ML20062B634

Forwards Request for Addl Info for Safety Review of OL Application.Response Requested within 10 Days

ML20062B634
Person / Time
Site:	Seabrook
Issue date:	07/27/1982
From:	Miraglia F Office of Nuclear Reactor Regulation
To:	Tallman W PUBLIC SERVICE CO. OF NEW HAMPSHIRE
References
NUDOCS 8208040644
Download: ML20062B634 (21)

Text

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g c . N . 'g' s JUL 2 71982 DISTRIBUTION:

, Document Control (50-443/444)

NRC PDR L PDR Docket Nos.: 50-443 PRC and 50-444 NSIC LB#3 Rdg.

William C. Tallman  : FMiraglia Chairman and Chief Executive Officer JLee Public Service Company of New' flampshire LWheeler Post Office Box 330 DEisenhut/RPurple Manchester, New Hampshire 03105 RLessy I&E

Dear Mr. Tallman:

ACRS (16)

Subject:

Request for Additional Information The NRC steff has determined 'that additional information is required for the safety review of the Seabrook operating license application.

Enclosed are the following Requests for Additional Information (RAIs):

Hydrologic and Geotechnical Engineering Branch (HGEB)(240.38-41)

Quality Assurance Branch (QAB)(260.28)

Auxiliary Systems Branch (ASB)(410.51-55)

ReactorSystemsBranch(RSB)(440.136)

OperatorLicensing. Branch (0LB)(610.1-3)

The staff is available to dis' cuss all of the above RAls as may be required to provide any necessary clarification. In most cases these RAIs have been discussed with your representatives in past meetings and are forwarded herewith to formally document,' staff requirements.

Your responsesto these RAIs should be forwarded to the NRC staff within 10 days of receipt of this request. The Seabrook Project Manager (Mr. L. Wheeler, 301/492-7792) is available to respond to any questions your staff may have.

. Sincerely, Original Signed By:

Frank J. fliraglia, Chief Licensing Branch No. 3

}

Division of Licensing

Enclosure:

RAIs as stated cc w/ encl.:

See next page 8208040644 820727 PDR ADOCK 05000443 A PDR o"* .. L.B#3fq.DL .

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8' NRC FORM 318 4tO.eOI NRCM O2dC OFFICIAL RECORD COPY * '*

• 32**2'

t William C. Tallman Chainnan and Chief Executive Officer

. Public Service Company of New Hampshire P. O. Box 330 Manchester, New Hampshire 03105 John A. Ritscher. Esq. E. Tupper Kinder. Esq.

P. opes and Gray Assistant Attorney General 225 Franklin Street Office of Attorney General boston, Massachusetts 02110 208 State house Annex Mr. Bruce B. Beckley. Project Manager Public Service Company of New Hampshire P. O. Box 330 Manchester. New Hampshire 03105 G. Sanborn U. S. NRC - Region I ,

631 Park Avenue Resident Inspector King of Prussia. Pennsylvania 19406 Seabrook Nuclear Power Station 4 c/o U. S. Nuclear Regulatory Commission P. O. Box *700 Seabrook, New Hampshire 03874 Mr. John DeVincentis. Project Manager Robert A. Backus, Esq. Yankee Atomic Electric Company O'Neill Backus and Spielman 1671 Worcester Road 116 Lowell Street Fanningham, Massachusetts 01701 Manchester, New Hampshire 03105 Mr. A. M. Ebner Project Manager Norman Ross, Esq. United Engineers and Constructors 30 Francis Street 30 South 17th Street Brookline. Massachusetts 02146 Post Office Box 8223 Philadelphia, Pennsylvania 19101 Karin P. Sheldon. Esq.

Sheldon. Hannon & Weiss Mr. W. Wright, Project Manager 1725 1 Street. N. W. Westinghouse Electric Corporation Washington. D. C. 20006 -

Post Office Box 355 Pittsburg. Pennsylvania 15230 Laurie Burt. Esq.

Office of the Assistant Attorney General Thomas Dignan. Esq.

