ML18017A083
| ML18017A083 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/20/1980 |
| From: | Parr O Office of Nuclear Reactor Regulation |
| To: | Curtis N PENNSYLVANIA POWER & LIGHT CO. |
| References | |
| NUDOCS 8003060599 | |
| Download: ML18017A083 (21) | |
Text
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'IVIIII.Nm7MIlIlBtIL3 I3PP DISTRIBUTION W/Rlt:LOSURE:
Doc et Fs es D.
Ross NRC PDR D. Vassallo Local PDR S.
Varga LHR 83 File F. Williams
- 0. Parr Docket Nos. 50-387/388 W. Houston V. Benaroya H. Kreger M. Ernst R. Denise OELD IE (3)
S. Miner M. Rushbrook R. Mattson S.
Hanauer J,
Knight R. Tedesco R.
DeYoung V. Moore BCC:
Mr. Norman W. Curtis Vice President - Engineering and Construction Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 4
Dear Mr. Curtis:
pEB g0 1980
SUBJECT:
SUSQUEHANNA STEAM ELECTRIC STATION. UNITS NOS.
1 ANID 2-REQUEST FOR ADDITIONAL INFORMATION Some of this review has been performed by the Savannah River Plant (SRP).
The questions in the Enclosure were originated by SRP.
Please contact us if you desire any discussion or clarification of the information requested.
Further, in our letter of February 8, 1980, the Instrumentation and Control Systems questions were numbered incorrectly.
Question 032.61 should have been numbered 032.64, 032.62 should have been 032.55 and so forth.
Sincerely, Qri"Iin"I C'"nQ4 bg O. D. Parr Olan D. Parr, Chief Light Hater Reactors Branch No.
3 Division of Project Management
Enclosure:
As stated cc w/enclosure:
See next page As a result of our review of your application for operating licenses for the Susquehanna Steam Electric, Plant, we find that we need additional information in the area of Reactor Systems.
The specific information required is listed in the Enclosure.
OFFICE P SURNAME OATE P
, LW~P:BC, ODParr/LLM 2/~
/80
,ASB.
VBenaroya 2/ jf0 /80
'I houston 2/ gg /80 NRC FORM 318 I9 76) NRCM 0240 4U.S. GOVERNMENT PRINTING OFFICE: 1979 289 369
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Mr. Norman W. Curtis FEB 20 ]ggp cc:
Mr. Earle M. Mead Project Engineering Manager Pennsylvania Power 8 Light Company 2 North Ninth Street Al 1entown, Pennsyl vania 18101
'ay
- Silberg, Esq.
Shaw, Pittman, Potts Trowbridge
'.1800 M Street, N.
W.
Washington, D. C.
20036 I
Mr. William E. Barberich, Nuclear Licensing Group Super visor Pennsylvania Power 8 Light Company 2 North Ninth Street Al 1 entow, Pennsyl vania '8101 Edward M. Nagel, Esquire General Counsel and Secretary Pennsylvania Power 8 Light Company 2 North Ninth Street Al 1 entown, Pennsyl vani a 18101 Bryan Snapp, Esq.
Pennsylvania Power 8 Light Company 901 Hamilton Street Al 1 entown, Pennsyl vania 18101 Robert M. Gallo Resident Inspector P. 0.
Box 52 Shickshinny, Pennsylvania 18655 Susquehanna Environmental Advocates c/o Gerald Schultz, Esq.
500 South River Street Wilkes-Barre, PA 18702 John L. Anderson Oak Ridge National Laboratory Union Carbide Corporation-Bldg. 3500, P. 0.
Box X
'Oak Ridge, Tennessee 37830 Mr. Robert J. Shovlin Project Manager Pennsylvania Power and Light Co.
2 North Ninth Street Al 1 entown, Pennsyl vani a 18101 Matias F. Travieso-Diaz, Esq.
- Shaw, Pi ttman, Potts 8
Trowbr idge 1800 M Street, N.
W.
Washington, D. C.
20036 Dr. Judith H. Johnsrud Co-Director
'nvironmental Coalition on Nuclear Power 433 Orlando Avenue State College, PA 16801 Mr. Thomas M. Gerusky, Director Bureau of Radiation Protection Department of Environmental Resources Commonwealth of Pennsylvania P. 0.
Box 2063 Harrisburg, PA 17120 Ms. Colleen Marsh Box 538A, RDP4 Mountain Top, PA 18707 Mrs. Irene Lemanowicz, Chairperson The Citizens Against Nuclear Dangers P. 0.
