ML20214L210
| ML20214L210 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 06/26/1986 |
| From: | Harrell E VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Casto C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| Shared Package | |
| ML20214L200 | List: |
| References | |
| RTR-NUREG-1021 NUDOCS 8612020599 | |
| Download: ML20214L210 (24) | |
Text
{{#Wiki_filter:. Umrnia i Attric AndItustr Comfwry \. % N wth Annait ett1Skstron iht Ofir eika 402
.\ftnensi timinia 2.1117 ENCLOSURE 3 VIRGINIA POWER June 26,1986 NORru cAROUNA POWER WEST VIRGINIA POWER Hr. Charles Casto Operator Licensing Section Division of Reactor Safety Region II U.S. Nuclear Regulatory Commission Suite 2900 101 Marietta St. , N.W.
Atlanta, Georgia 30323
Dear Mr. Casto:
In accordance with ES 201 of NUREG 1021 Virginia Electric and Power Company hereby makes official submittal of comments concerning the NRC examinations administered at the North Anna Power Station June 23 through 27, 1986. Detailed question-by-question comments on the written examinations are presented in Attachment 1. Concerning the simulator examination, the restriction of allowing only one instructor to operate the simulator and role play as other plant personnel when called upon by the candidates was very cumbersome, particularly when multiple malfunctions were initiated. It is recommended that more than one instructor be allowed to function in this capacity. Your consideration of these comments is requested. Very truly your E. Wayne rrell Station anager Attachment pc: Mr. W. L. Stewart Dr. B. L. Shriver Mr. L. L. Edmonds 20%h p V
9 4 ln 1 Attachmtnt I l- Page 1 06/26/86 NORTH ANNA POWER STATION COMMENTS ON WRITTEN NRC EXAMINATIONS ADMINISTERED ON JUNE 23, 1986 A. Reactor Operator Examination
- 1. Question 1.03 Comment: Actual answer is d as noted in the SRO Exam answer key for same question (#5.13) .
Recommendation: Ensure answer key is correct.
Reference:
Question #5.13 of SRO Exam.
- 2. Question 1.11 Comment: Question asks for safety limits while the answer key indicates the basis for the safety limits were required.
Recommendation: Accept either answer key or as follows:
- a. RCS pressure shall not exceed 2735 psig.
- b. Combination of thermal power, pressurizer pressure and highest operating loop (Tavg).
Reference:
North Anna Tech Specs (2.0 - Safety Limits).
- 3. Question 1.12 Comment: The greater portion of the neutron population at 36 hours af ter a trip is still fission neutrons. With a keff of 0.95 and a neutron population of 300, source neutrons contribute approximately 15 neutrons, spontaneous fission would provide a handful, but the predominant production of neutrons are still from fission. Additionally, the section referenced by the answer key does not address deuterium-neutron interaction.
Recommendation: Accept fission neutrons as the major contributor to the background neutron population.
Reference:
Above calculation.
- 4. Question 1.14 Comment: Exact answer is 21.955*F subcooled 1*F. The correct answer would not fall within answer key band. ,
Recommendation: Expand band of acceptable answer.
Reference:
Steam Tables L
;g O
n-
- Attachmant 1
.. Pags.2 . ,
06/26/86 , j.
- 5. Question 1.18a:
Comment: The question asks why "this action" (borating to Cold Shutdown) is - necessary. The answer key does not s respond to that question in total. Recommendation: Add the following to key: Ensures adequate shutdown margin in the event of an inadvertent uncontrolled cooldown.
Reference:
Westinghouse Background document, Volumn II-A for Rev. O, Page attached.
