ML18026A260

Forwards Response to 931123 RAI Re Personnel Access Questions Resulting from Postulated Loss of Spent Fuel Pool Cooling Events

ML18026A260
Person / Time
Site:	Susquehanna
Issue date:	01/04/1994
From:	Byram R PENNSYLVANIA POWER & LIGHT CO.
To:	Chris Miller Office of Nuclear Reactor Regulation
References
PLA-4069, NUDOCS 9401110318
Download: ML18026A260 (103)

Text

ACCELERATED DI TRIBUTION DEMONSTQWTION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9401110318 DOC.DATE: 94/01/04 NOTARIZED: NO DOCKET FACIL:50-387 Susquehanna Steam Electric Station, Unit 1, Pennsylva 05000387 50-388 Susquehanna Steam Electric Station, Unit 2, Pennsylva 05000388 AUTH. NAME AUTHOR AFFILIATION BYRAM,R.G. Pennsylvania Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION MILLER,C.L. Project Directorate I-2

SUBJECT:

Forwards response to 931123 RAI re personnel access questions resulting from postulated loss of spent fuel pool D cooling events.

DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR Submittal: General Distribution NOTES:

A RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL D PD1-2 LA 1 1 PD1-2 PD 1 1 CLARK,R 2 2 D INTERNAL: ACRS 6 6 NRR/DE/EELB 1 1 NRR/DORS/OTSB 1 1 NRR/DRCH/HICB 1 1 NRR/DRPW 1 1 NRR/DSSA/SPLB 1 1 NRR/DS SA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 1 0 OGC/HDS2 1 0 01 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 R

D D

D NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LIS15 FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED: LTTR 22 ENCL 20

Pennsylvania Power 8 Light Company Two North Ninth Street ~Alientown, PA 18101-1179 ~ 215/774-5151 Robert G. Byram Senior Vice President-Nuclear 21 5/774-7502 PAN y Director of Nuclear Reactor Regulation Attention: Mr. C. L. Miller, Project Director Project Directorate I-2 Division of Reactor Projects U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SUSQUEHANNA STEAM ELECTRIC STATION REQUEST FOR ADDITIONALINFORMATION ON LOSS OF SPENT FUEL POOL COOLING EVENTS Docket Nos. 50-387 PLA-4069 FILE R41-2 and 50-388

Dear Mr. Miller:

Attached is PP8rL's response to your November 23, 1993 Request for Additional Information concerning personnel access questions resulting from postulated loss of spent fuel pool cooling events.

Please contact Mr. James M. Kenny at (215) 774-7914 should your require additional information.

Very truly yours, R.. By Attachment cc:%RCDocumellt=Control-Desk=(original)tt NRC Region I Mr. G. S. Barber, NRC Sr. Resident Inspector - SSES Mr. R. J. Clark, NRC Sr. Project Manager - Rockville 0700tfa

.9401110318 940104

,PDR S., '..PDR ADOCK 05000387

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ATTACHMENT 1 to PLA-4069 ATTACHMENT1 RESPONSE TO QUESTION 1 OF NRC 11/23/93 RAI Page 1

0 h,

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ATTACHMENT 1 to PLA<069 QIIEHTION I Provide a detailed description of the modeling techniques used to generate the 4.22 Rem cited in the August 16, 1993 submittal as the dose received during the operation of the spent fuel pool emergency service water supply valves. Describe in detail the time-motion analysis performed to determine exposure time associated with required operator actions. Include the parameters and assumptions used to generate the contained and airborne source terms and calculate the respective dose components associated with each segment of the time-motion analysis.

RESPONSE 1 The following provides a complete response to question 1 of the November 23, 1993 NRC RAI concerning radiological evaluations for the Loss of SFP Cooling issue. As discussed in PPkL's May 24 and August 16, 1993 submittals, restoration of the normal SFP cooling system prior to boiling is expected. Consequently, PPAL would not expect it to be necessary to use ESW for make-up since the normal systems could be used or make-up from the non-accident unit could be provided. This response is separated into two parts. Section 1.0 provides a description of the modeling techniques and assumptions used for evaluating operator access to the ESW make-up valves. Also included in this section, is a discussion of how the results time-motion study were factored into the analysis. Section 2.0 provides summary tables of the doses (both airborne and contained) for each of the segments assumed in the calculation. These tables show doses for a DBA LOCA with 1% clad damage, 100% clad damage, and 100% fuel melt (i.e., Reg. Guide 1.3 source term).

1.0 Detailed Descri tion of Modelin Techni ues 1.1 Overview The TACT5 computer code is used to evaluate post-LOCA radiation sources inside the reactor building using the FSAR Chapter 15.6.5 DBA-LOCA activity flow path model with realistic estimates of containment leakage rates. Airborne activity concentrations in the reactor building and activity concentrations in the suppression pool water are evaluated for postulated cladding failure (NUREG-1465) and fuel melt (Regulatory Guide 1.3) source terms.

Using these post-LOCA source terms, radiation dose rates inside the reactor building from airborne activity and from suppression pool water contained sources are evaluated using the MICROSHIELD computer code.

Operator access requirements inside the reactor building were determined for establishing ESW makeup to the spent fuel pool under post-LOCA conditions. Operator access routes and missions were identified and divided into sequential segments for the purpose of evaluating operator access Page 2

ATTACHMENT 1 to PLA%069 doses. Operator access doses are computed by multiplying the radiation dose rates from both airborne and contained sources in a given mission segment by the time spent by the operator in that mission segment. The sum of the radiation doses for all segments of the mission provides the total mission dose.

1.2 Detailed Descri tion This analysis evaluates personnel access doses inside the reactor building for the following postulated LOCA initiated core damage cases: 1% clad damage, 100% clad damage, and 100%

fuel melt.

The clad damage cases are evaluated using assumptions consistent with the accident source terms described in NUREG-1465. The fuel melt case is conservatively evaluated using accident source terms consistent with USNRC Regulatory Guide 1.3 and the FSAR Chapter 15.6.5 DBA-LOCA licensing basis evaluation. For all cases, realistic estimates of the containment leakage rate are used.

For the activity flow path model used in this analysis, the activity concentrations inside the nodal volumes are calculated by dividing the activity in the node at the time of interest by the nodal volume and therefore are directly proportional to the activity source term released from the fuel into the containment and suppression pool.

For the postulated LOCA's that assume clad damage, the amount of activity released is directly proportional to the amount of clad damage. Therefore, a complete source term and dose analysis is performed for the 100% clad damage case and source terms and dose results for the 1% clad damage case are obtained by multiplying the 100% clad damage results by 0.01 . NUREG-1465 was used to supply release data for the clad damage cases. Since no specific numerical guidance is provided in NUREG-1465, the amount of clad activity released from the core which becomes airborne inside containment or remains in the suppression pool water is based on taking credit for fission product scrubbing and retention in accordance with Standard Review Plan 6.5.5. In addition, for the clad damage cases, an aerosol removal rate of 0.73 hr 'or particulate iodine and cesium is assumed based on information provided in NUREG-1465, Table 5.6 for the La Salle Nuclear Power Plant which is also a BWR Mark II containment design. No credit for aerosol removal was taken in the 100% fuel melt case.

Core activity release fractions for a LOCA with 100% fuel melt are based on the requirements of Regulatory Guide 1.3 and NUREG-0737 and are that 25% of the iodines and 100% of the noble gases are instantaneously airborne in primary containment and available for leakage and 50% of the core inventory of iodines and 1% of the particulate are released to the suppression pool water. Due to the number of particulate isotopes in the core and the number of dose calculations required, the particulate activity release from the core was not explicitly included in the suppression pool activity source term for the 100% fuel melt case. Instead, bounding dose calculations using post-LOCA suppression pool contained sources with and without particulate Page 3

ATTACHMENT 1 to PlA-4069 were used to determine a dose multiplier which was used to account for the dose contribution from 1% particulate in the suppression pool water. The iodine and noble gas activities released for the 100% fuel melt case are obtained from the FSAR DBA-LOCA analysis given in PP&L calculation FX-C-DAM-014. The bounding analysis for the dose contribution from particulate is based on post-LOCA suppression pool liquid activity source terms given in Susquehanna Project Bechtel Calculation 200-201.

For all clad damage and fuel melt cases, the isotopic chemical form of the activity released from the core is assumed to be as follows:

Io dines = 91% elemental

= 4% organic

= 5% particulate Cesium s = 100% particulate Noble Gases = 100% elemental This assumption is consistent with USNRC Regulatory Guide 1.3 for core fuel damage. For the clad damage cases, NUREG-1465 indicates that the chemical form for iodine entering containment is 95% particulate and 5% elemental, but without Ph control of the suppression pool water, a relatively large fraction of the particulate iodine dissolved in suppression pool water will be converted to elemental iodine. Therefore, the above chemical forms assumed for this analysis are conservative for the clad damage cases.

The primary containment design basis leakage rate is 1 %/day. For the FSAR licensing basis DBA-LOCA analysis this leakage rate is assumed for the duration of the accident. For this evaluation, a time dependent realistic containment leakage rate based on containment Integrated Leakage Rate Testing (ILRT) results and the calculated containment post-LOCA pressure response is used for both the clad damage and 100% fuel melt cases. The ILRT test pressure corresponds to the maximum calculated containment post-LOCA pressure with a design margin applied. The realistic containment leakage rate was calculated by reducing the ILRT measured leakage rate proportionately to the ILRT test pressure and the calculated containment pressure response for a LOCA. For this analysis, the most up-to-date leakage rate data is used so that the dose estimates reflect the most up-to-date containment leakage conditions. Therefore, the measured leakage rate of 0.606%/day from the Unit 1 ILRT performed 5/5/92 is assumed for this analysis. A leakage rate of 0.606%/day is also representative of typical measured leakage rates at SSES Units 1 & 2.

NUREG-0737 provides guidance for evaluating operator access to vital plant areas for post-accident operations. It states under Item (2) Systems Containing The Source that for post-LOCA accident operations, "Radiation from leakage of systems located outside of containment need not be considered for this analysis". Therefore, it is assumed for this evaluation that the leakage of post-LOCA containment airborne activity through containment penetrations that are water sealed Page 4

ATTACHMENT 1 to PLA<069 need not be considered. The realistic containment leakage rate used for both the clad and fuel damage cases are based on the measured leakage rates through containment penetrations that are not water sealed.

