Text

Responds to 940421 RAI Re Loss of Spent Fuel Cooling for Sses.Concludes Facility Has Multiple Means to Assure Makeup for Dba/Loca
	ML18026A263
Person / Time
Site:	Susquehanna
Issue date:	05/05/1994
From:	Bryam R; PENNSYLVANIA POWER & LIGHT CO.
To:	Chris Miller; Office of Nuclear Reactor Regulation
References
	PLA-4134, TAC-M85337, NUDOCS 9405170161
	Download: ML18026A263 (92)
	v • d • e

ACCELERATED D BUTION DEMONS TION SYSTEM

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ACCESSION NBR:9405170161 DOC.DATE: 94/05/05 NOTARIZED: NO DOCKET FACIL:50-387 Susquehanna Electric Station, Unit 1, Pennsylva 50-388 Susquehanna AUTH. NAME AUTHOR Steam Steam Electric Station, Unit AFFILIATION 2, 'sylva 05000387 05000388 BRYAM,R.G. Pennsylvania Power & Light Co.

RECIP.NAME RECIPIENT AFFILIATION

-MILLER,C. L. Office of Nuclear Reactor Regulation,.Direct r (Post 870411 R

SUBJECT:

Responds to 940421 RAI re loss of spent fuel cooling for SSES.

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Pennsylvania Power & Light Company Two North Ninth Street ~ Allentown, PA 16101-1179 ~ 610/774-5151 Robert G. Byram Senior Vice President Nuclear 610/774-7502 Fax: 610/774-5019 MAY 05 1994 Director of Nuclear Reactor Regulation Attn: 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 ELE<CTRIC STATION RE<SPONSE TO REQUE<ST FOR ADDITIONALINFORMATION, CONCERNING LOSS OF SPENT FUEL POOL COOLING Docket Nos. 50-387 PLA-4134 FILE R41-2 and 50-3SS

Reference:

NRC Letter, L W. Siiea to R. G. Byram "Rerprest for ArlditionalInformation (RAI) concerning Loss of Spent Fuel Pool Cooling, Susrluehanna Steam Electric Station, Units 1 anrl 2 PAC NO. MS5337)," dated April21, 1994.

Dear Mr. Miller:

In the above reference, the staff determined that the addition of make-up water to the spent fuel pool (SFP) from the emergency service water (ESW) system during a design basis loss of coolant accident (LOCA) and/or loss-of-offsite power (LOOP) event is a commitment documented in the licensing basis for the Susquehanna SES (SSES). Based on this determination, the staff requested information to demonstrate that this make-up can be supplied to the spent fuel pools from the ESW system under design basis accident (DBA) conditions (specifically, design basis LOCA).

As stated in previous PPkL.submittals, the engineered safety grade make-up system for fuel pool cooling for Susquehanna SES is the Emergency Service Water system, ESW is a system common to both Units with valves, provided to direct fuel pool make-up, located in each unit.

The use of ESW for supplying fuel pool make-up involves the manipulation of three manual valves (2") per loop of ESW. Susquehanna SES also has a common refueling floor with the spent fuel pools of both units hydraulically connected so that water from one unit's pool can flow to the other unit.

Initiation of ESW make-up to the non-accident SFP will result in both pools being filled when water level is raised above the height of the weirs in the spent fuel pool for the non-accident unit. With the pools isolated, water added to one pool will overflow to its skimmer surge tank; which when completely filled will overflow to the cask storage pit; which will overflow to the accident unit's skimmer surge tank; which in turn will overflow into the accident unit's SFP.

Instructions exist in the procedures to fill above the weirs, thereby completely filling the skimmer surge tank.

'405170161 .,940505",<

PDR 'DOCK 05000387 PDR'"

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FILE R41-2 PLA-4134 Mr. C. L. Miller In addition to ESW, there are several systems which can be used to provide water into the fuel pool based on system design and/or location. Specifically, make-up water can be added from the Condensate Storage, Tank (CST) or Residual Heat Removal Service Water (RHRSW).

Demineralized Water and Fire Water can also be called upon to add water as required. Thus, fuel pool make-up for both units can be accommodated in either Unit. In response to a LOCA event in one Unit, we would provide fuel pool make-up from the non-accident Unit and not require operators to be exposed to accident conditions.

However, PP&L has analyzed access to the ESW valves in the accident unit to address the unlikely situation that the non-accident unit ESW valves are unavailable and has concluded that even using a Regulatory Guide 1.3 source term, operator access is possible. Specific details are provided in the attached report. This report, as previously stated, also addresses all seven (7)

NRC's questions contained in the RAI, question by question. It should be noted that the specific scenario is not the same for each question. This is due to the nature of the questions asked within the RAI, which requested specific information for a seismic event and LOCA events both within and outside the SSES licensing basis. Therefore, the information provided as a response to one question should not be automatically applied to the next.

The actions required to support use of ESW in response to a loss of SFP cooling are identified in plant procedures generated to support operation of both SSES Units. Procedures enhancements have been made to provide greater attention to a loss of SFP cooling and address the need to restore cooling. These procedure enhancements are also incorporated into the operator re-qualification training program at SSES.

Therefore, PP&L concludes that:

~ With respect to makeup for a DBA/LOCA Susquehanna SES has multiple means to assure make-up.

~ The Emergency Service Water valves that support make-up to the Spent Fuel Pool have been determined to be accessible from the affected unit (i.e. (5 Rem) even if a Regulatory Guide 1.3 source term is assumed.

~ Emergency Service Water make-up to one Spent Fuel Pool will eventually fill both pools.

Therefore, access to the accident unit is not required to provide make-up to its Spent Fuel Pool.

Questions regarding this response should be directed to Mr. J. M. Kenny at (610) 774-7904.

Very truly yours, R.. B ram Attac ent

k, FILE R41-2 PLA4134 Mr. C. L. Miller cc: SRC Document Control Desk (original)

NRC Region I Mr. G. S.Barber, NRC Sr. Resident Inspector Mr. C. Poslusny, Jr., NRC Project Manager

ATTACHMENTto PLA-4134 RESPONSE TO NRC APRIL 21, 1994 RAI CONCERNING LOSS OF SFP COOLING SUSQUEHANNA STEAM ELECTRIC STATION UNITS 1 AND 2 May 5, 1994 I'9405170161

ATTACHME<NTto PLA-4134 This document provides PP8cL's response to the seven (7) specific questions identified in the enclosure to the NRC's April 21, 1994 Request for Additional Information (RAI) concerning a Loss of Spent Fuel Pool Cooling. It should be noted that the specific scenario is not the same for each question. This is due to the nature of the questions asked by the NRC, which requested specific information for both seismic and LOCA events. Therefore, the information provided in response to one question should not be applied to another question.

Page 1 of 25

ATTACHMENTto PLA-4134 N~RC UESTION 1 In a letter dated May 24, 1993, PP&L stated that "the normal SFP cooling system will automatically be shed from the plant electrical system, along with other non-safety-related equipment, to permit the start-up of the large emergency core cooling system (ECCS) pumps on the LOCA unit." In a letter dated March 21, 1994, PP&L provided clarification that indicated that the service water pumps are lost due to the auxiliary load shed feature and that the cooling function of the fuel pool cooling system will no longer be provided for the accident unit only.

The staff requests that you provide a comprehensive description of the response and operation of the spent fuel pool cooling and cleanup system following a LOCA. Please address the following issues as a minimum:

a. The March 21, 1994 letter indicated that loads such as the service water system can be restored 10 minutes after initiation of the event and that complete restoration of the service water system would take approximately one 12-hour shift. Please describe what activities, including relevant procedures, would be specifically required to restore service water to the fuel pool cooling system and the expected duration of these activities.

b. Describe the expected response and operation of the fuel pool cooling system following the loss of service water. Describe the expected system response, including system heat-up and any impact of system heat-up as well as expected or necessary operator manipulation of the fuel pool cooling system following the loss and restoration of service water.

RESPONSE TO NRC UKSTION 1a This response assumes a LOCA occurs while the plant is operating at 100% steady state power.

A Loss of Offsite Power (LOOP) is not assumed to occur coincident. As a result of the LOCA, a generator lockout will occur resulting in an auxiliary equipment load shed ("aux load shed").

The "aux load shed" will result in the trip of the Service Water (SW) pump supply breakers, thereby terminating cooling water flow to the SFP cooling heat exchangers. The Service Water pumps are supplied power from the 13.8 KV switchgear 1A101 and 1A102 which is located on the 699'levation of the turbine building. The pump supply breaker lock out relays have to be manually reset (they do not automatically reset). The supply breakers for the Spent Fuel Pool Cooling (SFPC) pumps willnot be tripped by the aux load shed and will continue to operate even though it will no longer be providing cooling to the LOCA Unit's SFP. The non-LOCA Unit's Service Water (SW) and SFPC systems will be unaffected by the LOCA Unit and will continue to operate.

The operations staff is trained and is sensitive to the need to get SW back as soon as possible, as it eases event recovery by making BOP systems available for use during the recovery.

Restoration of the SW system could begin as early as 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following a LOCA event and could take from 4-6 hours. The service water system restoration activities, except for venting of the fuel pool cooling heat exchangers and reactor building chillers, do not require reactor building access. Venting of these heat exchangers is done for optimizing system performance, not for waterhammer protection. Considering the potential post accident environment in the reactor building, optimization of fuel pool cooling heat exchanger and reactor building chiller operation Page 2 of 25

ATTACHMENTto PLA-4134 may occur at some later time. Area radiological assessments would be performed to determine accessibility. The recovery would be accomplished through implementation of procedures ON-135-001 "Loss of Fuel Pool Cooling/Coolant Inventory", ON-111-001, "Loss of Service Water",

and OP-111-001, "Service Water System" The principal actions to be taken to restore cooling

~

to the SFPC system are:

1) ON-111-001 "Loss of Service Water" would be entered upon loss of service water. It is not expected that ON-111-001 would be entered through ON-135-001 (Loss of Fuel Pool Cooling) since restoration of SW is desired for restoration activities that would occur prior to the need to reestablish fuel pool cooling. Instruction 3.3.2 of ON-111-001 "Loss of Service Water" requires checking of the Service Water pump supply breakers for trips and lockouts. The breakers are located on the 699'levation of the turbine building. The lockout relays would then be reset. This activity is not expected to take place prior to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the LOCA (to assure ECCS pump voltage would not be affected). The ON then directs pump restart to occur per OP-111-001 "Service Water".

2) With power now available to the pumps, Section 3.1 "Setup of Service Water System for Normal Operation" in OP-111-001 "Service Water System " would be implemented.