' Environmental Protection Division Ropes and Gray e One Ashburton Place 225 Franklin Street Boston. Massachusetts 02108 Boston. Massachusetts 02110 D. Pierre G. Cameron. Jr., Esq. Mr. Stephen D. Floyd General Counsel Public Service Company of New Hampshire Public Service Company of New Hampshire P. O. Box 330 P. O. Box 330 Manchester, New Hampshire 03105 Manchester. New Hampshire 03105 i

- Lf; CLOSURE

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240 HYDROLOGIC' AND E0TECHNICALEtiGIfiEERIU5B55CH' 240.38 In your response to Question 240.32 (Hydrologic Engineering Question (2.4.3) 240.02) you stated that the PMF on Hampton Harbor watershed combined with the PMH will increase the stillwater level at the plant site less than 0.1 feet above that calculated for the SPF combined with the PMH.

However, no detailed analysis has been provided to support this '

assertion. Provide detailed analysis supporting this contention.

240.39 Provide an evaluation of the effect on the wave overtopping rate  ;

(2.4.5) resulting from the increased Design Stillwater Level using the combined PMF/PMil rather than SPF/PMH event.

240.40 In your response to Question 240.34 (Hydrologic Engineering Question (2.4.5) 240.04) you indicated that wave overtopping will not cause significant erosion because of its short duration. Our analysis indicates that the peak wave overtopping rate of the vertical seawall is in excess of 1600 cfs for a period of about 0.2 hrs. We conclude that this could result in the loss of fill material behind the vertical seawall and adjacent to the two -class I electrical manholes (#13/14 and !15/16).

Discuss the consequences of this loss of fill material or describe the measures planned to prevent it.

240. 41 It is not apparent from our review of the ponding level on plant grade (2.4.2) that concurrent intense precipitation was included in your wave overtopping (2.4.5) runoff /ponding analysis. Therefore, provide a detailed analysis on the routing of the combined precipitation runoff from Probable liaximum Precipitation and wave overtopping runoff from the PMF/PMH event. .

a) If credit is taken for flow through the storm drainage system, provide justification that the stonn drainage system cannot become blocked during this event.

b) Identify the maximum water surface levels by location and elevation from the vertical seawall to the overflow weir (seawall).

c) Identify plant access openings and sill elevations that may be affected by the runoff on plant grade.

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260.0 Quality Assurance Branch 260.28 Section 17.1.2.2 of the standard format (Regulatory Guide 1.70) requires the identification of safety-related structures, systems, and components controlled by the QA program. You are requested to supplement and clarify the Seabrook FSAR in accordance with the following :

a) The following items do not appear on FSAR Table 3.2-1, Table <

3.2-2, Appendix 3H or Section 17.2.2.2. Add the appropriate items and provide a conmitment that the remaining items are subject to the pertinent requirements of the Operational FSAR QA program or justify not doing so.

1. Fuel assemblies

2. Core support structure

3. Control rods

4. Control rod drive mechanisns

5. Steam generator steam flow restrictors

6. Containment building polar crane

7. Cask handling crane

8. Spent fuel pool liner

9. Biological shielding within the primary auxiliary building '

and fuel storage building

10. Missile barriers within the primary auxiliary building, fuel storage building, and other buildings and structures as appropriate

, 11. Pressurizer PORY block valves

12. Fuel trans'fer system _and controls Refueling machine 13.

14. Spent fuel pool bridge and hoist

15. Containment interior concrete including emergency sump

16. Operators of safety-related valves

17. Supports for safety-related ducts, pipes , valves, motors , etc.

18. Motors for safety-related pumps

19. Containment emergency sump debris screen

20. Containment enclosure ventilation area ducting

21. Intake and discharge structures b) The following items are in Table 3.2-2 with no indication that 10 CFR 50 Appendix B applies. Provide a commitment that the pertinent requirements of the TSAR Section 17.2 QA program will be applied to these items during the operations phase or justify not doing so.

1. Diesel generator cooling water systems (p 25)

(a) Auxiliary coolant pumps (b) All remaining on-engine equipment and piping

2. Diesel generator starting systems (pp 25-26)

(a) All remaining equipment and piping

3. Diesel generator lubrication systens (p 26)

(a) Auxiliary lube oil pumps (b) All remaining on-engine equipment and piping

4. Diesel generator combustion air intake and exhaust systems (p 26)

(a) Intake silencers (b) Air intake filters

( c) Exhaust silencers -

5. Diesel generator fuel oil storage and transfer systems (pp 24-25)

(a) Fuel oil pumps (b) All remaining on-engine equipment and piping

6. liydrogen analyzer (p 5) c) Add the following items to sheet 4 of Table 3.2-1, "Onsite Power Systems" or justify not doing so.

