ML23151A556
| ML23151A556 | |
| Person / Time | |
|---|---|
| Issue date: | 07/25/1989 |
| From: | Chilk S NRC/SECY |
| To: | |
| References | |
| PRM-050-053, 54FR30905 | |
| Download: ML23151A556 (1) | |
Text
DOCUMENT DATE:
TITLE:
CASE
REFERENCE:
KEYWORD:
ADAMS Template: SECY-067 07/25/1989 PRM-050-053 - 54FR30905 - THE OHIO CITIZENS FOR RESPONSIBLE ENERGY: RECEIPT OF PETITION FOR RULEMAKING PRM-050-053 54FR30905 RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete
PAGE 1 OJ' 2 STATUS OF RULEMAKING RECORD 1 OF PROPOSED RULE:
PRM-050-53 RULE NAME:
THE OHIO CITIZENS FOR RESPONSIBLE ENERGY: RECEIPT OP PETITION FOR RULEMAKING PROPOSED RULB J'ED REG CITE:
.54FR30905 PROPOSED RULE PUBLICATION DATE:
07/25/89 ORIGINAL DATE FOR COMMENTS: 09/25/89 NUMBER OF COMMENTS:
EXTENSION DATE:
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5 1
FINAL RULE FED. REG. CITE: 59FR42182 FINAL RULE PUBLICATION DATE: 08/17/94 NOTES ON: PETITION IS CONCERNED WITH ATWS RULEMAKING PROCEEDINGS.
PETITION ATOS
- DENIED.
FILE LOCATED ON Pl.
RULB:
PRESS PAGE DOWN OR ENTER TO SEE RULE HISTORY OR STAFF CONTACT PRESS ESC TO SEE ADDITIONAL RULES, (E) TO EDIT OR (S) TO STOP DISPLAY PAGE 2 OF 2 HISTORY OF THE RULE PART AFFECTED: PRM-050-53 RULE TITLE:
THE OHIO CITIZENS FOR RESPONSIBLE ENERGY: RECEIPT OJ' PETITION FOR RULEMAKING OPOSBD RULE CY PAPER:
FINAL RULE SECY PAPER:
COIITACTll IUCJIABL LBSAR CONTACT2:
PROPOSED RULE SRM DATE:
FINAL RULE SRM DATE:
I I
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DATE PROPOSED RULE SIGNED BY SECRETARY:
07/18/89 DATE FINAL RULE SIGNED BY SECRETARY:
08/11/94 STAFF CONTACTS ON THE RULE JIAIL STOPI P-223 MAIL STOP:
PHONBI 492-7758 PHONE:
PRESS PAGEUP TO SEE STATUS OF RULEMAKING PRESS ESC TO SEE ADDITIONAL RULES, (E) TO EDIT OR (S) TO STOP DISPLAY
DOCKET NO. PRM-050-53 (54FR30905)
In the Matter of THE OHIO CITIZENS FOR RESPONSIBLE ENERGY: RECEIPT OF PETITION FOR RULEMAKING DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 05/30/89 07/22/89 S. HIATT ON BEHALF OF OCRE SUBMITTED REQUEST FOR ACTION UNDER 10 CFR 2.206.
(THE NRC IS TREATING A PORTION OF THE REQUEST AS A PETITION) 07/19/89 07/18/89 FEDERAL REGISTER NOTICE - RECEIPT OF PETITION FOR RULEMAKING 09/11/89 09/06/89 SUSAN HIATT, ON BEHALF OF OCRE, SUBMITTED FOUR 09/26/89 09/22/89 09/26/89 09/22/89 09/28/89 09/25/89 09/28/89 09/26/89 08/22/90 08/18/90 08/12/94 08/10/94 ADDITIONAL REPORTS WHICH ARE LABELLED AS EXHIBITS TO BE CONSIDERED AS PART OF THE PETITION RECORD COMMENT OF GENERAL ELECTRIC (P.W. MARRIOTT, MANAGER OF LICENSING) (
- 1)
COMMENT OF BWR OWNERS' GROUP (STEPHEN FLOYD, CHAIRMAN) (
- 2)
COMMENT OF MARVIN LEWIS (
- 3)
COMMENT OF PHILADELPHIA ELECTRIC COMPANY (E. P. FOGARTY, MANAGER) (
- 4)
COMMENT OF OCRE REPRESENTATIVE (SUSAN L. HIATT) (
LETTER FROM JOHN HOYLE, ACTING SECRETARY TO SUSAN HIATT, REPRESENTATIVE OF OCRE DENYING OCRE'S PETITION FOR RULEMAKING 08/12/94 08/11/94 FEDERAL REGISTER NOTICE ON THE DENIAL OF THE PETITION OF OCRE.
PUBLISHED ON 8/17/94 AT 59 FR 42182.
- 5)
AGENCY:
ACTION:
SUMMARY
DOCKET NUMBER PETITION RULE PAM,",.._~...
(ft L./ F fl 1 Dl!I/ o 5)
NUCLEAR REGULATORY COMMISSION 10 CFR Part 50
[Docket No. PRM 50-53]
Denial of Petition for Rulemaking Nuclear Regulatory Commission.
Denial of petition for rulemaking.
DOCKETED UStRC
- 94 I\\UG 1 2 A11 : 21 The Nuclear Regulatory Commission (NRC) is denying a petition for rulemaking (PRM-50-53) from Ms. Susan L. Hiatt on behalf of the Ohio Citizens for Responsible Energy, Inc. (OCRE).
The petition requested reopening of the rulemaking procedure that led to promulgation of 10 CFR 50.62, the 11Anticipated Transient Without Scram 11 (ATWS) rule. The principal basis for the OCRE request was the possibility that the ATWS analyses that formed the underlying bases of the ATWS rule were invalid because they did not appropriately account for the effects of large power oscillations, such as those that occurred during the March 9, 1988, instability event at the LaSalle County Nuclear Station (Unit 2).
The petition is being denied because the Commission has concluded, based on core stability analyses during hypothetical ATWS events, and based on recommended procedure changes at nuclear power plants, that large-amplitude power oscillations will not impact the core and containment response sufficiently to invalidate the assumptions and results of previous ATWS analyses that were the bases for the ATWS rule. The NRC has carefully considered the issues raised in the petition and has taken them into account in reaching its decision to deny the petition.
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ADDRESSES:
Copies of the petition for rulemaking and the NRC's letter to the petitioner, including attachments (SECY-94-123), are available for public inspection or copying in the NRC Public Document Room, 2120 L Street, NW.
(Lower Level), Washington, DC.
FOR FURTHER INFORMATION CONTACT:
Roy Woods, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 415-6622.
The Petition By letter dated July 22, 1988, Ms. Susan L. Hiatt, a representative of the Ohio Citizens for Responsible Energy, Inc., requested that the Director, Office of Nuclear Reactor Regulation (NRR), take immediate action with respect to boiling water reactors (BWRs) to relieve what she alleged to be undue risks to the public health and safety posed by the thermal-hydraulic instability of BWRs as revealed by an event at LaSalle County Station, Unit 2, on March 9, 1988.
The petitioner requested that the NRC conduct a rulemaking procedure under 10 CFR 2.802 to address:
- 1.
The possibility that the analyses used during the proceedings that promulgated 10 CFR 50.62 (the 11ATWS 11 rule) were invalid because they did not appropriately account for the effect of large power oscillations (those analyses were the underlying basis for the design requirements established in 10 CFR 50.62 to reduce the risk from ATWS events); and 2
- 2.
The appropriateness of the 10 CFR 50.62 requirement for automatic tripping of the recirculation pumps in response to designated ATWS signals.
In light of the potential consequences of large power oscillations, since tripping the recirculation pumps moves reactor operation into a state with high power-to-flow ratio where oscillations are likely, the petitioner requested that the pump-tripping requirement be reconsidered.
Staff Action on the Petition The staff has been reviewing generic concerns regarding the large power oscillations that were observed during the March 9, 1988, instability event at the LaSalle County Nuclear Station, Unit 2, since the event's occurrence.
That part of the effort that has focused on developing a response to the OCRE petition has concentrated on developing an improved understanding of BWR stability phenomena.
These staff [and associated Boiling Water Reactor Owner's Group {BWROG)] efforts have incl uded analytical studies of ATWS
~~enarios, stability sensitivity studi es, and the validation and verification of the analytical models and codes used for these studies. The primary objective was to determine if large-amplitude oscillations might impact the core and containment response sufficiently to invalidate the assumptions and results of previous ATWS analyses that were the bases for the ATWS rule.
With respect to OCRE's contention that the automatic tripping of the recirculation pumps in response to designated ATWS signals, as required by the ATWS rule, is inappropriate in light of the potent ial consequences of large power oscillations, the staff reviewed the advantages {related to decreased heat load on the containment) and d~sadvantages {related to exacerbation of 3
power osc;11at;ons) of the requ;rement that the rec;rculat;on pumps be tripped.
Reasons for Den;a1 The attachments to the NRC's letter to the petitioner (SECY-94-123) includes a deta;led presentat;on of the bases for the den;a1 of the petition.
In sumary, a substantial effort was necessary to develop computer codes to simulate the oscillation behavior of the modeled reactors and to validate and verify these codes to ensure that they give accurate predictions.
On the basis of ;ts review of TRACG code's qualifications for performing power oscillation analyses, the staff concluded that TRACG can serve as an adequate tool to est;mate qualitatively the global behavior of operating reactors during transients that may result in large power oscillations.
Although large power oscillations may increase the overheating and severity of fuel damage resulting from an ATWS event, the analyses indicate that core coolability and containment integrity can be acceptably maintained.
Therefore, the staff concluded that the ATWS analyses that formed the bases of the ATWS rule remain valid.
The staff's review of the advantages and disadvantages of the requirement that the recirculation pumps be tripped indicated that recirculation pump trip was appropriate and necessary to reduce heat load to the containment follow;ng an ATWS, and t hat the potentially adverse impact due to large power oscillations could be mitigated by revisions to the Emergency Procedure Guidelines (EPGs) that were recommended by the BWROG.
Revisions to the EPGs are:
prompt cessation of feedwater flow until water level is reduced to about one meter below the feedwater sparger, thus reducing core inlet 4
subcooling which dampens power oscillations; and earlier injection of boron in the presence of power oscillations, thus reducing power level, which reduces the adverse consequences of any remaining power oscillations. The staff concluded that these revisions are sufficient for mitigating the consequences of a bounding ATWS event with large oscillations.
On the bases of the above analyses and recommended procedure changes, the staff concludes that, although large power oscillations may increase the overheating and severity of fuel damage resulting from an ATWS event, core coolability and containment integrity can be acceptably maintained in a manner consistent with the assumptions and results of previous ATWS analyses that were the bases for the ATWS rule, and that, therefore, the requirements of the ATWS rule remain appropriate.
Because each of the issues raised in the petition has been substantively resolved, the NRC has denied this petition.
Dated at Rockville, Maryland, this !(~
day of August, 1994.
For the Nuclear Regulatory Commission.
John,
- Hoyle, Acti, g Secretary 5
OFFICE OF THE SECRETARY UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555 August 10, 1994 DOCKET NUMBER 5623 PET\\T\\ON RULE !..!PR;;;.M~-~
(S'i f tz 9P"I OS Ms. Susan L. Hiatt Representative of Ohio Citizens Responsible Energy, Inc.
for 8275 Munson Road Mentor, Ohio 44060
Dear Ms. Hiatt:
This letter is in further response to your petition of July 22, 1988, requesting that the Director, Office of Nuclear Reactor Regulation {NRR), take i11111ediate action with respect to boiling water reactors (BWRs) to relieve what
_ you allege to be undue risks to the public health and safety posed by the thermal-hydra~lic instability of BWRs as revealed by an event at LaSalle County Station, Unit 2, on March 9, 1988.
On August 26, 1988, you were informed by letter from the Director, NRR (Thomas E. Murley) that your request for immediate relief was denied because the allegations that form the basis for your petition did not reveal any new operational safety issue that posed an inrnediate safety concern for continued BWR operation.
You were also informed that your petition was being treated under 10 CFR 2.206 of the Commission's regulations and that appropriate action, that is, a formal decision, would be taken within a reasonable time.
On April 27, 1989, you were further informed by letter from the Director, NRR lfhomas E. Murley) that for the reasons set forth in Director's Decision DD-89-03 under 10 CFR 2.206 {enclosed with the letter), your petition had been denied.
However, you were also advised in the April 27, 1989, letter that your request to reopen rulemaking proceedings regarding anticipated transients without scram {ATWS) was being treated as a Petition for Rulemaking under 10 CFR 2.802 of the Commission' s regulations, and that it had been referred to the NRC Office of Research for appropriate action.
This letter notifies you that the Co11111ission has denied your petition for ATWS rulemaking under 10 CFR 2.802 of the Commission's regulations. This letter also provides you with a "simple statement of the grounds for denial" as required under 10 CFR 2.803 of the Co11111ission's regulations.
The principal basis for your petition for rulemaking under 10 CFR 2.802 was the possibility that the analyses used during the ATWS rulemaking proceedings that promulgated 10 CFR 50.62 (the *Arws* rule) were invalid because they did not appropriately ac~ount for the effect of large power oscillations {these analyses were the underlying basis for the design requirements established in 10 CFR 50.62 to reduce the risk from ATWS events). To address this concern in their programs for evaluating power oscillations during ATWS events, the NRC staff and the Boiling Water Reactor OWner's Group (BWROG) have reevaluated whether or not large-amplitude oscillations might impact the core and containment response sufficiently to invalidate the assumptions and results of previous ATWS analyses that were the bases for the ATWS rule.
Substantial effort was necessary to develop computer codes to simulate the oscillation behavior of the modeled reactors and to validate and verify these codes to ensure they will give accurate predictions. Utilizing the analyses performed by these computer codes, the staff has concluded that, although large power oscillations may worsen the overheating and severity of fuel damage resulting from an ATWS event, core coolability and containment integrity can be maintained and the radiological consequences will remain acceptable. Therefore, the staff concludes that the prescriptive requirements of the ATWS rule remain appropriate.
Your petition also raised the contention that the automatic tripping of the recirculation pumps in response to designated ATWS signals, as required by the ATWS rule, is inappropriate in light of the potential consequences of large power oscillations (tripping the recirculation pumps moves reactor operation into a state with high power-to-flow ratio where oscillations are likely), and that the ATWS rule's requirement that recirculation pumps be tripped following an ATWS event should therefore be reconsidered.
The staff's review indicated that recirculation pump trip is appropriate and necessary to reduce heat load to the containment following an ATWS, and that the adverse impact due to large power oscillations can be mitigated by revisions to the Emergency Procedure Guidelines (EPGs), consisting of i11111ediate action to reduce core inlet subcooling after confirmation of an ATWS event and earlier injection of boron in the presence of power oscillations.
The staff concluded that these revisions are sufficient for mitigating the consequences of a bounding ATWS event with oscillations, and that changes to the recirculation pump trip requirements vf the ATWS rule are therefore not necessary.
On the above bases, the staff concluded that large-amplitude oscillations will not impact the core and containment response sufficiently to invalidate the assumptions and results of previous ATWS analyses that were the bases for the ATWS rule; that changes to the pump-trip requirements of the ATWS rule are not necessary; and that therefore your petition should be denied.
While the Co11111ission deemed that an i11111ediate safety concern did not exist after the LaSalle event, the NRC staff did initiate a review of the technical issues involved.
Your petition st~mulated a more in-depth review of the issues. This review has resulted in an improved knowledge of this matter and in proposed revisions to the emergency procedures which will provide an even greater margin of safety.
We thank you for your interest in this matter. The staff will inform you of the status of the emergency procedure changes under separate cover.
I have enclosed a copy of the Commission Paper presented to the Commission in the process of obtaining Commission approval for denial of your petition for ATWS rulemaking.
It contains a more detailed presentation of the bases for our denial of your petition, and an attachment is a copy of a document that gives complete details of those bases.
Sincerely,
~~~
~ng Secretary
Enclosure:
Co111t1ission Paper with attachments
August 18, 1990 DOCKET NUMBER PETITION RULE PRM So-.s:3 cs--1 FR 3 o qos)
Docketing and Service Branch Office of the Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 Re: PRM-50-53
Dear Sir/Ms.:
DOCKEiED USNRC
- 90 AUG 22 P 2 :39 OFF!Cf.: OF SECi<ETAfiY OOCKU ING & SEf'Vlf.f BRANCl-i The petitioner, Ohio Citizens for Responsible Energy, Inc.
("OCRE") would like the rulemaking record in PRM-50-53 to reflect the following facts and comments:
- 1.
Simulation of BWR power oscillations without scram with the Brookhaven National Laboratory Engineering Plant Analyzer indicates sustained oscillations between 10% and 500% of full power; the Minimum Critical Power Ratio approaches unity at times (in the troughs of the oscillations), implying that fuel integrity might be challenged; centerline fuel temperature oscillates between 1200 F and 2300 F, while the clad wall temperature oscillates between 546 F and 566 F.
"Simulation of BWR Power Oscillations Without Scram," H.S. Cheng, Brookhaven National Laboratory, August 4, 1988.
Comment: what is the potential for fuel failure due to pellet-cladding interaction under these circumstances?
- 2.
Graphs presented to the NRC on December 6, 1988 by H.S.
Cheng and W. Wulff of Brookhaven of BWR simulations with the BNL Engineering Plant Analyzer show the MCPR to be less than 1.0 in the troughs of the oscillations for the ATWS sequences modeled.
The suppression pool temperature is predicted to reach 210 F within 12 minutes for the turbine trip without bypass event and 230 F within 15 minutes for the turbine trip with bypass but no feedwater pump trip.
- 3.
Material presented by Dr. Jose March-Leuba of the Oak Ridge National Laboratory -O the ACRS on May 23, 1989 indicates that average power increases due to power oscillations typically on the order of 1.5% of the peak osci l lation amplitude.
