ML20137F368

Deletion of Test 19 Core Power - Void Mode Response Test from Power Ascension Test Program, Safety Evaluation

ML20137F368
Person / Time
Site:	Hope Creek
Issue date:	08/20/1985
From:	Public Service Enterprise Group
To:
Shared Package
ML20137F347	List: ML20137F342 ML20137F349 ML20137F361 ML20137F368 ML20137F393 ML20137F427
References
PSE-SE-Z-003, PSE-SE-Z-3, NUDOCS 8508260203
Download: ML20137F368 (6)

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PUBLIC SERVICE ELECTRIC AND GAS-COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-Z-003 TITLE: DELETION OF TEST 19 CORE POWER - VOID MODE RESPONSE TEST FROM POWER ASCENSION TEST PROGRAM Date: AUG 201N5 1.0 PURPOSE

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The purpose of this Safety Evaluation is to evaluate the acceptability of deleting the core power-void mode response test from the Power Ascension Test Program.

2.0 SCOPE The test specifications are related to the Nuclear Core which is a subsystem of the Nuclear Boiler.

3.0 REFERENCES

1. General Electric Startup Test Specification, -

23A4137, Revision 0

2. Regulatory Guide 1.68, Revision 2, August 1978 -

3. Hope Creek Generating Station Final Safety Analysis Report Chapter 14

4. Letter, C. O. Thomas (NRC) to H. C. Pfefferlen (GE),

" Acceptance for Referencing of Licensing Topical Report NEDE-24011, Revision 6, Amendment 8, Thermal Hydraulic Stability Amendment to GESTAR II", April 24, 1985

5. General Electric Service Information Letter (SIL) 380

6. Hope Creek Generating Station Draft Technical Specifications 4.0 DISCUSSION There are no specific Regulatory Guide 1.68 requirements to perform stability testing during the Power Ascension Program. However, paragraphs 5.2, 5.v, and 5.h.h require the demonstration of acceptable control system responses during steady state and transient conditions. Test Number 19 (FSAR Figure 14.2-5), Core-Void Mode Response, measures the stability of the core power void dynamic response by moving a very high worth control rod one or two notches. In conjunction, Test Number 20 (FSAR Figure 14.2-5), Pressure Regulator, performs pressure regulator 8508260203 850821 PDR ADOCK 05000354 A PDR PSE-SE-2-003 1 of 3

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step changes to measure the core power void dynamic response. These tests are currently planned to be performed at Test Conditions 4 and 5 (FSAR Figure 14.2-4). It is proposed to delete the control rod movement tests at Test Conditions 4 and 5 while still maintaining the pressure regulator testing at Test Condition 5.

Response of the core power void mode is determined by analyzing test data and comparing to an acceptance criterion which' defines the required system performance.

The criterion requires that all system related variables must exhibit non-divergent behavior. System related variables are heat flux ind reactor pressure.

Extensive special testing of stability characteristics has been performed at several BWRs, including BWR/4

• plants similar to Hope Creek (Peach Bottom-2 and Browns Ferry). The test data has demonstrated the stability characteristics of BWRs over a wide range of conditions and has been reviewed along with extensive supporting analyses, as part of the staff's Safety Evaluation Report on core thermal-hydraulic stability (Reference 4).

For modern high power density reactors, control rod oscillator tests.are not desirable because of poor signal '

to-noise ratios in large reactor cores. Measurement of system stability will be provided using the small ,

pressure perturbation techniques. This has been verified a reliable technique to determine the reactor core i stability margins. Test Number 20, Pressure Regulator-testing, measures the system response to pressure disturbances caused by actions of the pressure regulator system. This testing yields valuable core stability data at the limiting high power / low flow condition encountered during normal operation (Test Condition 5). In addition to Test Number 20, normal observations of operational i

power maneuvers provide sufficient data to determine the normal stability characteristics and response of the system.

In addition to the. pressure regulator testing, Service Information Letter (SIL) 380 (Reference 5) provides' detailed recommendations for the monitoring of system behavior. These recommendations which are being incorporated by Hope Creek Generating Station, provide for monitoring of neutron flux characteristics during normal operation at high power low flow conditions and i

during abnormal operating conditions. In addition to the monitoring requirements, current Technical Specifications do not allow continued operation at natural circulation j

flow which is the least stable condition of the operating region.

PSE-SE-2-003 2 of 3

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5.0 CONCLUSION

As a result of the extensive testing and analysis of core thermal-hydraulic stability, it has been demonstrated that General Electric BWR fuel and core designs meet the stability criteria set forth in General Design Criteria 10 and 12 of 10CFR50, Appendix A. Based on the above discussion and the Staf f's Safety Evaluation Report (Reference 4), the proposed change will not affect any safety related systems or safe operation of the plant and therefore does not involve an Unreviewed Safety Question. System stability is adequate'ly measured during Test Number 20, Pressure Regulator, and has been extensively tested at several BWRs covering a wide range of designs. In addition, information on the system's stability is continuously provided by SIL-380 recommendations for the monitoring of neutron flux.

Therefore, Test Number 19, Core Power-Void Mode Response can be deleted from the Power Ascension Test Program.

The proposed changes do not affect the operation of Hope Creek and revisions to the plant Technical Specifications are not required.

6.0 DOCUMENTS GENERATED NONE

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7.0 RECOMMENDATIONS Revisions to Hope Creek's FSAR and Power Ascension procedures shall be made to reflect the deletion of the core power-void mode response testing as described above.

