ML20058A866
| ML20058A866 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 11/17/1993 |
| From: | Schrage J COMMONWEALTH EDISON CO. |
| To: | Murley T Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0519, RTR-NUREG-0800, RTR-NUREG-519, RTR-NUREG-800 TAC-M48058, TAC-M87720, TAC-M87721, NUDOCS 9312010272 | |
| Download: ML20058A866 (29) | |
Text
) Commonwrith Edicon f
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- - ' 7 1400 Opus Place
(,/ Downers Grove, Ilknois 00515 f-November 17,1993 Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington D.C.
20555 Attn. Document Control Desk
Subject:
LaSalle County Nuclear Station Units 1 and 2 Application for Emergency License Amendment to Facility Operating Licenses NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374
References:
- 1. NUREG-0519, March 1981, Safety Evaluation Report related to the operation of LaSalle County Station Units 1 and 2, Docket Nos.
50-373 and 50-374, Commonwealth Edison Company.
- 2. NUREG-0800, June 1987, Standard Rs 3w Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR edition.
Standard Review Plan (SRP) No.15.4.9, Spectrum of Rod Drop Accidents (8WR), Rev. 2.
- 3. Letter from J.A. Miller (CECO-LaSalle Site Engineering) to LaSalle Station Site Engineering Manager, dated September 1,1993 transmitting evaluation of the General Electric Co. analysis dated August 25,1993, from D.R. Rogers, General Electric - Nuclear Engineering concerning Revised Analysis of LaSalle Control Rod Drop Accident.
- 4. Memorandum, L.S. Rubenstein, NRC, to G.C. Lainas, NRC, February 15,1983, " Changes in GE Analysis of the Control Rod Drop Accident for Plant Reload (TAC-48058)."
- 5. General Electric document NEDE-24011-P-A-10-US, " General Electric Standard Application for Reactor Fuel (GESTAR-ll), Supplement for United States, dated March 1991.
- 6. Commonwealth Edison Co. Nuclear Fuel Services Report, NFSR-0075, Rev. O, " Control Rod Sequence Simplification",
December,1989.
- 7. Letter from R.J. Christensen (CECO Engineering) to J.A. Miller (LaSalle Site Engineenng) dated October 13,1993, Human Factors engineering review of the LaSalle County Station Units 1 and 2 License Change Amendment Request.
9312010272 932117 PDR ADDCK 05000373 f
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- Dr. Thomas E. Murley November 17,1993
- 8. Letter from W.E. Morgan to Dr. T. E. Murley dated September 10, 1993, LaSalle County Station Units 1 and 2 Request for Approval of an Unreviewed Safety Question concerning the Mechanical Vacuum Pump.
- 9. Letter from J.L. Kennedy to D.L. Farrar dated September 23,1993, Request for Approval of an Unreviewed Safety Question Concerning the Mechanical Vacuum Pump at LaSalle County Station, Units 1 and 2 (TAC NOS. M87720 and M87721) i Dr. Murley,
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Pursuant to 10 CFR 50.91(a)(5), Commonwealth Edison (CECO) requests NRC approval of an emergency amendment to Facility Operating Licenses NPF-11 and NPF-18.
The proposed emergency Laense Amendment requests NRC approval of an Unreviewed l
Safety Question concerning an existing condition for which operator action is proposed for tripping the mechanical vacuum pump in place of the automatic trip described in LaSalle UFSAR section 11.5.2.1.4 and NUREG 0519, section 15.3.4. The manual operator trip of the mechanical vacuum pump and closure of the isolation valve downstream of the pump on i
main steam line high radiation was never installed at LaSalle County Station (LaSalle).
The proposed emergency amendment request is subdivided as follows:
1.
Attachment A provides a description and safety analysis of the proposed
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License Amendment, including justification for approval as an emergency.
License Amendment.
2.
Attachment B provides CECO's evaluation performed in accordance with 10 CFR 50.92(c), which confirms that no significant hazards consideration is involved.
i 3.
Attachment C provides the Environmental Assessment.
l l
4.
Attachment D provides a copy of Reference (3); letter from J. Miller (CECO-LaSalle Station Engineering) to LaSalle Station Site Engineering Manager, dated September 1,1993, evaluating the General Electric analysis concerning the Revised Analysis of LaSalle Control Rod Drop Accident (General Electric to i
CECO letter dated August 25,1993, copy attached).
Commonwealth Edison respectfully requests review and approval of this proposed amendment to the LaSalle Station Facility Operating License by November 27,1993. This approvalis na.:essary prior to achieving the heating range during restart of LaSalle Station Unit 2 following the fifth refuel otage (L2R05).
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Dr. Thomas E. Murley November 17,1993 i
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This proposed amendment has been reviewed and approved by CECO On-Site and Off-l Site Review in accordance with Commonwealth Edison procedures.
I To the best of my knowledge and belief, the statements contained above are true and i
correct. In some respect these statements are not based on my personal knowledge, but obtained information furnished by other Commonwealth Edison employees, contractor employees, and consultants. Such information has been reviewed in accordance with i
company practice, and i believe it to be reliable.
Commonwealth Edison is notifying the State of Illinois of this application for
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amendment by transmitting a copy of this letter and its attachments to the designated state official.
- r 5
i Very truly you l
1 i
ohn chr e Nuclear Licensing Administrator i
Attachments:
I A.
Description and Safety Analysis.
l B.
Evaluation of Significant Hazards Consideration.
C.
Environmental Assessment.
D.
Letter from J. Miller (CECO-LaSalle Station Engineering) to LaSalle Station Site Engineering Manager, dated September 1,1993.
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cc:
J.B. Martin, Regional Administrator - Rlll J.L. Kennedy, Project Manager - NRR G.F. Dick, Project Manager - NRR A.T. Gody, Project Manager - NRR D. Hills, Senior Resident inspector - LaSalle R. Hague, Branch Chief - Rlli Office of Nuclear Facility Safety - lDNS l
Signed before me on this l'/ day of Nuw b,1993,)L-OFFICIAL SEAL by')ne uefC,c).
