ML20086R998

Final Technical Evaluation Rept - Wolf Creek Generating Station,Station Blackout Evaluation

ML20086R998
Person / Time
Site:	Wolf Creek
Issue date:	12/17/1991
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20086S004	List: ML20086R998
References
CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-91-1255, TAC-M68628, NUDOCS 9201020374
Download: ML20086R998 (28)

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Attachment 1

,, SAIC-93/1255 TECHNICAL EVALUATION REPORT WOLF CREEK GENERATING STATION STATION BLACK 0UT EVALUATION TAC No. 68628 EAIC Science Applications kstemationalCorpotetion An Employee-Owned Company Final December 17, 1991 Prepared for:

U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Contract NRC-03-87-029 p- .

Task Order No. 38 l ( 2Olo2D379 y 1710 Goodridge Drive. P.O. Box 1303. McLoan Virginia 22102 t703) 8214300 on wc mee av awas owe twooo sens cose ~umm u wes w s~os,s o,. ne o~,mo,wauseom sw.wrousw

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TABLE OF CONTENTS Section Etqg

1.0 BACKGROUND

..................... I 2.0 REVIEW PROCESS

................... 3 3.0 EVALUATION ,.................... 5 3.1 Proposed Station Blackout Duration . . . . . . . 5 3.2 Station Blackout Coping Capability . . . . . . . 9 3.3 Proposed Procedures and Training ........ 16 3.4 Proposed Modifications ............. 17 3.5 Quality Assurance and Technical Specifications . 17

4.0 CONCLUSION

S .................... 18

5.0 REFERENCES

..................... 22 APPENDIX A ........... .......... 24 l

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TECHNICAL EVALUATION REPORT WOLF CREEK GENERATING STATION STATION BLACK 0UT 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 se tton, 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 centainment 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).

i The staff issued Regulatory Guide (RG) 1.155, " Station Blackout," to provide guidance for meeting the requirements of 10 CFR 50.63 (7). Corcurrent with the developmant 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 dccument provides detailed guidelines and procedures 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 1

this method results ir a minimum acceptable SB0 duration. capability

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from two to 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> apw , , the plant's characteristics and vulnerabilities to the risk from station blackout. The plant's characteristics affecting-the required coping capabilii/ are: the redundancy i

of the onsite emergency AC power sources, the reliability'of onsite emeigency L

power sources, the frequency of loss of offsite power (LOOP), and the probable L -time to restore o'ffsite pows . -

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in order to achieve a consistent systematic response from licensees to the 500 rule and to expedite the staff review process, NUMARC developed two generic response documents. 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 Role for Plants Using Alternate AC Power," and

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

A plar.t-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 baseline assumptions used in NUMARC 87 00 are applicable to their plants and to verify the accuracy of the stated results. Compliance with the SB0 rule requirements is verified by review and evaluation of the licensee's submittal and audit review of the supporting documents as necessary, follow up NRC inspections assure that the licensee has implamented 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 documer.tation that support the licensees' submittals for several plants. These audits revealed several deficiencies which were not apparent from the review of the licensee submittals using the agreed upon generic response format. These deficiencies raised a generic question regarding the degree of the licensees' conformance to the requirements of the SB0 rule. To resolve this question, on January 4, 1990, NUMARC issued additional guidance as NUMARC 87 00 Suppleniental Quc!tions/ 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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.. 2.0 REY!EW PR00E55 The review of the licensee's submittal is focused on the following areas consistent with the positions of RG 1.155:

A. Minimum acceptable 500 duration (Section 3.1),

B. SB0 coping capability (Section 3.2),

C. Procedures and training for 500 (Section 3.3),

D. Proposed modifications (Section 3.4), and E. Quality assurance and technical specifications for SB0 equipment (Section3.5),

for the determination of the proposed minimum acceptable SB0 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 N5AC 10B 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 Regulatory Guide 1.130 provides a matrix for determining the required coping duration, for the 5B0 coping capability, the licensee's submittal 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 SB0 of acceptable duration which is determined above. The-review process follows the guidelines given in RG 1.155, Section 3.2, t>

assure:

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

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b. adequacy of the class lE battery capacity to wp;Mrt safe shutoown,

c. avai' lability of adequate compft.ssed air for tir operated valves neces u ry for ofe shutdeun,

d. adequasy of the ventilation systems in the vital and/or dominant areas that include equipment neces:,ary for safe shutdown of the pl ar.t ,

e. ability to provide sopropriate containment integrity, and

f. Ability of the plant to maintain adequate reactor coolant system inventory to ensure cort cooling for the r34uir2d coping duration.

