ML20058K225

Technical Evaluation Rept,Braidwood Station Units 1 & 2 Station Blackout Evaluation

ML20058K225
Person / Time
Site:	Byron, Braidwood
Issue date:	06/14/1990
From:	SCIENCE APPLICATIONS INTERNATIONAL CORP. (FORMERLY
To:	NRC
Shared Package
ML20058K167	List: ML20058K165 ML20058K225 ML20058L991 ML20058L996 ML20058M000
References
CON-NRC-03-87-029, CON-NRC-3-87-29 SAIC-90-1043, TAC-68515, NUDOCS 9006270317
Download: ML20058K225 (31)

Text

.. .._. _ _ . _ .. _ _ - ..- . - - - - - -

i ATTACHMENT'l TO ENCLOSURE 1 SAIC 90/1043 t

I l-TECHNICAL EVALUATIM REPORT I BRAIDWOOD STATIM UNITS 1 AND 2 STATIM BLACK 0UT EVALUATI M l

l. ,

,. i TAC Nos. 64515 and 68516 SAM Science Appilcations International Corporation An Empbyee-Owned Company Final June 14.' 1990

. +

t i

i

t-l Prepared for U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Contract NRC-03-87-029 Task Order No. 38 l

st O!!ce Box 1303,1710 GoodrWge Drive, McLean, Virginia 22102 (703) 8214300 QUQlo2.7697 8(pf' .,.

\ :'

lll-1 l TABLE OF CONTENTS Section g h

1.0 BACKGROUND

..................... 1 2.0 REVIEW PROCESS ................... 3 .

1 3.0 EVALUATION ..................... 6 Li 3.1 Proposed Station Blackout Duration ....... 6 i

3.2 Alternate AC (AAC) Power Source . . . . . . . . . 11 3.3 Station Blackout Coping Capability ..... . 14 3.4 Proposed Procedure and Training . . . . . . . . . 21 3.5 Proposed Modification ............. 22 3.6 Quality Assurance and Technical Specifications . 22

4.0 CONCLUSION

S .................... 24

5.0 REFERENCES

...................... 27 t

A s

\

4 o

1 11 l

1

e I

TECHNICAL EVALUATION REPORT

~

BRAIDWOOD STATION UNITS 1 AND 2 STATION BLACK 0UT EVALUATION

{ l.0 BACKGROUND On July 21, 1988, the Nuclear Regulatory Comission (NRC) amended its regulations in 10 CFR Part 50 by adding a new section, 50.63, " Loss of All Alternating Current Power" (1). The objective of this requirement is to i

assure that all nuclear power plants are capable of withstanding a station 7 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 comission 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). 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 WaterReactors,"NUMARC8700(8). This document provides detailed guidelines and procedures on how to assess each plant's capabilities to comply with the S80 rule. The NRC staff reviewed the guidelines and analysis methodology in ,

l NUMARC 87 00 and concluded that the NUMARC document provides an acceptable  ;

l guidance for addressing the 10 CFR 50.63 requirements. The application of this method results in selecting a minimum acceptable SB0 duration capability I

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

I 1

5

f  ;

1 i In order to achieve a consistent systematic response from licensees to l

the SB0 rule and to expedite the staff review process, NUMARC developed two  ;

generic response documents. These documents were reviewed and endorsed by the NRC staff (11) for the purposes of plant specific submittals. The documents are titled:

i

1. ' Generic Response to Station Blackout Rule for Plants Using ,

Alternate AC Power,' and l

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

A plant specific submittal, using one of the above generic formats, provides only a summary of results of the analysis of the plant's station blackout coping capability. Licensees are expected to ensure that the 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. Fn11ow 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 L 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 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 rest,1ve this question, on January.4,1990, NUMARC iss'ued additional guidance as h9 MARC 87-00 Supplemental Questions / Answers (13) 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, 2

i

- , , _ e 4.wv, -- , --,,-,-,r.---,.,-y-- -.----_ + - - - - , - . , . _ .

i

' O.

[ l 2.0 REVIEW PROCESS The review of the licensee's submittal is focused on the following areas j consistent with the positions of RG 1.155:  ;

a A. Minimumacceptable580 duration (Section3.1),  ;

o B. SB0copingcapability(Section3.2),

l C. ProceduresandtrainingforSB0(Section3.4),  ;

D. Proposedmodifications(Section3.3),and l

E. Quality assurance and technical specifications for 580 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 I power design characteristics, b) emergency AC power system configuration, c) determinationoftheemergencydieselgenerator(EDG)reliabilityconsistent 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 RG 1.155 provides a matrix for determining the required coping duration.

l For the SB0 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, to

, assure:

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

i

I-

)

b. adequacy of the class lE battery capacity to support safe  ;

shutdown, i

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.

i The licensee's submittal is reviewed to verify that required procedures (i.e., revised existing and new) for coping with 580 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 l 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 to determine whether it meets the criteria and guidelines of Section 3.3.5 of RG 1.155 lind Apper. dix B of NUMARC 87-00.

