ML20057A010
| ML20057A010 | |
| Person / Time | |
|---|---|
| Issue date: | 07/13/1993 |
| From: | Marcus G Office of Nuclear Reactor Regulation |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9309100278 | |
| Download: ML20057A010 (24) | |
Text
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July 13, 1993 l
, MEMORANDUM FOR: Document Control Desk
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Document Management Branch j
Division of Information Support Servicee Office of Information Resources Management j
FROM:
Gail H. Marcus, Chief Generic Communications Branch Division of Operatir.g Reactor Support t
Office of Nuclear Reactor Regulation
SUBJECT:
DOCUMENTS ASSOCIATED WITH THE PROPOSED GENERIC LETTER REGARDING LONG-TERM SOLUTIONS AND UPGRADE INTERIM CORRECTIVE a
ACTIONS FOR THERMAL-HYDRAULIC INSTABILITIES IN BOILING WATER REACTORS l
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The Reactor Systems Branch has prepared the subject draft generic letter. The l
Committee to Review Generic Requirements (C;UIR) has reviewed and endorsed this i
draft generic _ letter. The Generic Communications Branch (0GCB) is preparing to publish the draft generic letter in the Federal Reaister for public coment. This memorandum provides a compilation of the background material l
relevant to the subject generic letter that should be made available to the public. By copy of this memorandum we are providing the enclosed documents to the Public Document Room. The enclosures are (I) the draft generic letter as endorsed by the CRGR, (2) the CRGR Review Package,-and (3) the NRC approved j
safety evaluation report on "BWR Owners' Croup Long-Term Stability Solutions i
licensing Methodology" NED0-31960 and Supplement I.
We request that you provide us with the Nuclear Documents System accession number for this memorandum. This information can be provided to the listed l
contact by telephone or by E-Mail.
j CM ir41 sig::ed by:
I F
G il E. k: cus Gail H. Marcus, Chief Generic Communications Branch Division of Operating Reactor Support Office of Nuclear Reactor Regulation
Enclosures:
As Stated
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CONTACT:
Peter C. Wen, NRR j
504-2832 t
Distribution w/
Enclosures:
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ENCLOSURE 1 i'
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ALL HOLDERS OF OPERATINS LICENSES FOR EDILING W TO:
REACTORS (EWRs)
CTIONS FOR LONG-TERM SOLUTIONS AND UPGRADE OF INTERIM IC THERMAL-HYDRAULIC INSTABILITIES IN E0lLING SL'BJECT :
LETTER 93-XX)
PURPOSE i
letter to The U.S. Muclear Regulatory Commission (NRC) is issuing this gener c ctions to augment its it request that each addressee (1) take the appropr a e athermal-hydraulic resoective protecures and training for preventing describing the instabilities in its reactor and (2) submit to the NRC a plan l
tation long-term stability solution option it has selected and the imp emen tion systems to schedule it proposes for the modification of plant protec d 12 in Appendix A ensure compliance with General Design Criteria (GDC) 10 anl tions (10 CFR to Part 50 of Title 10 of the Cede of Faderal Recu a i
EACKGROUND ilities The possibility of power oscillations due to thermal-hydraulic instab have been in boiling water reactors (EWRs) and the consecuences of such events The staff evaluated thermal-hydraulic stability as of concern for many years.
i in Generic Cer.eric Issue B-19 ana recommenced closure actions for re ilities in Letter S5-02 "Long-Term Solutions to Thermal-Hydrauli pliance with Eoiling Water Reactors, and to irpose operating limitations, as appropriate, to ensure com GDC 10 requires that the reactor core te designed with limits will not be GDC 10 and 12.
appropriate margin to assure that specified fuel design h
ffects of exceeded during any condition of normal operation, including t e e GDC 12 recuires assurante that power anticipated operational occurrences.
ified acceptable oscillations which can result in conditions exceeding spec dily fuel design limits are either not possible or can be reliably and rea detectec and suppressed.
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On March 9. 1988, LaSalle Unit 2 experienced an instability event.
The ensuing work by both the staff and industry organizations have provided additional insight into thermal-hydraulic instabilities in BWR cores.
The LaSalle event is described in NRC Information Notice 88-39, "LaSalle-l Unit 2 Loss of Recirculation Pumps With Power Oscillation EveY,I" dated June 15, 1988. NRC Bulletin 88-07, also dated June 15, 1988, highlighted the i
generic concerns identified by the LaSalle event and requested all BWR i
licensees, regardless of BWR type or analytical core stability margin, to review the adequacy of procedures, instrumentation, and operator training programs to respond to power oscillations.
In response to these concerns, the
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BWR Owners' Group QWROG) initiated i project to investigate actions that
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should be taken to resolve the CUR stability issue.
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On October 29. 1988, tne General Electric Company (GE) notified the NRC under-10 CFR Part 21 that thermal margins might not be sufficient to prevent violation of the minimum critical power ratio safety limit for some BWR plants if a 10-percent average power range monitor (APRM) oscillation was used as a procedural action point for manual scram of the plant.
Based o~ this possibility, GE recommended stability " interim corrective actions" in a Novemoer 1988 letter to BWR utilities.
On December 30, 1988, the NRC issued Bulletin E8-07, Supplement 1. approving the proposed BWROG/GE interim operating recommendations and stating additional conditions. One of these conditions addressed the applicability of the experience-based stability exclusion boundaries defined in the interim corrective actions, and noted the need to reevaluate and justify these boundaries for cores that include new fuel designs. This bulletin also discussed long-term corrective actions.
l Such corrective actions might incluce hardware modifications or additions to f acilitate manual or automatic protective response to avoid neutron flux oscillations or to suppress oscillations should they occur.
Since it is possible for some oscillations to grow to levels exceeding NRC safety limits in the order of a minute, automatic protection action is generally indicated.
These actions are being defined by a greatly expanded post-LaSalle BW90G study i
4 to develop a generic resolution to the stability issue.
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5 In June 1991, the BWROG issued NED0-31960 (Ref. 1) which documented proposed long-term solutions to the stability issue as well as methodologies that have been developed to support the design of these long-term solutions.
Supplement I to I4EDO-31960 (Ref. 2) was issued in March 1992 and contained final methodology details and additional information requested by the NRC.
By a July 1993 letter from A. C. Thadani (NRC) to L. A. England (BWROG), the NRC transmitted its safety evaluation report on NED0-31960 and its Supplement 1 (Ref. 3) to BWROG.
Reference 3 describes the regulatory positions resulting from the staff review of the proposed solution concepts and associated methodologies.
This letter requests information about licensee plans for implementation of the approved solution concepts and about actions taken to ensure that interim stability protection is adequate until the long-term solution is implemented.
This resolution assumes the protection system will function when required and does not consider the combination of instability and anticipated transients without scram (ATWS).
That subject is being addressed by other f4RC and BWROG activities.
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NEED FOR MODIFICATION TO INTERIM CORRECTIVE ACTIONS I
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In early 1992, the BWROG, citing its continuing studies, provided its members additional guidance on implementation of the interim corrective actions attached to NRC Bulletin 88-07, Supplement 1.
In this guidance, the BWROG emphasized the need for caution when operating near the exclusion regions and 1
recommended reexamination of procedures and training to ensure that uncertainties in the definition of exclusion region boundaries were appropriately reflected.
The NRC staff considered this guidance in conjunction with an Augmented Inspection Team (AIT) review of a Washington Nuclear Power Unit 2 (WNP-2) 1992 instability event. The AIT report (Ref. 4) discusses that review and the BWROG guidance.
On August 15, 1992, Washington Nuclear Power Unit 2 (WNP-2) experienced power oscillations during startup.
The NRC evaluated this event, concluding that
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the primary cause of the oscillations was very skewed radial and bottom peaked axial power cistributions due to insufficient procedural control of control rod removal patterns during power ascension.
It was concluded from j
discussions with other licensees that similar procedural practjes were not unusual for some other BWRs, However, the vulnerability to instability was magnified in WNP-2 because the core loading consisted of a mixture of 9x9 and Bx8 fuel types which caused unbalanced flow and pressure drop characteristics.
The WNP-2 event is described in NRC Information Notice 92-74. "Pcwer Oscillations at Washington Nuclear Power Unit 2," dated November 10, 1992.
Most of the SWROG long-term solutions proposed in NED0-31960 (Ref. 1) involve suostantial modifications to the plant protection system hardware; these i
modifications are not expected to be ready for implementation until 1995-1996.
The staff review of analytical studies in support of these solutions, and the circumstances leading to the WNP-2 event, have identified the following vulnerabilities wnich reflect uncertainty in the effectiveness of the current interim corrective actions to provide protection until implementation of the long-term solutions can be accomplished:
(1)
Eulletin E8-07 Supplement 1, requested licensees of BWRs to take actions including a procedural requirement for a manual scram under all circumstances resulting in no recirculation pumps operating with the reactor in the RUN mode.
This action was not applicable to plants with effective automatic scram protection against out-of-phase regional oscillations.
Bulletin 88-07, Supplement 1 indicated that a flow-biased APRM scram circuit without a simulated thermal power monitor (filtered APRM signals) would provide such protection.
However, more recent analyses by EWROG show that the flow-biased APRM scram does not provide sufficient protection for the out-of-phase mode of instability, which can produce very large asymmetric oscillations before exceeding the average power scram set point.
The need for protection against out-of-phase oscillations remains under review for a few small core plants with small inlet orifices.
Therefore, prior to completion of the long-term solutions reviews, the only plants that qualify for an exception to the 1
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Supplement I requested procedural requirement for manual scram are the BWR2 plants, Oyster Creek and Nine Mile Point 1, which have quadrant-l based APRM scram protection.
(2)
Bulletin 88-07, Supplement 1, endorsed the experience-based power / flow
.l boundaries of the interim corrective actions based on tlfe%ssumption that other factors important to the core stability characteristics l
(e.g., radial and axial peaking, feedwater temperature, and thermal-hydraulic compatibility of mixed fuel types) were consistent with f
previous experience and the bounding values expected during normal l
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operation. The BWROG studies and the precautions recommended in the.
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j early 1992 letter to its members indicated that uncertainties existed in the definition of these boundaries (Ref. 4). The WNP-2 instability event and subsequent NRC evaluation determined that many licensees have j
given inadequate attention to'the impact on core stability of the reload core design and operating procedures for changing reactor power. The WNP-2 experience also highlighted the value of using on-line stability monitors as an operational aid to avoid unstable operation; the capability for on-line stability monitoring does not exist currently for f
most BWRs.
l RE0 VESTED ACTIONS 1.
Within 60 days of receipt of this letter, all BWR licensees, except for Consumers Power Company (Big Rock Point, which does not have the capability for operation under variable flow conditions), are requested to review tneir operating procedures and operator training programs developed for the interim corrective actions and modify them as appropriate to strengthen the administrative provisions intended to avoid power oscillations or to detect and suppress them if they occur before the long-term solution can be implemented. The experience gained at WNP-2 should be a primary guide in this review.
In doing this, each BWR licensee (except for Big Rock Point) should:
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For all reactors except Oyster Creek and Nine Mile Point Unit 1, j
include procedural requirements for a manual scram under all i
operating conditions when two recirculation pumps trip (or there are j
no pumps operating) with the reactor in the RUN mode;
)
b.
Ensure that f actors important to core stability chdTcteristics (e.g., radial and axial peaking, feedwater temperature, and thermal l
hydraulic compatibility of mixed fuel types) are controlled within appropriate limits consistent with the core design, power / flow j
exclusion boundaries, and core monitoring capabilities of the i
reactor in cuestion, and that these factors are controlled through procedures governing changes in reactor power, including startup and shutdown, particularly at low-flow operating conditions.
