ML20214L974
| ML20214L974 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 09/19/1986 |
| From: | Kingsley O MISSISSIPPI POWER & LIGHT CO. |
| To: | Munro J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML20214L754 | List: |
| References | |
| AECM-86-0297, AECM-86-297, NUDOCS 8612030241 | |
| Download: ML20214L974 (35) | |
Text
ENCLOSURE 3 A
MISSISSIPPI POWER & LIGHT COMPANY Helping Build Mississippi EdinEMidd5 P. O. B O X 164 0. J A C K S O N. MIS SIS SIP PI 39215-1640 September 19, 1986 O. D. KINGSLE Y, J R.
VICE PRES: DENT NUCLE AR OPERATIOpel U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, N.W., Suite 2900 Atlanta, Georgia 30323 Attention: Mr. John F. Munro, Operator Licensing Section
Dear Mr. Munro:
S'JBJECT: Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 Comments Regarding NRC Examinations s
AECM-86/0297 On September 8, 1986, NRC examiners gave written examinations to eleven (11)
Grand Gulf Nuclear Station (GGNS) Unit 1 candidates for NRC Operator and Senior Operator Licenses.
Upon completion of the examination, a copy of the examinations was supplied to Mississippi Power & Light with the request that the exam be reviewed and any appropriate comments be furnished to the NRC.
1 The attached coments are supplied pursuant to that request.
Should you require additional information, please contact myself or l
Mr. W. M. Shelley of GGNS Plant Staff.
Yours t uly, s-
,U ODK:vog Attachments f
l cc:
(See Next Page) 1 8612030241 861114 PDR ADOCK 05000416 V
PDR J13AECM86091901 - 1 Member Middle South Utilities System
t AECM-86/0297 Page 2 cc: Mr. T. H. Cloninger (w/a)
Mr. R. B. McGehee (w/a)
Mr. N. S. Reynolds (w/a)
Mr. H. L. Thomas (w/0)
Mr. R. C. Butcher (w/a)
Dr. J. Nelson Grace (w/o)
U. S. Nuclear Regulatory Comission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr. James M. Taylor, Director (w/a)
Office of Inspection & Enforcement U. S. Nuclear Regulatory Comission Washington, D.C.
20555 Mike Spencer (w/a)
Idaho National Engineering Laboratory P. O. Box 1625 Idaho Falls, Idaho 83415 Rich Miller (w/a)
Sonalysts, Inc.
P. O. Box 280 215 Parkway, North Waterford, Connecticut 06385 J13AECM86091901 - 2
s Attachment I to AECM-86/0297
" Comments Regarding NRC Operator Examinations" I.
REACTOR OPERATOR EXAMINATION Question 1.09 - The question asked the examinee to select the correct answer concerning rod motion, over a several hour period, four hours after the reactor was returned to criticality, following a scram from full power with equilbrium xenon conditions.
The answer key indicates item A as the correct answer. This is true if:
0 the several hour period does not take you past the xenon peak, and o
the increased burnout of xenon due to the reactor being critical is not taken into account.
If the several hour period takes you past the peak of xenon concentration, xenon would be adding positive reactivity as its concentration decreased.
In this case item B would be the correct answer.
We recommend that either ite-A or B be accepted as the correct answer since the length of 'several hours" is not clearly defined by the question.
(Reference GE BWR Academics, Chapter 6)
Question 2.02 - The answer key shows the fail position of the TBCW Makeup Valve as F0 (fail open). The actual fail position of this valve is FC (fail closed). Off Nonnal Event Procedure 05-1-02-V-9 revision 19 page 4 is enclosed as Attachment II for reference. We recommend that the answer key be changed to reflect FC for this valve.
Question 2.05 - Item D of the answer key indicates 3.0 psig as the correct answer. Technical Specification 5.2.2.c.2 states 3.0 psid and is enclosed as Attachment III for reference. We recommend the answer key be changed to reflect this answer.
Question 2.08 - The answer key for part A of the question does not include the following as possible correct answers:
o generator differential current o
generator ground overcurrent These two items will cause the generator lockout stated in the answer key. We recommend the above items be added as possible correct answers for part A of this question.
J13AECM86091901 - 4
)
- s
, to AECM-86/0297 The answer key for part B of the question does not include the following as possible correct answers:
o loss of excitation o
reverse power o
bus underfrequency o
generator overcurrent with voltage restraint We recommend the above items be accepted as correct answers for part B of this question.
Pages 3/4 8-5 and 3/4 8-7 of technical specification 3/4.8.1 is enclosed as Attachment IV for reference. Section 4.8.1.1.2.d.8 is reference for part A and 4.8.1.1.2.d.16 is referenced for part B.
Question 2.09 - Part C of the answer key states the difference between ATWS trips and RPT trips for recirculation pumps in a word discrip-tion. The following should also be considered as a correct answer:
o ATWS trips breakers CB-2 and CB-5 o
RPT trips breakers CB-3 and CB-4 resulting in high to low speed transfer We recommend the above be accepted as a correct answer for part C of this question.
Pages 29 and 44 of Lesson Plan OP-B33-1-501/Rev. 2, Recirculation System, is enclosed as Attachment V.
Question 2.10 - Item A of the question asked if HPCS SSW can be manually initiated and if so, how. The answer key states that it can not be manually initiated. The SSW System Operating Instruction, 04-1-01-P41-1, does provide instructions for manual initiation /startup of the HPCS SSW system. We recommend that a yes answer followed by the procedural steps given in the S0I in section 4.4.2 be accepted as a correct answer.
Section 4.4 of the SSW SOI is enclosed as Attachment VI for reference.
Question 2.16 - Part B of this question requires the operator to memorize valve numbers in order to answer the question. Memorization of valve numbers is not required by the GGNS job analysis for licensed or non-licensed operators. We recommend that part B of this question be deleted from the examination.
Question 3.03 - The answer key for part 3 of the question is correct if the effects of the rod pattern controller are not considered.
Considering the effects of the rod pattern controller (RPC), the answer is N0. With groups 1 and 2 fully withdrawn a single group 3 rod can not be fully withdrawn. The RPC will allow withdrawal of l
J13AECM86091901 - 5
~.
. c
, to AECM-86/0297 the group 3 rod to position 04 and then initiate a rod block. All group 3 rods will have to be withdrawn to position 04 before the RPC will allow withdrawal past position 04. This is also true for positions 08 and 12. We recommend the answer key be changed to reflect N0 for part 3 of this question. Table 3 of the Rod Control and Information System (RCIS) System Operating Instruction, 04-1-01-C11-2, is enclosed as Attachment VII for reference.
Question 3.06 - The answer key for this question is incomplete. The Control Rod Drive (CRD) system has components that can be operated from the Remote Shutdown Panels. We recommend that the CRD system be added as a possible correct answer for this question. Page 7 of the CRD System lesson plan is enclosed as Attachment VIII for reference.
Question 3.09 - The question requires the operator to list three (3) actions which will result in the Commands Disagree Lamp being lit.
