Text

Forwards Balance of Remaining Detailed Justifications for Power Ascension Test Mods for Consideration.Ge Technical Analyses,Util Safety Evaluation for Tests 1,3 & 5 & marked-up FSAR Pages Encl
	ML20138A656
Person / Time
Site:	Hope Creek
Issue date:	12/09/1985
From:	Corbin McNeil; Public Service Enterprise Group
To:	Adensam E; Office of Nuclear Reactor Regulation
References
	NUDOCS 8512120075
	Download: ML20138A656 (67)
	v • d • e

- - - -

% war m =4J-Public Service Electric and Gas Company Cerbin A. McNeill, Jr.

Public Senace Electric and Gas Cornpany P.O. Box 236 Hancocks Bndge, NJ 08038 609339-4800 Y

Vice President -

e Nuclear December 9, 1985

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Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission 7920 Norfolk Avenue Bethesda, Maryland 20814 Attention:

Ms. Elinor Adensam, Director Project Directorate 3 Division of BWR Licensing

Dear Ms. Adensam:

POWER ASCENSION PROGRAM CHANGES HOPE CREEK GENERATING STATION DOCKET NO. 50-354 Public Service Electric and Gas Company (PSE&G) hereby submits the balance of the remaining detailed justifications as identified below, pertaining to power ascension test modifications for NRC consideration.

1.

Test Number 1 - Chemical Radiochemical Test Simplification 2.

Test Number 3 - Elimination of Fuel-Loading Chambers During Fuel Loading 3.

Test Number 5 - Control Rod Scram Time Testing at Hot Standby with Full Reactor Scram Control Rod Time Testing Attachment I consists of General Electric Company's technical analysis and PSE&G's safety evaluation for each 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 Operation Review Committee (SORC) has reviewed each evaluation and concurs with the conclusions reached therein.

Attachment II consists of marked up FSAR pages applicable to the proposed testing modifications.

These FSAR changes g

will be included as an amendment to the FSAR pending approval y

for implementation from the NRC.

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12-9 ' Reactor Regulation For your information, two (2) other tests which were originally considered for inclusion in the power ascension program acceleration have been withdrawn from consideration after detailed review by the responsible organizations.

These tests were:

Test 22 - Turbine Valve Surveillance Test 30 - Loss of Offsite Power Test

-Submittal of these jusdEfi'ations to the-NRC completes c

this phase of the Power Ascension Test Program changes.

Twenty-six (26) justifications have been submitted to the NRC since August 21, 1985 however, PSE&G has not received official notification from the NRC regarding approval to proceed with the Power Ascension Program as justified in the safety evaluations.

As noted in all previous correspondence, these modifications impact finalizing the related power ascension testing detailed procedures, therefore, an expedited revies is again requested.

PSE&G is ready to meet with cognizant NRC personnel to discuss the proposed ~ modifications should you require additional information.

' Sincerely, Attachments 9

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Director of Nuclear 3

12-9-85 Reactor Regulation C

-D.H. Wagner.

USNRC Licensing Project Manager R.W..Borchardt USNRC Senior Resident Inspsector L.H.

Bettenhausen USNRC Chief - Operations Branch

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,a GENERAL ELECTRIC CO. TECHNICAL ANALYSIS HOPE CREEK GENERATING STATION TEST NUMBER 1 - CHEMICAL'AND RADIOCHEMICAL TEST SIMPLIFICATION - REDUCED NUMBER OF TESTS OBJECTIVE:

Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, i

paragraphs 4.h and 5.a.a require chemical and radiochemical tests and measurements to demonstrate the design capability of chemical control systems maintain reactor water quality within limits.

Test Number 1, Chemical and Radiochemical, demonstrates that the plant water chemistry and radiochemistry complies with these limits during the power ascension test program and also' demonstrates the design capability of the plant chemistry systems.

It is proposed to substitute plant surveillance procedures for the chemistry and radiochemistry monitoring requirements of Test Number 1 and to delete the integrated performance testing of the Reactor Water Cleanup (RWCU) System and condensate demineralizer system at Test Condition 3.

DISCUSSION:

Plant surveillance procedures are used to ensure that plant water chemistry and radiochemistry are within plant Technical Specification and GE Fuel Warranty limits as appropriate.

The procedures include all required measurements for power ascension testing and specify limits at least as restrictive as those of Test Number 1.

In addition, sampling frequencies specified by the plant procedures exceed the requirements of Test Number 1.

Nevertheless, performance of these procedures will be specified on the power ascension test sequencer at the test conditions required by the startup test specifications.

In addition to the chemistry and radiochemistry measurements of reactor coolant and auxiliary systems, Test Number 1 requires the RWCU to be taken out of service to demonstrate the integrated performance of the RWCU and demineralizer systems.

The no-RWCU-(no-Cleanup) test is planned to be performed at Test Condition 3 and 6.

Although performance of this test at Test Condition 3 provides an opporturnity to verify procedures and to obtain preliminary data, testing at Test Condition 3 is not required to operate at higher power levels or to be able to effectively perform the test at Test Condition 6.

Testing at Test Condition 6 will demonstrate the ability of these systems to adequately manage coolant chemistry at the most demanding plant operating condition (rated power / flow).

1

e CONCLUSION:

Performance of the plant chemistry and radiochemistry procedures provides the required data for demonstrating that water chemistry and gaseous and liquid effluent activities are within Plant Technical Specifications and GE Fuel Warranty limits during the power ascensi.on test program.

Performance of the no-cleanup test at Test Condition 6 provides the required data to assess the chemical control performance of the RWCU and condensate demineralizer systems.

These procedures and testing satisfy the objectives of Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraphs 4.h and 5.a.a.

Substitution of plant procedures for the chemistry and radiochemistry monitoring requirements of Test Number 1 and deletion of the no-cleanup test at Test Condition 3 will not affect any safety systems or the safe operation of the plant and thus does not involve and unreviewed safety question.

2 2 -

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PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-Z-Oll TITLE:

SIMPLIFICATION OF TEST NUMBER 1, CHEMICAL AND RADIOCHEMICAL Date:

NOV 121985 1.0 PURPOSE The purpose of this Safety Evaluation is to document the results of an evaluation performed to justify the substitution of plant chemistry and radiation protection surveillance procedures for the monitoring requirements of Test Number 1 and to justify the deletion of the RWCU and condensate demineralizer performance tests at Test Condition 3.

2.0 SCOPE The scope of this evaluation consists of the monitoring of radiolytic activity and control of reactor coolant system water quality during power ascension testing.

3.0 REFERENCES

1.

Hope Creek Final Safety Analysis Report (FSAR)

Chapter 14.

2.

GE Startup Test Specification No. 23A4137 Rev. 0 3.

NRC Regulatory Guide 1.68, Revision 2, August 1978 4.

Hope Creek Generating Station Draft Technical Specifications 5.

Draft Chemistry Frequencies, Specifications and Surveillance Procedure, CH-TI.ZZ-012(O) 6.

Draf t Radiation Protection Procedure Gaseous Effluent Surveillance, RP-ST.ZZ-004(O) 7.

GE Fuel Warranty 8.

BWR Water Chemistry Guidelines (EPRI Report April 1, 1984)

PSE-SE-Z-Oll 1 of 4

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4.0 DISCUSSION Paragraphs 4.h and 5.a.a of Appendix A to Regulatory Guide 1.68 require that tests be conducted to demonstrate the ability of chemical control systems and installed analysis and alarm systems to maintain water quality within design limits and to verify that reactor

. coolant chemistry does not' exceed design limits.

Test Number 1, Chemical and Radiochemical, demonstrates that plant water chemistry complies with GE fuel warranty and Technical Specifications requirements and demonstrates the ability of chemical control systems to maintain water quality.. It is proposed to substitute plant surveillance procedures for the chemistry monitoring requirements of Test Number 1 and to delete the testing of the RWCU system and the condensate demineralizer system at Test Condition 3'which is currently required by Test Number 1.

Plant surveillance procedure CH-TI.ZZ-012(O), Chemistry Sampling and Surveillance Procedure, which is based on the EPRI guidelines for BWR water quality (reference 8), is used to ensure that plant water chemistry meets the limits specified by the GE fuel warranty (reference

7).and Technical Specifications 3.4.4.

All chemistry measurements required for power ascension testing will be included in CH-TI.ZZ-012(O).

The chemistry limits of CH-TI.ZZ-012(O) are at least as strict as those of Test Number 1.

Plant procedures CH-TI.ZZ-012(O) and RP-ST.ZZ-004(O),

Gaseous Effluent Surveillance, monitor and analyze liquid and gaseous effluent activity in addition to monitoring the specific activity of reactor coolant.

These two procedures ensure compliance with Technical Specifications 3/4.4.5 and 3/4.11.

They will obtain all the effluent activity and reactor coolant activity data which would be required by power ascension testing.

The limits they impose on effluent and reactor coolant activity are at least as strict as the Test Number 1 limits.

In most cases, the sampling frequencies of CH-TI.22-012(O) exceed the requirements of Test Number 1.

However, the power ascension test sequencer will be used to ensure that all necessary chemical and radiochemical measurements are taken at each test condition prior to proceeding to the next test condition.

PSE-SE-2-011 2 of 4

[

.o Performance testing of,the RWCU and condensate demineralizer systems to demonstrate that-they meet design specifications is currently performed at Test Conditions 3 and 6.

However, since performance testing at full power, i.e., Test Condition 6, represents the most severe operating conditions for these systems, testing at low power levels (Test Condition 3) is less limiting and is, therefore, redundant.

