ML20065C961

Rev 2 to Technical Evaluation Rept Brunswick Steam Electric Plant Units 1 & 2 Station Blackout Evaluation, Final Rept

ML20065C961
Person / Time
Site:	Brunswick
Issue date:	08/24/1990
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20065C965	List: ML20065C961
References
CON-FIN-D-1311, CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-89-1156, SAIC-89-1156-R02, SAIC-89-1156-R2, TAC-68520, TAC-68521, NUDOCS 9009130008
Download: ML20065C961 (36)

Text

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SAIC-89/1156 Revision 2 I

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TECHNICAL EVALUATION REPORT BRUNSWICK $ TEAM ELECTRIC PLANT  :

UNIT Nos. 1 AND 2 l

$TATION BLACK 0UT EVALUATION  : i I i I c I ,

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TAC Nos. 68520 and 68521

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U.S. Nuclear Regulatory Consission Washington, D.C. 20555:

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i TABLE OF CONTENTS Section EASA 1.0- BACKGROUND ........................................... 1 2.0 REVIEW PROCESS ....................................... 3  !

3.0 ' EVALUATION ........................................... 6  ;

e- 3.1 Proposed Station Blackout Duration .............. 6 4 3.2 AlternateAC(AAC)PowerSource................. 9 3.3 Station Blackout Coping Capability .............. 12 3.4 Proposed Procedures and Training ................ 25 5 3.5 Proposed Modifications .......................... 26  ;

3.6 Quality Assurance and Technical Specifications .. 28 <

4.0 CONCLUSION

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5.0 REFERENCES

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TECHNICAL EVALUATION REPORT 1

BRUNSWICK STEAN ELECTRIC PLANT UNIT Nos. 1 AND 2 .

STATION BLACK 0UT EVALUATION

1.0 BACKGROUND

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

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 commission 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 Initiatives Addressing Station Blackout at Light Water Reactors,' NUMARC 87 00 (8). This document provides detailed guidelines and procedures on how to assess each plant's capabilities to comply with the i SB0 rule. The NRC staff reviewed the guidelines and analysis methodology in L 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 SB0 duration capability from two to sixteen hours depending on the plant's characteristics and j vulnerabilities to the risk from station blackout. The plant's characteristics affecting the required coping capability are: the redundancy of the onsite emergency AC power sources, the reliability of onsite emergency power sources the frequency of loss 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 SB0 rule and to expedite the staff review process, NUMARC developed two i

generic response documents. These documents were reviewed and endorsed by the NRC staff (21) 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 Station Blackout Response Power."

A plant-rpecific submittal, using one of the above generic formats, provides only a summary of results of the analysis of the plant's station blackout cop'.ng capability. Licensees are expected to en; ure that the baseline asn mptions used in NUMARC 87 00 are applicable to their plants and to verify th:t accuracy of the stated results. Compliance uith the SB0 rule requirements is verified by review and evaluation of the li ensee's submittal and audit review of the supporting documents as necessary. Follow up NRC inspections assure that the licensee has implemented the necessary changes as required to meet the SB0 rule.

In 1989, a joint NRC/SAIC team headed by an NRC staff member performed audit reviews of the 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 licensee's submittals using the agreed upon generic response format. These deficiencies raised a generic l question regarding the degree of the licensees' conformance to the

! requirements of the SB0 rule. To resolve this question, on January 4, 1990, I

NUMARC issued additional guidance as NUMARC 87 00 Supplemental Questions / Answers (22) 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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i 2.0 REVIEW PROCESS The review of the licensee's submittal is focused on the following areas consistent with the positions of RG 1.155:

t A. Minimum acceptable SB0 duration g$ection 3.1), '

B. SB0 coping capability (Section 3.2),  ;

C. Procedures and training for SB0 (Section 3.4).

D. Proposed modifications (Section 3.3), and E. Quality assurance and technical specifications for SB0 equipment (Section3.5).

P For the determination of the proposed minimum acceptable SB0 duration, the following factors in the licensee's submittal are reviewed: a)offsite .

powerdesigncharacteristics,b)emergencyacpowersystemconfiguration,c) '

determination of the emergency diesel generator (EDG) reliability consistent I with NSAC 108 criteria (9), 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 Regulatory Guide 1.155 provides a matrix for determining the required - ' ',

coping duration.

1 e L For the SB0 coping capability, the licensee's submittal is reviewed to assess the availability, adequacy and capability,of the plant' systems and

l. components needed to achieve and maintain a safe shutdown condition and j ' recover from an $80 of acceptable duration which is determined above. The

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L - review process follows the guidelines given in RG 1.155,' Section 3.2, to L fassure:

a.. availability of sufficient condensate inventory'for decay heat  :

removal, l 3 w . .. - - - = - - -- -

b. adequacy of the class IE battery capacity to support safe shutdown,

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 SB0 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, appropriate containment integrity 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 SB0 rule, are assessed for their adequacy.

The licensee's proposed use of an alternate AC power source is reviewed l to determine whether it meets the criteria and guidelines of Section 3.3.5 of RG 1.155 and Appendix B of NUMARC 87-00.

A normal SB0 review is limited to the review of the licensee submittal; it does not include a concurrent site audit review of the supporting documentation. Such an audit may be warranted as an additional confirmatory action.

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However, a limited number of concurrent site audit reviews were performed in order to obtain a benchmark for licensee conformance with the documentation requirements of the SB0 rule.