Standbv AC Auxiliary Power Systems (Class lE)

1. Diesel generator packages including auxiliaries (e.g. , governor, voltage regulatory and exci.tation system).

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2. Instrumentation, control and power cables (including underground cable system, cable splices, connectors and terminal blocks)

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3. Conduit and cable trays containing Class lE cables and their ,

supports and other raceway installations whose failure during a seismic event could damage other safety-related systems or components

4. Valve operators

5. Protective relays and control panels

6. Electrical penetration for containment - vital and non-vital including primary and backup fault current protective devices .

7. Emergency lighting battery packs -

8. AC vital bus distribution equipment DC Power Systems (Class lE_)

1. Cables i

2. Conduit and cable trays containing Class lE cables and their l supports and other raceway installations whose failure during l a seismic event 'could damage other safety related systems or l

components

3. Battery racks

4. DC switchgear, distribution panels and protective relays '

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' Provide a commitment that modifications of the site and roof drainage d) ,

systems, the seawall, retaining walls, and : revetmen ments of the operational QA program to ensure against increasing the _

I flood vulnerability of safety-related items.

Provide a commitment that the safety-related instrumentation and f e) controls (I&C) described in Sections 7.1 through 7.6 of the FSAR l

plus safety-related I&C for safety-related fluid systems will be subject to the pertinent requirements of the FSAR QA program..

f) Enclosure 2 of NUREG-0737, " Clarification of TMI Action Plan Require-ments," (November 1980) identified numerous items that are safety-l related or of such importance to safety that they should have the pertinent requirements of the FSAR Operational QA program applied.

These items are listed below.

or 17.2 of the FSAR or justify not doing so.

NU REG-0737 j (Enclosure 2) g Clarification Item a

• I.D.2

1. Plant-safety-parameter display console II.B.1

2. Reactor coolant system vents II.B.2

3. Plant shielding f

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C*G NUREG-0737 (Enclosure 2)

Clarification Item

4. Post-accident sampling capabilities II.B.3 j 5. Valve position indication I I . D. 3

6. Auxiliary feedwater system II.E.1.1

7. Auxiliary feedsater system initiation II.E.1.2 and flow

8. Emergency power for pressurizer heaters II.E.3.1

9. Dedicated hydrogen penetrations II.E.4.1 _

10. Containment isolation dependability ^

II.E.4.2

11. Accident monitoring instrumentation II.F.1

! 12. Instrumentation for detection of inadequate II.F.2 a

core cooling

13. Power supplies for pressurizer relief valves, block valves, and level indicators II.G.1

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14. Automatic PORV isolation II.K.3.1

16. PID controller II.K.3.9

17. Anticipatory reactor trip on turbine trip II.K.3.12

18. Power on pump seals II.K.3.25 j 19. Emergency plans III.A.l.1/III.A.2 l 20. Emergency support facilities III.A.l.2

} 21. Inplant 1 radiation monitoring 2 III.D.3.3 ^

22. Control room habitability

• III.D.3.4 l g) Section 17.2.2.2a should reference Table 3.3-1, Table 3.2-2, and 1

Appendix 3H for the identification of items controlled by the pertinent requirements of the FSAR Operational QA program.

. h) Section 17.2.2.2f references FSAR Section 12.5.3.8 regarding audits of the Health Physics program. Clarify the involvement of the PSNH QA organization in these audits. The QA organization should either perform the audits, furnish audit team leaders, or audit to verify the audits are in accordance with the commitments of the FSAR Operational QA program.