The licenseoi r suttoittsi is reviewed to verify that required procedures (i.e., l'avised existing an.: new) for coping with LB0 are identified anr,4 that appropriate opert, tor tra'ning will be provided.

lhe lis.ensee's submittal for miy proposed modifications to emergency AC sources, cordensate capacity, battery capacity, compressed air capacity, venti)atf or, systt.m. centainment isolation ;ystem integrity, and primary coulant make u9 capability is reviewed. Technical Specifications and quality assurance set forth by the licensee to ensure high reliability of the equipment, specif'.cally added or assigt,ed to meet the requirements of the SBD rule, are a sessed for their adequacy.

' This 500 evaluation is based on a review of the licensee's submittals cated April 17 1909 (1?),'and liarch 30, 1990 (13) and the information available in the plant Updned final safety Analysis Report (UFSAR) (14), it does not include a concurreat f.ite audit review of the supporting documentation. Such an audit may be warranted as an additional confirmatory action. This determination will be made and the audit may be scheduled and performed by the NRC staff at some later date.

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3.0 EVALUATICN During our evaluation, several questions were raised and transmitted to the licensee. The licensee was asked to provide a written response containing the information required to resolve the questions. By December 1991, we had not received the ret,uested informetton. The staff i

made the decision to review the plant coping cepability based on the i

available information contained in the licensee submittals and the plant UFSAR and not to wait for the licensee's response any longer. A list of the questions is givon in Appendix A.

3.1 Proposed Station Blat.kout Duration Licensee's Submittal The licensee, the Wolf Creek Nuclear Operatin0 Corporation, calculated (12 and 13) a minimum acceptable SB0 duration of four hours for the Wolf Creek Generating Station. The licensee stated that no modifications are necessary to att0in this proposed ".oping duration.

The plant factors used tu estimate the proposed SB0 duration are as follows:

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

a. Independence of offsite power group of "!!/2,"

b. Estimated frequency of LOOPS due to severe weather (SW) which places the plant in SW group "2,"

c. Estinated frequancy of LOOPS due to extremely severe weather (ESW) which places the plant in ESW Group "1," and 5

- d. Expected frequency of grid related LOOPS of less than cice per 20 years.

2. Emergency AC (EAC) Power Configuration Group The EAC power a.onfiguration of the plant is "C." Tht Wolf Cretk plant is equipped with two emergency diesel generators nich arn no mally available to the plant's safe shutdown equipment. One em.trgency AC power supply is sufficient to oper4 W the sai; shuttiown equipment following a loss of offsite pcwer.

3. Targit Emergen:y Diesel Generator (EDG) T,eliability Th2 licensee has selected (12) a target EDG rel'a' tiity of 0.95.

He selection of this target reliability ls l:ased on hav ng A nuclear unit average EDG reliability of greater than 0.95 for the last 100 demands, consistent with NUMARC 87 0). [

B Revicw of Licensee's Submittal y Factors which affect the estimation of the SB0 cooinc c'iratici, t.e: the independence of offsite power system grouping, the estimated frequency i of LOOPS due to SW and ESW conditions, the errac ad frequency cf ' rid 1 reiated LOOPS, the classification of EAC, and the selectior,i/ EDG target reliability.