The backpanel pushbuttons referenced in the answer key are for Instrumentation and Control personnel troubleshooting and testing and have no significance in the operation of the Rod Control and Information System (RCIS). The RCIS S0I was revised at which time the information concerning the pushbuttons was removed due to its lack of operational significance.
Pages 1 and 2 of 04-1-01-C11-2 of the RCIS I
S0I are enclosed as Attachment IX for reference. We recommend this question be deleted.
Ouestion 3.10 - This question requires the operator to have the Alarm Fesponse Instructions (ARI) memorized to answer the question. Conduct of Operations procedure, 01-S-06-2, sections 6.3.5 and 6.3.6 contain GGNS management's delineation of what procedures are required to be committed to memory.
The ARI's are not included in this list. We recommend that the answer key be expanded to accept answers concerned with decreasing or stopping the cooldown rate through adjustment of throttle valves in the RHR shutdown cooling lineup. The above referenced sections of 01-S-06-2 are enclosed as Attachment X.
[
Question 3.13 - The answer to part D of this question is given assuming a particular range of a multiple range instrument. We recommend that the answer key be expanded to accept correct meter / recorder indications for ranges other than the assumed range.
Question 4.03 - The question uses a reference to OP-IP-503 which is a lesson plan used in operator training and asks the operator to state when certain actions are performed. These actions are performed as governed by our Integrated Operating Instructions, which should have been the reference. We recommend that the wording of this question be changed prior to next use to reflect the proper procedure vice a lesson plan number as the reference.
I J13AECM86091901 - 6
i Attachment I to AECM-86/0297 Question 4.08 - The answer key gives seven (7) conditions for when an RWP is required.
Items f and g contain two separate cases for required RWP's. We recommend the answer key be expanded to the following:
f.
entrance into or work in a high radiation area g.
entrance into or work in a very high radiation area h.
entrance into or work in an airborne area 1.
entrance into or work in a potentially airborne area AI Page 25 of the Exposure and Contamination Control Procedure, 01-S-08-2, is enclosed as Attachment XI for reference.
t F
J13AECM86091901 - 7
. to AECM-86/0297 II. SENIOR REACTOR OPERATOR EXAM Question 5.02 - Source neutrons can be produced through cosmic ray interaction with core materials. The answer key does not state this as a valid answer. We recomend that cosmic rays be added to the answer key as a possible correct answer. See GE, BWR Reactor Theory, Chapter 2 for reference.
Question 5.03 - Using the equation provided on the equation sheet attached to the exam, the following answer would result:
K SDM = (1 Keff)/ eff Therefore... Keff = 1/(1 + SDM) = 1/(.3 + 1) =.77 f=
i.(1 effi) = 345 cps K
1 eff(f)
The answer key has 450 cps as the correct response. The difference is due to the value of K calculated in the answer key. The answer key eff of.7 and when used in the count rate formula solution produced a K resultsinthe450cp$hivenastheanswer.
f We recommend that either the 450 cps or the 345 cps be accepted as correct answers when the appropriate mathematical methods are shown j
on the candidates' answer page.
Question 5.11 - The Figure 1 supplied with the exam is not a BWR-6 power to flow map. Although this figure is from the GE Fundamentals material used at GGNS, we recomend that future exams use a BWR-6 specific figure.
Question 5.15 - The answer given in the answer key for item 4 is incorrect.
The point of CHF should be E on the diagram provided. We recomend that the answer for item 4 be changed to point E.
See GE, Heat Transfer & Fluid Flow, Chapter 8.
Question 5.20 - The answer key gives calculated values for the three requested answers with no margin for error, i.e.: +/- 0.
The values should be:
a) 1800 = 2000 X = 1333.3 gpm T2UD X
b)
(1800)2 = 1000 X = 444.44 psig (1200)2 X
c)
(1800)3 = 150 X = 44.44 Hp (1200)3 X
We recomend that there be an allowable +/- be added to the answer key.
J13AECM86091901 - 8
__________________________________________________J
- to AECM-86/0297 4
Question 6.01 - The answer given in the answer key is that the startup level control valve is in parallel with the # 5 and # 6 feedwater heaters. This is true when the general layout of the possible system flow paths is considered. When the valve is in use, the startup level control valve is in series with the # 5 and # 6 feedwater heaters.
Since there was no system lineup or plant conditions given in the question, we recommend that either item C or B be considered as correct responses to this question. Paga 39 through 41 of procedure 03-1-01-1, Cold Shutdown to Generator Carrying Minimum Load, is enclosed as Attachment XII for reference.
Question 6.07 - The answer key states that the APRM downscale rod block setpoint is 5%. Technical specification Table 3.3.6-2 shows a value of 4%. Also, the flow input to the flow biased network will provide a rod block at 108% (upscale). This is considered an APRM rod block (see lesson plan OP-C51-501). We reconnend that 4% be accepted as the APRM downscale rod block, and that the flow upscale high of 108% be added to the answer key. Technical specification table 3.3.6-2 is enclosed as Attachment XIII for reference.
Question 6.12 - Part B of the question asked what provides minimum flow for the LPCS system. The answer key states that the restricting orifice in the minimum flow line provides minimum flow for the system at 700 gpm. Due to the wording of the question, we recommend that the minimum flow valve operating at 1250 gpm be accepted as a correct response. Page 11 of lesson plan OP-E21-501 is enclosed as Attachment XIV for reference.
Question 6.13 - The answer key lists all possible correct answers except:
HPCS diesel generator output breaker control power failure o
We reconnend that this be added to the answer key as a possible correct answer. 04-1-02-1H13-P601-16A-H1, Alarm Response Instruction, is enclosed as Attachment XV for reference.
l Question 6.15 - Same coment as Question 2.16 Question 7.11 - The referenced procedure was included due to NRC's standard request for Technical Section procedures. This particular l
procedure is not taught to operators as it was not supported by the job analysis perfonned in conjunction with our accreditation effort.
We recommend that any of the criticality rules associated with this procedure be accepted as correct answers vice only those listed as
" General Rules", and that questions concerning this procedure no longer be used in future examinations. 09-S-02-100, page 3, Criticality Rules, is enclosed as Attachment XVI for reference.
J13AECM86091901 - 9
' to AECM-86/0297 Question 8.01 - The question required the examinee to list three (3)
LCO's associated with SRM's during refueling. Technical specifications only has one LC0 for SRM's during refueling which lists three (3) conditions that must be met in order to consider the SRM's operable. We recommend that the question be reworded such that it will not confuse or mislead the examinee as to the desired answer during future use of the question.
Question 8.13 - The answer key gives a list of events that require one hour notification. There is no reference listed in the answer key as to where the list of events was obtained. We recommend that the list of events in 01-S-06-5 be added to the answer key as possible correct answers. 01-2-06-5 Attachment III is enclosed as Attachment XVII for reference.
