ML20217E513

Application for Amend to License DPR-50,revising TS Section 3.1.2 to Incorporate New Pressure Limits for Reactor Vessel IAW 10CFR50,App G for Period of Applicability Through 17.7 Efpy.Request Also Submitted in Response to NRC

ML20217E513
Person / Time
Site:	Three Mile Island
Issue date:	03/23/1998
From:	Langenbach J GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20217E519	List: ML20217E513 ML20217E531
References
1920-98-20127, NUDOCS 9803310063
Download: ML20217E513 (8)

Text

l

. \

s. j l

GPU Nuclear. Inc.

{

g Route 441 South NUCLEAN 1"j$yo37.o4,o Tel 717 944-7621 1920-98-20127 March 23,1998 U. S. Nuclear Regulatory Commission Attention: Document Control Desk

' Washington, DC 20555

Subject:

Three Mile Island Nuclear Station, Unit 1 (TMI-1),

Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request (TSCR) No.270 RPV Pressure and Temperature (PT) Limits in accordance with 10 CFR 50.4 (b) (1), enclosed is Technical Specification Change Request No. 270. Also enclosed is the Certificate of Service for this request certifying

{

service to the chief executives of the township and county in which the facility is located, as well as the designated official of the Commonwealth of Pennsylvania, Bureau of Radiation Protection.

The purpose of this TSCR is to request that the TMI-l Technical Specifications Section 3.1.2, regarding reactor vessel pressurization heatup, cooldown, and inservice leak and hydrostatic test limitations, be revised to incorporate new pressure limits for the reactor vessel in accordance with 10 CFR50, Appendix G for a period of applicability through 17.7 effective full power years (EFPY).

This TSCR is also submitted in response to the NRC letter dated October 8,1997 which requested GPU Nuclear to reassess the reactor vessel materials reference temperature (RTwor and RTrrs) values. GPU Nuclear has reassessed, and thus is submitting this change for the TMI-l operating limits on the basis of NRC Regulatory Guide No.1.99 Rev.2, paragraph C. I.1, by selecting the appropriate material chemistry factor (CF) from Table No. I of the Regulatory Guide and commensurate higher margin term for the determination of the RTuo; and RTrrs values.

Pursuant to 10 CFR 50.91 (a)(1), enclosed is our analysis, applying the standards in 10 CFR 50.92 to make a determination of no significant hazards considerations. As stated ,

above, pursuant to 10 CFR 50.91(a), we have provided a copy of thi. letter, the proposed j  ;

9003310063 980323 PDR ADOCK 05000289 P pog- '[ 2

. . . . . ~ . . - .. .. . . .. . . .. .. .

- - - .. ]

-, 1920 98-20127 Page 2 l e

change in Technical Specifications, and our analyses of no significant hazards

- considerations to the designated representative of the Commonwealth of Pennsylvania.

Sincerely, D

. J. . Langenbach Vice President and Director, TMI

.YN ,

Enclosures:

(1) Technical Specification Change Request No. 270 (2) Certificate of Service for Technical Specification Change Request No. 270 cc: '

Administrator Region I

. TMI Senior Resident Inspector TMI-l Project Manager b

d

METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT !

Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 270 COMMONWEALTH OF PENNSYLVANIA )

) SS:

COUNTY OF DAUPHIN )

This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1. As a pan of this request, proposed replacement pages for Appendix A are also included.

All statements contained in this submittal have been reviewed, and all such statements made and matters set forth therein are true and correct to the best of my knowledge.

GPU NUCLEAR INC.

BY:

Vice President and Director, TM1 Sworn ang Subscribed to before me this 23^ day of alm c/v ,1998.

P

'- Notary Public une L E M m M%"T. ;;%

= = . - . , -

1920-98.-20127

~

..- Attachme*nt -

Page 1 of 5 :

L TECHN' ICAL SPECIFICATION CHANGE REOUEST (TSCR) No. 270 GPU Nuclear requests that the following changed replacement pages be inserted into the existing Technical Specification:

}'

Revised pagesc 3-3,3-4,3-5,3-Sa (Figure 3.1-1) and 3-5b (Figure 3.1-2).