This SB0 evaluation is based on a review of the licensee's submittals dated April 17,1989(10), and March 30,1990(16), the information available in the plant updated Fint1 Safety Analysis Report, (UFSAR) (12), a telephone conversation between NRC/SAIC and the licensee on February 26, 1990, and a 4

l

i 7

L follow up response from the licensee dated March 29,1990(17); it does not

}.

include a concurrent site audit review of the supporting documentation. Such an audit may be warranted as an additional confirmatory action. This determination would be made and the audit would be scheduled and performed by the NRC staff'at some later date.

9 I

t i

i e

1 i e t

0 I

l' la i

5 n

,----,,_.,.,_-,,--,..,.,.,.,._,.q..._,,,_ _ ._ , - , , . , , , , , , . _ ,

,. ,,-,,,.,a..

?

t

'{.

3.0 EVALUATION 3.1 Proposed Station Blackout Duration Licensee's submittal r

4 Thelicensee.CommonwealthEdisonCompany(CECO), calculated (10and16) a minimum acceptable SB0 duration of four hours for the Braidwood Station Units 1 and 2. The licensee stated that no modifications are necessary to attain this proposed coping duration.

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

l

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

I

a. Independence of offsite power group is "12,"

b. Estimated frequency of LOOP due to severe weather (SW) places Braidwood in SW group '2," ,

l

c. Estimated frequency of LOOP due to extremely severe weather

! (ESW) places Braidwood in ESW Group '1,* and  :

e. Expected frequency of grid related LOOP does not exceed once l per 20 years.

l '

2. Emergency AC (EAC) Power Configuration Group The EAC power configuration of the plant is 'C.' Each of the two l units at Braidwood are equipped with two emergency diesel generators which are normally available to the unit safe shutdown equipment.

6

9 l-One emergency AC power supply is sufficient to operate safe

,, shutdown equipment for each unit following a loss of offsite

! power.

3. Target Emergency Diesel Generator (EDG) Reliability i

The licensee has selected a target EDG reliability of 0.95 based on having a nuclear unit average EDG reliability of: a) greater

, than 0.90 for the last 20 demands for both units, and b) greater I

than 0.94 for the last 50 demands for Unit 1 (data for Unit 2 is

, not yet available due to limited operating history), consistent i with the NUMARC 87-00 selection criteria.

Review of Licensee's submittal Factors which affect the estimation of the 580 coping duration are: the  ;

independence of offsite power system grouping, the expected frequency of '

LOOP caused by grid related failures, the estimated frequency of LOOPS caused by severe weather (SW) and extremely severe weather (ESW) conditions, the classification of EAC, and the selection of EDG target reliability.

The licensee's estimation of LOOP frequency caused by SW condition is consistent with the guidance provided in NUMARC 87-00, Table 3 3, using multiple rights-of way transmission lines. The licensee's estimation of LOOP frequency caused by ESW condition places the site in ESW group "1."

This is different from the one provided in NUMARC 87-00, Table 3-2. ,-

DuringthetelephoneconversationonFebruary 26, 1990, the licensee stated that the Braidwood data in Table 3-2 of NUMARC 87-00 is not correct. The licensee added that Braidwood is located 5 miles south of

! the Dresden site and 15 miles east of the La Salle site. Both the La

] Salle and Dresden sites are categorized as ESW group "1." The licensee also stated that his consultant had contacted Mr. J. Flack of NRC and had confirmed that the Braidwood ESW group should be similar to that of 7

---.-.--,.,,---v-,~_---y --

-,-%, , . - - - - - - - - , - - - - - ~ , , - . , . , . _ - , - - <

, . i i Dresden(17). We believe the licensee's argument to be valid end the site should be categorized as ESW group 'I."

Our review of the Braidwood Station UFSAR (Section 8.0) indicates that. -

(see Figure 1). ,

1. all offsite power sources are connected to the plant through one switchyard, L

2. therearetwoemergencysafetyfeature(ESF)divisionsineach  ;

unit, and each division is normally powered from one system auxiliarytransformer(SAT),

3. the unit SATs (two per unit) are fed through one power line which connects them to the corresponding unit 345 kV ring bus (preferred power source for the emergency buses),

4. each SAT is capable of supplying the design basis accident loads of both divisions of one unit and the safe shutdown loads of both divisions of the other unit simultaneously, and

• i

5. upon loss of preferred power source, all safe shutdown buses can '

be connected to the alternate power source by a manual transfer.

