ML20023B883

Forwards Request for Addl Info to Complete Evaluation of Application for Ol.Fsar Should Be Amended to Include Info Requested.Response Requested by 830801

ML20023B883
Person / Time
Site:	Millstone
Issue date:	05/03/1983
From:	Youngblood B Office of Nuclear Reactor Regulation
To:	Counsil W NORTHEAST NUCLEAR ENERGY CO.
Shared Package
ML20023B340	List: ML20023B338 ML20023B883
References
NUDOCS 8305060648
Download: ML20023B883 (31)

Text

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NSIC MAY 31983 PRC System LB#1 Rdg i

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Docket No. 50-423 EDooli ttle DEisenhut/RPurple Mr. W. G. Counsil Attorney, OELD Senior Vice President ELJordan, DEQA:IE l

t Nuclear Engineering and Operations JMTaylor, DRP:IE Northeast Nuclear Energy Company Post Office Box 270 ACRS (10) i Hartford, Ccnnecticut 06101 i

Dear Mr. Counsil:

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Subject:

Request for Additional Information for Millstone Nuclear Power Station Uriit 3 i

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Enclosed are requests for additional infomation which the staff requires to complete its evaluation of your application for an operating license for Millstone 3.

These requests for additional information are the result of the l

staff's review of the information in your FSAR, You should amend your FSAR l

to include the information requested in Enclosure 1.

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Review of certain portions of your application has been delayed and questions concerning these portions are not included in this package. You will be advised of the schedule for transmittal of any questions resulting from this remaining review. Due to the expected number and nature of these questions, the staff expects that you will be able to provide your responses to these questions when 4

you respond to the questions in Enclosure 1; however, if necessary, you will be given ninety days from the date of transmittal to provide your responses.

i As you were advised in the letter from Darrell G. Eisenhut to you dated Janu-ary 31,1983 concerning acceptance of your application for docketing, only a i

single set of questions concerning your FSAR are being transmitted to you for i

responses. After the staff has reviewed your responses to these questions, a j

draft SER will be prepared to provide a basis for a series of meetings designed to close out open items.

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You should provide your response to the enclosed request for Additional infor-mation in the form of an amendment to the FSAR no later than August 1,1983.

l For further information or clarification, please contact the Licensing Project Manager, Elizabeth L. Doolittle (301/492-7134).

Sincerely, l

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3. J. Youngbloo4f j

8305060648 830503 B. J. Youngblood, Chief PDR ADOCK 05000423 Licensing Branch No.1 j

A PDR Division of Licensing e

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Mr. W. G. Counsil MAY 3 1983

Enclosure:

As Stated cc w/encls.: See next page 1

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l unc ronu ais oo-soj uncu ono OFFICIAL RECORD COPY-usoeo. mi-m

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u Mr. W. G. Counsil 3

Senior Vice President Nuclear Engineering and Operations i

- Northeast Nuclear Energy Company Post Office Box 270

' Hartford, Connecticut 06101 1

cc:

Mr. Harve L. Plasse, Manager R. W. Bishop, Esq.

City of Chicopee Electric Light Corporate Secretary Department Northeast Utilities Service Company 725 Front Street Post Office Box 270 Chicopee, Massachusetts 01014 Hartford, Connecticut 06101 Mr. Guy W. Nichols, President Mr. Richard T. Laudenat, Manager New England Electric System Generation Facilities Licensing i

20 Turnpike Road Fortheast Utilities Service Company' Westborough, Massachusetts 01581 i

Hartford, Connecticut 06101 Mr. Bruce R. Garlick Mr. J. C. Mattia Manager, Energy Supply Resident NRC Inspector Fitchburg Gas and Electric Light Office of Inspection and Enforcement Company U.-S. Nuclear Regulatory Commission l

655 Main Street Post Office Box 128 Fitchburg, Massachusetts 01420 Waterford, Ccnnecticut 06385 i

Mr. D. Pierre G. Cameron, Jr.

Mr. E. R. Foster, Director General Counsel Generation Constitution 45 Public Service Company of Northeast Utilities Service Company.

New Hampshire Post Office Box 270 1000 Elm Street Hartford, Connecticut 06101 Post Office Box 330 l

Manchester, New Hampshire 03105~

Mr. Bruce McKinnon Power Contracting Manager l

Gerald Garfield, Esq.

Massachusetts Municipal Wholesale i

Day, Berry & Howard Electric Company ~

One Constitution Plaza Post Office Box 426 Hartford, Connecticut 06103 Ludlow, Massachusetts 01056 John D. Fassett, Chairman &

Mr. Bruce H. Grier Chief Executive Officer U. S. Nuclear Regulatory Commission, The United Illuminating Company Region I 80 Temple Street 631 Park Avenue New Haven, Connecticut 06506 King of Prussia, Pennsylvania 19406 Mr. R. O. Bigelow Mr. Robert E. Busch Vice President-New England Power Project Manager-Millstone Unit 3-Company Northeast Utilities Service Company 20 Turnpike Road Post Office Box 270 Westborough, Massachusetts 01581 Hartford, Connecticut 06101 l

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ENCLOSURE 1 i

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 NORTHEAST NUCLEAR ENERGY COMPANY

'i DOCKET NO. 50-423 i

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s REQUEST FOR ADDITIONAL Ihf0RMATION L

MILLSTONE NUCLEAR POWER STATION, UNIT 3 d

DOCKET NO. 50-423 410.0 AUXILIARY SYSTEMS BRANCH

• 41 0.7 Provide the results of your analysis that missiles from the turbine

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(3. 5.1.1 )

driven auxiliary feedwater pump cannot damage other safety-related equipment. Provide a drawing of the moveable overhead missile shield over this pump and the provisions to ensure that this shield be in the proper place when the reactor is operating.

41 0.8 Verify that a seismic event will not result in gravity missiles j

(3. 5.1. 2) within the containment which could cause damage to essential systems required to assure a safe shutdown or result in unacceptable releases of radioactivity.

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410.9 Ve'rify that your analyses of interns 11y generated missiles have (3. 5.1. 2) considered the reactor head bolts as possible missile sources.

Verify also that secondary missiles, if any, generated by impact of the primary missiles identified in FSAR will not cause damage to essential systems required to assure a safe shutdown or result in unacceptable releases of radioactivity.

410.10 FSAR Section 3.5.2 is incorrect. As outlined in Regulatory Guide (3.5.2) 1.70, Se: tion 3.5.2 should describe the structure and system to be j

j f protected from externally generated missiles.

Verify that equipment, systems and components containing radio-active fluids are protected against tornado missile damage or assure L

that failure of unprotected components will not result in an unaccept-l i able release of radioactivity.

