ML19340D360

Forwards Response to NUREG-0737, Clarification of TMI Action Plan Requirements. Implementation of Desired Mods Will Cause Unavoidable Matl & Personnel Shortages,Which May Result in Delays in Implementation Schedules

ML19340D360
Person / Time
Site:	Point Beach
Issue date:	12/23/1980
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-1.C.5, TASK-1.D.1, TASK-2.K.2.13, TASK-2.K.2.16, TASK-TM NUDOCS 8012300404
Download: ML19340D360 (29)

Text

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O Wisconsin Electnc mwea couraur 231 W. MICHIGAN, P.O. BOX 2046. MILWAUKEE, WI 53201 December 23, 1980 Mr. Harold R.

Denton, Director Office of Nuclear Reactor Regulation U.

S.

NUCLEAR REGULATORY COMMISSION Washington, D.

C.

20555

Dear Mr. Denton:

DOCKET NOS. 50-266 AND 50-301 RESPONSE TO NUREG-0737 SCHEDULE REQUIREMENTS AS RELATED TO

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POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 This letter provides the response to the requirements of NUREG-0737, " Clarification of TMI Action Plan Requirements",

for Wisconsin Electric Power Company's Point Beach Nuclear Plant, Units 1 and 2.

Attached is our Schedule Table and Notes covering all the items specified in NUREG-0737, our implementation schedule, method and exceptions.

Each item is addressed relative to the requirements and schedules stated in NUREG-0737 with clarification added to provide as complete a response as possible.

Review of this response should be made with reference to prior Wisconsin Electric submittals, some of which are tabulated on page 11 of the attached table and notes.

Wisconsin Electric believes that backfitting an operating unit requires detailed review and evaluation of both the immediate and long-term impacts on all changes to the plant.

This requires that any backfitting be accomplished on a schedule which is plant specific and for only those items that contribute to and are appropriate and important to improved safety.

We are attempting to implement the NRC requirements in a manner that integrates the various changes into meaningful improvements in the plant and its operation.

To achieve this overall objective, some items do not meet the NRC schedule because they are integrated l

into a larger modification that has a later scheduled completion f gf O

date.

We believe in its review that the NRC will agree with the F

desirability of accomplishing the plant modifications in this J

manner rather than in a non-integrated, piecemeal approach that may detract from the overall safety of the plant.

/}

The work of providing several items is being performed by consultants and contractors who are furnishing certain engineering, construction and training services as required.

The completion J8012 3 o 0y0'[

y e

Mr. Harold R.~Denton December 23, 1980 of several items requiring physical plant modifications are of course dependent on suppliers' manufacturing schedules.

Since all licensees are attempting to meet the same set of implementation dates, some material and personnel shortages will be unavoidable.

These may result in some delays in implementation schedules, which are beyond our control.

Please advise if you have any questions in regard to this response.

Very truly yours, Fay,Dibector C. W.

Nuclear Power Department Attachments Subscribed and sworn to before me This 23rd day of December,,1980.

0m& ?.

G Notary Publie, State of Wisconsin My Commission expires 7"['[

Copy to: NRC Resident Inspector I

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4 RESPONSE TO NUREG-0737 POST-TMI REQUIREMENTS FOR OPERATING PIANTS j

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i Point Beach Nuclear Plant, Units 1 and 2 i

I Docket Nos. 50-266 and 50-301 Schedule Table and Notes I

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POST-TMI REQUIHtMENTS FOR OPEHATING RI; ACTORS Clariti-Implemen-PBNP cation ta tion Applica-PBNP Item Shortened Title Description Schedule bility Schedule Remarks I.A.I.1 Shift Technical Advisor 1. On duty 1/1/80 Yes Complete On duty since 1/1/80 -

Reference 1

2. Tech Specs 12/15/80 Yes 2/2/81 Note I.A.I.l.2

3. Trained per LL 1/1/81 Yes 3/1/81 Note I.A.I.l.3 Cat B

4. Describe long-1/1/81 Yes 3/1/81 Note I.A.l.l.4 i

term program 4

I.A.l.2 Shift Supervisor Delegate non-1/1/80 Yes Complete Reference 1 Responsibilities safety duties I.A.I.3 Shift Manning

1. Limit overtime 11/1/80 Yes 1/10/81 PBNP Approved Procedure 4.3, t

Implementa-Operations Division a

7 tion Date Personnel Assignments and Scheduling, Rev 0

2. Min Shift Crew 7/1/82 Yes N.A.

Note I.A.l.3.2 I.A.2.1 Immediate Upgrading

1. SRO Experience 5/1/80 Yes Comple t'e of RO and SRO Training

2. SRos be Ros 12/1/80 Yes Complete and Qualifications 1 yr

3. Three mo. trng 8/1/80 Yes Complete f Note I.A.2.1.1/4 on shift

4. Modify Training 8/1/80 Yes

CompleLg,

5. Facility 5/1/80 Yes Complete Note I.A.2.1.5 Certification I.A.2.3 Administration of Instructors 8/1/80 Yes Complete Note I.A.2.3 Training Programs Complete SRO Exam I'.A.3.1 Revise Scope and 1.

Increase scope 5/1/80 Yes Complete Criteria for Licensing 2.

Increase passing 5/1/80 Yes Complete Exams grade

3. Simulator exams 6/1/80 N.A.

Note I.A.3.1.3 N.A.

• Schedule not applicable to PBNP TBD a To be determined at a later date per the remarks i

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Clarifi-Imp l en.en-PDNP cation tation Applica-PBNP Item Shortened Title Description Schedule bility Schedule Remarks I.C.1 Short-Term Accident and 1. SB LOCA 6/1/80 Yes Completed Procedures Review 2.