If it is concluded that a near-term upgrade of core monitoring capability is f
called for to ease the burden on operators, determine the need to incorporate on-line stability monitoring or improved power distribution and thermal limits monitors, and inform the NRC of the i
schedule for such upgrades found to be necessary.
The procedural operation controls implemented for the interim corrective actions i
should be considered for retention as appropriate to compliment plant specific long-term solution approaches.
2.
By January 31, 1994, all BWR licensees, except for Big Rock Point, are requested to develop and submit to the NRC a plan for long-term stability corrective actions. including design specifications for any hardware modificatior.s or additions to facilitate manual or automatic protective response needed to ensure that the plant is in compliance j
with General Design Criteria 10 and 12.
An acceptable plan could provide for implementing one of the long-term stability solution. options proposed by the BWROG and approved by the NRC in Reference 3 or in subsequent documentation. The plan should include a description of the.
l action proposed and a schedule of any submittals requiring plant--
I specific design review and approval by the NRC and an installation schedule (if applicable).
The plan should also address the need'for i
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l near-term and long-term technical specification modifications.
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REPORTPlG REOUTREMENTS Pursuant to Section 182a of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f), each holder of a BWR operating license, excd$t"for Big Rock j
Point, shall:
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a 1.
Within sixty (60) days of the date of this letter:
i a.
Inform the ilRC, in writing and under oath or affirmation, of the licensee's plans and status with respect to the actions requested in l
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this letter, j
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b.
If the licensee does not plan to take an action requested in this j
letter, the reasons for not taking the action, a description of the 1
nature of any substitute action, and a schedule for completing or implementing the substitute action; j
2.
If the licensee plans to take an action requested, or a substitute action, within thirty (30) days of the completion of the action inform the fiRC, in writing and under oath or affirmation, of the action taken and verify its completion or implementation.
Each submittal shall be addressed to the U.S. fluclear Regulatory Commission, ATTil: Document Control Desk, Washington, DC 20555.
A copy shall also be submitted to the appropriate Regional Administrator.
This generic letter requests information that will enable the flRC to verify that the licensee is complying with its current licensing-basis regarding i
GDC 10 and 12. Accordingly, an evaluation justifying this information request is not necessary under 10 CFR 50.54(f).
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o REFERENCES i
1.
NED0-31960, "BWR Owners' Group, Long-Term Stability Solution' Licensing l
Methodology," May 1991.
2.
f1ED0-31960, Supplement 1, "BWR Owners' Group Long-Term Stability l
Solutions Licensing Methodology," March 1992.
l 3.
Letter from A. Thadani, f1RC, to L. A. England, Chairman, BWR Owners' '
Group, Acceptance for Referencing cf Topical Reports NED0-31960 and NED0-31960, Supplement 1, "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," dated July 1993.
l 4.
Letter from J. B. Martin, NRC, to A. L. Oxsen, Washington Public Power c pply System. "NRC Augmented Inspection of Washington Nuclear Project, u
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Unit 2",
September 29, 1992.
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PAPERWORK REDUCTION ACT STATEMENT I
This aeneric letter contains information collection requirements that are l
subject to Paperwork Reduction Act of 1980 (44 U.S.C. 3501'et seq.).
These requirements were approved by the Office of N:m ement and Budget, approval number 3150-0011, which expires May 31, 1994.
The public reporting burden for this collection of information is estimated to average 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information.
Send comments regarding this burden estimate or any other aspect of this collection of information, including suggestions for further reducing the reporting burden to the Desk-Officer, Office of Information and Regulatory Affairs. (3150-0011), NE0B-3019, Office of Management and Budget. Washington DC 20503, and to the U.S. Nuclear Regulatory Commission, Information and Records Management Branch, (MNBB-7714),
Washington, DC, 20555.
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9 Compliance with the following request for information is voluntary. The information would assist the NRC in evaluating the cost of complying with this generic letter.
(1) the licensee staff's time and costs to perform requestdd")rocedure reviews and implementation of changes; (2) the licensee staff's time and costs to prepare the requested reports and documentation; (3) the additional short-term costs incurred to address the changes, such as the costs of the corrective actions or the costs of down time; and (4) an estimate of the additional long-term costs that will be incurred as a result of implementation commitments.
j BACKFIT DISCUSSION l
l This generic letter defines the requested actions and reporting requirements to be met by all holders of BWR operating licensees, except for Consumers Power Ccmpany (Big Rock Point). in order to enhance the current interim l
l corrective action and to provide a long-term solution to the issue of thermal-l hydraulic instabilities in BWRs. The staff has concluded that these requested actions and reporting requirements are a backfit that is necessary to ensure compliance with GDC 10 and 12.
The basis for the determination is stated in the preceding discussion of the generic letter. Accordingly, pursuant to 10 CFR 50.109(a)(4)(i), a backfit analysis is not required.
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10 If you have any questions about this matter, please contact the technical contact listed below or the appropriate NRR project canager.
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James G. Particw, Associate Director for Projects Office of Nuclear Reuctor Regulation
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Technical
Contact:
Larry Phillips, NRR (301) 504-3232 1
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s DC[DSURE 2 s-E ENCLOSURE Resconse to Recuirements for Content of Package Sabmitted for r
CRGR Review (1)
The proposed generic requirement or staff position as it is proposed to be sent out to licensees _.
The preposed generic requirements are set forth in a i
draft generic letter (Enclosure 1).
The purposes of the generic letter are to:
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(a)
Provide guidance for the backfit modification of plant protection systems as a long-tern solution I
to assure compliance with the General Design Criteria (GDC) 10 and 12 as set forth in Appendix A of 10 CFR 50.
(b)
Provide guidance to strengthen the interin i
administrative controls requested in NRC Bulletin 88-07, Supplement 1.
j (c)
Require licensees to advise the staff of plans relating to the resolution of the issues.
(ii)
Draft staff papers or other underlying staff documents supporting the requirements or staff positions.
i (a)
"Long-Tern Solution To Thernal-Hydraulic Instabilities In Boiling Uater Reactors (Generic i
Letter 92-XX) t l
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2 (b)
Connission Papers providing Progress Reports on the Staff Review of Regulatory Implications of the Power Oscillations Event at LaSalle 2:
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SECY-92-386, November 16, 1992 I
SECY-91-323, October 11, 1991 SECY-91-090, April 3,
1990
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SECY-90-152, April 27, 1990 SECY-89-074, February 27, 1989 (c)
"NRC Bulletin No. 88-07:
Power Oscillations In Boiling Water Reactors (BWRs)," June 15, 1988, and "NRC Bulletin No. 88-07, Supplement 1,"
December 30, 1988.
l (d)
"EWR Owners' Group Long-Tern Stability Solutions Licensing Methodology," NEDO-31960, May 1991, and i
l Supplement 1, March 1992.
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(e)
Letter from A. Thadani (NRC), To C. Tully (BWROG),
transmitting the NRC " Safety Evaluation Report on "EWR Owners' Group Long-Tern Stability Solutions Licensing Methodology," NEDO-31960 and Supplement 1, December 1992, and attached ORNL Technical Evaluation Report, " Licensing Basis for l
l Long-Tern Solutions to BWR Stability Proposed by l
the BWR Owners' Group," ORNL/NRC/LTR-92/15, August 1992.
(f)
NRC Inforration Notice No. 88-39:
"LaSalle Unit 2 j
i Loss of Recirculation Pumps with Power Oscillation j
i Event," and No. 92-74:
" Power Oscillations at Washington Nuclear Power Unit 2."
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3 (iii)
Each proposed requirement or staff position shall contain the sponsoring office's position as to whether the proposal would increase requirements or staff positions, implement existing requirements or staff positions, or would relax or reduce exist $Eg requirements or staff positions.
l Licensee response to the proposed generic letter would implement staff positions and requirements pertaining to the stable cperation of BWRs as required by GDC 10 and 12.
(iv)
The proposed method of inplementation along with the concurrence (and any comments) of OGC on the method proposed, the concurrence of affected program offices, or an explanation of any nonconcurrence.
The commitments in licensee response to the generic letter would be implemented in accord with the schedule identified in the generic letter or as nodified by the
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licensees in the responses.
There will not be staff l
review of the responses on the "interin corrective actions" (ICA), but there would be regional inspections of the procedures implementing the ICA.
The hardware and software proposed to implement the "long tern solutions" till be reviewed by the staff generically and, in some cases, on plant specific basis.
These docunents have been reviewed by OGC.
Their suggestions have been incorporated.
OGC has no legal objections.
1 (v)
Regulatory analyses generally conforning to the l
directives and guidance of NUREG/BR-0058 and i
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b e
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(This does not apply for backfits that ensure compliance or ensure, define or redefine adequate protection.)
The "long-tern solution" and the augmentdY""interin I
corrective actions" which the generic letter requests to be selected and inplemented for each BWR are i
necessary in order to ensure compliance with the requirenents of GDC 10 and 12.
Therefore, a regulatory 4
analysis is not necessary.
(vi)
Identification of the category of reactor plants to
.nich the generic requirement cr staff position is to 1
apply.
All holders of EWR operating 'icenses, except Big Rock Point, are affected by this generic letter.
All affected BWRs are requested to select and implement a "long-term solution, and implement the augmented "interin corrective action" as indicated in the generic letter.
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- or backfits other than compliance or adequate protection backfits, a backfit analysis as defined in j
i The requirements of this generic letter are necessary to ensure compliance with GDC 10 and 12.
Accordingly, pursuant to 10 CFR 50.109, a backfit analysis is not required.
('/iii )
For each backfit analyzed pursuant to 10 CFR 5 0.109 ( a) (2 ) and (c)
(i.e.,
not adequate protection backfits and not conpliance backfits), the proposing office director's determination, together with the rationale
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for the determination based on the considerations of j
paragraphs (i) through (vii) above, that I
i (a) there is substantial increase in the overall 4 > 9P**
protection of public health and safety or the j
connon defense and security to be derived from the proposal; and (b) the direct and indirect costs of implementation, for the facilities affected, are justified in view j
cf this increased protection.
j
- ot applicable.
(ix)
For adequate protection or compliance backfits evaluated pursuant to 10 CFR 50.109 (a) (4 )
i (a) a documented evaluation consisting of:
i (1) the objectives of the nodification (2) the reasons for the nodification (3) the basis for invoking the compliance or adequate protection exemption.
(1)
The objectives of the long-tern solutions (LTS) are to provide automatic protection systems to prevent the occurrence or adverse consequences of large magnitude power oscillations in BWRs.
The augmentation of the interin corrective actions (ICA) is to improve operator action, for the same i
objective, for the interin period before the LTS are implemented.
The objectives and LTS are
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I discussed at length in the documents of section ii of this package, particularly the Commission I
papers of ii(b) and the LTS topical reports and corresponding SER of ii(d) and (e).
The ICA and i
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auamentations are discussed in the bulletins and j
information notice of ii(c) and (f).
l (2)
The LaSalle instability event of March 1988 i
f triggered an extensive analytical study and I
interaction.by the NRC (involving MRR and RES consultants) and the BWR Owners' Group (using GE l
j as a consultant for analysis).
This study, as documented, in part, in the above [ix(a)] reports, I
concluded that improvements in the protection j
system and/or restrictions in the operating region l
of the power / flow map are needed to prevent or 7
suppress large oscillations which could otherwise l
I result in exceeding fuel safety limits, in l
violation of GDC 12.