In a paper presented to the "BWR Stability Symposium" held in Idaho Falls, Idaho, August 10-11, 1989, Dr. March-Leuba states ~qat the average power increase is typically 1.5% to 2% of the.value of the peak power minus the steady state power.
"Under~tanding l
Acknowledg~ by card.................................,
I U.S. NUCLEAR RE(JULATORY COMMISS N DOCKETING & SERVICE SECTiON OFFICE OF THE SECRETARV OF THE COMMISSION Document Statistics Postmark Date -,-._,..;;.+-'-"- ---
Cop,es Received, _ _._ _ ___ _
Ada'I Copies Repnxb:ed --._ __
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the Boiling Water REactor Limit Cycle," p. 8.
- 4.
In NEDO-31708, "Fuel Thermal Margin During Core Thermal Hydraulic Oscillations in a Boiling Water Reactor,", General Electric states that at large magnitude oscillations (greater than 200% rated power) potential violations of the safety limit MCPR were predicted.
- 5.
In NEDO-31709, "Average Core Power During Large Core Thermal Hydraulic Oscillations in a Boiling Water Reactor,"
General Electric states that the increase in core average power over the period of increasing oscillation magnitude is less than seven percent of rated core power.
- 6.
In SECY-90-152, it is stated that large amplitude power oscillations, with power peaks between 500 and 2000 percent of rated power, have been calculated for ATWS events.
Such oscillations contribute to an average thermal powre increase of 1.5% to 2% per 100% of peak core power.
I.e., a 500% peak power oscillation could lead to an i ncrease of 7.5% to 10% in average thermal power in the core.
The proposed solution to this problem is said to involve revised operator actions, such as injecting boron earlier, reducing feedwater flow, and other measures to reduce core inlet subcooling.
- 7.
The summary in the Weekly Information Report for the week ending July 6, 1990 of the meeting with the BWR Owners Group held June 27, 1990 states that the BWROG has admitted that ATWS acceptance criteria (1979 Mattson l etter) could be exceeded (e.g., fuel temperature limits) and that significant fuel failure is likely.
Candidate revisions to operating procedures are modifications to initiation times for water level reduction, boron injection and depressurization.
Comment: Relying on operator actions, however they may be revised, may not compensate for basic physical limitations in system designs.
E.g., GE ATWS ana l yses (NEDO-24222) assumed that operators manually initiated the SLCS within 2 minutes. Is it realistic to assume that operators can and will in fact initiate the SLCS any sooner?
The existing SLCS in BWRs, even if actuation were instantaneous, falls far short of the SLCS recommended in Vol. 4 of NUREG-04 60 : 300-400 gpm capacity, recommended precisely to suppress power oscillations and their associated uncertainties.
Even with parallel two-pump operation, existing SLCS capacity is only 86 gpm.
It may be appropriate to require the use of enriched boron and/or other neutron poisons in the SLCS, in addition to revising operator procedures.
Automating the SLCS s hould also be reconsidered.
2
OCRE requests that the NRC respond to each of the above facts and comments in its final determination on PRM-50-53.
Respectfully submitted, Susan L. Hiatt OCRE Representative 8275 Munson Road Mentor, OH 44060
( 216) 255-3158 3
OCKET NUMBER
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PHILADELPklA ~LECTR1C COMPANY E. P. FOGARTY MANAGER NUCLIEAR SUPPORT DIVISION NUCLEAR GROUP HEADQUARTERS 955-65 CHESTERBROOK BLVD.
WAYNE, PA 19087-5691 (2 15 ) 640-6000
'89 SEP 28 P 2 :59 September 26r 1989 Mr. Samuel J. Chilk Secretary of the Commission
- u. s. Nuclear Regulatory Commission Washingtonr DC 20555
SUBJECT:
Comments Concerning the Nuclear Regulatory Commission 10CFR 50 Petition for Rulemaking by the Ohio Citizens for Responsible Energy for Responsible Energy Dear Mr. Chilkr This letter is being submitted in response to the Nuclear Reglatory Commission's (NRC's) request for comments regarding the 10 CFR 50 petition for rulemaking filed by the the Ohio Citizens for Responsible Energy (OCRE) on May 26r 1989r and published in the Federal Register (54 FR 30905r dated July 25r 1989).
Philadelphia Electric Company (PECo) appreciates the opportunity to comment on this petition for rulemaking which requests that the NRC reopen the Anticipated Transient Without SCRAM (ATWS) rulemaking proceedings.
This petition was filed with the NRC as a result of a power oscillation event that occurred at the LaSalle Unit 2 nuclear power plant on March 9r 1988.
The OCRE petition requested that the NRC take immediate action to relieve alleged undue risks to the public health and safety posed by the thermal-hydraulic instability of Boiling Water Reactors (BWRs).
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U.S. Nuclear Regulatory Commission Page 2 PECo does not support this petition for rulemaking on the following bases.
We consider that the current requirements established in 10 CFR 50.62 for reducing the risks from ATWS events for light water cooled nuclear power plants are adequate.
In addition, NRC Bulletin 88-07, "Power Oscillations in Boiling Water Reactors," required that all holders of operating licenses or construction permits for BWRs ensure that adequate operating procedures and instrumentation are available, and adequate operator training is provided to prevent occurrence of uncontrolled power oscillations during all modes of BWR operation.
Finally, the NRC is currently performing confirmatory analyses of ATWS events.
If evidence from these evaluations indicates that contradictions exist in the assumptions and results of previous ATWS analyses, reconsideration of the current ATWS rule may be warranted.
Until there is conclusive evidence to support any new findings, however, we are of the opinion that reopening the ATWS rulemaking proceedings is unjustified at this time.
If you have any questions, please do not hesitate to contact us.
Very truly yours,
DOCKE" u. 8 Marvin I. Lewis 7801 Roosevelt Boulevard Suite 62 Philadelphia, PA 19152 (215)624-1574 Secretary of the Commission LI. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Docketing and Service Branch Dear Secretary Chilk; "89 S[P 28 P3 :22 or,,,
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Please accept this letter as my comments on the Petition for rulemaking:
notice of receipt from the Ohio Citizens for Responsible Energy dated May 26, 1989. The Ohio Citizens for Responsible Energy has a long history of pointing out safety hazards of nuclear power.
They have been pointing out these hazards lon g before the disasters at Chernobyl and Three Mile Island proved the correctness of their stands.
This latest petition published in the July 25, 1989, Federal Register at Page 30905 continues a history of attempting to assure the safety of nuclear.
The present petition centers on the danger poised by "thermal hydraulic instability of BWRs." The dangers of thermal hydraulic instability were well known for many years. Richard E. Webb,
Ph.
D.,
discussed the possibility of runaway reactors in his classic book, "The Dangers of Nuc lear Power."
Dr. Webb estimated that a
reactor could experience a runaway reaction producing energies eleven times greater than the design basis. Chernobyl showed that Dr. Webb's figures were very possible: it happened.
The hydraulic instability accident is only one of a great number of possible and actual accidents which can endanger the health and safety of the public.
The recent findings of the inappropriately certified materials suggest that the reactors are not built as specified and designed. Problems with valves suggest that the reactors are sensitive to a
type of accident called "Interfacing Systems Loss of Coolant Accident,"
(ISL).
In response to the dangers of ISLs, I have sent the enclosed letter warning activists of the dangers of ISLs and the deficient manner with which the NRC has ap proached the subject.
The way that thermal hydraulic instability has been neglected by the NRC is only a symptom and not the disease. The disease is that the entire spectrum of possible accidents have been ignored by the NRC instead of giving these accidents proper airing and exhaustive repair.
0 '
1889
The deficient way that the NRC approaches these accident scenarios mirrors the way that the NRC approaches this rulemaking.
Despite the many deficiencies which have arisen in the nuclear industry, the NRC limits this petition far rulemaking an the broad and vital subject of Anticipated Transient Without Scram ta one small aspect: thermal hydraulic instability. The NRC assumes that the reactor will be built and maintained according ta NRC directives with defense in depth.
The recent findings of counterfeit materials, unqualified valves, discrepancies between as-built and design drawings ma ck any assumption that the reactors are built with defense in depth. The NRC continues ta approach all these woes with an attitude that accidents can't happen.
The history of TMI#1, Chernobyl, LaSalle and a thousand and one dangers do little ta make the NRC respond. The NRC seems ta hear only the primacy issue which is embodied in the introduction ta the 1954 Atomic Energy Act without considering the nine times that the "health and safety of the public" is referred ta in the same Act.
The Charter of the NRC also bears little weight upon the NRC when it refers ta the health and safety of the public.
My request is simple and is directed ta the NRC: What and haw will make the NRC respond ta protecting the health and safety of the public?
Marvin I. Le~i~, P. E.
7801 Roosevelt Boulevard Suit e 62 Ph i ladelphia, PA 19152 Dear Activist ;
The Chernobyl Accident started as a steam expl osion 45 seconds after coolant was injected onto a hot, uncovered core.
This accident is di fferent from any des i gn basis accident used in U. S. nuclear plants. The difference is obvious.
The accident descripLion used in the design of US plants assumes that a steam explosion will occur only after a core melt.
For the core to melt requires that the accident proceed for 2700 seconds (35 minutes) wi thout an adequate operator response.
Chernobyl experienced a steam explosion 45 seconds after voiding its coolant. 2700 seconds may give an operator time to avoid a core melt, but 45 seconds is not enougn time to stop an accident.
The Chernobyl Accident can happen at many U. S. nuclear plants. Many of the design differences between U. S. plants and Chernobyl have nothing to do with stopping a Chernobyl - like steam explosion. On the following pages, I describe a Chernobyl - like steam explosion which can happen at many U. S. plants. The r esult can be as disastrous as t he Chern obyl Accident.
This explana t ion involves only one set of deficienci e s.
Nuclear power plants have many deficiencies, an d thru a recen t ruling in the Limerick II Hearings, many nuclear license hearings can be r eopened. Please contact me for i nformation on how you can find contenti ons t o argue before the NRC, and stop th e plant ne a r you f ro m op er ating and endangering you and yours.
(215) 624 1 1
- Illustration I:
High pressure piping Valve closed The Chernobyl Connection Reactor
/
steam I ~,......_,..,_coolant ~,lAll' l
level I
1 Valve closed
~Low pressure piping I
/
\\
0 core High pressure reactor coolant system
- A light water reactor is producing electricity *
-Coolant covers the core and steam fills the remainder of the high pressure coolant system.
-Closed valves are shown as@, and open valves are shown as \\
Illustration II :
Reactor steam
,~-*- coo 1 ant -,~0 I level 1
I Valve opens I
I High pre s sure pipin g.
Olow pressure pipinq~
Valve closed 0
l_core I I )
_,./
- 2.
-The valve or valves separating the lo w pressure piping from the high pressure system reactor coolant system opens.
- The high pressure in the reactor surges into the low pressure piping.
-The high pressure reactor coolant system can wi thstand higher pressure than the low pressure piping.
-The low pressure piping has too thin a wall to contain the high pressure and breaks.
Illustration II I:
High pressure piping Valve closed Reactor
~
- ---~
steam
\\
coolant level -
c or e,
I Valve open
) Low pressure pipinq}f
-The coolant turns to steam and disch a rges thru the break in the low pressure piping until the core is uncovered.
- The core overheats due to the lack of coolant to cool it properly.
- 3.
Illustration IV :
Reactor coolant Valve closed levell j
(~tow pressure piping ressure I i Valve opens
' core./
1
-The valve on the low pressure piping closes.
- The valve on the high pressure coolant supply opens.
- The coolant under high pressure floods the hot, uncovered core.
-The coolant flashes explosively upon contact with the overheated core.
The effect of this energy release c an be as destructive as the steam explosion at Cherno byl.
- 4.
BWA DOCKET NUMBER PFT1T10f\\J RULt. P M ~0-.5.J Cs~-r~ 30 90.s)
OWNERS' GROUP Stephen D. Floyd. Chairman (919) 546-690 l BWROG-8970 September 22, 1989 do Carolina Power & light Company **~ l fSBJetOOill'f>S)~,2* Raleigh, NC 27602 Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C.
20555 Attention:
Subject:
Reference:
Docketing and Service Branch BWR Owners' Group Comments on the OCRE Petition on the ATWS Rulemaking (Federal Register Vol. 54, No. 141, July 25, 1989)
- 1)
- 2)
- 3)
NED0-31708, "Fuel Thermal Margin During Core Thermal Hydraulic Oscillations In a Boiling Water Reactor, June 1989 NED0-31709, "Average Core Power During Large Core Thermal Hydraulic Oscillations In a Boiling Water Reactor", June 1989 NED0-24222, "Assessment of BWR Mitigation of ATWS, Volume II (NUREG 0460 Alternate No. 3)",
February 1981 The BWR Owners' Group (BWROG) would like to take this opportunity to provide comments on the Ohio Citizens for Responsible Energy (OGRE) Peti-tion requesting NRC to reopen the ATWS rulemaking proceeding.
The BWROG has been performing extensive stability evaluations for over a year (References 1
and 2),
including the potential occurrence of oscillations without scram protection.
The latter evaluations were specifically performed in response to NRC questions regarding the possibility of increases in average core power associated with oscillations during a postulated ATWS.
The results of these evaluations have been discussed with the NRC on several occasions.
Based on this work, the BWROG concludes that oscillations will have no significant impact on ATWS average core power.
The existing ATWS rule (10CFR50.62) and related industry initiatives make the probability of an ATWS leading to significant power-flow oscillations extremely remote.
Additionally, it should be recognized that the insertion of a few control rods or a small amount of liquid boron will prevent oscillations or reduce the likelihood that they will grow to a significant size.
Furthermore, the specified ATWS actions to quickly inject liquid boron ensure that, even if oscillations occur, they would be short lived
- ics MISSION ION y
Postm,k D '
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Secretary of the Commission September 22, 1989 Page 2 and of no significance.
Finally, the possibility of power-flow oscilla-tions was recognized (Reference 3) during the ATWS rulemaking and the March 9, 1988 LaSalle Unit 2 event introduces no new information which justifies reopening this issue.
Because of this, the BWROG does not believe that reconsideration of the ATWS rule is appropriate.
The comments/positions provided in this letter have been endorsed by a substantial number of the members of the BWR Owners' Group; however, it should not be interpreted as a commitment of any individual member to this position.
Each member must formally endorse the BWR Owners' Group position in order for that position to become the members' position.
- Regards, Stephen D. Floyd, Chairman BWR Owners' Group 0918891/pc cc:
BWROG Primary Representatives BWROG Executive Oversight Committee BWROG Stability Committee GJ Beck, BWROG Vice Chairman L Phillips, NRC M Hodges, NRC F Miraglia, NRC R Warren, INPO R Evans, NUMARC R Galer, EPRI
r r MFN-069-91 September 22, 1989 PWM-89143 l MBE
. p M Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C.
20555 50-SJ GE Nuclear Energy r
f General Elect/IC Company 175 Curtner Avenue. San Jose, CA 95125
'89 SEP 26 P 4 : 1 6
- ~F FI
- ~*Gu,=..
Attention:
Docketing and Service Branch
Subject:
Comments on OCRE Petition Regarding ATWS Rulemaking (Federal Register Vol. 54, No. 141, July 25, 1989)
Reference:
NED0-24222, "Assessment of BWR Mitigation of ATWS, Volume II (NUREG 0460 Alternate No. 3)", February 1981 The purpose of this letter is to provide General Electric's comments on the Ohio Citizens for Responsible Energy (OCRE) Petition requesting NRC to reopen the ATWS rulemaking proceeding.
It is General Electric's view that there is no basis for the OCRE Petition.
The possibility of power/flow oscillations associated with a postulated ATWS event were fully recognized by NRC and the industry prior to (and during) the ATWS rulemaking.
Prior to the rulemaking, GE performed exten-sive analyses of BWR performance under ATWS conditions (Reference) and these analyses predicted the potential for power/fl ow oscillations under certain ATWS conditions.
The results of these analyses were clearly part of the NRC deliberations associated with the ATWS rulemaking and the March 9, 1988 LaSalle Unit 2 event does not introduce any new information which bears on the ATWS rule.
Beyond this, it is important to maintain the proper perspective relative to this situation.
First of all, ATWS is an exceedingly low probability event which encompasses a spectrum of postulated failures leading to the inabili-ty to insert some or all of the control rods.
With the successful inser-tion of only a few control rods (or a small amount of liquid boron) stabil-ity should not be an issue because of the stabiliz ing effect of the power reduction.
For the reasons stated above, the probability of unacceptable power-flow oscillations associated with an ATWS event is extremely low.
In swnmary, it is GE's opinion that the OCRE petition is without merit since stability was considered during the ATWS rulernaking and the LaSalle event adds no new information.
Furthermore, BWR system designs and inher-ent features are effective in preventing significant oscillations.
Based on GE's assessment it is concluded that with existing plant features and I, **
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Secretary of the Commission September 22, 1989 Page 2 specified operator actions there is a high probability that oscillations, in combination with ATWS, can be avoided entirely or short lived and of relatively low amplitude shoul d they occur.
Because of this, we do not believe the potential for power oscill ations justifies reconsidering the ATWS closure.
- Regards, P. W. Marriott, Manager Licensing and Consulting Services (408) 925-6948 M/C 682 0829891/HCPS cc:
L. Phillips (NRC)
W. Hodges (NRC)
A. Thadani (NRC)
L. Gifford (GE-Rockville)
September 6, 1989
("OCRE")
ON PRM-50-53 (54 FED. REG. 30905, JULY 25, 1989) C-uc*~' I. '14 The petitioner OCRE herein submits comments and reports to be made part of the record in this petition to reopen the ATWS rulemaking proceeding in light of the power oscillations occurring at the LaSalle-2 BWR following a dual recirculation pump trip on March 9, 1988.
The following enclosed reports should be made part of the record:
Exhibit 1: AEOD Special Report on the LaSalle-2
- event, AEOD Special Report No. AEOD/S803, dated June 8, 1988.
