ML17258A654

Forwards Updated Listing of Topics W/Identified Differences from Licensing Criteria & Brief Summary of Actual Identified Differences,Per 820310-12 Meeting in Rochester,Ny.Integrated Assessment Meeting Scheduled for 820402 in Bethesda,Md

ML17258A654
Person / Time
Site:	Ginna
Issue date:	03/17/1982
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
LSO5-82-03-078, LSO5-82-3-78, NUDOCS 8203220143
Download: ML17258A654 (22)

Text

'Oz March 17, 1982 Docket No. 50-244 LS05-82-03 078 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas I( Electric Corp.

89 East Avenue Rochester, New York 14649 cP>

Q CO C

" Beep~

Afpp~

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SUBJECT:

INTEGRATED ASSESSMENT MEETING AT NRC (BETHESDA)

Our letter dated March 9, 1982, subject "Integrated Assessment Meeting at Ginna,"

scheduled a meeting in Bethesda, Maryland, for April 2, 1982.

The purpose of this meeting is to review your proposals on the identified differences.

Enclosed is an updated listing of all topics with identi-fied differences from licensing criteria (Enclosure

1) and.a brief summary of the actual identified differences (Enclosure 2).

This list was discus-sed and updated with your staff during the March 10 - 12, 1982, meeting in Rochester, New York.

Sincerely, Original signed by

Enclosures:

As stated Dennis M.~Crutchfield, Chief.

Operating Reactors Branch No.

5 Division of Licensing cc w/enclosures:

See next page

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i Ginna Docket No. 50-244 Rev. 2/8/82 Nr. John E. Maier CC Harry H. Yoigt, Esquire

LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire

Avenue, N.

M'.

Suite 1100 Mashington, D. C.

20036 Nr. Hichael Slade 12 Tr ailwood Circle Rochester, New York 14618 Ezra Bialik

'ssistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New -York 10047 Resident Inspector R. E. Ginna Plant c/o U. S.

NRC

.1503 Lake Road

'ntario, New York

'.14519 Director, Bureau of Nuclear Oper'ations.

State of New York Energy Offioe Agency Building 2 Empire State Plaza

Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, New York 14519 V. S. Environmental. Protection Agency Region II Office ATTN:

Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I Office of Inspection and Enforcement 631 Park Avenue King of Prussia, Pennsylvania 19406 Dr, Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission'ashington, D. C.

20555.

Dr. Richard F. Cole Atomic Safety and Licensing Board.

U. S. Nuclear Regulatory Commission Washington, D. C.

20555

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GINNA I

'OPIC NO.

II-1.A II-2.A II-3.B TOPICS MHICH DO NOT MEET CURRENT CRITERIA OR E UIVALENT TITLE Exclusion Area Authority and Control Severe Meather Phenomena Flooding Potential and Protection Requirements 4

. II-3.B.1.

II<<3. C II-4. D III-1 III-2

- III-3.A I11-3, C III-4.A III-4. C III-5.A III-5.B III-7. R III-7.B III-8. A V-10; A Capability of Operating Plant to Cope Mith Design Basis Flooding Conditions Safety-related Mater Supply )Ultimate Heat Sink (UHS))

Stability of Slopes Classification of Structures, Systems and Corrrponents Mind and Tornado Loadings Effects of High Mater Level on Stiuctures

.'nservic'e Insp'ection of.Mater Control Structures Tornado Nissiles erne 1 1 Generated NQyi1 es Int

.y Effects of Pipe Break on Structures, Systems and Components Inside Containment Pipe Break Outside Containment RHR Heat Exchanger Tube Failures Seismic Design Considerations Inservice Inspection, Including Prestressed.

Concrete Containment with Either Grouted or Ungrouted Tendons Design

Codes, Desi.gn Cri-teria.,

and Loading'ombinations Loose Parts Monitoring and Core Barrel Vibration Program Reactor Coolant Pressure Boundary (RCPB) Leakage, Detection

TOPIC NO.

V-10. B

.YI-4 (Systems)

(Electrical)

YI-7.B VIII-3. B IX-3 IX-5 IX-6 TITLE RHR Reliability Containment Isolation ESF Switchover from Injection to Recirculation-Mode DC Power System Bus Voltage Monitoring and Annunciation Station Service and Cooling Water Systems Ventil'Ation Systems Fire Protection

GINNA Enclosure 2

TOPIC NO.

