ML19350E539
| ML19350E539 | |
| Person / Time | |
|---|---|
| Issue date: | 05/12/1981 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-1848, NUDOCS 8106230249 | |
| Download: ML19350E539 (12) | |
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ISSUE DATE:
5/12/81 I
MINUTES OF THE ACRS SUBCOMMITTEE MEETING ON REACTOR OPERATIONS WASHINGTON, DC APRIL 8, 1981 The ACRS Subcommittee on Reactor Operations held a meeting on April 8,1981 in Room 1167, 1717 H St., NW, Washington, DC.
The purpose of the meeting was to continue the Subcommittee's review of Rep. Udall's inquiries on anticipated transients without scram ( ATWS) which were prompted by the June 28, 1980 Browns Ferry 3 partial failure to scram The specific topic for this meeting was the Staff's ability to calculate ATWS consequences.
Notice of this meeting was published in the Federal Register on Friday, March 24, 1981. A copy of this notice is included as Attachment A.
A list of attendees for this meeting is included as Attachment B, and the schedule for the meeting is included as Attachment C.
A complete set of meeting handouts has been included in the ACRS Files. Attachment D is a list of the handouts and documents associated with the meeting.
There were no written statements or requests for time to make oral statements received from members of the public. The meeting was. entirely open to the public. TheDes,thn Federal Employee for the meeting was Richard Major.
f He noted the object 19 DISCUSSION WITH NRC STAFF 5
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Dr. Speis introduced the day's presentations.
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the meeting was to address the Staff's ability to calculate the cons g
quences of an ATWS. He noted that the Staff's assessment of the vendor's calculational methods is an aspect of this issue.
He also explained that he wanted to give the Subcommittee an overview of code development work to be done in the future.
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l 3r. Speis explained there is no regulatory requirement at this time neces-sitating the ability to calculate the consequences of ATWS events. Currently
. the only ATWS related requirements for applicants and ifcensees are admin-
' istrative with the exception of the requirement for a reactor pump trip in boiling water reactors (BWRs). The administrative measures include the development of emergency procedures and training operators to more readily recognize an ATWS event and take appropriate actions to bring the plant to a safe shutdown.
-The Staff has requested Commission approval to publish a proposed ATWS rule for comment. 'The proposed rule would require that licensees demonstrate by using an acceptable evaluation model that the calculated consequences of a i
postulated ATWS would be within the acceptance criteria specified by the rule.
Thus the rule would require licensees to have the capability to calculate the consequences of ATWS events.
The Staff would review models, but would not calculate ATWS consequences except to the extent necessary to audit the licensees' cal cul ations.
Dr. Speis stated that a calculation of ATWS consequences is not the basis for the proposed rule. The rule could have been written without any prior calcu-lation of ATWS consequences. However, both the Staff's and Industry's ATWS calculations have been helpful in establishing some requirements and calcu-lations do show the possible magnitude of the consequences of ATWS events and tha effect of variations in parameters on equipment failures. Prior ATWS calculations increased the Staff's understanding of ATWS events and aided in developing a proposed rule. A summary of the Staff's review of the industry's evaluations is presented in Volume 4 of NUREG-0460.
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. Dr. Speis noted the Staff's presentations would focus on the capability of BWR ATWS codes. He explained that he would not defend the adequacy of the models, but rather explain based on the Staff's current best knowledge the models capabilities. To date, evaluations have focused on the pressure rise in ATWS events, and the codes are most developed in this regard.
- However, long-term cooling must also be demonstrated, systems analysis to make this demonstration are less developed.
Dr. Odar of NRC-RES made a presentation on the Staff's assessment of the REDY and ODYN codes used by General Elactric (GE) to predict the early part of the transient response of a BWR to an ATWS event.
Although ODYN is a newer code, GE is currently using REDY for all ATWS analyses. The ODYN code' incorporates changes in both neutronic and thermal hydraulic modeling differing from the REDY code. The Staf f is reciewing the differences between the codes. The Staf f did note that ODYN has some limitations in long-term (cVer one minute) calculational capability.
Dr. Odar described the validation of specific models using plant startup 5
I tests.
