ML17276B019

Forwards Documentation Outlining Agreements & Commitments Reached at 820107 Meeting W/Nrc to Resolve Potential SER Open Items from Initial Test Program

ML17276B019
Person / Time
Site:	Columbia
Issue date:	01/19/1982
From:	Bouchey G WASHINGTON PUBLIC POWER SUPPLY SYSTEM
To:	Schwencer A Office of Nuclear Reactor Regulation
References
GO2-82-83, NUDOCS 8202020256
Download: ML17276B019 (50)

Text

SUBJECT:

Forwards documentation outlining agreements L'commitments reached at 820107 meetin'g w/NRC to resolve potential SER open items from ini.tial [test program.

DISTRIBUTION 'CODE:

B001S COPIES 'RECEIVED:LTR, P ENCL:J

-TITLE: PSAR/FSAR AMDTS and Related Cot respondence SIZE:,

NOTES:2 copies all matl:PP

~

)

REGULATORY 'RMATION DISTRIBUTION SY,

'4 (R IDS)

ACCE'SSION NBR: 8202020256 DOC ~ DATE: &2/01/19 NOTARIZED; NO 1'"l FACIL:50-397 WPPSS Nuclear Project<

Unit 2i Washington Public Powe

'AUTH ~ NAME, AUTHOR AFFILIATION BOUCHEYgG,D.

Washington.Public Power Supply System RECIP ~ NAME RECIPIENT AFFILIATION SCHWA'ENCERgA ~

Licensing Branch '2 DOCKET 05000397 05000397 RECIPIENT ID CODE/NAME ACTION:

'A/D LICENSNG LIC BR 02 LA INTERNAL: ELD IE/DEP/EPDB 35 MPA NRR/DE/EQB 13 NRR/DE/HGEB 30 NRR/DE/MTEB 17 NRR/DE/SAB 24 NRR/DHFS/HFEB40 NRR/DHFS/OLB 34 NRR/DS I/AEB 26 NRR/DSI/CPB 10 NRR/DS I/ETSB 12 NRR/DS I/PS 8 19 NRR/DSI/RSB 23 04 EXTERNAL; ACRS 41 FEMA"REP OIV 39 NRC PDR 02 NTIS

'COPIES LTTR ENCL 0

0 1

0 1

1 1

'0 3

3 2

'2 1

1 1

1i,i 1

1 1

1 1

1 1

1 1

1 1

1 1

1 16 16 1

1 1

1 1

1 RECIPIENT ID CODE/NAPE LIC BR Pg

~BC AULUCKgR~

01 IE 06 IE/DEP/EPLB 36 NRR/DE/CEB 11 NRR/DE/GB 28 NRR/DE/MFB 18 NRR/DE/QAB 21 NRR/DE/SEB 25'RR/DHFS/LQB32 NRR/DHFS/PTRB20 NRR/DSI/ASS 27 NRR/DS I/CSB 09 NRR/DS I/IiCSB 16 NRR/DSI/RAB 22 NRR/DST/LGB 33 BNL(AMDTS,ONLY)

LPDR 03 NSIC 05 COPIES LTTR ENCL 1

0 1

3'3 3

'2 2

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

6o TOTAL NUMBER OF COPIES REQUIRED'TTR

~

ENCL

A 4

E Washington Public Power Supply System P.O. Box 968 3000 George Washington Way Richland, Washington 99352 (509) 372-5000 January 19, 1982 G02-82-83 SS-L-02-PLP-82-002, Mr. A. Schwencer, Director Licensing Branch No.

2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C.

20555 REQEIQEQ 2 1S82e g

0 INmgzgqy~

tiC55~ g

Reference:

NUCLEAR PROJECT NO.

