ML20236T495

TS Change Request 256 to License DPR-16,permitting Alternative to Requirement to Perform Control Rod Drive Scram Time Testing W/Reactor Pressurized Prior to Resuming Power Operation.W/Certificate of Svc

ML20236T495
Person / Time
Site:	Oyster Creek
Issue date:	07/21/1998
From:	Roche M GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20236T487	List: ML20236T486 ML20236T495 ML20236T498
References
NUDOCS 9807280219
Download: ML20236T495 (7)

Text

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OYSTER CREEK NUCLEAR GENERATING STATION j

OPERATING LICENSE NO. DPR-16 I

TECHNICAL SPECIFICATION CHANGE REQUEST NO. 256  !

DOCKET NO. 50-219  :

Applicant submits by this Technical Specification Change Request No. 256 to the Oyster Creek Nuclear Generating Station Technical Specifications, modified page 4.2-1.

By: 0w MichaelB. Roche Vice President and Director Oyster Creek Swom to and Subsenised before me this 2l dday of 3v [ .1998.

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A Notary P6blic of Ndw Jersey i

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9807200219 990721 PDR ADOCK 05000219' P PDR_.

9 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTEROF l GPU NUCLEAR, INC. l DOCKET NO. 50-219 CERTIFICATE OF SERVICE I

This is to certify that a copy of Technical Specification Change Request No. 256 for the Oyster Creek Nuclear Generating Station Technical Specifications, filed with the U.S. Nuclear Regulatory Commission on 2l has this day of ~3dyj / 418 been served on the Mayor of Lacey Township, Ocean County, New Jersey by deposit in the U.S. mail, addressed as .

follows: )

The Honorable Louis A. Amato Mayor ofLacey Township 818 West Lacey Road Forked River, NJ 08731 i

By: f Michael B. Roche Vice President & Director Oyster Creek

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ENCLOSURE 1 OYSTER CREEK NUCLEAR GENERATING STATION OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 TECHNICAL SPECIFICATION CHANGE REQUEST NO. 256 Applicant hereby requests the Commission to change Appendix A of the above<aptioned license as follows:

1. Section to be Chanoe21: 4.2.C.1

2. Change Requested:

Replace old page: 4.2-1 with new pages : 4.2-1  !

3. Description of Chanoe:

The requested change, would permit an alternative to the requirement to perform Control Rod Drive (CkD) scram time testing with the reactor pressurized prior to resuming power operation. Instead, the change also permits: (1) scram time testing 1 with the reactor depressurized prior to resuming power operation, and (2) a second  !

scram time test with reactor pressure above 800 psig, prior to exceeding 40% reactor power.

Verification of the ability of the CRD System to insert rods into the reactor core within the required time to scram the reactor, could be performed in two steps, instead of the single step as presently required . Specifically, after each major refueling outage, all operable j control rods could alternatively be scram time tested as follows:

1; Prior to resuming power operation, scram time testing shall be conducted with the reactor depressurized. The 90% scram insertion time shall not exceed 2.2 seconds for each control rod tested, and 4

2. Prior to exceeding 40% reactor power operation, scram time testing shall be conducted with the reactor pressurized above 800 psig. The acceptance criteria of (existing) Technical Specification Section 4.2.C.2 (newly numbered section 4.2.C.3) shall be met for each control rod tested.

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, , 1940-98-20175 Enclosure 1 Page 2 Cold (Depressurized) Scram Time Test - A 90% scram test time of 2.2 seconds at 0 psig corresponds to about 3.2 seconds at 800 psig and 3.8 seconds at about 600 psig.

The Technical Specification Limiting Condition for Operation requirement is for the 90% insertion time average for all rods to be less than 5.0 seconds, with reactor l

pressure above 800 psig. Therefore, the cold (depressurized) 90% insertion time acceptance criteria of 2.2 seconds provides reasonable assurance that the 90% insertion time criteria of Technical Specification Section 3.2.B.3 is met. Additionally, below 40% power, sufficient inherent margin exists to core thermal operating limits (due to significantly lower heat flux), to offset the potential impact of any slow scram time.

There are three considerations associated with the cold (depressurized) scram time tests:

1. Wear of Control Rod Drives (CRD) - Cold (depressurized) scrams will subject the drive mechanisms to higher mechanical loads than during hot (pressurized) scrams. The CRDs are designed to accept these loads as documented in vendor Component Specifications. These specifications provide allowances for many startup and testing scram cycles. Additionally, Technical Specifications for other Boiling Water Reactors (BWR), which use the same CRD components, recognize the performance of depressurized (cold) scrams as a suitable way to verify rod functioning after i maintenance. The failure mechanism resulting from the higher loads during cold scrams is increased buffer seal wear or failure. This wear or failure of the buffer seal 1 I

would result in difficulty or inability to withdraw the rod subsequent to the depressurized scram. The safety function of the rod to insert on a scram signal, however, would be unaffected by this seal degradation. Therefore, there is no safety

, concern with wear on the CRDs during the performance of the cold scram test.