- 6. Question 2.01:
Comment: The first portion of the question, RHR discharging to l' the CVCS, occurs whether or not the RHR system is in operation. , With RHR in operation MOV 1720 A and B are open, the accumulator discharge M0V's are closed and HCV 1142 (discharge - to the CVCS) is open. With the RHR not in operation (above Mode 4) MOVs 1720 A -and B are closed, accumulator isolation M0V's are open and HCV 1142 (RHR discharge to the CVCS) is 10% ^ open. The question is referring to RHR in operations in which case the accumulator discharge valves are shut and MOV 1720 A and B are open, therefore, the answer given is not valid. ^> Recommendation: Deletion since statement in question is incorrect.
Reference:
OP 14.1.
- 7. Question 2.02c:
Comment: It is possible to physically rack _ in 15H7. However, since 15J7 ("C" alternate) is racked in, the breaker'(15H7) is electrically locked out from being closed. Recommendation: Accept as answer the following: 15H7 can be physically racked in with 15J7 previously racked in. However, 15H7 is electrically interlocked from closing since 15J7 is racked in.
Reference:
11715-Esk 5AN, SAP d G
.w 4-e --
y e, <--,.--,r
Atttchm:nt 1
, Paga 3 06/26/86 f
- 8. . Question 2.03 a & b:
Comment: Answer key is not complete. Both undervoltage and degraded voltage conditions are indicated for stub bus operation. Recommendation: Recommend accepting degraded or undervoltage situation for answer.
Reference:
11715-Esk-5AQ.
- 9. Question 2.06:
Comment: Question not operator performance based. Instrument technician's knowledge of component not required by operator. Operators are trained and guided by Abnormal Procedures to evaluate face of Radiation Monitor, not troubleshoot why it is not operating. Recommendation: Accept a channel failure as correct answer or delete the question.
Reference:
N/A
- 10. Question 2.09b:
Comment: Answer is upper range. Figure 200.1 notation (#) is for upper range. The status page shown has condition for that range. Recommendation: Change answer to upper vice full.
Reference:
Figure 200.1 and 200.2 of R0 and SRO Examination.
- 11. Question 2.10b:
Comment: NDT setpoints have been changed in the plant since the time the material was sent to the NRC. Recommendation: Accept either 140*F or the new setpoint of 185*F.
Reference:
1-0P-3.4, Page 8 of 16 (attached).
- 12. Question 3.10f:
Comment: Either Phase "A" or "B" will close Steam Generator blowdown trip valves. Recommendation: Accept either Phase A or Phase B.
Reference:
1-11715-ESK-6MA.
,; ~ .: 3 a Attachm:nt 1 Pags 4 06/26/86
- 13. Question 3.10d:
Comment: Charging pump recirc stop valvw (MOV 1373) receives no automatic signals. 4 Recommendation: Accept none as correct answer.
,v.)
Reference:
NCRODP-77.
- 14. Question 3.11c:
Comment: Urgent failure blocks rod motion in " Manual". Recommendation: Change answer key to " Rods will not move." -
Reference:
NCRODP 93.5. 4' 15. Question 3.13: Comment: Answer key is incorrect. Recommendation: Correct answer key to reflect Pt-402 feeds MOV
- 1700; PT 403 feeds MOV 1701. Therefore: MOV 1700 will not be able to open when required. MOV 1701 can be positioned as needed.
Reference:
11715-Esk-60T. 5
- 16. Question 3.19c:
Comment: Low Range Rad Monitor has been disabled due- to high background. Therefore, question is invalid. Recommendation: Delete.
Reference:
N/A
- 17. Question 3.20a:
Comment: Amber light is an indication of breaker disagreement and occurs with the control sv%tN red flagged and breaker open. Recommendation: Answer key tr ad ., g i above comtnent. 3
Reference:
NCRODP-90.1, Section II #5.a.1.c.
- 18. Question 3.20b:
Comment: Answer key is not all inclusive. Recommendation: Accept answer in key or accept "to ensure separate pre-designated flow paths to the Steam Generators in the event normal feedwater is lost and auxiliary feedwater is required."
Reference:
NCRODP-26
Attachm nt 1
. Pegn 5 06/26/86
- 19. Question 4.08:
, n. ;
- Comment: Answer is NO.