2 MINUTES OF THE NRC/WNP-2 MEETING TO RESOLVE POTENTIAL SER OPEN ITEMS FROM THE INITIAL TEST PROGRAM (01/07/82)

Letter, GD Bouchey to A Schwencer, "Request for Additional Information Regarding the Initial Test Program for WNP-2 SER Open Items", dated December 30, 1981 The referenced letter requested a meeting to resolve disagreements between the Supply System and the NRC on several of the issues forwarded with that letter.

The enclosed attachments (7 copies) represent the agreements and commitments reached at the resulting meeting (January 7, 1982).

Attachment 1 is a list of attendees, Attachment 2 is the NRC concerns as presented at the meeting by the NRC consultant and Attachment 3 is the agreed upon re-solution and commitments corresponding to the concerns in Attachment 2.

Please contact Mr. R.M. Nelson, Project Licensing Manager, WNP-2, if further clarification is required.

Very truly yours, G.

D. Bouchey Deputy Director, Safety and Security PLP/jca Attachments (3) cc:

WJ Apley Battelle R

Auluck -

NRC WS Chin BPA R

Feil NRC Site I

gnl

E

'E

/

!l

~

.,~

lr if l

I

WNP-2 FSAR,CHAPTER 14 NRC SER MEETING 1/7/82 ATTENDEES A. Wood Donald C. Fischer W. J. Apley.

Dave Whitcomb L. D. Kassakatis P. L. Powell G. L. Blackburn J. J."Bufis C.

M. Powers G.

K. Afflerbach J.

D. Martin R. A. Feil SS USNRC BNW SS SS SS Licensing SS SS SS SS SS NRC Attachment 1

S l

I

v v

~

4

" COWENTS ON WNP-2 RESPONSES 423.21 Answer does not state if the HPCS will be tested at normal operating pressure.

.Req:

Rea Guide 1.68,App A,l.h.l.c (p.1.68-9,10)

.423.30 1..'"::. OK 3a.

4. -.

6.

8.

10.

lib.

12.

13.

OK Answer states WNP-2 does not intend to do the Rod Sequence Exchange Startup Test.

Rea:

Reg Guide 1.68,App A,5.d (p.1.68-14)

OK

. Answer states WNP-2 does not intend to measure bypass valve capacity.

Req:

Reg Guide 1.68,App A,5.t(p.l.68-15) guestion stated that it was felt that the

+ 2 ~ criteria on total SRV flow was excessively

severe, given the +35%, -105 individual valve capacity criteria.

This is not a safety question and need not be pursued.

OK Answer states that the reactor scram and tlSIV

. isolation will not be performed from outs'de the control room, as this is contrary to WNP-2 written procedures.

Req:

Reg Guide 1.68.2,Rev 1, Section C.l 8 C.3

'(p. 1.68.2-2)

Answer states that a simultaneous trip of both recirculation pumps at 100K power will not be performed, as it will cause a scram and the same data is obtainable from other tests.

Req:

1.68,Rev l,App A,5.ii (p.1.68-16) or 1.68,Rev O,App A,0.2.r (p. 1.68-7) 14.

OK Attachment 2

0 page 2 of 2 423.33 423.39 423,41 423.41.3 423.41.4 423.42.7 Errata STP 8 24 has been added to Table 14.2-3, but the test conditions have not been specified.

Answer states that MNP-2 will not perform an analytical determination of the process variable-to-input response times for the Reactor Protection System (RPS).

Req:

1.68,Rev 1,App A,l.j (p.

1.68-9,10) l.

Answer appears to be OK but no draft of 14.2.12.1.37 was provided.

2.c Answer appears to be OK but no draft of 14.2.12.1.32 or 14.2.12.1.35 was provided.

2.d Answer states that test acceptance criteria for the hot pipe penetration testinq have not been developed to date.

2.e OK 2.g Answer appears to be OK but no draft of 14.2.12.1.48 was provided.

2.i Answer appears to be.OK but no draft of 14.2."12.1.17 was provided.

Answer states that the Control I Instrument Air System will be tested using an Acceptance Test.

