ML20028A288
| ML20028A288 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/12/1982 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20028A285 | List: |
| References | |
| NUDOCS 8211180056 | |
| Download: ML20028A288 (13) | |
Text
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UNITED STATES 8"
NUCLEAR REGULATORY COMMISSION
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING Af1ENDMENT fl0. 87 TO PROVISIONAL OPERATING LICENSE NO. DPR-21 NORTHEAST NUCLEAR ENERGY COMPANY MILLSTONE NUCLEAR POWER STATION, UNIT 1 DOCKET NO. 50-245
1.0 INTRODUCTION
AND DISCUSSION By letter dated October 15, 1982 (Ref. 1) Northeast Nuclear Energy Company (llflECO) proposed Technical Specification changes to allow plant operation at rated conditions following completion of actions planned for the outage and refueling (Reload 8) which began on September 5, 1982. The core (Reload 8) for fuel cycle 9 operation includes 192 new (unirradiated) 8X8 prepressurized retrofit fuel assemblies fabricated by General Electric Company. Seventy-two of the fuel assemblies contain 2.82 w/o 235U with 7 Gadolinia rods at 3 percent, identical to 128 new fuel assemblies used first~during the Reload 7 outage in the fall of 1980. The remaining 120 fuel. assemblies contain 2.83 w/o 235U with 7 Gadolinia rods at 4 percent.
g The,Gcreased number of new replacement fuel assemblies and the heavier 235U and Gd loadings are necessary to achieve the extended operating cycles desired by the licensee. The core for fuel cycle 9 has been calculated to produce 8022 MWD /ST (in contrast to 6580 for cycle 8). This burnup is equivalent to 18 months operation at rated power.
Additional information related to Reload 8/ Cycle 9 cperation and responses to NRC concerns was submitted by NNEC0 letter dated November 2, 1982.
This supplementary information (Ref. 9) was included in the staff evalu-ation of the proposed changes to the Operating Technical Specifications for the Millstone fluclear Power Station, Unit 1.
By separate letter, also dated October 15, 1982, NHEC0 submitted an Appendix (Ref, 11) that describes changes made to the Segmented Test Rod fuel assembly during Reload 8 outage.
2.0 EVALUATION OF PROPOSED TECHNICAL SPECIFICATION CHANGES Each of the items identified below (Ref. 1 Attachment 5) concerns changes beyond those described in General Electric Company proprietary topical report NEDE-240ll-P dealing with GE fuel which has been reviewed and approved by the NRC (Ref. 7).
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2.1 Reload 8/ Cycle 9 and Extended Load Line Limits The safety limit minim'm critical power ratio (MCPR) has been imposed to assure that 99.9 percent of the fuel rods in the core are not expected to experience boiling transition during normal and anticipated operational transients. As stated in Reference 6, the safety limit MCPR is 1.07.
2.1.1 Operating Limit MCPR Various transients could reduce the MCPR below the intended safety limit MCPR during Cycle 9 operation.
The most limiting events have been analyzed by the applicant to determine which event could potentially induce the largest reduction in the initial critical power ratio (ACPR). The dCPR values given in Section 9 (Ref. 2) are plant specific values calculated by using the ODYN methods. The calculated ACPRs were adjusted to reflect either Option A or Option B ACPRs by employing the conversion method described in Reference 8.
The cycle MCPR values are determined by adding the ACPRs to the safety limit MCPR. Section 11 (Ref. 2) presents both the cycle MCPR values for the non-pressurization and pressurization events. The maximum cycle MCPR values (Option A and B) in Section 11 are specified as the operating limit MCPRs and are incorporated into the Technical Specifications.
The maximum value of operation limit MCPR (OLMCPR) resulting from the limiting transient, the generator load rejection without bypass transients, is 1.48 for Cycle 9 as compared to.l.39 for Cycle 8.
The large difference of OLMCPR for this transient is due to the use of the ODYN methods compared to the REDY methods used for Cycle 8.
