ML20085M386

Final Technical Evaluation Rept,Sequoyah Nuclear Plant, Units 1 & 2 Station Blackout Evaluation

ML20085M386
Person / Time
Site:	Sequoyah
Issue date:	10/24/1991
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20085M389	List: ML20085M386
References
CON-FIN-D-1311, CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-91-1249, TAC-M68603, TAC-M68604, NUDOCS 9111080018
Download: ML20085M386 (32)

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_ Attachment 1

. SAIC 91/1249 TECilNICAL EVALUATION REPORT SEQUOYAll NUCLEAR PLANT, UNITS 1 AND 2 STATION BIACKOUT EVALUATION

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S:ience Applications kstomationalCapoentkwt An Employee-Owned Company Final October 24,1991 Prepared fon ,

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Contract NRC.03 87-029 Task Order No. 38 1710 Goh ,a. nye. PO Bon 1303. hiclean. Wyn:o 22102 (703) 821 43 3i

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l TAllLE OF CONTENTS Seetion yggs 1.0 B A C K G R O U N D . . .. . ... . .. .... .. . . . .. ... . . . . . ... . .. . . . . .. .. . .. .. ... ..... .. . . . .. .. . 1 2.0 R E V1 E W P R O C E S S .. . ...... . . .... ... ...... ... . .... .. .. .. ... .... .. .. .. . .. . . . ... . 3 3.0 E V AL U ATI O N . .. . . .. . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . .. . .. . . . . . .. . . . . . 5.. . . . . . . ..

3.1 Proposed Station Blackout Duration ..................... 5

'1 Station Blackout Coping Capability ....................... 9 3.3 Proposed Procedures and Training ........................ 19 3.4 Propose d M od ifications ............................................ 20 3.5 Quality Assurance and Technical Specifications . 21 4.0 CO N C L U S I O N S ....... ... . . . .. . ...... . ... .. .. . . ... . .. ........... ...... . .. ....... .. 22 5.0 R E F E R E N C E S . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 25. . . .. . . . . . . . . . . .

AppendixA.......................................................................................... 27 il

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, TECl!NICAL EVALUATION REPORT  :

SEQUOYAll NUCLEAR PLANT, UNITS 1 AND 2 '

STATION BLACKOUT EVALUATION

1.0 BACKGROUND

On July 21,1988, the Nuclear Regulatory Commission (NRC) amended its regulations in 10 CFR Part 50 by adding a new section,50.63,

• Loss of All Alternating Current Power" (1). The objective of this requirement is to assure that all nuclear power plants are capable of withstanding a station blackout (SBO) and maintaining adequate reactor core cooling and appropriate containment integrity for a required duration. This requirement is based on information developed under the c ammission study of Unresolved Safety Issue A 44,' Station Blackout" (2 6).

The staff issued Regulatory Guide (RG) 1.155, " Station Blackout," to provide guidance for meeting the requirements of 10 CFR 50.63 (7) Concurrent with the development of this regulatory guide, the Nuclear Utility Management and Resource Council (NUMARC) developed a document entitled," Guidelines and Technical Basis for NUMARC Irutlatives Addressing Station Blackout at Light Water Reactors," NUMARC 87 00 (8). This document provides detailed guidelines and procu. ares on how to assess each plant's capabilities to comply with the SBO rule. The NRC staff reviewed the guidelines and analysis methodology in NUMARC 87 00 and concluded that the NUMARC document provides an acceptable guidance for addressing the 10 CFR 50.63 requirements. The application of this method results in selecting a minimum acceptable SBO duration capability from two to sixteen hours depending on the plant's characteristics and vulnerabilities to the risk from station blackout. The plant's characteristics affeHng the re' quired coping capability are:

the redundancy of the onsite emergency AC power sources, the reliability of onsite emergency power sources, the frequency of l' .s of offsite power (LOOP), and the probable time to restore offsite power.

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in order to achieve a consistent systematic response from licensees to the SBO rule and to expedite the staff review process, NUMARC developed two generic response doeurnents. ,

These documents were reviewed and endorsed (9) by the NRC staff for the purposes of plant specific submittals. The documents are titled:

1. " Generic Response to Station Blackout Rule for Plants Using Alternate AC Power,"

and

2. " Generic Response to Station Blackout Rule for Plants Using AC Independent Statfor Blackout Response Power?

A plant specific submittal, using one of the above generic formats, provides only a summary of results of the analysis of the plant's station blackout coping capability.

Licensees are expected to ensure that the bueline assumptions used m Nt. ' MARC 87 00 are applicable to their plants and to verify the accuracy of the stated results. Compliance with the SBO rule requirements is verified by review and evalub .~.1 of the licensee's submittal and audit review of the supporting documents as necessary. Follow up NRC inspectiorts assure that the licensee has implemented the necessary changes as required to meet the SBO rule.

