ML19254D387
| ML19254D387 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 09/21/1979 |
| From: | NRC COMMISSION (OCM) |
| To: | NRC COMMISSION (OCM) |
| References | |
| NUDOCS 7910250418 | |
| Download: ML19254D387 (45) | |
Text
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,E W ASHINGTON, D. C. 20555 d-f
%, W Septe,ber 21, 1979 Docket Nos..
50-269, 50-270, 50-287, 50-289, 50-302, 50-312, 50-313, 50-346 FAC:LITIES:
Oconee Nuclear Station, Unit Nos.1, 2, & 3 (0conee)
Three Mile Island Nuclear Station, Unit No.1 (TMI-1)
Crystal River Nuclear Generating Station, Unit No. 3 (CR-3)
Rancho Seco Nuclear Generating Station (Rancho Seco)
Arkansas Nuclear One, Unit No. 1 (ANO-1)
Davis-Besse Nuclear Power Station, Unit No.1 (DB-1)
LICEN5EES:
Duke Power Comoany (DPC0)
Metropolitan Edison Company (Met-Ed)
Florida Pcwer Corporation (FPC)
Sacramento Municipal Utility District (SMUD)
Arkansas Power & Light Company (AP&L)
Toledo Edison Company (TECO)
SUBJECT:
SUMMARY
OF MEETING HELD ON SEPT MBER 13, 1979 WITH THE BABC0CK &
WILC0X (B&W) OW':ERS' GROUP TO DISCUSS ANALYSIS OF DESIGN AND OFF-NORMAL TRANSIENT'i AND ACCIDENTS AND OTHER B&W GENERIC REQUIREMENTS On September 13, 1979, members cf the NRC staff mot with the B&W Owners Group (TMI-2 Follow-up Subcommittea) and representatives of the B&W Company, in Bethesda, Maryland, to discuss the scope and schedule of the Owners' Group's program for complying with the requirements of 3ection 2.1.9 of NUREG-0578, (Analysis of Design and Off-Normal Tri nients).
In addition, several other items were discussed which relate to umpletion of the long-term portion of the Commission Orders of May 1979. is a copy of the meeting agenda.
A list of attendees is provided as Enclosure 2.
BACKGROUND During a meeting held on August 9,1979, between the NRC staff, tne B&W Owners' Group, end the B&W Company, the Owners' Group proposed an outline for providing the analyses, emergency procedure guidelines, and training needed to assure that reactor operators can recognize and respond to conditions of inadequate core cooling as well as otaer transients and accidents.
The summary of that meeting, dated August 24, 1979, provides the program description 6nd event tree methodology upon which the program is based.
The Owners' Group requested this meeting to upcate the staff on the progress made on the program since the August 9 meeting.
Additional agenda items were added to the meeting by the NRC staff.
1209 168 yg1opo 4 I T O
1
. DISCUSSION Due to the length of the discussion on the first agenda item, the remaining items were reordered for the meeting.
They are summarized in the order they were discussed.
Agenda Item 1:
ANALYSIS OF DESIGN AND OFF-NORMAL TRANSIENTS & ACCIDENTS Subsequent to the August 9, 1979 meeting with the Owners' Group, work has progressed steadily on developing operator, guidelines for inadequate core cooling and other transients and accidents, as required by Section 2.1.9 of NUREG-0578.
The Owners' Group presented the NRC staff with an update of the status of the program and requested NRC staff comments.
(1)
Inadeouate Core Cooling (ICC):
The objective of this program is to develop operating guidelines that will allow reactor operators to recognize and respond to conditions of inadequate core cooling under four conditions:
(a) Power Operation - DNB transient, (b) Loss of Reactor Coolant System (RCS) Inventory without RC Pumps Operating, (c) Loss of RCS Inventory with RC Pumps Onerating, and (d) Refueling.
An additional objective of the program is to identify any additional instrumen-tation which may be required to indicate inadequate core cooling under the conditions identified above.
The analysis approach, assumptions used in the analysis, criterion for defining inadequate core cooling, and possible detection methods for each of the four conditions listed above were discussed at the meeting.
Based on the material presented at the meeting, the NRC staff requested that the following additional items be included by the Owners' Group in its ICC program:
(a)
Power Operation - DNB transient: no additional concerns expressed; (b)
LOI without RC Pumps Operating - under detection methods, include the expected response of the excore detectors, (c) LOI with RC Pumps Operating: under detection methods, include the expected response of the RC loop flow instrumentation; and, (d) Refueling:
(i) review operating history in this area (St. Lucie 1 and Millstone 2);
(ii) consider the case of loss of inventory during refueling; (iii) under detection methods, consider the use of incore themocouples (when available) and possibly the response of the self-powered neutron detectors (SPNDs); and, (iv) a general concern was expressed by the staff, that the majority of the indications which would be used to detect ICC in this made are local 'ndications, i.e., not monitored from the control room.
\\
The handouts used during this presentation are included as Enclosure 3 to this summary.
(2) Anticipated Transients and Accidents:
The Owners' Group has developed a program for developing operator guidelines for various anticipated transients and accidents.
The program is entitied
" Abnormal Transient Operational Guidelines" (ATOG).
As discussed above, the purpose of this presentation was to update the staff on the progress made in this area since the August 9, 1979 meeting.
The presentation concentrated on three key elements of the program:
(a) Event Trees, (b) Safety Sequence Diagrams, and (c) Sys' :m Auxiliary Diagrams.
(a) For each transient, an event tree is developed.
The event trees provide a means of systematically determining plant conditions which can evolve during a postulated initiating event.
The event trees illustrate the operational sequence of events following a transient and take into account system malfunctions, component failures, and operator errors.
The event trees may then be used to determine specific sequences which require analysis consideration.
Event trees are useful in identifying the ultimate consequence of single and multiple failures as well as determining final pl?nt status and pointing out possible design deficiencies.
(b)
The building blocks for these event trees are safety sequence diagrams (SSDs). An SSD is a tool used for presenting system information fer each specific plant.
The SSDs are used to describe the plant specific systems, components, and terminology as well as identifying actions required by operators or systems.
The event trees, as well as the SSDs, take into consideration both safety and nonsafety-related equipment.
To help develop this area of the program, the Owners' Group, in conjunction with B&W, has hired a subcontractor (EDS Nuclear) who has specific expertise in the development and use of the safety sequence diagrams.
(c) System auxiliary diagrams (cause wheels) provide input information for determining corrective actions for the operating guidelines.
The cause wheels show the supporting systems essential to the operation of the systems which have a direct input to proper and desired plant response during a transient.
The cause wheels will also identify instrumentation required to verify proper operation of the supporting systems.
In their final form, the cause wheels may be a separate set of guidelines which are referenced. where applicable, in the abnormal transient operational guidelines.
If the ATOGs require that a certain system be placed into operation, and a malfunction of the system occurs, the cause wheels may be used by operators to rapidly identify the cause of the malfunction and determine what corrective actions should be taken to restore the system to an operational status.
1209 070
. The end result of this program will be for B&W to develop plant specific operational guidelines for use by the licensees, who will then develop detailed emergency procedures for a broad spectrum of abnormal transient events.
The only staff concerns that resulted from a review of the program were:
(a)
If various paths on the event trees are terminated and not analy:*ed, justification for such temination should be provided; and, (b)
The AT0Gs are designed to aid operators in recognizing plant conditions, based upon all available instrumentation, and then directing them to take actions which will place the plant in a stable and safe condition.
As such, the guidelines include safety-grade as well as nonsafety-grade instrumentation.
The staff is concerned that erroneous indications from nonsafety-grade instrumentation could lead operators to take improper actions.
The handouts used during this portion of the presentation are included as to this summary, Agenda Item 2:
SCHEDULE FOR SUBMITTING INFORMATION IN SUPPORT OF LONG-TERM REQUIREMENTS By letter from D. F. Ross (NRC) tn all B&W Operating Plants, dated August 21, 1979, each of the B&W operating plant licensees was sent a list of eight generic items, whose resolution is needed in order to satisfy the long-term portion of the Commission Orders of May 1979.
