IR 05000324/1997300
| ML20140C372 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/30/1997 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20140C344 | List: |
| References | |
| 50-324-97-300, 50-325-97-300, NUDOCS 9706090226 | |
| Download: ML20140C372 (285) | |
Text
I
!
l i
U. S. NUCLEAR REGULATORY COMMISSION
. REGION 11 l
Docket Nos.: 50-325, 50-324 l
License Nos.: DPR-71, DPR -62 !
l l
Report Nos.: 50-325/97-300, 50-324/97-300 Licensee: Carolina Power and Light Company Facility: Brunswick Steam Electric Plant Units 1 & 2 Location: Southport, NC
!
Dates: April 25 through May 2,1997 ,
Examiners: M2 George T. Hopper, Whief License Examiner James H. Moorman, Ill, License Examiner D. Charles Payne, License Examine'r l.
l
- Approved by:
Nf$
. Thomas A. Peebles, Chief, Operator Licensing and Human Performance Branch Division of Reactor Safety l
i
!'
!
9706090226 970530 PDR
Enclosure 1 V ADOCK 05000324 PDR ,
. - - -- -
- - .. -
.
EXECUTIVE SUMMARY 1
'
I Brunswick Steam Electric Plant Units 1 & 2
<
NRC Examination Report Nos. 50-325/97-300, 50-324/97-300 During the period April 25 through May 2,1997, NRC examiners conducted an announced operator licensing initial examination in accordance with the guidance of Examiner Standards, ,
, NUREG-1021, Interim Revision 8. This examination implemented the operator licensing l requirements of 10 CFR @55.41, @55.43, and @55.4 i l
Operations
Control room activities were observed during the examination validation week and i examination administration week. The operators were found to be attentive and professional in their duties (Section 01.1).
-
Five Senior Reactor Operator (SRO) candidates and four Reactor Operator (RO)
candidates received written examinations and operating tests. The licensee administered the written examination on April 25,1997, and the NRC administered the operating tests April 28 - May 2,1997 (Section 05.1).
.
Four SRO and three RO candidates passed the examination. Five of the seven candidates marginally passed the examination. One SRO and or.e RO candidate failed. (Section 05.1)
+ Candidate Pass / Fail SRO RO Total Percent Pass 4 3 7 77.8 %
Fail 1 1 2 22.2 %
+ Licensee examination preparation activities were considered good (Section 05.2).
- The examiners concluded that overall candidate performance on 11.3 operating test was weak. Candidates had difficulty prioritizing and correctly implementing the mitigation strategies of the emergency operating procedures (Section 05.3).
- The examiners identified a generic performance weakness in the area of procedural compliance (Section 05.3).
- One deviation concerning the description of the operation of the Automatic Depressurization System was identified (Section 08.1).
Enclosure 1 j
I Report Details Summary of Plant Status During the period of the examinations Unit 1 and Unit 2 were at 100 percent powe l. Operations 01 Conduct of Operations 01.1 Control Room Observation During validation and administration of the examination, the examiners observed the I conduct of operations by currently licensed operators in the control room. The ROs l were attentive to the evolutions in progress. The SROs limited personnel access for official business only, which contributed to a quiet, professional atmospher I 05 Operator Training and Qualifications )
i 05.1 General Comments NRC examiners conducted regular, announced operator licensing initial examinations during the period April 25 - May 2,1997. NRC examiners administered examinations developed by the licensee's training department, under the requirements of an NRC ,
security agreement, in accordance with the guidelines of the Examiner Standards !
(ES), NUREG-1021, Interim Revision 8. Three SRO instant, two SRO upgrade, and four RO license applicants received written examinations and operating test Four SRO and three RO candidates passed the examination. Five of the seven candidates marginally passed the examination. One SRO and one RO candidate failed. Two candidates (one SRO and one RO) were graded as marginal passes on the Job Performance Measure (JPM) portion of the operating test. One SRO candidate was a marginal pass on both the administrative portion and the dynamic !
simulator portion of the operating test. One RO candidate marginally passed the dynamic simulator. Yet a fourth SRO candidate was graded as a marginal pass on all three categories of the operating test. Candidate's are considered to have marginally passed if they receive an unsatisfactory grade on any one administrative topic area, complete only 80 percent of the JPMs successfully , or receive a grade of 1.8 to on any one competency during the dynamic simulator examinations. Detailed candidate performance comments have been transmitted under separate cover for managernent review and to allow appropriate candidate remediation.
05.2 Pre-Examination Activities The licensee developed the SRO and RO written examinations, three JPM sets, and four dynamic simulator scenarios for use during this examination. All materials were submitted to the NRC on time and were of good quality, meeting the guidelines specified in NUREG-1021. Most of the changes made to the written examinations were editorial in nature. Only two questions out of a total of 125 contained distractors which needed to be altered snd one additional question contained two potentially Enclosure 1
correct answers. The NRC reviewed the written examinations and found that 76 percent of the questions were written at the comprehension / analysis level. A total of 62 new questions had been develope The simulator scenarios were challenging and designed to ensure that each candidate could be adequately evaluated on a majority of the items listed in 10 CFR 55.45(a).
The examiners considered each scenario to be a challenging test of the candidates'
ability. Few changes were made to the content of each scenario. However, the NRC requested additional detail be provided in the operator activities section of some of the scenarios to clarify expected operator action .3 Examination Results and Related Findinas. Observations. and Conclusions a. Scooe The examiners evaluated the candidates' compliance with and use of plant procedures during the simulator scenarios and JPMs. The guidelines of NUREG-1021, Forms ES-303-3 and ES-303-4, " Competency Grading Worksheets for Integrated Plant Operations," were used as a basis for the evaluations, b. Observations and Findina Examiners identified numerous weaknesses in candidate performance during the operations portion of the examination. Details of the discrepancies are described in each individual's examination report, Form ES-303-1, " Operator Licensing Examination Report". The examiners identified several generic weaknesses which
, were o.f particular concer During performance of one scenario, the candidates responded to a recirculation line break Loss Of Caolant Accident (LOCA). Two out of three crews failed to initiate an emergency der:sssurization when required upon reaching Top of Active Fuel (TAF).
The crews div ..ot manually open Automatic Depressurization System (ADS) valves untillevel had decreased below the Minimum Zero injection Level (LLS). The scenario involved a loss of reactor coolant which exceeded the injection flowrate of the High Pressure Coolant Injection (HPCI) pump. Level decreased to TAF in approximately two minutes and ADS valves were not opened until four minutes after the start of the event. The candidates should have anticipated the need for an emergency depressurization, based upon the rate of level decrease, and opened the ADS valves when level reached TAF per step RC/L-28 of Emergency Operating Procedure EOP-01-RVCP. Emergency depressurization should have been manually initiated approximately two minutes sooner than it actually occurred. Examiners observed one crew conduct a crow brief as level continued to decrease before the directive was given to open the ADS valves. The order to open the valves should have been given immediately. While both crews were observed to have been aggressively performing the actions of EOP-02-PCCP, Primary Containment Control Procedure", addressing adverse containment conditions, the priority of ensuring adequate core cooling was neglecte Enclosure 1
.
I
During the performance of emergency depressurization, two of the three crews failed to ensure that Core Spray Loop A was injecting into the Reactor Pressure Vessel (RPV) as required by Procedure EOP-01-RVCP. Only two sources of Emergency Core Cooling System (ECCS) low pressure injection were available during the even Low Pressure Coolant injection (LPCI) Loop A, had been lined up earlier for suppression chamber spray, but was available for injection. Core Spray Loop A, was also available but had to be manually started and lined up due to a LOCA logic j failure. Both crews did not open the core spray injection valve (F-005A) until after the '
RPV had been depressurized. One crew took more than five minutes to realize that
- core spray was not injectin Adequate core cooling exists so long as RPV water level remains above TAF. If at i least one source of injection into the RPV is available, emergency depressurization is required when water level reaches TAF to maximize the iniection flowrate from all l operatina sources of iniection. The consequences of not depressurizing the RPV under conditions that require an emergency depressurization could include a loss of adequate core cooling or failure of the primary containment. Procedure EOP-01-RVCP required at least two low pressure injection systems be lined up to obtain i maximum available flow to provide core coolinS Ond restore vessellevel above TA The examiners noted that the most significant operator error identified in the BSEP !
Probabilistic Safety Assessment (PSA) is when operators fail to manually depressurize the reactor when required. Therefore, this is of concern because an increase in likelihood that open: tors fail to depressurize prior to reaching TAF would have a large ;
'
effect on the PSA r'sults and the plants's calculated Core Damage Frequency (CDF).
The examiners obs erved other misapplications of procedures or oversights by the candidates. None of the crews individually scramed a drifting rod as instructed by
'
Procedure APP-A-06 3-2, " ROD DRIFT" step 4.e. Two crews failed to follow a procedural precaution and secure the RCIC pump to prevent continuous operation at low speed despite this precaution being posted as an operator and on the control panel. One crew failed to inhibit ADS when reactor water level decreased below +45 inches as required by Procedure EOP-01 step RC/L-22, and allowed the reactor to automatically depressurize. This unplanned emergency depressurization may have altered the mitigation strategy of the event had the operators had time to recover
- HPCI and regain control of RPV leve Candidate performance on the walkthrough portion of the examination was poor. Four out of nine candidates marginally passed this portion of the examination. One other candidate failed. The examiners noted performance errors ranging from inattention to detail to gross conceptual errors. Examples of some of these errors included:
-
Racking in the wrong breakers to cross-tie emergency busses.
.
-
Using a number greater than 100 percent in an attempt to calculate alternate
,
power indication per Procedure GP-3 using bypass valve positio Demonstrated unfamiliarity with Diesel Generator controls and proper operation of the diese Enclosure 1
- . _ . . - - - . . - - .. - _ - - - . - - ~ = - .. , . .-- - . -- --
,
i
.
t -
Improperly venting the drywell instead of the suppression chamber given the l plant conditions imposed.