MARYEll_EN D LONG Notaryfublic
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"^'"' H+nc s1 An er enois uv convuon o,rwwomoim vwwweuwsw
f Attachment A Description and Safety Analysis of the Proposed License Amendment Desenpliottolthe Prooosed Change The proposed emergancy change concerns an existing condition for which operator action is proposed for tripping the mechanical vacuum pump in place of the automatic trip described in LaSalle UFSAR section 11.5.2.1.4 and NUREG 0519, section 15.3.4. Pending approval of this change, as a compensatory measure, the mechanical vacuum pump suction valve was taken Out-of-Service in the closed condition on Unit 1 and Unit 2. Also, a Caution Card was placed on the mechanical vacuum pump control room control switch to alert reactor operators to the concern.
Per an analysis performed by General Electric Co. (GE), transmitted to CECO by Reference 3, continuous operation of the mechanical vacuum pump (neither automatic trip nor operator action to manually trip) during the design basis control rod accident will result in exceeding the dose limit of SRP 15.4.9, which was the LaSalle licensing basis per NUREG 0519, section 15.3.4. Per the same GE analysis, it was determined that tripping the mechanical vacuum pump in less than or equal to 15 minutes after the release to the main condenser will assure that the dose limit of SRP 15.4.9 is not exceeded.
Descriotion of the Current Facility Ooerating License Reauirement The original design input from GE, as supplier of the Nuclear Steam Supply System (NSSS), included provision for closure of an offgas valve and trip of tho mechanical vacuum pump on high main steam line radiation from the Process Radiation Monitors and later from the Reactor Protection System (RPS) relay logic for reactor scram on main steam line high radiation. Commonwealth Edison Co. chose not to install the trip of the i
mechanical vacuum pump or the associated isolation valve. This decision was i
documented in FSAR amendment 57, dated July 1981, which deleted reference to main steam line high radiation trip of the mechanical vacuum pump and closure of its isolation valve from FSAR section 7.3.1.1.2.4.2. However, section 11.5.2.1.4, still stated that the trips occurred. Also, NUREG 0519 was received at about the same time as the FSAR amendment, but the discussion of the mechanical vacuum pump trips in section 15 of NUREG 0519 was apparently overlooked with respect to the change being made.
Bases for the Cunende.Quirement The bases for the Main Steam Line High Radiation scram and main steam line isolation is to j
limit the release of fission products from the reactor during the design basis control rod j
drop accident. The UFSAR section 15.4.9 and NUREG 0519 section 15.3.4 both assumed the same defined quantities of fuel failures, with a certain percentage of lodines and all of the Noble gases of fission products released from the fuel and instantly transported to the condenser. However, the UFSAR design basis control rod drop accident assumed a fuel peaking factor of 1.0 (design basis), versus a peaking factor of 1.5 used in the NU'9EG 0519 (licensing basis) accident analysis performed and accepted by the NRC for LaSalle.
The licensing basis of NUREG 0519 also included the following:
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8 Attachment A (cont.)
l 1.
" Detection of a high radiation signalin the main steam lines automatically closes the main steam isolation valves, shuts down thernechanical vacuum pump and closes the isolation valvedownstream of the pump."
i 2.
"The resulting doses shown in Tabie 15.2 are less than the acceptance cnteria of Section 15.4.9 of the Standard Review Plan and are well within the 10 CFR Part 100 guidelines."
Descriotion of the Need for Amending the Facility Ooerating License On August 27,1993, the 10 CFR 50.59 Safety Evaluation performed for this pre-licensing change resulted in an Unreviewed Safety Question (USO) determination. This was based on the GE analysis performed, which indicated that the SRP limits for dose consequences for the licensing basis control rod drop accident will be exceeded if the mechanical vacuum pump continues to run during the accident. In order for the dose consequences to remain less than the SRP limits for this accident, the GE analysis indicated that the mechanical vacuum pump must be manually tripped in less than or equal to 15 minutes of the accident. (The release begins at time zero, at w%n time the main steam line high j
radiation trip signals and alarms are also received.) The upstream valves that provide condenser suction for both the Off Gas system and the mechanical vacuum pump are subsequently closed by removal of power fuses to the valve control solenoids. This is a i
subsequent action, since tripping of the mechanical vacuum pump removes the ' driving.
force.for the release and with a vacuum in the condenser, no further release can occur until the condenser'is back at atmospheric pressure.
Without approval of the use of operator action to trip the mechanical vacuum pump within 15 minutes of the receipt of a High Main Steam Line Radiation annunciator and/or entry into the " Fuel Element Failure" abnormal procedure, LaSalle will be unable to start up a unit after the Off Gas system steam jet air ejectors stop functioning on shutdown and condenser vacuum is lost.
Descriotion of the Amended License / Technical Specifications There is no change required to the Technical Specifications. The trip of the mechanical vacuum pump within 15 minutes and subsequent closure of upstream isolation valves by An emergency amendment to the Unit 1 and 2 Facility Operating Licenses is required to l
i document NRC acceptance of operator action instead of automatic action to mitigate the consequences of the design basis control rod drop accident.
Bases for the Change to the Facility Ooerating_Lic.ense The requirement for immediate operator actions to trip the mechanical vacuum pump within 15 minutes of a main steam line high radiation trip signal is adequate based on the following:
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t Attachment A (cont.)
1.
LaSalle abnormal procedures were revised to require remote manual trip of the condenser mechanical vacuum pump (vacuum pump) as an immediate operator action, based on the determination that the mechanical vacuum pump was assumed to automatically trip. Information concerning the need to complete this action within 15 minutes to remain within the analysis limits for off-site dose is provided in the procedures. The " Fuel Element Failure" abnormal procedure includes subsequent operator actions requiring the closure of the upstream valves (two valves in parallel,1(2)N62-F300A/B).
2.
For the purpose of the analysis by GE, the mechanical vacuum pump trip was assumed to occur 15 minutes after the initiation of the event. This -
based on the assumption that all of the fission product gases and lodines a.
transported to the condenser instantaneously, which would cause Main Steam Line High Radiation trip and alarm at the same time. Operator action is not required during the first 15 minutes of the event because the activity released will remain less than the Standard Review Plan limit of 25% of 10 CFR 100 so long as the mechanical vacuum pump is tripped less than or equal to 15 minutes after the start of the accident.' The assumptions used and the results of the analysis are included as Attachment E.
t 3.
The assumptions of the Control Rod Drop Accident analysis are conservative with respect to the realistic or actual values or practice. A comparison of
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the conservative assumptions versus the more realistic case are as follows (even though not taken credit for in either the original or new analyses that have been performed):
4 a.