In 1989, a joint NRC/SAIC team headed by an NRC staff member performed audit reviews of tl e methodology and documentation that support the licensees'submittals for several plants. These audits revealed several deficiencies which were not apparent from the review of the !!censecs'submittals using the agreed upon generic response format. These deficiencies raised a generic question regarding the degree of licensees' conformance to the requirements of the SBO rule. To resolve this question, on January 4,1990, NUMARC issued additional guidance as NUMARC 87 00 Supplemental Questions / Answers (10) addressing the NRC's concerns regarding the deficiencies. NUMARC requested that the licensees send their supplemental responses to the NRC addressing these concerns by March 30,1990.

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l 2.0 REVIEW PROCESS The review of the licensee's submittalis focused on the following areas consistent with the positions of RG 1.155:

A. Minimum acceptable SBO duration (Section 3.1),

B. SBO coping capability (Section 3.2),

C. Procedures and training for SBO (Section 3.4),

D. Proposed modifications (Section 3.3), and E. Quality assurance and technical specifications for SBO equipment (Section 35).

For the detennh ation of the proposed min; mum acceptable SBO duration, the following factors in the licensee's submittal are reviewed: a) offsite power design characteristics, b) emergency AC power system configuration, c) determination of the emergency diesel generator (EDG) reliability consistent with NSAC 108 criteria (11), and d) determination of the accepted EDG target reliability. Once these factors are known, Table 3 8 of NUMARC 87 00 or Table 2 of RG 1.155 provides a matrix for determining the required coping duration.

For the SBO coping capability, the licensee's submi:tal is reviewed to assess the availability, adequacy and capability of the plant systems and components needed to achieve and maintain a safe shutdown condition and recover from an SBO of acceptable duration which is determined above. The review process follows the guidelines given in RG 1.155, Section 3.2, to assure:

a. availability of sufficient condensate inventory for decay heat removal, 3 ,

A b. adequacy of the class.1E battery capacity to support safe shutdowm,

c. availability of adequate compressed air for air operated valves necessary for safe shutdown,

d. adequacy of the ventilation systems in the vital and/or dominant areas that include equipment necessary for safe shutdown of the plant,

e. ability to provide appropriate containment integrity, and

f. ability of the plant to maintain adequate reactor coolant system inventory to ensure core cooling for the required coping duration.

The licensee's submittal is reviewed to verify that required procedures (i.e., revised existing and new) for coping with SBO are identified and that appropriate operator training will be provided.

The licensee's submittal for any proposed modifications to emergency AC sources, battery capacity, condensate capacity, compressed air capacity, ventilation systems, ,

containment isolation valves, and primary coolant make up capability is reviewed. Technical specifications and quality assurance set forth by the licensee to ensure high reliability of the equipment, specifically added or assigned to meet the requirements of the SBO rule, are assessed for their adequacy.

This SBO evaluation is based upon the review of the liceJtsee's submittals dated April 18, 1989 (12), and April 5, 1990 (13), a taphone conversation with the licensee on

- December 13, 1990, and the information av.ilable in the plant Updated Final Safety Analysis Report (UFSAR) (14). An audit may be warranted as an additional confirmatory action. This determination would be made and the audit would be scheduled and performed by the NRC staff at some later date.

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. 3.0 EVALUATION During our evaluation, several questions were raised, transmitted to the licensee, and l discussed with the licensee on December 13, 1990. During the telephone conversation .

was determined that an additional written response from the licensee was require i .r>

complete the evaluation. The licensee indicated that it would provide the requested information by early February,1991, By September,1991, we had not received the reque.<ted additionalinformation. The staff made the decision to review the plant coping capability based on the available information contained in the ',obmittals and the plant UFSAR and not to wait for the licensee's response any longer. A list of the questions and the licensee's understanding of the written response required is given in Appendix A. j l

3.1 Proposed Station Blackout Duration Licensee's Submittal The licensee, the Tennessea Valley Authority (TVA), calculated (12 and 13) a minimum acceptable station blackout duration of four hours for the Sequoyah Nuclear Plant (SON) site. The licensee stated (12) that no modifications are required to attain this coping duration.

The plant factors used to estimate the proposed SBO duration are:

1. OITsite Power Design Characteristics The plant AC power design characteristic group is "P1* based on:

a. Independence of the plant offsite power system characteristics of *11/2."

b. Expected frequency of grid related LOOPS of less than one per 20 years, 5

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. c. Estimated frequency of LOOPS due to extremely severe weather (ESW) which  :

places the plant in ESW Group "1," and - -

d. Estimated frequency of LOOPS due to severe weather (SW) which places the plant in SW Group "1."

2. Emergency AC (EAC) Power Configuration Group t

The EAC power configuration of the plant is *D." SON is equipped with four shared emergency diesel generators not credited as alternate AC (AAC) power sources,  ;

three of which are necessary to operate safe shutdown :quipment following a loss  !

of offsite power.