)
i .
Improperly performing procedure LEP-02, " Alternate Control Rod Insertion."
, Candidates scrammed rods simultaneously vice individuall Performing Alternate Emergency Depressurization procedure too slowly with l little concem for vessel level or the need for adequate core cooling, i
'
The examiners noted that candidates were unfamiliar with some of the procedures and encountered difficulty i., ensuring verbatim compliance. Most of the discrepancies noted could be attributed to the candidates failure to carafully read and follow procedural guidanc Conclusio3 The examiners concluded that overall candidate performance on the operating test was weak. Candidates had difficulty prioritizing and correctly implementing the mitigation strategies of the EOPs and were in sensitive to the need to keep the core covered in order to ensure adequate core cooling. Lack of procedural compliance contributed to many of the errors that were observed. . The weaknesses noted were ;
fundamental in nature and should not be attributed to subtleties or the level of difficulty of any part of the operating tes Miscellaneous Operations issues 08.1 Special Review of UFSAR Commitments Scope A recent discovery of a licensee operating their facility in a manner contrary to the UFSAR description highlighted the need for a special focused review that compares plant practices, procedures and/or parameters to the UFSAR descriptions. While performing the examinations discussed in this report, the inspectors reviewed the
' applicable portions of the UFSAR that related to the areas that were examined. The inspectors intent was to verify that the UFSAR wording was consistent with the observed plant practices, procedures and/or parameters, Observations and Findinos The inspector found that the system description for the Automatic Depressurization System on page 1.2.2-9 indicated that the system will function automatically in a LOCA situation in which the HPCI system fails to automatically maintain reactor vessel water level. Current procedural guidance contained in EOP-01-RVCP (step RC/L-22)
has the operators inhibiting ADS when vessel level drops to +45 inches, thereby defeating the automatic operation of the system. Consequently, the system will not operate as stated unless the operators do not perform the actions required by the EOPs. In addition, FSAR Chapter 15 page 15.OA.6-15, Event 34-Pipe Breaks Inside Primary Containment," described the equipment requirec to mitigate a LOCA which Enclosure 1
. . - _ . . _ ~ . _ _ . _ .. . _ _ . . _ _ _ . - . . . . _ - _ . _ _ _ _ . _ _ . . . _ _ _ . _ _
,
i
'
5 :
included the ADS system. This section again refers to the automatic operation of ADS i 4- and does not indicate that the system's automatic feature will be defeated and *
l operated manually by the operators during the blowdown phase of the even !
"
i I Conclusion
l
.' The inspectors concluded that the UFSAR wording in the items mentioned above were ,
?
inconsistent with current plant operating practices and procedures . These l discrepancies are collectively identified as additional examples of URI !
50-326,324/96-05-02, "FSAR Discrepancies." .
l V. Manaaement Meetinas
{
X1. Exit Meeting Summary At the conclusion of the site visit, the exarniners met with representatives of the plant staff listed on the following page to discuss the results of the examination The examiners asked the licensee whether any materials examined should be considered proprietary, No proprietary information was identifie I
,
Enclosure 1
. _ _
-. .
_ _ . - - _ .. . _ . _ . _ - ._._.__.. _ . _ _ . .. _ _ _ . _ . - -- - _ _._ _ _ __ .____
.
i :'
1 -
- 6
4 PARTIAL LIST OF PERSONS CONTACTED
- Licensee i ,
,
L. _ Dunlap, Training Manager (Acting) #
J. Gawron, Manager NAD j K.; Jury, Manager Regulatory Affairs 1 - W. Levis, Director of Site Operations 1
! R. Lopriore, Plant Manager i K. McCall, Supervisor, Operator initial Training
'
R. Mullis, Operations Manager l J: Rainsburrow, Operations Support Manager
-
G. Thearling, Regulatory Affairs
l NRC
!
].
'
E. Brown, Resident Inspector
i ITEMS OPENED, CLOSED, AND DISCUSSED
l Opened
i
- . None i
t i Closed
NONE i
l Discussed i
]. - 96-05-02 FSAR Discrepancies
i' NONE
.
A i' !
?
j 1'
i' l l
i 1 I 1 t i
1,
4 l e
.
, !
i l
-
Enclosure 1 k
.
w r - T'y a w y - - - - - - ,. .y 7'+- * =wm y-* - ---'-' - ' - s
.. ._ _ .___... _ ..-_ _ _... _ _ . __ _... _.._ . _ .. _ ._.. _ _ .._ . _ _ . ... _ _ __ _ _.- _ . _ _. - . _.. _ ..__ . _ _
l I
- )
!- 7 !
- ;
LIST OF ACRONYMS USED I e
J l ADS Automatic Depressurization System !
! CFR- Code of Federal Regulations =j j ECCS Emergency Core Cooling Systems '
j ES Examiner Standards (NUREG-1021)
l HPCI High Pressure Coolant injection
)
! LPCI Low Pressure Coolant injection
, NRC Nuclear Regulatory Commission
RO Reactor Operator j RPV Reactor Pressure Vessel SRO Senior Reactor Operator l
- TAF Top of Active Fuel 3- UFSAR Updated Final Safety Analysis Report
!
I.
l-a e
I
)
i
A l !
'
l l
l Enclosure 1
.. .- -~ . - - - . . - . - - - . . . - - . . _ _ - . - . . - . . - . - . . - - - - - . - - . -
!
t i
!
!
SIMULATION FACILITY REPORT !
i Facility Licensee: Brunswick Steam Electric Plant l Facility Docket Nos.: 50-325 an'd 50-324
Operating Tests Administered on: April 28 - May 2, j i
This form is to be used only to report observations. These observations do not constitute audit or inspection findings and are not, without further verification and review, indicative 'of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future evaluations. No licensee action is required in response to these observation While conducting the simulator portion of the operating tere, the following items were observed (if none, so state):
ITEM DESCRIPTION l
I NONE l
l
1
-
'
l
! l
!
l i
.
!
t Enclosure 2
, __
- -
. - -- .--.
~
SIMULATION FACILITY REPORT
- Facility Licensee
- Brunswick Steam Electric Plant
,
Facility Docket Nos.: 50-325 and 50-324 Operating Tests Administered on: April 28 - May 2, This form is to be used only to report observations. These observations do not constitute i
audit or inspection findings and are not, without further verification and review, indicative of noncompliance with 10 CFR 55.45(b). These observations do not affect NRC certification or approval of the simulation facility other than to provide information that may be used in future
, evaluations. No licensee action is required in response to these observation While conducting the simulator portion of the operating tests, the following items were observed (if none, so state):
ITEM DESCRIPTION NONE Enclosure 2
!
.
I WRITTEN EXAMINATION (S) AND ANSWER KEY (S)(SRO/RO)
Enclosure 3
J
,
<
!
,
t
,
s l
l h ( 1 i
l u .. ... -
_._____.._____;
_ _. . _ _ _ _..._ _.__ _ _ __._-- _--_ _ -_.__ _ __ ._ _-_
/
/
ES-401 Site-Specific Written Examination Form ES-401-7 Cover Sheet
,
l
!
U.S. Nuclear Regulatory Commission Site-Specific Written Examination i i
!
Applicant Information l
'
Name: Region: II Date: 04/25/97 Facility / Unit: Brunswick /1 & 2 License Level: Senior Reactor Operator Reactor Type: GE BWR-4 Start Time: Finish Time:
Instructions
]
l Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percen Examination papers will be collected four hours after the examination start __
Applicant Certification l All work done on this examination is my own. I have neither given nor received ai Applicant's Signature ,
i Results j Examination Value Points
!
Applicant's Score Points
-_
Applicant's Grade Percent i
=
l
.
a WRITTEN EXAMINATION GUIDELINES After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or given assistance in completing the examinatio . To pass the examination, you must achieve a grade of 80.00 percent or greate Every question is worth one poin . For an initial examination, the time limit for completing the examination is four hour . You may bring pens and calculators into the examination room. Use only black ink to ensure legible copies.
- Print your name in the blank provided on the examination cover sheet and the answer
'
sheet. You may be asked to provide the examiner with some form of positive
-
identificatio . Mark your answers on the answer sheet provided and do not leave any question blank. Use only the paper provided and do not write on the back side of the pages.
If you decide to change your original answer, draw a single line through the error, enter the desired answer, and initial the chang l If the intent of a question is unclear, ask questions of the NRC examiner or the i
designated facility instructor only.
4 Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately after the examination.
'
1 After you have turned in your examination, leave the examination area as defined by
'
the proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoke I 1. Do you have any questions?
,
l l
l
- "NRC 97-1 SRO, Rev 0" EXAMINATION ** ,
l QUESTION 1 POINT VALUE: 1.00 Unit One (1) is operating with the following plant conditions:
Core Thermal Power 2558 MWth Reactor pressure 1030 psig Core Flow 77 Mlbm/hr The SA'FETY LIMIT for THERMAL POWER is the MINIMUM CRITICAL POWER l RATIO (MCPR) shall not be less than: .07 .08 4 .09 ' .10 l
'
QUESTION 2 POINT VALUE: 1.00 A Unit Two (2) startup is in progress per GP-02. The Reactor is ,
critical with Reactor power at the point of adding hea .
l I
'
Coolant temperature is being raised to saturation conditions, with Reactor steam dome pressure at 0 psig. RWCU is in servic PT-01.7, Heatup/Cooldown Monitoring is being performe Which of the following conditions would result in UNSATISFACTORY Acceptance Criteria of the PT? G31-TI-R607, Pt 5 indicates 169*F C12-TR-R018, Ch 151 indicates 164*F Reactor water level is 200" G31-TI-R607, Pt 5 indicates 187aF C12-TR-R018, Ch 151 indicates 183 F Reactor water level is 210" G31-TI-R607, Pt 5 indicates 174 F C12-TR-R018, Ch 151 indicates 170*F Reactor water level is 200" G31-TI-R607, Pt 5 indicates 198aF C12-TR-R018, Ch 151 indicates 195*F Reactor water level is 210" PAGE 1
.- -- . . -. -
l l
l
- "NRC 97-1 SRO, Rev 0" EXAMINATION ** l
!