Based on the above activity flow pathways, the TACT5 code provides total activity in the reactor building and suppression pool as a function of time post accident. At the required evaluation time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post accident), isotopic activities were taken from the TACT5 output edit and divided by the appropriate dispersal volume to give activity concentrations.

Radiation doses inside the reactor building are evaluated for both" airborne activity and for suppression pool water contained sources. Operator access doses or area radiation dose rates are evaluated in each of the areas that require access to provide ESW makeup. Radiation doses are evaluated using the MICROSHIELD computer code with the TACT5 generated activity concentration source terms.

The dose rate from airborne activity inside the reactor building is actually an immersion dose.

However, since the MICROSHIELD computer code cannot calculate dose rates internal to the source, slab geometry is used. One half of the source volume is modeled as a rectangular volume source and a dose rate on the surface of this volume is calculated. The immersion dose is then calculated by multiplying the contact dose rate from half of the source volume by a factor of 2.

All immersion dose rates are conservatively calculated at the centerline of the room. Dose rates from the contained suppression pool liquid piping sources are calculated using cylindrical source geometry with receiver at side.

In order to evaluate operator access doses, operator walking rates and stair climbing rates are required. A time motion study was performed to verify operator access travel times inside the reactor building under LOCA conditions. An operator was dressed in protective clothing and wore a Self Contained Breathing Apparatus and actual transit times to valves located on elevations 670'nd 749'f the reactor building were measured. Access to these valves is required to provide ESW makeup to the spent fuel pool. Based upon the results of this time motion study, an operator walking rate of 200 ft/min and a stair climbing rate of 50. ft/min are conservatively assumed for operator ingress/egress dose calculations for access to the reactor building under LOCA conditions. Also as part of this study, using spare valves, the valve opening time for the 2 inch valves used for ESW system tie-in and flow control was measured to be 10 to 15 seconds, but for calculational conservatism, a valve opening time of 1 minute was used.

The incorporation of the time-motion study into the final calculation resulted in the preliminary dose to the operator increasing from 4.22 Rem. to 4.57 Rem., for access to the valves on elevations 670. The 4.22 Rem value was reported in PPkL's August 16, 1993 submittal.

Operator access doses are only evaluated for Unit 1 The operator access area locations for

~

Units 1 and 2 are identical except for the area containing the valves required for tie-in of the ESW system for makeup ( Valve Nos. 153500, 153501 and 253500, 253501). Based on the post-accident radiation levels given in Figures 18.1-3 and 18.1-4 of the SSES FSAR, the area containing the Unit 1 valves for ESW system tie-in has higher dose rates than the Unit 2 area.

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ATTACHMENT 1 to PLA-4069 Therefore dose rates calculated for the Unit 1 area will be conservative for the Unit 2 area.

Therefore, all of the calculated operator access doses for Unit 1 are applicable to Unit 2.

Operator access doses are evaluated at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA. For loss of spent fuel pool cooling, the fuel pool begins to boil in approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. For the condition where a Loss of Offsite Power (LOOP) is also postulated, power is expected to be restored within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Therefore, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was chosen as the latest time operator action could be taken to restore cooling and/or make-up to the pull and assure those actions would be completed prior to SFP boiling Some dose rates are also calculated at other times post-LOCA and to show that post-LOCA dose rates are decreasing for time periods greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

For the 100% fuel melt case, a dose factor is used to take into account the dose contribution from particulate in the suppression pool liquid (see Section 5.1, Assumption 5). The particulate dose factor is defined as the ratio of the dose from suppression pool liquid sources containing core iodine and particulate activity to the dose from suppression pool liquid sources containing iodine activity only. All doses from contained sources for the 100% fuel melt case are calculated using the MICROSHIELD computer code with a suppression pool source term that contains iodines only. The MICROSHIELD results for contained sources for the 100% fuel melt case are then multiplied by the particulate dose factor to take into account the dose contribution from particulate. This dose factor is calculated at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA and can only be applied to operator access doses from suppression pool liquid sources evaluated at this same time period post-accident.

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ATTACHMENT 1 to PLA-4069 2.0 Summ of 0 erator Doses for ESW Make-u Valves TABLE 2.1

SUMMARY

OF CALCULATEDOPERATOR ACCESS DOSES TO PROVIDE ESW MAKEUP TO SPENT FUEL POOL-

- REACTOR BUILDINGUNIT 1 ACCESS TO VALVES 153500 AND 153501 ELEV.

670'PERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 100% CLAD DAMAGE 1% CLAD DAMAGE (1) 100% FUEL MELT DOSE LOCATION/

SOURCE Dose From Dose From Dose From Dose From Dose From Dose From Airborne Contained Airborne Contained Airborne Contained Activity Sources Activity Sources Activity Sources Ingress/Egress To Valves 7.22-3 0.476 7.22-5 4.76-3 0.198 4.12 153500 & 153501 Operator Stay Time At Valves 7.45-3 4.21-3 7.45-5 4.21-5 0.204 0.0476 153500 Ec 153501 Totals 0.0147 0.480 1.47-4 4.80-3 0.402 4.168 Total Ingress/Egress 0.4833 4.83-3 4.318 Total Operator Stay Time 0.0117 1.17-4 0.252 TOTAL ACCESS DOSE 0.49 4.90-3 4.57 NOTES: (1) Post-LOCA radiation doses for the 1% clad damage case are obtained by multiplying the 100% clad damage doses by a factor of 0.01 .

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ATTACHMENT 1 to PLA-4069 TABLE 2.2

SUMMARY

OF CALCULATED OPERATOR ACCESS DOSES TO CONTROL ESW MAKEUP FLOW REACTOR BUILDINGUNIT 2 ACCESS TO HEAT EXCHANGER PUMP ROOM (1-514) ELEV.

749'PERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 100% CLAD DAMAGE 1% CLAD DAMAGE (1) 100% FUEL MELT DOSE LOCATION/

SOURCE Dose From Dose From Dose From Dose From Dose From Dose From Airborne Contained Airborne Contained Airborne Contained Activity Sources Activity Sources Activity Sources Ingress/Egress To Heat Exchanger 0.028 0.0288 2.80-4 2.88-4 0.704 0.292 Pump Room (11-514)

Operator Stay Time In Heat Exchanger 0.0225 2.77-4 2.25-4 2.77-6 0.615 2.65-3 Pump Room (11-514)

Totals 0.0505 0.0291 5.05-4 2.91-4 1.319 0.295 Total Ingress/Egress 0.0570 5.70-4 0.996 Total Operator Stay Time 0.0228 2.28-4 0.618 TOTAL ACCESS DOSE 0.0798 7.98-4 1.61 NOTES: (1) Post-LOCA radiation doses for the 1% clad damage case are obtained by multiplying the 100% clad damage doses by a factor of 0.01.

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ATTACHMENT2 to PLA-4069 ATTACHMENT2 RESPONSE TO QUESTION 2 OF NRC 11/23/93 RAI Page 9

ATTACHMENT 2 to PLA-4069 OV V ON Provide operator dose estimates for those operator actions needed to maintain the normal spent fuel pool cooling function under DBA accident conditions, assuming the normal spent fuel pool cooling system is operational following an accident. Consider those actions needed to restore normal spent fuel following automatic or manual load shed of the spent fuel pool cooling system.

Include the same level of detail as in your response to question 1.

RE<SPONSK 2 The following provides a complete response to question 2 of the November 23, 1993 NRC RAI concerning radiological evaluations for the Loss of SFP Cooling issue. For this evaluation, ESW was evaluated as the source of make-up water since it is safety-related. Other sources of non-safety related water could be used but were not evaluated in order to minimize the amount of calculations performed. As noted in the response to question 1, restoration of normal SFP cooling is the expected course of action for responding to a Loss of SFP cooling event. This response is broken into three parts. Section 1.0 provides a description of the modeling techniques and assumptions used for evaluating operator access to restore normal Spent Fuel Pool (SFP)

Cooling. Section 2.0 a description of the actions required to restore and maintain normal SFP Cooling. It is important to note that no time-motion study was performed for these actions. The timing is based on operator experience since these actions are performed on a regular basis.

Section 3.0 provides summary tables of the doses (both airborne and contained) for each of the segments/actions assumed in the calculation. It should be noted that several separate actions, at different locations are required to restore and maintain normal SFP cooling. These tables show doses for a DBA LOCA with 1% clad damage, 100% clad damage, and 100% fuel melt (i.e.,

Reg. Guide 1.3 source term).

1.0 Detailed Descri tion of Modelin Techni ues The calculation performed to determine the operator doses for restoration of normal SFP cooling is based on the calculation performed for the response to question 1. The only differences are associated with the stay times and location of the operators with regard to distance from contained sources. The results of the time-motion study performed for question 1 were used to determine operator transit times, while operator experience.was.used to determine the time to perform the actions.

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ATTACHMENT 2 to PLA-4069 2.0 Actions for Restoration of Normal SFP Coolin The actions are based on the following sequences of events and plant configuration':

TIME DESCRIPTION 0 hrs. Fuel Pools are isolated; Both are filled; U2 pool heat load = 8.2 MBTU/HR (just completed a 40 day outage); U1 pool heat load = 6.27 MBTU/HR (last outage began 135 days ago); Pool Temp = 110'F.

U2 LOCA/LOOP occurs; Loss of fuel pool cooling occurs to both pools; Reactor building HVAC recirculation system starts; SGTS starts.

ON-135(235)-001 LOSS OF FUEL POOL COOLING/COOLANT INVENTORY entered.

U1 controlled shutdown begins due to the LOOP condition.

24 Hrs. Access available to ESW in both Units.

Offsite power is restored.

Implement Off Normal Procedures (LOSS OF FUEL POOL COOLING) to check that the system can be operated and then implement OP-135(235)-001 to place fuel pool cooling into operation.

The ON-135(235)-001 (LOSS OF FUEL POOL COOLING) would have been entered at time 0 of the event and would have been implemented to check that the system can be operated. It would have been determined that it could not due to a LOOP. Once power is restored (assumed to occur at no later than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after event initiation) the ON provisions would guide the operator to determine that the system and the necessary support systems are available.

