ML20135H777

From kanterella
Jump to navigation Jump to search
Forwards Detailed Justifications on Power Ascension Test Mods & marked-up FSAR Changes Reflecting Proposed Mods.Encl Safety Evaluations Find That Proposed Mods Pose No Risk to Public.Expedited Review Requested.Fsar Will Be Amended
ML20135H777
Person / Time
Site: Hope Creek PSEG icon.png
Issue date: 09/20/1985
From: Mittl R
Public Service Enterprise Group
To: Butler W
Office of Nuclear Reactor Regulation
References
NUDOCS 8509240217
Download: ML20135H777 (40)
v • d • e


Text

Il l

,h

I

]-

Public Sarwce Electncand Gas L Cornoany 80 Park Plaza, Newark, NJ 07101/ 201430-8217 MAILING ADDRESS / P.O. Box 570, Newark, NJ 07101 Robert L Mitti General Manager Nuclear Assurance and Regulation September 20, 1985 Director.of Nuclear Reactor Regulation United States Nuclear Regulatory Commission

'7920 Norfolk Avenue Bethesda, Maryland 20814 Attention: Mr. Walter Butler, Chief Licensing Branch 2 Division of Licensing Gentlemen:

POWER ASCENSION PROGRAM CHANGES HOPE CREEK GENERATING STATION DOCKET-NO. 50-354 As committed in PSE&G's August 28, 1985 letter on the' subject topic, please find attached the following detailed justifi-cations on power ascension test modifications for-NRC's consideration:

1. Test Number 12 - Reactor Core Isolation Cooling System
2. Test Number.13 - High Pressure Coolant Injection System
3. Test Number 14A - Selected Process Temperatures
4. Test Number 9 - LPRM Calibration
5. Test Condition 4 - Natural Circulation Operation Attachment 1 consists of General Electric Company's technical analysis and PSE&G's safety evaluation-for each of the afore-mentioned items. The conclusion for each item shows'the proposed modifications pose no increase in risk to the health and safety of the public or an.unreviewed safety question.

The Station' Operations Review Committee (SORC) has reviewed each' evaluation and concurs with the conclusions reached therein.

Attachment _2 consists of marked up FSAR pages applicable to the proposed _ testing modifications. These FSAR changes will be included as an amendment-to the FSAR pending approval for implementation.from the NRC.

\

8509240217 850920 PDR ADOCK 05000354 r

fI g.

The Energy People A PDR J l{

9$4912 (3M) 4 84

6 Director of Nuclear 2 9/20/85 Reactor Regulation Test Number 2 - Radiation Measurements, has been withdrawn from consideration after performing the detailed review in support of the safety evaluation. The determination has been made by Public Service that this test will not be modified as part of the power ascension acceleration program. Test Number 1 - Chemical and Radiochemical Testing, and Test Number 10 - APRM Calibration are in the detailed I review process in support of the safety evaluation and will be submitted as soon as possible.

Since the enclosed modifications impact finalizing the related power ascension testing detailed procedures, an expedited review is requested. PSE&G is ready to meet with cognizant NRC personnel to discuss the proposed modifications should you require additional information.

Very truly yours, s .

(

Attachments C D.H. Wagner USNRC Licensing Project Manager A.R. Blough USNRC Senior Resident Inspector

1 ATTACIIMENT I (30 Pages) i

TEST NUMBER 12 - REACTOR CORE ISOLATION COOLING SYSTEM TEST SIMPLIFICATION REDUCED NUMBER OF TESTS OBJECTIVE:

a 1

Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, l_ paragraphs 4.k and 5.k require the demonstration of the

operability of steam-driven engineered safety features. Test Number 12, Reactor Core Isolation Cooling system (RCIC),

verifies the proper operation of the RCIC over its required operating pressure range. Testing is performed at low reactor t

pressures during heatup and at near rated reactor pressures during Test Condition 1. Testing of the RCIC is performed by flow injection into a test line leading to the condensate storage tank (CST) and by flow injection directly into the reactor vessel. Currently, tuning of the RCIC controllers is

! performed at both low pressures and near rated reactor i

pressure. It is proposed that reactor pressures be deleted.

the centroller tuning at low i

DISCUSSION:

Response of the RCIC system is determined by analyzing test

  • data and comparing to acceptance criteria which define the required performance of the system. The criteria place limits

-on the RCIC flow responEs times, RCIC turbine dynamic response

and speed and flow control loop stability. In addition, it is required that the RCIC turbine does not trip or isolate during automatic or manual start tests. Previous test experience from similar plants provides best estimates for initial controller settings; bench calibration of the controllers using these best estimates will be performed during the preoperational phase of l testing. Testing is performed at low reactor pressures and near rated reactor pressure to demonstrate compliance with i response and stability criteria over the expected range of operating conditions. Initially, a set' of CST injection tests are performed with manual and automatic starts at 150 psig and near rated reactor pressure conditions. This CST testing is done to demonstrate the general system operability and for

, making most controller adjustments. Experience at previous plants has shown that controller tuning at the low pressure 1

condition pressures.

does not result in optimum performance at higher Therefore, it is proposed that the turbine speed controller be CST injection mode.

first tuned near rated reactor pressure in the 1

Reactor vessel injection tests follow to complete the i

controller adjustments and to demonstrate automatic starting from a " cold" standby condition (" cold" is defined as a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any kind of RCIC operation). Since the reactor pressure vessel injection testing is the least stable 4

l

! I i

4

, , - -,, , - , , . ,..-. -- - - . . .-r_ - , , . ~ . - - - - ,- , , , - - .c, -- - --

.- - , - = .

operating condition, it is proposed that flow controller tuning

j. be performed. at these conditions near rated reactor pressure.

I The system can then be retested at lower pressures for verification of control loop settings. Af ter all final controller and system adjustments have oeen determined, a set of demonstration tests are performed with the final settings.

These demonstration tests provide the necessary information to demonstrate compliance with Regulatory Guide 1.68 objectives.

CONCLUSION:

Test Number 12, RCIC, requires a final set of demonstration tests to verify system performance. These tests demonstrate compliance with the objectives of Regulatory Guide 1.68, Appendix A, paragraphs 4.k and 5.k. Prior tuning of the control systems can therefore be done at the most efficient test conditions to minimize unnecessary testing of the system.

The proposed change does not affect any safety systems or the safe operation of the plant and therefore does not involve an unreviewed safety question. Test Number 12, RCIC, can

' therefore be simplified by deleting tuning of the controllers during low pressure of CST injection mode testing. Controller adjustments will be made during the rated reactor pressure ,

t testing of the reactor pressure vessel injection mode.

i i

f 5

i a

2 1

.- ~ , . . - ..

PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PS E-S E-2-010 TITLE: DELETION OF CONTROLLER TUNING AT LOW REACTOR PRESSURES FROM REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM TEST-TEST 12 SEP 201985 Date:

1.0 PURPOSE The purpose of this safety evaluation is to document the results of the evaluation of the proposal to tuning of the RCIC controllers at low reactor pressures during the power ascension test program.

2.0 SCOPE The area of concern is the performance testing of the RCIC system during the startup test program.

3.0 REFERENCES

1. NRC Regulatory Guide 1.68, Revision 2, August 1978.
2. General Electric Startup Test Specification 23A4137, Revision 0.
3. HCGS Draf t Technical Specifications.
4. FSAR Chapter 14.
5. Draft Hope Creek Power Ascension Test Procedure IC-LC-B D-0 0 2 ( O) , Loop calibration, RCIC Turbine Controller Tune-up.
6. Draf t Hope Creek Power Ascension Tes t Procedure IC-LC-BD-001(0), Loop calibration, RCIC Flow Controlle r Tune-up.