GE uses 10 rod groups for the analysis, LaSalle subdivides these into 12 groups. The smaller groups reduce radial peaking and incromental rod worths, resulting in lower fuel enthalpies.
b.
GE uses an adiabatic model to calculate the peak fuel enthalpy,.
Brookhaven National Laboratory (BNL) has analyzed for the NRC the CRDA using appropriate thermal-hydraulic feedback. BNL results show the peak fuel enthalpy to be well below 150 cal /gm for a 1.5%
4k rod worth compared to GE's analysis of 280 cal /gm for a 1.42 %
ak rod worth.
4.
Fifteen minutes for this operator action is reasonable time to respond to alarms based on licensed reactor operator training, including simulator training, and a Human Factors Task Analysis (Ref. 7). The LaSalle Licensed Operator Requalification Training scheduled from October 25,1993 through the week of December 6,1993 includes a startup drill. The drill scenario summary for this training in 1993 is as follows (LGP-1-1 is the LaSalle
" Normal Unit Startup" procedure):
1 This drill begins with a startup in progress with the reactor already around 60 psig. (The reactor was shutdown and kept hot (all rods in I
Attachment A (cont.)
and MSIVs closed).) Some control rods are already withdrawn but the reactor is not yet critical. Gland sealing steam is stillin service and the OFFGAS vacuum pump (hogger) is running, establishing vacuum.
When the reactor is just critical, but before the heating range is l
reached, a control rod, which was previously withdrawn but uncoupled and NOT checked at notch 48, will fall out of the core and insert sufficient reactivity to cause fuel damage. Steam line radiation will increase and stack radiation will subsequently increase. The crew is expected to secure and isolate the Off Gas vacuum pump to prevent an unprocessed release path out of the stack, within 15 minutes of the Main Steam Line High Radiation Trip alarm receipt.
A Human Factors Task Analysis has been performed by Commonwealth Edison and found acceptable assessing the actions to be performed by the control room operator. The Human Factors Task Analysis was performed by an experienced Human Factors Engineer (10 years experience) and included the following:
a.
A LaSalle Nuclear Station Operator (NSO, licensed reactor operator) was interviewed for the purpose of performing a task analysis j
regarding the circumstances surrounding a postulated Control Rod i
Drop Accident (CRDA) and subsequent receipt of a Main Steam Line l
High Radiation trip alarm. This interview was conducted to determine if any human factors issues would prevent an operator from tripping the Main Condenser Mechanical Vacuum Pump within 15 minutes of
-i the Channel A1/B1 (A2/B2) Main Steam Line High Radiation trip
- l alarm.
'i b.
With the addition of a caution regarding this potential event to the appropriate procedures prior to the steps that start the Mechanical I
Vacuum Pump, it was determined the NSO will have no difficulty in tripping the Mechanical Vacuum Pump within 15 minutes of receiving the Main Steam Line High Radiation trip alarm.
The Human Factors determination to accept the administrative solution to manually trip the pump is based on the following:
a.
A postulated sequence of events. _
b.
A minimum'of two NSOs associated with the unit in startup.
c.
Cautions and action steps added to applicable procedures.
The staffing levels during, Unit t. tart-up and the presence / addition of action steps and cautions in the appropriata p6&ec creates a situation where there is no undue task loading or time stressors t.q the operators. This was supported by the simulator drills that have been performed to trair operating crews on this accident i
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Attachment A (cont.)
i analysis. The Mechanical Vacuum Pump has been tripped well within 15 minutes j
during the simulator drills to date.
This change (no automatic condenser mechanical vacuum pump trip) does not affect in any -
way the probability that a CHDA will occur. The CRDA is a limiting fault frequency accident, and therefore not expected to occur during the life of the plant. -Technical Specification controls are provided to ensure this limiting fault frequency is maintained.
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The control rod drop accident is primarily a startup concern (control rods are being i
I withdrawn). As each control rod is withdrawn, it is verified to be coupled to its drive mechanism in accordance with Technical Specification 4.1.3.6:
l 4.1.3.6 A control rod shall be demonstrated to be coupled to its drive mechanism by observing any indicated response of the nuclear instrumentation while withdrawing the control rod to the fully withdrawn position and then. verifying that the control rod drive does not go to the overtravel position:
J a.
Prior to reactor criticality after completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity,
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b.
Anytime the control rod is withdrawn to the " Full out" position -
in subsequent operation, and Following maintenance on or modification'to the control rod or i
c.
control rod drive system which could have affected the control rod drive coupling integrity.
This required Technical Specification surveillance helps assure that control rods are coupled to the drive mechanism during startup, and thus minimizes the p'ossibility of I
a control rod drop accident.
During reactor shutdown, control rods are inserted, rather than withdrawn. Also, during unit operation above the low power setpoint of Rod Worth Minimizer (RWM),
I Technical Specification surveillance 4.1.3.1.2 is performed, which requires the following:
-l 4.1.3.1.2 When above the low power setpoint of RWM, all withdrawn control rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:
a.
At least once per 7 days, and b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is immovable j
as a result of excessive friction or mechanicalinterference.
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I Attachment A (cont.)
l Therefore, within the previous 7 days, every control rod that is not already fully inserted has been moved, and each control rod that is fully withdrawn receives a coupling check --
when it is returned to full-out after exercising per the above surveillance requirement. This -
'I surveillance, in addition to inserting control rods instead of withdrawing control rods (as -
during startup), minimizes the possibility of a control rod drop occurring during unit shutdown. Therefore, although not normally operated during shutdowns, operation of the l
mechanical vacuum pump during a shutdown is of minimal concern. However, the same immediate operator actions are required regardless of whether the unit is being started up or being shut down with the mechanical vacuum pump in operation.
Justification for an Emergency License Amendment l
t In accordance with 10 CFR 50.91(a)(5) a condition exists requiring an emergency License Amendment involving no significant hazards consideration. The justification is provided for the following conditions that are required to be met per 10 CFR 50.91(a)(5) to justify emergency approval of this License Amendment:
1.
The situation could result in the shutdown or derating of one or both station units,-
or the situation could prevent the resumption of unit operation or of ascension to full rated power.