Note that once it is confirmed that the pumps are operating satisfactorily, the various user chillers and heat exchangers are to be vented to assure optimal heat exchanger and chiller performance in Section 3.1.25. Venting of all heat exchanger and chillers would not be required since some are below the cooling tower basin elevation. The fuel pool cooling heat exchangers and reactor building chillers would be expected to need venting for optimum system performance. In order to perform this venting activity for the fuel pool cooling heat exchangers, the reactor building would be entered. These heat exchangers are located on the 749'levation. This venting would take 5-10 minutes per heat exchanger. The reactor building chillers are also located in the reactor building on would be vented. It would take between 4 - 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to implement Section 3.1 of the 749'nd OP-111-001.

The only activities listed above that require access to the reactor building are the venting of the SFPC heat exchangers and the Reactor Building chillers.

The dose the operator would receive venting the fuel pool cooling heat exchangers has not specifically been determined as it is largely dependent upon when the venting action is taken.

However, it is expected to be less than the maximum dose calculated to be received by an operator restoring SFPC as they are in the same general plant location. This dose has been calculated and is included in PLA-4069 dated 1/4/94.

Note that the preceding evaluation discusses the Unit 1 response and procedures. Unit 2 response would be identical except Unit 2 procedures would be utilized.

Page 3 of 25

ATTACHMENTto PLA-4134 RESPONSE TO NRC UESTION 1b As noted in the response to question la, the SFP cooling system would continue to operate following a LOCA without a LOOP and a loss of service water. The operators are not procedurally directed to trip the SFPC pumps following a loss of service water, unless service water cannot be restored. Since this question is concerned with restoration of service water, the actions for loss of service water without restoration of SFPC system will not be considered.

Since the SFPC system remains in operation, the only actions required will be periodic make-up to compensate for evaporative losses. Make-up would not be necessary until after service water is restored. An evaluation of the impact of operation of the SFPC system without cooling capability is provided below.

TIME ESTIMATE - FUEL POOL MAKEUP For this evaluation, the following is assumed

1) ESW is the only available makeup source.

2) Fuel Pool Cooling system is operating service water is not, thus no pool cooling is occurring.

3) Isolated fuel pool

4) LOCA

5) The pool decay heat will be assumed to be 8.2 MBTU/HR. This is the heat load utilized in the evaluation supporting response 2 in PLA- 4069 dated 1/4/94.

6) The evaporation rate is 5787 ibm, which is conservatively based on a heat load of 12.6 MBTU/HR.

OP-135-001 R17 and OP-235-001 R16 (Fuel Pool Cooling) identify in section 3.2.7 that skimmer surge tank level should be maintained between 67% to 90%. Operators make-up approximately 10% skimmer surge tank level once per shift, to maintain level at 90%. It will therefore be assumed that the tank is filled to the 90% level and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months later the event occurs before makeup is provided with 10 % evaporative loss occurring. Thus, the tank level is 80 %

full.

Until the Service Water system (SW) is restored to service, the Fuel Pool Cooling system pumps would remain in operation. The fuel pools would heatup due to the decay heat from the fuel and from the operating fuel pool cooling pumps. The heat added to the fluid by the operating pumps is the friction horsepower (FHP). FHP is the difference between the brake horsepower (BHP),

which is the power delivered to the pump shaft, and the hydraulic horsepower (WHP), which is the pump output. The pump efficiency (N,) is equal to WHP/BHP. From pump curves for the 1P211A and 1P211 B pumps at 600 GPM flowrate, it can be seen that BHP = 50 and Np =60%.

Based on this information, the heat added by the SFPC pumps is 152,700 BTU/HR. This is then added to the SFP decay heat of 8,200,000 BTU/HR, which greatly exceeds the heat added by the pumps. This combined heat load will result in a temperature increase of 67.2'F over the time period between the loss of Service Water and its restoration. Thus, the SFP will at most peak at 177'F.

Page 4 of 25

ATTACHME<NTto PLA-4134 During this heatup period of time, the pool evaporation will increase. Calculations shows that the fuel pool level and skimmer surge tank level would be effectively constant due to the thermal expansion effects of the heatup. Once service water was started and cooling restored, the level would drop due to thermal contraction. Assuming this contraction occurred instantly down to the pool temperature at the start of the event, the level loss would be 700 gallons. This will result in a reduction in tank level of 25.7 inches, or 9.7 % of tank level.

Using 10 % /12 hrs to be conservative; skimmer surge tank level drop as a function of time is:

Indicated Time hours Tank Level ~Activi

-12 90% Makeup provided.

80% Event occurs prior to makeup.

12 70% Begin restoring Service Water 15 66% Low level alarm 60% Service Water Restored 11% Pump trip setpoint.

The low level alarm is set for 66% (193"). The low level pump trip for NPSH pump protection occurs at the 11% level (30").

f Based on the above, ample time exists to provide makeup prior to reaching the low level pump trip which would not occur earlier than 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months after event occurrence. Therefore, actions for providing make-up via ESW are not required until 24 to 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months .

Page 5 of 25

ATTACHMENTto PLA-4134 UKSTION 2 In previous discussions with the staff, including discussions on July 8, 1993, PPAL indicated that one possible action for mitigation or preventing the spread of vapor from one (or two) boiling spent fuel pools entails shutting down the reactor building recirculation fans. Please describe your plans to provide procedures to perform these activities including guidance to the emergency response staff and/or system-specific procedures.

RKSPONSK TO UK<STION 2 EP-PS-102 "Technical Support Coordinator: Emergency Plan-Position Specific Procedure" Tab I presently addresses the Fuel Pool Boil event. The major task of this section is to "Determine if Fuel Pool boiling can be expected and initiate actions as necessary to prevent Fuel Pool boiling or to mitigate the consequence of Fuel Pool boiling." This procedure will be revised to reflect the following:

1. If a loss of fuel pool cooling event occurs and cooling cannot be restored, yet no source term is present (ie, seismic event with a LOOP, both fuel pools expected to boil, makeup available) then the following actions will be required.

(a) Isolate Zone(s) 1 and 2 from the Recirculation Plenum to preclude any spread of high temperature/high humidity environment to other areas of the Reactor Building, and (b) Shutdown the Reactor Building Recirculation and SGTS fan(s), and (c) Vent the refueling floor directly to atmosphere.

Actions (a) and (b) will be accomplished in the Control Structure.

Action (c) will be accomplished in the Reactor Building.

2. If a loss of fuel pool cooling event occurs and cooling cannot be restored and a source term is present (ie; LOCA/LOOP, one fuel pool expected to boil, makeup available) then the following actions are required.

(a) Shutdown the Reactor Building Recirculation Fan(s), and (b) Maintain SGTS in service, and (c) Any zone that automatically aligned to the Recirculation Plenum shall remained aligned.

Action (a) will be accomplished in the Control Structure.

The above procedure changes require additional evaluations to determine the appropriate time to turn off the recirculation fans. Due to COTTAP modeling constraints it will take approximately 4 months to complete the code revision and perform the analysis. It will take an additional month to prepare and approve the procedure change following the analysis.

Therefore, PPAL commits to completing the procedure changes identified above by October 1, 1994.

Page 6 of 25

ATTACHMENTto PLA-4134 UKSTION 3 The staff has reviewed additional portions of the licensing basis documentation. Section 9.2.5 of the Final Safety Analysis Report (FSAR) describes the emergency service water system (ESW) and Section 3.1.2.4.15 describes compliance of the facility design with General Design Criteria 44. Section 3 1.2.4.15 states:

~

"The emergency safeguard service water system, which comprises both the Emergency Service Water System and the Residual Heat Removal Service Water system, provides cooling water for the removal of excess heat from all structures, systems and components which are necessary to maintain safety during all abnormal and accident conditions. These include the standby diesel generators, the RHR pump oil coolers and seal water coolers, the core spray pump room unit coolers, [reactor core isolation cooling] RCIC pump room unit coolers, the [high pressure coolant injection] HPCI pump room unit coolers, the [residual heat removal] RHR heat exchangers, RHR pump room unit coolers, emergency switchgear and load center room coolers, the control structure chiller and the fuel pool make-up."

Section 9.2.1 of the SER describes the above function of the ESW system and states:

"The emergency service water system is an engineered safety features system designed to supply cooling water to the emergency diesel generators, residual heat removal pumps and to those rooms identified below that are required during normal and emergency conditions to safely shutdown the plant. The emergency service water system takes water from the spray pond (ultimate heat sink), pumps it to the heat exchangers which serves the above components or systems and returns it to the spray pond by way of a network of sprays.

The emergency service water system is required to supply cooling water to the residual heat removal pumps room coolers, residual heat removal pump bearing oil coolers,... and to the spent fuel pools as emergency makeup...

... Therefore a failure of the nonsafety related piping coupled with any single active failure of the safety-related Emergency Service Water System will not preclude one of the loops from performing its function. By providing this isolation capability and redundancy in components, we conclude that the requirements is General Design Criteria 44 "Cooling Water" are met, including the single active failure criterion."

The staff concludes that provision of make-up to the spent fuel pool during a design basis accident, including design basis loss-of-coolant accident, is within the design and licensing basis of the SSES facility. The staff review did not conclude that boiling of the spent fuel pool was necessarily implied but recognizes that make-up to the pool will be necessary to compensate for, at the very least, evaporative losses.

The staff requests that you provide information to demonstrate that make-up can be supplied to the spent fuel pool from the ESW system under design basis accident conditions (Specifically, design basis LOCA). The staff has noted the dose estimates that you supplied in previous correspondence regarding activities necessary to manipulate the ESW-fuel pool make-up valves under various accident conditions. The staff requests that you reanalyze the performance of these activates using design basis assumptions. Please specifically describe the correlation between the point in the accident sequence time line that make-up actions would be required, and times in the accident sequence when expected operator dose would exceed design basis limits.

Page 7 of 25

ATTACHMENTto PLA-4134 RE<SPONSK TO UKSTION 3 A radiation dose analysis was performed to evaluate personnel access doses inside the reactor building for providing Emergency Service Water (ESW) make-up to the spent fuel pool under DBA-LOCA accident conditions, without a LOOP. This analysis evaluates the adequacy of the reactor building radiation shielding design and addresses the operator access doses from contained radiation sources. Consistent with SSES FSAR Chapter 18.1.20, which was performed as a required response to Item II.B.2 of NUREG-0737, post-LOCA airborne radiation doses are not addressed in this calculation.

Operator access doses are calculated and based on point specific dose rates which are determined from the actual locations of the radiation source terms and the proximity of operator access routes in relation to those sources. Detailed access dose analyses were performed at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-LOCA for both ESW tie-in and ESW flow control missions. Access doses at other time periods were then evaluated by multiplying the dose results at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by the ratio of the radiation source term at the time period of interest to the source term at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Analyses were performed to determine the time post-LOCA at which ESW make-up should be initiated. It was determined that access can occur as early as 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the DBA-LOCA, and as late as 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months post-LOCA. To evaluate the accident sequence time line, 3 times were evaluated: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post LOCA, 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months post LOCA and 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months post LOCA.