3. Target Emergency Diesel Generator (EDG) Reliability The licensee selected a target EDG reliability of 0.975. The selection of this target reliability is based on having an average EDO reliability of greater than 0.95 for the last 100 demands, consistent with NUMARC 87 00, Section 3.2.4. 3 Review of Licensee's Submittal 9

Factors which affect the estimation of the SBO coping duration are: the independence

- of the offsite power system grouping, the estimated frequency of LOOPS due to ESW  ;

an'd SW conditions, the expected frequency of grid related LOOPS, the classification of EAC, and the selection of EDG target reliability. ,

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The licensee stated that the independence of the plant offsite power system grouping is . .

• I1/2." Upon review of the plant UFSAR:

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L 1.- All offsite power sources are connected to the plant through 'two electrically connected switchyards; ,

L 2. During normal power operation, the essential buses are powered from the unit main generator through the unit station service transformers;

3. Following a loss of power from the main generr. tor, an automatic transfer is

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provideri to the preferred 161 kV offsite power source through common station service transformers (CSSTs) A and C;

4. CSSTs A and C are each capable of supporting 'one shutdown board in each unit; s

5, ,Upon loss of either CSST A or C, there is a manual transfer to the alternate 161 kV offsite source through CSST B; L6. Each CSST has the capacity to power the required loads.

Based on these and the criteria stated in Table 5 of RG 1.155, we conclude that the:

plant independence of offsite power system group is "12."

Using Table 3-3 of NUMARC 87-ud, the expected frequency of LOOPS at SON due to SW condition.is group "2." Using Table 3 2 of NUMARC 87-00,11e expected frequency

- of LOOPS due to ESW conditions place the SON site in ESW group "2." The licensee's-estimates e.re not consistent with those provided in NUMARC 87-00. During the -

telephoot conversation on December 13,1990; the licensee sated that estimates of the -

site SW and ESW groupings were based upon 37 years of site sp cific data. Since the

- offsite power design characteristic will be 'P1" with either the licensee's .or the

'NUMARC estimates for the ESW and SW groupings, we consider the differences it the estimates to be of no concern.

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%e !!censee correctly classified the EAC configuration of SON as "D." At Sequoyah, there are four shared 4400 kW diesel generators, three of which are necessary to safely -

shut down the plant.

The licensee selected an EDG target reliability of 0.975 based on the EDG reliability data for the last 100 demands. Although this is an acceptable criterion for choosing an EDG target reliability, the guidance ir, RG 1.155 requires that the EDG reliability statistics for the last 20 and 50 demands also be ca'culated. Without this information, it is difficult to judge at this time how well ti.e E 20s t .ve performed in the past and if there should be any concern. ~he " ensee r.. ', to have an analysis showing the EDG reliability statistics for the last 20,50, and 100 demar,ds in its SBO submittal supporting documentation. Based on the information in NS/.C 108 (11),which gives the EDG reliability data at U.S. nuclear reactors for calendar years 1983 to 1985, the EDGs at Sequoyah experienced an average reliability of 0.991 per diesel per year.

An EAC classification of"D" requires the licensee to choose an EDG target reliability of 0.975. The licensee has committed to maintain this EDG target reliability in its submittal dated April 5,1990 (13).

Although the licensee is committed to maintain the target EDG reliability, it did not state whether the plant has any formal reliability program consistent with the guidance of RG 1.155, Section 1.2, and NUMARC 87-00, Appendix D. Since the information >

supporting the target EDG reliability is only available onsite for review, an audit may be required to ensure compliance.

With regard to the expected frequency of grid related LOOPS at the site, we can not confirm the stated results. The available information in NUREG/CR 3992 (3), which gives a compendium of information on the loss of offsite power at nuclear power plants in the U.S., indicates that Sequoyah did not have any symptomatic grid related LOOP 8

prior to the calendar year 1984, in the absence of any contradictory information, we agree with the licensee's statement.

Based on the above, the offsite power design characteristic for the Sequoyah site is "P1" with a minimum required SBO coping duration of four hours.

3.2 Station Blackout Coping Capability The plant coping capability with an SBO event for the required duration of four hours is assessed with the following results:

1. Condensate Inventory for Decay Heat Removal Licensee's Submittal The licensee stated (12) that 75,451 gallons of water per unit are required for decay-heat removal during a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO coping pericd. The minimum permissible condensate storage tank (CST) level per technical specifications corresponds to 190,000 gallons of water per unit, which exceeds the required quantity for copng with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SDO event. The licensee stated that no plant modifications or procedure changes are needed to utilize these water sources.

Review of Licensee's Submittal Using the expression provided in NUMARC 87 00, we have estimated that the water required for removing decay heat during the four-hour SBO would be ~77,000 gallons.- This estimate is based on 102% of the licensed core thermal rating of 3411 M W t.

