ML20199L456

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Proposed Tech Specs Revising TS Table 3.7-6, Air Temp Monitoring. Proposed FSAR Pages Describing Full Core off- Load Condition as Normal Evolution Under Unit 3 Licensing Basis,Included
ML20199L456
Person / Time
Site: Millstone Dominion icon.png
Issue date: 01/18/1999
From:
NORTHEAST NUCLEAR ENERGY CO.
To:
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ML20136C553 List:
References
NUDOCS 9901280015
Download: ML20199L456 (38)


Text

.. _ _ - _

.~ _ _..___ _

l l

l Docket No. 50-423 I

B17004 l

Millstone Nuclear Power Station Unit No. 3 License Amendment Request and Technical Specification Changes For Full Core Off-load Marked-Up Technical Specification Pane (s_)

January 1999 9,012e0015 99011e

,f' 4

PDR ADOCK 05000423 PM p

~

TABLE 3.7-6 (Continued)

AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (*F1 7.

FUEL BUILDING FB-02, Fuel Pool Pump Cubicles, El 24'6" s 4M-II ?

FB-03, General Area, El 52'4" 140+ /48 8.

FUEL OIL VAULT FV-01, Diesel Fuel Oil Vault 5 95 9.

HYDROGEN RECOMBINER BUILDING HR-01, Recombiner Skid Area, El 24'6" 5 125 HR-02, Controls Area, El 24'6" i 110 HR-03, Sampling Area, El 24'6" 1 110 HR-04, HVAC Area, El 37'6" 1 110 10.

MAIN STEAM VALVE BUILDING MS-01, Areas Above El 58'0" 5 140 MS-02, Areas Below El. 58'0" s 140 11.

TURBINE BUILDING TB-01, Entire Building 5 115

12. TUNNEL TH-02, Pipe Tunnel-Auxiliary, Fuel and ESF Bu.ilding 5 112
13. YARD YD-01, Yard s 115 MILLSTONE - UNIT 3 3/4 7-35 Amendment No. 77,100 assi

1 Docket No. 50-423 B17004 Millstone Nuclear Power Station Unit No. 3 License Amendment Request and Technical Specification Changes For Full Core Off-load Marked-Up Updated Final Safety Analysis Report Pane (s) l 1

January 1999

U.S. Nucirr Regulatory Commission B17004/ Attachment 5/Page 1 INDEX OF MARKUP OF PROPOSED REVISION Refer to the attached markup of the proposed revision to the Updated Final Safety Analysis Report (UFSAR). The attached markup reflects the currently issued version of the Updated FSAR.

The following changes are included in the attached markup.

Affected UFSAR Sections Section Title Paae Table 1.3-15 Comparison of Other Reactor Plant Systems 1 of 4 Table 1.9-1 Summary of Differences from SRP 14 of 16 Table 1.9-2 SRP Differences and Justifications 33 of 41 Table 3.6-3 Moderate-Energy Systems Outside Containment Remote from Essential Systems, Components and Structures 2 of 4 3.8.4.3 Loads and Loading Combinations 3.8-40 Appendix 3B, Appendix E, Environmental Design Conditions Chapter 9 List of Figures 9-viii 9.1.2.1 Design Bases 9.1-2 9.1.3.1 Design Bases 9.1-6 9.1.3.2

System Description

9.1 -6 9.1.3.3 Safety Evaluation 9.1-8 9.1.3.3 Safety Evaluation 9.1-9 Figure 9.1.7 Normal Refueling Figure 9.1.7A End of Cycle Full core Offload Figure 9.1-8 Emergency Core Offload Table 9-1.1 Fuel Pool Cooling and Purification System Principal Component Design Characteristics Table 9.1-2 Performance Characteristics of the Fuel Pool Cooling System j

9.4.2.1 Fuel Building Ventilation System - Design Bases 9.4-10 Details of Changes 1.

Table 1.3-15, Fuel Pool Cooling and Purification System (Section 9.1.3): Under the column labeled " Millstone 3", delete "1 1/3" for

  • Number of cores cooled" and replace with "151/2 (3048 fuel assemblies)."

Under the column labeled " Millstone 3", delete "140" for " Fuel pool temperature, normal ( F)" and replace with "150".

U.S. Nuclear Regulatory Commission B17004/ Attachment 5/Page 2 2.

Table 1.9-1, page 14 of 16, SRP Section 9.1.3, delete " Decay heat.. BTP ASB 9-2." and replace with " Decay heat removal is based on DECOR (based on ORIGEN2) computer code and credit for evaporative cooling instead of BTP ASB 9-2."

Insert the additional SRP Section 9.1.3 as shown below.

Specific SRP Summary Description Corn vonding SRP Section Acceptance Criteria of Difference FSAR Section 9.1.3 Ill.1.d - for maximum The maximum temperature 9.1.3.3 (Rev.1) normal heat load, the pool for a normal heat load is temperature should be kept 150 F.

at or below 140 F 9.1.3 Ill.1.h (ii) - maximum heat The decay time for the 9.1.3.3 (Rev.1) load is after 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> of maximum heat load is decay based on the heat removal capacity of the spent fuel pool heat exchangers and varies from 165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> to 349 hours0.00404 days <br />0.0969 hours <br />5.770503e-4 weeks <br />1.327945e-4 months <br />.

i 3.

Table 1.9-2, page 33 of 41, SRP 9.1.3, Section A: delete " Westinghouse generated curves," and replace with "the DECOR computer code (based on ORIGEN2) and credit for evaporative cooling," Number this difference "1" Insert the following as difference 2. "The maximum temperature for a normal heat load is 150 F, not 140 F as required by SRP 9.1.3 Paragraph lli.1.d."

Insert the following as difference 3: "The decay time for the maximum heat load in the spent fuel pool is based on the heat removal capacity of the spent fuel pool heat exchangers and varies from 165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> to 349 hours0.00404 days <br />0.0969 hours <br />5.770503e-4 weeks <br />1.327945e-4 months <br />, not 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> as required by SRP 9.1.3, Paragraph Ill.1.h(ii)."

Section B: delete " current NSSS curves utilized for the fuel that will actually be supplied to Millstone 3 " and replace with "the DECOR computer code (based on ORIGEN2) and credit for evaporative cooling in order to get a more accurate value of decay heat loads" Number this justification "1".

Insert the following as justification 2: "All SSCs associated with the Spent Fuel Pool have been evaluated and have been found to be acceptable for an increase over the SRP limit of 140 F. The decay heat of the fuel is removed and the water coverage of the fuelis maintained for all anticipated scenarios."

U.S. Nucl:ar Regul tory Commission B17004/ Attachment 5/Page 3 l

Insert the following as justification 3: "The decay time for the maximum heat load in the spent fuel poolis based on heat removal capacities that are dependent on the actual cooling water temperatures. Colder cooling water temperatures result in greater the heat removal capacities which permit larger heat loads to be placed in j

the pool and the shorter the decay time "

4.

Table 3.6-3, page 2 of 4, item " Fuel Pool Cooling and Purification", under column

" Location" add "FB" Under column " Temperature" change "140" to "150" 5.

