ML20052H057
| ML20052H057 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 05/12/1982 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 8205190325 | |
| Download: ML20052H057 (69) | |
Text
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
e 4.
s.
4 s
u h
t MAY 121982 m'*
A Docket Hos.: 50-413/414 D
'b O
Mr. William 0. Parker, Jr.
f
/
Vice President - Steam Production s
/
M P.O. Box 33189 Charlotte, North Carolina 28242
Dear Mr. Parker:
Subjec.t: Transmittal of Preliminary Draft SERs - Catawba Nuclear Station Enclosed for your review and comment are the preliminary draft SERs for the following areas:
1.
Thermal-Hydraulic Design (Enclosure 1) 2.
Fuel System and Nuclear Design (Enclosure 2) 3.
Fire Protection (Enclosure 3) 4.
Safe Shutdown for Fires, Appendix R (Enclosure 4) 5.
Pump and Yalve Operability Assurance (Enclosure 5) 6.
Seismic and Dynamic Qualification of Seisnic Category I Mechanical and Electrical Equipment (Enclosure 6)
Your attention is directed in particular to any open items contained within these preliminary drafts. A principal objective of this transnittal is to provide for tinely identification and resolution of any additional analysis, missing information, clarifications or other work necessary to resolve these items. Please contact the Staff's Project Manager, Kahtan Jabbour, regarding the need for any meetings and telephone conferences to this end.
Your comments, including schedules for completion of any further analyses or other work associated with resolution of the items requiring further evaluation, are requested within four weeks of this letter.
i l
Sincerely, 1
Original signed by Robert L. Tedesco Robert L. Tedesco, Assistant Director 8205190325 820512 for Licensing PDR ADOCK 05000413 Division of Licensing E
Enclosures:
- See Previous White
^* "#"
...g{LB#4 DL. B,.#.4
.. :.DL l omer)
)L. :.LB.#.4......
LA:DL.
!**=>
cc: See next page em)
- 5/6/82 5/7.../82 5/ /..f../82 5//}ydesco i
/82 sac row 3is oo,eo,wwcu e24o OFFICIAL RECORD COPY 4 - w32n2
.Q, o, y
- 1 DISTRIBUTION Docket File 50-413/414 i
NRC PDR L PDR CBerlinger TIC VBenaroya TERA ZRosztoczy LB#4 reading file Docket Nos.: 50-413/414 KJabbour MDuncan EAdensam RTedesco Attorney, OELD Hr. William O. Parker, Jr.
I&E Vice President - Steam Production JKramer P.O. Box 33189 RVollmer Charlotte, North Carolina 28242 RMattson RHartfield, MPA
Dear Hr. Parker:
ACRS (16)
Subject:
Transmittal of Preliminary Draft SERs - Catawba Nuclear Stati Enclosed for your review and comment are the preliminary draft S for the following areas:
1.
Thermal-Hydraulic Design (Enclosure 1) 2.
Fuel Systen and Nuclear Design (Enclosure 2) 3.
Fire Protection (Enclosure 3) 4.
Safe Shutdown for Fires, Appendix R (En osure 4) 5.
Punp and Yalve Operability Assurance nclosure 5) 6.
Seismic and Dynamic Qualification Seismic Category I Mechanical and Electrical Equip t (Enclosure 6)
In addition. Enclosures 4 and 5 con n items requiring further evaluation in the Auxiliary Systens and Mater als Engineering areas. A principal objective of this transnittal i to pec. de for timely identification and resolution of any additio analysis, missing information, clarifications or other work necessary to solve these items. Please contact the Staff's Project Manager, Kahtan J our, regarding the need for any meetings and telephone conferences t this end.
Your comments, incl ng schedules for completion of any further analyses or other work asso ated with resolution of the items requiring further evaluation, are quested within four weeks of this letter.
Sincere',y, Robert L. Tedesco, Assistant Director for Licensing Division of Licensing
Enclosures:
DL,;LB#4,j(y, LA:DL:LB#4 DL:LB,#4,
AD : L,/,DL,,,
"'c'>
See n7xtpage
.x m b ur.;e... @uncan....
EAdensam..
RTedesco,
9 sz. 6.,/82,,, st
./82,,.,,s/.,/82,
st.../82,.
NRC FORM 318 410,80t N RCM O2d C OFFICIAL RECORD COPY e m aw-32n2.
ENCLOSURE 1 CATAWBA UNIT 1 AND 2 Draft Safety Evaluation Report 4.4 Themal-Hydraulic Design The Catawba thermal-hydraulic design analyses were perfomed using the same analytical tools that are described in Section 4.4 of WCAP-9500. This methodology is known as the Improved Themal Design Procedure (ITDP) and a dilscription of thi: analysis technique is given in WCAP-8567.
The two major compor: ants of the ITDP are the WRB-1 Critical Heat Flux (CHF) correlation and the THINC-IV computer code. WCAP-8762 contains infomation on the development of the WRB-1 correlation while WCAP-9401 discusses the applicability of the correlation to the 17x17 Optimized Fuel Assembly (OFA).
A description of the THINC-IV program is given in WCAP-7956 and the design application of the code is presented in WCAP-8054.
All of the topical reports listed above have been reviewed and approved by the staff, therefore, the use of these documents in the Catawba themal hydraulic design is acceptable. Our review of Section 4.4 of the Catawba Final Safety Analysis Report (FSAR) was limited to deviations between the Catawba design and the WCAP-9500 design and to the plant specific information requirements identified in our approval of WCAP-8567.
In particular the following plant-specific infomation was required from the applicant:
1.
The plant-specific requirements listed in our safety evaluation reports on WCAP-8567 and WCAP-9500.
2.
Plant specific or generic margins used to offset the reduction in the departure from nucleate boiling ratio due to rod bcw.
3.
A description of the proposed Loose Parts Monitoring System used by the applicant.
4 A description of the instrumentation and the procedures needed to detect crud build-up.
=
a
. 5.
The documentation required by Item II.F.2 of NUREG-0737.
Any limitations to the Catawba core design which may result from the staff generic review of thermal-hydraulic stability will be accounted for by appropriate operating restrictions; hcvever, none are anticipated and nothing further on this subject is needed from the applicant.
4.4.1 Departure from Nucleate Boiling l
The use of the Westinghouse ITDP results in two different departure from j
nucleate boiling ratio (DNBR) limits. The initial DNBR limit accounts l
for uncertainties associated with the CHF correlation and DNB test procedures while the second limit accounts for uncertainties inherent in the ITDP.
Based on the information given in WCAP-8762 and WCAP-9401, the applicant has proposed a DNBR limit of 1.17 on the WRB-1 correlation. Since the staff has reviewed and approved WCAP-8762 and WCAP-9401, we conclude that the use of the WRB-1 correlation with a minimum DNBR of 1.17 is acceptable in the thermal-hydraulic design of the Catawba units.
The addition of the ITDP DNBR uncertainty increases the Catawba limiting DNBR, to 1.31 for a thimble cell and 1.33 for a typical cell. These values will serve as the basis of the Technical Specifications. Before the staff can determine the acceptability of the limiting DNBRs, we will require justification for the uncertainties, variances and distributions used in the ITDP in accordance with our SER for WCAP-8567.
The applicant has proposed a third and optional DNBR limit against which the results of transients are checked. These minimum DNBR values are 1.47 for a thimble cell and 1.49 for a typical cell. The thermal margin available between the plant-specific limits,1.47 and 1.49, and the DNBR design values, 1.31 and 1.33, can be used for flexibility in the operation and design analyses of future cores for the Catawba units.
t s
' A reload review of a pressurized water reactor, not of Westinghouse design, showed that the input parameters used in the themal-hydraulic analysis of the initial core did not bound future cycles. The Catawba design methodology is to select value; of input parameters which are expected to bound future cycles. When all of the reload related parameters for a given accident are bounded, the reference safety analysis is valfd; however, if a parameter is not bounded further evaluation is necessary. The staff concludes that this design approach is acceptable.
The staff also concludes that the use of a unifom core-exit pressure gradient is acceptable in the themal-hydraulic design of the Catawba units. This conclusion is based on a sensitivity study perfomed by Westinghouse, (Eicheldinger, November 2,1977) which showed that the effects of a non-unifom pressure distribution on the minimum DNBR are negligible.
l Based on our review of the information given in Section 4.4 of the Catawba FSAR, the staff concludes that the themal-hydraulic design methodology used by the applicant is acceptable; however, before the staff can determine the acceptability of the Catawba DNBR design values we will require the infomation needed to justify the uncertainties, variances, and distributions used in the ITDP.
4.4.2 Fuel Rod Bowing A significant parameter which affects the themal-hydraulic design of the core is fuel rod bowing within fuel assemblies. The Westinghouse methods for predicting the effects of rod bow on DNB are given in WCAP-8691, Revision 1, and are under review by the staff. The magnitude of rod bow as a function of burnup was evaluated based on interim methods which have been approved by the staff (Ross and Eisenhut, December 8,1976; Ross and Eisenhut, February 16, 1977; Meyer, March 2, 1978). The resultant reduction in DNBR due to rod bow is given in Table 4.4-1.
--e
,_,.n,
l Burnup DNBR Penalty l
(mwd /MTU)
(%)
0 0
3500 0
l 5000 0
10000 2.15 1
1 15000 4.64 20000 6.74 l
25000 8.59 30000 10.27 35000 13.07 40000 19.09 These values were presented to the applicant in a Regulatory Staff Position, (RSP).
In response to our RSP the applicant presented a relationship for determining the amount of thermal margin available to offset the DNBR reduction brsed on the difference between the plant-specific DNBR limit and the design value DNBR. This relationship is:
l Amount of Margin = (Plant-Specific DNBR Limit)-(Design Value DNBR)
(Plant-Specific DNBR Limit) i Applying this relationsiaip results in 10.9% margin for a thimble cell and j
10.7% margin for a typical cell.
For a region average burnup of 33,000 MWD /MTU the applicant calculated a gap closure of 84% and DNBR reductions of 11.1% and 13.6% for full flow conditions and a loss-of flow transient. Westinghouse does not consider the effects of rod bow for region average burnups greater than 33,000 MWD /MTU N
since beyond this 5urnup, F burndown effects preclude the fuel from 3H achieving the limiting value of F5H.
e a
a
' Using the available thermal margin and the DNBR reductions calculated at a burnup of 33,000 MWD /MTU the applicant has proposed an adjusted rod bow penalty of 0.4% for full flow conditions and 2.9% for the loss of flow transient. The applicant further stated that sufficient plant margin exists to offset the rod bow penalty for a loss-of flow transient.
