ML20040D003
| ML20040D003 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 01/15/1982 |
| From: | Tramm T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8201290467 | |
| Download: ML20040D003 (20) | |
Text
___.
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4 N Commonwoalth Edloon
) On2 First National Plaza. Chicag2, Illinois
-( O "j' Address Reply to: Post Office Box 767 -
(,e. Chicago, lilinois 60690 -
i January 15, 1982 i
Mr. Harold R. Denton, Director m
l Of fice of Nuclear _ Reactor Regulation 8
q U.S. Nuclear Regulatory Commission 9
i-Washington, DC 20555-S REC 3ft.g)
I M2 310ggg >.
Subject:
Byron Station Units 1 and 2
-1 Braidwood Station Units 1 and 2 {
T@gpg"g,.3 J
Advance FSAR Information re 4 g
.s NRC Docket Nos.- 50-454/455/456/45 q,f
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, Dea r Mr. Denton:
This is to provide advance. copies of information which will be included in the Byron /Braidwood FSAR in the next amendment.
Attachment A to this letter: lists the information enclosed.
One (1) signed original and fif ty.:nine ~ (59). copies o f this letter'are provided.
Fifteen (15) copies of the enclosures are 4
-included for your review 'and approval.
i
-Please address further questions to this of fice.
very truly yours, ff $Al-i p' Nuclear Licensing Administrator T.R.
Tramm i
Pressurized hater Reactors p
l Attachment i
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3129N fODE l-
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I 8201290467 820115 i
PDR ADOCK 05000454 A
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ATTACHMENT A LIST OF ENCLOSED INFORMATION I.
FSAR Question Responses New:
360.1 Revised:
022.22 022.72 Q6.2.4.10 110.37 110.63 212.155 II.
FSAR Text Changes Subsections 14.2.5, 14.2.11 Preop and Startup Tests:
Tables 14.2-21, 35, 75, 86, 89.
III.
Miscellaneous Items MEB Item 14 (revised response) l l
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B/B MEB Item 14 (Revised Response, 1/15/82)
A.
Inter-system LOCA/ Periodic Leakage Testing Periodic leakage testing of RCS pressure isolation valves i
identified as inter-system LOCA check valves will be done to demonstrate that each check valve leaks less than 5 gpm.
The measurement will be made using either the installed flow neter on the test lines or portable flowrate instrumentation.
A diagram showing the general test configuration is attached.
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r 0/B-FSAR prerequisites have been completed.
Test personnel will be gx instructed to initial and date the prerequisites included in-(
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each test procedure.
Data will be examined as each test
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proceeds and out-of-tolerance conditions will be-recorded and described in adequate detail to perndt post-test analysis.
Test data that is unsuccessful will be recorded, evaluated during post-test review, and resolved within the Quality Assurance program.
14.2.5 Review, Evaluation, and Approval of Test Results Init ial startup tests that fall within the scope of the Quality Assurance program will be subject to two stages of evaluation.
First, a detailed and comprehensive review by station personnel will be made.
The Station Nuclear Engineering Department project personnel will perform a second and final review and evaluation.
Modifications or rework of systems or equipment required to resolve deficiencies will be accomplished in accordance with controlled procedures.
Retesting, if required because of modification or rework, will be documented and filed with the initial test record.
The initial core loading procedure will specify the startup tests that must be completed prior to commencement of fuel load.
All testing identified as falling within the pre-operational test phase will be completed and the results g-~s evaluated prior to core load.
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Modification and rework on systems that is required to resolve test deficiencies is controlled by the On-site Review Board during post test review and by Project Engineering who has responsibility for final test acceptance and approval.
Project Engineering may specify additional test requirements to resolve test deficiencies prior to final test approval.
14.2.6 Test Records The initial startup test procedures and test data will be retained and maintained in accordance with the Quality Assurance program described in Chapter 17.0.
The original test records will be revicwed for completeness, identified, and indexed to establish them as part of a permanent record to be retained.
These records will include data sheets completed during the test.
14.2.7 Conformance of Test Program with Reculatory Guides Appendix A to the FSAR identifies those Regulatory Guides applicable to Byron /Braidwood and describes the anticipated degree of conformance to each.
14.;2.8 Utilir.ation of Reactor Ooeratino and Testina Excerience in Develaument of Test Procram x_/
The initial test program at Byron /Braidwood is similar to the programs condacted at the Quad-Cities, Dresden, Zion, and La Salle Stations.
