ML19329A576
| ML19329A576 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 03/04/1971 |
| From: | Schwencer A US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Deyoung R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 8001070573 | |
| Download: ML19329A576 (8) | |
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20545 March 4,1971 Richard C. DeYoung, Assistant Director for PWRs,
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Charles G. Long, Chief, PWR-2, DRL NOTES ON MEETING WITH DUKE P ANY ON TECHNICAL SPECIFICATIONS FOR OCONEE UNIT 1, DOCKET NO 0-269 )
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The meeting was held at Bethesda, January 28, 1971.
An attendance list is enclosed.
Our objective of finalizing the Technical Specifications was not ealized.
A discussion of outstanding areas is enclosed.
Major unresolved areas are (1) the staff bases for pressurization, heatup and cooldown limitations (fracture tougTness enacerr.9) i (2) reactor coolant leakage specification format, (3) auxiliary power degredations and bases, (4) res triction on cranes and hoists during refueling, (5) radioactive waste disposal.
Other areas requiring resolution are noted in the enclosed discussion.
Duke indicated that plant completion has slipped so that they will not be ready for fuel loading until late May at' the earliest with a more likely date being sometime in June.
Duke is not aware of any intervenors other than the North Carolina Municipalities on antitrust matters hearing).
(which is not expected to require a prelicensing
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A. Schwencer PWR Project Branch No. 2 Division of Reactor Licensing
Enclosures:
1.
Discussion 2.
Attendance List DISTRIBUTION:
DocketfilesT DRL Reading CMUpright, CO Region II RKlecker PWR-2 Reading NThomasson MMMann PAMorris RLBursey DRRoth
'FSchroeder
'TRWilson DRL Branch Chiefs RCDeYoung DRS Branch Chiefs ECCase, DRS I
Compliance (2)
-MSHildreth, CO, Hg.
y 8001070 6 [3 R
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~ DISCUSSION OF OUTSTANDING AREAS ' F THE OCONEE 0
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UNIT 1 TECHNICAL SPECIFICATIONS
_ GENERAL We used marked-up copies of the Oconee Tech Specs submitted as Amendment 24 dated December 14, 1970 (Page. references below refer to that document.)as a basis for discussion
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Rated Power Definition (p 15-2) - Resolved.
Duke agrees it is 2568 MWt core output, not " system" output.
We agreed that there should be no need to state in the definition that it is contingent upon "all four coolant pumps operating."
_ Containment Integrity (p 15-3) - Resolved.
Specification 15.1.2 A and emergency hatches remain closed except dur i
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or personnel passage through these hatches.
4 ns Heat Balance - Resolved.
that is to be compared to neutron power. Duke agreed that it is core thermal p We understand that certain since a secondary heat balance will be employed. computa Single Loop Operation - (p 15-21) - Resolved.
i revisions which included (1) adding a statementDuke agreed to proposed is authorized for testing only (2) timely notificatithat single loop single loop operation and stating that trip points shall on of tests and ent "to no higher than" rather than "at" 50% of rated power and 610' reactor outlet temperature.
F this Technical Specification (15.2.4.1 and two othersAlso, as pointed o 15.4.3) belong under limiting conditions of operation (15.2.4.2 and relocate them accordingly.
Duke will Reactor Coolant System Activity (p 15-24) - Resolved.
whichever is more limiting.were relaxed to permit activity to be 293 uCi/
Requirements X/Q _value is obtained.
The basis were expanded ' to show how ev Pressurization, Heatup and Cooldown Limitations (p 15-28) basis for the increased conservatism which we are
- Unresolved e
not " marry
(DRS has been asked to provide the basis input needed We will 1
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bukeLhas agreed for Specification 15.3.1.3.D2 to use an "integratethe 6
period of an additional 1.7 x 10We also requested that they give 15.3.1.1-1.
in proper consideration to use of capsule specimen test results in updating Figure establishing actual vessel exposure.