410.11 It is our position that the compartments which house the main steam (3. 6.1 )

lines, feedwater lines and the isolation valves be designed to i l consider the environmental effects (pressure, temperature, humidity) and potential flooding consequences from an assumed crack of one square foot. The essential equipment, including the atmospheric dump valves, main steam isolation valves and feedwater isolation valves and their operators, and the essential auxiliary feedwater

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pumps and associated equipment should be capable of operating in i

the environment resulting from the above crack.

Further, if this i ;

assumed crack could cause the structural failure of these compart-4 ments, then the failure should not jeopardize the safe shutdown of the plant.

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We, therefore, request that you submit a subcompartment pressure analysis to confirm that the design of the above compartments con-i forms to our position as outlined above. When you submit the results of your evaluation, identify the computer codes used, and the assumptions used for mass and energy release rates. The peak pressure and temperatures resulting from the postulated break of h

a high energy pipe located in these compartments are dependent

• This numbering system is a continuation of the numbers used in the OL Application Acceptance Review t.T

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.. REQUEST FOR ADDITIONAL-INFORMATION'.

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. MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET N0. 50-423

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410.0 AUXILIARY SYSTEMS BRANCH (CONT'D) onthemassandenergy,flowsduringthetimeofthebreak. There-fore, for the pipe break or crack analyzed, provide the total blow-down time and the mechanism used to tenninate or limit the time of blowdown flow so that. the environmental' effects will not affect j

safe shutdwon of the facility.

I Also provide a similar analysis for other com'partment's outside con-l tainment in the vicinity of safety-related structures, systems and components which house high energy lines such as CVCS charging, letdown and steam generator blowdown.

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410.12 Verify that the' fuel pool is not located in the vicinity of any -

(9.1.2) high-energy lines or rotating machinery"to ensure physical protec-

_ tion for the spent fuel from internally generated missiiles and the l

effects of pipe breaks.

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l 410.13 Provide the new fuel storage capacity in the spent fuel storage l

(9.1.2)

' facility and the arrangement for maintaining a subcritical array j

during all storage conditions.

l 410.14 Is there any portion of the spent fuel pool cooling and cleanup j

(9.1.3) system designed to nonseismic category requirements? If so, verify L.

that failure of the nonseismic Category I portion in an earthquake l

will not affect the operation of the cooling trains.

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410.15 With regards to the overall heavy load handling systems verify that l

(9.1.4 your design meets the guidelines of NUREG-0612. " Control of Heavy l

and9.1.5)

Loads at Nuclear Power Pla'uts," Phase I and II and provide suffi-cient information so that we can make an independent evaluation of i

how the guidelines of NUREG-0612 are met.

i 410.16 Describe the means provided for assuring that instrument air quality l

(9. 3.1 )

is within the necessary limits to assure proper functioning of all air operated valves and instrumentation in safety related systerr.s.

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410.17 Verify that adequate protection has been provided for safety-related.

as fire protection system and cooling water system) ping systems (such (9.3.3) equipment assuming a pipe rupture for nonseismic pi

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(such as tanks) located -in safety-related areas. This protection cannot assutne credit for nonseismic Category I sump pumps. Your response should include the time required for operator action if i

necessary to provide protection of essential equipment once indica-l tion from the Class lE level switches is given.

410.18 Some of the licensees have provided measures for detecting and cor-(9.4) recting dust accumulation on safety-related equipment.in order to assure their availability on demand.

Verify that dust accumulation-does not pose a problem in this plant.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET N0. 50-423 410.0 AUXII~ SYSTEMS BRANCH (CONT'D) 410.19 Describe the effect on the safety function of the essential HVAC (9.4) systems in the event of a single failure in a fire damper in the' i

ventilation system ducts.

It is our position that such a failure not compromise the safety function of the HVAC system.

410.20 Describe' the measures for assuring a proper operating environment (9.4.1 )

for essential control room and ESF switchgear room air handling units when the nomal control building HVAC system is not available during emergency conditions.

41 0.21 It is our position that these portions of the fuel building ventila-(9.4.2) tion system utilized during the emergency filtration modes be seismic Category I and that the system be designed so that failure of the nonsafety portion of the system will not compromise the operability of the safety-related portion. Verify that your plant complies with the above position.

410.22 Describe the means provided for isolating the waste disposal building ventilation system following a design basis event (such as SSE) in (9.4.9) order to prevent the release of potentially radioactive airborne contaminants through building openings.

410.23 Describe the purpose of the block valve located downstream of the (10.4.3 )

main steam pressure relieving bypass valve and the reasons why it is not locked open.

41 0.24 Describe the main steam isolation valve actuation mechanism and pro-(10.4.3) vide drawings showing the hydraulic and pneumatic operation system for this valve.

410.25 In the evaluation of potential flooding of essential plant areas as (10.4. 5) a result of a circulating water system failure, credit cannot be taken for operator action to stop the circulating water pump in 15 minutes to contain the spillage water in the turbine building to elevation 21 feet 6 inches.

Indicate the water level in the turbine building to which it will eventually rise and verify that this level of water will not affect any essential systems or components.

410.26 In order to meet Branch Technical Position ASB 10-2 for top-feed (10.4.7) design, commit to perform tests acceptable to NRC to verify that the unacceptable feedwater hamer will not occur using the plant operating procedures for normal and emergency restoration of steam generator water following loss of normal feedwater flow. This comitment should be reflected in the FSAR.

410.27 Provide a response to our March 10, 1980 letter to near-tem (10.4.9) operating license applicants concerning your EFW system design (TMI-2 Task Action Plan, NUREG-0737, Item II.E.1.1). This response should include the following:

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REQUEST FOR ADDITIONAL INFORMATION l

MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 410.0 AUXILIARY SYSTEMS BRANCH (CONT'D)

(a) A detailed point-by-point review of your EFW system design against Standard Review Plan Section 10.4.9, and Branch Technical Position ASB 10-1.

(b) A point-by-point view of your EFW system design, Technical Specifications and operating procedures against the generic short-tenn and long-term requirements discussed in the March 10,1980 letter.

(c) The design basis for the EFW flow requirements and verification that your EFW system will meet these requirements (refer to of the March 10, 1980 letter).

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423

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410.0 AUXILIARY SYSTEMS BRANCH - continued We have reviewed the applicant's submittal entitled, " Fire Protection Evaluation Report" on alternate safe shutdown capa-bility in the event of fires anywhere in the plant as defined in Appendix R to 10 CFR Part 50. The applicant has not pro-i vided sufficient information for us to evaluate Parts III.G and III.L of Appendix R.

The applicant should provide the following information:

410.28 The applicant should provide area arrangement drawings showing (SRP 9.5.1) the safe shutdown system (including cable routing) in order that l

we may review the results.

410.29 The applicant has stated that the unit is capable of remaining at hot shutdown status for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The applicant should confirm that the unit is capable of attaining cold shutdown status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of reactor trip using only onsite power.