Brunswick Steam Electric Plant was ore of the plants selected by the NRC for a concurrent audit review of the SB0 supporting documentation. This audit was performed by a joint NRC/SAIC team, headed by an NRC staff member, on June 26 30, 1989. The following evaluation was written in coordination with NRC staff and encompasses the review of the licensee's submittals dated March 3, 1989 (25), October 10, 1989 (10) and March 30, 1990 (20), and the site audit review.

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. l 3.0 EVALUATION 3.1 Proposed Station Blackout Duration I

Licensee's Submittal The licensee, Carolina Power and light (CP&L) Company, calculated (10, l 20 and 25) a minimum acceptable station blackout duration of four hours for the Brunswick Steam Electric Plant (BSEP) Unit Nos. I and 2. The licensee stated that no modifications are necessary to attain this proposed coping I duration, i I

The plant factors used to estimate the proposed SB0 duration are:

1. Offsite Power Design Characteristics The plant AC power design characteristic group is "P3** based on:

a. Independence of th~ e plant offsite power system characteristics of "I3,"

b. Expected frequency of grid-related LOOP events of less than one per 20 years,

c. Estimated frequency of LOOPS due to extremely severe weather (ESW) of 1.3E-2 per year (10) which places the plant in ESW group "S "

d. Estimated frequency of LOOPS due to severe weather (SW) of 2.5E 3 per year (10) which places the plant in SW group "2,"

and

e. Plant-specific pre hurricane shutdown rer,uirements and procedures that allow the inclusion of an "*" in the offsite AC power design characteristic group are currently in place.

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2. Emergency AC (EAC) Power Configuration Group The EAC power configuration of the plant is "C.* Each of the two units at Brunswick are equipped with two emergency diesel gener-ators which are normally available to the unit safe shutdown equipment. One EAC power supply is sufficient to operate safe shutdown equipment following a LOOP.

3. Target Emergency Diesel Generator (EDG) Reliability The licensee has selected a target EDG reliability of 0.975 based on having a nuclear unit average EDG reliability of greater than 0.95 for the last 100 demands consistent with NUMARC 87-00.

Review of Licensee's Submittal The factors which affect the estimation of the SB0 coping duration are:

the independence of offsite power system, the estimated frequency of LOOPS due to the severe and extremely severe weather conditions, the expected frequency of grid related LOOPS, the classification of EAC, and l the selection of EDG target reliability. The licensee's estimation of l the frequency of LOOPS due to ESW and SW conditions was based on the l data given in NUMARC 87 00. The independence of the offsite power l

system was correctly determined to meet the "I3' classification in i accordance with NUMARC 87 00, Section 3. The EAC classification is also consistent with the guidance.

The EDG operability data for the last 100 demands were included in the licensee's supporting documents, however, the data was not reviewed in detail. A cursory review of this data indicates that both units meet the unit average EDG reliability criterion (7) for the last 100 demands.

A ruiew of the information in the NSAC-108 indicates that the EDGs at l BSEP Unit I and Unit 2 experienced an average of 55 and 50 valid start l demi.nds per calendar year and had an average unit reliability of 0.974 and 0.955 per diesel per year, respectively. Using this data it appears 7

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that the target EDG reliability (0.975) selected by the licensee to be within the acceptable criteria provided in RG 1.155. However, the available information does not support the demonstrated EDG reliability of 0.975. Yet, we believe that this goal is achievable with the establishment of a formalized reliability program consistent with the guidance of RG 1.155, Section 1.2 and NUMARC 87-00 Appendix D.

During the site audit review the licensee stated that the present EDG '

reliability program is sufficient to satisfy the guidelines of RG 1.155, Section 1.2 and NUMARC 87-00, Appendix D. The licensee stated (20),

however, that a reliability program will be established in accordance with the final resolution of Generic issue B 56. This commitment on the part of the licensee appears to meet the requirements of the SB0 rule.

With regard to the expected frequency of grid related LOOPS at the licensee stated that the frequency of grid-related LOOPS is expected to be less than one per 20 years based on the historical evidence at the site. 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 U.S., indicates that BSEP did not have a grid-related LOOP up to 1984. In the absence of any contradicting information, we agree with the licensee's statement.

The licensee stated that a plant-specific pre hurricane shutdown requirements and procedures consistent with the guidance provided in NUMARC 87 00, Section 4.2.3, will be implemented at the site. This guidance requires that the plant to be in a safe shutdown condition two hours before the anticipated hurricane arrival at the site (i.e.,

sustained windspeeds in excess of 73 mph).

Based on the above, we concur with the licensee's classification of the plant offsite power design characteristic of "P3*" with a required coping duration of four hours.

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3.2 AlternateAC(AAC)powersource Licensee's submittal l

' The AAC power source is an existing class IE EDG that supplies power to the non-bla.ked out (NBO) unit. The AAC power will be supplied to the

' blacked out ucit via an existing cross tie between Unit I and Unit 2 4.16 kV emergency buses. Figure 1 (10) shows the configuration for the case in which EDG 1 is the only operating EDG. the licensee also provided (10) a separate figure for each single EAC power source.

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The licensee stated that the AAC power source is available within one hour of the onset of an SB0 event and has sufficient capacity and capability to operate the systems necessary for coping with an SB0 for the required duration of four hours. The licensee stated that the power source meets the criteria specified in Appendix B to NUMARC 87-00 and the assumptions stated in Section 2.3.1 of NUMARC 87 00, i

Review of Licensee's Submittal The licensee's proposed AAC power source configuration conforms to an acceptable configuration provided in Appendix C of NUMARC 87 00:

Configuration 28. The AAC power source, one of the site EDGs, meets all the required criteria in Appendix B of NUMARC 87 00. However, there are two items that require further explanations. They are:

o Paragraph B.9 of Appendix B states, "The AAC power source shall be

... capable of maintaining voltage and frequency within limits consistent with established industry standard that will not degrade the performance of any shutdown system or component. At a multi-unit site, except for 1/2 shared or 2/3 emergency AC power configuration, an adjacent unit's Class 1E power source may be used as an AAC power source for the blacked out unit if it is capable of' powering the required loads at both units."