Section 3.8.4.3, page 3.8-40: Delete the second sentence ("The spent fuel pool walls... Figures 3.8-79 and 3.8-80.") and replace with the following insert.

"The historical design of the spent fuel pool walls and mat considered the thermal effects based on the temperatures indicated in Figures 3.8-79 and 3.8-80. The analysis for classifying a full core off-load as a normal evolution evaluated the thermal effects based on temperatures indicated in Figure 3.8.82."

6.

Add Pigure 3.8.82 7.

Appendix 3B, Attachment 1, Appendix E Environmental Design Conditions:

Revised to show the updated environmental conditions associated with a normal fuel temperature of 150 F in accordance with calculation P(B)-1118, rev. O, CCN

1. Note 1 was revised to show that the expected post LOCA pool temperature transient is enveloped by the transient used to determine the EEQ environment.

Insert the following in pages 1 through 6 as marked.

Insect B: "and the increased area temperatures during filtration operation."

Insect C: "These temperatures account for the increased area temperatures during filtration operation."

Insert D: "The above postulated condition envelopes the actual expected post DBA pool heat up temperature transient. See section 9.1.3.3 for further information."

8.

Revise the List of Figums as follows:

Figure 9.1-7, delete Title and add " Bulk Pool Transient Temperature Plot (Full Core Offload)"

Figure 9.1-7a, delete Title and add " Bulk Pool Transient Temperature Plot (5mergency Core Offload)"

Figure 9.1-8, delete Title and add "Cooldown Curve for Normal Operation (4 Hours Less of Pool Cooling)"

Ado " Figure 9.1-20, Fuel Assembly Transfer Limit Verses CCP Temperature"

l i

U.S. Nucl:ar Regulatory Commission B17004/ Attachment 5/Page 4 9.

Section 9.1.2.1, page 9.1-2 second paragraph, last sentence, replace "2169" with "3048".

10. Section 9.1.3.1, page 9.1-6, number 13, replace "140 F" with "150 F".

I

11. Section 9.1.3.1, page 9.1-6, number 19, replace "2169" with "3048".
12. Section 9.1.3.1, page 9.1-6, add the following (Insert A) after item number 19.

"20. A full core off-load is designated as a normal evolution."

13. Section 9.1.3.2, page 9.1-6, Delete the first paragraph and second paragraph and replace with the following (Insert B):

"The spent fuel pool cooling system has been analyzed to remove the decay heat load of up to 3048 fuel assemblies and maintain a bulk pool temperature at or below 150 F using a single train of spent fuel pool cooling. A thermal-hydraulic analysis for these bounding heat loads was performed which provided bulk pool temperature curves for three scenarios, a normal full core off-load heat load (Figure 9.1-7), an emergency full core off-load heat load (Figure 9.1-7A) and a normal operation - loss of fuel pool cooling event (Figure 9.1-8).

These curves represent the analysis performed for cooling water to the spent fuel pool cooling heat exchangers (CCP) at 95 F, the upper design temperature limit. For normal and amergency full core off-loads, shorter core off-load times are permitted for lower CCP temperatures as shown in Figure 9.1-20."

"The thermal-hydraulic analysis assumes that outages for full core off-loads will have a minimum duration of 25 days from reactor shutdown to entry into Mode 4 following core reload and that a maximum of 97 fuel assemblies recently discharge will remain in the spent fuel pool. Refueling outages outside these conditions (less than 25 days or greater that 97 assemblies) will require specific calculations to show that the spent fuel pool decay heat levels are less than or equal to 21.1 x E6 BTU /hr."

l

" Cooling for the spent fuel pool consists of two cooling mechanisms. The first is the active cooling provided by the fuel pool heat exchangers. The spent fuel pool water flows from the fuel pool discharge through either of the two fuel pool i

cooling pumps and through the tube side of either fuel pool cooler, and then L

returns to the fuel pool. Table 9.1-2 lists the performance characteristics of the fuel pool cooling system. Cooling for the fuel pool coolers is provided by the l

reactor plant component cooling water system (Section 9.2.2.1). The second mechanism is the passive cooling provide by evaporative cooling from the surface of the pool."

U.S. Nuclear Regulatory Commission B17004/ Attachment 5/Page 5

14. Replace Figures 9.1-7,9.1-7A and 9.1-8 with the attached figures 9.1-7,9.1-7a and 9.1-8.
15. Section 9.1.3.3, page 9.1-8 & 9, following "These are:"in the fifth paragraph, delete subparagraphs "1.", "2." "3.", "4." and " Note 1" and replace with the following (Insert C):
1. Normal full core off-load (maximum bulk pool temperature - 150 F) - the full reactor core (193 assemblies) from the end-of life cycle is off-loaded to the spent fuel pool after one year of operation at full power. The core off-load rate and minimum fuel decay time prior to starting core off-load to the spent fuel pool is dependent on CCP inlet temperature to the Spent Fuel Pool heat exchangers as depicted in Figure 9.1.20.
2. Emergency full core off-load (maximum bulk pool temperature - 150 F) -

the full reactor core (193 assemblies) from the end of life cycle is off-loaded to the spent fuel pool after a previous outage lasting for 10 days with 36 days of operation at full power. The heat load to the spent fuel pool is fully bounded by the heat load for a normal full core off-load. The core off-load rate and rninimum fuel decay time prior to starting core off-load to the spent fuel pool is dependent on CCP inlet temperature to the Spent Fuel Pool heat exchangers as depicted in Figure 9.1.20.

I

3. Normal Operation / Loss of Pool Cooling - End of life core in the reactor i

vessel, the latest refueling load (97 assemblies) is in the spent fuel pool with 25 days (600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br />) of decay time. Following a design basis accident with loss of power, cooling to the spent fuel pool is lost for four hours due to dieselloading concerns before it is restored. Cooling to the pool is limited to the evaporative heat loss. Cooling water temperature prior to loss of cooling is assumed to be at the bounding limit of 95 F.

16. Section 9.1.3.3, page 9.1-9, delete second paragraph ("The decay heat load.."),

and replace with the following (Insert D):

"The decay heat in the spent fuel pool is the combination of decay heat from the previously discharged fuel assemblies and the decay heat from the most recently discharged fuel assemblies. The decay heat load for cycle 1 through cycle 5 discharged fuel was modeled on historical fuel discharge data. The projected fuel discharges for cycles 6 through the end of plant life are conservatively modeled at a bounding average batch burnup of 60,000 Mwd /MtU. The projected number of fuel assemblies discharged is conservative in that the most limiting scenario (yielding the largest discharge) was used. The most limiting scenario selected was a half core loading, consisting of alternating fresh fuel batches of 97 and 96 fuel assemblies per cycle. This resulted in discharging 1960 Millstone Unit 3 fuel

U.S. Nuclear Regulatory Commission B17004/ Attachment 5/Page 6 assemblies at the end of plant life (including the final full core discharge). In addition to the fuel discharged from MP3 operating cycles, an additional heat load from 1088 MP3 fuel assemblies with 60,000 Mwd /MtU burnup and ten years of decay is assumed to be in the pool to conservatively bound discharges of MP1 and MP2 fuel into the MP3 spent fuel pool. Therefore the total number of fuel assemblies considered in the heat load analysis was 3048 Millstone Unit 3 fuel assemblies."

l "A single active failure of the spent fuel cooling system was evaluated in the thermal-hydraulic analysis. The failure was assumed to disable the active train of l

cooling and 30 minutes was assumed in the T/H model to put the standby train into service. The T/H modelindicates that with the pool temperature at 150 F, the spent fuel pool bulk temperature would increase to approximately 155 F before cooling was restored.

