RC-05-0090, License Amendment Request Re Reactor Coolant System - Heatup/Cooldown Curves

License Amendment Request Re Reactor Coolant System - Heatup/Cooldown Curves

ML051790033
Person / Time
Site:	Summer
Issue date:	06/22/2005
From:	Archie J South Carolina Electric & Gas Co
To:	Martin R Document Control Desk, Office of Nuclear Reactor Regulation
References
RC-05-0090
Download: ML051790033 (22)

Text

Jeffrey B. Archie Vice President, Nuclear Operations 803.345.4214 June 22, 2005 RC-05-0090 A SCANA COMPANY Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Mr. Robert E. Martin

Subject:

VIRGIL C. SUMMER NUCLEAR STATION DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 LICENSE AMENDMENT REQUEST REACTOR COOLANT SYSTEM - HEATUP/COOLDOWN CURVES

Reference:

Jeffrey B. Archie, SCE&G, to NRC, RC-04-0170, October 22, 2004 pkursuant 1o 10 CFR 50.90, South Carolina Eiectric & Gas Company (SCE&G), acting for itseif and as agent for South Carolina Public Service Authority, hereby requests an amendment to the Virgil C. Summer Nuclear Station (VCSNS) Technical Specifications (TS).

The proposed changes will revise the heatup and cooldown curves located in TS section 3/4.4.9 Reactor Coolant System Pressure/Temperature Limits and the associated Technical Bases.

These changes are required based on the analysis and results of the last reactor vessel surveillance specimen that was removed and analyzed as detailed in the above referenced letter. The requested changes result in curves that are more stringent than those currently in the VCSNS TS. In accordance with Administrative Letter 98-10, actions have been implemented to procedurally revise acceptance criteria and incorporate the more stringent requirements into the applicable procedures as an administrative control.

Enclosed as Attachment IV is WCAP-16305-NP, Revision 0, 'V. C. Summer Heatup and Cooldown Limit Curves for Normal Operation", dated August 2004. This report details the development of the revised heatup and cooldown curves based on the data obtained from the analysis of the surveillance capsules removed from the VCSNS reactor vessel and updated calculated neutron fluences. The new VCSNS curves were generated using the 'axial flaw" methodology of the 1998 ASME Code, Section Xl through the 2000 Addenda. Included in this methodology is the use of the Kic stress intensity factors that were formally documented under ASME Code Case N-641.

The curves provided by this proposed change do not include instrument uncertainties. The curves that provide the operational limitations are located in plant operating procedures.

Instrument uncertainties, elevation differences between the relief valves and the reactor vessel beltline, and the effect of forced flow are factored into developing the operational limitations.

SCE&G I Virgil (. Summer Nuclear Station

• P.O. Box 88 *Jenkinsville, South Carolina 29065*T 1803) 345.5209 *wwwscano.om

Document Control Desk C-04-2135 RC-05-0090 Page 2 of 2 Attachment I provides the TS pages marked up with the proposed changes. Attachment II provides the retyped TS pages.

The VCSNS Plant Safety Review Committee and the Nuclear Safety Review Committee have reviewed and approved the proposed change. SCE&G has notified the State of South Carolina in accordance with 10CFR50.91 (b).

There are no other TS changes in process that will affect or be affected by this change request.

There are no significant changes to any FSAR or FPER sections.

There are no commitments resulting from this change request.

If you have any questions or require additional information, please contact Mr. Ronald B. Clary at (803) 345-4757.

I certify under penalty of perjury that the foregoing is true and correct.

