ML19210E253
| ML19210E253 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/26/1979 |
| From: | Vollmer R NRC - TMI-2 OPERATIONS/SUPPORT TASK FORCE |
| To: | Arnold R METROPOLITAN EDISON CO. |
| References | |
| NUDOCS 7912040153 | |
| Download: ML19210E253 (22) | |
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October 25, 1979 55
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Cocket No. 50-239 6
W" Mr. R. C. Arnold, Senior Vice President N
o-Metrepolitan Edison Coceany 260 Cherry Hill Road P arsippany, New Jersey 07054 Re: TMI-1 Restart Program - Request for Additional Information
Dear Mr. Arnold:
During our :neeting in Middletown on October 17, 1979, we discussed a nurcer of questions which the staff had thus f ar developed in reviewing the TMI-1 "R est art R eport. " These and others resulting from our further review of that reper,: through Amendment 2 are included in the enclosed request for additional information.
Since you have had :nost of the enclosed items available to you since October 17, and because of the schedular demands of the. Commission's Order of AugJst 9,1979, we will require comolete and adequate responses to the enclosed items by Novecter 7,1979, e th.e latest.
Earlier parti al submittals may expedite our review.
As indicated above, your submittal scheduled for the week of Octcber 22 has not yet been received and reviewed. We presume nuch 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 materi al.
i) p' Sincerely, m
pichard H..Yallmer, Of rector n
R Three Mile Island Support
Enclosure:
Request for Additional Information I
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N Mr. J. G. Herbein, Dr. Walter H. Jorden Yica President Muclear Opert.tions 831 W. Outer D M ye Metro;olitm Edison Coccany Gak Ridge Tennessee 37830 P.O. Icx a80 Middlatcwn, Pennsylvania 17057 Dr. Linda W. Little 50C0 Hermitage DMye Mr. E. G. Wallace, Licersing Manager Raleign, North Carolina 27612 Metro:olitm Edison Corpany 250 Cherry Hill Road Jecrge F. Trewbridge, Esq.
Pzrsi;pany, New Jerst.y 07054 Shaw, Pittman, Potts & Trowbridge 1800 M Street, N.W.
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G. P. Miller, Washington, D.C. 20006 Haager Succort Services and Logistics Met c:cli:an Ediron Coccany X arin W. Car:er, Esq.
250 Cherry Hill Acad 505 Executive House P zrsi:pany, Nen Jersey 07054 P.O. Box 2357 Harrisburg, Pennsylvani a 17120 Mr. J. L. Se'elinger, Super'n=endent, Unit 1 Honorable Mark Cohen Metro:olitm Edison Coccany 512 E-3 Main Capital Building 250 Cherry Mill Road Harrisburg, Pennsylvani a 17120 P esi:p ry, N=w J-ney 07054 Ellyn Weiss, Esq.
Mr. J. J. Colitz, Sheldon, Harnon, Roisman & Weiss Mancger Plant Engineeri,g 1725 I Street, N.W., Suite 506 Me*,re:olitan Edison Coco my Washington, D.C. 20006 2'20 Cherry Hill Rcad Pzrsi;pany, New Jersey 07054 Mr. Steven C. Shelly 304 S. Market Street Mr. I. R. Finfrock Mechanicsburg, Pennsylyania 17055 Jersey Central Power and Light Coc:pany Madis:n Avenue at Punch Scwl Road Mr. ' Thomas Gerasky Mer.-istcwn, New Jersey 07960 Sureau of Radiation 9totection Department of Environmental Resources Mr. R. W. Ccnrad P.O. Box 2063 Pe.'msylvani a Electric Comoany Harrisburg, Pennsylvani a 17120 1007 3rcad Street Johns.cwn, Pennsylvania 15907 Mr. Marvin I. Lewis 6504 Bradford Terrace J. 3. L ieberman, Esq.
Philadelphia, Pennsylvania 19149 Metro:ciitan Edison Cocpany 250 Cherry Hill Road Ms. J =.ne Lee Parsippany, New Jersey 07054 R.D. 3, Box 3521 Etters, Pennsylvania 17319 Ms. Mary Y. Southard, Chzfrperson, Citizens for a Safe Walter W. Cohen, Consucer Advocate Environnent
, Department of Justice e/o Metropolitan Edisen Ccmpany Strawoern Square,14th Floor 250 Cherry Hill Road Harrisburg, Pennsylvania 17127 Parsippany, New Jersey 07054
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R. C. Arnold Oceber 25, 1979 Robert L. Knuop, Esc.