410.30 The applicant should provide reactor coolant hot leg temperature indicator on the auxiliary shutdown panel for direct reading of process variables to control the reactor shutdown.

41 0.31 The applicant has stated that only active high/ low pressure inter-face is in the residual heat removal system with three three-phase motor operated valves in series and that a spurious operation is considered improbable. The applicant should clarify why a fire in the control room, where the control switches to these valves i

are probably located could not short the control supply for these valves.

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4 REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 I

280.0 CHEMICAL ENGINEERING BRANCH, Fire Protecticn Section 280.2 The fire protection program will' be reviewed to the guidelines of BTP CMEB 9.5-1 (NUREG-0800), July 1981.

Provide a comparison that

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shows conformance of the plant fire protection program to these guidelines. Deviations from the guidelines should be specifically identified. A technical basis should be provided for each deviation.

280.3 Provide the qualifications of the fire protection engineer responsible for tha formulation and implementation of the fire protection program.

(BTP CMEB 9.5-1, Section C.l.b) 280.4 Describe administrative controls that will be developed and imple-mented to comply with BTP CMEB 9.5-1 Section C.2.

280.5 Describe the plant fire brigade that will be provided to comply with BTP CMEB 9.5-1 Section C.3.

280.6 Describe the plant fire brigade equipment that will be provided to comply with BTP CMEB 9.5-1 Section C.3.c.

280.7 Describe fire brigade training program that will be provided to comply with BTP CMEB 9.5-1 Section C.3.d.

r 280.8 Substantiate the fire resistance capability of the fire barriers used to separate safety-related areas or high hazard areas. Verify that their construction is in accordance with a particular design that has been fire tested. Describe the design, the test method used and the acceptance criteria. Provide information for the following components (a) Fire Barriers, (b) Fire dampers, (c) fire doors, and fire barrier penetration seals around ducts, pipes. cables, cable trays and in other openings (e.g. concrete joints seals and fillers), and special treatment for other openings.

Describe the support structure for barriers that are not floors or ceilings; Describe how fire dampers are i

installed in the ventilation ducts that penetrate rated fire barriers l

of safety-related areas; verify that all seals are of the thickness j

specified in the fire tests; and verify that cables and cable trays l

are supported in a manner similar to supporting arrangements used in the fire tests.

(BTP CMEB 9.5-1 Section C.S.a) i

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 280.0 CHEMICAL ENGINEERING BRANCH, Fire Protection Section - continued 280.9 Describe how the fire doors will be kept closed ard supervised by one of the measures stated in BTP CMEB Section C.S.a.

280.10 Describe how fire protection has been provided for safe shutdown so that one train of systems necessary to achieve and maintain hot shut-down conditions from either the control room or emergency control sta-tion (s) is free of fire damage and that systems necessary to achieve and maintain cold shutdown from either the control room or the emer-gency t.ontrol station (s) can be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Provide an analysis whicn shows that one redundant train of equipment structures, systems, and cables necessary for safe shutdown can be maintained free of fire damage by either:

(a) Separation of cables and equipment and associated circuits of redundant trains by a fire barrier having a 3-hour rating. Struc-tural steel forming a part of or supporting such fire barriers should be protected to provide fire resistance equivalent to that required of the barrier; (b) Separation of cables and equipment and associated circuits of redundant trains by a horizontal distance of more than 20 feet with no intervening combustible or fire hazards.

In addition, fire detectors and an automatic fire suppression system should be installed in the fire area; or (c) Enclosure of cable and equipment and associated circuits of one redundant train in a fire barrier having a 1-hour rating.

In addition, fire detectors and an automatic fire suppression system should be installed in the fire area.

(BTP CMEB 9.5-1 Section C.S.b) 280.11 Identify those areas of the plant that will not meet the guidelines of Section C.S.b of BTP CMEB 9.5-1 and, thus alternative shutdown will be provided. Verify that all other areas of the plant will be in com-pliance with Section C.S.b of BTP CMEB 9.5-1.

280.12 Describe how redundant safety-related cable systems outside the cable spreading room are protected to comply with BTP CMEB 9.5-1 Section C.5.e(2).

REQUEST FOR ADDITIONAL INFORMATI0_N_

MILLSTONE NUCLEAR POWER STATION, UNIT _3 DOCKET NO. 50-4p 280.0 CHEMICAL ENGINEERING BRANCH, Fire Protection Section - continued 280.13 Verify that electric cable construction will pass the flame test in the current IEEE Std. 383 to comply with BTP CHEB 9.5-1 Section C.S.e.

280.14 Describe how hydrogen and other flammable gas lines which are routed a

thru safety related areas comply with BTP CMEB 9.5-1 Section 5.d(5).

280.15 It is our position that you comply with Section C.5.g(c) of BTP CMEB 9.5-1, in that a fixed emergency lighting system consisting of sealed beam units with individual (8-hour minimum) battery power supplies should be installed in all areas required for safe shutdown operations, including access and egress routes. Verify that you will comply with our position and specify the foot-candles provided at the floor level of access routes and at operational areas.

280.16 Describe how fixed repeaters installed to permit use of portable radio communication units will be protected from exposure fire damage to comply with BTP CMEB 9.5-1 Section C.5.g.

280.17 Describe the Class A fire detection system that has been provided to comply with BTP CMEB 9.5-1 Section C.6.a to protect all areas of the plant which contain or present an exposure fire hazard to safety related equipment and cables.

280.18 Describe the primary and secondary power supplies for the fire detec-tion systems provided to comply with Section 2220 of NFPA 72D.

(BTP CMEB 9.5-1 Section C.6.a(6))

280.19 Provide a plot plan of the site showing all fire protection water mains to include pipe sizes, valves, location of hydrants, and fire pump con-nections.

It is our position that the piping be arranged and valved such that no single break will cause the loss of both fire pumps or shut off all fire protection water to any area of the plant. Also, it is our position that standpipe and hose station be provided throughout the plant in accordance with NFPA 14, including permanent hose stations inside containment to meet Section C.7.a of BTP CMEB 9.5-1.

280.20 Verify that the fire pumps and their controllers are UL listed and installed in accordance with NFPA 20 requirements. The fire pumps start-up setpoints should be adjusted such that both fire pumps do not start simultaneously (at least a 5 to 10 second delay between pump start-ups is required by NFPA 20).

(BTP CMEB 9.5-1 Section C.6.b)

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET N0. 50-423 280.0 CHEMICAL ENGINEERING BRANCH, Fire Protection System - continued i

280.21 Verify that the fire pumps can provide, in accordance with BTP CMEB 9.5-1 Section C.6.b, the largest firewater flow and pressure (based on 500 gpm for manual hose streams plus the largest design demand of any sprinkler or deluge system in a safety related area as deter-mined in accordance with NFPA 13 or NFPA 15) with the largest fire i

pump out of service. Also, describe how the minimum firewater supply is physically dedicated in each storage tank.

