ML19210B972

Forwards Request for Addl Info Re Restart Program

ML19210B972
Person / Time
Site:	Crane
Issue date:	10/26/1979
From:	Vollmer R Office of Nuclear Reactor Regulation
To:	Arnold R METROPOLITAN EDISON CO.
References
NUDOCS 7911130101
Download: ML19210B972 (22)

Text

.

' October 26, 1979 Docket No. 50-289 Mr, R. C. Arnold, Senior Vice President Metropolitan Edison Coopany 260 Cherry Hill Road Parsippany, New Jersey 07054 Re: TMI-1 Restart Program - Request for Additional Information

Dear Mr. Arnold:

Ouring our meeting in Middletown on October 17, 1979, we discussed a nunber of questions which the staff had thus f ar developed in reviewing the THI-1

" Restart Report." These and others resulting from our further review of that r ; ort through Amendment 2 are included in tne enclosed request for additional information.

Since you have had most of the enclosed items available to you since October 17, and because of the schedular demands of the Commission's Order of August 9,1979, we will require complete and adequate responses to the cr.cloe,ed items by Novenber 7,1979, at the latest. Earlier partial submittals may expedite our review.

As indicated above, your submittal scheduled for the week of October 22 has not yet been received and reviewed.

We presume much of the missing information identified in our letter of October 2,1979, will be answered by that submittal, and that responses to some of the enclosed items will be included.

However, additional questions may be expected from our review of this material.

Sincerely,

'W~dsamt br:

m an d E. verhase Richard H. Vollmer Director Three Mile Island Support Enclosum:

Request for Additional Information 1323 099 JOI n

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R. C. Arnold Dr. Walter H. Jordon Mr. J. G. Hertein.

881 W. Outer Drive Vice Prasident Huclear Operations Oak Ridge Tennessee 37830 -

Metropolitan Edison Company P.O.. Box 480 Dr. Linda W. Little Middletown, Pennsylvania 17057 5000 Hermitage Drive Raleigh, North Carolina 27612 Mr. E. G. Wallace, licensing Manager Metropolitan Edison Company George F. Trowbridge, Esq.

260 Cherry Hill Road Shaw, Pittman, Potts & Trowbridge Parsippany, New Jersey 07054 1800 M Street, N.W.

Washington, D.C. 20006 Mr. G. P. Miller,

. Manager Support Services and Logistics Karin W. Carter, E::q.

Metropolitan Edison Co@any 505 Executive House 260 Cherry Hill Road P.O. Box 2357 Parsippany, New Jersey 07054 Harrisburg, Pennsylvania 17120 Mr. J. L. Seelinger, Honorable Mark Cohen Superintendent, Unit 1 512 E 3 Main capital Building M_.etrlo ol.ita.n Edison Company Harrisburg, Pennsylvania 17120

...i inii nuoc Parsippany, New Jersey 07054 Ellyn Weiss, Esq.

Mr. J. J. Colitz, Sheldon, Harmon, Roisman & Weiss Manager Plant Engineering 1725 I Street, N.W., Suite 506 Metropolitan Edison Company Washington, D.C. 20006 260 Cherry Hill Road Parsippany, New Jersey 07054 Mr. Steven C. Sho11y 304 S. Market Street Mr. I. R. Finfrock Mechanicsburg, Pennsylvania 17055 Jersey Central Power and Light Company Madison Avenue at Punch Bowl Road Mr. Thomas Gerusky Morristown, New Jersey 07960 Bureau of Radiation Protection Department of Environmental Resources Mr. R. W. Conrad P.O. Box 2063 Pennsylvania Electric Co@any Harrisburg, Pennsylvania 77120 1007 Broad Street Johnstown, Pennsylvania 15907 Mr. Marvin I. Lewis 6504 Bradford Terrace J. B. Lieberman, Esq.

Philadelphia, Pennsylvania 19149 Metropolitan Edison Company 260 Cherry Hill Road Ms. J ane Lee

}.323 lQQ Parsippany, New Jerse,y 07054 R.D. 3, Box 3521 Etters, Pennsylyania 17319 Ms. Mary V. Southard, Chairperson, Citizens for a Safe Walter W. Cohen, Consumer Advocate Environment Department of Justice c/o Metropolitan Edison Company Strawberry Square,14th Floor u

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NRC FORM 318 (9 76) NRCM G240 DU.S. GOVERNMENT PRINTING OFFICE: 1979-289-369