The study also concluded f
that interim procedural requirements were i
j necessary until the LTS were in place.
The recent WMP-2 instability event [ Reference ii(f)]
indicated the need for augmentation of existing interin procedural requirements [ Reference ii(c)].
(3)
The staff review indicated that with current protection systen action, it is possible to violate GDC 12 by exceeding fuel safety limits, particularly with possible nonsymmetric oscillations (from side to side across a core diameter) which reduce-the effectiveness of the existing APRM scram system.
It is possible to enter tnese situations both during maneuvers in the lower power / flow range and from events in the l
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i high power region leading to entry into unstable regions.
j (b)
In addition, for actions that were immediately
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1 effective (and therefore issued without prior CRGR l
review as discussed in III.C) tne evaluation shall document tne safety significance and appropriateness of the action taken and (if i
applicable) consideration of how costs contributed to selecting the solution among various acceptable i
alternatives.
Not applicable.
v (x)
For each evaluation conducted for proposed relaxations or decreases in current requirements or staff positions, the proposing office director's determination, together with the rationale for the determination based on the considerations of paragraphs (i) through (vii) above...
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Not applicable, i
(xi)
For each request f or information under 10 CFR 50.54 (f)
(which is not subject to exception as discussed in III.A) an evaluation that includes at least the following elements:
(a)
A problem statement that describes the need for the information in terms of potential safety benefit.
l (b)
The licensee actions required and the cost to develop a response to the information request.
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1 (c)
An anticipated schedule for URC use of the infor ation.
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(d)
A statement affirmina that the reauest does not i= pose new requirements on the licenTe'e, other l
than for the requested information.
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There are 2 areas of information required by the l
Generic Letter; (1.a) notification of the NRC of the licensee's plans and status with respect to the actions l
l requested, or (1.b) the reasons for not taking the l
action, a description of any substitute action and the l
schedule for completion er implementation of the i
i substitute action, and (2) informing the NRC within 30 days of the completion of the action taken and j
verification of its completion or implementation.
The j
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requested actions, in brief, concern (1) revisions and modifications of current procedures and training, scram actions, control of operating parameters and l
r examination of the need for near-term core nonitoring, i
(2) plans and schecule for long-tern stability corrective action and (3) documentation of evaluation i
of adequacy of stability related training programs, i
procedures and core monitoring.
The reporting l
I requirements are to inform the NRC that action
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j requested in the Generic Letter have been completed.
1 The cost and ranpower to provide these subnittals are estinated to be 5 person weeks per utility.
These requirements and requests do not impose new
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requirements on the licensee, other than the requested information.
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(xii)
An assessment of how the proposed action relates to the Connissicn's Safety Goal Policy Statement.
i This assessnent is not applicable since the proposed l
actions and requirements are necessary to ensure i
i compliance with GDC 10 and 12.
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W ASemvGTON. D.C. 0556 July 12,1993 L.
A.
England, Chairman BWR Owners' Group Gulf States Utilities River Bend Station North Access Rd.
St. Francesville, LA 70775 l
Dear Mr. England:
SUBJECT:
ACCEPTANCE FOR REFERENCING OF TOPICAL REPORTS NEDO-31960 AND NEDO-31960 SUPPLEMENT 1, "BWR OWNERS GROUP LONG-TERM STABILITY SOLUTIONS LICENSING METHODOLOGY" (TAC NO. M75928)
The staff has completed its review of the Topical Reports NEDO-31960 and NEDO-31960 Supplement 1," suositted by the BWR Owners Group (BWROG) by letters dated May 31, 1991 and March 16, 1992.
These reports describe and justify the use of several BWROG developed long-term solutions to BWR stability issues and the methodologies developed for evaluating appropriate setpoints and performance criteria for these solutions.
We find the solutions of NEDO-31960 and Supplement 1 to be acceptable for referencing in license applications to the extent specified, and under the limitations delineated in NEDO-31960 and Supplement 1 and the associated NRC technical evaluation.
The enclosed safety evaluation defines the basis for acceptance of these topical reports.
We do not intend to repeat our review of the matters found j
acceptable as described in NEDO-31960 and Supplement 1 when the report appears as a reference in license applications, except to i
assure that the material presented is applicable to the specific plant involved.
Our acceptance applies only to the matters described in the application of NEDO-31960 and Supplement 1.
In accordance with procedures established in NUREG-0390, it is requested that the BWROG publish accepted versions of this l
topical report, within three months of receipt of this letter.
The accepted versions shall include an "A"
(designating accepted) following the report identification symbol. D b i
o i ( L. A. England July 12, 1993 ) Should our criteria or regulations change so as to invalidate our i conclusions concerning the acceptability of the report, the BWROG or the applicants referencing the topical report will be expected to revise and resubmit their respective documentation, or submit justification for the continued effective applicability of,the topical report without revision of their respective.r-documentation. Sincerely, / {' 7 l ft 3,^. h ? A W, Ashok/C..Thadani, Director Division of Systems Safety and Analysis { i
Enclosure:
NEDO-31960 Evaluation i i I i l
i I v l l \\ j ENCLOSURE i I SAFETY EVALUATION REPORT ON f "BWR OWNERS' GROUP LONG-TERM STABILITY SOLUTIONS [ LICENSING METHODOLOGY _" l NEDO-31960 AND SUPPLEMENT 1 1 INTRODUCTION The Boiling Water Reactor Owners' Group (BWROG) has submitted'to the U.S. Nuclear Regulatory Commission (NRC) the Topical Report NEDO-31960, "Long-Term Stability Solutions Licensing Methodology," (Ref. 1) and Supplement 1 (Ref. 2) for staff j l review. The long-term solutions described in these reports. l i l consist of conceptual designs for automatic protection systems i t i aeveloped by the BWROG with its contractor, General Electric l Cc pany (GE), to either prevent stability-related neutron flux oscillations or to detect and suppress any oscillations that may occur. The reports also describe the methodologies that have been developed to establish setpoints and demonstrate the adequacy of the protection systems to prevent violation of the l t critical power ratio (CPR) safety limits in compliance with j l General Design Criteria (GDC) 10 and 12 in Appendix A to Part 50 i of Title 10 of the Code of Federal Reculations (10 CFR Part 50). i The BWROG has requested that the NRC. staff review these reports l l and approve the following: 1 The overall regional exclusion and detection and suppression methodology, including the overall treatment of uncertainties. The solution concepts and associated licensing approach for each option. The application of the methodology to the solution concepts, l
t s-1 l i 2 i r The BWROG/GE position concerning safety classification of i new and existing hardware; that is, the safety-l classification of all existing and interfacing equipment-t i i should not change when new or modified stability long-term i j solution hardware is installed. ? . >>w- + l l f The SWROG plans to proceed with the selection of options and 1 i specific hardware design based on NRC's approval of the proposed j i concepts and associated methodologies. i i 2 EVALUATION l ( The solution options proposed by the BWROG in NEDO-31960 are: I Exclusion Reaion A region in the high-power / low-flow area of the power / flow map outside of which instabilities are very unlikely is calculated for each representative BWR type using well-defined procedures. If the reactor is i operated within this exclusion region, an automatic i l t protective action is initiated to exit the region. This l action is based exclusively on power and flow neasurements; l the presence of oscillations is not required for its l l i .nitiation. Four solutions for the Type I option have been proposed by the EWROG, although not all have been completely developed: l 1 I-A Immediate protective action is taken when entering the exclusion region. This action can be either a scram or a select rod insert (SRI). i l I-B Same as I-A, but the protective action when entering l the exclusion region can be bypassed if a stability l j nonitor is operational and detecting sufficiently stable conditionc (e.g., a decay ratio less than 0.6). l I ?
h e 3 t j I-C Protective action is taken if the following two i conditions both exist: (1) the reactor is operating 1 inside the exclusion region and (2) an average power range monitor (APRM) oscillation (of small nagnitude) { is detected. .-~ q I-D A few small-core plants with tight inlet orifices have i a reduced likelihood of out-of-phase instabilities. ) i i ] For these plants, the existing unfiltered, flow-l j biased APRM scram provides sufficient protection. In i l addition, administrative controls are proposed to J naintain the reactor outside the exclusion region. j II Ouadrant-Based APRM Scran In a BWR/2, the. quadrant-I based APRM is capable of detecting both in-phase and I i 1 out-of-phase oscillations with sufficient sensitivity to j initiate automatic protective action to suppress the q oscillations before safety nargins are compromised. i i ] III iPRM-Based Detect and Suporess. Local power range monitor ] (LPRM) signals or combinations of a small number of LPRMs are analyzed on line by using three diverse algorithms. If l any of the algorithms detect an instability, automatic { protective acticn is taken to suppress the oscillations before safety margins are compromised. Two options have j been proposed by the BWROG: Option III and Option III-A. ] The main difference between the two is in the hardware implementation. Option III requires a new Class 1E computerized systen. For Option III-A, newly designed digital replacements of the existing APRM amplifier cards will be used and a smaller number of LPRM detectors in a revised configuration will be required. Conceptually, the j l algorithms are similar in both solutions. 1 i i i u 4 i ~
v 4 The NRC contractor, Oak Ridge National Laboratory (ORNL), I assisted the staff in reviewing the topical reports. ORNL has l provided a technical evaluation report (TER) that is included as l l. The TER describes the results of the staff's review of the functional performance criteria for the proposed protection systems and of the assumptions, principles? and models . inherent in the methodologies used to define protection system i stability boundaries and setpoints. The staff's evaluation of } hardware safety classification follows. i At a meeting on March 26, 1992, with the NRC staff, the BWROG proposed that for the long-term stability solution options 1 relying on "APRM flow biased scram" recirculation drive flow signals, the use of existing hardware be allowed in the new 1 protection system. The recirculation flow drive system, although highly reliable, is not designed to Class 1E standards. The staff, therefore, requested additional information on plant-specific arrangements of the existing recirculation drive flow ] instrument channels, channel integrity and independence, the failure rate data for each cenponent in the flow channels, and the failure indication alarms. The response, which the BWROG transmitted with a July 17, 1992, letter (Ref. 3) provides the l results of a survey among 9 licensees for 12 operating plants. l l In general, redundant flow channels exist in these arrangements. The failure history of the channel components (from eight BWR units covering 84 reactor-years) shows the failures to be random and the failure rate to be insignificant. For failure indication, the output signal from a flow channel is compared to the cutput signal from another flow unit. The comparator l activates an alarm when two flow signals differ more than the specified tolerance. Alarms also are activated when the comparator fails high or low. Isolators are provided between i flow units and between the comparator and the APRM circuitry and the alarm circuitry. However, the survey results indicate that many operating plants do not meet the configuration in BWROG l
