ML19323C891

Forwards Request for Addl Info for Continuing Review of Facility FSAR

ML19323C891
Person / Time
Site:	Fermi
Issue date:	04/25/1980
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Jens W DETROIT EDISON CO.
References
NUDOCS 8005190380
Download: ML19323C891 (12)

Text

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h, UNITED STATES 8005190380 y

g NUCLEAL 3 REGULATORY COMMISSION E

WASHINGTON, D. C. 20555

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APR 2 51930 Docket No: 50-341 Dr. Wayne H. Jens Assistant Vice President Engineering & Construction Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226

SUBJECT:

REQUESTS FOR ADDITIONAL INFG7J4ATION IN FERMI 2 FSAR As a result of our continuing review of the Final Safety: Analysis Report (FSAR) for the Enrico Femi Atomic Power Plant Unit 2, we have developed the enclosed requests for additional infomation.

Please amend your FSAR to comply with the requirements listed in the enclosure.

Our review schedule is based on the assumption that the additional infomation will be available for our review by June 4,1980.

If you cannot meet this date, please infom us within 7 days after receipt of this letter so that we may revise our scheduling.

Sincerely, 7j.

J dn' F. Stolz, Chief ight Water Reactors Branch No.1 Division of Project Management

Enclosure:

Requests for Additional Infomation cc:

See next page

a APR 2 51950 Dr. Wayne H. Jens cc: Eugene B. Thomas, Jr., Esq.

David E. Howell, Esq.

LeBoeuf, Lamb, Leiby & MacRae 21916 John R 1333 New Hampshire Avenue, N. W.

Hazel Park, Michigan 48030 Washington, D. C.

20036 Mrs. Martha Drake Peter A. Marquardt, Esq.

230 Fairview Co-Counsel Petoskey, Michigan 49770 The Detroit Edison Company 2000 Second Avenue William J. Scanlon, Esq.

1 Detroit, Michigan 48226 2034 Pauline Boulevard Ann Arbor, Michigan 48103 Mr. William J. Fahrner Project Manager - Fermi 2 The Detroit Edison Compar.y 2000 Second Avenue Detroit, Michigan 4822S t

Mr. Larry E. Schuerman Licensing Engineer - Fermi 2 Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 Charles Bechhoefer, Esq., Chairman Atomic Safety & Liccr. sing Board Panel V. S. Nuclear F.egulatory Commission Washington, D. C.

7.0555 Mr. David R. Schir.k Department of Oceanography Texas A & M University College Station, Texas 77840 Mr. Frederici. J. Shon Atomic Safety & Licensing Board '

Panel U. S. Nuclear Ragulatory Commission Washington, D. C.

20555 Mr. Jeffrey A. Alson 772 Green Street, Building 4 Ypsilanti, Mic.higan 48197

_ ENCLOSURE REQUESTS FOR ADDITIONAL INFORMATION ENRICO FERMI ATOMIC POWER PLANT UNIT 2 DOCXET NO. 50-341 Requests by the following branch in NRC are included in this enclosure.

Requests and pages are numbered sequentially with respect to previously transmitted requests.

Branch Page No.

Reactor Systems Branch 212-28 through 212-36 i

i

212-38 212.0 REACTOR SYSTEMS BRANCE 212.67A Review procedure III.5 of SRP Section 6 3 reccamends that prict to (6.3) installatica, representative active caponents used ir. the ECCS will be proof-tested taider environmental ecnditions and for time periods representative of the most severe operating conditions to which they may be subjected.

Insufficient information has been presented in the FSAR or in the respcnse to C212.67 to determine that proof-testing has been ccepleted for ECCS pumps. Provide a description of vendor in-shop testing and testing after installation.

Provide the expected service life of the ECCS paps and supply the maximun expected accuculated operating time for the ECCS punps during the life of the plant as indicated below:

a) In-shop testing b) Pre-operational testing c) Mcnthly testing d) Yearly testing e) Post-LCCA f) Shutdown 212.75 Provide characteristic curves for the RHR, HPCI, and.ccre spray punps.

(6 3)

Srake hercepower, efficiency, and NPSH should be included en these Curves.