R. C. Arnold October 26,1979 Robert L. Knupp, Esq.

Ms. Marjorie M. Aamodt Assistant Solicitor R.D. #5 Knupp and Andrews Coatesville, Pennsylvania 19320 P.O. Box P 407. N. Front Street Ms. Karen Sheldon Harrisburg, Pennsylvania 17108 Sheldon, Harmon, Roisman & Weiss 1725 I Street, N.W., Suite 506 John E. Minnich, Chairman Washington, D.C. 20006 Dauphin Co. Board of Comissioners Dauphin County Courthouse Earl B. Hoffman Front and Market Streets Dauphin County Commissioner Harrisburg, Pennsylvania 17101 Dauphin County Courthouse Front and Market Streets Robert' Q. Poll ard Harrisburg, Pennsylvani a 17101 Chesapeak Energy Alliance 609 Montpelier Street

• Ivan W. Smith, Esq.

Baltimore, Maryland 21218 Atomic Safety & Licensing Beard Panel U.S. Nuclear Regulatory Comission Chauncey Kepford Washington, D.C. 20555 Judith H. Johnsrud Environruntal Coalition on Nuclear Power

• Atomic Safety and Licensing Appeal 433 Orlando Avenue Board State College, Pennsylvania 16801 U.S. Nuclear Regulatory Ccmission Washington. 0.C. 20555 Ms. Frieda Berryhill, Chairlady Coalition for Nuclear Power Plant

• Atomic Safety and Licensing Board Postponement P anel 2610 Grendon Drive U.S. Nuclear Regulatory Comission Wilmington, Delaware 19808 Washington 0.C. 20555 Ho'lly S. Keck

• Docketing and Service Section Anti-Nuclear Group Representing York U.S. Nuclear Regulatory Comission 245 W. Philadelphia Street Washington, D.C. 20555 York, Pennsylvania 17404 John Levin Esq.

Pennsylvania Public Utilities Comission P.O. Box 3265 Harrisburg, Pennsylvant a 17120 1323 101 Jordon D. Cunningham, Esq.

Fox, Farr and Cunningham 2320 N. Second Street Harrisburg, Pennsylvania 17110 Ms. Kathy McCaughin Three Mile Island Alert. Inc.

23 South 21st Street Harrisburg, Pennsylvania 17104 o r rict );...........

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REQUEST FOR ADDITIONAL INFORMATION 1.

Restart Submittal Section 2.1.1.7.3 (Aux. Feedwater) describes the main feed pump differential pressure sensing equipment as control grade.

Provide safety grade automatic initiation for emergency feedwater.

2.

Provide a description, supplemented with sufficient electrical drawings, of how the manual EFW initiation and control to be added to the design can function in the presence of failures in the automatic initiai; ion and control portion of the design and vice versa.

3.

Provide the results of the detailed loading study on the diesel generators that confirm the acceptability of adding the AFW pumps.

Provide your schedule and a description of the actual testing planned.

4.

Restart Submittal Section 2.1.1.7.6 f (EFW) states that only one flow indicator is to be provided for each steam generator. This is unacceptable.

Describe redundant (diverse) means that can be used for this purpose in the exisitng design or modify your design to meet the single failure criteria.

For example, use of safety grade OTSG level indication as backup indication would be acceptable.

5.

Provide details of the provisions for startup and periodic functional testings the new EFW initiating circuits.

6.

In the Restart Submittal, Section 2.1.1.7.3 you indicate that the Controltron flow-sensing devices to be installed for emergency feedwater flow indication are safety related.

Describe what is meant by " safety related."

Indicate what qualification testing these instrumentations have received, and whether or not they are testable. We will require that these devices be safety grade for TMI-l (see Question #4).

7.

In the Restart Submittal, Section 2.1.1.7.7 you indicate that a functional test of the new manual control valve station and the emerge 7cy feedwater flow instrumenta-tion will be performed at cold shutdown condictions.

In order to properly assure manual emergency feedwater flow control, it is our position that a confirmatory test be perfortned during startup at low power to show adequate operator control under real dynamic conditions.

Provide a test plan for such a test including acceptance criteria, or alternatively provide test data from tests already conducted or actual system response to transients which demonsteate satisfacotry manual EFW control. Differances between tests and the actual system control following trcnsients would ave to be justified.

8.

Describe and justify the methe used to determine the minimum required emergency feed-water flow capacity (sizing criteria). Verify that the minimum emergency feedwater is consistent with your safety analyses for all anticipated accident and transient conditions assuming a single failure of any system component.

9.

