3F0997-30, Forwards Suppl Info to TS Change Request Notice 210 Re Proposed Changes to TS & Other Aspects of Licensing & Design Basis,To Address Mods & Procedure Changes Required to Mitigate Consequences of Certain SBLOCA

Forwards Suppl Info to TS Change Request Notice 210 Re Proposed Changes to TS & Other Aspects of Licensing & Design Basis,To Address Mods & Procedure Changes Required to Mitigate Consequences of Certain SBLOCA

ML20211D106
Person / Time
Site:	Crystal River
Issue date:	09/25/1997
From:	Holden J FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20211D112	List: 3F0997-30, Forwards Suppl Info to TS Change Request Notice 210 Re Proposed Changes to TS & Other Aspects of Licensing & Design Basis,To Address Mods & Procedure Changes Required to Mitigate Consequences of Certain SBLOCA ML20211D152 ML20211D161
References
3F0997-30, 3F997-30, NUDOCS 9709290050
Download: ML20211D106 (100)

Text

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Florida- -

Power ConPORAtl0N

• T *?* -

September 25,'1997 3F0997-3G U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

- Suppiement to Technical Specification Change Request Notice 210 ,

References:

1. FPC letter dated June -14,1997 (3F0697-10) " Technical Specification Change Request Notice 210"

2. FPC letter dated August 26,1997 (3F0897-25) " License Amendment Request 216, EDG Air Handling System"

3. FPC letter dated October 28,1996 (3F1095-22) " Crystal River Unit 3 Forced Outage"

4. FPC letter dated September 17,1997 (3F0997-31) " Request for

- Additional Information related to . Emergency Operating Procedures Technical Specifications Change Request Number (TSCRN) 210 (TAC g' No. M98991)"

l

Dear Sir:

g, .

Florida Power Corporation (FPC) hereby submits supplemental information as described in E Reference 1 which includes',11n part, a discussion of the modifications and procedures E necessary-to implement Technical Specification' Change Request Notice (TSCRN) 210. E-

'This supplemental information does not alter any conclusions of TSCRN 210. This E-information completes four of the commitments made in Reference 1. E-a Background , ,

.a  ?

~~

. In TSCRN 210, FPC - requested > approval of proposed changes to the Technical o,

.Q - Specifications, as well as other aspects;of the licensing and design basis, to address CRY. TAL RIVI # IN#80V FVu'" '"' * ****) W. Power Line Street .

• Crystal River. Morida 34428-6708 = 052) 795-M86

97 50g7g5 A Florda Progress Company

-p POR _ge

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U.S. Nucle:r Regulat::ry Commission 3F0997-30 Page 2 modifications and procedure changca required to mitigate the consequences of certain small break loss of coolant accidents (SBLOCA). The approval of TSCRN 210 is necessary to resolve identifed unreviewed safety questions (USQs) and to permit implementation of modifications and procedures required for plant restart. In order to maximize the time available for NRC review and to support the Crystal River Unit 3 (CR-3) restart schedule, FPC submitted TSCRN 210 as early as practical. This resulted in additional commitmen's as part of TSCRN 210 to provide supplemental information regarding several ongoing activities such as calculations, modifications and procedure changes. FPC and the NRC have held several meetings and conversations subsequent to the issuance of TSCRN 210 during which the NRC has requested additional information to support its review. FPC has provided additional information in Reference 4 and is providing its response to the NRC's other requests herein.

Supplemental Information The supplemer51information to suppod the NRC's review of TSCRN 210 is provided in the attachments hereto. This supplemental information does not alter FPC's "evious conclusioas or the basis for the conclusions provided in TSCRN 210. A summary of these attachments follows:

Attachment A - List of Commitments The attachment provides the list of commitments made in this submittal.

Attachment B - Calculations Consistent with Commitment 1 of Reference 1, FPC has completed the calculations for the emergency feedwater (EFW) block valve cycling and Control Complex Cooling, and has confirmed that the conclusions of these calculations support TSCRN 210. Additionally, FPC has confirmed that the required maximum accident loads on the EDGs are bounded by the lower limit of the emergency diesel generator (EDG) refueling interval surveillance test (i.e.,3300 kW) proposed by TSCRN 210.

Attachment C - Modifications Consistent with Commitment 5 of Reference 1, FPC confirmed that the modifix tions associated with TSCRN 210, except for impact on the EDG Air Handling System, will not involve an unreviewed safety question, and that no modification changes have been made which would alter tne Technical Specifications or Bases proposed by TSCRN 210. The attached Table 2 is a replacement for the Table 2 contained in TSCRN 210. Since the impact of the EDG uprate modification on the EDG Air Handling System involves an unreviewed safety question, a proposed amendment to l the CR-3 Operating License was submitted by License Amendment Request 216 i

(Reference 2).

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U.S. Nucle r R:platory Commission 3F0997-30 Page 3 Attachment D - Emeraency Operatino Procedures Consistent with Commitment 3 of Reference 1, FPC has confirmed that no changes have been made to the operator actions addressing a SBLOCA that would alter the Technical Specifications or Bases proposed by TSCRN 210. The attached Tables 3A and 3B are replacements for those contained in TSCRN 210. FPC has expanded Table 3B of TSCRN 210 to include a complete list of operator actions reqrired for mitigation of a SBLOCA. Both Tables 3A and 3B identify those operator actions that have been previously revieweo by the NRC. However, FPC requests that the NRC review these operator actions as an Integral part of TSCRN 210 in order to achieve a comprehensive review of the SBLOCA mitigation strategy.

Attachment E - Table Summary and Description of Init al Simulator Validations This Table includes the results of seven initial simulator validations of various Emergency Operating Procedure (EOP) scenarios, The results of these simulations indicate that operator response was within the time frame for required actions. This information was requested during meetings at the NRC's offices on September 18, 1997 to discuss FPC letter number 3F0997-31 dated September 17, 1997 (Reference 4).

Attachment F - Draft Anoendix to the Operatino License in a meeting held on June 24,1997, between the representatives of FPC and the NRC staff, the NRC discussed the possibility of incorporating one or more conditions associated with the issuance of TSCRN 210 into a new appendix to the CR-3 Operating License. The NRC requested FPC to propose a draft of such an appendix to facilitate issuance of the license amendment proposed by TSCRN 210, which is provided in this attachment, i Attachment G - EDG Test Plan As discussed in Reference 1, the EDGs are being modified to increase their service ratings. As part of the modifications, FPC will successfully complete testing in accordance with its written EDG test plan and obtain vendor certification to demonstrate that the EDGs are qualified to perform at their new service ratings specified by TSCRN 210 prior to entering Mode 4 from the forced outage initiated on September 2,1996. This attachment provides the test plan for the new ratings based on the installed modifications.

Attachment H - Update of Commitments Made in TSCRN 210 This attachment updates the list contained in Reference 1 and reflects the

- completion of those commitments associated with supplemental information and the

U.S. Nuclear Reguttory Commission 3F0997-30 Page 4 additional comrnitment associated with EDG testing -This attachment replaces the list of commitments contained in TCCRN 210 Attachment I- List of Acronyms and Abbreviations Used This attachment provides a listing of the acronyms and abbreviations used in the submittal.

FPC hereby incorporates the enclosed information into TSCRN 210. The above supplernental information demonstrates that TSCRN 210 is adequate as submitted, including the evaluation of the no significant hazards consideration and, when approved, will permit FPC to operate CR-3 within the approved design and licensing basis.

As the NRC is aware, FPC anticipates CR-3 will be ready to restart in December 1997.

FPC requested in Reference 1 that the NRC approve TSCRN 210 effective November 1, 1997, with an implementation period of up to 30 days. That schedule has not changed. To facilitate the NRC's approval of TSCRN 210, FPC suggests that a meeting be held approximately the first week of October 1997 to address the attached supplemental information.

If you have any questions concerning this supplementalinformation to TSCRN 210, please contact Mr. David Kunsemiller, Manager, Nuclear Licens:ng at (352) 563-4566.

Sincerely, fY

.J. Holden Director Site Nuclear Operations JJH/ma!

cc: Regional Administrator, Region ll Senior Resident inspector NRR Project Manager

7"

-l U.Sl NucleCr Regulatory Commtsion .  :

3F0997 30 ,

.Page5; s

l

Attachments;  !

!A. List of Commitments ~

B. Calculations

. C. Modifications _. . _

D Emergency Operating Procedures . i E. Table Summary and Description of initial Simulator Validations  :

+

F. Draft Appendix to Operating License

. G. EDG Test Plan H. Update of Commitments Made in TSCRN 210'

. l. List of Acronyms and Abbreviations Used 7

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F-U. S. Nuclear Regulatorv Commtsion 1 3F0997-30 21: Page8l ,

= STATE OF FLORIDA .

COUNTY OF CITRUS' John J. Holden sistes that he is the Director, Site Nuclear Operations for Florida Power Corporation;that he is authorized on the part of said company to sign and file with the Nuclear 1 Regulatory Commission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge,information,-

woc oelief.

A%4kitL JoYn J.Wolden, Director Site Nuclear Operations Sworn to and subscribed before me this 2h day of c+-[zo 1997, by John J. Holden.

/

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[ State of FloridagnatureMotary PM,ET,{j,Qg,,

/ ,,jaTl?og., g ,

E.[* .. '[.}

$\ #CC308952 jgE

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(Print, type, or stamp Commissioned Name of Notary Public) f

'Personallyi Produced

-O R-K wn Identification l

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT A l

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LIST OF COMMITMENTS l

ATTACIIMENT A LIST OF COMMITMENTS 1

i The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the submittal represent intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Manager, Nuclear Licensing of any questions regarding this document or any associated regulatory commitments.

ID Number Commitment ' ue Date D

3F0997-30-1 Florida Power Corporation (FPC) will Prior to entering Mode 4 successfully complete testing in from the forced outage -

accordance with its written EDG test initiated on September 2, plan and obtain vendor certification to 1996, demonstrate that the Emergency Diesel Generators are qualified to perform at their new service ratings specified by TSCRN 210.

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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 1 ATTACHMENT B CALCULATIONS i

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ATTACIIMENT B CALCULATIONS Many of the calculations to support TSCRN 210 were complete by the submittal date of TSCRN 210. Ilowever, to maximize the time available for NRC review and to support the CR-3 restart schedule, certain other calculations were sti ll pending completion at that time.

The anticipated conclusions of these pending calculations were used to support TSCRN 210.

To ensure that the conclusions of the calculations are valid, their inputs and assumptions were veriDed and subjected to interdepartmental reviews, except for the calculations involving Control Complex Cooling, EITV block valve cycling, and EDG loading, in TSCRN 210, FPC committed to:

• confirm that the calculations for EFW block valve cycling and Control Complex Cooling are complete and their conclusions support TSCRN 210, and

• confirm to the NRC that the expected maximum steady state accident loads on the EDGs are bounded by the lower limit of the EDG refueling interval surveillance test proposed by TSCRN 210.

Each of these calculations is discussed below.

Control Complex Cooline As described in the Safety Assessment for TSCRN 210, operation of the motor driven EFW pump is limited by EDG capacity limitations for certain SBLOCA scenarios. As such, the motor driven EISV pump (EFP-1) must be secured prior to loading the Control Complex Cooling chiller. In order to mitigate a SBLOCA involving a failure of the 'B' train DC electrical system, FPC would cross-connect the turbine driven EISV pump (EFP-2) to the 'A' train EFW flowpath prior to securing EFP-1. It has been determined that operators can complete this cross-connection in less than one hour. Therefore, FPC initiated calculations to-confirm that the Control Complex Cooling would not be required within the Orst hour of this postulated accident.

FPC has issued the calculations for the Control Complex Cooling issue. These calculations determined the transient temperature profile for the rooms in the Control Complex and compared these profiles to the maximum temperature for each room of the Control Complex.

This calculation concluded that the maximum temperature limits for each of the Control Complex rooms would not be exceeded for at least 60 minutes as long as the Control Complex and EFIC Room ventilation fans are operating within 30 minutes. The assessment of the EDG loads reflect: the operation of the ventilation fans concurrently with the motor driven feedwater pump. The operator action to ensure the ventilation is running is included in the list of operator actions discussed in Attachment D of this letter.

In summary, FPC has completed the calculations associated with the Control Complex Cooling and has confirmed that their results support TSCRN 210.

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U.S. Nuclear Regulatory Commission 3F0997-30 Attachment B Page 2 ,

EFW Block Valve Cycling As described in the Safety Assessment for TSCRN 210, the 'B' powered flow control valves, EFV-55 and EFV-56, would open and be unable to control EFW flow for certain SBLOCA scenarios with a loss of Battery 'B'. The OTSG would fill and the 'A' powered block valves, EFV-Il and EFV-32, would close once the steam generator overfill setpoint is reached. As the OTSG levels decrease to the overfill reset setpoint, the EFW block valves will open allowing EFW flow to the OTSGs. This cycling would be terminated by manual operator action when cross connecting EFW. FPC initiated calculations to determine the frequency of cycling and motor operator capability of the EFW block valves.