\

f The alte nate power source for each of the emergency buses is through the emergency bus of the opposing unit. The manuai line-up requires a total of four circuit breakers (two on each cross-tie) to be closed for establishing power to all emergency buses of the opposing unit. Using ,

the, guidance provided in Table 5 of RG 1.155, we conclude that the

licensee has correctly established the independence of plant offsite power grouping as "12.'

l EstablishmentoftheproperEmergencyAC(EAC)ConfigurationGroupis based on the number of EAC sources available and the number of EAC sources reqrtred to operate safe shutdown equipment following a LOOP.

8 1

Y i

tK nS e e UMe o

T2S cg in ta al m

er ho ..

~ cr sris 21 2mi h 4

t2 Tat I

i n5 Nsf U0 Uio ,

o.8 TE  :

U*  :

~ 22 4

ET O

N

_ T2

.] I M5 0

).

9 o

rv TE KE' 4

6 4 1 L- 219 1 I - T U246 I A4 1 - - W N

OE55 IS44 1

N N O

D TN33 I N AA T A TR: : A Y ST yx 7 T

N N I 8 S

C T 1 -

D I

W ea .

O O

S 1 W K

y c ey .

9 D I

2 8 A 4 5 4 R W:W:

1 1 8 4

3 Y- -

5 0

8 b G D

^ 1 E.

e A

r u

g b 1 i

V W- G F

O 1y Q .

. f Nrx 1 4

1

\ .

S U

B y

.1 I -

E U2 I I l I A4 1

V N ~

. OE 3 R IS E TN 4 1G AA 1 TR tN ST S iI U nA B UM

1. .

i Braidwood has two dedicated EAC sources with one required to operate after a LOOP, placing the plant in EAC Group 'C' (RG 1.155 Table 3) as I the licensee correctly identified.

4 The final characteristic needed to establish the duration of Braidwood's l required coping capability is the target EDG reliability. The licensee has selected a target EDG reliability of 0.95 consistent with the RG 1.155 and NUMARC 87-00 guidance. We were unable, however, to verify the demonstrated start and load run reliability of the plant EDGs. This .

information is only available onsite as part of the submittal's i

" r supporting documents. In response to the requirement for an EDG reliability program, the licensee stated during the telephone conversation on February 26, 1990, that a reliability program consistent with the guidance provided in RG 1.155 and NUMARC 87-00 is being developed. This statement was not documented in the licensee's '

submittal; however, it is committed to maintain the targeted EDG l reliability of 0.95.

1 With regard to the expected frequency of grid-related LOOPS at the site, we cannot confirm or reject 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., only covers the events prior to the calendar year 1984. Neither unit was in commercial operation before 1984.

During the telephone conversation on February 26, 1990, the licensee l stated that none of the CECO's nuclear stations has had any grid-related LOOP in the last 20 years. However, during a meeting on the

! Dresden and Quad Cities SB0 response on March 28, 1990, it was indicated j that' an event which had caused a reactor trip at La Salle, early that day, had also caused a voltage fluctuation resulting in an ESF actuation at Braidwood. A CECO staff member stated that a similar event occurred I'

in March of 1989, when lightning struck the Braidwood site but caused a i reactor at the la Salle site to trip. Both of these events indicate i

that a strong dynamic dependence exits between the offsite power

(

10 f

g- .

L characteristics of these two sites. Although no other information was available for each of these events, both events point to grid related dynamic instability and could be related to the way the offsite power lines are connected between these two sites. Figure 8.2.5 of the plant UFSAR indicates that the la Salle's 345 kV output is directly connected

_l: to the Braidwood's 345 kV ring buses.

I '

The licensee needs to evaluate these two events and verify that they are not symptomatic of underlying or growing grid instability between these two sites. Depending on the outcome of this evaluation the offsite power characteristic of the site could be affected (i.e. a change from I

'Pl'to'P3'). Based on the above, and pending verification of grid-related problems at the site, the AC power design characteristic of the Braidwood site is considered to be 'Pl" with a minimum required 580 1 duration of four hours, i 3.2 AlternateAC(AAC)PowerSource '

Licensee's submittal ,

~

ThelicenseestatedthattheAACpowersourceatBraidwoodStation(BS) will be the emergency AC power source from the non blacked out unit which meets the criteria specified in Appendix B to NUMARC 87 00 and the 3

assumptions in Section 2.3.1 of NUMARC 87-00. The AAC power source is available within 10 minutes of the onset of the SB0 event by manual operation of cross-tie breakers from the control room. The AAC configuration for the station is similar to configuration 2B of NUMARC

,; 87-00 Appendix C. Each of Braidwood's two units have two dedicated l' EDGs. Upon the loss of offsite power and failure of one unit's diesels to 6perate, either one of the other unit's diesels is capable of .

