ML19312C694
| ML19312C694 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 12/12/1969 |
| From: | Shcwencer A US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 7912190896 | |
| Download: ML19312C694 (19) | |
Text
Nd Sl: 2 6 f e,
[%
UNITED STATES ATOMIC ENERGY COMMISSION
,, g,
i g
WASHINGTON. D.C.
20545
- rn e December 12, 1969 Roger S. Boyd, Assistant Director for. Reactor. Projects,- Division of Reactor Licensing THRU: Charles G. Long, Chief, Reactor Project Br ch N. 3, Division of Reactor Licensing
- .' Z'" WITH DUKE POWER COMPANY ON OCONEE NUCLEAR' STATIO.t,aDOCKEIrNOS.
50
_ -269, 50-270, AND 50-267
SUMMARY
A meeting was held with Duke Power Company-November 18,- 19 ; 1969-to discuss the FSAR review areas of site, instrumentation and electrical, reactor physics, conduct of operations,-initial tests, internal vent valves, and steam generators. An attendance list is attached.
Prior to the meeting, reviewers in these areas-had-identified areas that will require further information.' Discussions were held.with each;. reviewer to identify areas that required discussion as opposed to straight forward information requests requiring no discussion.
An agenda was prepared and informally communicated to Duke in sufficient time for them to prepare for the meeting and bring the necessary technical personnel.
A brief general meeting for introductions and scheduling' discussion sessions included a presentation by Ray Maccary of an NDT'" Table A" prepared for Ocones Unit 1.
Except for brief su= mary sessions the remainder of the two days were devoted to concurrent discussions of agenda items scheduled to minimi::e manpower commitments.
D. Ross chaired discussions on site, conduct of operations and initial tests, and reactor physics. A Schwencer chaired i
discussions on instrumentation, electrical, internal vent values, and steam generators.
In a concluding discussion with the project staff Duke expressed concern on completion of review of balance of FSAR. We stated we are still working on an accelerated schedule from that discussed at cyr initial meeting but could offer no guarantees on completion. We expressed concern that Duke has not yet resolved the meterology problem posed by the valley drainage model which requires further justification if it is to be used.
Details of discussions are given below.
3 912 19 x
-s Roger S. Boyd 2
NDT TABLE A Ray Maccary, DRS, explained " Table A," an NDT requirement document he had prepared for Oconee Unit 1.
It is applicable to piping, pipe fittings, pu=ps and valves within the reactor coolant pressure-boundary. He stated that Duke must be in a position to certify -(to Compliance)* to Table A.
Duke incicated that both they and B&W~ had surveyed their systems and believe they either meet or exceed Table A requirements.
They-indicated that in some cases (such as certain B&W supplied valves) there would be certification sheets only in their possession or~in'B&W's possession.
In another area of concern, they were told that Table A would be applicable to Unit 2 and Unit 3 items that have already been purchased.
SITE 1.
Onsite Monitoring - There will be two monitoring points downstream of radwaste release.
One, at-the Highway 183 bridge, is inside the ex-clusion area.
Another is several miles downstream at-Highway 26 bridge.
There will be. a continuous water sampler at the-Highway 183 location.
Boiled down samples will be counted periodically.
During this part of the discussion we asked if Duke had verified the minimum dilution flow-from the
'Keowee tailrace with no hydro units operaring-(30 cfs was assumed in answer to Q.2.3, Supplement No. 1).- Duke indicated this could be done.
2.
Flexibility-in Environmental. Monitoring-We* pointed out there is a need to specify criteria for contracting or enlarging this program before the. fact.
(Duke-had already told the-Compliance inspector during a
. September 9-12 site visit that the statement in paragraph 2.7.3 of the FSAR was intended to cover only additional monitoring stations not fewer stations).
Penetration Room Leak ge - We noted that the FSAR contains very little 3.
t information cnr this engineered safety feature.: (Boundaries *are not well defined, design-data are not'given, discussion is not provided on doors or other openings, etc.) We learned they intend to operate at approximately -
0.5 in Hg vacuum. Vacuum relief will be provided at 2 in.Hg.
Alarms will annunciate at that level as well as at 0.1 in.Hg. - Air flow will be controlled
~
by a pressure signal, not flow.
(They implied that filter face velocity
.could be independently controlled.) We expressed concern on how they can and will assure themselves that they get a good vacuum distributed throughout the penetration room which has several doorways and 1s formed into " east" and " west" sections by a. constriction'at the-fuel pool.' They apparently
.had not considered this. Part of this concern-is-how they will assure integrity of room seals such as' around doors and at wall-joints, etc.' ' American Air Filter is supplying the filter package.
<3
a Roger S. Boyd '
3 We discussed how the air would be exhausted and filtered.
From Figure 6-5 it appears that a potential may exist for-ioss of air flow- (cooling)in j
one of he two filter trains cassing filter heating and potential de-s orp tion. Apparently Duke had not examined this as a possible mode of i
failure.' In pursuing charcoal filter heatup, we confirmed that they had t
calculated heat load on the basis of-the filter passing a 0.5% leak.
We then pointed out and they agreed that their proposed tech spec would permit a containment leak race up to 2.75%. -This is-1 1/2 times the containment design leak rate. We noted this would be an area requiring furthern consideration.
If we were to accept a leak rate above 0.5% then their heat load calculations are no longer valid'.- Also, the contribution of noble gases, now-based on a 0'.5% maximum-leak rate, would be invalid. I think we would have to have convincing information showing the 0.5% leak rate is unreasonable before relax 1ng' it at this stage.
4.