You indicate that the failure mode on loss of all air pressure for the emergency feedwater control valves will be changed to fully open in order to assure emergency feedwater flow when required. Verify that this change will not result in possible 1323 102

. overfilling of the steam generator with resulting adverse effects to the secondary side piping system. How long would it take to secure valves? Why is failing full-open best? Why not fail to some modulated position to minimize over-cooling? Assuming the control valves fail open, how much time is available for the operator to shutoff /or throttle flow? Identify the acceptance criteria used to establish this time period.

Provide your calculations.

4

10. The NRR Bulletins and Orders Task Force review of operating reactors in light of the accident at Three Mile Island Unit 2 recently identified additional requirements for auxiliary feedwater systems.

In addition to the requirements identified in this letter, other requirements which may be applicable to the subject facility are expected to be generated by the Bulletins and Orders Task Force. Such requirements are those resulting from our review of the loss-of-feedwater event and the small break loss-of-coolant accident. Our specific concerns include system reliability (other than auxiliary feedwater system), analyses, guidelines and procedures for operators, and operator training. The designs and procedures of your facility should be evaluated against the following requirements to determine the degree of conformance. Provide the results of this evaluation and an associated schedule and commitment for implementation of required changes or actions described in Appendix A.

We require that you provi6e redundant level indications and low level alarms in~

a.

the control room for the EFW system primary water supply to allow the operator to anticipate the need to make up water or transfer to an alternate water supply and prevent a low pump suction pressure conditions from occurring. The low level alarm setpoint should allow at least 20 minutes for operator action, assuming that the largest capacity EFW pump is operating.

b.

We require that you perform a 72-hour endurance test on all EFW system pumps,.

if such a test or continuous period of operation has not been accomplished to date.

Following the 72-hour pump run, the pumps should be shut down and cooled down and then restarted and run for one hour. Test acceptance criteria should include demonstrating that the pumps remain within design limits with respect to bearing /

bearing oil temperatures and vibration and that pump room ambient conditions (temperature, hufnidity) do not exceed environmental qualification limits for safety-related equipment in the room.

Question 8 previously submitted to you requests information on the basis for your c.

emergency feedwater system flow requirements.

Refer to Enclosure 1 for furthe'r details when responding to this question.

d.

We require that plants which require local manual realignment of valves to conduct periodic tests on one EFW system train and which have only one remaining EFW train available for operation, should propose Technical Specifica-tions to provide that a dedicated individual who is in comunication with the control room be stationed at the manual valves. Upon instruction from the control room, this operator would re-align the valves in the EFW system train from the test mode to its operational alignment.

1323 103

. We require that emergency procedures for transferring to alternate sources e.

of EFW supply be available to the plant operators. These procedures should include criteria to infom the operator when, and in what order, the transfer to alternate water sources should take place. The following cases should be covered by the procedures:

- The case in which the primary water supply is not initially available.

The procedures for this case should include any operator actions required to protect the EFW system pumps against self-damage before water flow is initiated; and,

- The case in which the primary water supply is being depleted.

The procedure for this case should provide for transfer to the alternate water sources prior to draining of the primary water supply.

f.

We require that you propose Technical Specifications to assure that prior to plant startup following an extended cold shutdown, a flow test would be perfomed to verify the nomal flow path from the primary EFW system water source to the steam generators. The flow test should be conducted with EFW system valves in their normal alignment.

We require that the EFWS should possess the capability to automatically terminate g.

auxiliary feedwater flow to a depressurized steam generator, and to automatically provide feedwater to the intact steam generator.

h.

We require that licensees having plants with unprotected normal EFW system water supplies should evaluate the design of their EFW systems to determine if automatic protection of the pumps is necessary following a seismic event or a tornado.

The time available before pump damage, the alarms and indications available t'o the control room operator, and the time necessary for assessing the problem and taking action should be considered in determining whether operator action can be relied on to prevent pump damage. Consideration should be given to providing pump pro-tection oy emans such as automatic switchover of the pump suctions to the alternate safety-grade source of water, automatic pump trips on low suction pressure or upgrading the nomal source of water to meet seismic Category I and. tornado protection requirements.

1.

Verify that at least one EFW system pump including its associated suxiliaries such as the turbine driven pump lube oil cooling system, and its associated flow path and essential instrumentation will automactically initiate EFW system flow and is capable of being operated independently of any alternating current power source for at least two hours. Conversion of direct current power to alternating current is acceptable.

j. We require that you evaluate the consequences of a postulated break in the steam line to the turbine-driven EFW pump to detemine the need to qualify the EFW system valves, valve actuators, and instrumentation for the environmental conditions resulting from such a high energy line break in order to maintain operability of the motor-driven EFW pumps and their associated flow trains.