Implementation of the ON's will assure the support systems are operable and function as required to support fuel pool cooling system operation. These procedures require a check to assure no system breach has occurred. When entrance is made to restore the system, it will be assumed that at this time any breach would be obvious and that no special. entrance to look for a system breach is necessary. The demineralizer portion of the system will not have to be inspected as it will be isolated from the cooling portion of the system by valves 15406/25406/05406 and 15444/25444/05444. These valves go closed on the loss of power at time 0 of the event. None of the other ON actions require entrance to the reactor building (except for than fuel pool cooling system start-up and fuel pool level makeup which will be discussed below) ~

A UNIT 2 LOCA IS ASSUMED FOR THIS CONDITION SINCE IT RESULTS IN THE WORST CASE RADIOLOGICAL CONDITIONS ~

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ATTACHMENT2 to PLA%069 Once the support systems are assured functioning and available, the fuel pool cooling system operation will be restored. The demineralizer function will not be restored until some time later when conditions have improved such that it can be inspected for possible system breach. It will be assumed however that an operator will go to the demineralizer panel OC207 on elevation 779'o determine that the fuel pool filter demineralizer subsystem is appropriately isolated. It is conservatively estimated that the operator willspend 10 minutes at the panel. The skimmer surge tank and fuel pool level will be more than adequate to support system operation as the pool swell due to heat up (including evaporative losses) will cause a slight increase in SFP level. It has been calculated that it will at most take 20 minutes to makeup the volume of water lost due to evaporation during the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in which it is assumed cooling is lost. This assumes one loop of ESW at 35 gpm make-up rate. Thus the pumps may be started once the bypass valve 153013 is closed, which takes at most 2 minutes (assuming the valve was full open at time 0 of the event). This valve is a manually operated valve and would not change position at time 0 of the event. Also, it takes approximately 5 minutes to turn on the three pumps and adjust the bypass valve 153013 open to pass the 1800 GPM fiow. Once the bypass valve is properly adjusted and the three pumps are operating, no other actions are required. Thus for Unit 1 in which the 153013 valve is next to the 1C206 panel, it will take one operator a maximum of 7 minutes to place the system in operation with three pumps excluding ingress and egress times. The dose associated with an ingress and egress time of = 6 minutes is calculated and reflected in the tables in section 3.0. On Unit 2, one operator will have to operate the 253013 valve (approximate 4 minute operation) and one will have to operate the pump control buttons (2 minute operation) as the valve is not in the vicinity of the panel.

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ATIACHMENT2 to PLA-4069 3.0 Summa of 0 erator Doses for Restoration of Normal SFP Coolin TABLE 3.1

SUMMARY

OF CALCULATED OPERATOR ACCESS DOSES TO RESTORE THE SPENT FUEL POOL COOLING SYSTEM REACTOR BUILDINGUNIT 1 ACCESS TO CONTROL PANEL 1C206/VALVE 153013 ELEV.

749'PERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 100% CLAD DAMAGE 1% CLAD DAMAGE (1) 100% FUEL MELT DOSE LOCATION/

SOURCE Dose From Dose From Dose From Dose From Dose From Dose From Airborne Contained Airborne Contained Airborne Contained Activity Sources Activity Sources Activity Sources Ingress/Egress To Control Panel 0.0272 0.0471 2.72-4 4.71-4 0.689 0.50 1C206/Valve 153013 Operator Stay Time At Control Panel 0.040 0.445 4.0-4 4.45-3 1.079 4.914 1C206/Valve 153013 Totals 0.0672 0.492 6.72-4 4.92-3 1.768 5.414 Total Ingress/Egress 0.0743 7.43-4 1.189 Total Operator Stay Time 0.485 4.85-3 5.993 TOTAL ACCESS DOSE 0.559 5.59-3 7.182 NOTES: (1) Post-LOCA radiation doses for the 1% clad damage case are obtained by multiplying the 100% clad damage doses by a factor of 0.01.

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ATTACHMENT2 to PLA-4069.

i<

TABLE 3.2

SUMMARY

OF CALCULATED OPERATOR ACCESS DOSES TO RESTORE THE SPENT FUEL POOL COOLING '.

SYSTEM REACTOR BUILDINGUNIT 2 MISSION 1, ACCESS TO CONTROL PANEL 2C206 ELEV.

749'PERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 100% CLAD DAMAGE 1% CLAD DAMAGE (1) 100% FUEL MELT DOSE LOCATION/

SOURCE Dose From Dose From Dose From Dose From Dose From Dose From Airborne Contained Airborne Contained Airborne Contained Activity Sources Activity Sources Activity Sources Ingress/Egress To Control Panel 2C206 0.0272 0.0471 2.72-4 4.71-4 0.689 0.50 Operator Stay Time At Control Panel 0.040 0.445 4.0-4 4.45-3 1.079 4.914 2C206 Totals 0.0672 0.492 6.72-4 4.92-3 1.768 5.414 Total Ingress/Egress 0.0743 7.43-4 1.189 Total Operator Stay Time 0.485 4.85-3 5.993 TOTAL ACCESS DOSE 0.559 5.59-3 7.182 NOTES: (1) Post-LOCA radiation doses for the 1% clad damage case are obtained by multiplying the 100% clad damage doses by a factor of 0.01.

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ATIACHMENT2 to PLA-4069.

TABLE 3.3

SUMMARY

OF CALCULATED OPERATOR ACCESS DOSES TO RESTORE THE SPENT FUEL POOL COOLING REACTOR BUILDINGUNIT 2 MISSION 2, ACCESS TO VALVE253013 PLATFORM ELEV. 762'-10" 'YSTEM OPERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 100% CLAD DAMAGE 1% CLAD DAMAGE (1) 100% FUEL MELT DOSE LOCATION/

SOURCE Dose From Dose From Dose From Dose From Dose From Dose From Airborne Contained Airborne Contained Airborne Contained Activity Sources Activity Sources Activity Sources Ingress/Egress To Valve 253013 0.0272 0.142 2.72-4 1.42-3 0.746 1.569 Operator Stay Time At Valve 253013 0.0396 0.538 3.96-4 5.38-3 1.079 6.08 Totals 0.0668 0.68 6.68-4 6.80-3 1.825 7.649 Total Ingress/Egress 0.169 1.69-3 2.315 Total Operator Stay Time 0.578 5.78-3 7.159 TOTAL ACCESS DOSE 0.75 7.50-3 9.47 NOTES: (1) Post-LOCA radiation doses for the 1% clad damage case are obtained by multiplying the 100% clad damage doses by a factor of 0.01.

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ATTACHMENT2 to PLA-4069.

TABLE 3.4

SUMMARY

OF CALCULATED OPERATOR ACCESS DOSES TO CHECK. DEMNERALIZER PANEL OC207 REACTOR BUILDINGUNIT 1 ACCESS TO PANEL OC207 ELEV.

779'PERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 100% CLAD DAMAGE 1% CLAD DAMAGE (1) 100% FUEL MELT DOSE LOCATION/

SOURCE Dose From Dose From Dose From Dose From Dose From Dose From Airborne Contained Airborne Contained Airborne Contained Activity Sources Activity Sources Activity Sources Ingress/Egress To Control Panel OC207 0.0243 0.0288 2.43-4 2.88-4 0.668 0.292 Operator Stay Time At Control Panel 0.0564 5.64-4 1.54 OC207 Totals 0.0807 0.0288 8.07-4 2.88-4 2.208 0.292 Total Ingress/Egress 0.0531 5.31-4 0.960 Total Operator Stay Time 0.0564 5.64-4 1.54 TOTAL ACCESS DOSE 0.11 1.10-3 2.5 NOTES: (1) Post-LOCA radiation doses for the 1% clad damage case are obtained by multiplying the 100% clad damage doses by a factor of 0.01.

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ATTACHMENT2 to PLA-4069.

TABLE 3.5

SUMMARY

OF CALCULATEDOPERATOR ACCESS TO PROVIDE MAKE-UP WATER TO THE SPENT FUEL POOL - '

REACTOR BUILDINGUNIT 1 ACCESS TO ROOM I-514 AND PANEL 1C206 ELEV.

749'PERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 100% CLAD DAMAGE 1% CLAD DAMAGE (1) 100% FUEL MELT DOSE LOCATION/

SOURCE Dose From Dose From Dose From Dose From Dose From Dose From Airborne Contained Airborne Contained Airborne Contained Activity Sources Activity Sources Activity Sources Ingress/Egress To Room 0.0241 0.0326 2.41-4 3.26-4 0.662 0.334 I-514 Dose To Check Water Level 0.00681 0.0567 6.81-5 5.67-4 0.186 0.632 At Panel 1C206 Dose Inside Room I-514 0.101 1.24-3 1.01-3 1.24-5 2.749 0.0118 Totals 0.132 0.0905 1.32-3 9.05-4 3.597 0.978 Total Ingress/Egress 0.0567 5.67-4 0.996 Total Operator Stay Time 0.166 1.66-3 3.579 TOTAL ACCESS DOSE 0.223 2.23-3 4.58 NOTES: (1) Post-LOCA radiation doses for the 1% clad damage case are obtained by multiplying the 100% clad damage doses by a factor of 0.01.

Page 17

~ ~

4 4.~'

ATTACKMfNT3 to PLA-4069 ATTACHMENT3 RESPONSE TO QUESTION 3 OF NRC 11/23/93 RAI Page 18

~~ ~

ATTACHMENT 3 to PLA-4069 S

QUES IQN Piovide operator dose estimates for those operator actions needed to maintain alternative spent fuel pool cooling functions under DBA accident conditions (i.e. use of accident and non-accident unit spent fuel pool cooling system to cool the accident unit fuel pool, etc.) assuming the accident unit normal spent fuel pool cooling system has failed as a result of a LOCA. Include the same level of detail as in your response to question 1 ~

RESPONSE 1 The following provides a complete response to question 3 of the November 23, 1993 NRC RAI concerning radiological evaluations for the Loss of SFP Cooling issue. This response is broken into two parts. Section 1.0 provides a 'discusses the availability of alternative SFP cooling under DBA (Reg. Guide 1.3) LOCA conditions. Section 2.0 provides a summary table of the doses (both airborne and contained) for RHR SFP cooling in the non-accident unit. This table shows doses for a DBA LOCA with 1% clad damage, 100% clad damage, and 100% fuel melt (i.e.,

Reg. Guide 1.3 source term).

1.0 Availabili of Alternative S ent Fuel Pool Coolin As discussed in PP&L's May 24 and August 16, 1993 submittals, RHR SFP Cooling in the accident unit and the Refueling floor are inaccessible for a DBA LOCA with an assumed Reg.

Guide 1.3 source term. PP&L has not performed a calculation for a Reg. Guide 1.3 source term for RHR SFP cooling mode, however, the dose rates for 100% clad damage are on the order of 50 to 440 Rem/hour. Based on these results, the dose rates for a fuel melt (Reg. Guide 1.3) would prohibit operator access. Under DBA LOCA conditions (i.e. 1% clad damage) operator access would be possible since the dose rates would be on the order of 0.5 to 4.4 Rem/hour.