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, Har:non, Roisman & Weiss 1725 I Street, N.W., Suite 506 Jchn E. Minnich, Chairman Washington, D.C. 20006 Dauphin Co. Board of Comissioners Dauphin County Courthouse Earl S. Hoffman Front and Market Streets Dauphin County Cocnissioner HarHsburg, P~ennsylvani a 17101 Dauchin County Cour. house Front and M arket Streets
'3.ob e-t Q. 'J oll ard Marrisburg, Pennsylvani a 17101 Chesa::eak Energy Alli ance 609 Montpelier Street
- Ivan W. Smith, Esq.
Baltimore, Maryland 21218 Atomic Safety & Licensing Board P anel U.S. Nuclear Regulatory Ccmission Chauncey Kepford Washington, D.C. 20555 Judith H. Johnsrud Environcental Coalition on Nuclear Pcwer
- Atomic Safety and Licensing Appeal 433 Orlando Avenue Board State College, Pennsylvani a 16801 U.S. Nuclear Regul atory Connission Washington, D.C. 20555 Ms. Frieda Serryhill, Chairlady Coalition for Nuclear Power Plant
- Atomic Saf'ety and Licensing Board Pestponecent P anel 2610 Grendon Drive U.S. Nuclear Regulatory Ccmission Wilmington, Delaware 198L3 Washington 0.C. 20555 Holly S. Keck
- 0ccketing and Service Section Anti-Nuclear Group Representing York U.S. Nuclear Regulatory Ccamission 245 W. Philadelphia Street Washington, D.C. 20555 York, Pennsylyani a 17404 Jchn Levin, Esq.
Pennsylvania Public Utilities Ccmission P.O. Box 3255 Harrisbufg, Penitsylvania 17120 Jordon D. Cunningham, Esq.
Fox, Farr and Cunninghas 2320 N. Second Street Harrisburg, Pennsylvania 17110 Ms. Kathy McCaughin Three Mile Island Alert, Inc.
23 South 21st Street Harrisburg, Pennsylvania 17104 I 4 () 8 2 O.
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u RE00EST FOR ADDITIONAL INFORMATI0'N Dg g
t I st 1.
Restart Subeittal section 2.1.1.7.3 (Aux. Feedwater) describes the m...
feed pug differential pressure sensing ecuipment 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 initiation and control portion of the design and vice versa.
3.
Provide the results of the detailed loading study on the diesel generators that confim the acceptability of adding the AFW pu=ps.
Provide your schedule and a description of the actual testing planned; 4.
Restart Submittal Section 2.1.1.7.5 f (EFW) states that only one flow indicator is o be provided for each steam generator. Tnis is unacceptable. Describe redundant (diverse) means that can be used for this purpose in the exisitng cesign or modify your cesign 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 EW 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 ficw indication are safety related.
Describe what is meant by " safety related." Indicate what
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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 emergency feedwater ficw instrumenta-tion will be performed at cold shutdown condictions.
In order to properly assure anual emergency feecwater flow control, it is our position that a confirmatory tast be performed 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 EW control. Differences between tests and the actual _ system control following transients would have to be justified.
8.
Describe and justify the mthod used to deternine the minimum required emergency feed-water flow capacity (sizing criteria). Verify that the minimum estrgency feedwa*ar 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 er:ergency feedwatar control valves will be changed to fully open in order to assure emergency feecaster flow when required. Verify that this change will not result in possible n
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overfilling f the steam generator with resulting adverse effects to the secondary side piping system. How long 'ould it take to secure valves? Why is failing full-open best? Why not fail to some modulated position to minimi:e over-cooling? Assur. ling the control valves fail open, hcw much time is availacie for the operator to shutoff /or throttle flow? Identify the acceptance criteria used
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to establish this time period. Provide your calculations.
- 10. The NRR Bulletins and Orders Task Force review of coerating 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. Tne designs and procedures of your facility shou': be evaluated against the following recuirements to deternine the degree of ccnfon.wi. Provide the results of this evaluation and an associated schedule and conn!cment for imolementatien of required changes or actions described in Appendix A.