1323 104 k.

We require that you provide automatic emergency feedwater initiation on low level indication in either steam generator in order to assure adequate steam generator inventory. The steam generator level sensing instrumentation shall be safety grade as required for all automatic emergency feedwater actuation signals.

11. The Restart Submittal Section 2.1.1.3.1.2 (Pressurizer Heaters) states that separation of Class lE and non Class 1E circuits will be in accordance with Regulatory Guide 1.75 wherever practicable. Provide the actual details of your design and augment your discussion with the details of how your new design maintains separation and isolation between the redundant Class 1E portion of the design.

.12. Provide sufficient electrical one line diagrams to facilitate our review of power assignments for ALL_ new and/or realigned equipment.

Provide a detailed list-ing of all the above equipment which gives power source, power requirements and demonstrates the ability of each power source identified to provide the additional power requirements without degradation.

13. The Restart Submittal (Section 2.1.1.2) states that each pressurizer safety valve at the power operated relief valve will be provided with elbow tap differential pressure transmitters to measure downstream discharge flow through each valve.

Provide the calculations, assumptions and descriptions of tests run by B&W which demonstrate that a satisfactory signal will be generated when any of the valves are open and conversely indicate positively that the valves have reclosed.

Provide a complete description of the accelerometers, method of mounting on the PORV, and the test desctiption, including relief valve used, conditions, etc.,

which demonstrate that this device is suitable for the function being performed.

We require that the indication of PORV and safety valve position be seismically qualified (rather than just mounted) and that the indication be environmentally qualified for the appropriate environment (any transient or accident which would cause the relief or safety valve to lift.)

Describe the tests that will be performed after installation to verify satisfactory operation of the new position indicators.

14. Your response (Section 2.1.1. 3.1. 2) indicates that Babock & Wilcox has recommended at least 126 kw of pressurizer be assured power within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after a loss of offsite power.

Provide or reference the calculations / tests which serve as the basis for this recommendation.

Include all assumptions, conservatisms and capacity margin available.

Address the situation for which the pressurizer may empty on a loss of offsite power, and therefore require more heater capacity to recover subcooling.

15.

(Section 2.1.1.2)

Include the use of the new pressurizer relief and safety valve indications in all appropriate procedures.

For example, EP-1202-29 does not direct the operator to refer to these indications to identify a stuck open pres-surizer safety valve. Submit revised procedures for our review.

1323 105

16. We note that you intend (Section 10.2.8) to participate in an Industry Program to test the pressurizer relief and safety valves (NUREG-0578, 2.1.2).

Provide the name of the industry group sponsoring the test program and a date when the program will be complete.

17.

The requirements of paragraph 2.1.3.b of NUREG-0578 did not intend that short-tenn modifications to existing instruments should be considered to the exclusion of existing, unmodified instrumentation (such as reactor coolant pump current and flow measurement) to detect inadequate core cooling. The Met-Ed response considers only the modified instrumentation.

Identify the existing, unmodified instruments which were ccasidered for detection of inadequate core cooling.

Describe those instruments which were selected for recognition of this condition.

Perform analysis and implement procedures and training based on those unmodified instruments as well as the modified instrumentation described previously.

18.

Identify those procedures which require the use of a) in-core thermocouples, b) wide range reactor outlet emperature measurement, c) reactor coolant saturation pressure margin, and d) other instrumentation identified in the response to Question

1. 17 above.

19. The number of operational in-core thermocouples available to assess core conditions is a function of core design (which may change during future reloads) and thermocouple reliability. State the minimum number of as-designed in-core thennocouples considered acceptable in the Met-Ed proposal.

State the minimum fraction of installed thermocouples which must be operational in the Met-Ed proposal. Show the analysis used to reach these conclusions.

a description of the proposed subcooling meter as requested in Appendix B.

Rest Wttal Section 2.1.1.5.1.3) explicitly state whether or not the non-ECCF s.,

t

'rvices for RCP operation will be upgraded to Seismic Category I

$1

'ec = rom BOTH pipe whip and jet impingement.

> ' 'ii Na. :y.mentify the valves referred to in the Restart submittal Section

1..l.1 hv penetration number and valve designation.

.ta=. Submittal Section 2.o.l.5.2) - Provide detailed descriptions for subparts

, 3, 4 5, 7, and 9.

Provide sufficient electrical drawings to allw an independen 3

evait ation for subparts 1, 2, 3, 4, 5, and 7.. Provide documentation to support subpart 6.

S.

icrovide the documentation that supports your statement that the RCP's can run for one week with only seal water providing the cooling for the pump seals.