The MNP-2 Control / Instrument Air System is non-safety related and the Acceptance Test meets the intent of Regulatory Guide 1.80.

Answer states that Regulatory Guide 1.108 is not applicable to llNP-2. It is (Rev 1 not Rev 0) and an abstract including the applicable testing will be required.

OK if note 10 is deleted from Table 14.2-4.

2.

3.

OK OK OK

.OK

IV

AGREEMENTS AND COMMITMENTS RESULTS FROM HNP-2 FSAR CHAPTER 14 NRC SER MEETING 1/7/82 Attachment 3

4

'l I

FSAR 14.2.12.1.14c(5) requires High Pressure Core Spray (HPCS) system verification of flow paths and flow rates.

The HPCS pump will be operated over the full spectrum of flows and expected discharge pressures, and the results compared to pump curves.

This test will show what the flow rate will be under any operating pressure.

4.0 NRC will evaluate rod sequence exchange tests at plants prior to Browns Ferry 1 and 2 and determine applicability to WNP-2.

(Cooper, Fitzpatrick, Hatch and Peach Bottom plants did not perform the subject tests.

WNP-2 contends that the value of the data from the tests is not commensurate with the costs of performing the tests:

reduction of power for an extended period.)

WNP-2 will commit to accomplish the subject tests at the first normally scheduled rod sequence exchange evolution.

(Subsequent communication with the NRC staff indicates that this position is acceptable.)

I l

t

8.0 WNP-2 will determine turbine bypass valve capacity and compare with developed acceptance criteria based on transient analysis assumptions.

0 r

E

423.30 10 OK

h

For purposes of this test, reactor scram and MSIV isolation will be accomplished from the control room..

A caution will be added to the WHP-2 emergency procedures to direct an operator to open the RPS MG set output'reakers in the event that reactor scram and MSIV isolation cannot be accomplished during control room evacuation.

13 The WNP-2 position is that the turbi~e trip at 100K power transient envelopes the simultaneous

. recirculation pump trip transient.

WNP-2 will expand this response to include engineering analysis to demonstrate that the two.pump trip is enveloped and not required.

(Response attached)

Flow coast down data will be obtained during the turbine trip at lOOX power test.

Subsequent NRC review of the attached response indicates this position is acceptable.

I

423.30.13 The design dynamic response of the plant to.a simultaneous trip of both recir-culation pumps at 100% power produces a reactor scram.

Our test program does

.not initiate transients solely to confirm that a reactor scram occurs when it should.

All data pertinent to a two pump trip from high power transient is obtained in conjunction with other tests (e.g.,

STI k'27 performed 9 lOOX power).

A two recirculation pump trip transient is performed in the test program from Test Condition 3 (approximately 100% reactor core flow and ap-proximately 75K reactor power).

The two pump trip is not done at Test Condi-tion 6 (approximately 100K reactor core flow and approximately 100% reactor power) because the turbine trip transient done from that condition is a more severe transient which still yields all the significant data which would be obtained from the pump trip transient.

The specifics of the comparison of the turbine trip transient with bypass capability and the two recirculation pump trip transient are presented in FSAR sections 15.2.3 and 15.3.1 respectively (attached).

Both these analyses start from lON of rated nuclear boiler steam flow and 100K core flow.

Tables 15.2-3 and 15;3-2 list the sequence of events for each of these transients.

At 0.01 second into the turbine trip transient the RPT (recirculation pump trip) circuitry initiates the tripping of both recirculation pumps and by 0.14 seconds the pump trip actuation is complete.

Also at 4.33 seconds into the recirculation pump trip transient a turbine trip is initiated by high reactor water level (L8).

Thus, both transients proceed toward the same final condition (i.e.,

a turbine trip).

-,As, is apparent from the information in these sections, the.turbine trip is a more severe transient.

For the.two recirculation pump trip, the turbine is 'available to take vessel steam flow during the first 4.33 seconds of the transient while the flow coast down is r educing reactor power from 105% down to about 40K at which time the turbine trip and scram takes place and only the turbine bypass capacity remains (see Figure 15.3-2).