The mission dose for access to Elevation 670'f the reactor building to tie-in the ESW system for make-up to the spent fuel pool was determined to be 7.27 Rem at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-LOCA. For operator entry at 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months post-LOCA, using the scaling procedure described above, the mission dose was found to be 4.8 Rem. If entry is delayed until 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months post-LOCA, the mission dose decreases to 2.75 Rem.

The operator mission dose for access to Elevation 749'f the reactor building to control ESW make-up flow to the spent fuel pool at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-LOCA is 1.41 Rem. Note that this is a separate and later action required after ESW tie-in has occurred. It will also be performed on a periodic basis for the duration of the event. Although not specifically evaluated at 40 and 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months , mission doses for control of ESW makeup to the spent fuel pool would decrease for time periods greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-LOCA.

The dose acceptance criteria in NUREG-0737, Item II.B.2 states that doses should not exceed 5 Rem to the whole body or its equivalent. With this limit in mind, and if conditions warrant, operator access to provide ESW make-up to the spent fuel pool should be delayed at least until 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months . Ample skimmer surge tank capacity exists to allow delaying operator access for make-up in order to minimize dose. However, this restriction must be weighed by a prudent ALARA review, and if necessary, entry earlier than 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months should be expected.

Susquehanna is designed such that initiation of ESW make-up to either SFP results in both pools being filled regardless of whether or not the pools are cross-tied. With the pools isolated, water added to one pool will overflow to its skimmer surge tank; which when completely filled will overflow to the cask storage pit; which will overflow to the opposite units'kimmer surge tank; which will in turn overflow to the opposite units SFP. Therefore, operator access to the LOCA unit is not required to assure make-up to the LOCA unit SFP and doses evaluated in this response would not be experienced.

Page 8 of 25

ATTACHMENTto PLA-4134 The following summarizes the assumptions and data used in the above mission dose evaluation:

1. This analysis is based on Design Basis Loss-Of-Coolant Accident (DBA-LOCA) conditions.

2. The activity source term for this analysis is based on the requirements on NUREG-0737, Item II.B. for post-LOCA liquid containing systems and is 50% of the core equilibrium halogen activity inventory and 1% of the core equilibrium particulate activity inventory released to the suppression pool water. Post-LOCA airborne radiation sources are not considered.

3. The equilibrium core inventory is based on power uprate conditions. The SSES power uprate core thermal power level for Design Basis Accident Analysis is 3616 MWt which is 105% of the uprated core thermal power.

4. Operator access doses are based on point specific dose rates which are determined from the actual locations of the radiation source terms and the proximity of operator access routes in relation to those sources. Detailed analyses were performed at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-LOCA. Access doses at other time periods were evaluated by multiplying the dose results at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by the ratio of the radiation source term at the time period of interest to the source term at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5. In order to provide ESW makeup to the spent fuel pool, operator access is required to valves in the Fuel Pool Cooling System to initially tie-in flow from the ESW system and then to control the ESW makeup flow rate.

System Tie-in:

UNIT 1: Open either valve 153500 or 153501.

UNIT 2: Open either valve 253500 or 253501.

ESW Makeup Flow Control:

UNIT 1: Control flow with valves 153090A&B or 153091A&B UNIT 2: Control flow with valves 253090A&B or 253091A&B Note: For flow control, both the 090 and 091 valves must be opened in each ESW supply line being used for makeup.

6. Two separate operator access missions are assumed in order to provide ESW makeup to the spent fuel pool under LOCA conditions. One operator access mission is required to tie-in the ESW system to the spent fuel pool and is a one time access requirement. The second operator access mission is required to control ESW system makeup flow and is a periodic access requirement following system tie-in. These missions are as follows:

Page 9 of 25

ATTACHMENTto PLA-4134 Access for S stem Tie-In One-time operator access to the following areas is required to tie-in the ESW system for makeup to the spent fuel pool:

UNIT 1: Reactor Building Elev. 670'-0", Equipment Area, Room 8 I-105 Open valve 153500 and 153501.

UNIT 2: Reactor Building Elev. 683'-0", Closed Cooling Water Heat Exchanger/Pump Room, Room 8 II-203 Open valve 253500 and 253501.

ESW makeup to the spent fuel pool can be provided by opening either valve 153500(253500) or 153501(253501). It is conservatively assumed for this dose analysis that both valves are opened by the operator during access to elevation 670'f the reactor building in order to make both ESW loops available for supplying makeup to the spent fuel pool.

Access for Flow Control Following ESW system tie-in, periodic operator access to the following areas is required to control ESW system makeup flow to the spent fuel pool:

UNIT 1: Reactor Building Elev. 749'-1", Fuel Pool Cooling Heat Exchanger Pump Room, Room 8 I-514 Only one ESW loop is required to provide make-up to the spent fuel pool.

Open fuel pool cooling and clean-up valves 153090 and 153091 in one ESW supply loop. Control flow with valves 153090 or 153091 in this loop.

UNIT 2: Reactor Building Elev. 749'-1", Fuel Pool Cooling Heat Exchanger Pump Room, Room 8 II-514 Only one ESW loop is required to provide make-up to the spent fuel pool.

Open fuel pool cooling and clean-up valves 253090 and 253091 in one ESW supply loop. Control flow with valves 253090 or 253091 in this loop.

Page 10 of 25

ATTACHMENTto PLA-4134

7. Plant operating procedures require that makeup to the spent fuel pool be made for skimmer surge tank levels between 67% and 90%. Following a DBA-LOCA, the sequence of events for losses from the spent fuel pool are:

Indicated Time hours Tank Level ~Activit

-12 90% Makeup provided.

80% Event occurs prior to makeup.

12 70%

15 66% Skimmer surge tank low level alarm setpoint.

24 60%

82 11% Pump trip setpoint.

The skimmer surge tank low level alarm is set at the 66% water level. The low level pump trip for NPSH pump protection occurs at the 11% water level. Ample time exists to provide makeup prior to reaching the low level pump trip which would not occur until approximately 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months after event occurrence. Therefore operator access to provide ESW make-up is not required prior to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months post-LOCA, but must be completed within 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months post-LOCA. Thus, Operator access doses are evaluated at 24 and 82 hours9.490741e-4 days 0.0228 hours 1.35582e-4 weeks 3.1201e-5 months post-LOCA.

8. A time motion study was performed to determine operator access and 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. Results of this time motion study are used to evaluate operator access doses.

9. Dose acceptance criteria for personnel access to a vital area during the course of an accident is given in NUREG-0737, Item II.B.2 and states that the doses should not exceed 5 Rem whole body or its equivalent to any part of the body for the duration of the accident.

Page 11 of 25

ATTACHMENTto PLA-4134 UKSTION 4 Earlier PP&L submittals have identified that a SFP with no operable cooling system may be cooled by natural convection through the cask storage pit by cooling systems associated with the remaining SFP. The staff requests that PP&L describe the basis (i.e., test results, calculational results, or operational experience) for concluding that this method of cooling is adequate. The staff also requests that PP&L estimate the temperature difference between the pools assuming that the decay heat load in the SFP without an operable cooling system is 6.2x10'TU/hr.

RKSPONSK TO UKSTION 4 In earlier submittals to the NRC, PP&L has identified that a SFP with no operable cooling system is adequately cooled by natural circulation through the cask storage pit using cooling from the other SFP. However, upon further review PP&L has determined that the outage unit's service water system is shutdown, but the SFPC system remains in operation during outage periods.

Therefore, even though no cooling is provided by the outage unit's SFPC system, it will aid in mixing the outage unit's SFP. The ability of the non-outage SFPC system to cool both SFPs is based upon plant operational data taken during refueling outages. The outage operations are controlled by TP-135(235)-011 which maintains the SFPC pumps in operation. The SFPC operating procedure, OP-135(235)-001 directs the operator to shutdown and isolate the SFPC system of a unit when the other unit is providing cooling to both SFPs via the cask storage pit.

While the configuration identified in the operating procedure is not benchmarked against plant operational data, PP&L believes that the results would be similar to those obtained during outages. The basis for this is explained in more detail below.

Test procedure TP-135(235)-011 is performed at each refueling outage to monitor fuel pool temperature and heat load. As noted above, this procedure only isolates Service Water to the SFPC heat exchangers and keeps the outage unit's SFPC pumps in operation. The data has been recorded for the past three (3) refueling outages and indicates that the temperature difference between the two SFPs has generally been less than one degree (1'F) throughout the duration of the outage. The in-plant tests have shown that this temperature difference between the two pools can be maintained with a heat load of approximately 20x10'tu/hr in the spent fuel pool without an operable cooling system. This data demonstrates that adequate thermal mixing occurs between the pools via natural circulation.

No significant difference is expected for lower heat loads. Mixing is inherent in the geometry of the fuel pool as the heated water rises from between the spent fuel bundles and mixes with the fluid above. The water exiting the spent fuel region is replaced by cooler fluid drawn down between the fuel and the pool walls, and a natural circulation flow is established. While a lower heat load will reduce the magnitude of the velocities induced by natural circulation through the spent fuel bundles, mixing will continue due to the nature of buoyancy and the hotter fluid will continually rise towards the free surface. The fuel pool cooling system takes a suction on the spent fuel pool via a weir at the free surface, drawing the warmest water from the pools. The in-plant tests have shown that this arrangement can effectively draw the hotter fluid from the opposite pool. A lower heat load will not affect the potential of the skimmer arrangement to

'draw fluid towards itself. As fluid is skimmed off of the surface, buoyancy acts to maintain a hot layer of essentially uniform thickness among all pools in communication. The temperature of one pool will not get significantly hotter than the other as long as the fluid near the top of both pools is in communication. Buoyancy will act to drive the hotter fluid to a stratified layer Page 12 of 25

ATTACHMENTto PLA-4134 from which the skimmer will draw its suction. The same physical process will result at lower heat loads. Therefore, the temperature difference between the pools with a decay heat load of 6.2x10~ BTU/hr in the spent fuel pool without an operable cooling system is expected to be on the order of 1 degree Fahrenheit.

During the in-plant tests, both fuel pool cooling systems were iri service, however, the cooling function of the reload unit's pool was removed by shutting down service water flow to its heat exchangers. The reload unit's fuel pool cooling system provided some mixing effect but no cooling to its pool. Operation of the fuel pool cooling pumps on the reload unit recirculated a portion of the fluid within the pool keeping it from migrating to the opposite pool for cooling.

While one of the fuel pool cooling return lines discharges towards the cross-tie point and could contribute to fluid transport to the opposite pool, the fluid must travel the width of the pool to reach the cross-tie point and the lateral velocity would diffuse allowing the fluid to be drawn downward by the natural circulation patterns. In addition, the fuel pools are a mirror image of each other with respect to the cross-tie point and the discharge momentum imparted to the fluid in both pools would tend to cancel each other. The less internal recirculation there is in the reload unit's fuel pool the more influence the opposite unit's cooling system can have on it and the greater the potential for fluid transport between the pools. Therefore, the cooling capability demonstrated by the in-plant tests should also result with no fuel pool cooling flow on the reload unit.