t i l s 5 ( Viewgraph 3/26-7, " Drive Flow Signal Path," which was shown at I the meeting on March 26, 1992. The staff will review the hardware design details on a plant-specific basis. In general, it finds the proposed concept to be may require modifications-for some pT5nts. acceptable, but 3 CONCLUSIONS i l The staff has reviewed the licensing basis for long-tern 1 l solutions to BWR stability proposed by the BWROG and adopts the i 1 "ecommendations described in the attached TER. The regulatory i l positions with respect to the specific approvals requested by the l BWROG are summarized below: l l (1) Methodoloov The exclusion region calculation methodology described in NEDO-31960 and its Supplement 1 is acceptable for defining the Option I-A exclusion region and the Options III and III-A exclusion boundaries outside of which the detect and I suppress action may be deactive ed. The overall treatment of uncertainties is acceptable for the selection of initial conditions and for the celection of oscillation contours and the treatrent of failed LPRM sensors for Options III and III-A. The methodology is acceptable for evaluating the protection provided by the Option II quadrant-based APRM scram. Specific procedures for application of the methodology consistent with documentation and calculations submitted for this review should be developed and documented by BWROG. (2) Solution Concents l (a) Options I-B and I-C have not been developed in l
1 +. l i 6 detail by the BWROG and, therefore, will not be i considered acceptable as long-term solutions until fully developed by the BWROG and reviewed and l approved by the staff. Option I-D is still under review and its acceptability as a long-term solution depends, to a large degree, on the details *of i calculations that are not yet available. Attachment 1 identifies some concerns about the option I-D reliance on predictive calculations to conclude that the out-of-phase mode of oscillation will be avoided. l To address these concerns and to provide reasonable l assurance that out-of-phase oscillations will be f l avoided by I-D plants, it may be necessary to incorporate strict operational controls on axial and radial power distribution and to enhance the capability to recognize operating conditions that are approaching inctability by other means such as on-1 I line stability monitoring. Core stability l sensitivities are illustrated by experience with the i instability event on August 15, 1992, at Washington 1 Nuclear Power Unit 2, in which oscillations developed outside of the stability exclusion regions because of a combination of fuel, core design and control rod patterns which resulted in conditions unfavorable to l core stability, and conditions that an NRC inspection team concluded to be vulnerable to out-of-phase ) instability (Ref. 5). The staff will evaluate the acceptability of Option I-D when the calculations for the lead plant are submitted. If the lead plant analyses are acceptable, the staff will evaluate 1 detailed calculations for all plants that may propose i Option I-D (e.g. Duane Arnold, Vermont Yankee, Monticello, and FitzPatrick). If individual plant analyses are inconclusive because of large uncertainties involving assumed operating
r ? ) 1 4 k 7 conditions, the quality of adlainistrative controls and available core monitoring to reduce instability { vulnerabilities will be considered in evaluating the Option 1-D acceptability for a specifio plant. J (b) The implementation of Option I-A is an ac*cIhtable long-term solution for any type of BWR, subject to the following conditions. l (i) Specific reload confirmation precedures should l be developed so that for every reload, the l 1 licensee can either confirm the applicability l of old exclusion region settings or set a new exclusion region boundary. { l i (ii) The exclusion boundary setpoints for this l option should be sufficiently bounding to avoid l change, on a cycle-by-cycle basis. Major i setpoint changes should be expocted only if the a fuel design changes significantly. i 1 l (iii) When establishing reactor trip setpoints for 1 i the power /ficw exclusion region scram, operational restrictions on other parameters important to stability (e.g., radial and axial power distribution during low flow power caneuvering) that are consistent with the assumptions of the exclusion boundary analyses should be addressed, including the need for technical specifications, and factored into tne setpoint evaluation. (iv) Select rod insert (SRI) may-be used in conjunction with Option I-A, but a full scram should occur if the reactor does not exit the i i s.- e., -n--m n w-ee- .r-- , ww .,m-,9 s-g y ,w,
l i i 8 region within a reasonable period of time (about a few seconds), i -(c) Option II is an acceptable long-term solution for t implementation in BWR/2s, which have quadrant-based For implementing Option II7* plant-l APRM scrams. specific analyses should show that the quadrant-based l APRM scram provides sufficient protection against out-of-phase instability modes to avoid the violation of CPR safety limits. I l r ? (d) Options III and III-A are acceptable long-term solutions for implementation in any type of BWR, subject to the following conditions: l (i) All three alcorithms described in NEDO-31960 l and supplement 1 should be used in option III I or III-A. These three algorithms are high LPRM oscillation amplitude, high-low detection i algorithm, and period-based algorithm. (ii) The validity of the scram setpoints selected should be demonstrated by analyses. These analyses may be performed for a generic representative plant when applicable, but should include an uncertainty treatment that accounts for the number of failed sensors permitted by the technical specifications of the plant's applicant. (iii) Implementation of Option III or III-A will require that the selected bypass region outside of which the detect and suppress action is j deactivated be defined in the technical specifications.
[ a,* 9 (iv) If the algorithms detect oscillations, an j automatic protective action should be initiated. This action may be a full scram or l an SRI. If an SRI is implemented with Option [ III or III-A, a backup full scram E6st take f effect if the oscillations do not disappear.in i a reasonable period of tine or if they reappear ( before control rod positions and operating conditions have been adjusted in accordance f with appropriate procedural requirements to permit reset of the SRI protective action. (v) The LPRM groupings defined in NEDO-31960 to provide input to the Option III or III-A algorithms are acceptable for the intended oscillation-detection function. These LPRM groupings are the oscillation power range monitor for Option III or the octant-based j arrangements for Option III-A. The l requirements for a minimum operable number of LPRM detectors set forth in NEDO-31960 are i acceptable. l (e) Options I and II do not protect the fuel against single-channel instability, and the protection provided by Options III and III-A for single-channel instability is not highly reliable. When implementing the long tern solution, a procedure to review the thermal hydraulic stability of lead use assemblies (LUA) in a core reload should be established. The review should ensure that inclusion of the LUA as proposed in the core reload is very unlikely to result in single-channel instability. l
y s i 10 (3) Safety Classification As a minimum, the recirculation drive flow channel should comply with the requirements of the Electrical and Electronics Engineers, Standard 279 (Re f*?"4 ), which include the single-failure criterion, component j quality, channel independence, and the cr.pability for test and calibration. Isolation devices are required to be qualified for their application. No credible failure at the output of an isolation device should f prevent the associated protection system channel from meeting the minimum performance requirements specified in the design bases. The plant-specific submittal should include the specification f documentation for the isolation device. In addition, l. because Solution I-A involves an automatic reactor scram function, any modification to the reactor l 6 protection system trip function requires a submittal i i to the NRC proposing a change in the technical specifications. The plant-specific technical specification change should include limiting f conditions for operation, action statements, allowable out-of-service times, surveillance tests, and test frequency commensurate with the importance to safety of the system. The detailed technical specification requirements should be addressed generically during review of the detailed hardware design. 4 PEFERENCES 1. NEDO-31960, "BWR Owners' Group,Long-Tern Stability Solutions Licensing Methodology," May 1991.
- - _ - ~ l l l v i l i 11 i l l l 2. NEDO-31960, Supplement 1, "BWR Owners' Group Long-Term l Stability Solutions Licensing Methodology," March 1992. [ 3. Letter from C. L. Tully (BWROG) to A. C. Thadani (NRC), j " Response to RAI on Stability Report NEDO-31'960*,'* dated June 5, 1992," July 17, 1992. 4. Institute of Electrical and Electronics Engineers, Standard I 279, " Criteria for Protection Systems for Nuclear Power Generating Stations." i 5. Letter from J. B. Martin, NRC, to A. L. Oxsen, Washington i Public Power Supply System, "NRC Augmented Inspection of Washington Nuclear Project, Unit 2", September 29, 1992. i l i l l 4 l l
Attach =ent 1 e ORNL/NRC/LTR-92/15 i Technical Support for the Reactor Systems Branch (L1697/P2) l Contract Program: Licensing Basis for Long-Term Solutions to BWR Stability Proposed I Subject ot' Document: by the BWR Owners' Group h Type of Document: Technical Evaluation Report 1 Autnor: Joss Staren-Leuba Date er' Document: August 1992 NRC hionitor - T. L. Huang. Otfice of Nuclear Reactor Regulation Prepared for U.S. Nuclear Regulatory Commission Office er' Nuclear Reactor Regulation under DOE Interagency Agreement 1886-8947-8A NRC FIN No. Ll697. Project 2 Prepared by Instrumentation and Controls Division OAK RIDGE NATIONAL LABORATORY managed by N1ARTIN N1ARIETTA ENERGY SYSTEh15. INC. for the U.S. DEPART 51ENT OF ENERGY under contract DE-AC05-540R21400 f
SUMMARY
This report documents the mam conclusions and recommendations derived from our review or the Boiling Water Reactor Owners' Group (BWROG) long-term solutions of the stability issue as described in NEDO-31960 and its Supplement-1 (references I and 2). Overall, this review is very positive. Our mam conclusion is that all three of the proposed solution types (1, II. and III) are tecnnically souna and, in our opinion. any of them will solve the stability issue if properly appiied. t Althougn not speci6cally related to the stability issue, the fact that implementation'of a new reactor protection tunction will most probably result in an increased number of c$,llenges t a reactor protec'. ion system may lead to a new safety problem unless the number of unnecessary challenges is mmtmued by design. It is recognized that the normal function of these solutions is to provide an automatic protection action (i.e., a scram or a runback) if either oscillations are detected ar the exclusion region is entered; however the implementation of this function must be performed carerully in orcer to minimize the number of unnecessary actuations while maintaining a very high l pronaciIity to penorm the intended safety funuion. Detaileo recommendations. including some qualifiers and reservations. are specided in the mam text of this report. A condensed summary of these recommendations tollows: Approse tne overall licensing methodology described in NEDO-31960 and its Supplement i j L for Solutions 1-A.11. and 111. This methodology includes the treatment of uncertainties and j the selection of initial conditions and calculation parameters. The approvai should be l conditioned to assure plant-speci6c consistency with the axial (2.0) and the radial (end-of-l cycle Haling) peaking factors assumed for the core power distribution calculation parameters. Do not approve Solution concepts I-B and 1-C because of its lack of detailed development 2. i and/or interest by the BWROG. l Do not approve Solution concept 1-D at this time until the tinal evaluations that NRC has 3 reauestea have been performed. This recommendation does not imply a rejection of l Solution i-D: ine approval of Solution 1-D depenas on the detatis or calculations that have not yet been performeo by the BWROG. l A select rod insert (SRI) is an acceptacle automatic protection action for any or the approved A solutions al-A. II. or ilD as long as a tull scram takes effect if either the oscillations do not disappear or the reactor does not exit the exclusion region within a reasonable period of time ta few seconds). The exclusion region must be examined prior to each plant operating cycle to assure consistency with the axial and radial power peaking distribution assumed in tt.e exclusion region coundary calculations. The BWROG must establish a critetta to limit radial and axial peaking factors during startup ( operations to those values considered for the analyses of the exclusion region. I The mam technical issue of signincant relevance that still remains to be solved is the reload-l dependent condrmatory analyses recuired to assert the applicability of the previous-cycle safety l ~ ld" exciusion regions to new types of fuel and settmgs and. in particular. the applicability ot o !oading patterns. The BWROG is aware of this problem and is currently developing a methodolog l l 1 l I i
t i for these cycle-dependent condrmations that is based on a " response sutiace" approach. The goal is that these condrmatory calculations should be expected to be positive most of the time; major setpoint changes should only be expected following sigm6 cant fuel design changes. He documentation of thi reioad-condrmation methodology is expected in Supplement 2 to NEDO-31960 that should be published in the spring of 1993. Supplement 2 wdl also contain a correlation to estimate the loss
- rttical-power-ratio margtn as a function of the power oscillation amplitude. His correlation is necessary to connrm the setpoints required for Solution III as well as the nonprotected region for Solution I-D.