These pages are attached to this change request.

II. REASON FOR CHANGE This change is submitted in response to NRC letter, dated October 8,1997 to James W. Langenbach' (GPU Nuclear),

Subject:

Three Mile Island Nuclear Generating Station, Unit 1 (TMI-1) Request for Additional Information Concerning Pressure-Temperature Limit Curves and Low Temperature Overpressurization Limits (TAC No. M99319). This letter is a closeout of NRC Generic Letter (GL 92-01) requesting GPU Nuclear to reassess the reactor vessel materials reference temperature (RT mand RTns) values. The reason for this request is that the basis of the present limit curves did not include the use of the chemistry factor ratio procedure per the provisions of NRC Regulatory Guide No.1.99 Rev. 2, paragraph C.2.1 and may thus have yielded non-conservative operating limits.

GPU Nuclear has reassessed, and thus submitting this change for, the TMI-l operating limits on the basis of NRC Regulatory Guide No.1.99 Rev. 2, paragraph C.I.1, by selecting the appropriate material chemistry factor from Table No. I of the Regulatory Guide.

This change is submitted in compliance with T.S. Section Nos. 3.1.2.4 and 3.1.2.5 which require an update of the operating P/f limits (Figures 3.1-1 and 3.1-2 in accordance with 10CFR 50, Appendix G, Section V.B for a change in service period (from 15.2 EFPY to 17.7 EFPY).

TMI-l expects to reach 17.7 EFPY near the beginning of 2002, after the completion ofits fuel Cycle 13 operation, and shortly after beginning operation from its Cycle 14 refueling shutdown.

111. SAFETY EVALUATION JUSTIFYING CHANGE The proposed Technical Specification revises the basis, and the provisions, including Figures 3.1-1 and 3.1-2, regarding the reactor vessel pressurization heatup, cooldown, and inservice leak and hydrostatic testing limitations, which are contained in Section 3.1.2. These pressure limits are established to provide protection for the reactor vessel from potential non-ductile failure during .

normal plant operating modes and are updated periodically to reflect the state of embrittlement at the end the period of applicability.

A. - Embrittlement Trend of Reactor Vessel Materialj

. The embrittlement trend of the reactor vessel materials has been established in accordance with Regulatory Guide 1.99 Revision-2, paragraph c.l.1, using Table 1. The material chemistry factor (CF) is obtained from Table 1, based on its Copper and Nickel chemical content. The mean initial RTmyrand its standard error have been previously established by Tech. Spec. Amendment No.176, having values of-7 and 20.6 respectively. Thus, based on the fluence trend at the limiting materials location, the embrittlement trend has been

calculated. The RTmand RTns results of this calculation, for the limiting TMI-l reactor

- vessel materials, at 17.7 EFPY (for RTmyr) and end-oflicense period (for RTns values) is presented in attached Table 1.

.(

4

1920-9&-20127 -

c.: t

A'tachme'nt~

Page 2 of 5 -

B. Operatina P/f Limit Curves The requirement for establishing operating pressure limits'is established by 10 CFR 50, App.

G. The methodology of establishing these pressure limits is specified by ASME Code,Section XI,' App. G,.1989 Edition (without Addenda) and have been determined in accordance with this Code by application of the B&W Reactor Vessel Owners Group (B&WOG) report BAW-10046A, Rev. 2. The results of the application of this methodology in the generation of-the proposed pressure limits is contained in FTl calculation No. 32-5001065-01,"TM1-1 P/f Limits," March 1998. The methodology requires as input the state of the material toughness, which is defined and indexed by the reference temperature (RTer). The methods, which the NRC finds acceptable, to determine the RTerare provided in Regulatory Guide No.1.99 Rev. 2. He method described by paragmph C.l.1 of the Regulatory Guide has been applied to the limiting material contained in the TMI-l reactor vessel (Linde80 weldments SA-1526 and WF-25).

The RTer values which have been utilized in the generation of the proposed operating P/f limits curves are 214*F and 160*F, at the 1/4T and 3/4T locations respectively. These values have been calculated, by GPU Nuclear calculation No. C-1101-221-ES20-013, Rev. O, to occur at the completion of 17.7 effective full power years (EFPY) of operation, and are expected to occur after the completion of fuel cycle No.13.