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410 AUXILIARY SYSTEMS BRANCH 410.51 At a meeting with the staff on June 23rd,1932, the applicant took the posi-tion that the staff's requirement for a source range neutron flux monitor (SRM) on the remote shutdown panels was not necessary, since the applicant meets the Appendix, R requirements for a " direct-reading of reactivity

} with an intermediate range neutron flux monitor (IRM) on the remote shutdown -

panels. -

In order for us to evaluate whether the IRM can adequately perform the j

functions expected of the SRM, the applicant should provide the following information:

a. Provide a diagram of the operable ranges of the SRM, IRM and power range

neutron fiux monitor (PRM) as a function of power level. Indicate the levels to be expected in a nonnal shutdown (normal T and K) as a function of time after shutdown (over several hours);

j b. State at what point on the IRM scale criticality would be expected to oc' cur for dilution starting at different times after shutdown;

c. Discuss the effect of reactor coolant temperature on IRM readings [ Lower temperature causes more attenuation. Sensors are calibrated for high temperature]; '

d. Discuss the response times of the operator during an increase in reactivity if the first alarm comes from the IRM vs SRM.

1 410.52 In Sections 3.1.1.4 and 3.1.2.1 of Fire Protection of safe _ shutdown capability, j the applicant assumes that the operator will trip the reactor, will trip all

)' four rea ctor coolant pumps and will close all four main steam isolation valves prior to evacuation of the main control room. Additional information c to verify this capability is required. It is our position that in the event I

of a fire which rapidly makes the control room uninhabitable allowing the operator only time to trip the reactor, that the capability to trio the four reactor coolant pumps (RCPs) and close the four MSIVs- _be provided out-side the main control room, in the event offsite power is maintained or icst. Verify that failure to trip the RCPs or close the MSIVs in the event of a control room evacuation does not result in an unacceptable plant condition, or verify whether the RCPs can be tripped and MSIVS closed outside the control room, that the delay in doing so will not result in a violation of l any of the criteria as listed in Section III.L of Appendix R to 10 CFR Part 50.

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A 410.53 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 toe applicant to review associated circuits of concern and the information to be provided for staff evaluation. The applicant shou,ld address Part II .C.

of the enclosure.

410.54 The applicant should commit to develop and implement alternate shutdown pro-cedures. These procedures should address manpower reo.uirements and manual actions to accomplish shutdown. A summary of these procedures should be provided for our review.

410.55 The applicant's submittal does not indicate whether repairs are required i to achieve safe shutdown. It is our position that systems and components '

used to achieve and maintain hot standby conditions must be free of fire damage and capable to maintain such conditions for the duration of the hot standby condition without repairs. Systems and components used to achieve and maintaiq cold shutdown should be either free of fire damage or the fire damage to such syste= should be limited such that repairs can be made and cold shut-

, down achieved within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Repair procedures for cold shutdown systems l must be developed and material for repair maintained onsite. It is our i position that electrical or pneumatic jumpers are not a suitable method of repair for cold shutdown.

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EICLO5URE ASSOC:ATED C2RCUIT GUIDANCE I. ~ INTRODUCTIOf-The following discusses the requirements for prctecting redundant and/or ziternative equipment needed for safe shutdonr, in the event

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The of a f req 0irements of Appendix E address hot shutdoon equipment which mu free of fire damage.

The foliewing_tequirements aisc apply to cold shutdown equipment ,

if thy licensee elects to de$onstrate that th.e equipment e_ be is t free of fir,e. damage. ,

Appendii. E does allow re'pairable damage to cold shu ecuipment. .

Using the requirements of Sections ))I.G and III . L cf App endix R, the' capa-bility 'to achieve het shutdown must exist given a fire i n any area of the plant in conjunction with a loss'of offsite power for 72 hou rs. Section Ill.G cf Appendix R provides four methods for ensuring that the hot shutdown capa-bility is protected from fires.

The first three options as defined in Section 131.G.2 provides methods for protectica- from fires of e quipment needed for hot shutdown:

1.

Redundant systems including cables, equipment, and associat d e circuits i

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may be separated by a three-hour fire rated barrier; or ,

2.

Redund nt systems .tncluding cables, equipment and associated circuits may be separated by a horizontal distance of more than 20 fe t vening ccmbustibles.

e with no inter-In addition, fire detection and an autcmatic fire suppression system are required; or,

3. .

Eadundant sys: ems including cables, eqcipment ar.d ed associat circuits may i

Le en:lcsed by a one-hour fire rated barrier.

In addition, fire detectors Ir.d an au cmatic fire suppression system areu red. req i

The last cption as defined by Section III.G.3 provides an alternative shutdown capability to the redundant trair.s damaged by a fire.

4. Alterdative shutdown equipment mus- be independent of the cables, equip-ment and associated circuits cf the redundant systems damaged by the fire.