We require that you provide redundant level indications and icw level alarms in a.
the control rocm for the EFW system primary water supply to allcw the operator to anticipate the nted to make up water or transfer to an alternate water supply and prevent a icw oump suction pressure conditions from occurring. The low level alam setpoir,+, should allow at least 20 minutes for operator action.
assuming that the lawt capacity EFW purp is operating.
b.
We require that you perfom a 72-hour endurance test on all EFW system pumps,,
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if such a test or continuous period of operation has not been accomplished to date.
Following the 72-hour puma run, the pumps should be shut dcwn and cooled down and then restarted and run for one hour. Test acceptance criteria should include demonstrating that the pumos remain within design limits wi2 respect to bearing /
bearing oil temperatures and vibration and that pump rocm ambient conditions (temoerature, humidity) do not exceed environmental qualification limits for
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safety-rtlated equipment in the room.
Question 8 previously submitted to you requests infomation on the basis for your c.
emergency feedwater system flow requirements. Refer to Enclosure I for further 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 wno 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 tnin from the test mode to its operational alignment..
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3 We require that errgency procedures for transferring to alterr. ate sources e.
of EW supply be available to the plant operators. These procedures should include criteria to infonn the operator wnen, and in wnat order, the transfer to alternate water sources should take place. The following cases should be covered by the precedures:
The case in which the primary water suoply is not initially available.
The procedures for this case should include any operator actions required to protect the EFW system pumos against self-damage before water flow
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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 prooose Technical Specifications to assure that criar to plant startup following an extended cold shutdown, a flow test would be performed to verify the normal flow path from the primary EW system water source to the steam generators. The flow test should be conducted with EW system valves in their normal alignment.
We require that the EWS 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 nonral EW system water supplies should evaluate the design of their EW systems to detennine if automatic protection af the pumos is necessary following a seismic event or a tornado.
The time available before pump damage, the alarms and indications available to the control room operator, and the time necessary for assessing the problem ~and taking action should be considered in detennining whether operator action can be relied on to prevent pu=o damage. Consideration should be given to providing pumo pro-taction by 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 uagrading the nor:al source of water to meet seismic Category I and _ tornado protection requirements.
- i. Verify that at least one EW system pu=a including its associated suxiliaries such r, the turbine driven pump lube oil cooling system, and its associated flow path and essential instrumentation will automactically initiate EW system flow and is capable of being operated independently of any alternating current ' power source for at least two hours. Conversion of direct current pcwer to alternating current is acceptable.
- j. W require that you evaluate the consequences of a postulated break in the steam line to the turbine-driven EW pump to detertine the need to qualify the EW -
system valves, valve actuators, and instrumentation for the environmental conditions resulting from sucn a nigh energy line break in order to maintain operability of the actor-driven EW pumos and their associated flow trains.
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We require that you provide automatic emergency feedwater initiation en icw level indication in either steam generator in order to assure adetuate steam generator inventor /. The steam generator level sensing instrumentation shall be safety grade ac required for all automatic emergency feedwater actuation signals.
- 11. The Restart Submittal Section 2.1.1.b.1.2 (Pressurizer Heaters) states that separation of Class 1E and non Class 1E circuits will be in accordance witn Regulatory Guide 1.75 wherever practicable. Provide the actual detafis 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 tne aoove equipment which gives pcwer source, pcwer recuirements and demonstrates the ability of each power source identified to provice the additional power requirements without degradation.
- 13. The Restart Submittal (Section 2.1.1.2} states that each pressurize safety valve and 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, asst. ptions 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
^0RV, and the test desctiption, including relief valve used, conditions, e tc.,
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 (inther than just mounted) and that the indication be environmentally qualified for the appropriata 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 reconnended' at least 125 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 4 loss of offsite power. Provide or reference the calculations / tests whieft serve as t'Te tasis for this reconnendation.. Include all assumotions, conservatisms and capacity margin available. Addrers the situation for which the pressuri:er 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 pressuri:er relief and safety valve irdications in all appropriate ' procedures. For examole, EP-1202-29 does not direct the operator to refer to these indications to identify a stuck open pres-
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surizer safety valve. Submit revised procedures for our review.
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- 16. We note that you intend (Section 10.2.3) to participate in an Industry Program to test the pressurizer relief and safety valves (NUREG-0578, 2.1.2).