I providing power for safe shutdown of both units for four hours, i

l Il o

l i-l

~e.. - - --,--n.--.nm , .,--n,...._,.,,,..,,.y-n,,_,.ca.,. . , ., .-- .n.. --.~

t Review of Licensee's submittal Except for the followinW concern, we agree with the licensee's statement l

that the AAC power source (site EDGs) meets the criteria in Appendix B

! of NUMARC 87-00:

i Paragraph B.9 of Appendix B states, " .... At a multi-unit site, except for 1/2 Shared or 2/3 emergency AC power configuration, an adjacent unit's Class It 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.'

During the telephone conversation on February 26, 1990, the licensee stated that all of the LOOP safe shutdown loads, except for the auxiliary feedwater (AFW) motor driven pump and the component cooling water pumps will be powered on the non blacked out (NBO) unit. The licensee added that the diesel driven AFW pump will be used instead of ,

the motor driven pump, and the component cooling system will not be required. The licensee also stated that the SB0 loads on the blacked-  ;

out unit will be almost the same as those on the NB0 unit. The ,

difference is the common equipment loads, such as main control room heating, ventilation and air conditioning (HVAC), that need to be powered from the NB0 unit. Based on this load management scheme and the capacity of the unit EDGs (with a 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of Sg34 KW each), the licensee claims that all the equipment needed for safe shutdown operation in both units can be powered from one EDG within 10 minutes. '

The guidance on the use of existing EDGs as AAC power sources at multi-L unit sites is documented in RG 1.155, Section 3.3.5, NUMARC 87-00, Sect' ion 2.3.l(3) and further detailed under question 3.4 and B.3 in NUMARC 87-00 Supplemental Questions / Answers which was reviewed and acceptedbytheNRCstaff(13). The SB0 rule states that at multi-unit sites where the combination of EAC power sources ' exceeds the afnimum redundancy requirements for safe shutdown (non-DBA) of all units, the i

12

e

'l e remaining EAC sources may be used as MC sources' provided that they ,

meet the applicable requirements.

The rule statement requires 'afn/aus redundancy. ' This means that in order for an EDG to qualify as an MC source there must be an EDG  :

! available in the N80 unit in addition to the number of EDGs required to meet the minimum EDG redundancy requirement for powering nonnal safe  ;

shutdown loads following a LDOP event. Thus, the EDGs in a two-unit site with two dedicated EDGs per unit would not qualify as MC sources.

• Two EDGs per unit would m e t on1I the minimum redundancy requirement, and there is no excess EDG, However, there are some plants at multi-unit sites which have EDGs that just meet the minimum redundancy but each EDG has sufficient capacity to power all the normal LDOP loads of the NB0 unit and also has sufficient  !

excess capacity to power the required safe shutdown loads of the SB0 unit. Recognizing the existence of this type of situation, the staff j has interpreted the 'l/teral' excess EDG redundancy requirement of the SB0 rule to allow large capacity EDGs to qualify as MC source, provided ,

other applicable requirements are met.

In order to take credit for this< interpretation, the NRC staff's basic position has been (14, 18 and 1g)'that:

1. no action should be taken that would exacerbate the already difficult situation in the NB0 unit. Any actions that make '

operator tasks more difficult such as load switching or i- disablement of information readouts or alarms in 'the control room are also considered to be a degradation of normal safe shutdown l ' capability for LDOP in the NB0 unit. And, 1

2. excess capacity of the EDG being designated as an MC source ll should not be the capacity made available by shedding or not l powering normal safe shutdown loads in the NB0 unit. Examples of lf l

.such loads are: motor driven auxiliary feedwater pumps; heating, 13

j. ventilation and air conditioning loads; the power supply of the plant computer; one or more sets of redundant instrumentation;

- etc. The shedding of such loads constitutes degradation of the normal safe shutdown capability of the NB0 unit.

It is not in the interest of safety to reduce the capability to handle various eventualities in one unit for the purpose of meeting the SB0 rule in another unit. Each unit must meet the SB0 rule on its own merits without reducing another unit's capability to respond to its own potential problems.

The excess capacity of the EDG in the NB0 unit that qualifies it as an AAC source is, therefore, 'only that available capacity within the normal continuous rating but above the EDG load represented by the complete contingent of safety related and non-safety related loads normally expected to be available for the LOOP condition.'