Therefore, use of the accident unit's RHR system for alternative SFP cooling is not an option if a Reg. Guide 1.3 source term is assumed, but is possible for the conditions expected in a DBA LOCA.

As reported in PP&L's May 24, 1993 submittal, the dose rate on the Refueling floor at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after a Reg. Guide 1.3 DBA LOCA is 79.8 Rem/hour. The doses for the 100% and 1 % clad damage conditions were reported to be 3.8 and 0.038 Rem/hour, respectively. Since the time to pull the cask pit gates is on the order of a shift, it is not possible to accomplish this activity and incur an acceptable dose to the operator for under Reg Guide 1.3 conditions with airborne radiation. Therefore, use of the non-accident unit's systems (normal SFP cooling and RHR) to cool the accident unit's SFP is not an option ifthe cask storage pit gates are installed and a Reg.

Guide 1.3 source term is present. These actions could be performed for the DBA LOCA for the 100% and 1% clad damage conditions.

The calculation used to determine the above dose rates is the same calculation described in the response to question 1.

Page 19

ATTACHMENT3 to PLA<069 2.0 Summa of 0 erator Doses for the Non-Accident Unit's RHR SFP Coolin The following table summarizes the dose that an operator would experience establishing RHR SFP cooling to the non-accident unit. While this is not specifically requested in the RAI, PP&L is providing this information to establish the accessibility of the non-accident unit even if the ventilation is not isolated from the accident unit. Non-isolation of non-accident unit is assumed for this case in order to maximize the dose in the non-accident unit. As noted in PP&L's August 16, 1993 submittal, the non-accident unit can be isolated from the reactor building HVAC recirculation plenum, thereby preventing the spread of radiation to the non-accident unit. The doses are based on the same calculation used to obtain the doses for the answer to question 2.

The time to stroke the RHR valves from full closed to full open is 2 minutes per valve and a separate operator would be sent to manipulate each valve.

Page 20

ATTACHMENT3 to PLA-4069 TABLE 2.1

SUMMARY

OF CALCULATED OPERATOR ACCESS DOSES FOR RHR FUEL POOL COOLING ASSIST FROM NON-ACCIDENT UNIT , REACTOR BUILDING UNIT 1 ACCESS TO VALVES 151060 AND 151070 PLATFORM ELEV.

705'PERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 100% CLAD DAMAGE 1% CLAD DAMAGE 100% FUEL MELT DOSE LOCATION/

SOURCE Dose From Dose From Dose From Dose From Dose From Dose From Airborne Contained Airborne Contained Airborne Contained Activity Sources Activity Sources Activity Sources Ingress/Egress To Valves 151060 & 0.026 5.17-3 2.60-4 5.17-3 0.714 5.17-3 151070 Operator Stay Time At Valves 151060 & 0.0231 6.67-3 2.31-4 6.67-3 0.634 6.67-3 151070 Totals 0.0491 0.0118 4.91-4 0.0118 1.348 0.0118 Total Ingress/Egress 0.0312 5.43-3 0.719 Total Operator Stay Time 0.0297 6.91-3 0.641 TOTAL ACCESS DOSE 0.0609 0.0123 1.36 NOTES: (1) Post-LOCA airborne radiation doses for the 1% clad damage case are obtained by multiplying the 100% clad damage doses by a factor of 0.01.

Page 21

il 01~4-1984 14( G~r P.O2 ENCLOSURE 1 Pennsylvania Power R Light Company VeO NOrth Ninth Stree ihllentOWn, PA tbtct.ttre o Stb/7744151 Robert 0. Syrem 2$ IM74-7502 Director of Nuclear Reactor Regulation Attention: Mr. C, L. Miller, Project Director Project Directorate 1-2 Division of Reactor Projects U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Cr SUSQUEHhNNh STEAM BLBCTRIC SThTION REQUEST FOR ADDITIONALINPORMATION ON LO55 Ot StENT WEL tOOL COOLINQ EVENTS Qodcat No@ 504tl and SMSI Dear Mr. Miller.

Attached is PP&?.'a reayonaa ta your November 23. 1993 Request for Additional Infortnation conccttting personnel access tiwsthns tisulting floe postulated loss of spent fuel pool cooling events.

Plass contact Mr. James M. Kenny at (215) 774-7914 shouM your requite additional infomtation.

Vcty truly yours, I

9<pllsppp6 Attachment cc: NRC Document Control Desk (oriiinal)

NRC Region l Mr. G, S, Barber, NRC Sr. Resident Inspector - SSSS Mr, R. J. Clek, NRC Sr, Project Manager -.Rockvtlle

AlThCHMENT 1 to pLA~

ATTACHMENT1 RESPONSE TO QUESTION 1 OF NRC 11/23/93 RAI Page 1

AlTACHMENT1 to PLA~

Provide a detailed description of the modeling techniques used to generate the 4,22 Rem cited in the Augus! 16, 1993 submittal ss the dose received dWng tba aeration of the spent Suel pool emergency service water supply valves. Describe in detail the time-motion analysis performed to determine exposure time eeoc lated with required operator actions. 1ncludc the parameters and assumptions used to Seneratc the contained and airborne source terms and calculate the respective dose components associated with each segment of the thne-motion analysis.

The following provides a complete response to question I of the November 23, 1993 NRC RAI concsrains radiological evaluations for the Loss of SFP Cooling issue. As discussed in PP&L's May 24 and August 16, 1993 submittals, restoration of the normal SFP cooling system prior to boQing is expccld. Consequently, PAL would not expect it to bc accessary to usa E8W for make-up since the noanal systetns could be used or make-up fhm the non-accident unit could be provided, This response is separated into two parts. Section 1.0 provides a description of the modeling techniques and assumptiona uasd for evaluating operator access to thc ESW make-up valves, Also included in this section, is a discussion of how the results time motion study were factored into the analysis. Section 2,0 provides sununary tables of the doses (both airbotna and contained) for each of thc segments assumed ln the calculatioL These tables show doses for a DBA LOCA with 19o clad damage, 1N% clad daautgc, and 10% tM tnelt {i,e., Reg. Guide 1.3 source tarn).

1.0 11 Qzmim The TACT5 computer code is used to evaluate post. LOCA radiation sources inside the reactor building using the CESAR Chsytcr 15.6.5 DBA-LOCA activity Qow path model with realistic

! bNtf~!

estimates of containment l<<akage ratea Airborne activity concentrations in the reactor building U  !! I p! I failure (NURE&14N) and heal molt (Regulatory Guide 1.3) source terms, hd!!! Nlhd %Id!

Using these post-LOCA source terms, radiation dose rates inside the reactor building horn airborne aotivity and &un suppression pool water contained sources ae evaluated using the MICROSHIELD computer coda Operator access requirements inside the mactbr building ware determined for establishing ESV mahrup to the spent fbe1 pool under post-LOCA conditions. Operator accus routes and misaom werc identKed and dlvidaf into s<<quential segments for th<<purpose of evaluating operator access Page 2

ATTACHMENT1 to PEA~9 doses. Operator access doses arc computed by multiplying ihc radiation dose rates from boih airborne and contained sources in a given mission segment by the time spent by the operator in that mission scgmcnt, The sum of the radiation doses for all segments of the mission provides the total mission dose.

1,2 This analysis evaluates personnel access doses inside the reactor building fnr the following postulated LOCA initiated core damage cases: 1% clad damage, 100% clad damage, and 100%

fuel iuclt.

The clad damage cases are evaluated using assumptions consistent with the accident source terms described in VUREG-1 465, The fuel melt case is cnnservatively evaluated using accident source terms consistent with USNRC Regulatory Guide 1.3 and the PSAR Chapter 15.6.5 DBA-LOCA licensing basis evaluation. For all cases, realistic cstimatos of thc containmcnt Icekego rate are used.

For the activity Rnw path madel used in this analysis, the activity concentrations inside the nodal volumes are calculated by dividing'the activity in the node at the time of interest by the nodal

~

volume and therefore are directly proportional to the activity source term rclcascd &om the fuel into thc containment and suppression pool, For the postulated LOCA's that assume clad damage, the amount of activity released is directly proportional to the amount of clad damage, Therefore, a complete source term and dose analysis is perfozmcd for thc lOON clad damage case and source tcrzns and dose results for thc I /o clad damage case azc obtained by multiplying the 100% clad damage results by 0,01, hKGKG-1465 was used to supply release data for the clad damage cases. Since no speciQc numerical guidance is provided in NURE6-1465, the amount of clad activity released &om the core which becomes airborne msidc containment or renudns in thc supprcoion pool water is based on taking credit for 55aton product scrubMng and retention in accordance with Standard Review Plhn 6,5,5. In addition, for the clad damage cases, an aerosol removal rate of 0.73 hz"'or particulate iodine and cesium is assumed based on infozznatioa provided in NUREG-1465, Table S,6 for the La Salle Nuclear Power Plant which is also a BWR Mark II containment design, No credit for aerosol removal was taken in the 18% fhcl melt case.

Coze activity release &actions for a LOCA with 15% fuel melt are based on thc requirements of Regulatory Guide 1.3 stnd NURBO-0737 and azc th¹ 25% of the iodincs and 100% of the noble gases are instantaneously airborne in priniary containznent and available for leakage and 50% of the core inventory of iodizls and 1% of the particulate are released to the supprcssioe pool water, Duc to the number of particulate isotopes in tbc core and the number of dose calculations required, the particul¹e ectivity release khan the core was not explicitly included in thc suppression pool activity source tenn for the 100% fbel melt case. Instead, bounding dose calculations using post-LOCA Nipyxession pool contained sources with and without particulate Page 3

ATTACHJHENT 0 to p~~

were used to determine a dose multiplia which was used to account for the dose contribution from 198 particulate in the suppression pool water, The iodine and noble gas activities released for tbe 100% Sml melt case are obtained &am the FSAR DBA.LOCA analysis given in PP8Q.

calculation FXMDAM<14. The Sounding analys'I for the dose coatributloa born perticulere is based on post-LOCA suppression pool liquid activity source terms given in Susquehanna Projoot Bechtcl Gdcuiation 21M-201.

