ML20083G311

Proposed Tech Spec Change Request 119 Re long-term Scram Sys Reliability

ML20083G311
Person / Time
Site:	Oyster Creek
Issue date:	12/21/1983
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20083G271	List: ML20083G264 ML20083G289 ML20083G311
References
NUDOCS 8401060161
Download: ML20083G311 (8)

Text

- _ _ _ _ _ _ _ _ _ _ _ _ _ _

l 0YSTER CREEK NUCLEAR GENERATING STATION PROVISI0tlAL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 TECHNICAL SPECIFICATION CHANGE REQUEST fl0. 119 Pursuant to 10CFR50.91, an analysis concerning significant hazards considerations is provided below:

1. Section to be changed:

3.1 and 4.2

2. Extent of change:

Scram Dump Volume (SDV) modifications were performed at the Oyster Creek Nuclear Generating Station in accordance with the BWR Owner's Group recommendations in response to IE Bulletin 80-17 and its supplements.

As a result of providing two SDV instrument volumes to monitor water accumulation, the high water level scram and rod block set points have been changed. Section 3.1 has been modified to reflect these changes.

In addition, the scram discharge volume drain and vent valves closure times in Section 4.2 has been changed to be in agreement with the BUR Owners Group recommendations.

3. Discussion:

On June 28, 1980, during a routine shutdown of the Browns Ferry Unit 3 reactor, a manual scram from approximately 36% power failed to insert about 40% of the control rods. Two additional manual scrams followed by an automatic scram were required before all control rods were fully inserted. The total time that elapsed from the initial scram until all rods were inserted was approximately 15 minutes.

Subsequent investigations by the licensee, General Electric Company, and the NRC staff narrowed the cause of the problem to an accumulation of water in the SDV header at the time of the first scram. It is believed that water accumulated because the SDV system venting and/or draining were obstructed. Furthermore, the accumulation of water was not detected by SDV level instruments which input to the reactor protection system.

It was believed that the SDV level instrumentation was designed to scram the reactor before water accumulated in the scram discharge volume that could hinder scram. As a result, two I&E Bulletins addressing both short and long term programs were issued.

The long term program (which TSCR #119 is in response to) addressed SDV system design. In order to improve the overall design of the SDV system an NRR task force has been working with a subgroup of the BUR Ouners Group to develop revised scram discharge system design and safety criteria. The NRC has endorsed the criteria developed by the BWR Owners Group. GPUN has designed a modification which will meet the criteria developed by the BUR Owners Group. The modification will ensure that i

8401060161 831221 DR ADOCK 05000219 PDR

there is sufficient volume available in the SDV to allow all 137 control rods to scram in tne event that plant conditions warrant this action. These hardware changes have necessitated changing some of the setpoints associated with the system. However, the safety function of the SDY and its associated instruments have not changed. Since the modification and associated technical specificants provide the redundant instrument volumes with necessary instrumentation and appropriate setpoints, the safety of the plant will be enchanced, thereby providing a greater degree of protection for the health and safety of the public.

4. Determination:

We have determined that the subject change request involves no significant hazards in that operation of the Oyster Creek Nuclear Generating Station in accordance with Technical Specification Change Request No. 119 would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3. Involve a significant reduction in a margin of safety.

3.1-4 isolation, initiate automatic depressurization in conjunction with low-low-reactor water level, initiate the standby gas treatment system and ,

isolate the reactor building. The scram function shuts the core down during I 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 l 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 setting is based on the design that 18.36 gallons (59 inches) of water, detected by either set of level instruments will permit the 137 control rods to scram. To provide further margin, an accumulation of 9 gallons (29 inches) of water detected in either instrument volume will(12.1 resultinches) in a rodofblock while an accumulation of 3.76 gallons waterand an alarm,ither detected in e instrument volume results 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 mitigating the transient and providing a margin.

1

3.1-7 TABLE 3.1.7 PROTECTIVE IllSTRUltEf!TATION REQUIREMENTS Reactor Modes Min. No. of !!in. No.of .

in which Function Operable or Instrument Must Be Operable Operatin(I Channels Per

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

• A. Scram Insert control rods

1. lianual Scram X X X X 2 1

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

3. High Drywell _< 2 psig X(u) X(u) X 2 2 Pressure

4. Low Reactor ** X X X 2 2 Water Level

5. a. High Water -< 18.36 gal . X(a) X(z) X(z) 2 4 4

Level in Scram Discharge Volume i

North Side

b. High Water -< 18.36 gal . X(a) X(z) X(z) 2 4 Level in Scram Discharge Volume South Side

6. Low Condenser > 23" hg.

X(b) X(b) X 2 2 Vacuum 4

~

i

) P

.. - - - - . . - . - ~_ .