The lis;ing included a status report of action taken on each item, as of that date.
The letter also requested that each licensee submit a schedule for completing the items for which licensee action was still pending.
A copy of this list is included as Enclosure 5 to this sumary.
As requested by the August 21, 1979 letter, each licensee submitted the requested responses.
The schedule for completing several of these items was not satisfactory to the NRC staff.
A discussion of each of the outstanding items took place at this meeting and is summarized below. shows a listing of each generic issue and the schedule requested by the NPC.
In addition, the submittal dates proposed by the licensees are also shown on Enclosure 6.
This listing was used at the meeting as a basis for the discussion.
(1)
Failure Mode Effects Analysis of the ICS:
The FMEA was submitted by B&W on August 17, 1979.
Each licensee endorsed this report as applicable to its facility.
Where appropriate, licensees identified plant specific differences from the generic repart.
A review of this report is currently underway by the B&O Task Force in conjunction with Oak Ridge National Laboratory.
(2) Operator Training and Drilling:
Licensee responses to this item are due September 21, 1979.
1209 071
(3) Vograde of the Anticipatory Reactor Trip to Safety-Grade:
By letter dated September 7,1979, each licensee was requested to review its schedule for installing the safety-grade trip.
If it could not install the safety-grade trip within approximately six months, licensees were requested to provide improvements in the current control-grade trip as an interim measure.
In addition, the letter forwarded a series of questions for which responses are needed prior to NRC approval of the proposed design.
Licensees will provide an expedited installation schedule for NRC review by September 28, 1979.
The licensees requested that they be allcwed to delay the response to improve-ments in the control-grade trip and the response to the r.equest for additional information until after the revised schedule is submitted.
The staff informed the licensees that, provided a much improved schedule for installation was submitted on September 28, 1979, a response to improvements in the control-grade trip would not be required at that time.
In addition, the staff required that a response to item 9 of the request for additional information be submitted on September 28, 1979. A delay in response to the first 8 questions was granted until after the revised installation schedule is received.
(4) AFW Reliability Study:
The schedule for submission of this item is considered satisfactory.
Therefore, no additional discussion was required.
(5) Thermal-Mechanical Reoort:
A full report concerning the thermal-mechanical conditions in the reactor vessel during recovery from small breaks, with an extended loss of all feedwater, was requested to be submitted to the NRC by October 15, 1979.
The licensees stated that this report could nct be completed until December 21, 1979.
No resolution was reached concerning a schedule improvement for this item.
(6)
PORV and Safety Valve Lift Frecuency and Mechanical Reliability:
The staff informed the licensees that additional information concerning the lift frequency of the PORV and Safety Valves would be required by October 15, 1979. A draft request for additional information was presented to the licensees at the meeting.
A copy of this draft request is included as Enclosure 7 to this summary.
Require-ment for determining the mechanical reliability of the PORV and safety valves has been superseded in scope and schedule by Section 2.1.2 of NUREG-0578.
Licensees will be required to comply with the schedule listed in NUREG-0578 for this item.
(7)
Small Break LOCA Analysis:
(a)
Item lA:
A benchmark analysis of sequential auxiliary feedwater flow to the steam generators following a loss of main feedwater was requested by the staff to be submitted by September 30, 1979.
The licensees' schedule for submitting this information is December 1, 1979.
Based on other higher priority items on this list, the staff accepted this submittal date.
(b)
Item 1B:
Justificatiun of relief and safety valve flow models used in the CRAFT-2 c-de will be submitted by the licensees on September 30, 1979, as requested.
1209 072
. (c)
Item 2A:
Justification of the 3 node steam generator model used in the CRAFT-2 analysis for small breaks will be submitted on September 30, 1979, as requested.
(d)
Item 28:
The staff requested that licensees provided, by September 30, 1979, the reactor system response to a stuck open PORV, for the case of a small break which causes the reactor sys em to pressurize to the PORV setpoint.
The licensees comitted to providing a statement that no small break with AFW will pressurize tne system to the PORV setpoint, and a qualitative assessment of this transient, by t~e requested date.
If analysis is required to confirm this assessment, it will be requested by the staff.
(e)
Item 3:
The staff requested that, by September 30, 1979, the licensees address the sources, effects, and operator actions regarding the presence of noncondensible gases within tne reactor coolant system following a small break LOCA.
The licensees stated that complete responses to these concerns may not be provided until October 31, 1979.
A conference call was conducted on September 14, 1979, between the staff and B&W, to resolve this issue.
B&W will provide the staff with as much informa-tion as possible on this subject by Sep,tember 30, 1979.
(f)
I_ tem 4:
By September 30, 1979, the staff requested that a CRAFT-2 simulation of the TMI-2 accident out to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> be performed.
The licensees stated that the simulation out to 100 minutes could be provided by September 30, 1979 and that the full 3-hour simulation could not be provided until July 1980.
A conference call was conducted on September 14, 1979, between B&W and the staff, to resolve this issue.
B&W stated that the code did not give realistic results past 100 minutes.
The staff will request that the first 100 minutes of the simulation be submitted by September 30, 1979.
Justification of why the code cannot predict the plant response out to 3 houFs also needs to be addressed.
(g)
Item 5:
The staff requested that an evaluation of the recent Semiscale small break experiment (S-07-108) be provided to the staff by September 1,1979.
The licensees will supply the requested information by September 30, 1979.
This date is acceptable to the staff.
(h)
Item 6:
By November 15, 1979, the licensees are required to provide pretest calculations for the LOFT small break test.
The licensees stated that the earliest this information could be supplied was January 15, 1980.
This is unacceptable to the staff since the LOFT test will have been completed and the results released by that time, the usefulness of pretest calculations will have been lost.
The staff informed the licensees that this should have number one priority on their schedules.
1209 173
. (8)
LOFW and Other Anticioated Transients:
Requirements under this item have been superseded by Section 2.1.9 of NUREG-0578.
The scope of this program (ICC and AT0G) was discussed under agenda item 1 of this meeting.
The schedule for completing tnese items appears in NUREG-0578.
(a) The schedule for completion of the ICC analysis and guidelines is October 31, 1979. The licensees stated that the analysis anc guidelines for the condition of " loss of inventory with RC pumps not operating" can be completed and submitted by the requested date. The cases for " loss of inventory with RC pumps operating" and the " refueling" modes cannot be submitted until December 14, 1979. The case of " Power Operation - DNB transient" will be submitted with the guidelines developed for the ATOG program.
At the present time, the staff finds this unacceptable.
Further resolution is needed on this issue.
(b) The schedule for completing the ATOG program is outlined in NUREG-0578.
Basically, the analysis and guidelines are required to be completed by January 1,1980 and the emergency procedures and training are to be completed by April 1, 1980.
The licensees have chosen ANO-1 as the lead plant for guideline preparation.
The licensees have proposed to meet with the NRC on October 18, 1979 to discuss progress on the program.
This would be folicwed-up by another meeting on January 8, 1980 in wnich they would discuss the analytical results for ANO-1.
The draft guidelines would be sent to AP&L by February 22, 1980 and draft guidel.ines would be sent to the remaining licensees by May 1, 1980.
No resolution was reached at this meeting concerning this schedule.
The scope of this program is very broad and complex and further consideration on extendh19 the completion dates is warranted.
Based on the detailed discussions at the meeting, coupled with the LOFT pretest calculations being given number one priority, the schedule for completing other items may be impacted.
Each licensee committed to rereviewing the schedule based on the information presented at the meeting and submitting a revised schedule to the NRC by September 21, 1979.
The staff will take further action on these items based on the revised schedule.
Agenda Item 3:
UPGRADE OF ANTICIPATORY REACTOR TRIP FOR LOFW & TURBINE TRIP This item was discussed in detail under agenda item 2 (3).
Therefore, no further discussion is provided under this item.
Agenda Item 4:
NON-LOCA TRANSIENT RESPONSE TO IE BULLETIN 79-05C On September 7, 1979, B&W reported an error in its generic report entitled " Analysis Summary in Support of an Early RC Pump Trip." The error is in Section III of the report, which presents an impact assessment of a RC pump trip for non-LOCA events.