However, during the turbine trip with bypass transient, the turbine is not available to take steam (beyond the bypass capacity) from the very start and the results from both the'essel pressurization and the core response perspectives are more severe (see Figure 15.2-3).

The following lists the significant parameters from the FSAR sections for each transient:

lMO RECIRC PUMP TRIP MCPR remains at 1.24 Neutron flux decreases Fuel sur face heat flux decreases Vessel dome pressure increases by only about 80 psig

423.30.13 (Cont'd)

TURBINE TRIP MITH'BYPASS MCPR falls to 1.18 Neutron flux peaks at 147.5%

Fuel surface heat flux peaks at 101.7%

Vessel dome pressure peaks at 1136 psig In addition, during the turbine trip transient all of the relevant recircu-lation system performance parameters will be automatically recorded to document the flow coastdown phenomenon and will be analyzed as appropriate.

Thus, the two recirculation pump trip from Test Condition 6 is not a necessary part of this test program and can therefore be excluded.

423.33 The subject table has been modified as attached.

0 k.

1

~

l

I

)

ThDM 14+2>>3

~

~

~

~

SThDILLITY TESTS TEST CONDITION hPPHOXIHhTB POHPiH i Hated 20 40 4o-vs

.qs-4 -<c 60-75100

@50.

'TBS1'O TEST TITLE hPPROXIMME CORE FLOH

% Hated 37 50 100 55 100 BC

~

22 22 23

~'3 29 Core PoMer " Void Node

Response

Press.

Hog+ Setpoint Changes Press.

Heg. Dackup Regulator

'H Systeme Hater Level Setpoint Change FH Systole lleater Loss Heoiroulation PloM Control System I

Tost Condition X

X X

X X

X X

X X

X X

X X

X X

X X

X X

~

X X

X 2

3 5

6'

~

\\

I

~

0 P

The Reactor Protection System (RPS) is designed to meet the intent of IEEE-279 for nuclear power plant protection systems.

The measurement of the response time as required by the Technical Specifications provides assurance that the protective functions ae completed within the time limit assumed in the accident analysis.

The time limits in the Technical Specifications are measurements from the input of the sensor to the reactor protection scram solenoids.

These limits will be verified in the Reactor Protection System Preoperational Test (14.2. 12. 1.18).

There are no known requirements for l<NP-2 to perform an analytical determination of the process variable-to-input response times and add the delay to the time measured in the reactor protection system preoperational test.

Therefore, we are not planning to include this in our surveillance of RPS sensor response.

OK

423.41 l., 2.c, 2.e, 2.g, 2.i.

OK

t I

1 II,

]

<~

I

423.41.2.d NRC provided the Grand Gulf coomitment on this issue for WNP-2 information.

WNP-2 will evaluate the Grand Gulf information and provide similar acceptance criteria specific to WNP-2.

I H

OK

c 5

423.41.4 NNP-2 will comply with Reg.

Guide 1.108 Rev.

1, 1977.

Test abstract (14.2.12.1.37) indicating compliance to Reg.

Guide 1.108 Rev. 1,1977 will be provided (indicating the 69/N start test).

attached.

~ ~

tat mamD Mo.. 12

. November 1980 14.2.12 1.37 Koss of Power and Safety Testing Preoperational Test a

Porose To verify the operaton of the 230/115KV, 6.9KVp 4.16KV and 4&OV distribution systems To verify the integrated ability of the plant electricaL distribution and safety systems to operate on normal and standby power sources during accident conditions.