Radiation Exposure of Station Personnel - The applicant stated that their calculations indicated less;than 1 remr.in-90-days would be received by station personnel under worst accident conditions.
5.
Interaction of Liquid. &. Gaseous Radwaste-Systems -- We were' concerned that malfunction of-bleadback valve WD-V66 could overpressure the liquid waste tanks.
Duke said this would be prevented by a relief valve that discharges through the gaseous release filter train.
We' did not-lobtain details of maximum' pressure that might-be reached on the-discharge capacity of the relief valve. Wa-said we would need additional information on this p rovision.
6.
Radmonitoring. Instrument Range-Inconsistencies Wa asked how-Duke justified radiation monitors whose ranges did nct' cover maximum acnident or technical specification radiation levels.- - As an example we cited' the moni-tor RIA-36 which has an upper range of 50 uC1/cc for reactor coolant 1 activity whereas Duke calculated 269 uC1/cc for the 1% failed-fuel assumption. Duke l
intends to substitute a less sensitive GM type instrument-for the NaI crystal type identified in the FS AR.
We asked what this might do to their ability to promptly-detect failed fuel by means of increase in primary activity.
From their response, it appeared that there-is no other failed fuel-detection means being considered at this-time.
)
7.
Reactor Coolant"S'torage Inter Connections- (3 Units)
- In our discussion over_ the potential' to make up from tanks of varying boron' concentrations, Duke pointed out that make up is actually from the lewdown storage tank (Figure 9-2) downstream of the interconnections between plants.
Since Duke has analyzed the case 'for dilution' water with zero boron concentration entering the letdown tank'we agreed that these intereennections present no unreviewed safety item.
i a
Roger S. Boyd 4
8.
Isolatien of Waste Gas Exhauster Line We noted that-Figure 11-3 shows an apparent single f ailure situation that would cause inadvertent discharge of gaseous waste through the waste gas exhauster-line (a similar situation
- exists on discharge from the waste gas tanksP.s Since this could constitute an uncontrolled release we will need to see-how-it-is prevented.
From the discussion and information contained-in Secion ll of the FSAR it is apparent they will have the ability to hold up release-by use of the waste gas tanks and need not use the waste gas exhauster except under very favorable meterological conditions.
9.
Moderator Dilution-Accident - The accident analysis-in-Section-14.1'.2.4.1 c :.tes an inflow of 500 gpm but does' not give-letdown flow to-letdown storage tank. Duke said this would be a maximum of.140 gpm.
They-did not feel this flow rate would tax the available tankage since it would-be terminated by a reactor trip.
10.
Ability to Min 1mize Offsite Liquid Release - We~ discussed in general the ability of the present system-designs to minimizer offsite releases.
From the discussion and the data contained in the-FSAR-it appears. evident that they ahve ample capacity to holdup releases containing significant levels of activity for extended periods of time- (over 60 days).
Except for showers and laundry which can be expected to contain only trace activity and reactor coolant bleed (expansion dilution: and partial drain) which is normally processed and held in the coolant bleed goldup tanks for return to the reactor coolant system, less than 4,000 ft waste *13generatedin 60 days.
This compares with a holdup capacity of 6,900 f t not counting the reactor building sumps.
11.
Refueling Accident Doses - We noted-that we-had some concern over the applicability of their accident model which assumes only 56 of the-208 pins of a fuel element fail. - Since failure of all 208 pins could result-in a site boundary. dose of approximately 500 Rem-to the. thyroid, we asked what.means might be possible to accomodate a 208 pin failure within part-100 guidelines.
Duke would prefer not to consider the addition of another set of charcoal filters.
They said it might be possible-to utilize the penetration room-filter system, but they had not considered this.- B&W didrnot: offer any experimental or calculational data to support their-56 pin-failure assumption.
12.
Containment Integrity During Shutdown.- Duke would like to keep the personnel and equipment hatches open-during refueling ~and maintenance shutdown. - We said it was not clear to us that it was necessary to keep them open except for those brief periods when material or personnel'were being moved in or out.
Further, we could not see the need for them.co be open at all when-irradiated fuel was being-handled.
It should be a. simple matter to keep one of the two personnel hatch-doors sealed and the equipment hatch closed with at least-four (equally spacedP bolts in place.
Duke
.~
Roger S. Boyd 5
could of fer no basis except " convenience"- fortwanting to-keep them open, and said they would:have to evaluate what-the effect on shutdern time would be if they were required to keep these-hatches closed.
REACTOR 1.
Core Enrichments - Oconee 1 has a three zone core. - The-first cycle enrichments have been set.- The reload enrichment-is.not yet definite but will be very close to the values used in the physics analyses. -Initial enrichments for Oconee 2 will be set in the Spring,1970 and for-Oconee 3 in the Fall, 1970.
The reloading patterns are known and fuel cycle calcu-l lations have been made for-Oconee 1 and work on-Oconee 2-fuel cycles has been started 2.
Reactivity Eigenvalue-Calculations
.Eigenvalue calculations as a-function of life for~ several conditions- (e.g. hot,100% power, poisoned,)- have been calculated for Oconee-1 cnd' results are available-if we ask-for -them.
They are, however, still labeled as " preliminary."- They are-2-dimensional PDQ-5 and -7 calculations.
The basic power scheme for Oconee 1 is 0.8 to 1.2% Ak/k reactivity held down by a Xenon transient control group- (always-designated as group 7) and an average of 0.2% ik/k held down-by partial-insertion of. group 6 J
for boron-dilution control. The rest of the excess reactivity at hot, rated power, poisoned conditions is held down by boron shim. During the last month ~ or so of life when boron has been reduced to its minimum of 17 ppm the transient Xenon group will be slowiy withdrawn.- This-increases the shutdown margin by about 1% Ak/k during-the last month of operation.
