ML20107F199

From kanterella
Jump to navigation Jump to search
Forwards Current List Re Status of Open & Confirmatory Items Identified in SER Sections 1.7 & 1.8.Resolutions to Items Also Encl
ML20107F199
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 02/15/1985
From: Mittl R
Public Service Enterprise Group
To: Schwencer A
Office of Nuclear Reactor Regulation
References
NUDOCS 8502260011
Download: ML20107F199 (14)
v • d • e


Text

i PS G Company Pudic Servee Electnc and Gas f

80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L. Mitti General Manager Nuclear Assurance and Regulation February 15, 1985 Director of Nuclear Reactor Regulation U.S.

Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, MD 20814

' Attention:

Mr. Albert Schwencer, Chief Licensing Branch 2 Division of Licensing Gentlemen:

SAFETY EVALUATION REPORT OPEN AND CONFIRMATORY ITEM STATUS HOPE CREEK GENERATING STATION DOCKET NO. 50-354 is a current-list which provides a status of the open_and confirmatory items identified in_ Sections 1.7 and 1.8 of the Safety Evaluation Report (SER).

Items identified as " complete" are those for which PSE&G has provided responses and no confirmation of status has been received from the staff.. We will consider these it' ems closed unless notified.otherwise.

In order to permit timely resolution of items identified as " complete" which may not be resolved to the staff's_ satisfaction, please provide a specific description of the issue which remains to be resolved.

[ Enclosed for.your review and approval (see' Attachment 3) are the resolutions to the SER items listed in Attachment 2.-

Should you have a_ny questions or require any additional,

'information on these items, please contact us.

Very truly yours, g'pl p"#mu nun, p

-Attachments PDR

-u C

D.

H. Wagner USNRC Licensing ~ Project Manager (w/ attach.)

. The Energy People -

'A.'R._Blough' mm uup er

.USNRC Senior Resident Inspector-(C/ attach'.)

~

ATTACHMENT 1 R.

L.

Mittl to A. Schwencer

, Item-No.

Subject Status ltr. dated OI-l

'Riverborne Missiles Partial Response 1/31/85 OI-- 2 '

' Equipment Qualification Pattial Response 2/1/85 OI-3 Preservice Inspection Program Partial Response.

2/14/85 OI-4 GDC :51 ~ Compliance Open

'LOI-5 Solid-State Logic Modules NRC Action OI-6'

~Postaccident Monitoring NRC. Action Instrumentation OI-7:

Minimum Separation Between Open Non-ClassLIE-Conduit and Class'IE' Cable Trays.

OI-8 1 Control of_ Heavy Loads-Completed 1/18/85 OI-9 l Alt'ernate and Safe! Shutdown-NRC Action OI-10 Delivery of: Diesel Generator-Closed-Amendment 8 Fuel' Oil and Lube. Oil

.-OI-.11'

. Filling".of Key Management'

.Open,

. Positions.

I ' Completed 1/7/85?

01-12 ^

~ Training? Program' Items;

!('h).! Initial Training' Program [ Completed

-12/28/84 P

( b)'

Requalifica't' ion Training; Completed :

11/7/85-,

Program-ic (c)1: Replacement Training'.
Completed

~1/7/85; y

Program

/(d)\\ iTMI Issues ; I. A'. 2.1,, _

Completed.

' l/7/8 5.

i

=I J A. 3.1,,fand II.B.'4

.n

.,)

~

c (e). Ncnli_ censed Traininig :

Completed _

1/,7/851

+

4 Program

...01-13P

Emergen'cy. - Dose L Assessment" Closed:-

L1/7/85:

[- [

Computer.Modell

,II/28/85l:

-(DI-145 l IProcedures GAnerstionjPackage;: Closed; r

n ; ".

S IOI-15!

9- [ Human FactorsfEngineering jOpen.

p -

m J.

~

> ?lh FMLP 85}27/10/1,-ar?