4.0 DISCUSSION Regulatory Guide - 1.6 8 (Revision 2, Augus t 1978 ), Appendix A, paragraph 4.k requires that the operability of the PSE-SE-2-010 1 of 4

s steam-driven engineered safety features be demonstrated, Paragraph 5.k requires a demonstration that ECCS high-pressure coolant ~ injection systems can start under )

sivalated accident conditions and inject into the coolant system as designed. Although not classified as an ECCS system, the RCIC system will be tested to verify that it meets the requirements of this paragraph. Test Number 12, Reactor Core Isolation Cooling (RCIC) System, verifies the proper operation of this system over its required operating pressure range as prescribed by GE

. Test Specification 23A4137. It is proposed that the tuning of the RCIC controllers at low reactor pressure during CST injection be deleted in Test Number 12.

Bench calibration of the controllers, and initial controller settings are performed during the latter stages of preoperational phase testing using data obtained from the testing experiences of similar plants.

During the startup test program, tests are performed at low reactor pressure and near rated reactor pressure to demonstrate compliance with the acceptance criteria of Test Number 12 (this defines the required system performance) over the expected range of operating conditions. The RCIC controllers will have been pre-set prior to this phase of testing. Normally, a series of CST injection tests are first performed (manual and automatic starts) at a reactor pressure of 150 psig and near rated reactor pressure conditions to demonstrate the RCIC system's general operability and for making most controller adjustments. Reactor vessel injection tests follow to complete the controller adjustments and to demonstrate automatic starting from cold standby conditions (" cold" is defined as a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any kind of RCIC operation). However, because controller tuning at low reactor pressure is not required to demonstrate RCIC system operability and because experience at other plants has shown that controller tuning at the low pressyre condition does not result in optimum performance at higher pressures, it is proposed that all controller tuning at low reactor pressure be deleted during CST injection mode. The RCIC turbine speed controller will be tuned at near rated reactor pressure during CST injection. Since reactor vessel injection by the RCIC system is the most rigorous operating condition of the RCIC system, the flow controller tuning will be performed at these conditions PSE-SE-2-010 2 of 4 r

s near rated re. actor pressure of Test Condition 1. The system will be retested at lower pressures for verification of control loop settings. After all final controller and system adjustments have been determined, a set of demonstration tests are performed with the final settings. These demonstration tests provide the necessary information to demonstrate compliance with Regulatory Guide 1.68 objectives.

5.0 CONCLUSION

Test Number 12, RCIC, requires a final set of demonstration tests to verify system performance. These tests demonstrate compliance with the objectives of Regulatory Guide 1.68, Appendix A, pragraphs 4.k and 5.k.

Since tuning of the control systems is not required at low reactor pressure to demonstrate RCIC system operability, it can be done at the most efficient test conditions to minimize unnecessary testing of the system.. The proposed change does not adversely af fect any safety systems or the safe operation of the plant and, therefore, does not involve an unreviewed safety question.

required.

A Technical Spe. +. fications change is not Test Number 12, Jh'I C , can, therefore, be simplified low pressure.by deleting tuning of the controllers during Flow Controller adjustments will be made during rated reactor pressure testing in the reactor vessel injection mode.

6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATION Revisions to Hope Creek's startup and power ascension test procedures shall be made to reflect the simplification of RCIC testing as discussed above.

8.0 ATTACHMENTS None r

PSE-SE-2-010 '

3 of 4 l

l

9.0 SIGNATURES Originator ( / 8 Verifier / -

i p (' S' Group Head SSE) S e e Sg P. E. 'Of2o/gs' Systems Analysis Group Head' k.,b kh ((b Site Engineering Manager [/1)(

12 9/, /C4 Da't!e / ~

PSE-SE-Z-010 4 of 4 e

, TEST NUMBER 13 - HIGH PRESSURE COOLANT INJECTION SYSTEM TEST SIMPLIFICATION - REDUCED NUMBER OF TESTS OBJ ECTIVE :

Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraphs 4.k requires the demonstration of the operability of steam-driven engineered safety features at less than 5% power.

Paragraph 5.'k requires demonstration of the HPCI system auto-start and reactor coolant system injection capability at 25 % to 50 % power. Test Number 13, High Pressure Coolant

Injection System ( H PCI) , verifies the proper operation of the i HPCI over its required operating. pressure range. Testing is i performed at low reactor pressures and near rated reactor pressure during heatup and at near rated reactor pressure
during Test Condition 3. Testing of the HPCI system is performed by flow injection into a test line leading to the condensate storage tank (CST) and by flow injection directly

, into the reactor vessel. Currently, tuning of the HPCI J controllers is performed at both low pressures and near rated reactor pressure. It is proposed that the controller tuning at low reactor pressures be deleted.

DISCUSSION:

Response of the HPCI system is determined by analyzing test data and comparing to acceptance criteria which define the required performance of tne system. The criteria place limits

, on the HPCI flow response times, H PCI turbine dynamic response and speed and flow control loop stability. In addition, it is required that the HPCI turbine does not trip or isolate during automatic or manual start tests. . Previous test experience from i similar plants provides best estimates for initial controller

settings.' Bench calibration of the controllers using these best estimate settings will be performed during the

{ preoperational phase of testing. Testing is performed at low l . reactor pressures and near rated reactor pressure to demonstrate compliance with response and stability criteria over the expected range of operating conditions. Initially, a set of CST injection tests are performed with manual and automatic starts at 200 psig and near rated reactor pressure conditions. This CST testing is done to demonstrate the general system capability and for making most controller i adjustments. Experience at previous plants has shown that controller tuning at the low pressure condition does not result in optimum performance at higher pressures. Therefore, it is proposed that the turbine speed controller be tuned first near rated reactor pressure in the CST injection mode.

i 1

i i

l

Reactor vessel injection tests follow to complete the controller adjustme'nts and to demonstrate automatic starting from.a " cold" standby condition (" cold" is defined as a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any kind of HPCI operation). Since the

' reactor pressure vessel injection testing is the least stable operating condition, it is proposed that flow controller tuning be performed at these conditions near rated reactor pressure.

The system can then be retested at lower pressures for verification of control loop settings. After all final system adjustments have been determined, a set of demonstration tests are performed with the final settings. These demonstration tests provide the necessary information to demonstrate compliance with Regulatory Guide 1.68 objectives.

CONCLUSION:

Test Number 13, HPCI, requires a final set of demonstration tests to verify system performance. These tests demonstrate '

compliance with the objectives of Regulatory Guide 1.68, Appendix A, paragraphs 4.k and 5.k. Prior tuning of the '

control systems can therefore be done at the most ef ficient test conditions to minimize unnecessary testing of the system.

i The proposed change does not af fect any safety systems or the safe operation of the plant and therefore does not involve an unreviewed safety question. Tes t Number 13, HPCI, can therefore be simplified by deleting tuning of the controllers during low pressure. Controller adjustments will be made during the rated reactor pressure testing of the CST injection mode for the turbine speed controller and the reactor pressure vessel injection mode for the flow controller.

4 2

, PU LIC SERVICE' ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-2-012 TITLE: SIMPLIFICATION OF POWER ASCENSION HIGH PRESSURE COOLANT INJECTION SYSTEM FLOW CONTROLLER TUNING -

4 TEST 13 Date: SEP 201985 1.0 PURPOSE The purpose of this Safety bvaluation is to document the results of the evaluation performed to ensure that simplification of Hope Creek's Power Ascension High Pressure Coolant Injection Flow Controller Tuning will not adversely affect reactor safety.

2.0 SCOPE ,

The Test Specifications are for the High Pressure Coolant Injection System.

4

3.0 REFERENCES

1. Regulatory Guide 1.68, Revision 2, August 1978
2. Hope Creek Final Safety Analysis Report (FSAR)

Chapter 14

3. General Electric Startup Test Specification, 23A4137, Revision 0
4. Hope Creek Generating Station Draft Technical Specifications
5. Draf t Hope Creek Power Ascension Test Procedure TE-SU-BJ-151(0), High Pressure Coolant Injection System Condensate Storage Tank Injection.