For ail clad damage and fbcl melt cases, the isotopic chemical form of the activity released from the core is assumed to be as allows:

Iodincs ~ 91% elemental

~ 4%i organic

~ 5% particulate Cesiums 10N4 particulate Noble Gases ~ 100% elenental This assumption is consistent with USNRC Regulatory &dde 1,3 for core fuel damage. For the clad daaage cases, NUREQ-1465 indicates that the chemical form for iodine entering containmcmt is 95% particulat and 5% elemental, but witbout Ph control of the suppression pool water, a relatively large &action of the particul¹e iodine dissolved in supIxelion yool water will be converted to eleInental iodine. 'Qierekre, the above chemical forms assumed for this anaiyis ale cohshrvative for the clat damage cases.

Tho primary ccntainmcnt design basis leakage rate is 1 &day. For the PSAR liccnins basis DBA-LOCA analysis this leakage rate ia assumed for the duration of the accident. Poc this evaluation, a one depmknt realists oontainaMstt leakage rata based oa containlnettt Integr¹ed Leakage Rata Tosting gLRT) jesuits and the calculasal costtaheeat poet-LOCA Ieeaam reapolm ia ueed Sc beth the clad damage aat IQ% M ddt orna The ILRT test pressure corresponds to thc maxhnum calcul¹cd containment poN LOCA pressure with a design margin applied. The realistic coatainment leakage rate waa calculated by reducing the ILRT measured leakage r¹e proportion¹ely to the ILRT test pressure and the calculated containment pressure response for a LOCA. Foe thh analyda, the most uy-to~ 1eaksge rate data ia used so that tho dose estimate>> reQect the most up-~late containment leakage coaditiona Tbetefore, the measured Icakago rate at 0.606&day 5om the Unit 1 ILRT performed 5/5/92 ia aouned for this analyia A leakage r¹e of O.N6&day is also representative of typical measured leakage rates at SSES Unita 1 4 2.

NUItE64737 provides guidance for evaluating operator acceN to vital plant mm for post-accident operations, It !ties under Itetn (2) Systema Containinl Tbe Source tha! for post-LOCA accident operationa, "IbdMon &om leakage ofsystetns located outside of containment baaed not be considered for this analysis". Tbaehx>>, it is ammed Sc thia evaluation th¹ the lealage of pner-I,OCA o~aha~ airborne activity though oorltallanltt yeaetratiooa that me ver oeelod Page 4

nssd zMst be oeasidecod. The rcaliltic coeaiarnent leakage rate used fbr both the clad and fuel damage cases <<re based on the measured leakage rates Qnaugh containment penetrations that are not water seals Based on the above activity flow pathways,. the TACTS code provides total activity in the reactor building and suppresaian pool as a function of time post accident. At the required evaluation time (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post accident), isotopic activities were taken &om the TACTS output edit and divided by the appropriate dispersal volume to give activity concentrations, Radiation doses insMe the reactor building are evaluated for both airborne activity and for suppression pool water contained sources. Operator access doses or area radiation dose ratci are evaluated in each of the areas that require access to provide BSW makeup, Rahstion doses are evaluated using the MIGROSKELD coaqtuter code with the TACTS generated activity concentration source terms, The dose rate &om airborne activity inside the reactor building is actually sn immsrsion dose.

However, since the MICROSHIELD computer code catuxrt calcuhrte dose rates hternal to the source, slab geometry is used. One half of the source volume is modeled as a rectangular volume source and a dose rate on the surhee of this volume is calculated. The inuncnion dose is th All immersion dose rates arc conservatively calculated at the centerline oC Lho @xnan, Dose

&om the mntained suppression pool lhiuid piping sources sre calctdated using cylindrical source

~

calculated by multiplying the contact dose rate 6am half of the source volume by a factor of 2.

geometry vdth receiver at side.

In order to evaluate operator acceN doses, operator walkhg rates and stir climbing rates are required. A tabac motion study was pcrfoanod to vmify operator accese travel times inside the reactor building under LOCA conditions. An opetahe waa dressed in peNctiva clothing and worl a Self Celled Srlathmg AIyardus and actual transit thaea to valves located on elevationa 670'nd 749'f the tea~ bi48ng were measeel. Anise to these valves is required to provide BSV maIkaup to Ow spent hei pool. Baad upon tbe results of this time motion study, an operator walking rate of 200 ft/min and a stair clhnbing rate ot 50. fthnin are IS l~

conservatively ammed for operator lngresa/egresa doss catcuistiona for access to the reactor buiMing under LOCA conditionL Also as patt of this study, using spare valves, the valve opeung thne for the 2 inch valves used for BS% system tie-in sad Qow coldrol was mealured to be 10 dl,baf M ldl ~%&~II fl The incoqecadoa of tho time.motion study hto the fhal calculation resulted in the preliminary dose to the ogler imeasing &om 422 Rem. to 4,57 Rem., for acoesa to the valves on elevations 670. The M2 Rccn value was reported in PPN 'a August 16, 1993 submittal.

OIerator accas doses are only evaluated for Unit 1. The operator access area locations for Units 1 and 2 are idccdcal eacept lbr the area containing the valves rcxpirN1 ibr tie-in of de ESW syltem fot makeup ( Valve NoL 153500, 153501 anal 253S00, 2S3501). Buon on the post-accident radiation levels given in Figures 1$ .1-3 and 11.14 of the SSES PSAR, the area containing the Unit 1 valves for RS% system tie in has higher dose rates than de Urtit 2 axea.

Page 5

ATTACHMENT1 to PUAOH Therefore doso rates calculated for the Unit 1 area w111 be conservative for the Unit 2 area.

Therefore, all of the calculated operator access doses for Unit 1 are apylicable to Unit 2, Operator aces'oses are evaluated at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA. Far lois of spent Swl pool oooHng, the fbel pool beiins to boB in appoximately 4I hours, For the condition where a Lou of CHMte Power (LOOP) is also post~i, povnc ls oxpaM to be restored within 24 bours, Therefote, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was chosen ss the latest time operator action could be taken to restore cooling and/or make up to the pull and assure those actions would be completed prior to SFP boiHng Some dose rates are aiio calculated at other times post. LOCA and to show that poet-LOCA dose rates are decreasing for time periods yeater than 24 boa For the 1$ % fidel melt ca+, a dose factor is used to take into account the dose contribution &om psrticulate in the suppression pool liquid (sec Section 5.1, Assumption 5). The particulate dose factor is defined as the ratio of the dnsa Rem suppceaaion pool liquid sources containing core am~~~~pool liqtnd~~conta S activity only. All doses &em contained sources for the 18% fbol melt case are calculated ushg the hQCROSHIELD computer code with a suppression pool source term th contains iodines only. The MICROSHIELD results for contained sources Sr the 18% heal melt ceo are then multiyHed by the particulate dose f ctnr to take into ~unt the dose contribution Crom particulate, This dose factor is calculated at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> post-LOCA and can only be applied to operator access doses kom suppression pool liquid sources evaluated at Ns satne thne period post~acctdent

ThBLE 2.1

SUMMARY

OF ChLCULATKDOPERATOR AUChSS DOSES TO PROVIDE ESW MAKEUP To SPECI'UEL POOL

- REACTOR BUILDINGUNIT I AOCIXS 1Q VALVES 153500 AND 153501 ELEV.

670'PERATOR ACCESS DOSES AT 24 HOURS POST-I.OCA (R)

IM% CXAD DAMAGE HC CLAD DAMAGE (1)

Sources Dose Fam Dose Fma Dose Fmra Dna Fam Dose Fam Dose Faxa Caaimxl Airborne hiram

~'T Coaamal Coataijsad bib'ctivity hctivity Scxxecs Activity Sources Vd 722-3 OA76 72? 5 4.76-3 O.1911 4.12 153500 4 153501 Oleralta Stay Tme ht Valves 7.45-3 heal-3 7.45-S heal-5 O.N76 153500 A 153501 T044 0.0147 OASO 1.47-4 4.80-3 O.402 4.16$

Teal 4.83-3 43]8 Total Qpaator Qay Tmc 0.0117 1.17-4 0252 0.49 4.9D-3 QQ~: (I) poat4A)CA tahatioa doses for tbc of OOI I< chd damage ca+ me obtaiaxl by aadtiplyiag the 1095 cia doge doses by o ~ y 8

ThBLE 22 SUMMhRY OF ChJA ULLED OPERATOR hCCESS BOSFA TO OONTROL ESW MAKEOP FLAY REA~R BUILIXNGUNIT 2 AOCKSS 'IQ HEAT EXCHANGER PUMP ROOM (I-514) ELEV.

749'PERATOR.

hOCESS DOSES hT 24 HOURS MST-LCICA (R) iN% CLhD DhhhLGE 1% CLaD OhMAGE (1) 1W% FUEL MELT Sources Dose Faan Doae Frcm Date Fnrn Dose Fxocn Dose From Dose Fern hirbcxm Cocancd Airborae Co&aiaed hirbaea: Coatauxxl Activity Soiacxa Activity herky Sources

~Togae l heap Iaogt gI-SI4) e 0.0288 2.854 0.704 0292 Qietatnr stay Time ln Heat Eacbmger 0.0225 0.615 2.65-3 Pmy Ipaaa (H-Sl4) 0.9291 S.Q5-4 2.91-4 1919 0.295 Total S.704 0.996 Total Oy~r Qay T~ 0.61$

0.079$ 7.9@4 1.61 post-L0CA radistioa doecg fm tbe 1% dad damage care axe obtained bY multiplVinS the 100Ye chd daaaage doses by a factor NOH%. (I) of 0.01.

01-04-1994 11'0 ATTACHMENT2 to PLA~

ATTACHMENT2 RESPONSE TO QUESTION 2 OF NRC i 1/23/93 RAI Page 9

GX-04-1984 11 I <0 Provide operator dose estnnates for those operator actions needed to maintain the normal spent gael pool cooling fbncdon under DBA accident conditions, assuming the nortnal spent fuel pool cooling systetn is operational foHowing an accident. Consider those actions needed to restore normal spent fbel following automatic or manual lpad shed of the spent fuel pool cooling systczn, Include the same level of detail as in your response to question l.

The following provides a complete response to question 2 of the November 23, 1993 NRC RAI concerning radiological evaluations for the Loss of SFP Cooling issue. For this evaluation, BSW was evaluated as the source of nuke.up water since it is safety related. Other sources of non-safety related water could be used but were not evaluated in order to mInimize the amount of calculations performed. As noted in the response to question 1, restoration of normal SPP cooling Is tbe expected course of actin for responding'to a Loss of SFP cooling event. This response is broken into three parts. Section 1.0 provides a description of the modeling techniques and assumptions used for evaluating operate access to restore normal Spent Fuel Pool (SPP)

CooHng. Section 2.0 a description of the actions required to restore and maintain nortnal SFP Cooling, It Is Imponaa to note that gg time. motion study was pccformod fot those ao6ans, The tiaung is based on operator eqertence since these actions are performed on a regula basis.