3.

Restart Submittal Section 2.1.1.5.5.2 states:

" Spurious initiation of an isolation signal will not introduce transients into the plant that are of significance. Thus, no new accidents /transicnts are introduced into the plant cesign."

Provide the detailed bases that support the above conclusions.

1323 106 26.

Your description of the Containment Isolation System does not specifically address if any of the isolated valves will automatically revert to their pre-isolated state (i.e., reopen). Address this aspect of your design.

27. (Restart Submittal Table 2.1-2) a.

Valves CM-V1, CM-V2, CM-V3 and CM-V4 do not receive a diverse safety grade ~ automatic isolation signal.

This is unacceptable. Modify your design accordingly, b.

Valve MU-V3 does not receive a diverse safety grade automatic isolation signal.

This is unacceptable. Modify your des _ign accordingly.

c.

Valves IC-V2, IC-V3, IC-V4 and IC-V6 do not receive a diverse safety grade automatic isolation signal unless a safety grade line break isolation signal is included in the design. Sufficient details have not been provided to ascertain if this design feature is incorporated in your design.

If the line break detection feature is incorporated in your design, provide the details supplemented with sufficient electrical drawings to allow an independent evaluation of whether or not the design meets IEEE Std. 279-1971.

If the line break detection feature is not-incorporated in your design, the design is unacceptable and must be modified accordingly, d.

Valves-RB-V2A and RB-V7 are listed with three identical options and accompanying note. The first option provides automatic isolation only on hi-hi containment pressure. This option is unacceptable. Options two and three are acceptable providing that the hi-containment pressure isolation signal is retained. Modify your description by deleting the multiple choice options and providing the specific details of your design consistent with the above evaluation.

e.

Valves MU-V33A, MU-V33B, MU-V33C and MU-V33D receive no automatic safety grade isolation signals. Contaiament hi-radiation is used to provide an alarm for subsequen remote-manual operator action. Provide the bases and rationale why diverse parameters are not included and why automatic isolation is not required.

In the absence.of acceptable justificaiton, we shall require single-failure-proof safety-grade automatic isolation of these valves.

28.

The TMI-2 accident highlighted the fact that relatively high levels of radio-activity can be released inside containment without an associated significant pressure rise.

In addition, a significant amount of reactor coolant was released through the PORV before the 4 psig setpoint was reached.

It appears that the 4 psig setpoint may be too high a level to properly initiate containment isolation. Evaluate the merits of lowering the containment hi-pressure setpoint to something like 2.0 or 2.5 psig, and report your findings.

29.

You state (Section 10.2) that item 1 of Bulletin 05A, which requires you to review a chronology of the TMI-2 accident so that an understanding of the events will ensure against such an occurrence at Unit 1, is "not applicable." Correct this statement and describe the adequacy of your review.

1323 107

30. Your response (Section 3.2)~ to the requirements of item 3 in Bulletin 05B states that you are still investigating the changes to the PORV and high pressure reactor trip set points.

Novide your comitment to make these changes or alternatives such that you respond to this bulletin item.

31. Your response (Section 2.1.1.1) to. Bulletin 05B, item 5, only addresses control-grade reactor trips. Provide for our review a design / schedule for implemen-tation of a safety-grade automatic anticipatory reactor scram prior to restart for loss of feedwater, turbine trip, and/or low steam generator level.

32. Your response (Section 3.1.1) to Bulletin 05A, items 3 and 4, and Bulletin 05B, item 1, indicates that procedures have been and are still being revised.

Provide the necessary procedures for our review and/or schedule for their completion.

33.

Your response (Section 3.1.1) to Bulletin 05C indicates that you are still evaluating this bulletin and will revise procedures and that supporting analyses will be submitted later. Provide a schedule for submitting this infomation.

Provide a safety-grade pump trip design description.

34.

Provide detailed scenarios of the two reactor trip / turbine trip events

(#11, #12) discussed in section 10.3.1.

Include (1) RCS temperature and corresnonding pressure vs. time; (2) effects of liquid relief on PORV in Reactor Trip #11; and (3) corrective action for mainsteam safety valves failure to reseat (redesign new valves, etc.?). Were these events reported as LER's? If so, provide copies.

35.

The small break LOCA analyses assumptions state that the TMI-l conditions are more conservative than the generic analysis assumption.

In this regard, justify an ESFAS trip for the plant at 1500 psig rather than 1600 psig as assumed in the generic analysis.

(The lower TMI-l pressure setting would result in later HPI initiation.)