IE Bulletin 79-05C required that each PWR licensee perform and submit a report of LOCA analyses for their plants for a range of small break sizes and a range of time lapses between reactor trip and pump trip.
In addition, each licensee was to develop new guidelines for operator action, for both LOCA and non-LOCA transients, that take into account the impact of RCP trip requirements.
The subject analysis was performed by B&W and presented as a generic report applicable 1209 074
The B&W operating plant licensees incorporated this report in their responses to IE Bulletin 79-05C.
IE Bulletin 79-05C requires that RC pumps be tripped immediately upon a reactor trip and an ESF actuation signal caused by RC system low pressure.
This action is required to protect the core for a certain spectrt.m of small break sizes.
Since certain overcooling transients can cause the same conditions without having a LOCA, licensees were asked to perform an assessment of the RC oump trip for this non-LOCA condition.
In its assessment, B&W presented an analysis of what it considered tne most limiting overcooling event, an unmitigated large steam line break (SLB).
This analysis shows that the excessive cooldown would produce void femation in the RC system hot legs, however, the size of the steam bubble volume and the duration of its presence was small and was not sufficient to affect the ability to cool the core on natural circulation.
The analysis shows a steam bubble volume of about 12 ft.3 in the hot leg attached to the pressurizer surge line and about 5 ft.3 in the other hot leg.
The duration is approximately 3 to 5 minutes.
(The volume of the " candy cane" at the top of the hot leg is 63 ft.3,)
In reviewing this analysis, B&W discovered an error in the conversion of the steam mass to steam volume.
The density used to convert the mass to volume was the average froth (two phase) density vice the actual steam density. When the proper values were used, the same analysis showed the vglume of steam inand about 150 f the hot leg with the pressurizer attached was about 250 ft.
in the other leg.
B&W stated that it then performed some hand calculations using a bubble rise model which showed approximately 400 ft.3 of steam in the hot leg with the pressurizer attached.
B&W was not sure of the steam volume in the other loop.
At that time, B&W stated that it would need more time to review the analysis and refine the model.
B&W committed to call back early the following week to review its results.
On September 12, 1979, another conference call was held between B&W and the staff.
B&W stated it had made refinements to its model and reran the analysis.
The changes included:
(1) modifications to the sensible heat of the S/G tubes, (2) incorporation of a phase separation model and (3) division of the hot leg into 2 nodes.
The results of this analysis showed approximately 400 ft.3 of steam in the hot leg with the pressurizer attached and none in the other hot leg.
The loop with no voiding remained between 40*F to 80*F subcooled.
The amount of steam in the hot leg with the pressurizer atteched was sufficient to retard natural circulation.
However, B&W stated any formation of voids would be temporary and the make-up water (from HPI) would collapse the steam bubble in approximately 9 minutes, allowing natural circulation to commence in that loop.
The loop without the pressurizer attached would not lose natural circulation during the course of the transient.
B&W also stated that it saw no reason to change the guidelines it had developed for the B&W operating
. plant licensees.
The revised analysis would be given to the licensees on Thursday, September 13, 1979, for submittal to the NRC.
1209 375
9-At this meeting, the staff presented an analysis done by Brookhaven National Labora tory.
The analysis was a simulated overfeed transient done with the IRT computer program.
The initial conditions assumed the reactor was at 100%
power,100% reactor coolant flow, and pressurizer pressure of 2158 psig.
At time "0" a turbine trip was initiated.
It was also assumed that the ICS failed in such a way that both steam generators continued to be fed by the main faec-water system.
The analysis showed a very rapid drop in RCS pressure and a void fraction of 28% in the " candy cane" within approximately 100 seconds.
This transient could be more limiting than the DESLB analyzed by B&W.
- However, there were questions concerning the number of single failures that would be necessary to produce this "run away main feedwater" transient.
B&W stated that it would review this analysis and advise the NRC of the results of its.
review in a timely fashion.
CONCLUSIONS:
Agenda Item 1:
ANALYSIS OF DESIGN AND 0FF-NORMAL TRANSIENTS & ACCIDENTS The Owners' Group will consider the staff comments made at the meeting.
Where appropriate, these concerns will be incorporated into the ICC and ATOG programs.
Further resolution is needed on the acceptability of the proposed schedule.
Agenda Item 2:
SCHEDULE FOR SUBMITTING INFORMATION IN SUPPORT OF LONG-TERM REQUIREMENTS There are a rignificant number of items for which no mutually agreeable schedule has been worked ou.
Based.upon discussions at the meeting and a clarification of NRC staff priorities, the licensees comnitted to rereview the entire schedule of out-standing items and resubmit a revised schedule for completing all items by Friday, September 21, 1979.
If this schedule is still not mutually agreeable, resolution will be escalated to upper level management.
Agenda Item 3:
UPGRADE OF ANTICIPATORY REACTOR TRIP FOR LOFW & TURBINE TRIP In accordance with our letter to all B&W operating plant licensees, dated September 7, 1979, licensees will submit an expedited schedule for installation of the safety-grade anticipatory trip by September 28, 1979.
Regarding our request for additional information concerning the design of the safety-grade trip, licensees will respond to question number 9 by September 28, 1979.
The remaining eight questions will be responded to by licensees as soon as the information becomes available.
Aaenda Item 4:
NON-LOCA TRANSIENT RESPONSE TO IE BULLETIN 79-05C Section III (Impact Assessment of a RC Pump Trip on Non-LOCA Events) of the B&W generic report " Analysis Summary in Support of an Early RC Pump Trip," will be 1209 076 revised by B&W and submitted to licensees for review on September 13, 1979.
In addition, B&W has comitted to perform additional analyses to ir.sure that the 12.2 ft.2 steam line break case is the worst case event for the non-LOCA analysis.
$0p'A R. A. Capra, B&W Project Manager Project Management Group Bulletins & Orders Task Force
Enclosures:
1.
Agenda 2.
List of Attendees 3.
B&W ATOG Presentation
- 5. to NRC letter of 8/21/79 6.
Schedule for long-tenn requirements 7.
Draft RAI-PORV & SV Lift Frequency cc w/ enclosures:
See attached 1209 J77
BABC0CK & WILCOX OPERATING PLANTS M,. William O. Parker Jr.
Vice President - Steam Production Duke Power Company P.O. Box 2178
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422 South Church Street Mr. William Cavanaugh, III Vice President, Generation and Construction Arkansas Power & Light Company
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Assistant General Manager and
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-Y Sacramento Municipal Utility District i
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6201 S Street I*
P.O. Box 15830 Sacramento, California 95813
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e Mr. Lowell E. Roe Vice President, Facilities Development Toledo Edison Company -
Edison Plaza 300 Madison Avenue Toledo, Ohio 43652 Mr. W. P. Stewart Manager, Nuclear Operations Florida Power Corporation P.O. Box 14042, Mail Stop C-4 St. Petersburg, Florida 33733 Mr. R. C. Arnold Senior Vice President Metropolitan Edison Company 260 Cherry Hill Road Parsippany, New Jersey 07054 Mr. James H. Taylor Manager, Licensing s
Babcock & Wilcox Company Power Generation Group P.O. Box 1260 Lynchburg, Virginia 24505 1209 078
y Ouke Power Company I
tir. William L. Porter Mr. Robert B. Borsum Duke power Conpany Babccck & Wilcox Post Of fice Box 2176 fluclear Power Generation Division 422 South Church Street Suite 420, 7735 Old Georgetown Road Charlotte, !; orth Carolina 28242 Bethesda, Maryland 20014 J. iiichael !!cGarry, !!I, Esquire Manager, LIS s
DeBevoise & Liberman NUS Corporation l
700 Shorehan Building 2536 Countryside Boulevard 806 15th Street, ff.W.
Cleans 6ter, Florida 33515 Washington, D. C.
20005 Office of Intergovernmental Relations 116 West Jones Street Honorable Janes !1. Phinney Raleigh, North Carolina 27603 County Supervisor of Oconee County Walhalla, South Carolina 29621 t
Director, Technical Assessment j
Division Office of Radiation Programs (AW 45?)