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The licensee did not indicate whether it planned to perform any cooldown of the primary system. The standard Westinghouse procedure ECA 0.0 calls for cooldown with a leak rate of the size postulated for an SBO event (110 gpm). Based on our analyses of other Westinghouse 4 loop 3411 MWt plants, SON will need ~170,000 gallons of condensate to remove decay heat and to cool down to a steam generator pressure of 260 psi. The licensee stated that technical specifications require a minimum of 190,000 gallons of water in each unit's CST be available to the auxiliary feedwater system. Therefore, we agree with the licensee that the site has sufficient condensate to cope with an SBO of 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> duration, even if a 'down is required.

2. Class 1E Battery Capacity t.icensee's Submittal The licensee stated (12) that the class 1E batteries are inadequate to meet SBO loads for four hours. Procedure ECA 0.0, " Loss of All AC Power," will be revised to include battery load stripping. The licensee stated (13) that the battery capacities were evaluated using a minimum temperature of 60 F, which is the minimum temperature expected for any event.

The licensee determined (12) that the non class-1E batteries are adequate to meet the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> coping duration based on an existing calculation that has not been upgraded to the present calculation standards. The SON battery evaluation did not require the diesel generator (DG) batteries to have the capacity to supply power to the diesels for the entire four hour coping duration. The licensee stated (13) that it is implied by the wording that the diesel battery must be capable of supplying sufficient current to flash the DG field at the end of a four hour event. Procedures are presently in place to dispatch Maintenance and Operation personnel to a failed DG for troubleshooting. No further evaluation of the procedure was made for SBO.

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.- The licensee added that no additional evaluations are believed to be necessary in accordance with 10 CFR 50.63, and none are planned. .

Review of Licensee's Submittal According to the plant UFSAR, the clas: lE batteries are designed to carry the f.ecessary loads for two hours. With load stripping, the licensee determined that the batteries have sufficient capacity to last for four hours. We did not receive any information on the licensee's battery capacity calculations, with the exception of the assumed electrolyte temperature (60 F). This temperature is consistent with that recommended by IEEE Std 485.

- De plant UFSAR provides insufficient information to evaluate the DG batteries.

The licensee needs to ensure that it has considered field flashing at the end of the four-hour SBO event when determining the adequacy of the DG battery capacity.

During the telephone conversation on December 13,1990, the licensee stated that loads that are not needed will be shed. These loads include the computers, some instrumentation, and non safety related loads. The licensee did not proside information ~o n the loads to be stripped nor on the assumptions it used in its determination of having 4-hour battery capacity. Without this informa: ion, we are

. unable to verify the adequacy of the be.tteries to carry the SBO loads for four hours, and, therefore, consider this to be an open item.

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3. Compressed Air Licensee's Submittal The licensee stated (12) that the air-operated valves relied upon to cope with an SBO for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> can be operated manually and that those valves requiring manual operation or needing back up sources for operation are identified in plant procedures. .The licensee added that the only air opc sted valves that are questionable for the SBO event are the air-operated, turbine driven auxiliary feed water (AFW) pump level control valves. The licensee stated that the amount of air available to operate these valves will be increased by the addition of compressed air or nitroges bottles. The control of these valves are presently manual, and operators are dispatched immediately when they recognize the need for steam generator level control. The dispatch time from the control room to the valves is 2 to 4 minutes, The licensee 5tated that modifications to add compressed-air back up supplies and the associated procedure changes will provide remote control of these valves for the 4-hour coping duistion requirement. During the telephone conversation on December 13,1990, the licensee stated that the modification to the compressed air system may be canceled. If it is canceled, then the AFW pump level control valves will continue to be manually controlled upon loss of compressed air.

Also during the telephone conversation, the licensee stated that the atmospheric relief valves (ARVs), which are normally air operated, can be manually controlled.

However, the licensee will take credit for the operation of tne steam generator safety valves during the 4-hour SBO event. ,

Review of Licensee's Submittal We did not receive any information on the licensee's final decision on whether or not the modification to add back-up supplies of compressed air will be performed, 12 S

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1 Based on the licensee's most recent staten.ent (during the telephone conversation),

it appears that the modification will not be made.

The AFW level control valves are normally closed valves which fail closed. These valves have accumulators which allow level control for a limited time, and are also equipped with a handwheel. If the modification will not be made, the operators will need to manually operate the AFW pump level control valves after the accumulators are exhausted. Since we do not know where the AFW level control valves are located, the licensee needs to ensure that the area (s) from which the valves will be operated are habitable (i.e., accessible, the temperature is reasonable, sufficient lighting and communications equipment exists). The licensee also needs to proceduralize the actions- necessary to operate these valves and to train the operators for the SBO scenario.

Should cooldown become necessary due to primary system leakage pursuant to the SBO procedure ECA 0.0, the licensee will need to operate the ARVs. During the telephone conversation, the licensee stated that the ARVs can be manually operated.

The licensee needs to verify that the area from which the valves must be operated is habitable (i.e., again the area is accessible, proper lighting and communications equipment need to be available, the temperature is reasonable).