The staff has reviewed the approach presented by the applicant and concludes that it is an acceptable means of offsetting the effects of rod bow. However, verification calculations performed by the staff, yielded different values of gap closure and DNBR reduction. For a burnup of 33,000 MWD /MTU the staff calculated a gap closure of 84.36% and a resultant DNBR penalty of 11.2%
and 13.74% for the two flow conditions. Therefore, our adjusted rod bow penalties are 0.5% and 3.04% for the full flow condition and the loss-of flow transient. The staff will ensure that the 10.95 margin for a thimble cell and the 10.7% margin for a typical cell are contained in the basis of the Technical Specifications and are incorporated into plant parameters as specified in the Technical Specifications. Additional plant specific margins used to offset the remaining rod bow penalty must also be included.
I 4.4.3 Instrumentation 4.4.3.1 Loose Parts Monitoring System The applicant has provided a description of the Loose Parts Monitoring System (LPMS) which will be used for the Catawba units. The design will include two transducers on the lower and upper head of the reactor vessel and one trans-ducer on the lower head of each steam generator. The LPMS will be capable of detecting a loose part having a kinetic energy of 0.05 ft-lb and the initial alarm setting will correspond to a 0.5 f t-lb impact. The staff has reviewed the applicant's LPMS and requested additional information on this system. The applicant has provided the information requested and it is presently under l
staff review. A preliminary assessment is that the LPMS is not in conformance with Regulatory Guide 1.133.
In addition to the information already supplied the applicant should commit to provide a final design report which contains the following:
i e-w--
9
,m,,,,-,.,w--
--,-e---
,y
,m,,
6-1.
An evaluation of the LPMS for confomance to Regulatory Guide 1.133, 2.
A descripticn of the system hardware, operation and implementation of the loose parts detection program after start-up testing. This should also include the baseline data and alarm settings.
3.
A description and evaluation of diagnostic procedures used to confim the presence of a locse part.
4.
A description of the operator training program.
Finally, it should be noted that the applicant's descriptfor of the LPMS was supplied in Section 7.8.8 of the FSAR. Regulatory Guides 1.133 and 1.70 state that the loose parts detection program should be described in Section 4.4.6 " Instrumentation Requirements."
4.4.3.2 Instrumentation for Detection of Crud Crud deposition in the core and an associated change in core pressure drop and flow have been observed in some pressurized water reactors, not of Westinghouse design. The staff requested that the applicant provide a description of the flow measurement capability of the Catawba units as well as a description of the procedures to detect flow degradation.
l The applicant responded that except for steam generator tube plugging, there have been no reports of a significant flow reduction in a relatively short period of time at any Westinghouse plant.
The staff will ensure that the Catawba Technical Specifications contain the requirement that the actual reactor coolant system (RCS) flow rate be verified to be greater than or equal to the minimum design flow rate at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
In addition, the applicant has stated that a calorimetric measure-ment will be perfomed at least once a month to detemine if the total RCS I
flow rate is within the region of acceptable operation.
8
.- n
,..--.----.-_.,.,--,,-,,,,,,--c-
-,, w,--.------ -
-e-,-------.----,.-.--g.
, -,- n
We will insure that this provision and a provision to perform a channel calibration using the colorimetric measurement technique at least once per 18 months are also included in the Technical Specifications.
With inclusion of the above requirements into the Technical Specifications, the staff concludes that our concerns en crud deposits in the core have been ddequately addressed.
4.4.3.3 Flow Measurement Uncertcinty In response to our question on crud deposits in the core the applicant provided a description of the calorimetric flow measurement system which will be used by the Catawba units.
Individual coolant loop flow is calculated using the steam generator output, corrected for pump heat input, and the enthalpy rise of the coolant. The total flow rate is then calculated by summing the individual flows. This flow rate is then used to calibrate the flow measured by the elbow taps located in each coolant loop.
By using a statistical error combination technique the applicant has proposed a total uncertainty of 1.5% on the calorimetric measurement. The staff has raised concerns on the values of the components comprising the calorimetric flow measurement uncertainties and how drift associated with the elbow taps is accounted for between channel calibrations.
We have brought these concerns to the attention of the applicant and are awaiting a response. Prior to issuance of our formal SER, the applicant should supply acceptable responses to these concerns in order to avoid impacting the ifcensing schedule.
4.4.3.4 ICC Instrumentation The applicant has submitted information in response to Item II.F.2 "Instrumen-tation for the Detection of Inadequate Core Cooling," of NUREG-0737. The submittal provides only partial information on the subcooling monitors and has no information on core exit thermocouples and reactor vessel level measurement.
f, The staff has reviewed the applicant's submittal and has found it incomplete.
We will require that the applicant provide the documentation itemized in Item II.F.2 of NUREG-0737. Acceptable documentation must be provided and approved prior to issuance of an operating license.
4.4.4 N-1 Loop Operation In response to a staff question regarding N-1 loop operation when oncreactor coolant loop is ouk of service, leaving only three loops available to supply to the core, the applicant stated that it does not wish to exercise the option to operate in the N-1 mode. The staff will require that Technical Specifications include appropriate provisions to ensure that this type of operation is prohibited.
4.4.5 Thennal-Hydraulic Comparison The thermal-hydraulic design parameters for the Catawba units are listed in Table 4.4-2 and these design values are compared with those for D. C.
Cook Unit 2.
The D. C. Cook design has been previously reviewed and approved by the staff.
The Catawba units are designed to operate at a higher total flow rate but have less effective flow for heat transfer, This is because the Catawba units are upper head injection (UHI) plants. The amount of bypass flow required for UHI plants is 7.5% while non-UHI plants, D. C. Cook, require bypass flow between 4.5% to 5.8%.
Additional differences between the two plants are a decrease in the active heat transfer area and a higher average and maximurn heat flux for the Catawba design. This increased heat flux coupled with a higher nominal inlet tempera-ture results in a decrease in the minimum DNBR. Therefore, the net change is a decrease in the thermal margin available to Catawba.
,-.y-.-e
~
i
-g-Both Catawba and D. C. Cook were designed using the ITDP. The similarities in the Catawba and D. C. Cook Unit 2 designs support the conclusion that the Catawba thermal-hydraulic design is acceptable.
4.4.6 Summary and Conclusion The Catawba thermal-hydraulic design was reviewed according to Section 4.4 of the Standard Review Plan (NUREG-0800). The scope of our review included the design criteria, core design, and the steady-state analysis of the core thermal-hydraulic performance. The review concentrated on the differences between the proposed design and those previously approved by the staff. The applicant's thermal-hydraulic analyses were performed using analytical methods and correlations that have been previously accepted.
Before the staff can determine if the core is designed with appropriate margin to ensure that it meets the requirements of General Design Criterion 10, 10 CFR Part 50, we will require justification of the variances, uncertainties, and distributions used in the ITDP. Also, the applicant should respond to our questions on the flow measurement uncertainties and effects of instrument drift on these uncertainties and supply the documentation itemized in Item II.F.2 of NUREG-0737.
Prompt and satisfactory responses will be needed to avoid impact on the licensing schedule.
i l
I l
~
a
- Table 4.4-2 Design Comparison Design Parameter Catawba D. C. Cook Unit 2 1.
Performance Characteristics:
Fuel Assembly Design 17x17 0FA 17x17 Reactor Core Heat Output (Mwt) 3411 3391 Nominal System Pressure, psia 2280 2280 Nominal Minimum DNBR Typical Cell 2.40 3.03 Thimble Cell 2.26 2.70 Critical Heat Flux Correlation WRB-1 WRB-1 Design DNBR for Design Transient Typical Cell 1.49 1.80 Thimble Cell 1.47 1.77 II.
Coolant Flow:
6 Total Flow Rate (10 lb/hr) 143.3 142.7 Effective Flow Rate for 6
Heat Transfer (10 lb/hr) 134.7 136.3 Average Velocity Along Fuel l
Rods (ft/s) 15.8 16.7 2
Effective Core Flow Area (ft )
54.1 51.1 I
l
. III. Coolant Tmperature, *F Nominal Reactor Inlet 561.6 541.3 Average Rise in Core 61.8 63.4 Pressure Drop Across 25.722.6 23.3 2.3 the Core, psia IV.
Heat Transfer,100 Percent Power Active Heat Transfer Surface 57,500 59,700 2
Area (ft )
2 Average Heat Flux (Btu hr-ft )
197,200 188,700 2
Maximum Heat Flux (Btu /hr-ft )
457,500 437,800 Average Linear Heat Rate (kW/ft) 5.44 5.41 Peak Linear Power resulting from 12.6 12.6 overpower trqpsients/ operator Errors (kW/ft) r i
o l
4'.4.7 References 4.4.7.1 NRC Reports NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.
NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," July 1981.
4.4.7.2 Westinghouse Reports WCAP-7956, "THINC-IV -- An Improved Program for Thermal-Hydraulic Analysis of Rod Bundle Cores," June 1973.
WCAP-8054, L. E. Hochreiter, " Application of the THINC-IV Program to PWR Design," September 1973.
WCAP-8567, H. Chelemer, et al., " Improved Thermal Design Procedure," July 1979.
WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation," July 1979, J. Skaritka, et al.,
WCAP-8762, "New Westinghouse Correlation WRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," July 1976.
WCAP-9401-P-A, M. D. Beaumont, et al., " Verification Testing and Analysis of the 17x17 Optimized Fuel Assembly," August 1981.
WCAP-9500, " Reference Core Report 17x17 Optimized Fuel Assembly," originally submitted July 30, 1979.
)
4'.4.7.3 Other Reference C. Eicheldinger (Westinghouse) letter to John F. Stolz (NRC), untitled letter on core exit pressure gradients, November 2,1977.
R. O. Meyer (NRC) memorandum to D. F. Ross, " Revised Coefficients for Interim Rod Bowing Analysis," March 2,1978 (Proprietary Infonnation, not publicly avail able).