1 J13AECM86091901 - 10
GRAND GULF NUCLEAR STATION OFF-NOR'!AL EVENT PROCEDURE l
Title:
Loss of Instrument Air lNo.:
l Revision:
19lPage: 4 l
l l 05-1-02-V-9 l l
l
(~')
ATTACHMENT II
\\s /
3.33 The makeup valve to the TBCW surge tank FAILS CLOSED. No TBCW cooling is available since PSW is lost. All TBCW temperature control valves FAIL OPEN. TBCW pump minimum flow valve FAILS CLOSED.
3.34 The makeup valve to the CCW surge tank and the CCW pump minimum flow valve FAIL CLOSED. No CCW cooling is available since PSW is Lost.
3.35 Plant Chilled Water is lost as the chillers trip due to loss of PSW.
Makeup to the chilled water expansion tank FAILS CLOSED.
3.36 Drywell Chilled Water is lost as the chillers trip due to loss of PSW.
Makeup to the chilled water expansion tank FAILS CLOSED.
3.37 Suppression Pool Cleanup is Lost.
3.38 The Drywell, Containment, and the Auxiliary Building Floor, Equipment, and Chemical drains are isolated. Also, the Condensate Clean Waste Pump is isolated on the suction and discharge, preventing pumping from the Condensate Clean Waste Tank.
3.39 Standby Liquid Control Storage Tank level indication is Lost.
3.40 All caustic transfer from the Makeup Water Treatment System is Lost.
Ok,)
3.41 Acid transfer from the Makeup Water Treatment System to SSW, Condensate Cleanup, and Circulating Water System is lost.
3.42 The mechanical vacuum pump cannot be started, but will not trip if already operating. This is because the air operated valves fail as-is.
3.43 Makeup to-the Demineralizer Water Storage tank is lost.
3.44 The Chlorination System is lost.
3.45 Liquid and Solid Radwaste systems are lost.
3.46 The Lube Oil Conditioning system is lost.
3.47 Loss of Auxiliary Steam.
d.0 IMMEDIATE OPERATOR ACTIONS 4.1 Start any available Air Compressor.
4.2 If Service Air is available:
l 4.2.1 Verify the Automatic Cross-tie valve is maintaining approximdtely 90 psig in the Instrument Air header.
O 05-1-02-V-9 TEXT
ATTACHMENT III 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.
LOW POPULATION ZONE 5.1.2 The low population zone shall be as'shown in Figure 5.1.2-1.
UNRESTRICTED AREA AND SITE B0UNDARY FOR GASEOUS EFFLUENTS AND FOR LIQUID EFFLUENTS 5.1.3 The UNRESTRICTED AREA AND SITE BOUNDARY for gaseous offluents and for liquid effluents shall be as shown in Figure 5.1.3-1.
The gaseous effluent release points are shown in Figure 5.1.1-1.
5.2 CONTAINMENT CONFIGURATION 5.2.1 The containment is a steel lined, reinforced concrete structure composed of a vertical right cylinder and a hemispherical dome.
Inside and at the bottom of the containment is a reinforced concrete drywell composed of a vertical right cylinder and a steel head which contains an approximately eighteen to nineteen foot deep water filled suppression pool connected to the drywell through a series of horizontal vents.
The containment has a minimum net free air volume of 1,400,000 cubic feet.
The drywell has a minimum net free air volume of 270,000 cubic feet.
DESIGN TEMPERATURE AND PRESSURE l
5.2.2 The containment and drywell are designed and shall be maintained for:
a.
Maximum internal pressure:
1.
Drywell 30 psig.
2.
Containment 15 psig.
b.
Maximum internal temperature:
1.
Drywell 330 F.
2.
Suppression pool 185 F.
c.
Maximum external-to-internal differential pressure:
1.
Drywell 21 psid.
2.
Containment 3 psid.
SECONDARY CONTAINMENT 5.2.3 The secondary containment consists of the Auxiliary Building and the Enclosure Building, and has a minimum free volume of 3,640,000 cubic feet.
GRAND GULF-UNIT 1 5-1 l
n m
Page 1 of 2 ATTACINENT IV ELECTRICAL POWER SYSTEMS
.(
SURVEILLANCE REQUIREMENTS (Continued) 5.
Verifying that on an ECCS actuation test signal, without loss of offsite power, the diesel generator starts on the auto-start signal and operates on standby for greater than or equal to 5 minutes.
The generator voltage and frequency shall be 4160 1 416 volts and 60 1 1.2 Hz within 10 seconds after the auto-start signal; the steady state generator voltage and frequency shall be maintained within these limits during this test.
6.
[0ELETED]
7.
Simulating a loss of offsite power in conjunction with an ECCS actuation test signal, and:
a)
For Division 1 and 2:
1)
Verifying deenergization of the emergency busses and load shedding from the emergency busses.
- 2). Verifying the diesel generator starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 10 seconds, energizes the auto-connected shutdown loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.
After energization, the steady state voltage and frequency of the emergency busses shall be maintained
(
at 4160 1 416 volts and 60 1 1.2 Hz during this test.
b)
For Division 3:
1)
Ver_ifying de energization of the emergency bus.
2)
Verifying the diesel generator starts on the auto-start signal, energizes the emergency bus with the permanently connected loads within 10 seconds and the autoconnected emergency loads within 20 seconds and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads. After energization, the steady state voltage and frequency of the emergency bus shall be maintained at 4160 1 416 volts and 60 1.2 4z during this test.
, m 8.
Verifying that all automatic diesel generator trips are automatically bypassed upon an ECCS actuation signal except:
a)
For Divisions 1 and 2, engine overspeed, generator differential current, low lube oil pressure, and generator L
around overcurrent.
b)
For Division 3, engine overspeed and generator differential _
current.
GRAND GULF-UNIT 1 3/4 8-5
Page 2 of 2 ATTACHMENT IV ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 16.
jerifying that the following diesel generator lockout features
' prevent diesel generator starting and/or trip the diesel generator only when required:
a)
Generator loss of excitation.
b)
Generator reverse power.
c)
High jacket water temperature.
d)
Generator overcurrent with voltage restraint.
j e)
Bus underfrequency (11 and 12 only).
f)
Engine bearing temperature high (11 and 12 only).
g)
Low turbo charger oil pressure (11 and 12 only).
h)
High vibration (11 and 12 only).
1)
High lube oil temperature (11 and 12 only).
j)
Low lube oil pressure (13 only).
L k)
High crankcase pressure.
e.
At least once per 10 years or after any modifications which could affect diesel generator interdependence by starting all three diesel generators simultaneously, during shutdown, and verifying that the three diesel generators accelerate to at least 441 rpm for diesel generators 11 and 12 and 882 rpm for diesel generator 13 in less than or equal to 10 seconds.
(
f.
At least once per 10 years by:
1.
Oraining each fuel oil storage tank, removing the accumulated sediment and cleaning the tank using a sodium hypochlorite or equivalent solution, and 2.
Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME Code in accordance with ASME Code Section XI, Article IWD-5000.
4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2 within 30 days.