12/z Z/5 Executed on //4~effrey B. Archi AMM/JBA/mb

Enclosure:

I. Evaluation of the proposed change Attachments:

I. Proposed Technical Specification Change - Mark-up

11. Proposed Technical Specification Change - Retyped 1II. List of Regulatory Commitments IV. WCAP-16305-NP, Revision 0 c: N. 0. Lorick NRC Resident Inspector S. A. Byrne P. Ledbetter N. S. Carns T. P. O'Kelley T. G. Eppink (w/o Attachments) Winston & Strawn R. J. White RTS (C-04-2135)

W. D. Travers File (813.20)

R. E. Martin DMS (RC-05-0090)

Document Control Desk Enclosure I RC-05-0090 Page 1 of 5 LICENSE AMENDMENT REQUEST TECHNICAL SPECIFICATION 3/4.4.9 AND ASSOCIATED BASES

1.0 DESCRIPTION

South Carolina Electric & Gas Company (SCE&G) requests an amendment to revise the Virgil C. Summer Nuclear Station (VCSNS) Technical Specification (TS) Section 3/4.4.9, Figures 3.4-2 and 3.4-3, and the associated Bases Section 3/4.4.9. The proposed change will:

a) Update Figures 3.4-2 and 3.4-3 to incorporate the latest available reactor vessel analysis and updated calculated fluences. The figures are applicable up to 56 Effective Full Power Years (EFPY).

b) Revise Bases 3/4.4.9 to delete outdated information and provide revised information applicable to the updated heatup/cooldown curves. The outdated information is being deleted and the revised information is contained in the referenced Westinghouse Topical Report. It was determined that this level of technical data is not required in the VCSNS TS Bases.

2.0 PROPOSED CHANGE

Specifically the proposed changes would revise the following:

2.1 TS Figures 3.4-2 and 3.4-3 These figures provide the VCSNS Reactor Coolant System heatup and cooldown limitations. The proposed changes incorporate the latest available reactor vessel information and updated calculated fluences, and also extend the applicability of the figures up to 56 EFPY. These proposed curves are more limiting than those currently contained in the TS.

2.2 TS Bases 3/4.4.9 The Bases are being updated to delete information that is no longer applicable and provide the reference to the Westinghouse Topical Report that contains the updated information.

3.0 BACKGROUND

Pressure/Temperature (PIT) limits are developed to satisfy 10 CFR Part 50, Appendix A, Design Criteria 14 and 31. These design criteria require that the reactor coolant pressure boundary be designed, fabricated, erected, and tested in order to have an extremely low probability of abnormal leakage, and of rapid or gross failure. The criteria also require that the reactor coolant pressure boundary be designed with sufficient margin to assure that when

Document Control Desk Enclosure I RC-05-0090 Page 2 of 5 stressed, the boundary behaves in a non-brittle manner and the probability of rapidly propagating fracture is minimized.

The requirements of 10 CFR 50, Appendix G, Fracture Toughness Requirements, describe the requirements for developing P/T limits and the basis for the limitations. The proposed VCSNS P/T limit curves were generated using the "axial flaw" methodology of the 1998 edition of the ASME Code, Section Xl through the 2000 Addenda. Included in this methodology is the use of the Kic stress intensity factors which were formally documented under ASME Code Cases N-640 and N-641. The method to predict the reactor vessel material irradiation damage is provided in Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials". Regulatory Guide 1.99, Revision 2, was used for calculation of Adjusted Reference Temperature (ART) values at the 1/4T and 3/4T locations, where T is the thickness of the vessel at the beltline region measured from the clad/base metal interface. The proposed P/T curves were generated using the most limiting ART values and the NRC approved methodology documented in WCAP-14040-NP-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves". This methodology determines pressure-temperature limiting curves in accordance with the requirements of Appendix G, 10 CFR Part 50, as augmented by Appendix G, Section Xl of the ASME Code.

The proposed P/T curves were generated based on the latest available reactor vessel specimen capsule analysis results and updated calculated fluences. This reactor vessel information was previously provided to the NRC in WCAP-16298-NP, Revision 0, 'Analysis of Capsule Z from the South Carolina Electric & Gas Company V. C. Summer Reactor Vessel Radiation Surveillance Program" by letter from Jeffrey B. Archie, SCE&G, to the NRC, dated October 22, 2004.