'ihl Recommendation: Change answer key to No vice Yes.
Reference:
1-0P-5.2, Page 5 of 10 attached.
-i B. Senior Reactor Operator Exam
.e g 1. Question 5.01:
'h' .
Comment: See Comment #1.14 on RO Examination.
- 2. Question 5.03:
Comment: See comment #1.11 on R0 Examination.
- 3. Question 5.05:
.. s Comment: See comment #1.18 on RO Examination.
- 4. Question 5.07a:
Comment: See comment #1.20a on R0 Examination.
- 5. Question 5.18c:
Comment: If C is constant, then, as the core ages one or more b
;u of the following have to vary:
a) Reactor Power decreases then Xenon decreases and differential boron worth is more negative. b) Rods- are withdrawn therefore less competition and differential boron worth more negative. c) Temperature decreases therefore differential boron worth more negative. Recommendation: This is not performance based. Boron does not remain constant as core ages. Accept either more negative as. ,- described above or less negative - answer key (more fission
- products, more competition and less negative given nothing else
- changes.
Reference:
N/A
- 6. Question 6.01c:
Comment: See Comment #2.02c on R0 Examination. I 7. Question 6.03b: Comment: See Comment #2.09b on R0 Examination. l
Attechm:nt 1
- , Page 6
- 06/26/86
- 8. Question 6.11:
Comment: See Comment #3.10 d and f on R0 Examination.
- 9. Question 6.12 e and d:
Comment: See Comment #3.11 e and d on R0 Examination.
- 10. Question 6.14:
Comment: See Comment #3.13 on R0 Examination.
- 11. Question 6.15:
Comment: Only two conditions are required to initiate the LHSI Swapover Sequence; SI Recirc Signal present and Low Low RWST Level. With these two signals present, LHSI Pump discharge value MOV1863A opens which allows the LHSI Pump Recire valves to close. This action allows MOV 1860 A & B to open and complete the swapover sequence. Recommendation: Accept two answers vice three since the Recirc ' valves are part of the swapover sequence.
Reference:
11715Esk 6 EP, EQ, ES, ET
- 12. Question 6.16:
Comment: See Comment #2.01 on R0 Examination.
- 13. Question 6.17c:
Comment: See Comment #3.19c on RO Examination.
- 14. Question 6.18 a and b:
Comment: See Comment #3.20 a and b on R0 Examination.
- 15. Question 7.07:
Comment: See Comment #4.08 on R0 Examination.
- 16. Question 7.27:
Comment: Answer key is not complete. Recommendation: Accept 100*F - 350'F if CRDM's are energized in addition to answer in key.
Reference:
1-0P-1.1, Page 6 of 12
r, ': , Attach::nt 1
. Ptga 7 06/26/86
- 17. Question 8.05:
Comment: New Tech Spec limits in place since information was sent to the NRC. Recommendation: Accept as answer the following: RCS Heatup - 60*F/hr. PZR Heatup - 100*F/hr.
Reference:
T.S. 3.4.9.1 and 3.4.9.2 (Units 1 & 2) .
- 18. Question 8.24a:
Comment: Question does not ask specifically as to whether 10CFR20 or Administrative limits are required. Recommendation: Accept either 3 Rem or 2750 Mrem as an answer.
Reference:
Section 2.3.4.4.b-5 of the Health Physics Protection Manual. References are attached as appropriate. l f l l l t . T l
~ _ _ _
_( l . i 'rx IV. ~ DISCUSSION OF SPECIFIC GUIDELINE STEPS, CAUTIONS AND NOTES Steps 1 and 2. Following stabilization of plant conditions and the determination that.;a
-plant cooldown is necessary during natural circulation mode, the coolant boron concentration should be corrected to provide for the required shutdown reactivity margin on a total coolant system mass basis. In other words. the RCS boron concentration'in the active portions of the system should be such that it provides the shutdown reactivity margin required by plant Technical Specifications when calculated on the basis of homogeneous distribution of boron within the total plant mass. This will provide ' easonable r assurance that even a f airly rapid temperature drop, which results in a large outsurge of relatively dilute pressurizer liquid into the active (loop) portion of the reactor coolant system, will not cause problems with loss of core shutdown margin. Without reactor coolant pump-driven pressurizer spray, no adequate means of mixing the loop coolant with pressurizer liquid exists. This means that the active (loop + core) portions of the system must be over-borated to some extent to provide for attainment of the required boron concentra-tion on an overall basis.