'II-l.A Difference Sumnar TITLE Exclusion Ar'ea Authority and Control The Exclusion Area Boundary (EAB) has been

changed, as subnitted by QGKE letter dated June 26, 1981.

This change is potentially significant enough to warrant' change to the Ginna Technical, Specifications to in-corporate the new exclusion area boundary map.

TOPIC NO..

II-2. A TITLE Severe Heather Phenomena Difference Summar TITLE

. Flooding Potential and Protection 9equirements

~

Difference Summar

~p 10 CFR 50 (GDC 2),

as implemented by Standard Review Plan (SRP) 2.4.10 and Regulatory Guide (RG) 1. 59 prescribes that the plant have ade'quate flood protection.

The water levels produced by a Probable Maximum Flood (PMF) on Deer Creek would cause water to pond 8'bove grade on the north side.

r 1,0'FR 50

('GDC 2), 'requires that the plant be designed to withstand the effects of natural phenomena.

The combined snow load for structural capability assessment'at Ginna is 100 lb/ft2.

Yarious safety related buildings were not constructed to withstand such a load'..

TOPIC; NO.

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2=3. B.

TOPIC;MO.

II-3.8.1 Difference Summar TITLE Capability of Operating Plants to Cope Mith Design Basis Flooding Conditions 10 CFR 50 (GDC 2),

as implemented by SRP 2.4.10 prescribes that the plant have adequate flood protection..The plant has no existing plans or technical specifications

'(TS} that relate to flooding from external sources.

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TOPIC NO.

II-3. C TITLE..

Safety-Related Water Supply t,Ultimate Heat Sink (UHS))

Difference Sugar

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10 CFR 50 (GDC 2),

as implemented by SRP 2.4.10 prescribes that the'plant have adequate flood protection.

An occurence of the Probable Haximum Flood on Deer Creek would inundate both the service water and circulating water pumps.

TOPIC HO.

TITLE II-4. D Difference Summar Stability of Slopes TITLE C

guality Group'lassificVt'ion of'tructures, Systems and Components III-1 10 CFR 50 (GDC 2),

as implemented by SRP 2.5.5'prescribes that the

'lant, be adequately protected aoainst failure of natural.or man-made

slopes.

The failure of the onsite slopes would affect safety-related structures.

/

P TX6'$C NO:

Difference Summar 10 CFR 50 (GDC T ),

as implemented. by Regulatory Guide 1.26, requires that; structures, systems and components important to safety be designed, fabri-

cated, erected and tested to ouality standards commensurate with the importance o

the safety functions to be performed.

'The follo~ing are deviations from current requirepents:-

1)

Cateoory C joints of vessels which'ould currently:be classified by

~

ASHE Section III, 1977 as Class 2 or 3 but built to ASHE Section III, 1955 as Class C do not satisfy current radiography requirements 2)

The regenerative heat;exchanger and the excess letdown'eat exchanger do not satisfy current radiography requirements because they are Class A

vessels built to Class C requirenents.

TOPIC HO.

III-2 Difference Summar TITLE Mind and Tornado Loading

'I 10 CFR 50 (GDC 2), as implemented by Standard Review Plan Sections 3.3.1 and 3.3.2 and Regulatory Guide 1.76 and 1.117 requires that the plant be

. designed to withstand the effects of n'atural phenomena.

The existing design and constriction of structures important to;safety for wind and tornado loadings does not meet current licensing criteria of remaining within Sta'ndard Review Plan stress limits.

TOPIC HO.

TITLE III-3. A'ffects of High Mater on Structures....

Dif, er ence Suima 10 CFR 50 (GDC 2),

as implemented by SRP 2.4.12 prescribes that the plant pe designed for 'groundwater problems.

Groundwater.'induced loads~

have not been considered for a groundwater elevation higher than eleva:

tion 250 ft. msl. It is not clear what groundwater elevation was used

-in the design of the diesel generator building.

Also, seismic Categ~or I structures, systems and equipment were pot designed for.flood dQe to H:

Deer Creek.