Box 377 RDgl
- Berwick, PA 18503
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ENCLOSURE RE UEST FOR ADDITIONAL INFORM.
ON 211.0 REACTOR SYSTEMS BRANCH Provide a realistic range and permitted operation band for the exposure dependent parameters in Tables 4.4-1 and 15.0-2.
In Table 15.0-2, provide assurance that values of parameters selected yield the most conservative results.
Uncertainty exists on the correct value of APRM neutron flux scram setpoint to, be used in transient analyses.
The value indicated as input for transient analysis in Table 15.0-2 is 125K, NBR.
- However, a value of 1205 NBR is indicated in Table 7.2-4 and 7.6-5.
Explain this discrepancy.
For the correct value of setpoint used in transient analyses, provide a breakdown of any uncertainty allowances that are added to the nominal value.
Provide a listing of the transients and accidents in Chapter 15 for which operator action is required in order to mitigate the consequences.
In the Chapter 15 time sequence of events or NSOA tables, provide the times of, and manual actions or auto-matic system changes that are required to place the plant in the final. stabilized condition (cold shutdown}.
The response to question 211.113 does not provide sufficient detail on non-safety grade equipment and components which mitigate transients and accidents.
Provide a table of the non-safety grade equipment and components assumed to mitigate consequences for each transient and accident in Chapter 15.
For those events where non-safety grade systems are used; provide the change in consequences or results when taking credit for safety grade equipment only.
The analysis of transients and accidents in.Chapter 15.0 does not state which of the RPS time response delays in Table 7.2-5 is used in the REDY computer code model (NEDO-10802).
For each transient and accident in Chapter 15.0, specify whether the sensor or overall delay time is used in the analysis and why the specified delay time is conservative.
Confirm the following items for all transients in Chapter 15.0 which require control rod insertion to prevent or lessen plant damage.
a)
All calculations were performed with the conservative scram reactivity curve Ho.
2 in Figure 15.0-2.
b)
The slowest allowable scram insertion speed was used.
a)
In Table 1 of Figure 5.1-3a (Nuclear Boiler), the relief valve spring set pressure at 1130 psig for safety/
relief valves B and E does not agree with a corresponding value of 1146 psig in Table 5.2-2 of the FSAR and in Table 1 of Drawing M-141, Rev. 9.
Correct this setpoint discrepancy for safety mode (mechanical) actuation.
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b)
For transient analysis, credit has been taken for safety/
relief valve actuation in the relief mode.
A more conservative approach would be to take credit for safety/
relief valve actuation in the safety mode, resulting in higher peak vessel pressures.
1)
What effect on MCPR and peak vessel pressure does credit for safety/relief valve actuation in the safety mode have on trans'ients analyzed in Chapter 152 2)
Are all equipment and components required for safety/
relief valve actuation in the relief mode safety'graders Modify Table 15.0-1 as follows:
a)
Give calculated values of MCPR instead of the entry >1.06.
b)
For the "feedwater controller failure at maximum demand" transient, correct'he discrepancy in values for maximum vessel
- pressure, maximum steam line pressure, and MCPR that exists between Table 15.0-1 and Section 15.1.2.3.3.
For transients and accidents in Chapter 15 in which it is stated that the operator initiates some corrective action, provide'ustification for any corrective actions by the operator prior to 20 minutes.
Discuss how the pre-operational and startup tests will be used to confirm flow parameters used in Chapter 15 analyses.
Provide details of any previous test of components in test facilities conducted to show satisfactory performance of the recirculation and feedwater flow control systems and respective pumps.
Oescribe how this information was used in Chapter 15 analyses.
Analyze the turbine trip and generator load rejection transient from a safe shutdown earthquake event.
Credit should not be taken for non-seismically qualified equipment or, any equipment contained in a non-seismic structure.
On page 4-7 of NEO0-10802, it is stated that the difference in trend of flow coastdown versus initial power between the analytical and experimental coastdown curves for Oresden Unit No.
2 (a BWR/3) in Figure 4-11 was due in part to differences between actual and computed jet pump efficiencies.
a)
How has this effect been treated in analysis of SSES transients involving flow coastdown with two recirculation pump trip (RPT)2 b)
Is this treatment applicable to Susquehanna which is a BWR/42 If so, explain how.