Therefere shutdown margins calculated at end of-life- (EOL) are-about 1% greater than those one month earlier.
A pending-internal B&W report-discusses 2-and 3-dimensional (PDQ & -7) flux shapes and eigenvalue calculations.
We-did not-discuss our possible need to see this report.
-3.
' Doppler-Coefficients - Doppler coefficients come-from internal sub-routines of relatively standard physics codes and are not verified experi-mentally. 'As a substitute, B&W makes sensitivitv analyses in accident and transient calculations to account for possible errors in knowledge of-Doppler
~-
coefficients.
Doppler coefficients have been calculated-for beginning of life (30L) and EOL, and Doppler feedback has been used in the thermal hydraulic analysis"of the core.
Roger S. Boyd 6
4.
Reactor Plant Dynamic Codes -- We asked about B&W's computational" ability to predict overall plant performance during abnormal transients.- Jim-Mallay answered that he had developed a digital-analog' hybrid-full plant simulator code and it was used for the steam line break accident and had the capa-bility to be used -for other abnormal t ansients-(e.g. rod" ejection,-loss of flow).
The code is referred -to in the B&W Steam-Generator topical, but is not fully described or even named outside of B&W's shopr -When we asked for the code's name, B&W responded that-they would prefer not to answer.*
5.
Moderator Reactivity-Coefficients - Isothermal temperature and uniform void moderator coefficients of reactivity are calculated-by-2-dimensional a
PDQ-5 codes.
Power coefficients are not calculated'because the: constant T-average control scheme makes, in B&W's thinking, power coefficients equiva-lent to Doppler-fuel coefficients. -We saidithis assumption would-be invalid for transients faster than.the~ response time in the Integrated Control System.
6.
Temeerature Dependsana n6_tha--Moderator.-Coefficient. +.B&W said they had calculated the positive moderator coefficients at-BOD.for several1 tempera-tures, but could.not say how-it behaved as a -function of temperature except that it was more positive at room temperature than at operating temperature.
They are studying this temperature coefficient-because of its-importance to understanding physics starc-up tests. -We mentioned that arial fuel expansion coefficients as well as Doppler coefficients may cause start-up test interpretations to be in error.
7.
Operation With Positive. Moderator: Coefficients. -We explained that opera-
. tion at power with a positive" moderator coefficient offered the possibility of sudden-insertion of reactivity-due to voiding. Voiding could be caused by any sudden decrease in primary system pressure-(e.g.- loss"of flow, loss of coolant,; rod ejection). We pointed out that-Figure 3-5 shows a potential for the sudden-insertion of- 0.75% ak/k if the core goes from zero to 27%
uniform voids and that the possibility of non-uniform-insertion of voids may show even: greater reactivity insertion potential.
- B&W-had not considered such reactivity insertions nor had they made scoping analyses of the safety problems related co operation with positive moderator coefficients. Ve noted elaborate Tech Spec-limitations were used on Connecticut Yankee. We indi-cated they should estimate the expected nuiper of-days of' operation with boron concentrations large enough to give positive moderator coefficients.
From a private-discussion with Mallay we learned that the lumped ~ parameter models incorporated-in the' code also form the-basis ~ for B&W's-Link training simulator being-installed this month in-Lynchburg, the physical models and plant equations may not be considered proprietary but the numerical schemes and code logic and programming-details are considered proprietary.
Roger S. Boyd 7
Further, we indicated that they should consider a change-in-the coefficients due to buildup of equiblibrium Xenon and-Samarium.
8.
Control; Rod--Grouping - The control rod" assemblies-(CRA's) vill-be regrouped once during life and the function- (shutdownr Xenon rransient,- dilution control) i of the groups will be changed several times-during core life.- An expected history of CRA grouping has' been established for-Oconee-1 and-is-being worked on for Oconee 2.
9.
Control P.od Worth - CRk worths, Xenon effects and power distributions are calculated by PDQ -5 and -7.
Most calculations are 2-dimensionair for selected configurations they are-3-dimensional; - A series of ejected-CRA worth calculations as a function of : lifetime for Oconee-1-have-been made.: Also, BOL and EOL calculations for the reactivity worth of a stuck-CRA have been made.
We suggested that minimum shutdown worths for-Oconee-2;-1st cycle might be inadequate.
From FSAR information the total-CRA worth ~is only-7.4% ak/k.
Of this 1.2%'is' in the core for transient control, 2.-3%- is the power-defect, and 1.7% is the stuck CRA worth.
This leaves only-2;2%-ak/k as the hot shutdown margin. Considering errors in-CRA worth and one stuck-CRA, the shutdown margin' may not give the required operating margin of -l%'ak/k.
B&W replied with two -important pieces of information, one good and one bad:
1.
The total CRA worths appearing in the FSAR-have~already been reduced by-10% to account for calculational errors. This-is not spelled out in the~FSAR.
2.
B&W has not considered the possibility that the stuck-CRA criterion 1-for shutdown must-be in addition to a withdrawn,-inoperable CRA and they-indicated that there were times in life that the 1%
ak/k margin would not' be met-if-both of ' these CRA's did not insert.
We did not pursue J:his matter because of: lack:of -detailed information on CRA worths as a function of life and becausan3&W"said it would-be easy to regroup j
rods to give increased shutdown margin. We will need more information to 4
understand-how~this increased margin 1s co be obtained.