1 TOPIC ttO.

TITLE III-3. C Inservice Inspection of'ater Control Structures Difference Summar 10 CFR 50 (GDC 45),

as inIplemented by Regulatory Guide 1.127 requir'es that the coolino water system shall be desioned to permit appropriate periodic inspection of impor.an.

components to ensure the integrity and'apability of the syste~.'he following are necessary for compliance with the intent of Reoulatory Guide 1.127:

1)

The inspection program now underway at Ginna-should be formalized so that standard report forms are suhnitted by competent and qualified inspectors to be reviewed by qualified',engineers.

2)

The licensee should develop a checklist for discharge canal inspections,

'including their frequency.

4-3)

The Deer Creek basin should be formally recognizt.'d as a water control structure and inspected accordingly on an annualbasis and follewing severe rains which cause flooding.

(a)

The Inservice Inspection Program for Deer Creek should 'be supple-mented by adding:

clogging of culve'rts by debris, slump conditions, soil creep, and bed load movement.

(b)

The wooded area downstream of the Visitors Center should be. cleaned out to initially establish adequate

'water conveyance during floods and a baseline for future inspection and maintenance.

. 4)

The Licensee should compile a comprehensive file of engineering drawings for safety-related water control structures to establish imediate post-s construction conditions.

5)

The routine inspection frequency is acceptable, but"special i'nspect$ ons

~ also must he performed after extreme events such as floods and seiches which may jeopardize the integrity of water control structures.

'The formal inspection program to be initiated at the R.

E: Ginna Plant, should incorporate such special inspections.

-',6)

The Licensee should develop a formal inspection program for water control structures that will result in the development of a comprehensive file, of appropriate inspection reports.

7g The L)cenEsee"s monAoririg program to be developed 'for'the revetment muW-

~,

'be approved by the HRC.

~P TOPIC NO.

III-4.A Difference Summar T1TLE Tornado Nissiles 10 CFR 50 (GDC 2),

as implemented by Regulatory Guide 1.117 prescribes struc-

tunes, sysiems and'components thai should be designed to withstand the effects of a tornado, in'eluding tornado missiles, without loss of capability to perform'heir safety function.

The'fo)lowing safety-related structur'es, systems and components were found to not be protected from tornado missiles:

1)

Component Cooling System 2)

Refueling Water Storage Tank

3)

Electrical Busses 14, 17 and 18.

4)

Service Mater System 5)

Diesel Generators and their Fuel Supply 6)

Relay Room

')

Yogin Steam and Feedwater piping between isolation valves and the containment penetrations..

1.

8)

The Top Surface of the Spent Fuel Pool is open and, there'fore, the internals are exposed 9)

Boric Acid Tanks TOPIC HO.

TITLE

~ ~

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III-4.C Internally Generated Hissiles

'L Di erence Sunmar 10 CFR 50,(GDC 4),

as implemented by SRP: S'ection 3.5.1.1: and prescribes that structures,.systens'and components important be designe'd to.wi.hstand'the effects of internally ge'nerated inside or ouisode of 'containment."

3. S.'1. 2:.;":-:

to safetg"-

missiles 8

The following are deviations or open items that have been identified:

1)

An evaluation of the piping and components associated with the ECCS accumu'ators with respect to missile'enera..ion and protection has not been co~Ipl eted.

2).

An evaluation of the effects of missile generation along the CVCS"let-down li.ne inside containment has not been completed.

I 3)

An evaluation of the potential effectscf an-unrestrained valve opera-tor associated with the steam generator blow'down system on safety re-lated components and systans has not been completed.

4)

The refueling water storaoe tank is inadequately protected from missiles.

TOPIC NO.

III-5.A TITLE '

~

Effects of Pipe Break on Structures, Systems

~ and Components Inside Contairrllent Difference Sugar ented b

SRP 3.6.2 prescribes hat structures, bed tth d

systems and components 1mport ant to safety es1gn ic ahd environmental effects of postulated pipe rup ure ui del ines that have keen identified following are deviations from review gu1 e ine

~

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d with the general assumptions of t 1's e

check valve in an incom1ng 1ne topic assessment was that a

'ded the check valve closes.-

Adequate assurance n in the event of a p1pe rea u

valve.