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! o Paragraph B.12 of appendix B states. * ... the AAC systes shall be demonstrated by the initial test to be capable of powering required shutdown equipment within one hour of station blackout

. event."

a Paragraph B.9 requires that in order for the adjacent unit's EDG to be an AAC power source it should have sufficient capacity to power all the required loads at both units. The guidance on the use of an existing i

EDG as an AAC power source at a multi-unit sites where the EDGs per unit just meet the minimum redundancy requirements is documented in RG 1.155, Section 3.3.5, NUMARC 87 00, Section 2.3.1(3), and under question 3.4 and B.3 in NUMARC 87 00 Supplemental Questions / Answers, and further explained in References 23, 24 and 26. .

The licensee proposes to use the AAC power source to energize the one -

division of battery chargers and to power portions of the 480 V emergency buses to ensure the operability of and position indication of certain AC powered containment isolation vales (CIVs) in the blacked out unit. Therefore, each EDG needs to carry both the normal LOOP loads in the HBO unit and the above selected loads in the blacked out unit.

A review of the plant UFSAR indicates that each EDG has a continuous rating of 3500 kW and a 2,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> rating of 3,850 kW. The UFSAR also gives a list of loads which are required for a normal cold shutdown of .

one unit. We reviewed the list against the NRC's guidance regarding the use of existing EDGs as AAC power sources, and found that each EDG will have at least 470 kW excess capacity, when compared to its 2,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> rating, that can used to support the selected loads in the blacked out l unit. The estimated $80 loads is less than 400 kW. Therefore we agree with the licensee that each EDG has sufficient capacity to power the l

required in both units.

l To comply with the paragraph B.12 the licensee needs to perform a single l test demonstrating that actions required for powering necessary shutdown 11

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equipment can be accomplished within one hour from the onset of an $80 event.

3.3 Station Blackout Copin9 Capability The plant coping capability with a station blackout for the required i

duration of four hours is assessed based on the following results:

1. Condensate Inventory for Decay Heat Removal Licenses submittal The licensee's analysis (Reference 18 modified by Reference 20) shows that 97,390 gallons of water are required for decay heat removal and reactor cooldown for the proposed $80 duration of four hours. The licenste used the expression given in Section 7.2.1 of NUMARC 87 00 to calculate of the required condensate volume for removing the decay heat from the reactor coolant system. The additional condensate volume required for cooldown and depressurization, and the associated shrinkage was also evaluated.

l The licensee used the total allowed leakage rate from Technical Specification 3.4.3.2 of 25 gpm and used the NUMARC 67-00

. recommended recirculation pump leakage of 18 gpm. The total required condensate inventory was calculated as follows: i Water Recuired for Volume f aa11ons) '

Decay heat removal 53,900 Depressurization 17,840 Shrinkage 11,960 -

RCS leakage 6,000 .- !

Recirculation pump seal leakage 7,690 1

97,390 12

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, The licensee stated (18) that the minimum available water in each unit's condensate storage tank (CST) per existing plant procedure 01-3.6 (17) is 103,738 gallons. Additionally, the licensee stated l (20) that the BSEP Technical Specifications require that a minimum t )

condensate storage tank inventory of 103.738 gallons to be reserved for HPCI and/or RCIC operations. This volume of water exceeds the quantity required to cope with an SB0 with a four hour duration. The licensee stated that no plant modifications or procedures changes are needed to use this water source.

Review of Licensee's Submittal Plant procedure 01 3.6 is a daily surveillance procedure for the radioactive waste related areas that includes a log sheet stating that the minimum CST level is 10 feet. We agree with the licensee's calculations indicating that 100,380 gallons of condensate are required to cope with an SB0 with a duration of four hours and that this 10 foot level ensures that 103,738 gallons of water are availabh for the RCIC or HPCI pump suction (18and20). However, if tank level is found to be less than 10 l feet, the operator is directed by 01-3.6 to submit a discrepancy form to the Radwaste Operations Specialist and inform Nuclear Unit Control Operator. We believe that this discrepancy form does not i

ensure prompt action t;. correct the deficient condensate inventory.

i In its supplemental submittal (20), the licensee stated that the CST level is covered by a technical specification. However, the licensee needs to verify that the CST technical specification is in effect during reactor operation or revise existing technical specifications to be consistent with the commitment to maintain adequate condensate inventory. Based on this commitment, we agree with the licensee that adequate condensate inventory is maintained ,

during plant operation for coping with and recovering from a 4-hour SB0 event.

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2. Class IE Battery capacity 1 i Licensee's submittal The licensee determined that class IE batteries at Brunswick were inadequate to meet station blackout loads for four hours. Modif.

4 ications and/or procedure changes are necessary to provide a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> capacity. The licensee plans to perform modifications and procedural revisions to permit the blacked out unit's battery charger to be powered from the non blacked out unit's EAC power source (which is the AAC source). The modification is described in some detail in the licensee's submittal (10) and Section 3.5 of 4

this report, i

Review of Licensee's Submittal The battery is required to cope with station blackout for one hour or until the cross connection to the unaffected unit is completed and the battery chargers are energized. RG 1.155 and NUMARC 87 00 require that the battery loads be analyzed for the ability to cope for one hour in this situation. With no operator action, the battery which is supplying the uninterruptible power supply (UPS)

, has the capacity to last 70 minutes. By transferring the UPS to a second battery bank (e.g., switching from battery A to battery B),

the batteries Mill last over two hours. Since a 70 minute battery life is clos 4 to the actual time to connect AAC power, addressing the timely transfer of the UPS from one battery bank to another in station blackout procedures would enhance the plant's ability to i cope with the S80 event.