Reports of diesel generator failures shall include the information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1 August 1977.
If the number of failures in the last 100 valid tests, on a per nuclear unit basis, is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.
b.b..3 For Tech Spec See TSPS EO2l GRAND GULF-UNIT 1 3/4 8-7
Pagn 1 of 2 OP-833-1-501/Rev. 2 ATTACHMENT V Page 29 of 96 OBJ.
TOPIC OF INSTRUCTION LEARNING ACTIVITY
- 1) Less than 22% feed flow for 15 secs.
(called the " Total FW Low Flow Interlock")
- 2) Reactor level less than 11.4".
- 3) Less than 7.4 F steam dome to recire loop differential temperatures.
Sb d.
E0C RPT signal from reactor Show Figure 13 protection system; during TCV closure most limiting pressurization deenergizes K48 transients at end of cycle, relays causing (Turbine Stop Valve Closure CB3+4 trip coils or Control Valve Fast to energize and Closure with no bypass) the picks up K47A and collapse of voids in the B relays causing core can add positive a transfer to reactivity faster than control slow speed.
rods can insert negative Associated white reactivity on a SCRAM lights go "out",
(for first few seconds) red lights come causing recirc pump transfer "on" on the induces voids in the core respective back to reduce the void collapse
- panel, and recovers loss of thermal margin from safety limits; signals are:
- 1) Turbine stop valve TSV closure closure ( 40 psig) deenergizes (oneoutoftwologic)
K50 relays causing identical actions except different sets of white lights go "out".
- 2) Turbine control valve closure ( 44.3 psig)
(2 out of 2 logic)
- 3) Bypassed less than 40% rated thermal power.
e.
EOC RPT causes CB3A/B and CB4A/B to open and normal transfer to slow speed.
L J
NTROLP - OP-B33-1-501
Page 2 of 2 OP-833-1-501/Rev. 2 ATTACHMENT V Page 44 of 96 OBJ.
TOPIC OF INSTRUCTION LEARNING ACTIVITY Sa 1.
ATWS RPT - causes CB-5 and Show Figure 24 CB-2 to trip on 1125 psig Note: ATWS trip Reactor Dome Pressure.
-41.6 logic can be inches Reactor Level.
bypassed with test switches on Leak a.
Limits the initial pressure Det. Cabinets Div.
rise and reduces Rx thermal 1 and 2.
power if an ATWS event occurs.
b.
ATWS = Anticipated Transient Any one signal in Without Scram.
the "A" or "C" logic will cause c.
Multiple failures have CB-2A and SA to prevented RPS from actuating.
trip.
Similarly any one signal in the "B" or "D" logic will cause CB2B and SB to trip.
Note:
it is possible if only one signal came in fro:n some inadvertent source that only one pump will trip.
2.
Suction valve less than 90% open.
3.
Discharge valve less than 90% open.
l 4.
CB-5 Handswitch to "STOP" or "STOP LOCK".
5.
Pump motor lockout relay activated.
t 6.
CB-3 opens.
7.
CB-4 opens.
f 8.
Incomplete sequence relay activates.
9.
Loss of control power during a start or transfer sequence.
6c H.
Low Speed Pump Trip to Off - Caused by:
Se l
l NTROLP - OP-B33-1-501
' GRAND GULF NUCLEAR STATION SYSTEM OPERATING INSTRUCTION I
Title:
Standby Service Water lNo.:
04-1-01-P41-1l Revision: 27lPage:
8 l
l System l
l l
l ATTACHMENT VI Page 1 of 2 4.3.2 iCont'd)
(2) SSW INBD OUTL FM DRWL PURGE COMPR B Valve F168B.
(3) SSW OTBD OUTL FM DRWL PURGE COMPR B Valve F160B.
g.
To cool Fuel Pool Cooling Heat Exchanger B from SSW B, refer to SOI 04-1-01-P42-1.
h.
To supply SSW B to CCW Heat Exchangers and the Drywell chillers perform the following:
1 I
I CAUTION l
1 I
I The following step will trip the plant chillers, if in l
1 operation.
l l
l
,(1) Close the PSW to Centrifugal Chillers Valve IP44-F115 (local Handswitch IP44 HS-M004).
(2) Open the SSW RTN FM DRWL CLRS/CCW HXS Valve IP44-F067.
(3) Close the PSW RTN HDR FM CTMT Valve IP44-F068 [10C].
(4) Close the PSW From CCW HX Valve IP44-F011 (local Handswitch IP44 HS-M000).
(5) Open the SSW OTBD SPLY TO DRWL CLRS/CCW HXS Valve IP44-F042.
(6) Open the SSW INBD SPLY TO DRWL CLR/CCW HXS Valve IP44-F054.
i.
To supply water to RHR H X "B", open the SSW INL TO RHR HX "B" P41-F014B and the SSW OUTLET FROM RHR H X "B" P41-F068B, and the P41-F006B, SSW PMP B RECIRC VLV, should be closed to prevent excessive system flow.
4.4 Manual Start of SSW Loop C 4.4.1 Prerequisites a.
SSW Loop C is in standby per Section 4.1.
l 04-1-01-P41-1 TEXT
.~
GRAND GULF NUCLEAR STATION SYSTEM OPERATING INSTRUCTION l
Title:
Stradby S2rvica Wster lNs.: 04-1-01-P41-1lR: vision: 27lPage: 9 l
l System l
l l
l ATTACHMENT VI Page 2 of 2 I
4.4.2
' Instructions l
l l
NOTE l
l l
l All operations are performed oh Panel 1H13-P870, Section 5, unless l
l otherwise noted. Panel sections are denoted by [].
l l
l a.
Start the HPCS SVC WTR PMP and immediately open the SSW LOOP C RTN TO CLG TWR A Valve F011C.
b.
Verify SSW LOOP C FLO of about 880 gpm and PRESS indicates about 80 psig.
4.5 SSW Loop A Return to Standby 4.5.1 Prerequisites a.
The following automatic initiation signals for SSW Loop A are cleared or reset:
(1) RHR pump running (2) LPCS pump running (3) RCIC turbine steam supply valve not closed (4) Reactor water level low (-150 in.)
(5) High drywell pressure (1.39 psig)
(6) Loss of offsite powr.r (7) Diesel Generator 11 running (8) Containment Spray "A" initiation.
l l
l NOTE l
l l
l LSS panel must be reset before SSW can be placed in standby following l l a LOCA or Loss of Offsite power.
l l
l 04-1-01-P41-1 TEXT
GRAND GULF NUCLEAR STATION SYSTEM OPERATING INSTRUCTION l
Title:
Rod Control and l No.:
04-1-01-C11-2 l Revision:
16 l Page: 33 l
l Information System l l
l l
ATTACHMENT VII Table 3 Rod Sequence Withdrau I.
Group 1 through 4 to 50% density A.
Any one of 1st four groups can be withdrawn.
1.
If Group 1 or 2 is first then Group 2 or 1 is' next.
2.