17. Section 9.1.3.3, page 9.1-9, third paragraph ("Following a design basis.. "), delete from to the third sentence ("If possible, pool cooling.. ") to the end of the paragraph and insert the following (Insert E):

"The loss of cooling to the spent fuel pool was evaluated for normal plant operation in Case 3 above where pool temperature rose to a maximum of 148.8 F. An j

additional analysis, which is outside the design basis of the spent fuel pool cooling system, was conducted as an input to the structural analysis of the spent fuel pool for the loss of spent fuel pool cooling during a full core off-load. The most limiting case occurs when the pool is at 150 F with the highest heat load in the pool. For this case, a pool temperature of 200 F would be reached after 4.41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br />. For all cases of a loss of pool cooling, the pool temperature is maintained below 200 F when pool cooling is restored after four hours."

18. Table 9.1-2, Replace table with the attached table.

)

19. Add Figure 9.1-20

(

l

20. Section 9.4.2.1, Table in item 5,.ename sub-items "a", "b" and "c" to read "b", "c" and "d". Insert F as new sub-item "a".

Insert F: "a. Pool water temperature 98 F 85 F is greater than 140 F"

U.S. Nucle:r Regulatory Commission B17004/ Attachment 5/Page 7 TABLE 9.1-2 PERFORMANCE CHARACTERISTICS OF THE FUEL POOL COOLING SYSTEM (ONE FUEL POOL COOLER OPERATING)

Operating Full Core Offload Emergency Full Core Normal Plant Condition Offload Operation 6

6 Heat Load 95 F CCP 36.08x10 Bounded by Full Core 21.1x10 8

Offload BTU /hr 90 F CCP 39.19x10 8

85 F CCP 42.30x10 8

80 F CCP 45.41x10 Required Duty of 95 F CCP 34.20x10e Bounded by Full Core 20.37x106 s

Offload One Fuel Pool 90 F CCP 37.31x10 8

Cooler BTU /hr 85'F CCP 40.42x10 8

80 F CCP 43.53x10 Heat Removal by 1.88x10 1.88x10 0.73x10 6

6 6

Evaporation BTU /hr Maximum 150 F 150 F 127.6'F Temperature Long Term Maximum Peak 155.7 F Bounded by Full Core N/A Temperature Offload Short Term Maximum Peak 200 F in 4.41 hours4.74537e-4 days <br />0.0114 hours <br />6.779101e-5 weeks <br />1.56005e-5 months <br /> Bounded by Full Core 148.8 F - Following Temperature (Condition is outside Offload a DBA with a four (Accidents)

Design Basis) hour loss of pool cooling Design Limits Maximum Long Term Temperature (Structural Requirement):

150 F Maximum Short Term Temperature (Structural Requirement):

200 F Maximum Temperature Loss of Pool Cooling:

155.7'F Flow Rates Tube Side - Fuel Pool Water (SFC):

3,500 gpm Shell Side - Reactor Plant Component Cooling Water (CCP): 1,800 gpm Cooling Water Reactor Plant Component Cooling (CCP) (Design):

95 F Temperatures Service Water (Cools CCP)(Design):

75 F I

i MNPS-3 FSAR

(

TABLE 1.3-15 COMPARISON OF OTHER REACTOR PLANT SYSTEMS Operatina Parameters Systems with Components Millstone 3 North Anna 1 and 2 Fuel Pool Coolina and Purification System (Section 9,1.3)

Fuel Pool Cooling Pumps:

Number 2

2 Design capacity (gpm) 3,500 2,750

- Design total head (f t) 92 80 l

Fuel Pool Coolers:

Number 2

2 Duty per heat 27,700,000 56,800,000 exchanger (Btu /hr)

Fuel pool cooling 3,500 2,750 i

flow (gpm)

Component cooling 1,800 3,350

- flow (gpm)

Number of cores cooled 1 1/3-IfYs ( 10VF F..IAs h u ) 1-1/3 n

Fuel pool temperature,.

W lyo 140 normal ("F)

Reactor Plant Comoonent Coolina System (Section 9.2.2.1)

.sRrctor Plant Component Co: ling Pumps:

Number 3

4 Design capacity (gpm) 8,100 8,000 Design total head (ft) 284 190 1

isusura '

1 of 4 tjovember 1994 l

1 MNPS-3 FSAR i

(

TABLE 1.9-1

SUMMARY

OF DIFFERENCES FROM SRP Specific SRP Summary Description Corresponding SRP Section Acceptance Criteria of Difference

_FSAR Section uo.i 0.1.2 Ill2.e - Evaluation of This evaluation has not 9'1 '2'3 (Rev. 3) lighter load drops at been performed.

maximum heights, 9.1.3 II.1.d (4) - BTP ASB

-Oscay imos.;.T.ra' E O

(Rev.1) 9-2 decay heat removal.

42rr sr. L3"..vlivu Dec.y h..t cea* val 8 s Eesel

- g=0 : :t =r/^r, ^^*

ori 000 M o.4 cred;4 Cr

)

i

-BP f.00 0-2.-

tit ar,h coohay mshel,(

4 red 9.I.1

-B P Ass T-2 9.1.4 111.6 - Evaluation of This evaluation has not (Rev. 2) -

lighter load drops at been perfonned, maximum heights.

i Eco4 (hseJ ea o A/GM.t) 4 comyster ceLs i

92.1 Ill.3.d - Location of No manual valve in si.i (Rev. 2) radiation monitors.

1 series with motor

[,.

operated valve.

1 92.2 (Rev.1)

II.3.e - Loss-of-cool Reactor-coolant-pumps 92.2 ant test for reactor-nave not been tested coolant-pumps.

for the 20-minute time requirement.

(9.4.1 11.4 - Compliance to The chlorine detectors 9.4.1.3 Rev. 2)

Reguhtory Guide 1.95.

are not Seismic Category 1.

9.4.5 11.4 - Protection from The bottoms of the fresh 9.4.5 (Rev. 2) dust accumulation.

air intakes are not all located at least 20 feet above grade elevation.

11.5 - Detection and Control Only normal building 9.4.5.2 of airbome contamina-ventilation is monitored.

tion leakage from the system.

lit.3.b - Tomado No protection of duct-9.4.5.4 protection.

work from negative pressure due to tomado.

1 lisSiws 14 of 16 Novombar 100A l

MNPS-3 FSAR TABLE 1.9-2

~

SRP DIFFERENCES AND JUSTIFICATIONS B.. Justification for difference from SRP Electrical intedocks and load paths prevent any load from being carried over the spent fuel pool with the new fuel handling crane. Spent fuel bridge and hoist only carries fuel assemblies at their normallifting height.

