ML20023C787
| ML20023C787 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 05/04/1983 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20023C785 | List: |
| References | |
| NUDOCS 8305180011 | |
| Download: ML20023C787 (19) | |
Text
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NUCLEAR REGULATORY COMMISSION
,j WASHINGTON, D. C. 20555 k...,.,/
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR RFACTOR REGULATION SUPPORTING AMENDMENT NO. 92 TO FACILITY LICENSE NO. OPR-56 PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY PEACH BOTTOM ATOMIC POWER STATION, UNIT 3 DOCKET NO. 50-278 1.0 Introduction The Philadelphia Electric Company (PECo or the licensee) requested (Ref.1) that the Technical Specifications (TSs) appended to Facility Operating Li, cense DPR-56 for Peach Bottom Atomic Power Station Unit 3 be amended to acconinodate the fifth refueling of the reactor.
Speci-fically, the requested TS changes were intended to accomplish the l
following:
l 1.
Identify the operating limits for all fuel types for Cycle 6 operations.
2.
Permit continued operation of a Pressurized Test Assembly (PTA) after reconstitution.
3.
Pennit operation with four new Lead Test Assemblies (LTAs).
I 4.
Incorporate Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for the new LTAs and extended exposure MAPLHGR limits for the PTA.
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- 5.. Permit operation with up to six General Electric (GE) hafnium Hybrid I Control Rods (HICRs).
6.
Modify bases to delete reference to a specific shutdown margin value provided by the Standby Liquid Control System.
An analysis of the safety considirations involved in the reactor refuel-ing and the Cycle 6 operating limits for all fuel types is set forth in Reference 2, which was filed along with other documents (Refs. 3,4) in December 1982. Other information (Refs. 5-10) relevant to the Cycle 6 reload had been provided earlier.
2.0 Fuel System Design
2.1 Background
The Peach Bottom 3 Cycle 6 core will contain 764 fuel assemblies of which 284 will be fresh reload 5 assemblies; The core composition is summa-rized in Table I.
Detailed descriptions of the four LTAs, the one PTA, 1
.and the 759 standard fuel assemblies are provided in References 6, 7, and 8, respectively. Since the standard fuel assemblies are comprised of a reviewed and approved design, this safety evaluation mainly addresses the -
four LTAs and the PTA, along with.six'GE HICRs described in Reference 3.
The fuel system design aspects of the six TS change objectives listed in Section 1.0 of this-safety evaluation are addressed in the following subsections.
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TABLE I PEACH BOTTOM UNIT 3 CYCLE 6 FUEL BUNDLES Fuel Type Cycle Loaded Number Irradiated P8DRB284H 4
263 P8DRB/99 5
216 NA 2
1 New P80RB284H 6
56 P8DRB299 -
6 224 PBLTA 1 6
,2 PBLTA 2 ~
6 2
Total 764 2.2 Operating Li'mits for Cycle 6 Fuel Types Information related to fuel system operating limits is contained in ~
Reference 2 and the re' lated TS changes were submitted with Reference 1.
4 Reference 2 contains analytical results of the safety considerations involved in the reactor refueling and Cycle 6 operating limits.
- Thus, core-wide critical power ratio changes (ACPRs) for several transients, including load rejection without bypass, loss of feedwater heating, feedwater controller failure, and rod withdrawal errors, are provided for the PTA and LTAs as functions of various input parameter assumptions.
i Minimum critical power ratios -(MCPRs) are listed in that ' report along with maximum linear heat generation rates (LHGRs) for the rod withdrawal error and misoriented bundle events. As discussed in Section 4.0, Thermal l
and Hydraulic Design, of this safety evaluation, the proposed operating, i
limits and corresponding TSs were reviewed and found acceptable.
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2.3 Operation with. Reconstituted Pressurized Test Assembly A PTA, described in Reference 7, was originilly inserted in the Peach Bottom 3 reactor during the Spring 1977, Reload 1 refueling.
The purpose
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of continued PTA operation is to obtain -fissicn gas measurements from the PTA in conjunction with an extended exposure program.