U. S. Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460 U. S. Environnental Protection Agency Region IV Office ATTil:
EIS COORDINATOR 345 Courtland Street, N.E.
Atlanta, Georgia 30308 U. S. Nuclear Regulatory Comission Region II Office of Inspection and Enforcement ATIN:
Mr. Francis Jape P. O. Box 85 Seneca, South Carolina 29678 1209 379
Arkansas Power & Light Company Phillip K. Lyon, Esq.
Director, Technical Assessment House, Holms & Jewell Division 1550 Tower Building Office of Radiation Programs Little Rock, Arkansas 72201 (AW-459)
U. S. Environmental Protection Agency Mr. David C. Trimble Crystal Mall #2 Manager, Licensing Arlington, Virginia 20460 Arkansas Power & Light Company P. O. Box 551 U. S. Environmental Protection Agency Little Rock, Arkansas 72203 Region VI Office ATTN:
EIS COORDINATOR Mr. James P. O'Hanlon 1201 Elm Street General Manager First International Building Arkansas Nuclear One Dallas, Texas 75270 P. O. Box 608
~~ ~~
Russellville, Arkansas 72801 Mr. William Johnson Director, Bureau of Environmental U. S. Nuclear Regulatory Commission Health Services P. O. Box 2090 4815 West Markham Street Russellville, Arkansas 72801 Little Rock, Arkansas 72201 Mr. Robert B. Bors'um Babcock & Wilcox Nuclear Power Generation Division Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Troy B. Conner, Jr., Esq.
Conner, Moore & Corber 1747 Fennsylvania Avenue, N.W.
Washington, D.C.
20006 Honorable Ermil Grant Acting County Judge of Pope County Pope County Courthouse Russellville, Arkansas 72801 1209 080
Florida Power Corporation Mr. S. A. Brandimore Mr. Robert B. Sorsum Vice President and General Counsel Babcock & Wilcox P. O. Box 14042 Nuclear Power Ger.eration Division St. Petersburg, Florida 33733 Suite 420, 7735 Old Georgetc.n Raad Bethesda, Maryland 20014 Mr. Wilbur Langely, Chairman Board of County Commissioners Citrus County Iverness, Florida 36250 Bureau of Intergovernmental Relations U. S. Environmental Protection Agency 660 Apalachee Parkway Region IV Office Tallahassee, Florida 32304 ATTH:
EIS COORDINATOR 345 Courtland Street, N.E.
Atlanta, Georgia 30309 Director. Technical Assessment Division Office of Radiation Programs (AW-459)
U. S. Environmental. Protection Agency Crystal Mall #2 Arlington, Virginia 20460 Mr. J. Shreve The Public Counsel Room a Holland Bldg.
Tallahassee, Florida 32304 Administrator Departnent of Environmental Regulation Power Plant Siting Section State of Florida "ontcomery Building 2562 Executive Center Circle, E.
Tallahassee, Florida 32301 Attorney General Department of Legal Affairs The Capitol Tallehassee, Florida 32304 1209 ]81
l Metropolitan Edison Company G. F. Trowbridge, Esquire Dauphin County Of fice Emergency Shaw, Pi ttman, Potts & Trowbridge Preparedness 1500 M Street, N.W.
Court House, Roon 7 Washington, D. C.
20035 Front & fiarket Streets Harrisburg, Pennsylvania 17101 GPU Service Corporation Richard W. Heward, Project Manager Mr. T. Gary Broughton, Safety and Department of Environmental Resources Licensing Manager ATTN:
Director, Of fice of Radiological 260 Cherry Hill Road Heal th Parsippany, New Jersey 07054 Post Of fice Box 2063 Harrisburg, Pennsylvania 17105 Pennsylvania Electric Company fir. R. W. Conrad Director, Technical Assessment Vice President, Generation Division 1001 Bread Street Office of R. diation Programs Johnstown, Pennsylvania 15907 (AW-459)
U. S. Environmental Protection Agency Mi ss flary V. Southard, Chairman Crystal Mall 72 Citizens for a Safe Environnent Arlington, Virginia 20060 Post Of fice Box OOE Harrisburg, Pennsylvania 17i08 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Dr. Edward O. Swar tz Board of Supervisors Londonderry Township Governor's Office of State Planning RFO=1 - Geyers Church Road and Development itiddletown, Pennsylvania 17057 ATTN:
Coordinator, Pennsylvania State Clearing ~ house U. S. Environmental Protection Agency P. O. Box 1323 Region III Of fice Harrisburg, Pennsylvania 17120 ATTH:
EIS COORDINATOR Mr. J. G. Herbeir Curtis Building (Sixth Floor)
Vice President 6th and Walnut Streets Philadel phia, Pennsylvania 19106 Metropolitan Edison Company P.O. Box 480 Middletown, Pennsylvania 17057 1209 082
'acramento Municipal Utility Page 1 of 2 Di stri ct Christopher Ellison, Esq.
David S. Kaplan, Secretary and Ofan Grueuich, Eso.
California Energy ' Commission General Ccansel 1111 Howe Avenue 5Z01 S Street Sacramento, California 95825 2
O. Box 15830 Sacramento, California 95813 Ms. Eleanor Schwartz California State Office Sacramento County 600 Pennsylvania Avenue, S.E., Rm. 201
?oard of Supervisors Washington, D.C.
20003
- '? 7th Street, Room 424 Sacramento, California 95814 Docketing and Service Section Office of the Secreta.ry U. S. Nuclear Regula ory Commissicn Washington, D.C.
20555 Michael L. Glaser, Esq.
1150 17th Street, N.W.
Director, Technical Assessmen' Washington, D.C.
20036 Divisicn Office of Raciation Programs Dr. Richard F. Cole (AU-459)
Atomic Safety and Licensing Board U. S. Environmental Protection Agency Panel Crystal Mall 72 U. S. Nuclear Regulatory Commission Arlington, Virginia 20460 Washington, D.C.
20555 U. S. Envi ronmental Protection Agency Mr. Frederick J. Shon Region IX Office Atomic Safety and Licensing Board ATTN:
EIS COORDINATOR panel 215 Fremont Street U. 5. Nuclear Regulatory Commission San Francisco, California 94111 Washington, D.C.
20555 Mr. Robert B. Borsum Timothy V. A. Dillon, Esq.
Sabcock & Wilcox Suite 380 Muclear Power Generation Division 1850 K Street, N.W.
cuite a20, 7735 Old Georgetown Road Washington, D.C.
20006 Ecthesda, Maryland 20014 James S. Reed, Esc.
Michael H. Remy, Eso.
Mr. Michael R. Eaton Reed, Samuel & Remy Energy Issues Coordinato.-
717 K Street, Suite 405 Sierra Club Legislative Office Sacramento, Califor,ia 95814 1107 9th Street, Room 1020 Sacramento, Cali fornia 95814 1209 J83
Page 2 of 2 Sacramento Municipal Utility District Atcuic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Atomic Safety and Licensing Appeal Eoard Panel l'. S. Nuclear Regulatory Commission Washington, D.C.
20555 fir. Richard D. Castro 2231 K Street Sacramento, California 95814 Mr. Gary Hursh, Esq.
520 Capital Mall Suite 700 Sacramento, Cal' fornia 95814 California Department of Health ATTN:
Chief, Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 95814 1209 084
Toledo Edison Company Mr. Donald H. Hauser, Esq.
Director, Technical Assessnent The Cleveland Electric Division Illuminating Company Office of Radiation Programs
?. O. Sox 5000 (AW-45g)
Cleveland, Chio 44101 U. S. Environmental Protection Agency Crystal Mall #2 Geralc Charnoff, Esq.
Arlington, Virginia 20450 Shaw, Pittman, Potts and Trowbridae U. S. Environmental Protection Agency 1500 M Street, N.W.
Federal Activities Branch Washing:en, D.C.
20035 Region V Office ATTN:
EIS C00RDINATOR Leslie Fenry, Esq.