+

Jy

-e s

. k?

r, 4

4

2 R.

L.

Mittl to A.

Schwencer Itcm No.

Subject Status ltr. dated C-1 Feedwater Isolation Check Open Valve Analysis C-2 Plant-unique Analysis Report Completed 1/8/85, 1/11/85,

& 1/31/85 C-3 Inservice. Testing of Pumps and

. Open Valves C-4 Fuel Assembly Accelerations Completed Amendment 8

~C-5 Fuel Assembly Liftoff

- Completed

. Amendment 8 C-6 Review of Stress Report Open C-_7 Use of' Code Cases Completed 12/17/84 C-8 Reactor Vessel Studs and'Fastners._ Completed 2/15/85 C-9:

Containment Depressurization NRC Review Analysis C-10 Reactor Pressure Vessel Shield NRC Review Annulus Analysis

' C-ll

. Drywell Head Region Pressure

'NRC Revi'ew

- Response Analysis.

~

C-12L Drywell-to-Wetwell Vacuum Breaker,;NRC Review

-Loads

-C-13

Short-Term 'Feedwater-System-

- Opens

' Analysis-LC floss-of-Coolant-Accident. Analysis,Open:

LC-15 Balance-of-Plant Testability.

l Completed-

. Amendment:8-Analysis'

~C-16:

-Instrumentation Setpoints DCompleted

2/15/85 w

C-17fs

' Isolation --Devices

,Open; fMRCRyview

C-18

~ Regulatory Guide l~.75 ~

'C-19 LReactor Mode.~ Switch'

OpenJ m

-e 3

' Engineered Safety'Featuresc cu Open1.

C-20 1

~

' MG Roset1 Controls.

t g

r

[MiP85127/1012-ar s

n y

3 R.

L. Mitti to A. Schwencer Item No.

' Subject Status ltr. dated C-21 High Pressure Coolant Injection Open Initiation

=C-22 IE Bulletin 79-27 Completed Amendment 8

C-23;

-Bypassed and Inoperable Status NRC Review Indication C-24 Logic for Low Pressure Coolant Open Injection Interlock-Circuitry C-25.

.End-of-Cycle Recirculation Pump Open.

Trip C-261 Multiple Control System Failures NRC Review

.C-27

' Relief Function of Safety / Relief Completed 2/15/85 Valves

,-C-28 Miin Steam Tunnel Flooding Open Analysis C-29; Cable Tray Separation-. Testing-Open 1

C-30.

.Use:ofEInverter:as. Isolation' Open'

' Device

_C-31'.

. Core Damage' Estimate Procedure.

Open-C Continuous Airborne! Particulate =

Open

Monitors-C-33l LOualifications ' of - Senior : Radlation 'Open.-

Protection Engine.er'.

cC-34

Onsite_ Instrument.Information Open

?c-35'

-Airborne Iodine Concentrationi

Open 1 Instruments-C-36.'

' Emergency [ Plan Items--

lPartiallResponse}ll/9/84,;

1/16/85, E

~#'

'2/7/85

~

IC-37l

'iTMIiItem II.K.3.'18-iOpen --

~

f M:P85)27/1033-ar J

,3 n

b.

3:

',} ; -

. i. i.[

'E i

7 4

ATTACHMENT 2 ITEM NO.

- SER-SECTION SUBJECT C-8 5.3.1.5 Reactor vessel studs and fasteners

' C-16 7.2.2.5 Instrumentation Setpoint C-27 7.7.2.2 Relief Function of Safety / Relief Valves-JES:mr.

' NN 13-la' YA.

p

/

s 1

t J

,.j L

h j.'

r' f;:.;.

j'

.c J

f k

s

^'-1 y,

'E

+

e L

2

_;ui

.:(

~

cl

,z..

3:

q j

'd

-C 1

T 9 :.

zj ll g/.