. >. ~ BACKGROUND l Following the March 1988 instability event in the LaSalle BWR. the BWROG initiated a task
- o mvestigate actions that industry should take to resolve the BWR stability issue as an operational concern. Through analysis.) the BWROG found that the current plant protection system, that is based i
an a scram on high average power range monitor (APRM) signal. may not provide enough protection agamst out-of-phase modes of instability; thus, the BWROG decided that a new automatic instability suppression funcuon was required as a long-term solution and that this tunction should have a rapid I and automatic response which does not rely on operator action. He BWROG does not plan to solve the stability problem on a " generic' basis. but it has l proposed three different options instead. It will be up to the individual licensees to choose which l solution will be implemented in their reactor. The options currently being considered by the BWROG are: Exclusion Recion. A region outside which instabilities are very unlikely is calculated for each i representative plant type using well-denned procedures. If the reactor is operated inside this exclusion region. an automatic protective action is initiated to exit the region. Dis action is l l based exclusively on power and now measurements. and the presence of oscillations is not reouired for its imtiation. Four concepts of type I have been proposed by the BWROG: Immediate protection acnon teither scram or SRI) upon entrance to the exclusion 1-A region. I \\ Same as I A. but the exclusion region can be bypassed if a stability monitor is I-B operational and detecting sufficiently stacle conditions (for instance. decay ratio less than 0.6) \\ 1-C Protection action is taken if two conditions are satis 6ed: (I) the reactor is operatmg inside the exciusion region (dedned similarly as m Solution I-A). A ND (2) an APRM oscillation tof small amplitudet is detected. 1-D Some small-core plants with tight inlet ori6ces have a reduced likelihood at out-ot-phase instabilities. For these plants it is claimed that the existing tiow-biased high APRM scram provides surncient protection. In addition. administrative controls are proposed to maintain tne reactor outside the exclusion region. 2 l
I I \\' l ? I 11 Ouadrant-Based APRM Scram. In a BWR/2. se quadrant-based average-power-range rnonitor is capable of detecttng both in-pnase and out-of-phase oscillations with surficient sensitivity to initiate automatic protective action to suppress the oscillations before safety t margins are compromised. Ill LPRM-Based Detect and Suceress. Local power range monitor (LPRM) signals or combinaions of a small numoer of LPRMs are analyzed on-line by using three diverse algorithms. If any of the algorithms detects an instability, automatic protective action is taken to suppress the oscillations before safety margins are compromised. T $ifferent options i have been considered by the BWROG: Solution Concept Ill, and Solution Concept III-A. The main differences between the two is in the hardware implementation: Solution III requires a new Class IE computerized system, and Solution III-A may use newly designed digital replacements of the APRM amplifier cards that will require a smaller number of LPRM detectors. Conceptually, the algorithms are (or may be) similar in both solutions. CONCLL*SIONS AND RECOMMENDATIONS Positive conclusions Overall, the BWROG has done an excellent job of addressing the stability issue in operating L reactors. He BWROG has recognized that a problem exists. and they have attempted to l solve it in a technically competent manner instead of performing analyses that would defend macu'on. 2. Tne three proposea solution types tl II. and IID are technically sound and, in our opinion, any of them will solve permanently the issue if pro,.erly applied. 3. The solutions can be implemented in existing reactors in a relatively straightforward manner without compromising their intended function. The analyses techmques proposed by the BWROG in their licensing metnodology appear to be A sufficient to verify the effectiveness of these solutions in the lead plants. l The proposed BWROG procedures to generate input data for exclusion region calculations i 4 appear to be sut5cienuy conservative enough. Even though these procedures do not call for j l absolutely bounding values for all parameters. the conservatism is derived from the fact that reasonably bounding values are used for all parameters at the same ume. In the real world, forcing one parameter towaras its bouncing limit is incompauble with having other parameters at their limit. He conservative nature of these procedures is veritied through the use of transient contirmatory analyses under expected operating conditions. which include startup. pump runbacks. and loss feedwater conditions. The application of Solution 11 to Oyster Creek has shown that the quadrant-based APRM 6. scram provides sutheient protectica for either in-phase or out-of-phase oscilianons m a BWR/2. 3 J
f e i Die three proposed algorithms for Solution 111 appear to be able to detect oscillations in a 7. manner reliably enough that automauc suppression action can be taken by the protection g j system. The good detection sensitivity of the period-based algorithm allows for fairly tight scram setpoints: therefore, this solution does not need to rely heavily on difficult calculations l to show sufficient safety margin under a wide range of conditions and fuel types. i f The arguments presented in NEDO 31960 about the expected oscillation modes are S. convincing, although they are not absolutely bounding in the case of the single-channel l oscillation. We acree tnat the most likelv oscillation modes will be either in-phase (or corewide) or out-of-phase (or regional). " Higher order regional modes are"not likely, becai ~ l of their increased eigenvalue separation, We however have some minor reservations with respect to the single-channel type of instability (see reservation 10 in the next section and j l r recommendation 12). i l Resen ations I i Even thougn there are only three general types of solutions (Solutions 1. II. and III). at least L l seven possible implementations (Solutions I-A. I-B. I-C, I-D. Ii, III, and III-A) have been proposed at one time or another. A more general type of solution that would apply to all l reactors would have been preterable. Some solutions (especially Solution I-A. regional exclusion with scram upon entry to the 2. region) will most probably be implemented with margins as tight as possible to avoid unnecessary scrams. This approach might result in cycle-dependent implementations that would require new safety-system setpoints based on cycle-specisc data for each reload. This is not a desirable feature in a long-term solution. I 3. Solution 1-B (exclusion region with bypass based on stability monitor) has not been developed l in detail. and it appears to have been abandoned by the BWROG. If solution I-B were stil! l l under considerauon. we would have reservations with respect to the ability of stability momtors to measure the decay ratio or the out-of-phase instability mode with sutticient accuracy to a' low a bypass of the exclusion region scram. ' For example, in the Ringanis-1 l tests.' the measured decay ratio was about 0.7 at 70% power and an instability was obsened l at 72.5% power. This event clearly casts a shadow on the viability of Solution I-B as an l l l option. t Solution I-C (delta APRM flux scrams has not been developed in detail by the BWROG. If some licensee would want to pursue this option, we would have to look in more detail at the l A scram setpoint. The methodology used to estimate this setpoint should be similar to the one used for Solution III. including uncertainties and failed LPRM signals. E Solution I-D (small cores with tight inlet ori6ces s relies too strongly on decay ratio calculations that predict that the oscillation mode is very likely to be corewide. In this j solution. the tiow-biased scram does not appear to give significant protection against out-of-l phase instabilities should they occur. Although these calculations will be documented in Supplement 2 of NEDO 31960 (due sprmg 1993). it is expected that there will be an area 4 i I t I i ,-,n - n..n,, -,..+, n ,n-
.s within the exclusion region wnere the flow-biased scram does not provide protection for out-of-phase oscillations. 6. We have some concerns about the methodology to estimate the stability of the out-of-phase (or regional) mode of oscillation. The BWROG proposes to use an acceptance region dermed in a two-dimensional plane with the FABLE-calculated corewide and hot-channei decay ratios as coordinates (see Fig. 5-1 of reference 2. NEDO-31960/S1). The applicability of this acceptance region to determine whether a reactor condition is likely to oscillate in-phase or out of-phase may impact the approval of Solution I-D. The two main cogrns that we have about the methodology that detines core-channel decay ratio acceptance criterra are: 6.1 Core-channel decay ratio acceptance criteria were developed by using test data and other calculations. In all these benchmark cases the actual radial and axial power shapes were used with FABLE to estimate the core and hot-channel decay ratio. The BWROG. however, proposes to distinguish between in-phase and out-of-phase oscillation modes based on this acceptance criteria but using the conservatively detined " procedure" power shapes as inputs instead of the best estimate shapes. Although we agree that the procedure power shapes result in a conservative exclusion region. they may bias the results towards the in-phase mode of oscillation by using nonconsistent axial power shapes (tlat for corewide and extremely bottom peaked for the hot channei). In summary, the data base used to develop the acceptance criteria do not envelop the conditions for the intended use (i.e.. cannot distinguish accurately between in-phase and out-of-phase modes). The out-of-phase mode of instability is a function of how strong the flow feedback is. 6.2 and that is represented qualitatively by the channel decay ratio in the acceptance l criteria. However, the out-of-phase mode is also a function of the eigenvalue l separation between the fundamental and first azimuthal neutronic modes. He eigenvalue separation is not included in the acceptance criteria, which represent only " typical" loading y wrns and core sizes. It is conceivable that other loading patterns might result in differe w acceptance criteria. Reducing the number or false positives ti.e. scrams when it was not required; for Solutmn ill 'LPRM-based detect and suppress > is crucial for the solution to work; however. the BWROG may take this false-scram avoidance to such an extreme that the solution will not work. Minor problems with electronic noise, controllers out of tune. or many other unknown i parameters may result in failure to scram when required. if this solution is not carefully l designed. His is the reason we recommend (as proposed by the BWROG) that several diverse algorithms be implemented simultaneously for Solution III. 5. Solutions III. I-C. and I-D depend partly on a correlation that relates the change in critical power ratio (CPR) caused by a neutron power oscillation. It is not clear that such a correlation exists or how many independent parameters it must contain. The BWROG has been working towards developing this correlation and is trying to detine it in a ecmservative manner. BWROG expects to complete this correlation development in February 1993. He correlation documentation will be included in the Supplement 2 to NEDO-31690 that is expected in the sprmg of 1993. 5 l
e The applicability of the delta-CPR correlation (see paragraph 8 above to new fuels or fuels 9. from different vendors is not clear. This pomt is being addressed by the BWROG. and a formal position is expected in Supplement 2 to NEDO 31960. Reactor operators have a large degree of freedom to chose control rod patterns and power 10. distributions during startup at low powers. Some of these " achievable" power distributions may result in instabilities outside the exclusion region. even if the reload condrmation procedures were successful. Criteria must be set by the BWROG to assure the operator that the reactor is within the limits where the Solution I exclusion region is appycable ~ Under normal conditions. single-channel instabilities are not probable because these 11. conditions are likely to induce an out of-phase instability before the single-channel instability develops. This argument, however. is based on the fact that many channels of the same type are loaded, and therefore if one channel is close to instability, many channels will also be unstable and are likely to produce a global out-of-phase oscillation. This is not the case. however, with lead use assemblies (LUAs). where perhaps only one channel of that type is loaded, if this LUA had stability enaracteristics quite different from those of the rest of the core. a single-channei instability in the LUA could be possible. For this reason we are recommending that the thermonydraulic stability of all LUAs be determined (see recommendation 13). I Recommendations Approve the overall exclusion region calculation methodology as described in NEDO-31960 1. and its Supplement 1. The results of these exclusion region calculations may be used as part of the implementation of Solutions I and III. Approve the overall treatment of uncertainties described in NEDO-31960 as it applies to the 1 selection of initial conditions for exclusion region calculations and its connrmatory runs. Pending review of the spect6c reload condrmation procedures tnat should be outlined in a f i l tion in any 3econd suppiement to NEDO-31960. approve Solution Concept 1-A or mp ementa BWR line with the following design 00jectives: Specide reload condrmanon procedures must be developed so that for every reload. 3.1 the licensee can either t1) condrm the applicability of old exclusion region settings, or i
- 12) set a new exclusion recion boundary.
