ML16148A257
| ML16148A257 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco, Crane  |
| Issue date: | 09/21/1979 |
| From: | Capra R NRC - TMI-2 BULLETINS & ORDERS TASK FORCE |
| To: | NRC - TMI-2 BULLETINS & ORDERS TASK FORCE |
| References | |
| NUDOCS 7910040028 | |
| Download: ML16148A257 (11) | |
Text
September 21, 1979 Docket Nos.: 502 O-270, 50-287, 50-
, 50-302, 50-312,.
0w-313, 50-346 FACI LITIES:
Oconee Nuclear Stati on, Unit-Nos. 1, 2, & 3 (Oconee)
Three Mile Island Nuclear Station, Unit-No.
1 (TMI-1.)
Crystal Rivet Nuclear Generating Stati'on9, Unit.No-. 3 (CR-3)
Rancho, Seco Nuclear Generating Station (Rancho..Seco,)
Arkansas Nuclear One, Unit No.
1.(ANO-1)
Davis-EBesseoNucleaPower
- Station, Unit No, 2,1& 3 (DB-1)
-LICENSEES:
Duke Power Colmpany DPCO LMetropolitan Edison Company (Met-Ed)
Florida Power Corporation (FPC)
Sacramentq Municipal Utility District'(SMUD)
Arkansas Power & Light Company (AP&L)
Toledo -Edison Company (TECO)
SUBJECT:
SUMMARY
OF-MEETING HELD ON SEPTEMBER 13, 1979 WITH THE BABCOCK &
WILCOX (B&W) OWNERS' GROUP TO DISCUSS ANALYSIS OF DESIGN AND OFF NORMAL TRANSIENTS AND ACCIDENTS AND OTE'R B&W GENERIC REQUIREMENTS On September 13 1979, members-of the NRC staff met with-the B&W Owners' Group (TMI-2 Follow-up Subcommittee) and representatives of the B&W Company, in Bethesda; Maryland, to discuss the scope and schedule of-.the Ownes' 'Group's program for.complying with the requirements of Section 2.1.9 of.NUREG-0578, (Analysis of Design and Off-Normal Transient). In addition, several other items were discussed which relate to completion of the long-term portion of the Commission Orders of May 1979. Enclosure1 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, the B&W Owners' Group, and 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 other transients and accidents. The summary of that meeting, dated August 24, 1979, provides the program description and event -tree methodology upon which the program'-is based. The Oiners' Grouprequested this meeting to update the staff on the progress made on the program since the August 9,meeting. (,j1L0 Additional agenda items were' added to the meeting by the'-NRC staff.
DATE NBC FORM 318 (9-76) NRCM 0240, u.s. GOVERNMENT PRINTING OFFICE: 197 "-
- 1
-2 DISCUSSION Due to the length of the discussion on the first agenda item, the remaining iten'swere reordered for the meeting. They-are summarized in the order they ere discussed.
Agenda Itei'l: 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 cool i'ng 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) Inadequate Core Cooling (ICC):
The objective o67f this program is to develop operating guidelines that will allow reactor operators to recognize and respond to conditions 6f inadequate core cooling under four conditions:- (a) Power-Operation -
DNB transient, (b) Loss of Reactor Coolant System (RCS) Inventory without RC iumps Operating, (c) Loss of RCS Inventory with RC Rumps Operating, 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 c6ditions identified above., The analysis approach, assumptions used inthe analysis, criterion for-defining inadequatecore 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:
N) f 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 thermocouples (when-available) and'possibly the response of the self-powered neutron detectors (SPNDs); and, (iv) a teneral concer was expresseJ !y the stafi, thiatt. he majority th f f i ft a i o w i fc w odar ese W to d i ec t.C C r
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NI.C FORM 318 (9-76) NRCM 0240 u.s. aoVERndMENT PRINTING OPPICE: 1978 - 265'- 769
-3 The handouts used during this presentation are included as Enclosure 3 to this summary.
,.Ant ci pAed Transients and Accidents:
The Owners' Group has developed a program for developing operator guidelines for various anticipated transients and accidents. The program is entitled "Abnormal Transient Operational Guidelines" (ATOG). As discussed above, the purpose of this presentation was to update the staff on the progress made i-n' 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) System Auxiliary Diagrams.
(a)
,For each transient, pn 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 plant.
status and pointing out possible design deficiencies.
(b) The building blocks for these event trees are safety sequengeediagrams (SSDs). ASSD is a-tool used forpresenting system information for 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, theOwners' Group, in.cond nction with B&W, has hired a subcontractor (EDS Nuclear) who has,specific expertise in the development and use of the safety sequence diagram.5.
.(c) System auxiliarydiagrams (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 op'eration 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
/ guidfelines. 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 anddetermine what corrective actions should be taken to restore the.
system'to an-operational status.
O 1rac FORM 318 (9-76) NRCM 0240
- U.S.
GOVERNmerN PRINTING OFFICE: 19708 265 78
4'.
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. i 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 analyzed, justification for such termination should be provided; and, (b) The ATOGs-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 hat erroneous indications from nonsafety-grade instrumentation coild. lead operators to.take improper actions..,-i The-handouts used during this portion of the presentation are included as Encldsure 4 to this' summary.
Agenda Item 2:
SCHEDULE FOR SUBMITTING INFORMATION IN SUPPORT OF LONG-TERM REQUIREMENTS By letter from D. F. Ross.(NRC). to 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. Theflisting 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 1ist is included as Enclosure 5 to this summary.
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. Enclosure 6 shows a listing of each generic issue and the schedule requested by the NRC. In addition, the submittal dates proposed by the licensees are also 'sthown on Enclosure 6. This 'isting 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, Ticensees identified plant specific differences from the generic report.- 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 ep Septe' b i-21, 7979.