In addition to the RTer values, the basis for the pressure limits include the assumptions regarding the rate of heatup and cooldown as well as plant restrictions regarding the number and combination of operating reactor coolant pumps in operation. These additional bases are contained in the basis section of the Technical Specification and are controlled by and monitored by the plant operating procedures. The heatup and cooldown rates have been established on the basis of the system component performance limitations as a function of the Reactor Coolant System (RCS) temperature or acceptable rates which can be controlled by the plant operation. Review of past plant performance provides confidence that these additional assumptions are valid inputs in the determination of the allowable pressure limits.

He resultant operating pressure limits for heatup and cooldown are less restrictive in the lower temperature range, below 275'F, than the present limits. This is due to the reduction in the conservatism, to more accurately reflect the plant performance, contained in the cooldown

- rates at the lower RCS temperatures. In the higher temperature range, the proposed pressure limits become more restrictive, by approximately 30 F for the heatup limits, and only a few degrees for the cooldown limits, than the present limits. The reason for the apparent small shift in the cooldown limits, in the higher temperature range is that for simplicity the present curve was generated by the lower bound of the heatup and cooldown limits to form a combined curve. ,

. In the higher temperature range, the limit curves for both heatup and cooldown are controlled by the' 1/4T RTer, in the cooldown case it is also controlled by the steady state conditions, i.e., no thermal stress. In the heatup case it is also controlled by the 1/4T steady state conditions; however, it is shifted as a result of the 1/4T temperature lag as compared to the RCS fluid temperature.- In the heatup case, the fracture mechanics method provided by BAW-10046A, Rev. 2 contains conservatism by not providing credit for the induced thermal compressive stress at the 1/4T location.

C. . Low Twiiwature Overoressure Protection Low temperature overpressure (LTOP) events are potential occurrences as a result of various 6 . abnormal plant maneuvers during the normal heatup and cooldown processes or a result of a

1920-98-20127 2 Attachme'nt'-

.Page 3 of 5 e component failure and which could result in an uncontrolled pressurization of the RCS above L the operating P/T limits.

In the BhW NSSS design and operation, a steam bubble is maintained in the pressurizer at all times. Thus, most potential abnormalities would result in either a very slow pressure rise or a very small pressure rise. In the low temperature range (wherein the RTer plus 50 is less than 275'F), the most severe pressurization event would be due to a failure of the normal RCS makeup co al valve to the full open position. While this is an unlikely event, it had been o analyzed for the present P/T limits. Since, in the low temperature range, the proposed pressure limits are less restrictive than the present limits, a reassessment of these potential events is not required, and the power operated relief valve pressure setpoint (485 psig) and the resettirg temperature (275*F), per T.S. 3.1.12, do not need to be revised.-

L in the higher temperature range, the high capacity mjection potential such as inadvertent HPI initiation or high capacity RCS makeup valve opening are prevented, by Technical Specification No. 3.1.12, which requires that the respective injection valves be closed and their motive power disabled.

Thus, there is no required change to the LTOP provisions, and the margin of safety is maintained.

D. Inservice Leak and Hydrostatic Test Limits The propose Inservice Leak and Hydrostatic Test (ISLH) pressure limits have been established in accordance with the same requirements, methodology and input assumptions as the operating pressure limits.

The ISLH limits are conservative, since they utilize the same input assumptions as for the operating limits. He conservatism is contained in the assumption of the heatup and cooldown rates, whereas, the performance ofISLH test would maintain the RCS at a relatively constant temperature. Hus, the thermal stress component would be absent.

The proposed ISLH pressure limits have been established on the basis of conservative input assumptions, and in accordance with an approved methodology. Therefore, the required margin of safety is maintained.

E. PTS Rule (RTmL 10 CFR 50.61 The PTS Rule,10 CFR 50.61, requires an update of the RTrrs value whenever a TSCR is submitted for the operating pressure / temperature curves. GPU Nuclear calculation (Calc. No.

C-1101-221-E520-013, Rev. 0) includes an assessment of the RTrrs values for the limiting materials at the end-of-license period (year 2014, with less than 28.8 EFPY).