We asked for a physical explanation"of why-Oconee - 3,~1st cycle has only 7.4% Ak/k total CRA worth while Oconee-1,"1st cycle has 10;6% and the equilibriam cycle-has 9.6%. _ Because Oconee-', -1st cycle will have once-i burned fuel elements in one zone from Oconee 1, we reasoned that CRA worths would be very similar to the 2nd-cycle of-Oconee 1 and therefore have a total CRi worth between 10.6% and 9.6%.
B&W replied that this apparent discrepancy could be explained. by the-fact that many of the CRA's would be placed"in the once-burned fuel elements.
j 1
l l
i J
Roger S. Boyd 8
10.
Control Rod Malpositioning - We asked what-instrumentation-in addition to CRA position indicators would be effective in detecting a misplaced-CRA.
They replied that the in-core flux detectors would detect malpositions, but asserted that the detectors have no safety function.
B&W said there were two independent rod position indicating schemes,' and we said it was our understanding that only one of these detected CRA position, and the other detected position of the drive mechanism relative to -its group. B&W said that if a Tech Spec on in-core detectors was contemplated-by-DRL they would prefer that it be in the form of a reduced.. power-limit to account for in-care instrumentation being inoperable or deficient.
11.
Xenon Instability Control - B&W-is developing a Xenon instability control manual for reactor operators.
It is based. on. detection of axial offset i
as measured by half length out-of-core flux detectors.
The change to half length out-of-core detectors will be documented to us by a pending FSAR amendment.
B&W said they expect Tech Spec limitations on axial offset.
B&W intends to use part length control rods routinely"for power shaping, not just for axial Xenon oscillations.
The predicted absence of azimuthal Xenon oscillation will be confirmed by in-core instrumentation measurements.
12.
Summary - In summing up the preliminary-discussions on* reactor physics the applicant and B&W reacted to our informal expressions of need for additional information by stating they wanted as little of the design information on the public record as possible. We noted that we have not completed our review of this part of the FSAR and would decide at a later date what information need be added to the public record, what could be received in a proprietary submittal, and what could be reviewed informally (e.g. by a visit to B&W).
VENT VALVES -(Proprietary Topical-BAW-10005)
Note: Bill Smith of B&W said none of the vent valve information discussed was proprietary except the Foster Wh2eler design report.
On that basis,
these notes are not ' considered to contain proprietary data.
1.
Design-Information - We noted that although the vent valves were an R&D development, no information was available giving the design basis and development
- effort which resulted in the prototype valves.
B&W made one copy of a Foster Wheeler proprietary-design report available at the meeting and indicated they would make this an appendix to their-Topical Report BAW-10005. Only minor modifications were made-during development, mostly relating ~to elimination of jack screw galling.
-~
Roger S. Boyd 9
2.
Venting Capacity - B&W had just submitted this-information en the Midland plant and provided us with a-handout copy- (their reply to-Midland Question 6.1.4).
3.
Materials - B&W's reply to Midland-Question-6.1;1 was cited for this information.
4.
Hinge Clearances - These clearances aremprovided-in' answer to-Midland Question 6.1.2.
A sesign criterion was a-Class-I fit or-looserr It was noted that the hinge pin is held captive-by a concentrically " staked" disc at each end.
5.
Design Basis-for.. Opening and: Withstand: Pressure: - The opening pressure 0.5 psi is-based on keeping approximately-1.5. feet of waterr above the core.
The withstand pressure (600 psi) was-derived: from calculating a-differential closing pressure of 585 psi for a-36" pipe-break at-full powerr
- BAW 10008 Part 1 gives'515 psi for this pressure-indicating a margin of conservatism in the 600 psi design (which was tested at-750 psi).
6.
Basis' for :Ocening. and. Closing. Forces During:-Inspection - B&W ! stated.
that while they-had measured 30 poundsr to =ove' the. valve off-its seat and-120 pounds to-hold-it in the full open position, they could permit 120 pounds and-540 pounds, respectively, based on criteria of-1/2 pound ps1 to start ~ movement and 1 1/2 pounds to hold full opent They plan, however to establish a lower set of limitsy taking into account " base line"-data on production valve performance.
7.
Partial Flow 1Perfcrmance B&W stated they-had" included this.information in BAW 10012.
8.
Plastic Deformation - The impact analysis showing plastic hinging and deformation of the disc is contained in the proprietary-Foster-Wheeler 1
report, Based on the discussion it did not appear that this impact would prevent the valve-from remaining open-following deformation.
B&W said that once the valve-opened, it would remainropenrand that repeated-impacta J
are not to be expected-during the course of an accident.
9.
- Vibration
-B&W stated that their calculations show-less than-l.0 mil of vertical vibratory movement pessible acr the: vent valve-locationr They i
assured.us that-the - ATL tests encompassed alll frequencies and amplitudes considered possibly present at the vent valve location. They monitored for resonance throughout the tests with stethoscope; strobe-lighty and' examina-tion of all recording traces. None was-found: "They expected none-because they said natural resonant requency. of the valve assembly is approximately 1500 cps, well above~ the excitation-frequencies used in the test.
1 l
1 l
1
Roger S. Boyd 10 10.
Demonstration:.ofnRemote-Inspectionnand:Ramovai-Capability -;We said we believed they should -demonstrate these: capabilities af terr the system had experienced-hot-flow conditions. -Duke-had-been thinking of-doing this cold. This can be resolved during our review of the pre-operation tests.
11.
Jackscrews - We expressed.a concern-that this mechanism might have a potential for galling and seizing or-loss:.of. sma11 parts 71nside:the core barrel. - B&W appears to have-been sensitive to these possible problems.
They had changed-thread type, material type;: and thickness of thread coating.
The jackscrew automatic-locking-deviteris covered-by a-housing.