4.0 DISCUSSION Reference 1, Appendix A, Paragraph 4.k requires that steam driven engineered safety features equipment be demonstrated operable prior to reaching 5% reactor power. Paragraph 5.k of Reference -1, Appendix A

.s PSE-SE-2-012

  • 1 of 3 i

S f .-

t f ,

~ .! , .

\(

[' requires that a demonstration that the high pressure coolant injection system can start under simulated accident conditions and inject into the reactor coolant

.)s system 'as designed be performed between 25 % and 50%

power. It is proposed to delete the tuning of the HPCI controllers at low reactor pressure.

Hope 2,

Creek's Power Ascension Test Number 13, Reference Paragraph 14.2.12.3.13 verifies system operability, fine-tunes the flow controller and demonstrates system performance range.

over the. required operating pressure CST injection testing is done during heatup to demonstrate power.

system operability prior to exceeding 5%

These tests consist of manual and automatic mode pressure.

starts at 200 psig ~and at. near rated reactor The HPCI controllers will have been pre-set during the HPCI preoperational test. The HPCI speed controller will be fine-tuned at near rated reactor

' pressure during the CST injection mode. HPCI performance will then be demonstrated at low pressure in the CST injection mode prior to exceeding 54 power.

Reactor vessel injection testing will be performed at

' rated pressure during Test Condition 3. These reactor vessel' injection tests will be used to fine-tune the flow controller since this is the most rigorous operating mode of the system. Testing during this p,hase'will also be done to demonstrate automatic starting from a " cold" standby condition (" cold" is

! operation).

defined as a minimum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any kind of HPCI Af ter the final controller and system adjustments are made, a series of demonstration tests will be performed to provide the necessary information

~

to demonstrate compliance with Regulatory Guide 1.68 power ascension testing requirements.

5.0 ' CONCLUSION The subject, test is simplified by bench calibration of l.

controllers'and tuning the system at the most efficient i test condition. The operability of the HPCI system is not-laffected by-these changes. No Technical Specifications ' change is necessary and safe operation of the plant is not affected. Based on the above, the simplification does not involve an unreviewed safety question. ,

\

4 PSE-SE-t-012 2 of 3

.y ,,

~

f. *

.l 6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATIONS Revisions to Hope Creek's FSAR and Startup test procedures shall be made to reflect the simplification of the High Pressure Coolant Injection test as discussed above.

8.0 ATTACHMENTS None 9.0 SIGNATURES 4 Originator // 9/ f Verifle r ,Mk.$ m. -

9/ ri Group Head (or 6 ) $u $ d k P. E / I 8[

Systems Analysis Group Head O.,. N . P.6 85-Site Engineering Manager h V 9

Date

~

45 N

PSE-SE-Z-012 3 of 3 i

t' W M 5

+4,

TEST NUMBER 14A - SELECTED PROCESS TEMPERATURES SUBSTITUTE WITH TECHNICAL SPECIFICATION SURVEILLANCE OBJECTIVE:

Regulatory Guide 1.68 (Revision 2; August 1978), Appendix A, does not specifically address requirements for measurement of selected process temperatures. Hope Creek 's Power Ascension Test Number 14A, Selected Process Temperatures, establishes low pump speed limits for the recirculation pumps to avoid coolant temperature stratification in the reactor pressure vessel (RPV) bottom head region and assures that idle recirculation loop temperature differentials are within Technical Specifications limits prior to restarting recirculation pump (s). It is proposed that the Technical Specification surveillance procedures associated with the determination of recirculation loop temperature differentials be substituted for this portion of the testing and that testing of the low pump speed limit be deleted.

DISCUSSION:

During initial heatup at hot standby conditions, the bottom drain line temperature and applicable reactor parameters are monitored as recirculation pump speed is slowly lowered to determine the proper setting of the low speed limiter. The coolant temperature in the bottom head region of the RPV may not warm up in phase with the saturation temperature of the coolant when the recirculation flow is at a low value.

Excessive themal stresses, temperature transients and neutron flux scrams may occur when . reactor recirculation flow is increased, sweeping this cooler water into the core.

Therefore, it is necessary that the minimum MG set speed be set at a point which will result in adequate mixing in the lower plenun. Testing at previous BWRs has demonstrated that the minimum pump speed limits currently established are sufficiently above the point where adequate mixing would not occur such that this testing can be deleted.

Following recirculation pump trips, Tes t Numbe r 14A monitors the bottom drain line temperature to determine if temperature stratification occurs in the idle' loop (s). Plant Technical Specifications place limits on the temperature differential between the idle loop (s) and the reactor pressure vessel steam space coolant temperature to prevent restart of idle recirculation loops with significantly cooler water.

Compliance with these Technical Specifications surveillance requirements will satisfy the objectives of Test Number 14A 1

. ~  :

to assure that idle recirculation loop temperature differentials are acceptable prior to restarting recirculation p ump ( s ) . Therefore, the portion of Test Number 14A associated with measuring the' recirculation loop (s) temperature differential can be substituted with the plant Technical Specifications surveillance procedure associated with idle Loop (s) startup.

CONCLUSION:

. Compliance with the Plant Technical Specifications surveillance requirements for idle recirculation loop startup, via existing surveillance procedures will satisfy the objectives of Test Number 14A, Selected Process Temperatures, prior to restart of idle recirculation loop (s). Testing.of the minimum recirculation p2mp speed to determine applicable stratification limits assures that proper performance of the recirculation system occurs'at low pump speeds. Based on the aoove discussion, the proposed change will not affect any safety systems or the safe operation of the plant, and as such, does not involve an unreviewed safety question. Therefore, the plant Technical Specifications surveillance procedures for idle recirculation loop startup can be substituted for that portion of Test Number 14A, Selected Process Temperatures, which measures the recirculation _ loop temperature differential prior to pump restart.

-2

PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-2-009 TITLE: SELECTED PROCESS TEMPERATURE TESTING, TEST NUMBER 14A, PARTIAL SUBSTITUTION WITH TECHNICAL SPECIFICATION SURVEILLANCE REQUIREMENTS AND PARTIAL DELETION SEP 131985 Date:

1.0 PURPOSE The purpose of this Safety Evaluation is to document the results of the evaluation of the proposal to partially substitute Draft Technical Specifications surveillance requirements for Selected Process Temperatures, Hope Creek Power Ascension Test Number 14A, and to delete the remaining portion of Test Number 14A f rom Hope Creek's power ascension test program.

2.0 SCOPE The area of concern for the proposed change is the adequacy of the power ascension test program.

3.0 REFERENCE

1. Regulatory Guide 1.68, Revision 2, August 1978.
2. Hope Creek Final Safety Analysis Report (FSAR)

Chapter 14.

3. General Electric Startup Test Specification, 23A4137, Revision 0.
4. Hope Creek Generating Station Draft Technical Specification.
5. Hope Creek Draft Power Ascension Tes t Procedure TE-SU.zz-161(0), Selected Process Temperature Testing PSE-SE-2-009 1 of 4

i

  • 4.0 DISCUSSION Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, does not specifically address requirements for the measurement of selected process temperatures. Hope Creek's Power Ascension Test Number 14A, Se lected Process Temperatures, establishes low pump speed limits for the l recirculation pumps to avoid coolant temperature stratification in the reactor pressure vessel (RPV) bottom head region and assures that idle recirculation loop temperature differentials are within Technical Specifications limits prior to restarting recirculation pumps. It is proposed that the Technical Specifications surveillance procedures associated with the determination of recirculation loop temperature differentials be I substituted for this portion of the testing and that testing of the low pump speed limit be deleted.