To verify that loss of a single AC or DC distri-bution system. division (exclusive of the HPCS diesel-generator and batteries) will not prevent the remaining systems from actuating during an accident condition ay~

g ppsS

h. ~1 The System Lineup Testsihave been cempieted and "

the TNG has reviewed an'd: approved the procedu're and the Startup Superintendent has approved the-initiation. of testing The 125V DC system.and the ECCS are available to support testing c

eneraL Test Methods an Acce tance Criteria, Verification of the 230/1'l5KV, 6..9KV, 4.16KV and 4&OV distribution system operability shall be demonstrated ky the-following:

(1) Demonstration of circuit integrity aid integrated operation of circuit 1xeakers, controls and interlocks, instrumentation, automatic transf'er features and. protective devices and alarms.

(2) Demonstration of proper system response to -a Loqs of the 230KV and tt5KF distribution systems independently and. simultaneously both with and without KOCA/Containment Isolation --

signals (3) Demonstration of proper system response -to -a

'oss of the 230/1 t5KV distribution systems --

and one individual standby DieseLMenerator"-

during an ECCS/Containment Isolation actuation.

14. 2-70

423.42.7 WNP-2 intends to meet the requirement of R.G. 1.68 by performing the MSIV full isolation at test condition 6.

WNP-2 will provide justification for note 10 by 1/9/82:

If sufficient justification is not provided, note 10 will be deleted.

(Note 10 has subsequently been deleted.

This issue is closed with this action.)

1

Subsequent to the 1/7/82 meeting the NRC requested a Supply System position on the simulated loss of all A/C power test to be performed at the LaSalle and Grand Gulf Power Stations, subject to the satisfactory safety evaluations.

In response, item IG1 of the WNP-2 FSAR Appendix 8 (Attached) has been modified to indicate the Supply System position.

4

~ ) ~

WNP-2 AM DMENT NO 17 July 1981 I G; 1 PREOPERATIONAZs AND COW-POWER TESTING PoSition (NUREG-0660)

The objective is to increase the capability of the shift crews to operate facilities in a safe and competent manner by assuring that training for plant changes and off-normal events is conducted.

Near-term operating license facilities will be required to develop and implement intensified training exec<<

cises during the low-power testing programs.

This may involve the repetition of startup tests on different shifts for training purposes.

Based on,experiences from the near-term operating license facilities, requirements may be applied to other new facilities or incorporated into the plant drill requirement (Item I.A.2.5).

Review comprehensiveness of test programs.

NRR will require new operating licensees to conduct a set of low-power tests to accomplish the requirement.

The set of tests will be determined on a case-by-case basis for the

'first few plants.

Then NRR will develop acceptance criteria for low-power test programs to provide "hands on" training for plant evaluation and off-normal events for each operating shift. It is not expected that all tests will be required to be conducted by each operating shift.

Observation by one shift of training of another shift may be acceptable.

NRR will develop criteria in conjunction with initial near-t'erm operating license reviews.

Licensees will (1) define training plan prior to loading fuel, and (2) conduct training prior to full-power operation.

Clarification None WNP-2 Position The Supply System is committed to meet the intent of NUREG-0660 by performance of a special low power test subprogram which provides supplemental operator training in the areas of response to abnormal plant conditions and fami-liarity with critical systems.

The special.subprogram will amplify the well-established training value of the present Startup Test Program (STP) through (1) instruction on-the con-tent, goals, and requirements of the existing program, (2) addition of selected special tests to the STP to demonstrate abnormal scenarios and use of critical sy'stems and/or emergency operating procedures to control them, and (3) utili-B.1-45

AMENDMENT NO 17 July 1981 zation of-the knowledge and experience gained during the STP in the training programs for future operators.

The overall Startup Test Program is outlined in Chapter 14 while the conduct of operations is discussed in Chapter 13.

.During the preoperational and power ascension test phases<.

the operations personnel will be intimately involved in the per-formance of the various test procedures.

With the impetus provided by the responsible test phase organization<

the operations staff is charged with establishing the required plant/system conditions, initiating and controlling the desired test transient and returning the plant/system to its normal condition.

The operations staff provides. the physi-cal ability to accomplish the Startup Test Program.