Page 13 of 25

A ATTACHMENTto PLA-4134 UESTION 5 Bechtel specifications for the watertight doors between the "A" core spray/reactor building sump room and other ECCS pump rooms reviewed during an audit on February 7, 1994, indicated that an unseating pressure of zero was specified for certain water tight doors. However, PP&L Calculation EC-035-0510 stated that the relevant watertight doors provide protection to 15 psid based on a pre-delivery hydrostatic test. Because the subject watertight doors were credited with preventing flooding of ECCS pump rooms in certain analyses, the staff requests that PP&L clarify the apparent inconsistency between the specification and the assumed performance of the watertight doors in Calculation EC-035-0510.

RE<SPONSK TO UK<STION 5 During fuel pool boiling scenarios, condensate from the 818'levation would drain to the basement of the reactor building. The areas into which this flow would drain are defined by the outline of the respective Unit's reactor building sump and "A" Core Spray pump rooms. Note that although there is a fire door separating these two areas, it is not watertight, and hence both areas would flood concurrently.

Watertight doors are located between the ECCS pumps rooms on the 645'levation to provide for flooding protection. These doors were procured for the Susquehanna Units under Bechtel Specification No. 8856-A-16 (Watertight Doors). In this document, both seating and unseating pressures for each door were specified. In addition, the specification also required that each door be designed and fabricated to withstand a test pressure of 1.25 times the specified pressure. Table 1 identifies the doors affected during the fuel pool boiling scenario, along with the appropriate information from Specification 8856-A-16.

Prom this table, it is seen that the lowest ~secified seating pressure is <0 paid, which corresponds to a water height of 23 feet. Therefore, this was the maximum allowable height of water used in the reactor building flooding assessment (PP&L Calculation EC-035-510, Revision 1).

In order for Door Ps 25 and 26 to contain the area flooded as described above, they are required to prevent leakage in the unseated direction. From the data in Table 1, it is seen that the original specification did not specify this as a requirement. However, all doors procured and delivered under this purchase order are of identical design.

The only difference between the doors which were required, or not required to be leak-tight in the unseated direction is the scope of post fabrication testing. All doors which were specified to be leak tight in both directions were hydrostatically tested in both directions, while those that were only specified to be leak tight in the seating direction were tested as such.

The original hydrostatic test reports were reviewed and indicate that the design of these doors is capable of withstanding 15 psi (1.25 x 12 psi) in both directions. Although the original purchase specification does not require that Door Ps 25 & 26 be leak-tight in the unseating direction, they are, by design, none-the-less watertight in both directions for pressures up to 15 psi (34.5 feet of water).

As a result, it is both conservative and reasonable to use a maximum flood height of 23 feet in both directions.

Page 14 of 25

\ I ATTACHMENTto PLA-4134 Table 1 - Watertight Doors Rated Pressure (psi)

Seating Unseating Unit Door 8 Areas Se grated Press Press 13 CS "A" Rm/ CS "B" Rm 10 10 25 RB Sump Rm/ RHR "A" Rm 12 0 2 12 CS "A" Rm/ Stair 202 10 0 2 14 CS "A" Rm/ CS "B" Rm 10 10 2 26 RB Sump Rm/ RHR "A" Rm 12 0 Page 15 of 25

ATTACHMENTto PLA-4134 UKSTION 6 The staff requests that PPAL evaluate the qualification of standby gas treatment system (SGTS) components within the control structure for operation in the environment created by ventilating the reactor building through the system for cases where one SFP is boiling and where two SFPs are boiling. The assumptions used in the evaluation should be consistent with previous evaluation of SGTS duct conditions during SFP boiling scenarios.

RKSPONSK TO UKSTION 6 The qualification of standby gas treatment system (STGS) components have been analyzed for the case with one SFP boiling (LOCA/LOOP) and for the case with two SFPs boiling (seismic event).

LOCA/LOOP PPkL has analyzed the impact on equipment qualification of higher room temperature due to loss of SFPC during a LOCA/LOOP. This case is based on Compartment Transient Temperature Analysis Program (COTTAP) temperature inputs. The conclusion was that temperatures in the Control Structure SGTS rooms do not exceed 104'F. All equipment in these areas is qualified for 104'F or higher in the Environmental Qualification Program.

Seismic Event PP&L has performed evaluations similar to the LOCA event for a seismic event. The conclusion was that, using a very conservative analysis, temperatures in the Control Structure SGTS room will not exceed 104'F at the point in time when SGTS capability will be lost due to condensation in the recirculation plenum as discussed in PLA-4133, dated 5/4/94. Allequipment in these areas is qualified for 104'1'r higher in the Environmental Qualification Program.

Page 16 of 25

ATTACHMENTto PLA-4134 UKSTION 7 In the March 7, 1994 RAI, the staff noted that the licensing basis of the plant credits the use of the SGTS during the boiling pool event following a seismic event. In the RAI, the staff asked a question to assess the potential to use the residual heat removal system to cool the spent fuel pools following a seismic event. The diesel generator loading associated with a seismic event and coincident loss of offsite power (LOOP) was presented in PP&L's letter dated March 25, 1994. The staff requests that PP&L perform an assessment of diesel generator loading considering the limiting single failure for a seismic event with a coincident LOOP. Justification for the assumed single failure should be provided.

The staff also requests that PP&L assess the ability of the diesel generators to accommodate the additional loading associated with operating one loop of the RHR system in the SFP cooling assist mode and one loop of alternate decay heat removal, as described in procedure ON-1(2)49-001, "Loss of RHR Shutdown Cooling," on the non-accident unit. The intent is to determine maximum diesel generator loading during a loss of coolant accident (LOCA) in the opposite unit coincident with a LOOP, assuming no single failure. If one of the diesel generator loading patterns presented in the FSAR is bounding, describe the basis for this conclusion.

RESPONSE TO UKSTION 7 This evaluation discusses the plant response to a seismic event concurrent with a LOOP, and the subsequent loss of Spent Fuel Pool Cooling (SFPC). Details regarding how the plant will be placed in safe shutdown for a seismic event are also provided. Such an evaluation is needed to determine the single failures that would impact use of RHR in the Fuel Pool Cooling mode concurrent with maintaining safe shutdown of both reactors. During such an event, the SFPC is assumed to fail due to the earthquake loading resulting in pipe stresses over allowables for both the SFPC and service water systems. It will also be unavailable due to the LOOP. Consequently it will be necessary to restore cooling prior to boiling of the SFPs in order to avoid the impact on SGTS discussed in PLA-4133, dated 5/4/94. Therefore, this evaluation only discusses the plant response and operator actions required to restore cooling to the SFPs via the RHR Fuel Pool Cooling Mode, and place both units in cold shutdown without exceeding any design limits.

The following assumptions are made for this analysis:

l. A Safe Shutdown Earthquake causes a LOOP which results in an MSIV closure event.

2. No Seismic Category I equipment fails as a result of the earthquake.

3. Seismic category I equipment is subject to random failures independent of the earthquake. Single failures are assessed per ANSI/ANS-58.9-1981.

Page 17 of 25

ATTACHMENTto PLA-4134

4. Equipment other than seismic category 1 equipment fails as a result of the earthquake.

5. Operator actions including manual in plant equipment manipulations are allowed provided that they are performed no sooner than 10 minutes after the earthquake and sufficient time is available to reliably execute the procedure.

6. Plant Configurations:

1. Both units in power operation without pool communication

2. Both units in power operation with pool communication,

3. One unit at power and the other unit in refueling without pool communication.

4. One unit at power and the other unit in refueling with pool communication.

7. Operating Restrictions:

1. Shutdown cooling and fuel pool cooling cannot be operated simultaneously on the same unit.

2. Division I of RHR cannot operate in suppression pool cooling with division II of RHR in fuel pool cooling.

3. Division I of RHR can operate in fuel pool cooling with division II RHR in suppression pool cooling.

The initiating event is an earthquake at time zero. The earthquake is assumed to cause a Loss of Offsite Power (LOOP). The immediate response of the plant to a LOOP includes:

~ reactor scram following loss of power,

~ MSIV isolation,

~

safety relief valve actuation following the MSIV isolation,

~ condensate/feedwater pump trip on loss of power and low suction pressure,

~ low low RPV water level following void collapse

~ reactor building isolation and SGTS initiation.

The above events lead to the following:

~ auto initiation and loading of the diesel generators,

~ auto initiation of RCIC and HPCI,

~ operator entry into the following procedures:

Page 1S of 25

ATTACHMENTto PLA-4134 SCRAM, ON-100/200-101 RPV Control, EO-100/200-102 Primary Containment Control, EO-100/200-103 The operator enters the above procedures based upon any of the following conditions: a scram condition, the RPV water level less than +13", the RPV pressure greater than 1037 PSIG, the drywell pressure greater than 1.72 PSIG and the suppression pool temperature greater than 90'F.

Entry into the Secondary Containment Control Procedure EO-100/200-104 is not expected until after 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the average reactor building temperature reaches 110'F. Ifoffsite power is not lost or recovered prior to the reactor building temperature exceeding the maximum normal temperature, the secondary containment procedure willnot be entered. The operator will execute these procedures concurrently. The more significant actions taken are described below:

SCRAM ON-100/200-101 Rev 1

~ ensuring a scram and the appropriate containment isolation,

~ ensuring diesel generator initiation and loading,

~ ensuring ESW initiation and proper operation,

~

placing the mode switch into shutdown.

The above actions are performed immediately.

RPV Control EO-100/200-102 Rev 5 Level Control

~ Restoring and maintaining the RPV water level between+13" and+54" using RCIC, HPCI, Core Spray and RHR as needed,

~

Resetting the generator lockout.

Pressure Control Preventing automatic actuation of SRVs by manual operating the SRVs and operating HPCI in the pressure control mode,

~ Maintaining the RPV pressure below the HCTL curve,

~

Depressurizing the RPV at less than 100'F/hr.

Once the level is controlled between the band+13 to+54 and the shutdown cooling interlocks have cleared, the procedure allows the operator to establish shutdown cooling.

Page 19 of 25

ATTACHMENTto PLA-4134 Prima Containment Control EO-100/200-103 Rev 5

~ Restoring and maintaining the suppression pool below 90 'F using suppression pool cooling,

~ Maintaining the RPV pressure and the suppression pool water level below the SRV tailpipe level limit.

The operator will take immediate action to restore the reactor and containment parameters to normal conditions. A significant amount of time exists to successfully perform these actions.