Following recirculation pump trips, Hope Creek's Power Ascension Test Number 14A monitors the bottom drain line temperature to determine if temperature stratification occurs in the idle loop (s). Hope Creek's Draft Technical Specifications Section 3.4.1.4 places a limit of 50*F on the temperature differential between the idle loop (s) and the' reactor pressure vessel coolant and a limit of 145'F on the temperature differential between the bottom head drain line coolant and the reactor pressure vessel steam space coolant to prevent restart of idle recirculation loops with significantly cooler water. Compliance with the Technical Specifications surveillance requirements associated with these limits will satisfy the objectives of Test Number 14A to assure that idle recirculation loop temperature differential are acceptable prior to restarting a recirculation pump. Therefore, the portion of Test Number 14A associated with measuring the recirculation loop (s) temperature dif ferential can be substituted with the plant Technical Specifications surveillance procedure associated with idle loop startup.

During initial heatup at hot standby conditions, the bottom drain line temperature and applicable reactor parameters are monitored as recirculation pump speed is slowly lowered to determine the proper setting of the low speed limiter.

The coolant temperature in the bottom head region of the RPV may not warm up in phase with the saturation temperature of the coolant when the 7

PSE-SE-2-009 2 of 4

recirculation flow is at a low valuo. Excessive thermal stresses, temperature transients and neutron flux scrams may occur when reactor recirculation flow is increased, sweeping this cooler water into the core. Therefore, it is necessary that the minimum MG set speed be set at a point which will result in adequate mixing in the lower plenum. Testing at previous BWRs has demonstrated that the minimum pump speed limits currently established are sufficiently above the point where adequate mixing would ~ occur such that this testing can be deleted at Hope Creek.

5.0 CONCLUSION

Compliance with the plant technical specifications surveillance requirements for idle recirculation loop startup, via existing surveillance procedures will satisfy the objectives of Hope Creek Power Ascension Test Numbe r 14A, Selected Process Temperatures, prior to restart of idle recirculation loop. Testing of the minimum recirculation pump speed at previous plants to determine applicable stratification limits provides assurance that proper performance of the recirculation system occurs at the currently set low pump speed

' limits. A change to Hope Creek's Technical Specifications is not required as a result of the change. Based on the above discussion, the proposed change will not adversely affect any safety. systems or the safe operations of the plant, and as such, does not involve an unreviewed safety question. Therefore, the plant Technical Specifications surveillance procedures for idle recirculation loop startup can be substituted for that portion of Test Number 14A, which measures the recirculation loop temperature' differential prior to pump.

restart and the testing of the low reactor recirculation pump speed limit can also be deleted.

'6.0 DOCUMENTS GENERATED None PSE-SE-2-009 3 of 4

.+

7.0 RECOMMENDATIONS Revisions to Hope Creek's FSAR to reflect the substitution of Startup test procedures shall be made to surveillance procedures required by the plant Technical Specification for the startup of an idle reactor recirculation loop for the applicable portion of Selected Process Temperature Testing, Tes t Number 14A, and the deletion of reactor recirculation pump minimum speed testing f rom Tes t Number 14A.

8.0 ATTACHMENT None 9.0 SIGNATURES Originator h CL Me J m e 6. 9/nA'r p D'a t h .

Verifier b,@"v.A_t;f'1 4/*/h' Group Head ( r 'bSE) I S u m h te l PE Y/II/ V Systems Analysis Group Head d.N. T[ 913[85 Site Engineering Manager [ bd sbP Date 413!?f '

PSE-SE-Z-009 4 of 4 i

.e

1 TEST NUMBER 9 - LPRM CALIBRATION TEST SIMPLIFICATION - REDUCED NUMBER OF TESTS OBJECTIVE:

Regulatory Gu ide 1.68 (Revision 2, August 1978), Appendix A, paragraph S.y requires that incore neutron flux instrumentation used to calculate thermal power level be calibrated as required and its operation verified. Test Number 9, Local Power Range Monitor (LPRM) Calibratlon, performs the required calibration of the LPRM instrumentation. In addition, the test currently plans to verify LPRM response during control rod movement at heatup conditions or at test condition 1. It is proposed to delete the heatup or test condition 1 testing of the LPRM response.

DISCUSSION:

Calibration of the LPRM system at several test conditions satisfies the objectives of Regulatory Guide 1.68, Appendix A, paragraph 5.y. The additional testing of Test Number 9, to verify the LPRM response during heatup is not required by the regulations and is often impractical to perform at these conditions. During heatup, the neutron flux level is low enough such that some of the LPRM detectors have signals which are less than the downscale value. Therefore, indication of the LPRM response is not available for the operator to verify the response. During control rod movement at power levels where sufficient indication of LPRM signals exists, indications of LPRM response can readily be observed by the operator. This response is often used as a confirmatory indication of control rod movement and provides sufficient demonstration of the LPRM

. system response when combined with the required calibration procedures.

CONCLUSION:

LPRM calibration during Test Number 9 demonstrates the acceptable performance of the LPRM system and satisfies the objectives.of Regulatory Guide 1.68, Appendix A, paragraph 5.y. The proposed change to delete the verification of LPRM response does not affect any safety systems or safe operation of the plant and therefore does not involve an unreviewed safety question. Test Number 9, LPRM Calibration, can therefore be simplified by deleting the LPRM response checks during heatup or at Test Condition 1.

1 1

I I

i PUBLIC SERVICE ELECTRIC AND GAS COMPANY

. HOPd CREEK PROJECT SAFETY EVALUATION No. PSE-SE-Z-013 TITLE: SIMPLIFICATION OF THE POWER ASCENSION LPRM CALIBRATION TESTING - TEST NUMBER 9 g,,,, SEP 19 HMS 1.0 PUR POS F, The purpose of this Safety Evaluation is to document the results of the evaluation of the proposal to simplify the LPRM calibration testing procedure during the power ascension test program.

2.0 SCOPE The LPRM system is part of the incore neutron monitoring system.

3.0 REFERENCES

1. FSAR Chapters 7 and 14
2. GE Startup Test Specification No. 23A4137, Revision 0
3. NRC Regulatory Guide 1.68, Revision 2, August 1978
4. Hope Creek Generating Station Draft Technical Specifications
5. Hope Creek Generating Station Draft Power Ascension Procedure, TE-SU.SE-111(O)
6. Preoperational Test Procedure SE-3, Power Range Neutron Monitoring System 4.0 DISCUSSION Calibration of the LPPM system at'several test conditions satisfies the objectives of Regulatory Guide 1.68, Appendix A, paragraph 5.y and Appendix C, paragraph 4.c. The additional testing of Test Number 9 to verify the LPRM response during heatup or Test Condition 1 is not required by the regulatory guide and can be accomplished by other means. During control rod movement at power levels where the flux is within the operating range of the.LPRMs, indication of LPRM response around the selected control rod can be readily observed by the operator on the four rod display .

PSE-SE-Z-013 1 of 3 O

9

This response is often used as a confirmatory indication of control rod movement and provides sufficient demonstration of the LPRM system response when combined with the calibration procedures. Also, indication is available to alert the operator to downscale readings of LPRM detectors. Installation and preoperational testing requirements are sufficient to ensure that the LPRM system is connected correctly and ready for operation.

The plant's high flux scram protection will not be compromised since the IRM system monitors neutron flux level until the APRM channels are verified to be operational and will. provide a scram trip output if required. Since the plant's high flux scram protection will be maintained and'since there is sufficient alternate means for the verification of LPRM flux response, the requirement of Test Number 9 for LPRM flux response verification during heatup or in Test Condition 1 can be deleted.