Exhibit 2: excerpt from the NRC Augmented Inspection Team report on the LaSalle-2 event.
Exhibit 3: letter report by the Advisory Committee on Reactor Safeguards, dated June 14, 1989, re Boiling Water Reactor Core Power Stability.
Exhibit 4: "A Report on Reactor Study Issue Number 25" prepared for the Ohio State University Expert Panel by Dr.
William P.
Stephany of Nuclear Education Training
- Services, Inc.
("NETS").
(This report is part of the comprehensive review of the 1975 General Electric Nuclear Reactor Study (commonly known as the Reed Report) commissioned by the Public Utilities Commission of Ohio.
The attachments to the report are not included herein.)
These reports support OCRE's position regarding the ATWS rulemaking in light of the LaSalle-2 power oscillation event.
The AEOD report, which was largely the basis for OCRE's 1988 petition in this matter filed under 10 CFR 2.206, states that the LaSalle event "necessitates that ATWS mitigation be reviewed in light of this event." Exhibit 1,
- p.
7, emphasis added.
The NRC's Augmented Inspection Team also expressed concern that, "in view of the large magnitude of the APRM oscillations in LaSalle, the AIT believes that the ultimate power level without scram is
- unknown, and that the 500 %
bounding level assumed in the ATWS investigation may not be bounding.
LPRM oscillation magnitudes more than seven times those of the APRMs have been observed in the case of regional oscillations."
Exhibit 2, p. 24.
These reports illustrate the 11 Susan L Hiatt 8275 MunlOll Rd.
Mentor, OH 44060 1
- 0. S. Nl!Cf r /\\f\\ ~f:-'IJI_ATGRY J.1n,W1!>Ki,
- DO, Kf. f 1, 1G /l. S_., /1 rF
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- ,'* -~ c-* 1::: ::* f!"TAI Y 0~ THE CON'-/ l'..i. ION Special Di*,~.6;.:t.,);l
/Ji 'l, fi2JD5
________ L!..S.o J
NRC Staff's concerns and advice on this matter, which OCRE is endorsing.
issue
- today, the Commission in the ATWS Staff made Significantly, the reason ATWS is still a safety despite the ATWS rule, 10 CFR 50.62, is that failed to follow the advice of its own Staff rulemaking process.
In NUREG-0460, Volume 4, the the following comments regarding ATWS in BWRs:
Several events are shown to have significant, periodic oscillations in neutron flux following an initial, large neutron flux spike.
The staff has never before encountered this type of accident behavior prediction, and so it has never been specifically considered in previous PCI evaluations.
- The combined effects of high neutron flux
- spikes, resulting in high cladding and boil temperatures, followed by oscillation in flux, fluid flow, etc., raise questions not only about fuel future, but also about the potential for loss of coolable (rod-like) geometry.
The 2200 F, 17 percent oxidation LOCA limits that GE proposes as evidence of coolable geometry are not applicable here because those limits address cladding oxidation and embrittlement effects only.
They do not address the potential effects of oscillating mechanical loads on wasted and collapsed cladding that might be "locked onto" the fuel pellets as a result of a BWR ATWS involving a high flux
- spike, nor do they consider center-melted oxide.
NUREG-0460, vo*lume 4, pp. A-87 to A-89.
The recommended mitigative measure, an automatic, high-capacity standby liquid control system (300-400*gpm), would eliminate or greatly reduce the oscillations.
NUREG-0460, Volume 4,
pp.
A-64, A-48 to -52, A-43, and 29.
Unfortunately, this measure was not incorporated into the final ATWS rule.
OCRE believes that incorporation of the automatic, high-capacity SLCS, along with the other provisions of the ATWS
- rule, 10 CFR 50.62, pertaining to
- BWRs, would allay any concerns about power oscillations resulting from the reci~culation pump trip, which is a necessary feature to quickly reduce power and reactor pressure to avoid failure of the reactor coolant pressure boundary.
It is significant that independent reviews of ATWS and the LaSalle event also point out the need to reconsider the ATWS rulemaking in light of the LaSalle oscillations.
Exhibit 4,
prepared by
- NETS, a
firm which provides consultants and services to the nuclear industry, concludes that "there is a
large gap between the ATWS prevention and mitigation recommendations stated in NUREG-0460 and the ATWS Rule stated in 10 CFR 50.62. In light of the recent LaSalle event, the 2
consequences of the recirculation pump trip ATWS mitigation feature do need to be reviewed, and the concerns expressed by the NRC Staff in Vol. 4 of NUREG-0460 also need to be looked at again."
Exhibit 4, p. 16.
The ACRS also recommends that "considerable attention be given in the longer term to the development of an improved understanding of the conditions that can lead to an ATWS compounded by core power oscillations."
Exhibit 3.
OCRE would urge the Commission to follow the Staff as given in NUREG-0460, the AEOD report, Augmented Inspection Report, as well as that reviewers such as the ACRS and NETS.
Respectfully submitted, Susan L. Hiatt OCRE Representative 8275 Munson Road Mentor, OH 44060
( 216) 255-3158 3
advice of its and the LaSalle of independent
1 e-\\C.H-t B fT 1
)
UNITED STATES
~
/
NUCLEAR REGULATORY COMMISSION ~--------
WASHINGTON. O. C. ~
JUN O 8 1988
'i MEVOR.ANDUM FOR:
Thomas E. Murley, Director Office of Nuclear Reactor Regulation Eric S. Beckjord, Director Office of Nuclear Regulatory Research F~CM:
Edward L. Jordan, Director Office fer Analysis and Evaluation.*.** ';.*
- *of Operational Data.* *.. ~* *;" _*,.*
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-:c - AEOD. CONCERNS REGARDING* THE t,'.ARCt-'-*,9.1988 POWER.
- ,.::': '=. -t. "'* * :"
SL2JECT:
- OSCILLATION EVENT AT LASALLE 2 --*:.;_.. : " --.-
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- Er.closed*is* an ;AEOD Special Report detailing our concerns**about the.'LaSalle 2 *_* -*.::::'.,:
pc....er*_osci 11ation event* of March 9;* 1988.
We oave revj_~wed ca1culation*s* *..:-._*:*--
- .. *::-, pr:rfo~d* by Brookhaven *~m,.the. BW8. Nuclear Pl ant':AnaJ_yz*et*/. as -we*l_f:'as *.:tne"0:**--::t ~8~~
_.'-1 icensee Is.,LER*.a'nd" other: foreign and U.S. Jnfonnatiori~::i.:A l_though *th.is. ts.. the-'*'-~-~-;~-..,...
.. */ffr.st~ "event of:thf(.typ( at: a.. domestic.. *reactor~
- sfmi lar:events have :QCftir"r:_ed*.. ~::...-
.:_-.:J_n foreign reactoz::~- *" Based on. this _reiiiew, we classi.fy:.this_ event as.. an.,... *. *
- .* * :f;a;portant ~precursor* :event **,dth
- sianificant safety ccincerns.:->":Our_most::?'.
,..... *~*-. _-::::::;'=_:signi f; cant:.conc'erns*.-an~ :aisoda'ted **recorrmendations. are--jles*c'r,ibe<[J:ie low-.;
- .
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- 1.
The LaSalle event raises questions about the adequacy of the analysis
.L ~,~,.
--... ::-used to meet the core stability r:~quirements*of GDC-12:"!hen b_oi:~. ------~--;-**
.. recirc:ulatlon* pumps are tripped.**The event*also points out the**'* :..
- 0 =-*:**-
. difficulties the operators fa_ce in rapid diagnosis of and response to an event which readily promotes significant complicating factors such as subsequent loss of feedwater heatjng and reactor water level fluctuations.
Simple and unambiguous procedure~* are needed to assure prompt proper operator response which ensures~campliance with GOC-12.
GE SIL 380 does not provide adequate guidance.
- 2.
During startup and shutdown, BWRs routinely enter regions of potential thermal-hydraulic-neutron kinetics instability. This operation can be avoided without large impact on plant operations by modifying plant operating procedures to increase recirculation flow slightly early in the startup and by inserting contro*l rods sooner during shutdown.
Several foreign reactors operate with power/flow operating restrictions that avoid the unstable region. Additionally, reduction or loss of forced recirculation flow during plant transients can result in the plant entering regions of potential instability. Prudent operator action is needed to restore stable plant operation and to avoid actions which could initiate events with more significant consequences.
For example, restart of recirculation pumps following loss of feedwater heating or MSIV closure could result in additional reactivity insertion while the reactor was exhibiting power oscillations.
- 3.
This event has implications regarding the reactor transient response to a recirculation pump trip during an ATWS.
In particular, the power oscilla-tions may substantially exceed previously predicted values and thus raise questions regarding previous fuel integrity evaluations.
Thomas E. Murley Conclusion The March:9 LaSalle event indicates serious deficiencies in the core stability analysis for LaSalle and perhaps other BWPs.
Further, such undamped power oscillations call for prompt operator recognition ard action, yet at LaSalle, operators were not trained to recognize or respond to such oscillations.
Adequate plant procedures did not exist at LaSalle, and few, if any, plant simulators in the U.S. are capable of w~deling these types of oscillations.
It is not at all clear at this time that we understand the nature and potential consequence of such power oscillations considering such factors as improper or no operator action, alternative core configurations and equipment failures, or divergent localized power oscillations. Since it will take time to thoroughly-analyze and understand the LaSalle event and its implications on other BWRs, we conclude tttat, at least in this interim period, action is*war.r.anted to minimize the po ten ti a 1 for core instability. _ Our reco1T1T1endations *in :this regard are* *. _** *.- *.-___ -
presented below.*
~- "
_ We anticipate_ a written response to these recommendations within -45.days* as*:.. c
- discussed in NRC Manual Chapter 0515~ * -.'_*
- _**:~----~-- *0 -~i-_ **:*-*
-,.-,_~--~---~
~ -
.i.
- _::.. "-*-.:<*-Recorrmendation to NRR
- >}:~_-
,--_-_ _..,:~_-.i\\t,~-_-:_:c-~~-_;_._~~~-:-,"~=- _.. --
)endi n*g *a fu 11 understanding of the _La Sa 11 e event "ana* its. implication~.* we ~t-;<':::"--~~,.~).'i~
~:~ {~ -__ :'.-_be 1i ~ye __.thtt \\1.1_*-. ~~~s ___ s~?~) d _b~.:.e~~: ~~t *.to:'.. __
-~ _.. __ / -) ( :~ ~i: *. -:~~ *_.:.'.
_* -; ~
~
- ". :_._. -~ ~;t
. -::c*;\\: -. -.: _ (a)* Imm.ediate ly 'insert *contra 1 rods.: to' below the 80% roi:t- -1 i ne *f,oll owing' :;..~~. y~~:J,,.:.._.;_.,;_f ?-~~
- ~.t--1..-
reduction or:lass of recirculation**.flow*or other transients*which resu1t**-;,_--, ___,..,,,.~--:...'..:.-*:*:
in entry_ into potentially unstable regions of the power/flow map.
(b) - Increase recirculation flow dur-ing *routine reactor *startups :ana insert some control rods prior to reducina recirculation flow below 50~ during shutdowns to avoid operation *;n potentially unstable areas of the power/
flow map.
(c)
Immediately scram the reactor if Ca) or (b) above are not successful.
Recorranendation to RES Review resolution of Gis B-19 and B-59 an~ tl'fS ~itigation in light of the LaSalle operating experience.
Please let me know if we can provide any clarification or additional assis-tance.
If you have questions regarding the enclosed Special Report, please call Jack Rosenthal on x24440.
~:nal Sii;,,,ed By:
E: D Jord¥1 Edward L. Jordan, Director Office for Analysis and Evaluation of Operational Data
Enclosure:
As stated Di5tribution:
See next page
- SEE PREVIOUS CONCURRENCE
- DSP:AEOD JKauffman :md 6/ /88
- DSP:AEOD
- OSP:AEOD
- OSP:AEOD
- D:DSP:AEOD Glanik 6/ /88 JRosenthal 6/ /88 VBenroya 6/ /88 TNovak 6/ /88 DD:AEOD CJHeltemes 6/ /88
- D :AEOD ELJordan 6/1 /88
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AEOD SPECIAL REPORT llNIT:
SPECIAL PEPORT N0.:
AE0D/SeO3 DC:CKET NO.:
LaSalle 2 SC-374
[)AT£:
JunE: 7. 1988 EVALUATOR/CONTACT:
J.Kauffl!l<ln/G.Lanik L 1 CENSEE:
SUBJECT:
Corr~cnwealth Edison AEOD CONCERNS REGARDING THE POWER OSCILLATION EVENT AT LASALLE 2 (BWR-5)
EVENT DATE:*. Marc_h 9, -1988 _:
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. The LaSdllt ~vent in.vclved 'power.*o~c*illations caused,.bi'neutron *flux/th~?inar *:-'.*...
. hydraulic* instabilitie$ of a maonitude that wt::rt::* not: predicted* by* d1::s19n.. *:_.:.-
- -.:~--*7'. :analysis; unan-ticipa_te1{b/'the*opt:rators, and poten.tia11.yJn-cont'lictwith:\\~~
- -:-.. *General Dt::sign Criterion (GDC} 12...,_Based on ven*dor analyse:*s, two NRC Gen*eric*:;/
- .-*
- _*.-'_* Issues.::(Gis} had prevjously*.-been r~solved conce*rning sta.bilfty'o(BWRs;.'aDcf':zJ
..:.,~.::-: ~-
- -;:. thfs evE:nt:raises que*stjons regaraing *the. ad~atiacy of those -resoi*utions*:*(.:::*\\~~~-
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- predi c.ted. thd f these* oscillations would* not oc*cu r*,"* littl c"i-'~_f'~~---
. :::-":t,.::.::
.,;: :*:J:: :r.,:::* guidance-and t_ra in i ng,. were provid,d". for opera tor detec;t i. ori_')nd re.spo*nse ~~ ;:;:.*<*~fS:~:t/::tP
,, :-'.:-'.i*.~;;-.::'.;.Furthe r ;., 'operat fan 'j ri_ unstable a*reas of the-~ BWR power/f:1 ow:'.ma p*tnas -::po-ten*t fa L::i~_.,;:_;,:..:_ J;'..,... :t
- \\I~~-*,::::;-:a~"'.~~-se _safety.'conseq_ue~ces ;::::_ Beca~~e. l.aSa 11 e t* s. core";_~as{~caTc~\\~_t:e*d'.*:to. -~~e"?:..~,;fj,t~~f;::.. )~
- -* -- *.
,..,this problem.. 0 **
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- *ir* lioht of tht-present* uncertaint1es,* we recorrmnd be re~u1red to implement proce~ures to
a)
Immediately 1nsert control rqds to b~low the 80t rod line following reauct1on or loss of r~circuJation flow or other transients which result in entry into potentially u_Q.stable regions of the power/flm,, map.
b~
Incr~&~t:: recirculation flow ouring routine reactor startups and insert scme control roas prior to reducing recirculation flow below 50~ during shutdnw,,s to avoid operdtion in potent1o.lly unstable areas of the; pm,:er/flow map; c) lmmedidtely scram the reactor if a) orb} abov1:: ore not successful in prevEnt1ng and suppressing ostillations.
We also recorrmend that NRR revisit Gis B-19 and B-59 and ATWS m1t19ation in light of the LaSalle operating experience.
Description of the Event (Compiled frow license~*s 50.72 report, ~arch 9, 1 S80, and references l through 5}.
~h1le performing the functional test on a differential pr~ssure s~1tch, an instrument maintenance technician inadvertently valved in the variable and reference legs with the equalizing valve open, thereby connecting the variable and reference legs.
Thfs initiated a "pressure equalization" between the variable and reference legs, and resulted in a high "indicated" level to the 8800170076 880608 rq ft>R Ftboct 050Mfl1r
f feedwater level control sy~tem, causing the feedwater pumps to begin reducing flow.
Realizing a valving error WdS w.ade, the reference leg was 1rrrned1ately isolated from the variable lea.
This resulted in a law "ind1cated" level
$p1ki:.
The levt'l spike caused uthe,-. ltve.l switches, util1z111a the same reference leg, to also actuate, including the trip of th~ reactor recirculation pumps from an Anticipated Transient W1thout Scram (hT~S) signal.
Due to the rapid power reduction from 84~ to approximately 40% caused by the trip of both recirculat1on pumps, feedwater heater high level alarms were received and heaters began automatically isolating.
This resulted in reduced feedwater temperature and the insertion of positive reactivity due to the negative moderator temperature coefficient. With feedwater level control -*
aaequately handling the: level tran:.ient, the lic~nsee tried *to re-establish*:.~*--
fet:?dwater hE:ating and to restart the recirculation pumps. _Atte.mpts to-restart'.-~ '.' _:_
the r:circulation pumps were unsuccessful.
-*-~
~
~
- With the unit in a h1gh control rod_ line c.ondition~*(power was ss;* prior. to: the.,.;_.. **--.. _:.;~
.. event) ano low flow cond_ition (natural c_irculation)_, the unit.-sta'rted,:.', ~- *1.*-::*. _: _-,-::1'
- exper.iencir.a nt!utron flux oscil-ldtions from rapid 'creation"and'co11apse *of *.~,<:.... :*"':
_ voids in th~ core reg10n.
Approximately 5 mrnutes into the event, mult)pfe_--_~.:
1
.~
<. high and low _alarms wer_e-recorded_-by th1= local power range:mo*nnors: (lPRMs)=. *".*-=*:*r:~,,t.i
--* *.: The *a*verage power* range monHors *(APRM} recorders were osci 11a*tinq _between 25'.t"."
- f*:- * *. __
"",**:,_----"*and 50'1'. of full power with an**approximate 2.to 3-second. per~foa-.-=***"Because_'of.~. -~.