LaSalle Unit 2 is currently in its fifth refuel outage (L2R05) and is currently scheduled to begin startup on November 27,1993. Approval of this License '
Amendment is required prior to use of the mechanical vacuum pump during startup.
If not approved in time, then Unit 2 startup would be delayed, at high cost to Commonwealth Edison.
2.
The circumstances leading to the need for an emergency License Amendment request could not be avoided.
j On September 10,1993, CECO submitted a request to the NRC for approval of the Unreviewed Safety Question (USO) conceming the Mechanical Vacuum Pump (Reference 8). CECO requested approval of the USQ by November 15,1993 in order to support the originally scheduled end of L2R05.
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1 The License Amendment request was not issued for public comment in accordance
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with 10 CFR 50.91 due to insufficient communication between CECO and the NRC j
with respect to whether a License Amendment was required for the USO. The fact that the request did not require a change to Technical Specifications also added an element of confusion.
On September 14,1993, LaSalle Unit 1 scrammed as a result of a loss of the Unit 1 System Auxiliary Transformer (SAT). This began a 22 day forced outage to repair -
various equipment. On September 23,1993, the NRC issued a letter (Reference 9) l
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Attachment A (cont.)
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based on discussions with NRR regarding the USQ for the Mechanical Vacuum Pump and the need to use the Mechanical Vacuum Pump during Unit 1 startup.
The letter allowed the use of the Unit 1 Mechanical Vacuum Pump to draw a vacuum on the main condenser during startup of Unit 1. This was based on a preliminary review of the LaSalle submittal (Reference 8). The NRC staff determined that LaSalle County Station, Unit 1, could be safely returned to operation with the operator actions and administrative controls that were in place.
i On October 20,1993, CECO and NRC representatives determined that a License
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Amendment would be required for approval of the USQ. Determination was made that the USO would need to be resubmitted in the form of a License Amendment.
At this time, both CECO and the NRC believed that another letter could be written I
to allow Unit 2 to use the mechanical vacuum pump during startup from L2R05.
Also, the NRC requested additional information concerning operator training.
The additional information requested by the NRC was based on a simulator drill that l
was planned for the Licensed Operator Requalification Module beginning the week of October 25,1993 and ending December 10,1993. Three weeks of operator I
training were completed before evaluation was made of the ability of the operating crews to carry out the required compensatory actions for the Control Rod Drop '
Accident.
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During preparation of a revised submittal to clarify the need for a License Amendment as the mechanism for NRC review and approval of an Unreviewed l
Safety Question, CECO determined that the amendment would be required for LaSalle Unit 2 to startup from its refuel outage. CECO made this determination on November 15,1993, with concurrence from the LaSalle NRR Project Manager. This License Amendment is necessary to support start-up following L2R05 on November 27,1993, therefore the circumstances require approval of the License Amendment on an emergency basis.
3.
The situation was not created by failure to make a timely application for a license amendment, because The original CECO submittal of the Unreviewed Safety Question (USQ) (Reference 8) requested approval prior to Unit 2 startup, which was originally scheduled for November 15,1993. The request included a finding of no Significant Hazards Consideration, but was not notarized as is required for an amendment to the License. On October 20,1993, CECO and NRC representatives determined that a l
License Amendment would be required for approval of the USQ. Based on 1
Reference 9, which allowed LaSalle Unit 1 to restart using the Mechanical Vacuum i
Pump during startup, CECO and the NRC believed that a similar letter could be issued by the NRC to allow LaSalle Unit 2 to startup from its refuel outage. At that time, CECO initiated efforts to revise the USQ/ License Amendment submittal. This included additional information in support of the License Amendment that was requested and obtained concerning licensed operator training and a Human Factors Task Analysis. The Human Factors Task Analysis performed concerned the
.l Attachment A (cont.)
i operatar actions required to mitigate the consequences of a Control Rod Drop i
Accident.
i During the On-Site Review of the revned submittal (November 12,1993), CECO -
questioned the need tor approval prior to Unit 2 startup, and discussed this with the NRR Project Manager. On November 15,1993, the NRR Project Manager prcvided confirmation that the License Amendment would be required prior to startup following the current Unit 2 refuel outage. Therefore, based upon the original submittal (Reference 8); the confusion surrounding the processing of the Unreviewed Safety Question; and the interim communications between CECO and the NRC, the situation was not created by a failure to make a timely application.
4.
The situation does not involve a significant hazards consideration.
The Significant Hazards Consideration, performed and attached as Attachment C,-
determined that the lack of an automatic trip of the Mechanical Vacuum Pump on l
Main Steam Line High Radiation does nat involve a significant hazards consideration.
Therefore, this emergency License amendment is justified in accordance with 10 CFR 50.91(a)(5).
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B 6
1 Attachment B Evaluation of No Significant Hazards Consideration Commonwealth Edison has evaluated the proposed change to the Facility Operating Licenses and determined that it does not represent a significant hazards consideration.
Based on the criteria for defining a significant hazards consideration established in 10 CFR -
50.92, operation of LaSalle County Station Units I and 2 in accordance with the proposed change will not:
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involve a significant increase in the probability or consequences of an accident previously evaluated because:
The lack of an automatic trip and isolation of the LaSalle Unit 1 and Unit 2 mechanical vacuum pumps does not change the accident initiators for a design basis control rod drop accident or the inventory of fuel fission products available for release during this accident. Therefore, the probability of the design basis control rod drop accident is not changed.
The lack of an automatic trip and isolation of the LaSalle Unit 1 and Unit 2 mechanical vacuum pumps does not significantly increase the consequences of the design basis control rod drop accident provided that the mechanical vacuum pump is tripped within 15 minutes of receiving the main steam high radiation trip alarms. Fifteen minutes for this operator action is reasonable time to respond to alarms based on licensed reactor operator training, including simulator training. The trip is accomplished with a hand switch located on the Main Control Room front panels. A Human Factors Task Analysis has been performed by Commonwealth Edison and found-acceptable assessing the actions to be performed by the control room operator. Also, the time that the mechanical vacuum pump operates during reactor startup, approximately 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, does not affect the probability of the design basis control rod drop accident.