1 280.22 It is our position that the reactor coolant pumps be equipped with an oil collection system in conformance with Section C.7.a of BTP CMEB 9.5-1.

Provide the design description of this system.

280.23 Verify that all cables in the control room meet the separation criteria t

and fire protection critiera detailed in BTP CMEB 9.5-1 Section C.7.b.

280.24 Verify that smoke detectors have been provided in all control room cabinets and consoles in accordance with BTP CMEB 9.5-1 Section C.7.b.

280.25 On page A-45 of the FPER, you state that primary fire suj-pression in the cable spreading room is provided by a total flooding CO2 extin-guishing system.

It is our position that the primary fire suppression l

in the cable spreading room be an automatic water system in conformance with BTP CMEB 9.5-1 Section C.7.c.

280.26 Verify that the loss of ventilation in the safety-related battery rooms is alarmed in accordance with BTP CMEB 9.5-1 Section C.7.g.

280.27 Identify all mechanisms by which fire or fire fighting activities may cause the simultaneous failure of redundants or diverse safety trains -

that have been considered in the design.

Describe the measures taken to preclude tuch failures.

Include consideration of other design basis events (e.g. seismic) simultaneously effecting fire protection system in several areas of the plant.

(BTP CMEB 9.5-1 Section C.l.c) i

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 1

DOCKET NO. 50-423 281.0 CHEMICAL ENGINEERING BRANCH, Chemical Technology Section 281.3 We are concerned that insoluble debris formed, under DBA conditions (6.1.2) inside containment from unqualified organic paint surfaces that do not meet the guidelines of ANSI N101.2 (1972) and Regulatory Guide 1.54 and from corrosion products from galvanized steel and zinc plants without qualified organic top coats that may adversely affect the performance of the RHR or the Containment Spray systems which take suction from the containment sump, r

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Provide the total amount (area and film thickness) of unquali-fied paint ar.d of zinc paints without qualified organic top coats inside containment on both NSSS and non-NSSS supplied components. We will use this information to estimate the amoont of solid debris that can be formed from the unqualified t

paint under DBA conditions which can potentially reach thi con-tainment sump, b.

Provide the results of an analysis that demonstrates that the i

insoluble debris from unqualifiad organic paints and corrosion i

products from galvanized steel and zinc paints without qualified i

organic top coat will not cause a sump debris problem or adversely affect the performance of the RHR or containment spray systems.

i 281.4 In order for the staff to estimate the rate of combustible gas (6.1. 2 ).

generation vs. time due to exposure of organic cable insulation to DBA conditions inside containment, provide the approximate total quantity (weight and volume) of organic cable insulation material used inside containment, including uncovered cable and cable in closed metal conduit or closed cable trays. We will give credit for beta radiation shielding for cable in closed conduit or trays if information is provided as to the respective quantities of cable in closed conduits or trays vs. uncovered cable.

281.5 Describe the samples and instrument readinos and the frequency of (9.1.3) measurement that will.be performed to monitor the water purity and need for spent fuel pit cleanup system demineralizer resin and filter replacement. State the chemical and radiochemical limits to be used in monitoring the spent fuel pool water and initiating corrective actions.

Provide the basis for establishing these limits. Your response should consider variables such as:

boron, gross gamma, and iodine activity, demineralizer decontamination factor and/or filter differential pressure, pH, and crud level.

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REQUEST FOR ADDITIONAL INFORMATI0_N MILLSTONE NUCLEAR POWER STATION, UNIT 3-DOCKET NO. 50-423 281.0 CHEMICAL ENGINEERING BRANCH, Chemical Technology Section - continued 281.6 You did not indicate in the FSAR that the chemical additive tank (9.3.2) in the CVCS system will be sampled.

Confirm that these tanks will be sampled according to Standard Review Plan 9.3.2.

281.7 The information you provided on the Post Accident Sampling Sys-(9.3.2) tem (PASS) is inadequate to demonstrate compliance with NUREG-0737 Item II.B.3.

Provide infomation that satisfies the criteria.

in the attachment.

1 281.8 Describe provisions for monitoring mixed bed, cation bed and (9.3.4) boron thermal regeneration demineralizers differential pressure to assure that pressure differential limits are not exceeded (Section III.G.l.b of Standard Review Plant 9.3.4).

1 281.9 The information that you have provided is insufficient for us to (10.3.5) evaluate the secondary water chemistry control program.

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a summary of operative procedures to be used for the steam generator secondary water chemistry control and monitoring program, addressing the following:

1.

Identify the sampling schedule for the critical chemical and other parameters and the control points or limits for these parameters for each operating mode of the plant, i.e. dry lay-up, cold shutdown, hot standby / shutdown, and power operation.

2.

Identify the procedures used to measure the values of the critical parameters, i.e. standard identifiable procedures and/or instruments.

3.

Identify the sampling points, considering as a minimum the stea:a generator blowdown, the hot well discharge, the feedwater and the demineralizer effluent. While some of this material-can be gleaned from Figures 10.3.2 and 10.3.3, the applicant should provide it, as well as the remaining information, in a more concise manner. We recommend a process flow chart similar to that in EPRI NP-2704-SR "PWR Secondary Water Chemistry Guide-lines."

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REQUEST FOR ADDITIONAL INFORMATION f

MILLSTONE' NUCLEAR POWER STATION, UNIT 3 i.

DOCKET NO. 50-423 e

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280.0 CHEMICAL ENGINEERING BRANCH, Chemical Technology Section - continued

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4.

State the procedures for recording and managment of data, defining corrective actions for various out-of-specification i

parameters. The procedures should define the allowable time j

for correction of out-of-specification parameters. We recom-l mend multiple levels of time allowable for providing correction based upon the amount of out-of-specification of the variable, i

(See EPRI NP-2704-SR above)

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Because of the significance of condenser in-leakage the chemistry program should include a corrective action provision such that j

a condenser inservice inspection program will be initiated if

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condenser leakage is of such a magnitude that power reduction 1

is required (action level 2 of the EPRI/SGOG guidelines) more j

than once per three month period.

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Identify (a) the authority responsible for interpreting the i

data and initiating action (b) the sequence and timing of j

administrative events required to initiata corrective action.

280.10 Explain how you prevent resin break through into the steam generators

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from the full flow demineralizers.

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280.11 The FSAR states on page 10.4-3, second paragraph that the tubesheet j

material in the condensers is aluminum-bronze. On page 10.4-19, j

second paragraph, line 10 the tube sheet material in the condenser

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is identified as copper-nickel. Verify the correct material. The information is needed for the dissimilar metal, junction capability evaluation.

j 280.12 Provide a description of any materials monitoring program for the (9.1.2) spent fuel pool.