Provide the nare 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-tem audifications 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 nudified instrumentation.
Identify the existing, unmodified instrumen*a which were considered for detection of inadequate core cooling. Describe those instruments which were selected for recognition of this condition. Perfor.n analysis.and implement procedures and training based on those unmodified instruments as well as the modified instrumentatien described previously.
18.
Identify those procedures which require the use of a) in-core themoccuoles, b) wide range reactor outlet emperature casurement, c) reactor coolant saturation pressure margin, and d) other instrumentation identified in the response to Question
- 17 above.
- 19. The nt.n.ber of operational in-core thernoccuples available to assess core conditions is a f.nction of core design (which may change during future reloads) and themoccuple reliability. State the minimum nunter of as-designed in-core themoccupies considered acceptable in the Met-Ed proposal. State the minimum fraction of installed ther.noccupies which must be operational in the Met-Ed proposal. Show the analysis used to reach these conclusions.
- 20. Provide a description of the proposed subccoling meter as requested in Appendix 3.
- 21. (Restart Submittal Section 2.1.1.5.1.3) explicitly state whether or not the non-ECCS support services for RCP operation will be upgraded to Seismic Category I AND protected frcm BOTH pipe whip and jet impingement.
- 22. Specifically identify the valves referred to in the Restart submittal Section-2.1.1.5.1.3 by penetration nuceer and valve designation.
- 23. (Restart Submittal Section 2.o.l.5.2) - provide cetailed description: for suboarts 1, 2, 3, 4, 5, 7, and 9.
Provide sufficient electrical drawings to allw an independem evaluation for subparts 1, 2, 3, 4, 5, and 7.
Provide documentation *w support subpart 6.
- 24. Provide the documentation that supports your statement that the RCP's can run for one week with only sea.1 water providing the cooling for the pump seals.
- 25. 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 / transients are intraduced into the plant design."
Provide the detailed bases that support the above conclusions.
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- 25. Your description of the Ccr.tainment Isolaticn System does not specifically addnss if any of the iso' aced valves will automatically res ar: to their pre-isolated state (i.e., reopen). Address this aspect of your design.
- 27. (Restart Submittal Table 2.1-2) a.
Valves ci-V1, CM-V2, ci-V3 and CM-V4 do not receive a diverse safety grade automatic
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isolation signal. This is unacceptable..%dify your design accordingly.
b.
Valve MU-V3 does not receive a diverse safety grade autcmatic isolaticn signal.
This is unacceptable. Modify your de. sign accordingly.
c.
Valves IC-V2, IC-V3, IC-V4 and IC-V6 do not receive a diverse safety grade autcratic isolation signal unless a safety grace lh break isolation U gnal is included in the design. Sufficient cetaih have not been provided to iscertain if this design feature is incorporated in your design.
If the line break utection feature is incorporated in your design, provide the details supplemencad with sufficient electrical drawings to allow an independent evaluation of whether or not the design meets IEEE Std. 279-1971.
If the line break htection feature is not incor; orated in your design, the design is unacceptable and.nu'st be modified accordingly.
d.
Valves-RB-V2A and RS-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-contailunent 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 evaluatien..
e.
Valves MU-V33A, MU-V333, MU-V33C and MU-V330 receive no automatic safety grade isolation signals. Containment hi-radiation is used to provide an alam for subsequent remote-manual operator acticn. Provide the bases and rationale why dive se 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.
- 23. 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 icwering the containment hi-pressure setpoint to something like 1.0 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.
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- 30. Your response (Section 8.2) to the requirements of item 3 in Bulletin 053 states that you are still investigating the changes to the PORV and high pressure reactor trip set points. Provide your comitment to make these changas or alternatives such that you respond to this bulletin item.
31. Your response (Section 2.1.1.'1) to Bulletin 058, 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 fsedwater, turbine trip, and/or low steam generator level.
- 32. Your response (Section 3.1.1) to Bulletin 05A, items 3 and 4, and Bulletin OSB, item 1, indicates that procedures-have been and are still being revised.
Provide the' necessary procedures for our review and/or sc.kedule 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 informa tion. Provide a safety-grade pump trip design description.
- 34. Provide detailed scenarios of the two reactor trip / turbine trip events (fil, #12) discussed in section 10.3.1.
