ML20043A900

Application for Amend to License NPF-47,revising Min Temp Vs Reactor Pressure, Curves Contained in Section 3/4.4.6 & Associated Bases

ML20043A900
Person / Time
Site:	River Bend
Issue date:	05/14/1990
From:	Deddens J GULF STATES UTILITIES CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20043A901	List: ML20043A900 ML20043A902 ML20043A905 ML20043A906 ML20043A908
References
RTR-REGGD-01.099, RTR-REGGD-1.099 GL-88-11, RBG-32835, NUDOCS 9005230259
Download: ML20043A900 (9)

Text

'

) .

GUXaF STA TES . UTZLXTIES COMPANY NWR tiiND ST AflON POST OFFict BOX 220 Ye5

- 57 FRANCISvtut, toutstANA 70776 ARE A CODE 604 636 6a94 - 346 BH51 1

May 14, 1990 RBG 132835 File Nos. G9.5, G9.42 U. S. Nuclear Regulatory Commission Document Control Desk Washington,'D. C. 20555

' Gentlemen:-

' River Bend Station - Unit 1 l Docket No. 50-458 Gulf. States Utilities Company (GSU). hereby file's an application- ,

to' amend the River Bend' Station - Unit. 1 Technical- j Specifications, Appendix A to Facility-Operating License NPF-47,  ;

pursuant to 10CFR50.90. 'This application is-filed to revise the i

" Minimum Temperature. vs . Reactor Pressure" curves ~ contained in j section 3/4.4.6 and the-associated bases. -This' includes : changes j to the . lead factors for the. reactor vessel-inside diameter and i 1/4 thickness depth, changes to the reactor vessel toughness -

i informationL and the " Fast Neutron- Fluence .(E > 1 MEV) at Vessel l I.D. as a Function of Service Life" ' curve.. These changes. are also requested in order to comply with the requirements of NRC l Generic Letter 11. To support the required reactor' coolant pressure boundary testing 'to be done near the end of Refueling .i Outage 3, approval of tthis License Amendment is- requested prior :l to October 1, 1990. The' attachment to this letter and the. 1 enclosure provide the justifications and' proposed revisions to i the Technical Specifications. l f

Your prompt attention to this application is appreciated. l 3

Sincere ,- l i

N "J. C. Deddens ]

Senior Vice President'  !

, River Bend Nuclear Group l TFP 0/ / b F/pg j Attachment  !

f.  !

"" 90$ /

l: 9005230259 900514  !

- PDR ADOCK 05000458 P PDC /[ 3 i

y jg , "

a- .

~

y

• A

• ^i q

l cc: U! S. , Nuclear Regulatory: Commission'.: i

.611.RyanLPlaza. Drive, Suitel-1000i a l Arlington,TXE76011 l t

..NRC-Resident Inspector. , <!

Post' Office' Box:1051- 1 St. Francisville,.LA4 70775.. I

Mr.. Walt Paulson:

.U.ES. Nuclear; Regulatory Commission' ji One! White. Flint Rock-North' y i

11555:Rockville P,ikez .

Rockville,MD120852- ,

~

4: <

Mr. William H. Spell', Administrator.

-Nuclear Energy'. Division . . _

. Louisiana. Dept.Lof Environmental; Quality; '

s P.'O.,BoxJ14690- ,

4 Baton Rouge,- LA:70898_ '

i u

il

.- fff P

q i

(!

..x dn i

l .. -

l

).[

1 P

l [l i

.j

! ~

ll s o ,

a f , 1 h s

- , , 9 ,_ , / . - lJ.'

i 1

s UNITED STATES OF AMERICA- )

NUCLEAR REGULATORY COMMISSION

' STATE OF LOUISIANA )

PARISH OF WEST FELICIANA -) 7 In the Matter of ) -

GULF STATES UTILITIES COMPANY ) )

-(River Bend Station - Unit 1)

AFFIDAVIT J. C. Deddens, being! duly sworn, states ,that he is'a-Senior Vice President of Gulf States Utilities Company; that he is authorized on the part of said company to sign and file

]

with the Nuclear Regulatory Commission the documents attached-l hereto; and. that. all such documents are true and-correct'to -

i l the best of his knowledge, information and belief.-  !

l l

J.