Our review of the licensee's proposed actions to meet the requirements of the SB0 rule (10 CFR part 50.63) indicates that the proposed load shedding scheme on the HBO unit is not consistent with the guidance stated above. The licensee has proposed to shed loads which will result in the degradation of the NB0 unit safe shutdown capability. If these loads were not shed, the AAC power source would not have sufficient capacity to power the equipment needed for safe shutdown operation of the blacked out unit. Therefore, the licensee's submittal does not conform with the requirements of the SB0 rule.

3.3 Station Blackout Coping Capability

The' licensee stated that since the AAC power source will be available within 10 minutes, the coping evaluations for class 1E battery capacity, compressed air, and containment isolation need not be addressed per 10 CFR 50.63(c)(2). The plant SB0 coping capability for the required four hours duration is assessed based on the following reviews

14 l

e I

1. Condensate inventory for decay heat removal i l <

Licensee's submittal '

y i The licensee stated that a total of 78,858 gallons of water is '

required per unit for decay heat removal without cooldown during i the 4-hour $80 event. The minimum permissible condensate storage  !

l tank level by technical specifications provides 200,000 gallons of water per unit, which exceeds the required quantity for coping l 3

with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SBO.

p Review of Licensee's subt.ittal l 4

i The condensate inventory needed to remove decay heat was estimated I using the information available in the plant UFSAR and the guidance provided in NUMARC 87 00, Section 7.2.1 assuming no cooldown of the primary system. Our calculations indicate that .

the licensee has properly evaluated the condensate requirements to 4 remove decay heat during a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SB0 evant. The licensee stated that, at a minimum, each unit would have 200,000 gallons of ,

condensate water. Therefore, we agree with the licensee that the site has sufficient condensate inventory to successfully cope with and recover from a 4-hour SB0 event.

E

2. Class IE Battery Capacity Licensee's submittal

, The licensee's submittal does not address the battery capacity.

During the telephone conversation on February 26, 1990, the

] i licensee stated that since the AAC is available within 10 minutes from the onset of an SB0 event, the battery calculations do not ll need to be addressed per guidance provided in NUMARC 87-00, Section 7.1.2, and 10 CFR Part 50.63 (c)(2). He added that no a

15 l

c i r i

~

i cross connect capability exits between the ESF divisions in each j unit. l Review of Licensee's submittal i l

Information in the plant UFSAR indicates that each division battery can support the connected loads for more than 30 minutes and can last for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the inverter loads were shed after 30 -  ;

minutes. The available information, however, is insufficient to be able to confirm the stated results. The licensee did not provide battery sizing calculation base on the assumptions that the AAC power source will charge one division of Class IE station batteries. Since our review indicates that the proposed AAC power i source does not meet the requirements of the 580 rule, the l sssumption that the AAC source will power one division of Cicss IE battery chargers within the 10 minutes is invalid. To conform with the guidance provided in NUMARC 87-00 Supplemental ,

Questiont/ Answers (13), the licensee needs to assess battery capacity, or provide charging capability to verify that all normal battery backed plant monitoring and electrical system controls

, remain operational for successfully coping with and recovering from an SB0 event.

3. Compressed Air Licensee's submittal The licensee's submittal does not address the compressed air system. During the telephone conversation on February 26, 1990, the licensee stated that no air operated valves are relied upon to I

cope with an SB0 event. Since an AAC power source will be available within 10 minutes, analysis of the compressed air system is not required.

16

-t.

I Review of Licensee's submittal

, The UFSAR states that failure of the compressed air systems will not prevent safety related components or systems from performing as intended. The' auxiliary feedwater control valves are air

{ operated. Upon a loss of air, these valves can be controlled locally (manually). The decay heat is released to the atmosphere 1

through a combination of atmospheric dump valves (ADVs) and main steam safety relief valves (MSSVs). The ADVs, (a total of four),

are electro hydraulically operated valves with manual backup l capability. The licensee stated that each ESF division powers two

, ADVs. After the AAC power source is established, the two ADVs on I the steam generators (SGs) powered from the available ESF division can be operated from the control room. The MSSVs ori the remaining two SGs would modulate to release the decay heat and maintain steam pressure near the lowest MSSV pressure set point.

)

Although this proposed decay heat release process produces an i asymmetric temperature in the reactor coolant loops, it will not I' affect the safety of the plant. We believe proper operation of the two ADVs will eventually stop the MSSVs modulation once the ADVs release capacity exceeds the decay heat generation rate.

However, the licensee needs to simulate this decay heat removal j operation scenario and train the operators appropriately.