4.0 TECHNICAL ANALYSIS

The proposed changes to the P/T curves reflect the results of the analyses performed on the reactor vessel surveillance capsules, most recently specimen Z, as part of the reactor vessel material irradiation surveillance specimen program. The analysis was performed and the calculations prepared using guidance contained within Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials", and Appendix G to 10 CFR 50.

Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels used for light-water-cooled reactor vessels.

The results of these analyses of specimen Z were provided to the NRC in a letter from Jeffrey B. Archie, SCE&G, to the NRC, dated October 22, 2004 as required by Appendix H to 10 CFR

50. The results were detailed in an attachment to that letter, WCAP-16298-NP, Revision 0, "Analysis of Capsule Z from the South Carolina Electric & Gas Company V. C. Summer Reactor Vessel Radiation Surveillance Program". Also provided at that time was WCAP-16306-NP, Revision 0, 'Evaluation of Pressurized Thermal Shock for V. C. Summer".

Document Control Desk Enclosure I RC-05-0090 Page 3 of 5 The new P/T curves for normal heatup and cooldown of the primary reactor coolant system have been developed using the methods as discussed in the attached WCAP-16305-NP, Revision 0, "V. C. Summer Heatup and Cooldown Limit Curves for Normal Operation". These curves provide the limits for operation of the Reactor Coolant System during heatup, cooldown, criticality, and hydrostatic testing. The curves are applicable to 56 EFPY and are being incorporated within the TS to preclude the necessity for a later TS change. These proposed curves are more limiting than those currently contained in the TS. Instrument uncertainties are not included in the TS figures; however, they are incorporated into the curves located in the plant operating procedures. The operational limit curves include the following effects:

1. Instrument uncertainties associated with the pressure and temperature measurements.

2. Pressure increases due to the elevation head differences between the pressure measurement location and the reactor vessel beltline region.

3. Pressure increases between the pressure measurement location and the reactor vessel beltline region due to Reactor Coolant System flow (i.e., form, friction, and velocity head effects resulting from Reactor Coolant Pump operation).

5.0 REGULATORY SAFETY ANALYSIS No Significant Hazards Consideration South Carolina Electric & Gas Company (SCE&G) has evaluated the proposed changes

-r to the VCSNS TS described above against 'the Significant Hazards' Criteria - of 10CFR50.92 and has determined that the changes do not involve any significant hazard.

The following is provided in support of this conclusion.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed changes revise the P/T limit curves to provide figures that reflect the results of the analysis performed on reactor vessel surveillance specimen Z.

This analysis was performed using NRC approved methodology as documented in WCAP 14040-NP-A, Revision 4, utilizing the 1998 ASME Code, Section Xl through the 2000 Addenda, Appendix G requirements. These curves provide the limits for operation of the Reactor Coolant System during heatup, cooldown, criticality, and hydrostatic testing. These curves are provided without instrument uncertainties included, however, the uncertainties are included in the curves provided in the plant operating procedures. The limits protect the reactor vessel from brittle fracture by separating the region of acceptable operation from the region where brittle fracture is postulated to occur. Failure of the reactor vessel is not a VCSNS design basis accident, and, in general, reactor vessel failure has a low probability of occurrence and is not considered in the safety analysis.

- - t Document Control Desk Enclosure I RC-05-0090 Page 4 of 5 Therefore, the change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed changes revises the P/T limits curves, Section 3/4.4.9, to incorporate the results of the analysis performed on reactor vessel specimen Z.

There are no physical plant design changes or significant changes in any operating procedures. This change adjusts the heatup and cooldown curves to reflect the shift in nil-ductility reference temperature of the reactor vessel as a result of neutron embrittlement. Therefore, the change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in margin of safety?

Response: No.

The proposed changes revise the P/T limits curves, Section 3/4.4.9, to incorporate the results of the analysis performed on reactor vessel specimen Z.