4

5.0 CONCLUSION

Compliance with plant surveillance procedures satisfies the requirements of paragraphs 4.h and 5.a.a of Appendix A to Regulatory Guide 1,68 and Test Number 1 with regard to the monitoring of plant water quality, reactor coolant radiolytic activity, and effluent activity.

Therefore, plant surveillance procedures can be substituted for the applicable sections of Test Number 1.

The performance of the RWCU system and condensate demineralizers is adequately demonscrated by performance of Test Number 1 at Test Condition 6 so the testing of these systems at Test Condition 3 can be deleted.

These changes will not adversely affect any safety systems or the safe operation of the plant so they do not consitute an unreviewed safety question.

The do not require a change to the plant Technical Specifications.

6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATIONS Revisions to Hope Creek's FSAR and Startup test procedures shall be made to reflect the substitution of plant surveillance procedures for the applicable requirements of Test Number 1 and the deletion of RWCU and condensate demineralizer systems testing at Test Condition 3.

8.0 ATTACHMENTS None PSE-SE-2-Oll 3 of 4 L

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GENERAL ELECTRIC CO. TECHNICAL ANALYSIS HOPE CREEK GENERATING STATION TEST NUMBER 3 - FUEL LOADING - TEST SIMPLIFICATION

- ELIMINATE FUEL LOADING CHAMBERS OBJECTIVE:

Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraph 2, establishes requirements for initial fuel loading to prevent inadvertent criticality.

Requirements for continuous monitoring of the neutron flux throughout the core loading must be established.

In addition, paragraph 2.a requires that the shutdown margin be verified for a partially and fully loaded core.

Test Number 3, Fuel Loading, provides the procedures to load fuel safely and efficiently to the full core size, and includes subcriticality checks, shutdown margin verification (including Test Number 4 - Full Core Shutdown Margin) and control rod functional testing (in conjunction with Test Number 5 - Control Rod Drives).

Currently, it is planned to use Fuel Loading Chambers (FLC's) to measure the neutron count rate during initial fuel loading because of their high sensitivity and ability to be moved near initial fuel loading and neutron sources.

It is proposed to simplify the fuel loading procedure by replacing the FLC's with the Source Range Monitor (SRM) instrumentation.

In addition, the startup sources will be positioned in their alternate locations (to be closer to the SRM detectors) and the fuel loading sequence will be modified such that initial fuel loading will begin between an SRM detector and a neutron source.

Fuel loading will continue in a spiral pattern around the initial SRM until the core is fully loaded.

DISCUSSION:

Requirements during fuel loading are established to preclude inadvertent criticality.

A thorough pre-fuel loading checklist is performed to verify that all systens required during fuel loading are operable.

Prior to fuel loading, all control rods are verified to be fully inserted and remain fully inserted throughout the proposed fuel loading, except during the functional /suberitical checks and the partial core shutdown margin test at which time rod movement is controlled by strict procedural guidance, refueling interlocks which prevent the withdrawal of more than one control rod at a time.

No control rod movement will be performed until the minimum count rate is achieved on at least one SRM.

1

Predictions of the core reactivity and shutdown margin have been prepared in advance for the Hope Creek Generating Station fuel design and support the safe fuel loading under suberitical conditions.

For the Hope Creek Generating Station initial fuel loading, it has been predicted that the effective neutron multiplication factor for a fully loaded core with all rods inserted is 0.92, and is 0.98 with the strongest worth control rod withdrawn (cold conditions).

These predictions have l

been performed with core physics calculation methods that have been extensively qualified and have been demonstrated to be highly accurate for criticality predictions based on tests / experiments, Monte Carlo benchmark calculations and Shutdown Margin (SDM) demonstration tests for initial and reload cores (Reference 1).

In addition, rigorous Ouality Assurance (OA) programs during fuel design and fuel and control blade manufacturing ensure that the fuel and control blades are manufactured as specified.

These OA programs (described in Reference 2) apply quality system elements necessary to provide assurance that systems and components meet the quality requirements of i

applicable codes, standards and regulatory agency requirements.

General Electric's OA program has been reviewed by the NRC and found to comply with all applicable requirements of Appendix B to 10CFR50 (Reference 3).

As a result, these pre-fuel loading precautions and measures provide significant assurances against inadvertent criticality during fuel loading.

For the proposed fuel loading procedure, the fuel will be initially loaded between a neutron source at its alternate location and the closest SRM.

Figure 1 shows the proposed fuel loading sequence.

Because of the fixed location of the SRM's 3

and the distance from the sources, the SRM count ratio will initially be less than 0.7 cps, the Plant Technical Specification minimum SRM count rate.

As fuel is loaded, neutronic coupling between the source and detector will occur and the count rate will increase.

The estimated minimum SRM count rate with 16 bundles loaded as a function of the source strength has been determined based on past BWR startup test results from seven plants using FLC's.

SRM count rates were also available for one of the seven plants.

The results are presented in Figure 2.

Figure 2 shows that, with appropriate conservatisms, a source strength of 500 curies is required to assure that the minimum count rate of 0.7 cps is achieved with the initial-16 bundles loaded.

The SRM data, for a startup where the alternate source location was used (source strength of 1855 curies), indicated an SRM count rate of 12 cps with six bundles loaded and 30 cps with 16 bundles loaded.

Similar performance is expected during fuel loading at Hope Creek 1

2 4

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Generating Station.

Nevertheless, an exemption to the Technical Specification requirement for a minimum count rate (0.7 cps) during the loading of the initial 16 bundles is required.

Since the SRM system is not safety related and no 4

credit is taken for the SRMs in the safety analysis for these conditions (Reference 4, paragraph 15.9.2.3.3), this exemption 2

will not adversely affect any safety systems or the safe operation of the plant.

To further support this exemption, a core reactivity calculation was performed for the initial 16 bundles loaded to demonstrate that even with all of the control rods withdrawn, the partial loading would remain subcritical, with significant margin.

A two dimensional (conservatively neglecting axial leakage) 16 bundle analysis was performed using a neutron transport Monte Carlo code.

The fuel bundle types and loading configuration of the 16 bundles were based on the proposed fuel loading procedure of Figure 1 and are shown in Figure 3.

For a j

moderator temperature of 20 C, the resulting effective neutron multiplication factor was 0.973i0.003 (one standard deviation) which includes the model critical benchmark bias for use with the_ENDF/B-IV cross section data.

This analysis demonstrates that the initial 16 bundle loading will remain subcritical by 2.7% Ag even if the control rods are withdrawn.

Thus further assurring that the SRM monitoring requirements can be exempted during this portion of the fuel loading procedure.

Similar exemptions of SRM monitoring requirements (for fewer bundles) during fuel loading have been approved by the NRC for several reload licenses (References 5, 6, and 7).

After the initial 16 bundles are loaded and an SRM is on scale, the fuel loading will continue in a spiral fashion as shown in Figure 1.

Since only one SRM will initially be on scale, a portable source will be used to periodically demonstrate operability of the SRM's located in areas with no fuel.

One of the SRM's will be required to maintain continuous visual indication in the control room until other SRM's are on scale.

Use of a portable source to demonstrate operability of SRM's has previously been approved by the NRC for other plants (References 5 and 7).

The portable source is widely used in the nuclear industry as a bugging source for detector calibration and is an easy device to operate with no complex or unsafe maneuvers required.

When 144 fuel bundles have been loaded (Step 36 in Figure 1),

_two SRM's will be surrounded by fuel and indicating greater than 0.7 cps.

At this time, a partial core shutdown margin test will be performed as required by Regulatory Guide 1.68.

The partial core configuration (fuel bundle types) for this off-center loading is almost identical to the partial core configuration resulting from a standard center spiral loading.

3 i

i m

4 s.

Core physics calculations for the off-center partial SDM test will be performed for the proposed fuel loading sequence.

After the partial core SDM test, the remaining fuel will be loaded based on the sequence shown in Figure 1 until the core is fully loaded.

During this portion of the fuel loading, at least two SRM's will be operable, with at least one providing continuous visual indication in the control room.

SRM's not surrounded by fuel will be periodically checked for operability using the portable source.

Once the core is fully loaded, the full core SDM test will be performed using the standard procedures. provides a summary of the major fuel loading steps.

+ -

Currently, Plant Technical Specifications require that at least two SRM's are operable and continuously visible in the control room.

The SRM's are required to be located in the quadrant where fuel loading occurs and in an adjacent quadrant.

For the proposed fuel' loading procedure, a Special Test Exception 4

(3/4.10.7, Special Instrumentation - Initial Core Loading) is required for the Technical Specifications to incorporate the changes to the flux monitoring requirements (Attachment 2).

Initially, continuous visual indication and minimum count rate requirements for the SRM's are exempted for the first 16 f uel bundles loaded since the core will be subcritical even with all control rods withdrawn.

Thereafter two SRM's, one in the quadrant where fuel loading is taking place and one in an adjacent quadrant, will still be required to be operable but one will be allowed to be demonstrated operable by using a portable neutron source.

This operability check will be periodically performed for those SRM's located in areas with no 1

fuel loaded.

As stated before, this method of demonstrating SRM operablity has been previously approved for other BWRs i

(Reference 5 and 7).

i The SRM system is not safety related and no credit is taken for the SRM's under these conditions (fuel loading with control rods' inserted, i.e.,

core subcritical) in the safety analysis i

of accidental positive reactivity insertions.

The SRM's provide indication of neutron flux changes as a matter of good practice during fuel loading and thus no safety requirements exist for an SPM detector.