Among the tests performed were recirculation pump trip-thermal margin tests and a turbine trip with bypass. He noted the test performed at the Peach Bottom Plant, which were turbine trip tests, showed that modeling of t
l pressure propagation in the steamline was necessary. REDY code calculations of the Peach Bottom tests had inaccuracies for both pressure rise and neutron flux rise. This was the reason General Electric produced the ODYN code.
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The ODYN code is able to model steamline dynamics, REDY could not do this acccurately. 0DYN also improves on the REDY code by modeling reactor core thermal hydraulics and neutronics more accurately. ODYN calculations were in good agreement with the Peach Bottom tests.
General Electric has performed full ATWS analyses using the REDY code.
Calculations have been presented in a proprietary report NEDE-24222. The transients analyzed, which are the most severe from the point of suppression pool temeratures during an ATWS were main steam isolation valves (MSIV) closure and the inadvertent opening of a relief valve.
A tur'bine trip with bypass is con'sidered the most severe ATWS transient from the aspect of a challenge to coolable core geometry.
The Staff has reviewed the GE analyses in NEDE-24222 and expressed concerns which are documented in NUREG-0460, Volume 4.
For the short-term transient behavior, the Staff requested further verification of REDY results using the ODYN code. For the long-term transient behavior, the Staff will require an assessment of: (1) an increased accuracy in the void sweep model in calcu-lating limit cycle oscillations (neutron flux / fluid flow oscillations which
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may have effects on fuel rod mechanical loads. These limit cycle osciliations are predicted to occur later in an ATWS event by the REDY code after the ij initial flux and pressure rise can be postulated to have created pellet /
cladding interactions which could have left the fuel in a weakened state.
i The oscillations in flux and flow that occur later in the event could produce some mechanical loading effects raising a concern over main +aining a coolable (rod like) geometry.)
(2) accuracy of calculations during natural circulation
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should be assessed; (3) non-equilibrium effects of HPCI should be assessed; and (4) the accuracy vf calculations for fuel clad temperatures during limit cycle oscillations should be assessed.
General Electric has stated that limit cycle oscillations can be avoided
-if standby liquid control system (SLCS) boron concentration is increased by a factcr of_ two, the pressure regulator set point is changed, and the MSIV are closed. The Staff requires elimination of these severe limit cycle oscillations to maintain coolable geometry and verification by models using separate effects tests as stated in Volume 4 of NUREG-0460.
A Staff review has indicated that'0DYN is acceptable for certain overpres-surization transients defined in Chapter 15 of the Safety Analysis Reports.
The Staff has not reviewed ODYN for ATWS analyses, however, the Staff believes it can be used for short term behavior where overpressurization occurs.
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Dr. Shier of Brookhaven National Laboratory (BNL) discussed models used to analyze BWR ATWS and half ATWS (which models one half of the core with rods in and the other half with rods out) calculations. The current codes used for BWR ATWS calculations are RELAP-3B a homogeneous equilibrium, thermal-hydraulic model, coupled to the BNL-TWIGL which is a two-dimensional cylindrical neutronics code.
- Validations of the RELAP-3B and BNL-TWIGL code have been based on the Peach Bottom tests. The results between the models and the actual tests show reasonable agreement.
. Dr. Shier identified the sourse of input parameters for both full and half ATWS calculations. For the full ATWS calculation the geometrical model (the Peach Bottom Plant) was supplied by GE.
The neutronics were derived from TWIGL for the initial 60 seconds of the transient. For the remainder of the transient, where space-time effects are less important, a point kinetics model is used. The thermal-hydraulics model uses 2 core channels with 10 nodes each and a bypass channel. The half-ATWS model used the same geometry as the full-ATWS modes. The half-ATWS neutronics are based on a point kinetics model and uses a GE power calculation as input into the system model., The thermal-hydraulics model uses 3 channels with 2 nodes each.
Mr. Shier summarized the Brookhaven work as being able to obtain suppression pool temperature and provide a system response to an overpressure transient.
The BNL work was characterized as providing an independent audit of the GE ATWS calculations.
Although BNL uses some input from GE, the use of different codes and preparation of additional input data separate the two calculations.
BNL has made a first attempt to.model a half-ATWS event.
Areas for improve-ment were identified including more accurate treatment of space kinetics, more realistic phase separation model, and improved level calculation.
Dr. Graves discussed results of GE's calculations performed at the request of the Staff for a half-ATWS.