- -~~-
-- limitations of the-APR~ recorders, the actuaf neutron flux *os'cTllatians~-.. ".*.::<
- _.'" _ --*(approximately ?Sr powe*rt*.w~re larger_.than the indications* of -the APPM -;**. ',
_ - -. ".. **-: recor.ders... _ The_ c9~t_ro 1 -rooi.n~9perators. were -ii] the *process*: b'f'.' *rr~nua l l'y:-_;-.:
- _:: :2 *~*-:.* :*scrarrming thE! unit, when.an automatic;-,scram occurred on upsc'ale* neutron.""trip-
- (1181 on-APRl'As)." lrrmediately prior to the scram, the operators noticed that a
.majoritv of the LPRM Hi alarms_were. lit.
The sEtpo1nt f_orAlre*LPRM. HL alarms*
is 105~* of full scale.:.'
Foreign nperating Exptrience.
A number of power oscillat1on.~venff have been reporttd by the NEA JPS system.
Power oscillations were reported in 1985 and 1966 at a foreign e~R-3 in i~S-677 and 6?1.
The osc1llations.:"were 14':'. pe:-ak to pe:ak aurinc r:atural circula-tH.rr, testing.
!n June 198? iri JRS'-22(,,
d foreign IW~- 4 reportt:C oscillations of 75r of tht "mean" flu>" durinq forced circult'.tion after rr,oving one: cor.trol rod.
Thf rl::actor tn~~Ed on APPi' i-:igh F1u> <:dlH five p:::-;~ half SCrdfl':-'had been reset.
Th~se power osc1llation~ had a 2.5 sec.one period.
- ~ response, operating limits were e~tabl1shed at that facility t0 prevent operation in the drca of instab1l1ties. Another event (IRS-220.2) at this reactor in January, 1983, demonstrated that it is pos~1ble to start these power osc1llat1ons from nc,nnal optrnting conditions.
IRS-363 reported that rn October, 1983, during testing at the same reactor, divergent, out-of-phase oscillaticr:s were experienced.
The report describing this event stated that this ~as "a potential GDC-12 violation." Again, operating restr1ct1ons were implemented that require rapidly maneuvering the reactor to a stable region following a
~,ngle rec1rculation pump trip.
Information received as followup to these events indicates that operating instructions were also cevelopea for loss of feedwater hedt1ng events, loss of all recirculation flow, and low recirculation flow cond1t1ons.
We have also rece1v~d information that follo~ina startup testing at yet another foreign E~R, ope*ating 1nstrucl1ons were i~plemented to prevent routine t!ntry into,potentially unstable <1reas.
In particular, guidance was developed ta prevent routine entrv into these dreas during reactor startups ano shutdowns, to require increased monitoring of APRMs and lPRMs in potentially unstable areas, and.to provide guidance for operator response to ctrtain transients such as loss of feedwater heaters and rec1rculat1on pump trips and restarts.
In sumnary, these foreigr1 plants have taken action to
~strict or prohibit operation 1n areas of instability.
Figure 1 is an example of operating restrictions during startup and shutdown in place at one foreign BkR.
U.S. Operating Experience Other than L~Salle, no events involvi~g d~vergir.6-power oscillatibns-at-BWRs
"'Ere identifie:d in the SCSS operating experienci::*data base.-_However,-startup
___ :;*~
testing and other testing have included inducing power osc11la*tions, -observing*"'*:>-*:,-.*:.-:_.;'
the reactor response, and test inc the: *effectiveness of osci llatio11 suppression ~-- :
-::-- :..~
mtthods.
Review of "the: data.'S~se Sl;Ce '19so did *capture 167. ~*.,*ents involv',ng a *.tr,p of_-* *:_,,
,*a ore or two recirculation pumps whilt the reactor was*criti,cal.-. *n,us;*;~hen -:-.::-*~ ~-.. ~
. combined with *rqutine sta*rtups ana_-shutdowns, ~it *1_s clear t~~t_ *BWRs are*
- -.:~:
~-
.; *~~--:
~
frequently opera*ted in poten"tially urlstable regions.~ The number of reported*~.~"~~ _::'... ~~ ~_.
- *.*events -1s low since there are*no reporting requirements"for -recirculation pump--,_.--
- _-,.* -trips, *unlE::ss it is in conjunction*with some other, reportaoTe condition:
- .~*._,..
- ::
- , *.. Scnall power oscil_la_~ions_are simi_larly not re~_ortab~e.
':*: ~-,. --* **
--:'c-...
- .?~"'
-~.... -;
~
~;;:-.....
1
~-\\,*;:. *,:::-_*_...t-:--.-;:_~-
--Related GDCs and Gf-s ---:::__:".,"'-- ~-~_--,:_*::,_ -.
._,, ~ :- _.s-:-=* -~
The LaSdlle event relates to two GDCs and two Gis*:.* - _-.*
-~.::
w..
~~...
"GENER~L DESftN.CRI~ERION 10- - ~eact;r-6esian~ -The reacto; core and associated cooldnt, control,- and protection* systems shal 1 be de$1qned with appropriate margin to assure that:*sp<:cified acc.tptable fu1:l design limit~ are not ~xceeoed during any condition of:--nonnal operdtion, including the effects of ar1ticipated or,uational ucc*urrenct:!s.
1 "GENEr.JIL OESIGt: CPITERlON lZ - Suppress1c,n of J;eactcr Power Osc1iiations.
The reactor core and assoc.iated coolar,t, contrvl, and protection systems shall be c1:s1gn1:d tu a!.sure thilt p0wer oscillctions...-hic.r c.an r~su:t 1ri c0:1ditfor1s exceeo1ng specified acceptdLlt fuel design limits ar~ rut possibl~ or can be relidbly and readily detect~d and suppressea."
GI 8-19:
"Thennal - Hydrculic Stab1lity" and GI ~-$S: "(N-1) lo('lp Operation in BWRs and PwRs".
These Gis were closed out by the issuance of Generic letters 86-C2 and 86-09.
Generic letter 86-02 stated that the approved GE and Exxon methods for calculation of core stahility decay ratio are ur.certain by 201 and 25~.
respectiv~ly, in pred1cting the onset of limit cycle oscillations (~ecay ratio
= 1.0).
The Generic letter noted, "... BWR 4, 5, and ts rr,ay not bE: cble to show compliance with GDCs 10 and 12 solely using analysis procedures to prove that thennal hydraulic instabilities are prevented by des1gn.
11 However, the Generic Letter concluded that BWR 1, 2, and 3s sh0uld have sufficient marain.
It also stated that for cores which do not meet the analytical criteria (decay ratio less than 0.8), the operating limits of GE Sil 380 would be sufficient to provide for detection and suppression of flux oscillations in operating regions of potential instability adequate to demonstrate compliance with GOC 10 and GOC 12 for cores 1oadtd with approved fuel designs.
Generic Letter 86-09 noted that the review of 8~~ (N-1) loop npf:ratio~ was ccr.,~il1cat1::a by potential thE::rmal-hyoraulic instatnlity and 2f!t purr.i:; vibration problems during single loop operdtlon.
In lo~ flow optrat1ng regions, it was necessary to dev~1op special operat,,,g procedurts to assure thdt GOCs 10 and 12 were satisfied in regara to, thennal-hydrdulic instabilities.
~lant Technical Specifications consistent with these procedures were accepted by the st.aff for r~dctors which were not oemonstrahly stable based on analyses using tht then approved analytical methods; details of the operating limitations were developed for GE SIL 380 and contributto to the resolution of GI B-19 *
. In addition*, tests at Brown's Ferry demonstrated that single loop operation
'had similar stability characteristics as two-loop.operatfon under the same
-~--*
- -. -.. ),pcwf::r/flow operating conditions~:~*.The tests* conf.irmed the staf.f'~s find1ng.. that
~~ _
-..,,, *Ttchn1cal Specifications.based on-GE, SIL 380 wlncti were* propo'sed' for-_fome B\\.:Rsc,:_~ * :..
- >.. :were*dppr,opriote f9r.the.detection ar.d suppress1~,n.uf thema*J hydrau*H;c,>:-~..,;_'.:.
_. -~*.:_
- _;*'***instabilit1es. *The staff*expected to approvE-sin,g1e**.1oop operat10ri*for.*.:*':. * *; **-.*...... **
':-'.: *.*- Jicensees_ who f?Ubmitted the appropriate ECC:S_aildlysis_:_.,
.-~.-'.':*,<~(:;.,,4
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- .' Relevant L i~erisfog" Actions...'*
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- ,,::::---t;:>-' :, ~: - _,,;. :.= - -
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?:. *\\. :::::._~~<:The fore i,gn'* e.ven r."*fovo l y"i rig oiJ t-'of -phd se, *-dfverge_n Lose Hl at i o.ns ; :resu_tt_ed_ fo* :-.:j.::.:::..:;;:~.~
- ,,..*,_:: :-::.issuance of a~board notificati.on.(No.84-062.) in March,, 1984;.::**,.stabil.i,ty testS': ___ i.>.-a:*::~J,..,_::;.;;
--d-£:m~!:S tra tt!-d_ tha. t;: II Ji mit. eye 1 e_ as*~ 1 J 1 at i,on s II*._cou 1 d occur.wi.th'i.n.. pe nni s~a-b 1 e.;;*..if{_j.J{:~7-i
-~o~~,-r~~f_ng_,~pa_ce_. be_fow fhe r?t_e.d,r9(Vne_ a_t_-na~~-~-~,1 ~~r~ul.rt,i9n~f,~ow-::-:-;}t,e,;:~;f:s/,X~ *-3:?l~£
--.1.::::,h1gh power l~vE:1 {J20.1).;.scram protection which'.'lS bas_eo on-APRM Sl'onaJs~would-:-:-.;,..".
- ,;;;.-:::*c8
.,,_,,,. no t"-ne*c.es s~-ri)y'~pre_v~n tvj O i afi on -0 f_ C ri ti Ceil b~a t.:nu! _*-.1 _ifn:i t'S~!~ f: 'sl(ch';'J oc:a*J~-;,::,:-:::"t
'<i-:7~?~
.-ii ns ta b't 1 it 1 es, ='\\o,c're".:"to. occur: -:---r he~:les t ~-demcn's tr'att.d tr.a 't-*Toca*-r: th 1: rrnii:l "~- ::i:. -~-,:.,;.,.-::4,:~-~ ;~*:: **'
- hvdraulic csc,lla.tions which arc out *of phase with the APR/1s could occur'.
I*t
- .: ~- -:.-=--*~w~s *unclear *a t'"that time* { 19E'4) how_'hioh' a, 1 oca 1' osci 11atio';* cou 1 d. reach"-~,""-::;*
~-
- .,,, 'btfore detectio~::-by *an_\\iperatrn<fcrew*\\js;ng the'n* curren(**rnon'itori-*
procedures. ** *.
T~,s bo~rd notification ~as mad~ atler the issudnce of GE S!L 38C. which,~
currently usec dS guidance to,operators for the~P type of events. Plant Technicbl S~ecificat1un changes Wfje_made for piants undtrco1nc l1cen~ing hearinGS to dddress the concerrs of this board ~otificatiun.
Prev1ous Vi::ndnr PecommE:ndation~
General Electric Co., h~d prev1ously identified in GE SIL 380 anc other dccuments th~: the cond1t1on of hiah rod line and loh flow was susc~pt1ble to neutron flu:i-/thermal-hydraulic oscillations.
Howt:!11er. based upon onalysis, Commonw~dlth Edi~on did not believe such oscillatic~s woulo occur at LaSalle, and as a result, the SIL was not implemented.
Because th1s event at LaSalle involved larce power oscillations, General ElE!ctnc Co. has issued Rapid lnfonnation CommuniCdtlOn Services lnfcrmation Letter {RICSIL) No. 006 Pev1sio11 1 pt:rtaining to BWP cart:" thennal hydraulic stability.
The RICSIL supplements GE SIL No. 380 P.~vision 1 on tre samt subject.
Concerns Regardino This Event f
- l.
I Stability analysis methods are highly uncertain. LaSalle 2's calculated decay ratio was approximately 0.Q for this fuel cycle.
This means that that tht transient reactor behavior that was observ~~ during this event was predicted not to occur.
The licensee's review of this ev~nt stated that the conditions present at the start of the oscillations appear to be only slightly more severe than the assumptions used to analyze the LaSalle decay ratio. There is also infonnat1on that indicates that the stability analysis for Vennont Yankee was shown by stability tests as non-conservative (Ref. 6),
- 2. *. LaSalle operators were not trained for this:type cf*event.~*Because_*~E".-.
--~.,:. -~ _ -*
,:analyses predicted that this event would not_,occur at.LaSa.lle, GE*SIL ~80-_*_: __ -_,_* :::':*
--.-"'*-:_-*-/'.was knowingly not in place and operators not' trafota on* _Gt. SIL 380 _at:_-:-.:-* -.*-f,.--
- ~*_.-: _
-~.. -: ~. * ~as~ 1_1-~~: as. a 11~~,ed**._by Generic. ~e_t~er:,_86:~o_::?:?f.~**'.'.~:.:! ~~- ::;:::~-:_;:
- -:. -.. ~~ _ *... * *. _*_,"f~:~, *)~ ~'-
"~:.' - 3 *. :**:GDC 12 may h~ve been. violated.*_ A,1though** chemi'stry s*amp*les' foT1owi_n-g ;t~e *-::-,
.,.::. ::,. * ~-; L.a Sa 11 e._ event.did_ r,ot __ disclose _any.fue 1 d~~;?*ge, ~ the.-evin'Cwfa\\ ~go teri t ia:J li~t-
- -:, a v 10 lat ion of GDC 12 in= th,n irndampened power o_sci lJ at io11s oc_c_u_,:-red :~a.n_d __.,.:';'::*:-~ -,.,._
- .. _'.-:.- -~:.: -. -~
no: procedures :or* methqds. were*-_; mp 1 emen ted t*o. -~f;' l i at?Jy* -~!Id r1:a q i1 i*de~tct'.~:: ':.:., _'
arid.su*ppress**these pow1=r-6si::illations*
- c*
,_._:*. *:._. '""..:: ;.<-*-3:..,::*
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- .:~,:/,., _~,"--~-i: :.-,-_;_;-~~~~~!r;... ~~:--:.:*~i
.iiP:.c.,.t:1~-~;i_ Otht!r:* sHR,s;may{ have _a: su~ce~t f b.i-1 i ty ta _uns_tatne-_powerj>s_ci) l' at 1 ons*:~t.~::::=J~~t~~r~-?~-i'1t~~~
,.. /.;_:?:~=s:.-~:<-:.ijecausJ!_:. _ana_}yses s*fmi lar_J_o. t_h~,,.a~es _useg at, 'LaSa 11 e ~!'e -~~ea_ ~t --?,~~-e-~.:~;j},::z.t=:r.z ~iii
';V:r.~_..-*:..,:tPlants:_to*:me~t GOCs. 1~-,a~d. ll,.*t~,s _tra_nsie~t-resp~nse-co~}.4.}J.f!=~cr_ ~!:J:.;;~}ii~};;,;~t l
~---~~:*":_._:,.l',-?*.:'.Other BWRs.:-w.1 th decar_ra vos:,;ress*.~ than, 0. 8.;,
- __ Hke;_taSa 1 Te;' ities_e *p_the_rr:--;;1.;_,~;,:~*:*:~>>~
'.',.:;t*,"
BWRs*ma"_y riot*tfave i'mple-rriE:i1ted'*proct-dur*e*s*to-"'riiia6ly dele*ct**a-nd *suppress,~-"-:'-*.<'at,':,~~
power oscillations.* At laSalle, thE: operator~. allowed)ledrl,r' ~-w~-ri:tin_!J~_f!_S -,.
.,*-*.,.-:~~--:.,**-:.."'.of unstdblt. operation.'-he'fore <;!ecfaing_to tak.eJc;.tion_to: shut.d_own the** __ ;;.-;,:~
~;;,
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- 5.
~-
'l GE SIL 380 Revision 1, eve-n"H i_mplemented,-is *inadequate-to ensure*
compliance with GDC-12.
This ra..ises the issue of the adequacy of GL 86-02 in assuring that GOC-12 is ~t for plants with predicte:-a decay ratios greater than O.~.
The SIL has a r,umber"of inadequacies:
r.
--APRM noi!,t" and not actual rapid power change~ 1s discussed as a
- result of f1ow instabilities.
--This noise is said tc, normally ranqe betwe~n 4-1,~ {peak-to-peak) of roted power, whereas LaSalle reported puwl.!r oscillations of nearly full scale (75r power).
--Some of the terms are not defined or cormronly undtrstood b1 utility operations personnel, e.g. "1imit cyele c,sc11lation." This makr:s it difficult to use as the basis for pperdtor gu1danc~ and procedures.
--Power oscillations ~ay not be r~&dily identifi~d and suppressed.
[luring an event with numerous fdilure~ afl(J alarms, it is net certain that operator attention will be promptly called ta power o?c1llations, especially s1nce the APR~ instruments typically hav~ largt osci11at1ons (noise up to 1oi under normal lOG'.t power steady state op1=ration) and the APRM recorders do not show the full magnitude of power oscillations
- 6.
due to til'DI:! delays.
as nonna 1.
i Operators might consider any indicated oscillations
--The basis for the proposed actions is apparently non-conservative or sensitiv~ t0 small parameter c~anges.
--Guidance is provided without explaining in detail why the actions are taken or th~ bases for the actions.
Even in the case where out-at-phase osc_illations were experienced, GE SIL 380 states that "very large margin to safety limits were maintained." This downplaying of the potential severity of thennal-hydraulic instabilities may mislead operators into thinking tha~ the stability concerns are not important.
Operator traini\\19
- on recognizing_ a*nd respo~1ding, to/power oscilla~i:cins :*.;;-. -
--~ ::,.* *. S*,
poor.