(

l-sj;: ap. wgg -

,e

1 e

4 0

.g 2

i 1

1 1

t E

i i

1 1

(

l ATTACHMENT 3 1

4 e

l t

I 4

i 9

i SER' ITEM NO. C-8 REACTOR VESSEL STUDS AND FASTENERS The reactor vessel studs and fasteners satisfy most of the recommendations of RG 1.65, " Materials and Inspections for Reactor Vessel Closure Studs."

The FSAR does not discuss the nondestructive examinations of the bolts and nuts, and the applicant.needs to confirm that the Code-specified inspections were performed.

This is a confirmatory issue.

RESPONSE

FSAR Section 5.3.1.7 has been revised to provide the infor-mation _ requested above.

4 7

m s

+

e M P85 29/0.6:1-cag.

I d

4

+~

Ju 24 850 2 7 7 3 4 8 4

HCGS FSAR 8/83.

5.3.1.7 Reactor Vessel Fasteners The reactor vessel closure head (flange) is fastened to the' reactor vessel shell flange by multiple sets of threaded studs and nuts.

The lower end of each stud is installed in a threaded Mole in the vessel shell flange.

A nut and washer are installed on the upper end of each stud.

The proper amount of preload can I

be applied to the studs by a sequential tensioning using hydraulic tensioners.

The design and analysis of this area of the vessel is in full compliance with all ASME B&PV Code,Section III, Class I requirements.

The material for studs, nuts, and washers is SA-540 Grade B24.

The maximum, reported ultimate tensile strength for the bolting material is less than the 170,000 psi limitation in Regulatory Guide 1.65.

Also the Charpy impact test recommendations in Paragraph IV.A.4 of Appendix G to 3

10 CFR 50 were not specified in the vessel order since the order was placed prior to Assuance of Appendis G to 10 CFR 50.

However,. impact data from the certified materials report shows that all bolting materials have met the Append G immet properties. Tlig nondesfresMs. annusemelsas Se riseestest of

  • D

'W Hf fenen werc e,4 e

A phosphate coating was applied to threaded arets of studs, nuts and bearing areas of nuts, and washers to act as a rust inhibitor and to assist in retaining lubricant on these surfaces.

5.3.1.8 SRP Rule Review-5.3.1.8.1 Acceptance Criterion II.2 SRP 5.3.1 acceptance criterion II.2 requires that the reactor.

vessel and its appurtenances be fabricated and installed in accordance with ASME R&PV Code, Section.III, Paragraph NB-4100.

The manufacturer or installer ct such components is required to certify, by application of the appropriate Code symbol and completion of an appropriate data report in accordance with ASitt B&PV Code,Section III, Paragraph NA-8000, that the materials used comply with the requirements of NS-2000, and that the fabrication or installation comply with the requirements of

.NS-4000.

The HCGS RPV and appurtenances were manufactured'in accordance with the 1968 edition of the ASME B&PV Code, Section'III, which does not have NB-designated subarticles.

In light of NCGS's compliance with 1968 ASME B&PV Code,Section III, and information i

l 5CA #4N 5.3-11 Amendment.1 c-8

. h SER ITEM NO. C-16

\\

INSTRUMENTATION SETPOINT The staff will confirm that the resolutions of the generic issues concerning the setpoint methodology are appropriate and successfully applied to the Hope Creek Technical Specifications.

- RESPONSE FSAR: Ouestion Response 421.18 has been revised to reflect the NRC acceptance of the setpoint methodology program.

The HCGS ' Technical Specifications. will be revised by 12/85 as required.

i 4

A.

J.

v

. M P85(29/06E2-cag!

J b

~

4

l' '

HCGS FSAR JAN 24 '850 2 7 7 312 4/84 OUESTION 421.18 (SECTIONS 7.2 AND 7.3)

Provide a detailed discussion on the methodology used to establish the technical specification trip setpoints and allowable values for the Reactor Protection System lincluding i

Reactor Trip and Engineered Safety Feature channels) assumed to j

operate in the FSAR accident and and transient analyses.