He pointed out key assumptions used in both the GE and BNL half-ATWS calculations and compared the results of both groups. The Staff requested GE to perfonn four sets of ATWS calculations.
One set assumes a MSIV closure event from rated power with a partial scram
similar to that which occurred at Browns Ferry 3.
The calculation assumes manual action to start the standby liquid control system (SLCS) within 10 minutes and manual action of the RHR system to begin pool cooling also within 10 minutes. Another set included the above sequence except SLCS is delayed 30 minutes. Two other calculations were made with the above parameters but with a partial scram that assumes half the core control rods fully inserted and half remain fully withdrawn. These calculations were perfonned using BWR-4 parameters and the REDY program together with GEs three dimensional BWR core simulator.
GE's calculations were compared to conditions observed at BF-3.
The model predicted 3% fission power after the partial scram versus an observed 1% fission power at BF-3.
(The Staff cautions that the BF-3 measurement is based on an operator's memory of a glance at the power meters.) Other s
cases considered included E0C conditions with a BF-3 type scram and natural j
circulation, and a case in which one half of the cores rods were fully out I
and the remaining rods fully inserted. These cases gave results 6f steady state powers of 10-20 percent respectively. For a full AhlS event, in which no rods scrammed and saturated coolant with no voids 's assumed, the amount of boron required to maintain the core just at criticality with zero fission power was 480 ppm.
Dr. Graves reported the results of GE's calculation for a half ATWS given MSIV closure from full power. The partial scram model assumes, the rods in one half of the core are fully inserted and the remaining rods are fully withdrawn.
Pool cooling is assumed to begin at 10 minutes into the transient. GE stated
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that initiation of.SLCS within about' 5 minutes is needed to keep the maximum pool temperature below 200 F. Dr. Graves compared the BNL calculation of 0
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pool temperature vs. time to that calculation done by GE. Maximum pool temperatures were about 10 F apart.
The Brookhaven model resulted in 0
lower pool temperatures which was probably due to a higher assumed mixing 0
ef ficiency. Peak pool temperatures were approximately 150 F.
Dr. Odar noted three new codes are being developed which will give the Staff important models for best estimate audit calculations of BWR ATWS calculational capability. These codes are TRAC, RELAP-5 and RAMONA.
i By the end of'FY-81 some calculation of a half-ATWS will be performed using RAMONA.
Additional experimental work to validate these newer codes will have to be performed in the future.
No future meetings are planned.
As a result of this meeting and the March 11, 1981 Reactor Operation meeting, a draft version of a reply to Congressman tidall l
will be produced to focus the Committee's deliberation on this subject. An initial session will be scheduled for the May 1981 ACRS meeting.
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NOTE: for additional details, a complete transcript of the meeting is avail-able in the NRC Public Document Room,1717 H St., NW, Washington, DC
'l 20555 or from Alderson Reporters, 300 7th St., SW, Washington, DC, i,
(202) 554-2345.
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MEETING R00h:
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i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS MEETING ON REACTOR OPERATIONS
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WASHINGTON, DC APRTL 8, 1981 ATTENDEES PLEASE SICN BELO'a'
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AFFILI ATION BADCE NO.
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a Attendees:
T. Speis, NRC/RSB W. M. Mathis, ACRS Member M. Hodges, NRC/RSB J. Ebersole W. Minners, NRC/ DST W. Kerr D. Diamond, BNL J. Ray M. Lu, BNL D. Ward S. Fabic, NRC/RES
- 1. Catton, ACRS Consultant W. Lyon, NRC/RES W. Lipinski "
F. Odar, NRC/RES R. Major, ACRS Staff D. Basdekas, NRC/RES F. Aguilar, EG8G Idaho C. Graves, NRC/RSB
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-R;vis;d 4/7/81 T*
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' TENTATIVE SCHEDULE FOR THE APRIL 8, 1981 ACP,5 SUBCOMMITTEE MEETING ON REACTOR OPERATIONS
.R00M 1167, 1717 H ST., NW, WASHINGTON, DC APPR0XIMATE TIME 1:00 p.m.
I.
Chairman's Opening Statement-a.
Discussion of Schedule b.
Meeting Goals 1:15 p.m.-
_II.
Introduction /0verview (T. Speis)
(15 mins.)