D. F. Ross and D. G. Eisenhut (NRC) renorandum to D. B. Yassallo and K. R.
Goller, " Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing in Themal Margin Calculations for Light Water Reactors," December 8,1976.
D. F. Ross and D. G. Eisenhut (NRC) memorandum to D. B. Vassallo and K. R.
Goller, " Revised Interim Safety Evaluation Report on the Effects of Fuel Rod Bowing in Themal Margin Calculations for Light Water Reactors,"
February 16, 1977.
L
ENCLOSURE 2 l'
4.2 Fuel System Design The Catawba Fuel Assembly Design aescribed in the FSAR is a 17x17 array of fuel rods having a diameter of 0.360 inches. This design appears to be the so-called Optimized Fuel Assembly (0FA) by Westinghouse and has been generically described in their Topical Report WCAP-9500, " Reference Core Report 17x17 Optimized Fuel Assembly." This topical report was reviewed and judged acceptable as a generic reference for the 0FA (Ref.1). Since the applicant has not used WCAP-9500 as a reference fo' r describing the Catawba fuel assembly, the applicant must confirm that the Catawba fuel assemby is the same as described in WCAP-9500. Assuming that the Catawba fuel assembly is the OFA, it is, therefore, generally acceptable. The applicant, however, must supply the following plant-specific information which was not supplied in WCAP-9500:
1.
Confirmation that the predicted cladding collapse time exceeds the expected lifetime of the fuel.
2.
Supplemental ECCS c.11culations using NRC-supplied LOCA cladding models.
3.
A determination that the appropriate seismic-and-LOCA forces are bounded by the cases considered in WCAP-9401.
4.
A description of plans for on-line fuel system monitoring.
5.
A description of plans for post-irradiation poolside surveillance of fuel.
Surveillance of the new Hafnium control rods is not required for Catawba unless surveillance of similar rods at Comanche Peak and the first of the SNUPPS units indicates that additional surveillance is warranted.
y 4.3 Nuclear Design The Catawba Units 1 and 2 power plants have a reactor based on the optimized fuel assembly design of WCAP-9500, which we recently reviewed. Our review of the Catawba nuclear design was therefore based on information contained in WCAP-9500, the Catawba Final Safety Analysis Report (FSAR), amendments thereto, and the referenced topical reports. Since the results of the review are essentially identical to that provided in Section 4.3 of the WCAP-9500 review, that material will not he repeated here and only a few changes or additions to that report will be noted.
4.3.1 Discussion The first cycle length for Catawba is about one-and-one half years.
Catawba Units 1 and 2 will use the improved load follow package. The Constant Axial Offset Control (CAOC) band will be +3 to -12 delta flux difference for this control mode. The analysis perfonned by Westinghouse has indicated that the peaking factor limit cannot be met at BOL of Cycle 1 due to the wide al band. This has resulted in limiting the width of the band to the value of 5% AI until 3000 tiWd/!1TU burnup for Catawba Units 1 and 2.
The 5% AI is the value previously justified by the CAOC analysis. These features will be incorporated in the Catawba Technical Specifications.
4.3.2 Evaluation Findings The Catawba nuclear design was reviewed according to Section 4.3.of the
. Standard Review Plan (NURtG-0800).
The applicant has described the computer programs and calculational techniques used to predict the nuclear characteristics of the reactor design and has provided examples to demonstrate the ability of these 4.3-1
i o
meh. hods to predict experimental results. The staff concludes that the infomation presented adequately demonstrates the ability of these analyses to predict reactivity and physics characteristics of the Catawba plant.
To allow for changes of reactivity due to reactor heatup, changes in operating conditions, fuel burnup, and fission product buildup, a significant amount of excess reactivity is designed into the core.
The applicant has provided substantial information relating to core reactivity requirements for the first cycle and has shown that means have been incorporated into the design to control excess reactivity at all times. The applicant has shown that sufficient control rod worth is available to make the reactor subcritical with an effective
~
multiplication factor no greater than 0.987 in the hot condition at any time during the cycle with the highest-worth control rod stuck in the fully withdrawn position.
On the basis of our review, we conclude that the applicant's assessment of reactivity co,ntrol requirements over the first core cycle is suitably conservative, and that adequate negative worth has been provided by the control system to assure shutdown capability.
Reactivity control require-ments will be reviewed for additional cycles as this infomation becomes available.
The staff concludes that the nuclear design is acceptable and meets the requirements of GDC 10, 11, 12, 13, 20, 25, 26, 27, and 28. This con-clusion is based on the following:
(1) The applicant has met the requirements of GDC 11 with respect to pronpt inherent nuclear feedback characteristics in the power operating range by:
a.
Calculating a negative Doppler coefficient of reactivity, and b.
Using calculational methods that have been found acceptable.
4.3-2
The staff has reviewed the Doppler reactivity coefficients in this case and found them to be suitably conservative.
i (2) The applicant has met the requirements of GDC 12 with respect to power oscillations which could result in. conditions exceeding specified acceptable fuel design limits by:
a.
Showing that such power oscillations are not~ possible and/or can be easily detected and thereby remedied, and b.
Using calculational methods that have been found acceptable.
(3) The applicant has met the requirements of GDC 13 with respect to provision of instrumentation and controls to monitor variables and systems that can affect the fission process by:
a.
Providing instrumentation and systems to monitor the core power distribution, control rod positions and patterns, and other process variables such as temperature and pressure, and b.
Providing suitable alanns and/or control room indications for these monitored variables.
(4) The applicant has met the requirements of GDC 26 with respect to provision of two independent reactivity control systems of different l
designs by:
I a.
-Having a system that can reliably control anticipated operational i
occurrences, b.
Having a systen that can hold the core subcritical under cold conditions, and l
c.
Having a system that can control planned, normal power changes.
i 4.3-3
( 5)' The applicant has met the requirements of GDC 27 with respect to reactivity control systems that have a combined capability in con-junction with poison addition by the emergency core cooling system of reliably controlling reactivity changes under postulated accident conditions by:
a.
Providing a movable control rod system and a liquid poison system, and b.
Performing calculations to demonstrate that the core has sufficient shutdown margin with the highest worth stuck rod.
(6) The applicant has met the requirements of GDC 28 with respect to postulated reactivity accidents by (reviewed under Section 15.4.8):
a.
Meeting the regulatory position in Regulatory Guide 1.77, b.
Meeting the criteria on the capability to cool the core, and c.
Using calculational methods that have been found acceptable for reactivity insertion accidents.
(7) The applicant has met the requirements of GDC 10, 20, and 25 with respect to specified acceptable fuel design limits by providing analyses demonstrating:
a.
That nomal operation, including the effects of anticipated operational occurrences, have met fuel design criteria, b.
That the automatic initiation of the reactivity control system as:ures that fuel design criteria are nut exceeded as a result of anticipated of systems and components important to safety under accident conditions, and c.
That no single malfunction of the reactivity control system causes violation of the fuel system limits.
4.3-4
.y 15.4.1 Uncontrolled Rod Cluster Control Assembly (Rod) Bank Withdrawal From Zero Power Conditions Discussion The consequences of an uncontrolled rod cluster control assembly bank withdrawal at zero power have been analyzed. Such a transient can be caused by a failure of the reactor control or rod control systems. The analysis assumes a conserva-tively small (in absolute magnitude) negative Doppler coefficient and a positive moderator coefficient.
Further, hot zero power initial conditions with the reactor just critical are chosen because they are known to maximize the calcuTated consequences. The reactivity insertion rate is assumed to be equivalent to the simultaneous withdrawal of the two highest worth banks at maximum speed (45 inches per minutes).
Reactor trip is assumed to occur on the low setting of the power-range neutron flux channel at 35 percent of full power (a ten percent uncertainty has been added to the setpoint value). The maximum heat flux is much less than the full-power value and average fuel temperature increases to a value lower than the nominal full power value. The minimum DNBR at all times remains above the limiting value (1.47 for the thimble cell and 1.49 for the typical cell with the WRB-1 DNB correlation).
Evaluation Findings We have reviewed this event according to the Standard Review Plan (NUREG-0800).
We have reviewed the reactivity worths and reactivity coefficients used in the analysis and conclude that conservative values have been used. We have reviewed the calculated consequences of this design transient and conclude that they are acceptable.
We, therefore, find that the requirements of General Design Criterion 20, which requires that protection be automatically initiated, and Criterion 25, which requires that a single failure of the protection system does not result in violation of specified fuel design limits, have been satisfied.
15-1
15.4.2 Uncontrolled Rod Cluster Control Assembly (Rod) Bank Withdrawal at Power Discussion The consequences of uncontrolled withdrawal of a rod bank in the power operating range have been analyzed. The effect of such an event is an increase in coolant temperature (due to the core-turbine power mismatch) which must be terminated prior to exceeding fuel design limits.
The analysis is performed as a function of reactivity insertion rates, reactivity feedback coefficients, and core power level. Protection is provided by the high neutron flux trip, the overtemperature AT and overpower AT trips, and pressurizer pressure and pressurizer water level trips.
In no case does the departure from nucleate boiling ratio fall below the limiting value (see 15.4.1).
Adequate fuel cooling is therefore maintained. The maximum power reached includ-ing uncertainties is 118 percent of full power, thus precluding fuel centerline mel ting.
Evaluation Findings We have reviewed this event according to the Standard Review Plan (HUREG-0800).
The basis for acceptance in the staff review is that the applicant's analysis method has been reviewed and approved, the input parameter have been found to be suitably conservative, and the results show that no fuel damage occurs.
The staff concludes that the calculations contain sufficient conservatism with respect to input assumptions and models to assure that fuel damage will not result from control rod withdrawal errors. The staff further concludes that the requirements of General Design Criteria 20 and 25 have been met.
15.4.3. Rod Cluster Control Assembly Malfunctions Discussion Rod cluster control assembly misalignment incidents including a dropped full length assembly, a dropped full-length bank, a misaligned full length assembly 15-2
and the withdrawal of a single assembly while operating a power have been analyzed by the applicant. Misaligned rods are detectable by:
(1) asymmetric power distributions sensed by excore nuclear instrumentation or core exit thermocouples, (2) rod deviation alarm, and (3) rod position indicators. A deviation of a rod from its bank by about 15 inches or twice the resolution of the rod position indicator will not cause power distributions"to exceed design limits. Additional surveillance will be required to assure rod alignment if one or more rod position channels are out of service.