230 South Dearborn Street r lle". Seney, He.ry and '::dge Chicago, Illinois 60504 u
3C0 Madison Avenue Toledo, Ohio 43504 Mr. Samuel J. Chilk, Secretary U. S. Nuclear Regulatory Comnission Mr. Rotert B. Borsum Washington, D.C.
20555
- abcocs & Wilcox Nu: lear Power Generation Division The Menorable Tim McCormack Suite 420, 7735 Old Georgetown Road Ohio Senate 5e besca, Maryland 20014 Statehouse Columbus, Ohio 43216 The Honorable Tim McCormack 170 E. 209th Street Euclid, Ohio 44123 Presicent, Board of Cour.ty Commissioners of 0::awa County Per: Clinton, Ohio 43452 Attorrey General Depart.en of Attorney General 30 Eas: Broad 5:reet Columous, Onio 43215 Harold Kahn, Staff Scientist Bruce Churchill, Esq.
Power Siting Commission Shaw, Pittman, Potts & Trowbridge 3
Eas: Broad Street 1800 M Street, N.W.
Col umous, Ohio 43215 Washington, D.C.
20036 Occketino and Service Section Atomic Safety & Licensing Board Panel Cfeice of :ne Secretary U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Wasnine:cn, D.C.
20555 Atomic Safety and Licensing Appeal Panel U. S. Nuclear Regulatory Commission Washington, D.C.
20555 1209 385
Toledo Edison Company Ivan V. Smith, Esq.
Atcmic Safety and Licensing Board Panel U. 5. Nuclear Regulatory Commission Washington, D.C.
20555 Or. Cadet H. Hand, Jr.
Direct:r, Sodega Marine Laboratory University of Califernia P. O. Box 247 Ecdega Bay, California 94923 Dr. Walter H. Jordan SSI W. Guter Drive Cak Ridge, Tenr.essee 37220 Ms. Jean DeJuljak 381 East 272 suclid, Ohio 44117 Mr. Rick Jagcer Industrial Comnission State of Ohio 2323 West 5th Avenua' Columbus, Chio 42216 Ohio Cepartrent of Health ATTN:
Director of Health 050 East T0wn Street Colu-tus, Ohio 43216 1209 086
ENCLOSURE 1 AGENDA FOR SEPTEMBER 13, 1979 MEETING WITH B&W OWNERS' GRCUP TIME SUBJECT LEAD ORGANIZATION 9:0u 7" OPENING lEMARKS NRC/0WNERS' GROUP 9:15 AM ANALYSIS OF DESIGN AND OFF-NORMAL TRANSIENTS & ACCIDENTS O'aNERS ' GRCU P
(
Reference:
Section 2.1.9 of NUREG-0578)
(1)
Inadequate Core Cooling (2)
Transients and Accidents 12:15 PM LUNCH
- 1:0C PM NCN-LOCA TRANSIENT RESPCNSE TO IE BULLETIN 79-05C B&W Discussion concerning error reported by B&W involving the impact assessment of a RCP trip during a large main steam line break accident
(
Reference:
" ANALYSIS
SUMMARY
IN SUPPORT OF AN EARLY RC PUMP TRIP" - August 1979)
- This is a tentative agenda item 2:00 PM UPGRADE OF ANTICIRATORY REACTOR TRIP FOR LOFW & TURBINE OPEN DISCUSSION TRIP
(
Reference:
Letter from R. Reid (NRC) to All B&W Operating Plant Licensees, dated September 7,1979)
(1)
Expedited schedule for installing safety-grade trip (2)
Interim improvements in control-grade trip (3)
Additic sl information required by NRC prior to approving design of safety-grade trip 3:00 PM SCHEDULE FOR SUBMITTING INFORMATION IN SUPPORT OF LONG-NRC TERM REQUIREMENTS
(
Reference:
Letter from D. Ross (NRC) to All B&W Operating Plant Licensees, dated August 21, 1979 and Licensees' responses dated August 31,1979)
Items requiring discussion NRC request date Licensee proposa 5.
Thermal-Mechanical Report 10/15/79 12/21/79 6.
Lift Frequency of PORV & SV requirement to be discussed at meet.
7.
Small Break LOCA 2B. Response to SBLOCA with stuck open PORY 09/30/79 02/01/80
- 38. Effects of NC gases 09/30/79 10/31/79
- 30. Operator actions to mitigate NC gases 09/30/79 10/31/79 6.
LOFT 11/15/79 05/01/80 4:00 PM APPR0XIMATE END OF MEETING 1209 087
ENCLOSURE 2 LIST OF ATTENDEES-9/13/79 ORGANIZATION NAME POSITION Duke Power Company R. L. Gill Oconee Licensing Engineer (Chairman, TMI-2 Effect Subcommittee, B&W 177 FA Owners' Group)
Metropolitan Edison Company None General Public titilities D.
Slear TMI-l Project Engineer Manager Services Corporation Sacramento Municipal Utility S. I. Anderson Nuclear Engineer, Licensing District Arkansas Power & Light Co.
D.
Williams Production Engineer, Nuclear Ops.
W.
Hinten Engineer D. G. Mardis Licensing Engineer W. C. Phillips Manager, I&C Engineering Toledo Edison Company T. J. Meyers Licensing Engineer Bechtel (for TECO)
B.
Novich Florida Power Corporation B.
Simpson Licensing Engineer Consumers Power Corporation D. M. Budzik Nuclear Engineering Babcock & Wilcox Company E. R. Kane Manager, Operating Plant Licensing R. E. Ham Product Line Manager, Eng. Services J. J. Kelley Engineer D.
LaBelle Manager, SafetypAnalysis Babcock-Brown Boveri Reaktor K.
Layer Resident Engineer NRC Staff T. M. Novak Deputy Director, B&0TF C. J. Heltemes Leader, Proj. Mgt. Group, B&OTF S.
Israel Leader, Systems Group, B&OTF Z.
Rosztoczy Leader, Analysis Group, B&OTF G.
Mazetis Sect. Ldr. Systems Group, B&0TF P. E. Norfan Alt. Ldr. Analysis Group, B&OTF F.
Ashe Systems Group, B&OTF G.
Kell ey Systems Group, B&OTF B.
Wilson Systems Group, B&OTF 1209 088
- w. L. Jensen Analysis Group, B&OTF G. M. Holahan Lessons Learned TF (Analysis)
P.
Tam ACRS
ENCLOSURE 2 (page 2)
ORGANIZATION NAME POSITION NRC Staff (cont.)
D. Garner Project Mgr. Rancho Seco (00R)
R. Woodruff Inspection & Enforcement S. Lewis Staff Counsel, OELD R. Capra B&W PM, B&OTF 1209 089
ENCLOSURE #3 ItMDECUATE CCRE C00'.IMG S BACKGROUND 6 PROGP#1 SCOPE O fE MODS ANhLYSISAPPROACH ASSLFPTIONS CRITERION FOR INADECUATE CORE COOLING DETECTION 0 SCIEDULE 1209 090
ENCLOSURE 3 page 2 INADEOUATE CORE COOLING I.
BACKG 0U'!D tiUREG-0578 PARA.2.1.3 DEVELOP PROCEDURES TO RECCGNIZE INADECUATE CORE COOLING WITH EXIST!M It'STP1] MENTATION BASED ON ANALYSES DESCRIEED IN SECTIO:: 2.1.9.
PROVIDE A DESCRIPTION OF ADDITIONAL INSTRUVEhTATIOR HEED B TO DETECT IfMDEQUATE CORE COOLING.
PARA.2.1.9 PROVIDE THE ANALYSIS, El'ERGENCY PROCEDURES AND TRAINING
!!EEDED TO ASSURE WAT THE REACTOR OPEPATCR CAN PECOGNT71 AND RESPOND TO C0f0lTICNS OF INADEQUATE CORE COOLING.
AUGU3T 9 fEETU;G, NRC AND EDV OWNERS GROUP BIE STAFF DESIRES ANALYSES FOR THREE SPECIFIC CONDITIONS:
- 1. LOSS OF RCS IINENTORY WITH RC PUMPS 2.