In this

fashion, the completion of, the Startup Test Program provides an unparalleled training opportunity for the operators.

The following outlines those additional actions the Supply System will implement to augment the extensive training bene-

'fits inherent in the existing STP program:

I.

Development and Implementation of a Training Course on the STP A.

General Classroom Instruction (Prior to testing) 1)

STP Overview a)

Organization, Delineation of Responsibilities, Goals b)

Administrative and Emergency Procedures c)

Preop and Power Ascension Test Schedule 2)

Review Selected STP Specifics, for example; a)

Pertinent Preop Test Purposes, Procedures>

Anticipated Results b)

Integrated System Cold Punctional Tests c)

Fuel Loading, Heatup, Power Ascension Test

Purposes, Procedures, Anticipated Results

'd) 'pecial Test Subprogram Test Purposes, Procedures, Anticipated Results 3)

Review Expected Utilization of STP Data F 1-46

1'

WNF-2 AM MENT NO 17 July 1981 a)

Documentation of Plant Safety

~ b)

Feedback/Confirmation of Anticipated Results B.

Test Phase Instruction Performed by Test Director on a Shi<t Basis (during, testing) 1)

Review of the Immediate Test Schedule 2) 3)

Discussion of the Impending Tests:

Procedures, Anticipated results, Precautions II Review/Disseminate Plant Response Data from Previous Shift(s)

C.

Post-STP Completion Instruction Performed by Test Director (following testing)

')

Review of the Actual STP Results vs. Anticipated Results 2)

- Review Plant Design Changes/System Modifications Required Development and Performance of a Special Test Subprogram A.

Additional RCZC System Tests 1)

RCZC Operat'ion Following Loss of AC Power to the System 2)

RCIC Operation to Prove DC Separation B.

Integrated Reactor Vessel Level Instrumentation Functional Test C.

Integrated Containment Pressure Instrumentation Functional Test D.

Simulated Loss of Control and Instrument Air Test E.

Repetition of Some Normal STP Tests', for example 1)

.Feedwater Pump Trip/Recirc Runback Demonstration 2)

Turbine Trip/Generator Load Rejection Within Bypass Valve Capacity 3)

Pressure Regulator Setpoint Changes

t~

I

,4

)

ll.

AMENDMENT NQ

$ 7 July 1981 4}

Recirculation Pump Trips 5)

RHR Steam Condensing Mode Operation 6)

Peedwater Level Setpoint Changes ZII. Utilization of the STP Data A.

Refine the WNP-2 Simulator Response

Models, as appropriate B.

Incorporate a Major Plant Transient

Response

Section in Operator Training Program, as appropriate C.

Update License Program Training and Requalification

Material, as appropriate.

~> Cic.x+&

It is anticipated that every

'ember of the opera-tions staff will obtain valuable knowledge and experience through participation in the WNP-2 Startup Test>Program.

Each will receive appropriate classroom iqstructiong~ through judi-cious scheduling of tests, most will

<Y$~eh, to a variety of plant/system transient responses (or review of results thereof} ~ aag~e training received will be con-tinually re-enforced through normal requalification program refinements.

Future license. candidates will also benefit from the training material upgrades resulting from the STP experience.

With this program outline, the Supply System is meeting the intent of NUREG-0660, Item Z.G.i.

Specific details of the training program, additional test procedures, and documen-tation methods will be developed and made available for on-site NRC Z&E review Additionally, the Supply System will review the results of the simulated loss of'll A/C power test to be performed at the La Salle and Grand Gulf Power Stations, subject to satisfactory safety evalutions.

The results and merits of performing these tests wi11 be reviewed, an analysis performed, and recommendations forwarded to the NRC as to whether or not the test, or some portion of it, should be repeated at WNP-2.

Based on the Supply System recommendations and the benefits realized on other BWR plants conducting the subject tests WNP-2 and the NRC will determine the scope of those portions. of the tests requiring performance at WNP-2.

r B'-48

1

~ &

il