Since the higher OLMCPRs obtained from the analyses are more restrictive for the Cycle 9 operation and since these limits will avoid violation of the safety limit MCPR in the event of any anticipated operational transient, the staff finds these limits to be acceptable.
The licensee has also submitted the revised power / flow map as shown in Figure 1 of Reference 4.
The proposed power / flow map is to allow power ascension along the 108% APRM rod block line to 100% power at 87% flow and to allow rated power operation at any flow between 87%
and 100%. An analysis of this extended load line limit 9
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M was performed for Cycle 8 in Reference 3 and. reverified for Cycle 9 in Reference 4.
Based on our review, we find in Reference 4 that (1) the calculated decay ratio for the reactor core stability is consistent with the result as included in Reference 3, which presents a decay ratio of 0.61 as compared to 0.6 for Cycle 9 reload operation (Ref. 2), and (2) the ODYN results from the generator load rejection without bypass event (most limiting transient) indicate that the power / flow s map at 100/100 point (licensing basis) is the most limiting point. The staff, therefore, concludes that the results of an extended load line limit analysis in Reference 4 are bounded by the licensing basis results of the supplemental reload submittal for Cycle 9 f
operation as included in Reference 2.
The staff has reviewed the OLMCPRs results (Ref. 2) and the results of the extended load limit line analysis discussed above and finds that the ODYN methods that were used and the results have shown an acceptable margin of safety from conditions which could lead to fuel damage during any anticipated operational transient.
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Section 2.1.2 of these Technical Specifications are being modified to include the extended operating power /
flow maps given in Figure 3.3.1 of the proposed Technical Specifications.
Section 3.11.C of the Technical Specifications are
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being modified to include the operating limit MCPR for Cycle 9 operation. Using Option A, the operating limit MCPRs shall be 1.45 for 8X8 and 8X8R fuel types, and 1.48 for P8X8R fuel. Using Option B', the operating limit MCPRs shall be 1.40 for 8X8 and 8X8R fuel types, and 1.43 for P8X8R fuel.
2.1.2 Thermal-Hydraulic Stability The results of the thermal-hydraulic analysis (Ref. 2) show that the maximum reactor core stability decay ratio is 0.6 for Cycle 9 as compared to 0.61 for Cycle 8.
Based on the evaluation results that (1) the calculated decay ratio for Cycle 9 is less than that for Cycle 8, and (2) the. decay ratio. compares favorably to the calculated value for several operating reactors which have been previousiy approved, the staff concludes that the thermal-hydraulic stability results are acceptable for Cycle 9.
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4-2.1.3 MAPLHGR Limits The licensee has submitted new and revised MAPLHGR limits for all Cycle 9 fuel types including the Segmented Test Rod Assembly (See Section 3.1). These limits were generated by methods (Ref. 9) submitted as part of this application.
Although the methodology used in generally applicable for these limits, the staff concluded that the effects of enhanced fission gas release in high burnup fuel (above 20 GWd/MtU) a
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were not adequately considered in the generic analysis.
In response to this concern, the General Electric Company requested (Refs. 13-14) that credit for. approved, but unapplied, emergency core cooling systems (ECCS) evaluation model 1
changes, and calculated peak cladding temperature margin, be used to avoid MAPLHGR penalities at" higher burnup. This proposal was found acceptable (Ref.15) provided that certain plant-t specific conditions were met.
In a letter dated November 2 1982 (Ref. 9) the licensee stated that tl.e General Electric ~
proposal was applicable to Millstone Unit 1.
On the basis of this finding, the staff concluded that the MAPLHGR limits proposed for Millstone Unit 1 Cycle 9 are acceptable and the Technical Specifications, Figure 3.11.1, as proposed are hereby approved.
e,y 2.1.4 Extended Load Line Analysis 4
The previous evaluation (for Cycle 8) applied to.th'e 100%
power /100% flow condition. The licensee has provided an analysis (Ref. 3 and 4) to justify the expansion of the operating region of the power / flow map for Millstone Unit l' to allow rated power operation at any flow between 87% and 100%. The approach used was to show the 100% power /100%
flow operating point is a more limiting condition than other i
conditions within the expanded operating region (See proposed -
Technical Specification Figure 3.3.7) for those transients and accidents sensitive to variations in power and flow.