The new P/T curves ensure that the 10 CFR 50 Appendix G, requirements are not exceeded during normal operation including Reactor Coolant System transients during heatup, cooldown, criticality, and hydrostatic testing. The new P/T curves were prepared, using approved industry methodology, for a projected reactor vessel neutron exposure of 56 EFPY. The proposed P/T limit curves reflect a shift of the limits in a conservative direction from the current requirements. Therefore, the change does not involve a significant reduction in a margin of safety.

Pursuant to 10 CFR 50.91, the preceding analyses provide a determination that the proposed TS change poses no significant hazard as delineated by 10 CFR 50.92.

6.0 ENVIRONMENTAL CONSIDERATION

This proposed TS change has been evaluated against criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21.

It has been determined that the proposed change meets the criteria for categorical exclusion as provided for under 10 CFR 51.22(c)(9). The following is a discussion of how the proposed TS change meets the criteria for categorical exclusion.

10 CFR 51.22(c)(9): Although the proposed change involves change to requirements with respect to inspection or Surveillance Requirements,

Document Control Desk Enclosure I RC-05-0090 Page 5 of 5 (i) the proposed change involves No Significance Hazards Consideration (refer to the No Significance Hazards Consideration Determination section of this TS Change Request);

(ii) there are no significant changes in the types or significant increase in the amounts of any effluents that may be released offsite since the proposed change does not affect the generation of any radioactive effluents nor does it affect any of the permitted release paths; and (iii) there is no significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Based on the aforementioned and pursuant to 10 CFR 51.22 (b), no environmental assessment or environmental impact statement need be prepared in connection with issuance of an amendment to the TS incorporating the proposed change.

Document Control Desk  !

Attachment I RC-05-0090 Page 1 of 7 ATTACHMENT I PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

Attachment to License Amendment No. XXX To Facility Operatina License No. NPF-12 Docket No. 50-395 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 B 3/4 4-6 B 3/4 4-6 B 3/4 4-14 B 3/4 4-14 SCE&G -- EXPLANATION OF CHANGES Pag -e Afrected Description of Change Reason'for Change Section 3/4 4-31 Figure 3.4-2 Replace heatup curve Revised P/T limits based on analysis of reactor vessel surveillance capsule.

3/4 4-32 Figure 3.4-3 Replace cooldown curve Revised P/T limits based on analysis of reactor vessel surveillance capsule.

B3/4 4-6 lB3/4.4.9 Added updated design basis Revised ASME Code information. reference.

B3/4 4-14 B3/4.4.9 Deleted unnecessary information Correct reference for P/T contained with in Bases section limit curves.

and updated applicable Bases information.

Document Control Desk Attachment I RC-05-0090 Page 2 of 7

.1 R4 Jj % A-k-,

T: I D%,.,r"L. St N

LA --

REATR COOLANT SYTIEM MATERIAL PROPERTY SASIS LlMITING MATERIAL- LOWER SHELL PLATES 09923-1, -2 LiMmNG .ART VALUES AT 32 EFPY:

2500 114T, 107 F 3/4T, 94 OF r LEAKEST IUMIT: I 2250

- HEATUP RATE: .

2000 i .UP TO 50 F/Hr. -

le 60

. -4 Ln 1750 - JUNACCEPTABI.

IOPERA~ioN I I.. EATUP RATE

.UP TO tO0 FA-r.

111 18 It i 1- _I!_

§ ti In!

l 1

CL) I I-F I 1 3.4 1500 -

-ACCEPTABLE

.OPERATION -

1250 LcrITr.

7-4, _

P-4 UwMIT  ! Itt "IC$

(Li ta 1000 I: ,F`OR50FA*. - -H H I

+ - - - - - I :'

I a

CrJT.U"W 750 . ran I OD F"~.'