Section 3,0 provides sudsy tables of the doses {both airborne and contahml) for each of the segnNnlactions assumed in the calculation. It should be noted that several separate actions, at different locations are required to restore and maintain tNnaal SFF cooling, These tables show doses for a DBh LOCA witb 1% cled *mage, 1554 clad damage, and 1$ % &el melt (i.e.,

Reg. Guide l.3 so+en tean).

1.0 The calculation perfornod to determine the operator doses for restoration of normal SFP cooling is based oa the calculation perfortned for the response to question l. The only differences are associated with the stay times and location of the oIerators with regard to distance from detertnIM parian the'Ictioea tN5& ~

contaitted soureeL Tbe Ieltlts of tbe time motion study ~Nrfosmed for question 1 were used to op~& %hQ0 operator eaq)aicaao was used to determine the time to Page lo

OL 04 1894 11148 AlTAcHMENT2 to plJwoae The actions are based on the following sequences of events and plant configuration'.

DESCRIPTION 0 hrs. Fuel Pools are isolated; Beth are filled,'2 pool heat load $ ,2 MBTU/HR (Iust completed a 40 day outage); Ul pool heat load 6.27 MBTU/HR (last outage began 135 days ago); Pool Temp 1104F.

UZ LOCA/LOOF occurs; Loss of fbel pvvl cvotinl occurs to both pools; Reactor building HVAC recirculation system starts; SGTS startL ON<<13S(23$ )M1 LOSS OF FUEL POOL COOLINO/COOLANT INVENTORY entered, Ul controlled shutdown begins due to the LOOP condition.

24 Hr!. Accus available to ES% in both Unita at&to power is restored, Qttylornent OffNonrtal Procedures (LOSS OF FUEL POOL COOLINQ) to check that the system can be operated and then implement OP-135(235)401 to place fidel pool cooHng into operation.

The ON-l3S(23001 g 088 OF FUEL POOL COOLING) would have been entered at time 0 of the event and would have been impletnented to check that the systetn can be operated. It would have been detacmined that it.could not due to a LOOP. Once power is restored (assumed to occur at ao later than 24 bours after event inititttion) tbe ON provisions wouM guide the operator to detertrtine that tho system and tbo ttocessary support systems ate available.

Itnplementation of the ON's willassure the ttupport systetns are operablo and Smction as required to support fho1 pool cooling systetn operation. Those procedures require a check to assure no isolated

m. Ih ~IN I~ff systetn breach has occurred. Wbea etttrance is tnado to restore the systetrt, it <vill be assunted that at this the argr breach would be obvious and that ao special eatrance to look NN044II00444. Ih M I NI lt I NISI f~ I 0

hem the cooling portion of tho systam by valva 15406/254ki/05406 and 0

for a sysuan brcach fh hf Ifmlt N

of the other ON actions teyire antrartce to tho reactor buHdhg (except St than fidel pool cooling, system start up and Set pool level make@ which vrN be discusecd bc4ow).

~

a asm 2 md 2! aseam caaI RMzaserlcM, cclrtzmMs ~

~ mza cNsozTLM a~ zT aeattscs m Page ll

81 94 19&4 ' i cO P. CU ATTAt:HMEMT2 to P~H Once tho support systems arc assured functioning and availablo, tho fuel pool cooling system operation wiH bc rcstorlxL The dcmincralizer function will not bc restored until some time later when conditions have improved such that it can be inspected for possible system brcach. It will be ssaunecL be+ever thar Ia operator willIo to the deminaraliiee yanel OC2M on elevarioet WF to determine that tho hNl pooL 61ter dcclrecraliser subsystem is appro1RICCLy isolated. It is conservgively estimated that the opener willspend 10 minutes at the panel. Tho skimuer surge tank and fuel pool level will be more than adequate to support system operation as thc pool swell due to heat up (including evaporative losses) wiH cause a sHght mxaaso in SFP leveL It has been calculated thit it will et moat taloe 2D minutes tn msheq) the volume of water lost duo to evaporation during the 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in which it is assurMd cooling is lost. This assumes onc loop of RSVP at 3S gpm mako.up rute. Thus tho pumps may bo started once tbo bypass valve 153013 is closed, which tales at most 2 minutes (assuming tbe valve was full open at time 0 of thc event), This valve is a manually operated valve and would not change position at time 0 of the event. Ahe, it takes approximately 5 minutes to turn on tbe three pumps and adust thc bypass valve 1S3013 open to pass thc 1800 GPM Qow, Once the bypass valve is properly a(usted and the throe pumps are operating, m other actions are required, Thus for Unit 1 in which the 193013 valve is next to the IC206 panel, it will tale one operator a maxitnum of 7 minutes to place the system in operation with three pumps cxciudini ingress and cycss times. The dose assoc}atsd with an ingress sad egress time of e 6 minutes ia calculated and rcQcctod in the tables in scctioa 3.0. On Unit 2, one operator will have to operate the 2S3013 valve (approximate 4 minute operation) end one will have to operate the puap control buttons (2 minuto oyoration) as the valve is not in the vicinity of tbe paneL Page 12

h1ThcHA4ENf g ~ ~any TABLE 3.1 SIJMMARY OP CAIZUIATEDOPERATOR h1X.'KSS DOSES TO RESTORE THE SPENT FUEL POOL COOING SYFHiM RBhCIOR KKlLDKNGUNlI' hCCESS 1Q 69NROL PANEL )C206AALVE l53013 KLEV. 74K OPERATOR hGC)%S DOSES AT 24 HOURS POST-LOCA (R) 1N% CXhD SAMhGE 1% CLAD SAMhGE () j 1&% HRL MELT DOSE LOCA'DON/

SOURCE Dose Fram Dose Faxa Dose Fred Dose Frori Oaoc Fran Dose From hxboroe Coecaimxl Airbrea: Coatained AiR~ G0QltslBcd Activity Socaccs Activity Sources Activity Sources T C 0.0471 Z72W 4.71-4 0.6&9

) C206/YaIve 1530) 3 Qgecgr Qay Tiae At G~ml Pand 0.040 OA45 4.45-3 4.914

)C206Nahe I530I3 I.079'.761 5.414 0.0743 7.43-4 ).)89 TagaI Opa~ QIy T~ 4.$ 5-3 $ 993 0559 $ 39-3 7.1N2 uf 0.0l.

~ ~ f0r Qx: 1% clad dsmagc cee ere oblained by maltiptyhg rhc l00$ dad daaagc dases by a D'or 3

Page l3

AlTAClgmr2 sa ~~yap The% 32

SUMMARY

OF AQXXJLAVEDOPERATOR ACCESS DOSES TO RESIORE THE SPENT FUEL POOL COGUNG SVSHQ4 REACMR BO1LD&IQ UMT 2 MISSIGbl l, Al3~8 TO CONTROL PANEL 2CM6 ELEV.

749'PERATOR hCCESS DOSES hT 24 HOURS POST-LOCh (R) l% CLAD Skh fAGK 1% CLAD DhMAGE (I) lM% FllKLMELT DOSE LOCATEÃ/

SOURCE Dose Fma Dose Fram Dme Fam Doee Fram Dose From Dose Fnlrr hihxm Coataiued hatbo~ Contained Airboca: Carrtamxl hctbrkg Sources. Activity Souces Activity Snxcea To Gyral tiead 2C206 0.0411 4.71-4 0.6$ 9 0.50 4.4$ -3 I 079 4.9I4 4.92-3 541 l.ls9 0.485 4.85-3 5993 TOYAL ACCIESS DOSE 5.59-3 V.II2 NOIES: (1) FMO-LOCh rttrhilborr doses Rc the 1% dad damnee case are obtaimxi by aulgj8yjgg tbe af OAI.

19'lad darxggp doses by a ~

Page 14

ATtACH~Z w ~~

ThBLB 33 SMAMllkYOF CALCULATEDOPERATOR hCCESS DOSES TO RESTORE THE SPENf FUEL POOL C0OLPg)

SYSTEM REhCMR BUILDINGUMT 2- MISSION 2, ACCF~ TO VALVE253013 PLATFORhf ELEV. 762'-10" OPERATOR ACCESS DOSES AT 24 HOURS POST-LOCA (R) 1N% CLAD Qkb%AGR 1% CLAD DAMlCE (I) 1% E%EL MELT DOSE LCXMTION/

SOURCE Doge Fam Boae Fnm Dose Fmn Dole Faxa Dose Fnln Boee Fmra Airbriae Coettlirred Ail&ma: Co&aned Air br'M: Contaaed Acthity Sasacs Activity Somxa Activity Sauces To Vahe 253013 O.l42 I AZ-3 0.746 ).569 Ope~ St¹y Tme At Valve 2530)3 OMS 538-3 ).079 0.68 6.$ 0-3 1.125 7.649 Total 0.169 1.69-3 2.315 5.78-3 7.159 0.75 7~3 9.47

~&@: (I) Pbo¹tgAKA agleam doses foe the 1% c)ad damage can: are obtiicecl by multiplyag the oE0.O).

IONt'lad r)ramage doscl by a ~

Page 15 O

TABLE 3.1

SUMMARY

OF CALCULATEDOPEgATQR A@CESS DOSES TO ~~@ ~M;IONIZER PANEL 0C207 REhClOR BUILDS'T I AOCESS TO PANEL OC207 FLED. Q I

779'2'ERhTOR Q

ACCESS DOSES hT 24 HO%K MST-LOCA (R) 1N% CLAD ShMAGE l% Cl~ ShMACE (I) II0% PIIM. MELT DOSE LOCA'HONJ'OURCE Dose Fram Sole Fam Dose Fmn Dose Dose Fam DoIe Freya .

hiahmm CoaCaiaed AItboaae Fax'oaCaincd Airboat Coasaiaed hctiviiy Smaoee Activity Soaxces Acciviy Soarces To Caael Pand OC?07 Q.9243 tklMS 2.43-4 2.N-4 0.66& 0.292 0.0564 0 5.64-4 T~

ToIsl

~

01~ Qay T~

0.6531 1.074 531-4 0.292 ITALNX~S DOSS 0.11 I.I0-3

, NQ'lpga: (I) ~LOCh zafwda) doses of tkOI.