Inadequate Core Cooling

a. Reanalyze 1/1/81 Yes 1/1/81 Generic procedures already and propose submitted guidelines j

b. Revise First Yes First Procedures retueling refueling outage after outage after 1/1/82 1/1/82

3. Transient; and accidents Note I.C. l. 3 a.. Reanalyze 1/1/81 Yes 1/1/81

]

and propose guidelines a

b. Revise First Yes First a

procedures refueling refua3Any outage after outage after 1/1/82 1/1/82 l

1.C.2 Shift and Relief Implement shift 1/1/80 Yes Completed l

Turnover Procedures turnover checklist I

l J.C.3 Shift-Supervisor Clearly define 1/1/80 Yes Completed Responsibility superv and oper responsibilities I.C.4 Control-Room Access Establish authority 1/1/80 Yes Completed 1

limit access

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I.C.5 Feedback of Operating Licensee to 1/1/81 Yes 1/1/81 PBNP Administrative implemant (Effective Procedure 3.15.7, Rev. O, procedures Date) approved 12/19/80, 2

" Procedure for Feedback of Operating Experience to Plant Staff" l

l Clarifi-Implemen-PUNP l

cation Lation Applica-PUNP Item Shortened Title Description Schedule bility Schedule Rema do s l

1.C.6 Verify Correct Revise 1/1/81 Yes Completed PDNP Administrative i

Performance of performance Procedure 4.13, Rev. 9, 1

Operating Activities procedures effective 6/20/80,

" Equipment Isolation Procedure" l

l.D.1 Control Room Design Preliminary TDD Yes TDD Note I.D.1 Reviews assessment and schedule for correcting deficiencies I.D.2 Plant Safety Para-

1. Description TBD Yes 7/1/81 meter Display Consolo (Projected) 2.

Installed TBD Yes 1/1/83 e

(Projected) > Note I.D.2 Y

3. Fully TBD Yes 7/1/83 implemented (Projected)

II.D.1 Reactor Coolant

1. Design ventc 7/1/81 Yes 7/1/81 System Vents 2.

Install Vents 7/1/82 Yes 7/1/82 (LL Cat D)

3. Procedures 1/1/82 Yes 1/1/82 II.B.2 Plant Shielding

1. Review designs 1/1/80 Yes Completed

2. Plant modifi-1/1/82 Yes 6/1/82 Note II.B.2.2 cations (LL Ca t B)

3. Equipment 6/30/82 Yes 6/30/82 qualification II.B.3 Post Accident Sampling

1. Interim system 1/1/80 Yes Completed

2. Plant modifi-1/1/82 Yes 1/1/82 Note II.B.3 cations (LL Cat B)

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Clarifi-Implemen-PUNP cation tation Applica-PBNP Item Shortened Title Description Schedule bility Schedule Remarks II.B.4 Training for

1. Develop training 1/1/81 Yes 4/1/81 Mitigating Core Damage program 2.

Impicment k Note II.B.4 program r

a.

Initial 4/1/81 Yes 7/1/81

b. Complete 10/1/81 Yes 10/1/81,j II.D.1 Relief and Safety Valve 1.

Submit program 1/1/80 Yes Completed Note II.D.1 Test Requirements 2.

RV and SV Testing (LL Cat B)

a. Complete 7/1/81 Yes 9/1/81 Note II.D.l.2a testing

b. Plant-10/1/81 Yes 1/1/82 Note II.D.l.2b 71 specific e

report 7

3. Block-Valve 7/1/82 Yes N.A.

Note II.D.1.3 testing II.D.3 Valve Position

1. Install direct 1/1/80 Yes Completed i

Indication indications of valve position

2. Tech Specs 12/15/80 Yes 2/2/81 Note II.D.3 II.E.1.1 Auxiliary Feedwater

1. Short term 7/1/81 Yes TBD Note II.E.1.1 System Evaluation

2. Long term 1/1/82 Yes TBD II.E.1.2 Auxiliary Feedwater 1.

Initiation System Initiation

a. Control grade 6/1/80 Yes N.A.

and Flow

b. Safety grade 7/1/81 Yes Original References 1, 2, and 3.

Plant Design

2. Flow Indica tion

a. Control grade 1/1/80 Yes 1/1/80

. Note II.E.1.2 I

b.

LL A Tech 12/15/80 Yes 2/2/81 Specs

c. Safety grade 7/1/81 Yes 7/1/81 j

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l Clarifi-Implemen-PDNP cation Lation Applica-PUNP Item Shortened Title Description Schedule bility Schedule Remarks II.E.3.1 Emergency Power for

1. Upgrade power 1/1/80 Yes Original References 1, 2, and 3.

Pressurizer Heaters Plant Design 1

2. Tech Specs 12/15/80 Yes 2/2/81 Note II.E.3.1.2 II.E.4.1 Dedicated Hydrogen

1. Design 1/1/80 Yes Original References 1, 2, and 3.

Penetrations Plant Design 2.

Install 7/1/81 Yes N.A.

II.E.4.2 Containment Isolation 1-4. Imp. diverse 1/1/80 Yes TBD Note II.E.4.2.1/4 Dependability isolation

5. Cntmt pressure setpoint

a. Specify 1/1/81 Yes 1/1/81 6 psig pressure Note II.E.4.2.5 l

b. Modifi-7/1/81 Yes N.A.

cations y

6. Cntmt purge 1/1/81 Yes completed Administratively closed.

valves i

7. Radiation 7/1/81 Yes Original Plant Refercnce Point Beach FFDSAR signal on design Section 4.2 and Fig. 5.2-8 purge valves

8. Tech Specs 12/15/80 Yes 2/2/81 Note II.E.4.2.8 II.F.1 Accident Monitoring

1. Noble gas 1/1/82 Yes 1/1/82*

]NoteII.F.1.1/2 monitor j

2. Iodine /

1/1/82 Yes 1/1/82*

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particulate sampling s

3. Containment 1/1/82 Yes 1/1/82*

i high-range I

monitor

4. Containment 1/1/82 Yes 1/1/82*

5. Containment 1/1/82 Yes 1/1/828 l

water level

6. Containment 1/1/82 Yes TDD Note II.F.1.6 l

hydrogen

• Schedule is based on delivery of equipment on schedule.