In addition, since AAC power source does not meet the requirements of the SB0 rule, the assumption that the AAC source powers two of j the ADVs thereby enabling the operators to modulate these valves from the control room is invalid. Therefore, the licensee needs to provide an alternate power source for the operation of the

! ADVs, and simulate the process to train the operators.

I 17

i

4. Effect of Loss of Ventilation L.censee's submittal I I

The licensee stated that the AAC power source provides power to

] heating,ventilationt..Jairconditio'ning(HVAC)systemsserving ,

dominant areas of concern to achieve and maintain safe shutdown I during an SB0 event. Therefore, consistent with the NUMARC 87-00, Sections 7.2.1 and 7.2.4 the effects of loss of ventilation '

j were not assessed.

Review of Licensee's submittal l f

During the telephone conversation on February 26, 1990,.the licensee stated that HVAC systems for the following areas will be available once the AAC power source is established:

I o control room (common to both units),

o auxiliary electrical equipment room, 1..

ll o miscellaneous electrical equipment room, o essential service water cubicle, j

o containmeat a1,' recirculation fans,

+

o charging pump ctibicie, o auxiliary feedwater pump arra, o diesel generator room, and o battery room.

18 l

l l' l

The licensee added that the component cooling water (CCW) system I

will not be operating in the blacked out unit. This system provides cooling water to several systems including containment penetrations. The plant UFSAR indicates a need for-44 gpm of CCW water flow through 29 penetrations during normal, or LOOP / hot ,

shutdown /cooldownconditions. Loss of this cooMng flow causes the local concrete surface temperature to rip od'aventually I approach the temperature of the piping which passes through the f

penetration. The maximum temperature could be as high as the steam line temperature which is around 550*F. The American t ConcreteInstitute(ACI 348 80) (15) states, 'for accident or short term period the temperature should not exceed 350*F for the surface." In response, the licensee stated that an analysis of the effect of loss of CCW will be performed for the SB0 unit (17).

Our analysis of capacity of the AAC power source indicates that "

L the available capacity is not sufficient to power the HVAC systems ,

L of the areas mentioned above. Therefore, the licensee needs to l l4 provide calculations for equipment operability in the blacked out  :

unit.

i 5. Containment Isolation .

e

!- Licensee's submittal lj The licensee's submittal does not address the containment isolation system. During the telephone conversation on February 26, 1990, the licensee stated that the AAC power source will be available within 10 minutes and has sufficient capacity to power l-the containment. isolation valves (CIVs) requiring closure capability during an SB0 event. Therefore, no analysis of CIVs is l

necessary.

l

l ll 19 a

1

r i'

Review of Licensee's Submittal During the telephone conversation on February 26, 1990, it became apparent that. if the AAC power were to be available, only one division of ESF buses would be powered. Therefore, not all the CIVs would be operational. However, our review indicates that the AAC power is not capable of powering all the proposed loads in the blacked out unit and the licensee needs to verify that appropriate f[

containment integrity can be assured during an SB0 event.

i I Reactor Coolant Inventory 6.

Licensee's Submittal 4

The licensee stated that the AAC source powers the necessary makeup system to maintain adequate reactor coolant system inventory to ensure that the core is cooled for the required coping duration.

Review of Licensee's submittal The reactor coolant system-(RCS) losses which the licensee needs to consider are:

1. 25 gpm per pump losses through reactor coolant pump seals j per NUMARC guidelines, I- 2. maximum allowable RCS leakage.per plant technical specifications.

l

4. Sraidwood has four reactor coolant pumps which will lose a total of 100 gpm seal leakage from the reactor coolant system (RCS).

l The licensee stated that the maximum allowed RCS leakage per technical specification is 11 gpm. Therefore, a total of 111 gpm f

20 l

r leakage from the RCS must be made up in order to maintain inventory.

1 I

Since the licensee's AAC power source does not conform to the NRC's guidance and the requirements of the SB0 rule, the assuniption that the AAC powers the make up system is invalid.

Therefore, the licensee needs to perform an evaluation showing j that, with an RCS leakage of 111 gps, the core will not be uncovered during an SB0 event.

[

3.4 Proposed Procedure and Training l '

Licensee's submittal i.

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

{ areas: ,

l 1. AC power restoration, and  !

o l

l 4

2. - Severeweather(tornado).

Procedures dealing with SB0 response have been reviewed and will be I modified to

1. start and load the AAC,

[ I. ensure operation of the diesel-driven AFW pump,

3. start a charging pump, and

. 4. restore offsite and EAC power when available.