QUESTION 3 'dOINT VALUE: 1.00 Unit One (1) is operating
/
with the following conditions:
Reactor Power 55% l Reactor Feed Pump 1A operating !
Reactor Feed Pump 1B idling Recirculation Pump Speeds 58% ;
Reactor Feed Pump (RFP) 1A trips and Reactor Level Hi/Lo alarm i Reactor Level drops to the scram setpoint and continues to lower to
{
+110 inches before the operator brings the idling RFP on line (30 l seconds after the trip of RFP 1A) to restore Reactor Leve l What should be the present status of the Recirculation Pumps? ) Running at 58% spee I Running on limiter # ] Running on limiter # Tripped on ATWS ARI/RP QUESTION 4 POINT VALUE: 1.00 Following unsatisfactory performance of PT-13.1, the crew has entered i AOP-0 One jet pump has been declared INOPERABLE, requiring a plant shutdown per GP-05 and Technical Specification Technical Specification Bases requires a plant shutdown due to the hazard in case of a Design Basis Accident associated with the increased blowdown area and the: reduction in core cooling from coastdown flow with a broken jet pump rise reduction in core cooling from coastdown flow with a broken jet pump standpip elimination of the capability of reflooding the core with a broken jet pump rise elimination of the capability of reflooding the core with a broken jet pump standpip PAGE 2 L
__ ._ - _ . _ _ . _ . - _ . _ _ _ . . _ . _ _ _ _ . _ . _ . _ . . . ~ . . _ . _ . . _ . _ . _ _ _ .
1 l l
<
l
I
.
,
- "NRC 97-1 SRO, Rev 0" EKAMINATION **
.
QUESTION 5 POINT VALUE: 1.00 Unit One (1) is operating at 27% power.during Unit startup. The i
Turbine Generator has been synchronized to the grid. A total loss of
,i Division'II DC Switchboard 1B results in a reactor scram. During this
- transient
57 control rods fail to fully insert Reactor pressure peaks at 1132 psig
-
Reactor water level lowers to +107 inches
.
BOP Buses fail to transfer to the SAT y . Diesel Generator 2 fails to start What is the expected status of the Alternate Rod Injection (ARI)
- system? ARI has:
t-4 auto initiated on high reactor pressure, j auto initiated on' low reactor water level, l not auto initiated but can be manually initiated.
i d.- not auto initiated and cannot be manually initiated.
.
l QUESTION 6 POINT VALUE: 1.00 i
l
,
'Following a line break on Unit Two (2) plant conditions are:
y i
Reactor pressure 450 psig Reactor water level +70 inches
'
Drywell pressure psig Drywell temp (average) 165 deg F i
i All Drywell Cooler Fans will:
! trip but can be restarted per SEP-10.
i l trip and cannot be restarted per SEP-1 auto start but can be tripped at the RTGB.
)4 auto start and cannot be tripped at the RTG b i'
i
PAGE 3
{
- . . .
_ . . - . _ . _ _ _ _ . . _ . _ _ _ _ _ . . _ . _ . . . _ _ , . _ _ _ _ _ _ - . . _ _ _ . _ _ . . _ _ _ _ . , _ - . _
l l-l l-
,
!
- "NRC 97-1 SRO, Rev 0" EKAMINATION ** ,
!
QUESTION 7 POINT VALUE: 1.00 Unit Two (2) is in OPERATIONAL CONDITION 5, with CORE ALTERATIONS in ,
progress and SECONDARY CONTAINMENT INTEGRITY establishe The Interruptible Instrument Air Header ruptures and the Interruptible i
Air Header isolation valves are closed. Non-Interruptible Air Header
! pressure is norma '
l l, How is. SECONDARY CONTAINMENT INTEGRITY affected?
l l
l Reactor Building Supply and Exhaust Fans remain in service, ;
! SBGT auto start >
L Reactor Building Supply and Exhaust Fans remain in service, SBGT remains in standb , Reactor Building Supply and Exhaust Fans trip and'SBGT auto e
'
starts to maintain negative pressure.
l
' Reactor Building Supply.and Exhaust Fans trip but SBGT must be manually started to maintain negative pressur i i
QUESTION 8 POINT VALUE: 1.00
,
- Unit Two (2) is operating at power with Diesel Geneator 3 (DG3) under clearance. A lockout of a BOP Bus initiates a transient resulting in a.
l reactor scram signal and an ATWS. Plant conditions:
Reactor power 10%
Bus E3 De-energized l
SLC Switch PUMP B RUN position l
Bus E7-E8 Cross-tie Breakers racked in
,
How will the SLC. system respond when the Bus E7-E8 Cross-tie Breakers are closed? SLC squib valve: A fires, no SLC pump start B fires, no SLC pump starts.
- A fires, one SLC pump starts.
,
!. B. fires, one SLC pump starts.
!
.
PAGE 4
_, > _. _. _ _ _ _ _ . . . _ _ _ . _ . .._..~ _ _ ..._. _ ._ _ .___.. _ _ _ _ ._ _. . . . . _ . . . . _ . _ . ~ _ _ _ . ~ . . . .
I i
'
n !
'
-
!
! !
1 1 ** "NRC 97 1 SRO , Rev 0" EXAMINATION . **
-
l t
'
.
i !
< ;
, QUESTION 9 POINT VALUE: 1.00 -8
A. Unit One (1) reactor scram occurs from 100% power. The operator'
completes the'immediate scram actions and notes the.following. Rod Worth j j Minimizer (RWM) display: i e i
ALL RODS IN: NO
<- SHUTDOWN: YES I j -RODS NOT FULL IN: 001 I j' When'the operator depresses the List Rods RWM key, rod 02-51 is
~
. displaye What'can be the furthest withdrawn position of rod 02-51 ;
to.cause the above display, and the basis for that rod position? ;
4 :
j ,. shutdown margin calculation .
j ,' maximum subcritical bank withdrawal position,
.
j j -t
. , maximum suberitical bank withdrawal positio l
- , maximum suberitical bank withdrawal positio ! l i l
. ,
- QUESTION 10 POINT VALUE
- 1.00 j e
During a low water level condition, CRD Flow maximization is being '
~
!. !
- implemented per SEP-09 with the Reactor Building accessible.
.,
e .The operator is directed to maintain. Charging Water Header pressure !
l' =950 psig while opening the Flow Control and Pressure Control valve ;
. .
) ~ This limitation will prevent pump: l
l l' trip on overcurrent protectio I
?
- trip due to low suction pressur '
< flowrate in excess of runout capacit ,
>
l -
d._ discharge pressure dropping below reactor pressur l
,
i i .
)
l i
r
..
- - PAGE 5
- t ;
. .+ . , . y -w ,, q~ q #
..9- m n -, -- 4 9 , 9r-
, . . . _ . - - , . . _ . . . - . _ _ . . . _ - _ _ _ _ . . _ _ _ _ _ . . _ . . _ . _ _ _ _ . _ _ _ _ _ . .
'
t
'
L l-
!>
'
l l l
'f
- "NRC 97-1 SRO, Rev 0" EXAMINATION ** ?
f l- i
I QUESTION 11 POINT'VALUE: 1.00 '
.
$
Unit One (1) is in OPERATIONAL CONDITION 5, performing COR !
!- ALTERATIONS.. A fuel shuffle 'is being performed (the core will not be l unloaded). The following'SRM~ indications are observed: }
SRM Channel A 4 cps
SRM Channel D 2 cps ,
. Per Technical Specifications, CORE ALTERATIONS may be performed in the following core quadrants: 1
- Northeast and Northwest onl l Northeast and Southeast onl I i Northeast, Northwest and-Southeast onl I i Northeast, Northwest and Southwest onl l
,
.
l l
!
l
!
i I l'
i l
'
l F
l J
$
PAGE 6 I
_
-,.
.
, . _ . . . _ . _ _. _ . . - - , ._i
.- - ._ . . . ~ . - . . . . . . ~ . . - ---...-.- - -- . . - - - . _ . ~ . ,
i
1
'
\
l
'
!.
t ** "NRC 97-1 SRO, Rev 0" EXAMINATION. **
I i
QUESTION 12 POINT VALUE: 1.00 During a Reactor startup following refueling, the operator is'
) performing PT-50.2, IRM Range 6/7 Overlap Determinatio The following i data is recorded:
Rance 6 Rance 7
i i j IRM A 50 4 f, IRM B 42 5 i IRM C 45 5 i IRM D 43 4 .
l IRM E 37 3 ! I IRM F 52 6 IRM G 44 4 IRM H 50 6 The Level 2 acceptance criteria of PT-50.2 is: satisfactory for all Division l'and II IRM ; unsatisfactory for at least one Division I IIU4 onl unsatisfactory for at least one Divivision II IRM only, unsatisfactory for at least one Division I and one Division II IR QUESTION 13 POINT VALUE: 1.00 1
i APRM Channel C has all associated LPRM inputs Operable, with the LPRM !
function switches in Operate. During performance of a MST, the LPRM function switches are placed to'and left in Bypass, one at a tim !
l
' What is the MINIMUM number of LPRM function switches that should have :
been placed in Bypass when a Neutron Monitoring trip signal (1/2 scram)
is received? l PAGE 7
-. . ._ .. .~. - ._ - _ - - - - .-. -, - . - . -. -
l..,
l' t I i
)
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
l l
~
i QUESTION 14- POINT VALUE: 1.00 Unit Two (2) is operating at 100% power.- The Standby Gas Treatment
[ System (SBGT) is in the standby alignment. A Trip of both Reactor
..
