 9.4.2.

The FHB is designed to prevent missiles from contacting the fuel. A more detailed discussion on missile protection is given in Section 3.5.

In addition, space is provided for storage of fuel during refueling inside the RCB for 64 fuel assemblies in four 4 x 4 modules having 16-in, center-to-center spacing (Figure 9.1.2-1A). These modules are firmly bolted in the floor.

9.1.2.2 Facilities Description. The FHB abuts the south side of the RCB and is adjacent to the west side of the MEAB of each unit. The locations of the two FHBs are shown in the station plot plan on Figure 1.2-3. For general arrangement of the spent fuel storage facilities, refer to Figures 1.2-39 through 1.2-48.

The spent fuel storage facilities are designed for the underwater storage of spent fuel assemblies and control rods after their removal from the reactor vessel. The spent fuel is transferred to the FHB and handled and stored in the spent fuel pool underwater. The fuel is stored to permit some decay, then transferred offsite. For a detailed discussion of spent fuel handling, see Section 9.1.4 The SFP is located in the northwest quadrant of each FHB. The floor of the pool is at El. 21 ft-il in., with normal water level at El. 66 ft-6 inches.

The top of a fuel assembly in a storage rack does not extend above the top of the storage rack which is El. 39 ft-10 in. maximum. The fuel assemblies are loaded into the spent fuel racks through the top and are stored vertically.

9.1-4 Revision 3

STPEGS UFSAR l 9.1.2.3 Safety Evaluation. Units 1 and 2 of the STP are each provided with separate and independent fuel handling facilities. Flood protection of each FHB is discussed in Section 3.4.1. A detailed discussion of missile protection is provided in Section 3.5.

The applicable design codes and the various external loads and forces considered in the design of the FHB are discussed in Section 3.8.4. Details of the seismic design and testing are presented in Section 3.7.

Design of this storage facility in accordance with GDC 62 and RG 1.13 ensures a safe condition under normal and postulated accident conditions. The K.,, of the spent fuel storage racks is maintained less than or equal to 0.95, even if ,

unborated water is used to fill the spent fuel storage pool, by both the designs of the fuel assemblies and the storage rack and the use of administra-tive procedures to control the placement of burned and fresh fuel.

Under accident conditions, the K.re is maintained well below 0.95 assuming 700 ppm borated water. The boron concentration of the water in the spent fuel pool is maintained at or above the minimum value needed to ensure that the rack Ne, is less than or equal to 0.95 in the event of misplaced assemblies in the close packed storage areas or in checkerboard storage areas. Con-sideration of criticality safety is discussed in Section 4.3.

The SFP is designed to maintain leaktight integrity. To ensure such integrity, the pool is lined with stainless steel plate, and plate welds are backed with channels to detect and locate leakage. Leakage entering these channels is directed to the Liquid Waste Processing System (LWPS) via the FHB sump. Should a leak be detected, either by a low-level alarm (setpoint: 6 in. below normal water level) or by the fuel pool liner channel leak detection method, the operator would initiate makeup to the spent fuel pool. Makeup capability is provided by permanently installed connections to: (1) the Demineralized Water System (DWS), (2) the Reactor Makeup Water System (RMWS),

and (3) the refueling water storage tank (RWST) in the Emergency Core Cooling System (ECCS).

? ccmp10tc lecc cf CFP cccling ic nct concidered c credibic event cince the ec=penente icvcived crc decigned tc CC 2 ccicmic Cctegcry I requiremente end eeuld Ec pcucred frc redund nt Engineered Scfety Fccturce 'Ecr} perce supplicc. Further, the cyctem: previding eccling crc redundent. Thereferc, ne cingle fcilurc uculd recult in : cc=pict: 200 Of fuel pec1 cccling. For :

mere detciled diccuccien cf CFP cccling, refer te rection 0.2.2

[Replacie with Insert]l{

9.1.3 Spent Fuel Pool Cooling and Cleanup System The Spent Fuel Pool Cooling and Cleanup System (SFPCCS) is designed to remove the decay heat generated by spent fuel assemblies stored in the SFP and/or the in-Containment storage area. A second function of the system is to maintain visual clarity and purity of the spent fuel cooling water and the refueling water.

9.1.3.1 Desion Bases. The SFPCCS design heat loads are given in Table 9.1-1. System capabilities to withstand natural phenomena and piping rupture are addressed in Chapter 3. The spent fuel pool cooling portions of the SFPCCS are designed to seismic Category I requirements, and are located in the FHB, a seismic Category I building. The spent fuel pool water purification Misc 45W5 018 002 9.1-5 Revision 3

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i INSERT #1 - UFSAR Section 9.1.2.3 (page 9.1-5) [

A complete loss of SFP water inventory is not considered a credible event since the components involved are designed to Safety Class 3, Seismic Category I requirements. System piping is arranged so that a loss of piping integrity, for - -;

all credible accidents, does not result in draining of the SFP below a minimum

• depth of 23 feet above the top of the fuel. Details of the system design are provided in Section 9.1.3.2. Loss of SFP heat removal is described in Section 9.1.3.3.

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. portions of the SFPCCS are not required for safety functions.and are not designed to seismic Category I requirements. .

9.1.3.1.1 Soent Fuel Coolina: The SFPCCS is designed to remove the  ;

amount of decay heat produced by the number of spent. fuel assemblies that are stored following refueling'. The system design incorporates two trains of  :

, equipment. Each train is capable of removing 100 percent of the normal ,

maximum design heat load and 50 percent.of the abnormal maximum design heat ,

load. The system can maintain the spent fuel cooling water temperature at or below the maximum allowable temperatures specified by Table 9.1-1. JAdd

,Ihsehpj#2[ This temperature is based on the heat exchangers (HXs) being supplied with component' cooling water (CCW) at the design flow and  ;

temperature. The flow through the spent fuel storage areas provides' sufficient mixing to maintain uniform water conditions.

If it i: ne:::::ry t: ::::c: : ::=;1:t: :::: fr:r th: ::::t:rfoga361;176 ors Offload, the system can maintain the spent fuel cooling water below the maximum allowable temperature specified by Table 9.1-1. Makeup water ,

requirements will be provided by either reactor makeup water, demineralized' ,

water, or refueling water. .The makeup flowpath from the reactor makeup water storage tank (RMWST) is seismic Category I. The flowpaths from.the demineralized water storage tank (DWST) and from the RWST are nonseismic Category I.

9.1.3.1.2 Dewaterina Protection: A depth of approximately 10 ft of water.over the top of the stored spent fuel assemblies will limit direct radiation to 2.5 mR/hr. System piping is arranged so that failure of any i pipeline cannot drain the spent fuel pool or the in-Containment temporary i storage area below a depth of approximately 23 ft of water over the top of the stored spent fuel assemblies. JAdd))nssrti)3[ Additionally, means are provided to detect component or system leakage. Refer'to Section 9.3.3 for the detailed description of leak detection via the floor drains. In addition, the water level instrumentation provides a means of leakage detection. ,

9.1.3.1.3 Water Purification: The system's demineralizers and filters are designed to provide adequate purification to permit unrestricted access for plant personnel to spent fuel storage areas and to maintain optical clarity of the spent fuel cooling water and the refueling water. The optical ,

clarity of the spent fuel pool surface is maintained by use of the system's .

skimmer pump, skimmer / strainer assemblies, and skimmer filter. The optical l clarity of the refueling cavity water is maintained by the reactor cavity filtration system. The Nuclear Steam Supply System (NSSS) vendor-recommended specifications and guidelines for the spent fuel pool water purity are provided in Table 9.1-4 and the monitoring frequency is provided in  ;

Table 9.3-3. j 9.1.3.2 System Descriotion. The SFPCCS, shown on Figures 9.1.3-1 and l 9.1.3-2 (piping and instrumentation diagrams [P& ids]), consists of two cooling trains @ith[a{d6mmonjeadey,twopurificationtrains, a surface skimmer loop l and a reactor cavity filtration system. i e

The SFPCCS removes decay heat produced by spent fuel after it is removed from the reactor. Spent fuel is removed from the reactor core during the refueling sequence and placed in the SFP, where it is stored until it is shipped offsite for reprocessing'or permanent storage. If, for some reason, it is desirable j l

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INSERT #2 - UFSAR Section 9.1.3.1.1 (page 9.1-6) j l

i Table 9.1-1 incorporates various SFP loading scenarios for 18-month fuel reload -l cycles. j INSERT #3 UFSAR Section 9.1.3.1.2 (page 9.1-6) l The return line contains an anti-siphon hole just below the low water level to [

prevent gravity drainage due to an open drain valve or due to all design basis credible breaks (non-seismic one-inch piping).

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• STPEGS UFSAR or necessary to delay the transfer of the spent fuel to the SFP, the in-Containment storage area can be used for temporary storage of up to one-third of a core. The system normally handles the heat load from one core region freshly discharged from the reactor. Heat is transferred from the SFPCCS through the HXs to the Component Cooling Water System (CCWS).

When the SFPCCS is in operation, water drawn from the SFP (and/or from the in-Containment storage area) by the SFP pumps is pumped through the tube side of the HXs, and then is returned to the spent fuel pool (and/or the in-Containment storage area). Each suction connection, which is provided with a strainer, is located at an elevation 4 ft below the normal water level (approximately 23 ft above the top of the fuel assembliec) . The return line contains an antisiphon hole near the surface of the water to prevent gravity drainage.

To maintain spent fuel cooling water purity, a bypass circuit composed of a demineralizer and a filter is connected to each cooling train. The demineral-izers are charged with either a mixed resin (cation and anion resin) or cation resin only, dependant on the type of contamination indicated by the required chemical analyses. While the heat removal operation is in process, a portion of the spent fuel cooling water is diverted upstream of each HX and passed through the purification circuit, returning downstream of the HXs. The demineralizers remove ionic corrosion impurities and fission products.

Filters are provided to remove any additional particulates and to prevent any resin fines from entering the system from the domineralizer discharge.

Transfer canal water may be circulated through the same purification circuits by removing the gate between the canal and the spent fuel pool. These purification Ir.7ps are sufficient for removing fission products and other contaminants which may be introduced into the spent fuel cooling water.

One purification loop may be isolated from the heat removal portion of the SFPCCS. By so doing, the isolated equipment may be used in conjunction with either the reactor coolant drain tank pumps or the refueling water purifica-tion pump to clean and purify the refueling water while spent fuel cooling and spent fuel cooling water cleanup operations proceed. Connections are provided such that the refueling water may be pumped from either the RWST or the refueling cavity through the demineralizer and filter, and discharged to either the refueling cavity or the RWST. Samples are periodically taken to determine the need for purification of the water as well as the purification efficiency.

To further assist in maintaining spent fuel cooling water clarity, the spent fuel pool is cleaned by a skimmer loop. Water is removed from the surfaces i via two skimiaer/ strainer assemblies located in the SFP. Water is pumped l through a filter by a skimmer pump and returned to the pool surface at a j single location remote from the skimmer / strainer assemblies. Piping for i future addition of a third skimmer / strainer for cleaning the surface of the l fuel trr.isfer canal water is also provided. l The SFP is initially filled with water having the same boron concentration as that in the RUST. Borated water may be supplied from the RWST via the SFPCCS return header, or by running a temporary line from the boric acid blending tee, located in the Chemical and Volume Control System (CVCS), directly into the pool. Domineralized water can also be added for makeup purposes (i.e., to  ;

replace evaporative losses) through a connection in the SFPCCS return header. l 9.1-7 Revision 3

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. STPEGS UFSAR t The water in the spent fuel pool may be_ separated from the water in the transfer canal by a gate. The gate is installed so that the transfer canal may be drained to allow maintenance of the fuel transfer equipment. The water in the transfer canal is first pumped, via a portable pump, into the spent fuel pool and then is transferred to the recycle holdup tanks in the Boron Recycle System (BRS) by a SFP pamp. When maintenance on the fuel transfer equipment is completed, the water is returned directly to the SFPCCS by the recycle evaporator feed pumps (BRS). A portable pump is again used to return water to the transfer canal.

When opent fuel assemblies are stored in the in-Containment storage area, either of the cooling trains may be utilized to remove the decay heat. The in-Containment storage area is sized to temporarily store one-third of a core.

The in-Containment storage area is directly connected to the refueling cavity and is filled with refueling water whenever the refueling cavity is filled.

Thus, during refueling, the in-Containment storage area is always ready for use. During refueling outages, the clarity of the water in the reactor cavity is maintained by the Reactor Cavity Filtration System. Water is removed from the reactor cavity pool through a submerged strainer located one foot above the reactor cavity floor, pumped through four cartridge-type filters, and returned to the reactor cavity pool.

9.1.3.2.1 Component Descriotion: The design codes and classifications of the components are given in Section 3.2. Equipment design parameters are given in Table 9.1-2.