According to the UFSAR, the power to the twu 4.M kV Enginee 1i Safeguards Features (ESF) buses come from tw indepadent and rtiundant sources of preferred offsite power through the electri nily connected switchyard. One preferred circuit from the switchyart is connected to 13.8/4.16 kV ESF transformer XNB01. The second prt;ferred circuit supplies power to the start up tran';r'ormer, which feeds the second 13.8/4.16 kV ESF transformer XHB02. The ESF tiansfonners supply power to ESF division 1 and 2, respectively. Either CSF bus can manually be connected to the opposite division's transformer. Therefore, we 6

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. conclude that Wolf Creek is in the independence of offsite power group ,

"!!/2." based on Table 5 of RG 1.155.

The SW caused LOOP f requency at Wolf Creek is based on the annual snowfall, tornado, the storm (wind velocities between 75 and 124 mph) frequency, and the number of rights of way on which offsite power traverses. The licensee's estimates of snowfall, storm frequency and the number of rights of way are consistent with the guidance of NUMARC 87 00. The licensee stated (12) that the tornado frequency is 0.0003282 per year based the methodology of NUREG/CR-4461 compared to 0.0003815 per year in NUMARC 87 00. Using the guidance and data provided in Section 3.2.1 of NUMARC 87-00, we have calculated the frequency of $W-caused loofs to be 1.01E 2 per year, which places the plant in SW group "3." Using the licensee's estimate of tornado frequency, along with the other NUMARC 87 00 data results in a calculated frequency of SW caused LOOPS of 9.45E-3 and places Wolf Creek in SW group "2." The licensee has not responded to a written request (see Appendix A) for information on the determination of tornado frequency and needs to explain how its approach is more representative of the Wolf Creek site than the data provided in NUMARC 87 00. Without a satisfactory explanation, we consider the site in SW group "3."

We agree with the licensee's estimated frequency of LOOPS caused by extremely severe weather (ESW) conditions which places the plant in ESW group "1." The classification is consistent with that given in NUMARC 87 00. Table 3-2.

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., covers only the events prior to the calendar year 1984. Wolf Creek was not covered by this document.

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Establishment of the proper Emergency AC (EAC) Configuration Group is based on the number of available EAC sources and the number of EAC sources required to operate safe shutdown equipment following a LOOP.  !

Wolf Creek has two dedicated EAC sources with one required after a LOOP, placing the plant in EAC Grcup *C" (RG 1.155 Table 3) as the itcensee correctly identified.

l The final characteristic needed to establish the duration of Wolf Creek's required coping capability is the target EDG reliability. The licensee stated (12) that the assignment of the EDG target reliability J of 0.95 is based on having an average EDG reliability of greater than 0.95 for the last 100 demands. Although this is an acceptable criterion for choosing an EDG reliability, the guidance of RG 1.155 requires that the EDG sti.tistics for the last 20 and 50 demands also be calculated.

Without this information, it is difficult to judge how well the EDG's have performed in the past and if there should be any concern. This information is only available onsite as part of the submittal's supporting documents. Although the licensee's data was not reviewed, we agree that the licensee can choose a reliability goal of 0.95.

The licensee has not responded to the requirement for an EDG reliability program, The licensee has comitted (13) to maintain the targeted EDG reliability of 0.95, however.

Based on an independence of offsite power group "Il/2," an SW group "3,"

and an ESW group "1," the offsite power design characteristic: of Wolf Creek is "P2." With this determination, in conjunction with EAC group "C" and a required coping duration of four hours, the licensee needs to comit to an EDG reliability goal of 0.975. Alter. natively, the licensee could stay with an EDG reliability goal of 0.95 and perform a coping analysis for an SB0 with a duration of eight hours.

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, 3.2 Station Blackout Coping Capability The plant coping capability with an SB0 =<ent for the required duration of four hours is assessed with the following results. If the licensee should increase its coping duration to eight hours, these results are no longer valid:

1. Condensate inventory for decay heat removal Licensee's submittal The licensee stated (12) that the Wolf Creek Technical Specifications requires a minimum condensate storage level of 281,000 gallons. The licensee stated that based on a plant-6pecific analysis,'IS1,000 gallons of water are required for decay heat removal for the required coping duration of four hours and that the plant specific analysis is more conservative than NUMARC 87 00, Section 7.2.1. The licensee stated that no modifications or procedural changes are necessary to use these water sources.