/

C/ Deddens n

Subscribed and sworn to before me, a Notary Public in and s for the State and ' Parish above= named, this /$IYAi' day of M ag , 1990 . My Commission expires with Life, h

01an d J.kket, Claudia F. Hurst Notary Public'in and for West Feliciana Parish,~ Louisiana i

e i

1 Gulf States Utilities Company L River Bend Station  ;

Docket 50-458/ License No. NPF-47 ,

1.-

Technical Specifications / Surveillance Requirements l_

Licensing Document Involved. .

J Changes to the RBS Technical Specifications are. requested by this l License Amendment Request. Affected-Tech. Spec.' sections include: r L A. Reactor Coolant - Pressure / Temperature Limits.3/4.4'6 .

I B. Bases Se: tion 3/4.4.6 The specific sections to be revised are as follows:

1. Figure 3.4.6.1-1 (p. 3/4-23)

2. Table 4.4.6.1.3-1 (p. 3/4-24)

3. Bases page 3/4 4-5 1

4. Bases'page 3/4 4 5. .

l-l

6. Bases Table.B Bases Figure 3/4'4.6-1 (p(p.

B 3/4.4.6-1 B 4-9)

. B 3/4 3/4 4-8) l Reason For Request This proposed- change is being requested in accordance. with 4 10CFR50.90. Changes to the " Minimum ' Temperature. Required vs. t Reactor Pressure" curves contained! in section- 3/4'.4.6 and the

! associated bases section are requested, This includes changes to. .

l the lead factors for the reactor. vessel- inside~ diameter and 1/4-thickness depth. changes to reactor ' vessel toughness information (Bases Table B 3/4.4.6-1), and the " Fast Neutron-Fluence (E>1 MEV) at Vessel .I.D. As ' A Function of Service Life" curve-(Bases' Figure 3/4.4.6-1).

These changes are requested in order to comply with t'he requirements of NRC Generic Letter 88-11. In Generic Letter 88-11', the NRC l advises that all licensees use the methods described in Revision 2 of Regulatory Guide'1.99 to predict the effect of neutron radiation on reactor vessel materials -as required- by -paragraph V.A. of 10CFR50, Appendix G. . Use of the Revision 2 methodology results in a revision of the adjusted reference temperature- for nil-ductility transition (RT-NOT) and the pressure-temperature limits contained in l Technical Specification 3/4.4.6. GSU provided its response to Generic Letter 88-11 (RBG-29292 dated 11/3/88) and-committed to revising the Pressure-Temperature curves and related information in the RBS Tech. Specs. and USAR prior to startup from RF-3.

I' i

i Page 1 of 6 .

1 Description Changes to-the RBS Technical Specif'ications proposed by the License Amendment Request are divided into =two basic areas:. 1)-

pressure-temperature curves and 2) neutron fluence as a function of service life. Discussion _of the regulatory basis and descriptions ,

of the changes- proposed for these. areas .is given below.

RBS-specific analyses .are provided in.. the . referenced reports (Attachments 1_and 2) and are not repeated here.

Regulatory Basis L 10CFR50, Appendix A, General-Design Criterion 14 " Reactor Coolant ll Pressure Boundary," requires that the reactor coolant -pressure.

L boundary. be designed, fabricated, erected, and tested in order to have an extremely low probability of- abnormal leakage, of rapid.

failure, and of - gross ruptu re. General- Design ' Criterion 31,.

-" Fracture Prevention of ' Reactor. Coolant. Pressure Boundary,"

requires, in part,. that; the. reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed _ under.

operating, mainter:ance, .and testing. the boundary . behaves in a nonbrittle manner ~and the -probability:'of rapidly propagating fracture is minimized. In ' order to -assess :the . structural integrity i

' of the reactor Reactor vessel

. Coolant .: GeneralBoundary,"

. Pressure Design criterion ~ 32,' _in requires ," Inspection part, .an -of appropriate materials surveillance program for the? . reactor ~ vessel.

beltline. region.