SRP 9.1.3 SRP TITLE:

SPENT FUEL POOL COOLING AND CLEANUP SYSTEM

~A. Actual difference between FSAR and SRP

% secote p4., co k (knA *n ORlGGW2&OCNA:N Se evoh* C**Ih '

($

f

l. Decay heat removalis based on Y!n1.F.eu,; ;;;re,-oM a.r.e;, not BTP ASB 9-2, as required by SRP 9.1.3, Paragraph II.1.d(4).

T use,t 2. ea J '3.

8. Justification for difference from gRPss'A ORIGEN1)and CeN

&#/ * *~ # # # 'I W

caces, ene,p+w c as

.a.n n/a 4 deayhed foJ

l. Decay heat removal analysis is based on $g,,,,," M S S S r e; ; C " ; ; d is; 7,e fu m.. ;. 4 Mi M e---fij ;

= ^ and is therefore more appropriate than the analysis recommended in BTP ASB 9-2.

{-

2.

.r.s.,+

2. 4,I 3

SRP 9.1.4 SRP TITLE:

/, LIGHT LOAD HANDLING SYSTEM (RELATED TO REFUELING)

)

A. Actualpifference between FSAR and SRP SRP 9.1.4, Paragraph 111.6, requires an evaluation of lighter load drops at maximum heights. Th evaluation has not been performed.

B. Justification for difference from SRP Electrical intedocks and load paths prevent canying any load over the spent fuel pool with the new fuel handling crane. Spent fuel bridge and hoist only carry fuel assemblies at their normal

- lifting height.

SRP 9.2.1

.SRP TITLE:

STATION SERVICE WATER SYSTEM (NUCLEAR SERVICE COOLING WATER SYSTEM) -

' A. Actual difference between FSAR and SRP k_

SRP 9.2.1, Paragraph lit.3.d, requires that radiation monitors be located on system discharge, and at components susceptible to the leakage, and that these components can be isol. tad by on

' automatic and one manual valve in series. There are motor-operated vtIves at the inlet and iss.awa 33 of 41 November 1994 l

l t

MNPS-3 FSAR

?,.

TABLE 3.6-3 r;

s

}

l MODER ATE-ENERGY SYSTEMS OUTSIDE CONTAINMENT REMOTE FROM l

ESSENTIAL SYSTEMS, COMPONENTS, AND STRUCTURES I

i l

Maximum Operatino Conditions l

Pressure System Location

  • Temoerature ('F)

(Psia) l' Fire Protection - Water AB, ABR, FB, 85 108 SB, TA, TB, WDB l

Fuel Pool Cooling and AB,FB 44e 150 130 Purification G:nsrator Hydrogen TB 100 0

cnd Co2 Hydrogen Gas AB,GSA,TB, 90 100 YA Instrument Air AB,ABR,ESB, 120 100 SB, TB, WDB i

Nitrogen Gas AB,ABR,NSP, 90 5

SB l

Office Building Chilled Water 1

Oxygen' Gas i

Primary Grade Water AB, FB, PWH, 45 100 i

ADB,YA Quench Spray ESB,YA 140 160 R:dioactive Solid Waste AB,FB 100 140 R: actor Plant Aerated AB,ESB,FB, 170 61 Dr: ins WDB l

Reactor Plant Gaseous AB 165 104 Drains.

'Rasctor Plant Aerated AB, WDB 120 0

Vents Rsactor Plant Gaseous AB, WDB 120 3

Vents I

j ss>s.ws 2 of 4 April 1997 l-

~... -.

MNPS-3 FSAR

{

Yard Structures

]

Vacuum Primino Pumnhouse (Nonsafety Related!

The vacuum priming pumphouse is a reinforced concrete structure located on top of the outfall structure. The area is approximately 40 ft x 35 ft with a floor O feet-6 inches above grade. A roof with insulation and 4-ply asphalt and gravel is provided at approximately 17 feet above grade.

2.

Miscellaneous Yard Tankage Boron recovery tanks, primary grade water tanks, demineralized water storage tank, refueling water storage tank, boron and waste test tanks, condensate storage tank, condensate surge tank and water treatment storage tanks are located on concrete pads with oil sand cushion O ft-6 inches above grade. The demineralized water storage tank is protected by 2 ft-O inch thick reinforced concrete walls and roof. The boron recovery tanks are enc!osed in a concrete and steel structure.

3.

Electrical / Conduit Manholes Electrical manholes are reinforced concrete structures constructed below grade with access through manhole covers at grade.

All other nonsafety related structures are located such that their failure does not damage safety related systems, structures, or components.

i 3.8.4.2 Applicable Codes, Standards, and Specifications Codes, standards, specifications, and NRC regulatory guides used in establishing design methods and material properties for Seismic Category I concrete and steel structures other than the containment are given in Section 3.8.1.2.

3.8.4.3 Loads and Loading Combinations All Seismic Category I structures other than the containment structure mat, shell, and dome are designed for the loads and load combinations in Table 3.8-3. Section 3.8.4.5.

describes allowablo stress levels.

For the spent fuel pool, the effects of loads imparted to the structure by the spent fuel racks as well as the effects of hydrostatic and seismically induced hydrodynamic loads are I658 considered in the design. O.c eper.t Sel pes' scella er.d ine; ese e'ee des,gned fe O,e...el d e!!m.i. t-d en tecupereR;ea ir.d:ceted ir. I'swee 3.300 and 3.0-03. The spent fuel pool 4%

walls and mat were also investigated for the revised thermal transient effects due to the Gl%f storage of higher enrichment fuels as shown in Figure 3.8-81. Utilizing the loads and load combinations in Table 3.8-3, the allowable stress levels described in Section 3.8.4.5 were satisfied.

1 sse wa 3.8-40 November 1997

Millstone Unit 3 Fuel Pool Temperature Transients 220 l

200

.200*E P_eak:Euil_ Core Offl ad Maximum _w/L,oss.offool_ Cool'ng I

l

\\

i\\

.eration w/ Loss of Pool Cooling

,1 \\

200*F Peak - Normal Op!

l I

---I 180 t

c l

\\

s

.g l---

--k-y--

y 160 I

150*F - Full Core Offload Steadv State Maximum E

148.8'F Peak -4 hour E 140

, Loss {of Pool Cooling

',- Normal Operation

\\

I

\\

127.6*F - Steady State Maximum Normal Operation

[

120 l

I 100

+

80 0

10 20 30 40 50 60 t

Time - Hours FSAR Figure 3.8.82 Fuel Pool Temperature Transients - Full Core Offload

MNPS-3 FSAR

'[

APPENDIX 3B ENVIRONMENTAL DESIGN CONDITIONS Appendix E Page 1 of 10 FUEL BUILDING -

. la Bevation 11 ft-O in. (Pipe Tunnel Area)

~

tl/$5 l ZQDg: FB-01 See Attachment 2, Appendix E, (Note 5) Page 1 of 2 Normal Envirar, erit (40-vr life)