Therefore, continuous visual indication of only one SRM does not adversely affect any safety systems or the safe operation of the plant since the core will be.subcritical during fuel loading and at least two SRM's will be demonstrated operable prior to and during fuel loading.

' CONCLUSION By locating the startup sources in alternate positions and modifying the fuel loading sequence, the SRM's will be capable of providing continuous monitoring of the neutron flux 4

~

throughout the core loading after the minimum count rate is initially achieved.

Monitoring requirements during loading of the initial fuel bundles (up'to 16 bundles) can be exempted because even with all of the control rods withdrawn, this configuration would be subcritical because of the high neutron leakage.

Pre-fuel load activities ensure that systems required for fuel loading are operable, control rods are inserted during the fuel loading except during special controlled tests and core reactivity predictions aid in the evaluation of the measured responses during fuel loading.

Together, these procedures and precautions provide assurance against inadvertent criticality.

Since the safety analysis does not require a minimum count rate on the SRM system under these conditions, the proposed change in the fuel loading procedures will not' adversely affect any safety system or safe operation of the plant and therefore does not involve an unreviewed safety question.

Test Number 3, Fuel Loading, can therefore be simplified by replacing the Fuel Loading Chambers with SRM's and using an off-center spiral loading sequence.

REFERENCES:

1.

" Steady State Nuclear Methods",. General Electric Company Licensing Topical Report, May 1985 (NEDO-30130-A).

2.

" Nuclear Energy Business Operation Quality Assurance Program Description," General Electric Company, March 1985 (NEDO-ll209, Revision 5).

3.

Letter, G.

G.

Zech (NRC) to J.

M.

Case (GE), "NRC Acceptance of Revised General Electric Quality Assurance Topical Report," April 19, 1985.

4..

Hope Creek Final Safety Analysis Report (FSAR) Chapter 15.

5.

Amendment No. 27 to Facility Operating License No.

DPR-63, Niagara Mohawk Power Co.,

Nine Mile Point Station 4

Unit No.

1, Docket No. 50-220, March 2, 1979.

6.

Amendment No. 66 to Facility Operating License No.

DPR-57, Georgia Power Co. et.

al.,

Edwin I.

Hatch Nuclear Station Unit No.

1, Docket No. 50-321, June 12, 1979 7.

Amendment No. 5 to' Facility Operating License No. NPF-29, Grand Gulf Nuclear Station Unit No.

1, Docket No. 50-416, October 12, 1985.

ATTACHMENTS:

1.

Summary of Major Fuel Loading Steps.

2.

Technical-Specification Change - Special Test Exception 3/4.10.7.

s

TES? NUMBER 3 - FUEL LOADlNG TEST $8MPLIFTCAT20N - ELIM!NATE FUEL LOADING CHAMBER $

Figure 1 - Proposed Fuel Loading Sequence M

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02 06 10 14 18 22 26 30 34 38 42 46 50 54 59 A = Alternate Source Locations

$ = SR1 Detectors (Notes Letters denote fuel loading regions) l l

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HOPE CREEK GENERATI!;3 STATION 11/19/85 l

TEST NUMBER 3 - FUEL LOADING TEST SIMPLIFICAT80N - ELIMINATE FUEL LOADING CHAMBERS Figure 2 - SRM Count Rate Estimate Startup Test Data g-36 -

SBN to Source distance = 43 cm d

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FCFE CREEK GENERATING STATION 11/19/85

TEST NUMBER 3 - FUEL LOAD!NG TEST $!MPLIFICATION - EL8MINATE FUEL LOADING CHAMBER $

1 Figure 3 - Fuel Loading Pattern li i 11 1 i ii ii ii ii N

T 55 55 55 56 55 SS 55 9 1 14 34 22 44 22 44 22 43 41 F

55 1 54 44 22 4 _4_22 44 22 44 4 _5 1

15 44 33 44 33 44~33 44 33 44 51 61 U 53 44 3S 44 33 44 33 44 33 44 35 YT u

15 34 33 44'22 44 22 44 22 44 33 43 51 47 F 44 44 33, 44 22 44 22 44 22 44 33 44 44 Tl m

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15 4__4 33 44 33 44 33 44 33 44 33 44 33 44 51 1 5 22 44 22 4 4~ 22 44 22 44 22 44 2 2' 44 22 51 39 -

15 22 44 22 44 22 44 22 44 22 44 22 44 22 51 15 44 33 44 33 44 33 44 33 44 3 3~ 44 33 44 51 35 ~

15 44 33 44 33 44 3__3 44 33 44 33 44 33 44 51 15 22 44 22 44 22 44 22 44 2 2' 4T 22 44 22 51 31 15 22 44 22 44 22 44 22 44 22 44 22 44 22 51 1 5 44 33 4T 33 44 33 44 3J 4f 33 44 33 44 51 27 -

15 4_ 4 33 44 3_ 3 44 33 44 33 44 33 44 33 44 51 15 22 44 22 44 22 44 22 44 22 44 22 44 22 51 23 ~

15 22 4 4_ 22 44 22 44 22 44 22 44 2__2 44 22 51 15 44 33 44 33 44 33 44 33 44 33 44 33 44 51 19~

1 5 34 33 44 33 4_ 4 33 44 33 44 33 44 33 43 5_1

._ 1 44 44 33 44 22 44 22 44 22 44 33 44 44 y 15 15 34 3__3_ 44 22 44 22 44 22 44 33 43 51 1 1 5 f~ 44 33 44 33 44 33 44 33 44 45 1 1 11 15 34 33 44 33 44 33 4__4 _33 43 5_1 1 54.44 22 44 2 2' 41 22 44 45 1

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= Analyzed Region HCPE CREEK GENERAT tiG STATION 11/19/85

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TEST NUMBER 3 - FUEL LOADING TEST SIMPLIFICATION - ELIMINATE FUEL LOADING CHAMBERS ATTACHMENT 1

SUMMARY

OF MAJOR FUEL LOADING STEPS 1.

Prerequisites j

a.

Neutron sources have been loaded into the alternate 1

. locations and blade guides have been oriented source as shown in Figure A-1.

b.

Nuclear instrumentation ( S RM 's a nd I RM 's ) has been installed and activated in the design configuration.

c.

The " shorting links" have been removed from the I

Reactor Protection System (RPS) and the non-coincident scram setpoints have been set.

d.

All control rods are fully inserted and have been functionally tested within approximately two weeks prior to the start of fuel loading and have been scram tested within, approximately six weeks prior to fuel loading.

e.

The operability of the SRM's has been verified.

f.

Fuel support castincs are installed and locations verified to assure proper orificing.

g.

All other systems required to be operable by Technical Soecifications shall be determined to be operable.

h.

The reactor mode switch is locked in the REFUEL mode during fuel loading.

2.

Procedure 4

i a.

Begin fuel loading following the steps shown in Figure A-2.

(Off-center loading).

b.

Fuel loading begins at cell location 18,47 and continues-until the first 16 bundles have been loaded (Steps 1-4, Figure A-2).

Perform SRM operability check for SRM adjacent to c.

the initial 16 bundles (16,45).

Perform channel checks for other SRM's according to Technical Specification requirements.

A portable source is to be used for this purpose when an SRM is not surrounded by fuel.

4 4

ATTACH:fENT 1 Page 1 of 2

ATTACHMENT 1 (Cont'd) 4

- d.

Continue fuel loading according to Figure A-2 until Region A is completely loaded (144 bundles).

During the process of fuel loading, SRM(s) count rate should be recorded.

This is used to generate 1/M plots.

Perform control rod functional checks and suberitical tests when the minimum count rate is achieved on at least one SRM channel.

i e.

Perform the partial core shutdown margin test after

. 144 bundles are loaded (Region A, Figure A-2).

Core physics calculations will be performed to support this test.

1 f.

Continue fuel loading in Region B.

The loading is continued according to sequential order of regions (Figure A-2) until the core is fully loaded.

g.

As soon as each SRM is on scale (0.7 cps or higher) during loading, SRM calibration is performed and i

documented.

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02 06 to 14 to 22 26 30 34 38 42 46 50 54 58 A

= Alternate Source Location e

= S191 Detector location N, / = Double Blade Guides I

I l

HOPE CREEK GENERATING STA!!ON 33f39f33

x TCST NUMBER 3 - FUELiLOAD8N8 s _.

TE$f,SIMPLIFICAf!ON - (LIMINAft FUEL 1.0ADING CHAMBER $

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02 06 10 14 18 22 26 37'334 38 42 46 60 54 59 A = Alternatak%arce, Locations

$ = SR1 Detectors (Notes Letters denote fuel loading regions) 1 HOPE CREEK GENERATING STAi!Ofi 11/19/85 1

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TEST NUMBER 3 - FUEL LOADING TEST SIMPLIFICATION - ELIMINATE FUEL LOADING CHAMBERS ATTACHMENT 2 TECHNICAL SPECIFICATION CHANGE Special Test Exceptions (NEW) 3/4.10.7 Special Instrumentation - Initial Core Loading Limiting Condition for Operation 3.10.7 During initial core loading within the Startup Test 2

Program the provisions of Specifications 3/4.9.2 may be suspended provided that at least two source range monitor (SRM) channels with detectors inserted to the normal operating level are OPERABLE with:

a.

One of the. required SRM channels continuously indicating

in the control room.

b.

One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant **.

c.

The RPS " shorting links" shall be removed prior to and during fuel loading.

d.

The reactor mode switch is OPERABLE and locked in the REFUEL position.

APPLICABILITY OPERATIONAL CONDITION 5 ACTION With-the requirements of the above specification not satisfied, immediately suspend all operations involving initial core loading.