If Group 3 or 4 is first then Group 4 or 3 is next.
B.
The first two groups (either 1 & 2 or 3 & 4) can be fully withdrawn continuously one group at a time in either single or gang mode.
1.
The first selected group must be fully withdrawn before the next group can start.
I C.
The next two groups (one group at a time) must be banked out (i.e., all rods move to one bank before proceeding to next bank).
1.
Bank positions are 4, 8, 12 for initial.startup, and will change as fuel depletes.
gg-2.
Rods must be notched out to position 12 then continuous (if desired) to full out.
3.
First group must be full out before next group can be withdrawn.
J 4.
Either single or gang mode.
m II.
Group 5 through 10 A.
Group 5 or 6 can be notch withdrawn to position 12 then notched to the required position.
1.
If Group 5 or 6 is first then Group 6 or 5 is next.
B.
Groups 7 and 8 or 9 and 10 can then be withdrawn.
1.
If Group 7 and 8 is first then Group 9 and 10 is next.
If Group 9 and 10 is first then Group 7 and 8 is next.
2.
Must be notch withdrawn to bank positions 4, 8, 12 then notched to the required position.
3.
Either single or gang mode.
04-1-01-C11-2 TEXT l
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ATTACHMENT Vill OP-C11-1A-501/Rav. 2 Page 7 of 31 08J.
TOPIC OF INSTRUCTION LEARNING ACTIVITY
- 4) Main pump delivers 10-12 psig.
- 5) Oil system cooled by 25 gpm (CCW) 3e e.
A pump controlled from Control Room panel P601 or Remote Shutdown Panel P150 (111' elevation, control building) Remote Shutdown Panel.
f.
B pumps controlled from P601 or Panel P151 (111' elevation, control building).
~
3c g.
Power supplies
- 3) Aux oil pump - ESF 120 VAC.
E.
Drive Water Filters 3a 1.
Remove impurities to protect CRD.
3b 2.
Two 100% filters a.
Manually cleaned b.
Location - 93' elevation; area 9, auxiliary building.
F.
Flow Control Station show Figure 3 3a 1.
Maintains approximately 68 g;,m flow.
2.
Description.
a.
Two air flow control valves F002A l
and B in parallel with one normally in use.
5 b.
Valve controlled manually or automatically by flow controller on l
P601.
l O
NTROLP - OP/C11-1A/501
. - w.
. -4.,
tren -
. GRAND GULF NUCLEAR STATION SYSTEM OPERATING INSTRUCTION l
Title:
Rod Control and l No.:
04-1-01-C11-2 l
Revision:
16 i Page: 1 l
l Information System l l
l l
ATTACHMENTIX Page l of 2 1.0 PURPOSE To provide instructions for the operation of the Rod Control and Information System (RC&IS).
TITLE PAGE 4.0 NORMAL OPERATIONS
~
4.1 RC&IS Startup 2
4.2 RC&IS Shutdown 3
4.3 RC&IS Rod / Rod Gang Notch Out 3
4.4 RC&IS Rod / Rod Gang Notch In 5
4.5 RC&IS Continuous Rod / Rod Gang Withdrawal 7
4.6 RC&IS Continuous Rod / Rod Gang Insertion 10 4.7 RC&IS Display Selection 12 4.8 RC&IS Mode Selection 15 4.9 RC&IS Rod Motion Selection 16 4.10 RC&IS Rod Pattern Control Selection 17 4.11 RC&IS Substitute Data 18 5.0 INFREQUENT OPERATIONS 5.1 RC&IS Rod Position Bypass and Control Rod Movement 19 5.2 Reactor Mode Switch Rod Block Verification 21 6.0 ABNORMAL OPERATIONS 6.1 RC&IS Rod Drive Bypass 22 2.0 ATTACHMENTS 2.1 Attachment I - None 2.2 Attachment II - None 2.3 Attachment III' Electrical Lineup Checksheet 2.4 Attachment IV - System Alarm Index 3.0 PRECAUTIONS AND LIMITATION 3.1 The RC&IS hardware will allow up to 8 control rods to have their position signals bypassed, via toggle switches in Rod Action Control Cabinets.
3.2 Care should be exercised when utilizing the position signal bypassed function to ensure the Rod Pattern Controller (RPC) constraints are not violated since the RPC does not affect' motion of the bypassed rods.
04-1-01-C11-2 TEXT
...m..,--
GRAND GULF NUCLEAR STATION SYSTEM OPERATING INSTRUCTION lIitle: Rod Control and l No.:
04-1-01-C11-2 l Revision:
16 l Page: 2 l
l, Information System ' l l
l l
ATTACHMENT IX Page 2 of 2 i
3.3 When reducing power toward the Low Power Setpoint (LPSP) the control rod pattern should be verified prior to, or at, the Low Power Alarm Point (LPAP) to ensure that all control ro,ds are within limits as required by the RPC.
If not, insert and/or withdraw blocks will be applied upon reaching the LPSP.
3.4 Each time a control rod is withdrawn to Notch position 48 (full-out), a coupling check is to be performed. Refer to Tech. Spec.
3.1.3.4/4.1.3.4.
3.5 Rod withdrawal is permitted only if both Rod Block Logic Circuits 1 and 2 are clear of all rod blocks.
3.6 A RC&IS failure that interrupts the dynamic 1 and 2 Signals (agree at all times) transmitted to the HCU's will prevent further control rod motion.
3.7 Failures consisting of short circuits, open circuits, loss of circuit continuity, loss of power or cards and instruments out of file will inhibit rod movement.
3.8 If a rod block signal is received during a rod withdrawal, the control rod is automatically stopped at the next notch position, even during a continuous rod withdrawal, jy-3.9 A control rod that has been Drive bypassed in the Rod Gang Drive System (RGDS) cannot be moved as long as the bypass is in effect.
4.0 NORMAL OPERATIONS 4.1 RC&IS Startup 4.1.1 Prerequisites a.
Electrical Lineup Checksheet (Attachment III) is completed.
4.1.2 Instructions a.
Depress the RESET pushbutton in the RGDS STATUS section on Panel 1H13-P653. This will sync the clocks.
b.
Verify that all FULL IN, green LED's on the Rod Display Module (RDM) 1H13-P680 are on for all control rods.
04-1-01-C11-2 TEXT
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mI
- GRAND GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE 5,
l
Title:
Conduct of Operations lNo.:
01-S-06-2 l Revision:.17 l Page: 22 l
l l
1 l
1 ATTACHMENT X 6.3.4 Normally, the Operator-at-the-Controls will be responsible for the 1H13-P680 panel and the Control Room Operator or other N0A will be responsible for the IH13-P601 and/or 1H13-P870 initially. Different assi'gnments will be made by the Shift Supervisor as necessary.
')
6.3.5 System Operating Instructions (SOI's), Alarm Response Instructions (ARI's), Off Normal Event Procedures (ONEP's), and Emergency Procedures (EP's), which are maintained in the Control Room, will be used by the Operations shift personnel to operate equipment and systems to return or maintain them and/or the plant in a safe condition.