In the event of a dropped assembly or group of assemblies the reactor will typically scram on a neutron flux negative rate trip, and analysis indicates that thermal limits will not be exceeded for the event.
If the rod locations' are such that the reactor does not scram, however, the automatic controller may return the reactor to full power and with a single failure the control could result in a power overshoot.
It is anticipated that a detailed analysis will show that if this occurs thennal limits will not be exceeded. However, that analysis has not been submitted as yet and it is thus assumed that departure from nucleate boiling could occur. The staff has accepted an interim position for operating reactors which consists of a restriction on operations above ninety percent power such that either the reactor is in manual control or rods are required to be out greater than 215 steps. This restriction will be applied to Catawba Units 1 and 2.
With this restriction thermal limits will not be exceeded. Approval of the analysis will result in removing the restriction.
For cases where a group is inserted to its insertion limit with a single rod in the group stuck in the fully withdrawn position, analysis indicates that departure from nucleate boiling will not occur. We have reviewed the calculated estimates of the expected reactivity and power distribution changes that accompany postulated misalignments of representative assemblies. We have concluded that the values used in this analysis conservatively bound the expected values including calculational uncertainties.
The inadvertent withdrawal of a single assembly requires multiple failures in the control system, multiple operator errors or deliberate operator actions combined with a single failure of the control system. As a result the single assembly withdrawal is classified as an infrequent occurrence. The resulting 15-3 w'
m w
transient is similar to that due to a bank withdrawal but the increased peaking factor may cause departure from nucleate boiling to occur in the region surround-ing the withdrawn assembly. Less than five percent. of the rods in the core experience departure from nucleate boiling for such a transient.
Evaluation Findings We have reviewed this event according to the Standard Review Plan (NUREG-0800).
We conclude that the analysis and calculated consequences of rod control cluster assembly malfunction are acceptable because they lead to no fuel damage, except for the inadvertent withdrawal of a single assembly for which the very limited DNB is acceptable for a fault of infrequent occurrence.
15.4.7 Inadvertent Loading of a Fuel Assembly into Improper Position Discussion Strict administrative controls in the fonn of previously approved established procedures and startup testing are followed during fuel loadings to prevent operation with a fuel assembly in an improper location or a misloaded burnable poison assembly.
Nevertheless, an analysis of the consequences of a loading error has been perfomed.
Comparisons of power distributions calculated for the nominal fuel loading pattern and those calculated for five loadings with misplaced fuel assemblies or burnable poison assemblies are presented by the applicant. The selected non-nomal loadings represent the spectrum of potential inadvertent fuel misplacement. Calculations included, in particular, the power in assemblies which contain provisions for monitoring with incore detectors.
As part-of the required startup testing, the incore detector system is used to detect misloaded fuel prior to operating at power. The analysis described above shows that all but one of the above misloading events would be detected by this test.
In the excepted case, an interchange of Region 1 and 2 assemblies near the center of the core, the increase in the power peaking is approximately equal to the uncertainty in the measurement of this quantity (5 percent). This 15-4
l
.?
uncertainty is allowed fer in analyses so that this misloading event does not result in unacceptable consequences.
Evaluation Findings We have reviewed this event according to the Standard Review Plan (NUREG-0800).
Based on our review of the analyses described above, we conclude that an improperly loaded fuel assembly or burnable poison assembly that could cause a significant safety problem would be detected by the operator with the instrumen-tation provided.
This satisfies the requirements of the Standard Review Plan Section 15.4.7 which
~
requires that any misloading that cannot be detected by the instrumentation provided in the core causes, when the core is operated in the normal mode, no fuel damage in excess of that for which offsite consequences will be more than a small fraction of 10 CFR Part 100 guidelines.
15.4.8 Rupture of a Control Rod Drive Mechanism Housing (Rod Cluster Control Assembly Ejection)
Discussion The mechanical failure of a control rod mechanism pressure housing would result in the ejection of a rod cluster control assembly.
For assemblies initially inserted, the consequences would be a rapid reactivity insertion together with an adverse core power distribution, possibly leading to localized fuel rod damage.
Although mechenical provisions have been made to make this accident extremely Onlikely, 'the applicant has analyzed the consequences of such an event.
' Methods used in the analysis are reported in WCAP-7588, Revision 1, "An Evaluation o' the Rod Ejection Accident in Westinghouse Reactors Using Spatial Kinetics Methods," which has been reviewed and accepted by the staff. This report demonstrated that the model used in the accident analysis is conservative relative to a three dimensional kinetics calculation.
15-5
Th'e applicant's criteria for gross damage of fuei are a maximun clad temperature of 2700 degrees Fahrenheit and an energy deposition of 200 calories per gram in the hottest pellet. These criteria are more conservative
- than those proposed in Regulatory Guide 1.77.
Therefore, they are acceptable.
Four cases were analyzed: beginning-of-cycle at 102 percent and zero power and end-of-cycle at 102 percent and zero power. The highest clad temperature, 7
2597 degrees Fahrenheit was reached in the zero-power beginning-of-cycle case and the highest fuel enthalpy,180 calories per gram was reached in the beginning-of-cycle full-power case. The analysis also shows that less than 10 percent of the fuel experiences departure from nucleate boiling and less than 10 percent of the hot pellet melts. Analyses have been perfonned to show that the pressure surge produced by the rod ejected is mild and will not approach the Reactor Coolant System emergency limits.
Further analyses have shown that a cascade effect, i.e., the ejection of a further rod due to the ejection of the first one, is not credible.
Evaluation Findings We have reviewed this event according to the Standard Review Plan (NUREG-0800).
The ejected rod worths and reactivity coefficients used in the analysis have been reviewed and have been judged to be conservative.
Also the assumptions and methods of analysis used by the applicant are in accordance with or are more conservative than those recommended in the Regulatory Guide 1.77.
Therefore, we conclude that this analysis is acceptable. The analysis of the dose conse-quences for this design-basis accident is described elsewhere in this Safety Evaluation Report.
We further conclude that the requirements of General Design Criterion 28 are met for this desigt. basis event.
Criterion 28 requires that reactivity insertion rates be limited so as to preclude extensive damage to the pres 3ure boundary and preclude non-coolability of the core.
Regulatory Guide 1.77 has an acceptance criterion of 280 calories per gram energy deposition and no criterion for clad temperature other than that implicit in requirements for fuel and pressure vessel damage.
15-6
. ew s'
.?
4 REFERENCES 4
- 1.
Letter from R. L. Tedesco, NRC, to T. M. Anderson, Westinghouse,
" Acceptance for Referencing of Licensing Topical Report WCAP-9500,"
dated May 22, 1981.
e O
e G
I i
e
(
l
Draft Safety Evaluation Report Catawba Nuclear St ation Units 1 and 2 Fire Protection Review Table of Contents I.
Introduction II.
Fire Protection Program Req ui r eme nt s A.
Fi re P rot ect ion P rog ram B.
Fire Hazards Analysis C.
Alternative Shutdown D.
Implementation of Fire Protection Programs III.
Administrative Controls IV.
Fire Brigade and Fire Brigade Training V.
General Plant Guidelines A.
Bui ldi ng Design B.
Safe Shutdown Capability C.
Alternate Shutdown C apabi li ty D.
Control of Combustibles E.
Electrical Cable Combustion, Cab'e Trays, and Cable Penetrations F.
Ventilation G.
Lighting & Communications VI.
Fire Detection
' Suppr ession A.
Fire Detection B.
Fire Protection Water Supply Systems
[
C.
Water Sp r i nk le r & Hose Standpipe Systems l
D.
Carbon Dioxide Suppression System E.
Portable E xt ingui s he rs
O e
1 Table of Contents (continued) '
VII.
Fire Protection f o r Specific Plant Areas A.
Containment 8.
Reactor Buildi ng C.
Control Room D.
Cable Spreading Rooms E.
Switchgear Rooms F.
Safety Related Battery Rooms G.
Diesel Generato r Areas H.
Remote Safety Related Panels I.
Other Areas VIII.
Conclusion i
w I
e we l
.i l
l l
l l
l l
l l
Chemical Engineering Branch /Fi re P rotect ion Sect ion
+
Fi re Protect ion Revi ew Catawba Nuclear St ation Uni ts 1 and 2 Docket Nos. 50-413/414 I.
Introduction We have reviewed the fi re prot ect ion p rogram for conformance with the Standard Review Pla n NUR EG-0 800 S e ct io n 9.5-1, Fire Protection dated July 1981.
This document, in BTP C MEB 9.5-1 incorporates the guidance of Appendix A to Branch Technical Position ASB 9.5-1 and the technical requi rements of Appendix R to 10 CFR 50.
The ap pli c a n t 's Fire P rotect ion Revi ew' t ransmi tted by letter dated December, 1977, with revisions dated June, 1979 and August, 1981, was in response to our request to evaluate their fire p rotect ion program against the guidelines of Appendix A to Branch Technical Position ASB 9.5-1,
" Guidelines for Fire Protection f or Nuclear Power Plants."
The applicant also provided an evaluation against the technical requi r eme nt s of Appendix R to 10 CFR 50 in the revised Fire Protection Review, dated October 23, 1981.
As pa rt of our r evi ew, we will visit the plant site to examine the relationship safety-related components, systems, and structures i-spe ci fic plant areas to bot h c ombu stib le
. ma t e ri als and to a s so ci a t ed fire detection and suppression systems.
The site visit has not been conducted up to this time because the construction of the plant has not progressed to the level where such a visit would be meaningful.
Our review included an evaluation of the automatic and manually ope rated water and gas fire s u p p r e s s i o'n systems, the fire detection systems, fire barriers, fire doors and dampers, fire prot ection administrative controls, and the fire brigade size and training.
The obj e ct ive of our revi ew is to ensure that in the event of a fire, pe rso nnel and plant equipment would be adequate to safety shut down the reactor, w
to maintain the plant in a safe shutdown condition, and to minimize the release of radioactive material to the envi ron-ment.
Because units 1 and 2 are of the same design, except as noted, the comments made in this report apply to both units.
II.
Fire P rot ection Requi rement s A.
Fire Protection Program The bases for the f ire prot ect ion p rogram i s desc ribed in the applicant's Fire Protection R e vi ew.