Rt Ae 1% cM daalge ~ mt obasinai by IrnsICiplymg da IONA oh'srnll@ ~g ~ a Page 16 0

ATFhCiaE~ g ~

ABLE 3.5 StjMMhRYOFCMCULATEDQPBRATORAOCESSTOPROVlljEMA1~-UP ly'Appal

- RFM!TQR. $ 81LIMNO UNIT 1 hCCESS TO 1INM 1-514 hND PANEL IC206 EL'. SP~FUH TOTE POO1 749'%RA&lR AOCBSS DOSES AT 24 HOURS NKI'-IANNA gt)

IIMCCLAD DhhklCR 1% CLADQA59LGE (1) IM% FllEL MIXT DOSE LOCATION/

SOURCE Dace Fmca Dace From Dace Faxn Dose Fmm Dose From Dose Fran

~T hirhxm GoahIined hirbarrs: CoIeaitljd hirborne Ctmained hcIIivity Sonr eel Acuity Soaxccs Actbity Souaxm m 06241 0.0326 2A1% 3264 0.662 1-5N Dose To Eback. %~ Lcvd 0.00681 6.81-5 5.67% 0.156 hI Pand 1C206 Dam haik Roau 1-514 0.101 1.24-3 1.01-3 1.Z4-5 OBI II 132-3 9.~

0.0567 $ 47% 0.996 Total Op~a Ray T~ 1.66-3 3.$ 79 NOTES: (l) ~LOCh gad~i dosee 5g'be 1% cad < eml8~ came are obtamas1 4 xn14~iag the lOMC eh@ d~p 4 b a4'O.OI.

Bsge 1?

ATI'ACHMENT3 RESPONSE 70 QUESTION 3 OF NRC 11/23/93 RAI

81~ 1394 P 12

~ J ATl'ACHMENT3 to NAAOH Provide operator dose estimates for those operator actions needed to maintain altera<<&re spent fidel pool cooliag hmctions under DBh accident ccnciitiens (i.e. use of <<ccideat aut non-accident unit spent fuel pool cooling system to cool the accident uait fool pool, etc.) assunung the <<ccideat uait aorjxll spoat fidel pool coolial system he filled as <<result of a LOCA. Include W same level of detail as in your response to question 1.

0 d rb ~ f aHRSFP thigh d ~

'Oe followiag provides a complete resporse to question 3 of the November 23, 1993 NRC RAI coacerniag radiological evaluations for the Loss of SFP Cooling issue. This response is broken into two parts, Section 1.0 provides a discusses the availsMity of altern<<tive SFP cooling unde

~)

doses for a DBA LOCA with 1% clad damage, 105i clad damage, snd 15% fbel melt (i.e.,

Reg. Quide 1.3 source term).

1,0 Tl'l DBA (Reg. Guide 1.3) LOCA cotgiitioaL Section 2.0 provides a Naaamy table of the doses k

As discussed in PPM.'s May 24 and August 15, 1993 sabmittals, RHR SFP Cooling ia tbe accidcrrt unit and the Refbeliag fhor are iaaeessible for a DBA LOCA with an assumed Reg, Guide 13 source tenn. PAL h<<s not performed a calculatioa for a Rag. QuMe 1.3 puree term for RHR SPP cooling mode, however, the dose rates for 100% clad damage Ne oa the order of

$ 0 to 440 Rcaulllur. S<<sad oa tbcse results, tbsj dose rates for a fbo1 molt @Log. Guide 1,$ )

would prohibit ope~r accesL Uader DBA LOCA conditions (i.e. 1% clad damge) operator acce<<a would be possible since the dose rates wouM be oa the order of 0,5 to 4.4 RNnthour.

Therefoee use of the aeidear uns RHR system for <<ltern<<tive SFP caallng ls not sn option if a Reg. (halte 1.3 souse term is assumed, but is possible for the cosditioas exyected in a DBA LOCA.

As reported ia PPN,'s May 24, 1993 subraittal, the dose rate oe the ReheUag 5oar at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> damage exxRCom pull Se ~plC

~

<<her ~ Res, (Idaho 1.3 MA LOCA h%S Reathea, Vise deese Ae the IONIC <<aL 1 % cl<<d apped to ba34aal LNIRemlbaa, rasyeetiveiy. Qaoe the thne te AN ietlr.the ocde of a shat, ft is ~ poedbie to awomryiish thh <<ctivity <<nd iacut at 'Nt 8 the operltoc for under Reg OuMe 13 ootMItiom with <<irwin<<

radhNadf ua of the memcM unit's systems (tlcmal SFP cooling <<nd RHR) to cool the <<ocidart unit's SPP is net m opt@a itthe cask storage pit I<<te<<are metall<<d <<ad a keg.

Gute 1,3 source tern is imleaL 'Hsa a:tioas oomph}be pe5xamd hr the DSA LOCA hc the IDOL asd 1% cd daaNIe ooa5thaa r

Tbe calcuia8aa used to detcemiae the above dose rates is the same calculstioa described in the re<<yea<<e to yee&m l.

Page 19

AtThCRMENY 3 to I'~H 2.0

~e fo))owiag table aummarina the dose that an opcmtor would experlEnce establishing RHR SFP cooling to the noa~cident unit. While this is not spcciQcaQy requested in the RAI, PPAL is pcevidinl this inforamtion to establish tbo eecelsibility of the non~dent uait Iten if the venti&on is gg isolated Bom the accident unit. Non-isolation of aon-accident unit is assumed for this case ia order to tnsxindxe the dose ia the non~ident unit, As noted in PPN.'s August 1t4 1993 subrnitta4 the noncident unit um be )solated &en the reactor building HVAC recircubLtion plenum, thereby preveatial the spread of radhtion to the aoa-accident uait, The doses aro based oa the same calcuMoa used to obtain the doses ibr, the answer to question'.

The titae to ¹roke the RHR valves 6am Ml cloaxl to fbi'pen is 2 minutes per valve and a separate opentor would be seat to manipuilte each valve.

ATTACHMENT' to ~~

ThSLE2.1 SUMhQiRY OF ChLCULATED OPERATOR hCXHSS DOSES Fok RHR FUEL POOL GOOf ING MON-hQQIHKI'NIT IEhCfOR BUILKMs UNIT hS8)ST PR~

1 hGCKSS TO VALVES 151060 hND )5)070 NA%0RM KLBV.

705'PMh1QR ACCESS DOSES AT 24 BOURS MST-DKA (R)

HL CX AD DhMAGR HMC FllBLMELT DOSE LOCA'GON/

SOURCE Dole Fran Dose Faxn Dose Fry Dose Fexn Dose Fam Boae Fern hislxxae Coal mal hirbana Oat ma} Air@~ Caauaiaed Activhy Soars@ hctiviy 84HIKcs Activity Sees fl)

To Vahes I51060 A O.OZS 5.17-3 5.17-3. 0.714 5 17-3 151070 Op~m Say Tha At Vabel 151060 R 0.0231 6.67-3 6.67-3 0.634 6.67-3 151070 0.0491 0.0III 49l-4 0.011I 0.011$

Total I 0.03l2 5.43-3 0.7)9 6.91-3 0.641

')OTAL hCCEaS DOSE 0.0609 0.0]23 NOTES: ()) Post-LOCA adxrm radikioa doses 6x the )% dad damage of 0.01.

~ age obtaig)ed byy zgg~ggygg tbe I~ c)~ d ~~ ~ b Page 21

Fuel Pool Cooling Issue Back round

~ Issue raised by two contractors

~ Contractors contend that a loss of FPC, concurrent with LOCA results in:

boiling SFP and fuel melt outside containment loss of all ECCS and failure of containment integrity

~ PP &L maintains that adequate capability exists to respond to a loss of FPC event

Fuel Pool Cooling Issue

Background

~ Loss of normal FPC system due to:

Seismic LOOP.

-'LOCA

~ Normal FPC system is first line of defense

~ Other systems available to provide:

makeup to SFP cooling to SFP

Fuel Pool Cooling Issue L A

~ Hydrodynamic Loads

~ FPC not.designed for hydrodynamic loads

~ PP&L assessment concluded normal EPC system may remain functional after LOCA

Fuel Pool Cooling Issue PP L Position

~ SSES Licensing Basis does not consider loss of FPC for other than Seismic events

~ Operators have time to react (50 to 130 hrs)

~ Safety-grade makeup source always available

~ Safety-grade cooling via RHR FPC mode

~ Boiling environment can be mitigated

e TYPICAL MARK II CONTAINMENT VENT CONTAINMENT PRDBSli AL WALL 'O

-~: .e.~

ai<'.i

' 'V.

.Qb o'

p.'A h .ii R R oP 0.."

~ ~jt 6>>

~C i:

r Ago

.. JET 0 WELL WETWELL VENT SUPPORT SA!

n COLUMN 0 $

" " 4s

'<.i4 i" "4'

~ o)

~ ~ ~

~ %p,+ ~ +~ ~ ~ J t~.%':)

~ t ~ ~ 4I ~

SUSQIJKHANNA CONTAINMKNTDESIGN

+ GE MARK II CONTAINMENTDESIGN - DRYWELL OVER SUPPRESSION CHAMBER

+ REACTOR COOLANT PRESSURE BOUNDARY COMPONENTS LOCATED IN DRYWELL

+ 87 DOWNCOMERS TO ROUTE NON-CONDENSABLES AND STEAM FROM THE DRYWELLTO SUPPRESSION POOL DURING A LOCA

+ INTERNAL DESIGN PRESSURE - 53 PSIG

- 'F

+ SUPPRESSION CHAMBER DESIGN TEMPERATURE 220

+ DRYWELL DESIGN TEMPERATURE - 340

'F

LOCA HYDRODYNAMICLOADS

+ DESIGN BASIS LOCA: POSTULATED DOUBLE ENDED BREAK OF THE RECIRCULATION SUCTION PIPE IN THE DRYWELL

+ PRODUCES HYDRODYNAMICLOADS DUE TO FLOW OF NON-CONDENSABLES AND STEAM FROM THE DRYWELLTO SUPPRESSION POOL VIATHE DOWNCOMERS

+ POOL SWELL

+ STEAM CONDENSATION LOADS - CONDENSATION OSCILLATION (CO) AND CHUGGING

+ DISCUSSION THAT FOLLOWS ARE THE DESIGN BASIS LOCA HYDRODYNAMICLOADS PREVIOUSLY APPROVED BY THE NRC - NO NEW INFORMATION