Unresolved - Ray Klecker and Dr. Mann Reactor Leakage _ (p 10-36) -
i to requested Duke to use the Point Beach Technical Specificat onsDuke will e We want to standardize.
extent possible.
We also said we wouldn' t accept
' Heach' Tech Spec for this purpose.
the 2.7 gpm as acceptable for offsite Part 20 doses.
(p 15-42) - We requested Duke to Moderator Temperature Coef ficient simply state that the max._ moderator coef ficient at full powerAny discu 4 ok/k/*F.
. shall not exceed +0.9 x 10 There should be no is calculated should be put in the bases.
p exceptions for which this value could be exceeded.
(p 13-47) - Unresolved - We noted their ble Features Engineered Safetyuse of the, term " safeguards" which has special meanin to other than nuclear power plant accidents.
the term " safety features" instead in the tech specs.
We also noted that only single instruments are associated with t r the BWST level, the core flooding tank pressure, and the reac oDuke s building emergency sump levels.
pressure channel to the CFT and are considering the feasibility (contro of adding a second BWST level instrument We agreed that a second emergency sump level indicator is not Either Duke will add a second BWST level instrument or h
tr required.the single instrument must be operational in order for t e reac o to be critical except for a brief maintenance period on the instrument.
Operational Safety Instrumentation (p 15-53) - Unresolved -
Specification B - Duke agreed to only one channel bypass key in control. room but now contrary to prior commitment during FSAR i
laced on the use review,- Duke wishes to relax the restrict ons pThere are 4 reactor protection chan of dummy bistable trip units.
flow each of which has 8 bistable trip units,1-power imbalance-comparator trip, 2-power-pump-comparator trip, 3-low-pressure t e trip, 4-pressure-temperature comparator trip, 5-high tempera ur
) trip, and trip, 6-high pressure trip, 7-high-power (neutron flux 8-shutdown bypass high pressure trip.
During the FS AR review Duke was.given th'e choice of (1) indicate by light on the control console the specific-protec i-(2) indicate by light on the control console just the pro-unit i
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tective channel that has a dummy bistable unit in place of one-of its bistable trip units and treat the entire protective channel as bypassed in determining minimum instrumentation requirements, or -(3) eliminate the dummy bistables.
It was pointed out by Dr. Mann that the bases for justifying the dummy bistable trip units has not been given.
As a minimum,- this basis must be added.
In addition, further review will be required.if we relax our position on the use of dummy bistables.
With regard'to failure of _ a control rod drive trip device the point was made that it is important both to eliminate a condition where one more failure can prevent a reactor scram and to verify that the remaining trip devices are still able to trip upon demand.
Duke does keep spare breakers available for this system but at the I
meeting was unwilling to commit ability to take all required
. j corrective action within 30 minutes as we proposed. We await Duke reply on this.
Reactor Building (p 15-69) - Resolved.
Duke agrees to adminis-tratively verify that manual containment isolation valves are closed prior to returning to critical af ter a refueling shutdown.
Auxiliary Pouer (p 15 Unresolved - Duke still wishes to have prior approval to degrade to complete loss of the Hydro Station substituting the Lee gas turbine during such an interval.
Our concern is that, should a distribution system outage cause loss 4
of the power grid, there would be only one source of power for l
Unit 1 shutdown loads.
This unique situation will exist until Unit 2 becomes available.
The November 13, 1970 basis for this specification (15.3.7) needs to be revised to account for the final version of this Technical Specification.
The staf f will write these bases since we do not intend to permit system degredation to the extent originally requested by Duke.
Fuel Loa 6.ng and Refueling (p 15-74) - Unresolved - Duke wishes to ignore ms.datory controls on the movement of the polar crane during fuel handting' operations.
Because of the demonstrated accident potential associated with cranes and hoists we cannot overlook this matter.