ATTACHMENT 2 Page 1 of 4

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ATTACHMENT 2 (Cont'd) 1 SURVEILLANCE REOUIREMENTS I

I 4.10.7.1 Within one hour prior to and at least' once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during the initial core loading verify that:

a.

The above required SRM channels are OPERABLE by:

1.

PerformanceofaCHINNELCHECK***

2.

Confirming that the above-required SRM detectors are at the normal operating level and located in the quadrants required by Specification 3.10.7.

b.

The RPS " shorting links" are removed.

c.

The reactor mode switch is locked in the REFUEL position.

4.10.7.2 Perform a CHANNEL FUNCTIONAL TEST within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start and at least once per 7 days during initial core loading.

4.10.7.3 For at least one.SRM channel,-verify that the count rate is at'least 0.7 cps ****:

?

i a.

Immediately following the loading.of the first 16 bundles.

b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter during initial

' core leading.

Up to 16 fuel bundles may be loaded without a visual indication of count rate.

The use of special movable detectors during CORE ALTEPATIONS in place of the normal SRM nuclear detectors is permissible as long as these special detectors are

_ connected to the normal SRM circuits.

- - Check may be performed by use of movable neutron source.

* * * :Provided signal-to-noise is > 2.

Otherwise, 3 cps.

ATTACHMENT 2 Page 2 of 4 h

ATTACHMENT 2 (Cont'd) i REFUELING OPERATIONS 3/4.9.2 INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.9.2 At least 2 source range monitor * (SRM) channels shall be OPERABLE and inserted to the normal operating level with:

a I

Annunciation and continuous visual indication in the a.

control room.

b.

One of the required SRM detectors located in the quadrant where CORE ALTERATIONS are being performed and the other required SRM detector located in an adjacent quadrant, and c.

Unless adequate shutdown margin has been demonstrated per Specification 3.1.1, the " shorting links" removed from the RPS circuitry prior to and during the time any control red is withdrawn.

l APPLICABILITY:

OPERATIONAL CONDITION 5.**

ACTION:

With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS and insert all insertable control rods.

SURVEILLANCE REOUIREMENTS i

4.9.2 Each of the above required SRM channels shall be demonstrated OPERABLE by:

a.

At least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:

1.

Performance of a CHANNEL CHECK.

2.

Verifying the detectors are inserted to the normal operating level, and 3.

During CORE ALTERATIONS, verifying that the detector of an OPERABLE SRM channel is located in the core quadrant where CORE ALTERATIONS are being performed and another is located in an adjacent quadrant.

ATTACHMENT 2 Page 3 of 4

e ATTACHMENT 2 (Cont'd)

.j The use of special movable detectors during CORE ALTERATIONS in place of the normal SRM nuclear detectors is permissible

]

as long as these special detectors are connected to the normal SRM circuits.

Not required for control rods removed per Specification 3.9.10.1 and 3.9.10.2.

- See Special Test Exception 3.10.7.

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ATTACHMENT 2 Page 4 of 4 t

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PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-2-028 TITLE:

TEST SIMPLIFICATION, ELIMINATION OF FUEL LOADING CHAMBERS DURING FUEL LOADING, TEST NUMBER 3 DEC 4 1985 Date:

1.0 PURPOSE The purpose of this Safety Evaluation is to document the results of the evaluation performed to ensure that the elimination of fuel loading chambers during performance of Test Number 3, Fuel Loading, will not adversely affect reactor safety.

2.0 SCOPE The scope of this Safety Evaluation is the adequacy of Hope Creek's power ascension test program as it concerns fuel loading.

3.0 REFERENCES

1.

Regulatory Guide 1.68, Revision 2, August 1978.

2.

Hope Creek Final Safety Analysis Report (FSAR)

Chapter 14 and 15.

3.

General Electric Startup Test Specification, 23A4137, Revision 0.

4.

Hope Creek Generating Station Technical Specifications.

5.

" Steady State Nuclear Methods," General Electric Company Licensing Topical Report, May 1985 (NEDO-30130-A).

6.

" Nuclear Energy Business Operation Ouality Assurance Program Description," General Electric Company, March 1985 (NEDO-11209, Revision 5).

7.

Letter, G.G.

Zech (NRC) to J.M. Case (GE), "NRC Acceptance of Revised General Electric Quality Assurance Topical Report," April 19, 1985.

8.

Amendment No. 27 to Facility Operating License No.

DPR-63, Niagara Mohawk Power Co., Nine Mile Point Station Unit No.

1, Docket No. 50-220, March 2, 1979.

PSE-SE-2-028 1 of 7

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9.

Amendment No. 66 to Facility Operating License No.

DPR-57, Georgia Power Co. et. al.,

Edwin I.

Hatch i

Nuclear Station Unit No.

1, Docket No. 50-321, June 21, 1979.

10.

Amendment No. S to Facility Operating License No.

NPP-29, Grand Gulf Nuclear Station Unit No.

1, Docket No. 50-416, October 12, 1985.

4.0 DISCUSSION Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A,

paragraph 2.(b) requires continuous monitoring of the neutron flux throughout the core loading so that all changes in the multiplication factor are observed.

Hope Creek's Technical Specification requires that at least 2 Source Range Monitor (SRM) channels shall be operable with annunciation and continuous visual indication in the control room during Core Alterations.

One of the required SRM detectors must be located in the quadrant where-Core Alterations are being performed and the other required SRM detector must be located in an adjacent quadrant.

Requirements to demonstrate that the SRMs are i

operable during core alterations include periodic performance of SRM channel checks and channel functional tests and verification that the channel count rate is at least 0.7 counts per second (cps) if the signal-to-noise ratio is greater than or equal to 2.

Otherwise, 3.0 cps are required.

Hope Creek's Technical Specification permits the use of special movable detectors (i.e.,

Fuel Loading Chambers, FLCs) during Core Alterations in lieu of the normal SRM nuclear detectors as long as they are connected to the 4

normal SRM channels.

FLCs made with B10 a re 6 x 103 times more sensitive than SRMs.

Use of these FLCs is often necessary during initial fuel load since the sb-Be neutron sources sometimes do not provide an adequate neutron population to result in the required count rate on the installed SRMs with no fuel in the core.

Hope Creek's Fuel Loading Procedure presently specifies the use of FLCs and installation of neutron sources in their normal location.

In accordance with the General Electric Startup Test Specificatio.ns, this procedure also

- PS E-S E-2 -0 29

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specifies that fuel be loaded in successive spiral rings about the core center.

The PLCs will be moved outward while maintaining continuous indication of the neutron flux.

It is proposed to simplify Test Number 3, Fue l Loading,

by eliminating FLCs, installing neutron sources in their alternate locations, and using a revised core loading sequence (Attachment 1).

Utilizing installed SRMs in lieu of FLCs will save the time associated with FLC installation, calibration, movement, and problem resolution.

Changes to the neutron source locations and fuel loading sequence will enhance the SRM count rate.

Specifically, the existing incore SRM detectors will be connected to the SRM channels instead of the FLCs.

Because the SRM detectors have lower sensitivity to neutrons than FLCs, the sources will be located closer to the SRM detectors by placing them at a length one control cell diagonally away from a SRM (Attachment 1).

Fuel will be loaded in the control cell between the SRM detector and the source, then in cells around the source as depicted in Attachment 1.

Thereafter, fuel will be loaded in spiraling rings around the already loaded cells.

As a result of this proposed fuel loading procedure, the count rate required by Hope Creek's Technical Specification to demonstrate the operability of SRMs will not be achieved until an intermediate point in the fuel load.

In addition, the requirement for a minimun count rate on a second SRM, located in a quadrant adjacent to the quadrant in which fuel is being loaded, will not be satisfied during the initial stages of the fuel loading procedure.

This is considered acceptable based on the following.

Requirements during fuel loading are established to preclude inadvertent criticality.

A thorough pre-fuel loading checklist is performed to verify that all systems required during fuel loading are operable.

Prior to fuel loading, all control rods are verified to be fully inserted and the refuel mode switch position interlocks are verified to be operable.

These interlocks are required by Technical Specifications to be verified operable at least once per 7 days during core alterations.

PSE-SE-2-023 3 of 7

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Predictions of core reactivity and shutdown margin have been prepared in advance for the Hope Creek fuel design and fuel loading.

They show that the core reactivity will be suberitical during fuel loading for the Hope Creek initial fuel loading.

Calculations show that the effective neutron multiplication factor for a fully loaded core with all rods inserted is 0 92 and is 0.98 with the strongest worth control rod withdrawn (cold conditions).

These predictions have been performed with core physics calculation methods that have been extensively qualified and have been demonstrated to be highly accurate for criticality predictions based on tests / experiments, Monte Carlo benchmark calculations and shutdown Margin (SDM) demona.tration tests for initial and reload cores (Reference 5).

't Rigorous Ouality Assurance (OA) programs implemented during fuel design and fuel and control blade manufacture ensure that the fuel and control blades are fabricated as specified.

These OA programs (described in Reference 6) apply quality system elements necessary to provide assurance that systems and components meet the quality requirements of applicable codes, standards and regulatory agency requirements.

General Electric's OA program has been reviewed by the NRC and found to comply with all applicable requirements of Appendix B to 10CFR50 (Reference 7).

As a result, these pre-fuel loading precautions and measures provide significant assurances 4

against inadvertent criticality during initial fuel loading.

The SRMs in the quadrant being loaded and in an adjacent quadrant will be continuously monitored.

Prior to loading fuel in the first quadrant, the SRM in that quadrant and in an adjacent quadrant will be demonstrated operable by the use of a portable source.