OU NA 13>
................................-........t...................................
DAT E 1aC FORM 318 (9-76) NRCM 0240 U.S. GOVERNMENT PRINTING OFPICE: 1978 - 265 - 769
-5 (3) Upgrade 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 inthe current control-grade trip.as an interim measure. In add itionthe letter forwarded a series of questions f Lrwhich responses are needed prior to NRC approval of theproposed esign.
Licensees will provide,anexpedited installation schedule for NRC review by September 28, 1979.
The licensees requested-that they be allowd to delay the response to ihpue ove ments in theucontl-grade trip andsthe response to the request for additional information until after the revised.schedule is.submitted.
The staff informed theicednseesthat, provided a much improved schedule for installationtwas ubmitted on Septembere28,q1
, a responseto improvemets in the control grade trip would not'be required at thattime.
In addition, thestaff required thatoa responseto ite 9 of the request for additionaluinformation be submitted.
on Se-ptember'28, 1979. A-delay in re-spons-e-to the first.8 'uestions was granted until after.the revised installation schedule is reeeived.
(4) *AFW Reliability Study: The schedule for submission of this'item is considered satisfactory.
Therefore, no additional d~iscussion was required.
(5) Thermal-Mechanical Report: A full report concerning the thermalji1echanical 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 not be completed until December 21, 1979. No resolution was reached concerning a schedule improvement for this item.
(6) PORV and Safety Valve Lift Frequdncy and Mechanical Reliability: The'staff informed..the licens es 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 yalves 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 i tem.
(7) Small Break LOCA Analysis:
(a) Item IA 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:
Justification of relief and safetyvalve flow models used in tie CRAFT-2 coc e will be sub itted by the itensees on
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MAC FORM 318 (9*76)
NRCM 6240 u.s. GOVERNuMENT P eNsrmo OFFCE:
1978 - 265 - 7s
.9
76
- (c) Item 2A:
Justification of the 3 node steam generator mode used in 7
,the CRAFT-2 analysis for small breaks will be submitted on
-September 30, 1979, as requested.
(d) Item 2B.:'
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 system to pressurize to -the PORV setpoint. The licensees committed to providing a statement that-no small break with AFW will pressurize the system.to the PORV setpoint, -and a qualitative assessment,,of this transient,.by the 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,- nd operator actions regarding the presence of noncondensible gases within the 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 September 30, 1979.
(f) Item 4:
By September 30, 1979, the staff requested that a CRAFT-2 simulation 'of the TMI-2 accidend out to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> be perfdned.
.-The licensees stated that the simulati6n out to 100 minutes could be pr yfded by September 30, 1979 and that the full 3-fpur 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 giverealistic~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 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> also needs to be addresse'd.
(g) Item
, The staff-requested that an evaluation of the recent Semiscale.
small break experiment,(S-07-lOB) 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 reiq ed to provide pretest calculations for the LOFT small-break-test.-,The licensees-st&ted that the earliest this information couldlbbe supplied was Janua,y 15, 1980.
This is unacceptable to the af since t 0FT test wi T have been chp-eted and h
-Finc LT fletda&F rsuIts--releas d.by-that-timy -the.usefulness-of..pretes.........
URNAM lculations w 11 have been ost.The staff informedthe.
Ocensees that iths shud~h ve number one priority onW Weir DATE ->........................
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- U.*S.
GOVERNMSENT PRINiTING oppiCsM: 1970 -
6 769
7 (8) LOFW and Other Anticipated Transients:. Requirements under this item have been superseded by Section 2.1.9 of NUREG-0578. The scope of this program.
(ICC :and ATOG) was discussed under agenda item 1 of this meeting., The schedule for completin these 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 and guidelines fdr the condition of "loss of.inventory with RC pumps not operating".can be completed and submitted by the requested dat'e., The cases for "ldss bf 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 dev'loped for the.ATOG program.
At the present time, the staff finds this unacceptable. Furthpr 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, 19 8. The licensees have chosen ANO-1 as the lead plant for guideline preparation. The licensees have proposed th-meet with the NRC-on October 18, 1979.to discuss progress on the program. This would be followed-up by another meeting on January 8, 1980 in which they would discuss the analytical results for ANO-1.
The draft guidelines would be sent to AP&L by-February 22, 1980 and draft guidelines would be sent-to the remaining,licensees by May 1, 1980.. No resolution was reached at this meeting concerningthis schedule. The scope of this program is-very broad and complex and further consideration on extending 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 theNRC by September 21, 1979..2The 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 Th1.'$'itpm.was discussed in detail-under agenda item 2 (3).
Therefore, no further disOussion 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 pres.ents an impact assessment of a-RC pump trip for non-LOCA events.
-T~f 0
epr of A nalyses -for their plants ior a range of smalT break-sizes and a ra ge o time bt C...
m t."
ditio c.iten:
weeft develop..new.-gu d perator..acti.o 1..for- -both4-O A. nd-nn-...........
traAsie tS, that take into account the impact of RCP trip requirements. Th
+
R 31
(-70 NRce 024o
- U.S.
ON MENT PRmy..N OPe.c.: 197..-
The B&W operdating plant licensees incorporated this report in their responses to IE Bulletin 79-05C.
IE Bulletin 79-O5C.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 spectrum 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 pump trip for this non-LOCA condition.
In its-assessment, B&W pesented an analysis of what it considered the most limiting overcooling eveht, an unmitigated large steam line break (SLB). This analysis shows that the, -excessive.cooldown would produce void formation in the RC system hot legs; however-, the size of the steam bubble volume-and the duration of its presence was smal' and- 'was nbt'sufficient'to affect the ability to cool
-the core on natural circulation. The analysis shos a steam bubble volume. of about 12 ft. 3 in'the hot-leg attached to the pressurizer surge lineand about 5 ft.3 in the other hot leg.