Once the operator has ensured a successful scram has taken place, the first actions are to ensure the RPV water level and pressure are being controlled. The RCIC and HPCI systems will automatically initiate and quickly flood RPV. Prior to the RPV water level reaching

+54", the operator will place the HPCI system in the pressure control mode to avoid RCIC and HPCI trip on high water level, and to augment the SRVs for RPV pressure control. After about 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months the HPCI system is sufficient to control RPV pressure without the use of the SRVs. The HPCI system can be used for up to 3.5 days to control RPV pressure if necessary. Ifthe water level is in the normal range the operator will commence a slow RPV depressurization at less than 100 'F/hr. Once the shutdown cooling interlocks have cleared the operator is permitted to place the unit into shutdown cooling or alternate shutdown cooling'. These interlocks should not be cleared prior to 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .

In the unlikely event that the RCIC and HPCI systems are both unavailable, the operator has 40 minutes to restore them to operation prior to having to initiate an emergency depressurization to allow either the core spray or RHR systems to be used for core cooling.

Once RPV water level is maintained above +13", the operator may enter shutdown cooling.

The operators will also be controlling the primary containment parameters, especially the suppression pool temperature and water level. The operator will initiate a loop of suppression pool cooling within the first 30 minutes. However the operator has over 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to initiate suppression pool cooling before exceeding the pool HCTL, which is the first critical parameter to be encountered.

Once RPV and Primary Containment control is established, the operator will then establish fuel pool cooling. For this scenario, iffuel pool cooling is established within 35 hour4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months s~, fuel pool boiling will be avoided. The operator will first attempt to establish normal fuel pool cooling. If this system is unavailable due to the unavailability of offsite power or seismic induced damage, the operator will then align the RHR system in the fuel pool cooling assist mode. It is estimated that this alignment will require about 8 to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . A single loop of RHR in the fuel pool cooling assist mode is more than capable of removing heat from both fuel pools, provided that the pools are connected.

Per GO-100/200-005 or ON-149/249-001 Attachment to PLA-4133, dated 5/4/94 Page 20 of 25

ATTACHMENTto PLA-4134 PLANT CONFIGURATION EVALUATION E<valuation of Configuration 1:

Both units at power; operation without fuel pool communication.

Review of the fuel pool risk assessment calculation'hows that there are 7 single failures in the RHR fuel pool cooling assist mode of operation. Additionally, loss of a single diesel will also prevent simultaneous operation of SDC and fuel pool cooling on both units. Therefore this case plus a single failure will lead to one of the two fuel pools boiling or a unit being unable to be placed into shutdown cooling.

Case Plant configuration 3 has a lower heat load, however the systems analysis and results are equivalent to configuration 1. Therefore, configuration 3 is not evaluated independently.

Probability of Fuel Pool Boiling, SA-TSY-001,11/23/93 Page 21 of 25

ATTACHMENTto PLA-4134 E<valuation of Configuration 2:

Both units at power; operation with fuel pool communication.

A review of the plant systems required to simultaneously operate Shutdown Cooling (SDC) or Alternate SDC and cool the connected fuel pools was performed. This evaluation revealed that the diesel generators and ESW/Spray Pond Network were controlling with regard to single failure.

It was concluded that a failure of a diesel generator can be tolerated, while the failure of a complete loop of ESW cannot. Additionally, the loading on the diesel generators with a single failure has been evaluated in a separate calculation and found to be within the continuous rating of the diesels. A summary of this calculation is provided as an attachment to this evaluation.

Both reactors and the fuel pools cannot be simultaneously cooled following an earthquake, ifthe spray pond bypass discharge valves, HV-01222A or HV-01222B fail to close when required by operator action. Closure of these valves is required to allow proper function of the spray pond spray network. It was determined based upon a review of ANSI/ANS-58.9-19S1'hat failure of these valves do not represent single failures.

These valves do not have to be closed until an RHR heat exchanger in the particular loop is placed into service. However, as stated above, operation of both these valves is not required until about 35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months after the earthquake. Additionally, it is reasonable to expect that these valves can be manually closed or repaired prior to fuel pool boiling. They are located in the ESW valve vault which is located outside the reactor building in the site yard next to the spray pond.

Furthermore the mean time to repair a valve is 5.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months '. Based upon these conditions these valves are excluded from single failure analysis per reference 5.

Therefore, both reactors and both crosstied fuel pools can be simultaneously cooled following a safe shutdown earthquake and a credible independent single failure (failure of a diesel generator).

American National Standards Single Failure Criteria for Light Water Reactor Safety-Related Fluid Systems, ANSI/ANS-58.9-1981. Feb. 17, 1981.

NUREG/CR-3154 Page 22 of 25

l ATTACHMENTto PLA-4134 ATTACHMENTTO RESPONSE 7 DIESEL GENERATOR LOADINGEVALUATION WITH SINGLE FAILURE Page 23 of 25

ATTACHMENTto PLA-4134 SPENT FUEL POOL COOLING DIESEL GENERATOR LOADING The purpose of this evaluation is to determine the Loading on the Diesel Generators for a Seismic Event with Loss of Spent Fuel Pool Cooling, an Extended Loss Of Offsite Power and single failure of one Diesel Generator at a time.

The following conditions were assumed for the attached Diesel Generator loading tables:

~ Unit 1 and Unit 2 at 100% Power

~ Seismic Event

~ Loss of Unit 1 and Unit 2 Spent Fuel Pool Cooling

~ Extended Loss Of Offsite Power

~ Reactor cooling provided via ALTERNATE DECAY HEAT REMOVAL MODE

~

Single Failure of one Diesel Generator at a time The Diesel Generator Loading was developed using FSAR Table 8.3-1 for the equipment KW rating and the assignment of ESF and selected non-ESF loads to the Diesel Generators. The 4160 VAC cable losses were considered in the Diesel Generator Loading.

Separate Diesel Generator loading tables were developed for Control Structure HVAC Train "A" running with Train "B" in Standby; and Control Structure HVAC Train "B" running with Train "A" in Standby. This was done since only one train of Control Structure HVAC is running.

Note FSAR Table 8.3-1a shows both trains of Control Structure HVAC running. This Table represents all loads connected to the Diesel Generator without regard to which loads are actually running.

Since the Diesel Generator Loading with the Control Structure HVAC Train "A" represents the most severe loading on the Diesel Generators, this loading was used as the base case for failure of the Diesel Generator A and B.

The loading tables were developed using Alternate Shutdown Cooling consisting of one RHR loop with one RHR pump in Suppression Pool Cooling and one Core Spray pump in Alternate Shutdown Cooling (See Procedure ON-149-001). This method of achieving Alternate Shutdown Cooling was used instead of one RHR pump, because it represents the worst loading on the Diesel Generators.

Page 24 of 25

ATTACHMENTto PLA-4134 oi For a single failure of a Diesel Generator the attached loading tables show that the Diesel Generators are capable of supplying the required loads to safely shutdown and maintain shutdown of Unit 1 and Unit 2 while the Fuel Pools are cooled by the Fuel Pool Mode of RHR. Also the continuous KW rating of the Diesel Generators are not exceeded for a single failure of a Diesel Generator.

The bounding condition for Diesel Generator Loading is as follows:

~

Loading 0 - 10 minutes LOCA (FSAR Table 8.3-2,3,4,5)

~

Loading 10 - 60 minutes Seismic Event (Attached tables 1,2,3,4)

~ Loading beyond 60 minutes Seismic Event (Attached tables 1,2,3,4)

A summary of the total Diesel Generator Loading for both the LOCA and Seismic Events is found in attached table 5 .

For failure of the Diesel Generator A or B, the following ESW valves must be manually opened to support operation of Fuel Pool Cooling mode of RHR:

DG A Failure - HV-01222A, HV-01224A1, HV-01224A2 DG B Failure - HV-01222B, HV-01224B1, HV-01224B2 These valves are located outside the reactor building and are Seismic Category L Page 25 of 25

TAI CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR A UNAVAILABLE CALCULATION I 002 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR 0 UNAVAILABLE DEMAND KW DEMAND KW DEMAND KW DEMAND KW 0.10 10.60 60 MIN 0.10 10-60 60 MIN 0-10 10.60 60 MIN 0-10 10-60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Loads Automatlcaly Started for LOOP Unit 1 Engineeml Safeguard Swit ar Er LC Room Unit Coolers 13 13 13 Unit 1 Battery Chargers, 125 VDC 19 19 19 23 23 23 20 20 20 Unit 1120V Instrument AC Dist Paneh Unit 1 Standby Lktuld Cont Tank Heater 10 10 10 Unit 1 Batt Char, 250 VDC 80 80 80 80 80 80 Unit 1 Pnmary Containment Isolation S tern M-G Set 13 13 13 Unit 1 Engineered Safeguard Load Center Transformer Losses 15 15 15 15 15 15 15 15 15 Unit 2 Engineered Safeguard Swit Er LC Room Unit Coolers 13 13 13 Unit 2 Batt Char, 125 VDC 19 19 19 23 23 23 20 20 20 Unit 2 Instrument AC Dist Panels 12 12 12 Unit 2 Standby Urtuid Control Tank Heater 10 10 10 Unit 2 Batt Char . 250 VDC 80 80 80 80 80 80 Unit 2 Compressor Motor for Emergency SWGR Er LC Room 48 48 48 Unit 2 Pnmary Containnent Isolation S tem M-G Set 13 13 13 PAGE 1

TAI CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR A UNAVAltABLE CALCULATION P 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D UNAVAILABLE DEMAND KW DEMAND KW DEMAND KW DEMAND KW 0-10 10 60 60 MIN 0-10 1O60 60 MIN 0-10 10.60 60 MIN 0.10 10-60 60 MIN LOADS MIN MIN BEYOND MIN . MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Unit 2 Engineering Safeguards Load Center Transformer Losses 15 15 15 15 15 15 15 15 15 Control Structure Battery Room Exhaust Fans 4.5 4.5 4.5 Diesel Generator Room Ventilation Fans 33 33 33 33 33 33 33 33 33 Diesel Generator Diesel Oil Transfer Pumps 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 Reactor Recirc Fans 61 61 61 Emer Sewice Water Pum 357 357 357 357 357 357 357 357 357 Standby Gas Treatment System Exhaust Fan 42 42 42 42 42 42 Control and Computer Room Air Condition Unit Pumps 66 66 66 Diesel Generator Starting Air Compressors 18 18 18 Control Structure Chiiled Water Circubrting Pumps 25 25 25 Control Structure Emergency Outside Air Supply Fans 17 17 17 Control Structure Water Chtaer Compressor 279 279 279 Control Structure Air Condition Unit Heat'oih 130 130 130 Standby Gss Treatment System E Room Exhaust Fans 4.5 4.5 4.5 4.5 4.5 4.5 Standby Gss Treatment System Equipment Room Heating Unit Hester Fans 4.5 4.5 4.5 4.5 4.5 4.5 Control Structure Air Conditioning Unit Fans 42 42 42 PAGE 2