211.148 (15.1.1.2.1) 211. 149 (15.1.1.2.3) 211.150 (15.1.1.3.2) 211.151 (15.1.2.2.1) f'15.'l.2.3. 1) 211.153 (15.1.2.3.1)
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- 15. 1.2.3.3.)
211'.155 (15.1.3.3.3.)
For the "loss of feedwater heating" transient, the sequence of events in Table
- 15. 1-2 for the limiting manual flow control mode is not described in sufficient detail to permit evaluation of transient results in Figure 15.1-2 and comparison wi.th NSOA events in Figure 15A.6-21.
No detail is presented jn Table 15.1-2 between 2 and 40 plus seconds.
Revise Table 15,1-2 to include NSOA events in Figure 15A.6-21 and additional detail between 2 and 40 plus seconds.
The thermal power monitor (TPM) is not included in the Susquehanna design per response to question 211.118.
However, it is indicated as the primary protection system trip for mitigating the consequencesof the "loss of feedwater heating" transient in Section 15.1.1.2.2.
What was used to scram the reactor in the manual mode?
Modify Figure 15A.6-21 and Sections 15.1.1.2.2.
and 15.1.1.2.3 accordingly.
This section states that input parameters and initial plant conditions for the "loss of feedwater heating" transient are in Table 15.0-1.
This should be changed to Table 15.0-2 in this section and in the corresponding sections of the remaining transients in Chapter 15 where this discrepancy occurs.
Correct discrepancies between events in Table 15.1-3 and NSOA Figure 15A.6-22 for the "feedwater controller failure at maximum demand" transient.
Table 15.1-3 does not include the initial core cooling and reactor vessel isolation events indicated in Figure,15A.6-22.
Explain the basis for the assumed feedwater flow controller failure at 1354 flow.
Is the indicated failure initiated at 0 seconds or does the failure begin at 0 seconds and increase to 135K flow at a later time.
If the former is true, correct figure 15.1-3 accordingly.
Correct the inadvertent combination of Section 15.1.2.3.2, beginning on page 15.1-7, with Section 15.1.2.3.1.
Provide justification that analysis of the "feedwater controller failure-maximum demand" transient at 105K NBR steam flow is more restrictive than at low power.
If so, delete reference to "low power" for NSOA event No.
22 in Table 15A.2-2.
If not, re-analyze and make appropriate corrections.
a)
It is not apparent from the text whether the "pressure regulator failure-open" transient is terminated by a low turbine-inlet pr essure trip or a LB trip.
Trips indicated in various sections of the text are summarized below:
Section 15.1.3.2.1.1 15.1.3.3.2 15.1.3.3.3.
Table 15.1-4 Low pressure at the turbine inlet Low pressure at the turbine inlet L8 trip Low pressure at the turbine inlet
(15. 1.3.3. 3) 211.156 (15.1.3.2.1)
Specify which trip is most restrictive on thermal margins and revise applicable tables,
- sections, and figures of the FSAR.
II b) It appears that less than the assumed ll55 NBR steam flow in Section 15.1.3.3.2 was simulated at the beginning of the transient in Figure 15.1-4.
Explain this discrepancy and make corrections, if necessary.
c)
Safety/relief valve (SRV) actuation for this transient in the relief mode is not included in Tables 15.0-1 and
- 15. 1-4 and Figure 15.1-4 for decay heat removal.
Please explain.
In Table 15.1-4, a)
Include safety/relief valve actuation times for the "pressure regulator failure-open" transient.
b)
Indicate the value of steam flow simulated at time
= 0, p):1 V.3.2) 211. 158 (15.1.4.3.1) 211.159 (15.2.1.2.1)
Specify the assumed'operating mode (manual or automatic) of the recirculation flow control system for the "pressure regulator failure-open" transient and provide justification that the. most conservative results on core thermal margins are obtained with the assumed operating mode.
A qualitative presentation of results for the "inadvertent safety/relief valve opening" transient is given because analyses from earlier FSAR's indicated this event is not limiting from a thermal margin standpoint.
a)
Provide supporting data that justifies this condition (i.e.,
referenced plant and MCPR).
For the "pressure regulator failure-closed" transient, correct the discrepancy that exists between the 5 psi setpoint difference for the backup pressure regulator in Sections 15.2.1.1.1 and 15.2. 1.2. 1 and a corresponding 10 psi setpoint difference in Section 10.3.2.