Include (1) RCS temperature and corresponding pressure vs. time; (2) effects of liquid relief on PORV in Reactor Trip fil; and (3) corrective action for mainsteam safety valves failure to reseat (redesign new valves, etc.7). 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-1 pressure setting would result in later HPI initiation.)
(p.8-16) Should you recc:nnend changing the ESFAS trip to 1600 psig, confirm that the accident and transient analyses for TMI-1 will still show acceptable consequences.
- 36. It is stated that a small break LOCA analysis performed at 2535 MWt with an HPI solit of 54%/36% is acceptable since the generic analysis is perfomed 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 confirm the adequacy of this design change and address the impact of this change on the cross connect modification a
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- 37. Your submit 1 states that cavitating venturis will ce added to the high pressure injection lines to eliminate the need for ocerator action on an HPI line break.
- 38. Subnit modified Technical Specifications to support the crocedural and design modifications required by the Order and Bulletins 05A, 05B and 05C.
- 39. NUREG-0578, Section 2.1.3.b, requires the de.velopment 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 c::erators 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 item.
- 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 nor ral reactor operations, anticipated transients, and effects of multiple equipclent failures and operational errors.
b.
Describe the nomal duties of the STE and his proximity to the control room, c.
Describe the responsibility and authority of the STE during off nomal si+wations.
43.
Provide the schedule for submittal of information required by NUREG-0578, Se: tion 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 Cen*ar.
- Information Request at 10/17 Meeting
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(Order Item 1(d)) Your response to this item incicates that procedures have been or are still being revised. Provide the procedures developed to define operator action during small break LCCA's.
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- 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 6.5.0.1 of the Operator Accelerated Retraining program (0ARP) states that an audit exam will be given to all licensed coerators. provide the cassing criteria for this examination and the retraining /reaudicing for tnase who may not achieve the passing score.
48.
(Order Item 1(e))* Outline how licensed operators will be provided with information related to procedure / design changes that are implemented after the completion of the related training
?.odul e.
49.
(Order Item 1(e))* Describe how the specific program objectives of the OARP (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 provida training in the OARP7 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 ecmpromise core cooling capability, especially natural circulation capability?
- 52. Bulletin 05A Item 4 Hcw does the OARP emphasize the use of various plant parameter indications in evaluating plant conditions?
- 53. Sulletin 05A Item 5 Your responsa 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.
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- Info tion requested at 10/17/79 meeting i
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_10 54 Bulletin 05A Item 5 Ca:: nit to perform cn incepencent valve alignmen-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 OT5G 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 ehgineered safety features.
- 57. Bulletin 05B Item 6 Outline the guidance provided operating personnel for pro =pt notification of the NRC.
- 58. Bulletin 058 Item 7 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 fras or returned to service.
Regulatory Guide 1.47 addresses visual indications for bypated and inoperable status of equip::ent.
Outline the Jesign features or procedural steps that provide visual indication at TMI-1.
51. Bulletin 05B Item 11 Your submittal indicates that the response to this ites 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 autcmated 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 mndify the reactor building spray system. Provide the detailed description and associated drawings related to this modification.
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Si The~ Radiation Protection and Chemistry Supervisor or some other individual kncwledgeable in health physics practices shall be presen; at ? ORC meetings whenever health physics related precedures or policies are reviewed. Revise Section 3.1 of the Restart Sul:mittal to incorporate this practice in the
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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 apolicable standard is ANSI /ANS-3.1-1978. Provide assurance that the minimum cualifications of the Supervisor-Radiation Protection and Chemistry comply witn 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 lesso~ s learned and other additional con-n siderations reflecting problem areas raised by the accident. Areas to be covered are characterized in Appendix C.
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APPENDIX A I N.
Basis for Auxiliary Feedwater System Flcw Requirements
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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 generator (s) to assure adequate ' removal of reactor decay heat are not well defined or documented.
We require that you provide the following AFWS ficw design basis infor-mation as applicable to the design basis' transients and accident conditiens for your plant.
1.
a.
Identify the plant transient and accident conditions considered in establishing AFWS flow requirements, including the folicwing events:
- 1) Loss of Main feed (LMFW)
- 2) LMP4 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
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Describe the piant protection acceptance criteria and correspor. ding technical bases used for each initiating event identified above.