Reference 3 provides the transient result comparisons between high and los flow initial conditions for various plant characterists. Peak vessel pressure and.ACPR are the significant parameters to be compared in evaluating the acceptability of the revised power / flow map. The analysis showed that m ak pressure is reduced slightly for low flow calculations of the MSIV closure transient.
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The oCPR shows a decreasing trend with decreasing flow for both the Load Rejection without Bypass and Feedwater Controller Failure events. Both trends are an indication i
that the 100% power /100% flow condition is the most limiting.
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s Refere1ces 6 and 7 require that the operational limit MCPR be increased for low flow conditions. This is done by the use of a flow factor, Kf, which is a multiplier applied to the MCPR operating limit at reduced core flow. The flow factor for Millstone 1 is identified in Figure 3.11.2 of the proposed Technical Specifications. The extended load limit line is applicable in conjunction with the flow factor multiplier so that flow excursions under manual and automatic flow control are determined not to exceed MCPR safety limits.
Based on our review of the results provided in Reference 3 and the Cycle 9 revision (Ref. 4), we conclude that the power / flow map proposed in the Technical Specifi-cation change is acceptable. The resulting operating limit MCPRs for ODYN Operations A and B are specified in Table 3.11.1 of the Technical Specifications and are acceptable.
2.1.5 Transient Analysis lhe licensee has reviewed those transients that are the basis for Cycle 9 for the Millstone Unit i license and has reanalyzed those transients that are critical with respect to safety margins and sensitive to the core reload e c.-
parameter changes. As noted previously, the ODYN code was used in the determination of Critical Power Ratios (CPRs) for the rapid pressurization transients. The REDY code was used for the rapid non-pressurization events. The most restrictive condition was calculated to occur as a result of a postulated Generator Load Rejection Without Bypass.
For this event, a limiting MCPR of 1.08 was predicted which is still above the MCPR safety limit of 1.07.
This is acceptable to the staff. The revised Technical Specifications contain dua.1 operating MCPR limits resulting from the use of the ODYN code. Option B requires implementation of a control rod scram timing program that specifies surveillance on greater than 15 rods every 120 days. This is consistent with the requirements identified in References 6 and 7, and implemented on recent reloads and is acceptable to the staff on this basis.
Reactor vessel overpressure protection was verified by an ODYN analysis of the closure of all main steam isolation valves (MSIV) with an indirect (flux). trip. MSIV closure is the limiting event with regard to vessel overpressurization.
At the end of Cycle 9, with all safety valves operating and an indirect scram, the peak vessel pressure was predicted
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J to be 1285 psig which is below the peak allowable ASME overpressure of 1375 psig.at the vessel bottom. This is acceptable. Based on previous sensitivity studies (Cycle 8), the staff concludes that. sufficient margin exists assuming the failure of one valve to open.
The Loss-of-Coolant Accident (LOCA) analysis was updated by the licensee to address the new fuel (Ref. 1, ).
In the staff safety evaluation of References 6 and 7 it was concluded that the continued application of the present GE ECCS-LOCA models to the 8X8 reload fuel is generically acceptable. Since the input parameters to the LOCA analysis are unchanged, the proposed MAPLHGR limits for the new prepressurized reload fuel are acceptable.
And finally, the licensee has agreed to modify the 1
Technical Specifications by March 15, 1983 to include limits appropriate to the turbine bypass system inoperable condition because of the possbility of exceeding the MCPR limit following a feedwater transient l
without turbine bypass capability. Because of the low probability of occurrence of such an event and the demonstrated high reliability of the Millstone Unit l-turbine bypass system, the staff has concluded that this proposal is acceptable.
2.2 Protective Instrumentation The proposed Technical Specification changes identify a change in the turbine low pressure setpoint to initiate MSIV closure.
The value of the setpoint was lowered from 880.psig to 825 psig.