Twenty-two fuel rods will be removed from the PTA and replaced with irradiated rods from an 80RB283 bundle (initially inserted as part of Reload 2), which is due to be discharged at end-of-cycle (E0C)-5 (Ref.17). The estimated PTA bundle average exposure at the Spring 1983 outage is approximately 30 GWd/MT.
An additional cycle of operation would extend the fuel bundle average exposure to approximately 35 GWd/MT with a peak pellet exposure approaching 46 GWd/MT.
An analysis'of the safety considerations involved in continuing the use of
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the PTA is set fotth in Appendix C of Reference 2.
As indicated therein, reconstitution of the PTA could res' ult in a slight increase in peak cladding temperature (PCT) due to stored energy and local power distribution effects.
The resulting increase in PCT on account of these effects is insignificant
( 10 to 20*F) (Ref. 2) compared tp the margin to the PCT loss-of-coolant-accident (LOCA) limit of 2200*F (the magnitude of the maximum PCT of the
' non-reconstituted PTA is 1923*F).
Since the enrichment of the replacement i
rods was selected to assure that the reactivity of the reconstituted PTA will not exceed that of the non-reconstituted PTA, since the peak linear heat generation rate of the reconstituted'PTA is well within the operating limit of 13.4 kW/ft, and since the LOCA limits are not exceeded, we conclude that continued operation of the PTA during Cycle 6 is acceptable.
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2.4 Lead Test Assembly Operation The TS changes reguested -by PEco in Reference 1 would permit operation 4
with four LTAs of fuel type P80QB326 (for fuel description, see GESTAR-II, Ref. 8). Two of the LTAs will utilize an improved pressure drop spacer (low AP spacer), while the other two LTAs will have the normal spacer provided for 8x8R fuel. The LIAs will also incorporate several other 7
. features similar to those submitted.for Browns Ferry 3 in the Fall of 1981.
Analyses of the safety considerations involved with the LTA program are provided in Refs. I and 6.
The proposed TS changes-incorporate MAPLHGR limits for the four new LTAs and extended exposure MAPLHGR limits for the one PTA.
Our review of those subjects is described in the following.
subsection (2.5) of this safety evaluation.
With regard to LTA unique inputs and analyses (described in Attachment 2 of Reference 6 and in Reference 20), both core-wide and localized transients and accidents were considered. The LTAs were stated (Ref. 6) to have been analyzed using GESTAR-II (Ref. 8) methods and to have met all applicable GESTAR-II approved criteria.
Except for the rotated bundle event, the calculated MCPRs did not violate the safety limit MCPR.
For that event, 8
however, special-loading surveillance should mitigate against the possibility of a misorienced bundle.
Since the number of LTAs (4) is small, since they have been designed and analyzed using approved methods, and since, except for the rotated bundle event, no design or operating limits will be exceeded, we conclude that there is reasonable assurance that the insertion and operation of the four LTAs will not pose an unacceptable risk to the public health and safety.
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We' expect to be 1; formed in a timely manner concerning the results of the measurements to be conducted on the LTAs. As indicated in Reference 6, those measurements, as currently envisioned, are to consist of overall bundle visual examinations, bundle and rod length measu'rements, rod integrity and profilometry measurements, corrosion thickness measurements, fission gas sampling, spacer spring relaxation and possibly garuna scans.
It should be noted that, while PECc has stated (Ref. 6) that GE will summarize-the results from the LTA program in GE's fuel experience reports "in a timely manner," those reports have had about a five-year periodicity.
Thus, in the interest of timeliness, we will. expect PECo to provide an, informal summary of the LTA examinations within six months following each refueling outage during their lifetime in reactors.
2.5 MAPLHGR Limits Analyses of the safety considerations involved in the proposed MAPLHGR limits for the four LTAs and. extended expos' re limits for the one PTA are u
provided in References 4 and 5.
Although the methodology used is generally applicable for these limits, we believe that the effects of enhanced fission gas release in high burnup fuel (above 20 mwd /kgU) were not adequately considered in the generic analysis.
In response to this. concern, GE requested (Refs. 9 and 10) that credit for approved, but unapplied,-
emergency core cooling system (ECCS) evaluation model changes, and calculated PCT margin, be used to avoid MAPLHGR penalties at higher burnup.