1 To borate, follow the normal procedures for boration using the VCT make-up control system set in the BORATE mode. Care must be taken to ensure homogeneity of boron in the coolant and, therefore, a slower. rate of boration may be necessary corresponding to the decreased flow rate of natural circulation. Af ter completing the baration, it is important that the operator deter-mine the system boron distribution by obtaining samples from available sample points, particularly the pressurizer liquid. The pressurizer liquid boron concentration will remain at or near the original coolant boron concentration prior to the loss of forced flow event. As mixing occurs in the active portions of the reactor coolant system, the boron concentration in the loop (s) with no charging connection should rise to meet the boron concentration in the loop with the charging connection. ES-0.2 L6603 4254B:1 *
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h' Y-" 'l <. I ' 1-OP-3.4 Page 8 of 16 02-14-86 (, Initials . 4.0 Procedure (cont.) ,
, .*t- 4.4.4 Open and place a Red Tag to the Shift Supervisor on the following breakers:
- l. 4.4.4.1 1H1-2SA4 MOV-RS-100A
.l 4.4.4.2 IJ1-2NC4 MOV-RS-100B.
i i 4.4.4.3 IJ1-2ND4 MOV-RS-101A 4.4.4.4 IH1-2SB4 MOV-RS-101B ,t
' 4.4.5 Close the LHSI Cold Les MOVs 4.4.5.1 MOV-1864A 4.4.5.2 MOV-1864B
.%, NOTE: Accumulator MOV's must be tagged to the Shift Supevisor
- ., a when Solid State fuses are removed.
4.4.6 If RCS temperature is to remain below 200*F AND if desired, remove the solid state protection fuses. l Enter in the Action Statement Status Log that the fuses shall be reinstalled and 1-PT-36.1A and IB satisfactorily completed prior to entering Mode 4.
',j y NOTE: PR7.R FORV setpoints change, with no alarms, as RCS temperature (T ) goes below 185*F.
NOTE: Maximum RCS (except the PRZR) cooldown is 50*F in any one period.
. NOTE: When makeup activities to the RCS are required, utilize 1-LOG-2A.
4.5 Commence or continue Reactor Coolant System cooldown to the desired temperature. Los at 30 minute intervals, Wide Range
. TH, P M Pren ure, and P M Temperature on Attachment #1. ' If containment vacuum is to be broken, init'iate required H.P.
4.6 i Release form. F *
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/ Pagn 5 of 10 3.0 03-18-86 Precautions'and Limitations (cont.)
Jh . 3.3.5 At anytime a coolant ptmp is running less than 400 #
/
! system pressure do not cycle MOV 1380 or MOV 1381 because the pressure transient in the seal water return line could force the #1 seal surfaces to touch and result in damage to same.
3.3.6 Insure seal water injection temperature < 130*F and seal leakoff between .2 gpm and 5 gpm. 3.3.7 Prior to initial operation or after the system has be.en depressurized, the Reactor Coolant Pump vapor seal stand pipe must contain sufficient water to insure that the #3 vapor seal shoes do not operate dry. 3.3.8 The No. I seal bypass valve should not be opened unless either the pump bearing temperature (seal inlet temper-( ature) or the No. I seal leakoff temperature approaches its alarm level. The No. I seal bypass valve should then be opened only if all of the following conditions are met: (NSD-TB-80-10)
- 1. Reactor coolant system is greater than 100 PSIG but less than 1000 PSIG.