FPC has issued calculations addressing this EFW Block Valve Cycling issue. These calculations address the number of cycles the valves would experience, the OTSG overfill and reset setpoints, and the time limits for cycling the valves. The time the EFW block valves can cycle has been conservatively limited to the design temperature of the motor operator, which is a function of the number of cycles, the stroke time, and the ambient temperature postulated in the vicinity of the valves, and reflects revised setpoints for the overfill and reset settings. FPC calculations are based on a motor stroke time which is conservatively greater than the stroke time acceptance criteria for the surveillance testing of the motor operated valve (MOV). The ambient temperature is based on the CR-3 Environmental Qualification Program, which documents the temperature postulated in the intermediate building after one hour assuming a loss of ventilation. The results of the calculations determined that the EFW block valves are capable of cycling for more than one hour. Consistent with the safety assessment for TSCRN 210, this is sufficient time to complete the manual operator actions to cross-connect EFW and terminate block valve cycling.

EDG Loadinn FPC is performing modifications to the EDGs at CR-3 which will increase the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> and 200 hour0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> service ratings. FPC proposed in TSCRN 210 a revision to the EDG refueling interval surveillance test to address the increased service ratings. TSCRN 210 proposes an increase in the minimum load for this test from 3100 kW to 3300 kW.

TSCRN 210 describes that accident loads include the automatically connected steady state accident loads and the required manually applied accident loads. However, steady state loads do not include loads imposed by the starting of motors such as during block loading, and short duration loads such as motor operated valves, battery charger surges, and short duration pump surge flows.

As discussed in the cover letter of TSCRN 210, a revision of the CR-3 EDG loading calculatior. is ongoing. Based on the work completed to date, FPC has confirmed that the maximum required accident loads on the EDG will be less than 3300 kW with voltage administratively controlled relative to nominal voltage. Loads imposed by the starting of motors are not included in the service ratings and are less than the EDG manufacturer limits of

U.S? Nuclear Regulatory Commission 3F0997-30 Attachment B Page 3 '

3910 kW for such loiiding, As stated in the cover letter for TSCRN-210,- thel calculation is .  ;

scheduled.to be completed prior to the implementation of the license amendment.resulting from TSCRN 210.

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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT C MODIFICATIONS l

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ATTACIIMENT C MODIFICATIONS As stated in TSCRN 210, plant modifications supporting TSCRN 210 are in various stages of implementation. As such, the 10 CFR 50.59 evaluations for each of these modincations were not yet completed. Ilowever, to support the ongoing NRC review, FPC identified the modifications to be completed this outage related to SBLOCA mitigation in Attachment F, Table 2 of TSCRN 210.

In the submittal for TSCRN 210. FPC committed to confirm that:

• the associated modifications do not involve an unreviewed safety question, and e no changes were made in the proposed modifications which would alter the Technical Specifications or Bases proposed by TSCRN 210.

Since TSCRN 210 was submitted, additional modifications were identified associated with the uprate of the Emergency Diesel Generators (EDG). These modifications involve Appendix R fire protection for the cooling fan cables as well as replacement of the radiators and associated air flow issues. The need for the fire protection of the cooling fan cables is necessery since both room cooling fans are required for the operation of the EDG, as discussed in FPC's August 26, 1997 letter (3F0897-25). The potential for the EDG rooms to exceed the design basis temperemres was discussed in LERs 50-302/97-013-00 and 50-302/97-19-00. These modifications have been added to Table 2 of TSCRN 210 and indicated by a revision bar.

FPC has reviewed the modifications identified in Table 2 against TSCRN 210 and has confirmed that these modifications have not been changed in such a manner that would alter the Technical Specifications and Bases proposed by TSCRN 210.

FMEA/ MAR 97-08-12-01 FPC informed the NRC in an October 28, 1996 letter (3F1066-22) of its plan to address eight design issues related to restart. One of these eight issues involves the Failure Modes and Effects Analysis (FMEA) of a loss of DC power. This FMEA has been completed and has identified a system interaction that could occur during a SBLOCA concurrent with a LOOP and a loss of the 'A' train DC power, which was reported to the NRC in Licensee Event Report 97-021-00. This system interaction l would affect the ability of the operator to bypass the engineered safeguards (ES) signal l and obtain manual control of the ES systems, which is required by the SBLOCA analysis m iSCRN 210. The resolution of the system interaction is a plant modification, Modification Approval Record (MAR) 97-08-12-01, which will restore the operator's ability in appropriate accident scenarios to bypass the ES signal and l obtain manual control of the ES systems. Accordingly, the proposed plant modification

has been added to Table 2 of TSCRN 210. This modification will correct the system interaction and will not alter any conclusions of TSCRN 210.

Except for MARS 97-08-12-01, 95-05-15-01/02/05, and 97-08-04-01, the 10 CFR 50.59 reviews for the modifications identified in Table 2 have been completed. MAR 96-10-05-01

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U.S. Nuclear Regulatory Commission 3F0997-30 Attachment C Page 2 '

involving the- EDG Air Handling System and the uprate=of the EDGs was determined to involve an unreviewed safety question. Consistent with 10 CFR 50.59, the proposed amendment to the CR-3 Operating License was submitted to the NRC via FPC's August 26,  :

1997 ' letter (3F0897-25). The completed 10 CFR^ 50.59 reviews for the remaining modifications were determined not to involve an unreviewed safety question.

The MAR for which the 10 CFR 50.59 reviews is not complete at this time, has been reviewed by the CR-3 Safety Analysis Group (SAG), who has concluded that this plant modification will

. not involve an unreviewed safety question. The CR-3 SAG is responsible for the review and approval of the plant modification 10 CFR 50.59 reviews.

FPC has modified Table 2 to add the recent Loss of 'A' Battery ES Modification MAR and to *

! reflect the above information for each of the related modifications. The revised Table also reflects the location of the new generator installed for the auxiliary feedwater pump (FWP-7)

'in response to a NRC question during an earlier meeting.

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U.S. Nuclear Regulatory Commission 3F0997-30 Attachment C Page 3 Table 2 Modifications MOD MAR Subject Description Alters TSCRN 210/

USQ l 96-11-01-01 ASV-2N EFIC Auto Restores the automatic opening of ASV-2(M, the steam admission valve to EFP-2 Does not alter Opening on an "A" EFIC actuation. This will restore the load sharing capability of the TSCRN 210/

Reinstallation Emergency Feedwater System for the LOCA concurrent with LOOP and loss of No USQ EDG-1B in order to reduce the load on EDG-1 A involved 2 96-10-02-01 Emergency Installs passive flow restricting devices on the discharge side of both EFP-1 and Does not alter Feedwater Cavitating EFP-2. This will prevent excessive pump flow resulting in possible failure TSCRN 210/

Venturis mechanisms of runout or inadequate NPSII available. No USQ involved 3 96-10-10-01 EFV-12 Valve Replace valve EFV-12 on the cross-tie piping between EFW train A and train B, Does not alter 96-10-104)2 Mods, MOV a manual operated gate valve, with a motor operated gate valve. This will TSCRN 210/

96-10-10-03 Installation, Conduit facilitate operator action to open this valve remotely and route discharge of EFP- No USQ Supports 2 tarough the cross-tie piping to the OTSGs. involved 4 97-01-(M-01 EFP-2 Flow Installs llow indication from the cavitating venturis installed downstream of EFP- Does not alter Indications 2. This control room indication of EFP-2 flow rate will be powered from the TSCRN 210/

opposite train ('A' side) to provide flow indication slwid a 'B' side failure No USQ disable its flow indication. This will provide feedback to the operator of flow involved froin EFP-2 when EFP-1 needs to be secured for EDG load management.

5 97-(M-01-01 EFP-1500 psig Trip Installs a control switch to allow operator action to defeat the automatic trip of Does not alter Defeat Switch EFP-1 (500 psig RCS pressure). Defeating this trip will allow EFP-1 operation TSCRN 210/

during a SBLOCA. This switch will allow continued EFP-1 operation when No USQ DIIP-1 A starts on a 500 psig actuation, after EDG-1 A load management by involved operator action.

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment C Page 4 Tabic 2 Modifications MOD MAR Subject Description - Alters TSCRN 210/

USQ 6 97 44-02-01 RW/SW Pumps Pull- Replaces existing control switches with a Pull-To-Lock switch on Nuclear Service Does not alter To-Lock Switches and Decay Heat Seawater pumps RWP-2A and -2B and Nuclear Services Closed TSCRN 210!

Cycle Cooling pumps SWP-1 A and -1B. This will prevent automatic restart of No USQ these pumps on subsequent Engineered Safeguards actuation signal facilitating involved EDG-1 A load management.

7 96-12-17-01 EDG Small Load This modification will remove the auto-start function from both nonsafety control Does not alter _

Reduction circuits of the Flush Water Pumps. This will prevent them from auto-loading TSCRN 210/

Modifications, DOP onto the EDGs. No USQ 2A/2B involved 8 96-10-05-01 Diesel Power Uprate Implements modifications to increase the service ratings of the EDGs. (1) The Do:s not alter Project combustion air flow rate will be increased by replacing nozzle rings in TSCRN 210/

turbochargers with larger ones, and (2) combustion air intercoolers will be USQ, see replacu! with a dual pass intercooler. letter dated Aug.26,1997 (LAR 216) 9 96-03-12-01 Emergency Diesel Installs more accurate power meters (kW indication) for EDGs-1 A and -1B. Does not alter and Generator Indication Accuracy was further improved by changes to CT/ pts. EDGs can be loaded TSCRN 210/

Associated Upgrade higher because of improved instrument accuracy. No USQ FCNs involved

U.S. Nuclear Regulatoiy Commission 31V)97-30 Attachment C Page 5 fable 2 Modifications MOD MAR Subject Description Alters TSCRN 273/

USQ 10 96 46-02-01 EFIC Integral Installs windup reset on integral controller on the EFIC system. ~his will Does not alter Windup Reset provide for faster response of EFW for control of flow to the OTSGs. TSCRN 210/

This reduces EFW flow and consequential EDG-1 A loadin upon No USQ initia lon. involved 11 97-03-01-01 Standby Generator Instals a new diesel generator (not safety-related) to provide an alternate Does not alter for FWP-7 Mciup power supply for FWP-7. The generator will be installed on the TSCRN 210/

berm, but below the postulated maximum hurricane flooding level. I No USQ involved 12 97-02-17-01 MUV-27 IIPI Changes the Engineered Safeguards automati actuation logic for the Does not aher Autoclosure normal Makeup supply valve MUV-27 to add automatic closure upon TSCRN 210/

receipt of a diverse containment isolation signal (which also initiates IIPI). No USQ ,

The purpose of the modifkation is to aid in IIPI flow balancing actions in involved the event of a broken IIPI line: MUV-27 must be closed to help ensure accurate IIPI flow indication.

13 97-08-12-01 Loss of 'N Ilattery Provide circuit changes to allow the operator to bypass the ES signal and Does not alter ES Modification obtain manual control of the ES systems. TSCRN 210/

Will not involve USQ 14 97-05-19-01 EDG Appendix R Reroute and protect cables for the room cooling fans and add switches for Does not alter Cables fans to .he remote shutdown panel. Provide conduits and supports for the TSCRN 210/

rerouted / protected control cables for the room cooling fans. No USQ involved

U.S. Nuclear Regulatorf Commission 3F0997-30

. Attachment C Page 6 Table 2 Modifications MOD MAR Subject Description Alters TSCRN 210/

USQ 15 97-05-17-02 EDG Appendix R Provide conduits and supports for the rerouted / protected control cables for Does not alter Cables the room cooling fans. TSCRN 210/

No USQ involved 16 97-05 EDG Radiator Replace the EDG radiators, including the fans (increase air flow rate for Does not alter 01/02/05 Replacement radiators. Modification af radiator fan drive to allow higher fan TSCRN 210/

horsepower for cold weather operation to be determined based on results Will not of EDG radiator text runs. involve USQ 17 97-04-03-02 EDG Building Adds registers to the engine room supply air ductwork to reduce pressure Does not alter IIVAC loss in the ductwork, rebalance system to redistribute the air in the engine TSCRN 210/

room, replace the ventilation system filters with ones that have a lower No USQ pressure drop. involved 18 97-08-04-01 Radiator Discharge Modifications to the EDG building to minimize recirculation of radiator Does not alter Air Recirculation discharge air. TSCRN 210/

Will not i involve USQ

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACHMENT D EMERGENCY OPERATING PROCEDURES i

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ATTACHMENT D EMERGENCY OPERATING PROCEDURES As stated in TSCRN 210, the affected emergency operating procedures (EOP) necessary to-

' implement TSCRN 210 are in various stages of revision. However, to support the NRC's

' review,- FPC identified in Tables 3A and 3B of TSCRN 210, (1) operator actions required to be completed within the first 20 minutes of the SBLOCA scenarios addressed by the solution sets (Table 3A), and (2) new operator actions required to be completed after 20 minutes of these SBLOCA scenarios (Table 38). Some of these actions are " defense in depth."