Clarifi-Implemen-cation tation Applica-PBNP j

Item Shortened Title Description Schedule bility Schedule Remarks II.F.2 Instrumentation for

1. Subcool meter 1/1/80 Yes Complete Note II.F.2.1 Detection of Inadequate 2. Tech Spec 12/15/80 Yes 2/2/81 Note II.F.2.2 Core Cooling (LL Cat A) 3 Install level 1/1/82 Yes 6/1/82*

Note II.F.2.3 instruments (LL Cat B) i II.G.1 Power Supplies for

1. Upgrade to 1/1/80 Yes Original Pressurizer Relief emerg sources Plant Design Valves, Block Valves,

2. Tech Specs 12/15/80 Yes 2/2/81 Note II.G.I.2 and Level Indicators i

II.K.1 IE bulletins 79-05, -06, -08 Dulletin Yes Bulletin specific specific 1

II.K.2 Orders on B&W Plants

8. Upgrade AFW See N.A.

i system II.E.1.1 8

9. FEMA on ICS 8/17/79 N.A.

1

10. Safety-grade 7/1/81 N.A.

trip

11. Operator Complete N.A.

training, drilling d

13. Thermal-mechanical report 1/1/82 Yes 1/1/82 Note 11.K.2.13

14. Lift frequency See N.A.

of PORVs and SVs II.K.3.7

15. Effects of slug Complete N.A.

I flow on OTSGS

16. RCP seal damage Complete N.A.

17. Voiding in RCS

a. Complete N.A.

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b. 1/1/82 Yes 1/1/82 Note II.K.2.17

19. Benchmark

a. Complete N.A.

analysis of seq. b. 1/1/82 Yes TDD Note II.K.2.19 1

AFW flow

20. System response Complete N.A.

to SB LOCA t

• Schedule is. based on delivery of equipment on schedule.

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Clarif1-Implemen-PBNP ca tion tation Applica-PBNP Item Shortened Title Description Schedule bility Schedule Remarks II.K.3 Final recommendations,

1. Auto PORV B&O Task Force isola tion 4

a. design 7/1/81 Yes TBD

b. Test /

lst refuel Yes TBD Note II.K.3.1 and Reference 4 i

install 6 mos af ter staff approval

2. Report on PORV 1/1/81 Yes 3/1/81 Note II.K.3.2 failures

3. Reporting SV and 1/1/81 Yes 3/1/81 To cover the period 4/1/80 RV failures and through 12/31/80 data will challenges be provided are part of the 1980 annual plant report to

5. Auto trip af RCPS a.

Propose 7/1/81 Yes TBD modifications Note II.K.3.5

b. Modify 3/1/82 Yes TUD Y

7. Eval of PORV 1/1/81 N.A.

opening probability 9.

PID controller 1/1/81 Yes Completed Controller change made upon initial notification Ly vendor prior to THI-2 (Reference 4) l

10. Proposed Plant Yes Original Reference 4 anticipatory specific Plant Design 4

trip modifi-cations

11. Justify use of Plant Yes N.A.

As part of the original certain PORV specific Plant design (different from 1HI-2), Point Beach has Copes-Vulcan PORVs which corresponds to the Westinghouse data base and, thus, no justification is needed.

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Clarifi-Implemen-PUNP cation tation Applica-PDNP Item Shortened Title Description Schedule bility Schedule Remarks II r.3 Final recommendations,

13. IIPCI & RCIC (Continued)

D&O Task Force init levels

a. Analysis 1/1/81 N.A.

b. Modify 7/1/81 N.A.

14. Iso condenser 1/1/82 N.A.

isol modifi-cation

15. Isolation of 7/1/81 N.A.

HPCI and RCIC modi f ica tion

16. Challenges and failures to relief valves

a. Study 4/1/81 N.A.

b. Modify 1st refuel N.A.

1 or 1 yr af ter i

approval

17. ECC system 1/1/81 Yes 3/1/81 To cover the report period I

outages 1/1/76 through 12/31/80, a report will be submitted within 60 days of the end of the period (Reference 4)

18. ADS actuation

a. Study 4/1/81 N.A.

b. Propose 4/1/82 N.A.

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mods

c. Modification 1st refuel N.A.

6 mo after sta f f. approval

19. Interlock 7/1/81 N.A.

.recirc pump modification

20. Loss of SVC 7/1/81 N.A.

21. Restart of CCS p

and LPCI

a. Design 1/1/81 N.A.

p

b. Modification 1st refueling N.A.

60 mo after staf f approval

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i Clarifi-Implemen-PUNP cation ta tion Applicab-PBNP Item __

Shortened Title Description Schedule bility Schedule Remarks II.K.3/

Final recommendations,

22. RCIC suction (Continued)

B&O Task Force a.

Verify 1/1/81 N.A.

procedures.

b. Modification 1/1/82 N.A.

24. Space cooling 1/1/82 N.A.

j for HPCI/RCIC modifications

25. Power on pump 5

seals a

a.

Propose 7/1/81 d.A.

mods 1/1/82 Yes 1/1/82

b. Modification 1/1/82 N.A.

7/1/82 Yes N.A.

27. Common ref.

7/1/81 h.A.

level

28. Qual of ADS 1/1/82 N.A.

8 accumulators

29. Performance of 4/1/81 N.A.

j isolation condensers 30 SB LOCA methods

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a. Schedule 11/15/80 Yes TBD outline

b. Model 1/1/82 Yes TBD Note II.K.3.30

c. New 1/1/83 or Yes TBD analyses 1 yr after staf f approval

31. Compliance 1/1/83 or Yes TBD Note II.K.3.31 with CFR 50.46 1 yr after staf f approval i

40. RCP seal See N.A.

damage II.K.2.16

43. Effects of See N.A.

slug flow II.K.2.15

43. Eval transient 1/1/81 N.A.

with single failure i

45. Manual depres-1/1/81 N.A.

suriz.ntion 4

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Clarifi-Implemen-PUNP ca tion tation Applicab-PHNP Item Shortened Title Desctlption Schedule bility Schedule Remarks II.K.3 Final recommendations

46. Micielson Complete N.A.

(Continued)

D&O Task Force concerns

57. Manual act TLD N.A.

of ADS III.A.l.1 Emergency Preparedness, Short-term Complete Yes Complete Short Term improvements III.A.I.2 Upgrade Emergency 1.

Interim TSC Complete Yes 3/1/81 Completion of temporary Support Facilities OSC and EOF instrumentation wiring and calibration is the only outstanding item (References 1, 2, and 3) 2.

Design TBD TBD TBD

,L 3.

Modifications TBD TBD TUD III.A.2 Emergency Preparedness

1. Upgrade emer-4/1/81 Yes 4/1/81 Note III.A.2.1 gency plans to App. E, 10 CPR 50

2. Meteorological 6/1/83 Yes 6/1/83 Note III.A.2.2 data III.D.l.1 Primary Coolarit

1. Leak reduction Complete Yes complete Currently changing to a Outside Containment yearly testing schedule for both units coincident with refueling outages (References 1, 2, and 3) 2.