TA CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR A UNAVAIlABLE CALCULATION P & 002 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D UNAVAILABLE DEMAND KW DEMAND KW DEMAND KW DEMAND KW 0-10 10 60 60 MIN 0-10 10.60 60 MIN 0-10 1o60 60 MIN 0.10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND ESW Pump House Supp Fan Control Suucture Compressor Oil Pump 1.4 1.4 1.4 Control Structure Chirr Emergency Condenser Circ Pumps 17 17 17 Control Structure Emergency Outside Air Supply Unit Heating Cogs 30 30 30 Standby Gas Treatment System Heater 90 90 90 90 90 90 Standby Gss Treatment Equipment Room Heater 30 30 30 30 30 30 Diesel Generator HVAC Panels 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.6 . 1.6 ESW Pump House HVAC Control Panels 3.8 3.6 3.8 Diesel Generator tube Oil Heaters Turbine Generator Turning Gear Oil Pumps 32 32 32 Essential Ughting 91 91 91 44 44 44 74 74 74 Turbine BuBdeg Cooang Water Pumps 13 13 13 Reactor Buikfing Closed Coogng Water Pumps 25 25 25 25 25 25 Control Suucture P Bevator 2.5 2.5 2.5 Reactor Bugding Service Bevator Standby liquid Ei Oxygen snd Hydrogen Analyzer Heat Trace Panels 19 19 19 PAGE 3

TAB CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR A UNAVAILABLE CALCULATION /I 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D UNAVAILASLE DEMAND KW DEMAND KW DEMAND KW DEMAND KW 0.10 1O60 60 MIN 010 10.60 60 MIN 0.10 1060 60 MIN 0.10 10-60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Makl Turbine LO Reservoir Vapor Extractor Main Turbine LO Reservoir Oil Mist EEmfnator 2.4 2.4 2.4 Reactor BuikEng Bectrical Equipment Room HILV Fans 10 10 10 Engineered Safeguards Transformer AuxrTiaries 3.7 3.7 3.7 30 KVA Transformer/Post Accident Vent Stack Monitoring and Sampgng Pumps UPS/SPDS Distribution Panels UPS/120V instrument AC Distribution Parvri 70 70 70 62 62 62 1030 999.3 999.3 1618.2 1587.5 1587.5 841.8 811.1 8'I 1.1 Unit 1 Safe Sbutdown Loads S'tarted RHR Purn 1B 1429 1429 RHR Pump Room Unit Coolers RHR Service Water Pum 18 463 463 Reactor Core Spr Pumps 8 552 552 Core Spray Pump Room Unit Coolers RHR Service Water Pump House Sup Fan 18 4.6 HPCI Pump Room Urvt Coolers 1.5 1.5 1.5 1.5 Unit 1 D II Coolers 64.6 64.6 64.6 1030 2997.4 2997.4 1682.8 1652.1 1652.1 841.8 1275.6 1275.6 PAGE 4

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TA CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR A UNAVAILABLE CALCUlATION F 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D UNAVAILABLE DEMAND KW DEMAND KW DEMAND KW DEMAND KW 0-10 10-60 60 MIN 0.10 10.60 60 MIN 0-10 10 60 60 MIN 0.10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Urdt 2 Safe Shutdown Loads Manually Started RHR SW Pump 28 463 Unit 2 D Coolers 64.6 64.6 64.6 RHR Pum 2D 1429 1429 Core S Pumps D 552 552 Core S Pump Room Unit Cooler RHR Service Water Pump House Fan 28 4.6 4.6 HPCI Unit Room Coohrs 1.5 1.5 1.5 1.5 RHR Pum Room Unit Cooler 2D 1030 3003.5 3466.5 1747.4 1716.7 1716.7 841.8 3269.1 3269.1 Fuel Pool Coolinp -RHR Assht RHR Pum 1C(45% of Ca 671 RHR Pump Room Unit Cooler RHR Service Water Pum 1A 463 RHR Service Water Pump House Fan 18 4.6 1030 3003.5 3471.1 1747.4 1716.7 2859.7 841.8 3269.1 3269.1 Non-ESF toads Started Unit 1 CRD Pump SGTS Stack Vent Vac Pum 0.9 0.9 PAGE 5

TAB CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR A UNAVAILABLE CALCULATION 6 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D UNAVAILABLE DEMAND KW DEMAND KW DEMAND KW DEMAND KW 0.10 10-60 60 MIN 0-10 10.60 60 MIN 0-10 10 60 60 MIN 0-10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Unit 2 CRD Pump RFPT Turrs Gear 4.8 4.8 Turbine Generator Bearing Aux Uft Pumps and Turning Gear 119 119 Stock Vent Vacuum Pump (Reactor Building) 1.2 1.2 Reactor Protection S tern M-G Sets 54 54 1030 3182.5 3650.1 1 747.4 1716.7 2859.7 841.8 3270 3270 4160 VAC Cable Losses 11.3 11.3 11.3 9.5 9.5 9.5 13.6 13.6 13.6 TOTAL DIESEL GENERATOR LOADING 1041.3 3193.8 3661.4 1756.9 1726.2 2869.2 855.4 3283.6 3283.6 Diesel Generator Rat'000 KW Continuous 4400 KW 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months PAGE 6

~ 4 'I TAB CONTROL STRUCTURE HVAC TRAIN 'A' E DIESEL GENERATOR B VNAVAItABLE CALCULATION I 2 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW UNAVAILABLE DEMAND KW DEMAND KW 0-10 10.60 60 MIN 0-10 1O60 60 MIN 0-10 'IO60 60 MIN 0-10 10-60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Loads Automatlcalty Staned for LOOP Unit 1 Engineered Safeguard Switchgear EB LC Room Unit Cookus 13 13 13 Unit 1 Batt , 125 VDC 20 20 20 23 23 23 20 20 20 Unit 1 120V Instrument AC Dist Panels Unit 1 Standby Uquid Cont Tank Heater 10 10 10 Unit 1 Batt Chargers, 250 VDC 44 44 44 44 44 44 Unit 1 Primary Containment Isolation S tern M-G Set 13 13 13 Unit 1 Engineered Safeguard Load Center Transformer Losses 15 15 15 15 15 15 15 15 15 Unit 2 Engineered Safeguard Switchgear EB LC Room Unit Coolers 13 13 13 Unit 2 Ban Char rs, 125 VDC 20 20 20 23 23 23 20 20 20 Unit 2 Insuument AC Dht Panels Unit 2 Standby Uquid Control Tank Heater 10 10 10 Ulst 2 Bane Char rs. 250 VDC 44 44 44 44 44 44 Unit 2 Compressor Motor for Emergency SWGR fk LC Room Cooing 48 48 48 Unit 2 Pnmary Containment Isolation S tern ther Set 13 13 13 PAGE 1

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TAB CONTROL STRUCTURE HVAC TRAIN 'A' E DIESEL GENERATOR 8 UNAVAILABLE CALCULATIONd 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW UNAVAILABLE DEMAND KW DEMAND KW 0-10 10.60 60 MIN 0-10 1 O60 60 MIN 0-10 10.60 60 MIN 0.10 10-60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Unit 2 Engineering Safeguards Load Center Transformer Losses 15 15 15 15 15 15 15 15 15 Control Stmcture Battery Room Exhaust Fans 4.5 4.5 4.5 Diesel Generator Room Ventaation Sup Fans 33 33 33 33 33 33 33 33 33 Diesel Generator Diesel 01 Transfer Purn 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 Reactor Budding Recirc Fans 61 61 61 Emerge Service Water Pumps 357 357 357 357 357 357 357 357 357 Standby Gas Treatment System Circulat'5 Exhaust Fan 42 42 42 42 42 42 Control and Computer Room Air Condition Unit Pumps 66 66 66 Diesel Generator Starting Air Com ressors 18 18 18 Control Structure Chiaed Water 25 25 Control Structure Emergency Outside Air Fans 17 17 17 Control Structure Water Ch5er Compressor 279 279 279 Control Structure Air Condition Unit Heating Cogs 130 130 130 130 130 130 Standby Gas Treatment ~tom Equipment Room Exhaust Fans 4.5 4.5 4.5 4.5 4.5 4.5 Standby Gas Treatment System Equipment Room Heating Unit Heater Fans 4.5 4.5 4.5 4.5 4.5 4.5 Control Structure Air Conditioning Unit Fans 42 42 42 PAGE 2

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TA CONTROL STRUCTURE HVAC TRAIN 'A' E DIESEL GENERATOR 8 UNAVAILABLE CALCIATIONP 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW UMAVAllABLE DEMAND KW DEMAND KW 0-10 10-60 60 MIN 0.10 to60 60 MIN 0-10 10.60 60 MIM 0.10 1060 60 MIN LOADS MIN MIN BEYOND MIN MIM BEYOND MIN MIM BEYOND MIN MIN BEYOND ESW Pump House Su Fan Control Structure Conpressor Oil Pum 1.4 1.4 1.4 Control Structure Chger Emergency Condenser Circ Pumps 17 17 17 Control Structure Emergency Outside Air Supply Unit Heabng Coils 30 30 30 Standby Gas Treatment System Heater 90 90 90 90 90 90 Standby Gas Treatment Equipment Room Heater 30 30 30 30 30 30 Diesel Generator HVAC Panels 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.6 ESW Pump House HVAC Control Panels 3.8 3.8 3.8 Pumps Essenthl ht'9 Diesel Generator Lube OI Heaters Turbine Generator Turning Gear Oil Turbine Buiik6ng Cooling Water Ptxn Reactor Busing Closed Cooang 32 13 32 79 13 32 79 13 46 46 75 75 75 Water Pumps 25 25 25 25 25 25 Control Structure Passenger Bevator 2.5 2.5 2.5 Reactor Br Service Bsvator Standby Uquid ET Oxygen and Hydrogen Analyzer Heat Trace Paneb 24 24 24 19 19 19 PAGE 3

7 TAB CONTROL STRUCTURE HVAC TRAIN 'A' E DIESEL GENERATOR 8 UNAVAILABLE CALCULATIONg 002 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW UNAVAILABLE DEMAND KW DEMAND KW 0.10 1O60 60 MIN 0-10 1O60 60 MIN 0-10 1O60 60 MIN 0-10 1O60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Main Turbine LO Reservoir Vapor Extractor Main Turbine LO Reservoir Oil Mist Egmfnator 2.4 2.4 2.4 Reactor Bug<frng Ehctrical Equipment Room H5V Sup Fans Engineered Safeguards Transformer Auxiliaries 2.8 0.9 3.7 30 KVA Transformer/Post Acck$ ent Ven! Stack Monitoring and Sampling Purn 30 35 35 UPS/SPDS Distribution Psneh UPS/120V Instrument AC Distribution Panel 70 70 70 62 62 62 993.1 968.3 968.3 1520.4 1492.5 1492.5 997.8 g67.1 967.1 Unit 1 Safe Shutdown Loads Started RHR Pump 1D 1429 1429 RHR Pum Room Unit Coolers RHR Service Water Pum 18 463 463 Reactor Core S ra Pum A 552 552 Core S pre Pump Room Unit Coohrs RHR Servke Water Pump House Fan 1A 4.6 4.6 HPCI Pump Room Unit Coohrs 1.5 1.5 Unit 1 64.6 64.6 64.6 15 993.1 1526.9 1526.9 1585 1557.1 1557.1 997.8 2869.6 2884.6 Unit 2 Safe Sbutdown Loads Started PAGE 4