212.76 Section 6 3.2.14 addresses NPSH censiderations for the LPCI and the

]

(6 3) ccre spray systems, but crJ.y cientiens the HPCI system. Provide NPSH calculatiens to suppcrt your claim that adequate NPSH is provided for HPCI operatien. Assurance should be previded that adequate NPSH will l

be available dtring switchover from the condensate storage tank to the suppressica pool.

)

212.77 Cescribe any ;recauticns (mechanical cr administrative) taken to (6.3) prevent vortex formatien and air ingestien during operation of the ECCS pumps.

The description should include both primary and seccndary water supplies (i.e., HPCI uses both the condensate stcrage tank and the suppression pool).

212,34A The response to C212 34 is incceplete. Section 6.3 2.2.7 describes the (6 3) leak detection system provided for the ECCS suction lines but does not address the possibility of ECCS impairment frem a reduced pool level.

Discuss preoperaticnal tests to be perfcrmed to demonstrate that there is to i=pairment of ECCS functicn due to lowered suppression pool level.

212.78 The response to Q212.39 requires additior-1 _larification. We are (6.3) concerned that administrative controls a.one may not be adequate to ensure that a system is operational.

Previde a list of those manual valves critical to the operation of the ECCS and indicate mich valves have positicn indicaticn in the control recm and which valves are centrolled cnly by administrative procedures.

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212-29 212.79 Table 6.3-10 ingicates that the maxima peak cladding temperature for (6.3) the C2A is 2084 F.

Figure 6 3-9a is intended to be a graphical representation of Table,6 3-10, but the PCT for the C2A is shown en the figure to be at: cut 21c0*F. Resolve this discrepancy.

212.80 Secticn 8.2.8 of the FSAR states that the suppressicn pool water is (6 3) ner= ally maintained at cendensate quality by circulating about 5C0 g;m thrcugh the Terus Water Management System and can be entirely replaced within a 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period.

'What ndministrative centrols er cther ccnstraints are prcvided to ensure that the suppressien pcol is not drained when the need for the ECCS systems could be required?

212.81 Scation 6.3 2.14.1 of the FSAR addresses the subject of pmp run-out (6 3) for the EHR pumps.

Provide assurr :es that the FPCI and/cr core spray pmps have been analyzed fer pa::p run-cut conditiens. Describe any tests and associated possible ECCS lineups to demonstrate desi;n adequacy for pmp run-out.

212.56A The respcnse to C212.56 is not complete. Provide calculations to support.

(6 3) the sizing of the relief valves installed en the RER and the Ccre Spray piping to protect against ever-pressurizatien.

212.62 Several plants have used sand-bags er sand-filled tanks as biolcgical (6.3' shielding inside centair=ent. In the event of a LCCA, these tanks er bags could be damaged and sand would be released. The release of sand inside centairment could result in damage to the ECCS pumps.

Identify any areas 4:ere sand-bags er sand-filled tanks are used for biolegical shielding. What precautions wuld be taken to prevent ECCS damage if sand, insulation, paint chips, or similar debris material were released within contain=ent?

212.83 Secticn 7.31.2.2.9 states that ACS safety / relief valve operability (6.3) will te menitcred by a temperature element installed en the valve discharge piping. Cperating experience has shown that a " false" temperature increase =ay be indicated even thcugh the valve has not operated. Justify use of the temperature elatent over a direct valve positien indication to assure safety / relief valve cperability. Discuss other instrument,ation, possibilities, (e.g., AP or acoustic).

212.24A The FSAR states that each of the safety / relief valves provided for (5.2.2) autcmatic depressurization is equippnd with an accumulator and check valve arrange =ent.

The response to previous request 212.24 implies that all fif teen safety / relief valves have accu =ulators. This is not clear from P&I diagram 5.1-3.

Please. clarify whether or not all safety /

relief valves have accumulators.

213-30 212.84 Provide the calculations to suppcrt your relief valve discharge (5.2.2) ccefficients and flow capacities.

212.85 Previde a discussicn of the nt=ber and type of operating cycles for (5 2.2) which each ccmpenent (such as valves, solenoids, vacut= breakers, etc.) in the everpressure protection system is designed.

212.86 Provide the results of the hydraulic calculations that shew the Mach (5.2.2) ntzber, pressure, and temperature at various locations fran upstream cf the safety / relief valves to the suppressicn pool at =axi=tm ficw conditions. The concern is related to the potential for the development of damaging shock waves to the discharge piping. Include the effects of suppression pool swell variations during the operation of the safety / relief valves.