Tech Specs 12/15/80 Yes 2/2/81 Noto III.D.l.1 III.D.3.3 Inplant lodine

1. Provide means to Complete Yes Complete Monitoring determine presence of radiciodine

2. Modifications 1/1/81 Yes 1/1/81 Note III.D.3.3 to accurately measure 12 III.D,3.4 control Room

1. Review 1/1/81 Yes TBD Note III.D.3.4.1 Habitability

2. Modification 1/1/83 Yes TBD Note III.D. 3.4.2 er

REFERENCES 1.

S.

Burstein (WE) to H. R.

Denton (NRC), December 31, 1979,

" Implementation of NUREG-0578" 2.

C. W.

Fay (WE) letter to H. R.

Denton (NRC), March 14, 1980, "Lmplementation of NUREG-0578" 3.

A. Schwencer (NRC) letter to S.

Burstein (WE), April 9, 1980,

" Evaluation of Compliance with Category "A" Lessons Learned Requirements" 4.

C.

W.

Fay (WE) letter to H.

R.

Denton (NRC), June 11, 1980,

" Implementation of Five Additional TMI-2 Related Requirements" 5.

C.

W.

Fay (WE) letter to H. R.

Denton (NRC), November 3,

1980,

" Status of Duty and Call Technical Advisor Training" 6.

C.

W.

Fay (WE) letter to H. R.

Denton (NRC), December 1, 1980,

" Revised Emergency Plan" 7.

C.

W.

Fay

(' E) letter to D.

G.

Eisenhut (NRC), November 3, 1980 W

" Operating Licenses DPR-24 and DPR-27, Interim Criteria for Shift Staffing" 8.

C. W.

Fay (WE) letter to H.

R.

Denton (NRC), September 22, 1980,

" Comments on Draft NUREG-0696, Functional Criteria for Emergency Response Facilities" 9.

S.

Burstein (WE) letter to H.

R.

Denton (NRC), October 20, 1979,

" Implementation of NUREG-0578" - including TMI Accident Review Task Force Report (Section 3.6.A) l

I.A.1.1.2

~ SHIFT TECHNICAL ADVISOR TECHNICAL SPECIFICATIONS Technical Specifications for the Shift Technical Advisor will be addressed by February 2, 1981 along with other TMI Technical Specification changes.

I.A.l.1.3 SHIFT TECHNICAL ADVISOR TRAINING Shift Technical Advisor training will be completed by March 1, 1981 with the exception of training for mitigating core damage as addressed in II.B.4.

I.A.l.l.4 SHIFT TECHNICAL ADVISOR LONG-TERM PROGRAM DESCRIPTION The Shift Technical Advisor long-term training program will be submitted by March 1, 1981.

See also Reference 5 for background information.

I.A.l.3.2 SHIFT MANNING - MINIMUM SHIFT CREW l

l Wisconsin Electric's position on shift crew staffing has been l

provided in Reference 7.

Current Point Beach staffing on shift is as follows:

1 - Duty Shift Supervisor (SRO) 1 - Operating Supervisor (SRO except for 1 RO currently) 2 - Control Operators (RO) l l

3 - Auxiliary Operators Point Beach is a two-unit, two-loop PWR of Westinghouse design with a single integrated control room.

We believe that our present staff is entirely adequate for thr.

I operation of the plant including short-term actions required to cope with accident conditions.

Ten years of safe and reliable operation of the Point Beach facility provide evidence in this regard.

We are, of course, modifying training as appropriate to meet new NRC requirements and to provide for replacement for trained personnel who may leave.

In some areas, such as training, we are increasing the staff.

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I.A.2.1.1/4 IMMEDIATE UPGRADING OF RO AND SRO TRAINING AND QUALIFICATIONS - ITEMS 1 THROUGH 4 The experience levels and training modifications have been made and all future applicants for RO and SRO licenses will meet these as stated.

No backfitting of experience or training will be attempted for previously licensed personnel.

See also Reference 4 for background information.

I.A.2.1.5 IMMEDIATE UPGRADING OF RO AND SRO TRAINING AND QUALIFICATIONS - FACILITY CERTIFICATION The highest level of corporate management responsible for plant operation and technically capable of certifying the training programs and facility is the Manager - Nuclear Operations Section.

This level of certification was accepted by the NRC in July 1980 on license applications.

I.A.2.3 ADMINISTRATION OF TRAINING PROGRAMS Training instructors at Point Beach either hold an SRO license or are directly supervised by an SRO licensed person.

I.A.3.1.3 REVISED SCOPE AND CRITERIA FOR LICENSING EXAMS -

SIMULATOR EXAMS Since Point Beach does not have a simulator, the date of October 1, 1981 is the effective date for initiation of the requirement for applicants to be examined on a simulator.

It is our intention to utilize other plant simulators to meet this requirement in the future.

Under these conditions, the role of the simulator in the examination process must be more clearly defined by the NRC before a definite schedule commitment can be made.

It is our assumption that the major use of a simulator will be in the area of transient and accident recognition.

Other examination requirements related to the control room, e.g.,

control board familiarity, can be performed at the plant in a manner similar to the present examinations.. _. _ - -

i I.C.l.3 GUIDANCE FOR THE EVALUATION AND DEVELOPMENT OF PROCEDURES FOR TRANSIENTS AND ACCIDENTS The Owners Group of Westinghouse utilities will submit by January 1, 1981 a detailed description of our program to comply with the requirements of Item I.C.l.

The program will identify previous owners Group submittals to the NRC, which we believe will comprise the bulk of the response.

Additional effort required to obtain full compliance with this item (with proposed schedules for completion) will also be identified, as discussed with the NRC on November 12, 1980.

I.D.1 CONTROL ROOM DESIGN REVIEWS t

The control room design reviews should be performed at a time when the TMI modifications are far enough along to be able to be evaluated as part of the control room' review, but early enough to change the design before the modifications are made.

For Point Beach, an initial review is presently scheduled for April-May 1981.

I.D.2 PLANT SAFETY PARAMETER DISPLAY CONSOLE i

Two projects are underway that relate to the safety parameter display console.