(Basis for Staff's ATWS Position in SECY-80-409; Future ATWS Evaluation Models to Verify Compli-ance with ATWS Requirements) 1:30 p.m.
III.
General -Description and Capability of Presently used Codes for ATWS Evaluations (Staff /BNL)
- RELAP-3B (W. Sheir, BNL)
(1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
- Models
- Validation
- Input Parameters
- Calculations / Applications
- Full ATWS
- Half ATWS
' REDY/0DYN (F. Odar, Staff)
(1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
- Models
- Validation
- Input Parameters
- Calculations / Applications 3:30 p.m.
BREAK 3:40 p.m.
IV.
Comparison of GE's and Staff Calculations for Partial ATWS Events (C. Graves, Staff) (30 mins.)
Short Term Plant Behavior Long Term Plant Behavior 4:10 p.m.
V.
Future Code Developments for ATWS Evaluations (Staff /BNL/INEL)
(75 mins.)
- Overview (F. Odar, Staff)
- RAMONA (D. Diamond, BNL)
- TRAC (F. Aguillar/INEL)
RELAP 5 (F. Aguillar/INEL) 5:25 p.m.
VI. Open Executive Session / Chairman's Closing Remarks FT1KHNENT C
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Reactor' Operations 4/8/81
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' LIST OF DOCLNENTS PROVIDED AT MEETING 1.
Tentative Meeting Schedule 2.
Slides used by F. Odar. NRC - Shoret Review of General Electric Full ATWS Analyses 3.
Slides used by W. Shier, BNL, - Current BNL BWR ATWS Capabilities -
Advisory Committee cn Reactor Safeguards Presentation - W. Shier, Department 'of Nuclear Energy, Brookhaven National Laberatory 4.
Slides used by C. Graves, NRC -~ 0verview GE Calculations on Partial
- ATWS, BNL Calculation on-Partial ATWS, and Comparison of GE and BNL Results 5.
Slides used by F. Odar, NRC - RES Plans for Development of System Transient Codes for ATWS Analyses ATTUSW l
_A 18421 24, 1981 / N tices Federal Register / V:1. 46, N2. 56 / Tuesd;y, Mitch a
,j obtained by a prepaid tiph:n2 call to Company.Philad:!phia Electric 3
2 Company.Delmarva Power and ught
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i Dated. March 17. test.
the cognizant Designated Federal
' Company and Atlantic City Electic Sn C Hqle.
Employee Mr. Richard K1.Ma or Company (the licensees) which revised I
II Advisory Committee Management Qfice (telephone 202/et4-1414]Tetween 8:
Technical Specifications for operation of 15 a.m. and 5.00 p.m EST.
the Salem Nuclear Generating Station.
lhw,,,o ca sarws zwso =l sus l have determined,in accordance with Unit No.1 (the facility) located in Salem j
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Subsection 10(d)of the Federal County, New Jersey.The amendment is I
%dvisory Committee on Reactor j
Safeguards, Subcommittee on Reactor Advisory Committee Act, that it may be effective as of the date ofissuance.
necessary to close some portions of this The amendment revises the o
Operations meeting to protect proprietary and Radiological Technical Specifications to The ACRS Subcommittee on Reactor Industrial Security inforraation. The incorporate new requirements related to Operations will hold a meeting on au'hority for such closure is Exemption reactor decay heat removal capability.
Wednesday. April 8,1981 in Room 1167 (4) to the Sunshine Act. 5 U.S.C.
The application for the amendment et 1717 H Street. N.W., Wash!ngton. DC.
The Subcommittee will review 552b[c)(4).
comp!ies with the standards and Congressman Udall's inquiries on Dated. March 17.19st.
requirements of the Atomic Energy Act t
' ATWS which were prompted by the g gg*
of1954. as amended (the Act). and the Commission's rules and regulations.The j
June 28.1980 Browns Ferry 3 partial
- N#""#8' "T*""
l Commission has made appropriate I
1" "'" * " * " "I 4 findings as required by the Act and the in accordance with the procedures Commission's rules and regulations in to
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' outlined in the Federal Register on
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October 7.1980. (45 FR 66535). oral orAdvisory Committee on Reactor CFR Chapter 1. which are set forth in the l written statements may be presented b Safeguards, Subcommittee on NRC license amendment. Prior public notice members of the public, recordings will Safety Research Program; Meeting of this amendment was not requ! red since the amendment does not insolve a I
h be permitted only during those portions The ACRS Subcommittee on the NRC significant hazards consideration.