In particular provide information on the frequency of inspection and type of-- samples used in 'the monitoring program.

l 281.13 Provide a list of the Codes and Standards used in the design and I

j (9.1. 2 )

fabrication of the spent fuel racks.

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POS1 ACCIDEfii % MPLl?iG SYS1LM t4UREG-0737, ll.li.3 EVALUA110fi CRITERIA GUIDELI!iES The post accident samoling = system will be ' evaluated for compliance with the criteria from NUD.EG-0737', II.B.3.

These eleven itemc have been-copied verbatim from NUREG-0737.

The licensees-submittal should include information equivalent to that which is nonnally provided in an FSAR.'

System schematics with sufficient information to verify flow paths should be included, consistent with documentation recairements in NUREG-0737, with appropriate discussion so that the reviewer can determine whether the criteria have been met.

Further inforriation pertaining to the specific clarifications of NUREG-0737 considered in the reviewers evaluation are listed below., which will be Technically justified alternatives to these criteria will be considered.

Criterion:

(1) The licensee shall have the capability to promptly obtain reactor L

i coolant samples and containment atmosphere samples.

The combined time allotted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision.is made to take a sample.

I Clarification:

Provide information on s =pling(s) and analytical laboratories locations including a discussion of relative elevations, distances f

and methods for sample transport.

Responses to this item should also include a discussion of sample recirculation, sample handling f~

and analytical times to demonstrate that the three-hour time limit will be met (see (6) below relative to radiation exposure).

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describe provisions for sampling during loss of off-site power (i.e. designate an alternative backup power source, not necessarily the vital (Class IE) bus, that can be energized-in sufficient time to meet the three-hour sampling and analysis time limit).

1 Criterion:

(2) The licensee shall establish an onsite radiological and chemical analysis capability to provide, within three-hour time frame j

established above, quantification of the following:

1 (a) certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core ~

damage (e.g., noble gases; todines and cesiums, and non-t s

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volatile isotopes);

(b) hydrogen levels in the containment atmosphere;

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(c) dissolved gases (e.g., H ), chloride (time allotted for l

2 analysis' subject to discussion below), and boron

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concentration of liquids.

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perform all or part of the above analyses.

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Tarification: 2 (a) A discussion of the counting equipment capabilities is needed, including provisions to handle samples and reduce background radiation to minimize personnel radiation exposures (ALARA).

Also a procedure is required for relating radienuclide concentrations to core, damage. The procedure should include:

1.

Monitoring for short and long 1tved volatile and non volatile radionuclides such as 133xe,131 1, 137 s C

Cs, 85 r, 140 a, and 88Kr (See Vol. II, part 2, 134 K

S pp. c24-527 of Rogovin Report for further information).

2.

provisions to estimate the extent of core damage based on radionuclide concentrations and taking into considera-tion other physical parameters such as core temperature data and sample location.

2 (b) Show a capability to obtain a grab sample, transport and analyze for hydrogen.

2 (c) Discuss the capabilities to sample and analyze for the accident sample species listed here 'and in Regulatory Guide 1.97 Rev. 2.

2 (d) provide a discussion of the reliability and maintenance information to demonstrate that the selected on-line instrucent is appropriate for this application.

(See (8) and (10) below relative to back-up grab sample capability and instrument range and accuracy).

Criterion:

(3)

Reactor coolant and containment atmosphere sampling during post accident conditions shall not require an isolated auxiliary system (e.g., the letdown system, reactor water cle.anup system (RWCUS)) to be placed in operation in order to use the sampling system.

Clarification:

System schematics and discussion's should clearly demonstrate

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that post accident samp1,ing, including recirculat. ion, from each sample source is possible without use of an i'sciated i-auxiliary system. It should be verified that valves which are not accessible after an accidsnt are environmentally qualified for the conditions in which they must operate.

Criterion:

(a) pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpressuri:ed reactor coolant samples.

The measurement of either total dissolved gases or H-gas in reactor coolant samoles is considered adequate. keasuringthe02 concentra-tion is recommended, but is not mandatory.

Clarifica ti on:

Discuss the method whereby t0tal dissolved gas or hydrpgen 1

and oxygen can be meltured and related to reactor coolant system concentrations. 4dditionally, if chlorides exceed 0.15 ppm, verificatiin that dT'ssolved oxygen is less than 0.1 ppm is necessary. Verification that dissolved oxy-

<0.1 ppm by measurement of a dissolved hydrogen residu,en is

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> 10 cc/kg is acceptable for up to 30 days after the accident. Within 30 days, consistent with minimizing personnel radiation dxposures (ALARA), direct _ monitoring for dissolved oxygen is recommended.

Criterion:

(5)

The time for a chloride analysis to be performed is dependent; upon two factors:

(a) if the plant's coolant water is seawater or brackish water and (b) if there is only a single,!

barrier between primary centainment systems and the cooling l_

water. Under both of the above conditions the licensee shall l' provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being taken. For all other cases, the licensee shall provide ;

for the analysis to be completed within 4 days. The chloride !

analysis does not have to be done onsite.

+

Clari fica tion:

BWR's on sea or brackish wr.ter sites, and plants which use sea or brackfish water in essential heat exchangers (e.g.

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shutdown cooling) that have only single barrier protection i

between the reactor coolant are required to analyze chloride within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. All other plants have 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to perform a chlorida analysis. Samples diluted by up to a factor of one thousand are acceptable as initial scoping analysis for

'1 chloride, provided (1) the results are reported as ppm C1 (the licensee should estao'lish this value; the number in tha blank should be no greater than 10.0 ppm Cl) in the reactor coolant system and (2) that dissolved oxygen can be verified at <0.1 ppm, consistent with the guidelines above in clariff-cation no. 4 Additionally, if chloride analysis is perfcrmed on a diluted sample, an undiluted sample need also be taken and retained for analysis within 30 days, consistent with ALAPA.

Criterion:

(5)

The design basis for plant equipment for reactor coolant and '

containment atmosphere sampling and analysis must assume that

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it is possible to obtain and analyze a sample without radiation exposures to any individual exceedf ag the criteria of GDC 19 L

(Appendix A,10 CFR Part 50) (i.e., 5 rem whole body, 75 rem extremities).

(Note that the design and operational review i

criterion was enanged from the operational limits of 10 CFR Part 20 (NUREG-0578) to the GDC 19 criterion (October 30, 1979 C

letter from H. R. Centon' to all licensees).

I Clarification:

Contistent with Regulatory Guide 1.3 or.l.4 source terms, provide information on the predicted personnel exposures based l

on person-motion for sampling, transport and analysis of all required parameters.

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Criterion:

(7)

The analysis of primarf' coolant samples for boren is required for PWRs.