Temoerature (Note 1)

M IN Range: 65-NMA: 85'F 113 of 13 llD 10 F MNE:

F 104*F I 8 hr I 10 cycles /yr

,+

ec MAE:

F 104*F 1 4 hr 1 1 cycle /40 yr Pressure: Atmosphere Relative Humiditv: 10-100 %

Radiation Dose trads) 40-vr life: 6.0 x.10' One Time Accident Environment due to Seoarate Events:

.(A)

DBA Event (Note 1) 120*F Temoerature:

_85'F I48 hr 1 Pressure: Atmosphere Relative Humiditv: 100% for 48 hr (B)

HELB vent (Note 2)

Max. Temp: 105'F (Note 4)

Max. Pressure: 16.3 psia Relative Humidity: 100% for 48 hr Profile FB-01, See Appendix E, Page 7 of 10 (C)

Fuel Handling Accident (Note 3)-

ly Accident Radiation Dose trads): 5.3 x 10'

(*)

Radiation Dose trads)- 40-vr life plus accident: 5.9 x 10" l Arraamra 3B-86 July 1993

MNPS-3 FSAR

(

APPENDIX 3B ENVIRONMENTAL DESIGN CONDITIONS Appendix E Page 2 of 10 FUEL BUILDING Zgag: FB-02 Elevation 24 ft-6 in. (G-H/50.6-52.8)

(Note 5)

Fuel Pool Cooling Pump Area Spent j 4116 13 Fuel Pool Cooler and Decontamination Rooms See Attachment 2, Appendix E, Pages 1 and 2 of 2 Normal Environment (40-vr life) 5 tJS BW 0 Temoerature (Note 1 11 3 Range:

65 - 12 F

NMA: 85'F 03 MNE:

  • F 104*F l8hr l

10 cycles /yr 13 IN lt's g op MAE:

F 104*F l 4 hr l

1 cycle /40 yr Pressure: Atmosphere Relative Humiditv: 10 -100%

~

Radiation Dose (rads) 40-vr life: 6.0 x 108 One Time Accident Environments due to Seoarate Events:

-(A)

DBA Event (Note 1) 120*F Temperature:

85'F ] 48 hr l Pressure: Atmosphere

)

Relative Humidity: 100% for 48 hr (B)

HELB Event (Note 2)

Max. Temp: 117'F (Note 4)

Max. Pressure: 16.3 psia j

Relative Humidity: 100% for 48 hr i

Profile FB-02, See Appendix E, Page 8 of 10 (C)

Fuel Handling Accident (Note 3)

Accident Radiation Dose (rad.s_1: 5.3 x 10*

I3 A

Radiation Dose (rads) vr life plus accident: 5.9 x 10' 3B-87 July 1993 l Aress.urs

MNPS-3 FSAR f

APPENDIX 3B ENVIRONMENTAL DESIGN CONDITIONS Appendix E FUEL BUILD!NG -

Elevation 24 ft-6 in.

Page 3 of 10 g3 Zgn.g: FB-04 New Fuel Pool and Receiving Areas g5 (Note 5) (G-H/52.8-53.8)

See Attachment 2, Appendix E, Pages 1 and 2 of 2 Normal Environment (40-vr lifel Temnerature (These temperatures are based on a maximu spent fuel pool temperature ofF

  • F) g 13 Range:

65 - 2 'F ggo NMA: 85 *F g.3 nSe LG MNE:h'F dih*F 104'F l8hr]

10 cycles /yr b

t inh *F MAE:

  • F 104*F l 4 hr l__

1 cycle /40 yr Pressure: Atmosphere Relative Humiditv: 10-100%

Radiation Dose trads) 40-vr life: 6.0 x 10' One Time Accident Environments due to Seoarate Events:

(A)

DBA Event (Note 1) 120*F Temoerature:

85'F l 48 hr{

Eressure: Atmosphere Belative Humiditv: 100% for 48 hr (B)

HELB Event (Note 2)

Max. Temp: 155'F (Note 4) '

Max. Pressure: 16.3 psia Relative Humidity: 100% for 48 hr Profile FB-04, See Appendix E, Page 10 of 10

\\

(C)

Fuel Handling Accident (Note 3) 3 Accident Radiation Dose trads): 5.3 x 10' Radiation Dose (rads) 40-vr life olus accident: 5.9 x 10' l was.wa 3B-88 July 1993

i MNPS-3 FSAR j

APPENDlX 3B j

ENVIRONMENTAL DESIGN CONDITIONS i

Appendix E Page 4 of 10 FUEL BUILDING -

Elevation 52 ft-4 in, and above ig Z2Gg: FB-03 See Attachment 2, Appendix E, Page 2 of 2 (4/sy)

(Note 5)

Normal Environment (40-vr lifm Temoerature (These temperatures are based on a maximum spent fuel pool temperature of

'F) ypyg7 Range:

65-

"F 3

NMA: 90'F lif-

[t$)

lo6 10b F

MNE:h F 956F l8 hr l

10 cycles /yr MDF b

MAE:

'F 95*F l 4 hr l

1 cycle /40 yr l

Pressure: Atmosphere L

Relative Humiditv: 10-100%

a Radiation Dose (rads) 40-vr life: 6.0 x 108 One Time Accident Environments due to Seoarate Events:

(A)

DBA Event (Note 1)

_120*F Temoerature:

85'F 148hrl Pressure: Atmosphere Relative Humiditv: 100% fgr 48 hr 1

(B)

HELB Event (Note 2)

AJnsa i

91 %

j Max. Temp: 125*F(Note 4) l Max. Pressure: 16.3 psia l

Relative Humidity: 100% for 48 hr j

Profile FB-03, See Appendix E, Page 9 of 10 (C)

Fuel Handling Accident (Note 3)

Accident Radiation Dose (rads): 5.3 x 10' (IN Radiation Dose trads) vr life ofus accidgat: 5.9 x 10' l

Ad w i

APnam n 38-89 July 1993 j ghg f-

MNPS-3 FSAR APPENDIX 3B ENVIRONMENTAL DESIGN CONDITIONS Appendix E Page 5 of 10 FUEL BUILDING -

Eievation 24 ft-6 in. Containment Tool og Zp_n1g: FB-05 Storage Room (G.6-H/50.6-51.2) g (glW)l (Note 5) See Attachment 2, Appendix E, Page 1 of 2 Normal Environment (40-vr life) f,956N Temoerature: (Note 1 Range:

65 - Q'F (O)l NMA: 85'F 12.3 gg3 F

MNE:

  • F 104*F l 8 hr l

10 cycles /yr I

d.S tS Qop MAE:h F 104*F l 4 hr l

1 cycle /40 yr Pressure: Atmosphere Relative Humiditv: 10-100 %

)

Radiation Dose (rads) 40-vr life: 6.0 x 108 1)* +4 (g/qq) One Time Accident Environment due to Seoarate Events: (Note 6)

Adam M/% (A)

DBA Event (Note 1) 120*F Temoerature:

85*F l 48 hr ]

Pressure: Atmosphere Relative Humiditv: 100% for 48 hr l

~ (4)+ 44 na m:.