Spent Ftel Pool Pumos The pumps are horizontal, centrifugal units, with all wetted surfaces being stainless steel. The pumps draw water from the spent fuel pool (and/or the in-Containment storage area) and deliver it to the HXs for cooling and to the purification trains for cleanup.

Spent Fuel Pool Skimmer Pumo This horizontal, centrifugal pump takes suction from the SFP via adjustable surface skimmer / strainer assemblies and from the fuel transfer canal and circulates the water through a filter and returns it to the SFP and the fuel transfer canal. All wetted surfaces of the pump are austenitic stainless 1 steel.

Refuelina Water Purification Pump This centrifugal pump is used to circulate water from the RWST through a SFP demineralizer and filter.

Spent Fuel Pool Heat Exchancers The HXs are the shell and U-tube type. Spent fuel cooling water circulates through the tubes while CCW circulates through the shell. Each HX is sized for 50 percent of the heat load. The decign hect lead Of the cFr Fxc ic baced en the decay heat gencrcted by ene third of 0 ccre placed ir the CFP chcrtly af ter recades-chutdecin daring a refueling Operatier .ith enc third Of c cere frer the previcuc refueling already r the pecl . (Repidceiith InseFt.1#4d Misc-95\95-0t 8.002 9.1-8 Revision 0

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INSERT #4 - UFSAR Section 9.1.3.2.1 (page 9.1-8) e The heat exchangers are designed to accommodate the Abnormal' Maximum heat ' i loads and temperatures given in Table 9.1-1.

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MISC 05W5418 002 ,

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. STPEGS UFSAR Scent Fuel Pool Demineralizers The two flushable demineralizers are designed to provide adequate spent fuel cooling water purity for unrestricted access of plant personnel to the spent

-fuel storage areas.

Soent Fuel Pool Filters [

A filter is located in each purification train, downstream of the demineralizer, to collect possible particulates and resin fines passed by the demireralizer. The filter assembly utilizes a disposable cartridge filter and is readily accessible for filter change.

foent Fuel Pool Skimmer Filter The SFP skimmer filter is used to remove particles swept from the spent fuel pool surface which are not removed by the skimmer / strainer assembly. The ,

filter assembly utilizes a disposable cartridge filter and is readily acces-sible for filter change.

Soent Fuel Pool Strainers A strainer is located in each SFP pump suction line from the SFP to prevent introduction of relatively large particles that might clog the spent fuel demineralizers or damage the SFP pumps.

Spent Fuel Pool Skimmer / Strainer Assemblies r

Two assemblies are provided. These assemblies make it possible to take .

suction from the pool surface and remove debris from the skimmer process flow.

Puel Transfer Canal Skimmer / Strainer Assembly (Future Expansion)

Piping is provided for future addition of one assembly which would take suction from the transfer canal surface. Debris would be removed via the i skimmer filter.  !

In-Containment Storace Area StraingI I

A strainer is located in the SFP pump suction line from the in-Containment storage area to prevent introduction of relatively large particles that might l clog the SFP demineralizers or damage the SFP pumps. i l

Reactor Cavity Filtration System i l

The Reactor Cavity Filtration System is a skid-mounted package system l including a horizontal, centrifugal pump with an electric motor driver, four I filter housings with cartridge-type filters, suction screen, and the necessary valves, instrumentation, and piping.

(AddLInsertl#5.)

l MISC-95i95418 002 9.1-9 Revision 0 l

INSERT #5 - Add New UFSAR Section 9.1.3.2.2 After Section 9.1.3.2.1 (page 9.1-9) 9.1.3.2.2 Snent Fuel Pool Cooline Durine Refueline Operation The normal method for refueling at STP is to perform a full-core offload.

Under the gui&mce provided in Standard Review Plan Section 9.1.3, this is the abnormal maximum condition. According to SRP Section 9.1.3, during a full-core offload, the abnormal maximum criteria applies and a single active failure need not be considered. For this condition, the SRP requires the temperature of ,

the SFP water to be kept below boiling and level maintained with the available normal systems in operation. Table 9.1-1 shows that this criteria is met for the STP Abnormal Maximum case.

Table 9.1-1 gives the SFP temperature limits for various fuel load and SFP cooling configurations. Cycle specific calculations will be performed, if necessary, to ensure that these SFP temperature limits are met. Cycle specific calculations are not required if the tytMcal STP fuel load conditions, specified in Table 9.1-1, represent the intended refueling operation.

• During full-core offload conditions at STP, two SFP cooling trains are administratively required to be available. At least one SFP cooling train will be available at all times backed by an on-site power source. The second cooling train will at least be functional, backed by either an on-site power source or a power source svailable from the switchyard.  ;

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STPEGS UFSAR 9.1.3.3 Safety Evaluation.

9.1.3.3.1 Av:.ilability and Reliability: The SFPCCS has no emergency function during an accident except to provide adegaate cooling to the SFP.

Since it is not necessary for automatic initiation post-accident, the SFP pumps are manually placed on the emergency power bus after completion of automatic load sequencing. In the event of failure of a SFP pump or loss of cooling to a SFP HX, the second cooling train would provide continued cooling of the stored spent fuel with the spent fuel cooling water at a higher equilibrium temperature (Table 9.1-1). [ Add [Inkertl#6 y A failure modes and effects analysis for the SFPCCS is given in Table 9.1-5.

9.1.3.3.2 Soent Fuel Storace Area Dewaterino: The most serious failure of this system would be complete loss of water in one of the storage areas.

To protect against that possibility, the SFP pump suction connections enter near the normal water level so that the storage areas cannot be siphoned. The eccling unter return line tc cach cterage are: centainc c anticiphen heir te prevent th peccibility cf ciphening,-[Addflnsertl#7d These design features assure that neither the SFP nor the in-Containment storage area can be drained more than 4 ft below the normal water level (normal water level is approximately 27 ft above the top of the stored spent fuel).

If a seismic event results in the failure of a non-seismic pipe in the reactor makeup water tank compartment, both reactor makeup pumps may be lost because of flooding in this compartment. As a result, makeup for the SFP from the RMWS would be lost. However, the following means may still be available to provide makeup to the SFP.

' Water from the RWST (seismic Category I) through the nonseismic Category I 1.

refueling water purification pump 1A and SFP demineralizer 1A to the SFP cooling return line.

2. Demineralized water through a 2-in. line connecting the demineralized water system to the SFP cooling return line.

3. Fire water from either one or both of the hose reel and hose cabinet located near the SFP.

In the unlikely event that all of these backup sources were not available, a seismically qualified makeup water source would be provided by connecting temporary hoses to the vent and drain valves located on the low head safety injection (LHSI) pump discharge piping. The LHSI pumps are located in the FHB at the lowest level. The hoses will be routed through building stairways and equipment hatches to the SFP. 7.ccuming c completc Iccc cf Srr cccling and a heat Iced ac deceribed ir 9.1.2.2.1 bel e t. , the '.catup time to bciling ic ,

e.pprcximately e heurc baced upcn unter lecc equir:1cnt to t ft cf pec1 levelv[ Replace withT Insert}#8J This allows sufficient time to route the hoses through the building.

9.1.3.3.3 Water Ouality: Whenever a fuel assembly with defective cladding is removed from the reactor core, a small quantity of fission products may enter the spent fuel cooling water. The purification loops provide a means of removing fission products and other contaminants from the water. By maintaining radioactivity concentrations in the spent fuel coolit.g water at 5 x 10 pCi/cc (S and y) or less, the dose at the water surface is 2.5 mR/hr or less.

Misc.95i95-ols 002 9.1-10 Revision 0

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i INSERT #6 - UFSAR Section 9.1.3.3.1 (page 9.1-10)

Table 9.1-1 shows various SFP loading scenarios for 18-month fuel reload cycles. Table 9.1-1 incorporates the routine refueling practice of full-core offload,18-month cycles.

INSERT #7 - UFSAR Section 9.1.3.3.2 (page 9.1-10)

The cooling water return line to each storage area contains an anti-siphon hole to prevent the possibility of siphoning for all design basis credible breaks (non-seismic one-inch piping).

INSERT #8 - UFSAR Section 9.1.3.3.2 (page 9.1-10)

For a complete loss of SFP cooling, the maximum heat load occurs for the STP Abnormal Maximum case listed in Table 9.1-1. This analysis assumes a SFP water level 4 feet below the normal water level, and that both SFP cooling trains simultaneously fail just when the SFP reaches its .

maximum temperature. This worst case scenario time-to-boil is calculated to be 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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9.1.3.3.4 Epent Fuel Pool Boiline Dose Analysis: In the event of a fire or moderate energy line crack in the FHB that disables both trains of SFP ,

cooling, the SFP temperature would begin to rise and, assuming no corrective action, would eventually boil. The following analysis examines the dose ,

I consequences of a loss of the SFPCCS and the use of the seismic Category 1 makeup water source (RMUS). {

It is assumed that a loss of the SFPCCS occurs after a refueling where a full core has been removed and placed into the SFP 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown. The -

heat loads supplied to the pool are comprised of the following sources: 1) the full core removed prior to the event; 2) 92 assemblies which have decayed 36 days after shutdown; and 3) spent fuel from the previous 20 refueling off-loads. The last full core offload fills the SFP to the maximum capacity of 1969 assemblies. For the purpose of this calculation, the pool is conserva- f tively assumed to boil instantaneously after the loss of the SFPCCS. This 3 [

loss of SFPCCS is assumed to occur at 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> after shutdown. Throughout  !

the event, the leakag)e rate (1.3 x 10-e sec~ .

rateiodine The for iodine is assumed available to beisthe for release based normal upon full the power ,

gap activity containing 10 percent of the rod inventory and the leakage occurs l from the defective 1 percent of the rods. The activity of the refueling water j prior to initiation of the event is assumed to be negligible.  ;

i Using these assumptions and those found in Table 9.1-6, the thyroid dose consequences of releasing the iodine as a result of SFP boiling are well below the dose requirements of 10CFR, Part 100.

9.1.3.4 Instrumentation Anolication. The instrumentation provided for the SFPCCS is discussed below. Alarms and indications are provided as noted.

9.1.3.4.1 Temperature: Instrumentation is provided to measure the l temperature of the water in the SFP and in the in-Containment storage area and  !

to give local indication as well as annunciation at the main control board j when normal temperatures are exceeded.  !

l Instrumentation is also provided to give local indication of the temperature of the spent fuel cooling water as it leaves each HX. .

9.1.3.4.2 Pressure: Instrumentation is provided to measure and give local indication of the pressures in the suction line of the SFP skimmer pump i and in the suction and discharge lines of the refueling water purification pump and of each SFP pump. Instrumentation is also provided at locations l upstream and downstream of each SFP filter, each SFP skimmer filter, and each  !

of the SFP demineralizer so that the pressure differential across these  :

filters can be determined.

9.1.3.4.3 Flow: Instrumentation is provided to measure and give local indication of the flow in the outlet line of each SFP filter. Instrumentation is also provided to measure discharge flow from the refueling water purifica-tion pump and to provide a low-flow alarm on the main control board in addition to providing local flow indication, 9.1.3.4.4 Level: Instrumentation is provided to give an alarm in the control room when the water level in the SFP or in the in-Containment storage area reaches either the high or low-level setpoints (6 in above or below normal water level).

9.1-11 Revision 3

. . 1 STPECS UFSAR i

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9.1.3.5 Tests and Insnections. Active components of the SFPCCS are in '

either continuous or intermittent use during normal system operation. The SFPCCS is included in the inservice inspection requirements described in l Section 6.6 and the inservice testing requirements described in Section 3.9.6.

9.1.4 Fuel Handling System 9.1.4.1 Desien Bases. The FHS consists of equipment and structures utilized in the transporting and handling of the fuel from the time it reaches the station until it leaves the station.

The following design bases apply to the FHS: ,

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1. Fuel-handling devices have provisions to avoid dropping or jamming of fuel assemblies during transfer operation.

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2. Fuel lifting and handling equipment will not fail in such a manner as to damage seismic Category I equipment in the event of an SSE. ,

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3. The Fuel Transfer System (FTS), where it penetrates the Containment, has provisions to preserve the integrity of the Containment pressure boundary, including a means to test for leak tightness.

4. Each machine used to lift spent fuel has a limited maximum lift height so that the minimum required depth of water shielding is maintained.

5. The cask handling components and the FHB layout limit vertical lift above the floor of the cask to less than 30 ft above the floor during any moving sequence.

6. The Spent Fuel Cask Handling System (SFCHS) utilizes a wet handling technique.

7. The pool gates are designed to maintain their integrity in the event of i an SSE.

I 9.1.4.2 System Descriotion. The equipment in the FHS is comprised of lifting equipment, handling equipment, an FTS, and the SFCHS. The structures associated with the FHS are the refueling cavity, refueling canal, and in.