He resultant values are:

Material Orientation Fluence Estimate at RTers 10 CFR 50.61 less than ' Value Screening Value Weld SA-1526 Longitudinal 7.06E+18 262 F 270"F Wcld WF-25 Circumferential 7.83E+18 268 F 300*F

. Since these values are below the 10 CFR 50.61 screening criteria, no further action is required.

L,- . - " - ---,

1920-91t-20127 i Attachme'nt -

Page 4 of 5 : s 11 IV. NO SIGNIFICANT HAZARDS CONSIDERATIONS

. GPU Nuclear has determined that the requested Technical Specification Change poses no significant hazard as dermed by NRC in 10 CFR 50.92. ,

l. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated. The design basis event related to this change is nonductile failure

- of the reactor coolant pressure boundary. 'Ihe updated pressure / temperature limits have been established in accordance with the requirements of 10 CFR 50, Appendix G.

Revision of these curves for an applicability period of 17.7 EFPY is based on maintaining the required design margin. Operation of the facility in accordance with the proposed (

amendment provides assurance of protection against nonductile failure of the reactor coolant pressure boundary for operation through 17.7 EFPY. Therefore, operation in accordance with the proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind _of accident from any previously evaluated. The design basis event related to the change is nonductile failure of the reactor coolant-boundary. The proposed amendment provides assurance of protection against nonductile failure of the reactor coolant boundary for operation through 17.7 EFPY and is unrelated to the possibility of creating a new or different kind of accident.

3. Operation of the facility in accordance with the proposed amendment would not involve any reduction in a margin of safety since the design methodology has maintained the existing margins.

V. IMPLEMENTATION TMI-l has already implemented administratively controlled pressure limits which maintain the more restrictive limits below an RCS temperature of 275 F, and above 275'F have conservatively shifted the pressure limits to the right by 30 F. The present pressure limits, including the administrative pressure limits assure adequately conservative limits for the protection against nonductile failure. However, in order to close out the RAI contained in NRC letter (TAC No. M99319), GPU Nuclear requests expeditious issuance of this amendment.

• - Implementation of this amendment into operating procedures and operator training will be achieved within 60 day from its issuance date.

b A

1920-98 20127 Aitachme'nt Page 5 of 5 3

TABLE 1 s

TMI-1 ASSESSMENT OF RTNOT g 17,7 EFPY and RTPTs @ End-of-License R.G.1.99 R-2, Position

1. 1 METHOD 1,2 RTNDT(u) -7

%Cu 0.34

%Ni 0.68 CF 220.6 ou 20.6 og 28 Margin 69.5 Fluence [ x 1E19]

WF-251/4T @ 17.7 EFPY 0.311 WF-25 3/4T @ 17.7 EFPY 0.113 WF-25 Cire. Weld @ EOL 0.783 SA-1526 Long. Weld @ EOL 0.706 Fluence Factor WF-251/4T @ 17.7 EFPY 0.6794 WF-25 3/4T @ 17.7 EFPY 0.4416 WF-25 Cire. Weld @ EOL 0.9314 SA-1526 Long. Weld @ EOL 0.9025 ARTuor WF-251/4T @ 17.7 EFPY 150 WF-25 3/4T @ 17.7 EFPY 97 WF-25 Circ. Weld @ EOL 205 SA-1526 Long. Weld @ EOL 199 ARTuoy 917.7 EFPY and RTers O EOL WF-251/4T @ 17.7 EFPY 212 WF-25 3/4T @ 17.7 EFPY 160 WF-25 Cire. Weld @ EOL 268 SA-1526 Long. Weld @ EOL 262

' The fluence / fluence factor differ from that reported in BAW-2222, but they are consistent with BAW 2108, Revision 1.

2 RTNDT(u) and ouare taken from NRC safety evaluation dated August 16,1993 and are consistent with those reported in BAW-2222. Weld and surveillance capsule chemistries per Framatome Technologies, Inc. (FTI) evaluation (July 10, 1997 letter from Matthew J. Devan to Mr. Barry Elliot, NRC). CF is from RG 1.99, Revision 2 Table.