CO I U

500 I sIoft Upl CRCAUYOLTY JMJT BASED ONJ INtSERVICE HYDROTATIC TEST 250 -

TDUIPERATURIE 4164'F FOR THE SERVICE PERIOD UiP TO W-0 EFTY I

0 -

4 I I I - II . I I 0 50 100 150 200 250 300 350 -400 450: 500 Moderator Temperatiure ( F)

Figuro 3.4-2 V. C. Summer Unit 1 Reactor Coolant System Heatup Limitations (Heatup ratos up to 50 and 100 0F/hr) Applicablo for the First 32 EFPY (Without Margins for Instrumentation Errors) 314 4-31 Amendment No. S-I1l 3 -,

SUMMER - UNIT 1

Document Control Desk Attachment I RC-05-0090 Page 3 of 7 g.E A cv C (R oL A V sI5TEr Y-MATERIAL PROPERTY BASIS: Limiting Material: Intermediate Shell Plate A9154-1 Limiting ARTValues @ 56 EFPY: 1/4T: 1530 F, 3/4T: 138°F 2500 2250 2000 1750

= 1500 CD -

IA e.

nw EDi 1 250 a)

.L B U 1000 750 500 250 0

.I 0 50 100 150 200 250- 300 350 400 450 500- 550 Moderator Temperature (Beg. F)

Figure 3,L4-V. C. Summer Reactor Coolant System Heatup Limitations (Heatup Rates of 50 and I 00°Fihr) Applicable for 56 EFPY (Without Margins for Instrumentation Errors) Using 1998 Appendix G Methodology S.!tinElRi - lJ ir1 I 3S )Lt t*-:3 I A via.

t .;

Document Control Desk Attachment I RC-05-0090 Page 4 of 7 tt9\v \ W;Ak, V IC I . ,3c{q 3 REACTOR CQOLANT SYSTEM 1ATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATES C9923-1 -2 LIMITING ART VALUES AT 32 EFPY: 1.4T. 107°F 3J4T, 94 °F 2500 2250 2000 1750 cn S-4 1500 U.

C) 1250 1000 U 750 500 250 0 I -. .: -.- . .- , - .. I 0 50 100 150 200!250 300 350 4 00 450 500 Moderator Temperature (oF)

Figure 3A-3 V. C. Summer Unit 1 Reactor Coolant Systcm Cooldown Limritations (Cooldown Rates of 0, 25. 50 and 100 6Flhr) Applicable for the First 32 EFPY (Without Margins -or Instrumentation Errors)

SUMMER - UNIT 1 :3f4 4-32 Amendment No. 53-S-113SJ3-

Document Control Desk Attachment I RC-05-0090 Page 5 of 7 ILt o4 Co L A.5 YSTE t

MATERIAL PROPERTY BASIS: lUmiting Material: Intermediate Shell Plate A9154-1 Limiting ART Vahies @ 56 EFPY: 114T: t53°F. 314T: 136°F 2500 2250

- 2000 I 1750 CD EL 1500 P-E C 1250

'a

n. ,

10 750 500 250~

0 0 so 100 150 200'- 250 300 350 400 450 - 500 550 Moderator Tomperature (la.% F)

Figure S.'-4;V. C. Summer Reactor 'Coolant System Cooldown LImitations (Cooldown Rates up to I00OF/hr) AppIlcable for 56 EFPY (Without Margins for Instrumentation Errors) Using 1998 Appendix G Methodology 5UtAY?- -VtfJ %'

SLJ'WX.1JtJT I 314 3 14 '-V3-3-2

Document Control Desk Attachment I RC-05-0090 Page 6 of 7 REAC:TQR OOOLIANT SYSTEN BASES -- . .- -!

3/4.4.8 SPECIFIC ACTMVITY . --

..The lmiltations on tho speciic activity of the prlmary coolant ensure that the rosuilting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dOses at the site boundary will not exceed an appropriately small fraction of Part iOo limits following a steam generator tube rupture accident hi conjunction with an assumed steady state ptimary-to-seoondaqy steam generator leakage rate of 1.0 GPM The vralues for the Imits on specific activity represent limits based upon a parametric evaluation by the NRC of typical stto locations. Thoso valuos are conservative In that specific site parameters of the Virgil C.

Summer site, such as sote boundary location and moteorological conditions, were not '

considered In this evaluation.