Few, if any, simulators used by utiHties are capable of modeling*::**.-:-'.~:-*-*:*
the type of ciscil1at.ions that-occurred at laSalle._.:,,.Si-nce*the eii,s.tJ_ng. **--:-_.f_.;.=;_
- * ~*guidanc1: in GE SIL 380 does not state 'that power osdllation:s~.frpl!! *:/~~:~ :: -~ -::=_:_:**. -."<ji
- .*_*:o to 120% power are possible and have been experienced, i_tjs li~ely_that:~ :.. s_ *:*
.. * *'. YHY few* licensed operator:-s 0r _trajning_ instructor.s Wtfr!:'!_f;:ve:n _awaJ:e J_tiat._': *
- ~*
.-'.. *. oscillatfcn of this magnftude.cou.ld.occu*r.. :rt._oper:.ator.actfori* *1s* necessary*
- ~-,* : _ _-.. t9_ensure compliance_ with.the GDCs, it is es*sential: th_at-li'cens~d*opt/rators.-'
-~:-
-<-~ be*.trained re9arding*.the.:.assumpti"ons 0
, conditfons,*'.lfmftatfons,-*et¢.-:.of*C;,:".
- -:-. ;.~ * ~s;-~ the opera ti ng_;c9ncerns.
- However\\~-~ imp le gu i.aarice *.:, such a*s: *;_:*i r:_edu_dion'\\-f. ~
- -~-: :*or loss.. of r_eci.~cu1ation flow.resulting in_,e_ntry into_ a pot_e_ntfalJy _':~~--
- ,.~:..:r unstab1e*area,.-,nsert control* rods to below. the 801 rod* line 11 *:-* that~::-:=
0 ** *.*::"'*ensures.avoidance of.the. unstablE!*.ofunanalyzed.. regions. is sfre-fe"rao1tt0tii**~~----y:
- ~. *: reliance. on. opera tor:lllmor,ic to. ensure* opera.ti on.-wfthi r,i-: a na l)'.ZEtd-'ieg i oris*.~p-*
--..:,::*.-:?l~~-.-~::-~*:'.!-*"//'.~~---;-~~~..:,.**~: *:,:,:~ii:-?~*:-.*.:_**-:":;,__*_.¥ ~-,.-f. -:.:?*-:~~~-/~;:.;_.*
7~
- 1mprop~*r operator-*dctfon_s could worsen the *everit.' The operators at
. LaSalle tried to restai:_.t,.rec.ircuJ.ation pump~-because their~trairfing**a*nd_ *,_ *
---.: :*procedu*res* alloweci" therr.::to _do so~..
- In this event.,'. wfth* a downcqme."r-f{Ued*f*.,*
... *:* *wfth cold.feedwater**and *an unstable* reactor, a. successful 'restart **of-*:.-..
/
~..
recirculation pumps would_ lead to* further rapid* reactivity insertfori-with -~.. ** * ----
potential adverse consequ~rices; We are also concerned abo.ut the effects that would have occurred if*addit;"ional reactivity insertion due to void collapse in response to a turbine trip or an HSIV closure had occurred during the power oscilldtion~. <Other operator actions, plant conditions, such as end of cyc1e or different power distribution, or plant transients may have resultea in fuel damage.
~
Sevtral calculations using the BWR Nuclear Plant Analyzer were performed by Brookhaven at AEOD request.
T~e simulation of the LaSalle event is shown in Figures 2 through 5.
By parametrically ir,creasing loop flow res1star.ces, it was possible to ~enerate power oscillations similar to those experienced at LaSalle.
Prelimir.ary results from these runs indicate that large reactivity changes occur during these evehts.
The power oscillations experianced at LaSalle are cyclic interactions of core void formation, flow, and neutron power.
The period of th~ oscilla-tions is about 2.5 seconds while the th~nnal time constant of the fuel is 5 to 7 seconds; and consequently, direct gamma heating of the coolant is the likely energy feedbacr fllt!Chdnism.
This phenomena apparently begins with thermal-hydraulic instabilities arising due to relatively large two-phase resistance in the core, while the driving head dnd flow rate are
., low due to loss of furced ~irculation.
Fonnation of voids then drives neutron power down which slows further void fonnation, resulting in lower two-phase flow resistance, and,increased natural circulation flow into the bcrttom of the core.
This col~ water increases core reactivity and results in a power increas~. The resultant voia fonnation continues the cycle of oscillation.
Large neutron power oscillations are the result of large reactivity changes.
Preliminary results from the Brookhaven analyzer indicate that large reactivity changes occur during these events.
Figure 4, for example, represents tht LaSalle bdse case, where the analyzer calculated 0.5 dollars total reactivity inserted just prior to the reactor trip *. -* *
- 8.
The LaSalle event fs an important.precursor event.* Although-the--:-~:-.*_';-_,:*. -
consequences of thi_s parUcular event were not serious, they could have.:_ ~ >. * < -~-.'/
been worse in other circumstances; First of all, the potential exists for: -
k,calized power oscillations where one half of the core oscillates 180 -.-
_.* degret:s c,ut of phase* w_ith the other* half; and iri that. case.the.APRH trip*.,... :*_,-*. c::,;;~
would not *trip the' reactor until* the amplitude of the* local power._*_-.--*:
- *. - *., - -~
oscillcitions was much greater.* An actual event of this type,.is noted.-irr -_ --._.: __ >::-.,..-*~-~
th~ foreign_opt:ra_ti_rig experience.'-_:-Secondly, the potential _-e*xists for:-;:-_-~:~-_':,~-~-:~.::--_~
- operator action or.plant equipment.failure to worsen the event; for-~':-.---_-
~.
example, restart. of *a recirculation pump or MSIV closure could* res.ult in - _ -
- . ___ *.~?
additional_r_eact_i_v_i_tj_i_n~ert_i?~--*"::*
~;_-> --~*-_.--:-0-~-
_, ___ -__ ---,.~--f h*
,f I
- 9. __ : Previous -e'f forts- *take~ fn _:reg a rd-tt A TWS mit i ga't 1 ori -may be.-:*tna-dequa te; '.~ \\_ - -.:.'* _-f -; _ ;'~. :-~:
-:f:* __ *. The action.of'tr.ipping recircula~ion pumps automatically and_induci_ng an*-_.:.,~*:-*:_-_~r:.:¥,:_:t evtnt similar tc the LaSalle cVent when it is*not clear.,.,here the*power -
__. ~::-.* ;,_,
oscillations would-~top dna whc1t the effect~ CJf these c,~cillation~ would.-----
. be* in the absence of an autorr,atic* scram, ntc.essitotE.s that:AT1r/S *:<-. -~-- __.
--:,.;~(.:;
mitigation be reviewed in light of this event.
---_ *~,- *
- _*--~-:;-.
The resolution of Gls B-19 and B~59 may be inadequate.
The analyses
- which fonn the technical bases for the resolution of thes~ issues have been chalienged.
The LaSallt ev-ent was predicted by analyses to be prevented by design, but it uccurrtd.
Potential Actions to Address the Problem
- 1.
We r~corrrnend that ewR licensees should b~ reouired to develo~ and implement procedures to:
a:
Imrr,ediately insert control rods to below the P.C'~ rod line following reduction or loss of recirculation flow or other transients which result in entry intc, potentially unstable rts1ons of the power/flow map.
b)
Increase recirculation flow during routine reactor startups and insert son~ control rods prior to reducing recirculation flow below soi during shutdowns to avoid operation in poter.tially unstable areas of the power/
f 1 ow map.
..., c)
Irrrnediate1y scram the reactor if a) orb) above are not succ1:ssful 1n preventing and suppresst~g oscillations.
1 We also recormiend that ffRR revisit Gis 8-19 and R-59 and AT~S ~itigation in light of the LaSa11~ ~perat1ng experience
- A
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.. REFERENCES
- 1.
Corrmonwealth Edison Company, LER 8~-003, Docket No. 50-374, dated April 7, 1988. *
- 2.
PN0-III-88-18, dated March 10, 1988.
- 3.
PN0-III-88-18A, dated March 17, 1988.
- 4.
PN0-III-88-188, dated March 25, 1988.
- 5.
NRR Event Followup Report 88-03, dated March 30, 1988.
, :* PDR)
- 6.
Memorandum from L.E: Phillips (NRC) to M.A. Ri-ng (NRC}..dated.April 7/-:._,.~;:~*
- 1988.. (Not available in POR), __.
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U.S. NUCLEAR REGULATORY COMMISSION REGION III Report No. 50-373/88008; 50-374/88008 Docket No. 50-373; 50-374 License No. NPF-11; N~F-18 Licensee:
Commonwealth Edison Company P. 0. Box 767 Chicago, IL 60690 Facility Name:
LaSalle County Station, Units 1 and 2 Inspection At:
LaSalle S1te, Marseilles, IL lpection Conducted:
March 16 through 24, 1988 Inspectors:
NRC Augmented Inspection Team Team Leader:
M. A. Rin; -r-~/
........_,.,,--1'(
- I Team Merr.bers:
R. A. Kopriva,;r,/li.,_111, L. E. Ph111*~/~?
P. Shemans~7 f~.,
Approved By:
W. L. Forney, Chief *,,
Reactor Projects Branch 1 Inspection Summary
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Date
~v/ 1--gi Date Date Ins ection on March 16 throu h 24, 1988 Re art No. 50-373/88008 DRP 50-374/88008 DRP Areas Insoected:
Special Augmented Inspection Team (AIT) inspection conducted in response to the dual recirculation pump trip and subsequent core power
- oscillations resulting in a reactor trip on March 9, 1988, at LaSalle, Unit 2.
The review included root cause determination, safety significance, performance of operators and equipment, adequacy of procedures, effects on the r~actor, reporting actions and potential generic implications.
j Results:
No violations or deviations were identified; however, the licensee has committed to procedure and Technical Specification changes as well as further study in the areas of inherent shutdown mechanisms, instrumentation capability and uncertainties in the decay ratio calculations.
The licensee's interim report, as required by the CAL, is included as attachment 5 to this report.
2
0 is needed to assess the nature and magnitude of neutron flux oscillations and the safety of restart after an instability event.
LaSalle and some other BWRs do not have high speed data recording instrumentation which can be committed for availability during plant operation.
- 4.
Oscillation Characteristics Some characteristics of the LaSalle neutron flux oscillations were atypical of previous events and have led to concerns about the applicability of previous safety analyses.
The magnitude of in-phase limit cycle oscillations previously observed on the APRMs during special stability tests and operating reactor events were typically in the range of 5~ to 15% (peak-to-peak) of rated power, and as high as 25%.
This compares to peak-to-peak values of about 100% at the time of the 118% neutron flux trip for LaSalle.
The estimated value of local power at the time of trip was greater than 310~ and LPRM readings indicate that the core power peak shifted and increased by 25%.
Even though the fuel LHGR limit of 13.4 kw/ft was not exceeded because of the thermal time constant of the fuel, the increased power peaking was unexpected based on Vermont Yankee stability tests, and was not factored into the generic safety evaluation performed by GE during review of the thermal hydraulic stability Generic Issue B-19.
The previous GE safety analyses considered several limiting moderate frequency transients which were initiated while the neutron flux was oscillating below the 120% scram setpoint, and included a rod withdrawal error with the flux oscillating up to the 120~ scram level. Additional analyses were performed to evaluate the impact of oscillations that approached 300% of rated neutron flux (e.g., regional oscillations) without scram prior to rod insertion and termination of the event. All of these analyses showed that significant fuel thermal margin existed to safety limits. While there are several aspects of these analyses which differ from LaSalle (initial power level and amplitude of the oscillations; no change in bundle peaking factors due to the event, etc.), the AIT agrees that they are sufficiently representative and conservative to demonstrate that no fuel thermal or mechanical limits were exceeded during the event.
However, reliable detection and suppression provisions are necessary to assure protection against future events which could involve regional oscillations to higher power levels.
The licensee was also asked to review the impact of the event on stability considerations addressed in the 1979 GE Generic ATWS report, "Assessment of BWR Mi~igation of ATWS'1 (NEDE-24222).
23 I
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r The report does specifically investigate the sensitivity and potential impact of limit cycle neutron flux oscillations up to 500% of rated bundle power following recirculation pump trip.
It was concluded that oscillations of this magnitude would not result in sufficient fuel clad temperature variation (130°F) to affect fuel integrity. It was further concluded that a loss of clad integrity due to prolonged exposure to limit cycles.was an acceptable consequence in view of the importance of the recirculation pump trip (RPT) to minimize the energy deposited in the suppression pool (thereby maintaining containment pressure within limits) during an ATWS event.
- 5.
In view of the large magnitude of the APRM oscillations in LaSalle, the AIT believes that the ultimate power level without scram is unknown, and that the 500% level assumed in the ATWS investigation may not be bounding.
LPRM oscillation magnitudes more than seven times those of the APRMs have been observed in the case of regional oscillations. The licensee reports that the BWROG is discussing this issue (inherent power limits) and the licensee will provide a status report on July 1, 1988.
Addi tior1a l Concerns Several additional concerns were presented to the licensee in the form of questions. These questions and the licensee's response are contained in Attachment 5 to this report.
- 8.
Recommendations The AIT recorrmends that the concerns identified in items IV.A.I through IV.A.5 of this report be examined by NRR for generic and LaSalle specific resolution.
In the interim, the AIT recommends that revised stability TS as discussed in IV.A.2 be developed for LaSalle Units 1 and 2 and the licensee be authorizied via letter to modify interim operating procedures provided they remain consistent with the new T.S.
The revised technical specifications and procedures should incorporate the changes surmiarized in Attachment 5 (Appendix A; Item 3), which include immediate insertion of high worth rods and observation of APRM/LPRM noise when no pumps are operating and power is above the 80% Rod Control Line.
The reactor is to be tripped immediately whenever instability is suspected. It is expected that the time available (greater than 5 minutes) to instability following a two pump trip transient is sufficient to permit manual power reduction, avoiding the need for reactor trip unless the core is unstable by a large margin.
Proposed procedures permit manual action for up to two minutes (prior to scram) to reverse operating actions which may result in small margins of instability when one or both pumps are operating.
V.
AIT CONCLUSIONS The AIT finds that the core power oscillations observed on LaSalle Unit 2 on ~arch 9, 1988, were initiated by a personnel error resulting in the 24 I j
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Ex.,H I B rr c(
A Report on Reactor Study Issue Number 25 Prepared for The Ohio State University Expert Panel Nuclear Education & Training Services, Inc.
Final Draft September 2, 1988 Dr. William P. Stephany
Issue 25 1
Description of the Issue The Status Report of Potential Safety Issues section of the Update Report entitles this Issue "PEAK PRESSURES IN ATWS CALCULATIONS FOR BWR/3 PLANTS" (Ref. 1, p. 219/459).
The Reed Report expresses concern about down time when backfit modifications are implemented for operating plants and, more generally, about the impact of anticipated required design modifications on plant availability.
Note that the BWR/6 design, specifically, was not being questioned.
This question for BWR/3 plants was raised by the Materials, Processes, and Chemistry Subtask group.
However, the Nuclear Systems Subtask group raised a broader issue related to ATWS for all BWR product lines that wasn't included in the issue title or description in the 1978 letters to the NRC from GE (Refs. 2, 3), the source for the pda.:e Report title.
That the ATWS event raised serious ques.:ions which apply to all BWRs as well as to PWRs was discussed in an NRC Staff review, also in 1978 (Ref. 6).
ATWS (Anticipated Transients Without Scram) events are significant in reactor safety considerations because of their potential for causing melting of the reactor fuel and the consequent release of a large amount of the fission product inventory.
Transients that isolate the reactor from riormal cooling systems have the highest probability of occurring and, in general, the severest potential consequences if the scram system fails.
Closure of valves in the feedwater or main steam lines, or tripping the main feedwater or condensate pumps, can isolate the reactor and interrupt the transfer of heat generated in the core.
Real or spurious signals indicating off-normal conditions or events, such as the loss of offsi te power, can initiate such transients.
N.ormal ly, the reactor protection system will scram the reactor, limiting the consequences of these events to moderate perturbations of the reactor power and pressure.
If, however, the scram function should not occur, transients which isolate the reactor pressure vessel could produce pressures which challenge the integrity of the vessel and connected piping.
Discussion of ATWS began in 1969 when the Advisory Committee on Reactor Safeguards opened a dialogue on the possibility of common mode failures reducing the reliability of protection systems to function properly during an anticipated uansient.
This resul.:ed in the NRC publishing WASH-1270, "Technical Report on Anticipated Transients Without Scram for Water-Cooled Power Reactors," in 1973, which required that ATWS events be considered in plant safety analyses.
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The manner in which ATWS events are treated in the design and safety evaluation of nuclear power plants has been controversial.
The nuclear industry criticized the NRC as being overly conservative.
The industry took the position that the high reliability of reactor protection systems makes the probability of an ATWS event negligibly small and not worthy of consideration as a design basis.
The ATWS question was identified as a Task Action Plan (TAP) item and determined to be an Unresolved Safety Issue (US! A-9).
The issue was formally resolved on June 26, 1984 with the incorporation of the "ATWS Rule" into 10 CFR Part 50, Paragraph 62.
The Perry Plant has incorporated the prevention and mitigation modifications required by 10 CPR 50.62.
There is, however, a concern raised by OCRE (Ref. 17, p. 7) because the Perry standby liquid control system is not automated, i.e., injection of the boron solution into the reactor requires manual initiation by the reactor operator.
The issue of automatic versus manual boron solution injec~ion has been a point of debate since its conception.
This is discussed in later sections of this Issue Report.
In addition to the standard review process of the plant's compliance with NRC requirements, the ATWS issue requires addressing the plant's response to the Salem ATWS event.
The Salem ATWS event is briefly described in Section 6.2 of this Report, with the documentation of the NRC review of the Perry Plant's response to the event provided in Attachment G.
This report concludes with a commentary on the implications of the recent LaSalle transient, during which that pl ant experienced instabilities and power oscillations following a trip of both recirculation pumps (Ref. 16).
The LaSalle even~ is relevant to this review of ATWS since a major feature of the BWR ATWS mitigation system is the ~ripping of both recirculation pumps to reduce reactor power.