1 This trip provides protection against fast reactor depressurization and the resulting rapid cooldown. The staff agrees with the licensee's conclusion that the vessel cooldown rate would not be affected by the setpoint change and is acceptable. The changes to the Technical Specifications as proposed by the licensee are acceptable.
2.3 Instrumentation That Initiates Rod Block l
There are six APRM inputs, designated 1, 2, 3, 4, 5 and 6, to the reactor protective system (RPS). These same inputs are also used for the APRM control rod withdrawal blocks (i.e., APRM Upscale and Downscale). For the RPS trip there are two channels, Channel A which uses APRM inputs 1, 2 and 3 and Channel B which uses APRM inputs 4, 5 and 6.
Since this system has a one out of two taken twice I
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Therefore, when an APRM is, bypassed for the RPS function, it is also bypassed for the rod withdrawal function.
The APRM rod withdrawal block function is also divided into two channels or strings of three-inputs each. However, any one of the six APRM inputs will result in a rod withdrawal block. The problem stems from the fact that the APRM inputs into the rod withdrawal-block strings are not the same as those used for the RPS function (i.e., one rod withdrawal _ block string uses 1, 3 and 5 APRM inputs and the other uses 2, 4 and 6 APRM inputs). The Millstone-1 T/S for rod withdrawal block states the minimum number of operable instrument channels / trip system (where a trip system has been interpreted as a string) is two.
If the right combination of APRM's is bypassed for the RPS function (e.g., inputs 1 and 5) this could result in only one APRM input into one of the rod withdrawal block strings which would be a violation of the T/S if interpreted as stated.
Based on discussions with the staff and licensee and the licensee's
' discussions with GE, it has been mutually agreed that this was not
2.4 Special Reports on Containment Leak Rate Tests The Standard Technical Specifications do not require a special report regarding the secondary containment leak rate tests and there are no unique circumstances to justify such a report. The staff agrees, therefore, that the Technical Specification may be changed as proposed -
by the licensee to eliminate this requirement.
2.5 Containment Systems r
The staff has confirmed that the proposed Technical Specifications in Section 2.5 of the licensee October 15, 1982 submittal are typo-graphical corrections and acceptable.
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2.6 Primary Containment Isolation The licensee noted (Ref. 9) that the proposed changes are undergoing further in-house review and has temporarily withdrawn the request for Technical Specification change approval. The licensee has concluded and the staff agrees that changes are not required prior to start up of cycle 9.
2.7 Scram Discharge System The requested changes were addressed in Amendment No. 86.
2.8 Reactor Protection System The licensee noted (Ref. 9) that the proposed changes are being re-evaluated and has temporarily withdrawn the request for Technical Specification change approval. The licensee has concluded, and the staff agrees, that changes are not required prior to start-up of cycle 9.
2.9 Fire Detection Instrumentation The staff approves the proposed Technical Specification changes which
- ": allow reduced frequency of heat detection testsin the condenser bay (a situation unique to Millstone-1) and correct the number of fire detection instruments. The reduced frequency of testing pneumatic rate-of-rise heat detectors is justified because:
o the air type heat detector is rugged and highly reliable, o
access to the heat detector sensors in limited due to radiation levels and the relatively inaccessible location, o piping leaks or control panel malfunction are annuniciated in the control room, and o of the total number available (36) vs the minimum number required (25) 1.e., the number o# sensors installed is more th'an 40% in excess of minimum requirements.
3.0 OTHER ISSUES 3.1 Segmented Test Rod Assembly In addition to the 71 standard 8X8, 148. standard 8X8R and 360 pre-pressurized 8X8R fuel assemblies, a previously-irradiated Segmented Test Rod (STR) Assembly will be inserted in the Cycle 9 core. The assembly is being used to test new fuel design concepts in an in-reactor environment. A description of the STR is contained in
Reference 10 and operation with this assembly was previously approved through Cycle 8.