(Note that Rev.$ 2 of the need for primary c,oolant bor,,jegulatory Guide 1.97 specifies' on analysis capability at BWR /

olants).

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clarification:

PWR's need to perform boron analysis. The guidelines for BWR's are to have the capability to perfom boron anal sis but they do not have to do so unless boron was injecte.

Criterion:

(8)

If inline monitoring in used for any sampling and analy-tical capability'specified herein, the licensee shall provide backup sampling through grab samples, and shall demonstrate

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i the capability of analyzing the samplics.

Established planning for analysis at offsite facilities is acceptable.

Equipment provided for backup samplir.g shall be capable of providing at lcast one sample per day for 7 days following onset of the accident, and at least one sample per week b

until the accident condition no longer exists.

Cla ri ficati on:

A capability to obtain both diluted and undiluted backup samples is required. Provisions to flush inline monitors to facilitate access for repair is desirable.

If an off-site 1.iboratory is to be relied on for the backup analysis, an l

explanation of the capability to ship and obtain analysis for one sample per weck thereafter until accident condition i

no longer exists should be provided.

. Criterion:

(9)

The licensee's radiolcgical and chemical sample analysis capability shall include provisions to:

4 (a)

Identify and quantify the isotopes of the nuclide categories discussed above to levels corresponding to the source tems given in Regulatory Guide 1.3 or 1.4 and 1.7.

Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduc-tion of personnel exposure should be provided. Sensi-tivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concen-tration in the range from approximately -lu Ci/g-to 10 Ci/g.

(b)

Restrict background levels of radiation in the radiolog-ical and chemical analysis facility from sources such that

.f the sample analysis will provide results with an acceptably small error (approximately a factor cf 2). This can be accomplished through the use of sufficient shielding.

around samples and outside sources, and by the use of a 9

ventilation system design which will control the presence of airborne radioactivity.

'! Clarification: (9) (a) Provide a discussion of the predicted activity in the samples to be taken and the methods of handling / dilution that will be employed to reduce the activity sufficiently to perform the required analysis.

Discuss the range of radionuclide concen-F 4

tration which can be analyzed for, including an assessment of, l-the amount of overlep.between post accident and normal sampling ~'

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background radiati,on levels and radiation effect will be on a sample being counted to assure an accuracy within a factor

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of 2.

j-i Criterion:'

(10')

Accuracy, range, and sensitivity shall be-adequate to provide' I pertinent data to the operator in order to describe radiolo-gical and chemical status of the reactor coolant systems.

Clarification:

Thc recemmended ranges for the required accident sample analyses are given in Regulatory Guide 1.97, Rev. 2.

The necessary accuracy within the recommended ranges are as follows :

- Gross activity, gamma spectrum: measured to estimate core damage, these analyses should be accurate within a factor of two across the entire range.

- Sa ron : measure to verify shutdown margin.

In general this analysis should be accurate within +5% of the measured value (i.e. at 6,000 ppm B the tolerance is 1300 ppm while at 1,000 ppm 8 the tolerance is + 50 ppm).

For concentrations below 1,000 ppm the tolerance band should remain at + 50 ppm.

- Chloride: measured to determine coolant corrosion potential.

For concentr, tiens between 0.5 and 20.0 ppm chloride the analysis should be accurate within + 10% of the measured n

value. At concentrations below 0.5 pp'm the tolerance band remains at + 0.05 ppm.

- Hydrogen or Total G'as: monitored to estimate core degrada-tion and corrosion. potential of the coolant.

An accuracy of + 10% is desirable between 50 and 2000 cc/kg but + 20% can be acceptable.

For concentration below 50 cc/kg the tolerance remains at + 5.0'cc/kg.

- Oxygen: monitored to assess coolant corrosion potential.

For concentrations between 0.5 and 20.0 ppm oxygen the analysis should be accurate within + 10% of the measured value. At

- l concentrations below 0.5 ppm the tolerance band remains at i

+ 0.05 ppm.

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- pH: measured to asstss coolant corrosion potential.

Between a pH of 5 to 9, the reading should be acc' urate within +0.3 pH units.

For all other ranges + 0.5 pH units is acceptable.

To demonstrate that the selected procedures and instrumentation will achieve the above listed accuracies, it is necessary to provide information demonstrating their applicability in the post accident water chemistry and radiation environment. This can be accomplished by performing tests utilizing the standar.d test matrix provided below or by providing evidence that the selected procedure or instrument has been used successfully in a similar environment.

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STANDARD TEST MATRIX FOR UNDILUTED REACTOR CCOLANT SAMPLES IN A POST-ACCIDENT ENVIRONMENT Nominal Constituient Cuncentration (pre)

Added as (chemical salt) i I-40 Potassium Iodide Cs+

250 Cesium Nitrate n

l Ba+2 10 Barium Nitrate La+3 5

!.anthanum Chloride Ce+4 5

Ammonium Cerium Nitrate Cl-10 B

2000 Boric Acid-Li+

2 Lithium Hydroxide

!!0' 150 NHf 5

K+

20 Ganma Radiation 104 Rad /gm of Adsorbed. Dose (Induced Field)

Reactor Coolant NOTES:

1) Instrumentation and procedures which are applicable to diluted samples l

only, shc61d be tested with an equally diluted chemical test. matrix.

The induced radiation environment should be adjusted commensurate I

with the weight of actual reactor coolant in the sample being tested.

2) ' For PWRs, procedures which may be affected by spray additive chemicals must be tested in both the standard test matrix plus appropriate spray additives. Both procedures (with and without spray additives) are required to be available.

s 3)

For SWRs, if procedures are veri.fied with poron in the test matrix, 'they i

n do not have to be tested without boron.

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4) In lieu of conducting tests utilizing the standard test matrix for instruments and procedures, provide evidence that the selected instrument or procedure has been' used successfully in a sinilar environment.

4 All equipmen't and procedures which are used for post accident sampling and analyses should be calibrated or tested at a frequency which will ensure, to a ~high degree of reliability, that it will be available if e

required. Operators should receive initial and refresher training in post accident sampling, analysis and transport. A minimum frequency for the above efforts is considered to be every six months if indicated by j

testing. - These provisions should be submitted in revised Technical Specifications in accordance with Enclosure 1 of NUREG-0737. The staff will provide mcdel Technical Specifications at a later date.

Criterion:

(11)~

In the design of the post accident sampling and analysis 4

capability, consideration should be given to the following -

i items:

(a) provisions for purging sample lines, for reducing plateout in sample lines, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose material in the RCS or contairnent, for appropriate disposal of the samples, and for flow restrictions.to limit reactor coolant loss from a rupture of the sample line. The post accident reactor coolant and containment atmosphere ~ samples

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should be representative of. the reactor coolant in the core area and the containment atmosphere following a transient or accident. The sample lines should be as short as possible to minimize the volume of fluid to be taken.