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Based on the licensee's calculation showing that each battery will last for greater than one hour and that the AAC source will be available within one hour to power one division of battery chargers, Brunswick conforms with the guidance provided in NUMARC 87 00, Supplemental Questions / Answers (22) regarding the 14 I

j availability of the normal battery backed plant monitoring and L

electrical system controls in the control room.

3. Compressed Air Licensee's Submittal The licensee has indicated that the air operated valves needed to 1

cope with a station blackout for a 4-hour SB0 coping duration have sufficient backup sources independent of the preferred and blacked

• out unit's class IE power supply. '

Review of Licensee's Submittal g The only air operated valves required to. cope with station black-i out are the scram valves and the automatic depressurization system (ADS) valves. The scram valves function normally without AC

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power. The licensee estimates that they have adequate back up supplies of air and nitrogen for several hundred cycles of ADS valves, which exceeds the amount necessary for the plant depressurization and cooldown planned for station blackout events. ,

We agree with the licensee that adequate compressed gas is  ;

available to cope with and SB0 and that no hardware or procedural modifications related to compressed air appear to be necessary.  !

'.. Effects of Loss of Ventilation '

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Licensee's submittal The calculated post-$50 steady state ambient air temperature for the plant areas containing SB0 equipment are.as:follows (10):

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Area Temperature ('F)

Eint], Initial High Pressure Coolant Injection 152 104 (HPCI) room Reactor Core Isolation Cooling 145 104 (RCIC) room Emergency Core Cooling System 131 104 (ECCS) Pipe Tunnel The assumption in NUMARC 87 00, Section 2.7.1 that the control room will not exceed 120*F during a station blackout was assessed.

The licensee concluded that temperature in the control room at BSEP, Units 1 and 2 will not exceed 120*F during a station blackout and therefore is not a dominant area of concern.

The licensee stated that during an S80 event control room heating, i ventilation and air conditioning (HVAC) will be available '

immediately if the AAC source is eitDer EDG 1, 3 or'4. Only in l the instance that EDG 2 is the AAC praer source will control room ventilation be unavailable until the electrical power cross connection is completed. Thelicenseeprovidedanalysis(18)  ;

indicating that' the control room will not exceed ll8'F ambient air temperature at the.end of one hour. The licensee stated that control room cabinets will be opened if operators are having L ,

difficulty establishing the cross connection necessary to ensure L

the restoration of ventilation within I hour, 1

i i The licensee stated that the EDG rooms are ventilated when the EDGs are running and are not ventilated if'the EDG is failed.

Switchgear rooms in the EDG building are ventilated from their related EDG, however, the room which houses the AAC cross connect l l

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,n switchgear is without ventilation during an $80 event. The licensee stated that all switchgear rooms will remain below 120'F.

The licensee stated that battery room ventilation remains available.

The licensee evaluated equipment operability in the HPCI and RCIC roo:., and concluded that all equipment was qualified to operate at the calculated steady state ambient air temperatures.

Review of Licensee's submittal The licensee's SB0 coping analysis (10) and the ambient temperature calculations were audited during the plant visit. The following summarizes the major areas which either need explanations or have concerns with:

o Control Poom The licensee initially compared the Brunswick control room to the control rooms at Robinson and Shearon Harris and concluded that the temperature will not exceed 120'F. The review team did not consider this to be adequate and requested additional information and analysis. In response, i the licensee performed an analysis of control room heat up during the first hour after the onset of an SB0 event. The licensee stated that the control room would remain below Il8'F, based on a conservative analysis, using heat loads from the ventilation design. Our review of this analysis revealed several errors. First, the licensee assumed the initial sheet metal temperature of all instrumentation cabinets to be 80'F. This is not conservative because I

during the site audit review many of the cabinets felt significantly warmer than the room ambient temperature.

Certainly the metal panels between cabinets, which do not 17 i

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'- l directly transmit with the 78'F ambient temperature, are initially above 80'F. '

4 Secondly, the licensee's heat transfer model has the final air temperature established to be 120'F and the final cabinet sheet metal temperature estimated to be 102'F.

Basic heat transfer laws require that the heat must travel from the hot electronics inside the cabinets to the cabinets themselves. The heat transfer from the electronics to the surrounding air occurs, but is secondary to the heat transfer to the cabinets. The cabinets will heat up and give off heat to the air. Therefore, the cabinets should be warmer than the air.

The licensee correctly modeled the fact that the air also receives heat from the lights and personnel in the room.

However, the licensee incorrectly used the NUMARC 87 00 room l heat up model. It does not appear that the NUMARC 87 00 model, which is an approximate steady state ambient model for air and wall temperature, can be used in conjunction with the time-dependent heat transfer model of the cabinets used by the licensee.

Our independent calculatioa did not validate the revised approach taken by the licensec, but confirmed that the control room ambient air temperature will be less than ll8'F after one hour of station blackout corG. ions. We used the COMPARE code to calculate the ambient air conditions. The ,

problem was set up such that the cabinets were heat sources with the appropriate heat capacity, and the people, process computer and lighting were heat sources with essentially no heat capacity. The heat sinks used were the walls, ceiling and air. The heat loads, sheet metal volume.and room areas ,

calculated by the li:ensee were used. The initial cabinet I metal temperature was conservatively estimated to be 98'F, ,

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and the other thermodynamic properties were based on similar

) calculations with standard references.