(p.8-16) Should you recommend changing the ESFAS

• trip to 1600 psig, confim that the accident and transient analyses for TMI-l will still show acceptable consequences.

36.

It is stated that a small break LOCA analysis performed at 2535 MWt with an HPI split of 64%/36% is acceptable since the generic analysis is performed at 2772 MWt and a 70%/30% split. Justify the statement.

(p.8-17).

a.

Provide a schedule for the HPI line break analysis which demonstrates that 250 gpm core makeup is adequate.

(p.8-17) b.

Provide drawings depicting the location of the venturis as well as the new cross connects.

c.

Describe the testing which will be done to confinn the adequacy of this design change and address the impact of this change on the cross connect modification 1323 108

37. Your submittal states that cavitating venturis will be added to the high pressure injection lines to eliminate the need for operator action on an HPI line break.

38. Submit modified Technical Specifications to support the procedural and design modifications recuired by the Order and Bulletins 05A, 05B and 05C.

39.

NUREG-0578, Section 2.1.3.b, requires the development of procedures to be used by the operator to recognize inadequate core cooling.

a.

Describe the guidance provided the operator to recognize inadequate core cooling.

Include existing and future instruments to be used and the expected instrument response under conditions of inadequate core cooling, b.

Describe the training provided licensed operators with regard to recognizing inadequate core cooling.

40. Your submittal indicates that the Lessons Learned requirements on Shift Supervisor Responsibilities (NUREG-0578, Section 2.2.1.a) will be provided at a later date.

Provide the schedule for completion of this itti.

41.

Paragraph 5.4.5 of the Met-Ed/GPU TMI-l Restart submittal indicates that the Shift Foreman should hold a Reactor Operator License, while Figure 5.2.1 indicates that a Senior Operator License is required. Since a Senior Operator License is necessary to direct the activities of Reactor Operators, revise paragraph 5.4.5 to require Shift Foremen to hold a Senior Operator License.

42. Paragraph 5.4.8 of the Met-Ed/GPU TMI-l Restart submittal outlines the educa-tional background requirements for a Shift Technical Engineer (STE).

• a.

Provide the details of the training the STE will receive to assure a thorough knowledge of normal reactor operations, anticipated transients,

, and effects of multiple equipment failures and operational errors.

b.

Describe the normal duties of the STE and his proximity to the control room.

c.

Describe the responsibility and authority of the STE during off normal situations.

43.

Provide the schedule for submittal of information required by NUREG-0578, Section 2.2.1.c Shift and Relief Turnover Procedures.

44.

Provide the schedule for submittal of information required by NUREG-0578, Section 2.2.2 (a, b, c) Control Room Access, Onsite Technical Support Center, and Onsite Operational Support Center.

• Information Request at 10/17 Meeting 23 109 45.

(OrderItem1(d)) Your response to this item indicates that procedures have been or are still being revised.

Provide the procedures developed to define operator action during small break LOCA's.

46.

(Order Item 1(e))* All licensed operators at B&W facilities have received a special exam on the TMI-2 accident; including transient effects, operator response, and related procedure / design changes. Provide a similar exam for the TMI-l licensed operators with a minimum passing grade of 90%. Also provide the details for retraining those individuals who may score less than the minimum.

47.

(Order Item 1(e))* Section 6.6.D.1 of the Operator Accelerated Retraining Program (0ARP) states that an audit exam will be given to all licensed operators.

Provide the passing criteria for this examination and the retraining /reauditing for those who may not achieve the passing score.

48.

(Order Item 1(e))* Outline how licensed operaturs will be provided with information related to procedure / design changes that are implemented after the completion of the related training module.

-4 9.

(Order Item 1(e))* Describe how the specific program objectives of the 0ARP (Section 6.2) will be factored into the future training and requalification of operators.

50.

(Order Item 1(e)) To what extent will outside organizations be used to provide training in the 0ARP? This includes organizations that may be providing training material, instructors, audit exams, etc.

51. Bulletin 05A Item 3 What guidance is provided operators to enhance core cooling in the event that voids in the primary system are large enough to compromise core cooling capability, especially natural circulation capability?

52. Bulletin 05A Item 4 How does the 0ARP emphasize the use of various plant parameter indications in evaluating plant conditions?

53. Bulletin 05A Item 5 Your response to these item is incomplete.

Outline how you intend to review all safety-related valve positions and positioning requirements to assure that valves are positioned in a manner to ensure proper operation of engineered safety features.

0

• Information requested at 10/17/79 meeting

54. Bulletin 05A Item 5 Comit to perform an independent valve alignment verification when returning the emergency feedwater system to operability after maintenance or surveillance testing.