As fuel is loaded, neutronic coupling between the source and detector will result in an increased count rate on the associated SRM channel.

The portable source will be used to periodically demonstrate operability of the SRMs located in areas with no_ fuel.

Use of a portable source to demonstrate operability of the remaining SRMs has previously been approved by the NRC (Reference 8 and 10).

The portable source is widely used in the nuclear industry as a bugging source for detector calibration and is an easy device to operate.

Its use does not pose a reactor safety concern.

PSE-SE-2-028 4 of 7 e

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Data from other BWR startups was used to generate a graph of estimated SRM Count Rate versus Source Strength with i

16 bundles loaded (Attachment 2).

This information indicates that a SRM count rate of 0.7 cps should be achieved well before 16 bundles have been loaded provided the source strength is greater than approximately 500 curies.

Since experience has shown at other BWRs that the signal-to-noise ratio will be greater than or equal to 2,'this count - rate should satisfy the existing Technical Specification requirement for demonstrating that a SRM is operable. shows that Hope Creek's presently ordered sources will have a 500 curie t

strength until approximately the middle of May 1986.

Although Hope Creek's startup schedule supports a fuel load date prior to this, Hope Creek has contingency plans to obtain sources of adequate strength if fuel is not loaded prior to this date.

To support the acceptability of loading up to 16 bundles i

without a continuous count rate on the SRMs, a core reactivity calculation was performed by General 2

Electric.

It was demonstrated that the initial 16 fuel bundles could be loaded and remain subcritical with margin even with all of the control rods withdrawn.

Specifically, a two dimensional (conservatively neglecting axial leakage) 16 bundle analysis was performed using a neutron transport Monte Carlo code.

The fuel loading configuration and bundle types of the 16 bundles were based on the proposed fuel loading map and pattern of Attachment 1 and 4 respectively.

With all control rods. withdrawn and a moderator temperature of 20 C, the resulting effective neutron multiplication factor was 0.973 i 0.003 (one standard deviation).

This includes the model critical benchmark basis for use with

- the ENDF/B-IV cross section data.

Based on the above,

- there is no safety concern for loading up to 16 bundles without a continuous count rate on the SRMs.

Similar exemptions of SRM monitoring requirements (for fewer bundles) during fuel loading have been approved by the NRC for several reload licenses (References 8,

9, and 10).

In addition, although safety analyses of accidental positive reactivity additions have assummed. as an initial condition that the neutron source level is above a specified nininum,

a. significant positive reactivity addition can only occur when the reactor is less than one control rod subcritical and therefore the minimum source level does not need to be observed during fuel loading PSE-SE-2-028 5 of 7

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o when all control rods are fully inserted (Reference 2, paragraph 15.9.6.2.3.3).

Thus, the plant Technical Specification requirement for SRM indication of neutron level during fuel loading is based on good operating practice rather tnan on any safety consideration.

Control rod drive functional checks and subcritical tests will be performed during fuel loading only after the minimum count rate is achieved on at least one SRM channel.

When 144 fuel bundles have been loaded (Step 36 in ), two SRMs will be surrounded b'y fuel and should be indicating greater than 0.7 cps.

Although only one SRM will be required to be indicating greater than 0.7 cps.

At this time, a partial core shutdown margin test will be performed as' required by Regulatory Guide 1.68.

The partial core configuration (fuel bundle types) for this off-center loading is almost identical to the partial core configuration resulting from a standard center spiral loading.

Core physics calculations for the off-center partial core shutdown margin test will be performed for the proposed fuel loading sequence.

After the partial core shutdown margin test, the remaining fuel will be loaded based on the sequence shown in Attachment 1 until the core is fully loaded.

Once the core is fully loaded, the full core shutdown margin test will be performed using the standard procedures.

provides a summary of the major fuel loading steps.

5.0 CONCLUSION

i

' Simplification of Test Number 3, Fuel Loading, by eliminating FLCs, installing neutron sources in their alternate locations, and using a revised core loading sequence as discussed above will not increase the probability of occurence or the consequence of a previously evaluated accident or malfunction of any safety system, or create a'different accident or malfunction.

In addition, the safety margin as defined in Hope Creek's Technical Specification is not reduced because the requirement for SRMs is based on good operating practice rather than on a safety consideration.

Consequently, an unreviewed safety question does not exist.

6.0 DOCUMENTS GENERATED None

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PSE-SE-2-028 6 of 7 i

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d 7.0 RECOMMENDATIONS Revisions to Hope Creek's Technical Specification and Fuel Load Procedure shall be made to simplify initial fuel loading by eliminating FLCs, installing neutron sources in their alternate locations and, using a revised core loading sequence.

8.0 ATTACHMENTS 1

1.

Proposed Fuel Loading Map.

2.

SRM Count Rate Estimate.

3.

Hope Creek's Sb-Be Neutron Source Strength.

4.

Fuel Loading Pattern.

5.

Summary of Major Fuel Loading Steps.

9.0 SIGNATURES Originator A 4. CAR __

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m 7

35 SRM to Source distance = 43 cm a

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16 bundles loaded M

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ATTACIIMENT 2 Page 1 of 1 l

ATTACHMENT 3 HOPE CREEK'S Sb-Be NEUTRON SOURCE STRENGTH DATE SOURCE STRENGTH (based on a 60 day half life) 10/6 7000 curies 12/16 3500 curies 2/6 1750 curies 4/6 875 curies 6/6 437 curies 8/6 268 curies 10/6 134 curies ATTACHME';

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ATTACHMENT 4 Fuel Loading Pattern W

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[1 55 55 55 66 55 55 55 Tl 1 14 34 22 44 22 44 22 43 41 F 55-

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U53 44 33 44 33 44 33 44 33 44 35 11 m

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2. 2 44 22 44 33 44 44 Il 15 34 33 44' 'J 3 44' 33 44 33 44 33 44 33 43 51 43 ~

15 44 33 44 33 44 33 44 33 44 33 44 33 4 4is1 1 5 22 44 22 44 22 44 22 44 22 44 22 44 22 51 39 ~

15 22 4 '4 22 44 22 44 22 44 22 44 22 44 22 51 15 44 33 44 33 44 33 44 33 44 33 44 33 44 51 35 ~ 15 44 33 44 33 44 33 44 33 44 33 44 33 44 51 15 22 44 22 44 22 44 22 44 2 2' 44 22 44 22 51 31' 15 22 44 22 44 22 44 22 44 22 44 22 44 22 51 1 5 44 33 44 33 44 33 44 33 4 4~ 33 44 33 4 4~5 1 27 ~

15 44 33 44 33 44 33 44 33 44 33 44 33 4451 1 5 22 44 22 44 22 44 22 44 22 44 22 44 22 b1 23 ~

1_ $ 22 44 22 44 22 44 22 44 22 44 22 44 22 51 1 5 44 33 44 33 44 33 4433 44 33 44 3344 51 19~ 1_ 5 34 33 443 3 44 334 43 3 44 334 4 3 3'4 3 S 1 1 44 44 334 4 22442 24 4 22 443 3 44!44 y 15 15 34 3344224 42244 224 43 3 4351 1 1 544 4'3 3 4 43 3j4 C 3 3 4 43 3.4 4 4 5 1

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02 06 10 14 18 22 26 30 34 38 42 46 50 54 SS 1 = 0.7 w/o U235 ( 92) 2 = 0.9 w/o U235 ( 132) 3 = 1.6 w/o U235 (160) 4 = 2.5 w/o U235 ( 308) 5 = 2.0 w/o U235 (72)

& = Altern. ate Source Location 9 = SNP1 Detmeter

= Analyzed Region j

ATTACIIMENT 4 Page 1 of 1

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SUMMARY

OF MAJOR FUEL LOADING STEPS 1.

Prerequisites Neutron sources have been loaded into the alternate a.

locations and blade guides have been oriented source as shown in Figure A-1.

b.

Nuclear instrumentation (SRM's and IRM's) has been installed and activated in the design configuration.

The " shorting links" have been removed from the c.

Reactor Protection System (RPS) and the non-coincident scram setpoints have been set, d.

All control rods are fully inserted and have been functionally tested within approximately two weeks prior to the start of fuel loading and have been scram tested within approximately six weeks prior to fuel loading.

e.

Fuel support castings are installed and locations verified to assure proper orificing.

f.

Operability of SRMs has been verified.

g.

The reactor mode switch is locked in the REFUEL mode during fuel loading.

h.

All other systems required to be operable by Technical Specifications shall be der <rmined to be operable.

2.

Precedure 9egin fuel loading following the steps shown in a.

Figure A-2. (Of f-cente r loading).

b.

Fuel loading begins at cell location 18,4 7 and continues until the first 16 bucdles have been loaded (Steps 1-4, Figure A-2).

Perforn SRM operability check for SRM adjacent c.

to the initial 16 bundles (16,45).

Perform channel checks for other SRM's according to Technical Specification requirements.

A portable source is to be used for this purpose when an SRM is not sarrounded by fuel.

ATTACPtEST 5 Page 1 of 4

a ATTACHMENT 5 ( Con t 'd )

d.

Continue fuel loading according to Figure A-2 until Reglen.A is completely loaded (144 bundles).

During the process of fuel loading, SRM(s) count rate should be recorded.

This is used to generate 1/M plots.

Perform the partial core shutdown margin test after e.

144 bundles are loaded (Region A, Figure A-2).

Core physics calculations will be performed to support this test.

f.

Continue fuel loading in Region B.

The loading is continued according to sequential order of regions (Figure A-2) until the core is fully loaded.

g.