Th1s 1s true prov>

must be 'demonstrated that these normal y open c

e their a,ssumed isolation function.

A h nistic evaluation was performed.

The line were all'elow'the=.criteria,-.so breaks I

p "A" accumulator line a mechan1s The d-at he loo compartment, where no adverse

- interactions would occur.

all 1 ocked d

ust on"-'the.reactor side of the norm y

lhe second po1nt 1s loca'e gust on-lve.'t this locat'ion no t

ovide this protect1on can be luatio s'.can be performed to emedial measui es o pro

'uld lead to a 'double-ended rupture do not ractur e mechan1cs eva d

d th Att h

t to E cl o e

exist as d1scusse 1n th 'u1 ance pl The effect of a br eak. i'in'he two inc accumu still under review by e

The 2)

'1nst'I Ument cll cul L,s ls ld affect safety-related.

since some jets cou d

'b d.'

0 bo ho ld. b equ1 pmen

~t analyses similar to those descr1 e

1n

'rovided.

evaluation of -the effects on cables and

)

For the letdown line, licensee evaluat1on o

cable trays is continuing.

Adequate protect1on or be provided.

m enerator blowdown lines is similar to item 7 ool hi t

/tern e

t' k 1s not 11m1t1ng with p

d ion capability.

The containment spr y y

1 t'd after the effects on the cabl e rays

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Pipe breaks were not postulated in the primary loop on the basis of the work done under TAP A-2.

Me concur with this approach.

However, %he SEP branch intends to evaluate the effects on safety-related equipment of jet loads resulting from the crack sizes associated with these analyses.

TOPIC NO.

III-5.B Difference Summar

'TITLE Pipe Break Outside Containment 10 CFR 50 (GDC 4); as implemented by SRP 3.6.1, 3.6.2,.BTP MEB 3-1 and BTP ASB 3-1, requires in part 'that structures, systems and components important

' "to safety be designed to accommodate the dynamic effects of postulated pip'e ruptures.

The following are deviations from review guidelines that have been identified:

1)

Because high and moderate energy, line breaks in the screen house could

'damage the power'upplies to all service water pumps, the licensee must

~ 'rovide protection for these power supplies in accordance with Standard

.Review Plan 3.6.1 consistent with the service water system modifications which must be performed in connection %ith 'crther ongoing SEP reviews-arid ":.

'-'the fire-protection review.

4e g

TOPIC NO.

TITLE II.I-6 Difference Summar Seismic Design Consideration The requirements of 10 CFR 50 (GDC 2) and 10 CFR 100., Appendix A as imple-mented by Regulatory Guides 1.26, 1.29, 1.60, 1.61, 1.92, 1.122 and SRP 2:5, 3.7, 3.8,

3. 9, 3.10 prescribe structures, systems and components that should be 'designed to withstand the effects of a postulated earthquake without loss of capacity to perform their safety function.

The'evaluation res~its are summarized below:

'L 1)

The structures were found capable of withstanding the postulated se'ismic event except two sets of steel bracings located in auxiliary and turbine building for'hich modifications are required.

2).

ESW Pump'perability is an open,.item-;,-

34-- RVS Tmk and other safety related tanks are open items...

4)

Control room electrical panel structuFal integrity is an open item.

5)

The functional integrity of electrical equipment is be'tng evaluated by testino through SEP Owners Group program.

6) gualification of electrical cable trays is being evaluated by testing through SEP Owners Group program.

TOPIC NO.

III-7. A TITLE Inservice Inspection, Including Prestressed Concrete Containments with'ither Grouted or Ungrouted Tendons:.

Difference Summar Regulatory. Guide 1.35, Revision 2 as interpreted in the Standard Technical Specifications requires that the licensee have ar, inspection program th'at will

~

detect any structurally significant. deterioration of Category I structures in order that the structures will be capable of performing their necessary func-tions.

e o

owing r

Th f 11 e deviations between.the tendon surveillance program tor Guide 1.35, at Ginna based on current Technical Specifications and Regulatory Gui e

'Revision 2:

0 1)

The acceptable lift-offrequirement does not.meet current criteria because the existing Technical Specification at Ginna require that the average. of the 14 tendon stresses be greater than a value constant with time.