5.0 CONCLUSION

LPRM calibration during Test Number 9 demonstrates the acceptable performance of the LPRM system'and satisfies the objectives of Regulatory Guide 1.68, Appendix A, paragraph 5.y and Appendix C, paragraph 4.c. The proposed change to delete the verification of LPRM response does not adversely affect any safety systems or the safe operation of the plant and, therefore, does not involve an unreviewed safety' question. Test Number

~9, LPRM calibration, can, therefore, be simplified by deleting the LPRM response checks during heatup or Test Condition 1.

6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATIONS The. FSAR and startup test procedures shall be revised to reflect the simplification of the power ascension LPRM calibration testing by the deletion of-the LPRM response check during heatup or Test Condition 1.

PSE-SE-2-013 2 of 3 e

. 8.0 ATTACHMENTS None 9.0 SIGNATURES Originator _

-M-I

)/).- \k,

?

Verifier (,/ , .. 9491(

l)a t Group Head [or' SE) 9 @%

Systems Analysis Group Head bL . / [f Da t'e Site Engineering Manager f];),k.,d.V

_..a

@ al., C4 v

'Da t s' PSE-SE-Z-013 3 of 3 i

- - - - - , - - - . . -s. , - r- - .

' TEST CONDITION 4 - NATURAL CIRCULATION OPERATION DELETE TEST CONDITION q OBJ ECTIVE :

Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraph 5.o requires that appropriate consideration should be given to testing at the extremes of possible operating modes for facility systems. It clarifies this. requirement by stating that testing under simulated' conditions of maximum and minimum equipment availability within systems should be accomplished if the facility is intended to be operated in these modes. Test

-Condition 4 defines the region of the power / flow map

.(Attachment 1) at natural circulation near the 100% rod line.

Testing of core performance and control system response is currently planned to be perform 5d in this region. It is proposed that all testing in Test Condition 4 be deleted.

DISCUSSION:

Operation at natural circulation conditions is not an intended mode of operation for a BWR. Natural circulation conditions can only be reached as the result of an abnormal operational transient (two recirculation pump trip). Hope Creek's Draft Technical Specifications (Section 3.4.1.1) require that with no reactor coolant system recirculation loops in operation, the operator must take immediate action to reduce thermal power to less tha'n or equal to that allowed by the Core- Thermal Power to Core Flow Map within two hours. This will reduce the core therna'l power to a level below that of Test Condition 4. .

Actionsjmustalsobetakentoplacetheunit in at least

-Startup:,within six hours and in. Hot Shutdown within the next six houts. Extensive special testing at natural circulation conditions has previously~ been pe.rformed at other BWRs: Peach Bottom 2, Vermont Yankee, Caorso, Leibstadt, and Browns Ferry.

These tests have thoroughly characterized the performance of the BWR at natural circulation conditions and confirm that plant operation under these conditions is acceptable.

CONCLUSION:

Testing of plant systems is performed over a wide range of operating conditions representing the extremes of possible e operating modes intended for the piant. These tests meet the objectives of Regulatory Guide 1.68, Appendix A, paragraph 5.. o . The proposed test change does not affect any safety system' or safe o' peration of the plant. Thus, it does not involve an unreviewed safety _ question. Test Condition 4 testing can therefore be deleted from Hope Creek's Power-Ascension Test Program (FSAR Table 14.2-5). g 1

l ATTACHM_ENT 1

i. -

necuaa essa:V 23A4137 BUSINESS O NaATMH85 GENERAL @ ELECTRIC ,, ,0. 6.,

cav 0 l -

kW'4 qs-w y y-tag e, i

  • 1 13 =

110 =

A. MATURAL CIRCULATION O t# "*

3. Iglesetsute RSC8RCULATION PutAP WWED C. AseALYTICAL LOIuWI LRet? OF *In.yg3 GAASTER POtuGR PLOur CONTROL 33 -

E AsaALYTN:AL UPPER LMes? OF GAASTWR POUWER PLOut CONTROL g .

,. /

/ p++

l N

/

f

  • Ig g

/

a -

) t Tes #

A 'f.

SS -

=

TC4 b

- f TC3 30 =

ATC2 teessiaeves POustR LINE 3- TYPtCAL STARTt# PATH ll TC1

/

~

CAVITATION REGION

, e , t , , , , ,  ;

e to so ao ao ao as a ao a too sio PeRCawT core PLOur  !

Figure 1.

Operational Power / Flow Map C Mao setA ( ASW. te/stl

g 3-PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-Z-008 TITLE: DELETION OF NATURAL CIRCULATION OPERATION TESTING, TEST CONDITION 4, FROM THE POWER ASCENSION TEST PROGRAM SEP 201985 Da te :

1.0 PURPOSE

' The purpose of this Safety Evaluation is to document the results of the evaluation performed to ensure that deletion of Natural Circulation Operation Testing, Test Condition 4, from the Hope Creek power ascension test program will not adversely affect reactor safety.

2.0 SCOPE The area of concern for this proposed change is the adequacy of Hope Creek's power ascension test program.

3.0 REFERENCES

1. Regulatory Guide 1.68, Revision 2, August 1978
2. Hope Creek Final Safety Analysis Report (FSAR)

Chapter 4 and 14

3. General Electric Startup Specification, 23A4137 Revision 0
4. Hope Creek Generating Station Draft Technical Specifications
5. NRC letter f rom Roger J. Mattson dated 2/27/84, regarding Board Notification - BWR Core Thermal Stability 4.0 DISCUSSION Regulatory Guide 1.68 (Revision 2, Augus t 1978), Appendix A, paragraph 5.o requires that appropriate consideration should be given to testing at the extremes of possible operating modes for facility systems. It clarifies this requirement by stating that testing under simulated conditions of maximum and minimum equipment availability PSE-SE-2-008 1 of 3

within systems should be accomplished if the facility is intended to be operated in these modes. Test Condition 4 defines the region of the power / flow map (Attachment 1) at natural circulation near the 100% rod line. Testing of core performance and control system response is currently planned to be performed in this region. The NRC letter dated February 27, 1984 (attachment No. 2),

addressed the acceptability of operating in this condition. It is proposed that all testing in Test Condition 4 be deleted.

Operation at aatural circulation is not an intended mode of operation for a BWR. Natural circulation conditions can only be reached as the result of an abnormal operational transient (two recirculation pump trip).

Hope Creek's Draft Technical Specifications (Section 3.4.1.1) require that with no reactor coolant system recirculation loops in operation, the operator must take immediate action to reduce thermal power to less than or equal to that allowed by the Core Thermal Power to Core Flow Map within two hours. This will reduce the core thermal power to a level below that of Test Condition 4.

Actions must also be taken to place the unit in at least Startup within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in Hot Shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Startup testing of BWRs with the same size reactor and orificing as Hope Creek (Susquehanna 1 and 2, Hanford, and LaSalle 1 and 2) has included natural circulation testing. In addition, extensive special testing at natural circulation conditions has previously been performed at other BWRs: Peach Bottom 2, Vermont Yankee, Caorso, Leibstadt, and Browns Ferry. All these tests have thoroughly characterized the performance of the BWR at natural circulation conditions. The cycle 1 stability analyses results, which characterize the thermal / hydraulic performance of the Hope Creek Reactor Core at natural circulation conditions', are presented in FSAR Chapter 4.4. The results show that there are no-safety concerns associated with the natural circulation mode.

5.0 CONCLUSION

S Testing of plant systems will be performed over a wide range.of operating conditions representing the extremes of possible operating. modes intended for the ' plant.

These tests meet the objectives of Regulatory Guide 1.68, Appendix A, paragraph 5.o. The proposed test change does not affect any safety system or safe operation of the PSE-SE-2-008 2 of 3 l

l

plant and a Technical Specifications change is not required. Based on the above, the proposed test change does not involve an unreviewed safety question.

Therefore, Test Condition 4 testing can be deleted f rom Hope Creek's power ascension test program (FSAR Table 14.2-5).