This proposal was found acceptable (Ref. 11) provided that certain plant-specific conditions' were met.
PEco has stated (Ref. 12) that the GE proposal.is applicable to both Peach Bottom Units 2 and 3.
On the basis of this finding, we conclude that the MAPLHGR limits proposed for Peach Bottom 3 Cycle 6 are acceptable..
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2.6. Operation with Hafnium (GE Hybrid I) Control Rods The TS changes requested by PECo would permit operation with up to six Type II (surveillance version) G6-HICRs. The HICR Type II test program is designed to provide pra-comercial test data for GE's new Type I (productionversion)HICR. An analysis of the. safety considerations involved in the HICR Type II test program is set forth in Reference 3 (NEDE-22290), which is a generic report describing the design and the analyses performed by GE to demonstrate the safety of both the Type I and Type II hafnium-hybrid control rods.
The principal objectives of the HICR are to (1) increase control rod assembly life and (2) eliminate cracking of absorber tubes containing baron carbide (B C). The major design changes that are intended to ensure 4
that those objectives are met are (1) the use of'an improved B C absorber 4
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rod tube material to eliminate stress corrosion cracking during the lifetime of the assembly and ('2) replacement of some B C absorber rods with solid 4
hafnium absorber rods.
In addition, there are other material and dimension-al changes, including a reduction in sheath wall' thickness and a change in f
the pin and roller materials from Stellite to other materials discussed in l
l Reference 13.
Other variables included the location of the hafnium rods, the type of tubing used for B C rods and the use of clad versus unclad Hf.
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4 Due to the cc.nplexity of the HICR test program (as evidenced by the large number of variables to be examined), a meeting (Ref. 4) was held with GE and PECo to discuss the program in Peach Bottom 3 as well as the overall R&D program, analyses, surveillance, etc. performed or underway by GE in support of the HICR design. The purpose of the meeting was actually twc,-fold: - -..
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1.
To. support the proposed amendment to the Peach Bottom 3 operating license to permit HICR use.
2.
To support the generic use of HICRs in BWRs.
Because the generic review is much broader in scope than could be accommodated by the tight schedule required for Peach Bottom 3 Reload 5, this safety. evaluation addresses only the issue: involving the six surveillance HICRs. The results of the generic review will be reported separately as a safety evaluation of the GE topical report, NEDE-22290 (Ref. 3).
With regard to the Peach Bottom 3 Cycle 6 use of the six Type II HICRs, the key issues co'ncerned the potential effects of the changes in component materials and dimensions. The safety considerations involved are discussed below for each desi~n change.
g Pins and Rollers - As indicated in EPRI NP-2329 (Ref. 13), the pin and roller materials currently in use in BWRs are cobalt-base alloys' (Haynes 25 andStellite3,respectively). Because cobalt-60 is an isotope that contributes significantly to plant radiation build'p, there is an incentive u
l to replace the cobalt alloys with non-cobalt alloys and thus reduce personnel radiation exposure during plant maintenance.
EPRI NR-2329 describes an extensive program at GE to qualify substitute non-cobalt alloy control rod pin and roller materials. Wear resistance measurements in a simulated BWR environment (excluding irradiation), coupled with impact strength and corrosion tests, indicate that the non-cobalt alloys have equivalent or better wear resistance, superior impact strength and similar corrosion resistance to the conventional cobalt alloys. Though the effects of
irradiation were not investigated in those tests, reactor tests have been initiated at a control' cell BWR and at a conventional core BWR. We conclude that the substitution of the non-cobalt alloys for Haynes'25 and Stellite 3 pins and rollers in the six Type Il surveillance HICRs is acceptable, based on the results of the tests described in EPRI NP-2329 and our expectation that (a) the surveillance described on page S-5 of EPRI NP-2329 will be carried out, (b) the results of that surveillance udll be reported in a timely fashion, and (c) surveillance of the six HICRs in Peach Bottom 3 i
will also be conducted and reported.
Control Rod Tubing fiaterial - As indicated on page 2-2 of NEDE-22290 (Ref. 3),
the B C absorber rod tubing for the Type I (production version) control rods 4
is a high purity Type 304 stainless steel, while the Type II (development) control rods will also contain some high purity Inconel 600 as an alternate absorber tube, material. Both of these alloys have undergone extensive qualification testing and evaluation including laboratory testing, correlation of field performance with intergrannular stress corrosion cracking suscept-ibility test,s, and assessment of archival materials.