10CFR50, Appendix G,Section IV specifies the. fracture toughness requirements for ferritic materials in - the= pressure - retaining, ,

components of the reactor coolant pressure boundary for testingLand operational conditions,-including operational occurrences. ' Appendix H of 10CFR50 requires a material surveillance program to monitor ~

changes in fracture - toughness properties of materials . in .the ,

beltline region, resulting from neutron irradiation throughout the.

service life. Regulatory Guide 1.99, developed in response- to the above regulations, describes general procedures acceptable to the NRC for calculating the effects of neutron radiation . embrittlement of the low-alloy steels used in reactor pressure vessels. 10CFR50, Appendix 11 and Regulatory Guide L1.99 specify .that the' material surveillance program must meet the requirement of ASTM E 185. -This standard requires, in_part,.that neutron dosimeters be included as part of the reactor material surveillance program.

The two measures of radiation embrittlement used in Reg. Guide-1.99 are obtained from Charpy V-notch impact tests. The adjustment of reference temperature, RT-NDT, is defined in'10CFR50, Appendix G as ,

the temperature shift in the Charpy curve for the irradiated material relative to that for the unirradiated. material measured at the 30-foot-pound energy level. The second measure is the decrease in the Charpy upper-self energy level which is defined in ASTM E 185. The current RBS USAR and Tech. Specs. for these values including the P-T curves were prepared in'accordance with Revision 1 Page 2 of 6

i of Reg. Guide 1.99. Revision 2, issued in May, 1988, updates. the calculational procedures -for the.- adjustment of, reference temperature; however, calculational procedures for decreases in upper-shelf energy are not changed. Thus, revision of.the reference temperature. and P-T curves is required _ in order. to maintain .;

compliance with the above listed regulations and GSU's commitment in response to Generic Letter 88-11.

Description of Changes An evaluation of the. impact resulting from~ application of Revision'2 i methodology was prepared by GE and is documented in report SASR 1 89-20,Rev.1(Attachment'1). This report also provides revised P-T curves for use in the RBS USAR and Tech Specs., and~ recommended r l USAR and Tech. Spec, changes. Justification for revision of the lead- l

' factors (ratio-of the . neutron flux, density at the surveillance ,

capsule specimen to the neutron flux density at-the reactor pressure .

vessel inside surface at the peak fluence location) is provided in a ~I GE letter included as part of Attachment 1. Specific-changestto the Tech. Specs. are described below, a.. Figure 3.4.6.1-1 (page 3/4 4-23) - - The upper portion of each curve is revised- as a result of change in adjusted reference temperature per Reg. Guide 1.99_,.Rev.2.

Notes'on thisfigurgarealsorevisedtoshow.achangeinamount g

of shift initial(48 F to111F),limitingmateriaf(

reference temperature (9 F'to -50 F)that plate to weld),-

, curves A, B, and C are valid for 2:EFPY 'of operation, and that i

curves A', B', and C' are valid for 8 EFPY of operation.,

l- (Note that the- first- reactor . material surveillance ,

specimens are to be removed.at 6 EFPY- which' allows for revision,'as necessary, of-the P-T curves beyond'8 EFP.Y).

b. Table 4.4.6.1.3-1 (page '3/4 4-24) - - The third column heading is revised to state lead factors at both the inside diameter (I.D.) of the vessel and at the 1/4 wall thickness (1/4 T) depth. .The-lead factors'are revised to 0.67/0.89-for each: surveillance capsule number. Revision-of the lead ,

factors results from-use o' more accurate ~two-dimensional i computational method;. Lead factors, computed by GE, relate the dosimeter flux to the flux at the. peak location in the vessel beltline region for 32. EFPY of reactor operation.