Feedwater Pumps results in the following plant conditions:
Reactor water level +80 inches l Reactor pressure 945 psig ;
Drywell pressure +0.3 psig Rx Bldg pressure -0.4" WC '
HPCI and RCIC have initiated to restore Reactor water leve All I systems respond as designed during the transien What operator action l is recuired concerning the SBGT System? 1 a .- Secure one SBGT fan per OP-1 ,
! Open Post LOCA Vent valves (SGT-V8 and V9). Open Primary Containment Suction valve (VA-2F-BFV-RB). l
! Restart Rx Bldg HVAC and secure both SBGT trains per SEP-04.
l QUESTION.15 POINT VALUE: 1.00 A condition has arisen on Unit Two (2) requiring BOTH of the Reactor ,
Building Vent Exhaust Radiation Monitors to be declared INOPERABL The Shift Superintendent has determined that Unit Two (2) must be ;
placed in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within the '
following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> What additional action is REQUIRED by Technical Specifications? Establish SECONDARY CONTAINMENT INTEGRITY with the Standby Gas System operating within one' hou Establish SECONDARY CONTAINMENT INTEGRITY with the Standby Gas System operating within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Place at least one INOPERABLE Reactor Building Vent Radiation channel in the tripped condition within>one hour.
, Place at least one INOPERABLE Reactor Building Vent Radiation
[
channel in the tripped condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i PAGE 8
__ . _ .
. . . .- - . . - . . ~ . - - - - . - --~- . - - - -, - - - - -
i d
.i
l
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
I i :
QUESTION 16 POINT VALUE: 1.00 i
Unit One (1) is operating at rated power,.when the loss of an i electrical power distribution system results in the following Group 1 PCIS status light indications on P601:
Inboard MSIV DC solenoid Out Inboard MSIV AC solenoid Out Outboard MSTV DC solenoid Out !
Outboard MSIV AC solenoid Lit j
-
J What power distribution system has been lost? I Division I A Division I D I Division II A Division II D l QUESTION 17 POINT VALUE: 1.00 ;
Unit'One (1) is performing Alternate Emergency Depressurizatio RCIC :
has been placed into pressure control to aid in Reactor pressure l reductio !
Circuit alterations have been performed per EOP-01-RVCP and EOP-01-SEP-10 for HPCI and RCI Which of the following conditions, by interlock, will remove the RCIC l System from the pressure control mode of operation? ; Reactor pressure lowers to 50 psi Drywell pressure rises to 2.5 psi Reactor water level lowers to +110 inche Suppression pool level rises to -23 inche PAGE 9
.. ._ _ _ _ _ - . - _ _ _ _ - _ - . . . . - - _ . . . . . . - - _ . _ . . _ _ _ _ ~m . _ - .
l I
'
l
- "NRC 97-1 SRO, Rev 0" EXAMINATION ** ,
,5 i
QUESTION 18 POINT VALUE: 1.00 Unit One (1) failed to scram. RHR is placed in Suppression Pool Cooling per the Hard Card without use of overrides. Reactor water level is then deliberately lowered to suppress powe Plant conditions are-
'
l D Reactor water level -45 inches (NO36/N037)
Reactor pressure 800 psig Drywell pressure 0.6 psig Suppression pool temp 130*F Suppression Pool Cooling valves have close Returning Suppression Pool Cooling to service requires: placing the Think Switch to Manual onl bypassing the 2/3 core height interlock onl ! placing the Think Switch to Manual, then bypassing the 2/3 core I height interloc ! bypassing the 2/3 core height interlock, then placing the Think Switch to Manua l
-l
. QUESTION 19 POINT VALUE: 1.00 Consider the following normal RHR suction valve interlocks for control of Rea'ctor water level: Shutdown Cooling suction isolation valves (F008/F009) will automatically isolate on low Reactor water leve . Shutdown Cooling pump suction valve (F006) cannot be opened unless Torus common suction' valve (F020) is close During Shutdown' Cooling operation from Remote Shutdown stations per AOP-32.0, which of the above (if any) interlocks are functional? only, only, both 1 and neither i nor PAGE 10
l
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
QUESTION 20 POINT VALUE: 1.00 Following a small line break, RHR Loop A is placed in Drywell and Suppression Chamber Spray using required overrides per SEP-02 and SEP-0 Plant conditions are:
Reactor Water Level +175 inches Reactor Pressure 900 psig Drywell Pressure 15.0 psig Reactor water level drops to -60 inche How will RHR Loop A and RHR Service Water (RHR SW) Loop A respond?
The RHR Loop A drywell/ suppression chamber spray valves: auto close, RHR SW Loop A pump (s) tri auto close, RHR SW Loop A remains runnin remain open, RHR SW Loop A pump (s) tri remain open, RHR SW Loop A remains runnin !
l QUESTION 21 POINT VALUE: 1.00 Unit Two (2) is operating at 100% power with an active 7 day LCO for RHR Pump 2A being under clearance. An AO reports that the breaker for RHR Room Cooler 2B at MCC 2XB is tripped magneticall What action is required by Technical Specifications and by OI-01.08, Control Of Equipment And System Status? I Room roolers are not required for RHR Operability, continue in the active 7 day LC Room Coolers are required for RHR Operability, however continue in the active 7 day LC Declare RHR Loop B Inoperable and place the unit in HOT SHUTDOWN within 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Declare RHR Loop B Inoperable and place the unit in HOT l SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i PAGE 11
,
. . ._ . _ . . - . _ . _ - ._ -_..____.m._ _ . . - . _ . _ _ . .
,
f
! l L !
l
,
'
- ' "NRC 97-1 SRO, Rev 0" EXAMINATION ** l
)
!
QUESTION 22 POINT VALUE: 1.00 L
During normal power operation of Unit Two (2), an ECCS Division I Trip I Cabinet Trouble alarm is received. Investigation shows that BOTH '
l - power supplies to the trip cabinet - (XU-63) have failed and all l
associated trip unit meters indicate downscale with no trip lights lit.
l A DBA LOCA then occurs resulting in Reactor water level rapidly l
'
dropping below the Top of Active Fuel and rapid Reactor .
depressuriztion. How will Division I Low Pressure ECCS (Core Spray 2A and RHR LPCI Loop 2A) respond?
!
! Core Spray 2A initiates, LPCI 2A fails to initiate.
i i Core Spray 2A and LPCI 2A both fail to initiat Core Spray 2A fails to initiate, LPCI 2A initiates.
l Core Spray 2A and LPCI 2A will both auto initiat QUESTION 23 POINT VALUE: 1.00 i Unit Two (2) HPCI has automatically initiated on a valid initiation
! signal. The operator observes the following indications: ;
i Steam Supply Pressure O psig Turbine Exhaust Pressure O psig Pump Discharge Pressure 0 psig Turbine Speed 1000 RPM, lowering HPCI TURB TRIP SOL ENERG NOT Alarming j Which of the following would explain the above indications? Isolation due to ruptured exhaust diaphra l
. Overspeed trip from speed feedback signal failur c Loss of oil pressure to the turbine control syste Loss of 125 VDC input to the 24/52.5 VDC power supplie ! l
5
PAGE 12
!
I ._ ___ _ . _ _ _ . _ .
. _ _ _ . _ . _ _ . _ _ _ . - _ . . _ . _ . _ _ . _ _ . _ . . . . _ _ _ _ _ . _ . _ ~ _ _ _ . _ ~.
t
!
l
>
!
l
- "NRC-97-1 SRO, Rev 0" EXAMINATION **
~
l QUESTION 24 POINT VALUE: 1.00 l
Unit One (1) has experienced a high Reactor pressure transient following a-Main Turbine tri Current plant conditions:
! 'All rods in
. Reactor Pressure 950.psig controlled by EHC Eleven (11) SRV's green indicating lights lit Eight ' (8) amber memory lights illuminated I Determine the extent of the pressure transien psig psig psig d .' 1152 psig I
i l
i l QUESTION 25 POINT VALUE: 1.00 l
l Unit One (1) is operating at power with Core Spray Pump 1B under l clearance. A Loss of Off-site Power occurs to BOTH unit Diesel L Generators (DGs) : 2 and 4 auto start, DGs 1 and 3 trip and lockou A stuck open SRV and loss of high pressure injection causes Reactor l water level to lower.- All available low pressure ECCS pumps have been 1 manually starte Plant conditions are: }
Reactor water level LPCI/ Core Spray initiation' signal just ,
received ;
Reactor pressure 650 psig l t Drywell pressure 0.6 psig I ADS Inhibit Switches AUTO Assuming NQ. operator action, ADS will: Auto initiate'in 1 minute 23 second !
.
, Auto initiate in.1 minute 45 second I
- Not' auto initiate due to lack of high drywell pressur !
- Not auto initiate due to lack of ECCS pump permissive.
l l \
l PAGE 13 l
. . . -- . -. - _ . ._
-- -_ -. . . ~ . .. . - . . - . . .. ~ ~..~ . _ . - . - . . - - .
,
!
!'
i I
i ** "NRC 97-1 SRO, Rev 0" EXAMINATION **
i <
'
i l
QUESTION 26 POINT VALUE: 1.00 )
Unit Two (2) . has a Loss Of Of f-Site Powe HPCI and RCIC both faile j Reactor Water Level dropped to +30 inches before CRD reversed the level ,
tren Current conditions are: ,
I Reactor water level +60 inches, rising Reactor pressure 800-1000 psig using SRVs ,
Drywell pressure 2.2 psig, rising i Average drywell temp 180*F, rising Generator primary lockout Tripped i
RBCCW Pumps are tripped, and Nuclear Service Water (NSW) cooling water valves (SW-V103/V106) are closed. What actions are required to restore RBCCW to control containment parameters? Reset the Primary Generator lockout, align RBCCW cooling to the conventional heade Reset the Primary Generator lockout, reopen NSW cooling water valves SW-V103/V10 Reset Core Spray initiation logic, align RBCCW cooling to the conventional heade Reset Core Spray initiation logic, reopen NSW cooling water valves SW-V103/V10 )
l
!