2. Risk of Stub Tube Leakage-When the reactor is pressurized, there is a large downward load on the stub tube equal to the system pressure times the cross-sectional area bounded by the stub tube outer diameter. During a scram, there is a small, momentary, upward load (rod scram reaction) which is reacted at the stub tube. Cold scrams may increase ,

the risk of stub tube leakage because, without the download due to reactor pressure, the . {

. momentary upward loading on the CRD stub tube puts the stub tube into tension. Any I flaws in the stub tube could grow and eventually result in a stub tube leak. The likelihood of there being flaws in the stub tubes, however, is very small, based on the l- extensive repair work on the stub tube surfaces performed prior to plant operation. The l integrity of the stub tube repairs is verified by the 1000 pound leak test performed during every startup of the reactor. This test, therefore, poses very minimal risk of stub tube leakage.

3. Operation of Ball Check Valve-The cold scram test can verify all aspects of the scram function, except for the operation of the ball check valve in the CRD flange.

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1940-98-20175 Enclosure 1 Page 3 This valve allows reactor pressure to provide the driving force for rod insertion after the accumulator drops below reactor pressure.

During the test at pressure the operation of the ball check valves are indirectly indicated by relatively slow rod insertion times. Plant procedures require an engineering evaluation of rods with insertion times outside an expected range. If this valve failed to

. reposition during a scram with system pressure above 600 psig, the control rod would fully insert, but at a much slower rate. This is acceptable because proper operation of the ball check valves will be verified during hot scram testing prior to exceeding 40%

power.

Hot Scram Time Test - The hot scram test is the same test currently required by the Oyster Creek Technical Specifications. The requested change allows the hot scram test to be performed prior to reaching 40% power, instead of being required prior to startup. The Standard Tech Specs for BWR - 4s and BWR - 6s require hot (pressurized) scram time testing after refueling prior to exceeding 40% reactor power operation. At 40% power, inherent margin will exist to the Minimum Critical Power Ratio (MCPR) operating limit to offset the potential impact of any slow scram times.

4. No Sinnifnnt Hmrds Determination Pursuant to 10 CFR 50.91, this Technical Specification Change Request has been determined to contain No Significant Hazards. These evaluations are specified in 10 CFR 50.92.

This request has been determined to involve No Significant Hazards in that it does not:

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l 1. Involve a significant increase in the probability of occurrence or consequences of an L accident previously evaluated; (or)

There will not be an increase in the probability of occurrence of an accident previously evaluated in the Safety Analysis Report (SAR) because the requested change provides additional assurance that the CRD System is able to perform its safety function, and therefore does not change the probability of occurrence of an accident.

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1940-98-20175 Enclosure 1 Page 4 There will not be an increase in the consequences of an accident previously evaluated in the Safety Analysis Report (SAR) because the requested change will ensure that the CRD System is able to perform its safety function, and therefore does not change the consequences of an accident.

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

The requested change will not create the possibility of a new or different kind of accident from any accident previously evaluated. The first issue associated with the

requested change is increased wear on the CRDs, resulting in increased buffer seal wear or failure. This wear or failure of the buffer seal would result in difficulty or inability to withdraw the rod subsequent to the depressurized scram. The safety function of the rod to insert on a scram signal, however, would be unaffected by this seal degradation. Therefore, there is no safety concern with the increased wear due to performance of the cold scram test.

The other consideration associated with the new requested change is the possible increased risk of stub tube leakage during the cold (depressurized) test. Without the download due to reactor pressure, the momentary upward loading on the CRD stub tube puts the stub tube into tension. Any flaws in the stub tube could grow and eventually result in a stub tube leak. The likelihood of flaws in the stub tubes, however, is very small, based on the extensive repair work on the stub tube surfaces

! performed prior to plant operation. The int 6grity of the stub tube repairs is verified by the 1000 pound leak test performed during every startup of the reactor. This test, therefore, poses very minimal risk of stub tube leakage.

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3. Involve a significant reduction in a margin of safety The change will not decrease the margin of safety as defined in the basis of any Technical Specification. This is because the requested change, like the existing Technical Specification test, provides assurance that the CRD System is able to perform l its safety function, and therefore does not change the margin of safety.

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I Enclosure II Changed Page 4