CONTROL STRUCTURE HVAC TRAIN 'A' E DIESEL GENERATOR B UNAVAILABLE CALCUlATION8 002 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW UNAVAILABLE DEMAND KW DEMAND KW 0-10 10-60 60 MIN 0-10 10-60 60 MIN 0-10 10-60 60 MIN 010 1O60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND RHR Pump 2C 1429 1429 RHR Service Water Pump 2A 463 463 Core Spra Pum 2A 552 552 Core S Pump Room Unit Cooler HPCI Unit Room Coolers 1.5 1.5 Unit 2 D 8 Coohrs 64.6 64.6 64.6 993.1 2541.9 2541.9 1649.6 3052.7 3052.7 997.8 2871.1 2886.1 Fuel Pool Coogng -RHR Asaht RHR Pump 1A(45% of Ca 671 RHR Pump Room Unit Coohr RHR Service Water Pump 1A 463 463 993.1 2541.9 3221.9 1649.6 3515.7 3515.7 997.8 2871.1 2886.1 Non.ESF Loads Started Unit 1 CRD Pump SGTS Stack Vent Vac Pump 0.9 0.9 0.9 Unit 2 CRD Pum RFPT T Gear 4.8 4.8 Turbine Generator Bearing Aux Uft snd T Gear 119 119 Stack Vent Vacuum Pump (Reactor

$ 8ding)

Reactor Protecthn S em MG Sets 54 54 PAGE 5

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TAB CONTROL STRUCTURE HVAC TRAIN 'A' DIESEL GENERATOR B UNAVAILABLE CALCULATlON f 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW UNAVAILABLE DEMAND KW DEMAND KW 0.10 10.60 60 MIN 0.10 1060 60 MIN 0-10 10-60 60 MIN 0-10 1O60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND 993.1 2719.7 3399.7 1649.6 3515.7 3515.7 998.7 2872 2887 4160 V AC Cable Losses 11.8 11.8 11.8 9.5 9.5 9.5 13.6 13.6 13.6 TOTAL DIESEL GENERATOR LOADING 1004.9 2731.5 3411.5 1659.1 3525.2 3525.2 1012.3 2885.6 2900.6 Diesel Generator Rating:

4000 KW Contiraous 4400 KW 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months PAGE 6

TAB CONTROL STRUCTURE HVAC TRAIN '8'N DIESEL GENERATOR C UNAVAILABLE CALCULATIONg 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW UNAVAIlABLE DEMAND KW 0-10 10.60 60 MIN 0-10 10.60 60 MIN 0.10 10.60 60 MIN 0-10 1O60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Loads LOOP Aut~ Started for Unit 1 Engineered Safeguard Switchgear 6B LC Room Unit Cooers 13 13 13 Unit 1 Bette Char, 125 VOC 20 20 20 19 19 19 20 20 20 Unit 1 120V Instrument AC Dist Panels Unit 1 Standby Uquid Cont Tank IAoter Unit 1 Battery Chargers. 250 VDC 80 80 80 80 80 80 Unit 1 Primary Containment Isolation S tern M-G Set 13 13 13 13 13 13 Unit 1 Engineered Safeguard Load Center Transformer Losses 15 15 15 15 15 15 15 15 15 Unit 2 Engineered Safeguard Swit ear Ik LC Room Unit Coolers 13 13 13 13 13 13 Unit 2 Batt Chargers, 125 VDC 20 20 20 19 19 19 20 20 20 Unit 2 Instrument AC Dist Panrris 12 12 12 Uist 2 Standby Uquid Control Tank Heater Unft 2 Batt Char, 250 VDC 80 80 80 80 80 80 Unit 2 Compressor Motor for Emergency SWGR Ik LC Room Cooing 48 48 48 Unit 2 Primaiy Containment isolation S tern M-G Set 13 13 13 13 13 13 PAGE 1

TAB CONTROL STRUCTURE HVAC TRAIN '8'N DIESEL GENERATOR C UNAVAILABLE CALCULATIONfr 002 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW UNAVAILABLE DEMAND KW 0.10 10-60 60 MIN 0-10 10.60 60 MIN 0.10 10.60 60 MIN 0.10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Unit 2 Engineerfng Safeguards Load Center Transformer Losses 15 15 15 15 15 15 15 15 15 Control Structure Battery Room Exhaust Fans 4.5 4.5 4.5 Diesel Generator Room VenQation Fans 33 33 33 33 33 33 33 33 33 Diesel Generator Diesel Oil Transfer Pumps 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 Reactor Braiding Recirc Fans 61 61 61 Emerg Service Water Pumps 357 357 357 357 357 357 357 357 357 Standby Gas Treatment System Exhaust Fan 42 42 42 Control and Computer Room Afr Condidon Unit Pumps 66 66 66 Diesel Generator Starting Air Com ressors 18 18 18 Control Structure Chgaxl Water Circubr ting Pum 25 25 25 Control Structure Emergency Outside Air Fans 17 17 17 Control Structure Water Chiber Com r 279 279 279 Control Structure Air Condition UNt Heating Cogs 130 130 130 Standby Gas Treatment System Equipment Room Exhaust Fans 4.5 4.5 4.5 Standby Gas Treatment System Equipment Room Heating Unit Heater Fans 4.5 4.5 4.5 Control Structure iur Conditioring Unit Fans- 42 42 42 PAGE 2

~ ~ ~ 7 TA CONTROL STRUCTURE HVAC TRAIN '8'N W DIESEL GENERATOR C UNAVAllABLE cALGUlATIDNrl 002 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW UNAVAILABLE DEMAND KW 0.10 10 60 60 MIN 010 10-60 60 MIN 0.10 10 60 60 MIN 0-10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND ESW Pum House Supp Fan 4.4 4.4 4.4 Control Structure Compressor Oil Pump 1.4 1.4 1.4 Control Structure Chger Emergency Condenser Circ Pum 17 17 17 Control Structure Emergency Outsxh Air Supply Unit Heating Crx1s 30 30 30 Standby Gas Treatment System Heater 90 90 90 Standby Gas Treatment Equipment Room Heater 30 30 30 Diesel Generator HVAC Paneh 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.6 1.6 ESW Pump House HVAC Control Paneh 3.8 3.8 3.8 3.8 3.8 3.8 Diesel Generator Lube Oil Heaters Turbine Generator Turning Gear Oil Pumps 32 32 32 32 32 32 Essenthl U ht 91 91 91 74 74 74 Turtxne Building Cooling Water Pumps 13 13 13 13 13 13 Reactor Buikgng Closed Coo5ng Water Pumps 25 25 25 25 25 25 Control Structure Passenger Ehvator Standby Uquid 8 Oxygen and Hydrogen Analyzer Heat Trace Psneh 24 24 24 19 19 19 PAGE 3

TAB CONTROL STRUCTURE HVAC TRAIN '8'N DIESEL GENERATOR C UNAVAILABLE CALCULATIONd 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW UNAVAILABLE DEMAND KW 010 10-60 60 MIN 0-10 10-60 60 MIN 010 10-60 60 MIN 0-10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Main Turbkxr LO Reservoir Vapor Extractor Main Turbine LO Reservoir 09 Mist Egminator 2.4 2.4 2.4 2.4 2.4 2.4 Reactor Buikgng Eectrical Equipment Room HILVSu Fans 10 10 10 Engineered Safeguards Transformer AuxBiaries 3.7 3.7 30 KVA Transformer/Post Accident Vent Stack Morxtoring and Sampgng Pum 30 35 35 UPS/SPDS Distribution Parxris UPS/120V Instrument AC Distribution Panel 40 40 40 62 62 62 1122.4 1096.7 1096.7 908 877.3 877.3 1450 1423 1423 Unit 1 Safe Shutdown Loads Started RHR Pump 18 1429 1429 RHR Pum Room Unit Coolers RHR Service Water Pum 18 463 463 Reactor Core Spray Pumps A 552 552 Core Spr Pump Room Unit Cooers RHR Service Water Pump House Sup Fan 18 4.6 4.6 HPCI Pump Room Unit Coolers 1.5 1.5 1.5 1.5 64.6 64.6 64.6 1122.4 1650.7 1650.7 908 2321.4 2321.4 1514.6 1952.1 1952.1 Unit 2 Safe Shutdown Loads Staned PAGE 4

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TAI CONTROL STRUCTURE HVAC TRAIN '8'N DIESEL GENERATOR C UNAVAILA6LE CALCUlATIONir 002 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW UNAVAILABLE DEMAND KW 010 10.60 60 MIN 0-10 1O60 60 MIN 0.10 10-60 60 MIN 0-10 1O60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND RHR Pum 2D 1429 1429 Unit 2 64.6 64.6 64.6 RHR Service Water Pump 28 463 463 RHR Servhe Water Pump House Sup Fsn 28 4.6 4.6 RHR Pump Room Unit Cooler 2C 1.5 1.5 HPCI Urit Room Coohrs 1.5 1.5 Core Spra Pum 28 552 552 Core S Pump Room Unit Coohrs RHR Pum Room Unit Cooler 2D 1122.4 1650.7 1650.7 908 3353.5 3353.5 1 579.2 3456.2 3456.2 Fuel Pool Cooang -RHR Assbt RHR Pump 1A f45% of Ca ) 671 RHR Pum Room Unit Coohr RHR Service Water Pump 2A 463 RHR Service Water Pump House Su Fan 1A 4.6 1122.4 1650.7 2799.3 908 3353.5 3353.5 1 579.2 3456.2 3456.2 Non-ESF Loads Started Urit 1 CRD Pump SGTS Stack Vent Vac Pum 0.9 0.9 0.9 Unit 2 CRD Pum PAGE 5

t ~ ~ I TAB CONTROL STRUCTURE HVAC TRAIN '8'N ~ DIESEL GENERATOR C UNAVAILABLE CALCUlATION4' DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW UNAVAILABLE DEMAND KW 0.10 10.60 60 MIN 0-10 10.60 60 MIN 0-10 10.60 60 MIN 0-10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND RFPT Tumi Gear 4.8 4.8 4.8 Turbine Generator Bearing Aux Uft Pumps and Turning Gear 119 119 119 119 Stack Vent Vacuum Pump (Reactor Buikf 1.2 1.2 Reactor Protection tern lkG Sets 54 54 54 54 1122.4 1828.5 2976.1 908 3532.5 3532.5 1580.1 3457.1 3457.1 4160 V AC Cabie tosses 11.8 11.8 11.8 11.3 11.3 11.3 13.6 13.6 13.6 TOTAL DIESEL GENERATOR LOADING 1134.2 1840.3 2987.9 919.3 3543.8 3543.8 1593.7 3470.7 3470.7 Diesel Generator Rating:

4000 KW Contsaous 4400 KW 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months PAGE 6

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TAB CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR D UNAVAILABLE CALCULATIONS 2 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW DEMAND KW UNAVAIlABlE 0-10 10.60 60 MIN 0.10 10-60 60 MIN 0-10 10-60 60 MIN 0-10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Loads Automaticidly Started for LOOP Unit 1 Engineered Safeguard Swit r Ei LC Room Unit Coolers 13 13 13 Unit 1 Battery Char, 125 VDC 20 20 20 19 19 19 23 23 23 UNt 1120V Instrument AC Dist Panels Unit 1 Standby Uquid Cont Tank Heater 10 10 10 UNt 1 Batt Char rs. 250 VDC 44 44 44 80 80 80 44 44 Unit 1 Primary Contaiisnent Isolation S temLHlSet 13 13 13 13 13 13 Unit 1 Engineered Safeguard Load Center Transformer Losses 15 15 15 15 15 15 15 15 15 Unit 2 Engkieered Safeguard Switchgear Ei LC Room Unit Coolers 13 13 13 Unit 2 Battery . 125 VDC 20 20 20 19 19 19 23 23 23 Unit 2 Instrument AC Dist Panels 12 12 12 Unit 2 Standby Uquid Control Tank Heater 10 10 10 Unit 2 Battery Char rs, 250 VDC 44 44 44 80 80 80 44 44 44 Unit 2 Compressor Motor for Emergency SWGR 8. LC Room Cooling 48 48 48 0 Unit 2 Pdmary Containment Isolation tern M-G Set 13 13 13 13 13 13 PAGE 1

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TAB CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR D UNAVAILABLE CALCULATIONC 2 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW DEMAND KW UNAVAILABLE 0- I 0 10-60 60 MIN 010 10-60 60 MIN 0.10 10-60 60 MIN 0.10 10-60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Unit 2 Engfneenng Safeguards Load Center Transformer Losses 15 15 15 15 15 15 15 15 15 Control Structure Battery Room Exhaust Fans 4.5 4.5 4.5 Diesel Generator Room Ventilation Fans 33 33 33 33 33 33 33 33 33 Diesel Generator Diesel Oil Transfer Pumps 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 2.5 Reactor Buikf Recirc Fare 61 61 61 Emer Senrice Water Pum 357 357 357 357 357 357 357 357 357 Standby Gas Treatment System Exhaust Fan 42 42 42 Control and Computer Room Air Condition Unit Pumps 66 66 66 Diesel Generator Starting Air Comp ressors 18 18 18 Control Structure Chilled Water Circdating Pumps 25 25 25 Control Svucture Emergency Outside Afr Supply Fans 17 17 17 Control Structure Water Chider Compressor 279 279 279 Control Structure Air Condition Unit Heat'ogs 130 130 130 Standby Gss Treatment System E Room Exhaust Fans 4.5 4.5 4.5 Standby Gas Treatment System Equipment Room Heating Unft Heater Fans 4.5 4.5 Control Structure Air Conditioning Unit Fans 42 42 42 PAGE 2

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TAB CONTROL STRUCTURE HVAC TRAIN 'A' DIESEL GENERATOR D UNAVAILABLE CALCULATION4 2 DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW DEMAND KW UNAVAILABLE 0-10 1 F60 60 MIN 0-10 10-60 60 MIN 0-10 10-60 60 MIN 0-10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND ESW Pump House Sup Fan 4.4 4.4 4.4 Control Structure Compressor Oil Pump 1.4 1.4 1.4 Control Structure Ctsger Emergency Condenser Circ Pumps 17 17 17 Control Structure Emergency Outside Alr Supply Unit Heating Cogs 30 30 30 Standby Gas Treatment System Heater 90 90 90 Standby Gas Treatment Eguipment Room Heater 30 30 30 Diesel Generator HVAC Panels 1.6 1.6 1.6 1.6 1.6 1.6 1.6 '.6 1.6 ESW Pump House HVAC Control Panels 3.8 3.8 3.8 3.8 3.8 3.8 Pumps Essential ht'2 Diesel Generator Lube Oil Heaters Turbine Generator Turning Gear Oil Turbine Bukgng Cooling Water Pum Reactor BtskEng Closed Cooling 13 32 13 32 13 32 91 13 32 91 13 32 91 13 44 44 44 Water Pum 25 25 25 25 25 25 Control Structure Passenger Elevator 2.5 2.5 2.5 Standby Uquid 5 Oxygen snd Hydrogen Analyzer Heat Trace Panebr 24 24 PAGE 3

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TAB CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR D UNAVAILABLE CALCULATIONd 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW DEMAND KW UNAVAILABLE 0-10 10.60 60 MIN 0.10 10.60 60 MIN 0-10 10.60 60 MIN 0-10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Main Turbine LO Reservoir Vapor Extractor Main Turbine LO Reservoir Oil Mist Efiminator 2.4 2.4 2.4 2.4 2.4 2.4 Reactor Building Electrical Equipment Room HE V Sup Fans 10 10 10 Engineered Safeguards Transformer Auxiliaries 2.8 3.7 0.9 30 KVATransformer/Post Accident Vent Stack Monitonng snd SampEng Pumps 30 35 35 UPS/SPDS Distribution Panels UPS/120V Instrument AC Distribution Panel 32 32 32 70 70 70 1014.1 989.3 989.3 922.4 891.7 891.7 1518.4 1490.5 1490.5 Unit 1 Safe Shutdown Loads Staned RHR Pum 18 1429 1429 RHR Pump Room Unit Coolers RHR Servke Water Pump 28 463 Reactor Core S ra Pum 18 552 552 Core S a Pum Room Unit Coolers RHR Servee Water Pump House Su Fan 28 4.6 4.6 HPCI Pum Room Unit CooIers 1.5 1.5 Unit 1 D II Coolers 64.6 64.6 64.6 1014.1 989.3 989.3 922.4 2887.8 3350.8 1583 1557.1 1557.1 Unit 2 Safe Shutdown Loads Started PAGE 4

I TAB CONTROL STRUCTURE HVAC TRAIN 'A' DIESEL GENERATOR D UNAVAILABLE CALCUlATION I 2 DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW DEMAND KW UNAVAllABLE 010 10.60 60 MIN 0-10 1O60 60 MIN 0-10 10-60 60 MIN 0-10 10-60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN BEYOND RHRPum 2C 1429 1429 Unit 2 D II Coolers 64.6 64.6 64.6 RHR Senrice Water Pump 2A 463 463 RHR Service Water Pump House Sup Fan 2A 4.6 4.6 RHR Pump Room Unit Cooler 2C Core S Pumps 2A 552 552 1.5 1.5 HPCI Unit Room Coolers 1014.1 2010.9 2010.9 922.4 2889.3 3352.3 1647.6 3059.7 3059.7 Fuel Pool CooEng -RHR Assist RHR Pump 1 A (45% of Caps 671 RHR Pump Room Unit Cooler RHR Service Water Pump 1A 463 RHR Service Water Pump House Fsn 1A 4.6 1014.1 2010.9 2695.5 922.4 2889.3 3352.3 1647.6 3059.7 3522.7 Non-ESF Loads Started Unit 1 CRD Pump SGTS Stack Vent Vac Pum Unit 2 CRD Pump RFPT Tu Gear 4.8 4.8 4.8 4.8 PAGE 5

TAB CONTROL STRUCTURE HVAC TRAIN 'A'N DIESEL GENERATOR D UNAVAILABLE CALCULATIONd ~ 2

,P DIESEL GENERATOR A DIESEL GENERATOR B DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW DEMAND KW UNAVAILABLE 0-10 10-60 60 MIN 0-10 10 60 60 MIN 0-10 10.60 60 MIN 0-10 10.60 60 MIN LOADS MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND Turbine Generator Beadn9 Aux Uft Pumps and T 'ear 119 119 119 119 Stack Vent Vacuum Pump (Reactor BuikEng) 1.2 1.2 Reactor Protection S tern M-G Sets 54 54 54 54 1014.1 2188.7 2873.3 922.4 3068.3 3531.3 1647.6 3059.7 3522.7 4160 V AC Cable Losses 11.8 11.8 11.8 11.3 1 'l.3 11.3 9.5 9.5 9.5 TOTAL DIESEL GENERATOR LOADING 1025.9 2200.5 2885.1 933.7 3079.6 3542.6 1657.1 3069.2 3532.2 Diesel Generator Rat 4000 KW Cotinuous 4400 KW 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months

Reference:

ON-149-001 Loss of RHR Shutdown Mode PAGE 6

tas

SUMMARY

OF DIESEL LOADING DG UNAVAILABLEAT A TIME. CALCUlATIONar 2 ~+

P DIESEL GENERATOR A DIESEL GENERATOR 8 DIESEL GENERATOR C DIESEL GENERATOR D DEMAND KW DEMAND KW DEMAND KW DEMAND KW 0-10 10.60 60 MIN 0-10 10.60 60 MIN 0.10 10.60 60 MIN 0-10 1O60 60 MIN TOTAL DG LOADING MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND MIN MIN BEYOND FSAR TABLE 8.3-2 DIESEL GENERATOR A UNAVAILABLE 3033.25 3302.05 3377.05 3256.88 2876.68 2968.38 3279.33 2286.53 2605.53 FSAR TABLE 8.34 DIESEL GENERATOR 8 UNAVAILABLE 3020.85 3264.95 3502.95 3584.48 2598.08 2764.78 2892.83 2852.13 2911.13 FSAR TABLE 8.3.4 DIESEL GENERATOR C UNAVAILABLE 3129.15 3268.35 3431.35 2912.75 3208.05 3283.05 3522.63 2532.63 2651.33 FSAR TABLE 8.3.5 DIESEL GENERATOR D UNAVAllABLE 3020.85 3264.95 3502.95 2981.15 3104.95 3050.95 3584.48 2598.08 2764.78 SEISMIC DIESEL GENERATOR A UNAVAltABLE 1041.3 3193.8 3661.4 1756.9 1726.2 2869.2 855.4 3283.6 3283.6 SEISMIC DIESEL GENERATOR 8 UNAVAILABLE 1004.9 2731.5 3411.5 1659.1 3525.2 3525.2 1012.3 2885.6 2900.6 SEISMIC DIESEL GENERATOR C UNAVAIlABLE 1134.2 1840.3 2987.9 919.3 3543.8 3543.8 1593.7 3470.7 3470.7 SEISMIC DIESEL GENERATOR D UNAVAllABLE 1025.9 2200.5 2885.1 933.7 3079.6 - 3542.6 1657.1 3069.2 3532.2 Diesel Generator Rating: 4000KW Continuous 4400 KW 2000 Hours Page 1

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