One project will develop the design of the system and demonstrate its effectiveness.

The second project deals with the procurement and installation of the hardware system.

By July 1, 1981, a detailed description of the system will be submitted to the NRC.

We project that tv January 1, 1983 our new computer system can be procureu and installed.

As with any new computer system, we foresee additional time to get the system completely operational.

It is our intention to integrate a number of requirements into a complete system resulting in a longer implementation time period but with an end result that will be superior to many independent, non-integrated fixes.

This system is dependent in part on the final requirements to be issued as NUREG-0696, for which a draft for comment has been issued.

Wisconsin Electric has submitted its comments (Reference 8) along with other interested parties and awaits the resolution of the items identified.

L II.B.2 PLANT SHIELDING The initial calculations for Point Beach were performed using the source terms specified in NUREG-0578.

The clarification -..

II.B.2 Continued in NUREG-0737 changed the assumptions used to develop the source terms and this impact is being evaluated.

It is estimated that the scope of this effort and anticipated shielding modifications can be determined by early February 1981 and the schedule for resolution submitted to the NRC by March 1,1981.

We are aware of the present differences between segments of the scientific and nuclear industry communities and the NRC concerning source terms specified in NUREG-0578.

To the extent that these differences remain unresolved, the final implementation of any modifications may be altered.

Because the final evaluation will not be completed until March 1, 1981 and the Unit 2 refueling follows shortly in the Spring of the year, our present schedule calls for the completion of shielding modifications by June 1, 1982, assuming equipment availability.

However, the NRC position as stated in NUREG-0737 is unclear, making an evaluation and implementation based on the current version difficult.

III.B.3 POST-ACCIDENT SAMPLING The final system configuration is under review at the present time relative to the changes to previous requirements and guidance.

The majority of all of the requirements are already met by Point Beach and minimum change to the current system and plant l

capabilities are expected.

Point Beach has previously described (References 1 and 2) the system configuration which has also j

been reviewed by the NRC (Reference 3).

Point Beach will continue to utilize the reciprocal agreement with the Kewaunee Nuclear Plant to provide backup analytical laboratory services.

We have elected not to provide the optional oxygen concentration analysis for reactor coolant samples.

Wisconsin Electric still objects to the requirement for chloride i

analysis for Point Beach.

The plant is not exposed to seawater l

or brackish water as part of the cooling system and has no known chloride source for the RCS.

This fact coupled with the difficulty l

of analysis makes the usefulness of any results dubious.

We request that this requirement be dropped for Point Beach and similar plants.

We are concerned with the post-implementation review for this l

item and we will, therefore, provide, for NRC review and pre-implementation concurrence, a description of the system and any proposed changes on a schedule consistent with completed imple-mentation by January 1, 1982.

II.B.4 TRAINING FOR MITIGATING CORE DAMAGE i

i NUREG-0737 establishes the current requirements for training for mitigating core damage.

The primary requirements are that a program be developed by January 1, 1981, that training be initiated by l

April 1, 1981, and that training be completed by October 1, 1981.,_. _ __

II.B.4 Continued Development of a detailed Point Beach training course on miti-gating core damage was originally delayed because it was unclear exactly at what depth the program should be taught, lack of pertinent technical information, and because qualified manpower was unavailable in the Point Beach training group that could be diverted from attending to the other training requirements resulting from TMI-2.

Those other training requirements include Shif t Technical Advisor training, additional simulator training and augmented licensed operator training.

It was anticipated that material from vendor courses would become available which could be adapted to Point Beach.

It was also anticipated that additional qualified training personnel could be made available to prepare the program.

To date, only t

limited material is available and there has not been an increase in available manpower.

Therefore, we find it impossible to meet the January 1, 1981 deadline using our own personnel and

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resources.

As an alternative, we have issued a purchase requisition for our nuclear steam supply system vendor to provide a program specific to Point Beach.

The NSSS vendor program is not scheduled to be ready until after March 1981.

Consequently, we will not be able to meet the April 1, 1981 deadline either.

However, at this time we see no reason why the initial training will not be completed before October 1, 1981, as specified.

The completion of the total training program for Shift Technical Advisors, for which a delay from January 1, 1981 to March 1, 1981 l

was previously requested (Reference 5), will also be affected l

by this delay of a program for mitigating core damage.

It is expected, however, that all training for Shift Technical Advisors, with the exception of this course, will be completed prior to i

the March 1, 1981 deadline.

l II.D.1 RELIEF AND SAFETY VALVE TEST REQUIREMENTS As a sponsor of the EPRI PWR Safety and Relief Valve Test Program, Wisconsin Electric intends to comply with the technical require-ments of NUREG-0578, Item 2.1.2.

By letter dated December 15,

1980, R. C. Youngdahl of Consumers Power Company has provided the current PWR Utilities' positions on NUREG-0737, Item II.D.1 clarifications.

Briefly those positions are:

A. Safety and Relief Valves and Piping - The EPRI " Program Plan for Performance Testing of PWR Safety and Relief Valves", Revision 1, dated July 1, 1980, does provide a program that satisfies the NRC requirements.

Discussions with the NRC Staff and their consultants are resolving specific detailed issues. l

II.D.1 Continued B.

Block Valves - The EPRI Program has not formally included the testing of block valves.

However, a small number of block valves have been tested at the Marshall Steam Station Test Facility.

The PWR Utilities and EPRI cannot provide a detailed block valve test program until results of the Wyle and Combustion Engineering relief valve tests are available.

Therefore, a block valve test program will not be provided before July 1981.

The PWR Utilities and EPRI believe that the proper operation of the TMI-2 and Crystal River block valves and other operational experience, plus knowledge of the Marshall tests, support the industry!c approach to block valve testing.

C.

ATWS Testing - PWR Utilities will not support additional efforts for ATWS valve testing until regulatory issues are resolved.

The major safety and relief valve test facility (Combustion Engineering) is nearing completion and some measures were taken to provide additional test capability beyond the current program requirements.

The NRC should i

recognize that results from the current program are likely to provide most of the information necessary to address ATWS events (i.e., relief capability at high pressures).

II.D.l.2a RELIEF AND SAFETY VALVE TEST REQUIREMENTS - COMPLETED TESTING It is unlikely that the EPRI testing program will be completed by July 1, 1981, even though their schedule indicates that this date will be achieved.