6.3.6 To ensure that emergency operations are initiated in a timely manner, all Licensed Operators shall be able to implement the j
immediate actions specified in the ONEP's and the entry conditions for EP-1 (05-S-01-EP-1), EP-3 (05-S-01-EP-3), and L
EP-10 (05-S-01-EP-10), without reference.
6.4 Operator License Requirements 6.4.1 The GGNS General Manager; Manager, Plant Operations; Manager, df Plant Maintenance; and Manager, Plant Support shall have acquired, at the time of appointment to the position, the experience and equivalent training normally required to be eligible for a Senior Reactor Operator License, whether or not the examination is taken.
6.4.2 The Operations Superintendent has overall responsibility for directing the licensed activities of all licensed operators and, therefore, shall be licensed as a Senior Reactor Operator.
6.4.3 The Shift Superintendents and the Shift Supervisors are i
responsible for directing the licensed activities of licensed operators and, therefore, shall also be licensed as Senior l
Reactor Operators.
6.4.4 Operators, who are responsible for performing licensed activities, are required to be licensed as Reactor Operators.
i l
l l
01-S-06-2 TEXT
^
GULF NUCLEAR STATION ADMINISTRAT2VE PROCEDURE lIitle: Exposure and Contamination lNo.:
01-S-08-2 l Revision: 10 lPage: 25 l
j Control l
l l
l
2.5 mrem / hour).
I c.
When work is to be done involving the opening of potentially highly contaminated systems.
d.
Jobs requiring man-rem expenditures >.1 rem.
e.
For any radiation or contamination-related job at the discretion of the Health Physics Section.
.I 01-S-08-2 TEXT
GRAND GUIE NUCLEAR STATION INTEGRATED OPERATING INSTRUCTIOM l
Title:
Cold Shutdown To Generator lNo.:
03-1-01-1 l Revision: 34 lPage: 39 l
l Carrying Minimum Load l
l l
l ATTACilliENT XII Page 1 of 3 Control Room Operator Initials /Date l
l l
NOTE l
I I
l Rejection rate from the vessel via the RWCU l l system may have to be increased as heatup
-l l progresses and swell causes RPV water level l l to increase.
l l
l 6.2.5 Have Chemistry verify that the feedwater meets required specifications for feeding the vessel.
/
6.2.6 Verify one condensate pump in operation. If vessel pressure is above 200 psi, a booster pump should also be in operation. If pressure is above 400 psi, one RFPT should also be in operation.
/
6.2.7 Ensure at least two heat exchangers in service per respective S.O.I. for-CCW & TBCW
/
r 6.2.8 Establish feedwater control with the startup level controller as follows:
1 I
l CAUTION l
l l
l Minimize the use of a condensate booster pump at l
l low pressures. Due to the excessive differential l
l l pressure developed across the startup level control l l valve, level instability may result.
l l
l I
l l
l l
NOTE l
l l
l The following operations are performed on Panel l l 1H13-P680 unless otherwise noted.
l l
l 6
l l
03-1-01-1 TEXT w
hRANDGULFNUCLEARSTATION INTEGRATED' OPERATING INSTRUCTION l
Title:
Cold Shutdown To Generator lNo.:
03-1-01-1 l Revision: 34 [Page: 40 l
l Carrying Minimum Load l
l l
l ATTACHliENT XII Page 2 of 3 6.2.8 (Continued)
Control Room Operator Initials /Date I
I l
CAUTION l
l l
l A rapid and large introduction of cold feedwater l l into the reactor vessel will cause a significant l l and rapid power increase.
l l
l a.
If the RFPT's are not in service (no steam to the turbines), place the RFPT minimum flow controllers in manual (Panel 1H22-P171), and open as required to maintain startup flow control valve differential pressure as low as possible while maintaining vessel level.
/
b.
Shut the FW CU RECIRC VLV N21-F510 by placing its controller to 0%.
/
c.
Close N21-F003, cleanup Recire Line Isol.
/
d.
Open the SU FCV OUTL ISOL VLV N21-F001 on Panel 1H13-P870.
/
e.
Shut the FW SU BYP VLV N21-F040
/
f.
Verify the RX WTR LVL SU CONT is in manual and at 0%.
/
I I
l CAUTION l
l 1
l Do not throttle B21-F065A or B. l l
l g.
Open the FW INL SHUT 0FF VLV B21-F065B and ensure B21-F065A is shut.
/
4 03-1-01-1 TEXT
GR$NDGUISNUCLEARSTATION INTEGRATED OPERATING INSTRUCTION l
Title:
Cold Shutdown To Generator lNo.:
03-1-01-1 l Revision: 34 lPage: 41 l
l Carrying Minimum. Load I
l l
l ATTACHliENT XII Page 3 of 3 6.2.8 (Continued)
Control Room Operator 1
Initials /Date h.
Open the RX WTR LVL SU CONT, N21-F513, as necessary to maintain reactor water I
level between 32-40 inches. Also adjust the RFPT minimum flow valve, as necessary.
/
1.
Balance the RX WTR LVL SU CONT, N21-F513 f
and place in Auto. Slowly adjust the tape setroint to 36".
/
A
~
l l
l NOTE l
1 l
l Monitor drywell and containment pressure l
l during heatup. Normally, drywell pressure l l should not increase significantly above l
l containment pressure. The drywell may be l
l vented, if necessary, to keep pressure < 1.0l l psig, per SOI 04-1-01-M14-1.
Refer to Tech l l Spec 3.6.1.9 before venting and log venting l l times in 06-OP-1000-V-0001. Containment l
l pressure will increase if the containment l
l ventilation supply and exhaust dampers are l l isolated.
l l
l I
l l
CAlfrION l
l l
l Opening MSIV's with pressure in the vessel l
l can cause a rapid drop in level.
Increase l
l RX level to approximately to 40" prior to l
l opening MSIV's.
l l
l 03-1-01-1 TEXT
ATTAC ENT XW i j
)
I
)
I TABLE 3.3.6-2 h
CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS 5
TRIP SETPOINT ALLOWABLE VALUE TRIP FUNCTION c.
k 1.
ROD PATTERN CONTROL SYSTEM a.
Low Power Setpoint 20 + 15 -05 of RATED THERMAL 20 + 15 -0% of RATED THERMAL POWER N
POWER b.
High Power Setpoint 5 70% of RATED THERMA.L POWER1 70% of RATED THERMAL POWER Q
~
2.
APRM a.
Flow Blased Neutron Flux-Upscale
< (0.66 W + 42%)T*
< (0.66 W + 45%)T*
D^
b.
T n". :
> 4% of RATED THERMAL POWER 13%ofRATEDTHERMALPOWER3 j c.
Downscale neut.ron tlux - Upscale Startup 5 12% of RATED THERMAL PCWER
$ 14% of RATED THERMAL POWER u.
3.
SOURCE RANGE MONITORS w
NA 5
D Detector not full in NA a.
