ML18026A327
| ML18026A327 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/20/1981 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| References | |
| NUDOCS 8103060507 | |
| Download: ML18026A327 (13) | |
Text
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~O Docket, Nos.:
50-387 and 50-388 DISTRIBUTION:
Docket Files jEB 20 tSSl LBPl Rdg DEisenhut JYoungblood RStark MRushbrook RTedesco
...,... SHanauer RVollmer TMurley Mr. Norman M. Curtis
..,,...,.gpss Vice President - Engineering,....,...,,.....,
.. RHartfield.,MPA and Construction
,,.....OELD Pennsylvania Power and Light. Company,., 0IE..(3)...
Two North Ninth Street Allentown, Pennsyl vania 13101 bcc:
TERA NRC/PDR L/PDR NSIC TIC ACRS (16)
Dear Mr. Curtis:
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Subject:
Susquehanna Steam Electric Station~/pits
.Nos.
1 and 2 - Request for Additional Information As a result of our review of your. application.for,.operating, licenses, for the, Susquehanna Steam Electric Plant,.we find that, we.aced additional information in the areas of Core Performance, Materials and.quality Assurance.
The enclosed questions were telecopied to. PPM. on February 6, 1981.,
Please provide your final responses by March 20,, 198$,,
If you desire any discussion or clarification of,the. information requested.,
please contact R. M. Stark, Project Manager, (301-492-7238)....
Sincerely, Orlgtnaf stgned by Robert L Tedesoo
Enclosures:
As stated cc w/encls.:
See next page Robert L. Tedesco, Assistant Director for Licensing Division of Licensing OFFICE P SURNAME)
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NRC FORNI 318110 801 NRCIJI 0240 OFFlCIAL RECORD COPY
~ VSGPO 1980-329 924
Mr. Norman W. Curtis Vice President. - Engineering and Construction Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsyl vani a 18101 CC:
Mr. Earle M. Mead Project Engineering Manager Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsyl vani a 18101 Jay Silberg, Esq.
Shaw, Pittman, Potts 5
Trowbridge 1800 M Street, N.
W.
Washington, D.
C.
20036 Mr. William E. Barberich, Nuclear Licensing Group Supervisor Pennsylvania Power
& Light Company
=
2 North Ninth Street Allentown, Pennsylvania 18101 Edward M. Nagel, Esquire General Counsel and Secretary Pennsylvania Power 8 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Bryan Snapp, Esq.
Pennsylvania Power 5 Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 Robert M. Gallo Resident Inspector P. 0.
Box 52 Shickshinny, Pennsylvania 18655 John L. Anderson Oak Ridge National Laboratory Union Carbide Corporation Bldg. 3500, P. 0.
Box X
Oak Ridge, Tennessee 37830 Gerald R. Schultz, Esq.
Susquehanna Environmental Advocates P. 0.
Box 1560 Wikes-Barre, Pennsylvania 18703" Mr. E.B. Poser Project Engineer....
Bechtel Power Corporation P. 0.
Box 3965 San Francisco, California 94119 Matias F. Travi eso-Di az,. Esq.
Shaw, Pittman, Potts 8
Trowbridge 1800 M Street, N.
W.
Washington, D. C.
20036 Dr. Judith H. Johnsrud Co-Director Environmental Coalition on Nuclear Power 433 Orlando Avenue State College, Pa 16801 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Department of Environmental Resources Coamonwealth of Pennsylvania P. 0.
Box 2063 Harrisburg, Pa 17120 Ms. Colleen Marsh Box 538A, RD¹4 Mountain Top, PA 18707 Thomas J
Halligan Correspondent The Citizens Against Nuclear Danagers P. 0.
Box 5
- Scranton, Pennsylvania 18501 Mr. J.W. Millard Project MAnager Mail Code 394 General Electric Company 175 Curtner Avenue San Jose, California 95125 Robert W. Adler Dept. of Environmental Resources 505 Executive House P. 0.
Box 2357 Harrisburg, Pennsylvania 15120
SDUSNANNA 1
Al<D 2 STAFF PVSITll Thermal-H draul ics Section, Core Performance Branch Jt't" 3
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230.1.
230 2.
2303.
The response to question 221.9 is unacceptable.
The applicanl Shpuld coowiit to sub(1(it. a rePort describing the computer program usgd for core ther(1>a'l-hydraulic analysis=prior to issuance of an operating license for Susquehanna.
The report should provide the code description, the calculational methods and empirical correlations
- used, a sample application and code verification through comparison with experimental data.
The response to question 221.2 is unacceptable.
guestion 2 requested the sen assumptions used for amount of crud used in design calcul t d
a ions an sitivity of CPR and core pressure drop to variations 'h rud present.