In addition, an extensive surveilTance program, including visual examinations, dimensional measurements, eddy current testing, neutron radiography, isotopic deter-minations, and steam. corrosion testing (see p. 5-10 of Ref. 3) is planned.
Based upon the information provided in Ref. 3 and in the meeting described in Ref. 9, we conclude that the use of the new absorber tube alloys is acceptable for the six Type II HICRs.
We expect to be infomed of the results from the-HICR surveillance program on the absorber tube materials as those results relate to the potential performance of the production version HICRs.
Absorber Material - As indicated in NEDE-22290 (Ref. 3), three of the B C 4
absorher rods per blade (12 in each control rod assembly) in the present BWR 2-4 0 lattice CRA design will be replaced with solid Hafnium rods.
In the Type I production ve'sion HICRs, the Hf rods are unclad and located r
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at the tip positions of each blade. The three main concerns related to the use of Hf rods involve (a) the increase in weight,'(b) the thermal expansion of Hf relative to the absorber cladding material, and (c) the corrosion res'istance of unclad Hf.
With regard to the increased weight resulting from the higher density of Hf relative to the B C it replaces, the reduction in blade sheath thickness 4
(and weight) compensates for the increase in absorber material weight. The resultant sheath thickness falls within the range of GE design experience, and the increased fuel channel clearance should reduce potential fuel channel interference.
From a mechanical design standpoint, therefore, there is reasonable assurance that the design changes rel&ted to the increased weight of the absorber material have been adequately accounted for in the six Type II HICRs. The plann6d surveillance of the HICRs should provide confirmation of this.
With regard to the thermal expansion and irradiation growth considerations, the coefficient of thermal expansion of Hf is approximately half that of Type 304 stainless steel and Inconel 600 (the B C absorber tubing materials),
4 and is comparable to an alternate cladding material used for some of the Hf rods in the Type II HICRs.
Inasmuch as only a few Type II rods will have the alternete cladding material, any adverse effects, which are not anticipated, should not be significant. The irradiation growth of hafnium is expected to be small'.- Bare hafnium absorber rods in the Peach Bottom 2~ reactor have shown virtually no change in length or diameter after 18 months service. Since i l
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dimensional measurements will be made of'the Hf rods at 18-24 month intervals as part of Type II HICR surveillance program, the irradiation 4
growth will readily be monitored.
With regard to the corrosion of hafnium in a BWR environment, there is significantly more information regarding.PWR use of hafnium (because of naval reactor use). GE did present some data (Refs. 3 and 4), however.
Those data showed that the corrosion behavior of hafnium in high tempera-ture water and steam is superior to that of Zircaloy-2.
In addition, an experimental, bare Hf control rod in Peach Bottom 2 has shown little corrosion after 1.5 years exposure (Refs. 14 and 15). The planned Type II HICR surveillance program is intended to include metallographic examinations of the Hf rod hydriding behavior and corrosion characteristics.
We conclude, therefore, that the corrosion behavior of the Type II HICR Hf rods has been ade'quately addressed for Peach Bottom 3 Cycle 6 operation.
2.7 Fuel System Design Conclusions We have reviewed the information submitted on the Cycle 6 operation of Peach Bottom 3, including the des,1gn, analysis, testing, and proposed surveillance of a PTA, four LTAs, and six Type II HICRs.
We find the Peach Bottom 3, reload 5 proposed refueling and related TS changes acceptable from a mechanical design standpoint.
3.0 Nuclear Design The nuclear design of the proposed reload was performed by the approved methods of Reference 8 including that of the LTAs.
The nuclear parameters for the reload are within the range of'those normally seen for BWR reloads and are acceptable.
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l 4.0 Thermal and Hydraulic Design The objective of the thermal-hydraulic review is to confirm that (a) the thermal-hydraulic design of the core has been accomplished using acceptable methods,-(b) the design provides an acceptable margin of safety from conditions which could lead to fuel damage during normal and anticipated operational transients, and (c) the design is not susceptible to thermal-hydraulic instability.