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from containment. The residues of sample collection should l

be returned to containment or to a closed system.

(b) The ventilatio'n exhaust from the sampling.st'ation should be filtered with charcoal absorbers and high-efficiency particulate air (3 EPA) filters.

1-Clarification: (ll)(a) A description of the provisions which address each of the items -in clarification ll.a should be provided. Such items, 1

+

as neat tracing and purge velocities, should be addressed. To demonstrate that samples are representative of core conditions

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a discussion of mixing, both.short and long term, is needed.

If a given sample location can be rendered inaccurate due to the accident (i.e. sampling from a hot or cold leg loop which I

may have a steam or gas pocket) describe 'the backup sampling capabilities or address the maximum time that this condition-can exist.

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SWR's should 'specifically epdress samples which are taken frce the core shroud area. and demonstrate how they are.repre-sentative of core conditions.-

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m Passive flow restrictors in the srple lines may be replaced by redundant, environmentally qualified, remotely operated isolation valves to ismit potential leakage from sampling lines. The automatic containment isolation valves should close on containment isolation or safety injection signals.

(11)(b)

A dedicated sample station filtration system is not required, provided a positise exhaust exists which is subsequently routed through charcoal absorbers and HEPA filters.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50+423 4 91.0 CORE PERFORMANCE BRANCH, Physics Section 4 91.1 Table 15.6-8 of the FSAR gives a value of 2.14 for the F (4.3.2.2) used in LOCA analysis.

Section 4.3 of the FSAR (which should (SRP 4.3.11.1) discuss all aspects of all power distributions used in Chap-ter 15 analyses, including in particular the power peaking factors used to satisfy LOCA analysis requirements) does not mention such a value.

Instead Section 4.3 presents only the standard Westinghouse discussion demonstrating that an Fo value of 2.32 can be maintained, using the standard Westing-house CAOC (with improved load follow) analysis, control and excore (split detector) surveillance. The only conclusion that one can draw from this information, as it stands, is that it will be necessary to derate the reactor to about 92 percent power.

If there is an alternate power distribution analysis, control scheme or surveillance system to be used with your reactor operations which will demonstrate that an' Fn of 2.14 can be maintained, please modify Section 4.3 (and other indi-cations of Fo and peak kW/ft such as Table 4.1).to present this new limit, and a discussion in detail of the modifica-tions involved to hardware, ana'iyses and operations, including a full uncertainty analysis. Topical reports may be subsitted and referenced, but modifications to Chapter 4 (and possibly Chapter 7) are required.

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4 REQUEST FOR ADDITIONAL INFORMATION MIU. STONE NUCLEAR POWER STATION, UNIT 2 00CXET NO. 50-423 a

240.0 ENVIRONMENTAL AND HYDROLOGIC ENGINEERING BRANCH, Hydrologic Engineering Section 240.1 Tne FSAR provides estimates of water lev 61s at various buildings (2.4.2.3) resulting from the local PMP runoff and Figure 2.4-7 indicates the (SRP 2.4.2) contributing areas to the storm drains.

Provide your detailed calculations and appropriate large scale drawings to verify that the local PMP runoff across final site grade, assuming the yards drains are ineffective, will not result in water levels in excess of door sill elevations to safety-related structures. Also verify that the effects of road crown elevations and security fencing are accounted for in the

analyses, i

240.2 Discuss the reason for the difference in PMH maximum stillwater level (2.4.5.2) values cited in the FSAR (19.7 feet MSL) and the PSAR, Amendment 14 (2.4.5.3) and the CP-SER (18.2 feet MSL).

l (SRP 2.4.5) 240.3 The correct reference for Initial P.ise in Regulatory Guide 1.59 is (2.4.5.2)

Table C-1 not Table 3.1.3-1.

(SRP 2.4.5) 2 240.4 FSAR Section 2.4.5.3 states that the maximum wave loading on the-front (2.4.5.3) face of the intake structure would occur with the slow translation

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(2.4. 5. 5) speed PMH at the time of peak surge of 19.7 feet MSL with a wave height (SRP 2.4.5) of 16.2 feet. Section 2.4.5.5 states that the maximum uplift pressure of 908 lb/sq ft on the pumphouse floor would occur during the. slow speed PMH with the maximum wave height of 16.9 feet and coincident PMH surge

]

level of 17.4 feet MSL.

a) Provide detailed discussion and calculations clarifying your conclusion that the maximum uplift floor pressures occurs when the PMH surge level is well below its maximum level.

b) The maximum uplift force (908 lb/sq ft) on the service water pump cubicles appears to have neglected to account for the fact that these cubicles are water tight whereas the adjacent areas in the intake structure would be flooded. This uplift force is also less than the associated clapotis force on face of the intake structure. Recent physical hydraulic model studies conducted for the Diablo Canyon Intake Structure indicated that uplift pressures were approximately 150% of the front face forces at the *,ame elevation.

(

Reference:

"The Investigation of Wave-Structure Interaction for the Cooling Water Intake Structure of the Diablo Canyon Nuclear Power Plant" by Fredric Raichlen for PG&E, December 1982). Either show that your analysis of the uplift forces has included the effects of the nonflooded service water pump cubicles, i

I

REQUEST FOR ACDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 240.0 ENVIRONMENTALANDHYDJ0LOGICENGINEERINGBRANCH,HydrologicEngineering Section - continued provide appropriatesanalyses that includes these effects, or show that it is not applicable. Additionally, either show that your analysis of the uplift force is conservative and the Diablo Canyon studies are not applicable or provide revised analysis which account for the relative increase in uplift forces over the appropriate front face forces.

240.5 Provide details of your analysis leading to your estimate of the maximum (2.4.5) water level within the intake structure during the postulated PMH.

(3.4.1.1)

Discuss (including appropriate references) the applicability of the method (SRP 2.4.5) you used (i.e., the pressure response factor to reduce the estimated pressure of the non-breaking wave or clapotis) in determining maximum water level within the intake structures. An alternate approach would be to utilize the design stillwater level of 19.7 feet MSL and the elevation of the orbital center of the 16.2 ft design wave (clapotis) for a range of peak spectural wave periods, T. The wave period associatedwithanyfullydevelopedwaveheig8tcanrangeuptoabout 1.35 T. Using this procedure we calculate:the water level elevation P

within the intake structure as 24.7 feet MSL for T = 9.0 seconds and 26.2 feet MSL for 1.35 T = 12 seconds.

EitherdeSignforamaximumwater level of 26.2 ft MSL wit 0in the intake structure or justify, including details of your analysis, a lower elevation.

240.6 Discuss the vulnerability of buried safety-related electrical and service (2.4.10) water conduits to exposure by wave induced erosion when water levels (SRP 2.4.10) including associated wave action exceeds El 14.0 ft MSL.