The description includes the protection of structures, systems, and components impo rt ant to safety.
The applicant states that fire protect ion policy and the fire protection program "will be d e li ne a t ed " in the
s 3-i St ation Fi re Pla n, without providing sufficient inf o rma tioa to verify compli ance wi th our guideli nes.
We will requi re that the applicant commit to comply with the guidelines in BTP CMEB 9.5-1 I t e m C.1. a in the development of the Station Fi re Pla n.
8.
Fire Haza rd Analyses The appli c a nt provided a fire hazard analysis with the Fire Protect ion Revi ew.
The analyses speci fied the combustible ma te ri als present in fire areas, identified safety related equipment, determined the consequences of a fire on safe shutdown capability, and summa rized available fire protection in accordance with BTP CMEB 9.5-1, Item C.1.b.
Our evaluation of the identified fire hazards is contained i n the balance of this report.
C.
Alternate Shutdown The applicant will ins tall a dedicated S t and by Shutdown System for the plant.
This capability is evaluated in Section V.C of this report.
D.
Implementation of Fire Protection Program The fire p rot ect ion p rogram f or bot h uni ts should be operational before initial fuel loading.
III.
Administrative Controls l
The administrative controls for fire protection consists of the fire protection program a nd orga niza tion, the fire brigade training, the controls over combu s t ib le s and ignition sources, i
l the prefire plans and p rocedures for fighting fires, and l
1 4
quality assurance.
The appli cant has stated in the Fire Protection Review that these controls will be delineated in "St ation Direct ives" wi thout providing sufficient inf ormation to ve ri f y compli ance wi th our guidelines.
We will require that the ap p li c a nt commit to follow the guidelines in BTP CMEB 9.5-1, Item C.2 r ega rdi ng administrative control in the f o rmulation of the fire protection station directives.
IV.
Fire Brigade and Fire Brigade Training The ap pli c a nt, has not provided description of the pla nt fire brigade, includi ng equipment and training to ve ri fy w
the guidelines contained in BTP CMEB 9.5-1 Item C.3.
We will r eq ui r e that the ap p li c a nt commit to follow the guidelines in BTP CMEB 9.5-1 Item C.3 in the es t ab li s hme nt e
and training of the fire brigade.
V.
General Plant Guidelines A.
Bui ldi ng Design Fire areas are defined by walls and f loo r / c eili ng a s se mb li es.
However, select areas, such as Fire Areas 1 ( Au xi li a ry B ui l di ng, Elevation 522) and 18 ( Auxili ary Buildi ng, Elevation 522) feature i
b ou ndijry penetrations by unp r ot ec t ed spiral stairways.
We will requiMe that these stairways be enclosed by 3-hour rated f ire walls y
c or{a ly with the guidelines of BTP CMEB 9.5-1, Item C.5, a. (1 ).
to
5-The applicant states that cable and cable tray penetrations of fire barriers wilL be sealed to provide protection equivalent to the rating of the original barrier.
The design of the pene-tration seats wilL meet the requirements of IEEE 634-1978.
To assure that the penetration seats are equivalent to the fire resistance of the original barrier, we wilL require that the seals meet the three-hour fire test requirement of ASTM E-119, in accordance with BTP CMEB 9. 5-1, It em C.S.a(3).
Door openings in fire rated barriers are provided with labeled fire doors.
The applicant identifies select openings, such as the hatchway in fire areas 39 and 40 ( Auxili ary Building, Elevation 543) and the doorways providing access to fire areas 9 and 10, (Auxiliary Building, Elevation 560) wh'ere unlisted but
" equivalent" fire doors wilL be installed.
We wilL require that these doors are three-hour rated, U.L.
Listed doors.
l The applicant has stated that 3-hour, U.L.
rated fire dampers are I
provided in ventilation ducts that penetrate rated fire barriers.
This commitment is in accordance with the guidelines of BTP CMEB 9.5-1 Item C.S.a, and therefore, is a<ceptable.
Interior wall and structural components, thermal insulation materials, radiation shielding materials, and sound proofing l
l
)
. a re noncombustib le.
Interior finishes have a flame spread rating of 25 or less and a smoke and fuel contribution of 50 or less in its end use configuration, as determined by the test method of ASTM E-84 We find this to be in accordance with the guidelines of BTP CMEB 9.5-1, Item C.S.a and i s, there-fore, a c cep t ab le.
The high vo lt age-high amperage load center transformers located in the Auxili ary Buildi ng are gas filled.
All other t rans f o rme rs located in safety related buildi ng a reas are dry type, air cooled.
There are no oil filled t rans f o rme rs loca ted within 50 f eet of the exterior wall of a bui ldi ng containing safety related eq ui pme nt.
Openings.in exterior walls of buildi ngs containing safety related systems which are exposed to fire hazards are closed with penetrator seats with a fire resistance equal to the rating of the ba r ri er.
This meets our guidelines in BTP CMEB 9.5-1, Item C.S.a and i s, t h,s r e f o r e, a c cep t ab le.
B.
Safe Shutdown Capabi li ty The information provided by the ap pli c a nt is insufficient to ve ri f y compli ance wi th our guidelines.
We will r >ui re the ap pli ca nt to provide a safe shutdown analysis in accordance l
with the guidelines of BTP CMEB 9. 5 -1, Item C.S.b.
l
7 C.
Alternate Shutdown Capability The applicant commits to install a completely independent
" Standby Shutdown System" to provide a means of b ringing the unit to a safe hot shutdown condition independent of loss of redundant safe shutdown capabi li ty.
A description has not been provided to ve ri f y the capabi li ty of the dedicated Standby shutdown System for achieving hot and cold shutdown.
We wiL L require that the Standby Shutdown System comply with the guidelines contained in BTP CMEB 9. 5 -1 Item C.S.c.
D.
Control of Conbustibles safety-related systems have been i sola ted or sepa rated from c ombu s t ib le mater.ials to the extent po s s ib le.
Safety related equipment is not exposed to the turbine generato r oil and hy d rau li c control fluid systems.
The reactor coolant pump moto rs feature an enclosure a round t he uppe r and lower oil pots to contain any oil spill.
We do not have a firm commitment 4
from the applicant that he will comply with NFPA 30, "F la mmab le and Combustible Liquids Code," in accordance with BTP CMEB, Item C.S.d(4).
We will req ui re that the ap pli ca nt comply with NFPA 30.
l A sepa rate buildi ng i s provided f o r bulk gas s to rage.
The ap pli ca nt has provided insufficient inf o rma tion concerning i
the design 'nd routing of bulk gas piping and the configuration
i s
8-of storage containers to ve ri f y compli ance with our guidelines.
We will requi re that the applicant comply wi th the guidelines contained in BTP CMEB 9. 5-1, I t em C.S.d.(5).
E.
Electrical Cable Construction, Cab le T r ay s, and Cable Penetrations The power, control and instrumentation cable used in Catawba is of an interlocked a rmo r design in a galvanized steet jacket.
All cables pass the IEEE 383-1974 flame test.
In addi tion, the ap pli ca nt has submitted samples of the c ab le for testing at Unde rw ri t e rs Laboratories in their " corner test" configuration.
When subjected to a 400,000 BTU /hr heat flux, t he cable exibited no tendency to propogate fire.
In ad di t ion the applicant has conduct ed t ests which demons trate that no fire propagation from c ab le to cab le or tray to tray occurs as a result of an electrically initiated fire.
We find this a c cep t ab le.
All cab le trays are constructed of galvanized steel.
Cable tray penetrations have a fire rating at least equal to the rating of fire barriers which they are penetrating.
Based on our evaluation, we find that the electrical cable construction, cable trays, and cab le penetrations meet the guidelines of BTP CMEB 9. 5 -1 Item C.S.c.
I e
m_
m
_9-F.
Ventilation Fire barrier openings are provided with fire dampers which close if a fire causes room t empe rature to exceed a set value.
Fresh air intakes to a reas containing s a f e ty related e q ui pm e nt or systems are located to reduce the possibility of contaminating the intake air with products of combustion.
We find this acceptable.
Charcoal filters have been provided with a fire suppression system in accordance wi th Regulato ry Gui de 1.52 "D es ign, T e s t i ng, and Maintenance criteria for Atmospheric Cleanup Ai r Filt ra tion."
We find t his ac cep tab le.
Where total f loodi ng gas extinguishing systems a re u sed, air intake and exhaust ventilation dampe rs are provided wi th mechanisms which close them upon initiation of gas flow.
We find this a c cep t ab le.
Based on our evaluation, we find that the ventilation system meets the guidelines of BTP CMEB 9.5-1, Item C.S.f, and i s, therefore, a c cep t ab le.
G.
Lighting and Communications E me rge ncy li gh ts wi th individual 8-hour battery pack power s upp li es are provided in the control room, Au xi li a ry Shutdown Panel area and in select locations r eq ui red for
l
. cold shutdown.
We wilL require that fixed emergency Lighting units with individual 8-hour battery power supplies be provided in atL areas that must be manned for safe shutdown and for access and egress routes to atL fire areas in accordance with BTP CMEB 9.5-1, Item C. 5.g. (1).
Emergency communication is dependent upon the station telephone system and the public address system.
An additional communications net, with a fixed repeater wiLL be i nstalled for the containment and would be operational only during outages and when the fire brigade enters the containment.
We are concerned that the system wilL not provide the flexible and reliable emergency communications capability necessary for effective fire fighting and safe shutdown operations.
We wiLL require e
that a portable radio communications system, independent of the normal plant communications system, be provided for use by the fire brigade and other operations personnel t
required to achieve safe plant shutdown in accordance with BTP CMEB 9.5-1, Item C.S g.(4).
l l
[
w-w, n
11 VI.
Fire Detection and Suppression A.
Fire Detection The fire detection systems consist of the detectors, associ ated elect rical ci r cui t ry, elect rical power suppli es, and the fire communication panets.
The types of detectors used a re ionization, rate-of-rise, f i xed t empe ra tur e, and combination f i xed t empe rature/ rate of ri se de tecto rs.
The systems provide audible and visual alarms locally and in the control room.
The fire detection system power is suppli ed from t he ba ttery-backed auxiliary cont rol powe r sys tem.