POOL SWELL LOADS

+ NON-CONDENSABLE AT THE DOWNCOMER EXITS GROW AND COALESCE INTO PANCAKE BUBBLE AND CAUSE RAPID RISE IN POOL

+ MAXIMUMPOOL SWELL HEIGHT IS 18.2'BOVE INITIAL POOL SURFACE

+ POOL SWELL ZONE EXTENDS FROM DOWNCOMER EXIT ELEVATIONTO MAXIMUMPOOL SWELL HEIGHT

+ POOL SWELL CAUSES

- IMPACT, DRAG AND FALLBACKLOADS ON COMPONENTS G, DOWNCOMER (I.E., PIPING, D BRACING, ETC.) LOCATED IN POOL SWELL ZONE AIR BUBBLE LOADS ON SUPPRESSION POOL STRUCTURE OL SWELL DOES NOT PRODUCE LOADS ON STRUCTURES OR COMPONENTS C LOCATED OUTSIDE THE POOL SWELL XONE

0 POOL SWELL LOAD NIETHODOLOGY WATlh 1LL!0 lt BASED ON GENERIC MARK II LOAD METHODOLOGY

~

METHODOLOGYAPPROVED BY NUREG-0487 AND SUPPLEMENTS) AND NUREG-0808

~

POOI SWELL ANALYTICALMODEL PSAIVI TO CALCULATE PRESSURE, POOL VEI OCITY, ACCELERATION AND HEIGTH I

POOL SWELL AlR BUBBLE LOAD ELdH II Kt IFR 0

%SKINT r

~

LOAD STATICALLYAPPLIED TO CONTAINMENTBOUNDARY IN ACCORDANCE WITH NUREG-0487 AIR BUBBLE LOAD DOES NOT AFFECT COMPONENTS LOCATED IN REACTOR BUILDING

POOL SWELL LOADS ON COMPONENTS

+ AFFECTS COMPONENTS LOCATED IN WETWELL BETWEEN DOWNCOMER EXIT ELEVATIONAND MAXIMUMSWELL HEIGHT(PIPING, SUPPORTS, BRACING, ETC.)

+ LOADS ON VERTICALLYORIENTED COMPONENTS ARE NEGLIGIBLE AND ARE NOT CONSIDERED FOR DESIGN

+ VELOCITY AND ACCELERATION VS. TIME AND ELEVATION CALCULATEDWITH PSAM

+ IMPACT, DRAG AND FALLBACKLOADS BASED ON ACCELERATION, VELOCITY, COMPONENT ELEVATION AND SIZE

+ LOADS ON COMPONENTS ARE NOT TRANSMITTED TO REACTOR BUILDING- LOCAL AFFECTS ONLY

LOCA STEAM CONDENSATION LOADS

+ BASED ON GENERIC MAIMII ACOUSTIC METHODOLOGY APPROVED IN NUREG-0808

+ ACOUSTIC MODEL OF SSES KITH CHUG AND CO 'SOURCES'LACED AT THE DOWNCOMER EXITS

+ CHUG AND CO SOURCES DERIVED FROM PLANT-UNIQUE GKM-IIMLOCA TEST PROGRAJVl CONDUCTED BY KRAFTWERK UNION (KWU)

+ GKM-IIMTESTS COVERED SPECTRUM OF LOCA BREAK SIZES

LOCA STEAM CONDENSATION LOAD

+ KWU SELECTED 4 CHUGS AND I CO PRESSURE TRACE FROM THE GKM-IIMDATA FOR SOURCING

+ DESIGN CHUG AND CO 'SOURCES'PPLIED TO ACOUSTIC MODEL OF SSES SUPPRESSION POOL

+ PRESSURE LOADS CALCULATEDAT POOL BOUNDARY FOR EACH SOURCE

+ PRESSURE TIME HISTORIES INPUTTED TO STRUCTURAL MODEL OF CONTAINMENT

+ DBA LOCA OCA DOES D NOT PRODUCE THE LARGEST CI-IUGGING AND CO LOAD

+ LOCA LOAD IS THE ENVELOP OF ALLCHUG AND CO PRESSURE TIME HISTORIES FOR ALL BREAK SIZES

l' JAERl COIlPARlSON EBS Sfdddl.

4 ANU 344 AT l,dA 4 4 JAKAT CUUAS AT l dll 04 ru 4I',

ZI 41 6l 8I

. FAEGUENCY <hz)

~

SSES LOCA STEAM CONDENSATION METHODOLOGY COMPARED W)TH JAERI TEST DATA

~

JAERI TEST FACILITYWAS A 20'LICE OF A MARK ) I CONTAINMENT- 7 DOWNCOMERS SIGNIFICANT CONSERVATISM IN LOCA LOAD

PRESENTATION OUTLINE

~ DEVELOPMENT OF LOCA RESPONSE SPECTRA IN THE REACTOR BlJILDING

- LOCA ANALYSIS OF THE CONTAINMENT

- LOAD TRANSFER

- LOCA ANALYSIS OF THE REACTOR BUILDING

~ COMPARISON OF LOCA AND SSE RESPONSE SPECTRA IN THE REACTOR BUILDING

LOCA ANALYSIS OF CONTAINMENT

~ PRESSURE TIME HISTORIES FROM THE ACOUSTIC MODEL ARE CONVERTED TO NODAL FORCE TIME HISTORIES

~ FORCE TIME HISTORIES ARE APPLIED TO A 3D FINITE ELEMENT MODEL

- MODEL DEVELOPED FOR HYDRODYNAMIC LOAD ANALYSIS

- MODEL CONSIDERS SOIL STRUCTURE INTERACTION

~ DYNAMIC ANALYSIS PERFORMED USING ANSYS PROGRAM

~ ANALYSIS RESULTS (ACCELERATION TIME HISTORIES) USED TO DEVELOP INPUTS FOR THE REACTOR BUILDING ANALYSIS

I'

~ y

~

) I

~ C IS ~

NOTE:

RPV X ~ AXIS IS IN PLANT EW AND Y - AX!S IN PLANT NS DIRECTION RPV SHIElD CONTAINMENT RPV PEDESTAL Rev. 9, 07/85 SINQUEHANNA STEAM ELECTRIC NTJ UNITS 1 AND 2 DESIGN ASSESSMENT REPORT 3-D CONTAZNMENT FZNZTE ELEMENT MODEL (ANSYS MODEL)

LOAD TRANSFER

~ NO DIRECT COUPLING OF THE CONTAINMENT AND REACTOR BUILDING EXCEPT AT BASEMAT

~ HORIZONTAL MOTIONS ARE FULLY TRANSFERRED; ACCELERATION TIME HISTORIES AROUND THE PERIMETER ARE AVERAGED o VERTICAL MOTIONS ARE TRANSFERRED TO THE REACTOR BUILDING THROUGH THE ROCK. LOAD ATTENUATION (BASED ON DISTANCE AMAY FROM THE SOURCE) IS CONSIDERED

LOCA ANALYSIS OF REACTOR BUILDING

~ "AVERAGED" TIME HISTORIES ARE APPLIED TO THE BASE OF STICK MODELS

~ 3 STICK MODELS ARE USED

- ONE FOR EACH DIRECTION

- SAME MODELS AS THOSE USED IN THE SEISMIC ANALYSIS

~ TIME HISTORY ANALYSES PERFORMED USING SECHTEL IN-HOUSE PROGRAMS

~ RESULTING NODE POINT TIME HISTORIES ARE CONVERTED TO NODE POINT RESPONSE SPECTRA

~ LOCA FLOOR ENVELOPING RESPONSE SPECTRA ARE DEVELOPED

)

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~ SCOPE OF EVALUATION

1. FUEL POOL COOLING PIPING LOCATED IN REACTOR BLDG. ABOVE ELEV. 749~-0" 6" TO 10" MAIN LINES ASME/ANSI NON-SAFETY RELATED, NON-SEISMIC PORTIONS ARE SAFETY RELATED, SEISMIC

2. SERVICE NTER PIPING LOCATED IN REACTOR BLDG. ABOVE ELEV. 719~-0" 4" TO 14" MAIN LINES ANSI NON-SAFETY RELATED, NON-SEISMIC

3. CONDENSATE PIPING LOCATED IN REACTOR BLDG. ABOVE ELEV. 683>-0" 1" TO 6" MAIN LINES ANSI: NON-SAFETY RELATED, NON-SEISMIC

~ HYDRODYNAMIC LOAD INPUT

- MAXIMUM HORIZONTAL ACCELERATION = 0.9 G'S

- MAXIMUM VERTICAL ACCELERATION = 0.5 G'S

- PEAK ACCELERATIONS OCCUR AT FREQUENCIES + 20 Hz.

PIPE STRESS AFFECTS

- NON-SEISMIC PIPING FREQUENCY TYPICALLY + 11 Hz.

- DISPLACEMENTS OF PIPE ANTICIPATED TO BE SMALL DUE TO HIGH FREQUENCY INPUT

- PREDICT ACCEPTABLE PIPE STRESSES DUE TO LOM DISPLACEMENTS

- REVIEM OF SAFETY RELATED ANALYSIS DEMONSTRATES LOM MAGNITUDE PIPE DISPLACEMENTS/STRESSES DUE TO HYDRODYNAMIC LOADS

~ PIPE SUPPORT ADEQUACY

- HANGERS ARE MAINLY COMPONENT TYPE SUPPORTS (SPRING CANS, STRUTS, RIGID RODS, KTC.)

- OTHER SUPPORTS COMPRISED OF STRUCTURAL STEEL MEMBERS

- PIPE SUPPORT CATALOG COMPONENTS TYPICALLY HAVE LARGE SAFETY FACTORS

- PREDICT SMALL PIPK SUPPORT LOAD INCREASES DUE TO INPUT MAGNITUDE

- REVIEW OF SAFETY RELATED ANALYSES IN THE R/B DEMONSTRATES THAT HYDRODYNAMIC LOADS ARK LESS THAN 25% OF DEADWEIGHT LOADS

~ SPATIAL INTERACTION/ANCHOR MOVEMENTS

- SMALL PIPE DISPLACEMENTS WILL MINIMIZE ADVERSE AFFECTS DUE TO:

DIFFERENTIAL ANCHOR MOVEMENTS IMPACT OF ADJACENT COMPONENTS/SYSTEMS SLIPPAGE OF PIPE OFF SUPPORTS