Review of Licensee's Submittal for calculating the condensate inventory requirement for Wolf treek during an SBO, the following shnuld be considered:

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Decay Helt - Using NUMARC 87 00. Section 7.2.1 and the maximum power level (102%) of 3636 MWt, we estimate that the plant would require 80,428 gallons of condensate to remove decay heat for four hours.

2. Sensible Heat The sensible heat removal _ from the primary and steam generator fluid and associated structure should be considered. According to the plant UFSAR, Table 6.2 1-43, the total stored energy in the primary system, excluding l that stored in accumulators, is 853.82 x 10' BTU. The 9

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average RCS fluid temperature is 592.7*f, according to Table 6.215 of the Wolf Creek UFSA' (14). If the average core temperature is neglected, and the entire RCS is assumed to be at 592.7'f, then, on the average,175.5 gallons of water at 125'F would be needed to cool down the RCS one degree Fahrenheit. The RCS is cooled down to about 280 psig or 410'F, according to ECA.0.0 procedural guidance. Therefore, about 32,064 gallons of condensate would be needed.

3. 11nm_ Gengrator WJter Mau .. Our calculation of tt.e additional condensate needed to restore the steam generator levels to hot zero power conditions indicates that approximately $2,800 gallons of condensate would be needed.

This calculation assumes that at the beginning of the incident the steam generator is at the low. low level, containing 82,000 lbm of water. At the end of the four hour cooldown to -260 psig (per ECA 0.0), the steam generator mass is equivalent to that at the hot zero power conditions

(- 3560 f t' of water).

Based on the above, a total of 165,292 gallons of water would be needed to ramove decay heat and cool down the RCS. Although this estimate is larger than that provided by the incensee, the condensate needed during a four hour SB0 is considerably less than the technical specification required condensate storage tank inventory of 281,000 gallons. Therefore, we conclude that the site has adequate condensate inventory to cope with an SB0 of four hours duration.

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,. 2. Class lE tattery Capacity Licensee's Submittal l The licensee stated (12) that a battery capacity calculation has  ;

been ' performed in accordance with NUMARC 87 00, Section 7.2.2 to verify that the class lE batteries have sufficient capacity to meet SB0 loads for four hours.

Review of Licensee's Submittal t The Wolf Creek DC power supply system consists of four separate.

class lE._125 V subsystems._ Each subsystem has a dedicated charger, invertor and battery, and Wolf Creek has a centrally f located battery charger and invertor that can be hooked to any  ;

subsystem. Subsystems 1 and 4 provide control power to AC load  ;

groups 1-and 2, respectively. Each subsystem provides Sital '

instrumentation and control power to similarly named channels of s the reactor protectier, and engineered safety features systems (14).

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According to the plant UfSAR, Section 8.3.2.1.2, the batteries are I sized to supply the necessary DC loads for a minimum of 200 minutes (3.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />)'. This is not adequate to cope with and recover'from an SB0 event with a duration of four hours. The i licensee has not responded to a written request for information '

describing how the class lE battery capacity was determined to be adequate for the four hour SB0 coping duration.  !

3. . Compressed Air  !

Licensee's Submittal Thelicenseestated(12)thatairoperatedvalvesrelieduponto

• cope with an SB0 for four hours can either be operated manually or 11 1

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have sufficient backup sources independent of the unit's preferred and class lE power sources.  !

l Reviaw of Licensee's Submittal The turbine driven auxiliary feedwater (AFW) pump steam supply valve, associated bypass valve and discharge control valves are ,

norm.11y closed, air operated valves which have to be operated during an SB0 event. Additionally, the steam generator power operated relief valves (PORVs) are air operated valves and are backed up by spring operated safety valves. According to the UFSAR, Section 9.3.1, the plant compressed air system provides a '

safety related backup supply of compressed gas (N ) for the PORVs and AFV valves that is designed for eight hours of operation without recharging. Based on this information, we consider the nitrogen backup to be sufficient for the operation of these valves, However, the licensee needs to verify that the duty cycle for the eight hour evolution bounds that of an 580 event. In addition, the licensee nee.is to verify that this backup supply allows remote operation of these valves from the control room.

