ML19347C735
| ML19347C735 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/31/1980 |
| From: | Baynard P FLORIDA POWER CORP. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19347C736 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-1.A.1.1, TASK-2.B.2, TASK-2.B.4, TASK-2.K.2.10, TASK-2.K.2.13, TASK-3.D.1.1, TASK-TM 3--30-QP, 3-0-30-QP, NUDOCS 8101050285 | |
| Download: ML19347C735 (14) | |
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'N' Mr. Darrell G. Eisenhut
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Director X.g]
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^5 Office of Nuclear Reactor Regulation M
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j U.S. Nuclear Regulatory Comission r
Washington, DC 20555 u
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 NUREG-0737, Post-TMI Requirements
Reference:
Florida Power Corporation (FPC) Letter, Baynard to Eisenhut, December 15, 1980
Dear Mr. Eisenhut:
The referenced F1orida Power Corporation (FPC) letter provided proposed implementation and licensing submittal schedules as requested by your letter of October 31, 1980.
This letter will address those items of the l
referenced FPC letter which indicated a December 31, 1980; January 1, 1981; or January 2,1981, submittal schedule.
' contains a summary listing of the above items.
- provides a discussion of the Enclosure 1 items which are not referenced to another letter.
j As always, we are prepared to meet with you and discuss our implementa-tion and/or submittal of these requirements.
Very truly yours, FLORIDA POWER CORPORATION l
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Patsy Y. Baynard Manager l
Nuclear Support Services PY8ekcW01(D8-3) 810,.105 oft 6 p
Enclosures General Office 3201 ininy-fourth stree t soutn. P O Box 14042, St Petersburg. Flor da 33733 e 813-866-5151
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STATE OF FLORIDA COUNTY OF PINELLAS P. Y. Baynani states that she is the Manager, Nuclear Support Services Department of Florida Power Corporation; that she is authorized on the part of said company to sign and file with the Nuclear Regulatory Com-mission the information attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of her knowledge, infonnation and belief, s
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.W 0 W. Y. IBdyaard
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Subscribed and sworn to before me, a Notary Public in and for the State and County above named, this 31st day of December 1980.
i h (lu aa u t & O 4 f
Notary Public
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I Notary Public, State of Florida at Large, My Commission Expires: June 8, 1984
-PYB/MAHNotary(08-3) 1
ENCLOSURE 1 Chronological Sumary of Items
Reference:
FPC Technical Specification Change Request Baynard to Reid, December 31, 1980 Item Title Description I.A.1.1 Shift Technical Advisor Trained per Category B of NUREG-0578, Lessons Learned I.A.1.3 Shift Manning Technical Specification (Reference)
II.B.2 Plant Shielding Plant Modifications Description II.B.4 Training for Mitigating Program Description Core Damage II.E.1.2 Auxiliary Feedwater System Technical Specification Initiation and Flow (Reference)
Indicat ion II.E.3.1 Emergency Power Supply for Technical Specification Pressurizer Heaters (Reference)
II.K.2.10 Safety-grade Anticipatory Design Description Reactor Trip I I.K.2.13 Thermal-Mechanical Report
- Report to be Provided by January 31, 1981 III.A.2 Emergency Preparedness Upgraded Emergency Plan and Meteorological System Description III.D.1.1 Primary Coolant Outside Submittal of Outstanding Containment Leakage Data and Technical Specification (Reference)
III.D.3.3 In-plant Radiation Technical Specification Monitoring (Reference)
PYBekcW01(D8-3)
ENCLOSURE 2 Item I.A.1.1 - Shift Technical Advisor As delineated and justified in our December 15, 1980, letter (Baynani to Eisenhut), FPC has developed and is presently implementing a program for permanent STA training, based upon the document included in NUREG-0737 (INP0 Guidelines, Rev.0, April 18,1980).
The permanent STAS are expec-ted to replace the present group of interim STAS by December 31, 1981.
By January 31, 1981, FPC will submit the current FPC " Nuclear Operations Technical Advisor Training Program", STA job description, STA qualifica-tion, and STA requalification.
Several options for the long-term STA program will also be discussed in the January 31, 1981, submittal, as specific guidance for long-term STA programs is nonexistent.
Item II.B.2 - Design Review of Plant Shielding and Environmental Quali-fication of Equipment for Spaces / Systems Which May be Used in Post-Accident Operations This Item does not require any analysis in addition to that which has been completed for fiUREG-0578, Section 2.1.6.b, and submitted to you in our letter dated January 1,1980, (Baynard to Denton).
Speci fically, our previous evaluation considered the areas which this Item explicitly says are to be considered for accessibility.
Under NUREG-0578, reactor coolant recirculation via the reactor building sump was assumed to contain 100/50/1 of the core inventory (i.e.,100%
of the noble gases, 50% of the halogens, and 1% of the other isotopes in the core).