In TSCRN 210, FPC committed to confirm that:

• the necessary procedure changes do not involve an unreviewed safety question, and e no changes were made to the proposed procedures which would alter the Technical Specifications or Bases proposed by TSCRN 210.

Consistent with this commitment, FPC has confirmed that no changes have been made to these operator actions that 'would alter the Technical Specifications or Bases proposed by

.TSCRN 210. To facilitate the NRC staff review, both Tables 3A and 3B have been annotated to identify the EOP steps that pertain to each action. The step numbers are as reflected in the draft revisions of EOPs submitted for review in Reference 4. FPC has expanded Table 3B of TSCRN 210 to include a complete list of operator actions required for mitigation of a SBLOCA.

! . Regarding whether the operator actions involve an unreviewed safety question, both Tables 3A and 3B identify those operator actions that have been previously reviewed by the NRC.

liowever, FPC requests that the NRC review the complete list of operator actions as an

, integral part of TSCRN 210 in order to achieve a comprehensive review of the SBLOCA mitigation strategy, l

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U.S. Nuclear Regulatory Commission 3F0997-30 Attachment D L Page 2 ,

Table 3A l Operator Actions Less Than 20 Minutes OA Operator Action Time Basis Prior NRC Referrnce Review 1 Tnp all running RCPs < 2 trenutes Required for Yes NRC letter to FPC dated 5/29/86 (Genenc Letter 86-05) refers loss of to B&W Owners Group (BWOG) studies which concluded that subcooling compliance with 10 CFR 50 46 is achieved if operator action to margin based on trip RCPs is taken within 2 minutes.

voiding condition of reactor See EOP-3. Step 2.1 and EOP.13 Rule 1,

• Loss of SCM*

coolant

, No changes were made that impact TSChN 210  !

! 2 If Subcoohng Margin (SCfA) is lost and ES has 10 minutes Required for Partial NRC letter to FPC dated 7Mi/79 (SER for Order dated 5/16/79 not actuated, initiate manual HPl and Reactor loss of based on TMI-2 Accdent) recognizes CR-3 revision to Building isolation and Cooling (RBIC) subcooling (Manual Enwpncy Procedure EP-106, which defines operator schon margin initiation in response to a spectrum of break sizes. States EP-106 was *

(precedes of RBIC " Judged to provide adeqde guidance to the operators to cope  ;

automatic has not with small break LOCA.* EP 106 (currently EOP-03.*Less of  ;

- isolates letdown (USQ6) initiation) been Subcooling Margin") contained guidance to initiate HPI and

-initiates HPI flow ' rariewed) ensure adequate HPI flow.

-isolates normal makeup (USQ6)

(contingency achons are provided in OA 4 See EOP-3. Step 2.1 and EOP-13 Rule 1. " Loss of SCM*

within 20 minutes if power is not available)

-isolates RCP seal control bleed off valves No changes were made that impact TSCRN 210

- actuates EFIC

-init:ates Emergency RB coohng 3 Ensure all four HPI injection valves are open 10 minutes Required only Yes NRC letter to FPC dated 5/29/79 ' Permanent Solution to 2- for loss of 1 train SBLOCA issue" recognizes operator achon to tum associated

- switch power supply for affected injection of Class 1E transler switch to open affected HPI valves by 10 minutes.

valves by manipulating switches in control power '

room See EOP - 3. Snep 3.3 No changes were made that impact TSCRN 210 i

U.S. Nuclear Regulatory Commission .;

3F0997-30 Attachment D - i

- Page 3 Table 3A -

Onerator Actions Less Than 20 Minutes OA Operator Action Time Basis - Prior NRC Reference Review ,

~

i 4 Isolate RCP sealinjechon (USQ6) 20 minutes Required to No FPC letter to NRC dated 2/28/79. answers a previous quest on  ;

maximize HPl of whether or not it was necessary to isolate any flow paths in i flow to reactor the makeup system after a LOCA. FPC refers to RCP seal injection and normal makeup and refers to a Gelbert Assooetes report that concludes adequate HPl flow is achieved without ,

these lines isolated. NRC letter to licensees with B&W (As a cordc@6cy action, if power is lost to designed systems (Genenc Letter 8405) dated S/29/86 stAes ~ -l MUV-27 (normal makeup) and MUV-18 (RCP the cooling water sources s@ porting the PCP with the  !

seal injection), transfer to an energized bus potential of being isolated are seal injechon, seal bleedoff, ,

and cicse valves) component coolmg water to seat hne coolers, and component cooling water to RCP motors and oil coolers The need to I isolate RCP Seal injection was discovered in 1995 to be necessary due to discovery that operators reiied on non-Reg a

Guide 1.97 instrumentation to measure this flow when -

determining HPI pump runout flow limits (see LER 95-026). I Seal injection isolation was also determined necessary dunng Refuel 10 in 1996 upon discovery that worst case instrument error may result in inadequate HP1 flow (see LER 96-006). .

FPC letter dated July 7,1997 (NOV 96-07) discusses the  !

additional need for closure of RCP t eal controlled bleed-off (CBO) valves after 90 seconds if seal injechon has not been i restored. See OA-2 See EOP - 3. Step 3.7 a

No changes were made that impact TSCRN 210 ,

e i

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U.S.- Nuclear Regulatory Commission 3F0997-30 Attachment D Page 4 Table 3A ODerator Actions Less Than 20 Minute _s OA Operator Action Time Basis Prior NRC Reference Review 5 Ensure adequate HPI flow (USQ6) (isolate a 20 minutes Required only Partial FPC letter to NRC dated 10/27/89 states HPl must be broken injection line using new isolation for break in HPI succeasfully balanced to support SBLOCA mitigation as criteria) line described in various B&W topical reports accepted by NRC.

Subsequent FPC letter dated 10/31/89 states that mitgation strategy employed from the late 1970's through the reviews done in response to NUREG 0737 releed on balancing HPl flow Requires bypassing ES to gain control of HPI for breaks in HPI injedion lines. These letters relate to LER valve.89-037, issued in November 1989 reportmg a design basis condition in which instrumentation used for balancing HPl flow was inadequate. NRC letter dated 12/20/89 confirmed verbal concu'rence to resume power operaton with the HP1 instrumentation problems. One condition was operator action for HPI flow balancing. NRC letter dated 2/17/95 from Gary Holahan to Ed Jacks (BWOG Operator Support Committee) states staff has completed its review of BWOG response to NUREG 0737 Item I.C.1 regarding EOP Guidelines and is finalizing an SER on the topic. Balancing HPI flows was a part of the ATOG/TBD guidelines incorporated into FPC procedures. FPC issued LER 96-007 on 3/15/96 to report another design basis condrtion involving HPI flow instrutnentation. The flow deiciencies desenbed therein were addressed by revised SBLOCA anafyses provided by Framatome Technologies in April 1996 which required isolation of the affected HPI line versus bafancing. Most recent FTl analyses have provided new isolation enteria.

See EOP - 3. Step 3.6.

EOP - 3. Step 3.9 is an *1f at any time

• step that addresses bypassing or resetting ES to ensure ES equipment is proper 1y liigned.

No changes were made that impact TSCRN 210

U.S. Nuclear Regulatory Commission

' 3F0997-30 Attachment D ' j Page 5. '

Table 3A Operator Actions Less Than 20 Minutes '

-OA Operator Action Time Basis Prior NRC Reference Revien i

6 Ensure adequate EFW flow (USQ6) 20 minutes Raise OTSG Yes B&W (7aylor) letter to NRC (Baer) dated 5/1/78 levels to ISCM provdes topical report 10104 *B&Ws ECCS .

setpoint (90%) Evaluation Model." which notes operator achon is necessary during early stages of the acodent to (EFIC was initiated in OA2; therefore, frutigate consequences and meet 10 CFR 50 46.

ensuring EFW r.ow is a confirmation step only) Auxiliary feedwater is assumed to be available. NRC letter to FPC dated 7E3/79 provdes a SER for actions ,

taken in res;mse to Commission Order dated 5/16/79. The SER states that a genenc review of This step manually raises OTSG levels to the B&W analyses entitled " Evaluation of Transient '

inadequate Subcooling Margin, ISCM level Behavior and Small RCS Evaks in the 177 Fuel Assembly Plant" resulted in a principle firuling that )

reconfirms SBLOCA analyses demonstrate a - t combination of heat removal by the steam generator

[

and the HPl system combmed with operator achon to , i ensure adequate core cookng These results are

{

opplicable to CR-3 consdenng the atsty to manually

~

start the redundant EFW pumps and HPI pumps from .i the contiof room, assuming failure of automatic EFW l

actuation. NRC letter to FPC dated 8/30/85 provides a SER for NUREG 0737 Item ll.K3.30, *SBLOCA Methods." Section Ill.5.a of the SER states *the timing of operator acbon to raise the secondary system water level to 93% was found not to be entical."

See EOP - 3. Step 3.8 t

No changes were made that impact TSCRN 210 .

f

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U.S. Nuclear Regulatory Commission 3F0997-30 Attachment D Page 6 Table 3B Operator Actions After 20 Minutes OA Operator Action Failure Prior Cycle Basis Reference Scenario NRC 11 Review Only 7 Start Control Complex Ventilaton in emergency LOEA No No To assure control room EvP - 3, Step 3.11 requires mode if fans are not already running LOBB operator dose is not exceeded concurrent pedwne& of EFP-2 and to provide control EOP-14. Enclosure 17, complex cooling Required to " Control Complex Emergency be accomplished within 30 Ventilation System."

minutes No changes were made that impact TSCRN 210 8 If at any time BWST is < 20 ft, transfer ECCS LOBA Yes No To ensure sufficent source of EOP - 3, Step 3.12 directs pump suction to RB sump LOBB borated water for injection by performance of EOP - 14, EFP-2 HPl!LPI. Depending on break Enclosure 19,"ECCS Suction size, action may be required Transfer."

between 25 minutes and 1-1/2 hours No changes were made that impact TSCRN 210 9 If "B' DC power is lost, crosstie EFP-2 to A train LOBB No Yes EFP-1 can only provide flow EOP - 3. Step 3.15 directs (EFV-12) for a speofic time penod, then pedwnes of EOP-14, EFP-2 must be aligned to Enclosure 11,"EDG A Load AND ensure sufficient margin is Management."

maintained on the 'A' EDG for Secure EFP-1 later adding of Control No changes were made that Complex Chiller impact TSCRN 210

4 U.S. Nuclear Regulatory Commission -

3F0997 Attachment D Page 7 Table 3fi

- Operator Actions After Than 23 Minutes OA Operator Action Failure Prior Cycle Basis Reference Scenario NRC 11 Review Only 10 Put EFIC in manual permissive LOBB .No Yes Required to prevent cychng of EOP - 3. Step 3.8 establishes the limited duty motors on the - appnMirty of EOP - 13 Rule AND EFW block valves - 3. 'EFW Control

• Also. EOP

- 14. Enc'osure 11 *EDG A Cbse EFW block valves (deenergized after Load Management " Step 11.4 closure) requires the normal EFP-2 discharge path to be isolated No changes were made that impact TSCRN 210

-11 Manage EDG load in order to extend EFP-1 EFP-2 No Yes Defense i.: Depth action for EOP - 3 Step 3.15 and EOP -

operation by - postulated single failure of the 8 Step 3.7 direct performance loss of EFP-2. These actions of EOP-14. Enclosure 11,

. Shutdown SWP-1 A & RWP-2A after extend the time EFP-1 is *EDG A Load Management."

verifying redundant pumps are operating available for OTSG coc ig and placing switches in Pull-to-Lock to No changes were made that prevent reactuation of pumps (EDG loading) impact TSCRN 210

. Place EFP-1 Tnp Defeat Switch in oefeat position to prevent automatic trip of EFP-1 wi RCS pressure of 500 psig i

l.

e U.S. Nuclear Regulatory Commission 3F0997-30 Attachment D

Page 8 Table 3B Operator Actions After 20 Minutes OA Operator Action Failure Prior Cycle Basis Reference Scenario NRC 11 Review Only 12 Start Control Complex Chitler if not already LOBA No No Required within 80 nunutes to EOP - 3. Otep 3.16 and EOP -

running LOBB ensure control complex S. Step 3.8 provide EFP-2 instrumentation remains within instnJctions to concurrently

, analyzed temperature ranges perform EOP - 14. Enclosure for instrument accuracy 18. " Control Complex Chiller ,

Startup" No changes were made that ,

ampact TSCRN 210 13 Isolate the RB sump by closing RB sump pump LOBA Yes No Required by IE Bulletin 79- EOP - 8, Steps 3.11 and 3_12 l o scharge valves, placing RB sump pumps in LOBB 05A to isolate systems utilized Pull-to-Lock, and closing waste gas header EFP-2 to transfer radioactive liquid No changes were made that isolation valves and gases from the impact TSCRN 210 containment. These 4

penetrations go to the waste gas header and the Miscetims Waste Storage tank. Isolating the RB sump penetration will maintain inventory in the containment "

4 tor possibie ECCS pump i suction for long term [

recirculation i

i

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment D Page 9 Tzble 3B Operator Actions After 20 Minutes OA Operator Action Failure Pnor l Cycle Basis R. "., dw Scenario NRi i 11 Renew Ont; 14 ff only EFP-2 es suppfyng feedwater to the LOBA W, No To maintam EFP-2 as an EOP - 8. Step 316 drects use OTSG. the RCS cooldown win ce stopped pior LOBB avadable source of feedwater of EOP-14. Erdosure 7, TFP-to reachng an EFP-2 operationallea nit. Manage and operate the pump witNn 2 Macing. .L*(see Steps up,4.w. of EFP-2 by doseg ASV-5 and ASV- anafyred regions Use of 7.12. 7.13. and 7.14) 204 on low OTSG pressure (Cyde EFW) and FVP-7 provides .Guc.-e

.estart EFP-2 when pressure increases resources avastable to This may involve entry mto w du = dunng a LOOP EOP -4 and retum to EOP - 8 No changes were made that (Megaton strategy includes operabon of desel impact TSC3 210 backed FWP-7 as a Defense in Depth acbon) .