UFSAR section 15.4.9 states that a rod drop does not exceed the 280 cal /gm design limit and failure of fuel cannot result naturally from a control rod drop accident. This determination was based on the following input parameters and initial conditions:
At the time of the control rod drop accident the core is assumed to be at a cycle point which results in the highest control rod worth. The core is also assumed to contain no xenon, to be in a hot-startup condition, and to have the control rods in sequence at a 50% rod density. The assumption to remove xenon, which competes well for neutron absorptions, increases the fractional absorptions, or worth of the control rods. The 50% control rod density assumption, (" black and white" rod pattern), which nominally occurs at the hot-startup condition, ensures that withdrawal on the next rod results in the maximum increment of reactivity.
D.
Attachment B (cont.)
The control rod drop accident analysis is performed as described in:
General Electric document NEDE-24011-P-A-10-US, " General Electric Standard Application for Reactor Fuel (GESTAR-II), Supplement for United States, dated March 1991.
if the worth of any control rod is determined to be greater than 1 %
Ak/k, a cycle specific control rod drop accident analysis is performed accordance with:
Commonwealth Edison Co. Nuclear Fuel Services Report, NFSR-0075, Rev. O, " Control Rod Sequence Simplification", December,1989.
j The analysis for each unit's current cycle performed per NFSR-0075 verifies that heat generated during a control rod drop accident is less than the 280 l
cal /gm design limit.
The assumptions of the Control Rod Drop Accident analysis are conservative
.l with respect to the realistic or actual values or practice. A comparison of j
the conservative assumptions versus the more realistic case are as follows l
(even though not taken credit for in either the original or new analyses that.
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have been performed):
a.
GE uses 10 rod groups for the analysis, LaSalle subdivides these into 12 groups. The smaller groups reduce radial peaking and incremental rod worths, resulting in lower fuel enthalpies.
b.
GE uses an adiabatic model to calculate the peak fuel enthalpy, _
Brookhaven National Laboratory (BNL) has analyzed for the NRC the
'i CRDA using appropriate thermal-hydraulic feedback. BNL results show the peak fuel enthalpy well below 150 cal /gm for a 1.5% Ak fod worth compared to GE's analysis of 280 cal /gm for a 1.42 % Ak rod
- worth, l
Based on the above, there is not a significant increase in the probability or consequences of the design basis control rod drop accident.
2)
Create the possibility of a new or different kind of accident from any accident
. previously evaluated because:
This change specifically affects the design basis control rod drop accident, and is the only low power event that involves release of fission products to the main -
condenser. The only difference between the accepted analysis and the new l
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analysis is the rate of release from the main condenser and a ground level release (original analysis) versus an elevated (from the station vent stack) release, for the i
new analysis. Therefore, the change does not create the possibility of a new or different kind of accident.
4 4
Attachment B (cont.)
l 3)
Involve a significant reduction in the margin of safety because:
The margin of safety that is affected by this change involves the radiological consequences of the design basis control rod drop accident. This margin of safety is based on the Standard Review Plan, section 15.4.9, which states that the calculated whole-body and thyroid doses at the exclusion area boundaries (EAB) and at the low population zone (LPZ) boundaries are well within the exposure guideline values in 10 CFR part 100, section 11, if the doses are less than 25% of the 10 CFR Part 100 exposure guideline values or 75 rem for the thyroid and 6 rem for j
whole-body doses. If the mechanical vacuum pump is manually tripped in less than or equal to 15 minutes after the receipt of the main steam line high radiation trip alarm, the analysis shows that the radiological consequences of the design basis control rod drop accident are less than 25% of the 10 CFR Part 100 exposure guideline values.
Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule,51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are and are not considered likely to involve significant hazards considerations. These proposed amendments most closely fit the example of a change whir:h may ely analyzed accident or may reduce in some way a safety margin, but where the results are clearly within all acceptable criteria with respect to the system or component specified in the Standard Review Plan section 15.4.9.
This proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings or a significant relaxation of the bases for the limiting conditions for operations.
Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed change does not constitute a significant hazards consideration.
Attachment C Environmental Assessment i
l Commonwealth Edison has evaluated the proposed amendment against the criteria for identification of licensing and regulatory action requiring environmental assessment in accordance with 10 CFR 51.20. It has been determined that the proposed changes meet the criteria for a categorical exclusion as provided under 10 CFR 51.22(c)(9). This l
conclusion has been determined because the changes requested do not pose significant hazards considerations or do not involve a significant increase in the amounts, and no significant changes in the types, of any effluents that may be released off-site.
Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure.
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Attachment D Letter from J.A. Miller (CECO-LaSalle Site Engineering) to LaSalle Station Site Engineering Manager dated September 1,1993; Evaluation of the General Electric Co. Analysis transmitted by letter dated August 25,1993, from D.R. Rogers, General Electric - Nuclear Engineering Concerning Revised Analysis of LaSalle Control Rod Drop Accident.
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September 1, 1993 In reply refer to CnRoN # 121932
Subject:
LaSalle County Station, Units 1 and 2 Lack of Automatic Trip of Main Condenser Mechanical Vacuum Pump Deviation Report No. 01-01-93-030 AIR No. 373-200-93-03002 System Codes: DIS (PR), N62 (OG)
Reference:
Letter dated August 25, 1993, from D. R. Rogers of GE Nuclear Energy to E. L. Seckinger transmitting an analysis of the LaSalle control rod drop accident with the Mechanical vacuum Pump in operation.'
To: SEC Manager The subject DVR described a design deficiency with the Main Condenser Mechanical Vacuum Pump (MVP), 1(2)OG02P, and the associated pump line isolation valves, 1(2)N62-F300A and B.
Contrary to UFSAR section 15.4.9, " Control Rod Drop Accident (CRDA)," and NUREG 0519, " Safety Evaluation Report Related to the Operation of LaSalle County Station Units 1 and 2," section 15.3.4, the MVP does not automatically shut down and the isolation valves do not automatically close on high main steamline radiation.
The only accident or transient which requires this trip is the CRDA.
At our request, GE Nuclear Energy performed a radiological analysis of the CRDA with the MVP running continuously and with it tripped after 15 minutes.
This analysis was performed in accordance with NEDO-31400 which has been approved by the NRC and with the same initial conditions and input parameters listed in UFSAR section 15.4.9 with the following exceptions:-
- 1. A radial power peaking factor of 1.5 was used to agree with the value used by the NRC in the SER.