Containment fuel storage area inside the RCB; and the fuel transfer canal, SFP, new fuel storage pit and inspection area, and cask loading pool and l decontamination platform in the FHB. The equipment is located in seismic Category 1 buildings.

9.1.4.2.1 New Fuel Handlinn: The new fuel arrives onsite either by rail or truck. The rail track and truck receiving areas are located on the ground floor of the FHB. When new fuel is delivered to the receiving area within the FHB, the shipping containers are unloaded from the transport vehicle and examined for shipment damage. The shipping containers are then i lifted to the operating floor by the FHB overhead crane. The shipping  !

containers can be placed directly in the new fuel inspection laydown area on i the operating floor or lowered through the equipment hatch in the operating ]

floor to the new fuel handling area below by means of the FHB overhead crane. l The shipping containers may be unloaded in either area. The shipping con- l tainers are placed horizontally on the floor. The containers are then stacked 9.1-12 Revision 0

O STPEGS UFSAR using the overhead crane servicing this area. One by one, these shipping containers are unstacked, their covers removed, and the pivotal, fuel support structure within the shipping container is elevated from the horizontal to the vertical position. In the new fuel handling area, this is accomplished using the new fuel handling area overhead crane. In the new fuel inspection laydown area on the operating floor, this is accomplished using the FHB overhead crane. The various clamping devices securing the fuel assembly to the support structure are then removed. The fuel assembly is lifted from the shipping container support structure. Inspection activities may now be conducted in either areas. Alternatively, inspection activities may be conducted at a later time following transfer of the fuel to the new fuel storage pit.

Following inspection, unacceptable new fuel assemblies are set aside for dispositioning. Acceptable new fuel assemblies are engaged by the new fuel handling tool, which is in turn attached to the hook of the FHB overhead crane. If the fuel was unloaded in the new fuel handling area, the new fuel assembly must be secured beneath the equipment hatch, released from the new fuel handling area overhead crane, and then engaged by the FHB overhead crane and lifted through the equipment hatch to the operating floor. Fuel assemblies are either inserted into the new fuel storage racks in the new fuel storage pit, placed in the new fuel elevator which is located in the fuel transfer canal, or placed directly into the SFP (Cycle 1 fuel only) . Those assemblies placed in the new fuel elevator are lowered t_o the bottom of the fuel transfer canal. They are then engaged by the spent fuel handling tool, which is in turn suspended from the fuel handling machine. The fuel handling machine either transfers the assembly to the spent fuel storage racks (before initial refueling) or to the FTS upender for transfer to the RCB for refueling operations. The upender pivots the assembly to the horizontal position and the FTS fuel container carries it through the fuel transfer tube to an upender inside the Containment.

9.1.4.2.2 Refueling Procedure: The refueling operation follows a detailed procedure to ensure a safe, efficient refueling operation. Prior to initiating refueling, shutdown conditions are as specified in the Technical Specifications. Criticality protection for the refueling operation, including a requirement for checks of boron concentration, is specified in the Technical Specifications.

Protection against uncontrolled rod cluster control assembly (RCCA) bank withdrawal from a suberitical condition is described in Section 15.4.1 and includes source-range, intermediate-range, and power range high neutron flux trips. The transient is assumed to be terminated by the power-range high neutron flux (low setting) reactor trip. Protection against uncontrolled boron dilution is described in Section 15.4.6.

The following significant points are assured by the refueling procedure:

1. The refueling water and the reactor coolant contain approximately 2,800 l ppm boron. This concentration is sufficient to keep the core approximately 5 percent ok/k suberitical during the refueling operations with all control rods removed and the core refueled to provide sufficient excess reactivity for operation to the next refueling outage.  ;

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STPECS UFSAR

2. The water level in the refueling cavity is high enough to keep the radiation levels within acceptable limits when the fuel assemblies are being removed from the core.

The refueling operation is divided into four major phases: (1) preparation, (2) reactor disassembly, (3) fuel-handling, and (4) reactor assembly. A general description of a typical refueling operation through the four phases is given below.

This description applies to rapid (unrodded) refueling which will normally be used during a refueling shutdown to maximize plant availability. The description also points out the different steps that would be included in a nonrapid (rodded) refueling operation, which is typically used for extended shutdowns involving nonroutine maintenance.

9.1.4.2.2.1 Phase I - Preparation - The reactor is shut down (rods in) and then simultaneously borated and cooled down to refueling shutdown condition. Following initiation of normal purge and a radiation survey of the Containment Building, refueling operations may proceed. The fuel transfer equipment and refueling machines are checked for proper operation. Each new fuel assembly is brought from dry storage in the FHB as described in Section 9.1.4.2.1. After transfer through the fuel transfer tube, the FTS fuel assembly container is pivoted to the vertical position by the in-Containment upender. The refueling machine transfers the new fuel into the in-Containment storage racks (shown in Figure 9.1.2.1.a). Refer to Figures 1,2-14 through 1.2-18 for general arrangement of the in-Containment fuel storage area.

When the Reactor Coolant System (RCS) has been cooled to 150*F, RCS draining is started. The RCC assemblies (control rods) are withdrawn to their full-out position, and each control rod's holdout device is activated to ensure that the rod is held in its withdrawn position inside its upper internals guide tube and reactor head pressure housing. As soon as RCS draining has lowered reactor coolant level to the reactor vessel nozzle centerline, degassing operations using the RCS Vacuum Degassing System (Setion 11.3) may be used, if necessary, to remove radioactive gases prior to head removal. ,

l For nonrapid refueling, the preparation for refueling is similar except that the control rods are not withdrawn from the core and the RCS temperature is reduced to 140*F before draining.

9.1.6.2.2.2 Phase II - Reactor Disassemb1v - The seismic tie rods

} attached to the missile shield are disconnected and stored. The insulation is removed from the vessel head flange area, and the Roto-Lok studs are detensioned and removed from the vessel flange. A stud hole plug is installed in each hole after the stud is removed to prevent entry of water. In addition, all flux mapping detectors and thimbles are retracted through the bottom of the reactor vessel. The refueling cavity is prepared for flooding by removing all tools, closing the refueling canal drain holes and by installing underwater lights. The reactor cavity is then flooded to 12 in.

below the top of the head flange. The upper head package (i.e., head, missile, cable bridge, upper internals, control rods, and rod drives) is lifted by the polar crane until the closure head guide pins are clear. Water i from the RWST is pumped into the RCS by the LHSI pumps, causing the water to I overflow into the refueling cavity. The vessel head is lifted in conjunction l with the water level in the refueling 9.1-14 Revision 4 l

. l STPEGS UFSAR j

. 1 cavity. When the refueling cavity is full, the upper package is moved to storage at the end of the refueling cavity opposite the refueling canal. If a radiation survey indicates the need, additional water shielding may be provided by pulling a vacuum through the reactor vessel head vent connection.

For nonrapid refueling, reactor disassembly includes additional steps starting with disconnecting the cables that run across the cable bridge. This allows the bridge to be lif ted by the polar crane separately from the reactor head. j The control rod drive shafts are disconnected leaving the rods in the core. 1 The upper internals assembly is then disconnected from the reactor vessel l head. After the cable bridge is lifted clear of the reactor head, the head is I lifted about 24 in. to permit a visual inspection of the RCC assembly drive shafts. This ensures that they are free in the control rod drive mechanism (CRDM) housings and were not raised with the reactor head. The refueling cavity is then flooded to a level just below the closure head. Simultaneous lifting of the reactor head and filling of the refueling cavity proceeds in the same manner as described for rapid refueling.

The vessel head is placed on a dry storage pedestal in a roped-off area of the ,

operating floor at the north end of the containment. RCC drive shafts are '

unlatched from their respective RCC assemblies. Finally, the upper internals are removed from the vessel by using the reactor internals lifting device suspended from the polar crane. The internals package is wet-stored on a stand in the north end of the refueling cavity.

9.1.4.2.2.3 Phase III - Fuel Handline - Fuel assemblies are removed from and inserted in to the reactor core by the refueling machine. The spent fuel assemblies are removed from the core in a sequence which is planned before each refueling. The positions of partially spent fuel assemblies are shuffled, and new fuel assemblies are added to the core.

The general fuel handling sequence is:

1. The refueling machine is positioned over a spent fuel assembly in the most depleted region of the core.

2. The spent fuel assembly is lifted by the refueling machine to a predetermined height sufficient to clear the reactor vessel and still leave sufficient water covering the spent fuel assembly to eliminate any radiation hazard to the operating personnel.

3. The fuel transfer car is moved into the refueling canal from the fuel transfer canal.

4 The FTS fuel assembly container is pivoted to the vertical position by the upender.

5. The refueling machine is moved from over the core to line up the spent fuel assembly with the fuel assembly container.

6. The refueling machine loads the spent fuel assembly into the fuel assembly container of the FTS transfer car.

7. The container is pivoted to the horizontal position by the upender.

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9. The fuel assembly container is pivoted to the vertical position by the .

upender. The spent fuel assembly is unloaded by the spent fuel handling l tool, which is suspended from the fuel handling machine hoist.  !

10. The spent fuel assembly is placed in the spent fuel storage rack after i

being transferred through the gate between the fuel transfer canal and the SFP. }

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11. ' Partially spent fuel assemblies are moved fitli@63biif[SFPfi{tt%FHB f 6E]to new positions in the reactor core, and new fuel assemblies are moked from the in-Containment storage racks to the core to replace the

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I' l spent fuel assemblies that were removed during the preceding fuel i

-handling steps.

12. This procedure is continued until refueling is. completed. I During nonrapid refueling, some of the spent fuel assemblies that are removed :j from the reactor core will contain a rod cluster control (RCC) (control rod) i element. Such assemblies are placed in the RCC change fixture by the ,

refueling machine. The RCC change fixture is located adjacent to the in-Containment storage racks. Here the RCC element is removed from the spent fuel assembly and deposited in a partially spent or new fuel assembly.  !

previously placed in the RCC change fixture. [

f Another step generally performed during either rapid or nonrapid refueling is [

the removal of an irradiated specimen from the reactor core for examination. l Also, remote boroscope and television camera inspections of the core and _l reactor vessel are performed using equipment suspended from'the refueling  !

machine.

6 9.1.4.2.2.4 Phase IV - Reactor Assembly - Reactor assembly, following l refueling, is essentially achieved by reversing the operations given in " Phase l II - Reactor Disassembly".

9.1.4.2.3 Soent Fuel Shioment: Spent fuel is shipped offsite either by  !

rail or truck in heavily shielded spent fuel shipping casks licensed for use by the Department of Transportation (DOT). General arrangements of the cask i handling area are provided in Section 1.2. l Upon receipt in the FHB, the spent fuel cask shipping vehicle (rail or truck)  ;

is braked and blocked in position for removal of the spent fuel cask. The l cask is then inspected for shipment damage and to ascertain the degree of l additional cleaning that will be required to remove road dirt or radioactive i contamination from the cask surface. The gate between the cask loading pool  :

and the decontamination area is removed using the 15/2-ton FHB overhead crane. l The upper and lower spent fuel cask impact structures are removed, and spent . )

fuel cask appurtenances are-disconnected. The spent fuel cask yoke is removed i from its storage area and attached to the spent fuel cask trunnions using the l 150-ton, overhead cask-handling crane. As the spent fuel cask is upended, the yoke is maintained in a vertical position by lateral movement of the overhead spent fuel cask handling crane trolley. (The reverse procedure is performed when the cask is loaded onto the shipping vehicle.) The spent fuel cask is then lifted clear of the shipping vehicle, moved over to the access bay, and lowered to the decontamination area, where any additional cleaning of road dirt or surface contamination is performed.

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STPEGS UFSAR Using the 150 ton, overhead cask handling crane, which is still connected to the cask, the spent fuel cask is lifted, moved horizontally, and lowered into the cask loading pool. The lifting yoke is removed and placed in its storage location. The spent fuel cask head is unbolted and moved, using the FHB overhead crane, to a temporary storage location. The gate between the cask loading pool and the decontamination area is replaced. The cask loading pool

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and channel are filled with borated water to the same level as the SFP (nominally El. 66 ft-6 in.). The two gates between the cask loading pool and the SFP are removed, by the 15/2-ton (15-ton main hook and a 2-ton auxiliary hook) FHB overhead crane, and fuel transfer is initiated. Loading of the spent fuel assemblies is accomplished in the following manner: the spent fuel assembly in the SFP is engaged by the long spent fuel handling tool, which is in turn attached to the fuel-handling machine; the fuel assembly is removed from the storage rack and transferred through the gate area into the cask loading pool and then lowered into the spent fuel cask. The height to which the spent fuel assembly can be lifted is restricted by the length of the spent fuel handling tool and the fuel-handling machine design in order to provide a minimum of 10 ft of shielding water above the spent fuel assembly. This spent fuel handling procedure is repeated until the spent fuel cask is full. The gates sealing the cask loading pool from the SFP are then replaced using the 15/2-ton FHB overhead crane.