I 15 If EFP-2 is not opeatmg e m a LOOP EFP-2 No Yes tf EFP-2 is net avadable. EOP - J. Step 3.15 daects corwitron with inadequate subccolmg. limit steps must be taken to ensure p-:n ' new of EOP - 14 cooldown pnor to EFP-1/LPI. EFP-1 operates as long as Endosure 11. EDG A Load needed M66(rp.m 4."(see Step 11.14)

No changes were made that impact TSCRN 210 16 Establesh RCS Cooldown useg TBVs or ADVs LOBA Yes No Irvtrates RCS cuwG.i n to EOP - 8. Step 3. t 7.

LOBB acfweve end pomt of event ErP-2 (start of decay heat) No changes were made that impact TSCRN 210

i.

. U.S. Nuclear Regulatory Commission 3f9997-30 Anachmem D .

Page 10 Table 3B Operator Actions ARw Than 20 Minutes

OA -C,_ A Acton Feelure Prior Cycle Bones Reference Sceneno NRC 11 i Rowlew Only

.17 Penoscany reevaluate HPt kne tweak cntena LOBA No No Requwed for spechc HPt kne EOP - 3 Step 321 transeons on RCS repressunzation LOBB pinch areas to ensure a to EOP-4 on inadequate heet

! EFP-2 twoken line wal be isolated d transler. EOP-4 Step 3.56 is solabon entene is not rnet an *tt at any twne' step and early in the evert while in requees closure of the allected

=

EOP-3 HPI tru, on ISCM 4

i M Were M M i wnpact TSCRN 210 l

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FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUhlBER DPR-72 ATTACHMENT E TABLE

SUMMARY

AND DESCRIPTION OF INITIAL SIMULATOR VALIpATrONS

U.S. Nuclear Regulatory Commission

+. '

3F0997-30 Attach 3nent E Page1 Time huJcuiSteps Based on Isss of SCM Operator Required ActualTime ActumCompleted(Mirrsec)DurmgSimulatorValmiatxas Required Time to 4 Action Complete 4 (Min) 7/30S7 8/5/97 8/6/97 8/12/97 8/13/97 9/15/97 9115/9 9/1987 9/19 S 7 2RO IRO 2RO 3RO 2RO See 2RO IRO IRO j 1SRO ISRO 2 SRO 2SRO ISRO scenarm 2 SRO 1SRO ISRO Scenario 1 Scenano2 Scenario 71 Scenarm Scenario Scenarm Scenarm Scenarm Scenare 69 75 88 89 97 98 4 Trip RCPs <2 Note 3 Note 3 Note 3 Note 3 Note 3 Note 1 0:06 00:25 0:01 Ensure llPI 10 3:10 2 20 2:00 421 1:47 3:05 2:52 Valves open Note 4 Note 4 Close MUV- 20 < 10:45 320 320 9:18 4:15 43 402  !

27 Note 4 Note 4 Isolatc high 20 14:30 15:15 SWO 13:13 N/A N/A N/A l flow line Close MUV- 20 10:45 6W0 6W0 1528 626 5:30 606 l

I8 Note 4 Ensure CC 30 23

13 18:45 1720 3237 15:54 13:35 14:55 Vent in Note 2 Emerg Mode Ensure CC 80 59:15 42:00 51:13 23:17 Note 4 1639 Chiller Note 4 running i Empty Box indicates time not captured Note 1: Confirmed trrpped I second after loss of SCM.

Note 2: Action was delayed due to damper indication problems resulting from battery failure. Indication is teing fixed in plant.  ;

Note 3: Pumps were stopped due to LOOP. Time not captured.

Note 4: Automatic actuation,confirmationof action only.

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l U.S. Nuclear Regulatory Commission 3F0997 30 Attachment E Page 2 ScenarioI 100% Power, MCC-3AB is aligned to the B Bus RCS leak develops in the B2 IIPI noule. l

+

leak starts at 50 gpm and increases to 200 gpm. Rx trip on low pressure or manual operator

- action. When Rx trips, a LOOP occurs including the 22KV backfeed to the air compressors. B EGDG fails to start due to loss of B Battery. EOP-02 is entered and immediate actions j performed. EOP-03 is entered on loss of SCM. The failed HPI line is isolated. EFW is cross- i t

tied to allow ' shutdown of EFP-1. The "A" chiller is started. Scenario ends when EOP-08 is =

entered. ,.

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. . .a . - - . . . _ , .....,_~ ,,.;.._..-._..__,.m.-. . - - . . . .. ~.__ . . , _ - - . - . , _ ,

U.S. Nuclear Reguttory CommUsion

' W7 30 l

>Alptiment E  ;

Pagu a i rq*g

.. 100% Power, MCC 3AB is aligned to the B Bus. RCS leak develops in the B2 IIPI nozzle. -

leak starts at 50 gpm and increases to 200 gpm. Rx trip on low pressure or manual operator action. When Rx trips, a LOOP occurs including the 22KV backfeed to the air compressors. B EGDO falls to start due to loss of B Battery. EOP-02 is entered atxi immediate actions i performed. EOP 03 is entered on loss of SCM. The failed IIPI line is isolated. EFW is cross-  :

tied to allow shutdown of EFP 1. The "A" chiller is started. Scenario ends when EOP-08 is '

entered.

i n

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- U.S. Nucicar Regulatory Commission 3F0997 30 Attachment E I

Page 4 Scenado # 71 Pzr Surer Line Runture Simulator Setus/ProgrammedFailures: l 100% RTP EOL No equipment OOS o Pressurizer surge line ruptures (full sheer).

o Rx is tripped by RPS on low pressure, o EFP 2 trips on overspeed, o When RCPs are tripped by operator, a LOOP occurs.

o Accccs to the AB is lost due to radiation levels. The PPO was not allowed back in to perform post event actions.

Scenario Challenses:

SCM will be lost and EOP 3 will be entered.

Alf EFW will tw lost due to the LOOP and the failure of EFP 2, but EFW is not required for LHLOCAs.

Time dependent actions for starting CREVs, and the Chiller will be balanced with ECCS transfer to the sump.

Expected procedure usage:

EOP-2 immediate actions will be performed.

EOP-3 will be entered due to a loss of SCM.

EOP 13 Rules will be used as appropriate.

EOP 8 will be used to perfonn ECCS suction transfer to sump, and boron precipitation.

6 i

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

U.S. Nuclear Regulatory Commission 3F0997 30 Attachment E Page5

$cenado # 69 LOCA Cooldown with loss of A Batters and Offsite Power Simulator Seto/Prosrammed Failu res:

100% RTP EOC Life No equipment OOS nS MCC 3AB will be powered from the A ES480 Volt bus.

o Small leak occurs in cohl leg (200 GPM).

o Rx la tripped by operator or RPS.

o Concurrent with Rx trip a loss of all off site power occurs.

o A Battery failure occurs (single failure) o The leak in the cold leg increase to 0.006 of a full cold leg break.

o Access to the AB is lost due to radiation levels. The PPO was not allowed back in to perfonn post event acilons.

o When EOP 8 is entered, the cold leg break will increase in size to 0.1 of a full break.

Scenado Challenses:

Loss of the A train will require transfer of ES status lights, transfer of ES MCC 3AB and closure of MUWIS and 27 to allow determinationof IIPI flows.

Loss of A Battery will also result in the use of the ES test switches to bypass ES to defeat the ES480v lockout. This will permit the starting of CREVs and the chiller.

When EOP 8 is entered the cold leg break will increase significantly. Rapid depressurizationof the RCS will occur. Building spray actuation may occur.

When LPl flow reaches 1400 GPM, a transition to the larEe break branch of EOP 8 will occur.

The ES test switches will prevent a reactuation requiring manual LPI actuation. The reactuation of ES will require restaning CREVs and Chiller.

Transfer if ECCS suctic n transfer to the sump will occur on the B train only.

The scenario will end when EOP-8 is exited and the TSC will be contacted.

Boron precipitation control will not be performed due to only one train of LPI and inability to open drop line.

Ewected procedure usace:

EOP-2 immediate actions will be performed.

EOP-3 will be entered due to a loss of SCM.

EOP-8 will be entered after completion of EOP-3.

EOP-13 Rules will be used as appropriate.

EOP-14 Enclosures 2,17,18,19 will be used.

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t U.S. Nuclear Regulatory Commission l 3IW7-30 ,

Attachnwnt E -

. Page 6 Stenado #75 Loss of SCM with no EFW and Degraded HM -

Simulator Setus/Prognavnmed Failures:

o 100% RTP.  ;

o EOL o FWP-7 R/T'd for motor rebuild i o ES MCC 3AB is powered from A ES 480 V Bus ,

o Units 1 and 2 steam unavailable A break occurs in the llPI line downstream of MUV-26 (App 200 gpm). When the Rx is  !

tripped, a loss of offsite power occurs and A EDG fails to start.

MUP lC is degraded to a maximum of 40% output. EFP-2 falls to start.

Operators perform the inunediate actions of EOP-02 then transition to EOP-03. ES MCC 3AB is transferred to B side and MUV 18 and 27 are closed. Operator isolates broken IIPI line to increase flow to the core but degraded MUP lC is not adequate to remove core heat. With EFP-2 unavailable, operators will transition to EOP-04 and attempt to regain OTSGs. After PORV is opened, ASV 50 is reset and OTSG cooling is established. Transition to EOP-08 is made. The scenario ends when MSIVs are isolated and operators are managing steam for EFP-2.

Scenado ChalltitEtn o Flowpath through EOP-03 with degraded IIPI o Transition to EOP-04 when lack of heat transfer is recognized o Transition from IIPI/PORV cooling to OTSG cooling o Control of cooldown and steam management for EFP-02 in EOP-08 t

U.S. Nuclear Regulatory Commission 3F0997 30 Attachment E Page 7 t

Scenado #88 SBLOCA resulting in loss of SCM and LOCA Cooldown Simulator Setun/Prverammed Failures:

o Mode 1100% FP o SBLOCA on "B2" IIPIline o Loss of offsite power o less of "B" Battery o Loss of Herm air compressors initially, only one RO and one SRO will be, present in the Control Room. Second RO will not be allowed to enter until 2 min after loss on' SCV.. Full E-Plan response is required for NSS ,

participation. STA will not be available until 10 minutes into the event. Time critical actions will be timed.

A SBLOCA develops on the "B2" IIPI line, When the reactor trips, a loss of offsite rower, B Battery and Ikrm air compressors occur. EOP-02 immediate actions are performed with transition to E0P-03 due to loss of SCM. Power is transferred to MUV-25 and 26. The high flow line is isolated and EFP 2 is cross connected to supply the A EFW train. Adequate SCM will not be regained and transition to EOP-08 will occur. Scenario ends when hold point in EOP-08 is reached for RCS pressure < 400 psi.

Scenado Challenges; o Flow path from EOP-02 to 03 to 08 o EFP 2 management o EDG 1 A load management

U.S. Nuclear Regulatory Commission ,

3F0997 30 Attachment E Page 8 Scenado #89 Larne OTSG ikbe Runture and LOCA Cooldown SLmulator Setue/ Programmed Fallut1; o Mode 1100% FP o 80 gpd tube leak increasing to 800 gpm in A OTSG o MUP-1 A OOS o MUP 1C trip Tube leak develops in A OTSG and rapidly increases to 800 gpm. The reactor is tripped and EOP-02 immediate actions are taken. SCM will be lost resulting in transition to E0P-03. MUP-IC trips on manual initiation of IIPI. E0P-03 actions will be taken with exit to EOP-08 since SCM will not be regained. A OTSG will fill to > 90% requiring TRACC isolation. Scenario crx!s when OTSG is isolated and RCS cooldown is in progress and under operator control.