14.2-3
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B/B-FSAR Preoperational testing will proceed concurrently with construction testing.as various systems reach completion and are
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turned over to the station stafr.
The principal milestones
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during this phase are expected to be the reactor coolant system hydrostatic test and the integrated hot functional test.-
The f ormer test is expected to be accomplished approximately 10
, months prior to f uel load.
Hot functional testing is expected to begin about 3 months before fuel load.
Tests of other systems will be scheduled as appropriate to support these tests.
A schedule for testing is provided in Figure 14.2-1.
Individual preoperational t'ests will be conducted as early in the 1
test program as practical and at no time will the safety of the plant be totally dependent on the performance of untested systems, components and features.
Core load will occur only after the satisfactory completion and approval of all preoperational tests.
Individual startup tests will be conducted after core load and test data obtained at each power test plateau will be evaluated and approved prior to increasing power load.
Any initial test schedule overlap at the Byron and Braidwood Stations will not result in significant divisions of responsi-bility or dilutions in the staff provided to implement the test program.
Preoperational test procedure drafts will be available for review ns by I&E inspectors at least 60 days prior to their intended use
(
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'and not less than 60 days prior to the core loading date for
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startup test procedure drafts.
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I 14.2-10
B/B-FSAR
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O TABLE 14.2-21 LEAK DETECTION SYSTEM (Preoperational Test)
Plant.iondition or Prerecuisite Prior to core load.
Test Objective l
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To ensure the containment sump and reactor cavity sump leak ~ detection systems are functional.
Test Summary Reactor cavity sump and containment sump level and flow monitoring instrumentation shall be functionally tested to verify proper operation.
Acceptance Criteria
) The leak detection system operates in accordance with Subsection 5.2.5.1.
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14.2-33 I
B/B-FSAR O
TABLE 14.2-35 AUXILIARY BUILDING HVAC (Preoperational Test)
Plant Condition or Prerecuis'ite Prior to core load.
Test Objective To demonstrate operation of the Auxiliary Building heating, ventilation, and air conditioning (HVAC) system.
Test Summary The system will be operated to check for leaks, demonstrate flows to the areas supplied by the system and to verify motor currents, speeds, setpoints, and check alarms.
Automatic l
operation of dampers will be demonstrated on high radiation signals.
The test will verify that auxiliary building exhaust fans automatically start on an ESF signal, and charcoal
( ) booster fans automatically start in response to high radiation signals.
Acceptance Criteria The Auxiliary Building HVAC system supplies ventilation in accordance with Subsections 9.4.5.1, 6.5.11.2, 9.2.7.3, respectively, and Regulatory Guide 1.52 (with comments and exceptions and as stated in FSAR Volume 14, Appendix A, pages A1.52-1 and A1.52-2).
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O 14.2-47
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B/B-FSAR or understood and considered not to adversely affect the '.afety of continued operations.
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The following reactivity addition sequence will be used to assure that criticality will not be passed through on a period shorter than approximately 30 seconds:
Nuclear monitoring data will be analyzed concurrent with RCS boron dilution to permit accurate predictions of the conditions under which criticality is expected to occur.
If, during RCS boron dilution, the nuclear monitoring data indicate a significant departure from expected response, dilution will be terminated until the source of the unexpected response is corrected, or understood and considered not to adversely affect the safety of continued operations.
When the Inverse Count Rate Ratio (ICRR) from any nuclear monitoring channel reaches approximately 0.1, the RCS dilution rate will be reduced to approx-imately 30 gpm, and nuclear monitoring ICRR data will be obtained and renormalized to 1.0.
Dilution at this new rate will be continued until criticality is achieved.
If criticality will be achieved by withdrawing control rods, the following will be followed:
When the ICRR reaches approximately 0.3 (after renormalization), the dilution will be terminated and approximately 15-30 minutes of waiting will be j
undertaken to permit RCS mixing.
Control bank D i
will then be withdrawn incrementally until criti-
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cality is achieved.
l Control bank D will be positioned as required to achie'te a stable startup rate of approximately 0.1 to 0.2 decades per rqinute and to allcu the neutrcn flux level to inc r e n.
until epproximatcly 1 x 10 " anp is indicated en the intermediate range nuclear channels.
l Control bank D position will be adjusted as required to.naintain criticality at che #1ux level established until the reactivity effects of RCS mixing are negligible.
Acceptance Criteria Thepgant is critical with the flux level at approximately lx10 amps on the intermediate range nuclear channels.