c. Bases Section 3/4.4.6 (page B 3/4 4-5) - - Several changes are made in the last' paragraph to reflect the use'of Rev. 2-of Reg. Guide 1.99.(the title of the Reg. Guide was also changed with issuance of Rev.2). 'The' reference 'to phosphorus content is' replaced with nickel content in accordance with Reg. Guide 1.99, .Rev.2. Reference to "end-of-life fluence" is replaced with " conditions at 8 EFPY" since the A', B', and C' curves have been calculated for 8 EFPY instead of 32 EFPY, These curves will be revised after results of the first surveillance capsule is Page 3 of 6

~

obtained.. This allows for the use of additional data in calculating the P-T curvesLbeyond 8 EFPY._

d. Bases- Section 3/4.4.6 (page B 3/4. 4-6) - - In the first -

paragraph, reference to the specific year edition (1973) of ASTM E185 is deleted since 10CFR50, Appendix-H allows the use.of any one of three= editions of this standard. The -

Reg. Guide revision _ number .i s also revised -in this paragraph.

e. Bases Table B 3/4.4.6-1 (page B 3/4 4-8) -- - Multiple changes are- made to ;this table as a result; of -the application'of Rev.'2 methodology.--The' column heading for.

P- (%) is replaced with Ni.(%)-to reflect the-change from the use of copper and phosphorus in Rev.-1 to -copper and :i nickel in Rev.2. Other column headings are revised-for clarity. ' Also,;"End-of-Life (E0L)"-is replaced with "32 EFPY" in the note.for this table.

Values of delta RT-NDT- and . maximum' RT-NDT:-for 0 vessel.

beltline plate material are revised to 75 F and 84 F respectively _as a: result of applying the . flux wire- (1 dosimetry-test results for neutron fluence and the Rev.2--

methodology. The, first value .is'the amount of shift-in nil-ductility transition reference temperature.over 32 EFPY and the'latter number is the adjusted: reference: temperature-for 32 EFPY of reactor. operations.- The nickel; content for the beltline plate material' is -0.63%.

Revision of . values for weld data reflect a change in the-heat and lot number that produce the limiting results with application of 'the = Rev. 2 methodology. The heat and lot numbers for the limiting weld material are SP6756 and 0342 respectively. Values for Cu, .Ni~, delta RT-NDT,-average upper shelf energy value, and maximum RT-NDT are 0.09%,

0.92%, 153 F, 97 ft.-lbs., and 103 0F respectively,

f. Bases Figure B 3/4.4.6-1 (page B 3/4 4-9) - - This figure-is revised based upon flux wire dosimeter test results and revised lead factors.. The flux wire dosimeter was removed after the first fuel _ cycle and analyzed by'.GE at its Vallecitos Nuclear Center. Results of their analysis are documented in report SASR 88-49 (attachment 2). Si nce' the vessel fluence is proportional to thermal power produced, the results'of flux wire dosimeter test.are used to provide

~

a calibration point of vessel fluence versus accumulated thermal power. A linear extrapolation provides an estimate of fluence at 32 EFPY.

No Significant Hazard Considerations The following discussion is provided to the NRC in support of "no significant hazards" per 10CFR50.92. Of the items to be changed by this license amendment request, only the pressure-temperature curves Page 4 of 5

are used as limiting' conditions of operation and as surveillance .

requirements. Operating limits for pressure and temperature are required for three categories of operation: . 1) hydrostatic pressure tests and leak tests (curves A and A'), 2) non-nuclear heatup/cooldown and low power physics tests (curves B and. B'), and r

3) core critical. operation (curves C and C'). The A, B, and C curves serve as baseline curves and are applicable 'up to 2 EFPY.

The A', B', and C' curves provide the pressure-temperature limits for their respective activities. through 8. EFPY, As required by Tech. Spec, surveillance statement 4'.4.6.1.3. these curves will-be updated as necessary based upon results -of the reactor. vessel '

surveillance specimen examinations.

, The revised pressure-temperature limits-are more' conservative'than l

that given by the_ original. Tech. Spec. curves. This is true- also when accounting for the revision of the lead factors. While the use, of the revised and more accurately calculated lead factcrs would by itself result in slightly less-conservative pressure-temperature limits, their application with the Reg. Guide 1.99, Rev. 2 methodology results in more conservative limits for temperature and pressure.for all operational conditions.