i Favor implementations of Solution I-A that are not expected to change the exclusion 32 boundary setpoints on a cycle-by-cycle basis. Connrmatory calculations should be expected to be positive most of the time; major setpoint changes should.uy a expected following signi6 cant tuel design changes. Select rod insert (SRI) may be used in conjunction with Solution I-A. but a full scram 13 must take etfect if the reactor does not exit the region within a reasonable period of time tot the order of a tew secondst. 6
j e Do not approve Solution Concept 1-B at this time. Solution I-B has not been developed in A detail by the BWROG. If a licensee chose to implement Solution I-B they would have to resolve the question of whether a noise-based stability monitor can provide adequate protection against instabilities in the out-of-phase mode. Do not approve Solution Concept 1-C at this time. Solution I-C F.as not been developed in 5. detail by the BWROG. Do not approve Solution Concept 1-D at this time until had plant connrmgton analyses are 6. performed. documented. and reviewed. It is expected Jiat Supplement 2 to NEDO 31960 (due spring 1993) will contain consrmation analyses for the Duane Arnold plant that will allow a detailed review and a final decision on the ecceptability of Solution Concept I-D. This recommendation does not imply a rejection r ( Solution I-D: the approval of Solution I-D i l depends on the details of calculations that have act yet been performed by the BWROG. Approve Solution Concept 11 for implementi on in BWRs with quadrant APRM scram ti.e. i 7 any BWR/2). Oyster Creek has already su';mitted technical specification changes that implement this solution (see Ref. 5). Approve Solution Concept 111 for imp'ementation in any BWR line with the following desigl 8. objectives: To avoid unexpecW problems. several diverse algorithms must be used to detect 6.! oscillations. Automatic protection action must initiate if either of the algorithms detects oscillations (i.e.. the algoritnm outputs are connected by a logical OR not a l logical AND). The three algorithms described in NEDO-31960 and its supplement may be used in 8.2 Solution 111. These three algorithms are t1) high LPRM oscillation amplitude. (2) high-low detection algorithm. and (3) period-based algorithm. Preferably. all three algorithms should be used. The licensees that implement these algorithms must demonstrate by analyses the 8.3 validity ot' the scram setpoints selected. These analyses may be performed on a representative-plant basis when applicable but must include an uncertainty treatment l that takes into account the number of failed sensors permitted by technical I _7ecifications. The scram setpoints should be selected such that at least one of the algorithms has a SA large probability of detecting the oscillation and initiating protective action to prevent violatan of any fuel safety criterion. If the alcorithms detect oscillations. an automatic protection action must be initiated. 3.5 This actn n may be a full scram or an SRI. If an SRI were to be implemented with Solution (.oncept 111. a full scram must take effect if either (1) the oscillations do not disappear i1 a reasonable period of time. or (2) the reactor remains inside the exclusion n tion as detined by the general re;ional exclusion methodology of Solution I-A. 7 l
1 The LPRM groupings denned in NEDO 31960 to provide input to the Solution III algorithms 9 j appear appropriate for the intended oscillation-detection function. These LPRM groupings are: the oscillation power range monitor (OPRM) for Solution III or the octant-based arrangements for Solution ill-A. The minimum requirements for operable number of LPRM detectors set in NEDO 31960 appear appropriate. Approve the overall treatment or uncertainties described in NEDO-31960 as it applies to the 10. selection of oscillation contours and failed LPRMs for Detect and Suppress Concepts rSolution III). Implementation of Solution III will require the documentation of the selection of the bypass I1. region outside which the detect and suppress action is deactivated. The BWROG must establish a criteria to limit the actual radial and axial peaking factors
- 2.
during startup operations to those values considered for the analyses of the exclusion region. This criteria must be based m parameters or information readily available to the operator in the control room. Defining inis crneria must be pan of the reload confirmation analyses. Establish a precedure to review the thermohydraulic stability of lead use assemblies (LUA). O Solutions I and 11 do not protect the reactor in the case of a single-channel instability, and the protection for Solution Ill is limited. These instabilities are not likely if many bundles of one i type are loaded in the core, but they could be possible if the wrong type of LUA were to be ^ loaded. Thermohydraulic stability analyses must be required during LUA review if Solutions i or 11 are used. REFERENCES i Genera. Electric Company, B%R Owners' Group Long-Term Stability Solutions Licensing 1. Methodo!ory NEDO-31960. May 1991. 2. General Ele:tric Company. EnR Owners ' Group Lone-Term Stability Solutions Licensing j Methodoloo NEDO-31960 Supplement 1. March 1992. i l 3. General Elects c Company. Fuel D:ermal Margin Durine Core D:ermal Hydraulic Oscillations in 1 Boiling %'ater Re.ictor. NEDO-31708. June 1989 B-G Bergdahl a td R. Oguma. "BWJ Stability investigation in Ringhals-1 Measurement Data A trom October 21.1989.' Proceedines of Die International Warnhop on Boiline Water Reactor Stabiliti, Holtsville N. Y. 17-l9 October i990, pp 142-159. OECDINEAiCSNI Report No 178 October 1990. Oyster Creek Suclear Generating Station. Technical Specification Change Request No.191. 5. Docket No..' 0-219. October 9.1991. 8 l l
.~ \\ i l APPENDIX A LAPUR AUDIT CALCULATIONS OF Solution i EXCLUSION REGION CALCULATIONS AUDIT r ALCUL ATIONS A series of audit calculations were performed with the LAPUR code to vgify the 'results presented by the BWROG that were based on FABLE /BYPSS calculations. All relevant input data used in the FABLE /BYPSS was made available for this review. and we set up LAPUR input decks + that were representative of the conditions modeled by FABLE. The main result of these calculations is presented in Table A.1 and Figs A.1 and A.2. We observe that the maximum difference between l LAPUR-and FABLE-calculated decay ratios is 0.09. This can be considered as excellent agre::. at and representative of the differences in modeling of both codes. This type of code-to-code benenmark is not as good as a code-to-data benchmark, but it assures tnat gross modeling errors or systematic errors in the preparation of the input decks have not occurred. Furthermore. it ensures that " data tudging' is not taking place to obtain desired results. because all the data has to be made available and is evaluated for expected value ranges. Both codes. LAPUR and FABLE have been benchmarked against data from actual stability tests with satisfactory results. In general, it is recognized that this type of frequency-domain codes has an accuracy better than 20%. Rus. if a decay ratio of 0.8 or smaller is calculated, it is highly probable that stable reactor operation will result. Decay ratios larger than 0.8 result in smaller probabilities of stable operation. Note, however, that large errors are possible if the proper data are not used as mput to the code. He 20% error quoted above is for detailed test benchmarks where extreme care is taken to reproduce the exact axial and radial power shapes, core pressure drops, and reactivity coerficients: calculations using approximate descriptions of the core operating condition are likely to result in larger errors. The axial power shapes assumed in tne BWROG analysis are: (1) fairiy uniform vend-ot-cycle Haling; snape to calculate tne corewide decay ratio, and (2) strongly bottom peaked (2.0 peaking at node 3'24) to calculate the hot-channel thermohydraulic decay ratio. It is well known that the high-power enannels (maybe 25% of the total number of channels) have the most influence in the stability l or the reactor. This is due to the fact that the adjoint 11ux and density reactivity coefficients are higher in the high-power channels. Furthermore. hot channels tend to have bottom-peaked power snapes. that may be more unstable. To test the validity of the uniform power shape assumption. we l ran two cases to determme corewide stability boundary: (1) with all channels having the same Haling power snape and (2) with a graded axial power shape so that the hot channels have a bottom-peaked (nape (2.0 at node 3/24). but the core average is the same as in case (1). The chosen power shapes are drawn m Figs. A.3 and AA for a BWR/3 and BWR15 respectively. The out-of-phase and hot-nannel decay ratio calculations were based on the graded power shapes of Figs. A.3 and A A. The out-of-phase decay ratio was calculated by LAPUR assuming an eigenvalue separation of St.00 between rundamental and first harmonic neutronic modes. This St.00 value is a representative, but j not bounding value of the eigenvalue separation. These results show that the uniform (Haling) power shape is more conservative at lower 110ws. but the use of bottom peaked graded shapes resuits in higher decay ratios at higher flows. 9 \\
Reload Confirmation Procedures he main technical issue of signi6 cant relevance that still remains to be solved is the reload-dependent con rmatory analyses required to assert the applicability of the previous-cycle safety settings and, in particular, the applicability of "old" exclusion regions to new types of fuel and loading patterns. He BWROG is aware of this problem and is currently developing a methodology tor these cycle-dependent connrmations that is based on a " response surface" approach. The goal is i ' hat these condrmatory calculations should be expected to be positive most of the time; major setpoint i changes should only be expected following signi6 cant fuel design changes.. The documentation of this reload-condrmation methodology is expected in Supplement 2 to NEDO-31960 tifai'should be puolished in the spring of 1993. Of particular concern is how the reload procedures will be used to evaluate startup power i l distributions. For example. the root cause of a recent instability event in a BWR/5 has been determined to be the extreme radial (1.92) and axial (up to 1.87) peaking factors during the startup. This extreme power distribution was apparently not covered by the standard exclusion region calculations. which assumed a more mild radial power peaking factor. Nevertheless. the operator was alowed to have that extreme distribution without violating any thermal limits. Figure A.5 shows a comparison of the equilibrium-cycle exclusion region for the Perry reactor (a BWR/6) and the exclusion region that results if the actual axial and power shapes from the recent BWR/5 event are used. As it can be observed. the standard BWR/6 exclusion region from l l l NEDO-31960 is not as conservative as the actual region. Therefore, we have recommended that the BWROG must establish a criteria to limit the actual radial and axial peaking factors during startup operations to those values considered for the analyses of the exclusion region. This criteria must be based on parameters or information readily available to the operator in the control room. i i 10 1
e Table A.I. LAPUR-FARLE!BYPSS benchmark / audit calculations j i i Corewide decay ratio Hot-channel decay ratio l Reactor Power Flow FABLE LAPUR FABLE LAPUR l tvne 4 e,) t e,) BWR/3 42 l30 l 0.77 0.68 0.34 l 0.28 l 1 52 l45 l 0.46 l 0.47 0.19 0.17 71 45 0.65 0.64 0.35 0.38 84 60 0.45 0.41 0.21 0.28 B W R/5 42 l30 l0.65 0.73 0.50 0.50 52 l45 l 0.39 l 0.37 0.34 0.32 i 71 { 45 } 0.56 l 0.50 0.55 l 0.62 l i i 84 l 60 l 0.30 l 0.32 0.31 l 0.39 l l i I l I l l 11 i i
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== f een E 9 8%.n m m' en ~ C O L e w~m ? O m ea ra "4 y ) 9 j M M f e m m a i ee m 5 .= L i m w w l l w rs m t a. w .t
===. h m eg = 4 Ta-O 1 m m' s=te W. m es m =
- Pe M
n. 20 wm g w end. W-Allb [ d8'8 e. g ou,t Sampg w P. lg. Vl A $= dubi w e w w a me h
- 7 **'
f N m'" O ~= ~ m. 4 .=
==- .m QC o ~ m. m El#. g" 'T O' C m _u.d m w = ea h O
- E m.
ami-Q .= m. m
- ?