Merely stating that "a conservative amount in e
of crud is deposited on the fuel rods and fuel rod spacers" does not begin to answer this question.
The question also asked for a discussion of how crud buildup in the core would be detected.
no discussion is provided.
'our response to question 221.13 is incomplete.
Since the operational 4.4-6 design guidelines are exceeded for some operating co d t F'hould be revised to show decay ratios as a function of rod n i sons, igure position, recirculation flow and power.
Figure 4.4-6 as currently presented is not sufficiently detailed for use in inferrin operational boundaries.
e n
n erring 230.4 Your response to question 221,15 is unacceptable.
You reference NEDQ-10958-A for a discussion of the uncertainties and their bases.
The staff evaluation of NEDO-10958 states "The estimated value of the uncertainties and the basis for the value depend th f
'g q
pment of each reactor and will be evaluated for each on e speci ic reactor at the time Technical Specifications are issued."
Information
. to support the uncertainty values for Susquehanna must be submitted prior to issuance of a safety evaluation report for Susquehanna.
p30.5.
The staff is performing a generic study of the hydrodynamic stabilit characteristics of 1.MRs under normal operation, anticipated transients.
and accident conditions.
The results of this study will be applied to the staff review and acceptance of stability analyses and 1 t 1
ow in use by the reactor vendors.
In the interim, the staff concludes that past operating experience, stability tests, and the for acce tin the Su inherent therttal-hydraulic characteristics of LMRs id b
p g
e usquehanna stability evaluation for normal operation prov e a as s
and anticipated transient events.
However, in order to provide additional margin to stability limits, natural circulation operation of Susquehanna will be prohibited until the staff review of these conditions is complete.
Any action resulting from the staff stud will be applied to Susquehanna.
s suy i
g ass(
230.6 ~
Because the Susquehanna stabi'lity analysis is for the first cycle only, a new analysis must be reviewed and approved hy the staff prior to second cycle operation.
230.7.
No analysis has been presented for MCPR limits or stability characteristics for one loop operation.
One loop operation will not be permitted until supporting analyses are provided and are approved by the staff.
230.8.
The steady-state operating limit for the Minimum Critical Power Ratio (MCPR) is 1.25.
This value is calculated based on REDY model described in NEDO-10802.
The results of three turbine trip tests performed at the Peach'ottom-2 have revealed that in certain cases the results predicted by REDY model are non-conservative.
The General Electric Company's new ODYN for use in transient analyses has been approved.
Accordingly, the applicant is required to reanalyze prior to criticality the following transie'nts with ODYN:
- 1) generator load rejection/turbine trip, 2) feed-water controller failure-maximum demand and 3) main steam isolation valve closure with position switch scram failure. If another event should be more limiting than those listed above, the other event should reanalyzed with ODYN.
The reanalyses should include CPR cal,culation and demonstrate that the operating limit for MCPR is not less than 1.25
Materials En ineerin
.Branch - Com onent Inte rit Section Paragraph IV.A.2.a, Appendix G, 10 CFR.Part 50, requires that a reference temperature, RTNOT, be determined for each ferritic material of the reactor vessel and that this reference temperature be used as a basis for providing adequate margins of safety for reactor operation.
Previously-submitted data are inadequate to define an RTNOT for the reactor vessel ferritic materials; therefore, supply the following additional information:
a) If b'oth CVN and dropweight tests were conducted for vessel beltline shell plates as stated in FSAR 3 5.3.1.5.1.2, supply the CVN test results in addition to the previously submitted dropweight test results (per response to guestion 121.2).
Calculate an RT for every shell plate, and explain in detail the method used to establish each RTNOT value.
b) If only dropweight tests were conducted for vessel beltline shell plates as stated in the response to guestion 121.2, explain in detail the method(s) used to establish an RTNO value for each vessel plate.
NOT c)
Supply both CVN and dropweight test results for, every other ferritic vessel plate not addressed by items (a) and (b).
This should include the upper shell and 'both lower and upper, vessel heads.
Calculate an RTNOT val ue for each pl ate and expl ain in detai 1 the method used to establish the RTNOT values.
d)
Identify every ferritic weld seam in the reactor vessel by weld wire, heat number, flux type, lot of flux and welding process.
This should include any ferritic weld in the beltline region, upper shell, and lower and upper vessel heads.
Submit CVN and dropweight test results in
'ddition to We previously submitted beltline weld data.
Calculate an RTNOT fol every ferritic wel d seam, and expl ain in detai 1 the method( s )
used to establish each RT>0T value.