The duration is.approximately 3 to 5"miani.wtes.
(The volume of the 'candy cane" at thertop 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 dnisity used to 'convertthe mass to volume was the average froth (two phase) density vice the actual s-team density. When the proper values were used, the same analysis showed the olumefbitsteanin the hot leg with the pressuriier attached was about 250 ft.- and about T50 ft.3 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 wouldi ed 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 theS/G tubes, (2) incorporation of a phase separation rmodel 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 ofsteam-in the hot leg with the pressutrizer attached was sufficient to retard natural circulation. However, B&W stated any formation of voids would be temporary and the mate-up water*.(from HPI-) would collapse the steam "bubble in approxirately 9 minutes, allowing natural circulation to commence in; that loop. The loop without the pressurizer attached would not lose natural cir ulation 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 plhnt :Lcensees..The revised analysis would-bea ven-to thp Iteensees 'on hursday \\eptember 3 -1979, for sbmittal to the NRC.
DARM
.18 (9.76) NRC.0240...........
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040 OvaNMUNT PRINTING OFFICE: 1978 2 65 -769 MW FORM 318
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-9 At this meeting, the staff presented an analysis done by Brookhaven -National Laboratory. The analysis was a simulated overfeed transient'done with the IRT computer program.- The initial-tonditions assumed the reactor was at 100%
power, 100% reactor cod1ant'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 feed 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 :-his transient could be more limiting.than the DESLB analyzedby B&W. 4-owever, there were questions concerning the number of singIe failures that would be necessar ' 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 ir a timely'fashion.
CONCLUSIONS:
Agenda Item 1:
ANALYSIS OF DESIGNAND-OFF-NORMAL TRANSIENTS & ACCIDENTS The Owners' Group will consider the staff comments made at the meeting. Where appropriate, these' conc rns 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 Ther are a significant number,of items for which, no mutually agreeable schedule has been Worked out. Based upon discussions at the meeting -and a clarification -of NRC staff priorities, the licensees committed to rerewitew 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 chedule for installation.of the safety-grade anticipatory trip by September 28, 1979. Regardingo9wr 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.
Agenda Item 4:
NON-LOCA TRANSIENT RESPONSE TO IE BUEEETIN 79-05C Section'I II (Impact Assessment of a RC Pump Trip on Non-LOCAEvents) of the B&W generic report "Analysis Summary-in Support of'an Early RC Pump Trip," will be I....
4.....
M~C FORM 318 (9-76) NECM '0240
- U.S.
GOViERNMEENT PRINTING OFFICE: 1 97 8 26
.7 69
11 10 revised by B&W and submitted to licensees for review on September 13,.1979.
In addition, B&W has committed to perform additional analyses to insure that the 12.2 ft. steam line break case is-the worst. case event for the non-LOCA analysis.
R. A. Capra, B&W Project Manager Project Management Group Bulletins & Orders Task Force
Enclosures:
- 1. Agenda
- 2. List of Attendees
- 4. B&W ATOG Presentation
-5. Enclosure.1 to NRC letter of 8/21/79 6..
Schedule for long-term requirements 7.', Draft RAI-PORV & SV Lift Frequency BF0T/ I B&0 T/F
~~...
DATE 9 u /79"
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0240
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1 DIC FORM 318 (9-76) NRCM 0240.
U.s. GovEERNMENT PRINTING OFFICE: 1978 -'265 - 769
BABCOCK & WILCOX OPERATING PLANTS Mr. William 0. Parker Jr.
Vice President -
Steam Production Duke Power Company P.O. Box 2178 422 South Church Street Charlotte, North Carolina 28242 Mr. William Cavanaugh, III Vice President, Generation and Construction Arkansas Power & Light Company Little Rock, Arkansas 72203 Mr. J. J. Mattimoe Assistant General Manager and Chief Engineer Sacramento Municipal Utility District 6201 S Street P.O. Box15830 Sacramento, California 95813 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 C14 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 Babcock & Wilcox Company Power Generation Group P.O. Box 1260 Lynchburg, Virginia 24505
Duke Power Company Mr. William L. Porter Mr. Robert B. Borsum Duke Power Company Babcock & Wilcox Post office Box 2178 Nuclear Power Generation Division 422 'South Church Street Suite 420, 7735 Old Georgetown Road Charlotte, North Carolina 28242 Bethesda, Maryland 200i4 J. Michael McGarry, III, Esquire Manager, LIS DeBevoise & Liberman NUS Corporation 700 Shoreham Building 2536 Countryside Boulevard 806 15th Street, N.W.
Clearwater, Florida 33515 Washington, D. C. 20005 Office of Intergovernmental Relations 116 West Jones Street Honorable Janes M. Phinney Raleigh, North Carolina 27603 County Supervisor of Oconee County Walhalla, South Caroliha 29621 Director, Technical Assessment Division Office of Radiation Prpgrams (AW-459)
U. S..Environmental Protection Agency Crystal Mall #2 Arlington, Virginia 20460 U. S. Environmental Protection Agency Region IV Office ATTN:
EIS COORDINATOR 345 Courtland Street, N.E.
Atlanta, Georgia 303Q8 U. S. Nuclear Regulatory Commission Region II Office of Inspection and Enforcement ATTN: Mr. Francis Jape P. 0. Box 85 SenEca, South Carolina 29678
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. 0. 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. 0. Box 608 Russellville, Arkansas 72801 Mr. William Johnson Director, Bureau of Environmental U. S. Nuclear Regulatory Commission Health Services P. 0. Box 2090 4815 West Markham Street Russellville, Arkansas 72801 Little Rock, Arkansas 72201 Mr. Robert B. Borsum Babcock & Wilcox Nuclear Power Generation Division Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Troy B. Conner, Jr., Esq.