Nuts holding the housing in place each havetonertack weld-(not in~high flux field) to prevent backing off.
The two jackscrewsrare synchronized, but B&W says they can be misaligned-by-two turns of a screw with no problems.
ONCE THROUGH STEAM GENERATOR "Proprietarv-Report-BAW-10002 Note: Bill-Smith of B&W advised the following areas of the-discussion were proprietaryr (a) boiling length, (b) excess surface and-foulingtfactors, (c) stability-discussion and analytical: stability computer model-development, and (d) feedwater nozzle design and performance.-- Since these notes-do not contain specific-design information concerning these areas, they are not consideredito reveal proprietary-information.
1.
Technology Base - B&W indicated ~ thatrsuccess-(economic) of their-design depended upon ability to transfer-heat within: thernucleaterboiling regime over a wide range of steam quality at relatively-lower velocities than.more conventional-heat exchangers.
(It was not clear-how this-is unique with the B&W design;) - They presented a curve shewing that ample-data-is available to ensure these conditions at high velocitiesr They also said that-based on their Research^ Center tests of 1-inch and: 3/4+1nch heated tubes they obtained reasonable confidence that these conditions could-be satisified at very low ~ velocities. These tests also established a relationship that
-indicated they could expect ~ to stay within the nneleate-boiling regime for qualities above 45%.
From the-limited discussion it appeared-that: heavy reliance-is being placed on empirical results obtained from-their-7-, ~19, and: 37-tube: bof.ler tests.
B&W apparently has aiso developed several computer programs to assist-in the design ofmthe steam generator and to predictrits-behavior under-dynamic conditions. -- These programs were not-described to us-in any-detail.
2.
Fuli-Scale Verification - B&W said thermocouples would'be placed at several' locations on the S-G, but did not: describe a planned-designiverification test program. We-indicated further information-in this area would-be needed.
O
s 1
Roger S. Boyd 11 3.
Extrapolation of.9bdai-Rasuits
-We noted:that the-iargest of their models had 37 tubes which represents-less. than-0.3 percentrof the tubes in the prcduction units.- They have attempted to-justify this extrapolation by use of' a " figure-of-merit" ratio- (inner tubesito total tubes)r -We said we needed the basis for use of this-figure:ofrmeritr - Also;-it was not clear that it-is applicable to ali aspectsrof:the: design;-including the transient performance and primary and secondary-flow maldistributions' that might exist-in-the-15,000-tube production units.
4.
Direct Feedwater-Heating- (B~- Sprav):
The key to this process was to adequately heat the-feedwater as it-f ails through a steam atmosphere (ex-tracted-from the-boiling section of-the-S-G)-in such a way as* to prevent cold shocking-the outer shell of the steam generatorr--- From the-discussion it appears that-they have evolved a satisfactory: design provided-fouling conditions do not-flood the nozzles; - B&W said-the nozzles can-be-inspected during plant shutdown.
-5.
Tube Fouling - We expressed concernr that-Duke have a good handle on steam generator " state-of-fouling" since:as: fouling-increases:(1) pressure oscillations-have-been shown to increase: and-(2): pressure drop-increases-decreasing the spray fall-height-for the: feedwater, which-if excessive could flood the nozzles. - We also asked what: cleaning method-has-been selected and what thickness of sacrificial metal they: intend to use to account. for possible metal loss-due to cleaning and possible mechanical abrasion at tube
]
supportrplates.
From the discussion,-it appeared that no allowance may have been made for* tube metal loss.
Also,-Duke appeared! reluctant to-discuss the possibility of verifying-integrity of-tube walls-(UT techniques are available) due-to the expected high radiation field at the-tube sheet access a re as.
l We pointed out - that chemical cleaning-holds: the: potential-for causing multiple tube-failures and,- for; that reason;:wer expect to-have: some assurances that-the cleaning process selected- (materials; temperatures,
' and cleaning
- times) are well proven and conservative assumptions are used
.egarding metal-loss.
4 6.
Transient Response We noted that!while: transient response curves are shown f or normal system-transients: nonet are: shown-for-the abnormal
- - - transient tests performed and that' we would need more-information here.
7; Basis-for Conclusions - We noted that:ini several areas of the report conclusions: are: stated withoutt a basis:being:available. to us.- We: cited a conclusion in paragraph 3 4.3 that"statedrthere were no-failures based on an " examination." However, since: no elaboration was provided,:we.have no clue as to the nature of the examination.--We said we expected the bases for such tenciusions to be in-the report.
H
- Roger S. Boyd 12
- 8..
Vibration Testing -- In discussing: this: test- (about-S- days-duration) we asked' on what-basis they concluded-it represented an adequate wear test i
when compared to a-40 year-plant life.- B&W felt that the amplitude of the test vibrations- (*~.05 inches) was far greater:than the tube would see in normal service and withstanding this amplitude showed the tube could survive lower amplitudes.- We said wa would need additional informationr on this test including nature of tube restraintsr and supports and some understanding 4
of what is-implied by "no consequential wear " : As part of this-discussion we learned thate in the actual steam generator: the tube supports will be solid plates-drilled and-broached to: leave:landsr for the tube support and passageways for water / steam flow.
The tube: support piates will-be the same throughout the length of the tube bundle; - All tubes-in the test models are made of-Inconel 600. With some minor: variations; probably trace im-purities, they are identical to the full scale tube material.
9.
Thermai-Fatigue - We noted they-had not addressed the subject of thermal fatigue at the liquid-vapor interface which will fluctuate-at a given load and shif t nominal level with change in load.