(B)

HELB Event (Not Applicable: Mild Environment)

Hl%

-(C)

Fuel Handling Accident (Note 3)

Accident Radiation Dose trads): < 1.0 x 103 Egjiat;on Dose (rads) vr life ofus accident: < 7.0 x 108 60 i

Atm.s q79 APP 38.MP3 3B-90 June 1994

ww u MNPS-3 FSAR y

i APPENDIX 3B ENVIRONMENTAL DESIGN CONDITIONS l

Appendix E Page 6 of 10 4

Mples on Fuel Buildina Environments 1.

Following a postulated DBA, if component cooling water is not supp 20 ours. This results in a one-time accident con e to the fuel g

shown. This assumes initial pool temperature of 150*F a

' y e

3/88 cooldown time subsequent to component cooling reinitiation.I Due to High Energy Line Break (HELB) of hot 2

3.

with or in close proximity to spent fuel assemblies. Fue ent in contact 4.

the increases to Tmax and Pmax. Equipment must be thermally s prior to

~

5.

based upon specific equipment locations and ope ex codes 6.

Zone FB-05 boundaries have been modified to provide an HELB seal this zone is considered a mild environment.shown are not sig

. Parameters W

)

s e

i Aeras ues 3B-91 June 1994

1 MNPS-3 FSAR CHAPTER 9 LIST OF FIGURES FIGURE NUMBER TITLE 9.1-1 PWR Spent Fuel Racks 9.1-2 Top View of 6 x 6 Rack Array 9.1-2A Region 1, Three of Four Spent Fuel Assembly Loading Schematic for A Typical EXE Storage Module S.1-3 Side View of 6 x 6 Rack Array 9.1-5 Adjustable Fuel Rack Leveling Pad 9.1-6 Fuel Pool Cooling and Purification System 9.1-7

- N;:~ e! " !ue"~2 _ Bolk Po.t Transred Temperdwe fl*I (Full co<e OKload) 9.1 -7A

-Ln.;u;Cyde c !! Ce'e O'See? Bulk Po.l TM-MT'ewde Pld CEmce<y e

Co,e off/.t) 9.1-8 C;2cr :=c; C^'= Of' Lc;L Co. tow, c g,, y,,,,I ope,,hm ( y he,, f,,,,f 9.1-9 Refueling Machine 9.1-10

. Spent Fuel Bridge and Hoisting Structure 9.1-11 New Fuel Bevator 9.1-12 Fuel Transfer System 9.1-13 Fuel Transfer System 9.1-14 Spent Fuel Handling Tool 9.1-15 New Fuel Handling Tool 9.1-16 Reactor Internals Lifting Device 9.1-17 Ouick Acting Stud Tensioner 9.1-18 New Fuel Pool Layout A@

9 1-19 New Fuel Rack Elevation View 4

9.1-20 Fuel Asse 4It TransL L:,;4 Ve<ses CCP Ten,,,e,,Le 9.2-1 Service Water System l

l eroc.MP3 9-viii March 1998

. ~.. -._

~

I MNPS 3 FSAR lb The racks are designated ANS Safety Class 3 and Seismic Category I and are designed to withstand normal and postulated dead loads, live loads, and loads caused by the operating basis earthquakes and safe shutdown earthquake events.

The design of the racks is such that K,,, remains less than or equal to 0.95 under all condi-tions, including fuel-handling accidents and the optimum moderation configuration. Due ;o the use of fuel barriers and the close spacing of the cells, it is impossible to insert a fuel

, assembly in other than design locations or between the rack periphery and the pool wall.

The racks are also designed with adequate energy absorption capabilities to withstand the impact of a dropped fuel assembly f om the maximum lift height of 5 feet over the top of r

l the racks. The fuel storage racks can withstand an uplift force equal to 2000 pounds.

All materials used in construction are compatible with the fuel building / vault environment a-all surfaces that come into contact with the fuel assemblies are made of annealed at,5tenitic stainless steel. All the materials are corrosion resistant and do not contaminate the fuel'ssemblies or vault environment.

a 9.1.2 Spent Fuel Storage 9.1.2.1 Design Bases The spent fuel pool, located in the fuel building,is designed to accommodate fuel racks (Figure 9.1-1) that store spent fuel assemblies. At the time of initial operation, installed capacity was at least one and one-third cores.

The spent fuel is stored in racks which are located under water in the spent fuel pool.

There are 756 fuel storage locations in 21 storage racks. Each rack consists of cells welded to a grid base and welded together at the top through an upper grid to form an j

integral structure (Figure 9.1-1). The vertical corners of adjacent cells are also welded N.n g together to form an integral structure. The spent fuel pool has the heat load design

> G a. )

capacity for G%G fuel assemblies (Section 9.1.3).

30't t l

The rack arrays (Figures 9.1-2 and 9.1-3) have a center-to-center spacing of 10.35 inches.

Each storage cellincorporates a neutron absorber and is composed of boron carbide in a homogeneous stable matrix. This materialis encapsulated in stainless steel for support but 1

is not sealed as it is compatible with the pool environment. The spacing and the design a the racks are such that there is a 95 percent probability that the effective multiplication factor, including uncertainties, does not exceed 0.95 at a 95 percent confidence level.

h@'l-lo9 For the storage of fuel assemblies with nominal enrichment levels between 3.85 and l

5.0 w/o Um, a regionalized fuel storage / pool configuration is implemented as follows:

High enriched, low (or no) burnup fuel is stored in Region I in a 3-out-of-4 array L

with the fourth storage location blocked. Up to a maximum of 100 storage locations will be blocked as shown in Figure 9.12A.

Low enriched, high burnup fuelis stored in Region 11 in a 4-out-of-4 array due to reactivity credit for burnup being taken into account as permitted by NRC Regula-tory Guide 1.13.

I oss un 9.1 2 April 1997

=c

,u

~

i 9, j, 3, l ( c on-f a v

  • d I60 13.

The temperature of the fuel pool water is maintained at or below MO8F for the normal operating condition of the spent fuel pool.

14.

tilM)

The temperature of the fuel pool water should not exceed a maximum temperature of 150*F for any long-term period. The rnaximum peak temper-ature the spent fuel pool water can reach is 2OO*F.

15.

Purity and clarity of the refueling cavity and fuel pool water is maintained to permit observation of fuel assembly handling during refueling operations.

16.

Filtration and ion exchange capability are pro.nded to remove suspended and dissolved radionuclides to allow access to required areas.

17.

The fuel pool cooling system and the service water makeup lines are safety related, Seismic Category 1, and are designated SC-3 and designed to the requirements of ASME Ill, Class 3.

18.

The purification system is not safety related and is designated nonnuclear safety (NNS).

c3 O W pq:)

19.

The spent fue. Dool cooling and, purification system is capable of handling the Ms/

accumulated oecay heat from 4MSspent fuel assemblies, which includes a l

full core offload of 193 fresh fuel assemblies.