Conner, Moore & Corber 1747 Pennsylvania Avenue, N.W.
Washington, D.C.
20006 Honorable Ermil Grant Acting County Judge of Pope County Pope County Courthouse Russellville, Arkansas 72801
Florida Power Corporation Mr. S. A. Brandimore Mr. Robert B. Borsum Vice President and General Counsel Babcock & Wilcox P. 0. Box 14042 Nuclear Power Generation Division St. Petersburg, Florida 33733 Suite 420, 7735 Old Georgetown Road Bethesda, 1--aryland 20014 Mr. Wilbur Langely, Chairman Board.of County Commissioners Citrus County Iverness, Florida 36250 Bureau of Intergovernmental Rel at ions U. S. Environmental Protection Agency 660 Apalachee Parkway Region IV Office Tallahassee, Florida 32304 ATTN:
EIS COORDINATOR 345 Courtland Street, N.E.
Atlanta, Georgia 3U308 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 PublAic Counsel Room 4 Holland Bldg.
Tallahassee, Florida 32304 Administrator Department of Environmental Regulation Power 'Plant Siting Section State of Florida Rontgomery Building
.2562 Executive Center Circle, E.
Tallahassee, Florida 32301 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304
Metropolitan Edison Company G. F. Trowbridge, Esquire Dauphin County Office Emergency Shaw, Pittman, Potts & Trowbridge Preparedness 1800 M Street, N.W.
Court House, Room 7 Washington, D. C. 20036 Front & Market 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, Office of Radiological 260 Cherry Hill Road Health Parsippany, New Jersey 07054 Post Office Box 2063 Harrisburg, Pennsylvania 17105 Pennsylvania Electric Company Mr. R. W. Conrad Director, Technical Assessment Vice President, Generation Division 1001 Broad Street Office of Radiation Programs Johnstown, Pennsylvania 15907 (AW-459)
U. S. Environmental Protection Agency Miss Mary V. Southard, Chairman Crystal Mall 02 Citizens for a Safe Environment Arlington, Virginia 20460 Post Office Box 405 Harrisburg, Pennsylvania 17108 Mr. Robert B. Borsu1 Babcock & Wilcox Nuclear Power Generation Divisionrce Suite 420, 7735 Old Georgetown Road Bethesda, Maryland 20014 Dr. Edward 0. Swartz Board of Supervisors Londonderry Township Governor's Office of State Planning RFD Geyers Church Road
- and Development Middletown, Pennsylvania 17057 ATTN:
Coordinator, Pennsylvania State Clearinghouse U. S. Environmental Protection Agency P. 0. Box 1323 Region III Office Harrisburg, Pennsylvania 17120 ATTN:
EIS COORDIN4ATOR NuclearGPowerGeineainDvso Curtis Building (Sixth Floor)M.J.GHeei 6 th and Walnut Streets Vice President Suiolte n
42,775do Georgetny Ra Philadelphia, Pennsylvania 191062 P.O. Box 480 Middletown, Pennsylvania 17057
acramento Mupicipal Utility Page 1 of 2 District Christopher Ellison, Esq.
David S. Kaplan, Secretary and Dian Grueuich, Esq.
Generl ConselCalifornia Energy Commission
.General Counsel 111HweAeu 6201 S Street Ho anue P. 0. Box 15830 Sacramento, California 95813 Sacramento County California State office
-1ar of pervisors 600 Pennsylvania Avenue, S.E., Rm,. 201 B;oard of Supevsr 827 7th Street, Room 424 Washington, D.C.
20003 Sacramento, California 95814 Docketing and Service Section Offi-ce of the Secretary U. S. Nuclear.Regulatory Commission Washington, D.C.
20555 Michael.L. Glaser, Esq.
Director, Technical Assessment 15 h Set
,.W0 Division Office of Radiation Programs Dr. Richard F. Cole (AW-459)
Atomic Safety and Licensing Board U. S. Environmental Protection Agency Panel Crystal Mall #2 Arlingto, 'irgini 2046 U. S. Nuclear Regulatory Commission Arlington, Virginia 2046020555 U. S. Environmental Protection Agency Mr. Frederick J. Shon Region IX Office Atomic Safety and Licensing Board ATTN:
EIS COORDINATOR Panel 215 Fremont Street San Framnict Calfonie911 U. S. Nuclear Regulatory Commission San Francisco, California 9411120555 Mr. Robert B. Borsum Babcock & Wilcox Tith D
Nuclear Power Generation Divisipn Suite 420, 7735 Old Georgetown Road Washington, D.C.
20006
£Ethesda, Maryland 20014 James S. Reed, Esq.
Michael H. Remy, Esq.
Mr. Michael R. Eaton Reed, Samuel & Remy Energy Issues Coordinator 717 K Street, Suite 405 Sierra Club Legislative Office Sacramento, California 95814 110.7 9th Street, Room 1020 Sacramento, California 95814
Page 2 of 2 Sacramento Municipal Utility District Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Atomic Safety and Licerising Appeal Board Panel U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Mr. Richard D. Castro 2231 K Street Sacramento, California 95814 Mr. Gary Hursh, Esq.
520 Capital Mall Suite 700 Sacramento, California 95814 California Department of Health ATTN:
Chief, Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498 Sacramento, California 9.5814
Toledo Edison Company Mr. Donald H. Hauser, Esq.
Director, Technical Assessment The Cleveland Electric Division Illuminating Company Office of Radiation Programs P. 0. Box 5000 (AW-459)
Cleveland, Ohio 44101 U. S. Environmental Protection Agency Crystal Mall 02 Gerald Charnoff, Esq.