B&W said they-had-looked into this and calculated that interface stresses will not exceed-20;000 psi.
i They stated Inconel 600 can withstand an-infinite number of cycles at 30,000 psi and therefore no problem exists.
10.
19-Tobe-Model-Tests - When we noted: ther treatment-in the report was very sketchy;-B&W said these tests were just:getting started when the topical was prepared:-- They said much new data has-been obtained from these tests, Further; a second-19-tube model-hast been* constructed and they expect to obtain additional-information from-both of these models^plus a short-section model which is being used for materials tests.
- 11.
Stabilitv-Studies We expressed concern on whether they had objectively predicted-instabilities
- and subsequently confirmed predictions by the tests.
Apparently they gave an outside' consultant the* necessary-information to simulate-S-G performance without telling-him: the nature of ther test results and' thereby obtained a form of objective confirmation.- We said we: would need seme assurance that they have a handle on the: causes of these-instabilities and a basis-for confidence in-their ability to* predict-its' exclusion.
in the full scale production
- units for Oconee.
INSTRUMENTATIOY.
1
- 1.
Seismic-Design Considerations
. B&W "piucked" the reactor: protection cabinets and-determined that resonant-frequency:is above the seismic range as given to them by Duke. - B&W modules will be shaker tested " live" over
)
. range 1-20 Hz91g and over range 0.1-1-G@4H.
ITE switchgear-has been tested for others-through-3g loading ~ with contacts being monitored. - The various
' l s
6
__y
- -w--
---e
Roger S. Boyd 13 process instrument sensors will also-ber tasted- " live";--A pressure trans-mitter has been tested over range-1-20 Hz@-10g for-2-1/2 hours and exposed to 100 Hz. 0 10g. ' It appeared from the discussions that-demonstration of seismic capabilities are being required for-instrumentation.
2.
Quality Assurance -- Mr.- E.- Patterson: ofr B&W-is a member of the standards group that developed-IEEE-279 and assuredr us:that all-B&W supplied items would meet that-document's-QA requirements:- As evidence of-how-QA is-handled on a specific-instrument they selected-theirrbistable: unit which-is used universally throughout the B&W-instrumentation;-- Ar Bailey Meter-Company production unit was examined.
B&W identified: their design-documents which included " General Design Specs for-Instrumentation C"92-18. "- A Fault analysis is performed as well as the expected performance tests-including environmental and transient exposure.
The-B&W-QA representative for Mechanical Systems, C Fletcher, stated: Bailey Meter-QA-is personally audited by him quarterly.-- Jim-Wells of Duke explained-how the equipment-is recieved and stored on site and Ollie-Bradham explained that-it would-be-installed and tested out-in' accordance with written procedures.
3.
Wire Run Separation & Fire Protection: -- These-items were-discussed in some detai1~ for the several types of cables- (power; controi; and-instru-mentation); - Typical specimens were shown. - All: powerr cables will have interlocked armor which Duke considers equivalent to a conduit-for fire protection purposes. -In general, it appears that-Duke-is executing acceptable designs-in this area and should-have no trouble-in-documenting this in the scope we feel necessary.
4.
Control-Room- & Equipment Room-Environment - Duke-believes it-incredible that air-conditioning to these areas willibe lost, except momentarily, due to standby capacity available. - However, they-have-looked at this and conclude they could operate continuously at 110* F'with an R.H. of 80%
or for-24 hours at an R.H. of 90%.
They have also looked at-loss of fans in the cabinets and found only one critical. componenti a power supply which, at nominal voltage would not overheat at 110* F ambient-for 1 112-hours.
It also appeared that means are available, through a reduced power bill, to prevent the room temperature from exceeding:110*- F.
It was-brought out that the printed circuits are not coated- (probably a good thing) and that adminis-trative procedures'are available-to preclude: cold surface condensation (sweating) on circuit boards.
Critical high-i=pedance circuit elements have been encapsulated.
B&W' had not tested' their: equipment under-dew point conditions.
During this discussion we: learned that air-conditioning-ducting is ' to be installed by a contractor, not Duke personnel.
Roger S. Boyd 14 5.
Safety Circuit Identification - Duke.has an extensive' color control scheme that will be applied to allt safety:egnipment; switch gear;- MCC's,
etc. to signify one of severni redundant:" chains" or-independent circuits.
All cables between such-items will have exteriors- (sheathes or armor) of the appropriate color.
Personnel will-be trained to recognize these iden-tifications and their significance.
" Green" personnel wfil not-be permitted to work on the systems.
6.
Plant-Shutdown Outside-Control Room:-- Dukerwill-have abiitty to shut down to hot standby outside the controir room with no prior action-before leaving the control room - We-learned that an: auxiliary shutdown panel will be provided for each unit- (by-B&W at-Duke's specific request per-B&W) somewhere outside the control room (s).
Duke'has not yet-decided-how this panel will be protected from unauthorized use.-- We-did not learn what has been added to this auxiliary panel or-how-it =eets " safety grade" requirements.
7.
- Emergency-Lighting and Offsite-Communications w We-learned that there are three-independent emergency-lighting. systems, one-de system and two ag systems. -We explained outside communications were of concern-from the stand-point of alerting the outside world of an emergency within a unit.
Duke uses a PBX with a 1-hour battery backup which connects-directly to their microwave system.
There is also a non-dial microwave circuit-from each control room tying the plant to offices in-Charlotte and Spartansburg.
Both systems
- are ac powered.
The direct-link microwave system-has an
- 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> capacity-battery-backup-further-backed up-by a propane gas generator with a 7 day fuel supply.- Also each unit-has artransmitter/ receiver operating at 49.98 and-47.84-Mega Hert: with sufficient 1 range: to reach central, a transmission substation.