4. EITects of Loss of Ventilation Licensee's Submittal The licensee stated that, to determine if the loss of HVAC during the SBO period will result in dominant areas of concern, it used among other sources a tranaent-method computer code obtained from Martin Marietta to assess the SBO heat loads in areas containing SBO response equipment. The licensee also stated that other analyses and plant-specific experience were used in determining dominant areas of 13

.; concern. During the telephone conversat ion on December 13, 1990, the licensee stated that it used the normal room temperature for its initial room temperature -

c ssumption.

.Sleam Driven AFW Pump Room The licensee stated that the tempuature in the steam driven AFW pump room is bounded by previous design basis analysis and will be below 120 F during the SBO period. Ventilation is provided to the AFW pump room by a vital battery-powered vent fan. The licensee stated that reasonable assurance of the operability of SBO response equipment has been assessed using Appendix F to NUMARC 87 00.

Control Room Complex The licensee stated (13) that the control room habitability was verified by a calculation using a verified computer code, which is documented in Engineering Calculation SON SBO 001. The licensee stated (12) that the control room temperature at Sequoyah does not exceed IWF during an SBO, and all equipment

'in the control room is qualified to operate continuously at this temperature. The i licensee also stated that the control room temperature did not exceed the maximum normal temperature value during the four hout SBO event. The licensee concluded that the control room is not a dominant area of concern and therefore no additional operator response, such as opening cabinet doors, is required.

West Valve Vauji The licensee stated that the west valve vault contains components that are required during a 4-hour SBO event. The licensee added that the temperature within these areas was monitored in 1988 when the room was in a worst-case scenario for heat loads, bounding the SBO loads. During the monitoring period, the non safety-related ventilation system failed and an assessment of equipment operability was based on a 97 F outside ambient air temperature and a recorded room temperature of 163.1*F. During the telephone conversation, the licensee stated that equipment 14

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[ .in the west valve vault is qualified for 185 F for eight hours. The licensee stated that'the assessment documents the operability cf all equipment within the room for -

six months.

Containment ,

Regarding the assumption that the loss of coolant accident /high energy line break  ;

(LOCA/HELB) temperature profiles envelope the SBO temperature profiles, the licensee stated (13) that the LOCA/HELB profile shows a very rapid initial increase while the SBO profile would be much more gradual. Even if the SBO curve exceeded the LOCA/HELB profile in the long term, the maximum temperature inside containment is expected to be from a main steam line break (327'F). De licensee stated that the maximum emironmental stresses on equipment would be from the rapid temperature increases associated with LOCA/HELB. De only equipment located inside containment and required for SBO is two level transmitters per unit. These transmitters are qualified in the environmental qualification binder to 370*F for 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The' licensee stated-(13) that, based on engineering judgment, the four hour ramp up during an SBO is not expected to exceed the equipment 5.5-hour rating. Also, the temperature of the level transmitters will lag the ambient temperature by several degrees during the SBO ramp up. The licensee concluded (13) that, given the above considerations, the. rigor to which this NUMARC assumption is documented is considered sufficient.

Review of Licensee's Submittal I

We' neither received nor reviewed the licensee's detailed heat up calculations.

During the telephone conversation and in a set of follow up questions, the licensee was asked to provide details of these calculations (seeA' ppendix A). We have not-received any detailed information on these calculations yet. l 1

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. The licensee needs to ensure that it has considered all areas which house SEO equipment as areas of concern, including the switchgear room, cable spreading room, -

etc. If any rooms were eliminated from consideration, the licensee needs to have justification for their exclusion in its SBO submittal supporting documentation.

Because we do not have any detailed information on the licensee's calculations, we are unable to concur with the licensee's conclusions in areas other than the containment and AFW pump room, which are discussed below.

Containment We compared the energy released during an SBO event to that released during a HELB and found that considerably less energy is released during an SBO event.

Although the containment coolers will not be available during the SBO, the ice in the ice condensers provide containment cooling. Since the only equipment in containment which is needed during an SBO event is qualified for 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> at a temperature of 370 F and the maximum HEII/i OCA temperature is 327*F, we agree with the licensee that the equipment 5.5-hour rating will not be exceeded during an SBO event.

Steam-Driven AFW Pumo Room The licensee stated that it has performed a calculation for the AFW pump room and

' determined that the temperature will be less than 120 F. With a battery-powered fan available, this temperature is reasonable. Therefore, we concur with the licensee's conclusion.

General Comment'on Initial Temoeratures The licensee stated that it used the normal room temperatures as the assumed initial temperatures in its heat-up calculations. This assumption is non-conservative.

However, if the licensee wishes to use these initial temperatures, then it must place administrative controls which ensure that the room temperatures will not exceed the assumed temperature under any circumstances.

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5. Containment Isolation Licensee's Submittal The licensee stated that the plant list of containment isolation valves (CIVs) has been reviewed to verify that the valves that must be capable of being closed or that must be operated (cycled) under SBO conditions can be positioned with indication independent of the unit's preferred and class 1E AC power supplies.