14.2-87a
D/B-FSAR TABLE 14.2-86 SHUTDOWN FROM OUTSIDE THE CO:ITROL ROOM (3
s_/
(Startup Test)
Plant Condition or Prerequisites Above 10% reactor power.
Test Obiective To demonstrate that the plant can be maintained in hot shutdown f rom outside the control room.
Test Summary When above 10% power, the plant will be tripped and the planc maintained in hot shutdown for a period of time from outside the control room.
To demonstrate that sufficient data can be taken outside the control room to verify a plant hot standby condition following shutdown, and to maintain a stable condition for 30 minutes.
Accentance criteria
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, Data obtained from outside the control room demonstrates:
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1.
The plant is tripped and in a stable hot shutdown condition fo,r at least 30 minutes.
2.
The capability for cold shutdown from the hot shutdown condition:
a.
Reactor coolant temperature and pressure can be lowered to permit operation of the core decay heat removal system.
b.
Operation of the decay heat removal system can be initiated and controlled.
c.
A heat transfer path to the ultimate heat sink can be established.
d.
Reactor coolant temperature can be reduced approxi-mately 50' F using the decay heat removal system at a rate in accordance with the Technical Specifications, Chapter 16.0.
OV 14.2-98
B/B-FSAR TABLE 14.2-89
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SO% LOAD REDUCTION (Startup Test)
Plant Condition or Prerequisites During power escalation.
Test Obiective To demonstrate manual and automatic plant response to a 50% load reduction.
Test summary Demonstrate plant response to a 50 % load reduction from approximately 100% reactor power.
With plant systems in their automatic modes of control, the 50% load reduction is initiated from the turbine control panel.
Acceptance Criteria a.
Reactor and turbine nust not trip.
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b.
Safety injection is not initiated.
c.
Neither steam generator relief valves nor safety valves should lift.
d.
Neither pressurizer relief valves nor safety valves should lift.
c.
No manual intervention should be required to bring plant conditions to steady state.
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I 14.2-101
B/B-FSAR O
Question 22.22 V
The FSAR Table 6.2-1 identify for each secondary side rupture case the single failure assumed.
RESPONSE
The single failure assumed in each of the secondary si62 rupture cases was a steamline stop valve failure.
This has been incorporated into the new results which were provided in Amendment 30.
In addition, the mass and energy releases given in Table 6.2-50 are based on the worst case failure.
This
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case is a failure of one MSIV to close and the loss of one containment spray train.
The feedwater and main steam isolation valve closure times associated with the mass and energy release data in Table 6.2-50 are given in Table 6.2-9.
In the case of a loss of one heat removal train (2 RCFC and 1 spray train) the peak containment pressure and temperature is 33 psig and 315' F respectively.
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022.22-1
B/B-FSAR QUESTION 022.72 "Concerning the containment isolation design of the hydrogen recombiner lines to and from containment:
a)
Verify that"Ihe following containment isolation valves have positive position indication in the control room and are remote manually operable from the control room in accordance with SRP Section 6.2.4 ll.5.c and ANSI N271-1976 Paragraph 4.2.2 and 4.2.3:
00G059 00G063 00G061 00G064 00G062 00G065 b)
Describe the isolation provisions for the hydrogen recombiner discharge lines (00G45B 3 and 00G43B 3).
Although the normally open valves (00G060 and 00G066) in these lines are supplied with power from emergency buses, they must receive an automatic containment isola-tion signal, be remote manually operable from the control room, and have positive position indication in the control room to be acceptable as containment isolation barriers."
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RESPONSE
The hydrogen recombiner containment isolation provisions will be revised to add an outboard isolation valve close to the containment on each of the four supply and return lines for each unit.
These valves will have position indication in the control room and will be able to be operated from the control room.
The outboard and inboard valve on each line will be powered from the same Class lE bus.
l The electrical feed to each recombiner and its associated c
l suction and discharge valve will be powered from the same Class lE bus.
Hydrogen recombiner 00G064 will be powered from bus E12.
The recombiner suction and exhaust valves and the unit crosstic valve will no longer receive containment isolation signals.
All valves associated with the hydrogen i
recombiner flow paths will be normally closed.
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O Q22.72-1 l
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B/B-FSAR
,New Ouestion (6.2.4.10) 1 s_-drovide the required information requested in FSAR Table 6.2-58 for the new process radiation lines penetrating containment.
RESPONSE
This infermation has been added to Table 6.2-58.