As a minimum, it usually takes at least two months to complete a report and to transfer the testing results to other parties.

It is our judgment that it will be no sooner than September 1, 1981 before we can complete a preliminary review of these generic test program results.

II.D.l.2b RELIEF AND SAFETY VALVE TESTING REQUIREMENTS - PLANT SPECIFIC REPORT The plant specific review will follow the preliminary review by three months.

This results in an estimated date of January 1, 1982 for submittal of the report.

This report will be delayed if the completion of the testing is delayed.

II.D.1.3 RELIEF AND SAFETY VALVE TEST REQUIREMENTS - BLOCK-VALVE TESTING l

l A small number of block valves are part of the EPRI testing i

program test configuration.

Block valve results that come out of the PORV test program will be utilized, but it is not our intention

II.D.l.3 Continued to do any block valve testing beyond that program.

Point Beach Nuclear Plant has been analyzed for a spectrum of break sizes that bound the size of a stuck-open PORV and block valve.

The analysis shows the Plant to be safe for this postulated event.

Refer to the response to II.D.l.

II.D.3 VALVE POSITION INDICATION - TECHNICAL SPECIFICATIONS Suggested Technical Specifications for TMI items will be addressed by February 2, 1981.

This item will be included in that submittal.

II.E.1.1 AUXILIARY FEEDWATER SYSTEM EVALUATION Based on the time of completion of the NRC evaluations of previous industry-wide submittals in response to IE Bulletin 79-06A and determination of the extent of the required modifications, the applicability and implementation schedule for the Point Beach Nuclear Plant cannot be determined at this time.

Upon issuance of the NRC evaluation recults, a determination of scope for Point Beach and projects of equipment delivery schedules will be used to establish an bnplementation schedule.

II.E.1.2 AUXILIARY FEEDWATER SYSTEM INITIATION AND FLOW -

FLOW INDICATION As previously described in References 1, 2, and 3, auxiliary feed-water flow indication is provided via pump discharge flow monitoring and steam generator level indication which was operational by January 1, 1980.

A final upgraded configuration of the pump discharge flow monitoring will be achieved by January 1, 1981 with safety-grade equipment whose qualification will be verified by a testing program now in progress.

Additionally, as committed to in Reference 1, indication of flow to the individual steam generators is being implemented.

This equipment will be safety grade and verified by the same testing program.

Consideration of seismic and environmental conditions in the system design has necessitated a delay from our previous commitment of January 1, 1981 for steam generator flow indication.

The projected completion date, however, is consistent with the July 1, 1981 requirement for safety-grade flow indication.

Technical Specifications for auxiliary feedwater flow indicaticn will be addressed by February 2, 1981 with the other TMI Technical Specification changes.

II.E.3.1.2 EMERGENCY POWER FOR PRESSURIZER HEATERS Technical Specifications for the emergency power for pressurizer

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heaters will be addressed by February 2, 1981 with the other TMI Technical Specification changes..

II.E.4.2.1/4 CONTAINMENT ISOLATION DEPENDABILITY - IMPROVED

' DIVERSE ISOLATION The addition of inside containment isolation valves on the let-down and seal water return lines (Reference 3) is in the process of being implemented utilizing the services of an outside consultant for design and equipment specification.

Ordering of the required valves should be completed by March 1, 1981.

Current industry estimates of delivery of this class of equipment is approximately 18 months.

Installation would then be scheduled for the next major outage for each unit.

The projected completion date for this work then becomes the Spring 1983 refueling outage for Unit 2.

II.E.4.2.5 CONTAINMENT ISOLATION DEPENDABILITY - CONTAINMENT PRESSURE SETPOINT We concur with the basic requirement that the containment setpoint pressure that initiates containment isolation of nonessential penetrations should be the minimum setpoint compatible with normal operating conditions.

However, pending additional evaluation, we cannot concur with the clarification statement that 1 psi above normal operation pressure is an appropriate minimum pressure setpoint and the margin should instead be related to instrume :

9nge rather than an absolute psi number.

The 1 psi value :

not aufficient to account for instrumentation drift and accuracy

.. well as abnormal operation scenarios which may increase contain-hent pressure above normal but for which isolation of nonessential penetrations is not desirable.

It should be noted that the same pressure setpoint that currently actuates containment isolation also actuates safety injection and trips the reactor.

We feel that it is inadvisable to run the risk of unwarranted challenges to the ECC system.

The setpoint for Point Beach Nuclear Plant (both Units 1 and 2) is currently set at less than the Technical Specification value of 6 psig based on original plant design.

II.E.4.2.8 CONTAINMENT ISOLATION DEPENDABILITY - TECHNICAL i

SPECIFICATIONS l

Suggested Technical Specifications for TMI items will be addressed by February 2, 1981.

This item will be included in that submittal.

II.F.1.1/2 ACCIDENT MONITORING - NOBLE GAS, IODINE, AND PARTICULATE It is anticipated that the gaseous effluent monitoring system changes required to meet the revised NRC requirements can be accomplished by January 1, 1982 for noble gas, iodine, and particulate monitoring.

However, several items may deJ ay or preclude completion of some modifications.

Equipment already ordered and constructed -_-.

k II.F.1.1/2 Continued to meet previous requirements will need to be modified or re-manufactured to meet the clarification items.

Isokinetic representative sampling may not be consistent with the requirement to locate the sampling equipment in an accessible location and also limit personnel exposures.

These conflicts may limit the ability to fully Englement monitoring at Point Beach as required.

II.F.1.6 ACCIDENT MONITORING - CONTAINMENT HYDROGEN The type of hydroger. monitoring systems presently available have been carefully reviewed for installation at the plant.

At this time, no equipment has completed its environmental qualification program to verify that it will operate in an accident environment.

Until these programs are completed, proper selection of a system cannot be made.

A schedule will be established following selection of a supplier.

The present manual means of sampling for hydrogen following an accident is judged to be sufficient as an interim measure until a qualified system can be selected and installed.

II.F.2.1 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE COOLING - SUBCOOLING METER The final subcooling system which will meet all of the NRC require-ments has been designed as part of the qualified instrumentation system being added to the plant.

This added system is intended to provide the reans by which the large number of instrumentation changes can be properly incorporated into the existing plant.