This Item reduces this requirement to 0/50/1 by making the assumption that the noble gases are released from the reactor coolant liquid when it is depressurized.
The result of this change is to reduce the dose rates from this source, thereby making the previous evaluation conservative.
j Item II.B.4 - Training for Mitigating Core Damage FPC has developed a training program, based on the INP0 Guidelines,
" Recognizing and Mitigating the Consequences of 52 vere Core Damage".
The initial training will be done for the Crystal River Unit 3 instruc-tional staff by the Babcock & Wilcox Company.
FPC will then write the l
site-specific caterials and implement Phase I (up to Recognition of Core 1
l J
ENCLOSURE 2 Item II.B.4 Training for Mitigating Core Dunge (Continued)
Damage) by June 1, 1981.
The following outline provides the topics to be covered in the Phase One Program.
TOPICAL OUTLINE OF DEGRADED CORE RECOGNITION AND MITIGATION PHASE ONE The following is a topical outline of the topics covered in Phase One of the Degraded Core Recognition and Mitigation Training Program:
I.
Recognition and Mitigation of conditions leading up to core damage.
a.
Core cooling mechanics.
b.
Potentially damaging operating conditions.
c.
Gas / steam binding affecting core cooling.
II.
Inadequate core cooling and core damage.
a.
Consequences of sustained inadequate core cooling.
b.
Progression of core damage.
III. An evaluation of incore and out-of-core nuclear instru-mentation for use as an aid in Recognition of Degraded Core Conditions.
IV.
The use of non-core instrumentation in evaluation of de-graded core.
a.
Response of radiation monitors.
b.
Fission product release and transport.
Item II.K.2.10 - Safety-grade Anticipatory Reactor Trip As requested in your December 20, 1979, letter (Reid to Hancock), and as committed in our December 15, 1980, letter (Baynard to Eisenhut), FPC hereby provides the additional information requested to be submitted prior to your approval of the final design for upgrade of the present control-grade anticipatory reactor trip to safety-grade.
Our responses correspond to the listing given in Attachment 1 of your_ December 20, 1979, letter.
j
ENCLOSURE 2 Item II.K.2.10 - Safety-grade Anticipatory Reactor Trip (Continued)
Item 1:
The final desiga submittal should include the final logic diagrams, electrical schematic diagrams, piping and instrumentation diagraas, and location l'ayout drawings.
Response 1:
Logic drawings attached are:
D 565544AY (4 of 4)
D 3040963LY (3 of 4)
D 8024296LY (2 of 4)
D 804780LDY Schematic drawings attached are:
0 8076417AY D 8076418AY D 8042368KY The above drcoings are for Subassembly "A" only; subassemblics "B",
l "C", and "D" are simil ar.
Instrumentation tubing / location drawings attached are:
l IC-308-861 IC-308-862 IC-308-863 IC-308-864 l
l Item 2:
For sensors located in non-seismic areas which have not previously contained RPS inputs, perform and submit an analysis which shows that the installation (including circuit routing) is designed such that the effects of credible faults (i.e., grounding, shorting, ap-plication of high vol tage, or electranagnetic interference) or failures in these areas could not be propagated back to the RPS and degrade the RPS performance or operability.
Response 2:
The sensors used for signal input to the RPS " Anticipatory Reactor Trip" (ART) will be pressure switches actuated by decreasing oil
ENCLOSURE ^
Item II.K.2.10 - Safety-grade Anticipatory Reactor Trip (Continued) pressure in each turbine trip oil system.
These pressure switches will be mounted in special stands located on - the turbine deck as identified on Sketch Loc.1.
Sensing line routing and floor stand detail s are shown on Drwings IC-308-681, 682, 683, and 684 (attached).
The new RPS inputs are electrically isolated from the RPS by con-tact buffers. As shown on the attached Bailey schematics, the con-tact buffer uses a transformer to power the contact buffer relays.
Thus, the combination of the transformer -and relays electrically isolate the pressure switch circuits from the RPS vital cabinet AC power and the RPS trip string.
The isolation level is up to 500 Volts.
Therefore, the effects of the credible faults such as grounding, shorting, application of high voltage, and electromag-netic interface will not be propagated back into the RPS or degrade the RPS performance.
In addition to the above protection of +.he RPS from the new inputs, i
the input circuits will be channel separ ted, per separation cri-teria, and routed in conduit to guard against credible faults or failures in the non-seismic qualified areas.
Item 3:
Submit " Seismic and Environmental Qualification Summary Reports" for the equipment which has not been previously submitted.
In ad-dition, we require that you demonstrate that the environmental test I
conditions bound the actual worst case accident conditions expected at the installed locations.
Response 3:
The sensors (Pressure Switches) to be used for signal input to the RPS will receive oil pressure signal s from non-safety-rel ated equipment located in a non-seismic structure, the turbine build-ing.