As soon as each SRM is on scale (0.7 cps or higher) during loading, SRM calibration is performed and documented.

ATTACHMENT 5 Page 2 nf'4 rf

4 ATTACIIMENT 5 Figure A Proposed Fuel I.cading Sequence i

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( Notel Letters denote fuel loading regions)

ATO.*.CDiENT 5 Page 3 of 4 i

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02 06 10 14 to 22 26 30 34 38 42 46 50 54 58 A

= Alternate Source Location G

= SRM Detector Wation

\\, / = Double Blade Gu.idas ATTACHMENT 5 Page 4 of 4 l

l a

i GENERAL ELECTRIC CO. TECHNICAL ANALYSIS HOPE CREEK GENERATING STATION TEST NUMBER 5 - CONTROL ROD DRIVE / HOT SINGLE ROD SCRAM TESTING IN CONJUNCTION WITH TEST NUMBER 30 OBJECTIVE:

Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraphs 2.b, 4.o, and 5.h require that control rod scram testing be performed at plant conditions that bound those under which the control rods might be required to function to achieve plant shutdown.

The control rods are required to scram within the specified times specified by Plant Technical Specifications and assumed in FSAR analysis.

Test Number 5, Control Rod Drive System, demonstrates the performance of the control rod drive (CRD) system over the full range of reactor operating conditions.

It is proposed to delete scram time testing of all but four CRDs at rated reactor pressure during Test Condition Heatup and to obtain the rated reactor pressure scram data from a planned full core scram at Test Condition 2 (approximately 30% power).

DISCUSSION:

Control rod drive scram timing tests are begun during preoperational testing and all later testing is an extension of the preoperational program.

Following fuel loading, with the reactor in the cold shutdown condition, all CRDs are individually withdrawn and scram time tested.

This test provides a data base from which four control rods are selected with scram times among the longest of those measured for which exhibit unusual operating characteristics.

The "four selected CRDs" must be compatible with the Rod Worth Minimizer (RWM),

Rod Sequence Control System (RSCS) and the CRD withdrawal sequence.

These four CRDs are individually scram time tested during heatup at approximately 600 and 800 psig reactor pressure.

At rated reactor pressure and low power level, all of the CRDs are again scram time tested.

From this data base, four CRDs are selected for additional individual testing and for monitoring during full reactor scrams during the power ascension test program.

It is proposed to only test four rather than all of the CRDs at rated reactor pressure during Test Condition Heatup.

Scram Times for the remaining CRDs would be obtained from the planned reactor scram during performance of the Loss of Offsite Power Test, Test Number 30, at Test Condition 2 (approximately 30% power).

This full reactor scram would provide scram times for all CRDs except those which were not at a fully withdrawn position prior to the scram.

Four CRDs would be selected from the data base of the fully scrammed CRDs for monitoring during future full reactor Scrams.

I

.,n--

Deferment of testing for all the CRDs from rated reactor pressure / low power to about 30% power is considered acceptable because (1) all CRDs are scram time tested at cold conditions, and (2) four CRDs are monitored during reactor heatup to rated pressure to determine the effects of increasing temperature and pressure.

The combination of this data provides an indication of the performance of all CRDs at rated temperature and pressure.

Increasing reactor power level should have little effect on CRD scram times, however scram times would be measured during the first planned, full reactor scram.

In the event of an inadvertent reactor scram, recording equipment will automatically start and provide a record of CRD scram performance.

No scram time measurements at rated reactor pressure are necessary for those CRDs not fully withdrawn prior to the scram at Test Condition 2 (approximately 37 CRDs).

Analysis shows that in the extreme case of 37 Control Cell Core (CCC) CRDs plus an additional eight inoperable CRDs not being inserted during a scram (see Figure 1 for CRD pattern), the effect on negative reactivily insertion is small (see Figure 2).

This is attributed to the even distribution of the 37 CCC and eight inoperable CRDs which are assumed to not scram.

The impact on FSAR safety analysis, shown in figure 3, is also small.

The Operating Limit Minimum Critical Power Ratio (OLMCPR) increases by 0.01, from 1.20 to 1.21, if~the average scran insertion times of the other 140 CRDs approaches Plant Technical Specification limits, (See Figure 3).

The increase is zero if the average time equals that experienced by similar operating reactors, (see Figure 3), since the limiting transient at this speed is not dependent on scram.

' Plant Technical Specifications would be changed to exempt the 37 CCC rods from scram time testing and to adjust the OLMCPR according to the measured scram times of the remaining CRDs (see Attachment 1).

Although the above approach would exempt the 37 CCC CRDs from scram time measurements at rated reactor pressure, insertion indication is provided for these CRDs during all full reactor scrams.

Because they are designated CCC rods, they will always be partially inserted in the core, except at end of cycle, and thus their insertion rates cannot be exactly compared to Plant Technical Specification limits.

Nevertheless, full insertion can be verified and estimates of insertion times can be obtained from data recorded during the performance of the loss of Offsite Power test at Test Condition 2.

2

CONCLUSION:

Compensatory actions such as scram time testing of four CRDs during heatup to show the effect of increasing reactor temperature and pressure, obtaining insertion times during a full reactor scram, and increasing the OLMCPR as required, provide sufficient assurance that safety margins are maintained with power ascension testing changed to delete scram time testing of all but four CRDs at rated pressure during Test Condition Heatup.

The proposed testing. change will not adversely affect any safety systems or safe operation of the plant and thus does not involve an unreviewed safety question.

4>

3

FIGURE 1

/*'

CONTROL ROD SCRA'1 DISTRIBUTION l

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FIGURE 3 - IMPACT ON NCPR OPERATING LIMIT t.M CCC and 8 inop reds not included..in scram. reactivity i

OPERATaseG t tusf FOR CYCLE 1

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Abevod I PROOF AND REV!EW COPY i

REACTIVITY CONTROL SYSTEMS CONTROL R00 MAXIMUM _$ CRAM INSERTION TIMES I

LIMITING CS SITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 5, based on de-energization of the scr:.m pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY: CPERATIONAL CONDITIONS 1 and 2.

ACTION:

a.

With the maximum scram insertion time of one or more control rods exceed-ing 7 seconds:

1.

Declare the control rod (s) with.the slow insertion time inoperable, and s

2.

Perform the Surveillance Rirquirements of Specification 4.1.3.2.c at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 :

seconds.

s Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

(

4.1.3.2 The maximum scram insertion time of the control rods

shall be demon-strated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control rod scram time tests, the control rod drive pumps isolated from the accumulators:

a.

For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER following CORE ALTERATIONS or after a reactor shutdown that is greater than 120 days.

~b.

For specifically affected individual control rods following mainte-nance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods, and c.

For at least 10% of.the control rods, on a rotating basis, at least once per 120 days of POWER OPERATION.

1

For the initial fuel cycle only, the 37 designated Control Cell Core control rods may be exempted from scram time testing under hot, pressurized conditions, provided they meet all other surveillance (test) requirements, and that all scram reactivity requirements of the plant safety analysis are met with no t

scram contribution from the Control Cell Core control rods.

Hope Creek ES:br:pc/N110578-3 3/4 1-6 11/6/85

PR00F AND REVIEW COPY REACTIVITY _ CONTROL $YSTEMS

=*

CONTROL R00 AVERAGE SCRAM' INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.3 The average scram insertion time of all 0PERABLE control rods from the i

fully withdrawn position *, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:

Position Inserted From Average Scram Inser-Fully Withdrewg t_fon Time (Seconds) 45 0.43 39 0.86 25 1.93 05 3.49 APPLICABILITY:

OPERATIONAL CON 0!TIONS 1.and 2.

j ACTION:

=-

With the average scram insertion time exceeding any of the above limits, be in at least HOT SHUT 00WN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.3 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.

For the initial fuel cycle only, the 37 designated Control Cell Core control rods may be exempted from scram time testing under hot, pressurized condi-tions, provided they meet all surveillance (test) requirements, and that all scram reactivity requirements of the plant safety analysis are met with no scram contribution from the Control Call Core control rods (special analysis and/or refer to in bases).

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Hope Creek l

ES:br:pc/N11057"-5 3/4 1-7 l

11/6/85 l

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PROOF AND REVIEW COPY i

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g POWER DISTRIBUTION LIMITS h

3/4.2.3 MINIMUM CRITICAL POWER _ RATIO LIMITING CONDITION FOR OPERATION s 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater f

than the MCPR limit 1.20*, times the K of Figure 3.2.3-1.

f APPLICA8ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With MCPR less than the MCPR limit times the K,shown in Figure 3.2.3-1 initiate corrective action within 15 minutes ahd restore MCPR to within the required ifmit with 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL FOWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .s

- s SURVEILLANCE REQUIREMENTS 3

4.2.3 MCPR shall be determined to be' equal to or greater than the MCPR limit times the K shown in Figure 3.2.3-1:

i f

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL PCWER increase of at least 15% of RATED THERMAL POWER, and

\\

Initially and at least once per 12 Hours when the reactor is operat-c.

ing with a LIMITING CONTROL R00 PATTERN for MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

MCPR Limit of 1.20 shall be increased to 1.21 for the initial cycle when CCC l

rods are exempt from het scram testing (per T.S. 3.1.3.2 exemption note) and CCC rods are not included in the average scram time.

l l

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Hope Creek ES:br:pc/N11057*-4 3/4 2-6 11/6/85

.~...