The acceptance criteria should address plant limits such as:
- Maximum RCS pressure (PORY or safety valve actuation)
- Fuel temperature or damage limits (CNS, PCT, maximum fuel centraltemperature)
- 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 justi'ication used with plant condition considered in 1.4 above including:
a.
MaxirJum reactor power (including instrument error allCWance) at the time of the initiating transient or accident.
b.
Time delay frem initiating event to reactor trip.
c.
Plant para =eter(s) which initiates AFWS ficw and time delay between initiating event and introduction of AFWS ficw into steam generator (s).
d.
Minimum steam generator water level when initiating event occurs.
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Initial steam generator water inventory and decletion rate before and after APWS flow com.ences - identify reactor decay beat rate used.
f.
Maximum pressure ;.t which steam is released from steam generator (s) and against which the AF4 pump =ust develop sufficient head, g.
Minimum number of steam generators that must receive AP4 flow; e.g.1 out of 2?, 2 out of 4?
h.
RC flow condition - continued operation of RC pumps or natural circulation.
- i. Maximum AFW iniet temperature
- j. Following a postulated steam or feed lina break, time delay assw.ed to isolate break and direct AP4 flow to intact steam generator (s). AFW pump flow capacity allowance to accennodate the time delay and maintain minimum steam generator water level.
Also identify credit taken for primary system heat removal due to blowdown.
k.
Volume arid 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 non::a1 blowdown following initiating event.
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used for cooldewn and AF4 ficw si:ing.
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Time at hot standby and time to cooldown RCS to RHR system cut in te=erature to size AF4 water source inventory.
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3.
Verify that the AFW pumps in your plant will supply the necessary
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flew to the steam generator (s) as deter nined by items 1 and 2 above considering a single failure.
Identify the margin in si:ing the pu=p ficw to allow for pump recirculation ficw, seal leakage and pump wear.
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Infor atten en the Subeccline "eter Plant Name:
Vendor:
Reference for Infor=ation:
Disclay Informtion 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 Dinlay Qualificatiens (seismic, environmental, IEEE279)
Calculator Type (process computer, dedicated digital or analog cale.)
If process c:mputer is used specify availability.
(% of time)
Single or redundant calculators Selection Logic (hignest T, lowest ortss)
Qualifications (seismic, environmental, IEEE279)
Calculational Technique (Steam Tables, functional fit, ranges)
Inout Temperature (RTD's or T/C's)
Tesperature (number of sensors and locations)
Range of temocrature sensors s
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at1 e)
Qualifications (seismic, envircnmental, IEEE279)
Pressure (specify instr'sent used)
Pressure (number of' sensors and locations)
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Range of Pr :ssure sensors e
Uncertainty *of pressure sensors (PSI at 1 e)
Qualifications (seismic, environmental, IEEE279)
Backuo Cacability Availability of Temp & Press Availability of Steam Tables etc.
Training of operators Procedures
- Uncertainty assessment must address conditions of forced flow and natural circulation.
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C-1 TENTATIVE'0UTLINE OF AREAS TO BE COVERED 1.
Policy -
Implementation Capability
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QA 2.
Management -
Responsibilities Communications Organizaticn - reporting chain Staffing - clerical. succor,t and a;oropriate specialists Qualifications of Staff - kncwledge of syscens, procacures Training - frequency, scoce Procedures 3.
Ccomunications -
Status of plant activities H.P. problem identification Changes in procedure Maintenance planning Changes in RWP work scope 4
Personnel Desi=etry -
Sadges, pocket dosimeters Calculations Sioassay Whole Body counting Review and evaluation 5.
Radiation Centrol -
Access centrol - RWP's Monitoring - Portable - Fixed Centrol procedures - Adm. - ALARA 6.
Centamination Control -
R/A Mat'l cont el - Posting, etc.
Mcnitoring Program
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Internal Dese Control -
Resp. Protection All-Monitoring Equipment Analysis - qual / quant.
8.
Design / Engineering reviews for Radiation Protection 4
9.
Training -
Plant Management H.P. Tech - Records 10.
Instrumentatien -
Capability to monitor for Iodines Selection, mainter.ance, calibration and nu=bers 11.
Internal QA Audits
- 12. Special Problems -
Beta Dosimetry
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