The spent fuel cask head is lifted from its storage area using the 15/2-ton overhead FHB crane and placed on the spent fuel cask. The cask head bolts that secure the cask head to the ensk are installed and tightened " finger tight" using long-handled tooling operated from the fuel-handling machine.

The cask loading pool level is lowered to at least 6 in below the bottom of the gate between the cask loading pool and the decontamination area. This gate is removed using the 15/2-ton overhead crane. Using the 150-ton cask-handling crane, the cask-lifting yoke is attached to the cask-lifting trunnions and the cask is transferred from the cask loading pool to the cask decontamination area. A radiological survey is made of the cask to determine the extent of cask decontamination necessary. The head bolts are torqued to the proper value, and various preshipment testing requirements characteristic of the spent fuel cask being used are completed. These tests may include a cask leak test and monitoring of cask coolant activity. The contaminated outside surface areas are manually decontaminated using high-pressure sprayers or by using scrub brushes with detergent, rinses, and wipes. Once the cask is decontaminated and has passed the preshipment testing requirements, it is lifted back through the access bay and returned to its horizontal position on the shipping vehicle. The cask holddown mechanisms are secured, the cask top and bottom impact limiters are replaced, and all appropriate appurtenances are reconnected. Shipping papers are completed and the cask is rencved from the FHB and released for shipment.

9.1.4.2.4 Coroonent Descriotion:

9.1.4.2.4.1 Refueline Machine The refueling machine (Figure 9.1.4-1) is a rectilinear bridge and trolley crane with a vertical mast extending down into the refueling cavity. The bridge spans the refueling cavity and runs on rails set into the edge of the refueling cavity. The bridge and trolley motions are used to position the vertical mast over a fuel assembly in the )

core. A long tube, with a pneumatic gripper on the end, is lowered down out of the mast to grip a fuel assembly. The gripper tube is long enough so that 9.1-17 Revision 0 i

STPECS UFSAR the upper end is still contained _ in the mast when the gripper end contacts the fuel. A winch mounted on the trolley raises the gripper tube and fuel cssembly up.into the mast tube. The fuel is transported while inside the mast tube to its new position.

The refueling machine gripper tube also includes a secondary gripper mechanism (located above the primary fuel assembly gripper) to remove and insert a thimble plug into a new fuel assembly.

The refueling machine drive uses a direct current control device that gives stepless, variable speeds from zero to ful1 speed.

All controls for the refueling machine are mounted in a console on the trolley. The brid5e and trolley are positioned by a servo system in relation to an X-Y coordinate grid pattern referenced to the reac;or core. Bridge and trolley position is indicated by an electric position-repeat back system.

Readout dials are read directly by the operator at the console. The drives for the bridge, trolley, and winch are variable speed and include a separate inching control on the bridge and trolley. The maximum speed is 60 ft/ min for the bridge and 20 ft/ min for the trolley and hoist. An auxiliary monorail hoist on the refueling machine uses a two-step magnetic controller to give hoisting speeds of approximately 7 ft/ min and 20 ft/ min for use in handling cecessory equipment.

Electrical interlocks and limit switches on the bridge and trolley drives prevent damage to the fuel assemblies. The winch is also provided with redundant limit switches plus a mechanical stop to prevent a fuel assembly from being raised above a safe shielding depth should the limit switch fail.

In an emergency, the bridge, trolley, and vinch can be operated manually using a handwheel on the motor shaft.

The refueling machine is provided with a television system which permits viewing of all fuel assembly positions and fuel movements to within 6 in. of the top of the core. This system includes a telescoping boom, a monitor, and a videotape recorder.

The refueling machine is designed to permit failed fuel detection by means of a sip test. A fuel assembly will be withdrawn inside the mast and held stationary above the core for a period of time. Doors at the bottom of the cast are shut to enclose the fuel assembly completely. The water surrounding the fuel assembly will then be circulated through detection equipment located en the crane trolley. Any activity increase will be evaluated to determine the severity of cladding failure.

9.1.4.2.4.2 Fuel Handline Machine - The fuel handling machine consists cf an electric monorail hoist carried on a wheel mounted bridge (Figure 9.1.4 2), which spans the SFP, fuel transfer canal, and cask loading pool.

The fuel handling machine is used exclusively for handling fuel assemblies and core components by means of handling tools suspended from the hoist. The hoist travel and tool lengths are designed to limit the maximum lift of a fuel cssembly or core component to a safe shielding depth.

The fuel handling machine has a two step magnetic controller for the bridge cnd hoist. The bridge speeds are 10 ft/ min and 30 ft/ min, and the hoist speeds are 7 ft/ min and 20 ft/ min. A hydraulic coupling is used in the bridge 9.1-18 Revision 0

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l STFEGS UFSAR drive to limit starting acceleration. The hoist trolley is manually positioned along the monorail by a chainfall.

9.1.4.2.4.3 New Fuel Elevator - The new fuel elevator (Figure 9.1.4-3) consists of a box-shaped elevator assembly with its top end open and sized to house one fuel assembly.

The new fuel elevator is normally used to lower a new fuel assembly into the g l fuel transfer canal, where the fuel handling machine can transport it to the FTS equipment for transfer into the Containment. The new fuel elevator can also be used to raise a new or spent fuel assembly provided administrative controls and procedures are utilized.

9.1.4.2.4.4 fuel Transfer System - The FIS (Figure 9.1.4-4) includes an ,

underwater, electric-motot-driven transfer car that runs on tracks extending from the refueling canal in the RCB through the fuel transfer tube and into the fuel transfer canal in the FHB, and a hydraulically actuated lifting arm (upender) at each end of the transfer tube. In the refueling canal the fuel container mounted on the transfer car receives a fuel assembly in the vertical position from the refvsling machine. The upender then lowers the fuel assembly to a horizontal position for passage through the transfer tube.

After passing through the tube, the fuel container is raised to a vertical position by the other upender for removal of the fuel assembly. The fuel handling machine lifts the fuel assembly out of the fuel container and moves it to the desired storage position in the spent fuel storage racks.

The transfer car is driven by a pusher arm connected to two continuous roller chains. The roller chains are driven by an electric motor mounted near the operating floor of the FHB next to the fuel transfer canal. They are connected to the chain-drive sprockets by a vertical drive shaft. The fuel container is center-pivoted, and the pivot structure serves to attach the container to the transfer car so they move together as a single unit. There is a mechanical stop at each end of the rail on which the car travels. Also, there is another mechanical stop for the vertical limit of travel when the fuel container is upended. Each of the upender lifting arms has its own hydraulic power unit end control console located on the operating floor of its i respective side of the Containment wall.

During reactor operation, the transfer car is stored in the fuel transfer  ;

canal. A blind flange is bolted on the refueling canal end of the transfer l tube to seal the Reactor Containment. The end of the tube in the FHB is l closed by a gate valve.

9.1.4.2.4.5 Rod Cluster Control Change Fixture - The RCC change fixture '

is located in the in-Containment fuel storage pit. The change fixture is used )

for periodic RCC element inspections and for transfer of the RCC elements from '

one fuel assembly to another, or transfer of secondary source assemblies from l one fuel assembly to another (Figure 9.1.4-5). During rapid refueling I operations, the RCC change fixture is not required for transfer of RCC I elements since the RCC elements are removed from the reactor core along with the upper head package. The major subassemblies which constitute the change fixture are the frame and track structure, the carriage, the guide tube, the gripper, and the drive mechanism. The carriage is a moveable container supported by the frame and track structure. The tracks provide a guide for the four flanged carriage wheels and allow horizontal movement of the carriage during the changing operation. The positioning stops on both the carriage and framt locate each of the three carriage compartments directly below the guide 9.1-19 Revision 4 l

STPECS UFSAR tube. Two of these compartments are designed to hold individual fuel assemblies while the third is made to support a single core component (e.g.,

RCCA or secondary source assembly). Situated above the carriage and mounted on the in-Containment fuel storage pit wall is the guide tube. The guide tube provides proper orientation of the gripper and core component as they are being raised and lowered. The gripper is a pneumatically actuated mechanism which engages the core component. It has two flexure fingers which can be inserted into the top of the core component when air pressure is applied to the gripper piston. Normally, the fingers are locked in a radially extended position. Mounted on the operating deck is the drive mechanism assembly, which is composed of the manual carriage drive mechanism, the revolving stop operating handle, the pneumatic selector valve for actuating the gripper piston, and the electric hoist for elevating the gripper.

The pneumatic gripper and winch lift the core component out of a spent fuel assembly and up into the guide tube. Then by repositioning the carriage, a new fuel assembly is brought under the guide tube and the gripper lowers and releases the core component ints the new fuel assembly. The refueling machine loads and unloads the fuel assrmblies into and from the carriage.

9.1.4.2.4.6 Spent _Eael Assembly Handling Tool - The spent fuel assembly handling tool (Figure 9.1.4-6) is used to handle new and spent fuel assemblies in the spent fuel pit. It is a manually actuated tool, suspended from the fuel handling machine. The gripping portion of the tool consists of a mechanical finger which engages the fuel assembly top nozzle when actuated.

The operating handle to actuate the fingers is located at the top of the tool.

When the fingers are latched, a pin is inserted into the operating handle, to prevent the fingers from being accidentally unlatched during fuel handling operations.

9.1.4.2.4.7 New Fuel Assembly Handline Tool - The new fuel assembly handling tool (Figure 9.1.4-7) is used to lift and transfer fuel assemblies from the new fuel shipping containers to the new fuel storage racks, to transfer fuel assemblies from the new fuel storage racks t o the new fuel elevator, or to directly place new fuel in the SFP (Cycle 1 only). It is a manually actuated tool, suspended from the FHB overhead crane, which uses four cam-actuated latching fingers to grip the underside of the fuel assembly top nozzle. The operating handles which actuate the fingers are located on the s;de of the tool. When the fingers are latched, a safety screw is turned in to prevent the accidental unlatching of the fingers.

9.1.4.2.4.8 Reactor Vessel Head and Uoner Internals Lifting Device -

The reactor vessel head and upper internals lifting device consists of a welded and bolted structural steel frame with suitable rigging to enable the crane operator to lift the head or the head and upper internals for storage during refueling operations. The lifting device normally remains attached to the reactor vessel head during plant operation. The missile shield and the control rod drive mechanism cooling shroud are attached to the head lifting device.

9.1.4.2.4.9 Reactor Internals Lifting Device - For rapid refueling, the upper internals are normally lifted in one lift as part of the upper head package. For nonrapid refueling, the head may be separated from the upper internals and lifted separately. In this case, the reactor internals lifting rig (Figure 9.1.4-8) is used to remove the upper internals.

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l STPEGS UFSAR 9.1.4.2.4.10 Reactor Vessel Stud Tensioner - The stud tensioners (Figure 9.1.4-9) are employed to secure or release the head closure joint at every refueling. The stud tensioner is a hydraulically operated device that uses oil as the working fluid. The device permits preloading and unloading of the reactor vessel closure studs at cold shutdown conditions. Stud tensioners minimize the time required for the tensioning or unloading operation. Three tensioners are provided and are applied simultaneously to three studs located 120 degrees apart. A single hydraulic pumping unit operates the tensioners, which are hydraulically connected in series. The studs are tensioned to their operational load in two steps to prevent high stresses on the reactor vessel flange region and unequal loading of the studs. Relief valves on each tensioner prevent overtensioning of the studs due to excessive pressure.

9.1.4.2.4.11 Soent Fuel Cask - The design of cask storage and cask  !

handling facilities is based on a design cask weighing up to 150 tons and measuring approximately 21 ft long by 10 ft in diameter. The shipping vehicle (rail car or truck) will take the cask to and from the FHB and will be equipped with a cask-cooling system and a storage area for the cask yoke. The FHB arrangement and cask-handling equipment are designed to preclude the occurrence of any accident to a loaded spent fuel shipping cask beyond the regulatory-specified design accident conditions for the cask. Overland offsite transportation of the cask will conform to transportation rules and regulations, 49CFR173.

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9.1.4.2.4.12 Fuel Handline Buildine Overhead Crane - This 15/2-ton overhead crane runs over the entire FHB area. Travel of this crane is shown on Figure 9.1.4-13. The design services of this crane include the following:

1. Transfer of new fuel assembly shipping containers from the shipping vehicle (truck or rail car) to the new fuel handling area.

2. Transfer of new fuel assemblies from new fuel shipping containers to new fuel elevator, new fuel storage pit, or SFP (cycle 1 only).