Th ACTION statement permitting POWER OPERATION to continuo for limited time periods with the primary coolant's specific activity greater than 1.0 microcurloslgram DOSE EQUIVALENT 1-131. but within the allowable limit show on Figure 3.4-1, accommodates possible lodino spiking phenomenon which may cwccur folowking changes In THERMAL POWER.

Roducing T,, to less than 5000 F prevents the release of activity should a steam

,ne~rator utbe rupture since the saturation pressure of the prImary coolant ls below the lift pressureollhe ainospheric steam reieleahves. .The surveNance requirements provide adequate assurance that excesse specific activity.levs In the primary coolant wig be detected Insufficent time to take corrective action. Information obtained on Iodine spiking wal be used to assess the parameters associated wvth splking phenomena. A reduction In frequency of Lsotopic analysos following power cihanges may be penrisbl If justified by t-e data obtained.

314.4. PReSSURE/TEMPPRATURE-LIMIT The temperature and pressure changes during heatup and cool down are imited by curves developed using the methodology from Westin house Topical Report, WCAP-14040-NP-A, updated to Include lhe requirements of the ZpSME Boiler and Pressure Vessel Code, Seclion Xl ppndbc Galong with ASME ode Cas&Ne e N

-C kfY 4; 2X 1 , 's - 6t6

1) -The reactor coolant temperature and pressure and system heatup and cooldowm rates (with the exception of the pressurizer) shall be limited In accordance wilth Figures 3.4-2 and 3A-3.

a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldownr rales between those presented may be obtained by Interpolation.

SUMMER- UNIT 1 B 3/4 4-6 Amendment No. 53,414AA,

Document Control Desk Attachment I RC-05-0090 Page 7 of 7 REACTOR COOLANT SYSTEM BASES PRESSUREtTEMPERATURE LIITS (Continued)

Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been caIculated as described In Westinghouse Topical Report, WCAAP-64e ,, W. C. Summor-1t-1 Heatup and Cooldown Curves for Normal OpQIFF.3° 5;020- fJ P ROJ-_b Transition temperature shifts occurring In the pressure vessel matedials c.

ra on exposure have been obtained disrctly from the reactor pressurevssel.

vurve CO program. Carpy test specimons from Capsule W inclcate that lii reglon shell plateoodo no. C9923-1. -2 are the limiting beitline mate for all heatup and oldowrn curves to be generated. Those materials exhlbit Iilng ART.

va07es of 1070t 1f4T and 940 F at 3f4T -at a calculated Inner surface f Mnce of 3.84 x 10'` nli 2 at 32 Y.

Allowable comb lions of tomperature and pressure for Pcifel temnperature.

change rates are below an The sight of the limit linos s on th heatup and cooldown curves. The reac o t not be made critical u pressure-temperature'.

combinations are to the right of criticalty limit line s In Figure 3.42. This is hI addition to other criteria which r e ins met before th ctor Is made critical9 as discussed in the following paragraph:s The leak test limit aurve 5how in 42 represents minimum temperature -

requirements at the leak lost pressure y applicab~a codes; The loak tse limit curve wau determried by methods of the nda' eview Plan'. Chapter 5.3.2 and Appendix t; of the ASIME Code, Sectlo

-The reactor must not be ma critical unlil pres temperature combinations are to the right of the critically lrIrbne shown tn Figure 3. The criticality rimit curve specifies pressure - temperatur lints for core operation to de additional margin during actual power producti as specifed In Appenx G to 1bF 50. The pressure

- temperature limits for / peration (excopt for low power physk tests) are that the reactor vessel must be altemperature equal to or herthan the imum temperature required f e inservice hydrostatic test. and at least 4 Igher than the minimum perrnissibl mperature in the corrosponding pressure - temp}ture cure for heatup and wn calculated as described in this technical basis. vertical lire drawn trom e oints on the pressure - temperature curve. Intersect a curve 40°F higher tht pres sure - lemporatr limit curve, constitutos the Omit f e operation fo e reactor vessel..:

Fl res 3.4-2 and 3.4-3 define lirits for Insuring prevention of nornductile fa re.