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i Therefore, as long as HVAC is restored within one hour of the onset of station blackout, the control room would not be a dominant area of concern. However, it is not clear from the licensee's submittal (10) whether HVAC or just ventilation is restored in this limiting case. If only ventilation (forced air, no cooling) is restored, additional l analysis addressing room temperature and equ4, pent i

operability in an ambient environment without cooling is required. Since no ventilation is available during the first hour of an $80, the licensee needs to provide an appropriate procedure which directs the operators to open the control room cabinet doprs within 30 minutes of the onset of SB0 to provide adequate air mixing to maintain internal cabinet temperatures in equilibrium with the control room temperature per guidance provided in NUMARC 87-00 Supplemental Questions / Answers, o HPCI Room The temperature calculations for the HPCI room assumed an initial temperature of 104'F. During the plant tour of June 30, 1989 the room temperature was 100'F. The licensee indicated that without manual operation of the room coolers, the room temperature would exceed 104'F. Currently a mer.o to the operations staff directs plant operators to use the-room cooling to maintain temperature below 104*F. As a minimum, the temperature is logged once per shift. A concern in this area is that the room could heat up for a period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> between operator visits. Plant personnel stated that this has not been allowed to occur during normal plant operation. However, a memo does not provide adequate assurance that the room temperature will be below 104'F at 19

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all times. Instructions for the control of the HPCI room 3 temperature should be formalized into a plant procedure.

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• Switchgear Room in the blacked-out unit, EDG building switchgear spaces apparently are not ventilated by equipment powered from the operating EDG during an SBO. The licensee calculated the ambient temperature rise during the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> $80 for each of four switchgear rooms. The rooms are not ventilated by their normal sources, but the licensee claims that they will

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receive a small amount of air flow from the pressurized ventilation header. (The ventilation for the running EDG l and it's associated switchgear room is operating.) The  ;

ventilation line up was not verified, but the licensee's approach and analysis seems to be acceptable. The warmest switchgear room cculd reach ll8'F, and all switch gear rooms could uceed 104'F during the event. Therefore, it is necessary to provide operators with procedural guidance to monitor room temperature and to open cabinet doors in the switchgear rooms within 30 minutes consistent with guidance in NUMARC 87 00 Supplemental Questions / Answers (22),

o Main Steam Tunnel This small room has a significant heat source (steam pipes) and no ventilation is available. However, the room was determined not to be an area of concern because the only l equipment in the room is the HPCI and RCIC steam line break detection instrumentation. This instrumentation is designed l-l to detect a steam line break represented by a high room temperature. Upon indication of a steam line break, the instrumentation provides a signal for the isolation of the HPCI or RCIC steam lines. Since no other accidents are assumed to occur during an SBO, this equipment 11 not 20 L

1

l i

required to cope with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO. Since failure of this i

f equipment could render HPCI or RCIC inoperable, the licensee stated that station blackout procedures will call for the '

de energization of this signal. A key operated switch, i

located in the control room is available for this purpose.

Proper implementation of these procedures should ensure that the steam line break instruments do not activate during an SBO. -

e Containment The containment and the drywell will heat up significantly during a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO, but the licensee stated (10) that the temperature profile is bounded by that of a large LOCA.

However, it is not clear whether the underlying assumptions used for the largo LOCA analysis are consistent with an $80 scenario. It is necessary for the licensee to explicitly verify the assumed ccnditions during LOCA, especially timing of the event, and operation of sprays, coolers and fans, to the SB0 scenario.

The major equipment operability concern associated with containment temperature rise is the operability of a reactor water level indicating systems. The manufacturer's data .

(15) for the reactor water level instrument, which is normally used to control reactor feed flow, indicates that

~

the reference legs required to provide reactor water level information will function properly at the elevated containment temperatures determined by the LOCA analysis.

The instrument is in the reactor building, where it is not exposed to high temperatures.

21.

5. Containment Isolation Licensee's Submittal The licensee reviewed the plant list of containment isolation valves to verify that valves which must be capable of being closed orbeoperated(cycled)understationblackoutconditionscanbe positioned (with indication) independent of the preferred an~d blacked out unit's class 1E power supplier The licensee has stated that Abnormal Operating Procedure, A0P-36.0, will be revised to sequentially supply power from the non-blacked out unit's EAC power source to motor control centers (MCCs) required to monitor position and/or close a group of identified valves.

Review of Licensee's Submittal The licensee identified 6 core spray and 12 residual heat removal motor operated valves that are essentially inaccessible and could be in the open position at the onset of the SB0 event. The four 480 volt MCCs that power these valves will be energized from the AAC source so operators can monitor the position and/or close this group of valves. This approach will ensure that appropriate containment isolation can be achieved and maintained in accordance with the guidance of RG 1.155, Section 3.2.7.

6. Reactor Coolant Inventory Licensee's Submittal The licensee stated that the expected rates of reactor coolant inventory loss under SB0 conditions do not result in more than a momentary core uncovery in an SB0 of four hours. Therefore, make-up systems in addition to those currently available under SB0 conditions are not required to maintain core cooling under natural circulation (including reflux boiling). Plant specific analyses 22-I

4 (11,12,19) were used to assess the plant's ability to maintain adequate reactor coolet system inventory without suppression pool boiling. The analysis has shown that the expected rates of the reactor coolant inventory loss under SB0 conditions do not result in core uncovery or suppression pool boiling in the four hours of SB0 duration. Additionally, the licensee provided analysis verifying that the suppression pool has sufficient volume for condensate addition when coping with a station blackout event. The licensee assumed that the only reactor coolant inventory losses are the ADS discharge to the suppression pool and the 25 gpm allowed by Technical Specification 3.4.3.2.