3. Transfer of spent fuel shipping cask head from the cask to temporary storage on the FHB operating floor and then back onto the shipping cask when the fuel loading is complete.

4 Replacement of safety injection and Containment spray pumps and SFP i cooling HXs.

5. Removal and replacement of pool gates for fuel transfer operations.

9.1.4.2.4.13 New Fuel Handline Area Overhead Crane This 5-tot.

overhead crane is used exclusively to handle new fuel assemblies and their .

shipping containers in the new fuel handling area.

9.1.4.2.5 Industrial Codes and Standards: The following industrial ,

codes and standards are used in the design of fuel handling equipnent.

1. Cranes: Crane Manufacturers Association of America (CMAA) specification No. 70. Class A-1,

2. Structural: ASME Code,Section III, Appendix XVII.

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3. Electrical: Applicable standards and requirements of the National Electric Code, National Fire Protection Association No. 70, and National Electrical Manufacturers Association standards MGI and ICS are used in-the design, installation, and manufacturing of all electrical equipment.

4 Materials: Materials conformed to the specifications of the American ,

Society for Testing Materials standard.  !

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'5. Safety: The design meets the applicable requirements of Section i 1910.179 of Subpart N of the occupational Safety and Health Act Code.  !

6. Others: American Institute of Steel Construction; American National l Standards Institute; American Society of Testing Materials; Institute of l Electrical & Electronic Engineers; National Electric Manufacturers Association; Occupational Safety & Health Administration; American Velding Society; Expansion Joint Manufacturers Association; ASME B&PV Code Sections VIII and XI; American Concrete Institute; Hydraulic i Institute Standards, j 9.1.4.3 Safety Evaluation. Design of the FHS in accordance with RG l 1.13 and with GDCs 2, 5, 61, and 62 ensures a safe condition under normal and l l postulated accident conditions.  !

9.1.4.3.1 Safe Handline:

9.1.4.3.1.1 Desien Criteria for the Refueline Machine and the Fuel- l Handline Machine - The primary design requirement of the machine is reliability. A conservative design approach is used for all load-bearing parts. Where practicable, components are used that have a proven record of reliable service. Throughout the design, consideration is given to the fact that the machine will spend long idle periods stored in an atmosphere of 80*F and high humidity. In general, the crane structure is considered in the Class A1, Standby Service, as defined by CMAA Specification No. 70.

All components critical to the operation of the machine and pqrts which could fall into the reactor are positively restrained from loosening.

9.1.4.3.1.2 Refueline Machine - The refueling machine design includes the following provisions to ensure safe handling of fuel assemblies:

1. Electrical Interlocks The electrical interlocks which ensure safe operation of the crane are designed to meet single-failure criteria.

2. Bridge, Trolley, and Hoist Drive Mutual Interlocks Bridge, trolley, and winch drives are mutually interlocked, using redundant interlocks to prevent simultaneous operation of any two drives, and therefore can withstand a single failure.

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3. Bridge Trolley Drive Cripper Tube Up Bridge and trolley drive operation is prevented except when'the " gripper tube up" position switches are actuated. The interlock is redundant and

.can withstand a single failure.

4. Cripper Interlock An interlock is supplied which prevents the opening of a solenoid valve in the air line to the gripper except when there is less than 600 pounds of suspended weight indicated on a load cell. As backup protection for ,

this interlock, the mechanical weight actuated lock in the gripper i prevents opening of the gripper under load even if air pressure is  !

applied to the operating cylinder. This interlock is redundant and can- .

withstand a single failure.

5. Excessive Suspended Ueight An excessive suspended weight load cell limit switch and a backup deflection-actuated limit switch prohibit raising the guide tube if the load suspended from it significantly exceeds the weight of the grf pper, a fuel assembly, and an RCCA. The interlock is intended to prevent inadvertent damage to a fuel assembly or adjacent components if the assembly becomes stuck during its removal. ,

6. Hoist-Gripper Position Interlock  ;

An interlock in the hoist drive circuit in the up direction permits the hoist to be operated only when either the open or closed indicating switch on the gripper is actuated. The hoist-gripper position interlock consists of two separate circuits that work in parallel such that one circuit must be closed for the hoist to operate. If one or both ,

interlocking circuits fail in the closed position, an audible and visual  ;

alarm on the console is actuated. The interlock, therefore, is not r redundant but can withstand a single failure since both an interlocking circuit and the monitoring circuit must fail to cause a hazardous condition.

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7. Bridge and Trolley Hold Down Devices  ;

Both the refueling machine bridge and trolley are horizontally j restrained on the rails by two pairs of guide rollers, one pair at each

• wheel location on one truck only. The rollers are attached to the bridge truck and contact the vertical faces on either side of the rail j to prevent horizontal movement. Vertical restraint is accomplished by .

antirotation bars located at each of the four wheels for both the bridge and trolley. The antirotation bars are bolted to the trucks and extended under the rail flange for the bridge restraint; the trolley ,

restraints extend beneath the top flange of the bridge girder which supports the trolley rail. Both horizontal and vertical restraints are adequately designed to withstand the forces and overturning moments  ;

resulting from an SSE. j J

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STPEGS UFSAR

8. Design Load The design load for structural components is their deadweight plus 5,700 pounds (three times the weight of a fuel assembly and RCCA).

9. Main Hoist Braking System The main hoists are equipped with two independent braking systems. A solenoid-release, spring set electric holding brake is mounted on the motor shaft. The brake operates normally to release upon application of current to the motor and to set when current is interrupted. The second brake is a mechanically actuated load control brake internal to the hoist gear box; it sets if the load starts to overhaul the hoist. It is necessary to apply torque from the motor to raise or lower the load. In raising the motor cams, the brake opens; in lowering, the motor slips the brake, allowing the load to lower. The brake actuates upon loss of torque from the motor for any reason and is not dependent upon any electrical circuits. On the main hoist, the motor brake is rated at 350 percent operating load and the mechanical brake at 300 percent.

The Main Hoist Braking System is supplied with redundant paths of load support such that failure of any one component will not result in free fall of the fuel assembly. Two wire ropes are anchored to the vinch drum and carried over independent sheaves to a load equalizing acchanism on the top of the gripper tube. In addition, supports for the sheaves and equalizing mechanism are backed up by passive restraints to pick up the load in the event of failure of this primary support. Each cable system is designed to support 13,750 pounds or 27,500 pounds acting together.

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10. Hoist Down Limit I A geared limit switch on the main hoist prevents lowering the gripper tube significantly below the position that would normally engage fuel assembly in the reactor core.

The working load of fuel assembly, RCCA, and gripper is approximately 2,850 pounds.

The gripper itself has four fingers gripping the fuel, any two of which will support the fuel assembly weight.

The gripper mechanism contains a spring-actuated mechanical lock which prevente ti,e gripper from opening unless the gripper is under a compressive load. '

The refueling machine gripper and hoist system are routinely load tested prior to refueling operations in accordance with the surveillance requireuents of the Technical Specifications.

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c STPEGS UFSAR 9.1.4.3.1.3 Fuel-Handline Machine - The fuel-handling machine includes the following safety features:

1. The bridge and hoist controls are interlocked to prevent simultaneous operation of both the bridge drive and the hoist. The interlocks _are redundant and can withstand a single failure.

2. A redunv. ant overload protection device is included on the hoist to limit the uplift force which could be applied to the spent fuel storage racks.

The protection device limits the hoist load to 125 percent (5,000 pounds) of the rated 2 ton hoist capacity. This device can withstand a single failure. A load monitoring device is provided between the hoist and spent fuel handling tool. By monitoring drag load when raising or lowering assemblies, it can be determined immediately if the assembly is hanging or experiencing unusual loads.

3. The design load on the hoist is the weight of one fuel assembly and RCCA (1,900 pounds), one failed fuel container (1,700 pounds), and the tool (400 pounds), which gives it a total weight of approximately 4,000 pounds.

4 Restraining bars are provided on each truck to prevent the bridge from overturning.

9.1.4.3.1.4 Fuel Transfer System - The following safety features are provided for in the FTS:

1. Transfer Car Permissive Switch The transfer car controls are located in the spent fuel pit area; therefore, conditions in the Containment are not visible to the operator. The transfer car permissive switch allows a second operator in the Containment to exercise some control over car movement if visible conditions warrant it.

Transfer car operation is possible only when both upenders are in the down position as indicated by the limit switches. The permissive switch is a backup for the-transfer car upender interlock. Assuming the fuel container is in the upright position in the Containment and the upender interlock circuit fails in its permissive condition, the operator in the spent fuel pit area still cannot operate the car because_of the transfer car permissive switch interlock. The interlock, therefore, can withstand a single failure.

2. Lifting Arm Transfer Car Position Two redundant interlocks allow lifting arm operation only when the transfer car is at the respective end of its rravel and therefore can 7

withstand a single failure.

Of the two redundant interlocks that allow lifting arm operation only when the transfer car is at the end of its travel, one interlock is a position limit switch in the control circuit. The backup interlock is a mechanical latch device-on the lif ting arm that is opened by the car moving into position.

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3. Transfer Car Valve Open Two redundant interlocks on the transfer tube valve permit operation of the transfer car only when the transfer tube valve position switch indicates the valve is fully open and, therefore, can withstand a single 1

failure. 1

4. Transfer Car Upender The transfer car upender interlock is primarily designed to protect the equipment from overload and possible damage if an attempt is made to move the car when the fuel container is in the vertical position. This ,

interlock is redundant and can withstand a single failure. The basic l interlock is a position limit switch in the control circuit. The backup interlock is a mechanical latch device that is opened by the weight of the fuel container when in the horizontal position.

5. Upender Refueling Machine The refueling canal upender is interlocked with the refueling machine.

Whenever the transfer car is located in the refueling canal, the upender cannot be operated unless the refueling machine mast is in the fully retracted position or the machine is over the core.

9.1.4.3.1.5 Fuel Handline Tools and Eculoment - All fuel handling tools and equipment which are used over the open reactor vessel are designed to prevent inadvertent decoupling from crane hooks; i.e., lifting rigs are pinned to the crane hook and safety latches are provided on books supporting tools.

Tools required for handling internal reactor components are designed with fail-safe features that prevent disengagement of the component in the event of operating mechanism malfunction. These safety features apply to the following tools:

1. Control Rod Drive Shaft Unlatching Tool The air cylinders actuating the gripper mechanism are equipped with backup springs which close the gripper in the event of loss of air to the cylinder. Air valves are equipped with safety locking rings to prevent inadvertent actuation.

2. Spent Fuel Handling Tool When the fingers are latched, a pin is inserted into the operating handle to prevent inadvertent actuation. The tool weighs approximately 400 pounds and is preoperationally tested at 125 percent of the weight of one fuel assembly and RCCA (1,900 pounds).

3. New Fuel Assembly Handling Tool When the fingers are latched, a safety screw is inserted to prevent inadvertent actuation. The tool weighs approximately 100 pounds and is preoperationally tested at 125 percent of the weight of one fuel assembly and RCCA (1,900 pounds).

9.1-26 Revision 0 A4 MiM ? M4k

r STPEGS UFSAR d

9.1.4.3.1,6 Overhead Cranes - Overhead cranes used in refueling and fuel handling operations include the polar crane (417/15-ton, Unit 1 and 500/15 ton, Unit 2), the 150-ton cask-handling crane, the 15/2-ton FHB crane, and the 5-ton new fuel handling area crane. These cranes are classified as non-nuclear safety (NNS) Class since they neither provide nor support any safety system function. However, during and after a seismic event, the cranes and their supports are designed to retain structural integrity and prevent collapse and damage to safety related equipment and structures. Operability need not be retained.

A report under separate cover has been submitted to the Nuclear Regulatory Commission (NRC) concerning control of heavy loads. This report contains details of crane / load combinations and the safeguards that prevent damage to spent fuel, the reactor core, equipment required for safe shutdown, and decay-heat removal.

1. Folar Crane The polar crane is used for general handling operations in the containment during refueling. These operations include:

a. Removal of the upper package

b. Removal of pumps, pump motors, and heat exchangers

c. Handling of pool gate

d. Handling of inservice inspection (ISI) rig

e. Movement of hatch covers.

A head drop analysis is discussed in Letter NS-CE-1101 (June 11, 1976) ,

and received NRC approval by letter on November 30, 1976. This crane is provided with seismic restraints to prevent derailment in the event of an SSE.