The Instrument uncertainties, effects .offorced flow from tho reactor coolant \

Pu s, and the elevation effect of the pressure sensors are incorporaled Into the as located In the plant operating procedures.

E -I SUMM5ER -UNIT 1 B 3,4 4-14 Amrendmont No. 5 ,

Document Control Desk 1 RC-05-0090 Page 1 of 5 ATTACHMENT If PROPOSED TECHNICAL SPECIFICATION CHANGES (RETYPED)

REACTOR COOLANT SYSTEM MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE A9154-1 LIMITING ART VALUES @ 56 EFPY: 1/4T, 153°F 3/4T, 138°F

Opetnlm Vern 5.2 Run-17345 Oper1imrrds Veason 5.2  : . ..  :- .... . .. ... w ...

2500 .

.000  :---Uacctabl - 1 -

- Aceptablc

- Operation

. ..  ! .- 1 .

1750

'ritical Limit So

Dog. ._ r Z!.,Umlt a.. Hoatup Rate OD c' . . 100 Dog. FIr

.1000 -

750 I

500' / . .

, - 0 , , ............... -

BoltupCriticalityUrnit bused gnt

..250 Tm. - insorvIco h~ydnrstatic I tempierature (210 F) fi servce, period up to 5' I0 . 50. 100 150 200 250 300 350 400 ::450 500 -550 Moderator Temperature (0 F)

FIGURE 3.4-2 V. C. Summer Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 50 and 100°F/hr) Applicable for 56 EFPY (Without Margins for Instrumentation Errors) Using 1998 Appendix G Methodology SUMMER - UNIT 1 3/4 4-31 Amendment No. 53,113, 433, 143,

REACTOR COOLANT SYSTEM MATERIAL PROPERTY BASIS LIMITING MATERIAL: INTERMEDIATE SHELL PLATE A9154-1 LIMITING ART VALUES @ 56 EFPY: 1/4T, 1530 F 3/4T, 1380 F 100 vi VC93krn*2 Mi.n:17345 Opevtmnadd Vers~on &

-2251 200(

Operation ___ =

• - 150

1250 150 7

GAMotdawn I It~

i~c4ra .a.

X .. 0

• 250 .

1-2511 MP.

.70 U ...QL u l

-IA) -U ~U zu U 300.. 350.. 400 .450 500 550 Moderator Temperature (0 F)

FIGURE 3.4-3 V. C. Summer Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1001F/hr) Applicable for 56 EFPY (Without Margins for Instrumentation Errors) Using 1998 Appendix G Methodology 3/4 4-32 Amendment No. 5 3 ,143, SUMMER - UNIT 1 133,1443,

REACTOR COOLANT SYSTEM BASES 314.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Virgil C.

Summer site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/gram DOSE EQUIVALENT 1-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Reducing Tvg to less than 500OF prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

314.4.9 PRESSURE/TEMPERATURE LIMITS The temperature and pressure changes during heatup and cool down are limited by curves developed using the methodology from Westinghouse Topical Report, WCAP-14040-NP-A, updated to include the requirements of the 1998 ASME Boiler and Pressure Vessel Code, Section Xl, through the 2000 Addenda, Appendix G, along with ASME Code Case N-641.

1) The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3.

a) Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.

SUMMER - UNIT I B 3/4 4-6 Amendment No. 53, 64, 143, 154,

REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS (Continued)

Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated as described in Westinghouse Topical Report, WCAP-16305-NP, Revision 0, UV. C. Summer Heatup and Cooldown Curves for Normal Operation".

SUMMER - UNIT 1 B 3/4 4-14 Amendment No. 53,143,

Document Control Desk Attachment III RC-05-0090 Page 1 of 1 ATTACHMENT III LIST OF REGULATORY COMMITMENTS There are no regulatory commitments created due to this License Amendment Request.

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