In the event of a power out age, the au x i li a ry control power sys tem is powered from'"
the emergency diesel ge ne ra to rs.
The fire detection systems are designed and i ns talled to conf orm to NFPA 720, 1975, "S t and a rd for the Installation, Maintenance, and Use of Proprietary P rot ect i ve S i gn ali ng Systems."
Fire detection systems wilL be installed in atL areas of the plant containing safety related system components and cabtes.
The systems wilL be continuously s upe rvi s ed to provide alarm and t r ou b le indication to the control room from atL detectors unde r single break or g round f ault c ondi tions.
Based on our evaluation, we find the fire detection systems
s 12 -
conf orm to the guidelines in BTP CMEB 9.5-1 Item C.6.a.
and are, therefore, a c cep t ab le.
B.
Fire Protection Water Supply Sys tem The water supply system consists of three fire pumps separately connected to a buried 12-i n ch c e me nt-li ned water main loop around t he pla nt.
All three fire pumps are 100% capaci ty electric driven, each rated 2500 gpm at 133 psig.
The three fire pumps have independent power suppli es a nr, controls.
Two fire pumps are suppli ed by separate station diesel ge ne ra to rs.
The fire pumps and c ont rolle rs are not Unde rw ri te r's Laborato ry Listed, but are installed and t ested in accordance with the intent of NFPA 20.
Two of the three fire pumps loca ted in the same bay of the intake structure are separated by a three-hour-fire-rated wall.
The other f i re pump is located in an adjacent bay of the intake structure.
e The fire protection water supply system is kept pressurized by one 200 gpm (stendby only) and two 25 gpm jockey pumps (one on standby) to p revent frequent starting of the fire pumps by maintaining pressure in the yard mains at 125 p sig.
A 5000 gallon pres surize r t ank is provided in the system to act as an I
accumulator or surge tank for the j o ckey pump.
The fire pumps are automatically started by low pressure wi th the set pressures i
(
i
' at which the pumps are activated s tagge red or manually started by the operator in the control room for each pump.
Once the fire pumps are started, they can only be shut of f manually.
Separate annunci ator ala rms on separate ci rcui ts are provided in the control room to moni tor the fire pump status, prime move r availabili ty, power failure, and failure of the fire pump to start.
The water supply for f ire p rot ect ion is taken f rom Lake Wyli e, which is of sufficient size to supply the anticipated fire flow.
Each fire pump has its own supply suction piping.
The greatest water demand f or the fixed fire suppression systems'-
is 3600 gpm, and coupled with 500 gpm f o r hose streams, creates a total water demand of 4100 gpm.
We find that the water supply system can delive r the requi red water demand with one pump out of service.
The ultimate heat sink f o r Catauba is the standby nuclear servi ce pond.
Fire hydrants are provided at intervals of 250 feet along the fire prot ection water supply loop.
Post indi cato r valves are provided to isolate sections of the fire loop f o r maint enance or repairs.
Standard hose houses are provided at alternate hydrants.
A single break in the water supply piping will not eliminate both the primary and secondary water suppression in any fire zone.
t
14 -
Valves in the f ire prot ect ion wa ter supply system are electrically supervised with alarms in the control room, or Locked in the normat operating position.
Based on our revi ew, we conclude that the water supply system meets the guidelines of BTP CMEB 9.5-1 Item C.6.b and i s, therefore, a c cep t ab le.
C.
Sprinkler and Standpipe Systems The sprinkle r sys tems and standpipe hose systems are independently connected to looped fire protection heade rs so as to prevent single failures from impairing both the primary and backup fi re prot ection sys tems out side contain-ment.
w We do not have a firm commitment from the appli c a nt that-he will comply with NFPA 13, " Installation of sprink te r Systems."
We will require the appli cant to comply with NFPA 13 in accordance with BTP CMEB 9.5-1, Item C. 6. c. (1 ).
The areas that are being equipped with automatic water suppression systems are:
RHR Pump Rooms 100, 104, 105, 109, 110 and connecting corridors.
Fire Areas 2&3 (rooms 250 & 260).
Centrifugal Charging Pumps, rooms 231, 230, 241 and 240.
15 -
Component c oo li ng pumps and cable concentration areas.
Reactor building annulus.
Fire Area RB-2, pipe corridor.
Manual pre-action for the lower containment filt ers.
Reactor coolant pumps.
In the Fire Hazards Analyses, t he appli cant identified fire areas containing safe shutdown related equipment that are not prot ect ed by an automatic sprinkle r sys tem.
Fi re prot ect ion f or these areas consists of automatic fire de tecto rs, manual hose stations and portable fire extinguishers.
The bounda ries of these areas are composed of three-hour fire rnted construction.
C ab le is of a galvanized steel interlocked a rmor design di scussed in Section V.E of this report.
Th e dedi ca ted Standby Shutdown System is available to achieve safe shutdown in the event of a fire i n a ny of these areas.
This is an acceptabte deviation from the guidelines of BTP CMEB 9.5-1, Item C.S.b(2).
Int erio r manual hose stations are provided and equipped to reach any plant location wi th at least one effective hose stream.
Each hose station is provided wi th 100 f eet of 1 1/2 inch hose with a spray nozzle to provide adequate coverage.
We find that the hose stations meet our guidelines of BTP CMEB 9.5-1 Item C.6.c and are, therefore, a c cep t ab le.
. The appli cant has not identified seismic design of St andpipe Systems which is recommended in BTP CMEB 9.5-1, Item C.6. c. (1 ).
For plants with construction permits issued prior to July 30, 1976, the guidelines contained in Appendix A to BTP ASB 9.5-1 have no requirement for seismic design of standpipe systems.
Therefore, this is an acceptable deviation f rom t he guideli nes of CMEB 9. 5 -1, Item C. 6. c. (1 ).
D.
Carbon Dioxide Suppression System Low pressure carbon dioxide, automatic total flooding, and Local application sys tems are provided f o r p rima ry protection in the diesel generato r set rooms and the auxiliary feed water pump pits.
The systems art activated by c ros s :oned de tecto rs which ala rm and annunci ate in the control room.
The carbon dioxide systems may also be activated manually.
The appli cant has not committed to designing and i ns t al li ng the systems in e
accordance with NFPA St anda rd No. 12, "C a rbon Di oxi de Extinguishing Systems."
We will require that the applicant comply with the guidelines of NFPA 12 in ac cordance wi th BTP CMEB 9.5-1 Item C.6.c.
E.
Portable E xt i ngui s h e rs Po rt ab le f i re ex t ingui she rs are provided to conform with the guidelines of NFPA Standard No. 10, " Portable Fi re Ext ingui shers,"
1978.
We find this a c cep t ab le.
Based on our review we c o n c lu de that these ex t ingu i s h e rs meet the guidelines of BTP CMEB 9. 5 -1, Item C.6.f and a re, therefore, a c cep t ab le.
o e
e 17 VII.
Fire Protection f or Speci fic P la nt Areas A.
Containment Containment fire protection features include manual hose stations; automatic fire detecto rs; a fixed, automatic s pr i nk le r system for the pipe corridor; and a fixed, manual pre-action sprinkler system for the lower containment filt ers.
The reactor coola nt pump motors are manu f actured with an enclosure around the upper and lower oil pots to contain any oil spill and direct it to piping which goes to a drain tank.
The oil collection system for the reactor coolant pumps are designed to withstand the design basis seismic event.
Heat sensing cable de t ecto rs are i ns talled a round bot h the pumps and motors.
g Based on our evaluation, we conclude that the fire protection for containment meets the guidelines of BTP CMEB 9. 5 -1, Item C.7.a, and i s, therefore, a c cep t ab le.
B.
Reactor Building The reactor building is separated from adjacent bui ldi ng s by three-hour rated ba rriers.
A fixed automatic sprinkler system protects the annulus.
Ad di tional fire p rot ec t io n includes automatic fi re detecto rs, manual hose stations and po rt ab le fire ex t ingui s he rs.
i Based on our evaluation, we conclude that the fire protection f or this area meets the guidelines of BTP CMEB 9.5-1, Item C.7.a, and i s, therefore, a c cep t ab le.
C.
Control Room The control room is s ep a rat ed from all other areas of the plant by three-hour rated a s sembli es.
All ventila tion duct s penetrating these barriers have three-hour-fire-rated dampers.
M e c h a n i c a '. and electrical penetrations in rated fire ba rriers are sealed with an approved t hree-hour barrier.
Access to the control room is through U.L.
approved, three-hour rated fire doors and frames.
Ionization smoke detectors and rate-of-rise / fixed temperature heat detecto rs are installed on the c e ili ng and inside the main cont rol boa rd cons o le s.
AL L detecto rs are ala rmed a nd annunci ated in the control room.
The control room ventilation i nt ak e s are equipped with smoke detectors which alarm in the control room and automatically close the intake in the event of radioactive contamination.
I Smoke is prevented from e nt e ri ng the control room from other areas because the room is maintained under a positive pressure.
If it is necessary to e x h au s t smoke from the room, a purge fan is provided to purge smoke to the au xi li a ry bui ldi ng exhaust system f or di scharge through the station vent.
9
. AlL cables that enter the control room terminate in the control room.
Only powe r and control cabtes essential for operation of lighting and HVAC ducts are located in the concealed c e ili ng space.
Redundant safety related cable is located below the raised floor in the control room.
The cable is of an interlocked a rmor design as desc ribed in Section V.E of this report.
The unde rf loor area is not pr ot ec t ed by a fire suppression system a s requi red.
Prot ction i s achieved by physical separation of the cable, supplemented by automatic fire detectors, m a nu a l hose stations and po rt ab Le fire extinguishers.
The Stanaby Shutdown System (di scussed in Sect ion V.C) is available to achieve safe shutdown in the event of a fire in the control room.
This is an acceptable deviation from the guidelines of BTP CMEB 9.5-1, Item e
~~
C.7.b.
D.
Cable Spreading Rooms There are two c ab le spreading rooms (one for each unit).
Each cable spreading room contains cabtes from redundant divisions.
The rooms are s ep a r a t ed from the remainder of i
the plant and each other by three-hour fire-rated walls and f l oo r/ c e i l i ng a s s emb li es.
Al L penetrations through fire-(
rated barriers are fitted with three-hour fire dampers and/or penetration seats.
l
{
t I
' The cables are of an int erlocked a rmo r, galvanized steel design as described i n Section V.E.