2. Cask-Handling Overhead Crane  :

This 150. ton crane is provided for handling the spent fuel shipping cask. Crane design and building arrangement preclude travel of this crane over the SFP; consequently, the shipping cask cannot be lifted or dropped over the spent fuel racks. This crane is designed to maintain ,

its structural integrity and hold its load under the dynamic loading  !

conditions of the SSE. Building arrangement and lifting rig design i prevents this crane from lifting the cash higher than 30 ft above the i floor. The spent fuel cask drop accident is discussed in Section -

15.7.5.

l 9.1-27 Revision 0 m _ - - . _ _ . _ --- _ __ _ _ _ _ - - --

f^ l

. 1 I

STPEGS UFSAR l

3. Fuel-Handling Building Overhead Crane The 15/2-ton capacity crane is to be used for general handling operations in the FHB. These operations include:

a. Movement of new fuel assemblies

b. Removal of pumps and heat exchangers

c. Handling of the pool gates

d. Movement of the cask head

e. Movement of hatch covers This crane is designed to maintain its structural integrity under the dynamic loading of the SSE. The crane will retain its load under such dynamic loadings. ,

This crane main hoist is also provided with a redundant reeving system.

With this redundancy, the crane can withstand a single failure without drcpping its load cad therefore meets the intent of RC 1.104 A more detailed description of compliance of the 15/2-ton FHB crane with RG 1.104 is given in Table 9.1-3.

4 New Fuel Handling Area Overhead Crane The 5 ton new fuel handling area overhead crane is used for movement of new fuel assemblies within the new fuel handling area. Dropping of new fuel assemblies due to SSE induced dynamic loading of the crane will not result in an offsite radiological hazard. The crane travels over no safety-related equipment.

9.1.4.3.2 Seismic Considerations: The safety classifications for all fuel handling and storage equipment are listed in Table 3.2.B 2. SC 1, 2, and 3 equipment is designed to withstand the effects of an SSE without loss of capability to perform its safety function. Further, the combined normal and SSE stresses are limited to the allowable stresses as defined by ASME Code,Section III, Appendix XVII-2110. SC 1 and 2 equipment is designed to withstand the forces of an Operating Basis Earthquake (OBE), with the combined normal and OBE stresses being limited to the allowable stresses, as defined by ASME Code,Section III, Appendix XVII. For SC 3 equipment, consideration is given to the OBE only insofar as failure of the SC 3 equipment might adversely affect SC 1 or 2 equipment.

For NNS equipment, design for the SSE is considered if failure might adversely affect a SC 1, 2, or 3 component. Design for OBE is considered if failure of the NNS component might adversely affect an SC 1 or 2 component.

9.1 4.3.3 Containment Pressure Boundarv Interrity: The fuel transfer tube, which connects the refueling canal (inside the RCB) and the STP (outside ,

the Containment), is closed on the refueling canal side by a blind flange at all times except during refueling operations. Further discussion on the fuel transfer tube can be found in Section 3.B.2.1.3.3.

9.1-28 Revision 0

STPEGS UFSAR l 9.1.4.3.4 Radiation Shieldinc: During all phases of spent fuel transfer, the gamma dose rate at the surface of the water is 2.5 mR/hr or less. This is accomplished by maintaining approximately 10 ft of water above the top of the fuel assembly during all handling operations.

The two machines used to lift spent fuel assemblies are the refueling machine and the fuel handling machine. The refueling machine contains positive stops that prevent the top of a fuel assembly from being raised to within approximately 10 ft of the normal water level in the refueling cavity. The hoist on the fuel handling machine moves spent fuel assemblies with a long-handled tool. Hoist travel and tool length likewise limit the maximum lift of  :

a fuel assembly to within approximately 10 ft of the normal water level in the SFP.

9.1.4.4 Tests and Inspections. As part of normal plant operations, the fuel handling equipment is inspected for operability before each refueling operation. During the operational testing of this equipment, procedures are followed that will verify the correct performance of the FHS interlocks.

9.1.4.5 Instrumentation Reauirements. A description of the instrumentation and controls is provided in Section 9.1.4.3 for the refueling machine, the fuel handling machine, the FTS , and the SFCHS .

6 9.1-29 Revision 0

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TABLE 9.1-1 '

SPENT FUEL POOL HEATUP ANALYSIS RESULTS FOR 18-MONTH RELOAD CYCLES SPENT FUEL FUEL LOAD TYPICAL MAXIMUM ALLOWABLE MAXIPUM POOL MAXIMUM POOL MAXIMUM-POOL MODE (SRP 9.1.3 Criteria) STP FUEL LOAD POOL TEMP. TEMPERATURE TEMPERATURE' HEAT LOAD (SRP 9.1.3 Criteria) (1 COOLING TRAIN) (2 COOLING TRAINS) (10* Stu/hr)C#3 SRP 1 LOAD a 150 HRS

. NORMAL 1 LOAD a 1 YR N/A 140*F 142*F 124*F 23.6 MAXIMUM 1 LOAD a 400 DAYS STP 88 ASSEMBLIES a 150 HRS NORMAL N/A 88 ASSEMBLIES a 1 YR 140*F 155*F 131*F 31.4 MAXIMUM ACTUAL OR 88 ASSEMBLIES a 2-28 YRSD2 y

, STP 1 FULL-CORE G 100 HRS RAPID N/A 88 ASSEMBLIES a 1 YR N/A 202*F 155'F 62.5 REFUELING ACTUAL OR 88  !

ASSEMBLIES a 2-28 YRS"3 SRP 1 FULL-CORE a 150 HRS ABNORMAL 1 LOAD a 36 DAYS N/A NO BOILING f5 '3 189'F 148'F 53.6 $:

MAXIMUM 1 LOAD a 400 DAYS 'o w

STP 1 FULL-CORE a 150 HR$

(n Y ABNORMAL N/A 88 ASSEMBLIES a 36 DAYS NO BOLLING [5 6) 205*F 156*F 63.1 ,

y MAXIM.M 88 ASSEMBLIES a 1 YR g ACTUAL OR 88 g3 ASSEMBLIES a 2-28 YRSt ,i i

NOTES:

1. SRP = STANDARD REVIEW PLAN (NUREG-0800), SFP = SPENT FUEL POOL, FHB = FUEL HANDLING BUILDING, MAS = MECHANICAL AUXILIARY BUILDING, LHS! = LOW PEAD SAFETY INJECTION

2. MAXIMUM HEAT LOAD BASED ON FUEL EXPOSURE OF 40000 EFFECTIVE FULL POWER HOURS AND SRP ASB 9.2 DECAY HEAT FORMULATIONS.

3. MAINTAINS FULL CORE DISCHARGE CAPABILITY, i.e.,1776 FUEL ASSEMBilES IN SFP AND 193 LOCATIONS AVAILABLE -

4. ALL FUEL STORAGE LOCATIONS ARE FILLED WITH SPENT FUEL, i.e.,1%9 FUEL ASSEMBLIES ARE IN THE SFP. ,

5. IN THE EVENT OF FIRE OR MODERATE ENERGY LINE BREAK IN THE FHB THAT DISABLES BOTH TRAINS OF SFP COOLING, THE SFP MAY EVENTUALLY BOIL. MAKEUP WATER CAE BE PROVIDED VIA THE REACTOR MAKEUP PUMPS, IN ADDITION, SFP MAKEUP WATER CAN ALSO BE SUPPLIED USING LOCAL HOSE STATIONS IN THE FHB. DETAILS PROVIDED IN SECTION 3.3 OF THE FIRE HAZARDS ANALYSIS REPORT. ,

6. IF BOTH REACTOR MAKEUP WATER PUMPS ARE UNAVAILABLE BECAUSE OF FLOODING IN THE MAB, THE SEISMIC CATEGORY- LHSI PUMP WOULD BE AVAILABLE FOR REFILLING THE SFP BY CONNECTING TEMPORARY HOSES TO THE VENT AND DRAIN VALVES LOCATED ON LHS! PUMP DISCHARGE PIP!NG SO THAT REFUELING WATER COULD BE DELIVERED TO THE SFP.

MISC-95\95-Ols.002

-r - . - - - . . - . - - - - , ,, - _ _ _. ._ , _ , - ,--, ,,-, , , - , - . . , , . - , - ,, -,,e

4

- STPEGS UFSAR TABLE 9.1-2 SPENT FUEL POOL COOLING AND CLEANUP SYSTEM COMPONENT DESIGN PARAMETERS Spent Fuel Fool Cooling Pump 2

Number t Design pressure, psig 150 Design temperature, 'F 200 Design flow, gal / min 2,500 200 Design head, ft Material Stainless steel Spent Fuel Fool Skimmer Pump 1

Ndmber Design pressure, psig 150 200 l Design temperature, 'F Design flow, gal / min 100 50 Design head, ft I

Material Stainless steel Refueling Water Purification Pump 1

Number Design Pressure, psig 150 Design temperature, 'F 200 200 Design flow, gal / min J i

Design head, ft 200 i

Material Stainless steel ]

9.1 31 Revision 0 4

c.

STPEGS UFSAR 4

TABLE 9.1-2 (Continued) i SPENT FUEL POOL COOLING AND CLEANUP SYSTEM COMPONENT DESIGN PARAMETERS Spent Fuel Pool Heat Exchanger Number 2 Design heat transfer, Btu /hr 9.1 x 10 8 Shell Inh.e Design pressure, psig 150 150 Design temperature, 'F 200 200 Design flow, lb/hr 1.5 x 10 8 1.4 x 10 6 Inlet temperature, "F 105 120 Outlet temperature, 'F 111.1 113.6 Fluid circulated Component Spent fuel cooling cooling  ;

water water Material Carbon steel Stainless steel i Spent Fuel Fool Demineralizer  !

Number 2 Design pressure, psig 300 Design temperature, *F 250 P

Design flow, gal / min 250 Resin volume, ft8 75 Material Austenitic stainless steel 9.1 32 Revision 0

e STPEGS UFSAR a.

TABLE 9.1 2 (Continued)

(

SPENT FUEL POOL COOLING AND CLEANUP SYSTEM COMPONENT DESIGN PARAMETERS Spent Fuel Fool Filter Number 2 l

Design pressure, psig 300 l

l Design temperature, 'F 250 1

Design flow, gal / min 250 Filtration requirement 98% retention of j particles above 5 microns Material, vessel Austenitic stainless steel Spent Fuel Pool Skimmer Filter Number 1 Design pressure, psig 300 Design temperature, 'F 250 Design flow, gal / min 250 Filtration requirement 98% retention of )

particles above 5 i microns Material, vessel Austenitic stainless steel ,

Spent Fuel Pool Strainer Number 2 Design pressure, psig Not applicable Design temperature, 'F 200 Design flow, gal / min 5,000 Material Austenitic stainless steel 9.1 33 Revision 0

^

y ,.

ri STPEGS UFSAR

..e-4 g TABLE 9.1-2 (Continued) i 1

SPENT FUEL POOL COOLING AND' CLEANUP SYSTEM l COMPONENT DESIGN PARMETERS l t

t Spent Fuel Pool Skimmer / Strainer. Assembly  !

Number 2 ,

Design pressure, psig 50-Design temperature, 'F 200 Design. flow, gal / min 50 Material Austenitic  ;

stainless steel l

In-Containment Storage Area Strainer ,

i Number 1 Design pressure, psig Not applicable  !

Design temperature. *F 200 [

Design flow, gal / min 2,500 l Material Austenitic f stainless steel j l

Reactor Cavity Filtration System  !

Number 1 Design Pressure, psig 150  !

> s Design temperature, *F 200  !

Design flow, gal / min 250 t Design head of pump, ft 48 i Filtration requirement of filters 10 micron Material Austenitic )

stainless steel j l

9.1-34 Revision 0 !

4

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+

STPECS UFSAR TABLE 9.1 3 FHB 15/2-TON CRANE - COMPLIANCE WITH REGUIATORY GUIDE 1.104 STPEGS Meets the STPEGS Intent of Rev. O STPEGS Not Applicable by l

Complies and Complies with Takes Either Rev. O or Regulatory with a Proposed Revision Exception the Proposed Revision Position Rev. O Dated January 1978 to Rev. O Dated January 1978 C.1.a X 1.b(1) X (2) X (3) X 1 (4) X 1.c X 1.d X 1.e X 1.f X C.2.a X '

2.b X I 2.c X j 2.d X l 1

C.3.a X 3.b X 3.c X 3.d X 3.e X 3.f X 3.g X 3.h X 3.1 X  !

3.j X l 3.k X j 3.1 X {

3.m X l' 3.n X 3.o Note 1 3.p X 3.q X 3.r X 3.s Note 2 3.t X 3.u X

1. Controlled plugging measures shall be provided so that if the operator reverses a drive while it is in motion, the torque during reverse shall be automatically controlled to a predetermined torque limit during deceleration.

2. The crane is designed to CMAA standards (i.e, a 5:1 minimum factor of safety for each component) for 15-ton lifts.

9.1-35 Revision 0

s.