Because of this location, and the monitoring of non-safety-related equipment performance, seismic qualification of the oil pressure monitoring function is not feasible nor is the environmen-tal. qualification to conditions more limiting than that of nomal operation necessary.
The worst case accident conditions which would initiate a turbine trip could not significantly alter. the re-motely located pressure switch environment.
In addition, the need for seismic and environmental qualification is moot because these anticipatory trips are backed up by fully qualified RPS trips.
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ENCLOSURE 2 1
Item II.K.2.10 - Safety-grade Anticipatory Reactor Trip. (Con inued)
However, recognizing the need for highly reliable switches, we are providing ASCO safety-related pressure switches which are water-tight, with an operating range of 6 to 200 psig hydraulic oil.
Qualifica',on by ASCO is currently underway to IEEE 323, 1974, and IEEE 344, 1975.
Completion of this testing is expected by mid-1981.
Item 4:
Assure that the ARTS testability includes provisions to perfom channel functional tests at power. Testing of this curcuitry is to be included in the RPS monthly surveillance tests.
Response 4:
Channel tests capability of the ARTS at power are provided for from the contact buffers to a trip actuation.
This test will be incor-porated into the existing monthly RPS channel tests.
In addition, i
the refueling tests of the RPS will also incorporate ART testing from the pressure switch input to trip actuation.
Item 5:
Include in the final design submittal, the RPS checkout procedure which wil demonstrate both the operability of the new trip circui-try and the continued operability of the previous RPS.
Response 5:
- -1 provides the functional test description.
Item II.K.2.13 - Thermal-Mechanical Report In response to a July 12, 1979, letter from your D.
H.
Ross, to J. H. Taylor, of Babcock & Wilcox, FPC has participated in the prepara-tion of a generic report to address the concern ~ of exceeding the frac-ture mechanics acceptance criteria of the reactor vessel by excessive cooling from high pressure injection flow (without reactor coolant loop flow) during small breaks. A status of this evaluation was presented to members of your staff in a meeting held on May 23, 1980.
FPC has reviewed this generic report from B&W.
Upon resolution of minor technical issues, but' no later than January 31, 1981, FPC will submit this report for your review.
__j
I ENCLOSURE 2 Item III.A.2 - Improved Licensee Emergency Preparedness -- Long-Term 1
As delineated in emergency procedures EM-100, " Emergency Plan" and EM-204, " Dose Assessment by Use of Meteorological Overlays", a procedur-al approach will be taken for timely estimates of whole body dose to personnel downstream of an accidental reiease of radioactive noble gas-es.
To determine the magnitude of the postulated dose, gas release mon-itor readings in conjunction with meteorological overlays will be used.
4 The overlays represent various combinations of meteorological conditions 1
that have been observed at Crystal River Unit 3.
Meteorological meas-urements (i.e., wind speed, wind direction, and temperature difference) are indicated in the Control Roan.
Consistent with the existing technical specification (reference Techni-cal Specification 3/4 3.3.4), the meteorological instrumentation neces-sary to determine projected doses will be maintained operable.
This commitment meets the intent of your alternative to Item (3) and will, therefore, be used until our long-term plans for upgrade of meteorologi-cal assessment and dose projection system can be integrated into on-going measurement, assessment, and communication system.
You will be advised of our long-term plans by April 1,1981.
Item III.D.1.1 - Primary Coolant Outside Containment FPCs letter of February 11,1980, (Baynard to Denton) provided the leak-age test data for the Liquid Waste Disposal System, Waste Gas Disposal System, the Makeup and Purification System, and the High Pressure Injec-tion System.
Leakage data was not provided for the following systems:
Decay Heat Removal System Reactor Building Spray System Liquid Sampling System.
These systems have been leak tested and found to be within the accep-tance criteria as delineated in the above letter._ The Waste Gas Dispos-al System was also retested and found to meet the acceptance criteria.
l l
PYBekcW01(D8-3) l.
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ATTAODIENT 5-1 Page 1 of 4 FIELD FUNCTIONAL TESTING The f ollowing procedure is to assure the proper functioning of the field change f ollowing installation into the existing NI/RPS subasssembly.
A.
Verify subassembly is deenergized.
B.
Perform continuity tests to verify wiring connecting the prewired & tested module mounting case to the NI/RPS subassembly.
C.
Connect jumper leads across the field contact input terminals to simulate closed field contacts.
NOTE:
Clip leads are preferable as later they will be opened to simulate field contact opening and terminals will have approximately 120 volts interrogating voltage on terminals.
D.
Ref erring to the cabinet layout f or the subassembly being changed, reinsert modules into their proper positions in the module mounting case.
E.
Reenergize subassembly and place in manual bypass.