LO3S OF RCS !!NENTORY WITHOUT RC PUMPS
- 3. TRANSIENTS IN hMICH Di3 OCCURS THESE GUIDELINES SHOU 3 BE DEVELOPED FOR ALL MODES OF OPERATION, I.E.,
POWER OPERATION HOT SHIJILG U REFUELING 1209 J91
ENCLOSURE 3 page 3 INADEC'JATE CORE C00LI!;G II.
PROGPA*4 SCCFF, ENELOP OPE?ATING GUIIELINES THAT WILL ALLOW THE REACTOR A.
TO RECOGNIZE AiO RESPCID TO CCICITIO ;S OF I?ADECMTE CORE C00_!i 3 U;OER TH: FCLLO',lII;G CO:OITICI;S:
- 1. POER OPERATICM - D!3 IRANSIEIJT
- 2. LOSS OF RCS IINENTORY WITHOUT RC P3?S
- 3. LOSS OF RCS ITNEi1 TORY WIT 11 RC PC?S
- 4. REFUELIT,'G PP.0 VIDE RECOWBCATIONS FOR NN ADDITIONAL IIISTRLPETATICN B.
REQUIRED TO INDICATE INADEC'ITE CORE COOLING UiOER THE CCNDITI LISTED AE0VE.
1209 092
ENCLOSURE.3 page 4 lime 0BE COE C00Lif1G II.A.1.
POWER OPERATION - DNB TPANSIENT ANALYSIS APPROACH ALL CO?F)lTIONS IN EE CORE WILL EE ASSUMED NOMINAL N3 THE FOLLO',ll!'G TWO PARAMETERS WILL BE VARIED, ONE AT A TIME, UNTIL DiB OCCURS:
- 1. CORE FLCW
- 2. CORE POWER PEAKING FACTORS ASSLI4PTIONS NON MECHANISTIC REDUCTICN IN CORE FLOW NOT DETECTED LY LOOP FLON MEASURFFENTS NON MECHNIISTIC INCREASE IN RADIAL POWER PEAK - E,0TH SY91ETRIC AND ASSYiO'ETRIC FEAKING INCREASES WILL SE CONSIDERED, CRITERION FOR INADEOUATE CORE COOLING 1.0 DNDR DEIECTION (POSSIBLE MTDODS)
(CORE EXIT T/C - HOT LEG RWS) > 3CF CORE EXIT THERICCOUPLES = C49F (SATURATION TEMP. AT 220 (CORE EXIT T/C - COLD LEG RTD) > 75F POWER DISTRIBUTION ICASURED BY SPND'S EXCEEDS DESIGN VALUES ROD POSITI0tt INDICATION LETDOWN LINE RADIATION MONITOR 1
1209 393
ENCLOSURE 3 page 5 IlMDE00 ATE CORE C00LIl1G II. A. 2 LOSS OF RCS INVENTORY WITHOUT RC PUTPS ANALYSIS APPROACH REDUCE RCS IINENTORY TO iliE POINT WHERE THE CORE BECCVES UNCOVERED.
CALCULATE THE DIFFERENCE BETiEEN STEAM AND CLADDING TEMPERATURES FOR VARIOUS DEGREES OF CORE 'JNCOVERY.
ASSUMPTIONS NON MECHANISTIC REDUCTION IN RCS IfNENTORY DECAY HEAT - 200 SECONDS, 1.2 X ANS CRITERION FOR INADEGUATE CORE COOLING HIGil CLADDING TEMPERATURE DETECTION (POSSIELE METHODS)
HOT LEG RTD'S > SATURATION CORE E.i711EiF0 COUPLES > SATURATION
'1209 094
ENCLOSURE 3 page 6 IIMDECUME CORE COOLING II. A. 3 LO'.o GF RCS I NE. TORY WITH RC PUT?S ANALYSIS APPRCACH REDUCE RCS ITNENTORY TO THE POII1T WHERE THE Cl1JDING AND FLUID TEMPERATURES DIVERGE.
ASSUT?TIONS TON MECHANISTIC REDUCTION IN RCS ilNENTORY DECAY HEAT - 200 SECONDS - 1.2 X ANS CRITERION FC9 It4ADECUATE CORE COOLING HIGHCl>DblNGTEfERATURE DETECTICN (POSSIELE METICDS)
HOT' LEG RTD'S > SATURATION CORE EXIT THERiDCO'JPLES > SATURATION COLD LEG RTD'S 1 SATURATICM LOW RC PUMP CURRENT 1209 395
ENCLOSURE 3 page 7 IllAECBE COE CCOLlilG II. A. L}
REFUELIf!G
- LOSS OF DECAY HEAT RETCVAL SYSTEM N!ALYSIS APPROACH THE tit'E UNTIL THE CORE IS UNCOVERED WILL BE CALCULATED FOR THE CASE WHERE THE LOCPS ARE DRAIN ~D TO THE RV FlJJ;GE AND THE RV HEAD IS REMOVED CUT THE REFUELING cat lAL IS NOT FLOCCED.
ASStMPTIONS DECAY liEAT - 20 HRS. - 1.0 X ANS CRITERION FOR INADECUATE CORE COOLING CORE UNCOVERED DETECTICN RV HEAD ON - HOT LEG RTO TEl'PERATURE INCREASE RV HEAD OFF - RCS LEVEL INDICATION STEAM FORMATION CONTAlilMENT RADIAT10tl MONITORS DHR SYSTEM, TEMPERATURE 1209 096
ENCLOSURE 3 page 8 INADEQUATE CORE C00LIi1G SCHEDULE SEPT.
OCT.
?!OV.
DEC.
JAN.
GUIDELINES FOR LOSS OF RCS IfNEllTORY W/0 RC PUMPS I
GUIDELINES FOR LOSS OF RCS INVENTCRY WITH RC PUMPS ON AflD DURING REFUELING RECOMMENDATIONS FOR ADDITIONAL INSTRUMENTATION 1209 J97
ENCLOSURE 4 s
o EVEfiT TREES PURPOSE Systematically determine various plant conditions which can evolve following a postulated initiating event.
OBJECTIVES 1.
Illustrate operational sequence following:
a.
postulated event b.
system malfunction c.
component failure d.
operator error 2.
Pinpoint specific sequences requiring analysis considering:
1.
probabilities b.
ultimate consequences c.
number of successive failures 3.
Identify consequences of multiple failures 4.
Determir.e final plant status 5.
Detect obvious design deficiencies 1209 098
ENCLOSURE 4 page 2 s
SAFETY SE0VENCE DIAGRAMS PURPOSE To present, in a logical format, system information for each specific plant.
- BJECTIVES 1.
Organize and present raw data 2.
Describe plant a.
systems b.
Components c.
terminology 3.
Identify actions of system / operator during event a.
safety related b.
non-safety related 4.
Highlight plant specific differences 5.
Provide " building blocks" for Event Trees 6.
Detect obvious design deficiencies
~
i209 099
Safety Sequence Diagram (Continued)
ENCLOSURE 4 page 3 I
If1FORMATI0ft SUMMARIZED
- All systems involved in achieving a safety function
- Systen major components
- Component actuation logic
- Setpoints
- Redundancy
- Parameters monitored
- Component f unctional inter - rel ationships
- Plant specific terminology
- Input references
- Operator actions 1209 100
ENCLOSURE 4 page 4 LOSS OF FElISATER f
S t! Af t SAFLTY YALVES REACTIV!TY CONTROL AUXILIARY TLLDWATER
/
.a SECONDARY SYSTDt a
STI'A't 1SOLAT10N GEN!RATO.4 LEYLE
/
i SECONDARY T IC?l
/ PRESSURE /
PRESSURE /
LEVEL LEVEL
\\ SECONDARY SECCNDARY
,PRLS5URE/\\
LEVFL
)
SECONDARY /
PRESSURE /
\\
LEVEL SAFETY SEQUENCE
's PRIMARY DIACRAM PRESSURE /
\\ LEVEL
\\ PRDIARY STABLE
/(
)/x
,,ASLE EVENT TREE DDR iDRE 1209 101
ENCLOSURE 'page 5 SYSTEM AUXILIARY DIAGRAM i
(CAUSE WHEELS)
PURPOSE To provide input information for determinina corrective actions for the operating guidelines OBJECTIVES 1.