Review of Licensee's Submittal The licensee stated that reactor is to be cooled down and depressurized to 150 psig by releasing steam to the suppression pool. The HPCI and RCIC systems are available to inject water into the reactor throughout the event. They are steam driven systems that depend upon DC power for actuation and control. The condensate water usage calculation took into account both the water released via SRV and/or ADS, and the water required for the cooldown. The suppression pool will be heated by the steam released from the reactor via the ADS valves and through the HPC1 and/or RCIC exhaust. The licensee calculated a final suppression rool temperature of 188'F. However,.'our review does not agree with the licensee's results for the following reasons:

1)- The licensee used NUREG/CR 4041 (13) for the decay heat cal-culations. We used ANS 5.1/N18, which was used in NUMARC 87-00 and the BTP ASB 9-2 (14), and is incorporated into the H NRC's Standard Review Plan. The licensee used 100 days at 100% power to calculate the decay heat load. Brunswick sould (and may have) exceed this on their operating cycle.

Our calculation resulted in somewhat more decay heat than 23

___._:---__--_L______.-_____-__----__________

] ,

that calculated by the licensee. This resulted in a non conservative error'of about 7'F.

The core average effective full power days (EFPD) should be calculated, or a decay heat value based on infinite operation be used. EFPD can be calculated from: ,

EFPD = B

• MTU / Q  ;

where:

• l B , Average end of cycle fuel burnup (MW days /MTU) t l

MTU -- Metric Tons of Uranium <

Q =

P1 ant MW-(thermal) j t

2)~ The licensee. credited excessive natural convection heat

transfer from the torus wall to the drywell. Some credit is {

. appropriate, but the calculational method may not be L

conservative. The maximum impact that this can have on the

. final suppression pool temperature is about 3*F.

.  ; i 3)- An apparent small arithmetic error in the licensee's calculstion of total suppression pool-water inventory resulted in a conservative difference of about 40

) Our-calculations indicate thst th'e: suppression ~ pool- i temperature will exceed the licensee calculated,188'F by e about:10'F. However, this temperature willl not exceed' the i structural design _ temperature of 220*F. lit 4should be noted- I

~

thatthefinal1temperaturecalculated'bythereviewersL(195' -

j

. )'

J200*Fidepending onttheitorus' heat' transfer assumptions);is 1 approaching the technicaU specifihation limit. We-are concerne'd that"if _the operators;should destatet slightly fron-J the planned coping _' approach, the suppression pool! 1 itemp'erature will exceed;200*F. ;This could jeopardize the -

..operabili,tyofsystemsneededtocopewithSB0l(e.g.,fRCIC)

, , 2, 24  ;

q V p 's ,

m t

e. .

3  ;- .

< 9t j :b_ _ i _ MEL ' ' ' ' ' ' -

~ ~ ~

i l

and to recover from the SB0 after power is restored (e.g.,

RHR). One likely cause of deviation is the reactor pressure gauge used to monitor the depressurization. The error in l this gauge could allow operator to add significantly more I heat to the suppression pool during the controlled cooldown.

Other gauge errors and the possibility of the operator I continuing the cooldown past 150 psig goal could contrioute to the problem. Although no action is required, it should be noted that the suppression pool heat capacity is marginally acceptable.

i ,

The licensee correctly calculated (19) that the suppression pool level will rise 20.5 inches (from the highest allowed level) to a distance of.7.5 inches below the torus center line. However, they need to verify that this level will not affect the operation of l: the HPCI or RCIC systems. The concern is that during the

)- intermittent operation of HPCI or RCIC, suppression pool water L

could backflow into and flood the HPCI and RCIC turbines when the turbine is not running. This would prevent further operation of p the turbine.

L 3.4- - Proposed Proceoures and Training 1.

t E Licensee's submittal L

- The licensee stated that the following. procedures have been reviewed and modified..where necessary, to meet guidelines in NHARC 87-00:  ;

l ,

H 1. 'AC power restoration procedures, 2.. Severe weather procedures, and

3. Station blackout response procedures.

L The licensee indicated that the loss of'offsite power emergency j procedure and the associated abnormal operating and operation procedures are being revised to level.that exceeds the procedural guidelines of-L 25-k e

i 9-

~

NUKARC 87-00. The plant hurricane procedure has been reviewed and determined to be consistent with the guidelines of NUMARC 87-00.

i The licensee stated that existing plant policies and guidelines for training will be used upon issuance of new and revised procedures. The licensee also stated that these policies will ensare that operators are adequately prepared to cope with station blackout events.

Review of Licensee's Submittal i

We did not examine the affected procedures or training. With the 1 exception of the hurricane procedere, none of the SB0 procedures have been implemented. These procedures are plant specific actions concerning the required activities to cope with an SBO.

1 The licensee appears to have identified the procedures that need to be modified and/or implemented to cope with SBO. It is the licensee's responsibility to revise and implement these procedures, as needed, to mitigate an SB0 event and to assure that these procedures are complete and correct in their contents and the associated training neede are carried out accordingly.