55.

Bulletin 05B Item 1 Your procedure 1102-16, Natural Circulation, includes anticipatory filling of the OTSG prior to securing the reactor coolant pumps.

Submit the analysis performed to provide guidance as to the expected system response.

56.

Bulletin 05B Item 2 State the procedural guidance and training provided operating personnel with respect to the override /

tennination. of engineered safety features.

57. Bulletin 05B Item 6 Outline the guidance provided operating personnel for prompt notification of the NRC.

58.

Bulletin 05B Item i Identify the procedural controls that have been implemented to assure proper positioning of the emergency feedwater system manual valves and manually operated motor-driven valves.

59.

Bulletin 05B Item 9 Identify the procedures and guidance provided operators with respect to resetting containment isolation.

60.

Bulletin 05B Item 10 Your response indicates that shift relief procedures will be used to notify reactor operating personnel when safety-related systems are removed from or returned to service.

Regulatory Guide 1.47 addresses visual indications for bypassed and inoperable status of equipment.

Outline the design features or procedural steps that provide visual indication at TMI-1.

61.

• Bulletin 05B Item 11 Your submittal indicates that the response to this item will be provided at a later date.

Provide the schedule for completion of this item.

62.

Bulletin 05C Item 1 Your response did not include Section B of this item.

State your intention to provide two licensed operators in the control room at all times during operation.

63.

Your submittal states that you will provide an automated switchover to recirculation on the emergency core cooling system. Provide a detailed description and associated drawings related to this modification.

64.

Your submittal states that you intend to modify the reactor building spray system.

Provide the detailed description and associated drawings related to this modification.

1323 111

65. The Radiation Protection and Chemistry Supervisor or some other individual knowledgeable in health physics practices shall be present at PORC meetings, whenever health physics related procedures or policies are reviewed, Revise Section 3,1 of the Restart Submittal to incorporate this practice in the PORC Quorum description.

66. The ANSI Standard (.N18,1-1971) referenced in Section 5.4.10 of the Restart Submittal has been substantially revised. The current applicable standard is ANSI /ANS-3.1-1978.

Provide assurance that the minimum qualifications of

.the Supervisor-Radiation Protection and Chemistry comply with those stipulated in ANSI /ANS-3,1.

67.

The licensee has not described the proposed radiation protection program plan.

We will require a detailed description, analogous to recent FSAR submittals, taking into account 1.essons learned and other additional con-siderations reflecting problem areas raised by the accident. Areas to be coverc1 are characterized in Appendix C.

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A-1 APPENDIX A Basis for Auxiliary Feedwater System Flow Requirements As a result of recent staff reviews of operating plant Auxiliary Feedwater Systems (AFWS), the staff concludes that the design bases and criteria provided by licensees for establishing AFWS requirements for flow to the steam generatcr(s) to assure adequate removal of reactor decay heat are not well defined or documented.

We require that you provide the following AFWS flow design basis infor-mation as applicable to the design basis transients and accident conditions for your plant.

1.

a.

Identify the plant transient and accident conditions considered in establishing AFWS flow requirements, including the following events:

1)

Loss of Main Feed (LMFW)

2) LMFW w/ loss of offsite AC power 3)

LMFW w/ loss of onsite and offsite AC power

4) Plant cooldown

5) Turbine trip with and without bypass

6) Main steam isolation valve closure

7) Main feed line break

8) Main steam line break

9) Small break LOCA

10) Other transient or accident conditions not listed above 1323 113

A-2 b.

Describe the plant protection acceptance criteria and corresponding technical bases used for each initiating event identified above.

The-acceptance criteria should address plant limits such as:

- Maximum RCS pressure (PORV or safety valve actuation)

- Fuel temperature or damage limits (DNB, PCT, maximum fuel central temperature)

- RCS cooling rate limit to avoid excessive coolant shrinkage

- Minimum steam generator level to assure sufficient steam generator heat transfer surface to remove decay heat and/or cool down the primary system.

2.

Describe the analyses and assumptions and corresponding technical justification used with plant condition considered in 1.a. above including:

a.

Maximum reactor power (including instrument error allowance) at the time of the initiating transient or accident.

b.

Time delay from initiating event to reactor trip.

c.

Plant parameter (s) which initiates AFWS flow and time delay between initiating event and introduction of AFWS flow into steamgenerator(s).

d.

Minimum steam generator water level when initiating event occurs.

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A~-3 e.

Initial _ steam generator water inventory and depletion rate before and after AFWS flow commences - identify reactor decay heat rate used.

f.