240.7 Discuss measures taken to prevent wave scour of erodable soil behind (2.4.10) the seawall and around the circulating and service water pumphouse.

(SRP 2.4.10) 240.8 You state that you have determined the coefficient of permeability cr (2.4.13) givenforgeachsand(l.gx10goilsamplesfgomthesite.

hydraulic conductivity, k, for The values (SRP 2.4.13) to 1.7 x 10- cm/sec) and backfill (1.7 x 10- to 3.4 x 10 cm/sec), however, appear to be at least two orders of magnitude lower than average values for the materials described. Provide detailed descriptions of the tests perfonned.

Discuss the apparent discrepancy between the results and the values for similar materials (medium beach sand) cited in the literature (see, e.g., " Groundwater Hydrology" by D. K. Todd).

240.9 In your discussion of the flood protection of the service water cubicles (2.4.14) credit was taken for the watertight steel doors. Are these doors nonnally (SRP 2.4.14) closed and secured? If so, what alerts operators if they are not secured?

If they are not normally secured, discuss flood protection procedures to be taken to secure service water cubicles prior to the arrival of a surge level including concurrent wave action that will exceed El.14.0 feet MSL.

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION, UNIT 3 DOCKET NO. 50-423 730.0 GENERIC ISSUES BRANCH 730.1 The Atomic Safety and Licensing Appeal Board in ALAB-444 determined that the Safety Evaluation Report for each plant should contain an assessment of each significant unresolved generic safety question.

It is the staff's view that the generic issues identified as " Unresolved Safety Issues" (NUREG-0606) are the substantive safety issues referred to by the Appeal Board. Accordingly, we are requesting that you provide us with a summary description of your relevant investigative programs i

and the interim measures you have devised for dealing with these issues pending the completion of the investigation, and what alternative courses of action might be available should the program not produce the envisaged resul t.

There are currently a total of 27 Unresolved Safety Issues. We do not require information from you at this time for a number of the issues sir.ce a number of the issues do not apply to your type of reactor or because a generic resolution has been issued.

Issues which have been resolved have been or are being incorporated into the NRC licensing guidance and are addressed as a part of the normal review process.

4 However, we do request the information noted above for each of the issues listed below:

s 1.

Water Hammer (A-1) 2.

Steam Generator Tube Integrity (A-3) 3.

Steam Generator and Reactor Coolant Pump Support (A-12) 4.

Systems Interaction (A-17) 5.

Seismic Design Criteria (A-40) 6.

Containment Emergency Sump Performance (A-43) 7.

Station Blackout (A-44) 8.

Shutdown Decay Heat Removal Requirements (A-45) 9.

Seismic Qualification of Equipment in Operating Plants (A-46)

10. Safety Implications of Control Systems (A-47)

11. Hydrogen Control Measures and Effects of Hydrogen Burns on Safety Equipment (A-48)

12. Pressurized Thermal Shock (A-49)

REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET N0. 50-423 231,0 GEOSCIENCESBRANCH,GE0LOGYSECTIM 231.1 SRP 2.5.1 Additional offshore seismic reflection profiles have been run since the 1978 Weston Geophysical Study made for the New England, Power Company (1976), which indicate that the New Shoreham fault could extend 7 kilometers closer to the site than was detemined by the study. These lines were the product of a joint effort by the U. S. Geological Survey (Preliminary Open-File Report 83-XXX) office at Woods Hole and the The staff concluded Connecticut Geological and Natural History Survey.

that this fault was not capable (NEPC01 & 2, SER,1978). The new lines appear to confirm that the fault is truncated by Upper Wisconsinan glacial outwash, which appear to agree with the findings in the Weston Geophysical report.

(a) Assess the recent USGS data and determine if it does agree with previous applicant and staff positions regarding the age of last movement on this fault.

(b)

If applicable, incorporate by reference and support the position on the New Shoreham fault non-capability as proposed by the applicant for NEPC0 1 & 2.

Incorporate the references into the FSAR.

(c) Provide any other new information that could modify the conclusions of applicant and staff for NEPC0 1 & 2.

Discuss.

231.2 SRP 2.5.1.1 The FSAR refers to an unnamed north-south trending fault, which cuts the Honey Hill fault, located approximately 10 miles northeast of the site.

Based on an oral communication Goldsmith (1973), it is stated that the fault is a normal fault related to Triassic tectonics. Update the study of the fault, citing any recent publicaticns pertaining to it, and a) provide data supporting the characterization of the fault and a Triassic age limitation for it, b) discuss the extent of the fault and its significance to the plant site and c) provide supporting data for your co'nclusion that the fault terminates approximately 10.5 miles northeast of Millstone Point.

231.3 SRP 2.5.3.2 The discussion and maps of faulting in the section on " Surface Faulting" is difficult to follow, with the data and analyses in four different reports on faulting in the plant excavations, dating from 1975 to 1982, and in FSAR section 2.5.4 under the geotechnical engineering heading,

" Stability of Subsurface Materials and Foundations." There is generally

REQUEST FOR ADDITIONAL INFORMA. TION MILLSTONE NUCLEAR POWER STATION UNIT 3-DOCKET NO. 50-423 231.0 GEOSCIENCES BRANCH, GE0 LOGY SECTION - continued poor correlation between individual fault discussions and the particular report and section that contains the data and analyses supporting the conclusions. Reports (NNEC0, 1975, 1976. 1977, and 1982) should either be appended to the FSAR or adequate cross-referencing should be established with section 2.5.3.

In particular, reference the pertinent

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reports for each fault or fault system that is being discussed.

The discussion under the heading " Geologic Mapping During Construction" in the geotechnical engineering section 2.5.4 should either be placed in the geology section 2.5.3 or an adequate system of cross-referencing should be established with section 2.5.3.

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REQUEST FOR ADDITIONAL INFORMATION MILLSTONE NUCLEAR POWER STATION UNIT 3 DOCKET NO. 50-423 231.0 ' GEOSCIF:4CES BRANCH, GE0 LOGY.SECTION - continued i

References 4

New England Power Co..(NEPC0) 1976. Charlestown Preliminary Safety Analysis Report, Appendix 28.

Marine Geophysical Survey, New Shoreham Fault Investigation by Weston Geophysical.

1 U. S. Geological Survey, Preliminary Open-File Report 83-High-Resolution Seismic-Reflection Profiles and Sidescan-Sonar Records Collected on Eastern Long Island Sound and Block Island Sound by U. S.

j Geological Survey, R/V Asterias Cruise AST 83-3, by Sally W. Needell and Ralph S. Lewis New England Power Company (NEPC0) 1978 Safety Evaluation Report by the i

U. S. Nuclear Regulatory Commission in the Matter of NEP Nuclear.

i Generating Station Units 1 and 2.

NUREG-0424 l

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