Otherwise, the licensee needs to ensure habitability (temperature, lighting, and communication) of the areas where these valves will be operated during an SB0 event.

4. Iffects of Loss of Wntilation Licensee *s Sabmittal The licensee stated (12) that the only dominant area of concern (DAC) is the steam driven AFW pump room, for which the calculated ateady state ambient air temperature is 150*F. The calculation of '

, heat transmission into this room was based on the restriction of heat flow due to insulatior . rom three high energy pipes. The licensee stated that a he c generation value equal to twice the maximum heat loss allov

6. Reactor Coolant Inventory Licensee's Submittal The licensee stated (12 and 13) that the ability to maintain adequate reactor coolant system (RCS) Inventory was assessed in a plant specific analysis using the metnod described in NUMARC 87-00, Section 2.5.2. The analysis shows that expected rates of -

reactor coolant inventory loss under SB0 conditions do not result i.' core uncovery for an SB0 of four hours. Therefore, the ,

licen;ae concluded that no additional makeup systems are required to cope wnh an SB0 with a duration of four hours.

Review of I.icensee's Submittal We performed an independent analysis of the RCS inventory using ,

- the information in the plant UFSAR and the submittals. According to the plant UFSAR, the RCS water volume at 100% power is 11,393 L - ft'. Using the assumed RCS leakage of 112 gpm and a final RCS pressure of 280 psia and temperature of 410*F (per ECA 0.0) , we found that 5242.43 ft of water will remain in the RCS at the end 3

of four hours. This exceeds the RCS inventory required to keep l the core covered since the reactor vessel water volume is 3700 ft'. Therefore, we agre with ine licensee's statement that no additional takeup capabilnies are necessary to cope with an SB0 -

of four hours duration.

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NOTit "The Z}_spm RCP seal leak rtt3 was agreed to between NUMARC and the staff pending resolution uf Generic Issue (GI) 23.

If the final resolution of GI-23 defines higher 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 addressing conformance to the SB0 rule."

3.3 Proposed Procedures and Training Licensee's submittal The licensee stated (12) that plant procedures have been reviewed and modificed if necessary to meet the intent of the guidelines in NUMARC 87-00, Section 4 in the following areas:

1. Ac power restoration, and

2. Severe weather.

Review of Licensee's Submittal We neither received nor reviewed the affected procedures or training.

These procedures are plant specific actions concerning the required activities to cope with a 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 that the associated training needs are carried out l accordingly, 16 l

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l 3.4 Proposed Modifications Licensee's submittal The licensee stated (12) that no modifications would be required to cope l with an SB0 with a duration of four hours. l Review of Licensee's submittal l Our review of the licensee's submittals hus identified several c w. erns (see Sections 3.1 and 3.2) which may require modification (s) to cope with an SB0 event of four hour duration. It is the licensee's responsibility to ensure that any modifications comply with the SB0 guidance.

3.5 Quality Assurance And Technical Specifications The licensee did not provide any information on how the plant complies with the requirements of RG 1.155, Appendices A and B.

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, r .' L 4_ . 0 ' 00NEOS!0NS-Based on our review of the licensee's submitthis, and the information

'available in the UFSAR for the Wolf Creek Plant, we find the submittal conforms.with the requirements of the SB0 rule and the guidance of RG 1.155 with the following exceptions:

1. Proposed Station Blackout Duration

a. levere Weather Groupina The licensee, using a sitt specif'ic tornado frequency and the data given in Table 3 3 of NUMARC 87-00, classifies the site as SW group "2." We have used the data and guidance provided in-Section 3.2.1 of NUMARC 87-00 and calculated a severe weather