Arlington, Virginia 20460 Shaw, Pittman, Potts and Trowbridge U. S. Environmental Protection Agency 1800 M Street, N.W.
Federal Activities Branch Washington, D.C. 20036 Region V Office ATTN:
EIS COORDINATOR Leslie Henry, Esq.
230 South Dearborn Street Fuller, Seney, Henry and Hodge Chicago, Illinois 60604 300 Madison Avenue Toledo, Ohio 43604 Mr. Samuel J. Chilk, Secretary U. S. Nuclear Regulatory Cormmission Mr. Robert B. Borsum Washington, D.C.
20555 Pabcock & Wilcox Nuclear Power Generation Division The Honorable Tim McCormack Suite 420, 7735 Old Georgetown Road Ohio Senate Bethesda, Maryland T0014 Statehouse Columbus, Ohio 43216 fThe Honorable Tim McCormack 170 E. 209th Street Euclid, Ohio 44123 President, Board of County CoUissioners of Ottawa County Port Clinton, Ohio 43452 Attorney General Uepartment of Attorney General 30 East Broad Street Columbus, Ohio 43215 Harold Kahn, Staff Scientist Bruce Churchill, Esq.
Power Sitinq Commission Shaw, Pittman, Potts & Trowbridge 361 East Broad Street.
1800 M Street, N.W.
Columbus, Ohio 43216 Washington, D.C.
20036 Docketina and Service Sectio'n Atomic Safety & Licensing Board Panel Office of the Secretary U. S. Nuclear Regulatory Conmission UrU.
S. Nuclear Regulatoy Commission Washington, D.C.
20555 Washington, D.C.
20555 S
Atomic Safety and Licensing Appeal Panel U. S. Nuclear Regulatory Commission Washinoton, D.C.
20555
Toledo Edison Company Ivan W. Smith, Esq.
Atomic Safety and Licensing Board Panel U. S. Nuclear Regulatory Commission Washington, D.C.
20555 Dr. Cadet H. Hand, Jr.
Director, Bodega Marine Laboratory University of California P. 0. Box 247 Bodega Bay, California 94923 Dr. Walter H. Jordan 881 W. Outer Drive Oak Ridge, Tennessee 37830 Ms. Jean DeJuljak 381 East 272 Euclid, Ohio 44117 Mr. Rick Jagger Industrial Commission State of Ohio 2323 West 5th Avenue Columbus, Ohio 43216 Ohio Department of Health ATTN:
Director of Health 450 East Town Street
- Columbus, Ohio 43216
ENCLOSURE AGENDA FOR SEPTEMBER 13, 1979 MEETING WITH B&W OWNERS' GROUP TIME SUBJECT LEAD ORGANIZATION 9!00 AM OPENING REMARKS NRC/OWNERS' GROUP 9:15 AM ANALYSIS OF DESIGN AND OFF-NORMAL TRANSIENTS & ACCIDENTS OWNERS' GROUP
(
Reference:
Section '2.1.9 of NUREG-0578)
(1)
Inadequate Core Cooling (2) Transients and 'Accidents 12:15 PM LUNCH
- 1:00 PM NON-LOCA TRANSIENT RESPONSE 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 PUMR TRIP" -
August 1979)
- This is a tentative agenda item 2:00 PM UPGRADE OF ANTICIPATORY 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) Additional 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 proposae
.5. Thermal-Mechanical Report 10/15/79 12/21/79
- 7. Small Break LOCA 2B. Response to'SBLOCA with stuck open PORV 09/30/79 02/01/80 3B. Effects of NC gases 09/30/79 10/31/79 3D. Operator actions to mitigate NC gases 09/30/79 10/31/79
- 6. LOFT 11/15/79 05/01/80 4:00 PM APPROXIMATE END OF MEETING
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 Utilities D. Slear TMI-1 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.
Hinton 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 Ltne Manger,Eng., Services J. J. Kelley Engineer D.
LaBelle Manager, Safety Analysis Babcock-Brown Boveri Reaktor K,
Layer Resident Engineer NRC Staff T. M. Novak Deputy Director, B&OTF C. J. Heltemes Leader, Proj. Mgt. Group, B&OTF S.
Israel Leader, Systems Group, B&OTF Z.
Rosztoczy Leader Analysts Group, B&OTF G.
Mazetis Sect. Ldr. Systems Group, B&OTF P. E. Norian Alt. Ldr. Analysis Group, B&OTF F.
Ashe Systems Group, B&OTF G.
Kelley Systems Group, B&OTF
'B. Wilson Systems Group, B&OTF W. L. Jensen Analysis Group, B&OTF G, M. Holahan Lessons Learned TF (Analysis)
P.
Tam ACRS SytmsGouB&T
ENCLOSURE 2 (page 2)
ORGANIZATION NAME POSITION NRC Staff (cont.)
D. Garner Project Mgr. Rancho Seco (DOR)
R. Woodruff Inspection & Enforcement S. Lewis Staff Counsel, OELD R. Capra B&W PM, B&OTF
ENCLOSURE #3 INADEQUATE CORE COOLING 0
BACKGROUND 0 PROGRAM SCOPE 0
METHODS ANALYSIS APPROACH ASSUMPT IONS CRITERION FOR INADEQUATE CORE COOLING DETECTION 0
SCHEDULE
ENCLOSURE 3 page 2 INADEQUATE CORE COOLING B
I3ACKGROUND NUREG-0 578 PARA, 2.1.3 DEVELOP PROCEDURES TO RECOGNIZE INADEQUATE CORE COOLING WITH EXISTINGINSEUMENTATION BASED ON ANALYSES DESCRIBED IN SECTION 2.1,9.