'f The thermal-hydraulic review includes the following areas:
(1) safety limit MCPR, (2) operating limit MCPR, (3)-thermal-hydraulic stability, and (4)-
changes to Tables 3.5.K.2 and 3.5.K.3 and Figures 3.5.K.1 and 3.5.K.2 of the TSs.
The licensee has submitted the analysis report for Cycle 6 operation at rated core flow conditions (Ref. 2).
Discussion of our review concerning the thermal-hydraulic design for Cycle 6 operation follows:
4.1 Safety Limit MCPR The safety limit 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 8, the safety limit MCPR is 1.07.
The safety limit MCPR of 1.07 is used for Peach Bottom 3 Cycle 6 operation.
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4.2 Operating Limit MPCR The' most limiting events have been analyzed hy the licensee to determine which event could potentially induce the largest reduction in the initial critical power ratio (ACPR). The ACPR values given in Section 9 of Reference 2 are plant specific values calculated by including the ODYN Methods. The calculated ACPRs.are adjusted to reflect either Option A or Option.B ACPRs by employing the conversion methods described in Reference
Section 11 of Reference 2 presents both the cycle MCPR values for the pressurization and non-pressurization transients. The.
maximum cycle MCPR values (Options A and B) in Section 11 are specified as the operating limit MCPRs and incorporated into the TSs. Since the approved method was'.used to determine the operating Hmit MCPRs to avoid violation of the safety lir.it MCPR'in the event of any anticipated transients, we conclude that.these limits are acceptable.
4.3 Thermal-Hydraulic Stability l
The results of thermal-hydraulic analysis (Ref. 2) show that maximum reactor l
l core stability decay ratio is about 0.98, which is comparable to the calculated value for Peach Bottom 2 Reload 3, which has been.previously approved. Since operation in the natural circulation mode is prohibited by TS 2.1.A.4, there will be added margin to the stability limit. The therefore conclude that the thermal-hydraulic stability results are acceptable for Cycle 6 operation.
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- 4.4 Changes to the Technical Specifications Figures 3.5.K.1, 3.5.K.2 and Tables 3.5.K.2, 3.5.K.3 of the TSs have been modified to ~ include the operating limit MCPRs for Cycle 6 operation.
Using Option A, the operating limit MCPRs would be 1.33 for Cycle 6. fuels at burnup conditions from BOC to 2000 MWD /t before EOC, and 1.39 for PTA, P8X8R fuel types and 1.40 for LTA at burnup conditions from 2000 MWD /t before EOC to EOC. Using Option B, the operating limit MCPRs.
would be 1.26 for Cycle 6 fuels at burnup conditions from BOC to 2000 MWD /t -
before EOC, and 1.27 for PTA, PSX8R fuel types and 1.28 for LTA at burnup conditions from 2000 MWD /t before EOC to EOC.
4.5 Thermal and Hydraulic Design Evaluation Summary We find that approved thermal-hydraulic methods have been used and that
-results of ana' lyse,s support.the proposed limit MCPRs, which avoid violation of the safety limit MCPR for design transients.
We conclude that this core. reload will not adversely affect the capability to operate Peach Bottom 3 safely during Cycle 6 operation and that the revised Figures 3.5.K.1, 3.5.K.2 and Tables 3.5.K.2, 3.5.K.3 of the TSs discussed above are acceptable.
5.0 Transients and Accidents As described in Section 2.4 above, the analyses of the transients and accidents have been performed with the approved methods of Reference 8, and with the exception of the fuel misorientation event, meet all acceptance criteria.
The fuel misorientation event is discussed below.
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4 The effect of the presence of the HICRs on the results of these events is expected to be negligible for the followin,g reasons:
1.
Only six of the HICRs are present, 2.
The nuclear characteristics of the hybrid rods resemble closely those for standard rods, and 3.
The scram speeds are identical.to the standard rods.
5.1 Fuel Assembly Misorientation for Lead Test Assembly
- When analyzed with standard procedures (NED0-24011-A-US, Section S.2.5.4.2),
the misorientation of one type of LTA (PBLTA1) can lead to a MCPR value of 1.06 when the core is operated at the proposed operating limit. The licensee states that the proposed operating limit MCPR need not be altered to accom-modate this event since special precautions will be taken to prevent it. We find this ppsition to be acceptable for the following reasons:
1.