. frequency'which places the-site in SW group "3." The licensee was '

asked to provide information showing how-its approach is more representative. of the Wolf Creek site than'the' data provided in NUMARC. -The licensee has not responded to this request. Without a satisfactory expl'1ation, we consider the licensee to be in SW

-groupf"3."

b. '{DG Reliability Go(),

'The licensee stated . (12)~-that it would commit to an -EDG.

reliability goal of-0.95. _ J3ased on an indepe.1dence of offsite

_pcwer groups"!!/2," an SW group."3" (see item a. above), and an ESW group "1," the offsite:poweridesign-characteristics of Wolf

. Creek is "P2." With this determination, iniconjunction with EAC

. group "C" and a required coping duration'of four' hours, the licensee needs'to commit to an EDG reliability goal of 0.975.

Alternatively,-the licensee could stay with an'EDG. reliability-goal of 0.95 and re-submit a coping analysis for an SB0 with a ,

duration of~eight hour;;.

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2. Ece'*gency Diesel Generator (EDG) Reliability The licensee needs to have an analysis showing the EDG reliability statistics for the last 20, 50, and 100 demands in its 580 submittal supporting documents. In addition, the licensee has committed to maintain the plant's target EDG reliability, and needs a formal EDG reliability program consistent with the guidance of RG 1.155, Regulatory 1:sition 1.2, and NUMARC 87-00, Appendix D.

3. Class-lE Battery Capacity The licensee stated that a battery capacity ,:alculation has been performed to verify that the class-lE batteries have sufficient capacity to meet SB0 loads for four hours. Our review of the p', ant UFSAR, Section S.3.2.1.2, indicates that the batteries are sized to supply the necessary DC loads for a minimum of 200 minutes (3.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />). This is not adequate to core with and recover from an SB0 event with a duration of four hours. The l licensee was askt.d to provide an analysis that shows that the l

l class-lE batteries have adequate capacity to support the SB0 loads l for the durat.on of four hours. In the absence of any response from the licansee, we can not confirm the licensee's statement.

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4 Compressed Air Our review of the plant UFSAR indicates that a backup supply of compressed air (nitrogen) is available for the operation of FORV's and AFW valves for a duration of eight hours. We consider the nitrogen backup to the compressed air system an adequate source of compressed air for the SB0 coping duration. However, the licensee needs to verify that the duty cycle for the eight hour evolution bounds that of an SB0 event. In addition, the licensee needs to '

verify that this backup supply allows remote operation of these valves from the control room. Otherwise, the licensee needs to 19

, ens'Jre habitability (temperature, lighting, and conrnunication) of the areas where these valves will be operated during an SB0 event.

5. Effects of Loss of Ventilation Thelicenseewasaskedtoprovidethecalculationpackag'e'5forthe control room and the AFW pump room as well as documentation of the review process used to determine that no additional plant areas are potential dominant areas of concern. The licensee has not provided this information. Therefore, we are not able to confirm that there is reasonable assurance of the operability of equipment needed to cope with an SB0 event.

6. Containment Isolation Our review of the containment penetrations and associated containment isolation valves (CIVs) identified four penetrations which do not meet the CIV exclusion criteria of NUMARC 87-00, Section 7.2.5. The CIVs for penetrations 13, 14, 15, and 16 meet the intent of SB0 requirements if they are closed during normal operation per procedure and are only ongd for surveillanc? (i.e.

cold shutdown and/or refueling) during Modes 5 and 6. Otherwise, The licensee needs to list these valves in an appropriate procedure and identify the actions necessary to ensure that these valves are fully closed, if needed, upon the loss of AC power.

The valve closure needs to be confirmed by position indication (local, mechanical, remote, process information, etc.).

7. Proposed Modifications t

Our review of the licensee's submittals has identified several concerns (see items 1 through 6 above) which may require modification (s) or changes to existing equipment to enable the l

plant to achieve a four hour SB0 coping duration l

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. 8. Quality assurance and Technical Specifications The licensee needs to provide information on how the plant complies with the technical specification and quality assurance requirements 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 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. % clear Regulatory Comission, " 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 Comission, " Reliability of Emergency AC Power System at Nuclear Power Plants," NUREG/CR 2989, July 1983.

