ML20084Q351

Forwards marked-up Replacement Pages 3.1-7,3.1-11 & 3.1-14 to Tech Spec Change Request 119 Re Final Trip Settings for Mod of Scram Discharge Vol as-built Sys Testing,Per 830713 Application to Amend License DPR-16

ML20084Q351
Person / Time
Site:	Oyster Creek
Issue date:	05/15/1984
From:	Fiedler P GENERAL PUBLIC UTILITIES CORP.
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8405210223
Download: ML20084Q351 (5)

Text

T GPU Nuclear Corporation Nuclear  ::= sea 88 Forked River.New Jersey 08731-0388 609 971-4000 Writer's Direct Dial Number:

May 15, 1984 Director, Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Oyster Creek Nuclear Generating Station Docket No. 50-219 Technical Specification Change Request No. 119 Pursuant to your correspondence of October 1, 1980 concerning long-term scram system reliability, the Oyster Creek Nuclear Generating Station scram discharge volume has been modified. Our letters of July 13, 1983 and December 21, 1983 provided Technical Specification Change Request (TSCR) No. 119 in support of this modification and our December 21, 1983 letter indicated that upon completion of as-built system testing, the final trip settings of Table 3.1.1 would be submitted.

Our previous letters were providea to initiate, in accordance with the Confirmatory Order dated June 24, 1983, NRC review of the proposed TbCR in order to ensure its approval prior to plant start-up from the current outage.

At the time of our latest letter, the scram discharge volume modification had not progressed to the point wnere as-built scram discharge instrument volumes corresponding to the instrument setpoint levels could be measured. As a result, scram discharge instrument volume design estimates were provided in order to facilitate NRC review.

The final setpoint values provided on attached Tecnnical Specification pages 3.1-7, 3.1-11 and 3.1-14 ensure that even considering the maximum design scram valve leakage, sufficient scram discharge volume will be available to permit 137 control rods to scram, if required. Since there is sufficient free volume remaining in the scram header piping to accommodate a scram with the setpoints established, the Significant Safety Hazards Analysis, provided by our submittal of December 21, 1983, is not altered. In addition, line No. 5-on page 3.1-7 of Table 3.1.1 which was forwarded by our July 13, 1983 letter accounted for both new scram discharge instrument volumes (SDIV). As such, the number in the column headed ' Min. No. of Instrument Channels Per 8405210223 840515 PDR ADOCK 05000219 ie P PDR I GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation

, to Director U.S. Nuclear Regulatory Commission ,

Page 2 May 15, 1984 Operable Trip System' was four (4). Line No. 5 on Page 3.1-7 of Table 3.1.1 was modified in our submittal dated December 21, 1983 when each SDIV was accounted for separately. Consequently, the number in the above-referenced column should have been changed to two (2) for each SDIV, but was not.

Therefore, please replace pages 3.1-7, 3.1-11 and 3.1-14 of TSCR No. 119 with the attached Technical Specification pages. Should you have any questions, please contact Mr. Michael Laggart at (201) 299-2341.

Very truly yours, I

Ta Pet IU Fiedler r Vice President and Director Oyster Creek i

PBF/ dam Enclosures cc: Administrator  :

Region I ,

U.S. Nuclear Regulatory Commission '

631 Park Avenue King of Prussia, PA 19406 NRC Resident Inspector Oyster Creek Nuclear Generating Station Forked River, NJ 08731 l

1 9

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3.1-4 isolation, initiate automatic depressurization in conjunction with low-low-low-reactor water level, initiate the standby gas treatment system and isolate the reactor building. The scram function shuts the core down during the loss-of-coolant accidents. A steam leak of about 15 gpm and a liquid leak of about 35 gpm from the primary system will cause drywell pressure to reach the scram point; and, therefore the scram provides protection for breaks greater than the above.

High drywell pressure provides a second means of initiating the core spray to mitigate the consequences of a loss-of-coolant accident. Its set point of 2 psig initiates the core spray in time to provide adequate core cooling. The break-size coverage of high drywell pressure was discussed above. Low-low water level and high drywell pressure in addition to initiating core spray also causes isolation valve closure. These settings are adequate to cause isolation to minimize the offsite dose within required limits.