The licensee also stated that no plant modifications were determined to be required to ensure that appropriate cocarnment integrity can be provided under SBO conditions. The licensee stated that several valves were identified that require manual action to confirm their closure to ensure appropriate containment intearity can be provided under SBO conditions and that a procedure, ECA 0.0, will be revised to list these valves.

Review of Licensee's Suba ittal During the telephone conversation, the licensee stated that the closure of manually controlled valves will be proceduralized. We did not receive any information on whether the licensee used any exclusion criteria in addition to those given in RG 1.155.

Upon review of the plant UFSAR, Table 6.2.41, we found that there is one valve which cannot be excluded by the five criteria given in RG 1.155. This valve which is on the chemical and volume control system (CVCS) line requires manual action if it needs to be closed during an SBO event. The licensee needs to include the manual closure of this valve in an appropriate procedure and ensure that the valve is accessible.

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6. Reactor Coo ant Inventory Licensee's Submittal The licensee stated that the ability to maintain adequate reactor coolant system (RCS) inventory to ensure that the core is cooled has been assessed using a generic analysis listed in Secti on 2.5.2 of NUMARC 87-00 which is applicable to the specific design of SQN. The expected rates of reactor coolant inventory loss under SBO conditions do not result in core uncovery in an SBO of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and therefore make-up systems in addition to those currently available under SBO conditions are n3t required to maintain core cooling under natural circulation. During the telephone conversation with the licensee, it stated that the assumed leak rate was 110 gpm (25 gpm per pump and technical specifications leak rate of 10 gpm). The licensee stated (13) that it has information from Westinghouse that the expected leakage is lower than the postulated 25 gpm, assuming that the secondary sealing O rings maintain their integrity, although the required coping duration can be met aith margin using the 25 gpm leak rate.

Review of Licensee's Submittal The licensee's use of a generic analysis without specific justifications for its applicability to the plant is not acceptable. We performed an independent evaluation of the RCS inventory using the available information in the plant UFSAR. Using a postulated leak rate of 110 gpm (25 gpm per pump seal per NUMARC 87-00 guideline and the Technical Speci5 cations maximum allowable leakage of 10 gpm as the li.:ensee indicated), the total leakage from the RCS during the 4-hout SBO event is 26,400 gallons or ~3500 ft). Upon review of the UFSAR (Section 5.1), we found that the total RCS vc ume to be 11,892 ft', leaving an RCS 3

volum of ~8000 ft without any cooldown. If the primary system is cooled down folloving ECA 0.0, the RCS volume will be ~5000 ft' at the end of the SBO event, 18

which is sufficient to keep the core covered. Therefore we concur with the licensee that sufficient RCS inventory exists to keep the core covered, and natural circulation, ~

through reflux boiling, will keep the core cooled.

NOTE:

The 25 com RCP seal leak rate was agreed to between NUMARC and the NRC staff pending resolution of Generic Issue (GI) 23. If the final resolution of GI-23 defines higher RCP seal leak rates than assumed for the RCS inventory evaluation, the licensee needs to be aware of the potential impact of this resolution on its analyses and actions add:essing conformance to the SBO rule.

3.3 Proposed Procedures and Training Licensee's Submittal The licensee stated that the following plant procedures have been reviewed per guidelines in NUMARC 87-00, Section 4:

1. Station blackout response guicelines,

2. - AC power restoratica, and

3. Severe weather.

The licensee stated that these procedures have been reviewed and the changes necessary to meet NUMARC 87-00 guidelines will be implemented.

,, Review of Licensee's Submittal We neither received nor reviewed the affected SBO procedures. These procedures are plant specific actions concerning the required activities to cope with an SBO. It is the licensee's responsibility to revise and implement these procedures, as needed, to mitigate 19

an SBO event and to assure that these nrocedures are complete and correct, and that the associated training needs are carried out accordingly. -

V 3.4 Proposed Modincations Licensee's Submittal

.The licensee initially stated that a back up supply of compressed air will be added to support the AFW pump level control valves (See Compressed Air Section). Howevu,

during the telephone conversation, the licensee stated that this moCficauen may not be made.

Review of Licensee's Submittal The licensee did not provide justification for the reversalin its decision to add a back up supph d compressed air to support the AF pump level control valves. Presently,

,. operators are instructed to manually control these valves following a loss of AC power.

e i ,

The licensee also stated that instead of using the ARVs, it is taking credit for the steam

- generator safety valves. Pursuant'to ECA 0.0 and the postulated leak rate, cooldown -

' l' will be necessary and the licensee will not be able to cool down with the safety valves.

Therefore, the licensee will need to operate the ARVs. Since both the ARVs and the AFW level control valves will have to be manually controlled under SBO conditions, the licensee needs to verify that the manual controls for these valves are accessible and the -

area in which they are located is haMable during an SBO event. In addition r review has identined several concerns @;. ecti?n 3.2) which may require modifications for i .their resolution.-

20 j

4

. 3.5 Qeality Assurance and Technical Speelfications The licensee did not provide any information on how the plant complies with the requirement of RG 1.155, Appendices A and B.

