ML20059L122
| ML20059L122 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 01/25/1994 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20059L121 | List: |
| References | |
| NUDOCS 9402030192 | |
| Download: ML20059L122 (3) | |
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UNITED STATES j h;m i
NUCLEAR REGULATORY COMMISSION rg j
WASHINGTON, D.C. 20555-0001
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 61 TO FACILITY OPERATING LICENSE NO. NPF-57 PUBLIC SERVICE ELECTRIC-& GAS COMPANY ATLANTIC CITY ELECTRIC COMPANY HOPE CREEK GENERATING STATION DOCKET NO. 50-354
1.0 INTRODUCTION
By letter dated February 2, 1993, as supplemented by letter dated November 16, 1993, the Public Service Electric & Gas rampany (the licensee) submitted a request for changes to the Hope Creek Gelerating Station, Technical Specification (TS). The proposed changes extend the period of time permitted to reduce the average power range monitor (APRM) and rod block monitor (RBM) setpoints when the plant enters single loop operations (SLO). The November 16, 1993, letter provided clarifying informaticn that did not change the initial proposed no significant hazards consideration determination.
Other changes incorporate core values determined since startup testing:
(1) flow differential for SLO, (2) the minimum thermal power corresponding to no thermal stratification, and (3) core flow with both recirculation pumps at minimum speed.
Long-term single recirculation loop operation at Hope Creek Generating Station was approved by Amendment 3 (See Reference 2), so these changes are updates to a permitted mode of operation.
2.0 EVALUATION 2.1 Sinale looo Goerations Technical Specification 3.4.1.1, Action (a)(1)(e) currently requires reduction of the APRM scram and rod block setpoints to their single loop values within 4 Fours of entering single loop operation. Otherwise the plant must be in hot shutdown in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The proposed change still allows 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to reduce the APRM scram trip setpoints, the APRM control rod block setpoints, and the RBM setpoints to the applicable single. loop values. However, this change requires-that after the initial 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period, instruments be placed in the tripped condition if their setpoints have not been reduced.
Six additional hours are then allowed to adjust the affected channels and place them in operation.
Essentially, the change permits an additional 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to effect the setpoint reductions after entering single loop operation.
The proposed change separately addresses the APRM scram, APRM control rod block, and RBM trip requirements. There is a slight difference in the actions m2oggjigjg.
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. prescribed following the initial 4-hour period for these instruments. The APRM scram actions require that the affected trip system be put in the tripped condition if setpoints are not reduced after the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. The APRM rod block and RBH sections specify that at least one affected channel be placed in the tripped condition if the 4-hour period is exceeded.
Chapter 7 of the Hope Creek Final Safety Analysis Report (FSAR) describes the reactor protection system (RPS) and the rod block portion of the rod control system. The RPS consists of two trip systems.
Each receives inputs from two channels, with APRM instrumentation assigned to each of these trip channels.
Putting one trip system in a tripped condition effectively reduces the RPS coincidence requirement to one-out-of-one from two-out-of-two.
This ensures that a channel with the adjusted setpoints is providing protection. The APRM rod block and RBM action places at least one of the affected channels in the tripped condition, thereby imposing a rod block since any tripped condition input to the rod motion inhibit logic prevents rod movement.
Therefore, by requiring the setpoints of at least one trip system or function to be adjusted to the single loop values, a scram or rod block will be imposed if safety setpoints are reached.
Certain instrumentation guidelines require automatic changes to more restrictive setpoints if operation under conditions requiring them is a likely planned mode of operation (see Paragraph 4.15 of IEEE Standard 279-1971, Reference 3).
However, the staff position for boiling water reactors is that automatic transfer is not required to eu ure a safe and expedient implementation of the proper setpoints. The basis of this staff position allowing manual setpoint switching requires administrative controls to ensure that the more restrictive setpoints are activated in the time specified by the TS.
The proposed change seeks to extend the setpoint switching period by 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
However, it does so by ensuring that at least one protective channel is available with the reduced setpoints after the initial 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period.
Channels not adjusted after the initial period are placed in the tripped condition, thereby ensuring that plant protection is afforded by the more restrictive setpoints.
Thus, plant safety is not adversely affected by the proposed change since more restrictive setpoints are in effect in the time currently required.
Positive-means of implementing the lower setpoints are still provided by the proposed TS requirements. As presented, the changes to provide additional time to reduce protective system setpoints for single recirculation loop operation are acceptable.
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2.2 Power Ascension Proaram Data The values in the TS for delta-w for single-loop operation, minimum thermal power without temperature stratification, and core flow with both
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- l recirculation pumps at minimum speed are initial values established prior to startup testing.
The proposed updates seek to incorporate final values.
a.
Delta-w for single-loop operation Delta-w is defined as the difference in indicated drive flow (in percent of drive flow which produces rated core flow) between two-loop and single-loop operation. A conservative value has been determined based upon calculations following General Electric's (GE) guidelines using plant test data.
This value is acceptable.
b.
Minimum power corresponding to conditions in which no temperature stratification occurs Power ascension testing verified power and flow conditions during single-loop operation at which no temperature stratification occurred. Although the test data did not establish actual minimum power and flow values for the onset of stratification, the data demonstrated that the surveillances required to check for stratification are not required above the proposed TS values of 38 percent thermal power and 50 percent of rated core flow. These proposed updated values of single loop power and flow limits of thermal stratification are conservative relative to actual plant test data and are acceptable.
c.
Core flow with both recirculation pumps at minimum speed TS Figure 3.4.1.1-1 is the power to flow limit used to specify actions needed to detect or avoid limit cycle neutron flux oscillations.
TS 3.4.1.1 contains Actions and Surveillance Requirements (SR) dealing with operation in the power to flow region attributed to these oscillations. The. current TS stipulates neutron monitoring noise level surveillances to be performed between core flow at minimum recirculation pump speed up to the maximum flow defining the limit region. Corrective action to avoid oscillations is taken when flow is equal to or below that value.
The proposed change revises this threshold from 39 percent to 40 percent of rated core flow, consistent with NRC Bulletin 88-07, Supplement 1, " Power Oscillations in Boiling Water Reactors" (Reference 4).
As this is the value specified in the Bulletin for the required actions, and the change to a higher flow is more conservative, the change is acceptable.
2.3 Technical Soecification 3.0.5 The licensee proposed the addition of Limiting Condition for Operation (LCO)'
3.0.5 and the associated bases of NUREG-1433, " Standard Technical Specifications General Electric Plants, BWR/4," to the Hope Creek Generating Station TSs.
The licensee requested this change in order to clarify the intent of LC0 3.0.2 to prevent misinterpretation of.the LCO and associated Bases.
The new LC0 states, " Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its
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