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2 Concern Described in the Reed Report General Finding 16 in the Reed Report expresses concern that regulatory requirements resulting from ATWS review wil I cause a few percent unavailability in operating plants during the year when the required backfit modifications are implemented (Ref. 1, p. 263/459).
The Materials Subtask Report notes that peak pressure calculations were made under postulated ATWS conditions for a number of plants, with the results showing that the reactor pressure vessels were not challenged:
Calculations of peak pressures under anticipated transient without scram (ATWS) conditions have been made wt thin the past year for various BWRs.
Peak pressures in the 1600 to 1650 psig range have been calculated for certain BWR-3 plants and considerably lower values for other BWRs.
These pressures are well within the capacity of the vessel.
(Ref. 1, Vol. 2, Materials, Processes and Chemistry Subtask Report, p. 16)
This same Subtask report recommends that GE I s evaluation of vessel integrity under ATWS conditions be documented (Ref. 1, Vol. 2, Materials, Processes and Chemistry Subtask Report, p. 18).
The concern stated above is directed to BWR/3s.
The Nuclear Systems Subtask Report explains the basis for a more general concern, extending beyond the BWR/3 product line:
Consideration of the ATWS event arose from concern by the Advisory Committee on Reactor Safeguards (ACRS) and the Nuclear Regulatory Commission (NRC} as to the possibility of common mode failures which might disable the reactor protection (scram} system when called upon to act.
The NRG established a position (Document WASH-1270) which required water reactors to accommodate this event without compromising the pressure vessel and containment.
The NRC position for plants about to go into operation is (essentially} that backfit will be required.
For alread.v operating plants the type of backfit to be required will be determined on a case by case basis.
While maintaining that this event is too remote to require consideration, GE developed and submitted to the NRG a method for protection against it which consists of:
- 1) Trip of the recirculation pumps to promptly reduce the reactor power level.
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- 2) Automatically initiated injection of liquid boron poison to shut the reactor down.
The injection would be earlier and at greater rate than the manually initiated system now existing on all plants.
- 3) Use of high pressure emergency core cooling systems pipes {ECCS) to inject some of the boron.
This method for protection is now under review by the NRC.
It has not yet been accepted.
The recirculation pump trip is already, or will be, installed on all BWR/4 and later models.
This function is not expected to contribute significantly to plant unavailability.
If required and approved by the NRG.
the automatic. fast injection of liquid boron is expected to be installable during a regular refueling outage.
If it requires additional outage time this would add 2%/week to plant unavailability, one time only.
(This estimate is based on a prellminar.v evaluation.)
Inadvertent boron injection is a serious threat to plant availability.
Preliminary estimates, assuming very carefully designed systems, indicate that one such event is to be expected during a plant lifetime and that this would require several weeks of cleanup.
One event in 30 years with a 30 day cleanup means one day per year unavailability.
(Ref. 1, Vol. 2, Nuclear Systems Task Final Report, pp. 32 -
- 33)
The Nuclear Systems Task Final Report indicates that the recirculation pump trip was already installed or would be installed on all BWRs that were currently being bui It ( BWR/4), and in all future models.
Automatic, fast injection of liquid boron was anticipated as another feature that would need incorporation.
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3 General Electric Company Statements on the Reed Concern This Reed Issue is expressed in the Status Report of Potential Safety Issues section of the Update Report as follows:
The Reed Report noted that. under certain assumed adverse conditions, pressures in the 1600-1650 psig range had been calculated for certain BWR/3 plant transient events.
Discussion and Status The Reed Report commented on calculations for BWR/3 plants which show possible peak pressures in the range of from 1600-1650 psig for anticipated transients without scram (ATJiS) events.
The report further noted that since these calculated anticipated transients without scram overpressures would occur at elevated temperatures when the reactor pressure vessel materials are in a ductile state, they would be expected to have no serious damage consequences on the vessel.
However. it was further observed that analyses should be performed to verify this expectation.
A recent analysis for a BWR/3 plant, using state-of-the-art analytical tools, calculates a peak pressure less than the values identified above.
General Electric has developed design modification alternatives intended to comply with new NRC requirements issued* since the Reed study was prepared.
NRG has given generic approval of these alternatives under the ATWS Rule.
These alternatives have been selected by the plant owner as, appropriate for each BWR type and class.
Each plant will receive NRG approval of the design selected to comply with the ATWS Rule requirements.
Solutions *have been identified and are being implemented.
This issue is satisfactorily resolved with the NRG.
(Ref. 1, p. 219/459)
The Executive Summary of the Update Report states that GE submitted a study to the NRC in 1976 showing that the BWR scram system possesses high inherent reliability and that, since the probability of an ATWS event is so low, the additional automatic boron injection system required by the NRG is not necessary.
Paraphrasing the Executive Summary, the ATWS issue was resolved by the 1984 NRC ruling requiring that the BWR have an automatic recirculation pump trip, an alternate rod insertion system, and manual initiation of the liquid boron system (Ref. 1, p. 37/459).
The body of the Update report adds that BWR plants are implementing these measures on a schedule agreed to with the NRC, and summarizes the present GE position:
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Consistent with the Reed review concern, GE initiated a study of the present BWR scram system, including potential common cause fai 1 ures.
This study. submitted to the Nuclear Regulatory Cammi ssi on in September 1976, confirmed the previous GE position that the scram system possessed high inherent reliability, therefore obviating the need for an automatic boron injection system.
This same study also identified areas where the scram system reliability could be even further improved if deemed advisable.
General Electric's position regarding ATWS was that the probability of the ATWS event was so low that no further plant modifications were required.
The NRG issued a rule on July 26.
1984 defining the requirements for resolving the ATWS issue.
For the BWR this rule requires automatic recirculation pump trip. an alternate rod insertion system and manual initiation of the liquid boron system with increased flow capacity (equivalent to operating both pumps simultaneously).
The NRG has concluded that this capability in conjunction with updated emergency operating procedures provides an acceptable solution to this concern.
(Ref. 1, pp. 114/459 - 115/459)
The 1976 GE report referred to is the "General Electric AT~vs Report,"
GE Response to NRC Status Report and ~RC Letter of April 7, 1976 (June 30, 1976).
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4 NRC Statements on the Issue 4.1 Background Information The requirements for BWRs contained in the "ATWS Rule" are:
Each boiling water reactor must have an alternate rod injection (ARI) system that is diverse (from the reactor trip system) from sensor output to the final actuation device.
The ARI system must have redundant scram air header exhaust valves.
The ARI must be designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation device.
Each boiling water reactor must have a standby liquid control system (SLCS) with a minimum flor'I capacit_v and boron content equivalent in control capacity to 86 gallons per minute of 13 weight percent sodium pentaborate solution.
The SLCS and its injection location must be designed to perform its function in a reliable manner.
The SLCS initiation must be automatic and must be designed to perform its function in a reliable manner for plants granted a construction permit after July 26, 1984, and for plants granted a construction permit prior to July 26, 1984, that have already been designed and built to include this feature.
Each boiling water reactor must have equipment to trip the reactor coolant recirculating pumps automatically under conditions indicative of an ATWS.
This equipment must be designed to perform its function in a reliable manner.
(Ref. 4, p.50-32C)
The Construction Permit for the Perry Plant was issued by the NRC in May, 1977.
Therefore, given that the plant was not designed and built to incorporate the automatic standby liquid injection feature, according to the Code of Federal Regulations, it is not required.
The Standard Review Plan succinctly covers ATWS.
The most current version of the Standard Review Plan contains Revisj on 1 of Section 15.8, "Anticipated Transient Without Scram," revised as of July, 1981 (Ref. 5).
This section is included as Attachment A to this report.
The review procedure pertaining to BWRs requires, in addition to more general considerations, that:
there is a recirculation pump trip initiated by high pressure or low level in the reactor vessel emergency procedures are provided for a failure to scram after specified transients NETS, Inc.
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the emergency procedures for ATWS events describe the actions to be taken, including manually scramming the reactor, reactor core isolation cooling, and high pressure core spray.
The actions are to include the prompt initiation (manual initiation implied) of boration by actuation of the standby liquid control system if methods to quickly scram the reactor fail.
(Ref. 5, pp. 15.8-3 to 15.8-3)
Note that this review procedure does not explicitly call for the automatic standby liquid control injection feature, but looks to the emergency operating procedures to assure boron injection is manually initiated in a timely fashion in the event that manual scram of the reactor cannot be quickly achieved.
The major general review of the ATWS is provided in NUREG-0460 (Ref.
6). This document provides the detailed review by the NRC Staff of the analyses provided by each reactor vendor in support of their positions to accommodate an ATWS event.
The first two volumes of the report were issued in April, 1978, and reviewed, among others, the GE ATWS response provided in the 1976 General Electric ATWS Report.
These two volumes evaluated the information provided by all vendors, and came to the general conclusion that ATWS events presented and unacceptably high risk to the public.
Corrective measures, including the installation of new systems or modifications of existing systems, were seen as necessary to reduce the risk of severe consequences arising from possible ATWS events.
Volume 3 to NUREG-0460, published in December, 1978, re-examined the earlier approach to resolving the ATWS problem and recommended more specific ATWS prevention and mitigation modifications as well as certain generic safety analyses.
Generic questions and information guidelines were provided to all NSSS vendors in a letter dated February 15, 1979, from R. Mattson, NRC.
- However, the Three Mile Island accident in March, 1979, resulted in diversion of NRC and industry resources from the ATWS issue, with the result that the industry's response to the questions posed in the Mattson letter fell far short of what was expected.
Volume 4 of NUREG-0460, published in March, 1980, presents a criticism of the NSSS vendor responses to the Mattson letter, extends the requirements for ATWS prevention and mitigation previously stated in NUREG-0460, Volume 3, and contains the proposed resolution to the unresolved safety issue (TAP A-9).
The body of NUREG-0460, Vol. 4, is included as Attachment B to this report; the NRC Staff evaluation of the GE response to the Mattson letter is provided as Attachment C.
The areas judged by the NRC staff to be inadequately addressed by GE in its response to the Mattson letter are:
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Most BhW designs, but BWR/4 in particular, are calculated by GE to produce severe power/flow oscillations several hundred seconds following a turbine trip ATWS event.
Oscillations have been observed in an opera ting BWR.
The staff is concerned with the prediction capability of the codes, the impact on plant control systems, and the fuel integrity after the core undergoes these oscillations.
-~~-
The reports do not address the ability of equipment needed for safe shutdown to withstand the ATWS transients, including the oscillations.
Insufficient information on the reactor coolant pressure boundary components integrity/operability is provided.
ATJt'S containment loads are not shown to be bounded by design basis loads.
Many questions remain on radiological evaluation if containment is not isolated early in the ATWS event.
Design information on prevention and mitigation systems has been inadequately addressed.
The above concerns notwithstanding, the current information provides some degree of assurance that the consequences from ATWS events in BWRs are likely to be acceptable if the mitigating systems function as presumed and the oscillations do not cause unforeseen damage.
(Ref. 6, Vol. 4, pp.8 - 9)
The recommendations of NUREG-0460, Vol. 4 (See Attachment B) for GE plants, specifically applicable to Perry based on the date of issuance of its construction permit, are:
Alternate rod insertion Recirculation pump trip Automatic, high capacity (300-400 gpm) poison injection system Changes in logic to reduce vessel isolations and p~rmit feedwater runback Modification of the scram discharge volume Analyses to verify adequacy of the modifications Provisions for quick closure of containment isolation valves if fuel failure occurs.
If a plant cannot conform to these requirements, a study of alternative modifications (optimization study) is to be submitted to the NRC, to show how the plant will achieve a level of safety equivalent to that represented by the above recommendations (Attachment B, p. 29).
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4.2 NRC Comments on Reed Issue 25 The 1987 NRC review of the Reed Report states the present status of this issue as follows:
Such pressures [1600-1650 psig] resulting from a transient event could occur only at elevated temperatures when the pressure vessel material is in a ductile state and is thus less subject to damage by an overpressure event.
- Further, more refined calculations by GE using better analytical methods demonstrate that peak pressures in such an event would be far less than the 1600 to 1650 psig estimated in 1975.
Interim resolution of the ATWS issue was provided by improved procedures and operator training, and through implementation of certain hardware modifications (e.g., recirculation pump trip).
The ATWS issue was finally resolved when NRG issued the ATWS rule (Title 10 of the Code of Federal Regulations, Section 50.62 (10 CFR 50.62), in July, 1984.
In response to this rule, plant-specific measures, including hardware modifications will be made in some plants.
In October, 1986.
the NRG accepted the GE licensing topical report NEDE-31096-P, "Anticipated Transients Without Scram; Response to NRG ATWS Rule. 10 CFR 50. 62, 11 which means that licensees may now reference this report in their plant-specific actions.
(Ref. 7, p. 24) 5 Plant/General Electric Company Position The description of the Perry Plant I s ATWS prevention and mitigation systems is provided in the Plant's USAR (Ref. 8), which is included as Attachment D to this report.
The presentation begins with the statement that a generic assessment of BWR ATWS mitigation is presented in the GE Topical Report "Assessment of BWR Mitigation of ATWS," NEDE-24222 (December, 1979).
This report was included in the NRG review contained in Attachment C to this report.
Perry incorporates the following features designed specifically to mitigate an ATWS event (Ref. 8, pp. 15C-4 to 15C-7:
See Attachment D. ).
Redundant Reactivity Control System (RRCS). ATWS detection sensors ( four reactor pressure vessel dome high pressure sensors and four low vessel level sensors) and actuation logic to automatically initiate alternate rod insertion, recirculation pump trip, and feedwater runback.
These sensors provide indication and alarm in the control room.
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Alternate Rod Insertion System (ARI).
This provides an electrically diverse scram logic, independent of the reactor protection system, to vent the scram discharge air header through valves separate from the reactor protection system scram valves.
It consists of redundant scram pilot air header exhaust valves which are actuated by the ATWS detection sensors.
Recirculation Pump Trip (RPT).
The recirculation pump motors are tripped by ATWS signals from the RRCS logic.
Feedwater Runback.
Runback of feedwater flow occurs if a high pressure signal and confirmed failure to scram is indicated by the RRCS logic.
Standby Liquid Control System (SLCS).
The standby liquid control system is capable of delivering a minimum flow rate of 86 gpm of 13 weight percent sodium pem:aborate solution to the reactor vessel through the high pressure core spray piping.
The system is manua'11y actuated from the control room.
Scram Discharge Volume Modifications.
The scram discharge volume has been modified to minimize the probability of the failure to scram due to unavailability of this volume.
The results of the plant I s ATWS analyses are provided in the same section of the USAR and summarized in Table 15C-8.
(See p. 15C-22 of Attachment D.)
For all the events analyzed, the USAR assumes that the ARI function does not operate and that the standby liquid control system is manually initiated 120 seconds following control room indication of an ATWS event.
As shown in Table 15C-8, the MSIV closure ATWS produces the highest vessel pressure (1304 psig} and the highest averag~ heat flux (151% NBR).
It should be noted that the design pressure of the reactor vessel is 1250 psig.
The pressure safety limit is 10% above this value, i.e., 1375 psig.
The 1375 psig value-is considered as the acceptable maximum pressure limit for the reactor vessel during transients (Ref. 20, US.C-78).
The conclusion presented in the USAR based on this analysis is that Perry can withstand the consequences of an ATWS event (Attachment D, p. 15C-10).
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6 NRC Evaluation 6.1 General ATWS Issues Section 7.2.2.3 of the Perry SER (Ref. 9, pp. 7-5 to 7-6) discusses the Perry ATWS recirculation pump trip design, approving it based on its similarity to that of Grand Gulf and Hatch, which were previously reviewed and accepted by the NRC staff.
As a confirmatory issue, the Plant was required to provide a detailed description of its ATWS recirculation pump trip (Confirmatory Issue 26).
This was satisfactorily resolved in Supplement 4 to the SER (Ref. 10, p. 7-2).
Because of the significance of the ATWS recirculation pump trip to the mitigation of an ATWS event, the discussion presented in SSER 4 is included with this report as Attachment E.
Section 7.2.2.4 of the SER discusses Perry's modification of the scram discharge instrument volume level monitoring system to assure effective insertion of the control rods during a scram, based on the Brown's Ferry event.
Revision of the FSAR to provide a complete description of this modification was required as Confirmatory Issue
- 27.
This was resolved in Supplement 2 to the SER (Ref. 11, p. 7-1).
Supplement No. 1 to the SER addresses the standby liquid control system changes intended by the plant, leading to Outstanding Issue 20.
This Issue was resolved in Supplement 3 to the SER (Ref. 12).
Because of the significance of the standby liquid control system to the mitigation of an ATWS event, the discussion presented in SSER 3 is included with this report as Attachment F.
Section 15.6 of the SER explicitly addresses ATWS. Therein it is stated that Perry will be subject to any rulemaking by the NRC that may result from the NUREG-0460 reviews.
The only rulemaking to which the plant has had to comply is that contained in 10 CFR 50. 62.
The SER required that an emergency procedure be developed for the ATWS e*vent and provided specific guidance on the content of the procedure.
The discussion concludes:
Early operator action as described above, in conjunction with the recirculation pump trip, would provide significant protection for some ATWS events, namely those which occur (1) as a result of common mode failure in the electrical portion of the scram system and some portions of the drive system.
and (2) at lower power levels where the existing standby liquid control system capability is sufficient to limit the pool temperature rise to an acceptable level.
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The GE Owners Group is currently developing a set of reactivity control guidelines, which will incorporate the above steps for mitigating ATWS events.
The applicant's proce~ure program for mitigating ATWS will be reviewed under the emergency operating procedure program as described in Section 13.5.2.
The result of the staff review will be reported in a supplement to this report.
(Ref. 9, p. 15-18)
The review of the emergency operating procedure program is documented in Supplement 8 to the SER (Ref. 13, pp. 13-2 to 13-7).
The issue addressed is the long term upgrading of the plant's emergency operating procedures in accordance with TM! Action Plan Item I. C. 9.