STPEGS UFSAR TABLE 9.1-3 (Continued)

FHB 15 /2-TON CRANE - COMPLI ANCE VITH REGUIATORY CUIDE 1.104 e i STPEGS Meets the STPEGS Intent of Rev. O STPEGS Not Applicable by Complies and Complies with Takes Either Rev. O or l Regulatory with a Proposed Revision Exception the Proposed Revision Position Rev. O Dated January 1978 to Rev. O Dated January 1978 .

i C.4.a X i

4b X X  !

4.c X

4.d {

C.S.a X  ;

5.b X i

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9.1-36 Revision 0 i 4

, STPEGS UFSAR TABLE 9.1-4 NSSS VENDOR RECOMMENDED SPECIFICATIONS AND CUIDELINES FOR SPENT FUEL POOL UATER PURITY Specification Parameters Scric Acid, ppm B =2800 l Chloride, ppb s150 Fluoride, ppb s150 Guideline Parameters pH @ 77*F 4.0 - 4.7 Aluminum, ppb s500 Calcium + Magnesium, ppb s500

)

Magnesium s250 i

i l

l 9.1-37 Revision 4

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i TABLE 9.1-5 I SPENT FUEL POOL COOLING AND CLEAN-UP SYSTEM FAILURE MGsES AND EFFECTS ANALYSIS

• i Plant Method Faiture Effect Description Safety Operating Failure of Failure on System Safety j of Conconent Function Mode ** Mode (s) Detection Function Capability General Remarks SFP Cooling Circulate the 1-6 One pung Status None - A redundant Punps water to the faits to monitoring cooling train is (Typical) spent fuet. provide , avaltable which will pool adequate _ Temperature provide adequate cooling flow indication of SF? of the SFP water Class 1E AC Provide 1E 1-6 Loss one Bus taider- None - A redwident power Power power to the train of voltage alarms train exists to power the (Typical) - punps power to redundant ptmp y its ESF status -4' w associated monitoring of 8 pump pumps _m i m Pung status gs tights m' Channett III Loss of DC ESF monitoring on None - Redundant trains Provide DC 1-6 DC Power control power power UPS failure, DC provide system capability (Train 8) trouble starm ESF monitoring for ptmp (not rtaining, no controt power)

Channet IV DC Provide DC . 1-6 - Loss of DC ESF monitoring on None - Redundant trains Power control power power' UPS failure DC provide system capability (Train C) trouble alarm ESF monitoring for ptmp (Pat

! running, no

e control power) is d,. (A.dd tnsert".#.91 o

• Single failure only applies to normal modes of SFPCCS operation.

D ** Ptont Modes o 1. Power Operation 3. - Hot Stanchy 5. Cold Shutdown

2. Startup 4. Hot Shutdown 6. Refueling MISC-95\95-018.002 l

__ _ _ _ - _ _ _ _ _ _ _ . - - ..-_._--.m_ .m - - . _ _ . . - . _ . . _ ._. -, _ . _ - -. __.._.-.----._.-...a---.--_-_-_._._......__-.__.. ._

. -.-;e- ,

UFSAR - INSERT #9 DESCRIPTION OF PLANT MET 110D OF I AILURE EnTCT ON SYSTO1 COMPONmi SAFETY FUNCTION OPERATING FAtWRE MODE (s) FAIWRE DETEC110N SAFETY FUNC110N GENERAL REMARKS MODE" CAPABi1IIY MOVs 0447 A 0032 Circulate CCW water 1-6 MOV fails closed. Loss of CCW Flow Alann NONE. Frequency of occunence is minimal since (Nonnally Open) to SFP Cooling IIcat (Fails to Reopen at Main Contml Board The valve can be cpened the valves are normally open and close Exchangers A & B. (Full Power following an ESF (non-Class IE). either manually or only on an ESF signal. After receiving an to Signal) locally by an operator. ESF signal Emergency Operating Note: Refueling liigh SFP Temperature Procedure instructs the operator to open The FMEA of with Alarm at Main Control the MOVs within an allotted time frame.

MOVs 0447 & full-core Board (non-Class IE).

0032 is applicable ofiload) Failure can be prevented by checking to to the SFPCCS. ascertain the valves reopen after closing This does not due to an ESF signal. If the valve fails to address the ESF reopen manua!!y after receiving an ESF function of the signal, the valve can be reopened locally, valves.

The consequences of loss of SFP cooling are mitigated by administratively providing makeup water through:

(1) Reactor makeup water system, (2) Water from RWST (3) Demineralized water, or (4) Fire water.

Misc-95\95418 002

.. - . - _ . _ . . - . _ .-. . . . . - _ .. . - - . - ~ . . . - . . - - . . . . - - - - - . . . _...- -

. . - . . _ . . , . . . . _ _ _ ~ . -...--- -_- -

4 4 '

- STPEGS UFSAR Table 9.1-6 SPENT FUEL POOL BOILING i

Parameters ,

Core Thermal power, Mut 3,800  :

Decay assumed for the last full core offload 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> l3 11ours of reactor exposure per as'sembly 40,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Decay heat generator methodology SRP 9.1.3 ,

Volume of water in spent fuel pool, ft 3 47,105

+

(water level 3 ft below normal water level with 10% of the pool volume assumed to be - 3 stainless steel)

Dispersion Factors (x/Q) Table 15.B-1

\

Dose Consecuences Exclusion Zone Boundary (0-2 hours) 1.3 x 10"3 Thyroid, rems 3

Low Population Zone (0 30 days) 7.1 Thyroid, rems l l

i 9.1-39 Revision 3

i I

, STPEGS UFSAR 1

9.2.2 Component Cooling Water System 9.2.2.1 Desien Bases. The CCWS meets the requirements of 10CFR50, l Appendix A, General Design Criteria (GDC) 1, 2, 3, 4, 5, 44, 45, 46, 54, 56 and

57. STPECS Units 1 and 2 have separate but identical CCWSs. l The CCWS is designed to: 1

1. Provide cooling water to various nuclear plant components during all modes of plant operation. This includes plant equipment required for safe  :

shutdown and ESF equipment required after a postulated DBA.

2. Provide an intermediate fluid barrier between potentially radioactive

)

systems and the ECWS to reduce the possibility of leakage of radioactive i contamination to the outside environment. '

3. Perform its cooling function following a DBA with offsite or standby power sources, automatically and without operator action, assuming a single active or passive failure.

4 Provide cooling water at 60*F to 105'F temperature during normal operation.

The maximum temperature during DBA is 120.5*F (refer to Table 9.2.5-5 for temperature for the individual scenarios).

5. Conform to seismic Category I requirements and safety classifications as indicated on Figures 9.2.2-1 through 9.2.2-5 and in Table 3.2.A-1.

6. Permit periodic inspection of important components and periodic and functional testing to assure the integrity and operability of the system.

See Sections 3.9.6 and 6.6.

In addition, the CCWS is protected from the effects of tornado loadings,  :

missiles, flooding, pipe whip, and jet forces from pipe breaks. See Sections 3.3.2, 3.4.1, 3.5, and 3.6.

9.2.2.2 System Descriotion.

9.2.2.2.1 Descrintion: The CCWS consists of three separate redundant trains, each with a pump, HX, associated piping, and valves, that service two RCFCs, Residual Heat Removal (RHR) HX, and RHR pump, as shown on Figures 9.2.2-1 through 9.2.2-3. The three trains are connected to a common header which services other equipment as shown in Figures 9.2.2-4 and 9.2.2-5. In addition, a :

compartmentalized surge tank is used to accommodate the water thermal expansion l and contraction, and a chemical addition tank is used to balance the water j chemistry (Table 9.2.2-2) of the system. 1 A CCW RX bypass line is provided to maintain 60*F minimum CCWS temperature. This line is only used when the ECW temperature is very low.

For heat removal following a DBA, all three CCWS trains will operate if {

available, but two trains are capable of performing the heat removal function.  :

Except for the seal water HX, reactor coolant pump (RCP) lube oil coolers and )

thermal barrier, RCP motor air coolers, RHR pump seal coolers, centrifugal I charging pump (CCP) supplementary coolers, CCP lube oil coolers, and positive  !

displacement pump supplementary cooler, the remaining equipment is isolated by 1 valves which close on an SI signal. Flow to the RCP lube oil coolers, thermal l barrier, and notor air coolers is automatically isolated upon reaching the j Containment pressure HI-3 setpoint. An SI signal opens the pneumatic valve '

(closed during normal operation) to provide cooling water to each RHR HX. Also, an SI signal shifts the cooling water supply to the RCFCs from the chilled water system to the CCWS by closing the chilled water and opening the CCUS motor-operated supply and return valves.

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. STPEGS UFSAR Cooling water to the spent fuel pool (SFP) HXs is manually restored within eght heurc yhe(pime[fppm{ permitted [bf]plaiitIEsergene(Operatiing(Procedtipes following the SI-induced isolation. j Upon a LOOP signal, the RCB chilled water supply and return valves are closed automatically. The CCWS motor-operated valves (MOVs-0057, 0069, 0148, 0136, 0210, 0197) to and from the RCFC remain closed. The CCWS motor-operated valves (MOVs-0235, 0236, 0393, 0297) to the letdown and excess letdown heat ,

exchangers remain open. Operator action from the main control room is required to restore flow to the RCFCs by closing MOVs-0235, 0236, 0393, 0297 and opening MOVs-0057, 0069, 0148, 0136, 0210, 0197 within 30 minutes after LOOP.

The following components are cooled by the OCWS:

1. ESF Loads

a. RHR HXs

b. RCFCs

2. Non-ESF Loads f

a. RCPs - lube oil coolers and thermal barrier

b. RCP motor air coolers

c. RHR pumps - seal coolers

d. Centrifugal charging pumps (CCPs) - lube oil coolers

e. Excess letdown HX

f. Reactor coolant drain tank (RCDT) HX

g. Seal water HX

h. Boron Recycle System (BRS) recycle evaporator package

i. Boron Thermal Regeneration System (BTRS) chiller unit

j. Post-accident sampling system (PASS) sample coolers, primary j sampling coolers, boric acid sample coolers, and radiation monitor sample coolers

k. Liquid Waste Processing System (LWPS) evaporator package

1. SFP HXs

m. Letdown HX

n. CCP supplementary coolers

o. Positive displacement pump supplementary cooler.

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o STPECS UFSAR The maximum heat loads and required flow rates for each component durin5 the different modes of operation are listed in Table 9.2.2-4.

Each CCW HX is designed to meet the normal operational heat loads. The design

~ maximum CCW outlet temperature from the CCW HX is 105*F. During normal  !

operation, one CCW pump, one compartment of the CCW surge tank, one ECUS pump, and one CCW HX are in use to provide a source of cooling water. The system has sufficient capacity to meet the required heat removal rates for the remaining operating conditions such as startup, shutdown, and recirculation.

The CCWS is required to remove residual and sensible heat from the Reactor Coolant System (RCS) through the RHR HXs from approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after normal shutdown (when the RHR HXs are placed in service) until the reactor coolant temperature reaches 150*F. Operation of the system is then continued in order to maintain cold shutdown. The CCW flow from each RHR HX is indi-I cated on the main control board and displayed through the QDPS. The position I of each CCW block valve in the outlet line of each RHR HX is monitored by l open/close indicating lights on the main control board and by the ESF Status Monitoring System. Since the RHR HX does not need to function until after recirculation switchover is completed, following a design basis IDCA, the operator would have sufficient time to open the CCW block valve by using the manual switch in the control room, if the SI signal failed to actuate the valve. If this attempt to open the block valve fails, the corresponding SI train should be shut down and the core cooling requirement will be provided from the two remaining SI trains. The CCW trains should also be shut down because of flow considerations (minimum pump flow). The normal cooldown heat load is removed by all three CCW pumps and HXs. If one of the three trains is inoperative, the cooldown function is still effective, although the cooldown period is extended beyond the three train cooldown time of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. However, this does not affect the safe shutdown and cooldown of the reactor.

Pumps, safety-related valves, and instrumentation required to operate each CCW train are provided with power from their respective standby power sources if normal and offsite power fail. Two of the three cooling trains are available l

within 60 seconds for a safe shutdown and cooldown of the reactor in case of a IDOP and a simultaneous failure of one standby power source.

Each CCW pump is connected to a separate ESF bus which is powered from either offsite or standby power sources. The pump control logic provides for automatic starting of the standby pumpf s) in the event of low system pressure or low ECW pressure during normal operation.