The hydrogen and process radiation monitoring containment isolation valves are powered from Class lE sources, have position indication in the control room and can be operated from the control room.
The primary mode of operation on containment isolation is automatic with the secondary mode remote manual operation from the control room.
Both process radiation containment isolation valves are located out--
side containment and are powered from opposite C, lass lE sources.
Each redundant hydrogen monitoring line is supplied with two containment isolation valves in series outside containment.
Power to the valves in series in each line is powered from the same Class 1E bus.
All these valves are' located outside containment so that the failure of any electrical bus will not negate the capability of opening the valves for monitoring.
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B/B-FSAR
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(4)
A list of snubbers on systems which experience
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sufficient thermal movement to be measured will be included in the test program when the design is final.
Cold and hot positions are indicated on all snubbers.
(5)
A description of.the thermal monitoring program will be included in the test program.
- Briefly, the thermal monitoring program consists of a visual verification of proper snubber movement, as indicated on the snubber, from room temper-ature to maximum operating temperature.
This will be done for those systems whose normal operating temperature exceeds 250* F.
If maximum operating temperature is not attained during testing, the expected amount of movement will be calculated by multiplying tne movement indi-cated on the snubber by the ratio of the temper-ature rise to the maximum operating temperature to the temperature rise ;o the test temperature (AT /AT Extrapolated results will be evaluated to $ssubb' snubbers remain within their stroke capabilities.
(6). If vibration levels are noted beyond the accep-
,-s) tance level of (3) above, attempts will be
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made to identify and eliminate the source of-the vibration to the extent that the acceptance levels are met.
If this is not possible, an analysis of the vibrational effects on the piping will be performed, which may or may not necessitate the addition of corrective restraints to limit stress and fatigue levels to within design limits.
If,no snubber trarel is observed during the thermal monitoring program, the snubber will be removed and tested to assure that the " drag" effect on thermal growth is within design limits.
If it.is not, the snubber will be repaired or replaced.
If it is, the piping will be examined for evidence of constraint differing from design.
N Q110.37-3
r-D/B-FSAR O)
"(a)
During initial system heatup and cooldown, at
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specified temperature intervals for any system which attains operating temperature, "erify the snubber expected thermal movement.
" (b)
For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement.
"(c)" Verify the snubber swing clearance at specified heatup and cooldown intervals.
Any discrepancies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
"The above described operability program for snubbers should be included and documented by the pre-service inspection and pre-operational test programs.
"The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion.
This test program should be specified in Chapter 14 of the FSAR."
b)
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RESPONES The: snubbers intended for use on safety-related systems f
at Byron Station were procured under Specifications F-2771 and F-2805.
The preservice inspection of the snubbers will be based on the following:
1.
Qualified Hydraulic and Mechanical Snubbers were procured from Boeing Engineering Company and ITT Grinnell.
Qualifi-cation Criteria is addressed in each respective specifi-cation.
2.
Design Verificaiton is achieved through documents on file at Byron, and the equipment supplier assuring that snubbers were manufa"tured to applicable codes and standcrds.
3.
Verification of operability is as addressed in specifi-cations.
Documentation addressi.lg Operability Testing is on file at the msnufacturer's site.
i 4.
Snubbers will be installed as identified on the applicable l
Sargent & Lundy design drawings for component supports.
If installation cannot be made per thece drawings, Sargent &
l g-~g Lundy component support installation guidelines and tcl-crances will be used.
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Q110.63-2
B/B-FSAR l
(M)
C 5.
Preservice Visual Examination:
A documented visual l
examination will be performed on all safety-related i
snubbers no longer than 6 months prior to hot-functional i
testing.
This visual examination will verify:
a.
No visible indications of damage or potential degraded i
operability.
b.
Attachments are in place and appear secure.
t c.
Snubber position, as documented during thermal covement verification, has changed, indicating that snubbers have freedom of movement.
d.
Verification that hydraulic snubber leak ratas are i
less than or equal to design parameters.
6.
During hot-functional testing, snubber thermal movement will be monitored and documented as described in the response to Question 110.37.
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0110.63-3
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B/B-FSAR
/~N QUESTION 212.155 6
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" Provide more detailed information on the RWST heating system than is given in your response to Q212.50.
Spec-ifically address:
a)
Seismic Design b)
Single Failure Criteria c)
Power Supplies d)
Piping and Control System Diagrams"
RESPONSE
The RWST heating system consists of an Electric-to-Water heat exchanger with a water circulating pump, piping and valves, as shown in Figure 6.3-1.