!
l l
!
!
!
PAGE 14
. . . . _ _ . .- _ .__. _ . . - _ __ . . _ _ _ . . _ , _ . . _
.
- "NRC 97-1 SRO, Rev 0" EXAMINATION ** i
!
-QUESTION 27- POINT VALUE: 1.00
'
.
.
' Unit Two (2) is operating steady state at rated power with the following Electro Hydraulic Control (EHC) conditions:
Reactor pressure 1005 psig EHC Pressure Setpoint 920 psig PAM pressure 950 psig .
Pressure. regulator A In control ,
Pressure regulator B 5 psig bias
,
The Pressure Averaging Manifold (PAM) pressure input to Pressure Regulator A fails low. The PAM pressure input to Pressure Regulator B is unaffecte How will the EHC System respond? Pressure. regulator B takes control and stabilizes PAM pressure at 945 psi Pressure regulator B takes control and stabilizes PAM pressure ,
at 955 psi ~ Control valves close, reactor pressure and neutron flux rise and the reactor scrams, Control / Bypass valves open and steam line pressure lowers to the Group 1 isolation setpoin i l
l l
,
l
!
I i-
!
l PAGE 15 ;
i
)
_ _ . . . . _ . _ . . _ . _ ._ _ _ _ _ _ _ _ _ _ _ . _ _ . _ . _ .._ _ __..- _ ._ .
i
'
l
. . <
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
i l
QUESTION 28- POINT VALUE: 1.00 Unit one (l') is operating at 100% power with the'following Condensate system alignment:
Condensate pumps A & B Running Condensate pump C Standby / Auto Condensate Booster pumps B & C Running Condensate. Booster pump A Under clearance 4KV BOP Bus 1D trips and locks out due to a bus fault. The transient results in a Reactor scram. What is the availability of the Condensate system to provide makeup to the Reactor vessel?:
Condensate pump:
, A is available, no Condensate Booster pump is availabl eB is available, no Condensate Booster pump is availabl ,
l A is available and Condensate Booster pump C is availabl ! B is available and Condensate Booster pump C is availab'l .
QUESTION 29 POINT.VALUE: 1.00 l Unit Two (2) is operating at 100% powe The Digital Feedwater Control-System (DFCS) is aligned as follows:
i Master Controller Auto, set at 187 inches !
Level instrument N004A 187 inches !
Level instrument N004B 187 inches l Level instrument N004C failed downecale Mode Select switch 3 Element !
Level Select switch Level A !
Level.-instrument N004A fails downscal Assuming no operator action, i Reactor water level will: rise and flood the main steam line !
i drop to the low level scram setpoin rise resulting in a main turbine and' feed pump tri i
! remain at 187" with level instrument N004B in contro PAGE 16 ,
l
. _- _ _ _
. _ , _ _ _ _ . . _ . _ - _ .. .. .- -.~. _ m . . . _ .., _ _ . - .. .. _ _ _
<- ,
, . .
>
l i I l
'
i
- ** "NRC 97-1 SRO,'Rev 0" EXAMINATION **
i k
.
QUESTION 30 POINT VALUE: 1.00 Unit Two (2) .is in operation at 100% powe The following annunciator and conditions exist:
A-07, 1-2 RFP FW CONTROL SIGNAL FAILURE I Amber light above A RFPT lockout switch out Amber light above B RFPT lockout switch lit f j
A Reactor Recirculation Pump trip Reactor power lowers and Reactor l water level begins to rise uncontrollably.. How can the operator i restore Reactor water level to the normal band? I Operate RFP A MSC control in the Lower directio Operate RFP B MSC control in the Lower directio ;
I Place RFP A-MGU in Manual and lower output deman Place RFP B MGU in Manual and lower output deman l QUESTION 31 POINT VALUE: 1.00 The Control Building Ventilation system has initiated in the radiation protection mode. The alignment of system controls is:
!
Emergency Filtration Fan A Pref i Emergency Filtration Fan B Stby
. Emergency Filtration Fan A has been running for 10 minutes when a high temperature is sensed in the Train A charcoal bed. Assuming the radiation initiation signal is still present, Emergency Filtration Fan A will: immediately trip and Fan B will immediately auto star remain running since the high temperature trip is bypasse ; immediately trip and Fan B will start after a 10 second dela remain running since the high temperature provides alarm onl l i
e l
PAGE 17
- "NRC 97-1 SRO, Rev 0". EXAMINATION **
QUESTION 32 POINT VALUE: 1.00 A. Loss of Off-Site Power has occured. Secondary Containment. isolate Reactor Building Ventilation was restarted by guidance of EOP-03-SCCP using SEP-0 Plant conditions are:
Reactor Water Level is +150 inches, slowly rising
'
Drywell Pressure is 1.2 psig, slowly rising CAC Vent Purge Isol Ovrd (CAC-CS-5519) is in OVERRIDE Reactor Building Vent Rad Monitors have been reset PCIS Isolation Reset push buttons on P601 have been depressed Which of the following would cause the Reactor Building to re-isolate? Drywell pressure rises above 2.0 psig,
~ Reactor level drops to the Top Of Active Fue Main Stack Radiation Monitor exceeds the Hi-Hi setpoin Reactor Building Vent Exhaust temperature exceeds 140' QUESTION 33 POINT VALUE: 1.00 ,
l Following a Loss of Off-Site Power, Diesel Generator #1 (DG1) is running in AUTO, tied to Bus E1. DG1 parameters:
Kilowatt load 3500 KW Terminal Voltage 4160 Volts Reactive load 1300 KVAR Frequency 60 Hz Off-Site Power has been restored and the BOP buses energized from the SAT. The BOP bus to El Master / Slave breaker is still ope .The~ operator depresses the DG #1 CONTROL ROOM MANUAL push button'on ,
the RTGB. DG #1 frequency will be approximately: I H i H ' H H PAGE 18 l
_ _ _..._-- _ .. _ . _ ___ _ _.- _ . _ . _ .-.- _ _ - ~ > - - - - -
'
'
l
'
,
j ** "NRC.97-1 SRO, Rev 0" EXAMINATION **
!
QURCTION 34 POINT VALUE: 1.00 The following sequence of events occurs on Unit One (1):
.
Time = 0 seconds Off-site power is lost
. Time = 5 seconds A LOCA signal is recieved Time = 10 seconds Diesel generators energize thier respective E Buees The Motor Driven Fire Pump normal feeder breaker from: Bus El closes-at Time = 25 seconds, Bus E2 closes at Time = 25 seconds, Bus El closes at Time = 30 second Bus E2 closes at Time = 30 second j i
I QUESTION 35 POINT VALUE: 1.00 UnitLOne (1) is in an outage with the SAT-energized and the UAT in backfeed alignment. All backfeed selector switches are in the BACKFEED position. Electrical system alignment:
BOP Buses 1C/1D Powered from UAT E Buses E1/E2 Powered from BOP Buses DGs 1, 2, 3, 4 Operable in standby alignment ,
" !
A sudden' fault pressure occurs in the' Main Power Transformer resulting I in a Backup Main Generator lockou How vill the electrical ;
distribution system respond? I i BOP buses 1C/1D transfer to the Sl1T, four DGs receive an auto i start signa I BOP buses 1C/1D transfer to the f,AT, no DGs receive an auto j start signa i-BOP buses 1C/1D are de-energized., four DGs receive an auto
~ start signa i BOP buses IC/1D are de-energized, DGs 1 and 2 only receive an auto start signa .l
l PAGE 19
- . .- -.
. - . __ . ~ ~ . . . - _ - . _ . - - . . -..- - . _. - -. - . ..-- . . ~ - _ - . _
I l
,
!
.
,
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
' l i
! QUESTION 36 POINT VALUE: 1.00 4 l
.i '
- Unit One - (1) is in a refueling outage with the SAT de-energized and a !
,
UAT backfeed in service. Unit Two (2) is operating at 100% power. The l following Diesel Generator start times to rated speed are recorded !
during PT-12.8:
4-DG1 9.8 seconds
-
DG2 10.2 seconds- I i DG3 9.3 seconds l DG4- 9.6 seconds l 1 l
What is the maximum time Unit Two (2) may continue POWER OPERATIONS without entering a Technical Specification shutdown Statement? i 5-I 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> i
!
<
b.' 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> )
'
!
- day !
- day l l
i
.
QUESTION 37 POINT VALUE: 1.00 i i
A "250 V BATT A GROUND" alarm has been received on Unit One (1).
'
The j following readings are reported from tht. Battery Room: )
.
,
f P Bus milliamps ;
j N Bus milliamps j
Charger 1A-1 135 volts, in float l
Charger 1A-2 135 volts, in float Per OP-51 and AI-115, the ground is on the:
l P Bus, action level 1 applie b P Bus.,. action level 2 applie I N Bus, action level 1 applie d. <N Bus, action level 2 applie ;
!
PAGE 20
-
i
!
!
! l l
'
!