Scenado Challengtti o Flow path from E0P-02 to 3 to 8 o TRACC isolationof A OTSG

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U.S. Nuclear Regulatory Conunission i 3F0997 30 Attachment E.

Page 9 l t

Scenario 197 Cold Leg Break (Full Shear)  !

Sankger Setus/Prosrammed Failures:

I o 100% FP, EOL o Nothing OOS l b

o Cold leg break (full shear) o DliV-42 fails to open during sump swapover l Break is inserted. Rx trip by RPS. AB access is lost due to radiation levels which won't allow PPO in to perform post trip actions. EOP-02 immediate actions will be performed with transition to EOP-03 due to loss of SCM. EOP-08 will be transitioned to from EOP-03 to perform sump swapover.

Scenario Challenges:

I DilP 1 A, BSP 1 A and A Train MUP must be tripped during sump swapover due to failure of DlIV 42

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U.S. Nuclear Regulatory Commission 3F0997 30 Attachment E Page 10 Scenado #98 Letdown Line Runture Downstream of MUV-49 Singulator Setun/ Programmed Failures; o 100% FP, MOL o Nothing OOS o Minimum crew staffing (1 SRO,1 RO) o Letdown line rupture in AB immediately downstream of MUV-49

- Break is inserted. Rx trip by RPS or operator. Access to AB is 1;ost due to rad levels so PPO  ;

can't enter to perform post trip actions. EOP-02 immediate actions performed. EOP-03 entered due to loss of SCM. Performanceof Rule I will isolate the leak. EOP-02 will be reentered after completionof EOP-03 Scenado Chattenern isolation of leak will cause SCM to be regained and ES will have to be bypassed to throttle llPI

FLORIDA POWER CORPORATION l l

CRYSTAL RIVER UNIT 3 l

DOCKET NUMBER 50-302/ LICENSE NUMllER DPR-72 ATTACIIMENT F DRAFT APPENDIX TO OPERATING LICENSE

ATTACIIMENT F DRAFT ADDITIONAL CONDITIONS  !

t i

in a nweting held on June 24, 1997, between representatives of FPC and the NRC staff, the ,

. NRC discussed the possibility of incorporating one or more coixlitions associated the issuance of TSCRN 210 into a new apperxlix to the CR-3 Operating License. The draft appendix attached hereto addresses the testing of the EDGs resulting from the uprate modifications.

This draft is provided at the request of the NRC in order to facilitate the issuance of the license ameruiment proposed by TSCRN 210. a v

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U.S. Nuclear Regulatory Commission 3F0997 30 Attachment F Page 2 1 DRAFT ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE No. DPR-72 Amendment Number Additional Conditions Applicability Date To Be Florida Power Corporation (FPC) will successfully Prior to entering MODE 4 Determined complete testing in accordance with its written EDG from the forced outage -

test plan and obtain vendor certification to demonstrate initiated on September 2. ,

that the Emergency Diesel Generators are qualified to 1996.

perform at their new service ratings specified by this Amendment, at which time this condition shall no longer be applicable.

j

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 ATTACIIMENT G EL)G TEST PLAN

Attachment G EDG Test Plan ,

Florida Power Corporation (Fit) is modifying the Crystal River Unit 3 (CR 3) emergency diesel generators (EDG) to increase their intermediate power ratings to add margin above design basis accident loads. The maximum 30-minute power rating of the EDGs is not changing. FPC has identified a rigorous test plan for the EDGs to confirm functionality, ,

engine qualification, and to confirm that EDG room cooling is adequate.  ;

The EDG test program includes multiple stans and loadings. During these runs, numerous  !

parameters will be monitored to ensure each aspect of the modifications is fully tested. Each EDG will be run for approximately 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, in addition to FIC's testing, the EDG vendor performed load testing on an EDO of the same type for the full duration of the new intermediate ratings,3200 kW for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> and 3400 kW for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. The EDG verxior has also performed cn analysis to confinn that FPC's EDGs, as modified, will be capable of operating at the new higher ratings.

The changes required for the FPC EDGs to meet the new ratings are minimal because most of the required modifications were installed in 1990, The post modification testing in 1990 assured proper perfonnance and confirmed qualificaticn of the EDG at its curren ratings. In order to establish the new ratings, only two stationary parts had to be replaced on the EDG engine. The power uprate does not require generator inodifications. In addition to the modifications required for the power uprate, the engine cooling radiators are being replaced due to low cooling air flow rate to the radiator cores discovered during EDG testing and to facilitate future plans.

FPC has taken additional measures to ensure its test plan is adequate to provide reasonable.

assurance that the EDGs will perfonn as designed. Two independent third party consultants i have reviewed the EDG test plan. Both concurred that the EDG test plan was adequate to demonstrate qualification and operability of the EDGs. In addition, FPC discussed the test plan with several IEEE 387 Board Members who concurred with FPC's EIX) test plan-as being adequate to demonstrate qualification and operability.

L)uring a meeting on June 24 1997, the NRC requested FPC to provide the test plan for the EDG modifications. This submittal provides:

1. A summary of FPC's licensing basis including commitments to applicable standards and '

guidelines.

2. A discussion of the modifications made to the CR 3 EDGs.

3. Basis for EDG qualification.

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U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 2

4. A discussion of the post modification testing program including a sum. nary of the EDG starting and loading used in the determination of qualification. The test p!an is based on the testing required to demonstrate that the modifications will perform as designed and to confinn adequate room cooling.

1. Licensingjais l CR-3 was designed, constmeted and licensed before either IEEE 387-1977 Reg. Guide 1.108 (the predecessor to Reg. Guide 1.9), or Reg. Guide 1.9 were issued. The EDGs for the plant were selected, installed and licensed in accordance with the design requirements or NRC Safety Guide 9. " Selection of Diesel Generator Set Capacity for Standby Power Supplies."

Although CR 3 is not required to comply with IEEE 387 or Reg. Guide 1.9, performance of the analyses and . testing described in this test plan demonstrates that FPC is meeting the technical content and intent of both those industry and licensing guidelines.

2. Stilttttlary of hdjBradons to CR 3 I:mergnisy Diesel Generators The following is a summary of the changes (modifications) to the FPC EDGs being installed in 1997 associated with the new service ratings. The modifications are listed according to the Modification Approval Record (MAR) packages issued for their installation.

2.1 MAR 96-10-05 150 kW UpIrmis_dTClpftart item D-6A)

A. Increased intermediate power ratings.

11. Replaced engine combustion air intercoolers with dual pass coolers and modified kical piping.

C. Changed out turbocharger noule ring with one which slows down the rotor.

D. Increased the minimum amount of fuel oil to be stored in the day tank and the storage tanks.

E. Raised instrumentation setpoints for day tank level instruments.

F. Increased the minimum storage requirement for lube oil.

G. Removed damper from bulkhead and close opening with a plate.

II. Require operation of two room cooling fans.

2.2 Mall 97 05-10 Appendix R Cable Modifications (FPClestart item D-29S)

A. Reroute / protect control cables for the room cooling fans and add switches for fans to the remote shutdown panel.

US. Nuclear Regulatory Commission 3F0997 30 Attachment G Page 3 2.3 bMR 97-95-17 Apoendix R Conduit and Conduit Supoort hiodifications (FPC Rrstart item D-29s)

A. Provide conduits and supports for the rerouted / protected control cables for the room cooling fans from the above hfAR.

2.4 MAR 97-0515 Emercency Diesel Generator Radiator bicchanical Reolacement ilTC Restart item D-29R)

A. Replaca radiators in::luding radiator fans (increase air flow rate of radiators).

H. Replace missile shield / flood barrier to reduce pressure drop / improve air flow rate of the ale ;;o!ng to the radiator.

C Reroute fuel transfer piping.

D. Add antifreetc to rodiator.

E. I.ucted minimum design temperature which in reased required fan horsepower durin; m rene cold weather operation (150F ambient).

F. hioved cagine room's exhaust dampers from one ,ide of bulkhead wall to the other so the dampers do not interfere with new radiators and reroute associated control air.

G. Added separate surge tank for the intercooler cooling system.

2.f l. JAR 97-05-15 Electrical /l&C Work for Emercency Diesel Generator Radiator Heplacement (FPC Rystart item D-29R)

A. Removal and reinstallation of electrical /I&C construction interferences.

H. Reroutes to match new positions of components.

C. Added lights inside the radiator compartment.

D. Added remote level instrumentation for surge tanks.

2.6 hf AR 97-0515 Emercency Diesel Geusrator Radiator Fan Drive Modification (FPC Restad item D-29R)

A. Modification of radiator fan drive to allow higher fan horsepower for cold weather operation to be determined based on results of the EDG radiator test runs.

2.7 MAR 97M03 Emercency Diesel Gengrator Huildine HVAC hbxlifications for thq_Radiater.Bepjacement (FPC Restart item D-29R)

A. Adds registers to the engine room supply air ductwork to reduce pressure loss in the ductwork.

H. Rebalance system to redistribute the air in engine room.

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U.S. Nuclear Regulatory Commission 3F0997 30 Attachment G Page 4 C. Replace the ventilation system filters with ones that have a lower pressure drop to reduce the pressure drop of the system.

2.8 MAR 97-08-N Correction of Radiator Discharce Air Recirplilation (FPC Restart hgin D-611)

A. Modifications will be made to the EDG building to minimize recirculation of radiator discharge air.

2.9 hiAR 96-03-12 kW Meter Uperade A. Reduce burden on CT/ pts to improve main control board kW meter accuracy.

2.A Discussion of EDG Modificatiem 2.A.1 The new power ratings will be as follows:

Upper Limit Duration Old Ratine New Ratine Chance Continuous 0 to 2,850 kW 0 to 2,850 kW Unchanged 2,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 2,851 to 3,000 kW 2,851 to 3.200 kW Increased 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> 3,001 to 3,250 kW 3,201 to 3,400 kW Increased 30 minutes 3,251 to 3,500 kW 3,401 to 3,500 kW Unchanged As documented in the previous section, the EDG intermediate ratings were increased by installing two stationary bolt on replacement parts which have no contact to moving parts on the engine. The new parts can be confirmed to be functioning properly by monitoring turbocharger RPM, blower and intercookr discharge temperatures, and cylinder firing pressure.

In order to provide adequate heat removal for the EDGs at the new ratings, the cooling system had to be modified. The radiator modification significantly improves the engine cooling by increasing the cooling air flow rate and significantly increasing the size of the radiator cores for improved heat transfer. A larger radiator fan is being installed: however, most of the moving parts associated with the radiator remain unchanged.

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page5

2. A.2 Missile Shield and Radiator Modifications Other modifications are being made in addition to the engine modifications discussed above.

The missile shield on the exterior of the EDG building is being modified to reduce its resistance to air flowing through it to the radiator, the engine combustion air, and the engine room cooling system. The missile shield does not connect directly to any EDG component, but it influences proper EDG performance.

The engine radiator is also being replaced with one of the same configuration but larger heat transfer capability. The new radiator provides more cooling air flow because its cores are significantly larger than the o'iginal units and have a lower pressure loss for air flowing through them. A new fan designed for the higher flow and lower pressure loss is also being installed. The new fan uses the same nominal horsepower as the original fan. The new fan is designed for the conditions of higher flow at a lower pressure rise so it uses the same horsepower as the original fan during standard operating conditions. The new radiator is the same configuration as the original radiator so the radiator fan drive is effectively the same as the original fan drive. The new radiator fan drive uses the same basic parts, so there is no significant impact of the new radiator fan drive when compared to the original radiator fan drive. The primary differences are the length of the drive shafts, changing the gear ratio in the right angle gear box to reduce the speed of the fan, and the diameter of the fan is increased. The new fan is different than the previous fan, but it will be tested for proper operation during the 3 day EDG functional test.

The radiator replacement has been fully analyzed by the vendor. Coltec report VTS-985-970714-OlR, dated 7/31/97 documents the analysis. This analysis for the radiator and its interface with the engine is referenced in the following evaluation of the modifications. The radiator analysis plus the EDG power uprate analysis VTS-985-961108-0lR, Revision 2, dated 5/28/97, documents the vendor's confirmation of the EDG's capability to perform up to 2955 kW. The vendor report will be updated after testing of the new radiator to confirm the radiator's capability to adequately cool the engine at the new ratings.

2. A.3 Imnlementing the_llse of Antifree7e During evaluations for replacement of the radiators, the potential of freezing the radiator during the first several minutes of operation was identified, The planned corrective action is to add antifreeze to the radiator water on a year-round basis. The use of antifreeze reduces the heat transfer capability of the fluid. Ilowever, the new radiator and radiator fan are adequate even when accounting for this effect. The air flow through the fan has been increased to provide more heat transfer to account for the reduction in heat transfer due to the antifreeze.

The increase in air flow, in conjunction with the reduced differential pressure, causes the fan horsepower to closely match the fan horsepower required for the old fan. Coltec analysis VTS-985-961108-OlR, Revision 2, dated 5/28/97, documents the vendor's confirmation of the

U.S. Nuclear Regulatory Commission 3F0997 30 Attachment G Page 6 EDG's capability to use antifreeze in the engine and radiator. Similar engines have been using antifreeze for years with no known adverse conditions.