Show supporting systems essential to the operation of the system having a direct input to plant' response.
2.
Identify instrumentation required to verify proper operation of the supporting systems I209 1og
System Auxiliary Diagram (Continued)
ENCLOSURE 4 page 6 8
INFORMATION SUMMARIZED
- Supporting systems and interdependence
- Power supplies
- Actuation parameters and in s trumen ta tion
- Valves actuated (including failure position)
- Logic and setpoints Safety qualifications
- Required operator actions
- Verification instrumentation
- Output actions and signals
- References i
i i
e 1209 103
ENCLOSURE 4 page 7 I
I I
i ATCG PROGRAM REVIEW DATES l.
LEAD PLANT EVENT IREES 10/18/79 LEAD PLANT SSD'S LEAD PLANT ANTICIPATED ANALYSES 2.
LEAD PLANT COMPLETED 1/08/80 ANALYSES LEAD PLANT CAUSE WHEELS 3.
LEAD PLANT DRAFT 2/22/80 GUIDELINES 4.
BALANCE OF OPERATING 5/01/80 PLANT DRAFT GUIDELINES I209 104
ENCLOSURE 5
'TEMS RELATED TO THE LONG-TERM PORTION OF COMMISSION ORDERS GENERIC TO ALL B&W OPERATING PLANTS Direct Renuirements of the Commission Orders:
1.
Failure mode and affects analysis of the integrated control system.
B&W has indicated that this report will be avaiiaoie for our. eview by August 20, 1979.
By August 31, 1979, each licensee should endorse this report, or indicate the degree to which it is not applicable.
Fol i.,,'i ng our staff review of this report, any system or prucedural changes necessary will be sent to each licensee.
2.
Continued operator training and drilling.
Each licensee shall document the steps it has taken to insure that continued operator training and drilling incorporates the necessary lessons learned from TMI-2 and assures a continuing high state of preparedness.
This shall be submitted to the NRC by September 21, 1979.
Pending Commission action regarding improvements in the Operator Licensing Program, this requirement may be keyed to an upgrade in the initial training and requalification program by licensees.
3.
Upgrade of. the anticipatory reactor trip to safety-grade.
Each licensee has submitted a preliminary design for implementing a safety-The grade reactor trip upon loss of main feedwater and/or turbine trip.
staff is evaluating these proposals at the present time.
Staff conments will be issued to each licensee by August 31, 1979.
In light of the rect.
failure of the control-grade trip at ANO-1, accelerated installation sch should be developed.
4.
Auxiliary / emergency feedwater system reliability upgrade.
We be.n.a The long-term provisions of the Orders vary on this requirement.
that the most efficient way to fully define the needed improvements " to and perform the AFW/EFW system reliability study discussed in our July August 9,1979 meetings with the Owners' Group.
By August 17, 1970
..e expect a letter from B&W outlining in detail the scope of the study and the schedule for completing the study.
By August 31, 1979, each licensee will submit a letter to the NRC committing to the proposed schedule and study, or provide an alternative.
The study for the lead plant (tent tiv,,)
Rancho Seco) will be available for our review in draft form by Septene 1979.
The studies for the remaining plants will be available in dra
r by October 22, 1979.
The final report will be published by Dece..be
', 9 1209 105
ENCLOSURE 5 page 2 Recuirements Develooed Durino Our Staff Evaluations of Licensees
- Compliance witn the Commission Orders:
5.
A detailed analysis of the thermal-mechanical conditions in the reactor vessel during recovery from small breaks with extended loss of all feedwater.
This issue was identified in the staff evaluations for Rancho Seco, Davis-Besse 1, and Crystal River 3.
However, it is also applicable to Cconee and Ar':ansas Nuclear One 1.
Our request for additional information on this subject was sent to Mr. J. H. Taylor (B&W) frcm Mr. D. F. Ross (NRC) by letter dated July 12.
In a letter from Taylor to Ross dated August 3, 1979, B&W stated:
" Prior to responding to your letter (dated July 12), we feel it is essential to have discussions with our utility customers.
Following this discussion, we will provide you with a schedule." We desire this schedule from the B&W utilities by August 31, 1979.
Note:
It appears to us that the concern is valid for Davis-Besse, but to a lesser degree due to the significantly lower shutoff head of the HPI pumps.
5.
PORV and safety valve lift frequency and mechanical reliability.
This item is discussed in Section 8.4.6 of NUREG-0560 and endorsed in the staff's evaluation for each plant.
This requirement has been superseded in scope and schedule by recommendation 2.1.2 of NUREG-05?8.
Licensees will be directed by letter to take further action on this matter in the near future.
7.
Small Break LOCA Analysis.
This item is discussed in Section 8.4.2 of NUREG-0560 and endorsed in the staff's evaluation for each plant.
Most of this work has been completed
'for the B&W plants.
However, additional information is still required before the staff can issue its evaluation (NUREG-0565
" Staff Report on Generic Evaluation of Small Break Loss-of-Coolant Accident Behavior for Babcock & Wilcox Operating Plants").
Attachment A to this enclosure is a listing of the specific information needed.
We plan on issuing NUREG-0565 in late September 1979.
By August 31, 1979, provide a schedule for the submission of items 1 through 5 of Attachment A such that the information will be received in time to support the publication of NUREG-0565.
8.
Analysis for Loss of Feedwater and Other Anticipated Transients.
This item is discussed in Section 8.4.1 of NUREG-0560 and endorsed in the staff's evaluation of each plant.
Some of this work has been completed; however, the scope and schedule of this requirement has been superseded by recommendation 2.1.9 of NUREG-0578.
In a meeting with the staff on August 9,1979, B&W and the B,&W Owners' Group presented a program by which they intend to satisfy this requirement.
Subject to incorporation of the comments given by the staff at the August 9 meeting and additional comments discussed with B&W by phone (Z. Rosztoczy (NRC) and E. Kane (B&W))
on August 14, 1979, the staff expects the proposed program and schedule for completing this item will be acceptable.
By August 31,1979, each utility should provide a written program outline and schedule for completion of this item.
1209 106
ATTACHMENT A to ENCLOSURE 5 LISTING OF OUTSTANDING ITEMS RELATED TO B&W SMALL BREAX ANALYSIS 1.
Requests made at a meeti rg in Bethesda, April 25, 1979:
A.
Provide a benchmark ar.alysis of sequential auxiliary feedwater flow to the steam generators following a loss of main feedwater.
This ar.alysis was provided in a letter from J. Taylor (B&W) to R. Mattson (NRC) dated June 15, 1979.
However, in this analysis the TRAP-2 c:de with a 5 node stecm generator model was utilized.
All small break analyses presented to the NRC have been performed using the CRAFT-2 code with a 3 noce steam generator model. We require a benchmark analysis for sequential auxiliary feedwater flow also be performed using CRAFT-2 witn a 3 node steam generator representation.
B.
Provide justification of relief and safety valve flow nodels used in the CRAFT-2 code.
2.
Analytical conccens raised by the ACRS (ECCS Subccmmittee) June 19, 1979:
A.
Provide justification that the 3 node steam generator model used in the CRAFT-2 analysis of small breaks is adequate for the prediction
~~
of steam generator /h at transfer.
h Provide the reactor system response to a stuck open PORY for the case of a small break which causes the reactor system to cressurize to the PORV setpoint.
3.
Regarding the presence of noncondensible gases within the reactor coolant system following a small break LOCA:
A.
Provide the sources'of noncondensible gases in the primary system.
Discuss the effect of noncondensible gases on: - m; (1) condensation heat transfer, (2) system pressure calculations and (3) natural circulation flow.
~'
C.
Describe any operator actions and/or emergency procedures necessary to preclude introduction of significant quantities of noncondensible gases into the primary system.
D.
Describe operator actions to be taken in the event of a significant accumulation of noncondensible gases in the primary system.
Provide a CRAFT-2 simulation for the first three hours of the THI-2 accident.
The first 20 minutes of this analysis was provided in the " Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plant" (May 7, 1979).