The cable spreading rooms are not protected by a fixed fire suppression system as required.
Fire protection consists of automatic fire detecto rs, supplemented by manual hose stations and po rt ab le fire extinguishers.
In the event of a fire in the cable spreading rooms, the Standby Shutdown System (di scussed i n Section V.C) is available to achieve safe shutdown.
This protection provides an acceptable level of safety a nd is an acceptable deviation from the guidelines of BTP CMEB 9. 5 -1, Item C.7.c.
E.
Switchgear Rooms There is one switchgear room f or each division with complete divisional separation.
Switchgear rooms are separated from each other and the remainde r of the plant by three-hour rated watls and floor / ceiling a s semb Li es.
Automatic smoke detectors that alarm Locally and in the control room are provided.
Po rt ab le fire extinguishers and manual hose stations are provided in and adjacent to each switchgear room.
- 21 Based on our evaluation, we conclude that the fire protection for the switchgear rooms meets the guidelines of BTP CMEB 9.5-1, Item C.7.e, and i s, therefore, a c cep t ab le.
F.
Remote Safety Related Panels separate rooms are provided for the au x i li a ry shutdown panets for each division to achieve complete divisional separation.
The rooms are separated from the rema inde r of the plant and each other by ceili ng / floor assemb Li es, walls and doors with fire resistance ratings of three-hours.
Each room is provided with automatic smoke detectors that alarm locally and i n the control room.
Based on our evaluation of the information submitted, we find t he fi re prot ection f o r this w
area to be in accordance with BTP C MEB 9.5-1, I t em C.7.e.
and i s, therefore, acceptabte.
G.
Safety Related Battery Rooms The plant battery rooms are separated from each other and a
the remainder of the fire area by three-hour rated fire ba rri e rs.
Each ba ttery room i s equipped wi th redundant exhaust ventitation to preYent the buildup of hydrogen.
Ionization smoke detectors are provided in each room.
There are air f low moni to rs which alarm in the control room to monitor loss of ventilation in each battery room.
j
22 -
The applicant has not provided t he necessary i nf orma tion on divisional separation wi thin each battery room.
'We will report on this issue in a subsequent safety evaluation report.
H.
Diesel Generator Areas Each diesel generator is located i n a di f f e rent fire area separated by three-hour fire rated walls and f loor/ceili ng as semb li es.
All cable and piping penetration through the fire rated ba rriers are fitted with three-hour rated penetration seals.
The diesel fuel oil day t ank is located wi thin the emergency diesel generator reom which is separated f rom ot her pla nt areas by three-hour fire rated ba r ri e rs.
Each 600 gallon di esel fuel oil day tank i s contained by a dike around the tank.
The walls extend above piping a
to the tank to block oil spray from the di es el generator room.
Each diesel fuel oil day tank is protected by (L,
total flooding CO2 extinguishing system.
The diesel fuel oil storage tanks are buried.
The day tank can be i so la t ed from the main f uel oil t anks by means of a valve located e x t e ri o r to the diesel gene ra to r room.
i Each diesel gene rator room i s prot ect ed by an automatic l
total flooding carbon dioxide extinguishing system.
Heat de t ecto rs actuate the carbon dioxide system.
Upon actuation of the carbon dioxide fire suppression system, t he dieset buildi ng ventitation system is automatically de-energized and the outside air and exhaust dampe rs are closed.
Ventilation system is automatically de-energized a nd the outside air and exhaust dampe rs are closed.
Based on our evaluation, we conclude that the fire prot ection f or the di esel ge ne rato r rooms meets the guidelines of BTP CME 8 9.5-1, Item C.7.1, a nd i s, therefore, a c cep t ab le.
I.
Other Plant Areas e
~~
The appli cant's Fi re Haza rds Analysis addressed othe r plant areas not specificalLy discussed in this report.
We find that the fi re prot ect ion f or these areas is in accordance with the guidelines of BTP CMEB 9.5-1, Item C.7.
and i s, therefore, a c cep t ab le.
. VIII.
Con c lu s io n The technical requirements of Appendix R to 10 CFR 50 and Appendix A to ASB 9.5-1 have been incl;ded i n BTP CMEB
- 9. 5 -1.
No deviations from the requirements of Appendix R have been identified.
The following deviations from the guidelines of BTP CMEB 9.5-1 have been approved:
Automatic Sp ri nk le rs for Safe Shutdown Areas (Section VI.C)
Seismic Design of Standpipe Systems (Section VI.C)
Fixed Suppression System f or Under Floor Area in the Control Room (Section VII.B)
Sprinkle r Sys tem f o r Cable Spreading Rooms (Section VII.C) s.
The f ollowing items remain open:
Station Fi re Plan (Section II.A) e
~~
Admi ni s t r a t i ve Controls (Sect ion III)
Fire Brigade and Fire Brigade Training (Section IV)
Integrity of three-hour Fire Barriers (Section V.A)
Ve ri f i c a t ion of Cable Penetration (Section V. A)
Verification of Fire Doors (Section V.A)
Safe Shutdown Analyses (Section V.8)
Description of Standby Shutdown System (Section V.C)
Control of C ombu s t ib le s (Section V.0)
- Design of Bulk Gas System (Sect ion V.D)
Eme rgency Lighting (Sect ion V.G)
Emergency Communications (Sect ion V.G)
Design of Sprinkler Systems (Section VI.C)
Design of Ca rbon Dioxide Systems (Section VI.D)
Divi sional Sepa ration in Batte ry Rooms (Sect ion VIII.F)
The applicant has been informed of the necessity of the resolution of all open items so that all fire prot ection features can be implemented prior to fuel loading.
We will r ep o rt our review of these unresolved items in a subs eque nt safety evaluation report.
h w.
F
+--wr-
--7 m - - - - + -- --
ENCLOSURE 4 CATAWBA NUCLEAR STATION DRAFT SAFETY EVALUATION REPORT INPUT CONCERNING APPENDIX R, SAFE SHUTDOWN FOR FIRES 9.5.1.6 Safe Shutdown for Fires ( Appendi x R)
We have reviewed the applicant's submittal entitled
" Response to Appendix A to Branch Technical Position APCSB 9.5-1 (August 1981 revision)" with respect to the safe shutdown capability in the event of fires anywhere in the plant as defined in Appendix R to 10 CFR Part 50.
We conclude that the applicant's response as identified in Appendix D to this study is unsatis-factory.
The applicant has not provided sufficient
^
information for us to evaluate Parts III.G and III.L of Appendix R.
The applicant should provide the folLowing additional information:
1.
The applicant should describe the methodology used to verify that proper separation (fire protection) is provided for the safe shutdown capability in accordance with Appendix R, Part III.G.
The applicant should provide area arrangement drawings showing the safe shutdown system (including cable routing) in order that we may review the results.
o s
2.
The applicant should address the means provided for assuring the function of the safe shutdown capability when considering fire induced failures in associated circuits.
The enclosure provides the staff concern with associated circuits.
The enclosure also provides guidance needed by the applicant to review associated ctreuits of concern and the information to be provided for staff evaluation.
The applicant should address Part II.C. of the enclosure.
3.
The applicant should described in detail the design capability of the dedicated Standby Shutdown System (SSS) for achieving hot and cold shutdown in accordance with Parts III.G and lII.L of Appendix R.
This dis-cussion should include the equipment which compro-mises the SSS necessary for performing various safe shutdown functions, alL required support equipment, and the instrumentation available for monitoring shutdown.
, _._w
-~
ENCLOSURE ASSOCIATED CIRCUIT GUIDANCE I.
INTRODUCTION The following discusses the requirements for protecting redundant and/or alternative equipment needed for safe shutdown in the event of a fire.The requirements of Appendix R ad' dress hot shutdown equipment which must be_
free of fire damage. The following,r.equirements also apply to cold shutdown equipment if the licensee elects to dedonstrate that tiie equipment. is 'to be free of. fir,e. damage. Append 6 R does allow repatrable damage to cold sh ecuipmenf..,
~
Using the requirements of Sections III.G and III.L of Appendix R
, the capa-bility 'to achieve hot shutdown must exist given a fire in any area of plant in conjunction with a loss of offsite power for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Section III.G of Appendix R provides four methods for ensuring that the hot sh bility is protected from fires.
The first three options as defined in Section III.G.2 provides methods for protection frcm fires of eqtripment need or hot shutdown:
1.
Redundant systems including cables, equipment, and associated circu i
may be separated by a three-hour fire rated barrier; or, 2.
Redundant systems including cables, equipment and associated circui be separated by a horizontal distance of more than 20 feet with no in I
vening combustibles.
In addition, fire detection and an automatic fire suppression system are required; or, 3.
Redundant systems including cables, equipment and associated circtiits be enclosed by a one-hour fire rated barrier.
In addition, fire detectors and an automatic fire suppression system are required.
2 The last option as defined by Section III.G.3 provides an alternative shutdown capability to the redundant trains damaged by a fire.
4.
Alternative shutdown equipment must be independent of the cables, equip-ment and associated circuits of the redundant systems damaged by the fire.
II. Associated Circuits of Concern The following discussion provides A) a definition of associated circuits for Appendix R consideration, B) the guidelines for protecting the safe'shutdcwn capability from the fire induced failures of associated circuits and C) the in-formation required by the staff to review associated circuits. It is important to note that our interest is only with those circuit (cables) whose fire-induced failure could affect shutdown. ~ Guidelines for protecting the safe shutdown capability from the fire-induced failures of associated circuits are provided. These guidelines do not limit the alternatives ava lable to the licensee for protecting the shutdown capability. All proposed methods for protection of the shutdown capability from fire-induced failures will be evaluated by the staff for acceptability.
A.
Our concern is that circuits within the fire area.will receive fire damage which can affect shutdown capability and thereby prevent post-fire safe shutdown. Associated Circuits
- of Concern are' defined as those cables (safety related, non-safety related, Class lE,' and non-Class lE) that:
- The definition for associated circuits is not exactly the same as the definition presented in IEEE-384-1977.
J
's
.___m
. 1.
Have a physical separation less than that required by Section III.G.2 of Appendix R,.and; 2.