.{-

, 21

-f i

. , - , . , --.-,w ,.c. .-- - , - - - - - , - . . - ,

d

/

The licensee stated that procedure changes required to conform with the NUMARC guidelines will be implemented within one year of the notification from'the NRC staff per 10 CFR 50.63(c)(3).

Review of Licensee's submittal Jt-The affected proci.iures were not submitted by the licensee for the NRC staff review. We' view these procedures as plant specific actions concerning the required activities to cope with an SBO. We believe that {

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 and that the associated training needs are carried out accordingly. -

3.5 Proposed Modifications ~

Licensee's submittal ,

The licensee did not identify any needed modifications to equipment at Braidwood to cope with a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SB0 event .

I Review of Licensee's Submittal -

No equipment modifications were discussed in the licensee's submit.^

Our analysis indicates that some modifications in terms of changm procedures and/or coping approach are necessary to conform.with tl3 requirements of the SB0 rule.

3.6 Quality Assurance and Technical Specifications Tha licensee did not provide any information on how the plant complies with the requirement of RG 1.155, Appendices A and B. The licensee stated that all the SB0 equipment is covered by the normal plant quality assurance program and technical specifications. The licensee needs to verify that the 22 l

l w _ . - _ _ _ - se

e ,

.j

't v .,,

=, j '

I'SB0 equ'ipment is covered by appropriate QA and technical specifications-

~

~

i

< programs consistent with the guidence of RG 1.155, Appendices A:and BL .

t u

d.

d.

m  :

+

.{,

l-k 4

l J. ., l

.(

.h  ; ,

).

1 -l

4-l

.l l

.j ,' s s ;;

4 l

^(i

(

i 23 r

t;{r i

it.

\

)l__ i 7 .

. t ,

-. _ _ - - - _ _ _ _ - _ _ - - _ _ _ _ - _ _ _ - - - - _ - - - - _ _ _ - _ - - - _ _ - - - - _l

4.0 CONCLUSION

S Based on our review of. the licensee's submittals, a telephone

, conversation between NRC/SAIC and the licensee, and the information available in the UFSAR for Braidwood station, we find the submittal- does not conform

.with the SB0 rule for the following reasons:

l.

1. Emergency Diesel Generator Reliability Prograa p The licensee's submittals do not document the conformance of the I plant's EDG reliability program with the guidance of the RG 1.155, Section 1.2 and NUMARC 87-00, Appendix D. The licensee stated that plant engineering is in the process of developing a reliability program consistent with the above guidance. The licensee is comitted, however, to maintain the targeted EDG reliability of 0.95.

tj ,

2. Alternate AC power source '

=

The licensee's proposed load shedding of the non-blacked-out (NB0) j, unit is not in conformance with the requirements of the SB0 rule t -- and the guidelines provided in RG 1.155, NUMARC 87-00 Supplemental Questions / Answers. The proposed load shedding scheme will result

~

in the degradation _of LOOP safe shutdown capability of the NB0

=

unit. This excess capacity made available by proposed load

_[ stedc'ing could not be credited as an AAC source for the blacked out unit (see the discussion under the AAC power source in Section V 3.2). Therefore, the AAC power source does not have sufficient I capacity to power the proposed safe shutdown equipment in the q blacked out unit.

_t Class IE Battery Capacity

[ ,

3.

The plant UFSAR indicates that each class IE battery will last for four hours if the inverter loads on this battery are shed after 30 24

~l l M

r" W

l minutes into the accident. . To comply with the NUL *C 87 00 Supplemental Questions / Answers the licensee needs to assess the battery capacity to verify that all normal battery backed-monitoring and electrical system controls are operational during-the entire duration of an SB0 event and subsequent recovery.

.i

'4. Compressed Air lI  !

'I The licensee stated that the diesel-driven auxiliary feedwater  ;

pump will be used in conjunction with the manual operation of the feedwater and steam release valves to control = the decay heat removal process. Since the AAC power source does not conform to j-_ the requirements of the SB0 rule,.and it may not power the ADVs, the licensee needs to provide an alternate means for modulating I these valves and simulate the process to train the operators accordingly.

5. Effects of Loss of Ventilation Our review indicates that the AAC power source does not have-sufficient capacity to load all'the proposed equipment in-the ,

l blacked-out unit (see item 2 above). The licensee needs to verify that the operability of SB0 equipment will not be degraded, and to l 4 perform a committed analysis to evaluate the effect of loss of j component cooling water flow on the integrity of the~ containment .

l e penetrations. -l 1 ,

1 L 6. Containment . Isolation Our review indicates that the AAC power is not capable of powering all the proposed loads in the blacked out unit. Therefore, the licensee needs to verify that appropriate containment integrity

_b c:n be assured during an SB0 event.