211.160 (15.2.1.3.3) 211.161 (15.2.2.3.2)
It is stated that the pressure disturbance in the reactor vessel
.from failure of the primary pressure regulator in the closed mode is not expected to exceed flux or'ressure scram trip set-points.
Explain the bases for this conclusion.
In the evaluation of the "generator load rejection" transient, a full-stroke closure time of 0.15 seconds is assumed for the turbine control valves (TCV).
Section 15.2.2.3.4 states that the assumed closure time is conservative compared to an actual closure time of more like 0.20 seconds.
However, in Figure 10.2-2,
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(15.2.2.3.2) 211.162 (15.2.2.3.3) 211'; 163 (15.2.2.4.1) 211.164 (15.2.3.2.1.3) 211.165 (15.2.3;2.1.3) 211.166
.(15.2.3.2.3.1)
Turbine Control Valve Fast Closure Characteristic, an acceptable apparent non-conservative discrepancy and the effect it has on analyses in Chapter 15 requiring TCV closure.
Explain why vessel steam and bypass flows in Figure 15.2-1 drop to zero at approximately 37 seconds instead of zero at 45-plus seconds from a L2 vessel level isolation in Table 15.2-1.
Ouring the "generator load rejection with bypass" transient, it is stated that peak pressure remains within normal operating'ange.
Explain how this is accomplished since safety/relief valve actuation in the relief mode occurs from the pressure increase.
Correct NSOA Figure 15A.6-31, Protection Se uence Hain Turbine
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h>s error occurred during revision of this figure per guestion 211.110.
Mould a turbine trip coupled with failure of the operator to put the mode switch in the startup position before reactor pressure decays to <850 psig (action (5)) be more restrictive on thermal margins than the "turbine trip with bypass failure" transient analyzed in Section 15.2.3.3.3.2?
This section addresses the effect of single failures and operator errors for turbine trips at power levels
>67%.
a)
What is the basis for power levels
>67%?
b)
Explain the discrepancy with NSOA Figures 15A.6-26 and 15A.6-31 which refer to power levels
>30%.
211.167 Ouring the "turbine trip with bypass" transient, explain (15.2.3.3.3.1) why vessel steam and bypass flows in Figure 15.2-3 drop to zero at approximately 37 seconds instead of zero at 45-plus seconds from a L2 vessel level isolation in Table 15.2-3.
211'.168
.(15.2.4.5)
This section includes a detailed discussion of activity above the suppression pool, activity releases to the environs, and offsite radiological doses for HSIV closure transients.
Explain why this information was n'ot included in corresponding sections of other events in Chapter 15 requiring SRV actuation.
For instance, the "generator load rejection with bypass failure" transient clearly has a higher peak vessel pressure and longer blowdown.
t,
. 211.'169 (15.2. 4. 2.1) 211.170 (15.2.5.2.1) 211.171.
(15;2.5.3.3)
Table 15.2-5 does not list all signficant events up to 40 seconds for the "closure of all MSIV" transient.
Include the following items:
a)
Significant actions associated with attainment of applicable vessel setpoints.
b)
Recirculation pump runback if it wassimulated in the analysis.
Include the time at which the turbine stop valves are closed in Table 15.2-10.
This section states that the turbine bypass valve. and main steam isolation valve closure would follow the main'urbine and feedwater turbine trip about 5 seconds after they initiate during the "loss of condenser vacuum" transient.
Based on this, the bypass valves should close at approximately 5.01 seconds instead of 12.1 seconds in Table 15.2-10 and Figure 15.2-6.
Explain this apparent discrepancy.
211.172 Add the following items to Table 15.2-12 to be consistent with (15.2.6. 2. 1. 1)
, Figure 15A. 6-28 for the "loss of auxiliary power transformer" transient:
a)
Safety/relief valve actuation b)
Reactor vessel and containment isolation 211.173 (15.2.6.2.1.2)
Add the following items to Table 15.2-13 to be consistent with Figure 15A.6-29 for the "loss of all grid connections" transient:
211. 174 (15.2.7.2.2) a)
Reactor vessel and containment isolation b)
Initiation of the standby AC power system IC is indicated in the "loss of feedwater transient" that credit is taken for safety/relief valve operation with "low setpoints" to remove decay heat since bypass valves become ineffective with HSIV isolation.
Specify the value of the low set points used in the analyses.
What are the consequences
. if the safety function of SRY is used?
(See f211'.139).
211. 175 (15. 2.9.2.1.1)
For the "failure of RHR shutdown cooling" transient, the FSAR considers alternate shutdown cooling methods in the event the residual heat removal (RHR) system in the suction line may not be used because of valve failure.