I
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
l QUESTION 38 POINT VALUE: 1.00 I
Unit One (1) is operating at 100% powe The UPS system is in its normal alignment for both units. A total loss of UPS occurs on Unit One (1), followed shortly by a spurious Reactor scra Plant conditions:
Reactor water level lowers to +135 inches Reactor pressure 950 poig controlled by EHC APRM recorders on P603 indicate 100% power Digital Feedwater controller displays are blank How will the loss of UPS affect the plant during this transient?
a. Reactor power cannot be determined to be less than 3% from P60 Reactor feed pumps will not respond to the reduced Reactor water leve c. EHC pressure control will be lost as the main turbine coasts down to zero spee d. Reactor Building HVAC is lost until the main stack rad monitor is transferred to Unit Tw l QUESTION 39 POINT VALUE: 1.00 An Auxiliary Operator has received 1.95 rem TEDE for the current yea The Ao is needed to perform work in a 25 mrem /hr fiel The work is expected to last 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> In accordance with NGGM-FM-0002, Radir. tion Control and Protection Manual, the worker requires:
a. approval from the Manager - E&RC to exceed the annual administrative dose limi b. approval from the Plant General Manager to exceed the annual administrative dose limit.
l c. approval from the Site Vice President to exceed the annual administrative dose limit.
,
d. no special authorizations since the annual administrative limit should not be exceede PAGE 21
. . - .. . . _ . . . _ - _ - - _ - - - -
. . . . - - . . . . - . ... . - - ..- . _ .~... - . _ .
t
,
!
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
l l
QUESTION 40 POINT VALUE: 1.00 I Both Units have lost off-site power. The only available Diesel .)
Generator is DG2. Buses E2 and E4 cannot be cross-tie Unit Two (2)
UPS has been de-energized for DC load strippin What instruments are available to monitor Reactor Water Level on the Unit Two (2) RTGB? Fuel Zone' indicator NO36 onl Fuel Zone indicator NO37 onl Fuel Zone indicator NO36 and Narrow Range indicators N004A/B/ ~ l Fuel Zone indicator NO37 and Narrow Range indicators N004A/B/ QUESTION 41 POINT VALUE: 1.00 1
.
~
You are escorting a visitor with a red badge in the protected are It becomes desired to temporarily give up your escort dutie How may this be accomplished? You and the visitor must exit the protected are You may turn over escort duties to a security guard onl You may turn over escort duties to any other qualified escort, and notify security at the access point you entere You may turn over escort duties to any other qualified escort, ,
and notify security at the secondary alarm statio i
-
!
I l
l t
, i f
PAGE 22
-- , _
_ = _ _ _ . _ -. , . _ .
.. .. - . . . _ - . - . - _ ._ -..
,
"NRC 97-1 SRO, Rev'0" EXAMINATION F
- ** ,
QUESTION 42 POINT VALUE: 1.00 I-During operation of Unit One (1), a line break in the drywell results l in a Reactor' Scram. All control rods fully insert and immediate '
operator actions are complete Reactor Water Level rapidly drops off-scale low on the Fuel Zone instruments prior to Low Pressure ECCS initiating and restoring adequate core coolin Technical Specifications requires that the NRC Operations Center be ,
notified within one hour, and the: ! Plant Manager - Brunswick Nuclear Plant within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Plant Manager - Brunswick Nuclear Plant within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Vice President - Brunswick Nuclear Plant within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Vice President - Brunswick Nuclear Plant within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> QUESTION 43 POINT VALUE: 1.00 Which one of the following sets of conditions meets the requirements for. OPERABILITY of the Standby Liquid Control System for Unit Two (2) ?
TANK SOLUTION LEVEL CONCENTRATION TEMPERATURE '3150 gal 16.0% 70 F gal 14.5% 75 F gal 13.0% 80*F gal 15.0% 70'F PAGE 23
... . .. _-. . . . . . - . . . . . - . . . - - - - . - - - - . ..- - . .. - - . > - - - - - -
!
,
'
,
,
i
- "NRC 97-1 SRO, Rev 0" EXAMINATION ** !
i QUESTION 44 POINT VALUE: 1.00 A Temporary Procedure Change is developed and designated as Revision To-Follo This Temporary Change receives:
Interim approval on March 2nd Final approval on March 6th
)
What is the last date this Temporary change may be used WITHOUT, receiving any allowable extention(s) ? May 1s May 5t May 31s June 4th.
t
!
QUESTION 45- POINT VALUE: 1.00 Which of the following describes the level of use of Emergency l Operating Procedures? Reference use procedure Continuous use procedure Information use procedures.
- Exempt from level of use requirements.
I l
I
l
,
!
4 l
s
PAGE 24
_ ,, _ ., . .. . ._. . _ . . _ _ _ . _ _ ._ .
_ _ . . _ . . . ..__._. ,
'
i a
<
i I
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
,
!
QUESTION 46 POINT VALUE: 1.00 Per PLP-21, which of the following would be an UNACCEPTABLE method of
.
performing independent verification by use of RTGB indications?
Independent verification of a Core Spray System:
a. pump breaker closure using red light indication.
- b. pump breaker rack in status using green light indicatio injection valve opening using system flow indication mete injection valve standby position'using green light' indicatio ,
!
- QUESTION 47 POINT VALUE
- 1.00 l l
,
A Station Blackout requires simultaneous execution of the Emergency i
- Operating Procedures, Emergency Plar Procedures and AOP-36.2.
The Work Control Center SRO reports to the Control Room functioning in a dual role as the Shift Technical Advisor (STA).
] Per OI-01.01, since the STA holds an active license, the STA may be
', allowed to:
' function as the Site Emergency Coordinator.
. install circuit alterations required by the EOPs.
- silence and acknowledge, but not reset annunciators.
i
!
d. direct other operators in AOP-36.2 activities while the Unit
! SCO directs EOPs.
f i
)
<
^
PAGE 25
- _
.. . . . . ,_ . . . .._- _ _. . - . - - . _ _ . . . . - - ~ - - , - _ . - . - _ - -
!v l
l
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
QUESTION 48 POINT VALUE: 1.00 j During RTGB walkdown in preparation for shift turnover, the oncoming i Reactor Operator notes that annunciator A-5 2-2, Rod Out Block:
i l Alarm is sealed in l Has a yellow dot affixed to the window l-j If the signal causing this alarm clears, then comes back in without
the alarm being reset, the subsequent alarm condition is indicated by ,
j' the alarm window flashing:
L slowly with an audible alar ,
l l
' rapidly with an audible alar '
j slowly, then rapidly without an audible alarm.
!
a l rapidly, then slowly without'an audible alar l l QUESTION 49 POINT VALUE: 1.00 A plant shutdown is in progress per GP-05 for refuelin Current plant conditions are:
Reactor power 20% !
l Drywell oxygen 19.0%
!
Drywell entry may he made with the approval of the Manager-E&RC and the authorization of the Shift Superintendent only if reactor power is '
reduced at least-l %, oxygen concentration is acceptabl '
l %, oxygen concentration must be raised, j %, oxygen concentration is acceptabl ;
, d. '15%, oxygen concentration must be raised.
!
!
!.
!
l l
l PAGE 26
,
-
. . . _ _ . - ~ = - . .
.
-. ... . . . - . ~ . .
,
L e , i I l t
i l
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
'
!
QUESTION 50 POINT VALUE: 1.00 Which of the following valve / actuator types.may be used'as a clearance
. boundary isolation component, provided the associated restrictions of
! AI-58 are satisfied?
! Solenoid operated ball valve marked as fail-closed on the l~
print.
I Motor operated globe valve normally used as a flow control
! valv Pressure balanced diaphragm operated pinch valve not equipped with a handwheel.
l Double acting cylinder operated butterfly valve not marked as l fail-closed on the print.
i l
- QUESTION 51 POINT VALUE
- 1.00 l
l
'
A clearance request has been received for a fluid system with the following normal operating parameters:
!
l Pressure 475 psig
! Temperature 175oF Per AI-58, the clearance: 1 ( May use single valve boundary isolation.
l
! Must use dual valve boundary isolation due to pressure only.
l l Must use dual valve boundary isolation due to temperature only.
l
'
- Must use dual valve boundary isolation due to pressure and !
temperatur I i
!
i l
PAGE 27
! , _ - , __ _ ._. ._
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
QUESTION 52 POINT VALUE: 1.00
, 1 During accident conditions on Unit One (1), the following plant conditions exist:
"
Reactor water level -7G inches (Fuel Zone) 1 Reactor pressure 1100 psig Drywell average temp 190*F Drywell ref leg area temp 270*F Injection sources None available Under these conditions, peak fuel clad temperature will not exceed: *F, provided Reactor water level remains above -80 inche *F, provided Reactor water level remains above -90 inche F, provided Reactor water level remains above -80 inche F, provided Reactor water level remains above -90 inches.
<
QUESTION 53 POINT VALUE: 1.00 Following accident conditions, the crew is executing the Reactor Vessel Flooding Procedure, EOP-01-RXFP. Plant conditions are:
Control rods Fully inserted Reactor water level Unknown The operator is directed to control injection flow to the Reactor to maintain at least SRV/ ADS Valves open and Reactor pressure: ; above the Minimum Alternate Flooding Pressur ; above the Minim'.m Alternate Flooding Pressur ; at least 50 psig above suppression chamber pressur ; at least 50 psig above suppression chamber pressur :
PAGE 28
. _ ._ .. . _ . _ __ _ .. .._-m.. __ ..._ . ___ _ .___ _ . . _ . _
i i
i
<
I
-
f
.
'
.
- "NRC 97-1 SRO, Rev 0" EKAMINATION **
L
!
l !
'
QUESTION 54 POINT VALUE: 1.00 i
A-heavy influx of marsh grass on the Circulating Water Screens has 1
.!
'
caused a loss of all Circulating Water pumps and a reactor scram.
'
Plant conditions are: :
i
",
!
c Group 1 isolated !
! Condenser vacuum is 0" Hg !
Turbine speed.is 500 rpm, dropping '
EHC Electrical Malfunction in alarm due to loss of the'PM l
\
The marsh grass is now cleared and the Circulating Water System has !
Is the Main Condenser available as a heat sink?
'
been restarte l
, No , the MSIVs are close !
i j No, the EHC system is not availabl '
, No, the condenser is not under vacuu !
k i Yes, .all required systems are availabl !<
.