Have one of the following:
a.
a common power source with the shutdown equipment (redundant or alternative) and the power source is not electrically protected frcm the circuit of concern by cuerdinated breakers, fuses, or similar devices (see diagram 2a), or b.
a connecticn to circuits of equipment whose spurious operation would adversely affect the shutdown capability (e.g., RHR/RCS isolation valves ADS valves, PORVs, steam generator atmospheric dump valves, instrumentation, steam bypass, etc.) (see diagram 2b), or c.
a common enclosure (e.g., raceway, panel, junction) with the shutdown cables (redundantandalternative)and, (1) are not electrically protected by circuit breakers, fuses or simi-lar devices, or (2) will allow propagation of the fire into the common enclosure, (see diagram 2c).
l l
O i
EXAMPLES OF ASSOCIATED CIRCUITS OF CONCERN
~
i l
l 1
i Fear. AftE A I
i FIRE AREA Fspg 8BEA
)
N
\\
\\
\\
.fl
\\i
\\
]l i
N
\\
l x
\\
d!
(
.tl l
,\\
l
\\
s l
\\
\\
N l
N N
ol N
N m
I N
8 3
\\
x
\\
')
.l
\\
4
\\
k I
N i
x 3
g
'a 1
j
\\
il
$1 Fl j
\\
.)
~.,
y i_
\\
P s
l l l
\\
l
(
<l N
l
\\
\\
l I-
\\NNNNkNN
\\\\\\ \\\\\\$ N NN
\\'N\\\\\\\\Y
~
i i
g i
coanos 3
- vwve, EuctosaeE l
As's P n - nus E op,,,ed wAo,e sp4m The area barriers shown above meet f
the appropriate sub-paragraphs (a-f) opt < ht coalet affec/
of section III.G-2 of Appendix R.
i skrsddo an Diagram 2A Diagram 28 Diagram 2C l
i i
h i
W a
~
.=
I 4
' B.
The following guidelines are for protecting the shutdown capability from fire-induced failures of circuits (cables) in the fire area. The shutdown capability may be protected from the adverse effect of damage to associated circuits of concern by the following methods:
1.
Provide protection between the associated circuits of corcern and the shutdown circuits as per Section III.G.2 of Appendix R, or 2.
a.
For a comon power source case of associated circuit:
Provide load fuse / breaker (interrupting devices) to feeder fuse / breaker coordination to prevent loss of the redundant or alternative shutdown power source. To ensure that the following coordination criteria are met the 'following should apply:
(1) The associated circuit of concern interrupting devices '
(breakers or fuses) time-overcurrent trip characteristic for all circuits faults should cause the interrupting device to interrupt the fault current prior to initiation of a trip of any upstream interrupting device which will cause a loss of the ecmon power. source, (2) The power source shall supply the necessary fault current for sufficient time to ensure the proper coordination without loss of function of the shutdc n loads.
l l
l w
W __
. +.,
I l The acceptability of a particular interrupting device is considered demonstrated if the following criteria are met:
(1) The interrupting device design shall be factory tested to verify overcurrent protection as designed in accordance with the applicable UL, ANSI, or ND% standards.
(ii) For low and medium voltage switchgear (480 V and above) circuit breaker / protective relay periodic testing shall demonstrate that the overall coordination scheme remains within the limits specified in the design criteria. This testing may be performed as a series of overlapping tests.
(iii) iloided case circuit breakers shall periodically be manually exercised and inspected to insure ease of operation. On a rotating refueling outage basis a sample of these breakers shall be tested to determine that breaker drift is within that allowed by tha design criteria. Breakers should be tested in accordance with an accepted QC testing methodology such as MIL ST') 10 5 D.
(iv) Fuses when ur.ed as interrupting devices do not requ re i
periodic testing. Administrative controls must insure that replacement fuses with ratings o'ther than those selected for proper coordination are not accidentally used.
l l
b.
For circuits of equipment and/or components whose spurious operation would affect the capability to safely shutdown:
I
e a
. (1) provide a means to isolate the equipment and/or components from the fire area prior to the fire (i.e., remove power cables, open circuitbreakers);or
~
(2) provide electrical isolation that prevents spurious operation.
Potential 1 clation devices include breakers, fuses, ampli-fiers, control switches, current XFRS, fiber optic couplers, relays and transducers; or (3) provide a means to detect spurious operations and then proce-dures to defeat the maloperation of equipment (i.e., closure of the block valve if PORY spuriously operates, opening of the breakers to remove spurious operation of safety injection);
c.
For corren enclosure cases of associated circuits:
(1) provide appropriate measures to prevent propagation of the fire; and (2) provide electrical protection (i.e., breakers, fuses or similardevices)
C.
INFOP.MATION REOUIRED The following information is required to demonstrate that associated circuits will not prevent operation or cause maloperation of the shutdown method:
_ L._.
~, -
a.
Describe the methodology used to assess the potential of associated citcuit adversely affecting the shuttown capability. The description of the methodology should include tha methods used to identify the circuits which share a common power supply or a common enclosure with the shutdown system and the circuits whose spurious operation would affect shutdown. Additionally, the description should include the methods used to identify if these circuits are associated circuffs of concern due to their location in the fire area.
b.
Show that fire-induced failures (hot shorts, open circuits or shorts to ground) of each of the associated circuits of concern will not 4
prevent operation of cause maloperation of the shutdown method.
2.
The residual heat removal system is generally a low pressure system that interfaces with the high pressure primary coolant system.
To preclude a LOCA through this interface, we require compliance with the recommendations of Branch Technical Position RSB 5-1.
Thus, the interface most likely consists of two redundant and independent motor operated valves.
These two motor operated valves and their associited i
cables may be subject to a single f.fre hazard. It is our concern that i
this single fire could cause the two valves to open resulting in a fire initiated LOCA through the high-low pressure system
(
interface.
To assure that this interface and other high-low pressure interfaces are adequately protected from the effects of a single fire, we. require the following information:
Identify each high-low pressure interface that uses redundant a.
electrically controlled devices (such as two series motor operated valves) to isolate or preclude rupture of any primary coolant boundary.
. -. - ~
~ :
..A c
- -- f.
- .~
l
.8.
b.
For each set of redundant valves identified in i., verify the redundant cabling (power and ccntrol) have adequate physical separation as required by Section III.G.2 of Appendix P..
i i
c.
For each case where adequate sap: ration is r.ct prcvidel, sh:t thIt l
fire induced failures (hot short, open circuits or short to ground) of the cables will net cause maloperation and result in a LOCA.
l 9
I l
l
3(
ENCLOSURE 5 s
Equipment Qualification Branch Input for Safety Evaluation Report Catauba 3.9.3.2 Pump and Valve Operability Assurance The staff has reviewed the applicant's pump and valve operability assurance program as discussed in Section 3.9.3.2 of the FSAR and compared this information with Section 3.10 of the Revised Standard Review Plan.
Based on our reviews the applicant has provided information to define how active pumps and valves are generally qualified with respect to operability.
However, and in particulari for those components where qualifi-cation and/or operability assurance is by analysis aloner some question remains as to the confidence level assured by this methodology.
The necessity for additional component testing is being considered and can not be established without an inspection at the plant site.
Thereforer for the staff to determine the adequacy of the implementation of the applicant's pump and valve operability assurance programs an on-site audit of the equipment and supporting documentation is required.
The on-site audit will include a plant inspection to observe the as-built configuration and installation of the equipment.
Also during the audit the staff will review qualifying documentation, for example, test reports and analysis, which are described in the applicant's program.
Thus our overall review includes an FSAR review and an on-site a u d'i t of the equipment.
Both phases of the staff review must be determined acceptable to arrive at a favorable conclusion on the applicant's overall pump and valve
'N..
. operability assurance program.
The applicant has been requested to provide information on the completion status of the equipment documentations and on-site installation of the equipment.
Before the audit is conductede 85 to 90 percent completion should be attained for both the equipment documentation and the on-site installstion of the equipment.
Once the applicant has indicated that his work is substantially completer the staff will conduct an on-site audit shortly thereafter.
Because of the limited number of equipment that can be audited within a reasonable times the audit results must provide a high degree of confidence that the implementation of the applicant's program is acceptable.
k.
- k..
ENCLOSURE 6
-1 3.10 Seismic and Dynamic Qualification of Seismic Category I Mechanical and Electrical Equipment Our evaluation of the adequacy of the applicant's program for qualification of safety-related electrical and mechanical equipment for seismic and dynamic loads consists of (1) a determination of the acc*ptability of the procedures used, standards folLowed, and the completeness of the program in general, and (2) an onsite audit of selected equipment items to develop the basis for the staff judgment on the completeness and adequacy of the inplementation of the entire seismic and dynamic qualification program.
The Seismic Qualification Review Team (SQRT), which consists of reviewers from the Equipment Qualification Branch (EQB) and consultants from Brookhaven National Laboratory (BNL), has reviewed the methodology and procedure of equipment seismic and dynamic qualification program contained in the pertinent FSAR Sections 3.2 3.9.2, 3.9.3, and 3.10.
The SQRT has concluded that the information reviewed in general meets the intent of the current Licensing criteria as described in Regulatory Guides 1.100 and 1.92, and the Standard Review Plan Section 3.10 (NUREG-0800).
However, the methods and procedures described in FSAR Section 3.9.2.2 concerning single frequency testi single axis test, and multi-axis and multi-frequency test are not totally consistent with these current Licensing criteria.
The applicant is required to rectify this in his FSAR.
The SQRT will follow the applicant's effort in this area closely, and wiLL confirm its implementation during the onsite audit.
L.
~
'o 2-NRC is currently pursuing rule making activities in the area of environmental qualification of electrical equipment and accredita-tion of testing laboratories.
The extent to which the proposed rules may apply to this plant in the area of equipment aging prior to seismic testing in an accreditated laboratory will be reviewed and reported in a future supplement of this safety evaluation report.
In our communication with the applicant, we indicated that a substantial portion (85%-90%) of the equipment must be qualifiedi documented in an auditable manner, and installed onsite before an onsite audit by the staff can be performed.
We also indicated to the applicant the type of information necessary for us to select the equipment items for a detailed onsite review.
Whenever the applicant? indicates that his work is substantially completes the staff will then conduct an onsite audit shortly thereafter.
We shall report the results of our audit and the followup and resolution
~
of our concerns described above in a future supplement to our SER.