8.0 ATTACHMENTS NONE 9.0 SIGNATURES Originator b , ftw [/Date I Verifier CT23,;2- CL h f/2*/f5<

Date Group Head (or SSE) $m k 5'/M/ B I Systems Analysis Group Head -d,N. M I 18 5 Site Engineering Manager [b)J) [~ f[3 i'

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?I PSE-SE-2-003 3 of 3

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TEST NUMBER 25 - TURBINE TRIP AND GENERATOR LOAD REJECTION DELETION OF TURBINE TRIP TEST AT TEST CONDITION 3 OBJECTIVE:

Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraphs 5.1.1 and 5.n.n require that a Turbine Trip and Generator Load Rejection be performed at 100% power to demonstrate that the dynamic response of the plant is in accordance with design requirements for turbine trip and full load rejection. These tests may be combined if a turbine trip is initiated directly during the generator load rejection instead of tripping.from secondary effects such as a turbine overspeed trip. Test Number 25, Turbine Trip _and Generator Load Rejection, is currently planned to be performed at three conditions during the power ascension test program; (1) a generator load rejection during Test Condition 1 or 2 (within the bypass capacity of the plant); (2) a turbine trip during Test Condition 3 (approximately 75%

power); and (3) a generator load rejection at Test Condition 6 (approximately 100% power) . It is proposed to delete the Turbine Trip test at Test Condition 3 and change the Generator Load Rejection test at Test Condition 1 or 2 to a Turbine Trip test. This proposed testing will demonstrate that Regulatory Guide 1.68 objectives are met.

DISCUSSION: -

Response of the system during a turbine trip and generator load rejection is determined by analyzing test data and comparing to acceptance criteria, level 1 and level 2, which define the required system performance. Level l criteria require; proper operation of the turbine control and stop valve closure times with respect to the bypass valve opening time, adequate bypass valve response times, proper feedwater control system response to prevent flooding of the steam lines, that recirculation flow coastdown following protective trips is within design values, acceptable vessel dome pressure and simulated heat flux response, and require proper operation of the low-low set pressure relief logic for the safety / relief valves. Level 2 criteria require; that no MSIV closure occur during the first three minutes of the event, that vessel dome pressure and simulated heat flux changes do not exceed predicted values, for the generator load rejection within bypass capacity calculated is greater than or equal to the assumed value in FSAR analysis, that low water level recirculation pump trip, HPCI and RCIC are not initiated, that feedwater level control avoids loss of feedwater because of high level trips and require that safety / relief valve discharge temperatures remain within acceptable limits.

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The generator load rejection (Test Condition 6) and the turbine trip (Test Condition 1 or 2) provide data to demonstrate that the level 1 and 2 criteria are met during a turbine trip. The turbine bypass system performance will be verified at a lower power level by changing the proposed generator load rejection at Test Condition 1 or 2 to a turbine trip. Because of the enhanced data acquisition systems available, integrated system respon'se to a turbine trio can be obtafned f rom the Generator Load Rejection test at test Condition 6.

Control systems which regulate the long term operation following the transients are separately tested during the power ascension test program.' Feedwater and level control system turning in Test Number 23 (ST Feedwater System Response) will ensure proper water level control. High and low water level trip avoidance will be verified in the Generator Load Rejection test at test Condition 6. Pressure control tuning during Test Number 22 (ST Pressure Regulator testing ) will ensure that the MSIV closure trip on low turbine inlet-pressure is avoided during the transient.

CONCLUSIONS:

The turbine trip test has.been previously demonstrated to be a mild transient event and poses no serious threat to the '

core and reactor integrity (reference Chapter 15 section,

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15.2.3). In addition, the transient results from a generator load rejection at full power are more limiting than the results from a turbine trip at Test Condition 3 (see Table 1). Based on the above discussions, the proposed change will not affect the safety related systems or safe operation of the plant and therefore does not involve an unreviewed safety question.

Current testing of the generator load rejection at 100%

power, satisfies the requirements imposed by Regulatory Guide 1.68 (Revision 2), Appendix A, paragraphs 5.1.1 and 5.n.n.

In addition, the proposed Turbine Trip test within bypass valve capacity (Test Condition 1 or 2) provides additional verification of the response of the protective systems and also provides demonstration of the bypass system's capability to avoid scram at low power levels. Therefore, the turbine trip at Test Condition 3 can be deleted with the added change that the generator load rejection at Test Condition 1 or 2 will be changed to a turbine trip test.

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TABLE 1 j STARTUP DATA COMPARISON i Grand Gulf-1 Kuo Sheng 1 Kuo Sheng 2 LaSalle 2 Susquehanna.1 LR TT LR TT LR TT LR TT LR TT 4

Power / Flow 99.7 72.1 97 74 95.8 75.1 95.7 69 99.9 75 97.8 96.4 99.4 100 99.1 99.9 93.1 93 98.8 100 l

Initial Dome 1023 989 985 978 981 972 999 967 1010 980 Pressure (psig)

I Peak Dome 1111 1044 1045 1009 1052 1004 1078 1030 1075 1043 Pres. (psig)

Dome Press 88 55 60 31 71 32 79 63 81 63 j (psi)

Heat Flux 0 0 0 0 0 0 0 0 0 0

(% NBR)

LR = Load Rejection TT = Turbine Trip i

thj j

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