Cur-

". rent criteria requires that each tendon fall within acceptance limits that vary with time.

2)

Tendons which are found to be unacceptable a.re not handled as required in Section 7 of Regulatory Guide 1..35, Revision 2.

3)

Regulatory Guide 1.35, Revision 2 requires inspections and mechanical

.tests be performed on one unstressed wire per tendon per inspection, 4)

Ginna should include in its inspection.report wire breakage and filler

'rease.

'TOPI C HO.

Difference Summar TITLE Design

Codes, Design Criteria and Loading Combinations

'L

~ '10.CFR 50 (GDC 1, 2 and 4),

as interpreted by Standard Review'Plan 3.8, required

~'he plant to be designed and contructed to various design codes, criteria, loads

.and load combinations.

The fo1lowing. are u..eas where-differences exi~s between the plant design and current licensirig criteria; V

1)

Code changes have been identified in the following 'structural elements.

{See table next page. frog SEP Topic III-7.B issued 12/30/81.)

0

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2) 'oad and Load Combinations 3)

A ther al discontinuity exists in the liner plate at the point where tt6 insulation stops.

This will cause high thermal stresses in the liner during postulated LOCA temperatures and could result in the liner. buckling and failing.

TOP.IC t'0.

III-8. A Difference Summar TITLE Loose Parts Monitoring and Core Barrel Vibration Program The requirements of 10 CFR 50 (GDC 13),

as implemented by Regulatory Guide 1.133, Revision 1, and SRP Section 4.4 prescribe

.a loose parts monitoring program for the primary system of light-water-cooled reactors.

Ginna does not have a loose parts.monitoring program that mee s the criteria of Regulatory Guide 1.133.

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Structural Elements to be Examined members Desicned to rate in an Inelastic Re ime gev Code AISC 1980 Old Cede A.ISC 1963 Code Chan e Affecti These Klements Spacing of lateral bracing 2

9 2 ~ 8 Short Brackets and Corbels havi.ng a shear span-to-depth ratio of unity or less PCI 349-76'1.13 ACI 318-63 Shear Halls used as a

primary load-carrying meter ACI 349-76

11. 16 ACI 318-63 Precast Concrete Structural

Kiements, vhere. shear is not a member of diagonal tension Concrete Recions Sub'ect to Hich Temperatures A I 349-76

11. 15 ACI'49-76 ACI 318-63 ACI 318-63 Time-dependent and

. po'si 'on-dependent tempera ure va iations V

Columns ~ith 5"liced Re infor cemen.

subject to stress reversals;

~ 'y in compression to 1/2 fy in tension Pppendix A

ACI 349-76

~ tw 7.10.3 PCI 318-.63 Steel Kmbedments used to transmit load to concrete Containment and Other Klerents,, transmittina In-o}ane shear ACI 349-76 Appendix B

BapV Code Section III, Div. 2, 1980 CC-3421.5 ACI 318-63 P "I 318-63 Reeion of shell carrying concentrated forces normal to the shell surface (see ease study 13 for details)

BCPV Code, ACI 318-63 Section III, 1707 Div. 2 g 1980 CC-3421.6

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Structural Elements to be Examined Sc IL!as a.

Ccnposite Beams Nev Code AISC 1980 Old Code AISC 1963 Code Change hffectin These Elements'I 1.

Shear'onnectors in composite beams 1.11. 4 1.11.4 2.

Composite beams or g irder s vith formed steel deck 1.11.5

',:b.

Hybrid Girders Stress in flange Ceo=ress ion Elements With:vid D-to-thickness ratio hicher than speci-fied in 1.9.1.2 Tensicn Members When load is transmitted'y bolts or rivets

1. 10 ~ 6

" AISC 1980 1.9.1.2 and Appendix C-JrISC 1980 1.14.2.2 1 10.6 AISC'963

1. 9.1 4

AIS 1963 Connections a.

Beam encs vith top flange coped, if subject to shear AISC 1980 1.5.1.2.2 AISC 1963 b,

Connections carry'ng moment or festrained member connection

1. 15.5. 2 1.15 5.3 1.15.5.C

~Double dash

(

) indicates that no prov'isions vere provided in the older:-code.