Duke plans to purchase. one or more cars and a boat with this same X/R equipment.
The boat and car will also have a powered bull-horn to warn personnel within the exclusion radius of a need to evacuate.
8.
Core-Injection Bypass - B&W stated the three: ESF reactor coolant sensors are-independent of the four-RPS' reactor trip sensors shown on FSAR Figure 7-1.
Figure 7-17.shows that:the RPS and-ESF pressure sensors are positioned-in each of the hot legs to give-further separation.
B&W is preparing a revision to Figure 7-3.
Using the block diagram prepared for this revision, B&W showed that there are S-LP and 3 HP bypass switches.
The purpose of these-bypass switches is to permit normal cooldown and heatup of the reactor coolant system without activating
- the HP or-LP-injection systems.
From a discussion'of the HP system-it was apparent that B&W did not intend for the operator-to' be able-to readily remove the bypass-below 1900 psi once initiated.- They stated,- however, they were sure the bypass could be removed-below 1900 psig "some how" if it were necessary to do so.
s t'
i-F Roger S. Boyd 15 Our concern was that the bypass would be removed at 1900 psig, well above the 1500 psig trip point for activating the-HP-injection system.' B&W l
explained that the' wide range pressure-instruments used-for-this purpose j
require a substantial margin above-the set point to ensure prevention of HP injection system actuation. While:we could agree to some margin being necessary,' it was not clear that a:400 psi margin-is required or that the diverse containment pressure channels will be tested for proper operation l_
prior to-bypass. B&W made the point that rods could not be pulled' until the reactor coolant system was at 530* F and* nominal operating pressure
}-
. (2200 psig).
l 9.
Instrument Status Indicators - B&W noted-that trip status -oftali protective l
channels' are displayed to the operator.r-Bypasses are at-least indicated by lighted switches.
There remains some-doubt-in our minds as to what l
indications are available to the~ operator to warn him that one or more channels have administrative 1y-been put-inr a-test or maintenance condition
..t' and then supplied with an intentional " cheater circuit" which-defeats the protective function of the chanpel until removedv - The applicant claims.that only one* channel at a time (of 4 or ofL3 as applicable) would be so compromised for test or maintenance.
q 10.
Emergency-Safety Feature Control-Circuits'- B&W clarified-for us that.no j
ESF trip or RPS' trip can be automaticall:p cleared.--Once tripped, reset
]
must be manual. Our concern arose in connection with the controi circuit I
for solenoid operated valves appearing on Figure 7-4.
j 11.
Rod Drive Control - B&W stated the feature that limitedtcontrol rod drive speed to 30 -in./minuta was the use of synchronous programmer motors locked to the plant's 60 Hz. f requency.- Our subsequent examination of the i
i FSAR-indicates there may1 e' an intervening static-inverter supplying this b
" plant frequency." We-intend to pursue this further in our-formal questions -
to Duke.- We understand there is also a slow speed programmer motor intended -
to-be used only for jogging single - rods.
Based on the-limited discussion i
at the meeting,. it is expected that we will ask to see additional circuit details on the rod drive control system.
4 t
t
- 12. - Testing of Reactor. Building Spray
- Actuation Pressure-Switches - Duke indicated that these switches would be periodically tested by applying a calibrated test pressure to each switch.- The switches are-inside con--
tainment, but will be accessible during reactor operation.- (Duke plans
~
to go inside' containment routinelyrabout"once per shift to take readings,etc.).
- 13. -Integrated-Control System- (ICS) - During our ' review of the-FSAR we i
- noted that, for the emergency-feudwater systemr to"be-lined up, the ICF i.
had to sequentially open and _closa several valves.- B&W said they-had checked
- this 'out as wel1~ as completely ' reexamining the ICS and are satisfied._that it is not required to operate (even for the emergency-feedwater system) l-f g
t-n y y
-,+r..
~
--+y
,v..-
,_,,3 m,-
m,,-', -.r-~,
c-v
Roger S. Boyd 16 in order for any RPS or ESF system to meet required performance.- For the emergency feedwater system they said all valves could be lined up manually if ICS failed and that there is ample time to do this-(23 minutes per FSAR paragraph 14.l.2.8.3).
14.
Neutron Detectors B&W plans to use.all West {gghouse detectors-for the RPS channels. Source range detectors will use a-B licing (instead of BF3 gas).
The intermediate range detectors will be standard-Westinghouse design CIC's.
The power range detectors will be UCIC's.
B&W has changed from three 4-foot sections in one can to two 6-foot sections in one can.
They have apparently retained the ability to monitor-flux in each section (for use in detecting abnormal flux patterns).
15.
Design of Reactor ~ Protection Circuits - Maximum ~use is made of integrated circuits and solid state components, except several mechanical relays are employed.
Selected Type 709 IC operational amplifiers are standard.
Bailey Meter is supplying all B&W RPS instruments.
16.
HP and LP1 Injection. 6With andaWithout-Reactor Trip)
-We carefully pointed out we vere not taking a position; but were seeking ~information on whether-B&W's LOCA analyscs must take credit-for a reactor trip.
B&W stated that they take credit for a reactor trip in their analyses-for small size breaks.
They said they have not analyzed-LOCA consequences without taking credit for a reactor trip, since redundant reactor-low pressure trip channels are available (separate from the-HP and LP-injection-low pressure trip channels) to trip the reactor.
CONDUCT OF OPERATION AND INITIAL TESTS-1.
Single-Unit We stated that, unless there-is clear' evidence to the contrary, we-believe Oconee Unit 1 should have-five men per shif t including three licensed operators. After reaching" commercial" operation Duke might be able to justify elimination of one unlicensed operator except for startups and scheduled shutdowns.