PROVIDE A DESCRIPTION OF ADDITIONAL INSTRUMENTATIO NEEDED TO DETECT INADEQUATE CORE COOLING, PARAI 2.1.9 PROVIDE THE ANALYSIS; EMERGENCY PROCEDURES AND TRAINING NEEDED TO ASSURE THAT THE REACTOR OPERATOR CAN RECOGNIZE AND RESPOND TO CONDITIONS OF INADEQUATE CORE COOLING.
AUGUST 9 MEETING, NRC AND BBW OWNERS GROUP TIE STAFF DESIRES ANALYSES FOR THREE SPECIFIC CONDITIONS:
- 1.
LOSS OF RCS I1NWENTORY WITH RC PUMPS
- 2. LOSS OF RCS INVENTORY WITHOUT RC PUMPS
- 3.
TRANSIENTS IN WHICH DNB OCCURS THESE GUIDELINES SHOULD BE DEVELOPED FOR ALL MODES OF OPERATION; I.E.,
POWER OPERATION HOT SHU1DOWN REFUELING
ENCLOSURE 3 page 3 INADEQJATE CORE COOLING II.
PROGRAM SCOP-.
A. DEVELDEDEEDI INIELINES THAT WILL ALLON THE REACTOR OPERATOR TO RECOGNIZE AND RESPOND TO CONDITIONS OF INADEQUATE CORE COOLING UNDER THE FOLLQOWING CONDITIONS:
- i.
POWER OPERATION - DNB TRANSIENT
- 2.
LOSS OF RCS INVENTDRY WITHOUT RC PUMPS
- 3.
UOSS OF RCS INVENTORY WITH RC PUMPS
- 4.
REFUELING B.
PROVIDE RECOMENDATIONS FOR ANY ADDR I
USA-1E E
NIATL REQUIRED TO INDICATE INADEQUATE CORE COOLING UNDER THE CONDITIONS LISTED ABOVE.
ENCLOSURE 3 page 4 INADLEOUAE CORE COOLING II.A.1, POWER OPERATION - DNB TRANSIENT ANALYSIS APPROACH ALL CONDITIONS IN THE CORE WILL BE ASSUMED NOMINAL AND THE FOLLOWING TWQ PARAMETERS WILL BE VARIED, ONE AT A TIME; UNTIL DNB OCCURS:
- 1.
CQRE FLOW
- 2.
CORE POWER PEAKING FACTORS ASSUMPTIONS NON MECHANISTIC REDUCTION IN CORE FLOW NOT DETECTED BY LOOP FLOW MEASUREENTS NON MECHANISTIC INCREASE IN RADIAL POWER PEAK - BOTH SYMMETRIC AND ASSYMMETRIC PEAKING INCREASES WILL BE CONSTDERED.
CRITERION FOR INADEQUATE CORE COOLING DNBR 1.0 DETECTION (POSSIBLE METHODS)
(CORE EXIT T/C -
HOT LEG RTD's) > 30F CORE EXIT THERMOCOUPLES = 649F (SATURATION TEMP.
AT 2200 PSIA)
(CORE EXIT T/C -
COLD LEG RTD)
> 75F POWER DISTRIBUTION MEASURED BY SPND'S EXCEEDS DESIGN VALUES ROD POSITION INDICATION LETDOWN LINE RADIATION MONITOR
ENCLQSURE 3 page 5 INADQUATE CORE COOLING II.A.2 LOSS OF RCS INVENTORY WITHOUT RC PUMPS ANALYSIS APPROACH REDUCE RCS INVENTORY TO THE POINT WHERE THE CORE BECOMES UNCOVERED.
CALCULATE THE DIFFERENCE BETWEEN STEAM AND CLADDING TEMPERATURES FOR VARIOUS DEGREES OF CORE UJNCOVERY.
ASSUMPTIONS NON MECHANISTIC REDUCTION IN RCS INVENTORY DECAY HEAT - 200 SECONDS, 1.2 X ANS CRITERION FOR INADEQUATE CORE COOLING HIGH CLADDING TEMPERATURE DETECTION (POSSIBLE METHODS)
HOT LEG RTD S > SATURATION CORE EXIT THERMOCOUPLES > SATURATION
ENCLOSURE 3 page 6 IIiADEQUATE COE COOLING II.A.3 LQSS OF RCS INVENTORY WITH RC PUMPS ANALYSIS APPROACH REDUCE RCS INVENTORY TO THE POINT WHERE THE CLA)DING AND FLUID TEMPERATURES DIVERGE.
ASSUMPTIONS NON MECHANISTIC REDUCTION IN RCS INVENTORY DECAY HEAT - 200 SECONDS - 1,2 X ANS CRITERION FOR INADEQUATE CORE COOLING HIGH CLADDING TENPERATURE DETECTION (POSSIBLE METHODS)
HOT LEG RTD S > SATURATION CORE EXIT THERMO)COUPLES SATURATION COLD LEG RTD'S ? SATURATION LOW RC PUMP CURRENT
ENCLOSURE 3 page 7 INADEQUATE CORE COOLING II.A.4 REFUELING
- LOSS OF DECAY HEAT REMOVAL SYSTEM ANALYSIS APPROACH THE TIME UNTIL THE CORE IS UNCOVERED WILL BE CALCULATED FOR THE CASE WHERE THE LOOPS ARE DRAINED TO THE RV FLANGE AND THE RV HEAD IS REMOVED BUT THE REFUELING CANAL IS NOT FLOODED.