If reactor power is below the bypass setpoint (typically 20%
for f actory checkout), place the subassembly power range channel into test and increase the simulated power to above the bypass setpoint.
Check that the bypass bistables output state lights are dim, and reset the output memory.
Allow thirty minutes warmup of subassembly.
F.
Module Interlock Test:
1.
Reaove the MPP "A" Tripped contact buffer.
Check test trip light on reactor trip module is bright.
Reinsert module and reset reactor trip module.
Check test trip light dim.
2.
Repeat Test F-1 for MFP "B" Tripped contact buffer 3.
Repeat Test F-1 for Main Feed Pump Trip Bypass bi s t a bl e.
4.
Repeat Test F-1 f or Turbine Tripped contact buffer 5.
Repeat tes t F-1 f or Turbine Trip Bypass bistable.
l
ATTACINENT 5-1 page 2 of 4 G.
Test Trip Test.
1.
Depress test switch S-1 on MFP "A" Tripoed contact buffer and hold.
Check DS-1 of f and DS-2 on.
Check test trip lamp bright on reactor trip module.
Release S-1 and check test trip dim on reactor trip modul e.
Depress test switch S-2 on contact buffer.
Check DS-1 on and DS-2 off.
Reset reactor trip module and 2.
Repeat Test G-1 on MFP "B" Tripped contact buffer.
3.
Repeat Test G-1 on Turbine Tripped contact buffer.
H.
Trip Tests:
Prior to beginning the trip tests, if reactor power is below the bypass setpoint (typically 20% f or f actory checkou t), place the subassembly power range channel into test and increase simulated power to above the bypass setpoint.
Check that the bypass bistables output state lights are dim and reset the output memory.
1.
Main Feed Pump Trip a.
Remove the j umper simulating the MFP"A" field contact.
Check that on the MFP"A" Tripped contact buffer DS-1 is off and DS-2 is on.
Verify the channel trip light on the reactor l
trip module remains dim.
Check annunciator and computer output terminals are open f or the MFP"A" Tripped function and closed for i
the MFP Trip function.
l b.
Remove the jumper simulating the MFP"B" field contact.
Check that on the MFP"B" Tripped contact buffer DS-1 is off and DS-2 is on.
Check the channel trip light on the reactor trip module is nou bright.
Check annunciator and computer output terminals are open for both'the MFP"B" Tripped function and the MFP Trip function.
c.
Reconnect jumpers simulating MFP"A" &
"B"'
l field contacts.
Check that the contact buf f ers do not change state.
d.
Reset the MFP"A" & "B" contact buffers by depressing test switch S-2 on the contact buffer modules and verify the output contacts are now closed.
Reset the reactor trip module.
N
ATDK3 BENT 5-1 pqm 3 of 4 2.
Turbine Trip a.
Remove the jumper simulating the Turbine Tripped field contact.
Check that on the Turbine Tripped contact buf f er DS-1 is of f and DS-2 is on.
Check the channel trip light on the reactor trip module is now bright.
Check annunciator and computer output terminals are open for both the Turbine Tripped f unction and the Turbine Trip function.
~b.
Reconnect jumper simulating the Turbine Tripped field contact.
Check that the contact buffer does not change state.
c.
Reset the Turbine Tripped contact buffer by depressing. test switch S-2 on the contact buf f er module and verify the output contacts are now closed.
Reset the reactor trip module.
I.
Bypass Tests:
From previous tests, the subassemblies' power range is either above the bypass set point or simulated above the bypass setpoint.
The output state lamps on the two bypass bistables should be dim.
1.
Verif y the annunciator and computer output terminals are closed for both the MPP Trip bypass and Turbine Trip Bypass f unctions.
2.
If power range is not in test, place in test and reduce simulated power below the bypass setpoint.
Check that the output state lamps on the bypass bistables are bright.
Check l
the annunciator and computer output terminals are open for both the MFP Trip Bypass and I
Turbine Trip Bypass f unctions.
i I
3.
Depress the tes t swi tches f or. both MFP"A" and l
MFP"B" Tripped contact buffers.
Check that the channel trip lamp on the reactor trip l
module remains dim.
Check the annunciator and computer output terminals are still closed for the,MFP Trip f unction.
4.
Reset the MPP"A" &
"B" Tripped contact buf f ers by depressing test switch S-2 on the contact buf f er modules.
ATTAQBETF 5-1 3
PmLm 4 of 4 5.
Repeat steps I-3 and I-4 for the Turbine Trip function.
J.
Remove power f rom the subassembly.
K.
Remove the jumpers simulating field contacts.
L.
Referring to the field change supplied external connection drawing for the subassembly being modified, connect field wiring to designated terminals.
M.
Return subassembly to normal operation using standard site procedures.
N.
It is recommended that the normal site functional tests be performed on the subassembly to verify the field change did not ef f ect the balance of the subassemblies functions.
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