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TOPIC NO.

TITLE

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Reactor Coolant Pressure Boundary Leakage Detection

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Review Criteria 50 GDC 2 and 30),

as implemented by SRP 5.2.5 and Regulatory Guide t f le kage from the reactor coolant pressure 1.45 requires the measurement o

ea ag

-nd states de-(

PB) t the containment and inter facing systems and s

a sign criteria for the systens employed for such.

,is em lo ed for measurement of leakage from.he RCPB to the 'con-e 1.45 states that

1) system should be an air

't monitor that is SSE qualified, 2) a minimum borne parts ula e

ad o t v y

of.two others should be present which are OBE qua s ie an

't to de.ect leakage of 1

gpm wst in our.

1 k

ho ld 1

d o

fo ctivit flow, level, pressure, temperature,.etc.

and f

measurement of sntersystem ea age s

o s t s sho ld 1) have al s and indica-

"""""'"""" '"'."'."'}'b..'d'1,'"".t.bl..-.-l b-"d d-iht normal operation, and have their availa

.y sn Difference Summar id lines that have been the deviations-from-review gui e.

nes

~

The following surmarszes sdentifi all of the recomn nded types of leakage detec'tion systehs 1

k o f.

th' ool t bo d

td th f 1't, th t

to the containment have b

o p

'n the leakaoe detection systems for the detection b

d 1

k 1 te.

h' R

1 to G

d Therei'ore, we cannot determine the extent to w sc egu ification. 3.4.4.6 and the corresponding sur-')

t nda d

ech al pec -

n.

illance requirem nts concernirig the op coolan

- pressure bo y

a Tichnical Specification

3. l. 5. 3 and FSAR the current basis for Ginna Technsca peci ic ised to state that the sensitivsties o

t leakaae detection systems.

prressure boundary to containmen e

Y'ity?nd

'im re u'ired to achieve sensitivity

~h

~ore t d temine the contribution of

. is incomple e.

Therefo this technique to the overall leak detec ion e

'TOPIC NO.

,Y-10. A Difference Summar TITLE RHR Heat Exchanger Tube Failures

. ~ I SRP 9.2,"1 requires that the service water system include the capability for detec.ion and control. of radioactive leakage into and out of the system and prevent accidental releases to the environment.

The.Service Mater System does not have a.radiation detector.

TITLE RHR Reliability TOPIC NO.

V-10. B

~ ~

'Di,ference Summar

~ 10 CFR 50 (GDC 19 and 34),

as implemented by SRP 5.4.7, BTP RSB 5-1 and Regu-latory Guide 1.139, require that the plant can be taken from normal operating conditions to'old shutdown using only sa ety-grade

systems, assuming a single fail.ure and utilizing either onsite or offsfte power through,the. use of suit-able procedur s.

The Ginna plant has safe.y-grade plant systems capable of safe shutdow'n under'hese conditions; however, the plant operating procedures ely upon other non-safety grade systems and: do-~ot.specify how the coo'Ldm'a=;.'.;-

would be accomplished by

~he operator in the event of fai1ures,in non-safeiy "

'rade systmE....Al'so, while we have concluded. that the OPS and -RHR:relief va>wes:

=

provide sufficient RHR system overpressure protection, ho'wever, the present tech-nical specifications would allow o]5eration of".the RHR without enabling the OPS.:

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TOPIC HO.

TITLE YI-4 Difference Summar Containment Isolation Systems 1)

The isolation valving arrangements do not meet the requirements of 10 CFR 50 (GDC 55 or 56),

as implemented by SRP 6.2.4 'from the stand-point of valve location for"penetrations 112, 120b, 12lc, 12ld, 123,'24b,'29,

140, 202,

203a, 203b,

205, 206a,

207a, 210, 304, 305a,

305c, and 332a.

2)

The isolation valving arrangements do not meet the requirements of 10 CFR 50 (GDC 55 or 56), as implemented by SRP 6.2.4 from the stand- '

point of valve number for penetrations

100, 102, 105, 106, 108, 109,'10a, and 110b.

3)

The isolation.valving arrangements differ from the explicit requi'rement's of 10 CFR 50 (GDC 55,'56 and 57),

as implemented by SRP 6.2.4 from the

~

standpoint of valve type by using a check valve outside containment:for penetrations 105, 109, 12la, and 129.