Duke indicated that-this was near their estimate of requirements, however I
they suggested that shift size be reduced upon the completion of the appropriate power testing at a~ given power-level rather than " commercial operation'.""This would allow the use of a'four-man shift at a power-level less than 100% provided the testing at that-level had'been completed and some restraint prevented further escalation to full rated output with a four man crew. We agreed this approach seemed reasonable provided the exact details could-be finalized prior to approval of technical specifications.
Duke intends"to have at leatt one man per shift qualified'to perform the duties of a-Health-Physics" Technician.
Roger S. Boyd 17 2.
Dual-Unit - Duke made a short presentation stating that-Geonee 1 & 2 controls had been laid out in such an arrangement that'fourrmen per shift could effectively control both units from a common control" room.- Such a practice was common-in their newer dual-unit coal-fired stations now in operation.
Our position was that perhaps as many as eight men per shift could-be re-quired unless justification = for-less could-be provided:--Duke-had not carefully analytti their manpower requirements-for the worst case situation and was not prepared to-fustify their-four man proposal; - They informed us that they-intend ~to carefully examine shift manpower requirements and will want to-discuss this with us at a-later-date. -This must be" resolved in a timely fashion to insure selection and training of the necessary personnel-for operation of Unit 2.
During the general discussion regarding-dual-unit operation; Duke stated their intent to-have one operator monitor the operation of-both units if the remainder of the crew was occupied with other-duties outside of the control room.- We indicated such operation would-be unacceptable.-- (This subject cust-be addressed specifically in-the technical specifications.)
3.
Cross-Licensing.of. Operators
. Duke-intends that all' 11 censed personnel will hold licenses-for all three units at the Oconee-Station.- -This arrange-ment will allow the maximum flexibility-in shif t staffing.
4.
Emergency Plan - A general description of the overail emergency plan
.i was given-by the applicant. A preliminary copy of the plan was-left with us.
Duke was advised to expect a request for-formal submittal of the plan in-cluding medical" preparedness arrangements.
i 5.
Industrial Security - The applicant outlined'the security measures to be taken during operation of Unit-1 while the* remaining untts are still under construction.' These measures will include a perimeter-fence with gate guards, closed circuit TV for-back-shif t gate monitoring and adminis-trative control for operating equipment.--We said that-Duke should evaluate the sensitive areas of their plant and provide suitable security measures co prevent simple acts ~of sabotage. ' We discussed the general scope of the evaluation required.
Duke inquired and we put off for later resolution how this evaluation would be reviewed by the regulatory staff; -We did say; however;~ that it would be reviewed.
The point of" concern; obviously, was minimizing potential compromise of-sensitive data.
6.
Startup Organization - Duke expects ten men to-be " cold-licensed" l
~ for-initial plant startup. - Of these-ten men; two will-be management per-sonnel, leaving eight-licensed operators,to supervise plant operations until J
r
,np
.,.. - - - - + -,,, --
Roger S. Boyd 18 additional operators are obtained. - Duke ~is -planning on these men working a seven day week hours on,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> off-during this period. - B&W will not supply startup operating personnel-but will provide eight to twelve people to assist in test evaluation and technical assistant.--Supervision and performance of all test work will be done-by-Duke.
7.
Initial ~ Testing and Power Escalation - We toid-Duke we-believe they should provide the following information for each test which could have safety significance:
(a) test objectivei- (b) plant prerequisites, (c) general description of method including major steps, and- (d) acceptance criteria with allowable deviations; Duke said this-information would be available prior to testing; however, they objected to a-formal sub-mittal because it would be voluminous and subject to change;~ Duke offered to prepare the above information for a representative test-for-discussion at a later meeting at which time they would also' identify the safety areas believed significant.
l
- A-Schwencer
- Reactor:P.rojects Branch-No. 3 Division of Reactor-Licensing
Attachment:
List of Attendees Distribution:
AEC Attendees P. A. Morris F. Schroeder T. R. Wilson DRL Branch-Chiefs S. Levine D. Skovholt R. C. DeYoung Docket files l
DRL Reading RPB-3 Reading Orig: A. Schwencer 1
l
ATTENDANCE LIST DUKE OCONEE MEETINGS DOCKET NOS. 50-269, 50-270 and 50-287 November 18 - 19, 1969 AEC
. DUKE. POWER. COMPANY B&W C. Long A. Thies W. Smith A. Schwencer P. Barton D. Montgomery D. Ross C. Wylie J. Mallay
- B. Cady (19)
W. Owen R. Craig
- T. Novak (18, 19)
J. Smith
- C. Fletcher (18)
- 0. Parr (18)
R. Wells
- E. Patterson (18)
- R. Pollard (18,19)
W. Parker
- J. McCreary (18)
- K. Wichman (19)
C. Price
- R. Abbot (18)
- J. Knight (19)
B. Rice
- T. Schuler (18)
- M. Dunnenfeld (19)
W. Foley
- G. Snyder (19)
- H. Richings (19)
J. Elliott
- B. Mcdonald (19)
- R. Turner (19)
- R. Waterfield (18)
J. Hall
- J. McGough (19)
L. Lewis
- J. Taylor (19)
- J. Buzy (19)
K. Canady
- C, Russell (18)
- P.
Collins (18,19)
D. French
- W. Brunson (18)
- 0. Bradham
- R. Bybee (18)
J. Hampton
- N. Hennessy (19)
S. Nabow C. Sansbury T. Wyke H. Lark Part time (date participated)
Part time - observers only (date present)