2. A.4 Implementine a lower Minimum Design Temperature for Outside Air FPC has conservatively reduced the design value for the minimum outside air temperature to 15'F. This affects the design value for the radiater fan horsepower. The fan effectively moves the same CFM even though the density of the air changes. As the air gets colder, it becomes more dense and the required fan horsepower increases. With the minimum design temperature lowered, the theoretical fan horsepower required increases. On the average, the minimum design temperature of 15'F occurs I hour per year per the Crystal River Nuclear Plant " Environmental and Seismic QualiGcation Program Manual" zone OS.

The fan horsepower required for the 15'F condition is 264 horsepower per Coltec report VTS-985-970714-OlR, dated 7/31/97. This is 34 more horsepower than was previously analyzed prior to the uprate and the installation of the new radiator. This equates to an increase of 0.7% above the previously analyzed maximum engine horsepower of 5012. This is considered a negligible change which has been confirmed by the vendor. The vendor notes this condition occurs during the best operating conditions of the engine, when it is cool and drawing very dense combustion air for power operation. As soon as the engine warms up, in approximately 15 to 30 minutes, the low temperature condition is eliminated because the radiator is heating the air, in summary, a 0.7% increase in engine horsepower prior to the uprate is negligible because the engine is capable of meeting this short duration requirement as con 0rmed by the vendor.

2. A.5 Elimination of Recheulajion of Radiator Discharce Air During 1997 surveillance testing of the FPC EDGs, a portion of the radiator discharge air was identified as recirculating back to the radiator room intake structure. This recirculation was measured during the testing as affecting the supply air temperature by as much as 15'F. The condition increases the supply air temperature to the radiaior, combustion air, and the engine room cooling. Appropriate rnodincations will be installed to minimize the recirculation of radiator discharge air and confonn with required design. The planned modifications are to the building. The affect on the supply air temperatures will be reduced to a negligible in0uence as determined by wind tunnel testing of the revised con 0guration. Functional testing will validate the wind tunnel testing and confirm proper EDG performance.

2. A.6 Engine Room Ventilation Modi 0 cations The cooling to the EDG engine room is being increased for two reasons. First, the heat load to the engine room is increasing due to the engine and generator releasing more heat to the room due to the increased intermediate ratings. The increased heat loads are documented in I

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U.S. Nuclear Regulatory Commission 3F0997 30 Attachment G Page 7 Collec report VTS 985 970714-OlR, dated 7/31/97. The second reason is testing performed in 1997 identified the heat load in the engine room at the present EDG ratings is higher than was previously documented. ,

The now rate for the cooling air to the room will be increased approximately 50%. The  !

increase will be accomplished by requiring both ventilation supply fans for each engine room to operate. There are two ventilation supply fans for each room. These fans are not driven by  !

the engine of the EIX3. Prior to this modincation, only one of the two fans was required to operate. Both fans have always automatically started when the EDG started. The procedures .

allowed one of the fans to be secured after both fans were conurmed to be operating.

Therefore, the only operating change is to eliminate the steps in the operating procedures allowing one fan to be secured. (Control cables for some of the fans are being rerouted and switches being added to comply with Appendix R protection requirements, but the operation of the fans is not effected). The cooling capability is tied to outside air temperature and the ventilation calculation has confirmed two fans are required only when the outside air temperature is above 85'F. (This value was conservatively determined and the number may change based on testing after all the modincations are installed).

The use of two fans is not considered a signincard reduction in the EDG system reliability, but it is considered an Unreviewed Safety Question (USQ) since it can increase the potential for equipment malfunction. The USQ has been submitted for NRC approval by FPC's letter number 3F0897-25 dated August 25,1997.

The supply fan Hiters are being replaced with ones having a lower pressure drop and grills are being added to the ductwork to reduce back pressure in the system. These changes are being implemented to maintain the present two fan air now when the new radiators are installed.

The radiators will slightly block a portion of the engine room exhaust damper. The ventilation system resistance is being reduced by the modifications discussed above to offset the additional resistance of the partial blockage of the exhaust damper. General system balancing will also be performed.

Engine room and component temperatures will be monitored during testing to confirm adequate cooling. The evaluations will account for the actual outside air temperature during testing.

3. Basis for Emernency Diesel Generator Oualificall0B The EDGs will be qualified after the mod'ncations based on testing performed by the EDG vendor, analysis by the vendor and FPC, similarity of the components being replaced, and site testing of the EDGs after the modincations.

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U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 8 3.1 Oualification of the Engine hiodifications and the Power Uprate The engine modifications and power upate have been qualified by the vendor and will be confirmed by site testing. The vendor performed the qualification for Baltimore Gas and Electric (BG&E) by testing a similar EDG at or above 3300 kW for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> and at or above 3500 kW for 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. The Coltec qualification test reports are included as an enclosure to this test plan. The applicability and capability for the FPC EDGs to be uprated using the BG&E configuration was analyzed by the EDG vendor. Refer to section 3.4 below for a comparison of the FPC's EDG to the BG&E EDG, The FPC EDG power uprate analysis is documented in Coltec's report VTS-985 961108-0lR, Revision 2, dated 5/28/97.

This analysis documents the vendor's confirmation that the FPC EDG and its support systems will perform as required at the new ratings. The report accounts for the test EDG not having a radiator fan for engine cooling by reducing the rated power output of the FPC EDG by 100 kW. The 100 kW accounts for the horsepower used to power the crankshaft driven radiator fan. The report analyzes the FPC EDG support systems, including the radiator, and confirms the FPC EDG and its support systems capability at the new ratings.

Electrical systems and protection systems were evaluated and confinned acceptable for the new intermediate ratings. This was expected since the maximum rating is not increasing. Even though the maximum rating is not increasing, FPC did extensive testing of the exciter in March 1997 and reconfirmed the capability of the component. The generator capability was reconfirmed in Collec's report VTS 985 961108-0lR, Revision 2, dated 5/28/97.

In addition to our internal review of the EDG modifications and test program, FPC requested assistance from numerous industry experts. Specifically we consulted with Coltec (the EDG vendor), the Woodard Corporation, MPR Associates Inc., and two IEEE 387 Board Members.

These personnel reviewed the engine modifications and the power uprate. All agreed that the proposed modifications and test plan for the CR-3 EDGs are acceptable and clearly demonstrate qualification and operability of the EDGs.

In summary, the EDG modifications for the engine modifications and power uprate are qualified by the qualification testing performed by the vendor. The vendor, Coltec, confirmed the similarity of the FPC EDGs to the BG&E EDG tested.

3.2 1992 Modifications by Baltimore Gas and Electric (BG&E)

The following discussion provides additional information for understanding the similarity between the BG&E EDGs and FPC's EDGs.

In 1991, BG&E needed to increase the ratings for their EDGs at Calvert Cliffs. At that time, the BG&E EDGs were units of the same type as the pre-1990 modified FPC units. Ilowever,

U.S. Nuclear Regulatory Commission 3F0997 30 Attachment G Page 9 they needed higher intermediate ratings than the FPC units had been qualified and modified to in 1990. The EDG vendor suggested implementing the same changes FPC had installed in 1990 plus two other changes. The additional changes were to use dual pass combustion air intercoolers and to change the nozzle ring in the turbochargers to a pitch that allows the turbocharger rotor to operate at a lower RPM. These changes have been implemented at FPC in 1997 and are the subject of discussion above.

3.3 Discussion of HG&E Intermediate Rating Power Testing HG&E required a rigorous and comprehensive qualification of the EDG modifications by the vendor. The vendor performed qualification tests on BG&E's spare EDG after modifying it to the new configuration. This test EDG was outfitted with the same configuration as the FPC EDGs except for the two additional changes. The vendor then performed power qualification testing on the spare EDG for a total of 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br />. The spare EDG was operated for 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> at 3300 kW and 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> at 3500 kW. The test was for the full duration of the ratings.

The testing is documented in Collec report titled 200 Hour Rating Qualification Test Report and Coltec report titled Results of 2000 llour Test at 385 BilP/Cyl. on 38TD81/8 O. P.

Engine, Report #R-5.08-0236 (9/13/94).

3.4 Comnarison of FPC Emercency Diesel Generator to the HG&E Emergency Diesel Generator Tested The BG&E test EDG was the same configuration as FPC's EDGs except for two minor engine differences. Therefore, the new intermediate ratings for the HG&E EDGs are applicable to FPC's EDGs after the new intercoolers and nozzle rings are installed on the FPC engines.

Documentation supplied in Table I confirms the BG&E EDG is the same configuration as FPC's EDG. This information has been confirmed by the EDG vendor. The FPC's EDG engines were modified in the first quarter of 1997 to install the :ew turbocharger nozzle rings and the new intercoolers. The functional testing of the modifications confirme<1 the performance of FPC's EDGs matches the test resub of Coltec's testing of the BG&E test engine and also the test results of testing by HG&E at their site. Numerous parameters were measured during the testing. The key component parameters monitored are identified in the attached Table 2. The comparison of the test results confirm that FPC's EDGs have the same performance parameters as the qualified BG&E engines.

3.5 EDG Modification impact Analysis FPC performed an analysis of the impact of all the modifications on each of the EDG subsystems. Evaluations were performed on the potential impact that the changes could have on EDG starting and power operation. Also, the testing required to demonstrate that the EDOs L

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U.S. Nuclear Regulatory Commission 3F0997 30 Attachment G Page 10 are not adversely affected was identified. This analysis is included as Table 3. Impact of 1997 Uprate Modification to EDG reliability for Starting and Operation.

3.6 Oualification of The Radiator Modifications The new radiators are larger but are of the same general configuration and utilize essentially the same drive components for the radiator fan. The EDG vendor, Coltec, provided the new radiators. The vendor analyzed the new radiators and confirmed the radiators will perform the required function. Since the radiators are of the same general configuration and perform the same function, the new radiators are considered a component replacement. The testing discussed in this submittal and the analysis performed by the vendor ensure qualification of the EDGs. Quatincation was confirmed by the EDG vendor, two independent third party reviews, and discussions with five members of the 1995 IEEE 387 Board. All the modifications to be implemented durine the radiator outage were discussed with all five IEEE 387 Board Members. The Board Members agreed that the radiator modifications would not impact EDG qualification.

3.7 Summary The maximum rating for the FPC EDGs is not increasing. The BG&E Calvert Cliffs spare EDG, matching FPC's units, was power tested by the vendor for 2200 hours0.0255 days <br />0.611 hours <br />0.00364 weeks <br />8.371e-4 months <br /> total at the new intermediate ratings. In addition, the two BG&E EDGs of the same type at Calvert Cliffs have been site tested and put in servi:e with the modifications and new ratings. The changes to the BG&E EDGs have been in operation for more than three years. Therefore, the design has been thoroughly tested. The only changes that were required to make the FPC's EDG configuration match the qualified configuration of the BG&E EDG was the replacement of the turbochargers nozzle rings and replacement of the intercoolers.

4. EDG Post-Modification Test Platl FPC is implementing the test plan discussed below. The testing confirms functionality, qualification and operability of the EDGs. The EDG testing objectives are:

1. Demonstrate the engine operates within specified limits at new power ratings. The key engine parameters are: turbocharger RPM, intercooler discharge temperature, firing pressures. Other EDG parameters to be monitored include jacket water and lube oil temperatures.

U.S. Nuclear Regulatory Commission 3F0997 30 Attachment G Page 11

2. Demonstrate generator will perform as required at .iew power ratings. Main parameters to monitor included generator stator and field temperatures and also the exciter amps at the required generator output kW / kVA. t-

3. Demonstrate new radiator will perfonn its function at new ratings. The prime parameters to be monitored are: cooling air flow rate / radiator fan performance, heat transfer across radiator, heat transfer between engine components and radiator.

4. Demonstrate radiator discharge air recirculation has been eliminated by confinning the air  ;

temperature in radiator room is approximately the s4mc as the outdoor temperature.

5. Demonstrate the engine room and component temperatures within criteria. The room exhaust air temperature and the temperature rise across the room will be monitored, in addition, the temperature of various components and cabinets will be monitored. The tests evill be run a sufficiently long time to achieve temperature stability during data gathering.

The test results of the monitored parameters will be compared to the analysis and allowable operating parameters defined by the verdor to validate proper performance and assure similar results to the qualified BG&E engine.

The following is a general schedule for testing the emergency diesel generator radiator following the radiator modifications:

A .- Unloaded Maintenance Run Slow Start with incremental increase in speed (500 to 900 RPM)

a. Inspect for leakage, multiple para.neters checked

b. If major leakage, maintenance required, then repeat Unloaded l

Maintenance Run

c. Stop Engine with Overspeed Trip Test Duration: Approximately 30 Minutes B. Slow Start and Load (2625 to 2825 kW)

Record Radiator Data (Fl:..v. dp)

Record 50+ parameters Shutdown engine Evaluate data for changing Radiator Fan Blade Pitch Adjust Fan Blade Pitch Duration: Approximately 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />

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

U.S. Nuclear Regulatory Commission 4 ,

-i 3F0997-30 :.