25 j-

n

7. Quality' Assurance and Technical specifications s

-s The licensee needs to verify that-the SB0 equipment is covered by.

l- an appropriate QA and technical specification program consistent

.with'the guidance of RG 1.155 Appendices A and B.

8. Brid-Related Instability .

Lt -

We are aware of two events where a reactor trip at the La Salle site caused momentary voltage fluctuation resulting in an ESF actuation at Braidwood. Both of these events point to a dynamic.

dependence between these two sites. The licensee needs to verify that these events are not symptomatic of underlying or growing 3

grid instability between these two sites. Depending on the outt.ome of this-evaluation the offsite power characteristic of the site could be affected, i.e. a change from "P1" to "P3."

1 1 '

wi =

4

~

26 l-

II 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, i 2. - U.S. Nuclear Regulatory Comission, " Evaluation of Station Blackout Accidents at Nuclear Power Plants - Techni.s1 Findings Related to l Unresolved Safety Issue A 44," NUREG-1032, Baranowsky, P.W., June.1988. ,

t

3. U.S. Nuclear Regulatory Comission, " Collection and Evaluation of T-Complete and Partial losses of Offsite Pwer at Nuclear Power Plants,"

NUREG/CR-3992, February 1985.

L

4. U.S. Nuclear Regulatory Comission, " Reliability of Emergency AC Power -

System at Nuclear Power Plants," NUREG/CR-2989, July 1983.

E 5. U.S. Nuclear Regulatory Commission, " Emergency Diesel Generator i Operating 2xperience, 1981-1983," NUREG/CR-4347, December 1985.

l-

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 Regulatory i L Research, " Regulatory Guide 1.155 S'ation Blackout," August 1988. ~

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

g

9. Nuc1 ear Safety Analysis Center, "The Reliability of Emergency Diesel 4 Generators at U.S. Nuclear Power Plants," NSAC-108, Wyckoff, H.,

September 1986.

L

10. Richter, M. H., letter to T. E. Murley, Director of Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission, "Dresden Station 27 1

--v

fg)

! Units 2 and 3, Quad Cities Station Units 1 and 2, Zion Station Units 1 and 2, La Salle County Station Units 1 and 2, Byron Station Units'1 and 2, Braidwood Station Units 1 and 2, Response to Station Blackout Rule,

( NRC Docket Nos. 50-237/249, 50-254/265, 50 254/265, 50-295/304, 50-

-373/374, 50-454/455, and 50 456/457," dated April 17, 1989.

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

i 12. Byron /Braidwood Stations, Updated Final Safety Analysis Report.

1 g 13 Thadani, A. C., letter to A. Marion of NUMARC, " Publicly-Noticed Meeting L December 27, 1989," dated January 3, 1990 (confirming "NUMARC 87-00 Supplemental Questions / Answers," December 27,1989).

l

14. Rosa, F., letter to Duquesne Light Company - 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 6,

! 1990.

I

15. ACI 349-80, " Code Requirements for Nuclear Safety Related Concrete Structures (ACI 349-80)," American Concrete Institute, Dated April 1981.

16. Richter, M. H., letter to T. E. Murley of U.S. Nuclear Regulatory

[ Commission, "Dresden Station Units 2 and 3, Quad Cities Station Units 1 and 2, Zion Station Units 1 and 2, La Salle County Station Units 1 and 2, Byron Station Units 1 and 2, Braidwood Station Units 1 and 2, Supplemental Response to Station Blackout Rule, NRC Docket Nos. 50-237/249,50-254/265,50-254/265,50-295/304,50-373/374,50-454/455,and 50-456/457," dated March 30, 1989.

17. 'Schuster, T. K., letter to T. E. Murley of U.S. Nuclear Regulatory Commission, " Byron Station Units 1 and 2, Braidwood Station Units 1 and 28

.O.

t[

[. 2. Response to Station Blackout (SBO) Questions, NRC Docket Nos. 50-454/455 and 50-456/457,"-dated March 29, 1990.

i. 18. Tam P. S., Memorandum for, " Daily Highlight- Forthcoming Meeting with

, NUMARC on Station Blackout (SBO) Issues (TAC 40577)," (provi('ing a Draft

[ Staff Position Regarding Use of Emergency AC Power Sources (EDGs) as Alternate AC (AAC) Power Sources, dated April 24,1990), dated April 25, I 1990.-

~

19. Russell, W. T., letter to W. Rasin of NUMARC, " STATION BLACK 0UT," dated.

O June 6, 1990.

7

]. ,

a q .i i

o a

s 29