In the analysis, valves in the automatic depressurization system (ADS) were used to transfer fluid (steam, water or a combination of these) from the reactor vessel to the suppression pool.
The RHR system removes the added heat by removing cooling water from the suppression pool and injecting it into the reactor vessel.
We require that you perform a test or cite previous test results to demonstrate that the AOS valves can discharge the fluid under the most limiting conditions when the fluid is all water.
Show that the
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(15.2.9.2.1.1)
'211".176 (15.3.1.3. 3.2) alternate method is a viable means of shutdown cooling by comparing the system hydraulic losses with the available pump head.
Hydraulic losses should be provided for each system component and, wherever possible, should be derived from experimental results.
Table 15.3-2 indicates that zero vessel steam flow does not.
occur until after 46 seconds for the "trip of both recircu-lation pump motors" transient.
However, Figure 15.3-2 indicates zero steam flow occurs at approximately 36 seconds.
Explain this discrepancy.
211.177 (15.3.1.3.2)
In the analysis of one and two recirculation pump trip events in Sections 15.3.1, a minimum design rotating inertia was used to obtain a predicted rate of decrease in core flow greater than.
expected.
Specify the.inertia value used for each transient in Chapter 15 and the basis for selection.
In the selection
- basis, include the effect on MCPR and reactor vessel pressure.
211.178 (15.3.1.3.3.2) 211. 179 (15.3.3.3. 3) 211. 180 (15.4. 4) 211. 181 (15.4.4) 211.182 (15.5.1)
. Include relief valve flow in Figure 15.3-2.
a)
Table 15.3-3 indicates that zero steam flow should not occur until after 41.7 seconds for the "seizure of one recirculation pump" transient.
However, Figure 15.3-3 indicates zero steam flow at approximately 35 seconds.
Explain this discrepancy.
b)
Include relief valve flow in Figure 15.3-3.
The narrative on page 15.4-13 discussing the "abnormal star tup of an idle recirculation pump",transient states, "The water level does not reach either the high or low level set points."
Table 15.4.3 indicates a low level trip occurs 22.0 seconds after pump start.
Figure 15.4-6 indicates a low level trip occurs approximately 23.5 seconds after pump start.
Further:
a)
Table 15.4-6 indicates a low level alarm 10.5 seconds after pump start while Figure 15.4.6 indicates this alarm occurs about 11.5 seconds after the pump starts.
b)
Table 15.4-6 indicates vessel level beginning to stabilize 50.0 seconds after the pump starts.
Figure 15.4-6 shows no such indication.
Resolve these discrepancies.
Identify the diffuser flow units in Figure 15.4-6 (and also in Sector 2 of Figure 15.4-7).
If this is X.
flow, explain why diffuser flow 1
drops to zero about 30 seconds after the pump starts.
The narrative of page 15.5-3 discussing inadvertent HPCI startup and Table 15.5-1 both indicate full HPCI flow is
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(15.5. 1) established at approximately 195 of rated feedwater flow in one second.
Explain why the curve of feedwater flow in Fjqure 15.5-1 does not show this change.
211. 183 (15.6. 1) 211.184 (4 6)
The FSAR indicates that the inadvertent relief valve opening transient is analyzed in Subsection 15.1.4.
- However, no analytical data (curves) are provided in Subsection 15.1.4.
Supply necessary information so that this transient can be evaluated concerning a decrease in reactor coolant inventory.
A number of inconsistencies exist among narrative descriptions,
- tables, and figures in Appendix 15A relative to the control rod drive system.
Please resolve the following:
a)
Table 15A.6-2 indicates that event 7 can occur in states C
8 D.
Figure 15A.6-7 indicates applicability to states A, 8, C, D.
The narrative on page 15A-35 indicates any state.
b)
Table 15A.6-2 indicates event 16 can occur in states A, B, 5
C.
The narrative and Figure 15A.6-16 indicate applicability in states A & B only.
c)
Figure 15A.6-17 and the narrative on page 15A-39 indicate event 17 is applicable in states C 5 D.
The definition indicates that it is not applicable in state C.
d)
Figure 15A.6-25 does not indicate event 25 is applicable to state D only.
e)
Figure 15A.6-28, Table 15A.6-2 and the narrative on page 15A-44 for event 28 are inconsistent for a'pplicable states.
f)
The narrative on page 15A-50, Table 15A.6-4 and Figure 15A.6-40 for event 40 are inconsistent for applicable state.
