ML19256A572

Summary of 781219 Meeting W/Licensee & Bechtel Re Mod of Control Bldg.Discussed Preliminary Test Results of Shear Wall Specimens,Proposed Mod & Security Aspects During Mod. W/Encl List of Attendees

ML19256A572
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	12/28/1978
From:	Trammell C Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 7901090039
Download: ML19256A572 (25)

Text

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December 28, 1978

's *,.. f DOCKET N0.:

50-344 LICENSEE:

Portland General Electric Company (PGE)

FACILITY:

Trojan Nuclear Plant

SUMMARY

OF MEETING HELD ON DECEMBER 19, 1978 TO DISCUSS MODIFICATION OF THE TROJAN CONTROL BUILDING On December 19, 1978, the NRC staff met with representatives of Portland General Electric Company (PGE) and Bechtel to discuss future modification of the Trojan Control Building.

All infor-matior. furnished was classified by PGE/Bechtel as preliminary at this time.

A list of attendees is attached (Attachment 1).

Highlights of the meeting are summarized below.

The meeting was devoted to four subjects:

preliminary test results of representative Trojan Control Building shear wall specimens; the proposed Control Building modification; a report to be submitted describing the proposed modification; and security aspects during the modification.

A test program using representative shear wall specimens is currently underway, and will be completed by the end of January 1979. The purpose of this program is to establish the material properties to be used for the existing Control Building shear walls as input to the final building modification.

Results achieved thus far indicate that the Modified Schneider Criteria (used for the interim operation evaluation) are conservative, and do not invalidate assumptions made for this evaluation. Three tests remain to be conducted. Attachment 2 summarizes the test results achieved to date.

The proposed modification will consist of re-routing the railroad tracks which currently pass through the north end of the Control Building and (1) building two new shear walls of reinforced concrete across the railroad bay from el. 45' to el. 77', and (2) adding a 3-incn-thick steel plate to the outside of tne Control Building west wall from el. 65' to el. 93'.

This will stiffen the Control Building in the N-S direction, and is expected to restore the OBE capacity of the Control /. auxiliary / Fuel Building ccmplex 7901090039 I

pg f

Meeting Summary for Trojan Control Building December 28, 1978 to the licensed value of 0.15g. The modification is expected to raise the N-S building complex frequency (first mode) from 6.8 H Z

to about 7.4 H '

Z All Class I equipment, systems and components will be invetigated and appropriately qualified to FSAR criteria (0BE and SSE) as part of the modification.

If the modification causes an upward frequency shift which causes a change in floor response spectra outside of current spectra for safe-shutdown systems and systems required to maintain off-site doses to within 10 CFR 100 guide-lines for the SSE, these systems would be modified before the new walls were constructed, thereby maintaining their seismic (SSE) qualification.

Although preliminary, PGE stated that it appeared that the modification could be completed without restrictions on plant operation.

No preparatory work is planned, although some unrelated modifications (fire protection, security) may go forward.

PGE will keep the NRC informed of such work.

The Reactor Safeguards Licensing Branch will be visiting the Trojan site on January 4,1979, and will familiarize themselves with security aspects involved with the modification.

The final meeting w.bject was a preview of the modification report Table of Contents, and a discussion of the information PGE plans to submit to NRC.

See Attachment 4.

This report is scheduled to be submitted about mid-January 1979, with a supplement in mid-February 1979 describing the final three wall specimen tests.

f-J:.. w,

.'%t Charles M. Trammell, Project Manager Operating Reactors Branch #1 Division of Operating Reactors Attachments:

1.

List of Attendees 2.

Testing Program Results 3.

Modification 4.

Table of Contents -

Topical Report

Meeting Sumary for Trojan Control Building Docket File NRC PDR Local POR ORB 1 Reading NRR Reading H. Denton E. Case V. Stello D. Eisenhut B. Grimes D. Davis D. Ziemann P. Check G. Lainas A. Schwencer R. Reid T. Ippolito V. Noonan J. McGough Project Manager OELD OI&E(3)

ACRS (16)

C. Parrish NRC Participants TERA J. R. Buchanan Licensee

Meeting Summary for Trojan Control Building cc: Mr. H. H. Phillips Ms. Nina Bell Portland General Electric Company 532 SE 18th 121 SW Salmon Street Portland, Oregon 97214 Portland, Oregon 97204 Mr. Stephen M. Willingham Warren Hastings, Esquire 555 N. Tomahawk Drive Counsel for Portland General Portland, Oregon 97217 Electric Company 121 SW Salmon Street Mr. Eugene Rosolie Portland, Oregon 97204 Coalition for Safe Power 215 SE 9th Avenue Mr. J. L. Frewing, Manager Portland, Oregon 97214 Generation Licensing a.nd Analysis Portland General Electric Company John H. Socolofsky 121 SW Salmon Street Oregon Department of Energy and Portland, Oregon 97204 Oregon Public Utility Comissioner Department of Justice Columbia County Courthouse State Office Builsing Law Library, Circuity Court Rcom Salem, Oregon St. Helens, Oregon 97501 Maurice Axelrad, Esq.