212.87 The narrative on page 15B.4-24 states that the water level does not (15B.4.4) ' reach sicher the high or low level setpoints. Curve 1 of Figure 153.4.4-1 indicates the low level trip point is reached at time T=22 seconds.

Explain.

212.88 Identify the units of flow in sector 2 and sector 3 of Figures 153.4.4-1 (153.4.4) and 153.4.5-1 and in sector 2 and sector 3 of all figures of Appendix 153.

(15B.4.5) 212.89 Tou state that the quantitative analyses of the spectrun of pipe breaks (153.6.5) is covered in Sections 6.2, 6.3, 7.1, 7.3, and 8.3.

Ecwever, most of the information provided applies only to the DBA line break. Provide a list of the pipe size and break locations that were analy=ed for LOCA inside cent Inment.

212.90 Cescribe at precedures are available to the centrol r0cm (5.2.7) operater convert various irdicater readouts to a eccuen leakage equivalent, such as sliens per minute, er pounds per hour.

212.91 Do the primary centainment radiatien menitors have a built-in (5.2.7) check source to pennit operability testing? Cescribe your methed of calibrating the centainment radiation monitcring system.

212.92 Cescribe your methed of testing and calibrating the drywell (5.2.7) at=ospheric cceler cendensate menitoring instrtmentatien.

l 212.93 The drywell equipment drain sump collects identified leakage (5.2.7) frcm "het sources, such as the reacter vessel head flange, valve stem packings, and pu=p packing glands. This leakage

=ay flash into steam which =usn be cendensed to reach the sump in ceder to be measured. 'ahat assurance is there that the steam will be cendensed fer leak detection menitcring purposes?

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l

218-31 4

I 212.94 verify that the calibration of the leak detection level (5.2.7) senscrs is performed during nomal plant operation. Note that per SRP 5.2.5, the leakage detection systems should be equipped with provisions to pemit calibration and cperability tests during plant operation. Also, the testing and calibration should be in ccepliance with I'.EE Standard 279-1971. Discuss how you intend to comply with the r.bove requirenents.

212.95 Is the RCIC turbine speed centrol system a safety grade (5.5.6) design (i.e., Seimic Category I)?

212.96 Show that the ;reoperational initial startup test pregrams (5.5.6) for the RCIC system meet the intent of Regulatcry Guide 1.68.

212.97 Cescribe the design features and operating precedures that (5.5 6) precluce water hammer effects at the ptzcp discharge of the RCIC system.

212.98 In the steam condensing cperaticn of the RCIC system, (5.5.6) turbine speed centrol is transferred frcm the flew control mode to the heat exchanger level centrol mcde. Is this transfer autcr::atic er manual?

j 212.99 Provide a realistic range and a permitted cperating band for the l

( 15E.0) exposure dependent paraneters in Tables 4.4-1 and 155.0-1.

In Table 153.C-1, previde assurance that values of parameters selected yield the most consenative results.

212.100 The correct value of APRM neutron flux scram setpoint to be

( 155.0) used in transient analyses is not clear. The value indicated as input fer transient analysis in Table 15E.0-1 is 122.4% NER.

Hewever, a value of 120% NER is indicated in Tables 7.2-1 and 7.6-10.

Explain this discrepancy. For the correct value of setpoint used in transient analyses, provide a breakdown of sny uncertainty allowances that are added to the ncminal value.

i 212.101 Fcr each transient and accident analyzed in Section 15E.0 include (152.C) secticns en the following items:

a)

Identificatien of Cperatcr Actions b) Effect cf Single Failures or Operator Errors Specify the transients and accidents in Section 152.0 for Q11ch operater acticn is required in crder to mitigate the event censequences. In those cases, provide justificatica for any ccrrective actices by the cperater af ter transient initiation.

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212-32 l

212.64A In Cuestien 212.64 and its respense, we noted that seme analyzed (15B.0) transients and accidents take credit for ncnsafety-grade systems er cc=penents. The main ecccern is that these events are using less reliable systems to shcw that the acceptance criteria are =et.