NOT e)
Submit the correlation data used to establish an RTH0T value of no less than -50'F when dropweight results are not available for we)d material.
This data should include weld wire and flux types, welding process,,
and heat treatment for each correlation weldment specimen.
Explain in detail the analysis used to establish the -50'F value.
121.10 Paragraph IV.A.3, Appendix G, 10 CFR Part 50, requires that materials for
- piping, pumps and valves meet the impact energy requirements of Paragraph HB-2332 of the ASi1E Code.
Materials for bolting must meet the requirements of Paragraph NB-2333 of the AStlE Code.
To demonstrate compliance with Paragraph IY.A.3, supply all impact test data for the ferritic materials of these components, Identify each material by its ASME specification, heat or lot number, and dimensions when applicable.
If any of the above data are not available, submit data from the literature and/or further tests, and analyses to demonstrate compliance with Appendix G.
121. 1I Paragraph IV.B, Appendix G, 10 CFR Part 50, requires the reactor vessel beltline'aterials have a minimum upper shelf energy of 75 ft-lbs in the transverse direction.
Insufficient data have been supplied to demonstrate that all the beltline plates and welds meet this upper shelf requirement.
Submit the following information to demonstrate compliance with Paragraph IV.B:
a)
Impact energy data for all beltline plates (21-1, -2, and -3 of Unit Ho.
1 and 21-1, -2, and -3 of Unit No. 2) that will demonstrate that the plates in the vessel beltline will have 75 ft-lbs (in the transverse direction) for unirradiated material or that the upper shelf energy will not fall below 50 ft-lbs at the design fluence level. If these data are not available, submit data from the literature and/or further tests on similar base metal, and analyses used to define the upper shelf energy level.
b)
Impact energy data for the following beltline weld materials that. will demonstrate that the weld seams in the vessel beltline will have 75 ft-lbs for unirradiated material or that the upper shelf energy will not fall below 50 ft-lbs at the design fluence level.
These welds, identified by
lot number/heat number are:
629616/L320A27AG, 411L3071/L31IA27AF, J417B27AF/412P3611, C109A27A/09N057 and E204A27A/624263, ff these data are not available, submit data from the literature and/or further tests of weld material of the same weld wire and flux type, and analyses used to define the upper shelf energy level.
>2> ~ >2 The materials surveillance program uses three specimen
- capsules, that should contain reactor vessel steel specimens of the limiting base material, weld metal and heat-affected zone material.
To help demonstrate compliance with Appendix H, 10 CFR Part 50, provide a table that includes the following information for each specimen:
(1)
Actual surveillance material; (2) origin of each surveillance specimen (base metal:
heat number, plate identification number; weld metal: weld wire, heat of filler material, production welding conditions, and plate material used to make weld specimens);
(3) test specimen and type; (4) fabrication history of each test specimen; (5) chemical composition of each test specimen.
Pl ovide the location, lead factor and withdrawal time for each specimen capsule calculated with respect to the vessel inner wall.
The above informa-tion should be submitted in tabular form as illustrated in Enclosure l.
121.13 Paragraph III.Aof Appendix G, requires that ferritic materials of the reactor coolant pressure boundary be impact tested by means of Charpy V-notch and dropweight (when required by the ASHE Code) tests.
Supply the impact test data for the vessel
- nozzles, flanges and shel'1 regions near geometric
discontinuities to demonstrate compliance with Paragraph III.A.
Each component material must be identified by heat number and location within the reactor coolant pressure boundary.
Impact test data should include test temperatures, CYN energy, and/or mils lateral expansion.
EIICLOSURE 1
P Number Time(EFPY)
Location Com Os i tion Factor
-Haterials
& Orientation REACTOR VESSEL HATERIAL SURVEILLANCE PROGRAH Ca sule Azimuthal Withdrawl Lead Surv'eillance Specimen Type Number Chemical Fabrication IIis tor Base Haterial 8
IIAK
- heat no.
- identification code no.
X Cu.
I P
) ~
Weld Hetal
- weld wire
- heat of filler material
- flux type
- lot of flux
- base material combination
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REQUEST FOR ADDITIONAL INFORMATION Sus uehanna Steam Electric Station 260.0 Qual it Assurance Branch 260.1 Section 17.1.2.2 of the standard format (Regulatory Guide 1,.70) requires the.
identification of safety-related structures,
- systems, and'components (Q-list) controlled by the QA program.
You are requested to supplement and clarify the Q-list in Table 3.2-1 of the FSAR in accordance with the following:
a.
The following items from the Q-list need expansion and/or clarification as noted.
Revise the list as indicated or justify not doing so.