This new system consists of five instrument racks and an auxiliary safety instrumentation panel (ASIP) for each unit.

The ASIP is located 16 the control room at a location which allows for easy viewing of the panel-mounted display devices.

Subcooling display meters will be located on the ASIP.

The operability of the final subcooling meter is dependent upon the delivery, installation and operational checkout of the new racks and panels.

The new panels are each fifteen feet long and their installation in the control room must be carefully planned and implemented.

We are proceeding on a best effort basis and expect to have the panels installed during late 1981 and their associated systems operational in early 1982.

The subcooling capability described in References 1 and 2 is judged to be sufficient until the final system is installed in the plant.

II.F.2.2 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE i

COOLING - TECHNICAL SPECIFICATIONS 1

l Suggested Technical Specifications for TMI items will be addressed by February 2, 1981.

This item will be included in that submittal.

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II.F.2.3 INSTRUMENTATION FOR DETECTION OF INADEQUATE CORE

. COOLING - INSTALL LEVEL INSTRUMENTS The reactor vessel level system design and description has been provided in References 1 and 2.

The detailed design of the system is in progress.

The instrumentation system which will process the sensor signals and display the resulting feactor vessel level will utilize the instrumentation racks and auxiliary safety instrumentation panel described in II.F.2.1.

A report detailed the planned instrumentation for monitoring of inadequate core cooling will be sent to the NRC by April 1, 1981.

The planned schedule for having the vessel level systems operational is January 1, 1982 for Unit 1 and June 1, 1982 for Unit 2.

This is based on delivery of equipment on schedule to meet the Unit 1 Fall 1981 and Unit 2 Spring 1982 refueling outages.

II.G.1.2 POWER SUPPLIES FOR PORVS, BLOCK VALVES, AND LEVEL INDICATORS - TECHNICAL SPECIFICATIONS Suggested Technical Specifications for TMI items will be addressed by February 2, 1981.

This item will be included in that submittal.

II.K.2.13 THERMAL MECHANICAL REPORT - EFFECT CF HIGH PRESSURE INJECTION ON VESSEL INTEGRITY FOR SMALL-BFTAK LOSS-OF-COOLANT ACCIDENT WITH NO AUXILIARY FEEDWATER To completely address the NRC requirements of detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with an extended loss of all feedwater, a program will be completed and documented to the NRC by January 1, 1982 by the Owners Group of Westinghouse utilities.

This program will consist of analysis for generic Westinghouse PWR plant groupings.

Following completion of this generic program, additional plant specific analyses, if required, will be provided.

A schedule for the plant specific analysis will be determined based on the results of the generic analysis.

II.K.2.17 POTENTIAL FOR VOIDING IN THE REACTOR COOLANT SYSTEM DURING TRANSIENTS The Owners Group of Westinghouse utilities is currently addressing the potential for void formation in the reactor coolant system (RCS) during natural circulation cooldown conditions, as described in I

Westinghouse letters NS-TMA-2298 from T. M.

Anderson (Westinghouse)

II.K.2.17 Continued to P. S. Check (NRC).

We believe the results of this effort.will fully address the NRC requirement for analysis to determine the potential for voiding in the RCS during anticipated transients.

A report describing the results of this effort will be provided to the NRC before January 1, 1982.

II.K.2.19 SEQUENTIAL AUXILIARY FEEDWATER FLOW ANALYSIS The transient analysis code, LOFTRAN, and the present small break evaluations analysis code, WFLASH, have both undergone benchmarking against plant information or experimental test facilities.

These codes, under appropriate conditions, have also been compared with each other.

The Owners Group of Westinghouse utilities will provide, on a schedule consistent with the requirement of Task II.K.2.19, a report addressing the benchmarking of these codes.

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II.K.3.1 INSTALLATION AND TESTING OF AUTOMATIC POWER-OPERATED RELIEF VALVE ISOLATION SYSTEM Based on the previous operating history at the Point Beach Nuclear l

Plant during which no PORV has failed open, we see no need for an automatic isolation system.

We believe that such a system would l

add unnecessary complexity to a well-designed, functional system i

and may, in fact, contribute to a reduction in safety and operator attention to the PORV status.

Such a system may also reduce the j

number of options available to the operator and, thus, limit the accident response capability for the plant.

However, per II.K.3.2, t

i we will participate in a further study of the need for such an automatic isolation system and may modify this position based on the l

results of the study.

II.K.3.2 REPORT ON GVERALL SAFETY EFFECT OF POWER-OPERATED RELIEF VALVE ISOLATION SYSTEM The Owners Group of Westinghouse utilities is in the process of developing a report (including historical valve failure rate data and documentation of actions taken since the TMI-2 event to decrease the probability of a stuck-open PORV) to address the NRC concerns expressed in Item II.K.3.2.

However, due to the l

time-consuming processing of data gathering, breakdown, and evaluation, this report is scheduled for submittal to the NRC i

on March 1, 1981.

As required by the NRC, this report will be used to support a decision relative to the necessity of incor-porating an automatic PORV isolation system as specified in Task j

Action Item II.K.3.1.

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v II.K.3.2 Continued For the Point Beach Nuclear Plant specifically, several post-TMI actions have been taken to reduce the probability of a small break LOCA caused by a stuck open PORV.

In addition to training of operators relative to the TMI accident sequence and added precautions for operating procedures, where appropriate, the circuitry for automatically opening the PORVs on high RCS j

pressure has been changed from 1-of-l to a 2-of-2 actuation i

logic.

Coincidentally, both Point Beach units are currently operating at 2,000 psia which substantially increases the margin to the setpoint for PORV opening.

It should be noted that the original plant design for Point Beach included direct positive position indication (on the main control board) for the PORVs and that the ind.: cation equipment has been upgraded over the years prior to TMI-2.

Also, the basic design of the PORVs in the Point Beach Nuclear Plant is quite different frcm that of the TMI-2 PORV (pilot operated).

II.K.3.5 AUTOMATIC TRIP OF REACTOR COOLANT PUMPS DURING LOSS-OF-COOLANT ACCIDENT The Owners Group of Westinghouse utilities' resolution of this issue has been to perform analyses using the Westinghouse small break evaluation model (WFLASH) to show ample time is available for the operator to trip the reactor coolant pumps following certain size small breaks (see WCAP-9584).