5
< 1.5 x 10 cps Y
b.
Upscale
< 1 x 10 cps c.
Inoperative RA NA d.
Downscale 1 0.7 cps 1 0.5 cps 4.
- a. - Detector not full in NA NA b.
Upscale
< 108/125 of full scale
< 110/125 of full scale c.
Inoperative NA NA d.
Downscale
> 5/125 of full scale
> 3/125 of full scale 5.
Water Level-High
< 32 inches
< 33.5 inches 6.
REACTOR COOLANT SYSTEM REC 1RCULATION FLOW I a.
upscaie s108xofratedriow]
s 111% of rated fiow 7.
REACTOR PODE SWITCH SHUIDOWN POSITION NA NA
- The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W) and the ratio of FRACTION of RATED THERMAL POWER to the MAXIMUM FRACTION of LIMI The trip setting of this function must be maintained in accordance with Specification 3.2.2.
(T factor).
~
ATTACHMEbnr XIV OP-E21-501/Rev. 2 Page 11 of 27
(
OBd.
TOPIC OF INSTRUCTION LEARNING ACTIVITY
overload protec-5.
tion.
D.
Interlock:
Component
_ Interlock Injection Auto open on initiation Auto open can be Valve F005 and power available manually overridden No auto close by placing H/S Manual opening with H/S to close. Valve prohibited above 50 psig will not reopen line pressure after 15 on auto until min TD.
To protect piping the initiation and pump from high pressure.
signal is reset.
Test Return to Auto Close on LPCS
(
Supp Pool initiation F012A LPCS Pump Auto start on initiation Auto start can be signal and LSS panel overridden by sequence on permissive placing H/S to stop.
Pump will not auto start until initiation i
is reset.
Auto Open when flow < 7 Minimum Flow Valve F011 1250 gpm and pump breaker closed.
Auto Close > 1250 gpm.
00SVC Norm /
Prevent continuing alarms Inop H/S in Inop pos. for testing LPCS Room Auto starts on LPCS pump Cooler start PIS N652, 145 psig pump run permissive N653 to ADS logic LPCS/RHR Initiates LPCS and RHR A A Man. Init
' Initiates Div 1 LSS 4
NTROLP - OP/E21/501 REV. 2
ATTACII)f ENT XV GRAND CULF NUCLEAR STATION ALARM RESPONSE INSTRUCTION l 04-1-02-1H13-P601-16A-H1 l
l l Rev.:
10 i Page 1 of 1l l
l RPCS SYS l
l NOT READY FOR l
Safety Related l
AUTO START l
Alarm Device E22-KA-L614 I
l 1.0 POSSIBLE CAUSES 1.1 Voltage regulator not in AUTO 1.2 Engine lock-out not reset 1.3 Generator lock-out not reset 1.4 UNIT MODE SEL. SW not in AUTO 1.5 RPCS motor breaker in lower or control power failure f ewer or control power failure)I 1.6 Diesel generator. outnue bremh-e in 1.7 Undervoltage circuit control power failure 2.0 AUTOMATIC ACTION u
2.1 None 3.0 IMMEDIATE OPERATOR ACTION 3.1 Dispatch an operator to determine cause of alarm i
4.0 SUBSEQUENT OPERATOR ACTION 4.1 If alarm cannot be cleared, declare HPCS system out of service.
4.2 Verify compliance with Technical Specifications section 3.5.1, 3.5.2, 3.5.3, 3.8.1.1 b
i i
l l
ATTACllMENT XVI Page 1 of 4 PLANT OPERATIONS MANUAL Yalume 09 09-S-02-100 S:ction 02 Revision 3 Date 6-18-83 TECHNICAL SECTION INSTRUCTION CRITICALITY RULES SAFETY RELATED Prepared:
1 f
wv\\
Rsviewed:
o Technical e'v Indepen
[
TedinR @ Superi Qendpnt l
Document Control CO:11 ROLLED COPJ List of Effective Pages:
g Pcge l
1-3 GGl:S list of TCN's Incorporated:
- 1) e,
%Icag g,gA y Y
e j "., 'D t o p,. !Ifty SVIE7 Rsvision TCN No._
%. en #830. j* 'h1 f
c ? o '-.,
4E '%gf O
% al-3
- d he '=
0 None
- ):o as~, 13Pp3
- y hn*
f "I?;S$, n II: y,
Ts 1
None
,q
- 3
"'"atyRI.[I i
/
4,,
I 1:;
1 GREND GULF NUCLEAR STATION TECHNICAL SECTION INSTRUCTION Title Criticality Rules No.: 09-5-02-100 Revision: J Page:
1 ATTACHMENT XVI I~ age 2 of 4 1.0 PURPOSE To establish rules that 'will prevent mishandling of fuel.
2.0 REFERENCES
2.1 GE Services Information Letter, SIL No.152, Criticality Margine for Storage of New Fuel 3.0 DEFINITIONS None 4.0 PREREQUISITES None 5.0 PRECAUTIONS 5.1 The type of portable fire extinguishers used around fuel must be of pressurized water type, UL2-1A.
5.2 The nozzles of the fire hoses in the fuel handling area shall be streaming nozzles only prior to storing fuel.
5.3 In the event of a fire in the area where fuel is stored on the refueling floor. the fuel shall be immediately covered with a fire retardant Cover.
6.0 METHOD 6.1 New Fuel 6.1.1 New fuel assemblies shall be stored in the shipping boxes, the new fuel storage vault, or the Auxiliary Building spent fuel pool, or containment fuel pool. Fuel stored in these locations shall be stored in approved racks and approved containers.
6.1.2 When stored in the new fuel storage vault or in the spent fuel pool, assemblies shall be placed only in racks especially made for storing fuel. Fuel shall not be moved through or placed in t
=_
GRAN GULF NUCLEAR STATION TECHNICAL SECTION INSTRUCTION
Title:
Criticailty Rules No.: 09-5-02-100 Revision: 3 Page: 2 ATTACH 11ENT XVI Page 3 of 4 the aisles between fuel storage racks at the level of the storage racks.
n.
If the spent fuel pool is not flooded when new fuel is in storage, the fuel should be covered by a fire retardant material to prevent possible inundation by low density ffre extinguisher foam or water mist.
4 b.
The new fuel storage vault NFV should always be dry. New fuel stored in the vault must be covered with a fire retardant cover that would preclude criticality due to inundation by low density water, fog or spray from a fire hose.
c.
The metal NFV hatch covers must be placed over the entrance to the vault at all times after the conclusion of all fuel inspection and handling activities involving the new fuel vault.
6.1.3 When stored in shipping boxes, the shipping boxes must not be stacked more than four high.
6.1.4 When handling metal shipping boxes, the following shall apply:
a.
Any damaged shipping boxes shall be segregated from all other boxes.
b.
Any work, other than that associated with fuel handling, shall be prohibited in the storage area unless specifically authorized by Reactor Engineering.
c.