Note: The peaking factor used in the UFSAR analysis was 1.0 and results in lower off-site dose levels.
- 2. Atmospheric dispersion factors calculated in accordance with Regulatory Guide 1.145 were used.
- 3. While the MVP was running, the leak rate from the condenser was equal to the MVP flow rate (2850 cfm) and was an elevated release.
The GE analysis shows that if the MVP runs continuously after a CRDA, the two-hour whole body dose at the Exclusion Area Bounaary (EAB) will exceed the limits of NRC Standard Review Plan.
(SRP) 15.4.9, but not 10CFR100 limits.
The two-hour thyroid dose e
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4 at the EAB, the two-hour whole body and thyroid doses at the Low Population Zone (LPZ), and the 30-day whole body and thyroid doses at the EAR and LPZ are within the SRP 15.4.9 limits.
The SRP 15.4.9 dose limits are 25% of 10CFR100 limits.
If the MVP is tripped within 15 minutes af ter a Main Steam High Radiation trip, the off-site doses are reduced by at least 30%, and the two-hour whole body dose at the EAB drops within SRP 15.4.9 limits.
The calculated off-site doses are listed in the referenced GE letter.
The Plant Support Group of the Site Engineering and Construction (SEC) Department has reviewed and concurs with the GE analysis.
Therefore, SEC has determined that this "de facto" change to the facility as described in the SAR represents an unreviewed safety question because operator action is required to meet the off-site dose limits of SRP 15.4.9 following a CRDA.
SEC recommends the following corrective actions:
- 1. Request NRC approval for replacing the automatic trip of the MVP and associated isolation valves after a High Main Steam Line Radiation trip with administrative controls since operator actions are not required for 15 minutes and can be easily accomplished at Control Room panel 1(2)N62-P601.
- 2. Revise station procedures LOA-NB-08, "Puel Element Failure", LOA 1 (2) H13-P603-B305, " Channel Al/B1 Main Steam Line High Radiation", and LOA 1(2)H13-P603-B309,
" Channel A2/B2 Main Steam Line High Radiation", to require the operators to verify that the MVP is shut down within 15 minutes after a Main Steam High Radiation trip.
- 3. Revise the applicable sections of the UFSAR to agree with the as-built configuraticn of the plant with respect to the MVP and to revise the CRDA analysis.
Regulatory Assurance is currently preparing the appropriate documentation for submittal to the NRC.
To assist in the completion of Item 1, SEC has prepared a Safety Evaluation for on-site review.
Iten 2 is complete.
SEC has initiated the required procedure changes, and operating has approved them.
Item 3 is scheduled to be completed within approximately two months and requires the submittal of additional analysis by GE.
Because of the large unexplainable difference in results between the SER and UESAR CRDA analysis, SEC has detenmined that a new CRDA analysis with immediate tripping of the MVP is required.
The completion of this item is being tracked by the subject AIR. --
A copy of the referenced GE document is available through the subject CHRON entry.
If you have any questions, please contact Ed Seckinger at extension x2005.
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f J. A. Miller Station Site Support Eng. Supervisor ELS\\
Attachment cc: J.V. Schmeltz (w/o att)
J.E. Lockwood (w/att)
T.A. Hammerick (w/att)
E.L. Seckinger (w/o att)
V.K. Gilautra (S&L) (w/o att)
NEDCC (w/att)
CHRON (w/att)
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?EY.$U?E.w ussus August 25. !$93 cc: $. 3. Wang DRF A00 05497 E. L. Seekinger Ceesonwealth Edison corpany LaSalle County Conrating Station Marssilles. IL 61341
Subject:
Revised Analysis of LaSalla control Rod Drop Accident With Mechanical Vacuta Pump In Operation
References:
- 1. Telecopy, E. L. Sackinger (Ceco) to 0. R. Angers (CE NE).
8-19-93.
- 2. Letter, D. R. Rogsrs (GE NE) to E. L. Seckingar (Ceco),
"LaSalls Control Rod Drop Accident With Machanical Vacuus
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Paep In Operation", 6 15-93.
Catr Mr. Seckinger:
In responsa to your request (Reference 1), the analysis of off-site radiological centequences for a Control Rod Drop Accident (CROA)ly without automatic isolation of the Hechanical Yacuus Pump (HYP) previous reported in Refersnes 2 has baan revised to reflect an assumed radf el power peaking factor of 1.$ and revised Chi /Q values for ground level release from the turbine butidtr4 Tw analysis cases were considered:
(1) no trip of the NYP assumed and (2) r.anual trip of the MVP assmed to occur 15 minutes after the CRDA. The results of our analysis for tha two cases are attached.
Please call if you have any questions concerning this analysis.
Very truly yours.
4. A. %rs
- 0. R. Roge Plant Performance Analysis Proj.
(408)925-1935
ANALYSIS OP CollTROL 203 OROP ACCIODIT EAD10 LOGICAL gggrayDr.ES WITH NECHARICAL VActABI Pl5IP ePERAT!ke An analysis of potential off site redjelegical consequences of a Centrol.
Red Drop Accident-Vacuus Pusp (MVP) h(CRDA) without automatic isolation of the Mechanical as been performed for LaSalle Count request of Commonwealth Edisse Ceepsay (Reference 1). y 5tation at the q
Two analysis cases were requested. These were:
1 o
Case 1.
He trip of the MVP occurs: 1.s., the MVP runs continuously after.4 CRDA.
case 2.
The Mp is manually tripped off 15 minutes after-the CADL occurs.
o The assuptions for the existing Design.tasis CRDA analysts as described i
in Section 15.4.g of the LaSalle UFSAR were augmented by data required l
to include the effect of the MP (References 2 and 3) and modified in I
accordance with Reference I to include..a radial peakin instead of 1.0 and revised ground level Cht/Q values. g factor of 1.5 g
The assumptions for. the analysis am suemarized in Table 1.
Cassistent-I with the UFsAR analysis, 770 fuel rods were assuesd to fall in a CRDA.
j Failure of 170 rods was assumed to apply for est fuel bundles having 42 fuel rods per bundle. The assettens required to evaluate the fission product activity released free tse feel and trans were taken free Section 15.4.9.4. Section 15.4.9. ported to the condenser s.1, and Table 15.4-8 of the LaSalle UFEAA. Applicatten of a radial poner peaking faster of i
1.5 as specified in Reference 1 is consistent with the outdance of to the condenser is assumed to be ava(ilable for release to theReference 4). - A current Standard teview Plan 15.4.9 i
envireneant at time = 0.