p of the meeting when a tra iscript is being ; Safety Research Program will hold a The Commission has determined that f
4 kept, and questions may be asked only meeting on April 8,1981 in Room 1046.
the issuance of this amendment will not
{ by members of the Subcommittee,its 3 consultants, and Staff. Persons desiring 1717 H Street. N.W., Washington. DC.
result in any significant environmental
,Y The Subcommittee will be considering impact and that pursuant to to CFR
, to make oral statements should notify p the Designated Federal Employee as far [r predecisional budget informa i 51.5(d)(4) an environmental impact
."8 statement or negative declaration and in advance as practicable so that
. Research Program Plan as requested by the Nuclear Regulatory Commission. In environmentalimpact appraisal need k appropriate arrangements can be made not be prepared in connection with f to allow the necessary time during the B
i l order to perform this review, the ACRS must be able to engage in frank issuance of this amendment.
meeting for such statements.
0 For further details with respect to this j The entire meeting will be open to discussion with members of the NRC action, see (1) the application for public attendance except for thosesessions during which the Subcommittee. Staff.F amendment dated October 15,1980 (2) f Amendment No. 34 to Ucense No. DpR-finds it necessary to discuss proprietary end Industrial Security Information. One,, in public session.I have determined, therefore, that it is 70, and (3) the Commission's related Safety Evaluation. All of these items are I
g or more closed sessions may be necessary to discuss such informaUon.
necessary to close this meeting to available for public inspection at the y extent practicable,these closed sessions l prevent frustration of th f(Sunshine Act Exemption 4).To the ACRS' statutory responsibilities. He Commission's Public Document Room.
3 a
' authority for such closure is Exemption 1737 H Street. N.W., Washington, D.C.
and at the Salem Free Public Ubrary,
- will be held so as to minimize 6 inconvenience to members of the public., 9(D) to the Government in the Sunshine 112 West Broadw sy. Salem, New Jersey.
( Act(522b[c)(9)(B)).
Further information can be obtained A copy ofitems (2) and (3) may be
{ in attendance, a ne agenda for subject meetina shall, by a prepaid telephone call to theobtained upon request addressed to the J.
3 jr be as follows:
Designated Federal Employee, Mr. Sam U.S. Nuclear Regulatory Commission.
- - [ telephone 202/634-3267)
Washington, D.C. 20555, Attention:
Wednesdoy. April 8,1931-IS0p.m.
fDurahmween 8:15 a.m. and 5:00 p.m est.
Director, Division of Ucensing.
,. until the conclusion of business.
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During the initial portion of the Dated at Bethesda, Maryland, this 6th day meeting. the Subcommittee, along with
, Date. htarch M.1981.
of March,19et.
' any of its consultants who may be John C. f toyle, For the Nuclear Regulatory Commission.
. present, will exchange preliminary Advisory Comminee Aravagement Officer.
Steven A.Varga, views regarding tritters to be tra oo-sa:4 nw s-a-c ass..;
Chief. Operating Reactors Bmnch & J.
y considered during the balance of the
- Division of Ucensing.
supeo cooe neo.ewar p.am_a.:stn.:s.am.aes i meeting.
s He Subcommittee will then hear
- r (DocketNo.50-2721 suneo coot neo-es-as D presentations by and hold discussions with representatives of the NRC Staff, Publ*c Service Electric'A Gas Co. et at.;
e j
their consultants, and other interested
. Issuance of Amendment to Facility Wckat No. H5) l
) Opeamg hse
%e U.S. Nuclear Regulatory Wisconsin Pubbe Service Corp. et af.;
ID ur her info tion reg d n topics
~ to be discussed, whether the meeting
, Commission [t te Commission) has Notice of Issuance of Amendment to has been cancelled or rescheduled. the. issued Amendment No.34 to Facility Facility Operating Ucense ne Chairman's ruling on requests for the g OperatingUcense No.DPR-70.lasued is ne U.S. Nuclear Regulatory opportunity to present oral statements f Public Service Electric and Gas and the time allotted therefor can be ATTACHMENT A
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