5. U.S. Nuclear Regulatory Comission, " Emergency Diesel Generator Operating Experience, 1901-1983," NUREG/CR-4347, December 1985.

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

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

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

9. Thadani, A. C., lettee to W. H. Rasin of NUMARC, " Approval of NUMARC Documents 'n Station Blackout (IAC-40577)," dated October 7,1988.

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).

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

September 1966.

12, 8ailey, J. A., letter to V. S. Nuclear Regulatory Commision Document Control Desk, " Docket So. 50-482: Response to Station Blackout Rule (NRC TAC No. 68628)," dated April 17, 1989.

13. Rhodes, F. T., letter to U. S. Nuclear Regulatory Commission Document

-Control Desk, " Docket No. 50-482: Supplemental Response to Station Blackout Rule (NRC TAC No. 68628)," dated March 30, 1990.

14. Wolf Creek Plant, Updated Final Safety Analysis Report, i

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1 APPENDIX A

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A

' June 10,1991 QUESTIONS ON THE WOLF CREEK SBO SUBh11TTAL' WITH EMPHASIS ON THE FOLLOWING AREAS

1) Severe Weather (SW) Group Classincation The Wolf Creek submittal estimation of the frequency of LOOPS due to severe weather places the plant in SW Group 2. An estimate using the data provided in NUMARC 87 00 and transmission lines on multipic rights of ways would place Wolf Creek in SW group 3. This is important because being in SW Group 2 allows Wolf Creek to choose an EDG reliability target of 95%, while SW Group 3 would require a target of 97.5%. The diffetence in estimates is caused by different estimates of the annual expectation of tornados of sevurity f2 or greater. Explain the reason (s) for the difference between your estimate and that given in NUMARC 87-00. If the analysis package provides complete answers to these questions, please provide the package.

2) Class 1E Battery Cnoncity The UFSAR states that the plant has four 125 V Class IE ba:teries that are designed to function for 200 minutes (3.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) without charging. Explain what loads will be stripped, and when they will be stripped, to ensure that the batteries will last for the four hour SBO without charging. Describe the load profile, method and assurnptions (e.g., temperature factor, design margin, aging factor) which were used to determine that the battery capacity is adequate for four hours. If a calculation package provides complete answers to these questions, please provide the package.

3) Effects of Loss of HVAC More detailed explanations of analyses of the effects ofloss of ventilation are needed in order to perform a meaningful review. If calculation package (s) provide complete answers to these questions, please provide the package (s).

Our initial concerns are:

a) - What are the areas chosen as areas of concem? What areas were considered, but dismissed as being of concern?

b) What methods were used to calculate the temperature rises.

c) State the major assumptions made in each room analysis (i.e., heat loads initial temperatures, invertor efficiency, material properties, timing) 25

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d) When are the corridor doors and designated cabinets identified in your submittal going to be opened? What procedure govems this win?

e) What is the expected temperature of the containment during an SBO?

Does it pose equipment operability problems?

4) Containment Isolation Explain the containment isolation valve (CIV) analysis and how CIVs are treated in SBO procedures. Ple:..: address the treatment of RHR suction isolation, sump isolation and Containment Spt.ty suction isolation valves.

S) Condensate inventorv How was the condensate inventory required for coping with an SBO determined? What were the major assumptions in the analysis? Did the analysis take into account water used for (1) decay heat removal, (2) sensible heat removal from the RCS fluid and structure, (3) sensible heat removal from the steam generator fluid and suveture, and (4) restoration of steam generator water level? What were the contributions to the total water required of the four above mentioned areas. If a calculation packege provides complete nnswer to these questions, please provide the package.

6) Compressed Air What is the basis for the statement that Wolf Creek has adequate compressed air to cope with a station blackout with a duration of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Detailed information may be required in areas other than those mentioned above. If a telephone conversation is scheduled, hr.ve knowledgeable personnel

.nd supporting documentation available for all areas of the SBO analysis and submittals.

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