Maximum pressure at which steam is released from steam generator (s) and against which the AFW pump must develop sufficient head.

g.

Minimum number of steam generators that must receive AFW flow; e.g.1 out of 2?, 2 out of 47 h.

RC flow condition - continued operation of RC pumps or natural circulation.

i. Maximum AFW inlet temperature j.

Following a postulated steam or feed line break, time delay assumec.to isolate break and direct AFW flow to intact steam generator (s). AFW pump flow capacity allowance to accomodate the time delay and maintain minimum steam generator water level.

Also identify credit taken for primary system heat removal due to blowdown.

k.

Volume and maximum temperature of water in main feed lines between steam generator (s) and AFWS connection to main feed line.

1.

Operating condition of steam generator normal blowdown following initiating event.

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A-4 Primary and secondary system water and metal sensible heat m.

used for cooldown and AFW flow sizing.

n.

Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory.

3.

Verify that the AFW pumps in your plant will supply the necessary flow to the steam generator (s) as determined by items 1 and 2 above considering a single failure.

Identify the margin in sizing the pump flow to allow for pump recirculation flow, seal leakage and pump wear.

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Information on the Subcooling Meter Plant Name:

Vendor:

Reference for Information:

Display Information Displayed (T-Tsat, Tsat, Press, etc)

Display Type (analog, Digital, CRT)'

Continuous or on Demand Single or Redundant Display Location of Display Alarms (include setpoints)

Overall uncertainty ( F, PSI)

Range of Display Qualifications (seismic, environmental, IEEE279)

Calculator Type (process computer, dedicated digital or analog calc.)

If process computer is used specify availability.

(% of time)

Single or redundant calculators Selection Logic (highest T, lowest press)

Qualifications (seismic, environmental, IEEE279)

Calculational Technique (Steam Tables, functional fit, ranges)

Input Temperature (RTD's or T/C's)

Temperature (number of ansors and locations)

Range of temperature sensors 1323 117

B-2 Uncertainty *of temperature ~ sensors (*F at 1 o)

Qualifications (seismic, environmental, IEEE279)

Pressure (specify instrument used)

Pressure (number of sensors and locations)

Range of Pr.:ssure sensors Uncertainty *of pressure sensors (PSI at 1 a)

-Qualifications (seismic, environmental, IEEE279)

Backup Capability Availability of Temp & Press Availability of Steam Tables etc.

Training of operators Procedures

• Uncertainty assessment must address conditions of forced flow and natural circulation.

1323 118 t

~

TEtiTATIVE OUTLINE OF AREAS TO BE COVERED 1.

Policy -

Implementation Capability QA 2.

Management -

Responsibilities Communications Organization - reporting chain Staffing clerical.suppor,t and appropriate specialists Qualifications of Staff - knowledge of systems, procedures Training - frequency, scope Procedures 3.

Communications -

Status of plant activities H.P. problem identification Changes in procedure Maintenance planning Changes in RWP work scope 4.

Personnel Dosimetry -

Badges, pocket dosimeters Calculations Bioassay Whole Body counting Review and evaluation 5.

Radiation Control -

Access control - RWP's fionitoring - Portable - Fixed Control procedures - Adm. - ALARA 6.

Contamination Control ~

R/A Mat'l control - Posting, etc.

fionitoring Program Surface - Air Clothing Control Procedures / Boundary 1323 l19

7.

Internal Dose Control -

Resp. Protection Air Monitoring Equipment Analysis - qual / quant.

8.

Design / Engineering reviews for Radiation Protection 9

Training -

Plant - Management H.P. Tech.- Records 10.

Instrumentation -

Capability to monitor for Iodines Selection, maintenance, calibration and numbers 11.

Internal QA Audits

12. Special Problems -

Beta Dosimetry Interface with Unit 2 1323 120

o R. C. Arnold October 26, 1979 Distribution:

j Docket No. 50-289 NRC PDR Local PDR SEP/TMI r/f PSB/TMI r/f-P-433 TMI Site r/f NRR r/f HDenton ECase DEisenhut RTedesco BGrimes RYollmer JCollins HSilver SMiner TJCarter ftiulkey, ELD JTourtellotte. ELD Attorney / ELD IE(3)

RReid CNelson DDil anni RIngram, LA JRBuchanan, NSIC

-TERA ACRS(16)

FAllenspach JGilray WHaass Applicant's Attorney of Record Parties to Proceeding SfrWBrtfy

<T fe.: zWr "

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4 D U S. GOvCrtNVENT PRINTING OFrlCE: 1979-249-269 une coeu 3ts (9 76) NRCM O240