3.5 Proposed Modifications Licensee's submittal The' licensee indicated that modifications to the inter-tie control power circuitry are necessary to ensure the prompt availability of the AAC source. The modification will isolate the LOCA interlocks and permissives that might inhibit operation of the cross connection breakers during station blackout conditions. During normal operation, control power to the brekkers will be de energized by installing a local key-locked switch positioned to remove control power ~. This will permit the breakers to be maintained in the racked in position during normal operation with no potential for spurious operation due to an electrical 26

i l .

fault control room fire. Inadvertent operation of the breakers will be prevented by controlling access to the keys for the key-locked switch.

The cubicles will be conspicuously labeled such that no switches are to be operated to energize breaker control circuits except during an emergency sach as a station blackout. During su:h an emergency, the key-locked switch will be repositioned to energize the control power circuit to allow operation of the breakers.

The licensee stated that the equipment required for a SB0 will be affected by the modifications are the emergency switchgear buses for both units: El and E2 in Unit 1, E3 and E4 in Unit 2.

As part of the modification, the licensee stated that unit specific performance testing will be accomplished to ensure proper operations of each respective unit's breakers subsequent to the modifications.

The modifications and at: xiated procedure changes will be completed within 2 years after the notification provided by the Director, Office of Nuclear Reactor Regulation in accordance with 10 CFR 50.63(c)(3).

In addition to the modifications associated with establishing an AAC

r. power source, the licensee has committed (10) to provide an AC-independent containment pressure indicator.

Review of Licensee's Submittal The modifications are currently at the planning and design stage. If i properly implemented, the modifications could meet 10 CFR 50.63 without violating any General Design Criterion.

I l

27 L

l

4

.s 3.6 Quality Assurance And Technical Specifications Ouality Assurance The licensee provided a list of equipment required to cope with a station blackout. This list is based on preliminary design of modifications and expected versions of procedures. According to the licensee the only equipment that is not covered by the Nuclear Safety Class (NSC) QA systems is a DC powered reactor water level indicator which measures the vessel level between 150" to 210". The licensee has stated (18) that this component, or an equivalent, will be placed under a QA program meeting guidance of the RG 1.155, Appendices A and B.

Th- ist of equipment required to cope with SB0 was reviewed. All equipment, except for the above mentioned reactor vessel water level indicator, that is being used to cope with 580 is already incorporated into an appropriate QA program.

l^

lechnical Soecifications l The review of the proposed modifications and equipment needed to cope with station blackout indicates that most equipment used to cope with a station blackout (EDGs, HPCI, RCIC, batteries, chargers) is already -

covered by technical specification. The exception is that the licensee needs to verify the commitment that the CST level is included in their technical specifications during operation in Modes I and.2.

-28.

a

4.0 CONCLUSION

S Based on our review of the licensee's submittal and the related supporting documents and discussions during site audit review for the Brunswick Steam Electric Plant Units 1 and 2, we find that the submittal conforms with the requirements of the SB0 rule and the guidance of RG 1.155 with the following exceptions:

1. Alternate AC power source NUMARC 87-00, Appendix B provides guidance on demonstrating that the buses used in the cross connection bt' ween units are capable of powering the required equipment within one hour.

A test should be performed in accordance with the requirements of paragraph B.12 of NUMARC 87-00, Appendix B. ,

2.

Effects of Loss of Ventilation

a. {DG Buildina Switchaear Rooms I

The licensee calculated the final EDG switchgear room (4 rooms) 1 temperatures to be between 104*F and 119'F. Therefore, it is l

necessary to either: (1) provide operators with procedural guidance to monitor room temperature and to open cabinet doors in the switchgear rooms within 30 minutes (NUMARC 87-00, Supplemental Questions / Answers), or (2) assess the equipment in the switchgear rooms to determine if it is designed to operate in the ambient air temperatures expected during an SBO.

Procedural controls should include requirements for operators to monitor room temperatures and open cabinet doors.

L 29

.t.

b. Containment /Drywell The licensee stated that several containment temperature analyses, with their results adjusted for Brunswick containment size and heat load proved that the Brunswick containment will not exceed the equipment's qualification. The licensee needs to verify that the assumptions in the LOCA analysis are consistent with that of ,

the SB0 scenario, and the SB0 equipment remains operable for the duration of time that it must function (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) during an SBO.

c. Control Room t

The licensee provided a calculation for the ambient air i

L temperature rise in the control room during the first hour after the onset of an SB0 event. The results of this calculation indicate that the control room air temperature will remain below ll8'F using a conservative heat load from the control room HVAC design. We found the analysis to have several problems, however, our independent analysis confirmed that the control room temperature remains below ll8'F after one hour into an SB0 event.

Since no ventilation and/or air conditioning will be available during the first hour of an SB0 event, the licensee needs to establish a procedural guidance which directs the operators to open the control room cabinet doors within 30 minutes of the onset of an SB0 per guidance provided in NUMARC 87-00 Supplemental Quntions/ Answers.

3. Reactor Coolant Inventory

a. Suppression Pool' Heat up i

The licensee made several errors in the calculation of suppression pool temperature.

These errors (as described in Section 3.3) underestimate the final suppression pool temperature by about l l

10'F. The calculated final suppression pool temperature is within 30 1

l v 1 l

.. ~

,/

5*F of the 200*F operating limit for RCIC.

(This limit also applies to the RHR system which is used during recovery.) There is a ccncern that depressurizing the plant slightly below the planned 150 psig (because of existing uncertainty in pressure instruments), or a delay in restoration of AC power, could result in a suppression pool temperature exceeding the 200'F limit and rendering decay heat removal system inoperable.

b. Suooression Pool level The licensee correctly calculated (19) that the suppression pool level will rise 20.5 inches (from the highest allowed level) to a dt.:tance of 7.5 inches below the torus center line. However, they need to verify that this level will not affect the operation of the HPCI or RCIC systems. The concern is that during the intermittent operation of HPCI or RCIC, suppression pool water could nckflow into and flood the HPCI and RCIC turbines when the turbine is not running. Tiiis could render decay heat removal system inoperable.