POWER DISTRIBUTION LIMITS 9

3/4.2.3 M!N! MUM CRITICAL POWER RATIO (Optional-00YN Option B)

LIMITING CON 0! TION FOR OPERATION 3.2.3' The MINIMUM CRITICAL POWER RATIO (MC than the MCPR limit shown in Figure 3.2.3 p) sh'all be equal to or greater Irt.imes the K, shown in Figure 3.2.3-2, (provided that the end-of-cycle recirculation pump trip (E0C-RPT) system is OPERABLE per Specification 3.3.4.2), with:

m (t,y,. Tg)

(:~t 3.2.3-1 dill be.Simd@~

t, - 1, 4 u p,r U a CF M $

t=

6 where:

l

~

(0.86) Seconds, control red average scram insertion tg = time limit to-notch (39) per Specification 3.1.3.3, b

8 = (0.588) + 1.65 3 (0.052),

1 n

N I

g i=1

(

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i t,y,,

4,7 N;t, n

l I

N j

i=1 q

l 11 n = number of surveillan'.e tests performed to date in cycle,

$h Ng = number of activr, control rods measured in the i surveillance *,est, tg = average scram time to notch (39) of all rods seasured th in the i surveillance test, and total number of active rods measured in Specification Ng = 4.1.3.2.a.

APPLICABILITY:

3 OPERATIONAL CONDITION 1, yben THERHAL POWER is greater than or equal to j

(25)% of RATED THERMAL POWER.

(7 GE-STS (SWR /4) 3/4 2-6b

.... ~....

IMDM W

POWER DISTRIBUTION LIMITS i

LIMITING COWITION FOR OPE _ RATION (Continued)

ACTION

~

(a. With the end-of-cycle recirculation pump trip F ' stem inoperable per specification 3.3.4.2, operation may continue 4 id the provisions of specification 3.0.4 are not applicable providet that, within one hour, MCPR is determined to be greater than or equal to the MCpR limit shown in Figure 3.2.3-1 E0C-RPT inoperable curve, times the Kf shown in

[

Figure 3.2.3-2.)

b.

With MCPR less than the applicable MCPR limit shown in Figures 3.2.3-1 l

and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than (25)% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 1*_

4.2.3 MCPR, with:

l s

t = 1.0 prior to performance of the initial scram time measurementi a.

for the cycle in accordance with Specification 4.1.3.2, or b.

I as defined in Specification 3.2.3 used to determine the limit within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram tima surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or, greater than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.-2.3-2:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of j

at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is c.

operating with a LIMITING CONTROL ROD PATTERN for MCPR.

l e-GE-STS(ShR/4) 3/4 2-7b

PUBLIC SERVICE ELECTRIC AND GAS COMPANY HOPE CREEK PROJECT SAFETY EVALUATION No. PSE-SE-I-027 TITLE:

SUBSTITUTION OF INDIVIDUAL CONTROL ROD SCRAM TIME TESTING AT HOT STANDBY WITH FULL REACTOR SCRAM CONTROL ROD TIME TESTING AT TEST CONDITION 2 Date:

// l27 /gf f

/

1.0 PURPOSE The purpose of this safety evaluation is to document the results of the evaluation performed to ensure that substitution of individual control rod scram time testing at Hot Standby with full reactor scram control rod time testing at Test Condition 2 will not adversely affect reactor safety.

2.0 SCOPE The scope of this safety evaluation is the adequacy of Hope Creek's power ascension test program as it concerns control rod scram time testing.

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.

General Electric Preoperational Test Specification, 22A2271, Revision 2 4.0 DISCUSSION Regulatory Guide 1.68 (Revision 2, August 1978), Appendix A, paragraphs 2.0, 4.0, and 5.h require that control rod scram testing be performed at plant conditions that bound those under which the control rods might bn required to function to achieve plant shutdown.

The control rods are required to scram within times specified by Hope Creek's PSE-SE-Z-027 1 of 4

Draft. Technical Specifications and assumed in Hope Creek's Final Safety Analysis Report (FSAR) analyses.

Control rod scram time testing is performed before fuel is loaded as part of the Preoperational Test Program and

'at several conditions after fuel is loaded as part of the Power Ascension Test Program.

Specifically the Preoperational Test Program includes scram time testing all control rods individually and simultaneously with zero psig reactor vessel pressure.

Test Number 5 of the Startup Test Program includes scram time testing all control rods individually with zero psig reactor vessel pressure, 4 selected control rods with 600 psig and 800 psig reactor vessel pressure, all control rods individually at rated pressure and-4 selected control rods at-Test Condition 2, 3, and 6 during planned reactor scrams.

The 4 control rods selected for the additional 4

monitorin'g are selected based on slow scram times or z

unusual operating characteristics identified by the testing at zero reactor pressure or rated reactor pressure.

In addition, the 4 selected control rods must also be compatible with the Rod Worth Minimizer, Reactor Sequence Control System, and control rod withdrawal sequence.

It is proposed to substitute scram time testing all control rods individually at rated reactor vessel pressure during Test Condition Heatup with scram time testing control rods during a full reactor scram at Test Condition 2.

The "four selected control rods" will be scram time tested at rated reactor vessel pressure however, as will be done at 600 psig and 800 psig reactor vessel pressures.

The proposed testing will provide scram times from a fully withdrawn position for all control rods except for approximately 37 to 45 control rods which will be i

partially inserted prior to the scram to control reactor power and flux shapes.

Absence of full scram time data for these control rods at rated pressure is considered acceptable for three reasons.

Firstly, all control rods are scram time tested at cold shutdown conditions.

In I

addition,.the scram times of four control' rods will be l

checked at 600 psig, 800 psig,-and rated reactor pressure during initial reactor.heatup to determine the effects of increasing temperatures and pressure on scram times.

The combinat' ion of this data will provide an indication of the performances of all control rods at~ rated temperature-and pressure.. Based on previous startup experience, increasing reactor power level.should have little effect on control. rod' scram times.

PS E-S E-Z-0 27_

2 of 4

Secondly, analysis shows that in the extreme case of 37 Control Cell Core (CCC) control rods plus an additional eight control rods not being inserted at all during a scram (See Figure 1 for assumed control rod pattern), the effect on negative reactivity insertion is small (Figure 2).

The impact on the PSAR safety analyses (Figure 3) is also small.

Specifically, the Operating Limit Minimum Critical Power Ratio (OLMCPR) will only increase by 0.01, from 1.20 to 1.21, if the average scram insertion times of the 140 control rods which fully scram approaches Hope Creek's Draft Technical Specification Limit ('Ei).

There will be no OLMCPR increase if the average scram times equal that experienced by similar operating reactors (T ).

g Lastly, scram times will be recorded for the control rods which will not be fully withdrawn during the full reactor scram test at Test Condition 2.

Although these times will not be from a fully withdrawn position, this data will provide an indication of the proper operation of these control rods.

Thirty seven of the control rods which will not be fully withdrawn are Control Cell Core control rods which, for the majority of'the cycle, will be at an intermediate position.

It is acceptable to defer scram time testing of all-control rods from Hot Standby to Test Condition 2 based on the scram time testing which will be' performed prior to Test Condition 2 as discussed above.

In addition, plant Technical Specification 3/4.1.3.2 allows reactor operation up to 40% of rated thermal power following core alternations before requiring all' control rods to be scram time tested.

In the event of an inadvertent reactor scram, recording equipment will automatically start and provide a record of control rod scram performance.

5.0 CONCLUSION

Substitution of individual control rod scram time testing 4

at Hot Standby with full reactor scram control rod time testing at Test Condition 2 as discussed above will not increase the probability of occurence or the consequence of an accident or malfunction of any safety system and will not adversely affect the safe operation of the plant.

In addition, the safety margin as defined in Hope Creek's Draft Technical Specification is notfreduced.

Consequently, an unreviewed safety question does not exist.

PSE-SE-2-027 3 of 4

6.0 DOCUMENTS GENERATED None 7.0 RECOMMENDATIONS Revision to Hope Creek's FSAR, Draft Technical Specification and power ascension test procedures shall be made to reflect the substitution of single control rod scram time testing at hot standby with full reactor scram'.

control rod time testing at Test Condition 2.

~

8.0 ATTACHMENTS t

Figure 1 Control Rod Scram Distribution Figure 2 Control Rod Hot-Scram Reactivity Addition Figure 3 Impact on Operating Limit Minimum Critical Power Ratio (OLMCPR) 9.0 SIGNATURES Originator (b S. @ O _

86.