The entire heating system is Category II (non-safety-related).
a.
Seismic Design
' None of the electrical power, control or instrumentation circuits as designed to meet Seismic Category I requirements.
The heating system is connected to the RWST with two l
fs inch Category II piping.
Protection against inadvertent
(
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draining of the tank in the event of a pipe rupture
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is provided by a Category I stand pipe on the return side of the system, and a Category I manual shut off valve at the Category I/II interface on the supply side.
If a break occurs in the Category II portion of the system, alarms will result from drainage into the Auxiliary Building sumps and also from the RWST level when the tank is drained to the low level alarm point.
Drainage from the low level alarm point to the minimum Technical Specification tank level through a fully severed two inch line would require in excess of 45 minutes.
This.
would afford sufficient time to take action to close the isolation valve.
In the event a line break occurs on the suction side of the heating pump downstream of the manual isolation valve, the maximum flow from the line would be 260 gpm.
Based upon the times required to complete switchover to recirculation following a LOCA as given in Tables 022.25-1 and 022.25-2, the line break would not impair the function of the RWST.
b.
Single Failure Criteria
(~s t:one of the electrical power, control of instrumentation
(_)
circuits as designed to meet Single Failure Criteria requirements.
Q212.155-1
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B/B-FSAR
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c.
Power Supplies Elec'tric power for this system is derived from the following non-safety-related (non-Class lE) buses:
Pump Motor (power and control) - 480 BMCC #134V5 (LAP 48E)
Heater (power) - 480 V Switchgear 134Y (lAPl7EN).
Heater ACB Control - 125 Vdc Distribution Panel 114 (lDC06EB).
Instrumentation - Miscellaneous Control System Panel ilPA20JC and 480 V MCC 134V5 (LAP 48E).
l d.
Piping and Control Svstem Diagrams Refer to Figure 6.3-1.
l To prevent the RWST vent from freezing during cold weather, heat tracing will be added to the portion of the vent pipe which is external.
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0212.155-2
CYRON-FSAR QUESTION 360.1 "At the meeting on December 22, 1981, on geotechnical questions, your geologic consultants stated that the glacial till overlying the bedrock at the site is now considered to be entirely from the Illinoian stage of the Pleistocene, and that no Wisconsinan till is present in the site locality.
Please discuss this determination providing the basis for it, the absolute age with support-ing evidence, and all relevant references."
RESPONSE
The FSAR states that three glacial tills have been identified at the plant site (Byron Station Subsection 2. 5.1. 2. 3.1. 2).
These till deposits are not consistently present throughout the site, having irregular thicknesses and distributions.
~Furthermore, at no one place at the site are all Quarternary units present in a complete stratigraphic section.
The oldest glacial till at the site is the Ogle Till Member of 'the Glasford Formation and is considered to be Illinoian in age.
This unit was observed in the trench wall at the solution basin recently studied.
The FSAR states that the other two till units are the Argyle and Esmond tills in the castern portion of the site by unnamed preglacial deposits.
(~Nx_)
In the western portions of the site, where the Argyle is present, it directly overlies weathered bedrock or residual soil (Byron Station Subsection 2. 5.1. 2. 3.1. 2).
The Esmond Till is separated from the underlying Argyle Till by the Morton Loess.
The Esmond Till is present only in the extreme eastern edge of the site and is not present west of the cooling towers.
Unpublished studies performed in the past year or so by the Illinois State Geological Survey (ISGS) reinterpret the age of the Esmond and Argyle tills.
These studies now show that the Esmond correlates with the Sterling or Radnor tills of the Illinoian Stage.
Furthermore, these units have been stratigraphically miscorrelated in the past and will be redesignated in future nomenclature (Drs. John Kempton and Leon Follmer, ISGS, personal communication).
The present understanding of glacial stratigraphy in northern Illinois shows that no Wisconsinan' tills occur at the site.
The nearest known Wisconsinan till is found in southeastern Ogle County in the Bloomington Moraine.
We understand that the reinterpretation of glacial deposits at the site was forwarded to Dr. Ina Altermann of the NRC by Dr. John Kepton of the ISGS.
t This reevaluation is of technical significance but, except
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for nomenclature, does not change the overall interpretation of the site geology in the FSAR.
Revisions to Subsection 2.5.1.2.3.1.2 and several fiaures will be made in the next amendment to update glacial stratigraphy.
Q360.1-1