ASSUMPT IONS DECAY HEAT - 20 HRS,
- 1.0 X ANS CRITERION FOR INADEQUATE CORE COOLING CORE UNCOVERED DETECTION RV HEAD ON -
HQT LEG RTD TEMPERATURE INCREASE RV HEAD OFF -
RCS LEVEL INDICATION STEAM FORMATION CONTAINMENT RADIATION MONITORS JHR SYSTEM, TEMPERATURE
ENCLOSURE 3 page 8 INADEQUATE CORE COOLING SCHEDULE SEPT.
OCT.
NOV.
- DEC, JAN.
GUIDELINES FOR LOSS OF RCS INVENTORY W/O RC PUMPS GUIDELINES FOR LOSS OF RCS INVENTORY WITH RC PUMPS ON AND DURING REFUELING RECOWMENDATIONS FOR ADDITIONAL INSTRUMENTATION
ENCLOSURE 4 EVENT TREES PURPOSE Systematically determine various pl.ant conditions which can evolve following a postulated initiating event.
OBJECTIVES
- 1. Illustrate operational sequence followinq:
- a. postulated event
- b. system malfunction
- c. component failure
- d. operator error
- 2. Pinpoint specific sequences requiring analysis considering:
- a. probabilities
- b. ultimate consequences
- c. number of successive failures
- 3. Identify consequences of multiple failures
- 4. Determine final plant status
- 5. Detect obvious design deficiencies
ENCLOSURE 4 page 2 SAFETY SEQUENCE DIAGRAMS PURPOSE To present, in a logical format, system information for each specific plant.
OBJECTIVES
- 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
Safety Sequence Diagram (Continued)
ENCLOSURE 4 page 3 INFORMATION SUMMARIZED All systems involved in achieving a safety function System major components
- Component actuation logic Setpoints Redundancy
-Parameters monitored Component functional inter.-relationships Plant specific terminplogy Input references Operator actions
ENCLOSURE 4 page 4 LOSS OF FEEDWATER ST EAM SAFETY VALVES REACTIVITY AUXILIARY CONTROL FEEDWATER SECONDARY STE SYSTEM GENERATOR ISOLATION LEVEL SECONDARY SYSTEM ISOLATION PRESSURE/
PRESSURE/
LEVEL LEVEL.
SECONDARY SECONDARY PRESSURE/
LEVEL SECONDARY PRESSURE/
SAFETY SEQUENCE PRIAR DIAGRAM PRESSURE/
LEVEL PRIMARY STABLEE EVENT TREE
ENCLOSURE 4 page 5 SYSTEM AUXILIARY DIAGRAM (CAUSE WHEELS)
PURPOSE To provide input information for determininq 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
.System Au-xiliary Diagram (Continued)
ENCLOSURE page 6 INFORMATION SUMMARIZED Supporting systems and interdependence Power supplies Actuation parameters and instrumentation
- Valves actuated (including failure position)
Logic and setpoints Safety qualifications Required operator actions
-Verification instrumentation Output actions and signals References
ENCLOSURE 4 page 7 ATOG PROGRAM REVIEW DATES 1,
LEAD PLANT EVENT TREES 10/18/79 LEAD PLANT SSD's LEAD PLANT ANTIQIPATED ANALYSES
- 3. LEAD PLANT DRAFT 2/22/80 GUIDELINES
- 4. BALANCE OF OPERATING 5/01/80 PLANT DRAFT GUIDELINES
ENCLOSURE 5 ITEMS RELATED TO THE LONG-TERM PORTION OF COMMISSION ORDERS GENERIC TO ALL B&W OPERATING PLANTS Direct Requirements of the Commission Orders:
- 1. Failure mode and affects analysis of the integrated control system.
B&W has indicated that this report will be available for our review by August 20, 1979. By August 31, 1979, each licensee should endorse this report, or indicate the degree to which it is not applicable. Following our staff review of this report, any system or procedural 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 grade reactor trip upon loss of main feedwater and/or turbine trip. The staff is evaluating these proposals at the present time. Staff comments will be issued to each licensee by August 31, 1979. In light of the recent failure of the control-grade trip at ANO-1, accelerated installation schedules should be developed.
- 4. Auxiliary/emergency feedwater system reliability upgrade.
The long-term provisions of the Orders vary on this requirement. We believe that the most efficient way to fully define the needed improvements is to perform the AFW/EFW system reliability study discussed in our July 19 and August 9, 1979 meetings with the Owners' Group. By August 17, 1979, we 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 (tentatively Rancho Seco) will be available for our review in draft form by September 17, 1979.
The studies for the remaining plants will be available in draft form by October 22, 1979. The final report will be published by December 3, 1979.
EN!OSURE 5 page 2 Recuirements Developed During Our Staff Evaluations of Licensees' Compliance with 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 River3. However, it is also applicable to Oconee and Arkansas Nuclear One 1. Our request for additional information on this subject was sent to Mr. J. H. Taylor (B&W) from 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 t6 the significantly lower shutoff head of the HPI pumps.
- 6. 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-0578. 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 spccific 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 thrpugh 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.
AT@HMENT A to ENCLOSURE 5 LISTING OF OUTSTANDING ITEMS RELATED TO B&W SMALL BREAK ANALYSIS
- 1. Requests made at a meeti'rg in Bethesda, April 26, 1979:
A. Provide a benchmark analysis of sequential auxiliary feedwater flow to the steam generators following a loss of main feedwater. This analysis 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 code with a 6 node steam generator model was utilized. All small break analyses presented to the NRC have been performed using the CRAFT-2 code with a 3 node steam generator model.
We require a benchmark analysis for sequential auxiliary feedwater flow also be performed using CRAFT-2 with a 3 node steam generator representation.
B. Provide justification of relief and safety valve flow models used in the CRAFT-2 code.
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/Neat transfer.
(
Provide the reactor system response to a stuck open PORV for the case of a small break which causes the reactor system to pressurize 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: -'-/3 (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 systen.