-3.1-7a j- .

Reactor Modes Hin. 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 *

7. High Radiation <t10 x normal X(s) X X 2 2 in Main Steam Fackground Line Tunnel

8. Average Power ** X(c,s) X(c) X(c) 2 3 Range Monitor (APRM)

9. Intermediate ** X(d) X(d) 2 3 Range Monoitor (IRM)

10. Main Steamline ** X(b,s) X(b) X 2 4 Isolation Valve Closure

11. Turbine Trip ** X(j) 2 4 Scram

12. Generator Load ** X(j) 2 2 Rejection Scram

3.1-11 Reactor Modes  !!in. No. of Min. No.of .

in which Functripn 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-

1. SRM Upscale 55 x 105 X X(1) 1 3(y)

2. SRii Downscale >100 cps (f) X X(1) 1 3(y)

3. IRll Downscale >5/125 fu11 scale (g) X X 2 3

4. APRM Upscale **

X(s) X X 2 3(c)

5. APR!l Downscale >_.2/150 fullscale X 2 3

6. IR!l Upscale 5108/125 fullscale X X 2 3

7. a) water level 4 9 gallons

- X(z) X(z) X(z) 1 1 per high scram instrum, discharge volume volume North b) water level < 9 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- $ 10 x Normal During Startup and 2 2 ation in !!ain background Run when vacuum pump 1 Steam Tunnel operating 4

f

O 3.1-11a .

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

• M. Diesel Generator Time delay Consider con-Load sequence Timers after energi- tainment zation of relay spray loop inoperable and comply with Spec.

3.4.C (See note q.).

1. Containment 40 sec + 15% X X X X 2(m) 1(n)

Spray Pump

2. CRD pump 60 sec + 15%

X X X X 2(m) 1(n) Consider the pump inoper-able and comply with Spec. 3.4.D (See Note q)

3. Emerg. Service 45 sec. + 15%

X X X X 2(m) 1(n) Consider the Water Pump (r) loop inoper-able and comply with Spec. 3.4.C (See Note q)

4. Service Water 120 sec. + 15% (SK1A) X X X X 2(o) 2(p) Consider the Pump (aa) 10 sec. + 15% (SK2A)

(SK7A)

(SK8A)

5. Closed Cooling 166 Sec. + 15%

X X X X 2(m) 1(n) Consider the ,

Water Pump (bb) pump inoper-able and comply within '

7 days (See Note q)

4.2-2 F. At specific power operation conditions, the actual control rod configuration will be compared with the ex)ected configuration based upon appropriately corrected past data. T11s comparison shall be made every equivalent full power month. The initial rod inventory measurement performed when equilibrium conditions are established after a refueling or major core alteration will be used as base data for reactivity monitoring during subsequent power operation throughout the fuel cycle.

G. At power operating conditions, the actual conrtrol rod density will be compared with the 3.5 percent control rod density included in Specification 3.2.B.6. T11s comparison shall be made every equivalent full power month.

H. The scram discharge volume drain and vent valves shall be verified open at least ora per 31 days, except in shutdown mode *, and shall be cycled at least one complete cycle of full travel at least quarterly.

I. All withdrawn control rods shall be determined OPERABLE by demonstrating the scram discharge volume drain and vent valves OPERABLE. This will be done at least once per refueling cycle by placing the mode switch in shutdown and by verifying that:

a. The drain and vent valves close within 30 seconds after receipt of a signal for control rods to scram, and

b. The scram signal can be reset and the drain and vent valves open when the scram discharge volume trip is bypassed.

Basis: The core reactivity limitation (Specification 3.2.A) requires that core reactivity be limited such that the core could be made subcritical at any time during the operating cycle, with the strongest operable control rod fully withdrawn and all other operable rods fully inserted. Compliance with his requirement can be demonstrated conveniently only at the time of refueling. Therefore, the demonstration must be such that it will apply to the entire subsequent fuel cycle. The l

i demonstration is performed with the reactor in the cold,

! xenon-free condition and will show that the reactor is sub-critical at that time by at least R + 0.25% 4 k with the l

highest worth operable control rod fully withdrawn.

1

• These valves may be closed intermittently for testing under l administrative control, l

l l

l