A A W 8 m c - C w w 4 - rs Z = O c"e) Qw - o W = er e w' m. asw PR. sp=* w mum n .rn m p a.= m, ~: e Pg am /** 2 = en S't W gar e, e-rg. c my (. WER w af w g m
== W .sd A w W II D so w 8 o C j 6 N N m O e ge e.g p es' p
== e C 2 g. m s** es mw m d'k gem P$ ~ f b.= N.= = f8 P8
- 0
- g
.==,
- 7, m
.y c ** 4 5 10 M I S
v i i t j l ~ 1 I Table A.6. LAPUR calculations for a typical BWR/3 f (DR = decay ratio NF = natural frequency of oscillation) Corewide* Out-of-phase Hot-channel Corewide" l Flow Power (Mib/h (MW) NF NF NF NF } DR (Hz) DR (Hz) DR (Hz) DR . (Hz) l l 10 l 800 l l 0.93 l 0.31 0.80 0.27-1.13 0.31 l 20 l800 0.82 0.26 0.36 0.34 0.32 0.29 0.64 0.34 20 l 1000 l 0.65 0.38 0.48 0.34 0.85 0.38 l 20 l1200 l l 0.95 0.41 l0.67 l 0.37 - 1.07 l 0.41 l 29.4 l1060 l 0.68 l 0.32 l0.28 0.41 0.28 0.36 0.50 l 0.42 l l ! 29.4 l 1200 l 0.79 l 0.34 l l l l l l ! 29.4 l1500 l l l 0.79 l 0.50 0.55 0.45 0.86 l 0.50 f l 13% 0.47 0.37 l0.14 0.48 l0.17 l0.42 l 0.29 l 0.48 l 44 l1783 l0.64 0.44 0.45 l0.59 0.38 0.54 0.51 l0.58 f 44 r i l2000 0.69 0.43 0.66 0.62 0.49 0.57 0.64 0.61 44 l 2500 l0.85 0.51 l1.05 0.67 0.84 l 0.62 0.98 0.66 f 44 50 l2500 l l 0.82 l 0.70 0.59 l0.65 l 0.73 0.70 50 l2600 l l !0.85 l 0.70 l0.66 0.66 l0.80 0.70 50 l3000 l 0.79 l 0.57 l [ l l l 58.7 l2109 l 0.41 l 0.50 l 0.22 l 0.68 ! 0.28 ! 0.66 l 0.27 l0.66 l 58.7 l3000 l l l0.64 l0.84 l 0.50 l 0.77 0.51 l 0.81 58.7 l 3:00 l l l0.82 l0.85 l 0.59 l 0.79 l0.62 0.84
- Using average axial power shape for all channels.
" Using graded power shapes of Fig. A.3. 18
l l Table A.7. LAPUR calculations for a typical BWR/5 (DR = decay ratio. SF = natural frequency of oscillationi Flow Power Corewide' l Out-of-phase l Hot-channel Corewide" f (Mlb/h (MW) ) NF NF NF NF DR (Hz) DR (Hz) DR i (Hz) DR (Hz) l l 10 l500 l 0.76 l0.17 l 0.52 0.22 0.74 - l 0.20 ha35 l '0.22 10 600 1.06 0.I8 0.74 0.24 0.96 0.24 20 800 0.76 0.24 l 0.49 0.31 0.40 0.27 0.65 0.31 l 20 l1000 l0.96 l 0.27 l0.84 0.35 0.63 l 0.31 0.93 0.34 32.5 l1396 l0.73 l0.35 l 0.68 0.46 0.50 l 0.42 0.72 l0.45 ! 32.5 l1500 l 0.78 l 0.36 l0.83 l0.47 0.57 l 0.43 l0.82 l 0.46 l 32.5 l1600 l0.83 l 0.38 l l l f r 48.7 2200 l l 0.72 0.63 0.53 l 0.59 0.57 0.61 f l2360 0.50 l 0.49 l0.87 0.65 l 0.62 l0.60 0.69 l0.63 f 48.7 l0.62 0.80 0.64 48.7 l2500 l 0.55 0.50 l 0.99 l0.66 0.70 I 48.7 3000 l 0.71 l 0.55 l 48.7 l3500 0.94 0.60 l l 3 l 0.68 .l 0.69 0.51 l 0.65 0.49 0.67 55 2500 l i 55 l2700 l l l0.84 l 0.71 l0.61 l0.67 0.62 0.69 l l 55 l3000 l 0.50 l 0.57 l1.07 l 0.75 l 0.78 l 0.70 l0.84 l0.72 l [ ! 55 l4000 l0.82 I0.66 l l l l l l l 0.74 l l 65 l2791 l 0.32 l 0.58 l 0.32 0.78 0.39 l 0.75 0.23
- Using average axial power shape for all channels.
~ Using graded power shapes of Fig. A.3. l l l 19
m r.a> a 1 + o O 4 0 N ) \\. i i i i 1 l 6 0 m i er. =w i, < >W y , i 8 m w M l W W \\ = 4' i O s \\ + see ,,s l -e;s = N i i e me. m y M m I f e M W ~ O ~ .J ~ ~ a mm> 7 i ~ N W a N w e f 4 I N sE I l ~ \\ 7 a .I w \\. t I N 7 A.4 3 M C j W 7 N
- J l
T M 7 g s 1 1 ,4 .s A
- )
= 5GS M C "J t "J s -s m 4 I e emW Wem h / se en a s1 m M M a W l l e e e h I l 1" l ese M g h N N q g 99* w d d J l w = ~~=
- s m
~ =s
- 4
. *i, * "'*, i i i
I s 4e e = I' W s s 2 -* l 7 ~ ~ p W hN%e \\ < f C
=
m ^ s, .Y... ) = W
- uun=~
hur e M Ys M* O w, e r =M 4 ~ eu v [ s
- =
I ~ O A I 3 ~ g "b = l' i N ~ ~ 4*9 +t \\_ % O v.+ 1 i ? = ~ 'J
- J o
e*e S + 4 -m b 7 e.e, E 9 N F4. ~ 4 5 e s~ N = =o W
- IS e#
~ = = A' 4 N Y ( = m,,
e 1 l I Y 1 1 O. e i l i l 1 4 s I t + t 1 i a m ! w 1 .x.e U, N i 4 4 e >>s. i 3 l ~ \\ s w.m ( 4 M e m ts g 's 'w$t j t = s i "'3 I s i \\ 4 I t M**= i M i I j j 1 i ~ l = h i s ~ ~- l 7o W, W i s l _3 J s 6-am, 4 .m Y M 1 m ,g i -sv i o i r i w
- ll b
i w i Q I d O i 's I ++- M =t -mas N g ( .2 1 a 4 M + \\.m% G,e-m
- )
~ s / emM e M -uD temp 6 ..J99 Wup ane om 4 - M ii, f M C_ v.s g F I 6.
== elr*A
- =
ed.e 2 M 9 'm M O J w e w a a A.n.R M M R.md W .m e.m.> g V m -..
1 1 i a. l 1 7 g 'J O 3: ~ ~ ~ ~O' .I 1 J '.T m N i 1 1 1 1 N ~ .>h }. ~ C 9 k, m m e" ~ -+- + w d x emus T .O m ff' ' ~. -f .w o -s .d. 2, + m / ~ ps fI,. += +< .n y a., , s j ^% 8 s 9J s D 8 T O ,/ s / ,,, [ + = 3
- )
-.. +- 2; + + q x f j .~ m 3 _r s f. A, Y, 3 N ...&' ~... ..e,.... ...w + ~~% N i ( j s /*\\ ' ,J. g ) M / .s s, ~ I h =. .V.
- /
N, \\r = g .g 'l . s amo am y =_ s
===,es. / see ...m m M ~ .r. i = .~ /. x . s 'Y ~ N ~ m. .m '}. e +7 'p. / N A -e-- s /...,....#, ,...w.., av .mp s s n M M g / y ,F" ?% ~ s 'r, s 5_ Nx _ a.a. W ens a a. es. 7*
- =*
M
- d e=
m W +4 i e snap emmy m
- um ene W
W ' enf W W p'
- alFII,
y '*'*a9 euded w p emmiums e m emmer
v-4- o i 9 t 9 t ) h ~4 i O ? 3 D w .o .i a 1 1 Z 4 1 1 1 1 ![ g =>ew N y . ~ a y .L i i l l } ^ C -s .C_ ^ tN .Z s'~ g +
====.s -~ r r + P.' 1, ,\\,7 o.s . /. - V, 8 T' 7_ aj 9 .s /k y ' + = M. + -......... -......< D -f; t 4 = ? = Q ); f ? s' t 2.( s ' 4 ~ p ,g' e, m
- 4 '.I
.S.. Ti e s f I a f g 71 L' l Ej ...wJ%. L..- 7 '/ j ..L...,... s u J~.+ = l k
- 4 s
1 5 I f' f (
- f. /
A;. m, = ,( m .e N$ N = .io = J y J. m W-3 [g f#. '~' N .&,. %b o ~ l 7 g a l r m f. r- ++ e..- w /, b i f => f / s n. .? .-f......r..... .. Y' '. l y e 4- [ 1 / r 'n 4 ,? s.- .s' r m 4.wa \\ \\, \\.. g o s U n iP h C_ i ^. 'D ~~ rJ ~ 2 -7 ^' J g m W W / r / m -.a. i I L w --w:r ,qps -a-.- w-g -.y .gw-- .w
I C@r r'j EO-C Cle / ,/, 3 ...... ExC! Ceg. ,/ \\ . >w / / ./ i ./ _.f... ...._...q n 7 /I i t / b
- nst:0
.y ,/ CO nC i *. On5 /
- n L _
i ] 1 / g .v ] / ~ /,/ s' l } / / L APUR-C 1cuiatea .. - - - -Peg -Or j -. ~... ~p . /.. Exct ~o .ctuat Conc s j 'y '
- _-- > L.-s i
f -_. m -- i t i i ( -n cu 20 30 40 50 50 i ~ L a h'
- ~o)
Figure A.5. Comparison between ecuilibriurn-cycle exclusion region and the exclusion regien ter the speciti operaung conditions ot a BWR/5 instability event. 24 I I
l i W i l 1 APPENDIX B Solution I-D REQUIREMENTS l l l Ris section de6nes a minimum set of items that will have to be provided to complete our review of Solution I-D. Rese items should be provided for the initial Solution I-D fead plant (Duane j Arnold). t Describe how the exclusion region for administrative control actions will be calculated and 1. dedned. Desenbe in detail the proposed administrative control actions if the reactor enters the 2. exclusion region. 4 Describe any hardware or administrative control rod block functions that will be associated i 3. with the exclusion region. Speci5cally, describe how these functions are calculated and l denned and what type of automated or operator action is required. Describe in detail the information that the operator relies on to provide these administrative 4 controls. In particular. describe how the information is presented to the operator and its " safety classi6 cation ~ (i.e. Class 1-E or not). Explain why this safety classitication is adeau' ate. 5. Describe what indications the operator would have in the control room if a power oscillation j (either in-phase or out+f-phases were to develop. Describe the operator actions required under these circumstances. Provide analyses showing the area inside the exclusion region where the tiow-biased scram 6. does not provide protection for out-of-phase instabilities. These calculations determine the nonprotection line, which is de6ned as the line in the power-flow map below which the flow-I biased scram does not provide automatic protection. Two lines must be de' Sed: The nonprotection line at the 954 probability level with the initial CPR at technical 6.1 speci5 cation limits. He nonprotection line at the 50% probability level with the expected initial CPR. 6.2 Provide reasonably bounding analyses showing that oscillation; it. tire out-of-phase mode are highly unlikely in Solution I-D plants operatmg below the 50% .el nonprotection line. Rese calculations must be performed along the 50%-leve: nonprotection line and include at least the followmg cases: .1 Calculations of core and hot-channel decay ratios using the standa*d BWROG procedures for exclusion region calculations (NEDO 31960). These calculations must j show that the core decay ratio is significantly larger than the hot-channel decay ratio j so that the predicted mode of oscillation for these conoitions is in-phase. Provide documentation of the racial power distribution (in particular the hot-channel peaking i ~-.