For penetrations 12la and 129 the nitrogen pressure regulating valve is

.::'not an adequate isolation valve.

The ical.ation provided does not meet the requiremeii%s.of 10 CFR-50 (GDp.':

55, 56 and 57),

as implemented by SRP 6.2.4 from the standpo'int of valve"

~ 'ctuation for penetrations 1/2, 120b, lac, 121d, 123, 201; 203,'05,

.206a,

207a, 209,

305a, 308 311, 312, 3l5, 316,

318, 320, 323, and 332a..

5) 10 CFR 50 (GDC 57 ), as implemented by SRP 6.2.4 was used to judge the acceptability of the isolation provisions for lines 301 and 303 (auxiliary steam heating to containment) since a closed system was identified inside

.." 'ontainment..

The.l.icensee.

should verify that this.,portion of.th system is of safety grade design to assure that the use of.GDC 57 is appropriate.

6)

The ESF reset pushbottons are inadequately protected from accidental actuation.

A TOPIC HO.

VI-7.B

'Dif erence Summar TITLE ESF Switchover From Injection to Recirculation Node

, 1)

Item 19 of SRP Section 6.3 states that the complete sequence of ECCS operation from injection to.long term core cooling (recirculation) should be examined to,see that a minimum of manual action is required, and that where manual action is needed a sufficient time (greater than 20 minutes is available for the operator to respond.

The current Ginna procedures for switchover from -injection to recirculation do not meet current NRC criteria for operator actions.

E

'2)

Branch Technical Positions ISCB.20 has not been satisfied because of the short tirrie.'(ll minLtes) 'Qiat is available for the operator to detect and

'correct a

ailure to follow procedures and his reliance on a single alarm

, to. a1ert'im to such an error.

TOPIC NO.

TITLE VIII-3. B -

DC Power System Bus Vo1 tagegonitoring -and knnunciat4qa

~ '

Di terence Sarsmar.

10 CFR 50.55a (h) as implemeqted 4y SRP 8.3.:Z,'and Regulatory Guide 1.47 requires

~

that the dc power system be monitored to the-extent that it is shown ready to

'erform its'ntended function.

The Ginna control room has no indication. of battery current, charger output current, charger output voltage, battery hi'gh discharge rate, bus under/over voltage, or battery or charger breaker/fuse status TOPIC NO,

'X-3 TITLE.

~ a Station Service and Cooling Water Systems'i fference Summar 10 CFR 50 (GDC 44),

as implemented by SRP 9.2.1 and SRP 9.2.2 requires.

a system to transfer heat from structures, systems and componets important to safety to an ultimate heat sink.

The technical specifications allow the plant to be operated with only two out of foui sei.vice pump's which, since two pu~ms are needed to handle post-accident heat loads, renders the system vul-nerable to a single failure. " There ss no redundant leveT 'indication forrtre CCW Surge Tank.

The failure of various non-seismic tanks could cause t'loodingl of variotis safety related equipment io the auxiliary building.

TOPIC NO.

. IX-5 TITLE

'Yentilation Systems Difference Summar 10 CFR 50 (GDC 60),

as implemented'by Standard Review Plan 9.4.5 requires that the plant include a means to suitably control ".the release of radioactive mater.-

. ials i'n gaseous and liquid effluents.

Current 'criteria requires that the capagility exist 'to d'irect ventilation air from areas of low radioactivity to'reas of progressively higher radioactivity.

There are two scenarios which could possibly violate this requirement, both of which occur with the"main exhaust fans shut-'down when offsite power is..not available and the plant is operating on emer-gency diesel power.

TOPIC NO.

IX-6 TITLE Fire Protection

'ifference Summar 4

10 CFR 50 (GDC 3),

as implemented by 10 CFR 50.4S and Appendix R requires that strictures-,

systems.'and "components 'important to safety shall be.'designed:a7zd:

loca<ed to minimize, consistent with other safety requirements, the probability

.and effect of fires.

Ginna cannet reach coke shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, as required by Appendix R, in zone ABRH, since a fire there could cause the loss of both RHR pumps.

0 h

)