As a minimum, there must be a clear requirement which prohibits any polar crane movement over the open core and refueling canal'when a fuel assembly is attached to or being transported by either of the fuel handling bridges located inside p
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- the ' reactor' building.
If possible, this should be accomplished by physical interlocks, otherwise administrative control in the form of direct supervision should be employed.
Radioactive Waste Disposal (p 15-76) - Unresolved -
The project leader was requested to seek: full compliance in this Specification to the 'new release requirements covered by Amendments to 10 CFR 20 and 10 CFR 50 which were published December 3,1970 in the Federal
. Register (Volume 35, ' o. - 235 on pages 18385 through 18388)
N e f fec tive ~ January 2, 19 71.
to be
- The_ applicant-felt that he was in " substantial" agreement with this new requirement and expressed the opinion that the present specifica-tion wording was largely that supplied e1rlier by the staff.
The one apparent major Duke objection to fuli requirements published December 3,1970 it 'ompliance with the new she need to pass all i
reactor building purge through the installe '3 EPA and charcoal filter system even if there were_ no "measureable" activity being exhausted.
They contend this will unnecessarily use up these expensive filters They also objected to increasing gaseous waste holdup for 20 days rather than 10 days as propased.
~60 day holdup capacity and therefore could notWe pointed out they have in e for this objection.
understand the basis They also objected to placing on the record the estimated quantities of each of the principal radio-nuclides expected to be released i.
annually in the liquid and gaseous effluents, contending that it could be derived from the material already in the FSAR (in our judgement, this is true only if we assume 1% failed fa el is expected)
- Also, through discussion, it was established that the low activity s
waste tank _ and condensate tes t tank activity in Specification 15.3.9.B.9 is meaningless in that the restricted area dose, level specification 15.3.9.B.5 would be exceeded before reaching the tank activity levels
.pe rmi tted.
Therefore we intend to delete 15.3.9.B.9.
Table 15.4.1-1 (p 15-88) - Unresolved - Duke was requested to conform with other.PWR Tech Specs on frequency of instrument surveillance.
We did not accept the unproved premise that B&W instrumentation L
is;better than that provided by other vendors and thus warrants relaxed surveillance.
is obtain~d.
This is a matter to consider af ter experience e
- On reactor: heat balance the procedures for obtaining this heat balance
' have not been written. We suggested that, as'a minimum, there must be a daily comparison of neutron flux measurements and heat balance
- me'asuremen ts.
to maximize agreement between heat balance data.We do not intend th
- for "drif t" should be given as the basis of recalibration.However, criteria Duke
.was in general' agreement with this approach.
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UTable.15.4.1-2 (p 15-89) _ :. Unresolved - We recommended addition of 1 appropriate testsf on -(l).- fire pumps and power supplies, (2) con--
tainment isolation ~ trip, (3) service water system, and (4) spent fuel cpoling. syn tems.
Table 15.~4.1-3 (15-90) - Unresolved - We recommended that E
' determination be started when gross activity in the reactor coolant exceeds 24 pCi/ml because 240 uCi/ml is much too close to the Tech
- Spec limit of 293 pCi/ml to be reasonable.
Ne = also recommended that unit vent gaseous release be tested for
- particulate.as well as iodine and thac a 50% increase in gross release rate within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> be reason for determining the release rate-increase for iodines and particulates provided the gross release rate' exceeds 1% of the maximum allowable release rate.
ECCS 'and 'RB Cooling System Tes ting (p 15-93) - Unresolved - We
-l noted that the motor-operated valves in the core flooding tanks should i
be exercised for as short a time as practicable to determine mechanical integrity.
Duke did not feel these valves are engineered safety feature valves and, therefore had not intended to verify their operability.
Based on further discussion, Duke will cover these valves to indicate functional verification upon return to pressure af ter cold shutdown.
A basis will be added to address motor-operated core flooding tank valves.