These changes are to be made in order to make the'RBS Tech. Specs.

I conform to changes in the regulations (Regulatory. Guide 1.99). The

-proposed changes are, therefore~, fully within. the regulations and maintain the margin of safety required for reactor vessel materials.

-Operation of RBS in accordance with the changes proposed. in this amendment involves no significant- hazards based upon the evaluations -

given below.

a. Changes made through this request are 1.n accordance with the applicable design, material, and construction standards.

These revisions-are the result of changes in calculational-methodology for shift in nil-ductility transition reference temperature promulgated by issuance of Revision 2 of-Regulatory Guide 1.99. The changes, recognizing the use of revised and slightly less conservative lead factors, result in more conservative limits for temperature and pressure for ,

all operational conditions (hydrostatic and leak testing, non-nuclear heatup/cooldown and low level physics testing,  :

and core critical operations). Consequently, no significant-increase in the probability or the consequences of an accident previously evaluated results from these changes,

b. No structures, systems, or components are added or ' deleted by these changes; thus the possibility of additional single failures resulting in a new or different kind of accident from that previously evaluated is not introduced as a result of these changes. The changes provide more conservative limits for temperature and pressure at higher system pressures and no limits are removed or made less conservative. Therefore, these changes would not create the possibility of a new or different kind of accident from any accident previously evaluated. ,

Page 5 of 6

c. A larger shift in RT-NDT results from application of the Regulatory =

Guide 1.99, Rev. 2 methodology. Consequently, the _ revised pressure-temperature curves to. be used for RBS Tech. Spec.3/4.4.6 -

are more conservative. Thus, . additional margin is provided to assure that when the reactor pressure vessel is stressed.under operating, maintenance, testing and postulated accident conditions, the reactor coolant pressure. boundary behaves in'a nonbrittle-manner and the probability of rapidly propagating; fracture is minimized. Therefore, these changes- would not involve a significant reduction in margin of_ safety.

The proposed. amendment will not Lincrease the possibility .or 'the consequences. of a-previously evaluated event and will not crease a new or 4 different kind of accident from any- previously evaluated.. Also, the results of. this proposed change are_ clearly within' all' acceptance criteria with respect to system-components and. design -requirements. The ability of the reactor pressure vesse'l

~

to perform as ' described in the=USAR is3 mainteined and therefore, the proposed change does not. involve- .a significant . reduction in margin ~ of safety. GSU proposes that no_ s signir'icant reduction in margin of safety. GSU proposes that' no significant hazards are' involved for these changes'.

Proposed ticense Revision Marked-up pages of the RBS Technical ' Specifications are provided in Enclosure 1..

Schedule For Attaining Compliance

. River Bend Station is currently.E in compliance - with the' applicable' Technical Specifications. However, USV:has~ committee to revision of the-pressure-temperature' curves given in the- RBS Tech.' Specs. prior to startup from the third refueling outage scheduled-for September, 1990-(reference.

RBG-29292 and TRAC item 07130).

In order to support the required. system pressure testing (Reactor Coolant Pressure Boundary) to be done near the end of Refueling Outage 3, approval of this License Amendment Request by October 1, 1990 isirequested.

NOTIFICATION OF STATE PERSONNEL A copy of the amendment application and this. submittal is being provided-to the State. of Louisiana, Department of Environmental Quality-Nuclear Energy Division, ,

. 1 ENVIRONMENTAL IMPACT APPRAISAL Gulf States Utilities Company (GSU)_ has reviewed the proposed license amendment against the criteria of 10CFR51.22' for' environmental-considerations. As shown above, the proposed changes do _ not- involve' - a '!

significant hazards consideration, nor increase the types and amounts of-effluents that may be released offsite, nor significantly increase ,

individual or cumulative occupational radiation exposures. Based on'.the '

foregoing, GSU concludes that the proposed changes meet the criteria given n in 10CFR51.22(c)(9) for a categorical exclusion from the requirement for an Environmental Impact Statement.

l

)

Page 6 of 6

- _ _ - .