Director, Oregon Department Lowenstein, Newman, Reis of Energy Axelrad and Toll Labor and Industries Building, Rm.111 Suite 1214 Salem, Oregon 97310 1025 Connecticut Avenue, NW Washington, D. C.

20036 Dr. Hugh C. Paxton Mr. David B. McCoy 1220 41st Straet 348 Hussey Lane Los Alamos, New Mexico 87544 Grants Pass, Oregon 97526 Marshall E. Miller, Esq., Chairman Ms. C. Gail Parson Atomic Safety and Licensing Board 800 SW Green #6 U. S. Nuclear Regulatory Commission Portland, Oregon 9752g Washington, D. C.

20555 William Kinsey, Esq.

Dr. Kenneth A. McCollom, Dean 1002 NE Hollad:y Division of Engineering, Portland, Oregon 97232 Architecture & Technology Oklahoma State University Stillwater, Oklahoma 74074 Mr. John A. Kullberg Route One Box 250Q Sauvie Island, Oregon 97231

Meeting Sumnary for Trojan Control Building cc:

Gregory Kafoury, Esq.

Counsel for Columbia Environmental Council 202 Oregon Piuneer Building 320 SW Stark Street Portland, Oregon 97204 Robert M. Johnson, Esq.

Assistant Attorney General 100 State Office Building Salem, Oregon Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Atomic Safety and Licensing Appeal Board Panel V. S. Nuclear Regulatory Commission Washington, D. C.

20555 Docketing and Service Section Office of the Secretary U. S. Nuclear Regulatory Commission Washington, D. C.

20555

ATTACHMENT 1 LIST OF ATTENDEES TROJAN CONTROL BUILDING MEETING DECEMBER 19, 1978 PGE W. Lindblad D. Broehl T. Bushnell R. Johnson L. Erickson NRC C. Tramell K. Herring V. Noonan G. Arndt T. Cheng J. Gray R. Clark K. Wichman BECHTEL J. O' Leary W. White B. Sarkar M. Celebi F. Meyer R. Anderson LOWENSTEIN, NEWMAN, REIS, AXELRAD AND TOLL M. Axelrad SHAW, PITTMAN, POTTS AND TROWBRICGE E. Blake B. Churchill

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khCl1bef k ORAFT TOPICAL REPQdI_

0U STRENGTHENING OF THE TROJAN CONTROL BUILDitJG 1

INTRODUCTION 1.1 PURPOSE 1.2

SUMMARY

1.3 ORGANIZATION 1.4 DEFINITIONS OF THE NOTATIONS US 'D

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2 BACKGROUND 2.1 CONTROL EUILDING SEIS4IC DESIGN 2.2 EVALUATION OF THE EXISTING Bu!LDING 3

GENERAL DESIGN 3.1 CRITERIA 3.2 CONTROL BUILDING P0DIFICATIONS 3.3 SEISMIC ANALYSIS 4

SAFETY EVALUATION 4.1 SAFETY EVALUATION (STR UCTURAL) 4.2 OPERATIONAL CONSIDERATIONS

f ed 5

CONSTRUCTION PROGRAM DRAFT 5.1 CONSTRUCTION SEQUENCE AND PROCEDURES 5.2 BECHTEL QUALITY CONTROL PROGRAM 5.3 BECHTEL--PGE INTERFACE 5.4 SCHEDULE 6

0UALITY ASSURANCE

~.-

6.1 PGE QUALITY ASSURANCE 6.2 BECHTEL CUALITY ASSURANCE APPENDIX A:

SHEAR WALL SPECIMEN IESTING PRCGRAM APPENDIX 3.

SEISMIC OUALIFICATION OF Eau!PMENT COMPONENTS AND PIPING

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D' N ','

8 parc 'b

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UNITED STATES e

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[v f' t NUCLEAR REGULATORY COMMISSION kJ j

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WASHINGTON. D. C. 20555

's E

s, DEC 15 80 Docket flos. :

50-329 50-330 fir. S. H. Howell, Vice President Consumers Power Company 212 West Michigan Avenue Jackson, Michigan 49201

SUBJECT:

STAFF POSITI0ftS AND REQUESTS FOR ADDITI0f1AL I'NFORMATION (PART 2)

My letter of December 11, 1978 forwarded part I of our requests for additional information anc our positions that differ from those in your FSAR.