The review of the Perry emergency operating procedure development and implementation plan concluded that the plan met the existing requirements (Ref. 13, p. 13-7).
6.2 NRC Review of the Perry Response to the Salem ATWS Event On two occasions (February 22 and 25, 1983), Salem Unit 1 failed to scram automatically due to failure of both reactor trip breakers to open on receipt of an actuation signal.
In both cases the unit was successfully tripped by manual action.
The failure of the breakers has been identified as excessive wear due to improper maintenance of the undervol tage relays which receive the *trip signal from the protection system and cause mechanical action to open the breakers.
Generic Letter 83-28, "Required Actions Based on Generic Implications of Salem ATWS Events," was issued to all utilities on July 8, 1983.
Perry's response to this issue is identified in Supplement 7 to the SER.
In the letter from W. Butler, NRC,
-co M. Edelman, CE!, dated October 10, 1986, Perry's response to specific action items was approved.
The details of the approval are provided in Supplement 10 to the SER (Ref. 15, pp. 7-2 to 7-5).
There were still specific items under review and it is stated in SSER 10 that the results of this review will be documented in a future supplement to the SER.
However, since an operating license has been issued, no further SER, supplements will be issued.
The only further documentation we have regarding this issue is an onsite inspection report for the period May 23 through June 10, 1988 of Perry's implementation of Generic Letter 83-28, stating that no deviations were identified.
The items relative to this issue, with the exception of the Generic Letter 83-28 itself,
~'
are contained in Attachment G to this report.
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7 The LaSalle Event The March 8, 1988, transient at LaSalle Unit 2 resulted in an operational state of the reactor very similar to what would occur following a recirculation pump trip in an ATWS condition.
The NRC I s Office for the Analysis and Evaluation of Operational Data (AEOD) report on the LaSalle event recommends that ATWS mitigation be reviewed in light of the power oscillations that the plant experienced following the trip of the two recirculation pumps (Ref. 16). In the words of the evaluator:
Previous efforts taken in regard to ATWS mitigation may be inadequate.
The action of tripping recirculation pumps automatically and inducing an event similar to the LaSalle event when it 1s not where the power oscillations would stop and what the effects of these oscillations would be in the absence of an automatic scram. necessitates that ATft'S mitigation be reviewed in light of this event.
(Ref. 16, p. 7 )
In the NRC Staff review of the GE response to the February, 1979, Mattson letter regarding ATWS, the staff expresses concern for issues directly related to instabilities (See Attachment C.).
In the context of GE's ATWS analysis, the following concerns are expressed by the staff:
- 1)
In terms of long term behavior of the plant following an ATWS event, there are some concerns about limit cycle oscillations (Attachment C, p. A-61).
- 2)
The REDY code calculated that there are limit cycle oscillations on an average core basis.
Translation of the co.re average cycles into those cycles occurring in individual fuel channels is very difficult.
REDY cannot be used to predict fuel failures resulting from local oscillations (Attachment C, p. A-63).
- 3)
- 4)
Several events are shown to have significant, periodic oscillations in neutron flux following an initial, large neutron flux spike.
The effect this has on PCI failures has never been considered (Attachment C, p. A-87).
Oscillations in neutron flux and fluid flow that occur later in the event could produce some unique mechanical loading effects that have not been modeled (Attachment C, p. A-88).
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- 5)
The combined effects of high neutron flux spikes, resulting in high cladding temperatures and boiling, followed by oscillations in flux, fluid flow, etc., raise questions not only about the fuel future, but also about the potential for loss of coolable geometry (Attachment C, p. A-89).
8 Letters Received by PUCO from Hiatt and Thorpe Nuclear Education & Training Services, Inc., has been asked to evaluate the concerns expressed in letters (Refs. 17, 18, 19) to the PUCO from Ms. Susan L. Hiatt of the Ohio Citizens for Responsible Energy and Mr. Sheldon Thorpe of the Northern Ohio Section of the American Nuclear Society.
Mr. Thorpe's letter did not contain comments related to Reactor Study Issue 25.
The instability question is raised in the OCRE analysis of the safety implications of the Reed Report, dated August, 1987, transmitted with Ms. Hiatt's first letter.
Her second letter transmitted information
~regarding the LaSalle event.
Based on the LaSalle event, the issue of iinstability in the BWR is certainly significant, especially when
~coupled with the ATWS recirculation pump trip mitigation feature.
Regarding Perry not installing an automated standby liquid control system, which question was specifically raised in the OCRE review, it is obvious that Perry had a strong preference to maintain manual control of the SLCS, leaving the decision to inject up to the control room operators.
According to 10 CFR 50.62, they were able to do this.
As is clearly stated in NUREG-0460, the NRC Staff reviewing the ATWS analyses also perceived automatic actuation of the standby liquid control system to be a valuable feature.
We, however, do not know of any plants that have installed it.
From a plant operations viewpoint, it is not desirable because the possibility of unintentional injection would require extended shutdown for cleanup. of the reactor vessel.
Further, injection during a LOCA event when ECCS water is being supplied to the core would dilute the boron such that its effect would be minimized.
It would seem to us to be sufficient to have manual initiation capability with well trained operators clearly directed by procedure to use the system when required.
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9 NETS Evaluation of the Issue Based on the information at our disposal, there is a large ~ap between the *ATWS prevention and mitigation recommendations stated in NUREG-0460. ~nd the ATWS Rule stated in 10 CFR 50.62.
In light of the recent LaSalle event, the consequences of the recirculation _pump trip ATWS mitigation feature do need to be reviewed,* *and the concerns expressed by the NRC Staff in - Vol. 4 of NUREG-c:Q460 *a-Isa- *need *to---be -looked at again.
At the time this draft is being written, we are requesting further information from Perry regarding -the reasons for* selection of,the manua'.l"tTather<Wrthan-"autoraatic µ-initiation **of
- s*ysYe111.'"- This is viewed as a relatively insignificant matter as compared with the possible effects due to limit cycie oscillations occurring during an ATWS event.
Perry meets the current NRC requirements mitigation.
The issue is closed in this sense.
the instabilities experienced by LaSalle Unit 2 the recirculation pump trip ATWS mitigation evaluated.
pertaining to ATWS However, in light of the consequences of feature need to be NETS, Inc.
Report on Reactor Study Issue 25 Page 16
References
- 1.
"12 Years Later... An Update Report on the Nuclear Reactor Study," Vol. 1 and 2, (GE, July, 1987).
- 2.
Letter dated March 22, 1978, from Glenn G. Sherwood (GE) to Dr.
Roger J. Mattson (NRC).
- 3.
Lette*r dated May 26, 1978, from Glenn G. Sherwood (GE) to Dr.
Roger J. Mattson (NRC).
- 4.
Title 10, Chapter 1, Code of Federal Regulations, Part 50 (USNRC, December 31, 1987).
- 5.
NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition (June, 1987).
- 6.
NUREG-0460, Anticipated Transients Without Scram for Light Water Reactors, USNRC: Vol. 1-2 (April, 1978); Vol. 3 (December, 1978);
Vol. 4 (March, 1980).
- 7.
NUREG-1285, NRC Staff Evaluation of the General Electric Company Nuclear Reactor Study ("Reed Report") (July, 1987).
- 8.
Perry Nuclear Power Plant, Updated Safety Analysis Report for Unit 1.
- 9.
NUREG-0887, Safety Evaluation Report Related to the Operation of Perry Nuclear Power Plant, Units 1 and 2, Docket Nos. 50-440 and 50-441 (May, 1982).
- 10.
NUREG-0887, Supplement No. 4 (February, 1984).
i'l.
NUREG-0887, Supplement No. 2 (January, 1983).
- 12.
NUREG-0887, Supplement No. 3 (April, 1983).
- 13.
NUREG-0887, Supplement No. 8 (January, 1986).
- 14.
NUREG-0887, Supplement No. 7 (November, 1985).
- 15.
NUREG-0887, Supplement No. 10 (September, 1986).
- 16.
"AEOD Concerns Regarding the Power Oscillation Event at LaSalle 2 (BWR-5)," AEOD/S803 (June 7, 1988).
- 17.
Letter dated June 14,
- 1988, from Susan L.
- Hiatt, OCRE Representative, Ohio Citizens for Responsible Energy, to Mr.
Andrew Grandjean, Chief, Nuclear and Gas Pipeline Safety Division, Public Utilities Commission of Ohio, transmitting "OCRE, Inc., Analysis of the Safety and Regulatory Implications of the Reed Report," dated August, 1987.
NETS, Inc.
Report on Reactor Study Issue 25 Page 17
- 18.
Letter dated June 24, 1988, from Sheldon Thorpe, American Nuclear Society, Northern Ohio Section, to Mr. Andrew Grandjean, Public Utilities Commission of Ohio, transmitting recommendations regarding review of the Reed Report, statements related to radioactive material releases, and statements regarding containment performance.
- 19.
Letter dated June 14,
- 1988, from Susan L.
- Hiatt, OCRE Representative, Ohio Citizens for Responsible Energy, to Mr.
Andrew Grandjean, Chief, Nuclear and Gas Pipeline Safety Division, Public Utilities Commission of Ohio, transmitting NRC Information Notice 88-39 and NRC Bulletin No. 88-07.
- 20.
NRC Safety Evaluation Approving the General Electric Topical Report, Generic Reload Fuel Application (NEDE-24011-P) provided as pp. US. C-5 to US. C-89 of the General Electric Standard Application for Reactor Fuel (Supplement for the United States),
NED0-24011-A-8-US (May, 1986).
NETS, Inc.
Report on Reactor Study Issue 25 Page 18
Appendices A. Section 15.8 of NUREG-0800 B. Main Tex-t from NUREG-0460, Volume 4 C. BWR Section of Appendix A of NUREG-0460 D. Appendix 15C from Chapter 15, Perry USAR E. Section 7.2.2.3 of SSER 4 F, Section 9.3.3 of SSER 3 G, Documentation Pertaining to Perry's Response to the Salem ATWS Event NETS. Inc.
DOCKET NUMBER
. ETITION RULE PRM S 0-J 5 3 (S'-IFR3o105)
L'OCr,[ H.Q U,Nr
'89 JUL 19 P6 :21
- 1F ;
- 1 tcc*1
- NUCLEAR REGULATORY COMMISSION' 10 CFR PART 50
[Docket No. PRM-50-53]
The Ohio Citizens for Responsible Energy; Receipt of Petition for Rulemaking AGENCY:
Nuclear Regulatory Commission.
ACTION:
Petition for ru lemaking:
Notice of receipt.
[7590-01]
SUMMARY
The Nuclear Regulatory Conmission (NRC) is publishing for public conment a notice of receipt of a petition for rulemaking dated May 26, 1989, which was filed with the Commission by the Ohio Citizens for Responsible Energy (OCRE).
The petition was docketed by the Commission on May 26, 1989, and has been assigned Docket No. PRM-50-53.
The petitioner requests that the NRC reopen the ATWS rulemaking proceeding.
This request was one portion. of a request by OCRE that NRC take a number of actions to relieve alleged undue risks posed by the thermal-hydraulic instability of boiling water reactors (BWRs).
DATE:
Submit comments by September 25, 1989.
Comments received after this date will be considered if it is practical to do so, but the Commission is able to assure consideration only for comments received on or before this date.
1 Pu J /,sleJ I" tlr
/; Jt.-o.l R.eJ *,s-/-er On
?-'J..S-8Cf
'11; fiiUCLEAR REGULATORY COMMISSIOl'o 0OO'H 1NG 0, 5r.1v: CE ~.~CT ION t) ~:
- I t -:s Postm~rk Da* _ N JJ Copie5 ::-:.2, e,_*
)
Adi C::
Special Di. r*k, --1
ADDRESSES:
Submit written comments to the Secretary of the Cofllllission, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, Attention: Docketing and Service Branch.
For a copy of the petition, write the Regulatory Publications Branch, Division of Freedom of Information and Publications Services, Office of Administration, U.S. Nuclear Regulatory Coflllliss1on, Washington, DC 20555.
The petition and copies of comments received may be inspected and copiea for a fee at the NRC Public Document Room, 2120 L Street NW, Lower Level, Washington, DC.
FOR FURTHER INFORMATION CONTACT:
Michael T. Lesar, Acting Chief, Rules Review Section, Regulatory Publications Branch, Division of Freedom of Information and Publications Services, Offjce of Administration, U.S. Nuclear Regulatory Con111ission, Washington, DC 20555, Telephone:
301-492-7758 or Toll Free:
800-368-5642.
SUPPLEMENTARY INFORMATION:
On July 22, 1988, the Ohio Citizens for Responsible Energy (OCRE) filed a request for action under 10 CFR 2.206 with the Nuclear Regulatory Co11111ission (NRC).
This petition was referred to the Director, Office of Nuclear Reactor Regulation (NRR) for consideration.
Under the provisions of 10 CFR 2.206, a person may request that the NRC institute a proceeding to modify, suspend, or revoke a license or take any other action that may be proper under 10 CFR 2.202. Section 2.202 authorizes the Executive Director for Operations and the Office Directors with program authority to take this type of action by serving an order to show cause on the licensee. This type of proceeding, generally
referred to as a petition for a director's decision under 10 CFR 2.206, involves licensing actions and pertains to a specific facility or class of facilities.
OCRE requested that the Director, NRR, take immediate action to relieve alleged undue risks to the public health and safety posed by the thermal-hydraulic instability of boiling water reactors (BWRs} revealed by the power oscillation event at LaSalle Unit 2 on March 9, 1988 (LaSalJe*Event). The petitioner specifically requested the NRC to order all BWR licensees to (1) place their reactors in cold shutdown, (2) develop and implement specified operating procedures relating to the thermal-hydraulic instability issues, (3) demonstrate that certain specified training has been provided relating to these procedures, (4) demonstrate the capability of instrumentation related to power os*cillations, (5) develop simulators capable of modeling power oscillations similar to those occurring at LaSalle as well as out-of-phase power oscillations, (6) report to the NRC all past and future incidents in whtch recirculation pumps have tripped off, or in which power oscillations were
- e involved, and {7) submit to the NRC justification for continued operation of BWRs.
In addition, the petitioner requested that the NRC reopen Generic Issues B-19 and B-59, and also the Anticipated Transients Without Scram {ATWS) rulemaking proceeding.
The petitioner sugge~ted that resolution of the ATWS problem depends on measures other than tripping the recirculation pumps to rapidly reduce reactivity. In this regard, the petitioner specifically suggests the use of an automatic, high-capacity standby liquid control system.
Finally, the petitioner suggests that the NRC reconsider the use of the End-of-Cycle Recirculation Pump Trip on BWRs.
The Director, NRR, acknowledged receipt of the request for action in a letter to the petitioner dated August 26, 1988.
The Director informed the 3
petitioner that (1) the request for immediate relief was denied because the allegations that formed the basis for the petition did not reveal any new operational safety issues that posed an immediate safety concern for continued BWR operation, (2) the petition would be treated under 10 CFR 2.206 of the Commission's regulations, and (3) appropriate action would be taken within a reasonable amount of time.
On April 27, 1989, the Director, NRR, responded to the OCRE request for action in a Director's Decision under 10 CFR 2.206.
The NRC denied the petitioner's requests for action pursuant to 10 CFR 2.206 because the issuance of proceedings under 10 CFR 2.202 is appropriate only where substantial health and safety issues have been raised.
The NRC determined that the petitioner's request did not raise any substantial health and safety issue.
In the Director's Decision (DD-89-03) the NRC denied all of the petitioner's requests, except for the request to reopen the ATWS rulemaking proceeding, which would be more properly treated as a petition for rulemaking under 10 CFR 2.802.
A copy of the Director's Decision (DD-89-03) was filed with the Secretary of the Commission for the Co1T111ission 1s review as directed by 10 CFR 2.206 and notice of this action was published in the Federal Register on May 10, 1989 (54 FR 20218).
Under 10 CFR 2.802, an interested person may petition the Commission to issue, amend, or rescind any regulation.
Because OCRE's request to reopen the ATWS proceeding could ultimately result in an amendment to the regulations in 10 CFR Part 50, the NRC is treating this portion of OCRE's request separately as a petition for rulemaking under 10 CFR 2.802.
The ATWS portion of OCRE's request was docketed as a petition for rulemaking under 10 CFR 2.802 on May 26, 1989 (Docket No. PRM-50-53).
4
As background information, the NRC publishea a final rule on June 26, 1984 (49 FR 26036), that was intended to reduce the risks of an ATWS event for a light-water-cooled nuclear power plant.
An ATWS event takes place if an abnormal operating condition (anticipated transient) occurs at a nuclear power plant which could cause the reactor protection system to initiate a rapid shutdown (scram) of the reactor but the reactor shutdown system fails to function.
The final rule, codified in 10 CFR 50.62, required the installation of certain equipment in nuclear power plants and encouragea the development of a reliability as5urance program for the reactor trip system.
The NRC is currently performing confirmatory analyses of ATWS events.
If evidence from these evaluations, or colTiTlents received as a result of this notice, show contradictions of assumptions and results of previous ATWS analyses, it may be appropriate for the CoD111ission to reconsider the current ATWS rule.
Dated at Rockville, Maryland, thf s /'l ': day of J,J7 1989.
or the Nuclear Regulatory Co1T1T1ission.
Secretary of the Commission.
5
. ~:,
- .Jr.... - *........ *---***-----**---
Before l'-L i:.
- as; M,W 26 A 8 :53
- a9 MAY 30 P12 :20 ONITE00J~ATES or_ AMERICA NUCLEAR lEQ-W~TOkY 00MMlSSlON 3!{t, NL*:
Director c fie* cf uelear
'f'.
JUL 2 e 1988 ~..
& P. f. IT.
BOSTON !01S N co. (P119rim Nuclear Power lt&t1on, Docket No.