1)
Clarify that the Control Rod Drive System includes the scram accumulators.
2)
Clarify that discharge piping fill lines and jockey pumps are included in the HPCI, RCIC, RHR, and Core Spray Systems.
3)
Clarify that the Emergency Core Cooling and RCIC Systems include the mechanical vortex suppression devices.
4)
Identify the "equipment associated with a safety action" as regards the Leakage Detection System.
For example, it is not clear that post-LOCA ECCS Leakage Detection Systems are included.
b.
Tne following items do not appear on the 'Q-list.
Add the following items to the list or justify not doing so.
1)
ESSM Spray Pond Emergency Spillway.
2}
Site grading.
3)
Roof scuppers and parapet openings.
4)
Pressure resisting doors.
5)
Meteorological data collection programs.
6)
Refueling Interlock System.
7)
8)
Primary Containment Vacuum Relief System - instrumentation and controls.
9)
Standby Gas Treatment System - instrumentation and controls.
10)
Missile barriers for safety related equipment.
ll} Steam lines to the HPCI and RCIC turbines along with the associated valves and res'traints.
12)
Equipment and drain floor - piping and containment isolation val ves.
13) quencher and quencher support.
14)
Downcomers and braces.
15)
Primary Containment Purge System.
16)
Primary Containment Ventilation System - piping and containment isolation valves.
17)
Onsite Power Systems (Class lE) a)
transformers b}
valve operators c) protective relays and control panels 18)
Engineered safety features DC equioment - protective relays and control panels.
19)
Biological shielding within primary containment, reactor build-ing, and control building.
20)
Nuclear boiler system instrumentation piping beyond the outermost isolation valve.
21)
Drywell cooling system piping and valves for coolers Y-414A and B, V-415A and B, and Y-416A and B.
22)
Mainsteam system piping to turbine stop valves and branch line piping up to and including first valve.
23)
Spent fuel pool lines.
24)
Radiation monitoring (fixed and portable).
25)
Radioactivity monitoring (fixed and portable).
26)
Radioactivity sampling (air, surfaces, liquids}.
27)
Radioactive contamination measurement and analysis.
28}
Personnel monitoring internal (e.g., whole body counter) and external (e~.,
TLD system).'9)
Instrument storage, calibrati'on, and maintenance.
30)
Decontamination (facilities, personnel, and equipment).
31)
Respiratory protection, including testing.
32)
Contamination control.
33)
Feedwater sparger s.
34)
Safety-related masonry walls (see IE Bulletin No.. 80-11).
35)
Measuring and test equipment used for safety-related structures,
- systems, and components.
36)
Expendable and consumable items necessary for the functional performance of safety-related structures,
- systems, and components (i.e., weld rod, fuel oil, boric acid, snubber oil, etc.).
Enclosure 2 of NUREG-0737, "Clarification of THI Action Plan Requirements" (November 1980) identified numerous items that are safety-related and appropriate for OL application and therefore should be on the g-list.
These items are listed below.
Add these items to the g-list and/or indicate where on the g-list'hey can be found.
Otherwise justify not doing so.
NUREG-0737 (Enclosure 2)
Clarification Item 1)
Plant-safety-parameter display console.
2)
Reactor coolant system vents.
3)
Plant shielding.
4)
Post accident sampling.
5)
Valve position indication.
6)
Dedicated hydrogen'penetra'tions.
7)
Containment isolation dependability.
8)
Accident monitoring instrumentation.
I.D.2 II.B.l II.B.2 II.B.3 I I.D.3 II.E.4.1 II.E.4.2 I
I.F.l')
Instrumentation for detection of inadequate II.F.2 core-cool ing.
10)
& RCIC initiation 1 evel s.
11)
12)
Challenges to and failure of relief valves.
13)
ADS actuation.
14)
Restart of core spray and LPCI.
15)
RCIC suction.
I I.K.3(13)
II.K.3(15)
II. K. 3(16)
II.K.3(18)
II.K.3(21)
II.K.3(22)
16)
Space cooling for HPCI
& RCIC.
17)
. Power on pump seals.
18)
Common reference level.
19)
AOS valves, accumulators, and associated equipment and instrumentation.
20)
21)
Emergency support facilities.
22}
Inplant I2 radiation monitoring.
23)
Control-room habi tabi1 ity.
II.K.3(2e)
II.K.3(25)
II.X.3(27)
II.K.3(28)
III.A. 1.1/ III.A.2 III.A.1.2 III,D.3.3 III.D.3.4 d.
The instrumentation and control systems and components must be identified on the 9-list (FSAR Table 3.2-1) to the same scope and level of detail provided in Chapter 7 of the FSAR.