6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATIONS Revision to Hope Creek's FSAR and Startup Test Procedures shall be made to delete Natural Circulation Operation Testing, Test Condition 4, from the power ascension testing program.

,8 . 0 ATTACHMENTS

1. Hope Creek's Operational Power / Flow Map.
2. NRC letter ~(Reference-5).

9.0 SIGNATURES Originator _ fba. ,e _ &__ f Ar. 9/.2s/rs-Verifier _, , d .>c; M

!j ll L ~l ' ' } n / Da f e Group Head (o'r SSE)

$m fdE P.E . 9l2e/ g Da te '

Systems Analysis Group Head b.d, N hb. 9/fo[86 Site Engineering Manager Cdd _A T U/ /CS~

'Dat6 PSE-SE-Z-008 3 of 3

ATTACMMENT 1 NUCLEAR ENeser i

i

, SUSINESS OPetafl0NS GENERAL $ ELEC7AlC 3 5 137

.e o. 6 asv 0 1 j

i IO r

--- r im=

1m=

  • i 110m A, -,

SW " 8. asATustAL csACULAThess 3 Reesseseuel 28CIRCULAfl010 PUID even '

C. l AfaALYTICAL LaugWt Lee TOF g . S. MASTWA POWWA PLOW G3 BET 1004 s. .

AdeALYTICAL Uf84ft Latet? 0F

  • enesTRA P0tuGA PL0er C0erTWOL 1 1

/s ,

g. .

= - N

.l /w -

? 38 e- b

,,0* / \

TCS 3

TCE .

3m esnotesues posete Liese TYPEAL STARMM PATM CAWITATIORI

, P Atescoe e a a e se as a

e '

as e es as a

se a e ses ito PEACWrf COAS PLdM Figure 1.

Operational Power / Flow Mgy.

e

.-e . - - - - - - . , -

, - . . - , _ , . , , , , . . . . , , . , _ , _ . , , _ , _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _m_,,

I

. ce+;5scss.= 5

((

i

  • .'**v.,'.

t,

. , . . i f* i . i ,*. L.*:i r13 371. 713

, s sM .'1UCt.!A.4 .4 E 3ULA 70.W CO*.f.'.ll53;ce r.._g .....

..m-meros 3. c. :: sis

,agg 2 7 'i4

" I ** *, TCT! .?;.S : -

T -

i-1

,_ . ;-gy.

  • f'stston of i.? dhsing '

71t:M: *

.tager J. Mattson. Ofrectar 3tytaten of Systems Integratfon SU3J!CT:

3 CARO NOTIFICATION - SWR CCRC nfERMAL HYORAULIC $

?.

fins *se "e iffestisn .I De staff has teen inform Sy General t1ectric

31) of -ecent thermal hydesulte stanf tf ty tests at a foreign co(actor, a Mf1ts; satse reac:or with relatively htgn sewer density, uhtch demen.

1 strated the occurrence af limit cycle neutron flua csct11 l 4tt7er s t  :

natural ctreulatten and severst gercent above the rat

  • , f -

ocerating stata testad and the oscillations were osserva41e .on the A P.t.w i

1 s and su;:ressed through control rod insertion.

, than those previously observed fn other(2 sta6111ty ant.

tests.t nation of the detaf f ed data of tats test shoved !?st some LP'.'s oscillated out of shase with the APRM signal ar.d at sn spetitude as great as six sfcas the core average messured by the A7.t"1. ,

These data are important since they confirm the poss but 3tates. net obsersed in stat 14r stability tafts ;erfsrmed in the J 'attad D.e staff f s ;rtsently working wtth ac31Icants and wf11 the l'IR Cwners Grous to revf ew the standard Technical !cect Pf estica assure that they propefly protect against the Sotential fer insta=

httftfes.

If chantee are to be recutred, they duld fallcw staff C.184 reefew. procedures for generic Technical Spect fication changes, inciacin 2.

Relevamew and *sterfalf tr Stabt1f ty tests on a !!!R oscilittfens can occur w/4 reactor defenstested that ifmit cycle ested r:4 Ifne at naturst cir:ithin -er91ssable ocerating saace tal tw 13e 21stian F1:w. The hi-h zewer 'evs' (120 :er: ant! scetn erotsc .ise which is 3ssed sn L"M sf psis uo M no r.ecessaril/ :revent vfolttfon af celtical hea: ff us (C4) limits if local 19stabf110tes occur.

Con set: 1. Sci ,est, ,1t :c.:3  !

X.231.it i

+ '

~~ .h . -

i

)

, a m en 495; s- 3-a4;.2:c4Pm

~.25<.~. .~

/. ic99355C553= 6

[ . -

2.1. Itseneet 7 . ii 2 ' .::*.

are ssf;st art <rcen is iecrease fi **e*'"t* * / **14;

.,f u ei;*se ::nor : ansi ty sne 41*311i*se 13: Ita:t * * :/ .

23

  • scisted of:S  ::r:J:13n:s 49* :* !s 4

e ::a:!: tens ' vat s is:^;el::iss? Irs 9tv'F1 1 eI't* ^.aI :t* :tt :ars.

escer than 31: Sal asst 114:fsas wf u!acat tr.stast!!:tes t ::s f:rti;s tas:

see net ef eer. . 611 *.41 1 24.?st 'es in : nase

ntti care stantiIty sehavior can se :e::ee =rs. ,

4 afsted and showing ese14fnee tatt local estillattsasthe nonstaff teltavesofthat characteristtc the 3asw Intersetfen the Ape .

si;nf ficante to tasse %svf at aftignals may escur is relevant to all SWE de c.9ersc: eristics as tyof *feg sy fner sewer donat:y and lower casetag as 524/4. WR/1 Inc gua/g test;ns.

2. t!st**espea sf *es- S'fd s tbn o

ne testare waich demonstrated zu: sd :stse .v'uat 't local 3e mel hytetulfe esst11st'ons

  • e &#t.*s can 2 tur.

sessi:llity sf local ssettisti:ns accur-ing fa the stint .vntch .Sf3 entses the

stght sfgnetsnot ane/arse statted t.7RM utscale 3y mealarme. 4etrat:r without went torfag of '.psy 1 hfgn e Ietal asst 11stion c:uld gems it 13 unclear befers at thisby detection tfas annewofertSer ) .

ustag current eenf tsrf ag 2rstedures. -

State we cann ",

Itattfag tegnitude for such an estillation, we can.et seedtet the that CHP Ifstts would not to esteeded. net be certate we will eseetne SWA l'echafcal Seest ficatians to assure that theyIt is for sesvita adeguate assurance that osersting regtens of :stans castable :tegnf tude are detected and suppressed. -

~

j i The staff eens1sdes .that the n'em tafernetten fyes tas forefga statt11ty tests does not pese sa fausediate safety concern for continued SWR operatten prise to orderly emeestnetten and sessible change of *echafcal igetifftatiens for the esasons which fallew:

(s) Current 3WR Technfe41  ;

eserstten uctor tsadf(4esiftsattens eas of natural ctreulatten elaceorrestrfc:fons single on less low with operation statt1fty susa sortfn thatitthe frequency very low. of oeeraston fn regions doetgas fty aest eserating reestore are sufffatently stableIn additton, tre care

  • fa the'Iese statte ser'stssihte eserating regimes. . Even tft power fevel AMM saram protesties would be cases. j I adequate in (4)
  • he sainitude of ther=tal hydraullt instant 11ty induced nousess flua essf11att:ns f s considerstly ht yter than the esaillations .

the fuel car 41 stes c:nstant.in the averste clattf at heat flua totause of General flestrfs Caetany

-, -..-.~ - ,- - -- -- --- -,, ,_,.--.. -. --. -- __ _ -__ _-- --- -

-...~.a .. .- - -- --  ;

~~- -

.. -= -  :--i . a

i. ....s
  • k

?. .