'
,
. .
QUESTION 55 POINT VALUE: 1.00
- .
i i A Reactor scram occurs on low Reactor water level following a loss of -
- feedwate The following sequence occurs
- !
i
0100 Reactor scrams
0115 Reactor Vessel Control Procedure is enterred -i
0130 Circuit alteration installed per RVCP for HPCI l t- 0145 EOPs are exited !
l 0200 HPCI alteration restored per SEP-08 !
Y l
The plant is at 900 psig with the Mode Switch in SHUTDOWN. The SRO should initiate a WR/JO to functionally test the altered HPCI circuit I
and. initiate:
I an active LCO with a start time of 013 '
i b. 'an active LCO with a start time of 014 ;
J a tracking LCO with a start time of 013 , a tracking LCO with a start time of 0145.
4 PAGE 29 ,
.
. . - - .
. .___. . _ _ , . . ._ ._..____ ._,. ._ _ . _ _ _ _ - _ . . _ .
-
i
!
i
!
! !
.
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
I
<
QUESTION 56 POINT VALUE: 1.00 -
t Following a loss of all high pressure injection on Unit one (1), seven [
ADS valves have been manually opened to restore adequate core coolin i-Plant conditions are now: j i
i Reactor water level +25", N026A/B :
Reactor water level +140", NO36/37 Reactor pressure l
,
'
25 psig ~
Drywell average temp 155*F -
, Drywell ref leg area tem *F i Reactor water level may be determined using: *
, N026A/B onl NO36/37 only.
- Both NO26A/B and NO36/3 j
. .
e
- Neither NO26A/B nor NO36/3 {
-
!
QUESTION 57 POINT VALUE: 1.00 ,
A Unit Two (2) reactor scram has occurre Seven control rods failed
- to fully insert and are between positions 08 and 18. Conditions are
- '
'
,
All APRM Downscale lights are LIT -
l MSIVs are open !
!. fotal Steam Flow 3.6 E6 lbm/Hr, dropping '
Reactor Pressure 900 psig, dropping Narrow Range Level Instruments (N004s) +155 inches, rising
.
'
Master Feedwater setpoint at +170" '
Two Reactor Feed Pumps in operation I
~
With current plant conditions, the operator is required as an
,
IMMEDIATE action to: , trip the Main Turbin ;
- ' trip one Reactor Feed Pum i place the Mode Switch to SHUTDOW f- enter Alternate Control Rod Insertion.
'
[
PAGE 30
,
!
!
-- , - - . ..
. . . . ~ . . . . ~ . _ - . . _ - , - - . - -.. . . - . . . ~ . , - . - - - - . - - - .
l l
!
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
!
l
I QUESTION 58 POINT VALUE: 1.00 ;
Following a group 1 isolation and a reactor scram,.the operating crew is performing the Reactor Scram Procedure, EOP-01-RS Plant l conditions are:
l l Reactor water level 195 inches, slowly rising (N004s)
Reactor pressure 800-1000 psig, controlled by SRVs :
Drywell pressure 1.0 psig, slowly rising l Suppression pool temp 94*F, slowly rising !
Suppression pool level -27.5", slowly rising The operating crew is recuired to enter EOP-01-RVCP and execute 1 l concurrently with the scram procedure if:
l drywell pressure rises to 1.5 psig.
i
' ! reactor water level rises to +230 inche suppression pool temperature rises to 111' suppression pool level rises to -26.5 inche I
QUESTION 59 POINT VALUE: 1.00
l Following a Unit One (1) Reactor scram, the crew has entered and is l executing EOP-01-RSP, Reactor Scram Procedure. Plant conditions are:
l Reactor water level- 220", rising l
! Reactor pressure 945 psig, stable j l MSIVs Open i :
Per EOP-01-RSP, the MSIVs must be manually closed if Reactor water level cannot be maintained below: " "
- ' "
,
- "
'
<
!
l l PAGE 31 i
..- , _ . , ., _ _
.. .
- . . - . . . . . . . . - . - . . . ~ - . . - _ _ ~ -. - - - - ~..,~ . - ~ - .
- - - _ - - . -
!
!
- "NRC 97-1 SROJ Rev 0" EXAMINATION **
( QUESTION-60 POINT.VALUE: 1.00 The entry conditions for Unit One (1) EOP-01-RVCP, Reactor Vessel Control Procedure for Reactor pressure and water level are Reactor pressure is greater than: psig, or Reactor water level less than 153".
7 psig, or Reactor water level less than 166", psig, or Reactor water level less than 153". psig, or Reactor water level less than 166".
QUESTION 61 POINT VALUE: 1.00
)
During an ATWS on Unit One (1), all Bypass Valves were. full open !
together with one SRV for pressure contro Conditions required injection to the Reactor vessel to be terminated and prevente Current 1 conditions are:
I
'
Reactor power 22%, dropping i Reactor water level +125 inches, dropping Reactor pressure 952 psig Suppression pool temp 115 'F, dropping Drywell pressure 0.5 psig, dropping Bypass valves 3 1/2 open SRVs All Closed SLC tank level 66%
What action is required? Continue to lower level until Reactor power is below 3% or Reactor water level reaches TA Continue to lower level until Reactor power is below 3% or Reactor water level reaches LL Establish a level band no higher than +125 inches and no lower than TAF.
, Establish a level band no higher than +125 inches and no lower l
than LL4..
I l
l PAGE 32
l
.- -
.- . -
.- - . - . , - _ ,.
- - --
l
'!
l l
- "NRC 97f1 SRO Rev 0" EXAMINATION
, **
QUESTION 62 POINT VALUE: - 1.00 During an ATWS on Unit Two (2), a Safety Relief Valve fails ope The following plant conditions-exist:
Reactor Power APRMs downscale Reactor water level -60 inches (Fuel zone)
. Reactor pressure' 700 psig, lowering Drywell ref leg area temp 187'F HPCI NOT available RCIC, CRD, SLC Injecting Assume Reactor water level remains constant. The Reactor must be
- Emergency Depressurized if Reactor pressure drops to: psig, RCIC may inject during the depressurizatio psig, RCIC may inject during the depressurizatio , psig, RCIC must be terminated prior to opening ADS valve .- psig, RCIC must be terminated prior to-opening ADS valves.
i
!
QUESTION 63 POINT VALUE: 1.00
Following an incomplete Reactor scram, the operating crew is executing
- EOP-01-LPC, Level / Power Contro A decision step is reached asking i "Is The Reactor Shutdown?".
. Which of the following conditions would satisfy the definition of l " SHUTDOWN" as it applies to the Reactor?
., All operable APRMs indicate downscal The Reactor is subcritical on range 6 of IRMs.
i~
! The entire SLC Tank has been injected to the Reactor.
!
- Hot Shutdown Boron Weight has been injected to the Reacto '
.
I !
l PAGE 33 ,
.
4' .
.
ep - e+- >- p w q .- -m .y--s .---g-p .a- + = , aw., . -+sW-7--g .p--y - + .e, w r.-w_q g- -.-w
. . . . - . . . . . . . - . .. . . , -- - -- - - .- - . _ - -
!
i i
!
- "NRC 97-1 SRO, Rev 0" EKAMINATION **
,
i i
,
QUESTION 64 POINT VALUE: 1.00 l i
During severe accident conditions on Unit Two (2) ,- the following ;
sequence of events occurs: '
0100 Reactor scrams 0630 Reactor water level undetermined '
0700 Reactor flooding conditions initially established 0930 Minimum core flooding interval is satisfied 1600 Injection to-the Reactor is terminated At what time must injection be re-established if Reactor water level cannot be determined? l l ' QUESTION 65 POINT VALUE: 1.00 Following a reactor scram and a group 1 isolation, SRVs are being used to maintain reactor pressure 900-1000 psi Which'of the following conditions requires ALL group 1 isolations to be defeated and the reactor vessel rapidly depressurized to the main condenser? Suppression Pool Level is +4' 6" Suppression Pool Temperature is 95'F l Suppression Pool Level is -l' 6" i Suppression Pool Temperature is 170*F Suppression Pool Level is -4' 3" Suppression Pool Temperature is 156*F l
d ". Suppression Pool Level is -8' 1"
-Suppression Pool Temperature is 105'F
PAGE 34 i
_, ._ _ . ,
. - . - - _ _ . _ . _ _ _ . _ . . _ _ . _ _ _ . __ _ _ _ _ . .__ . _ . . . _ . .
!
>
!
!
!
!
- "NRC'97-1 SRO, Rev 0" EXAMINATION ** !
i
-
l
>
QUESTION 66 POINT VALUE: 1.00 i
!
Following a large Recirculation line rupture, EOP-01-PCFP, Primary i
,-s Containment-Flooding Procedure, is being.exectuted. The following :i
, indications are available: I v '
,
"
CAC-LI-2601-1 +5.9 feet CAC-PI-1257-2A 23 psig l CAC-PI-1230 21 psig CAC-PI-4176 25 psig CAC-PR-1257-1 22 psig l'
What-is Primary Containment water level? +14.5 feet l l +9.9 feet I +7.6' feet
!
' +4.6 feet QUECTION 67 POINT VALUE: 1.00 During accident conditions, the operating crew is executing the Primary Containment Flooding Procedure, EOP-01-PCF This procedure requires that the Inboard Steam Line Drain Isolation
. Valve (B21-F016) be disabled (breaker opened) in the:
l open position prior to Primary Containment water level reaching
'
i 21 fee l l open position prior to Primary Containment water level' reaching l 23 feet.
i
! closed position prior to Primary Containment water level L reaching 21 feet.
l j closed position prior to Primary Containment water level
- reaching 23 feet.
!-
i I:
i
!
PAGE 35 l l
!
.- . ._- _ - . ,,.