It is permissible to make the drywell pressure instrument channels inoperable during performance of the integrated primary containment leakage rate test provided the reactor is in the cold shutdown condition. The reason for this is that the Engineered Safety Features, which are effective in case of a LOCA under these conditions, will still be effective because they will be activated by low-low reactor water level.

The scram discharge volume has two separate instrument volumes utilized to detect water accumulation. The high water level is based on the design that the water in the SDIV's, as detected by either set of level instruments, shall not be allowed to exceed 29.0 gallons; thereby, permitting 137 control rods to scram. To provide further margin, an accumulation of not more than 14.0 gallons of water, as detected by either instrument volume, will result in a rod block and an alarm. The accumulation of not more than 7.0 gallons of water, as detected in either instrument volume will result in an alarm.

Detailed analyses of transients have shown that sufficient protection is provided by other scrams below 45% power to permit bypassing of the turbine trip and generator load rejection scrams. However, for operational convenience, 40% of rated power has been chosen as the setpoint below which these trips are bypassed. This setpoint is coincident with bypass valve capacity.

A low condenser vacuum scram trip of 23" Hg has been provided to protect the main condenser in the event that vacuum is lost. A loss of condenser vacuum would cause the turbine stop valves to close, resulting in a turbine trip transient. The low condenser vacuum trip anticipates this transient and scrams the reactor. The condenser is capable of receiving bypass steam until 7" Hg vacuum thereby initigating the transient and providing a margin.

3.1-11.

Reactor lades Min. No. of Min. No.of in which Function Operable or Instrument Must Be Operable Operating Channels Per

[ tripped] Operable Action Function Trip Setting Shutdown Refuel Startup- Run Trip systems Trip Systems Required

• K. Rod Block No control rod with-drawals per-mitted

1. SRM Upscale di 5 x 105 cps X X(1) 1 3(y)

2. SRM Downscale 21 100 cps (f) X X(1) 1 3(y)

3. IRM Downscale 2L 5/125 fullscale(g) X X 2 3

4. APRM Upscale ** 2 3(c)

X(s) X X

5. APRM Downscale 3: 2/150 fullscale X 2 3 (c)

6. IRM Upscale 25108/125 fullscale X X 2 3

7. a) water level del 4 gallons X(z) X(z) X(z) 1 1 per high scram instrum, discharge volume volume North b) water level [ 14 gallons X(z) X(z) X(z) 1 1 per high scram instrum, discharge volume, volume South L. Condenser Vacuum Pump Insert Isolation Control Rods

1. High Radia- IE10 x Normal During Startup and 2 2 ation in Main background Run when vacuum pump 1 Steam Tunnel operating Amendment No. g g ,63'

3.1 ,

TABLE 3.1.1' PROTECTIVE INSTRUMENTATION REQUIREMENTS- .

/ .

/ Reactor Modes Min. No.'of Min.'No.of

, /. in which Function! Operable or Instrument . , .

Must Be Operable- Operating' Channels Per

- [ tripped] Operable' Action Function Trip Setting Shutdown Refuel Startup Run Trip systems Trip Systems Required

• A. Scram. Insert f

control rods-

1. Manual Scram. X X X X 2 1

2. High' Reactor ** 2 2 X(s) X X Pressure .

3. High Drywell 5 2 psig X(c) X(u) X 2 2 Pressure-

4. : Low Reactor ** X X X- 2 2 Water Level ,

4

5. a. High Water < 29 gal. X(a) X(z) X(z) 2 2 Level in Scram Discharge Volume -

North Side

b. High Water 5 29 gal. X(a) X(z)- X(z) 2 2 Level in Scram -

! Discharge Volume i South Side

-l 6. Low Condenser > 23" hg. X(b) X(b) X 2 2 l

Vacuum Amendment No. 2ff, X g

.. . . . . . _ _ _ - .