Containment Leakage Tests _ (p 15-102) - Unresolved - We require all applicants to determine that containment leakage measured at less than-accident pressure be the lesser of two fractions of allowable
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accident condition leakage as oetermined by (1) the ratio of leakages measured at these two pressures prior to initial unit operation,
.(L
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), or (2) by the square root of the ratios of the two pressures, pm (P /p jl/2,
Duke resists use of the L f method as beinr potentially misleading tm Lpm or even indeterminate where actual leakage measured is extremely low magnifying measurement errors perhaps even to the point of resulting in apparent negative leakage (in leakage).
Our. position is that actual test data indicates thpg containment leakage paths are 'too complex to make the (P /p )1 model alone p
an infallat'+ extrapolation tool. Containments have been known
. to leak :even less as pressure _is increased. At this point in time we are.willing to permit the reduced pressure test provided both methods of extrapolation are used and the most conservative results
-accepted.. Otherwise, a full accident condition pressure will be acceptable for._the periodic integrated ' leak test.
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6 Reactor Building Hydrogen Purge System (p 15-108a, - Unresolved -
We stated than an in-place system test with the portable unit hooked-up should be performed initially and during each refueling
. period to verify full inventory and functional performance of the complete system. We also stated 'that the hydrogen concentration instrumentation accuracy should be ' verified initially and periodi-cally.
Further we note that the FSAR (p 14A-14) indicates that hydrogen concentration samples can be obtained both from the main reactor and from the purge line to obtain representative conditions.
The bases should include justification for the hydrogen concentration measurement techniques to be employed. If moisture content is sig-nificant (as it may well be) means for taking this into account should be included also.
j Emergency Feedwater Pump Tests - p15-117) - Resolved - Duke has agreed to specify a minimum operating time to assure that the i
pump has reached operating conditions during the periodic test.
Table 15.4.11 Resolved - Duke has agreed to minor changes and corrections to this table which (1) will require gross alpha and beta activity analysis on a scheduled basis regardless of activity 90 and 1131 level and (2) add K to the gamma analysis for fish and milk samples and (3) require Keowee River water samples to be analyzed monthly.
Reactor Design (p 15-127) - Resolved - We informed Duke that these Technical Specifications cannot apply to other "similar" design reload fuel because this_ is, at present, an unreviewed safety matter.
Therefore, Specification A.4 was eliminated.
Fuel Storage '(p 15-129) - gesolved - It was agreed for Specification 15.5.4.B.2 that the spent fuel pool should be filled with borated
- water with a minimum concentration of 1800 ppm boron when fuel is present.
General Office Review Committee (p 15-132) - Resolved - It was agreed that this Committee shall have only one member who is also a member of the Oconee Station staff. Duke prefers that person to be the Superintendent of Oconee.
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I-Station Operating' Procedures - Resolved - In Specification 15.6.2. A,
" procedures" will be stated to include applicable check-off lists, and ~ instructions.
In Specification C, shift supervisor approval of' temporary minor changes to written procedures shall be written also.' 'In Specification D, selected drills will be conducted quarterly.
Radiological Controls - (p 15-141) - Unresolved - Duke has been told to respond to the January 10, 1971 letter (on the subject'of an exemption to CFR Part 20 for the use of respiratory equipment) with.the data requested therein.
We can then procede by incorporating the exemption in this Technical Specification as initially issued if response is timely.
If not, the exemption and credit for use of this equipment will have to be covered by an amendment.
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ENCLOSURE 2 ATTENDANCE LIST 1-28-71 Duke Power Company Paul Barton William Parker Edward Smith L. Lewis L. Snow K. Canady
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B. Rice i
B&W R. Straub AEC M. M. Mann, DR R. C. DeYoung, DRL*
R. Klecker, DRL C. G. Long, DRL*
A. Schwencer, DRL N. Thomasson, DRL*
M. S. Hildreth, C0 Hg.
C. M. Upright, CO Region 11 R. L. Burrey, DRL D. R. Roth, DRL
- Indicates Part Time t
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