If caly safety rade systems er ccmpenents were used, analysis results may show that receptance criteria are net met, a) Identify each nemally cperating system for which credit has been taken for each transient and accident analyzed in Secticn 152.0.

b) Provide a table of the ncnsafety-grade systems and cccpenents assmed to mitigate censequences for each transient and accident. For each system er ccmponent, reference er provide a descriptien of ics design basis and feacuras.

c) Podify NSCA drawings in Appendix F.6 to include nensafety-grade systems er czponents which mitigate consequences of transients and accidents, d) Provide the A(ACPR) and A(a peak vessel pressure) for each transient and accident that takes credit for specific ncnsafety-grade systems er eccpenents that sculd result if only safety-grade systems or ccupenents were assuned in the analysis.

212.102 In the analysis of transients and accidents in Sechien 15B.0, there (152.0) is no descriptien of the RPS ti=e respense delays used in the REDY analytical tredel (NECC-10802). For each transient and accident in Secticn 152, specify whether the senser er everall delay ti=e is used in the analysis and why the specified delay time is censervative.

212.103 Ccnfirm the fellowing items fer all transients and accidents in (153.0)

Section 15B.0 which require centrol red insertion to prevent er lessen plant danage!

a) All calculatiens were perfor=ed with the censervative scram reaccivicy curve No. 2 in Figure 153.0-2.

b) The slowesc allowable scram insercica speed was used.

212.104 Fcr transients and accidents analyzed in Secticn 153.0, credit was (152.0) taken fcr safety / relief valve (SRV) actuatien in the safety =cde, based en the respcase to C212.20. For the pilet-cperated Target Rcck safety / relief valves used in the Enrico Femi 2 design, credit for caly 3/4 cf the rated capacity (11 cr less valves) is allowed for safety mode actuatien in acccrdance with ASME E & PV Ccde Section III, NE-70CO. Credit for more than 3/u of rated capacity is assumed fer scme transients in Table 153.0-2.

a) Re-analy:e all transients and accidents in Section 15E.0 that take credit fer =cre than 3/4 cf ratec SRV capacity taking credit for enly 3/4 cf the rated espacity.

b) Explain why SRV actuatien in the safety mcde is indicated as relief valve ficw instead of safety valve ficw in the figures asscciated with each transient and accident.

212.105 Ciscuss hcw the pre-cceratienal anc startup tests will be used to (153.0) confin ficw par =neters used in Secticn 152.0 analyses. Previde details. cf any previous test of ecmpenents in test facilities conducted to shew satisfactcry perfer=ance of the recirculatien and feedwater ficw centrcl systems and respective pumps. Describe how this infcmaticn was used in Secticn 153.0 analyses.

212-33 212.106 Cn page 4-7 of N~c4-1Co02, it' is stated that the difference in (152.0) trend of flew coastdown versus initial power between the analytical and experimental coastdown curves for Cresden Unit No. 2 (a EWR/3) in Figure 4-11 was due in part to differences between actual and et=puted jet pt=p efficiencies.

a) How has this effect been treated in the analysis of Enrico Fenti 2 transients involving ficw coastdown with tse recirculation ptmp trips?

b) Explain why this treatment is applicable to Enrico Fermi 2 which is a B'a'd/4.

212.107 Table 153.0-1 does not contain all of the input paramaters used in the (15B.0)

RIDY computer code. For each transient and accident analyzed in Section ISB.O. provide the following:

a) A list of all input parameters.

b) Justification that the input parameters are conservative.

1!2.108 The assumed feedwater flow controller failure at 1,20% flow appears low (153.1.2 compared to a failure value of up to 146% flow used in other FSARs.

.3.2)

Explain the basis for the assu=ed feedwater flow c'entroller failure at 120% flow.

212.109 Safety / relief valve (SIV) actuation for this transient is not included (153.1.3 in Tables 153.0-2 and 153.1.3-1 and Section 153.1.3.5 of the FSAR for 3.3) decay heat renoval. Eowever, it is included in Figure 153.1.3-1.

Explain this discrepancy.

212.110 The scean flow increase of 130% in Section 153.1.3.3.2 is not simulated (153.1.3 for the " pressure regulator failure-open to 130%" transient at time = 0

.3.3) in Figure 153.1.3-1.

In addition, a pressure increase is indicated during the initial portien of the transient when a pressure decrease should first occur. Please explain.