II. Associated Circuits of Concern The following discus'sion provides A) a definition of associated circuits for Appendix R consideration, B) the guidelines fer protecting the safe'shutdtwn ,

, capability from the fire-induced failures of associated circuits and C) the in-formation reguired by the staff to review associated circuits. It is important to note that our interest is oniy 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 -

i available 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 .w.ill receiv.e f. ir.e dama.c.e 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,Ciass lE,' and non-Class lE) that:

'The definition for associated circuits is not exactly the same as the definition presented in IEEE-334-lg77.

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1. Have a physical separation-less than that required by Section III.G.2 of Appendix R, and;

2. Jiave. one of the following:

a. a co=on power source with the shut own equipment (redur. dant or alternative) and the power scur:e is not ele rically protected .

from the circuit of concern by cc:rdinated breakers, fuses, or

• similar devices (see diagram 2a), or .

b. a connection to circuits of equipment whose spurious operation would adversely affect the shutdown capability (e.c., RHR/RCS isolation valve:, ADS valves, PORVs, steam generator a= spheric dump valycs, instrumentation, steam bypass, etc.) (see diagram 2

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c. a comon enclosure;(e.g., raceway, panel, junction) with the shutdown cables (redundant and alternative) and, .

(1) are not electrically protected by circuit breakers, fuses or simi-lar devices, or propagation of the fire into the cc. men (2) will allow en:losure, (see diagram 2:).

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B. The following guidelines are for protecting the shutdown caoability 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 concern and -

the shutdown circuits as per Section III.G.2 of Appendix p., or

2. a. For a common power source case of associated circuit:

Provide load fuse / breaker (interruptint devices) to feeder fuse / breaker coordination to prevent loss of the redundant or alternative shutdown power source.

1

! To ensure that the following coordination criteria are met the 'fo')owine T should apply:

(1) The associated circuit of concern interrupting devices '

(breakers or fuses) time-overcurrent trip characteristic ~

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l for all circuits faults should cause the interruotinc t

device .to interrupt the fault current prior to initiation of a trip of any upstream interrupting device which will cause a loss of the conaan 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 shutdown loads. .

I

I The acceptability of a particular interrupting device is censidered demonstrated if the following criteria are cet:

(i) The interrupting device design shall be factory tested to verify over:urrent pr:tection as designed in a :ordance with the appiitable UL, AH5I, or NEMA standards.

(ii) For low a'nd medium voittge switchgear (430 V and above) circuit breaker /prote:tive relay periodic testing shall demonstrate that the overall coordination scheme remains within the litits specified in the design criteria. This testing may be performed as a series of overlapping tests.

(iii) P.alded case circuit breakers shall periodically be raanually exercised and inspected to insure ease of operation. On a rotating refuelin; cutage basis a sa= pie of these breakers ~

shall be [ tested to determine that breaker drift is within that allowed by the design criteria. Breakers should be tested in accordance with an accepted QC testing methodology such as MIL STD 10 b D.

(iv) Fuses when used as ir.terrupting devices do not recuire periodic testing. Administrative controls must insure i

that replacement fuses wit.h ratings o'ther than those selected for proper coordination are not accidentally used.

I

b. For circuits of equipment and/or cc penents whose spuricus cperation would affect the capability to safely shutdown:

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-E-i (1) provide a means to isolate the equipment and/or components from the fire area prior tc the fire (i.e., remove power cables, open circuit breakers); or (2) provide electrical isolation that prevents :purious operation.

Potential isciation devices include breakers, fuses, ampli-fiers, control switches, current XFRS, fiber optic couplers, relays and transducers; or (3) provide a means to detect spuricus operaticas and then pro:e-dures to defeat the maleperation of equipment (i.e., closure of the block valve if PORY spuriously operates, opening of the breakers to Femove spurious operation cf safety injection);

c. For common enclosure cases of as:ociated circuits: ~

(1) provide appropriate measures to prevent propagation of the fire; and i.

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! (2) provide eie:trical protection (i.e., breakers, fuses or l

similar devices) l C. INFORMATICH REGUIRED The fo'llowing information is recuired to demonstrate that associated circuits will not prevent operation or cause caloperation of the shutdown method:

. , _ . . . - . . .. .. . . .. . = . = = .