Part 2 of our requests and positions is contained in hereto.

We w;11 need response and resolution to Enclosure 1 by January 19, 1979.

if you cannot meet this date, inform us within seven days after receipt of this letter so that we may revise our schedule accordingly.

Should you desire clarification of Enclosure 1, please contact us.

Sincerely, t en ga, 1

Light Water Reactors \\S nch J Division of Project Management

Enclosure:

As stated cc: Listed on following page 4(

7901090g$

consumers Power Company ccs:

Micnael I. Miller, Esq.

Isha:n, Lincoln & Beale Suite 4200 One First flational Plaza Chicago, Illinois 606/0 Judd L. Bacon, Esq.

Consumers Power Company 212 West Michigan Avenue Jackson, Michi gan 49201 Mr. Paul A. Perry Secret a ry Consumers Power Conpany 212 W. Michigan Avenue Jackson, Michigan 49201 My ron M. Cherry, Esq.

Une IBM Plaza Chicago, Illinois 60611 Mary Sinclair 5711 Summerset Dri ve Midland, Michigan 48640 Frank J. Kelley, Esq.

Attorney General State of Michigan Environmental Protection Division 720 Law Building Lansing, Michigan 48913 Mr. Windell Marshall Route 10 Midland, Michigan 48640 e

e-e-enemee H

e f

ENCLOSURE 1 STAFF POSITIONS (Q-2s) AND RECUEST FOR ADDITIONAL INFORMATION PART 2 MIDLAND PLANT UNITS 1 & 2 These positions and requests for additional information are numbered such that the three digits to the lef t of the decimal identify the technical review branch and the numbers to the right of the decimal are the sequential request numbers.

The number in parenthesis indicates

.e relevant section in the Safety Analysis Report. The, initials RSP indicate the request represents a regulatory staff position.

Branch Technical Positions referenced in these recuests can be fcund in " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," NUREG-7 5 /087.

211-1 211.0 REACTOR SYSTEMS BRAth,_

211.176 Your response to request 211.147 provides the initial conditions (15.2) for BAW-10043 to show that it " brackets" the Midland units.

(5.2.2)

It is not clear that BAW-10043 bounds the Midland units.

(RSP)

Comparison of parameters from the Midland FSAR and your response is as follows:

BAW-il043 Midland FSAR Core Power (MWt}

3105 (1125) 2452 (1005)

Pump Heat (MWt) 16

,16 6

6 RCS Flow Rate ( /hr) 137.9 x 10 126.3 x 10 Pressurizer Code Safety Valve Capacity (Ib/hr) 690,000 595,690 Secondary Safety) Valve Capacity (lb/hr 13,680,000 12,484,520 The effects of less flow and relief valve capacity are not obvious relative to the lower power level.

Submit a plant-specific overpressure valve sizing calculation for Midland.

Also, it is our position that the analysis assume that the reactor scram is initiated by the second safety grade signal from the reactor protection system.

Your above analysis for Midland should be performed accordingly.

211.177 Your response to request 211.131 does not satisfy our concern (6.3) with respect to the detection and isolation of passive ECCS (5.4.7) failures during the long-term cooling phase after a LOCA.

(RSP)

Although the test report addressin, injection pump seals indicates that seal integrity was maintained for the conditions under which they were tested, we do not concur with your proposal for LPI seal leakage of 500 ml/ min to serve as the bounding leak rate for a passive failure following a LOCA (valve stem packing or pump seal failure). Operating data indicates that leak rates in excess of your proposal have occurred.

" Bounding" leak

211-2 rate assumptions on the order of 30-50 gpm have been accepted by the staff in the past; therefore, we require that you show that the ECCS equipment layout, room water level detectors and airborne radiation monitors in the Midland plant meet the criteria listed in request 211.47 assuming leakage rates of this magnitude, or revise your design accordingly.

211.178 Your response to our position in request 211.129 dces not provide (6.3) assurance that the single vent on the BWST is adequate since:

(RSP) 1.

It is on top of the tank and would be susceptible to blockage due to snow buildup.

~

2.

No heat tracing is provided on the vent.

3.

Your response does not describe the " screen inside the BWST" which is heated.

We require that a BWST vent configuration be provided which will preclude vent blockage due to icing or snow accumulation.

Revise your design accordingly.

211.179 Your response to request 211.126 states that flow indication (6.3) in the " dump-to-sump" lines is not necessary. Our position is (RSP) that the operator must be provided with flow indication to confirm that at least the minimum required dilution flow exists subse-quent to a LOCA.