///2 >/6 Verifier

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Systems Analysis Group Head

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I Site Engineering Manager _

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FICURE 3 - IMPACT ON tiCPR OPERATING LIMIT

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O ATTACHMENT II I

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lx l

l lz l

l (5) metween 80 and 90 percent there-1 power, l SG l Arms Calibration l

lz la lx la l

lx lx l

l and near 100 percent core flow l st l Process coputer lx l

lx(3) l lx l

lx l

l l

l 12 l acic l

lz la l

l l

(6) mas pu nuncut capabtitty a ascsec pump l 13 l meet l

lx l l

la l

l l

ambeck must ham already been mpleted i 14 l s t.cted process Temp l

lx l l

lx lx(4) l lx(4) l l

l 15 l meer savel aof tog Top l

lx l l

lx l

l la l

I (7) me.ctor power between s0 eaa

percent l 15 l sveta= mesmeton lx l

la l

l lx l

l l 16 l Txp moertainty l

l l

l la l

l (8) meactor power between 45 and 65 mrcent l 17 l core performance l

l lx lx lx la la la l x l

l to I steam praenetton l

l l

l l x l

(9) menctor po=or betwees. 75 ame 90 r* rcer.

l se l core sw-void made aceponsal l

l l

l la la l

l l

l 20 l Prenews mogelator l

l lx lx la lx lx lx l

l (10) At mesimum peer that will act ch e scram l 21 l mes sy -sespoint changes l lx la la lx la la lx l

l l 2s l pees are-soes ow u.atine l

l l

lx(5) l l

(111 perform between test consitions 1 -ad 3 l 2s l peesueter peep Trip l

l l

lats) l lst ?) 'L l 21 l mes pu amant capability l

l' l

l l

(12) anactor po=r het-on 40 ame 55 par. ent l 22 i Twbine m1 smettlance l l

l x(181lx(12)lat e) l l

l gx(9) lxt :0 ) g g

l 23 l ms1T punctsonal seet l

lx l

l in(13)l l

l (13) me ctor power between 60 ans e5 perc ent l 23 l ImIT m 11 Isolation l

l lx(20)l:

lx(20)l l

la(20)l l

(14) metusen test conditions 2 and 3 l

l l

l lx l

l l 24 l Belief miose l

lE l

l 25 i Twhane Trip a toma l

l l

lx(15) lst '63 l l

lati?)l l

l l Sejectson l

l l

l (15) Generator load rejection, within bypaan l 26 j shetdoen Outside CaC l

l l

lx l

l l

l valve capacity l 27 l Reciraalation Flow Control l l

l lx(I4Il l

lx( 18)l l

l l 28 l Rectro-one pump Trip l

l l

l lx l

l (16) meactor power between 60 and 80 percent l M l spT Trip 'hso pumpe l

l l

l lx(193l l

l l

l at core flow L 95 percent - turMne trip l M l Baciro S t e perfosmance l l

l lx la lx l

lx l

l l M l Socirc pump ammheck l

l l

l lx l

l l

(17) load rejection l M l Socirc sys cavitatica l

l l

l lx l

l l

l 30 l tese of Offette Pwr l

l lx l

l l

(10) metween test conditions 5 and 6 l 39 l Pipe Vibratloa l

lx lx la lx l

l lx l

l l 29 l ascito Flow Calibration l

l l

l lx l

l la l

l (19) >$04 power and 195 core flow, and performed l 32 l NN33 l

lx l l

l l

before Turbine Trip & toad aejection l 33 l Bila l

l l

lx l

l l

lu(211l l

l 34 l orrwell a seman Tenel l

lx lx l

lx l

l lx l

l (20) check saw est pointe during major scram l

l cooline l

l l

l seste no,E caEEn l 35 l Gesecus madeste l

l lx l

fx l

l lx l

l GEIAE aATieIG STATION l 38 l SACS perfosmance l

l l

l lx l

l (21)' aerformed &aring cooldown from test F180AL SAFETV AsaALYS4S aEPOaT l do l confirmatory In-Plant Test l l

l lx l

l l

l condition 6 rsaa 3/7 (22) The test number correlates to FSAn section g

14.2.12.3.x there a le the indicated test number.

FtGuaE 14.26 10,06/06

)

/

HCGS FSAR 5/85

+

their recommendations.

This report must discuss alternatives of action, as well as the concluding recommendation, so that it can be evaluated-by all related parties.

During performance of.startup tests, technical specifications override any test in progress if plant conditions dictate.

14.2.12.3 Startup Test Procedures 14.2.12.3.1 Chemical and Radiochemical." : iter: :=d S-.-1 Oy;te;;

Objective [

a.

st provides verification of the sample systems' ability 1.

Maintain quality control o ant systems' chemistry and ensure that sampling

ent, procedures, and analytical techniques supp T-a s rser Osw o e sru r+r TH+ Pea ro e w A ucc eF ref L l2+4eron W A T&rt C L-&A du p Ado tod eFDCA r+

4F wide,e xe sqpa Sytr&ws.

l G

l 14.2-153 Amendment 10 l

)

HCGS FSAR 10/84 q

)

required to demonstrate that fluids meet quali ifications and process requirements 2.

Monitor fuel integrity, tion of filters and domineralizers,. condenser tube

city,

~ operation of the offgas system and s J.

separator-dryer, and tuning of system moni j'

i 4'

b.

Prerequisites 1 {I 5A=

Intrument calibration and preoperational testing of

[i t

05::10:1, ::dicti:n, nd

dicch::ical senite ; have been completed.

o23 9C

, a s-R u> c a A WO coajo sa sn r c-c.

Test Method 0+*w(-ad u s'c syrreu r g$9 u

b.r

$ p,j ior to fuel loading, a complete set of chemical and e

ra chemical samples are taken to ensure that all o

6 *1 4

sample tations are functioning properly and to

}ji 4

determine e initial concentrations.

During reactor wo heatup, subs ont to fuel loading, samples are taken and measurement ade at each major power level plateau 3$

to determine-the c ical and radiochemical quality of a

reactor water and rea e feedwater, amount of a

3 e

radiolytic gas in the at gaseous activities after S

4 the air ejectors, decay tim n the gaseous radweste lines, and performance of filt and demineralizers.

{.<)3 c

Baseline data for the main steam cess radiation g

monitoring subsystems and the offgas nitoring 4 g g.

subsystems is also taken at each major er level 3

ay plateau.

Adjustments are made, as require to ge33 monitors in the liquid waste management syste LWMS),

liquid process lines, and offgas treatment syst d.

Acceptance Criteria Level 1:

vor APPuc A Soc l

cal and radiochemical, and water of quality factors are within the technical specifications and fue requirements.

Gaseous, particulate, and liquid e ctivities shall conform with Technical Specifications.

~

14.2-154 Amendment 8

1 HCGS FSAR 10/84 d.

Acceptance Criteria Level 1:

The shutdown margin measurements shall verify that the core remains subcritical with the most reactive control rod withdrawn and all other control rods fully inserted by at least 0.38% AK/K.

Level 2:

Criticality should occur within 1.0% AK/K of the predicted critical.

14.2.12.3.5 Control Rod Drive Sys, emq a. Objective s L

y G l?

The test objective is to obtain the baseline data for $g % 2 the CRD system, and to demonstrate that the system ( 5 operates over the full range of primary coolant gM %j conditions, from ambient to operating. o9 i a T 39 b. Prerequisites EI o' 1 $ m g$ Preoperational testing of the CRD system has been completed and the system is ready for operation. g g d}1 ) =gg $ c. Test Method o 4 3 w 3 I The startup tests performed on the CRD system are an EI extension of the preoperational tests. Initial post 4 d {vc fuel load tests with zero reactor pressure include 2a 3 position indication, normal insert / withdraw stroking, S 's y $ friction testing, and scram testing. Coupling checks L s o are verified using station operating procedures. q 2o,g Following initial heatup to rated reactor pressure, the b 334 friction and scram test is accomplished. Tell;;ing q e 3 initi:1 h::t;;, the four slowest CRDs are measured for 9gG5& scram times following planned reactor scram as detailed on Figure 14.2-5. In addition,-proper response of the CRD flow control valve will be verified. 14.2-157 Amendment 8 -tw.-,--- -,,aw---,,ev-ww-,ww,,rww-,+w

HCGS FSAR 5/85 s -s -d. Acceptance Criteria ~ Level 1 ) The withdrawal speeds and scram times shall meet the requirements of the GE startup test specifications, Level 2 Th's. friction test results shall meet the requirements of the GE startup test specifications. 14.2.12.3.6 Source Range Monitor Performance i I a. Objective The test objective is to demonstrate that the neutron i sources, SRM instrumentation, and rod withdrawal sequences provide adequate information to achieve criticality in a safe and effici,ent manner. i b. Prerequisites Fuel loading is complete, neutron sources have been installed, and all control rods have been inserted. The CRD system is operational. i i i I i l l 14.2-158 Amendment 10

/ HCGS FSAR TABLE 15.0-4 Required Operating Limit MCPR Values OLMCPR OLMCPR Pressurization Events (Option A)(1) (Option B)(1) Load rejection with bypass 4-4+ t.lf, h e6-1.07 Load rejection without bypass 4-44-l.20 4-tt f.17-Turbine trip with bypass hi l. t 35 htf I. o 75~ Turbine trip without bypass F-F71.sq Fr&P 1.si Feedwater controller failure j 'with bypass 4-44 1.2.1 1.17 1.19 Nonpressurization Events OLMCPR Loss of feedwater heating, MFC 1.20 Rod withdrawal error (RBM=106%) 1.20 (1) Option A and B include adjustment factors as specified in Reference 15.0.2. O

M" e. FIGURE 3 - IMPACT ON NCPR OPERATING LIMIT t.% CCC and 8 inop mds not included..in scram yeactivity w ,s OPT 51ATING L9MsT FOR CYCLE 1 ROO WIIHEM4 AWAL (5184064 OR t 055 04 i L t l>WA TEI) 86( AT E R ' - ~ / - ./

5 FWCF 3

W*ay, Ass g t e ^* '"'* - / / g, Ait a coa' f'- u o E / / u- / " s g s'/ / / - us / / / / gM n $'** / / z LRNBP /v / gc / po ",_ g'g e4',p ' 3 u_______.- TTNBP ss#- gg - y \\ 8 g / ,p#"p,,/ E -a tR c. fee * '" ' g 3 g. sm m y z 2o -j >cm Q ">4 S 9 os a ' Ave 'n ' Avr " 'n ' AVE " 'A n, m - M AVEf t AGE SCf4 AM SPE F D E 3 I. a = P C L 4 _}}

ML20138A656

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ML20138A656

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Dear Ms. Adensam:

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5.0 CONCLUSION

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