1,

a. Describe the methodology used to assess the potential of associated citcuit adversely affecting the shutdown capability. The description of the methodology should include the 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 7

methods used to identify if these . circuits are associated circuiys of concern due to their location in the fire area.

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b. Show that fire-induced failures (hot shorts, open circuits or shorts to ground) of each of the associated circuits of concern will not prevent operation of cause maloperation of the shutdown method.

, 2.

! The residual heat removal system is generally a low pressure system l

that interfaces with the high pressure primary coolant system.

To 4

i preclude a LOCA through this interface, we require compliance with the recommendations of Bran:h Technical Position P.SE 5-1.the Thus, i

i interface most likely consists of two redundant and independent notor operated valves.

These two motor operated valves and their associated cib'les may be subject to a single f. ire-hazard.

It is our con:ern that l

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 -

t pressure interfaces are adequately protected from the effects of a single fire, we require the following information:

a.

Identify each high-los pressure interface that uses redundant electrically controlled cevices (such as two series motor tocrated valves) to isolate or ore:Waw- "

c .

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b. For each set of redundant valves identified in a., verify the redundant cabling (power and centrol) have adequate physical separation as required by Se:tien III.G.2 of Appendix P..

c. For each case wher.e adequate sc;; ration is r.:t previde_, sh:.. th:t fire induced failures (hot short, open circuits or short to tround) '

of the cables will net cause maloperation and result in a LOCA.

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440 REACTOR SYSTEf15 BRANCH 440.136 The recent steam generator tube rupture (SGTR) event at R. E.

(15.6.3) Ginna Plant and previous SGTR events at other PWRs indicate the need for a more detailed review of the analysis of this accident. Our review of Seabrook FSAR section 15.6.3 (SGTR) and your response to AEB Question 450.4 on this subject resulted in several questions and a need for the following additional information and clarification.

(1) FSAR Section 15.6.3 indicates equalization of primary and secondary pressure 30 minutes after the SGTR event, with consequent termination of steam generator tube leakage. However, Figure 1 of your response to Question 450.4 indicates a minimum primary pressure of 1700 psia at approximately 600 seconds, followed by a rise to 2100 psia at 1800 seconds. Explain this discrepancy and

modify your.ana' lysis of this event acccrdingly, including consideration of longer leak times if indicated by these results.

(2) Demonstrate that your assumption of secondary relief actuation at 1236 psia (

Reference:

Table 2 of your response to question 450.4) is conservative from a radiological standpcint in view o# the fact that the set points for the atmospheric dump valve and the lowest safety valve are 1135 psia and 1185 psia, respectively.

~,

1. 80 (3) Clarify whether you have analyzed a case which considers the radiological effects of a SGTR with the highest worth control rod stuck out of the core, with equilibrium iodine concentration, including the effects cf any additional fuel failure caused by this event. (

Reference:

SRP Section 15.6.3, Subsections II (1) &

III.7)

(4) Discuss whether as a result of possible modification of your "

analysis, including consideration of longer leak times as discussed in item (1), liquid can enter the main steam lines and what the effects would be on the integrity of the steam piping and supports.

Consider both the liquid dead weight and the possibility of water hammer.

(5) Table 1 in your response to Question 450.4 (Sequence of Events) does not provide all the information requested. Provide the time of turbine trip and loss of offsite power, the setpoints for system actuations, and operator action times. Clarify the flow termination time for main feedwater, which is indicated at 302 seconds in the table while the text indicates that main feedwater flow is terminated by the safety injection signal which occurs at 555 seconds.

(6) In view of the fact that the emergency feedwater turbine drive steam flow cannot be terminated from the control room, provide the results of activity and dose calculations from the turbine steam exhaust for the duration of the tube leak.

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610 OPERATOR LICEt1SItiG BRAf1CH 610.1 Deleted.

610.2 Reference FSAR page 13.2-2, Item 4 (Replacement Training)- Provide details of replacement training program and how this replacement training is applied to operators with different previous experience.

610.3 Reference FSAR page 13.2-7, Item 13.2.1.3-6 (On-the-Job Training)-

Provide documentation of conformance with the requirements of H. R. Denton's letter of March 28, 1980 on Qualification of Reactor Operators (see Enclosure 4 of the letter).

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