212.111 With regard to the " pressure regulator failure open to 130%" transient:

(153.1.3

.3.2) a)

Specify the assened operating mode (zanual or automatic) of the 1

recirculation flow control system and provide justification that 1

the most conservative results on core thermal margins are obtained

{

with the assuned operating mode.

b) Explain the discrepancy between the high flux peaks shavn in Figure 153. 1.3-1 and the sna11 value on Table 153.0-2.

212.112 The sequence of events in Table 153.1.3-1 is.not present in sufficient (153.1.3 detail to follow the process variations in Figure 153.1.3-1.

For 2.1) exanple, neither safety / relief valve actuation nor recirculation pump i

rusback, nor initial core cooling are included. Provide a more detailed Table 153.1.3-1 between 0 and 40 seconds.

1 212.43A The response to Q212.43 is not acceptable. Provide a quantitative (15B.1.1 analysis for operation with partial feedwater heating.

.1.1)

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1 212-34 212.42A The response to C212.42 is not acceptable. Previde the folicwing (155.2.

infermation:

2 3 2) a) Confirm quantitatively that evaluatien of this transient at icwer power with Nu stroke turbine centrol valve (TCV) closure does not result in a mere restrictive AMCPR and peak l

vessel pressure.

b) In Section 10.2, a TCV full are closure time of 0.18 to 0.22 seconds is indicated. The maximun fun are closure time should not exceed 0.18 seconds for transient analysis. Explain ey i

the previous analysis used a conservative 0.15 seconds for j

fun stroke closure and the current analysis uses a ncn-conservative value of 0.20 seconds. Provide an analysis with a N11 stroke TCV closure time of 0.18 seconds.

c) Confim quantitatively that evaluation of this transient at 4

All pow with partially cpen TCVs does not result in a more restrictive AKPR and peak vessel pressure.

212.113 For the " turbine trip" and "MSIV closure" transients, a full stroke (153.2.

closure time of 0.20 secends was used for the turbine stop valves 3 3 2)

(TSV) in Section 15B.2 3 There is no descriptien of an acceptable TSV Nu stroke closure time span in Secticn 10.2.i Similar inferma-tion was presented for the turbine centrol valves. Previde the following infor. nation:

a) The acceptable full stroke closure time span for the TSV.

b) Verificaticn that the TSV fun stroke closure time used for transients in Section 153.2 3 and 153.2.5 was the minimm value of the ti=e span cr less. If not, re-analyze these transients with a TSV full stroke elesure time less than er equal to the minimta value of the acceptable time span.

212.114 Provide the following information regarding the " turbine trip with (15B.2.3 partial steam bypass failure" transient:

J

.3.3) a) The EPR for the " turbine trip with partial steam bypass failure" 4

l transient is specified as 1.14.

However, this dces not agree with a value of 1.18 in Table 15B.0-2.

Explain this discrepancy.

b) Provide analytical results (figures) for the " turbine trip with steam bypass" transient. Revise Table 15B.2 3-1 to include the same level of detail as included in Table 153.2 3-2.

212.115 Curing the " turbine trip with partial steam bypass" transient,

)

(155.2.

explain why feedwater ficw in Figure 153.2 3-1 decreases to zero 2 3 2 2) price to the !.8 vessel level setpoint trip of the feedwater ptaps i

at 31.7 ' seconds in Table 153.2 3-2.

212.116 Include the time at which the foncwing eccur in Table 153.2.5-2 (153.2.

a) Turbine stop valves are closed.

5.2.1) b) Feedwater pumps are tripped.

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212.117 It is indicated that credit is taken for safety / relief valve operation (15B.2.7 with " low setpoints" to remove decay heat since bypass valves become

.2.2) ineffective with MSI7 isolation. Specify the values of these " low setpoints" and provide justification if they are lower than the values in Table 15B.0-1.

212.113 For the " trip of one recirculation pump" transient, the text indicates (15.3.1 that no scram occurs and core flow and power level should stabilize at

.2.1.1) a new equilibrium condition. In Table 153.3.1-1 and Figure 153.3.1-1, a scram is indicated via an L3 initiated turbine trip. Correct this discrepancy and update Table 153.3.1-1 and Figure 153.3.1-1 accordingly.

212.119 Tor the " trip of both recirculation pumps" transient, an L8 initiated (15.3.1 turbine trip scran occurs. Typically, this results in the pressure

.2.1.2) relief, reactor vessel isolation, and initial core cooling safety

actions, a) Explain why these safety actions do not occur for the Enrico Fermi 2 design.

b)

Since the scram safety action occurs for this' transient, provide an NSOA figure for event 27 in Appendix F.6. ' Include the safety actions in step a) above if they occur and update the text description for event 27.