After the W P path is isolated, the condenser-is assumed to release activity to the environment at ground level at the -
rate of 1 Wday. Revised Cht/Q values' from Reference I were applied for ground level rolesse of-activity from the turbine building.
j Assumptions for the MVP flow rate (8850 cfm) he Refersace 2 discussio and condesser free values.
J (130.000 cubic feet) are in acconlance with t sad Reference 3 data transalttal. The MVP flew rate the perferesace of the pump at atmospheric pressere. which is based en
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constant, even though paup perfomance any decrease.~ ls assumed to be under conditlens of l
high vaceus. For these assuptions. the MVP removes activity from the 1
condenser at a cesstant rate of approximately 3,157 f/ day.' Activity removed by the Mp is released to the environment via the plant stack.
Chi /0 data applicable to the stack elevated release point was provided j
by CECO (References 2,3).
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The analysis was performed with tho' CONAC03 rectalogical consequence
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esalvationcode(Reference 5). An older version of this code (CONACO!)
is referenced in Section 15.4.9 of the UF54R. $ lace speated data is used in the curr9nt version (including Regulatory Guide 1.100 thyroid inhalation dopo : onversion facters and revised values of fission product specific fr.
nos in Curies /Mw) the calculatten bases are not completely c-ottant with those of the existing UF5AR analysts.
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The results obtained for I hour dese et the Esclusion Area Sevadary and 30 day dose et the Low Popglatten tone for the two cases are presented in Table 2.
The calculated thyroid and whole body deses with MP i
omration terminated at 15 sinutes post-tecident (Case 2 are all within tie NRC acceptance criteria of 75 ren thyreld and 4 res u)kole body as stated in Standard Review Plan 15.4.9 (Asference 41. The standard Review Plan criteria are egelvalent to Its of the ;;0CFR100 lietts. For Case 1. in which WP cperation was assumed to continue without terutnation after the accident, the calculated EA4 uncle body dose would exceed the Standard Review Plan criterios of 6 res.- while the other calculated doses would remain within the standard Reetow Plan criteria.
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Prepared by:
Verified by:
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a$s he &
j 'idy O. R. Aoters ~
- 5. P7 Wang Plant Performance Analysis Proj.
Plant PerforKnes Analysis Proj.
WC 4st (408)S25-1935 WC 449 (404)925-1594 ll
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Rafareneas
- 1. Telecopy. E. L. Seekinger (Ceco) to D. A. Rogers (GE NE), 8-19 93.
1.
Telephone Olscussion, E. L. Seekinger (CECO) and D. R. Rogers (GE.Nf), 6-21-93.
3.
Letter, J. A. Miller (CECO) to D. R. Rogers (GE-lE), *La$alle County i
Station, Units 1 and R. Ellainatten of Main steen Line Radiation Monitor Trips control and Drop Acchent Analysis" CHROII i 119450, 6 25-33.
4.
NUREE-0800 Rev.1, USNAC 5tandard Review Plan, Sectica 15.4.s 1x 5 'Rediological Consequences of Control Rod Drep Accident
(
), July 1981.
5.
NEDD E12431, " Radiological Accident Evaluation The.CONAC03 Code' N. A. Careway, V. D. Nguyen, P. P. Itancawage. December 1981.
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.TASLE 1 A5SLftPfft)N1 FOR ANALYSIS OF CRDA VITH MVP OPERATINA Reactor Power (WWt) 3454 Number of Failed Fuel Rode 770 sumber of Fuel mode per Bundle 1
Number of Fuel Bandles la Core 764 a
Radial Power Peaking Factor 1.s Fusi Rod Plenum Activity Fractions (4):
Noble gases 10 Iodinea 10 Fraction of Fuel Malted in Failed Rods (t) 0.77 j
Activity Released from Malted Fuel (t):
Noble gases 100 Iodines Sc Decay Prior to Release from Fuel (min) o Fraction of Released Activity Transported te condenser (tJ Noble gases 100 Iodines 10 Fraction of Condenser Iodine Activity Airborne (4) 10 Leak Rote from condenser (%/ day) 1 NYF Flow Rate (cfa) 2850 condenser Free Volume (cu. ft.)
130,000 MVP Per16d of Operation (min)
Variable 3
chi /Q for Elevated Release (sec/s ):
l o - 1/2 hr - EAB/I.P5 8.4E-5/5.92-6 1/2 - 2 hr - EAS/LPE 2.6E-6/1.6E-6 i
8 hr 2
LPE 9.23-7 a
- 24 hr us 5.5E-7 j
p 1
4 day -
LPS 2.53-7
-l 4
= 30- day -
LPE 8.2E-4 j
3 Chi /Q for Ground Level Release (sec/m ):
)
0 2
hr - ZAE/LPE 5.1E-4/1.ot-5 3
A hr LP5 1.05 4 8 - 24 hr LPE 5.7E-6 1
4 day -
LPS 2.65-6 4
= 30 day -
LPE 6.55-7 4-1
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a TAtLE t OFF-11TE B311 Frail 12RA WIfW llWP 6 MEAT!Dal er (RIM)
FAR (thel LM (10 div) l h afP omaratten Thyroid tlhala tady Thwretd. Wha' a and, 1.
Centinuous 72.1 8.7 8.8 0.59 2.
Il Minutes 41.3 5.9 4.5 0.83 3RP 15.4.9 Limit 75 6
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NUCLEAR REOULATORY COMMlWCW an w essvest a,e,asm s
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ftp 15 33 -
MD e ulDUM FOR: 8. C. tainesIng Reactors, BLAssistant Mroctor for Operat Flol:
L. 8. bbenstainIsat Systems. 35!
Assistant Director for Cers and F
SUBJECT:
CHANGES IN St ANALYSIS OF THE CONTROL 20 DR0p' ACCIDENT FOR PLANT RELO4DS (TAC 544058)
In a letter of February Id.1982 from R. Engel (CE) to D. Vassalle (kitl General Electric presented its proposal to delete tw Control Rod Drop Accident (CRDA.