The CCW surge tank is partitioned into thr.. equal volumes by internal baffles. Each compartment is connected to the inlet piping of one of the CCW pumps. The surge tank is sized to accommodate the water thermal expansion and contraction and the inleakage or outleakage in the system until the leaking component is isolated. The internal baffles are designed to provide separation between redundant CCW trains, so excess leakage from a pipe break in one train does not affect the operability of the other trains. The surge tank is located at the highest point in the system to ersure the required NPSH for proper operation of the CCW pumps and to provide a path for collecting entrapped gases. A radiation monitor activates an alarm if the CCW radioactivity increases beyond a preset level (see Section 11.5) due to inleakage from systems being cooled. Makeup to the CCWS is added automatically from the Demineralized Water System (DWS). Low surge tank level 9.2-12 Revision 2

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initiates makeup; high surge tank level terminates makeup. As a backup by manual valve alignment, makeup water can be obtained from the Reactor Makeup Vater System (RMVS) which is a safety-related and seismic Category I An makeupalarm is source. A high-high surge tank level alarms in the control room.

also indicated in the control room if the level in the surge tank continues to decrease due to inadequate makeup. Further decrease in the surge tank level closes the valves on the lines to some non-ESF equipment (BTRS chiller unit.

LMPS evaporator package, sample coolers, radiation monitor sample coolers, l

1etdown HX, excess letdown HX, BRS evaporator package, and RCDT HX).

l If the level in the surge tank continues to decrease, each CCW train is '

automatically isolated from the other trains by closing the motor-operated supply and return valves located upstream and downstream of the common supply and the return headers, respectively. Simultaneously, the pneumatic cross-connect valves (FV 4656 and FV-4657) in the CCP and positive displacement pump coolers supply and return headers (CCW train A and CCW train B from CCW train ,

l C) and motor operated valves MOV 0768 and MOV 0772 are closed to prevent loss of more than one CCW train at a time. As soon as the CCW common header is isolated, the low pressure switch located in the common header will automatic-ally start the standby CCW pump (s).

1 CCW supply to the RCPs is terminated as a result of the isolation of the CCW common header. However, the seal water to the RCPs from the CCPs is main-tained since CCW supply to the CCP lube oil coolers is not isolated. Should the water level fall below the top of the surge tank dividing baffles, an alarm is indicated in the control room. Low-low surge tank compartment level initiates another alarm in the control room. The operator may trip the pump to respond to these alarms.

Self-actuated, spring-loaded relief valves are provided for lines and com-ponents which could be pressurized beyond their design pressure due to improper operation or failure of the physical barrier separating the CCW from other high pressure systems.

The system, except for NNS portions, is designed to seismic Category I requirements and is located inside seismic Category I structures. A failure i in the NNS portion would not affect the operation of the remainder of the system since low (makeup inadequate) surge tank level isolates the NNS section. The system equipment and piping are classified as SC 2 and 3 and I

NNS, as shown on Figures 9.2.2-1 through 9.2.2-5. The codes and standards to which the system is designed are listed in Table 3.2.A-1.

Table 9.2.2-2 specifies the water chemistry of the system. The system is initially filled with water from the demineralized water storage tank (DWST) through the surge tank. The corrosion inhibitor is added during the system filling via the surge tank manhole. During normal operation the corrosion inhibitor concentration is maintained via the chemical addition tank utilizing the CCW pump discharge pressure for injection to the surge makeup header.

9.2.2.2.2 Coroonents:

1. Heat Exchangers The CCW HXs are of the shell and straight tube type. The ECW circulates through the tubes while CCW circulates through the shell. Titanium tubes are 9.2-13 Revision 2

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used to provide maximum corrosion resistance to the brackish ECW. The shell is carbon steel. The shell-side (CCW) pressure is maintained higher than the tube-side pressure to prevent leakage of ECW into the system. Design

, parameters are listed in Table 9.2.2-1,

2. Pumps  ;

The CCW pumps are motor-driven horizontal, double-suction, dual-volute centrifugal pumps with antifriction bearings. Design parameters are listed in Table 9.2.2-1.

3. Surge Tank The CCW surge tank is constructed of carbon steel. The tank is partitioned to form three separate volumes in the lower portion of the tank. The upper portion of the tank is open to each compartment. Design parameters are listed in Table 9.2.2-1.

4. Chemical Addition Tank The chemical addition tank is a vertical cylindrical tank constructed of carbon steel. Its design parameters are given in Table 9.2.2-1.

5. Valves -

The valves used in the CCWS are constructed of carbon steel. Self-actuated, r spring-loaded relief valves are provided for lines and components which might be pressurized beyond their design pressure by improper operation or malfunc.

tion. Table 9.2.2-5 lists those relief valves in the CCWS for which Code Case N-242-1 is permitted. This Code Case provides alternate rules which may be used for the acceptance of metallic materials which were not manufactured or supplied in complete conformance with the rules of NCA-3800 (or NA-3700) and which are used in the construction of items for which the applicable code is Winter 1973 Addendum or later. The valves are listed by tag number.

6. Piping All CCW piping is carbon steel with welded joint end connections, except at components which might be removed for maintenance. In this case, flanged connections are used.

9.2.2.3 Safety Evaluation.

9.2.2.3.1 Availability and Reliability: The normal operation of each CCW pump and HX train is rotated to monitor operational capability and balance operating time.

The operability of safety-related valves and equipment, as required during various modes of operation of the system, is tested by simulated signals corresponding to each mode on a routine basis.

The CCWS performs vital cooling functions during and after a postulated DBA, and therefore is designed to meet the single-failure criterion. Sufficient cooling capacity is provided to fulfill all system requirements under accident  !

conditions, assuming a single failure in the CCVS.

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o STPEGS UFSAR To meet the ESF cooling requirements, the CCW pumps are automatically sequenced on standby power in the event of a 140P. The electric power supplies to the pumps, valves, instrumentation, and control cabling for each cooling train are physically separated. Power is supplied to each pump from an independent ESF bus.

The CCW pumps, HXs, SC 2 and 3 valves, and piping are designed to seismic Category I requirements and are located in seismic Category I structures.

They are protected against internally generated missiles, pipe whip, jet forces, and flooding due to pipe rupture. The CCUS equipment is located in j

areas shielded from radiation and is accessible for maintenance or inspection during power operation.

! A failure mode and effect analysis (FEMA) for the CCUS is given in Table 9.2.2-3.

The CCUS design-incorporates automatic isolation of cooling water to the RCPs by the Containment isolation phase B signal. Loss of cooling water to che l RCPs due to a spurious isolation signal or the operator closing a single isolation valve is avoided by parallel Containment isolation valves, as shown on Figure 9.2.2-5. The cont.rel logic for the valves is designed so that a spurious isolation signal in any one actuation train does not isolate flow; however, isolation is nasumed on a penetration basis even assuming the single failure of one of the three actuation trains.

9.2.2.3.2 Leakare Minimization. Detection. and Isolation: To minimize the possibility of leakage, welded construction is used throughout the CCWS where practicable.

Makeup to the system is automatic. The surge tank level is monitored by the main plant computer and the QDPS. Opening of the valve providing tank makeup is alarmed by the computer to give an indication of system leakage. If the normal source of makeup (the DWS) fails or is inadequate, the surge tank level l

instrumentation provides alarms and actuations as discussed in Section 9.2.2.2.1.

I Should a large inleakage into the CCWS develop, the level in the surge tank will rise, and the high-high-level condition will be annunciated. If a p leaking fluid is radioactive, then a high radiation level will be alarmed.

Refer to Section 11.5 for further information on the process radiation monitor. If the level in the surge tank continues to increase, the CCW will discharge to the CCW sump through the rpen vent. The vent is designed to accommodate the maximum inleakage due to the rupture of one RCP thermal barrier, which is estimated at 275 gal / min. The increased pressure due to a rupture of the RCP thermal barrier will close two active self-actuated pressure regulated valves located at the CCW outlet from the RCP thermal batrier and isolate the RCS inleakage. In addition, a high flow or high temperature in the return line from the thermal barrier, indicating failure of the RCP thermal barrier pressure retai..ing boundary, will close the CCW motor-operated valve downstream of the two pressure regulated valves; an alarm in l the control room alerts the operator as well. The portion isolated is designed for RCS pressure and temperature. 1 The operator, by checking flow and temperature readings against normal values, can locate the affected portion of the system and isolate this portion by 9.2-15 Revision 0 L-- --- - - - _ _ _ _ _ _

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• STPEGS UFSAR closing the appropriate remotely operated or manual valves. Very small leaks will be' detected by periodic inspection of the system piping and valves.

The relief valves on the CCW lines to the various HXs are sized to relieve the volumetric expansion occurring if the shell side of the HX is isolated and high-temperature process fluid flows through the HX tubes.

9.2.2.4 Tests and Inspections. During preoperational testing, the following. vill be checked:

1. Calibration of all instrumentation

2. Actuation of automatic controls at their proper setpoints

3. Actuatien of alarms at the'setpoints

4. Operation of power-operated valves

5. Operation of CCW pumps and checking of flow and discharge pressure

6. Checking and adjusting required flow to each component serviced by CCUS

7. System water chemistry All the above functions are checked periodically during the life of the plant.

Inservice inspections of ASME B&PV Code,Section III, Class 2 and 3 components are performed in accordance with ASME B&PV Code,Section XI.

9.2.2.5 Instrumentation Avolication. The Engineered Safety Features Actuation System (ESFAS) for the CCWS is discussed in Section 7.3.1. Controls for remote manual operation of each CCW pump and selection of standby CCW ,

pumps are provided in the control room. Remote manual control of )

pneumatically operated valves under normal operating conditions is provided in I the control room. Pneumatically operated valves are not required for control I

purposes and fail in the safe position on LOOP. Remote manual control of all motor-operated and solenoid-operated valves necessary for post-LOCA cooling, for surge tank makeup, for maintenance of the CCWS trains, and for Containment isolation is provided in the control room. Ti'e valves that are power locked j out will require that the motor control center (MCC) breaker be closed to  !

enable remote operability. CCW pump operating status and valve position .

indicating lights are provided in the control room. For valves that are power l locked out the indicating lights are not illuminated and control room indica-tion is given via bypass / inoperable for valves not in the power locked out position. Provisions are made for local indication and control in the switchgear room for each CCW pump and selected valves in the CCWS.

The temperature in each CCW main loop is monitored by the QDPS. The tempera-ture at each CCW pump discharge is also monitored by the plant computer. The CCW return temperatures from all equipment are monitored through the plant computer except for the sample coolers, boron recycle evaporator, positive displacement pump supplementary cooler, CCP lube oil coolers, LWPS evaporator package, and CCP supplementary coolers. Temperatures for this equipment, with the exception of the sample coolers, are monitored locally.

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Flows through the CCW main loops, RHR RXs, and RCFCs are displayed in the control room on panel indicators and logged by the QDPS while the remaining equipment, with the exception of the letdown HX and sample coolers, has local flow indications or test connections. The temperature difference in conjunc- i tion with the flow data from individual components is used to monitor the l particular component's performance. Each CCW main loop outlet pressure is )

indicated in the control room through indicators and the QDPS. The radiation j level in the CCWS is available in the control room through the Radiation '

Monitoring System (RMS) display. 1 Local pressure gauges are provided on the suction and discharge lines of each CCW pump. Local level gauges are provided for each compartment of the surge tank. Surge tank level indication is displayed in the control room through indicators and Le QDPS. High and low levels in the surge tank are alarmed in the control room. Measurements taken downstream of each CCW HX, which indicate system malfunction, are annunciated in the control room. Local flow indication is provided on each RCP lube oil cooler and motor air cooler out-let. Low flow condition at aach RCP lube oil cooler and motor air cooler outlet is annunciated in the control room.

9.2.3 Makeup Demineralized Water System 9.2.3.1 Desien Baset. The Makeup Demineralized Water System (HDWS) uses ozonated and filtered service water as influent and removes the ionic impurities of the raw water to provide high-purity demineralized water suitable for use in the primary and secondary cycles of the plant. Sodium hypochlorite is used as a backup in the event the ozone generator fails.

The purity of the water produced is suitable for preoperational tests, wet layup, startup, and normal operation of the plant primary and secondary systems. The system is designed for a specific nominal capacity, and transient demands for abnormal operating conditions are allowed for by the buffer effect of the storage volumes of the Demineralized Water Distribution System. For a more complete description of the Demineralized Water Distribution System and supply interfaces with other plant systems, refer to Sections 9.2.6 and 9.2.7 and Figures 9.2.6-1, 9.2.6-2, and 9.2.7-1. The capability of the MDWS to store, handle, and dispense any chemicals used in the demineralization and regeneration process is described in Section 3.6 of the Environmental Report.

Piping, ion exchangers, and associated equipment of the MDWS are constructed of corrosion-resistant material or lined carbon steel to prevent contamination {

of the water by corrosion products. The equipment is designed in accordance '

with ASME B&PV Code,Section VIII, Division 1.

The MDWS is designed to produce demineralized water to meet NSSS vendor recommended guidelines as follows:

Specific Conductivity"' Less than 0.1 micro mhos/cm at 25*C 0xygen Less than 100 ppb Suspended Solids Less than 50 ppb Total Silica, SiO 2 Less than 50 ppb 9.2-17 Revision 0 t