L Attachment G -

! Page~ 12; 3 C. Slow Start and Load (2625 to 2825 kW)._ ,

iRecord Radiator Data (Flow, dp)-

i Record 50+ parameters '

. Shutdown engine

~ Evaluate data for changing Radiator Fan Blade Pitch i Adjust Fan Biade Pitch Durationf Approximately 2 to 3_ hours .

D. Slow Start and Load (2625' to 2825 kW) l

- Record Radiator Data (Flow, dp);.

Record Room Ventilation Fan Data (Dual / Single Fan, Clean / Dirty Filter)

- Record 100+ parameters t

Duration: Apisoximately 24+ hours E. Fast Start and Load (2625 to 2825 kW)

Modified Missile Shield Record Radiator Data r Record Room Ventilation Data Record 50+ parameters

- Duration:. Approximately 1 to.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, F.- Slow Start and Load (2625 to 2825 kW)

SP-354A Final operational run with test equipment removed from engine room g Record 50+ paranneters Duration: Approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> G. Review tess results and perform additional testing if necessary.

Summary of 1997 EDG Starts and Loads Aher Radiator Modification

.  ?

The following table summarizes the starts and rated load runs for the EDGs. This table L . includes normal surveillance runs as tvell as those done for post-maintenance testing. The l EDGs Lhave passed all surveillance tests since the completion of the turbocharger _and e iniercooler modifications. Throughout the modification- process, the EDGs have met the' reliability goals of the EDG surveillance program.-

1 I No; of Engine No. of Rated Load Estimated No. of L

e EDG Train Starts (No Load) Runs - Operating Hours "A" Train 3"' 12'" 64 L "B" Train 4"' 9"' 50

~(1). Includes only those starts and rated load runs since installation of new turbochargers and intercoolers.

+

Y 'Tgr-F#FT sf 1D'1y $ =- esueq= WwW r 7'.m. ..en 'k ta- -r-

• w---T-r af av tr-'w-e- --'NT' +ws w -- er e er 's a--u - * '-r'

U.S. Nuclear Regulatory Commission 3F0997-30 .,

-- Attachment G Page 13'-

The following is a general summary for the Power Uprate testing:

'A . Slow' Start and Load (2625 to 2825 kW)

Record 100+ parameters Duration: Approximately 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> B. Power Testing at 2000 Hour Rating (3100 to 3175 kW)

Record 100+ parameters -

Duratiom Approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> -

C. Power Testing at 200 Hour Rating (3300 to 3375 kW)

Operate during dual and single HVAC fans to confirm adequate cooling during both conditions Record 100+ parameters.

Duration: Approximately 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> D. Full load rejection test (greater than 3300 W') .

E. Largest single load rejection test - shutdown F. Hot Start and Load (2625 to 2825 kW)

Record 50+ parameters Duration: Approximately l' hour ' shutdown G. Fast Start and Load (2625 to 2825 kW)

Record 50+ parameters Duration: Approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> - shutdown -

H. As a contingency, additional hot starts or engine ambient starts and loaded run testing may be performed based on system performance results.

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U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 14 Summary of EDG Starts and Loads For Power Uurate The following table summarizes the additional starts and lated load runs that are planned as part of the post modification testing planned after all EDG modifications are coinplete, prior to entering Mode 4.

No. of Engine No. of Rated Load Estimated No. of EDG Train Starts (No Load) Runs Operating liours

" A" Train N/A 3 63 _

"B" Train N/A 3 63 The order and exact durations of the testing may vary. All the tests will be performed and the basic intent of the durations will be maintained, The kW loading values account for instrument inaccuracy of 25 kW, so the test load is 25 kW less thcn the rating to prevent exceeding the rating. An operating band is provided to account for instabir af the grid and operator control. The power level is maintained in the continuous rating of ~ EDG when not performing a specinc power test to maintain the heat load in the room. The heat load is maintained to monitor IIVAC cooling capability during an extended run.

Prior to restart FPC w;11 perform the combined ES actuation with simulated LOOP test which requires the EDG to automatically load. This test may not be performed at the same time as the other EDG testing because the ES components must be made available to perform the test.

Other plant restraints may prevent some of the ES components from being available concurrently with the EDG test.

After the testing is complete for each EDG, FPC will perfonn evaluations of test results with the vendor and confirm all systems function as required for the new ratings.

Summary of the 1997 Power Uprate Testina Proaram Iuls Description Start and load run Start and run at 2625 to 2825 kW for I hour Fast start Time to start & reach voltage and frequency Combined ES Act Simulated LOOP with de-energizing safety bus. EDG auto-start and auto-load sequencing, load > 5 minute Single load reject. Single largest rejected. voltage & frequency monitored Full load rejection Reject maximum load. voltage maintained & no overspeed Endor. & margin At least 22 hrs @ 2625 to 2825 kW & 2 hrs @ 3325 to 3375 kW, see note below Ilot restart Full temp. restart - voltage, frequency & time

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 15 Note for Endurance and Margin test - The design of the CR-3 electrical system does not allow loading of the diesel to pre-specified loads in a LOOP con 0guration (i.e., not connected to the offsite grid). Therefore, while voltage and frequency will be monitored, they will be significantly innuenced by the offsite grid. The voltage and frequency monitoring done during the load rejection tests serves to assure these parameters are being properly controlled.

The following table provides a summary of all starts and rated load runs completed or planned from the completion of the intercooler and turbocharger modiDeations until the end of the power uprate testing. The test program involves multiple starts and extended loaded runs as summarized in the table below. These tests will provide the data necessary to ensure qualification.

Summary of 1997 EDG Starts and Loads Estimated No.

EDG Train No. of Engine No. of Rated of Operating Total Starts /

Starts (No Load) Load Runs Hours Loads "A" Train 3 15 127 18/15 "13" Train 4 12 113 16/12 Conclusion A rigorous and comprehensive test plan is being implemented to confirm the EDGs are functioning as designed. The necessary EDG systems will be tested and monitored to confirm proper function. In conclusion, FPC is implementing a test plan which will confirm the qualification of the EDGs.

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 16 Table 1 BG&E/FPC EDG PARTS COMPARISON BG&E PART DESCRIPTION EEC 38TD8-1/8 Engine make/model same 12 # Cylinders same 16609598 Piston Assy - Upper same 16609599 Piston Assy - Lower same 16609595 Cylinder Liner / Belt Assy same 16600739 Crankshaft - Upper same 16600738 Crankshaft - Lower 16600998

16604650 Block same 16600737 Con Rod - Upper same 16600736 Con Rod- Lower same 16603864 Camshaft same 16605621 Bearings same 16501476 Blower same 16611024 F.1. Pump 16611017"'

166N922 L.O. Pump 16605139"'

16610746- J.W. Pump 16600637"'

16601667 1.C. Pump same 16600208 Fuel Pump / Drive same 16611593 Turbo Assy - 23" same 16402002 Governor 167N844"' .

16605593 Vertical Drive 16600727"'

TG2DJ Generator Model same 00604217 Generator Stator same 006N189 Generator Rotor same 11905506 Exciter same Discussion of differences:

(1) -_ ' Lower crankshaft and L.O. Pump on FPC has provision for power take off.

(2) Fuel Injection Pump on FPC is a "back header" vs. BG&E " front header". Both pumps have 5/8" piunger/ barrels and deliver same qualities of fuel to engine.

(3) Pump casting rotated to modify position of inlet / outlet. No other difference.

(4) BG&E uses EGBIO vs. FPC uses UG-8. No difference in load carrying capability.

(5) BG&E shafts incorporate end plate for easier maintenance. FPC shafts have " shouldered" ends.

No difference in load bearing or wearing components.

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 17 TABLE 2 COMPARISON OF KEY ENGINE PARAMETERS TESTING FOR ENGINE POST MODIFICATION BG&E 3GAE Engine as Engine at Calvert Coltec Cliffs 'A' Engine at FPC 'B' Engine at FPC 3054kW 3250kW 2690kW 3190kW 2730kW 3145kW htmg Pressure for Cylinder 1 1250 1340 1200 1340 1220 1380 2 1238 1290 1200 1340 1200 1330 3 1210 1300 1100 1360 1200 1330 4 1230 1340 1100 1340 1210 1340 5 1240 1340 1100 1380 1250 1380 6 1240 1280 1200 1340 1230 1360 7 1190 1300 1180 1320 1220 1340 8 1240 1330 1200 1360 1240 1370 9 1200 1300 1180 1340 1230 1380 10 1210 1260 1200 1320 1220 1350 11 1210 1320 1200 1360 1250 1380 12 1270 1340 1200 1360 1240 1360 Turbo Charger RPM Control Side 17861 18050 16545 17925 16600 17959-Opposite 17568 17700 16320 17700 16421 17721 Control Side Inter-Cooler Discharge Temp 101 IM IN 108 110 115 Fuel Rack 8.0 8.0 new 6.5 new 8.0 new 7.0 new 8.0 old 7.0 old 8.0 old 7.0 old 8.0

1 U.S. Nuclear Regulatory Conumssion  :

3F0997-30 Attachment G .

Page 18 -

TABLE 2 COMPARISON OF KEY ENGINE PARAMETERS TESTING FOR ENGINE POST

• MODIFICATION BG&E BG&E Engine at 'A' Engine at FPC *B' Engine at FPC Engine at Calvert Coltec Cliffs ,

3054kW 3250kW 2716kW 3191kW 2730kW 3145kW '

!!daust Temp for cyl('F) 1 853 820 860 890 840 900 2 879 860 850 890 840 900 3 716 700 710 740 720 750 4 8% 760 860 880 790 830 5 832 800 870 900 820 860 l 6 706 660 730 750 720 760 i

7 888 830 840 880 850 890 8 851 830 850 890 880 920 9 706 680 670 700 680 720 10 861 850 800 830 850 890 11 824 820 750 790 820 860 12 689 680 700 720 720 750 Turbo CS 904 860 900 930 900 940 ,

Turbo OCS 861 840 930 960 880 920

(

U.S. Nuclear Regulatory Commis.sion 3R)997-30 Attachment G Page 19 i TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION ,

Sub-system l Changes l Impact to Staning l Impact to Power Operation l Required Testing Starting None - _

None None -

no modifications; function and No change to air start system operation unchanged and no: involved when engine  ;

is running Lubrication None None Piping reroutes will be morutored for lube

a. System function, operation, a. No impact a. No impact oil flow and vibration, and flow rates are unchanged b. Pipe routing changes have b. Piping reroutes will not affect

b. Piping teroutes to match new negligible impact on flow and lube oil ilcw or EDG l radiators pressure drop operation; wi'.1 be monitored

c. Conservatively increased c.No impact - lube oil is stored in for vibration, minimum storage volume of drums outside of the EDG building c. No impact -lube oil is stored lube oil, not required per in drums outside of the EDG vendor since oil consumption building not increased l Fuel a. Raised setpoints on day tank System effectively unchanged a. liigher day tank level a A b. calibration testing will be performed for pump start & stop and improves NPSII for pump on setpoint changes also low and high level a. liigher minimum day tant level operation d. piping will be monitored for vibration alarm does not affect starting b.Iligher storage tank level and confirm ,

b. Increased minimum amount b. lligher minimum storage tank level improves NPSli for pump "

of fuel to be stored in day does not affect starting operation tanks and main storage tanks c. Fuel flow rate for start unchanged c. Engine driven fuel oil pump t

c. Flow rate into engine d. Rerouted transfer pipe has no flow rate is actually

' l increases for 200 and 20(X) impact on EDG start unchanged since the pump is t hour ratings but no change oversized and the extra fuel i for maximum rating so flow is being recirculated back maximum flow rate does not to the day tank; the injector change pumps are also oversized and inject only the required fuel

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 20 TABLE 3 TMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION Sub-system Changes l Impact to Starting impact to Power Operation l Required Testing i d.Rernuted fuel oil transfer based on the fuel rack (pump

. piping plunger helix) position and the maximum power level has not changed; The fuel oil transfer pumps are capable of about 100% more flow than required based on pump performance tests, the transfer pumps are qualified for continuous operation and will operate for longer durations at intermediate ratings but about the same numter of cycles based on the new positions of the on/off switches in the day tank There has been no increase for the maximum flow rate since the maximum kW rating has not changed (the 1.6%

horsepower / fuel oil flow rate variation accounting for the new ambient minimum design temperature and the use of antifreeze in the radiator is a negligible impact)

U.S. Nuclear Reguictory Commission 3F0997-30 Attachment G Page 21 TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION Sub-system l Changes - l Impact to Staning l Impact to Power Operation l Required Testing Combustion air a. Combustion air intercooler No impact for starting No negative affect on running Confirm improved cooling by intercooler; has been changed out with confirm turbocharger RPM acceptable, more efficient dual pass units No changes to blower which provides No changes to blower confirm combustion air delta P acceptable

b. Nozzle rings in air for staning (including missile barrier influence) turbochargers replaced with i rings which slow down the a. Cooler air from intercooler does a. Cooler air from intercooler rotor not occur until after engine has improves general engine

c. Air flow rate is to increase been running and cooler air would performance by providing slightly for a given power improve general engine cooler, denser air to the level due to more efficient performance combustion chamber combustion air intercooler - b. Turbocharger nozzle ring change b. Turbocharger nozzle ring

d. Pressure drop of supply air does not affect starting since change improves the i to inlet of filter should turbocharger not used for starting turbocharger life by slowing reduce due to missile shield c. Starting air flow rate does not down the rotor while still tnodification and increase change appreciably providing adequate  !

due to higher radiator air d.No impact for the missile barrier combustion air to the engine.

flow rate - total change change because air flow has not The denser air from the should be less than 0.25" been established during starting intercooler reduces the water so influence effectively e. No functional changes pressurization requirement for negligible compared to the turbocharger thus allowing change in barometric it to operate at a lower RPM pressure for a given power level

e. No changes have been made No moving parts have been to the system - function and changed so there is no operation unchanged required break in time No physical changes to the combustion air supply piping and the slight increase in flow

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 22.

TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION Sub-system  ! Changes l Impact to Starting j 1mpact to Power Operation l Required Testin;:

rate of combustion air does t

. not have a significant impact on pressure drop which has been confirmed for power levels up to 3100 kW and will be rechecked during the 3400 kW run i Exhaust None Confirm delta P for exhaust is acceptable f

a. Air flow rate is to increase a. Slight increase in air flow will not a. Slight increase in flow rate of slightly for a given power be in effect during start exhaust gases due to the slight '

level due to more efficient increase in combustion air combustion air intercooler - flow rate does not have a b.No changes have been made significant impact which has to the system - function and been confirmed for power operation unchanged. levels up to 3100 kW and will  !

be rechecked during the 3400 kW run Generator None None None Monitor temperature of field, stator, and No changes have been made to bearing to assure proper function.  ;

the system - function and Extrapolate test results to design  ;

operation unchanged - conditions as necessary.

intermediate power levels increased but maximum rating remains unchanged Coltec evaluated the generator and confirmed it has sufficient capacity.

t x

)

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 23 TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION Sub-system - l Changes l Impact to Starting l Impact to Power Operation Required Testing Excitation None None None Monitor field current to assure remains No changes have been made to below manufacturer's limit of 54 amps the system - function and continuous. Extrapolate test results to operation unchanged - design conditions as necessary.

intermediate power levels increased but maximum rating remains unchanged in addition to the analysis by Coltec, the exciter was monitored during initial functional testing of the engine modification in Feb. and March 1997 and confirmed to operate with sufficient margin.

Voltage None None None ,

regulation No changes have been made to the system - function and operation uxhanged -

intermediate power levels increavd but maximum rating remains unchanged Governor None None a. Ma*.imum single load EDG must a. Perform loss of single largest load test No changes have been made to support as identified in the Tech b. Perform loss of maximum load; confirm the system - function and Specs increased EDG does not trip on overspeed operation unchanged b. Maximum load EDG must support as idenufied in the Tech Spec 4 increased but not ateve previous maximum load rating Auxiliary Lights and switch were added None None electric to the radiator compartment

U.S. Nuclear Reguictory Commission 3F0997-30 Attachment G Page 24 TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION Sub-system l Changes l Impact to Staning l Impact to Power Operation l Required Testing '

Crankcase None None None ventilation No changes have been made to the system - function and operation unchanged Control and a. Improve main control board a. None - no changes to the primary a. None -no changes to the a. Perform calibration checks protection kW meter accuracy by control system, provides more primary control system, b. Perform calibration checks systems removing burden on cts accurate kW reading to the operator provides more accuiate kW

b. Relocated auxiliary gauge b.None - function and operation of reading to the operator panel to match radiator instruments remain the same b.None - function and operation compartment and readouts of instruments remain the for level of the surge tanks same added Cooling air and a.Must run both fans per room No changes to ventilation system for Room cooling is improved a. Verify required flow rates achieved -

ventilation instead of starting both and starting a, b, & d. Changes are to b. & c. Perform system balance system securing one - increasing Always had both fans start when maintain or increase cooling d. Confirm filter dP required air flow. EDG stans of EDG to assure design limits General - Component temperatures will be

b. Adding grills to supply a, b, c, d, & e - changes do not a e mamtained monitored to confinn proper cooling ductwork irapact EDG staning capability c. Adjusting grills is to improve

c. Adjusting grills to local cooling, component redistribute air flow (new temperatures will be balancing) monitored to confirm proper

d. Replacing filters with ones cooling having lower pressure drop e. Always had both fans start c.Two fan runnin*, when EDG starts; runniag one concurrently draws more or two fans has no impact on current than current due to EDG stan or load carrying one fan operating multiplied capability by 2.

U.S. Nuclear Regulatory Commission 3F0997-30 Attachment G Page 25 TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION-Sub-system l Changes l Impact to Staning l Impact to Power Operation l Req sired Testmg Primary a. Pumps and cocling water a. No impact a.No impact Momtor piping for vibration; Cooling pumps and flow rates are b.No impact - all flow bypasses the b. Positive impact - larger monitor air flow rates for proper CFM; unchanged radiator cores during the first few radiator cores provide monitor EDG for prom 1:coling;

b. Radiator cores replaced with minutes of operation; improved cooling monitor fan and fan drive train for larger, more efficient ones c. No impact - clutch not engaged for c. & d. rerouted jacket water vibration;

c. Rerouted jacket water to start and combustion air cooling Monitor fan performance - delta P and flow match new cores d. No impact - clutch not engaged for piping has no impact on EDG rate;

d. Rerouted corroustion air stan which will be confirmed by monitor dP across missile barrier coolant piping to match new c. No impact - drive train not engaged the 3 day test. Will monitor gyrform frequency recovery testing; cores during stan of engine until engine for vibration and proper verify proper operation of surge tank level

e. Radiator far. drive train is up to 450 RPM due to clutch cooling alarms.

replaced but most assembly. The new drive train is c.No impact - see discussion components are the same or the same configuration and uses the under starting vety similar - the following same basic parts, so there is no f. No impact - No influence on is a summation of the significant impact of the drive train starting or operation of EDG differences: changes when compared to the as long as change in heat

1) Engine drive shaft - no present radiator drive train. transfer capability is  ;

change; floating 1)No impact - negligible increase in accounted for. The new -

shaft / couplings - same except inertia, will perform frequency larger cores and the fan air 26.5" longer for relocated respons: testing to confirm flow rates account for this. .

fan due to extended radiator 2)No impact g.No impact - the two compartment. 3)No impact subsystems still operate and

2) Clutch support bearing - no 4)No impact function the same; have the change confirm no impact due to inertia capability for independent control now i

1 U.S. Nuclear Regulatory Commission

' 3F0997-30 Attachment G Page 26 TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION l Sub-system Changes l 1mpact to Starting l Impact to Power Operation Required Testing -

3) Centrifugal clutch - no 5)No impact - negligible increase in h. Positive effect - 7he planned change except potentially to inertia and turns at slower speed so reduced pressure drop through change 3 of the 6 shoes from have offsetting influence, will ' the missile barrier will reduce

aluminum to cast iron so it perform frequency response testing the required horsepower for can transmit more to confirm no impact due to inertia. the radiator fan. This horsepower 6)No negative impact -improved reduction has conservatively 4)Right angle gear drive - reliability not been accounted for in the same snake and effectively 7)No horsepower change for new fan fan analysis discuss'on above.

the same model except for normal operating conditions Testing will confirm the changed the gear ratio to 1:1 from the nominal 180 HP. Due to reduction in pressure drop and from 1:1.25 so fan tip the new minimum design will be accounted for later.

velocity is minimized. temperature limits and also

5) Double universal joint accounting using antifreeze in the assembly - compressed radiator, the fan horsepower can length of the shaft increased increase to a maximum of 264 from 62 to 75.5 inches horsepower for approximately the

6) Fan drive bearmg assembly - first 10 minutes of operation. As the new bearing assembly is the engine increases in much stronger with a larger temperature, the horsepower to diameter &ft and the use of drive the fan will reduce to 234 tapered roller bearings horsepower when the outside air :

instead of ball bearings. temperature remains at 15 F.

s U.S. Nuclear Regulatory Commission 3F0997-30 -

Attachment G -

Page 27 -

. TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION Sub-system - l Changes l Impact to Starting impact to Power Operation l Required Testing -

7)New larger diameter fan is The horsepower requirement reduces

' designed to move mere air at as the outside air temperature the same horsepower increases. He maximum because the new radiator horsepower change from nominal is core has less pressure drop 264 - 180 HP = 8411P. The so the fan can move more air engine is producing approximately at a lower delta P. New fan 5012 HP at 3500 kW (including the is easier to adjust blade pitch normal fan horsepower of 180) so Increased horsepower this variation is 1.7% of the required to drive fan for low standard engine horsepower. The '

temperature operation due to variation from the previous fan revised design point maximum horsepower of 234 is

8) Fan discharge duct modified only 34 horsepower more or 0.7%

to match new size and The maximum horsepower position of fan relative to requirement occurs when the existing hole in roof engine is drawing cold, Jense air 9)New upper compartment for combustion air and cooling so housing for radiator to match the engine capability is increased larger cores and fan for this condition compared to a 95 F day. Therefore, there is no significant affect on required engine horsepower due to the fan change-out or operation in cold weather. Coltec confirmed, in their report VTS-985-970714-0IR, that the co!d weather fan horsepower is compared to normal fan horsepower is insignificant. _

U.S. Nucleer Regulatory Commission 3F0997 Attachment G :

Page .28 TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION Sub-system l Changes l Impact to Starting l Impact to Power Operation Required Testin,$

f. Antifreeze adoed tojacket r :d intercooler water The fan drive train changes has reducing heat exchange negligible impact on the mR2 of the capacity EDO assembly, especially since the

g. Eliminated tie between jacket fan does not engage until the engine water and combust. ion air RPM is greater than 450. The new cooling water and added fan rotates at a 20% lower RPM second surge tank for than the old fan which reduces the combustion air cooling water inertia influence of the fan. Coltec system confirmed, in their report VTS-

h. Missile barrier modified to 985-970714-OlR. that the mR2 ,7 2

reduce pressure drop for the new fan is 2347 lb-ft and the cooling air flow to the mR* of the old fan is 782 lb-ft*.

radiator flowever, the inertia is not significant compared to the engine generator assembly which has a inenia of aR* = 39,000 lb-ft2 ,

This is caly a 4% change.

Frequency recovery testing will be performed as part of the overall testing to confirm there is no inHuence due to this.

i i

w U.S. Nuclear Regulatory Commission -

3F0997-30 Attachment G Page 29 TABLE 3 131 PACT OF 1997 UPRATE AIODIFICATIONS TO EDG RELIABILITY FOR STARTLNG AND OPERATION -

~

Sub-system l Changes - l Impact to Starting l Impact to Power Operation Required Testing 8 & 9) No impact .lmpact covered

- by fan discussion above

f. No impact -No influence on starting or operation of EDG as long as change in heat transfer capability is accounted for. 'The new larger cores and the fan air flow rates account for this.

g.No impact -the two subsystems still operate and function the same; two stirge tanks were installed to :

minimize the tank size required and to simplify piping.

h. Positive effect - The planned reduced pressure drop through the missile barrier will reduce the required horsepower for the radiator fan. This redt.ction has conservatively not been accounted ,

for in the fan analysis discusskna above. Testing will confirm the reduction in pressure drop and will '

be accounted for later.

d

U.S. Nuclear Regulctory Commission 3F0997-30 Attachment G Page 30 TABLE 3 IMPACT OF 1997 UPRATE MODIFICATIONS TO EDG RELIABILITY FOR STARTING AND OPERATION 3

Sub-system l Changes l Impact to Starting l Impact to Power Operation l Required Testing Secondary N/A -- do not use secondary N/A N/A cooling cooling - use radiators Engine a. Horsepower oatput increased a & b - None a & b. New intermediate Monitor firing pressure for 3400 kW test; for increased intermediate ratings were qualified 'oy already done at lower power levels with ratings c.See combustion air discussion Coltec on an EDG of the same acceptable results above type for the full duration of

b. Minor horsepower increase the rating. Maximum kW (1.6%) for fan drive during rating for EDG unchanged.

Iow temperature operation No physical changes to engine other than those discussed in

c. Combustion air system the combustion air system changes discussed above above which improve performance. The higher mass flow rate of combustion ,

air reduces peak firing pressure which reduces engine stresses ar a specific power level. The horse-power has a negligible power increase for the radiator fan associated l with the new minimum ambient air temperature and using antifreeze