factor) used in these calculations, and justify why the chosen peaking factors are conservative. 7.2 Calculations of core and hot-channel decay ratios using conservatively defined bottom-peaked power shapes that are more representative of startup conditions than the standard BWROG procedure. Dese calculations must include axial and radial power shapes representative of (1) normal startup and (2) operation with failed feedwater heaters. Document the actual power shapes used and justify their conservatism. t LAPUR CALCULATIONS RELATED TO Solution I-D t A series of calculations have been performed with the LAPUR code to contirm the validity of the BWROG claim that small cores with tight inlet ori6ces are not likely to have out-of-phase instabilities. He results of our anwyses show that indeed (as claimed by the BWROG) small cores and tight inlet ort 6ces are benencial for the out of-phase mode. However, this benedt does not appear to be sufficient to completely discard the possibility of out-of-phase instabilities in these types of reactors: therefore we have requested that the BWROG perform the calculations described in the preceding section. Table B.1 shows some of the results of these analyses. t Effect of tirbt inlet ori6ce For the calculations presented in Table B.I. we prepared a representative LAPUR input deck ishown in Table B.2) with a single thermohydraulic region and calculated the corewide and out-of-phase decay ratios as a function of the inlet restriction coetficient to simulate the differences between Solution I-D plants and others. In plants where solut.on I-D is applicable, the inlet restriction coet6cient is of the order of 35 to 40 velocity heads. while other plants have values of the order of 25 i to 30 velocity heads: for example. Duane Arnold (the proposed lead plant for Solution I-D) has an
- niet ori6ce diameter of 2.09 inches. compared to 2.43 inches for LaSalle. We have to note inat the conditions (especially the axial power shaDel chosen for these analyses are not representative of l
normal operation. but they are achievable and not necessarily bounding; these conditions were chosen because they tend to excite the out-of-phase mode more than the corewide. Two main conclusions j can be crawn from the results in Table B.1: (1) ne smaller inlet ori6ce by itself does not preclude the possibility of out-of-phase instabilities. For example, at 35 velocity heads, the out-of-phase mode is predicted to be unstable (decay tano greater than 0.8) even at large eigenvalue separations of $1.5. Q The smaller ori6ce by itself does not guarantee that the corewide mode will dominate and become unstable before the out-of-phase mode does. For example. at 35 velocity heads. the out-of-phase decay ratio is 0.90 at 51.0 suberitical, while the corewide decay ratio is only 0.84. In summary, even tnough smaller (tight) inlet or:6ces are bene 6cial and tend to stabilize the out-of-phase mode. increasing the ori6ce coerncient by about 10 velocity heads reduces the out-of-26 i
l a t I I I phase decay ratio by only 10% to 20% depending on the actual circumstances. Therefore, tight inlet orince plants are less likely to have out-of-phase instabilities, but given that it only results in a 10% j to 20% reduction, this erfect by itself is not sut5cient to preclude out-of-phase instabilities. i i Effect of smaller cores Smaller cores affect the stability of the out-of-phase mode by increasing the neutron leakage on the core periphery. Larger leakage rates tend to increase the eigenvalue separation between the tundamental and nrst azimuthal harmonic; the larger the separatir. the more sta6Te'the out-of-phase mode (see Table B.1 for an example). Our evaluation analyses tu.ng the LAPUR code indicate that the net effect of reducing the core size in half is to reduce the out-of-phase decay ratio by 10% to 15%. His evaluation assumes constant loading patterns and fuel types: positive or negative changes of larger magnitude can be achieved by altering the loading patterns or fuel type. Herefore, we conclude that the net effect of the small core size by itself (although beneticial) is not sufficient to preclude out-of-phase instabilities in Solution I-D plants. In tirst approximation tassuming a homogenous, cylindrical cores, the eigenvalue separation of the first azimuthal mode, p,. is given by 2 p* MB (B 1) vEf where D is the diffusion coef5cient, rE,is the fission cross section. and AB' is the difference in ~ geometric buciding between the fundamental and the first azimuthal modes. The geometric buckling in a cylinder is approximately proponional to the inverse of the radius square and, therefore, is somehow inversely proponional to the number of bundles in the core. Consequently, if core A has half the number of bundles as core B, core A should have approximately twice the eigenvalue separation of core B. From Table B.I. we observe that doubling the eigenvalue separation results in a reduction of decay ratio of the order of 10% to 15%. The eigenvalue separation. however. depends on many more parameters than just the core size. For instance, super low leakage loading patterns (SL'P) have very low leakage and result in signincantly lower eigenvalue separation than in a core the same size with a conventional loading l pattern. Another parameter that affects the eigenvalue separation is the tission cross section (see Eq. (B-1)]; therefore, fuels with high enrichment (to allow tor longer refueling cycles) should result in smaller eigenvalue separation that can negate the advantages of the small core. In summary, the core size is an important parameter that atfects the eigenvalue separation, but it is not the only one. It is, thus, hard to justify what the eigenvalue separation of a Solution I-D really is. t l 27 i
Oceratine emerience An additional argument against Solution 1-D is the fact that Swedish reactors (for example, Ringahls-1 (see Reference 5)] have experienced out-of-phase instabilities. Swedish BWRs have very tight inlet orifices and have relatively small cores (for instance Ringahls-1 has only 648 fuel l bundles t. .n Table B.I. LAPUR-calculated decay ratios as a function of inlet orince size Inlet restriction coefficient (velocity heads) 25 vh 30 vh 35 vh 40 vb 45 vh 50 vh Average-channel decay ratio l0.62 l 0.57 0.52 l 0.48 0.44 0.41 Corewide moae decay rano l0.86 l0.85 0.84 l0.83 0.82 l0.81 l l Out+f-phase mode p = - 50.5 1.08 1.02 0.96 0.92 0.87 0.83 i decav ratio. if o = -51.0 1.06 0.97 0.90 l0.83 0.77 0.71 eigenvalue separation IS p = - 51.5 1.01 0.90 0.81 0.73 0.66 0.61 l t' l l I l l l l 28 I
e, e t I,ible18.2 1.Al'llR5X input lin Solution 1-1) ana!yses I 102 (N 1 Al'ilH5X Xcs (nu los HW(llui 5,'l 1 i) 20 1 1 102 W 977 0. 490 0,1000 0, 20 lb. 0 0,0 0. 0 63, t 0 21 2 I i 36 1,24.O,l.O.O,0,I 0 22 3 1 01 1.25 23 4 I OI t $ 2 4, 15 24, 15 24, 15 24 IS 24 15 24, 15 24 15 24 15 24, 15 24 15 24 15 24 15 24, 15 24 24 8 43 15 24, 15 24, 15 24, 15 24 15 24, 15 24 15 24 25 15 24 15 24, 15 24, 45 53 1 0 125 5 26 0 95, I 60 I 80, 1 70, I 55, I 45, 1 30 1 20, 1.15, i 10 1.00, 0 95, 0 92. 0 90 1. 1 27 0 86, 0 83. 0 80, 0 78, 0 72, 0 67, 0 62 I 10 42 0 50, 0 40, O 20, 0 00 28 7 I I 44tn I, 1 29 9 1 0 5586 i, 7tv4 30 10 l 00373 1, 30 0 31 ll 1 0 0813 l.-0 280 32 13 I O1356 1, 0 33 14 1 0 0l14 l. 764 34 ^5 1, I 1, 62 35 I 16 i l I, I I 36 li 4 i i 29 i a t 29 37 88 1 40 I 218 90 53 89 29 1 1
t 'oc, 50 8 7 2 8 8 3 7 2 0 5 0 4 5 - 1 6 0 2 0 0 5 4 0 3 2 0 l 4 1 0 5 00 I 7 5 si 0 00 l 3 se 3 s a ly t 0 1 0 I r o a 8 5 u 2 1 0 I a 0 1 2 i 0 I 5 ) 2 6 1 l 6 0 O 2 i 0 I 0 5 o i 1 1 t l 1. 1 0 u 0 1 0 5 o 0 S 2 03 004 i 8 4 0 0 r 2 2 2 . 3 2 3 7 o f 0 t 3 up O i W 5 R U 8 l' l A. 5 0 l 4 4 1 i 0 i t 3 2 I 8 1 l 1 I 3 2 l e l 4 0 t b 0 1 a 8 4 9 T 5 l 1 4 6 4 i 0 3 i 1 I i 3 m 0 6 1 il ) 2 0 i 0 l ( I l 4 i 5 5 S E 1 8 i 1 7 G 6 0 l 0 l 2 O 9 0 R W 3 7 ) H i 1 L 1 0 1 0 1 r 1 0 ei 0 l l i es ) a 5 5 4 1 0 2 C I 7 I I u 9 I 1 1 1 I 0 1 E 2 0 9 E 1 5 t 5 0 2 2 c 2 9 1 I l W 6 1 4 I 0 3 ( 0 1 1 2 3 1 5 6 1 4 0 1 l i l o b b 4 l'. l I I 0 1 2 7 8 9 0 1 I l' 1 I 6 7 8 9 1 1 1 1 1 1 2 2 3 4 5 4 0 A. 1 5 1
v ORNLINRC/LTR-92/15 INTERNAL DISTRIBUTION 1. J. March-Leuba 12. B. ChexaL. Advis,og 2. B. Damiano 13. V. Radeka. Advisor 3. H. R. Brasbear 14 R. M. Taylor. Advisor 15-16. Central Research Library 4 B. G. Eads 17. Y-12 Technical Reference Section 5. D. N. Fry 18-19. Laboratory Records Department 6. D. W. Mcdonald 7. D. R. Miller 20. Laboratory Records-Record Copy 21. ORNL Patem Section S. R. E. Uhrig 22. I&C Division Publications Office 9. J. O. Stiegler 10. D. G. O' Conor 11. C. E. Pugh i EXTERN AL DISTRIBUTION 23. D. E. Bessette. NLN-353 NRC/RES. Washington DC 20555 l 24. E. S. Beckjord NLS-007. NRC/RES. Washington DC 20555 25. Amira Gill. WFN-8E23. NRC/NRR. Washington DC 20555 25. S. Fabic. MNBB-3701. AEOD/DOA Washington DC 20555 l
- 6.
W. S. Farmer. NLS-217B. NRC/RES. Washington DC 20555 27. T. L. Huang, WFN-8E23. NRC/NRR. Washington DC 20555
- 3.
R. C. Jones. WFN-8E23. NRC/NRR. Washington DC 20555 l
- 9.
T. E. Murely. WFN-1:G18. NRC/NRR. Washington DC 20555 30. L. E. Phillips. WFN-8E23. NRC/NRR. Washington DC 20555 31. H. J. Richings. WFN-8E23. NRC/NRR Washington DC 20555 32. B. W. Sheron. NL-007. NRC/RES. Washington DC 20555 l 33. H. H. Scott NLN-353. NRC/RES. Washington DC 20555 34. L. M. Shotkin. NLN 353. NRC/RES. Washington DC 20555 l 35. A. C. Thadani. WFN-8E23. NRC/NRR. Washington DC 20555 l 36. NRC Central File 37. Assistant Manager for Energy Research and Development. DOE-OR. P.O. Box 2001. Oak Ridge. TN 37831-8600 38-47. Office of Scientine and Technical Informanon. U.S. Department of Energy. P.O. Box 62. Oak Ridge, TN 37831 31}}