4. Quality Assurance and Technical Specifications The li ensee has committed to provide an appropriate QA program for the DC_ powered reactor water level indictor, which measures the vessel level 150" and 210", consistent with the guidance of RG 1.155, Appendix A.

i-L The licensee stated (20) that the BSEP Technical Specifications condensate storage tank minimum inventory is 103,738 gallons,

' which is adequate to cope with an SB0 with a duration of four- {

1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. However, the licensee needs to verify that the CST level I

is included in their technical specifications during operation in l Modes 1 and 2.-

1 1

31 L

1

.g: *

}

l

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, a

2. U.S. Nuclear Regulatory Comission, " 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 Comission, " Collection and Evaluation of Complete and Partial losses of Offsite Power at Nuclear Power Plants," t NUREG/CR-3992, February 1985.

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

$.. U.S. Nuclear _ Regulatory Comission, " Emergency Diesel Generator.

Operating Experience,- 1981-1983," NUREG/CR-4347, December 1985. t 6.- U.S. Nuclear Regulatory Comission, " Station Blackout Accident ' Analyses '

(Part-of NRC. Task Action Plan A-44)," NUREG/CR-3226, May 1983.

i

7. -U.S. Nuclear Regulatory Comission' Office of Nuclear Regulatory Research, " Regulatory Guide 1.155 Station Blackout," ? August 1988.

p, 8..  : Nuclear Management and Resources Council',:Inc., " Guidelines a, '

L . Technical Bases for NUMARC Initiatives Addressing; Station Blackout at i

y Light Water Reactors,"- NUMARC 87-00,: November 1987. ~!

'F

9. Huclear Safety Analysis Center, "The Reliability of- Emergency Diesel: 't GeneratgrsLat' U.S. Nuclear Power Plants," NSAC-108, Wyckoff, H.,'

September 1986.' '

r l: .c. '32 '

.J, y,

' 7,6 y ,

'{.

• ' .~

v

10. Cutter, A. B., Letter to the Document Control Center of the U.S.

Nuclear Regulatory Comission, " Brunswick Steam Electric Plant, Unit l Nos. I and 2 Docket Nos. 50 325 & 50 324/ License Nos DPR 71 & DPR-62

} Revised Response to Station Blackout r a le," (NRC TAC Nos. 68520 and 68521) Carolina Power and Light Serial; NLS 89-256, dated October 10, l 1989.

L 11. Carolina Power and Light, " Condensate Inventory for Decay Heat Removal Calculation," BSEP calculation 8542 M-01, Rev. 1, October 1989.

l

12. Carolina Power and Light, " Suppression Pool Temperature Calculation,"

BSEP calculation 8542-M 02, December 1988.

l

13. U.S. Nuclear Regulatory Comission, " System Analysis Handbook,"

NUREG/CR-4041, November 1985.

i

14. U.S. Nuclear Regulatory Comission, " Standard Review Plan for Light Water Reactors," NUREG 0800, July 1981.

15. Carolina Power and Light, BSEP/Vol. VI/EOP-01-UG, Rev. 9, pages 72-74.

16. U. S. Nuclear Regulatory Comission, "The Impact of Mechanical and Maintenance Induced Failures of Main Reactor Coolant Pump Seals on -

Plant Safety," NUREG/CR-4400, December 1985.

-17. Carolina Power and Light, "Radwaste Auxiliary Operator Daily Surveillance Report," BSEP Operating Instruction 01-03.6, Volume VII, Rev. 2. June 1989,

18. Floyd, S. D. (CP&L), Memo to A. S. Gill (USNRC), " Additional Information for Station Blackout Review of Brunswick Plant," October 9, 1989, 33.

.- i h ,* '

19. Floyd, S. D. (CP&L), Memo to A. S. Gill (USNRC), " Additional Information for Station Blackout Review of Brunswick Plant," November 13, 1989.

20. Cutter, A. B., letter to the Document Control Desk of the U.S. Nuclear Regulatory Comission, " Brunswick Steam and Electric Plant, Units Nos.

I and 2, Docket Nos. 50 325 and 50 324/ License Nos. OPR-71 and DPR 62, Supplemental Response to Station Blackout Rule," March 30, 1990.

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

22. 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).

23. Rosa, F., Memorandum to the Docket Concerning Beaver Valley Units 1 and 2, " Meeting Summary - Meeting of February 22, 1990, on Station Blackout Issues (TAC 68510/68511)," Docket Nos. 50-334 and 50-412, dated March l 6, 1990.

24. Russell, W. T., letter to W. Rasin of NUMARC, " Station Blackout," dated June 6, 1990.

25. Cutter, A. B., letter to the Document Control Desk of the U.S. Nuclear Regulatory fomission, " Brunswick-Steam and Electric Plant, Units Nos.

l' I and 2, Docket Nos. 50 325 and 50-324/ License Nos. DPR 71 and DPR 62, -

Supplemental Response to Station Blackout Rule," March 3,1989.

26. Tam, P. S., Memorandum for, " Daily Highlight- Forthcoming Meeting with NUMARC on Station Blackout (SBO) Issues (TAC 40577)," dated April 25, 1990, (providing'a Draft Staff Position Regarding Use of Emergency AC PowerSources(EDGs)asAlterna*.eAC(AAC)-PowerSources,datedApril 24,1990).

34-l

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