In addition, the Owners Group is supporting a best estimate study using the NOTRUMP computer code to demonstrate that tripping the reactor coolant pump at the worst trip time, after a small break, will lead to l

acceptable results.

For both of these analysis efforts, the Owners Group is performing blind post-test predictions of LOFT experiment L3-6.

The input data and model to be used with WFLASH on LOFT L3-6 has been submitted to the Staff on December 1, 1980 (NS-TMA-2348).

The information to be used with NOTRUMP on LOFT L3-6 will be submitted prior to performance of the L3-6 test as stated in letter OG-45 dated December 3, 1980.

The LOFT prediction from both models will be submitted to the Staff on February 15, 1981 given that the test is performed on schedule.

The best estimate study is scheduled for completion by April 1, 1981.

Based on these studies, the Owners Group believes that resolution of this issue will be achieved without any design modifications.

In the event that this is not the case, a schedule will be provided for potential modifications.

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W II.K.3.25 FINAL RECOMMENDATIONS, B&O TASK FORCE - POWER ON PUMP SEALS A complete loss of alternating-current (AC) power for a unit at Point Beach may result in a loss of power to the reactor coolant pumps (RCP), charging pumps, and component cooling water (CCW) pumps as well as a reactor trip.

Since Point Beach is a dual unit, charging and CCW pump loads can be picked up by the unaffected unit within one-half hour, thus precluding an extended challenge to the RCP seals.

Additionally, both the charging and CCW pumps can be powered from the plant emergency buses with proper consideration for loading per plant procedures.

This provides for seal and pump protection for the cases where AC power is lost to one or both units.

It should be noted that upon loss of AC power, the RCPs stop rotating and water flow past the seals is minimized.

This would result in minimum RCS water upflow (and resulting RCS inventory loss) during the short time that charging or CCW flow could be lost.

Normal seal water return for nominal operating conditions (2,250 psia) or hydrostatic test conditions (2,350 psia for 1-1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />) is 1-2 gpm.

This would result in a maximum RCS inventory loss over a two-hour period of 480 gallons, which is not significant relative to total RCS initial inventory (<0.1%).

Although Wisconsin Electric will pursue a determination of RCP seal survival capability on a generic basis through the Utility Owners Group of Westinghouse plants, no plant modifications are planned for Point Beach, t

II.K.3.30 FINAL RECOMMENDATIONS, B&O TASK FORCE - SMALL BREAK LOCA METHODS The present Westinghouse small-break model used to analyze Point Beach Nuclear Plant Units 1 and 2 is in conformance with 10 CFR Part 50, Appendix K.

However, Westinghouse has indicated that they will, nevertheless, address the schedule for completion of a new model by January 1, 1982.

Westinghouse has provided a detailed outline of the scope and schedule of this effort by means of a direct letter to the NRC.

II.K.3.31 FINAL RECOMMENDATIONS, B&O TASK FORCE - COMPLIANCE WITH 10 CFR 50.46 If the results of the new Westinghouse model (and subsequent NRC review and approval) indicate that the present small-break LOCA analysis for Point Beach Nuclear Plant Units 1 and 2 are not in conformance with 10 CFR 50.46, a new analysis utilizing the new I

and approved Westinghouse model will be submitted to the NRC in accordance with a mutually agreeable schedule.

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O III.A.2.1 EMERGENCY PREPAREDNESS - UPGRADED EMERGENCY PLANS TO APPENDIX E, 10 CFR 50 Copies of the revised PBNP Radiological Emergency Response Plan, as well as copies of the state and local plans, will be submitted on or before January 2, 1981.

The PBNP Plan to be submitted will be essentially complete except for the fif teen-minute notification and Table B-1.

Wisconsin Electric is currently reviewing its position with respect to the requirements for fifteen-minute notification and will provide further discussion along with the January 2, 1981 submittal.

We concur with the functional requirements of Table B-1 but will propose an alternate method for fulfilling these functions in terms of manpower; again, further discussion will be provided along with the January 2 submittal.

Implementing procedures will be submitted on or before March 1, 1981, and will include a description of methods to assess and monitor potential offsite consequences of an emergency.

The upgraded emergency response plans will be implemented on or before April 1, 1981.

III.A.2.2 EMERGENCY PREPAREDNESS - METEOROLOGICAL DATA Conceptual plans for upgrading meteorological capabilities will be provided by January 9, 1981.

III.D.l.1 PRIMARY COOLANT OUTSIDE CONTAINMENT - TECHNICAL SPECIFICATIONS Suggested Technical Specifications for TMI items will be addressed by Feb'ruary 2, 1981.

This item will be included in that submittal.

III.D.3.3 INPLANT IODINE MONITORING Point Beach Nuclear Plant meets the current requirements for in-plant monitoring of iodine, noble gas, and particulates through the use of both fixed and portable instrumentation.

Recent equipment purchases have added to previous analysis and sampling capabilities.

Silver-zeolite filters are now in use on sampling l

equipment to provide augmented iodine monitoring.

l Inplant counting room capabilities are supplemented by a reciprocal j

agreement with the Kewaunee Nuclear Plant.

The new Technical l

Support Center under construction will also have an additional low-background, low-contamination, counting and lab facility.

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e' III.D.3.3 Continued An upgrading of the plant radiation monitoring system is also in progress which will include additional iodine, particulate, and noble gas monitoring equipment.

Post-accident monitoring and display capabilities will be integrated with the instrumen-tation changes discussed in II.F.2.1.

Additional monitoring will be provided for areas within the plant where personnel may be present during and after an accident, such as the control room and TSC.

Changes identified as a result of the control room habitability review (III.D.3.4) will be included in the radiation monitoring system upgrade.

For background information, see References 1 and 2.

III.D.3.4.1 CONTROL ROOM HABITABILITY - REVIEW Re-evaluation of the PBNP control room habitability is being conducted by our consultant in connection with the implementation of other Lessons Learned items.

We will be unable to complete the evaluation by January 2, 1981.

However, an initial report of assumptions and early findings will be submitted by January 9, 1981, along with a schedule for completion of the evaluation as early as possible in the new year.

III.D.3.4.2 CONTROL ROOM HABITABILITY - MODIFICATION Modifications, if required, will be discussed in the final submittal of the evaluation in early 1981.

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