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4.0 CONCLUSION

S Based on our review of the licensee's submittals and the information available in the LFSAR for Sequoyah Nuclear Plant, Units 1 and 2, we find that the submittal conforms with the requirements of the SBO rule by following the guidance of RG .155 with the following exceptions:

1. Offslie Power Characteristics

a. Severe Weather (SW) Grouping Using Table 3 3 of NUMARC 07 00, the expected frequency of LOOPS due to severe weather pece the site in SW group "2." ' n its submittal, the licensee stated that if site spec 0c data is used, its SW group is "1."

h. .E3tremelv S.tu.re Weather (FSW) Groaoing Us:ng Table 3 2 of NUMARC 01'-00, the excerted frequency of LOOh due to ESW corsditions place the SON site in ESW group "2." In its submittal, the licensee sta'.ed that if site specific data is used, its ESW group is "1."

These differences in the SW an.1 ESW groupin c not affect the offsite power design characteristic for the site, and, therefore, are of no concern. They are, however, different from the poupings given in NUMARC 87-00.

!, Emergency Diese: Generator Reilability Program The licensee's submittal does not document the conformance of the plant's EDG reliability program with the guidance of RG 1.155, Section 1.2 and NUMARC 87 00, Appendix D. The licersee, however, is committed to maintain the target EDG reliability of 0.975.

22

3. Clar 1E Battery Capacity According to the information available in the plant UFSAR, the class 1E batteries hwe sufficient capacity to support the necessary loads for two hours. The licensee determined that, with load stripping, the batteries have suf6cient capacity to last for four hours. The licensee did not provide information on the loads to be stripped nor en the assumptions it used in its determination of having 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> battery cape. city.

Without this infonnation, we are unable to verify the adequacy of the battery capacity, and, therefore, consider this to be an open item. In addition, the licensee needs to ensure that it har, included field flashing at the end of the four-hour SBO event in its calculation of the adequacy of .hn DG batteries.

4. Compressed Air Our review identified that both the ARVs and the AFW flow control valves are air operated and require manual operation during an SBO event. Therefore, the licensee needs to verify that the areas in which these valves are located are habitable (i.e., accessible, sufficient lighting and communications equipment exist, the area temperature is reasonable). The licensee also needs to proceduralize the actions

- necessary to operate these valves and to train the operators for the SBO scenario.

5. Effects of Loss of Ventilation The licensee needs to ensure that it has considered all areas which house SBO equipsnt as areas of concern. The licensee needy to have justification for all rooms which were eliminated from consideration in its SBO submittal supporting documentation. In addition, the licensee assumed the normal room temperature for the initial temperature in its calculation. If the licensee wishes to use these initial temperatures, then it must place administrative controls on the room temperatures 23 1

which ensure that the room temperature does not exceed the assumed initial temperature under any circumstances.

During the telephone conversation and in a set of follow up questions, the licensee was asked to provide details of its heat-up calculations. We have not yet received any detailed information on the calculations. Because we do not have any detailed information on the licensee's calculations, we are unable to concur with the licensee's conclusions in areas other than the containment and the AFW pump room.

6. Containment Isolation Upon review of the plant UFSAR, Table 6.2.41, we found that there is one valve which cannot be excluded by the five criteria given in RG 1.155. This valve which is on the CVCS line requires manual action if it needs to be closed during an SBO event. The licensee needs to include the manual closure of this valve in an appropriate procedure and ensure that the valve is accessible.

7. Proposed Modifications The licensee did not provide justification for the reversal in its decision to add a back-up supply of compressed air for the AFW level control valves. Presently, operators are instructed to manually control these valves. In addition, our review has identified several concerns which may require modifications for their resolution (see Section 3.2).

8. Quality Assurance and Technical Specifications The lice.oee did not provide any information on how the plant complies with the requirement of RG 1.155, Appendices A and B.

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5.0 REFERENCES

1. The Office of Federal Register, " Code of Federal Regulations Title 10 Part 50.63," 10 CFR 50.63, January 1,1989.

2. U.S. Nuclear Regulatory Commission, " Evaluation of Station Blackout Accidents at Nuclear Power Plants - Technical Findings Related to Unresolved Safety Issue A-44,"

NUREG 1032, Baranowsky, P. W., June 1988.

3. U.S. Nuclear Regulatory Commission, " Collection and Evaluation of Complete and Partial Losses of Offsite Power at Nuclear Power Plants," NUREG/CR-3992, February 1985.

4. U.S. Nuclear Regulatory Commission, " Reliability of Emergency AC Power System at Nuclear Power Plants," NUREG/CR 2989, July 1983.

5. U.S. Nuclear Regulatory Commission, " Emergency Diesel Generator Operating Experience, 19811983," NUREG/CR-4347, December 1985.

6. U.S. Nuclear Regula ary Cornmission, " Station Blackout Accident Analyses (Part of NRC Task Action Plan A-44)," NUREG/CR-3226, May 1983.