3 N Gc RT "A" 9.1.3.2

System Description

T N s s e T-

B '

1.u) Th3 spent fuel pool has been designed to hnid T tc 2100 luci assemblies. A heat load c;lculation using this 1 performed which provided decay heat load curves for

[ lthree see

, o normal refueling heat load (Figure 9.1-7), end-of-cycle full-core off-load (Fi

.1-7A), and an emergency full core offload heat load (Figure 9.1-8).

J Th3 spent fuel pool water flows from the fuel pool discharge throimb =1theroTthe two fuel) 6 pool cooling pumps and through the tubge sid"th.-iUET podi~ cooler, and then returns to theduel pool. Table 9.1-

- tN p1Tormance characteristics of the fuel pool cooling o

gy; tem.

r the fuel pool coolers is provided by the reactor plant component ing water system (Section 9.2.2.1).

Th3 purification system consists of two purification pumps, two purification prefilters, one coarse filter, one purification demineralizer, and one postfilter. This equipment is not saf ty related.

The purification system provides means for filtering and demineralizing the following areas:

1.

The fuel pool water to improve optical clarity for ease of underwater fuel handling and to reduce radioactive contamination in the water 2.

The reactor cavity water during a refueling operation to improve optical clarity for ease of underwater fuel handling, and to reduce radioactive contamination in the water est en 9.1-6 July 1996

MNPS-3 FSAR crrangement be needed, a temporary flanged spool piece can be inserted in the line to cnable one of the fuel pool purification pumps to pump the water within the spent fuel shipping cask storage area either through the prefilters or through the prefilters.

demineralizer, and postfilter to the boron recovery tanks (Section 9.3.5). Administrative,

procedures are followed to assure that the' cask storage area gate is inserted in the transfer slot in the wall separating the fuel pool from the spent fuel shipping cask storage area before pumping commences. However, the design of the gate is such that even with the gate open, the fuel pool cannot be drained below the top of the active fuel region of the fuel assemblies.

Piping, valves, and components of this system making contact with the fuel pool water are custenitic stainless steel which is corrosion-resistant to the boric acid solution.

A sample connection is provided downstream of the fuel pool demineralizer for sample removal to check the gross activity, particulate matter, boric acid concentration, and component performance.

9.1.3.3 Safety Evaluation Two full-size fuel pool cooling pumps and two full-size fuel pool coolers are provided to ensure 100-percent redundant cooling capacity. This portion of the system is Seismic Category I and Safety Class 3. The Seismic Category I cooling portion of the fuel pool cooling and purification system is independent of the nonseismic purification portion. Fa-ilure of the purification portion in an earthquake does not affect the operation of the cooling trains.

Erch pipe which enters the fuel pool has either a 1/2 inch vent hole drilled into the pipe to cet as an anti-siphoning device or terminates at an elevation above these vent holes.

These provisions prevent siphoning of the fuel pool water to uncover the spent fuel (see Figure 9.1-6).

One pump and one cooler are sufficient to rnaintain the pool temperatures as indicated in Ttble 9.1-2.

An evaluation of the capabilities of the spent fuel pool cooling system has been performed l for normal and abnormal conditions, gg.g The decay heat loads were calculated for a number of pool operating conditions at the end GI41.

of pool life. These are:

g NSERT C.

%~N al refueling (2.177 x 10' Btu /hr, Figure 9.1-7)--one refueling loa

.f (1/2 c 6 assemblies) transferred from the reactor vess e spent fuel pool startin en days after shutdown.

kq 2.

Normal plant operation (1.859 r, Figure 9.1-7 at 600 hrs)-one refueling load (1/2 cor le spent fuel po r 25 days of decay time.

This conserv y models the decay heat load in the fuel pool during ant operation (reactor coolant system operating, turbi line).

/Q su wa 9.1-8 May 1997

MNPS-3 FSAR End-of-cycle full-core of f-load (3.479 x 10' Btu /hr, Figure 9.1-7Al--on i

ore (193 assemblies) off-load after one year of power operati oto 1).

Fue sfer from the reactor starts eleven days after own.

4.

Emergency full-co f-load (3.505 x 10' r, Figure 9.1-8F-a full-core h.2.a.)

of f-load commencing 3'ht att e start of the previous refueling (note 1). This corresponde a

normal maximum heat load" as defined in SRP 9.1.3 and i, as the decay load for all full-core off-loads occurring I ian one year after a previous

. ling shutdown.

ote 1:

A rnaximum of six full-core off-loads are ass -

to occur over the life of the plant (Conditions 3 and 4 combi To determine the decay heat load, the fuel bundles were analyzed as being transferred to the spent fuel pool at an average rate of three bundles per hour over the time it takes to fqi,,p) off-load the assuraed fuelload.

g gg 3 NssrT t)

L heat load for the older spent fuelin the pool was modeled as one refu o

batch 1 year refueling batch at 3 years old, and the remaini ing batches at 18-month intervals for e Qnd 3 above. For Cae e older spent fuel was modeled as one refueling batch 18 mb'rith t e remaining refueling batches at 18 month intervals over the life pool. The

- Id and fresh fuel heat loads were added to this basel' oad. For conservatism, the fuel es used were e

consistmu-witfi'3~4-month operating cycles.

Following a design basis accident with loss of power, the reactor plant component cooling water system is

. Oailable to cool the spent fuel pool coolers until approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after ib secdent at which time cooling will be restored. Power from the emer-g gg I d%

4 g ency generators is not immediately available due to loading considerations.l Hgwaver;'iii h+[b

y thTtvent.Qol cooling is not available at this time, a loss of coolin.

son has been performed whic that it would take at least 1 ore the spent fuel pool would reach its design tempe is provides sufficient time to initiate poo cooling. Once the coolinc ed, the te e decreases to 150 Fin less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Thie io forms one of the bases for the qua i f the pool structure

.8.4.3) and conservatively envelops expected pool transiants.

9.q For a safety grade cold shutdown with a normal plant operation spent fuel pool heat load up to and including Plant Cycle 20, the temperature of the fuel pool water is maintained at or below 140*F. Plant Cycles 20 through the end of plant life could result in short-term gg.qs fuel pool water temperature transients above 140*F, but less than 150 F. This condition SMg results from higher reactor plant component cooling water supply temperature of approxi-mately 113'F which is permitted during a safety grade cold shutdown (see Section 9.2.2.1.2).

Redundant safety grade fuel pool temperature indication is provided on the main control g

board. Redundant safety class 3 level switches are connected to the fuel pool which alarm U-37 in the control room. They are set to provide indication before the water level f alls below dD 23 feet above the top of the fuel racks. Piping penetration are at least 11 feet above the top of the spent fuel so that failure of inlets, outlets or accidental piping leaks cannot reduce the water below this level, m em 9.1-9 March 1998

i

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200 400 GOO 800 1000 TIME AFTER REACTOR SHUT 00WN,HR Je le.+e oJ Re p l* *c l

l.