We require that the analysis be extended for a period of three: hours in order to evaluate the ability of the CRAFT-2 code to evaluate the sequential reactor coolant pump trips and the subsequent The period in which natural circulation was lost in the primary system.
analysis should include at least curves for the following paramaters: pressure, temperature, void fraction, and flow in the reactor coolant loops.
1209 107
ATTACHMENT A page 2 to ENCLOSURE 5 5.
Perform an evaluation of the recent Semiscale small break experiment (S-07-108) witn your small break computer program.
This request was sent to D. Holt (Chairman, B&W Owners' Group Subcommittee on THI-2 Follow'up)'from D. Ross on July 16, 1979.
Copies of this letter were sent to all B&W Licensees.
6.
Pretest calculations of the Loss of Fluid Test (LOFT) small break tests shall be performed as means to verify the analyses performed in support of small break emergency procedures and in support of an eventual icng-term verification of compliance with Appendix K to 10 CFR Part 50.
This item is discussed in recommendation number 2.1.9 of NUREG-0578.
i The items 1-6 in this attachment appear to be the type of information that in the past would be generated by B&W and sent to us.
However, in consideration of the revised working relationship, we require each utility to separately be responsible for supplying the above information.
O 6
9 O
o 1209 108
ENCLOSURE 6 SCHEDULt FOR THE LONG-TERM ITEMS RELATED TO THE COMMISSION ORDERS OF PAY 1979 Revised 9/12/79
[
NRC REQUEST B&W LICENSEES p
1.
itEA-IC5 RECEIVED 08/17/79 08/17/79 INITIAL SUEMITTAL. COMPLETE I REVIEW UNDERWAY WITH ORNL i
2.
OPERATOR TRAIN & DRILL DUE 9/21/79 NA 09/21/79 ! PESPONSES DUE 09/21/79 e
i NA NONE l RESPONSES DUE 09/28/79 3.
UPGRADE OF RX TRIP RAI9/07/79l 4.
AFW RELIABILITY STUDY l
DRAFT-RANCHO SEC0 09/14/79 09/14/79 09/14/79 l SCHEDULE SATISFACTORY CRAFT-ALL OTHERS 10/22/79 10/22/79 10/22/79 FINAL REPORT 12/03/79 12/03/79 12/03/79 5.
THERMAL-MECHANICAL RPT l
ANAL. WCRST CASE BK.
12/04/79 i
FIf;AL REPORT 10/15/79 12/21/79 12/21/79 IMPROVEMENT IN iCHEDULE NEEDFD 5.
Pun'. AND d/
LiiT ~ REQ.
10/15/79 NONE NONE REQUIREMENT; DISr.USSED 09/13/h MECH. I:EL.
PRCG. DESCRIPTION 01/01/80 NONE NONE SCHEDULE IAW flVREG-0578 CCMPLETE TESTING 07/01/81 NONE NONE SCHEDULE IAW NUREG-0578 7.
5"ALL BREAK LOCA lA BENCHMARK SAFW FLOW 09/00/79 12/01/79 12/01/79 SCHEDULE SATISFACTORY TMSTI FIC Ai10N OF PCi<'t 09/00/79 09/30/79 09/30/79 SCHEDULE SATISFACTORY
& SV FLOW MODELS 7X~3: NODE 5/G MODLL t09/00/79 09/30/79 09/30/79 S'CHEDULE SATISFACTORY F RE57045t io 5BLOCA 09/00/79 l12/30/79 SCHEDULE IMPROVEMENT NEEDED WITH STUCK OPEN PORV REQUEST OPTION 3 BY 09/30/79 OPTION 1 09/30/79 OPTION 2 12/30/79 l OPTION 3 02/01/80 lA 50UIICE5 0F NC GA5ES l09/00//9 09/30/79 l 09/30//9 SCHEDULE SATISFACTORY TEFFECTS OF NC GA5r.5 l09/00/79 10/31/79 10/31//9 iSCHEDULE IMPROVEMENT NEEDED
'3C~0A TO PRECLUDE nc i09/00/79 09/30/79 09/30//9 ! SCHEDULE SATISFA ' TORY 30 0A TO MITIGAIE NC 109/00/79 10/31/79 10/31/79 lSCHECULE IMPROVEMENT NEEDED T C A AFT-2 TMI sir 4UL.
3 h0URS 09/00/79 07/00/80 DISCUSSION REQUIRED TO DETERMI1 100 MINS 09/30/79 09/30/79 nr 100 MINS. SIM. IS SATISFACTi
- dE?I5CALE 09/01/79 1 09/30/79 i 09/30//9 iSCHEDULE SATISFACTORY o LCFi t11/15/79 I05/01/80 0b/01/do SCHEDULE IMPROVEMENT NEEDED 3.
LOFW & OTHER ANT. TRANS A.
ICC LGSS OF INV-RCPS OFF 10/31/79 10/31/79 NONE DISCUSSION REQUIRED TO DETERMIin LOSS OF INV-RCPS ON 10/31/79 12/14/79 NONE IF SCHEDULE IS SATISFACTORY C f.B 10/31/79 12/14/79 NONE REFUEL 10/31/79 12/14/79 NONE 1209 109
ENCLOSURE 6 page 2 ITEM NRC REQUEST B&W LICENSEES
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B.
ACCIDENTS & TRANS DUE 9/14/79 DISCUSSION REQUIRED T0 ANAL (SIS & GUIDELINES 01/01/80 DETERMINE IF SCHEDULE EMERGENCY PROCEDURES 04/01/80 IS SATISFACTORY PEET NRC-GEN. ANAL 10/18/79 MEET NRC-DISCUSS RESULT 01/08/80 ORAFT GL TO LEAD UTIL 02/22/80 DRAFT GL TO ALL UTIL 05/01/80
- ALL L7CENSEES RESPONDED TO SCHEDULE IN LETTERS DATED 8/31/79 FOR CERTAIN ITEMS, DB-1 RECUIRES AN ADDITIONAL TWO WEEKS FOR PLANT SPECIFIC REVIEW.
I209 110
ENCLOSURE 7 ORAFT. 9/12/79 DRAFT REGUEST FOR ADDITIONAL INFORMATION ON PORV ACTUATION AND REACTOR TRIP FREQ To help the staff in evaluating the probability of a small break loss-of-coolant acciden, and the impact of the increased number of reactor trips expected for the Babcock & Wilcox operating plants, as a result of the revised setpoints for the high pressure reactor trip and PORY actuation, provide the following information:
1.
Provide a complete listing of PORV openings for each facility wnich occurr3d prior to the revised setpoints for PORV actuation and high pressure-reactor trip.
This listing should include the following items'.
a.
the facility at which each event occurred, b.
the cause of each event, c.
the initial power level prior to che transient which caused the FORV to
- open, d.
indicate which of these transients caused the reactor to trip on hign RCS pressure and/or caused the safety valve (s) to lift, and e.
if the present setpoints for high pressure trip and PORV actuation were in af fect at the time of each of these transients, indicate whether any or all of the fcilowing would have taken place:
(1) reactor trip, (2)
PORV actuation, and (3) lifting of the safety valve (s).
2.
Provide a complete listing of reactor trips for each facility which have occurred subsequent to the revised setpoints for PORV actuation and high pressure reactor trip.
This listing should include the following items:
a, the facility at which event occurred, i209 1ii
ENCLOSURE 7 page 2 c.
the fnitial power level prior to the transient which caused the reactor
- trip, d.
indicate which of these transients caused the PORV and/or safety valve (s) to open, and e.
i f the old (pre-TMI-2) setpoints for high RCS pressure and PORV activation were in affect at the time each of these transients, indicate whether any or all of the following would have taken place:
(1) PORV actuation, (2) reactor trip on high RCS pressure, and (3) lifting of the safety valve (s).
3.
Tae lowering of the high pressure reactor trip setpoint has increased the number of expected reactor trips at each facility.
Discuss how much the frequency of reactor trip has increased based on the lower trip setpoint.
This discussion should include a breakdown of both primary and secondary induced transients.
1209 112 e
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