7. U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research,

" Regulatory Guide 1.155 Station Blackout," August 1988.

8. Nuclear Management and Resources Council, Inc., " Guidelines and Technical Ba:,:s for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," NUMARC 87-00, November 1987. l 1

1 i

l 25 l

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1 l

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9. Thadani, A. C., Letter to W. H. Rasin of NUMARC, " Approval of NUMARC Documents on Station Blackout (TAC-40577)," dated October 7,1988.

i

10. Thadani, A. C., letter to A. Marion of NUMARC," Publicly Noticed Meeting December 27, 1989," dated January 3,1990, (confirming "NUMARC 87 00 Supplemental Questions / Answers," December 27,1989).

11. Nuclear Safety Analysis Center, 'The Reliability of Emergency Diesel Generators at U.S. Nuclear Power Plants," NSAC 108, Wyckoff, H., September 1986.

12. Fox, C.H., Jr., letter to U. S. Nuclear Regulatory Commission Document Control Desk, "TVA's Station Blackout (SBO) Evaluation Results Pursuant to 10 CFR 50.63 for the Browns Ferry and Sequoyah Nuclear Plants," dated April 18,1989.

13. Wallace, E.G., letter to U. S. Regulatory Commission Document Control Desk, "Sequoyah (SON) and Browns Ferry (BFN) Nuclear Plants - Supplemental Submittal on Station Blackout (SBO) Implementation of NUMARC 87 00 Guidance," dated April 5, 1990.

14. Sequoyah Nuclear Plant Updated Final Safety Analysis Report, Amendment 8, April 15, 1991.

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I 4

APPENDIX A The following pages contain questions sent to the licensee regarding its SBO submittals.

This appendix also contains the licensee's understanding of the additional information requested during the telephone conversation which it was supposed to provide by early February,1991, 27

)

I

• /

• *"/ N M I INFORMATION REQUESTED BY NRC

,. STAfr0N BLACKOUT ($50) TELECON DECEMBER 13, 1990 l

,p' A telecon was held on December 13, 1990, between Nic and their contractors and' ivA.

.h This discussion was to provide NRC information regarding WA's SB0 submittal to NRC dated April 18, 1989, and supplemented by letter dated April 5,1990.

The NRC questions dated October 11, 1990 (attached) were supplied to TVA on November 8,1990, by telecopy, and was the scope for the telecon.

TVA provided NRC with the answers and information regarding these questions and additions 1 related areas during the telecon. However, so!ce specific areas could not be sufficiently answered with the WA personnel involved in the discussions and/or without some level of research. Therefore, the following information will be informally supplied to the NRC by early ysbruary 1991, to support approval of TVA's $30 submittal to NRC.

Quantion la No additional information requested.

Question 2: No additional information requested.

Question 3: Provide list of breakers shed during SBO.

Provide by these load sheds.

list of instrumentation disabled Provide battery loading information (lead Profile) during 350 with shedding.

Provide a brief discussion (paragraph) that supports that loads removed are not required for a 830 safe shutdown.

Question 4: Provide a description of how the ninAs rr code proces. works.

Provide initial conditions and assumptions R' CEIV D along with justification used for the MIDAS evaluations. JAN ,4990 Provide information that supports the weet valve vault temperature studies e bounding a 650 event with steam relief. y ON 3 Provide information on the containment /

analysis details for EELBs such that a S50 event would be bounded. Include assumptions (i.e. initial temperature RCS isakage, operating equipment, etc).

Question 5: Provide a justification that WCAP 10541 Revision 1 is valid for 8QM 550. .

RECE1 VEE Provide information on length of natural circulation (S50), volume of RCS and JAN / he, conrparison of percent volume needed for natural circulation and percent volume lost for c. four hour 350. f*]

Additional question on containment isolation valves:

Provide a discussion (paragraph) of the changes to procedures that will cices or verify closed the containment isolation valves during a 550.

December 18, 1990 0979y

.4 l

QUESTIONS-0N SEQUOYAH SB0 SUBMITTAL ,

WITH EMPHASIS ON THE FOLLOWlHG AREAS

!) Justify the discrepancy between the estimated ESW and SW assignments in the submittal and that given in NUMARC 87-00.

2) Explain the "1" grouping.

3) Explain what will be stripped and when it will be stripped to ensure that the batteries will last for the four hour SBO.

4) Loss of HVAC (detailed explanation is needed):

a) Explain the non NUMARC 87-00 methodology (MITAS 11 co( -)used to calculate the temperatures, b) How are the' dominant areas of concern (DAC's) chosen?

c) What are the assumed initial temperatures?

d) Explain the containment heat-up calculations.

e) What is in the west valve vault- and why do the heat sources in this arca during the plant operation bound the 580 condition?

5) RCS make-up: it i s not clear from the submittals whether the technical specification allowable RCS leakage is considered. Explain.

October 11, 1990

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