FIGURE 9.1-7 NORMAL REFUELING MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT

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200 400 600 800 1000 TIME AFTER REACTOR SHUTDOWN,HR D e k -le a n d Ret "e I

FIGUF.E 9.1-7A END OF CYCLE FULL CORE OFFLOAD MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT

.TiltK 199:

150 N\\

145

\\

140 135 130 1

S N 8.125 N

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H g120-ca 115 110 105 100 0

200 400 600 800 1000 1200 1400 1600 1800 Time After Previous Outage Shutdown [ Hrs]

BULKPOOLTRANSIENTTEMPERATURE PLOT (En1ERGEucY CoAe oFFLon0 1

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O 500 1000 1500 2000 2500 TIME AFTER NORMAL REFEULING SHUTDOWN, HR he} ac(

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C-QA FIGURE 9.1-8 EMERGENCY CORE OFFLOAO MILLSTONE NUCLEAR POWER STATION UNIT 3 FINAL SAFETY ANALYSIS REPORT JUNE 19

i l'

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t 125 120 600 620 640 660 680 700 720 740 760 780 Time After Reactor Shutdown [ Hrs]

COOLDOWN CURVE FOR Nogin4L OrgAArtav h-HOUR LOSS l

OF POOL COOLING)

Fg u re i

m

. m m.=.

m-

~..

m m ----

Figure 9.1-20 Fuel Assembly Transfer Limit Verses CCP Temperature 200 190

--4--?-

-?--4-- ?--:

-F-----

180 r-----,---'

170 r-

_8 Fuel Movement

_ ;_ _ __.______[________._.___i__..___,__

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iso S

Not Permitted l

-- ~---- ----

g 150 m

p 140 CCP-c-

,E. 130 Temperature 1----.

s80*F

_.CCP___L_.

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____l___<__,._

jE 120 Temperature 110 6

s 85 F,'
CCP;

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g 100

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Tem' erature PJ-p

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- $ 95 E

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90 100110120130140150160170180190 200 210 220 230 240 250 260 270 280 290 300 310 320 330 J40 350 360 TIME AFTER REACTOR SHUTDOWN (HOURS)

MILLSTONE UNIT 3 FULLCORE OFFLOAD

f l

I l

i i

MNPS-3 FSAR

~

l Rep

TABLE 9.1-2 PERFORMANCE CHARACTERISTICS OF THE FUEL POOL COOLING SYSTEM (ONE FUEL POOL COOLER OPERATING)

Emergene uit Core End-of-Cycle Full-Off-loa i

Normal Refueling Normal Plant Core Off load (Ab mal Maximum Op: rating (Fuel Shuffle)

Operation (6 events w/cond 4) 6 ents w/cond 3)

Ccndition Cond.1 Cond.2 Cond.3 ond.4 i

Hrt Load 2.177 x 10' 1.859 x 10' 3.479 x 10' 3.505 x 10' BTU /hr R: quired Duty of 2.177 x 10' 1.859 x 10' 3.479 x 10' 3.505 x 10'

!Jne Fuel Pool Cool;r BTU /hr Maximum 130'F 125'F a0*F 150'F Tsmosrature Lcng Term MIximum Peak 140'F 140*F 155'F 150*F (No Failures i

Tamp 3rature Assumed) j Short Term Maximum Peak 140*F 1

'F 155'F 150'F (No Failures Tempirature Assumed)

(Accidents)

Design Limits Maximum Lo Term Temperature (Structural Requirement): 150'F Maximum hort Term Temperature (Structural Requirement): 200'F Maxi m Temperature Loss of Pool Cooling (Cond.1-3): 155'F l

Flow Rates Tu Side Fuel Pool Water (SFCl: 3,500 opm

. eli Side - Reactor Plant Component Cooling Water (CCP): 1,800 ppm Colling Water Reactor Plant Componer!t Cooling (CCP):' 95'F Temperatur Service Water Temperature (Cools CCP): 75'F Not Conditions (Cond.) refer to descriptions in Section 9.1.3.3.

t i

ssi 2mn 1 of 1 April 1997

MNPS-3 FSAR vibration monitor (common).

o Power not available status lights are provided on the rear of the main control board for

cach motor control center.

All radiation monitor alarms annunciate in the control room.

Fuel Building Venti!ation System 9.4.2 The fuel building ventilation system (Figure 9.4 2) removes heat generated by equipment cnd water vapor from fuel pool evaporation, prevents moisture condensation on interior walls, provides a suitable environment for equipment operation and personnel. It also limits potential radioactive release to the atmosphere during normal operation or anticipated cperational transients, and following a postulated fuel handling accident (FHA).

9.4.2.1 Design Bases The fuel building ventilation system is designed in accordance with the following criteria.

1.

General Design Criterion 2. as related to the system being capable of withstanding the effects of earthquakes.

2.

General Design Criterion 5, as related to shared systems and components impor-tant to safety.

3.

General Design Criterion 60 and Regulatory Guides 1.52 and 1.140, for design testing and maintenance criteria for atmosphere cleanup systems.

4.

General Design Criterion 61 and Regulatory Guide 1,13, for fuel storage and radioactive control.

5.

Outdoor air design temperatures are listed under design weather data in Section 9.4. The fuel building ventilation system is designed to maintain the following ei.y space temperatures during normal operation.

Mh Maximum Minimum Space Space Soent Fuel Pool Area Temperature Temoerature

-d "M'g ----) u.

P<.,

. w., + e t e e ruo t< <

y9 af:

WScp L p. Pooi wa't'eYiemydrat$r'e 95*F 85*F i

is greater than 100*F c p. Pool water temperature 95'F 65*

is less than 100*F d p. All other areas 104*F 65*F 6.

Air flow is directed from areas of lower potential radioactivity to areas of higher potential radioactivity.

esa uP3 9.4-10 February 1995

i Docket No. 50-423 B17004 l

l l

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1 l

Millstone Nuclear Power Station Unit No. 3

. License Amendment Request and Technical Specification Changes For Full Core Off-load Retvoed Technical Specification Pane (s) l i

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l January 1999 l

TABLE 3.7-6 (Continued)

AREA TEMPERATURE MONITORING 8EE8 TEMPERATURE LIMIT (*F)

7. FUEL BUILDING FB-02, Fuel Pool Pump Cubicles, El 24'6" 1 119 FB-03, General Area, El 52'4" 1 108
8. FUEL OIL VAULT FV-01, Diesel Fuel Oil Vault 5 95 9.

HYDROGEN RECOMBINER BUILDING HR-01, Recombiner Skid Area, El 24'6" 1 125

' HR-02, Controls Area, El 24'6" S 110

'HR-03, Sampling Area, El 24'6" s 110 HR-04, HVAC Area, El 37'6" s 110

10. MAIN STEAM VALVE BUILDING MS-01, Areas above El. 58'0" s 140 i

MS-02, Areas below El. 58'0" 1 140 1

11. TURBINE BUILDING TB-01, Entire Building s 115
12. TUNNEL TN-02, Pipe Tunnel-Auxiliary, Feel and l

ESF Building s 112 4

13. 188Q YD-01, Yard s 115 i

MILLSTONE - UNIT 3 3/4 7-35 Amendment No. pg, Jpp, 0619

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