10*213)
CAROLINA,o RI LIGHT co. (lrun,wick Station, Dn1tl 1 an4 2, Docket Noa. 50*324 and 50*325)
CLEVELAND E !CTRIC %tiVM1NATlNG COMPANY, Zt AL. (ferrr Nuclear lower flantf Unit 1, Docket Ho. 50*440)
COMMONW!ALT' !01SON co. (Dr11dtn Nuclear tower tlant, Unit* 2 and 3, Cock t No,. !0*237 and 50*24tJ, (Quad Cit111 Muclear tower Plant Unit* 1 and 2, Cocket Noa.
50*254 -and 50*215>,
(LaSalle Cc nty ltation, Onita 1 and 2, Cocket Noa. 50*373 and 50*3'74) f.-.
/ r*
CONSUM!lS P WER CO. (819 lock toim, Docket *110. s*0*155)
DETlOit ZDI ON co. (Fermi Onit Z, Docket Ho. IO*J41 GEORGIA PCWtR co. (Hatch Nuclear Power tlant, Onitl 1 and 2, 0ocket No. f0*321 an4 SO*JIIJ GULF STATES,DT%~1T%1S co.
(liver lend ltation, Oocket No.
50*451>
JLLtNOII to l co. (Clinton *~clear tover t1ant, Oocktt No.
10-,11, XOWA EL!CTR C LlOHT I,owia co.
ccuane Arnold Nuc1aar Pover tlant, Dock t No. SO*JJ1)
CENERAL t~! lC OtlLlTlll (Oyattr Creek ltatlon, Docket Ho.
10*21t) 1,0NG JSLAND LIGHTING co. (Shdrahaa Nuclear Power tlant, Docket No. 10*J22)
M?SSlSSlPPl *POWER I LIGHT CO.
(Grand Gulf Nuclear Station, Docket No. o-,i*>
NEBRASKA,u LlC POWER DllTllCT (Cooper Station, oocket No.
50*291)
NIAGARA MOH wx POWEi coi,. (Nin* N11* toint, Onit 1 and 2, co=k*t NOi, 50*220,nd 50*410).
, ~,....., rr-
~1s ION r
I i.11,..._ _ _ _
WORTHtAST O 1L%%1ES (M1111ton, Unit 1, Oooklt No. 50*245)
- oaTHZM IT TIS POWER co.
(Mont1o*11o Huclear tower,1ant, Docket *o*
0*213)
PEHNSY1.VAN1 fOWER I LIOHT co.
(lu quehanna lt*am 11ectr1c Station, on t1 1 and 2, oocket Mo. IO*Jl7 and SO*JII>
PH%U01tPH%
EL!CTR%C co. (Peach lottom Nuclear ltation, Onitl 2 and 3, Co ket Noa. 10-2,, and so-2,1,,
(Limerick *~clear Power tlant gn1t 1, Docket Ko. 50*352J POWER AUTHO %TY or THI ITA!I or nw YORI,~--* a. fit1pau-1ck lt&tion, 00 k*~ No. SO*JJJ)
PUILtc*11~v CE ZLJ:CTIIC I Gll co.
(Hope Creek Genaratin; Station, oo k*t Ko. 50*35')
~ENN!SS!E V LL!Y AOTHOklTY (lrown* Perry N~elear ltation, Onit 1, 2, and J Docket No1. 10*251, so-2,0, and 10-2,,1 VZMONT YAN El NUCLIAk POWll c01,.
CVermont rank** Nuclear Power Jlant Docket No. 10*271)
WASHINGTON Ul~1C POWER IOPPLY IYSTIM (WNP Unit 1, Docket No.
50*317)
PETI 10N nNDER 10 era 2.201 UQUIITINC Wl\\C ACTION iD UV IVZ OND0E RISK POSED IY BOILING WATER JlEACTOR THERKAL*HYDRAULIC %NSTAl1LlTY Pur uant to 10 era 2.201, the Ohio, Cit11an for ae1pon1ible 1ner91, %no.,*ocu*J herbf petition,*
th*
~1rector, fflc* of Nuolear leaetor le;ulat1on, to take immadiat*
- tion to ~*11*v* u.ndYe ri1k1*to the public health and* aafety poee4 bf -the thermal*hydrau11c inat&~lll~f of bo111n; vat r reactor,,** r1ve1lad bf th* power 01=111at1on event at ta all* an1t 2 on Haroh 1, 1111
- J%.
>>ESCIIP 10N or PET%T%0NER Pet1t1 ner OCIE 11 a private, nonprofi~ corporation
- orvan11*d u der ~* law, of the state of Chio.
ocu apacia111** n r***areh and advocacy on i11ue1 of nuolear reactor **f ty and haa ** it1 9oa1 t.he promotion and e
application of th* h~9he1t atandard1 of afety to
- uob 1ac11iti***
roa ll!tlEr A.
litnific nee of tht.Lala11**2 lvent on Marc,, 1111, while operatin; at 1,1 power, ~*
i.asa11* Unit 2 reactor underwent a dual recirculation pump t.r1p on* fa~** 1 w level uan1i1nt 11,nal cauaed by* technician~*
valvin; erro ~n t.he r1actor pr111ur* ve1111 level **nain; 11n*** a1,u1 ant natural c1rcu1at1on operation at the u.ncban;ed
- ttl flow can rol line yielded powr 01c111ation, which want on for aeveral in~t** until t.ht' reactor automatically crammed on APJUC h19b f1 11111). It i~ e1timate4 that 1oca1 n*~tron flux variation, & 1ar9* ** JODI ooourrtd <*** Attachment 1,.
Tb*
plant* opera or; rather than craMing~ the ~**ot.or, ware inatead tryi 9 co re tart ch* recirculation pumpa.
Had tbtf aucca*de4 in r* tartlng th* pump, 1t i* likely that t.b,e even~
would have vor**n*~ due to the in11rt1on of additional
~*activity, Th* *vent ii deaer1bed further in Attachment 1, *n article *PP* rin; in th* J"lY 4, 1t88 i u* of Juldt N1e,, Information Notice 18-Jt, Attachment I, lullatln 11*07, an4 A tachment 4, June I, 1918 Memorandum for Thoma, 1.
Nuley and ltck~ord, Cffioe1 of Muclear aaactor
~*;ulation d ****arch, from Zdvard L. ~ordan, Cffio1 for Analy111 and Evaluation of Operational Data, and t.he eftoloas4 I
uo, Sp*eialj Report, A!OC/1103.
Attachint1 1 throu9h
- cl*arly indicate that the t.&1&111 l
ever.t ha** icua 1afety implication, for &11 bo11int water
. I
%eactor1. Jwtr o cillation ver* not prtdieted to 'occur at usa11a, th, c&1cu1ated decay ratio for La.8&11* wa1 0.1.
in fact, *~S*l,* W&I con1id1rad to bl on* of the llOlt table BWRI in th*
CCU try (Attachment 1).
Other IWR, v1th h19h*r c:aloulate4 ratio, &re even more *~*ceptib1* to power cac111at1on when operated in un1tal:)le re;ion
- ror example, th* calcula ed deoay zat1o tor th* Perrr Muclaar tower 11ant 11 o.,, (Upd&t d &af*ty An&lflil ieport, P* 4.,-13). th* NRC baa concluded.
hat the decay. ratio dat*rmined by
.tic1n1ln9 ea1culation ia not a ~*liable in41cator of core atu11ity.
the ttachment 1tat1 that there 11 a ~u11t1on o!
compliance it.h aener&1 01,19n Criterion 12 of Appen41x A to 10 era iart 10 GDC 12 require* t.ha,. the reac:tor eor* and a11ociated oolant, control, and pro\\eotion 1y1t1m1 be d111tntd.
- o a, to*** ure that ?OWtr 01c111at1on vh1ch* can %IIU1t in a
condition, cttdLnt epecified acceptable fuel deai;n 1Lmit
- are not po* ibl* or can b* reliably detacte4 and 1uppr*11ad.
~b* LaSalle vent clearly demo~1trate1 that the deaitn and an*1Y*** of
- reactor cannot be re11*4 upon to en1\\ll"e that oac111ation are not po111b1e. It a110 1how14 dtf1o1encie in operator t.r ining to detect and *~ppr**~ 01cL11atlona. General Zl*ctric'* uidanoe 1n ll~ JIO aevi11on 1 ii inadequate to
- ensure com liance with GDC 12 (lttacbent *>*
,1ant intrumenta ion **Y not detect power 01cllla~iou.
UkK 1Ji1trument* typ1;a11y ahov 1ar9* nol** fluctuat1ona.
Sf out*of*pha1 01cillation1 occurre4, the avera;* 119na1 ** tbt 1 A1JUC trip f nction would b* below the trip level, even tbou9h local 01011 ation* could be 1ar;e enou1h to oau11 fut1 d&uf**
~h* LaSalle 01cillation1 var* a110 too fa1t to ~* ~ocurate1y mea ured i, in1u-utD*ntation availel* to th**
opt1"&tor1.
(Attachment 1 and*>*
th* Al D Z'ICOIU\\lftdl that in t.he interim, IWl liotl\\1111 ahould bet quired to develop proc*d~r** to avoid un1te1a operation C ttaohfflant 4, UOD Special a.port at 1-1,.
1100 ndl that the nc re*evaluat* the sa o1utlon o!
Gentrio 111 ** 1*1t, therma1,Hfdraulio lt&~11ity, and 1*11, H*1 Loop Cpeziat on in an. ant tn1, &nd tJ\\t ATWS 11iti9at!.on
- ure of trippin9 tb* reclrou1at1on pump, C*** 10 era 10.1210) cs, *
- The NRC Hae railed to Take Appropriate lt9µ1atory Action I
ha* thorou;hly documented the 11r1ou1 imp11cat19n of the LaSalle event and b&I mad*
om* 1pecific to th* Office ct NM and UI.
_Unfortumat1ly, it appear, at u00 11 advio1 ha1 fallen on deaf ear1.
l.&1&11* evt i11ued *~11et1n 11*07 on ~wi* is, 1111.
After the 119n1f1cant 91n1ric 1mp11cat1on1 of t.b*
the HRC outline, thr** *r1qu11ted* action* for an 11c1n1t I and announc11 it1 intent to puraue the ~euinln; 111u11 with th* an own1r1 Group.
the r1qu11tad action, include *n* rin; that 1icen11d op*rator1 and lhift techn1oa1 Advi1or1 ar
- thoroughly briefed* re91rdin; the tala11* event y1 of receipt of the bulletin, and, vith1n 10 day1, v1rifyin9 t
- adequ~cy cf procedure, and uainin1 pro9rama to
- la* awar power o cil Ca> plant condition* which can r1ault in (b) action* which can,, ~aken to avoid auch plant ond1t1on1, (c) bow to reco;ni1a th* on11t of power c c111ation, and (d) aotlon1 which can be taken in re1pon11 to power 01011
~* adequac ct the intrumentatlon which 11 relia4 upon ~f th*
operator
- th* lioen**** thin have an additional 30 day1 following
- completion of tsh*** action* to confirm by letter
~at they h ve b**n comp1ete4. fh* evaluation or the adequacy of p;oc1dur,, tra1n1n; pro;r~ma, and lnl\\rumentation 11 not ~o 1,e it to ~* NRC, nor &&de ava11ab1* ~o th* p~1lo, lnat to be k*~t on !11* at th* plant aita, for a p1rlo4 a
of at 1aa1t two year* !h* apecific recoanmend1tion1 111.&d* bf
~ECD ar* no required or even **ntionad.
Givan h* vr*v* insp11eation1 of the LaSalle event, **
thoroughly ocumented.in th* UOO report, th* action requ**t*d in th* Bull tin are in1uffic1ent. the *ac muat take a9;r1111va enforcement action to *n1ure the bea1t.h and aaf1ty of th*
public. ** au** it £1 probable that moat, if not &11, IWR are tn a 1tate of noncompliance v1t.h ace 12, iha 01reotor, n1 abould.imme 1ata1y take t.h* action* outlined below.
- a. ~h* Dir,=tor, NRR, 1hou1d i11u1 an order requirin9 all General Elecyric an lic*n**** to place t.hir reactor, in cold abutdown uz;t11 perm111lon to reatart ia 9iv1n, baed on completion a!ncS implementation of actJon11, C, and D below.
- ~h* Dire tor, KU., ahou14
£11u1 an order requir1nt all General Zl* tria awa 110101111 io d*v*lop and implement I
procedure, to CaJ 11Mltd1ate1r in11rt control rode,o b11ov the IOI rod 11nJ fo11owin9 redQ~\\ion or 1011 of r*circu1ation' flow or other tra 1ient1 Which r11uli in entry into potent1&11f 11n1t1~1* r* ion of ~* pGwer/f1ov *map, Cbl incr****
%ecir~u1atio tiov durin9 routine reaotor atartup* An4 1n11rt
- o~* control rod prior ~o reducin9 r1ciiculation flow below SOI during* utdown to avoid operation in pot*ntl&11Y un1ta~l*.*
- area, pow1:/flow aap1 (e) iaun*diatelf *c~am the r11=tor if (a) or () al:>ove are not *ucc111ful £n prev1nt1n9 and auppre11ln1 01cillation1. ~h* lic1n**** aha11 *~bait th***
procedu.r**
o th* HRC for review and approval.
- c.
ctor, NM, 1hould iu* an order requirin; all General El* tr1c an 11cen**** to d1fflon1trat* that all 1ioent4 oparator1 a 4 lhi!t Technical Advi1or1 bav* >>**n tr&1n*4 to.
th*** pro~* ur*** Additionally, th* order 1hould require an.
l~o*n****
en1ur1 that all licenatd op1rator1 and lhitt feohnica1 dviaora have race1ved t.ra1n1nt in ant have exhibite.cl, through t11tln1 and
- xam1natlon, docum1nt1d 1imd1r1tan4i 9 of the nature and dan9ar1 of povr 01c111ation1, the therm&l hydra~11e eonditlon1 cau11n9 power 01c1111tion1, how auch co dition can~* avoided, how power 01c111atlon1 can
~* dtt*~t*, and the need to promptly terminate power oac111at1on throg9h acrammin1 the reactor.
Tbe 1io*n**~*
aha11 1ubm t evidence that the above M& ur** have b**n completed,
- z:e1ult1 of approval.
ncluding th* content of tra1nin9 prc9ram1 and the
- t1n; and examination1, *o the HlC for review and D.
ctcr, HlR, hou1d t11ue an order r*q~irin9 &11 G*n~r*l 11 ctrio IWA llc*I'****
to.
d1mon1irai1 that.
in1trum*nta ion raadllf av&11ab1* to oontrol room operator* i1 capable o! detecting and alertin1 tbe op*~*tora to power
01cillation1( including out*of*pb*** or a1yfflfflitrica1 power 01c111atLonj*
!he 11cen****
aba11 aubmit evidence demon1tr*t1 that 'their intrWDentation ***t*
th***
requiremant1 tot.he HRC !or review and approval.
Z, the ~irectcr, MM, *hou14 £11ue an order requ1r1n; General Z1eotric.Ja 11cen**** to develop aimulato~* capable of ode11n; poJer 01cillation1 auch ** tho** ooQurrint at Lala11*
I and a110 ou 1
- of*pha
- or a1ymm1trica1 power 01ci11at1ou.
(tb1 lican****, Gen*r*.1 Zlact.ric, or t.ha IWl own1r1 Group
- r develop one auch 1lmulator and **nd operator* from 'their facilitie1 ~r trainint on it.J After thil 11mulator 11 I
develop_td *r built, a1~ IWl 1101n1e11 11\\llt include 1im1.a1ator training on rover 01c111ation1 for &11 lieen11d operator,, the MRC 1hou14 valuate t.ha capab11iti** of the aimulator and th*
e adequacy of
- trainin9 on 1,.
- r. !ht Director, IOUl, abould i11ua an order to all a1nara1
- 11ectr1c I 11o*n**** r1qulr£n1 them to report to the Mac all t**t in~id* ta in Which t.h* r*=1rcu1atlon pump hava tripped
_off, and th, cau1ativ* event,ud ~**ult1, and a11 LnaLd1nu 1nvo1v1nt,o.wer oac111atlon,1 1:h* MP.C'* order ahould *.*tabllah reportint J*qulr1m1nt1 tor all futur~
.t.no1denta
- or reeiroulati n pwnp trip and/or power O}C1~1ation1.
- o. the C~ffl 111ion ahould reopen Gentr1c 111ue1 1*11 and 1*51
),...
r~1*suk1n9. ~h* A'tWI reao1ut1on 1ho~ld d*peftd on maa u~** o *r. than tripping th* rec1rc~1at1on pump, to rapidly reduce rea.c iv1ty, auc*b aa an automatic, hi;hvcapacity atan~y liquid cont ol ay tem. 1.rhe Comm1aaion ahou14.a110. r*con1id*r
~* u11 of
- lnd*o!*Cycle aec1rculation,wnp trip on 1n1.
Th* Directo, RllR, *hou14 i11u* &n order requirinf all G*neral
- Electric aw lic*n**** within 30 day1 to aumit t.Q *th* nc for zeview and pprova1 a ju1t1!101tion for aontin\\.1e4 operation in 119ht of th* unre101ve4 Ntter &nd t!\\1 reopened v*n*rlc i11u11 and ATWI r 1emakin;.
apprcpriate_juatification for eontinue4 operation *hall have it facilit op~ratin; lic*n** 1u1p1nd1d until thl1 i11u1 11 r1101ved.
- s. ~h* Dir ctor, NllR, *hould i11u1 an order requi&"'inf &11 Gen*r*l Ile tr1c an lican**** v1thin one rear to *~m,, a repor\\ :to th* mu: for r*vi1w and ap~roval demonat.ratlnf compliance 1th GDC 12.
Any 1lo*n*** falling *o eitab1iah co1npU,anc* i\\b GDC 12 1ba11 have it* tao11ltf, operatinf
. hlpttCtf\\111)' 1ubra1t.Ud,
~
- -;X~
lutan 1,. Hiatt OCR& iepr***ntattva 8275 M\\lftlOft J.oad Mentor, OH 4,o,o
- (311) 2,,-:1111
- J.0* I