Revise your design and response accordingly.

211.180 Your response to request 211.106 states that the alarm provided (5.2.5) in the control room to detect a reactor building sump level (RSP) increase corresponding to 1 gpm leak within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> will be generated by the plant computer. Since the plant computer may not be available during plant operation, we require that an alarm be provided in the control room which will be available at all times. Revise your design accordingly.

211.181 Your response to request 211.113 states that extended operation (6.3) of the Decay Heat Removal pumps at flows less than 800 gpm would result in damage to the pumps.

(This was your basis for not using continued recirculation through the DHR heat exchanger and recirculation line to protect the DHR pumps from closure of a suction valve). Confirm, with basis, that the low pressure injection system will perform its function in the piggyback mode, since the LPI (DHR) pump flow will be less than 800 gpm.

211-3 211.182 Your response to request 211.103 does not meet our requirements (6.3) with respect to check valve leak testing.

The proposal to test (RSP) two valves in each of the Core Flood and Low Pressure Injection Lines is acceptable for these systems, however, we require that at least two check valves in each of the high pressure injection lines be tested also. This should be done by classifying these valves as AC in accordance with Section XI of the ASME Code.

Modify your response accordingly.

211.183 The response to request 211.152 does not satisfy our concern (15.4.5) that dilution events could occur at rates less than the makeup (RSP) flow rate setpoint, and would not be detected.,Although these events would take longer than 30 minutes to reach criticality, no indication would be provided from the high makeup flow alarm to alert the operator to terminate the event. We require that the operator have adequate time after indication of the event in accordance with the following criteria:

Plant Condition Time Prior to Criticality Af ter Indication Refueling 30 minutes Startup, cold shutdown, hot standby, and power operation 15 minutes Provide assurance that these criteria are met or revise your design accordingly.

211.184 During the recent review of the loss-of-offsite-power preopera-(15.2) tive test procedure for another plant, a concern arose regarding the control of 0TSG level by the auxiliary feedwater system during the event.

Specifically, overcooling of the primary system could result from feeding the OTSG with the cold auxiliary feedwater. The cooldown could be large enough to empty the pressurizer and cause a steam bubble to form in the hot leg high points, which could impede natural circulation and core cooling. Address this concern for the Midland units. Provido the results of an analysis of a loss-of-offsite power assuming the worst-case initial conditions (low power appears to be worst since programmed steam generator level is lowest).

Include plots of steam ge serator level, reactor coolant system temper ature, and pressurizer level.

Discuss your assumptions regarding auxiliary feedwater control. Show that MDNBR will remain above 1,30 and core cooling will not be impaired.

211-4 211.185 Your response to requast 211.157 regarding worst case single (150) failure for a main steam line break is insufficient. The analysis (15.1. 5 )

should consider the following:

(RSP) 1.

Inadvertent atmospheric dump valve opening 2.

Steam flow through all unisolated lines down stream of the MSIV's (Unit 2).

Table 10.3-5 indicates that all lines are not isolated after a steam line break assuming the single failure of one MSIV.

3.

Process steam cross-connect valves opening, (see request 211.160 unless power will be removed.

Provide your basis for stating that one upI pump is the worst single failuce with respect to overcooling.

Provide the worst single failure with basis for the worst DNSR main steam line break.

211.186 Confirm that the bounding Midland Chapter 15 accidents and (150) transient analyses have considered all events which could occur (15.1. 5 )

in Modes 1, 2, 3, and 4 as defined in FSAR Section 7.7.1.6.2.2.

We require that all allowable modes of operation be considered in your safety analysis and be specifically defined in the 'iidland Technical Specifications. We also require that all modes which are physically possiDie but which have not been considered in your safety analyses (e.g., Unit 1 NSSS supplying she Unit 2 turbine) be identified and be specifically prohibited by the Midland Technical Specifications.

O

9 321.0 EFFLdENT TREATMENT SYSTEMS ERANCH 321.6 Justify your position that the proposed extrucer/eva:Or2 tor (11.4) has the capacity for tne combined input from Midland Plant, Unit Nos. 1 and 2.

Your estimates of annual quantities of solid waste and your comparison to other plants in Tables 11.4-1 and 11.4.5 of the FSAR is at variance with the capacity of the proposed solidification system.

You should consider the data from operating PWR's sucn as is given in NUREG/CR-0144, "A Review of Solid Radioactive Waste Practices in Light-Water-Cooled Nuclear Po';er Plants,"

ONRL/NRC, October 1978, and the expected feedrate for the VRS-T120 extruder / evaporator as recommended in Amendr.ent 1, Table III, of the Topical Report ';PC-VRS-001 (Revision 1),

May 1978.

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