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from the standard EE-BWR reload package for Banted Position Withdraal se:
(RPW5) plants. His uns based on a statistical analysis of resvits fra previous reload analyses. His proposal is similar to previous proposals to modify reload svtsittals for tb red withdrawal at power and mislocated assably events. Those previous proposals were accepted arith, however questions about future reload characteristics. Dose questions have been ans,wered and reviewed. The proposal for the CRDA submittals have been reyfewed and the SER indicating that the pro-posal is acceptable is enclosed.
This review also indicates that the responses about the future refoed characteristics is acceptahis for both the previous events and the CRDA as wil.
His completes our efforts for TAC No. 48058.
L. S. Ibibenstein. Assistant Director for Core and Plant Systees. D51 Emelesure.
As stated ec: R. Mattson 5.[isenhet L Capra D. Vassa11e i
V. Roony Reactor Physics Section
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Contact:
N. Richings
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2-29445 1
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a-gv4LUAT10N W TWC E PROPOSAL 10 j
Gn E TNC CtDA ANALY$lt FR(pt SPW5 PLANT t[ LOAD phae!5 T
In a letter of February 24,1982 (Raference 1) Eeneral Electric (GE) proposed to
_ delete the analysis of the control red drop sceident (CRDA) free the standA reload pactage for plants using the banked positten erithdravel. seguence (tpus) for sentrol rod withdrawal. This proposal is based on a statistical analysis by st of the resvits of past CRDA analyses for plaats using SpW5. sad the conclusion that with a 95/95 probability / confidence level Oe. peat b1 enthaply would be (such) less than 280 cal /pn (the staff limit) with the marians control red incremental reactivity worth. The GE proposal ses to delete the j
CaDA analysis from the reload sutraittats beginning April 1982. h staff.
nyiew (hasis ta be discussed) ins favorable, for present populations of -
nactor nloads, although no formal report was written at the time.
This modification of an event analysis for reloads via e statistical analysis of previous results is siellar to two past SE proposals for mioed submitt41 modiff-cations. These proposals tere (1) fbr discontinuing plant-cycle specific mis-located tondte analyses (Novanber 1980, see Refarence 2), and (2) for a change in the analysis of the control rod withdrawal at power event (May 1981, su Reference 3). The staff reviews found both of these modifications to be
^ !
acceptable (see References d.5.g. and 7) for present populations' ef reloads, although the nylevs contimed with questions related to fkture populations of nisads. These questions more mspohdad to by SE (References 8 sne 9). The staff has nyiewed these responses and finds then inforestive and acceptable and concludes that for the presently fbreseeable fkture the characteristics of the reload populations appear to be cenpatible with those sensidered in the statistical analyses.
For the statistical analyses of the CRBA analysis SE has empiled the result:1 from abost so atoads for plants using the BpW5 (or these estag stattar I
patterns for part of the withdrawa1). The control ved reactivity'eerths used j
in CRDA analyses at significant points la the startup sequesice tere examined.
and the statistics of rod worth, facluding average and variance, sere j
2
.. g..,
o 1
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produced. Ilest of.the (relevant) red morths are ender 15 a k and all including a statistically detamined 95/95 probability / confidence level en wil ender 1.85 At. All ref morths used result in peak enthalpfes mall ender 280 cal /p-3 esing the St (Iptc approved) calculational mathed.
c The staff review of these reselts indicate that the red worths and resulting t
peak enthalpfes am fa an expected range based es past reviews and staff t
consultant (SNL) calevistfons. The results are acceptabfe. In addition staff
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(BNL) calculations of the CitDA (e.g., see Reftronce 10) with a given red morth erd estag appropriate themal.hdraulle feedback (en11ts the SE analysts aAich.
i fs. very conservatively, adiabatic) results in such fo.or peak enthalpfes.
e.g.. well under 150 cal /p for a 1.51 Jt rod and wil unde'r 200 cal /p for e-2.05,d k rod even when using conservative assumptions on the intifal thermal.
{
hdravitc stats.
Thus it is concluded that for present populations of reloads the expected relevant' rod worths and resulting peak enthapfes are such that it is reason-i
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able not to include the CitDA analysis in reload sulaf ttals. Furthermore, j
based on the review of.information from answers to questions froithe two l
preyfous statistical analyses about the characteristics of fkture i
populations as w11 as the significant margin to Itaf ts indicated by. the BNL' cateulations, there is no apparent need at this time to further exaafne the' j
future population charactaristics. - The elfsinatfon of cycle speciff,e citDA j
analyses fras 8PWS plants is acceptable.
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F v. e s.
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m REFCREWl}
- 3. Letter. L E. Deel. St. to 5. 5. Vassalle. RC. February 24,1982
{
- Control Rod Drop Accident".
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- 2. Letter L E. tapel. St. to 7. A. Ippof f to.' MC. November-14.1980 i
- mange in teneral Electric Methods fbr Analysis of Risfecated Sundle Accident *.
1 y
- 3. Letter. R. E. Enge1. St. to D. B. Vassallo. RC. May 18,1981
- Chany In teneral Electric Methods for Anatysis of Control ReIW1thd Error".
- 4. Namorande. W. 7. Johnsten to T. A. Ippelf te. April 14,1931.
'*many in General Electric Analysis of Mistocated Sundle Accide j
Letter. L. 5. Rubeastein, MC to R. E. Enge1. Sr. May ts.1981,.
5.
' Chang in General Electric Analyses of Nistocated Bundle A ccident*.
- 6. Memoranda, L. S. h6enstein to T. M. Nevak. November 33. 1981.
' Evaluation of the Statistical Analysis of the RNE far Natch 18 Letter. L. S. bbenstein MC. to.R. E. Engel. St. November i
7
- 25. 1981, i
'Chany in General tiectrfc Analysfs of Red Wfthdrawal Error *,
8.
Letter, t. E. Engel. St. to D. 8. Vassalle. NRC. March E3.1982
' Chang in General Electric Methods for Analysis of Nfslocated Sundle Accident".
9.
Letter. 8. E. Engel. W. to L. S. Rubenstein. MC. April 5.1982
- Responses to Questions Rega.rding Change in General Electric Rathods' for Analysis of. Control Rod Withdraust greer".
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- 10. Menorands. N. J, Richings to E. hfel. April 35.1980. *1he 98 RDA es Viewd Through SNL Feedback'.
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