Include the following information:

1.

The trip setpoint and allowable value for the technical specifications.

~

The safety limits necessary to protect the integrity of the 2.

physical barriers which guard against uncontrolled release of radioactivity.

The safety limits should be the limits established for licensing purposes, for example the technical specification safety limits on minimum critical power ratio (1.06), and reactor coolant system pressure t

l (1325 psig).

The values assigned to each component of the combined 3.

channel error allowance (e.g., modeling uncertainties, analytical uncertainties, transient overshoot, response time, trip unit setting accuracy, test equipment accuracy, primary element accuracy, sensor drift, nominal and harsh O

environmental allowances, trip unit drift), the basis for these values, and the method used to sua the individual l

errors.

Where zero is assumed for an error a justification l

that the error is negligible should be provided.

i 4.

The margin (i.e., the difference between the safety limit and the setpoint less the combined channel error allowance).

5.

Identify any trip for which the setpoint and allowable value in the technical specifications will be assigned best estimate values and for which you do not have an analysis of errors and/or uncertainties to confirm that the trip function will occur before the actual value of the measured parameter exceeds that assumed in the plant safety analysis.

l Provide justification for this nonanalytical approach.

RESPONSE

I Public Service Electric and Gas is currently participating with a number of other utilities and the General Electric Company in a tesseerd o p --. <,. 4a m.....a.-- -it' t'; -- It:"' :::::: in; the methodology used to establish'the technical specification trip setpoints and allowable values for the reactor protection system.

All the is. sues r,aised by this question are. be.ing covered.by it::: ;:::ri i i d s Q,

m.. M t.., s w e.

d i m i n c ~ ~s a f p s ~ A n d w m & a s h 1

42i.is-i amendment s

.s c 4 irem 4

M a

i-un m m

  • a r-f) #1 M P1 #1 1 a HCGS FSAR " " "5 4/84 e'

ue i

be li o

HC as pr 1

ci all he a

req tw b

u i

pts

.ma sehe4 le. -for=<rds41s k Nad Sinh ra Refeue e< l.

me. see. the-SRys 'is a reference to -the nonsafety" function jofl the a

.s; y

l

/

i LMiP85.29/06L3-cag

/

4

- fJ

, Al i

Q'

o SER ITEM NO. C-27 (CONT'D)

RESPONSE (CONT'D) manual relief mode and not to the qualification of the elec-trical. components.

Since the components and power supplies j

of both the manual relief function and the automatic depres-

- surization function are safety grade, no isolation problems exist between these functions.

FSAR Table 440.33-1 has been revised to clarify this item.

l

.)

a

+1 1,{

s n

s' 2l l

. I p,

-..};

lL

?

.g

p. 4 r

.n,

M < P85 :.29/06 ' 4-cag.

~"?6 s

11 :

o

,1

y

+

7

~

i t,-

p s

2, -

' ;m. ' M eb h%. d a a s

j

j i

)

Js 24 '85 a 2 77 3 4 7 y,

\\

TABLE 440.33-1 (cont)

Page 2 of 2 INFREQUENT EVENTS 15.2.2 Load rejection without bypass Relief valves 15.2.3 Turbine trip without bypass Relief valves l

dIMITING EVENTS 15.3.3 Recirculation pump seizure Level-8 turbine trip and feedwater pump trip, i

turbine bypass relief l

valves 15.3.4 Recirculation pump shaft break Level-8 turbine trip and feedwater pump trip, turbine bypass, relief i

valves i

es sesenesal l

O)

" Relief M lves" refers t nonsafet -related 1.._i__ : ti ti ; "-

relief mode of th SRVs. Al Np does M ser;W a l

q in "ik h adie.IA e.M c--/ d

.m e - -t us

-f T.

ya i

l l

l 1

i CA t TEN O~

l Amendssnt 3 l

i

-

Retrieved from "https://kanterella.com/w/index.php?title=ML20107F199&oldid=4298852"