The modifications made to this assembly during the Reload 8 outage are described.in Reference 11. The modifications, which consist of replacing several fuel rods, result in no significant changes in the thermal, mechanical or nuclear characteristics of the assembly. The licensee has stated that operation with this assembly does not affect the results of the safety analyses reported in Reference 10. The staff agrees with this conclusion.
3.2 Low Pressure Core Spray Sparger Crack Clamp Visual inspection of the core spary sparger using an underwater TV camera revealed a crack near the sparger-to-junction box "D" weld..
No indicationsof cracks were found anywhere else in the spray system.
The crack was not observed during the inspection of the sparger in 1980. The licensee attributes this to the pool television quality at that time. The views of the welds were not clear enough to see the details revealed during the 1982 outage.
During a meeting with the staff on October 29, 1982, the licensee ee described the crack at the sparger weld connection at junction box "D" and the clamp that was installed to limit the loss of spray water if the crack should progress further. Similar clamps
.have been reviewed and approved by the staff for other BWRs'where the indications of sparger cracks has been detected. The staff has concluded that the clamp design and installation clearances are adequate to limit the maximum opening of a 3600 crack and flow rate through the crack when emergency core spray core cooling is required.
By letter dated November 10, 1982 the licensee presented the results of analysis that show the capability to deliver core spray water through the spray header nozzles, with allowance for the maximum crack opening, remains in excess of design requirements. On this basis the staff conc'urs that the clamp provides adequate assurance of sufficient core spray water to meet emergency core cooling require-ments if the maximum amount of water is lost through a postulated crack.
3.3 Ultrasonic Inspection of the Reactor Coolant Recirculation Piping The licensee described, at a meeting with the staff on October 29, 1982, inspection of the coolant recirculation piping welds using ultrasonic techniques.
Because through wall cracks were detected in other BWR recirculation systems, i.e., visual observation of leaks, during a hydrostatic test after the recirculation system welds,had been inspected l
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100% using ultrasonic techniques, the validity of ultrasonic tests at Millstone-1 was re-examined. The Millstone-1 ultrasonic test inspectors travelled to the Battelle Laboratory in Ohio to demonstrate the sensitivity of ultrasonic tests for indications employed during the Millstone-1 in-service inspection of the recirculation piping. This was achieved by detecting indications in a pipe with known indications provided by Battelle.
It was concluded that the ultrasonic techniques used at the Millstone-1 plant were adequate to detect the Battelle calibrated sample indications and on this basis the ISI performed by the licensee was determined by the staff to be acceptable.
3.4 Isolation Condenser Steam Line Crack During an October 29, 1982 meeting with the staff, the licensee reviewed the results and corrective measures following the augumented in-service inspection of the isolation condenser steam supply piping outside the dry well. Selection of welds for inspection was based on flVREG-0313 Revision 1 guidelines. One weld (CAC-F-21), a reducer to pipe weld (type 304) circumferential linear indication in heat affected zone (on the pipe side), was found to contain a reportable linear indication. This indication exceeded the ASME XI
"" (IWB-3514-3) acceptance standards. The licensee believes the indication was due to intergranular stress corrosion cracking and completed a repair program to remove the unacceptable indication.
A replacement spool piece of SS type 304K (low carbon) was installed employing heat sink welding. The staff agrees that the replacement piece of pipe is more corrosion resistant and on that basis acceptable.
3.5 Accidental Drop of Underwater Welder into Reactor Vessel During the October 29, 1982 meeting with the staff in Bethesda, Maryland, the licensee described the circumstances and consequences of the accidental drop of the special underwater welder into the reactor. This occurred at a time when all the fuel had been temporarily removed from the reactor vessel to the spent fuel pool and replacement of jet pump support beams was in progress. The pre-outage plans included replacement of all jet pumps beam supports (in response to IE Bulletin 80-07). This work required the use of a rotary service platform at the top of the reactor vessel opening which rides a track at the vessel flange and a special underwater welder for use during the replacement of the 20 beam assemblies on jet pumps. On October 10, 1982 the platform e
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accidentally tipped into the top of the reactor vessel and the welding machine along with a number of other items that were on the platform fell into the vessel. A main steam line plug was struck by the platform allowing water to fill the steam line and flow out of the openings where S/R valves had been removed, into the dry well and torus.