- 2. No. I seal leakoff valve is open.
- 3. No. I seal leakoff flowrate is less than 2
one GPM.
- 4. Seal injection water flow rate to each pump is greater than six GPM.
3.3.9 Maintain flow through #1 seal bypass line until the j i
'A conditions of 3.3.8 cannot be met.
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g , 7 e y;' 1-L5-86 o . l REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 1 l 3.4.9.1
- The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines showa on Figures 3.4-2 and 3.4-3 during heatup, cocidown, criticality, and inservice l
1eak and hydrostatic testing with: l a. A maximum heatup of 60*F in any one hour period. { b. A maximum cooldown of 100*F in any one hour period. c. A maximum temperature change of less than or equal to 10*F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves. APPLICA5ILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains i acceptable for continued operations or be in at least HOT STAN08Y within the next6hoursandreducetheRCSTl'I0hoursand respectively, within the followin . pressure to less than 200*F a , SURVEILLANCE REQUIREMEffTS i ' 4.4.9.1.1 The Reactor Coolant System temperature anu pressure shall be deter-mined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall i be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. shall be used to update Figures 3.4-2 and 3.4-3. The results of these examinations NORTH ANNA - UNIT 2 3/4 4-26 Amendment No. 60 n--w --n,w---,--,w,. -,w, ,_ -n--.,--,_,,,n --,,.,,.,-,-,-- ,-.r._- _ _ _ --- - - - - - - . - - - , --- ---~
. o 9.y3..; 1-15-86 REACTOR COOLANT SYSTEM 3/4.4.g PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shewn on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
- a. A maximum heatup of 60*F in any one hour period.
- b. A maximum cooldown of 100*F in any one hour period.
- Note APPLICABILITY: At all times.
ACTION: With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T and pressure to less than 200'F and 500 psig, respectively, within the folio Sg30 hours. SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Cdolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inseivice leak and hydrostatic testing operations. 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined to detemine changes in material properties, at the intervals required by 10 CFR 50, Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.
- Specification 3.4.9.1.c was erroneously deleted by Amendment 74. A Tech. Spec. change is being submitted to correct this error. The Specification read as follows:
"A maximum temperature change of less than or equal to 100F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves."
NORTH ANNA - UNIT 1 3/4 4-26 Amendment No. 74
6
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. 8-21-80 REACTOR COOLANT SYSTEM
,, PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be Ifmited to:
a. A maximum period, and heatup of 100*F or cooldown of 200'F, in any one hour b. A maximum spray water temperature and pressurizer temperature differential of 320'F. APPLICA81LITY: At all times. ACTION: With the pressurizer temperature Ifmits,in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the cut-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANOBY within the next 6 hours and reduce the pressurizer pressure to less than 500 psig within the following 30 hours. l SURVEILLANCE REQUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours during auxiliary spray operation. I i NORTH ANNA - UNIT 2 3/4 4-29
$ f ,! 41 T R $' '
11-26-77 I REACTOR COOLANT SYSTEM PRES $URIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer tenperature shall be 1tmited to:
- a. A maximum heatup of 100*F or coofdown of 200*F, in any one hour period, and
- b. A maximum spray water temperature and pressurizer temperature differential of 320*F.
APPLICABILITY: At all times. ACTION: With the pressurizer temperature limits in excess of any of the above limits restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-s limit condition on the fracture toughness properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANOBY within the next 6 hours and reduce the pres-surizer pressure to less than 500 psig within the following 30 hours. I i SURVEILLANCE REQUIREMENTS f l 4.4.9.2 The pressurizer temperatures shall be determin.. to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water tenperature differential shall be determinec to be ' within the limit at least once per 12 hours during auxiliary spray operation, i I i h NORTH ANNA . UNIT 1 3/4 4-30 __}}