Empty boxes, wooden or metal, shall be clearly marked empty.
d.
If any box is opened, the contents thereof shall be removed as soon as possible or the box shall be reclosed.
6.2 Irradiated Fuel r
6.2.1 Irradiated fuel assemblies shall be stored in the spent fuel storage pool and placed only in racks made especially for storing fuel. During outages, fuel may be stored in the containment fuel pool.
i l
-..,-,.,,..n,...c.._,..-.__,...n,.
.,,, +,
SRAND GULF NUCLEAR STATION TECHNICAL SECTION INSTRUCTION
Title:
Criticality Rules No.: 09-5-02-100 Revision: 3 Page: 3 ATTACl!!!ENT XVI Page 4 of 4 l
6.2.2 Defective fuel shall be placed in a defective fuel storage container or stored in a control rod rack until shipment off
, alte.
6.3 General 6.3.1 No more than one fuel bundle should be suspended above the fuel storage array and this at a height no greater than 6 feet for a 4
prolonged period of time.
i 6.3.2 Damaged or dropped' fuel, either new or irradiated, shall not be moved until all factors have been studied and the situation fully analyzed to preclude further damage or contamination.
6.3.3 A fuel array of up to three fuel bundles outside of a normal storage area or outside a normal shipping container should be maintained with an edge-to-edge spacing of 12 inches or more from all other fuel.
6.3.4 A fuel array of four or more fuel bundles outside of the normal fuel storage areas or properly designed fuel shipping container is prohibited.
6.3.5 No more than two fuel bundles should be allowed in or around a fuel prep machine at any time.
This fuel should be separated from the main body of stored fuel by at least 12 inches.
6.3.6 Load handling over the fuel storage areas is to be limited to one 7
fuel assembly or weight equivalent per crane unless the fuel storage area is empty.
An exception to this equirement is a properly designed fuel shipping container suspended over the fuel l
shipping container storage area.
6.3.7 Fuel assemblies shall be stored in such a manner that water will l
drain freely from the assemblies in the event of flooding and subsequent draining of the fuel storage area.
l l
7.0 DOCUMENTATION None l
l i
I.
_ _. - ~
-,.,___.m..
m
_,.. ~,. - _ _
ATTACIlli3NT XVil Page 1 of 2 GRAND GUIE NUCLEAR STATION ADMINISTRATIVE PROCEDURE l 01-S-06-5 l Revision 14l l Attachment IIIlPage 1 of 6 l Page ___ of NRC NOTIFICATION REQUIREMENTS s
7 I.
Immediate Notifiestion (within one hour of occurrence) 1.
The declaration of any of the emergency classes specified in the licensee's approved emergency plan. 50.72(a)(1)(i) a.
The notification.will normally be made and documented in accordance with 10-S-01-6.
2.
The initiation of any nuclear plant shutdown required by the Plant's
, Technical Specifications. '50.72(b)(1).(1)(A) (See. Attachment V Page 1 for
' examples)'
3.
Any deviation from the Plant's Technical specifications authorized pursuant to 10 CFR 50.54(x). 50.72(b)(1)(1)(B) 4.
Any event or condition during operation that results in the condition of the plant, including its principal safety barriers, being seriously degraded; cr results in the plant being:,50.72(b)(1)(li) (See Attachment c
V Page 4 for examples) s.
In an unanalyzed condition that significantly compromises plant safety. 50.72(b)(1)(ii)(A) b.
In a condition that is-outside the desjgn; basis of the plant; or 50.72(b)(1)(ii)(B)
.2.,
In a condition not covered by the plant's operating and emergency c.
procedures. 50.72(a)(1)(fi)(c) 5.
Any natural phenomenon or other externai condition that poses an actual threat to the safety of the plant or significantly hampers site personnel in the performance of duties necessary for the safe operation of the plant. 50.72(b)(1)(iii) (See Attachment V Page 6 for examples.)
6.
Any event that results or should have resulted in Emergency Core Cooling System (ECCS) discharge into the reactor coolant system as a result of valid signal. 50.72(b)(1)(iv) (See Attachment V Page 8 for examples) l 7.
Any event that results in a major loss of emergency assessment capability, offsite response capability, or communications capability (e.g.,
j significant portion of Control Room indication, emergency notification system, or offsite notification system). 50.72(b)(1)(v) (See Attachment l
V Page 18 for discussion).
8.
Any event that poses an actual threat to the safety of the plant or significantly hampers site personnel in the performance of duties necessary for the safe operation of the Plant including fires, toxic gas releases or radioactive releases. 50.72(b)(1)(vi) (See Attachment V page 17 for examples) 1 01-S-06-5 ATT III
ATTACllMENT XVII Page 2 of 2 GRAND GULF NUCLEAR STATION ADMINISTRATIVE PROCEDURE l 01-S-06-5 l Revision 14l l Attachment IIIlPage 2 of 6 l Page ___ of ___
NRC NOTIFICATION REQUIREHENTS 9.
Any Licensed Haterial event which causes.or threatens to cause exposure of the whole body of any individual to 25 rems or more of radiation; exposure of the skin of the whole body of any individual of 150 rems or more; or exposure of the extremities of any individual to 375 rems or more of radiation.
Any Licensed Material event which caused or threatens to cause loss of one working week or more of the operation of any facilities affected.
10.CFR 20.403 (a)(3) 11.
Any Licensed Material event which caused or threatens to cause damage to property in excess of $200,000. 10 CFR 20.403(a)(4) 12.
Immediately after determining the loss or theft of licensed material in such quantities and under such circumstances that it appears that a substantial hazard may result to persons in unrestricted areas.
10 CFR,20.402(a)(1)
=
['
II.
The NRC regional office (Region'II) shall immediately (l' hour) be notified by telephone and telegraph, mailgram or facsimile of the following occurrences:
(10 CFR 20.205(b)(2) & (c)(2))
l 1.
The detection of removable radioactive contamination in excess of 0.01 l
microcuries (22,000 DPM) per 100 square centimeters of packages surface on the external surfaces of the package, or l
2.
If radiation levels are found on the external surface of the package in excess of 200 millirem per hour, or at three feet from the external surface of the package in excess of 10 millirem per hour.
l NOTE l
l l
l The final delivering carrier must also be notified immediately of this event.l l
a III. Four-hour reoorts (within four hours of occurrence).
1.
Any event, found while the reactor is shutdown, that had it been found l
while the reactor was in operation, would have resulted in the plant, including its principal safety barriers, being seriously degraded or being in an unanalyzed condition that significantly compromises plant safety.
50.72(b)(2)(1) (See Attachment V Page 4 of examples) l l
2.
Any event or condition that results in manual or automatic actuation of any engineered safety feature (ESF), including the Reactor Protection System (RPS). However, actuation of an ESF, including the RPS, that results from, and is part of, the preplanned sequence during testing or reactor operation need not be reported. 50.72(b)(2)(ii) (See Attachment V Page 8 for examples) 01-S-06-5 ATT III A~