The licensee reported that all items dropped into the reactor vessel were recovered, inspection revealed no damage to the vessel or internals and spilled water was cleaned up. The platform was restored to its working position, a replacement underwater welder was obtained and the work was completed without further incident. Long term modifications to the platform to prevent a repeat occurrence are being evaluated. On the basis of the information provided by the licensee,the staff concluded that there is no unreviewed safety consideration and the installation of 20 new beam assemblies was ~ completed and reconstitution of the core completed without further incident.
We, therefore, conclude that the proposed Technical Specifi-cations addressed in this evaluation are acceptable.
4.0_Ef{V.Ij0NMENTALCONSIDERATION We have determined that the amendment does not aut'iorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR 551.5(d)(4), that an environmental impact statement -
or negative declaration and environmental impact appraisal need not be prepared in connection with the issuance of this amendment.
5.0 CONCLUSION
We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a sigoificant increase in the probability or consequences of an accident previously evaluated, does not create the possibility of an accident of a type different from any evaluated previously, and does not involve a significant reduction in a margin of safety, the amendment does not involve a significant hazards consideration; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; and (3) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
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_CKNOWLEDGEMENTS A
6.0 The following staff members have contributed to this evaluation:
J. Shea J. Voglewede B. Sun W. Hodges
.l M. McCoy 2
Dated: November 12, 1982 5
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I REFERENCES 1.
Letter from W. G. Counsil (NU) to D. M. Crutchfield (NRC), Millstone Nuclear Power Station, Unit 1, Provisional Operating License No. DPR-21 Reload 8/ Cycle 9 License Amendment Submittal dated October 15, 1982.
2.
Y1003J01A44, Supplemental Reload Licensing Submittal for Millstone Unit 1 Reload 8, June 1982.
(NNEC0 submittal dated October 15, 1982 )
3.
NE00-24366, Extended Load Line Limit Analysis (Millstone Point Nuclear PowerStation, Unit 1), September 1981.
(NNEC0 submittal dated October 15, 1982 - Attachment 3) 4.
NE00-24366-1, Supplement 1 to Extended Load Line Limit Analysis Millstone Point Nuclear Power Station, Unit 1, (Reverification for Cycle 9), June 1982.
(NNEC0 submittal dated October 15, 1982 - )
5.
Proposed Technical Specification changes for Millstone Unit 1, Reload 8.
(NNEC0 submittal dated October 15, 1982 - Attachment 5) 6.
NED0-240ll-A-4, General Electric Boiling Water Reactor Generic m,3eload Fuel Applications, January 1982.
7.
Letter from D. G. Eisenhut (NRC) to R. Gridley (GE) dated May 12, 1978.
8.
Letter From R. Buchholz (GE) to P. Check (NRC), Response to NRC Request for Information on ODYN Computer Model, dated September 5,,
1980.
9.
Letter from NNECO to NRR " Additional Information Regarding the Reload 8/ Cycle 9 License Amendment submittal dated November 2, 1982.
10.
"STR Bundle Submittal, Millstone-i Segmented Test Rod Bundle,"
General Electric Company Rdport NZD0-20592, 1974 (with Supplements 1-5) 11.
"STR Bundle Submittal, Millstone 1 Segmented Test Rod Bundle,"
General Electric Company Report NE00-20592-6 (Supplement 6)
July 1982.
(License Submittal dated October 15,1982) 12.
W. G. Counsil (NU) letter to D. M. Crutchfield (NRC) dated May 20, 1981.
13.
R. E. Engle (GE) letter to T. A. Ippolito (NRC) dated May 6, 1981.
14.
R. E. Engle (GE) letter to T. A. Ippolito (NRC) dated May 28, 1981.
15.
L. S. Rubenstein (NRC) memorandum for T. M. Novak (NRC) on "Extention of General Electric Emergency Core Cooling Systems Performance Limits" dated June 25, 1981.
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