3. at tfsenhut _ *

, g 3. 7 ;g

,cantat* ens !De Pt cast waere the sutran dius 3scf1? stas us ts.120 sercent of ested (APS trfs :3191) t adicata test,- :*a 5 sortent of its ested value.surfsta heat flus has a :eak estillatten sm .

satsts ta safety Dus alarge 1!.st:s even if very,  ::sstssrsole neutn:n 9ar flu: tfn esstliatfons occur.

. (c) The esat11ations are reedt' and can be easily suppressee, by inserting control reds. detecta (d)  ;

Geners) Electrfe Cemoany is in the 2 recess of provfdfag to all for thernal nyeraulte instantt tttes and ren the act i should be taken sa suppress such estillations If they sim14 occur.

pendfag WR actfans.eners have teen .1tada

  • aware of the pretles tad of

.' 4 Relaston to 7eelects "~

. ~

i .

The foreign stahtitty test retults relate to all WA reseters. It i -

is recommended that aopropriate boards he notiffed of the new .

information and of the staff's slans to wert with the W4 app 1f.

cants all NRsand 1teensees properly protect to agatast assure the thatpotentia the Technical Specificattens for r instabilf ties.

di -

A

!sger J. M tson Ofrector ~

Olviston, of Systems Integration 1

9 i I l

l l

r.-,v,, , , , . , . , , _ _ - . , - , , -.gw,.y,.-g,,, gr.ew-y,-,,. q, ,,.y me m m v , e

AW& neat 1 (7 Pages)

HCGS FSAR 10/84 above the reactor pressure to simulate the largest expected pipeline pressure drop. This CST testing is done

__ur__

to demonstrate general system operability,2.? frr

___&__i3__

. _ . . . . . . , ..___ __.._______ _24.._u___

Reactor vessel injection tests follow to cr 7 1 -te th; controller adjustments and to demonstrate automatic starting from a cold standby condition. " Cold" is defined as a minimum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any kind of RCIC operation. Data will be taken to determine the RCIC high steam flow isolation trip setpoint while injecting at rated flow to the reactor vessel.

After all final controller and system adjustments have been determined, a defined set of demonstration tests must be performed with that one set of adjustments.

Two consecutive reactor vessel injections starting from cold conditions in the automatic mode must satisfactorily be performed to demonstrate system reliability. - Following these tests, a set of CST injections are done to provide a benchmark for comparison with future surveillance tests.

After the auto start portion of certain of the above tests is completed, and while the system is still operating, small step disturbances in speed and~ flow command are input (in manual and automatic mode respectively) in order to demonstrate satisfactory stability. This is to be done at both low (above minimum turbine speed) and near rated flow initial conditions to span the RCIC operating range.

l A demonstration of extended operation of up to two hours (or until pump and turbine oil temperature is stabilized) of continuous running at rated flow conditions is to be scheduled at a convenient time during the startup test program.

Depressing the manual initiation pushbutton is defined as automatic starting or automatic initiation of the RCIC system.

d. Acceptance Criteria Level 1:

l

1. Following automatic initiation, the pump discharge flow must be equal to or greater than rated flow  ;

as specified in Section 5.4.6 within the time i l

specified by the GE startup test specification.

14.2-165 Amendment 8 l

l

x HCGS FSAR 5/85

1. By flow injection into a test line leading to the condensate storage tank (CST), and-
2. By flow injection directly into the reactor vessel.

The earlier set of CST injection tests consist of manual and automatic mode starts at 200 psig and near rated reactor pressure conditions. The pump discharge pressure during these tests is throttled to be 100 psi above the reactor pressure to simulate the largest expected pipeline pressure drop. This CST testing is done to demonstrate general system operability, end-4ev i

--ui 7 mme+ emm*,mii

, -ajuet ente. I Reactor vessel injection tests follow to ^ perform

^=r'-*^ *h" l

controller adjustments and to demonstrate automatic starting from a cold standby condition. " Cold" is defined as a minimum 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> without any kind of HPCI operation. Data will be taken to determine the HPCI high steam flow isolation trip setpoint while injecting at rated flow to~the reactor vessel. Depressing the manual initiation pushbutton is defined as automatic starting or automatic initiation of the HPCI system.

After all final controller and system adjustments have-been determined, a defined set of demonstration tests must be performed with that one set of adjustments.

Two consecutive reactor vessel injections starting from cold conditions in the automatic mode must satisfactorily be performed to demonstrate system reliability. Following these tests, a set of CST injections are done to provide a benchmark for comparison with future surveillance tests.

i After the auto start portion of certain of the above

'"v, rests is completed, and while the system is still operating, small step disturbances in speed and flow command are input (in manual and automatic modes respectively) in order to demonstrate satisfactory stability. This is to be done at both low (above minimum turbine speed) and near rated flow initial conditions to span the HPCI operating range.

A continuous running test.is to be scheduled at a convenient time during the startup test program. This demonstration of extended operation should be for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or until steady turbine and pump conditions are i

14.2-167 Amendment 10

HCGS FSAR 10/84 reached or until limits on plant operation are encountered.

d. Acceptance Criteria Level 1: l
1. Following automatic initiation, the pump discharge flow must be equal to or greater than the rated flow, and within the time specified in Section 6.3.2.2.1.
2. The HPCI turbine shall not isolate or trip during automatic or manual start tests.

Level 2: l

1. The speed and flow control loops are adjusted to meet the decay ratio specified in the GE startup test specification.
2. The turbine gland seal system is capable of preventing steam leakage to the atmosphere.
3. The delta-pressure setpoints for HPCI steam supply line high flow shall be calibrated to technical specification requirements using actual flow conditions.
4. In order to provide overspeed and isolation trip avoidance margin, the transient start speed peaks must not exceed the requirements of the GE startup test specification.

14.2.12.3.14 / h Water Level Reference Leg Temperatures

a. Objectives

. stablish low speed limits for the recirc to avoid coolant temperature l stratification in essure vessel 1 (RPV) bottom head region '

l .

l .

l l 14.2-168 Amendment 8 i

I i

HCGS FSAR 10/84 l, /. To ensure that the measured bottom head drain temperature corresponds to bottom head coolant temperature during normal operation.

J,/ To measure the reactor water level instrument reference leg temperature and recalibrate the affected indicators if the measured temperature is different than expected.

b. Prerequisites The plant is in a hot standby condition. System and test instrumentation have been installed.
c. Test Method ,

ng initial heatup at hot standby conditions, bott ain line' temperature and applicabl ctor parameters monitored as the recir on pump speed is slowly ed to deter the proper setting of the low speed limi parameters above are also monitored durin O to determine if the idle 1 n

rature stra circulation pump trips and to assure that i ation occurs in p-to-bulk

/

cool emperature differentials are within nical -

cification limits prior to restarting the pump The bottom drain line temperature and applicable parameters are monitored when core flow is 100% of rated flow.

A test is also performed at rated temperature and pressure under steady state conditions to verify that the reference leg temperature of the level instrumentation is the value assumed during initial calibration. Recalibration will be performed if necessary, j d. Acceptance Criteria l

bev.Li : l  !