There are only four LTAs - only two of which are a concern for this event.
2.
The licensee proposed to initiate special procedures for the LTAs during this cycle to prevent misorientation.
3.
The calculation of MCPR for this event tends to be conservative and the variation from the safety limit is small.
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6.0 Technical Specification Change PECo, wishes to delete the reference to a five percent shutdown margin in the I
bases to the TSs for the Standby Liquid Control System (Specification 3.4).
This would bring the specification into correspondence with that in the Standard Technical Specifications (for BWRs, Specification 3/4.1.5). The actual value of the shutdown margin is,provided for each cycle as part of.-
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the supplemental reload licensing submittal for the 660 parts per million of boron which is cited in the bases to the specification.
We find this change to be consistent with the approved reload licensing procedures of Reference 8 and therefore acceptable.
7.0 Sumary
.We conclude that the fuel system design, nuclear design, thermal-hydraulic design, transient accident analyses, and associated proposed TS changes for Peach Bottom 3 Cycle 6 operation are acceptable.
8.0 Environmental Consideration
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We have detemined that the amendment does not authorize a change.
in effluent types or total amounts nor an increase in power level
-and will no't result in any significant environmental impact.
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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 statenent, or negative declaration and env_ironmental impact appraisal need not be prepared in connection with the issuance of this amendment.
9.0 Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of an accident previously evaluatdd, l
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 16-
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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
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issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.
Dated: llay 4,1983 The following NRC personnel have contributed to this Safety Evaluation:
M. Tokar, W. Brooks and S. Sun.
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l 0.0 References 1.
E. J. Bradley (PEco), letter to H. R. Denton (NRC), December 30, 1982.
2.
R. A. Browning, " Supplemental Reload Licensing Submittal for Peach Bottom Atomic Power Station Unit 3 Reload 5," General Electric Report j
Y1003J01A54, Decenber 1982.
3.
A. N. Tschaeche, " Safety Evaluation of the General Electric Hybrid I Control Rod Assenbly " General Electric Company Proprietary Report NEDE-22290, December 1982.
, 4.
"Erratta and Addenda Sheet No. 5" to NEDO-24082, October 1982.
5.
" Loss-of-Coolant Accident Analyses for Peach Bottom Atomic Power Station Unit 3," General Electric Company Report NEDO-24082, December 1977.
6.
R. A. Blough (PECO), letter to J. F. Stolz (NRC),
Subject:
Peach Bottom Unit 3 Lead Test Assemblies, December 20, 1982.
7.
NED0-21363-4, Supplement 4. January 1977.
8.
" General Electric Standard Application forlReactor Fuel," GESTAR-II, General Electric Report NEDE-24011 (latest approved version).
9.
R. E. Engel (GE), letter to T. A. Ippolito (NRC), May 6,1981.
- 10. R. E. Engel (GE), letter to T. A. Ippolito (NRC), May 28, 1981.
- 11. L. S. Rubenstein (NRC), memorandum for T. M. Novak, " Extension of General Electric Emergency Core Cooling Systems Perfomance Limits',"
June 25, 1981.
12, S. L. Daltroff (PECo), letter to J. F. Stolz (NRC), July 15, 1981.
N V.
- 13. P. Aldred, "BWR Control Rod Cobolt Alloy Replacement," Executive Summary, EPRI Report NP-2329-SY, March 1982.
- 14. A. N. Tschaeche (GE), letter to M. Tokar-(NRC) with GE proprietary document titled " Hybrid Control Rod Licensing Presentation to the Nuclear
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Regulatory Commissiou, March 23,1983," March 24,1983.
- 15. " Proposed Peach Bottom Atomic Power Station Unit 2 Alternate Absorber Control Blade Test Program " General Electric Company Report NEDO-24231, Rev.1. January 1980.
- 16. " Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors," NED0-24154P, October 1978.
- 17. J. W. Gallagher (PECo), letter to J. R. Stolz (HRC), April 6, 1983.
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