212.120 For the " recirculation pump seizure" accident, coincident loss of (153.3.3 offsite power is not simulated with the assumed turbine trip and

.2.2) coastdown of the undamaged pu=p.

The current analysis also takes credit for nonsafety-grade equipment (LS trip) to terminate the event.

Reanalyze this transient assuning coincident loss of offsite power and the use of only safety-grade equipment (see Q212.64A).

212.121 Regulatory Guide 1.45 recemends that "The leakage detecticn systems (5.2.7) shculd be capable of performing their functiens fc11cwing sei:=rie events that do not require plant shutdown. The airborne particulate radicactivity mcnitcring system should remain functional when subjected to the safe shutdown earthquake." It is cct clear frem the FSAR whether the abcve criteria are met er not. Discuss the seismic capabilities of the leakage detection system ccmponents.

212.122 Cn Figure 5.5-16, sheet 6 of 6 (Amendment 3), the following valve (5.5.7) positiens are stated for RCIC operatien in the steam condensing cede:

Valves FC51, FC52, F053, 026, F607 cpened Valves FCO3 and F047 closed 1bwever, the figure itself shchs:

Valve F003 cpen. Valves FC47 and F607 are net indicated en the figure. Further, the pcsitiening of Valve F606 is not addressed.

Ccerect the discrepancies.

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212.123 Anend the ISAR as inficated belw.

1.

Figure 6.3-3 sheet 3 shers a pressure switch on the discharge of each of (6.3) the core spray pc=ps. m se ;ress= e switches are used as a permissive in the.05.legic.

Stee:s 4, 5, and 6 of Figure 6.3-3 show only one pressure switch,per loep.

""he pres.ure switch identification numbers also changed in going f:cm staat 3 :o sheets 4, 5, and 6.

Clarify.

2.

Table 6.3-12 sca:e:s that the lov va. e level (L2) initiation signal for (6.3)

EPCI activation is set at 10.26 feet above the top of the active core.

Figure 5.1-3 states that EPCI ini:iation occurs at water level L2 (10.75 fee: above the top of the active core)'. Resolve this inconsistency.

3.

The following er: ors were found in the FSAR and should be corrected:

(6.3) a) Normalized core average flow shou *d be a dimenstionless quantity; Figure 6.3-12 speci.fies psia u=its for the normalized core average flow.

b) Section 5.3.3.7.3 of the FSAE incorrectly speci ies that the n:aximun pe:mitted pealc c* adding ta=perature is 220 ? rather than 2200 7.

c) Section 6.3.2.4 of the FSAR identifies Dresser as the safety / relief valve supplier. Fer=1 safety / relief valves are Target Rock two-stage valves.

4.

Correct the subsectica references en page 15.1.6-1.

Inadvertent HPCI (15.1) star: is covered i= Subsectica 15.3.5.1 (not 15.3.4.2).

Startup of an idle recircula:ic= p sp is c:vered in Subsection 15.B.4.4 (not 15. 3.5.1).

5.

Correc: the foll:ving inco=siste:cy:

(15.1)

On page 15.1-6 y:u s:a:e that, " Maintaining a MCPR greater than 1.07 is a sufficient, b se not a necessary :endition to ensure that no fuel da= age CCCurs."

On page 153.0-7 ycu sente that, % 4--=4"4"g a MCPR greater than 1.06 is a suffician:, but not a necessa_y condition to ensure that no fuel da=a3e Occurs."

6.

.1dd :he i=itial ecre coo 2=g saft:. a:tiens to Tables 153.2.5-2 and (153.2.5.2) 15'3.2. A-1 :o be censis:e:: vich Figures F.6-9 and 7.6-7, respectively.

7.

Correct the folic-dig:

(153.2.6.2) a) Add the reac::: vessel a:d cen-='-ent isolation safety actions to Figura 7.6-15 for _he "1 css of all grid connections."

b) Add the folleving ite=s to Table 133.2.5-1 to be consistent with the proposed revisie: :o Tig:re 7.5-15 for the " loss of all grid connections"

a=sient:

1)

Rastora 'en of.AC pcver

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2) Extended =cre coolins D

3)

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