' l

1. The reactor r tion pumps shall not be started unless the loo delta-temperatures and steam done to bottom drain retures are within the technical specification O

14.2-169 Amendment 8 ,

l____ l l l l l l l l l (1) Test conditions refer to plant coMitions l l ITESTj lOPEN lHEATI l l l l l on Figure 14.2-4 l l l Lo.1 TEST FAME l VESSEL l UP l 1 l 2 l 3 l 4 l 5 l 6 lhaRRANTY l l(22)) l l l l l l l l l (2) Perform Test 5, timing of 4 slowest control l l l l l l l l l, j l ru.ls, in conjunction with expected m: rams l 1 l Chemical an! Radiochemical lX lI lX l lX ll lX lI l l l 2 l LMiation Meastrument lX lX lX l lX l l lX ] l (3) Dynamic System Test Case to be completed l 3 l Fuel loating lX l l l l l l l l betwen test conditions 1 and 3 l

l 4 l Full Core Shutdown Margin lX ] l l l l l l l p,yggg) l 5 j control bd Drive lI lX lX(2) llIIII lXIII l l lx(2) l l (4) w n _ , _ m . . , _, m ; y. m. m s1 l 0 l SRM Performance lI l l l l l l l [ l ct.d t i:M l 6 j IRM Performance l lX lX l l l l l l l l T l lpm Calibration l lX lX l ]X l lx l (5) Between 80 arvi 90 percent thermal power, l l l 10 l Arm Calibration l lX lX lE lI l lE lX [ [ and near 100 percent core flow l 11 l Process Camputer lX lX lXI33 l lI l lX l l l l 12 l RCIC l lX jX l l l l l l l (6) Max W Raout Capability & Recirc Pump l 13 l TPCI l lX l l lx l l l l l Runback must have already been completed l 14 l Selected Process Temp l lX l l lI l1r6M l @l l l 14 l Water Invel Ref Iag Temp l lX l l [X l l lx l l (7) Reactor power between 80 and 90 percent i 15 l Sy' tan Expansion lX [E [X l lI l l lE l l l 16 l T1P thcertainty l l l l lX l l lE l l (8) Raactor power between 45 and 65 percent l 17 l Core Mrformance l l lX lI lX fb lX lx l I l l 18 l Steam Production j l l l l l l l ] 3 l (9) Reactor power between 75 and 90 percent l 19 l Core Pwr-Void Mode Resp >nse l l l l l lT- lX ] l l l 20 l Pressure Regulator l l lX lI lT lA. lI lX l l ( 10 ) At maximum power that will not cause scr am l 21 l Feed Sys-Setpoint Changes l lX lX lE lx lb lI lE l l l 21 l Feed Sys-Ines FW Heating l l l l- l l l jX(5) l [ (11) Perform between test conditions 1 and 3 l 21 l Feedwater Pump Trip l l l l l l l lE(6) l g l 21 l Max W Runout Capability l l l l l l l lX(II l l ( 12 ) Reactor power between 40 and 55 percent l 22 l Turbine valve Surveillance l l l lg(9) [g( 10 ) g l l 23 l MSIV Functional Test l lX lIIIII]X l II2 3lX(8) 1 ll lIII3Il l l (13) Reactor power between 60 and 85 percent l 23 [ MSIV Full !aalation l l l l l l [ [x l l l 24 l Czlief Valves l lX l X(20)lx ]xt20) l lg(20)l l ( 14 ) setween test conditions 2 and 3 l 25 l Turbine Trip & Loaw! l l l l X( 15) l g( 16) l l l ggggggg g j l Rfjection l l l l l l l l l (15) Cenerator load rejection, within bypass l

l 26 l $ hutdown Outside CRC l l l lX l l l l l l valve capacity l 27 l Cecirculation Flow Control l l l lE(I4Il l lX(18)l l l l 28 l Recirc=One Pump Trip l l l l lI l l lx l l (16) Reactor power between 60 and 80 percent l 28 l RPT trip-Two Pumps l l l l lX(19)l l l l l at core flow > 95 gercent - turbine trip l 28 l Rectre Systen Performance l l l lx lX lse- l lx l l l 28 l Recirc Pump Runback l l l l lx l l l l l (17) kad rejecticn l 28 l Recirc Sys Cavitation l l l l lX l l l l l l 30 l Loza of Of f site Pwr l l lx l l l l l l l (18) Between test corrtitions 5 and 6 l 31 l Pipe Vibration l lI lX lE lx l l lX l l l 29 l Recirc Flow Calibration l l l l lI l l lX l l (19) >5 0% power and _>95 core flow, and performed l 32 l RWCJ l lX l l l l l l l l before Turbine Trip & Ioad Rejection l 33 l RNR l l l lX l l l [X(2IIl [

l 34 l Drywell & Steam Tmnel l lI lX l lE l l lX l l (20 ) Check SRV set points during ma jor m-ram l l Cooling l l l l l l l l l l tests HOPE CREE K l 35 l Gaseous Radusste l l lE l [x l l lx l l GENERATING STATION l 38 l SACJ Performance l l l l lX l l lx l l (21) Performed during cooldown frces test FINAL SAFETY ANALYSIS REPORT l A0 ] Confirmatory In-Plant Test l l l lX l l l l l condition 6 l

FSAR 3/7 (22) The test number correlates to FSAR Section 14.2.12.3.x there x is the indicated test T WMM M MMN numbe r.

FIGURE 1424 A_ :10.05/8'

r r, l

  • -f 1 s

~

' 110 ~~ A. NATURAL CIRCULATION ,

. . B. MINIMUM RECIRCULATION PUMP SPEE .

C. ANALYTICAL LOWER LIMIT OF -

MASTER POWER FLOW CONTROL 90 ~~ D. ANALYTICAL UPPER LIMIT T.c..e OF MASTER POWER FLOW CONTROL ,oo

/ -

80 -

T.C..$ g*

k '

T.'c.e 3 70 ~

~

,4 A i (b 60 ~

~

s' 1 # //

14J /

3 A l

~ ~

o S O

d o Q. i, n00

, gog - .

> 40 ~

~ '

2 w -

30 ~

~

NT.c.. -

6 ~

iminewn unt -

IL 20 ~ , T.c..

. CAL STAATUP PATM N .

, mEmou ur f I f I f I f

d. 0 ~. i a i 6 e i i i e i s 0 10 20 30 40 50 60 70 80 90 100 11 0 PERCENT CORE FLOW TEST CONDITION (TC) REGION DEFINITIONS TEST CONDITION NO. POWER. FLOW MAP REGION AND NOTES *

'1 BEFORE OR AFTER MAIN GENERATOR j SYNCHRONIZATION BETWEEN 5.% AND 20.% THERMAL i POWER 4VITHIN f10.% OF M4 SET MINIMUM I

OPERATING SPEED LINE IN LOCAL MANUAL MODE.

2 AFTER MAIN GENER ATOR SYNCHRONIZATION BETWEEN THE 45.% AND 75.% POWER ROO LINES BETWEEN M.G SET MINIMUM SPEEDS FOR LOCAL MANUAL AND MASTER MANUAL MODES THE LOWER POWER CORNER MUST BE LESS THAN BYPASS V ALVE CAPACITY.

3 BETWEEN THE 45.% and 75.% POWER ROD LINES -

CORE FLOW BETWEEN 80% AND 100.% OF ITS RATED VALUE.

Deleted.

5 WITHIN +0. 4% OF THE 100.% POWER ROD LINE -

WITHIN 5.% OF THE ANALYTICAL OF THE LOWER LIMIT OF MASTER FLOW CONTROL 6 WITHIN +0,4% 0 F RATED 100.% POWER - WITHIN

+0,-5% OF RATED 100.% CORE FLOW RATE.

HOPE CREEK GENERATING STATION FINAL SAFETY ANALYSIS REPORT OPER ATIONAL POWER / FLOW MAP l

FIGURE 14.2 4 Amendment 1,8/83

s l ,

w 9

Additions To Pace 1.8-42 in FSAR Appendix A . .

Paragraph 5.0 No Startup tests will be performed with the reactor operating at Natural circulation conditions since this is not an intended mode s of operation for the plant as directed by Technical Specifications, 5

i

\

W . \ .

4 l

-t 1

s

. _ . ,. - r, ~ , .,

Retrieved from "https://kanterella.com/w/index.php?title=ML20135H777&oldid=3375364"