)
Provide a CRAFT-2 simulation forithe first three hours of the TMI-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 period in which natural circulation was lost in the primary system. The analysis should include at least curves for the following paramaters:.pressure, temperature, void fraction, an4 flow in the reactor coolant loops.
A1HMENT A page 2 to ENCLOSURE 5
- 5. Perform an evaluation of the recent Semiscale small break experiment (S-07-10B) with your small break computer program. This request was sent to D. Holt (Chairman, B&W Owners' Group Subcommittee on TMI-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 long 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.
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.
ENCLOSURE 6
-CHEDULE FOR THE LONG-TERM ITEMt RELATED TO THE COMMISSION ORDERS OF MAY 1979 Revised 9/12/79 TEM NRC REQUEST B&W LICENSEES
- 1.
FlEA-ICS RECEIVED 08/17/79 08/17/79 INITIAL SUBMITTAL COMPLETE REVIEW UNDERWAY WITH ORNL
- 2. OPERATOR TRAIN & DRILL DUE 9/21/79 NA 09/21/79 RESPONSES DUE 09/21/79
- 3. UPGRADE OF RX TRIP RAI 9/07/79 NA NONE RESPONSES DUE 09/28/79
- 4. AFW RELIABILITY STUDY DRAFT-RANCHO SECO 09/14779 09/14/79 09/14/79 SCHEDULE SATISFACTORY DRAFT-ALL OTHERS 10/22/79 10/22/79 10/22/79 FINAL REPORT 12/03779 12/03/79 12/03/79
- 5. THERMAL-MECHANICAL RPT ANAL. WORST CASE BK.
12/04/79 FINAL REPORT 10/15/79 12/21/79 12/21/79 IMPROVEMENT IN SCHEDULE NEEDED
LIFT FREQ.
10/15/79 NONE NONE REQUIREMENTS DISCUSSED 09/13/711 MECH.
REL.
PROG. DESCRIPTION 01/01/80 NONE NONE SCHEDULE IAW NUREG-0578 COMPLETE TESTING 07/01781 NONE NONE SCHEDULE IAW NUREG-0578 7.'
SMALL BREAK LOCA 1A BENCHMARK SAFW FLOW 09/00/79 12/01/79 12/01/79 SCHEDULE SATISFACTORY TB JUSIlFICATION OF PURV 09/007/9 09/30/79 09/30//9 SCHEDULE SATISFACTORY
& SV FLOW MODELS 2A 3-NODE S/G MDL 09/00179 0
9 SCHEDULE SATISFACTORY, 2BTRESPONSE TO SBLOCA 09/00/79 12/30/79 SCHEDULE IMPROVEMENT NEEDED WITH STUCK OPEN PORV REQUEST OPTION 3 BY 09/30/79 OPTION 1 09/30/79 OPTION 2 12/30/79 OPTION 3 02/01/80 WA SOURCES OF NC GASES 09/00/79 09/30/79 09/30/7 SCHEDULE SATISFACTORY MBEFECIS OF NC GASES 09/00779 10/31/79 10/3179 SCHEDULE IMPROVEMENT NEEDED JCU TO PRECLUDE-09/0779F
.09/30/79 09/30/79 SCHEDULE SATISFACTORY A TO MITIGATE NC 09/00/79 10/31/79 TU777W SCHEDULE IMPROVEMENT NEEDED 3 HOURS 09/00/79 07/00/80 DISCUSSION REQUIRED TO DETERMIN 100 MINS 09/30/79 09/30/79 OF 100 MINS. SIM. IS SATISFACT CA1SC ALEC N
09/07/9 09/30/79 0/79 SCHEDULE SATISFACTORY 6
0 FT 11/1/79 05/01/80 05/017/80 SCHEDULE IMPROVEMENT NEEDED"
- 8.
LOFW & OTHER ANT. TRANS A. ICC LOSS OF INV-RCPS OFF 10/31/79 10/31/79 NONE DISCUSSION REQUIRED TO DETERMINI LOSS OF INV-RCPS ON 10/31Y79 12/14/79 NONE IF SCHEDULE IS SATISFACTORY DNB 10/31/79 12/14/79 NONE REFUEL 10/31/79 12/14/79 NONE
ENCLOSURE 6 page 2 ITEM NRC REQUEST B&W LICENSEES B. ACCIDENTS & TRANS DUE 9/14/79 DISCUSSION REQUIRED TO ANALYSIS & GUIDELINES 61/01/80 DETERMINE IF SCHEDULE EMERGENCY PROCEDURES 04/01/80 IS SATISFACTORY MEET NRC-GEN. ANAL 10/18/79 MEET NRC-DISCUSS RESUL 01/08/80 DRAFT GL TO LEAD UTIL 02/22/80 DRAFT GL TO ALL UTIL 05/01/80
- ALL LICENSEES RESPONDED TO SCHEDULE IN LETTERS DATED 8/31/79 FOR CERTAIN ITEMS, DB-1 REQUIRES AN ADDITIONAL TWO WEEKS FOR PLANT SPECIFIC REVIEW.
ENCLOSURE 7 DRAFT -
9/12/79 DRAFT REQUEST FOR ADDITIONAL INFORMATION ON PORV ACTUATION AND REACTOR TRIP FREQUENCY:
To help the staff in evaluating the probability of a small break loss-of-coolant accident 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 PORV actuation, provide the following information:
- 1. Provide a complete listing of PORV openings for each facility which occurred 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,
- open,
- d. indicate which of these transients caused the reactor to trip on high 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 affect at the time of each of these transients, indicate whether any or all of the following 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,
- b. the cause of each event,
ENCLOSURE 7 page 2
- c. the initial 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. if 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. The 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.