211.185
-(3.13.1) 211.186 (3.13.1)
Regulatory Guide 1.29, Section C.l.e, specifies that portions of the steam systems of boiling water reactors extending from the outermost containment isolation valve up to but not including the turbine stop valve, and connected piping of 2 1/2 inches or larger nominal pipe size up to and including the first valve that is either normally closed or capable of automatic closure during all modes of normal reactor operation, be classified Seismic Category I.,
You state on page 3.13-10 that your equivalent portion of the steam system is non-Seismic Category I.
Justify your design deviation from the above require-ments.
a)
Item (5) on page 3.13-11, discusses those portions of structures,
- systems, or components (SSC) whose continued function is not required but whose failure could reduce the functioning of items important to safety.
Provide a
list of these SSC.
211.18?
(3. 13. 1) 211.188 (3.2.1)
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(3.2.2) 211.189 (3.2.2) 211. 190 (3.2.2) 211.-191 (3.2.2) b)
Regulatory Guide 1.29, Section'.4, requires that Appendix B
of 10 CFR 50 should be applied to the above SSC.
Provide justification for not including such items in the 10 CFR 50 Appendix B Quality Assurance Program.
a)
Provide a list of those structures, systems and components which form interfaces between Seismic Category I and non-seismic Category I features.
b)
Provide justification for not adhering to 10 CFR 50, Appendix B for such items (item (6), page 3.13-11).
In Table 3.2-1, fill in the following information, where mi ssing:
(1)
Principal construction codes and standards (most pages).
(2)
Page 18, Main Steam System:
Pressure
- vessels, heat exchangers (all information).
(3)
Page 1, Nuclear Boiler System:
Air supply check valves (safety class).
The RHR pump return line as shown on P 5 I Diagram M-151 (Figure 5.4-13) penetrates into the Suppression Chamber as a Safety Class 2, Qualtiy Group B line (pipe 18"-GBB-109).
After pene-tration, the quality group classification is changed to D.
Standard Review Plan Section 3.2.2. states that changes in quality. group classification are usually permitted only at valve locations, with the valve assigned the higher classi-fication.
Demoristrate that the safety function of the system is not 'impaired due to the fact that quality group 'classification changes at a point where no valve was located.
The RHR containment spray line piping (within isolation valve) is listed as Quality Group A, Safety Class I, Seismic Category I (Table 3.2-1, page 4).
In Figure 5.4-13 (P
8 ID M-151) this line is indicated as 12" GBB-118, i.e. Quality Group B.
Resolve this inconsistency.
Table 3.2-1, page'10, lists piping and valves forming a part of containment boundary of the Reactor Building Closed Cooling Water System as Quality Group B, Safety Class 2, Seismic Category I.
Penetration of primary containment for this piping is not shown on any of the relevant P
E I Diagrams.
Show the above piping and valves on appropriate P
8 I Diagrams and indicate the classification of this piping.
Since the initial discovery of cracking in boiling water reactor (BWR) control rod drive return line (CRDRL) nozzles, General Electric (GE) has proposed a number of solutions to the problem.
One solution GE has proposed is a system modifi-cation that involves total removal of the gCDRL and cutting and capping of the CRDRL nozzle.
It appears
.frgm your response to 211.7 that SSES plans this modification.'he staff asked for more information on the impact of this modification on your plant and also required a
SSES comnitment to preoperational testing to verify performance of the modified CRD system in guestion 211.43.
When you respond to 211.43, you should address the applicable items and staff concerns specified in the letter from D. Eisenhut, NRC, to R. Gridley, GE, dated January 28, 1980,. on the subject of control rod drive return line (CRDRL) removal and capping CRDRL nozzles.
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 February 20, 1980 MEMORANDUM FOR:
Victor+ena~rya Wayne Hoosfon FROM:
SUBJECT:
Olan D. Parr REVIEW OF QUESTIONS FOR SUSQUEHANNA, UNITS 1
AND 2 Reactor Systems Branch has forwarded to DPM what they refer to as the first set of additional questions pr pared by the Savannah River Laboratory in support o
the RSB review of the Susquehanna FSAR.
RSB c utioned that Questions 211.168 and 211.85 thru 11.191 should be endorsed by AAB and ASB.
Your oncurrence of the attached letter which forwards t$ e questions to the applicant will serve as that ehdorsement.