_. .m __ .. . ... _ _ _ _ . - _ . . . _ _ _ - . _ - _ . _ _ _ . _ . - _ .__ _ ..._ . .-
'
-
l I i
- I j ** "NRC 97-1 SRO, Rev 0" EXAMINATION ** j l
QUESTION 68 POINT VALUE: 1.00' j i
A seismic event has occurred that has resulted in a Loss of Off-site l Power and high power ATWS condition The SLC Storage Tank outlet line completely severed at the tank during
-
i the earthquake. The SLC tank is EMPTY making the SLC pumps unavailable'for boron injectio Which system should be selected for alternate boron' injection? j
, CRD
, Condensate
4
.
QUESTION 69 POINT VALUE: 1.00
.i
, A Condensate header rupture in the cable spread area of the Control
,
Building has resulted in a loss of all UPS and RPS powe !
-
- Plant status is as follows
- '
. Blue scram lights 137 illuminated I
IRM Indications 50 on Range 10
, What method of EOP-01-LEP-02, Alternate Control Rod Insertion, would be
,
MOST effective in inserting the withdrawn rods?
. Vent the scram air heade ' Vent the overpiston area of control rods.
1 Scram individual rods with the scram test switche ~
Insert control rods with the Reactor Manual Control Syste !
d i !
- )
0 ;
!
PAGE 36
!
.-.
i l
l
- "NRC 97-1 SRO, Rev 0" EXAMINATION ** i i
i l
QUESTION 70 POINT VALUE: 1.00 During a low reactor water level condition, Alternate Coolant Injection using demineralized water is being aligned using the HPCI system. A i valid HPCI isolation signal is present, resulting in an automatic closure signal to the HPCI Injection valve (E41-F006). l How is the HPCI Injection Valve (E41-F006) opened to provide injection to the reactor?
E41-F006 is opened from the: RTGB after placing the HPCI ASSD Interlock Defeat Switch on the l RTGB to BYPAS MCC by placing the breaker's NORMAL / LOCAL switch to LOCAL to bypass valve interlock RTGB and the breaker at the MCC is opened by an AO when the ;
valve indicates full ope l RTGB after jumpers are installed to bypass the valve auto closure interlocks.
QUESTION 71 POINT VALUE: 1.00 Following a loss of drywell cooling, a small steam leak in the drywell results in the following containment conditions:
Drywell pressure 9 psig, rising Suppression chamber pressure 8 psig, rising Suppression pool level +2 feet Average Drywell temp 270 F, rising The crew is directed to initiate drywell spray to control drywell temperature. Under current plant conditions, drywell spray may: be initiated, all required conditions are me NOT be initiated, suppression pool level is too hig NOT be initiated, suppression chamber pressure is too low, NOT be initiated, conditions are in the UNSAFE region of the Drywell Spray Initiation Limi PAGE 37
. - . - . . . _ . - . - . . . .. . . .- - - . ~ _ - -. -
i
.
- <
j I
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
$ QUESTION 72 POINT VALUE: 1.00 The Suppression' Chamber Spray Initiation Pressure is in the .i
,
Suppression Chamber and is based on: ' .7 psig; intrusion of air into primary containment due to i Reactor Building-Torus vacuum breaker operatio .7 psig; the lowest suppression chamber pressure that RHR system logic will allow sprays to be initiated.
' psig; 95% of the noncondensibles in the drywell have been transferred to the suppression chamber airspace.
4 psig; the highest pressure that initiation of sprays will prevent exceeding the Pressure Suppression Pressure !
l Limit.
l QUESTION 73 POINT VALUE: 1.00
,
'
During'an accident on Unit One (1), the following primary containment and plant conditions exist:
l Reactor pressure 798 psig Suppression pool level -42 inches Suppression pool temperature 171*F
, Suppression chamber pressure 17 psig
' Current conditions are in the:
- SAFE region of all Containment Limits.
4' i' UNSAFE region of the Heat Capacity Level Limi UNSAFE region of the Heat Capacity Temperature Limit.
, UNSAFE region of the Pressure Suppression Pressure Limit, i
- ..
a
.
PAGE 38 i
- - .. .. . . - - . - - . ~ - - .. . - . - - - . - . - . . . - - ~ . . . - - . ~ . . -
!
'
i
.
I l
'
'** "NRC 97-1 SRO, Rev 0" EXAMINATION ** *
,
"
!
,
QUESTION 74 POINT VALUE: 1.00 l
, +
!
During accident conditions on' Unit Two (2), plant conditions are:
, Reactor level Below TAF j React or pressure 500 psig !
HPCI and RCIC injecting at rated flow :
- Suppression pool level -6.6 ft ;
i Drywell Temperature 310*F (Average)
Drywell H2 concentration 5.1% (Compensated) i s Drywell O2 concentration 6.1% (Compensated) j The operator should perform Emergency Depressurization and: . Vent the drywell irrespective of radiation releas ! Initiate drywell sprays irrespective of adequate core coolin Vent the suppre;sion chamber irrespective of radiation ~ releas Terminate HPCI injection irrespective of adequate core coolin QUESTION 75 POINT VALUE: 1.00 A primary system discharging into Secondary Containment has resulted in ,
one area exceeding the Maximum Safe Operating Radiation Level, but j within the EQ envelop. .The radiation level-in this area is l subsequently reduced below the Maximum Safe valu j A second area subsequently exceeds its Maximum Safe Operating' Radiation Leve What action is required by the Secondary Containment Control Procedure? ; Shutdown the Reactor per GP-0 Scram the Reactor and initiate a cooldown s100 F/ Hou I Scram the Reactor and initiate a cooldown >100*F/ Hou I
PAGE 39
., _ _ . . - - _ _ _ _ . _ _ . _ _ _ . - . . _ _ __ _ _ - . _ . _ _ ~ . . _ _ . - . _ . _ . _ . _ . - . _ _ _ _ . _ _ _ _ . _ . _
.
!
!
!
l
'
i
!
- "NRC 97-1 SRO, Rev 0" EKAMINATION ** !
I
QUESTION 76 POINT VALUE: 1.00 ;
!
While performing PT 9.2, HPCI OPERABILITY TEST, the HPCI. steam supply j line ruptured. HPCI failed to automatically isolate and attempts to j manually isolate HPCI are unsuccessful, i i
l The following Steam Leak Detection NUMAC channels are in alarm:
l B21-XY-5949A, Channel A3-3, reading 303*F B21-XY-5949B, Channel A3-3, reading 298oF 1 B21-XY-5948A, Channel A5-1, reading 301'F l B21-XY-5948B, Channel AS-1, reading 296*F f No other channels are in alar What action is recuired to be be taken? Scram the reactor and commence a cooldown at normal rate Shutdown the reactor using GP-05 or scram the reactor as directed by the Shift Supervisor, Scram the reactor and emergency depressuriz Scram the reactor and rapidly depressurize to the main L condenser.
!-
!
l r
'
i l
l
',
.
'
,
PAGE 40
)
. .- . - -. -.1
. . _ . _ _ . _ . - . _ . . . _ _ . _ _ _ . . _ . . . . _ _ _ . _ . _ _ . . - . _ _ . _ . . _ _ . . . _ . . _ _ . .
,
.
[
i :
,
- "NRC 97-1 SRO, Rev 0" EKAMINATION ** '
1 I l
.
QUESTION 77 POINT VALUES 1.00 '
i
'
Unit Two .(2) is operating at power when a rupture of RWCU piping I downstream of the Non Regenerative Heat Exchangers occurs. RWCU l
.
Inboard Isolation Valve (G31-F001) and Outboard Isolation Valve !
j (G31-F004) BOTH fail in the open positio Plant conditions:
'
Rx Bldg 50' temp 135*F Rx Bldg 20' temp 105'F I S Core Spray Room Flood Level Hi Hi alarm aealed in S RHR Room Flood Level Hi alarm sealed in The' operating crew is required to enter EOP-03-SCCP and: continue attempts to isolate the leak, commence an immediate plant shutdown per GP-0 continue attempts to isolate the leak, scram the reactor when 50' temperature exceeds 140* immediately scram the reactor and consider anticipation of emergency depressurizatio immediately scram the reactor and open seven ADS valves for emergency depressurization i
!
i l
I PAGE 41
,
- l
"NRC 97-1-SRO, Rev 0" EKAMINATION ** '
QUESTION 78 POINT VALUE: 1.00 Following a small line break in the drywell HPCI/RCIC. auto initiat A HPCI steam leak in the HPCI_ steam tunnel results in an isolation of
_
HPC MSIVs are closed and Feed Pumps are NOT available. -Plant conditions are: 1 Reactor Water Level 90' inches, steady Drywell' Pressure 15 psig RCIC Flow 500 gpm RCIC Steam Tunnel 195*F, steady Reactor Bldg Vent Rad Hi Alarm cleared The 30 minute RCIC steam tunnel leak detection isolation timer is running. What required operator action could maintain RCIC availability? Restart Reactor Building Ventilation per SEP-0 Restart Reactor Building Ventilation per OP-3 Turn' power off to steam leak detection NUMAC modules, Install a circuit alteration to defeat steam leak detectio i l
I
i
)
!
!
PAGE 42 i
,
>>
-. _ _ . . . _ . _ _ . _ . _ . _ . - . _ . . _ . -_m._ _ . - _ _ _ _ - . - .- ._ .. _, _m _ - - _ _ _ .
'
J
,- :
,
i
- . "NRC 97-1 SRO, Rev 0" EXAMINATION **
i QUESTION 79 POINT VALUE: 1.00 !
!
Following core damage, an unisolable steam leak in the Turbine Building i requires declaration of a General Emergency due to loss of three out of ;
three fission product barrier , ,
The crew is executing EOP-04-RRCP, Radiation Release Control Procedur Field surveys and Off-Site dose projections -(PEP-03.4.7) are.being performe :
!
When, per EOP-04-RRCP, is Emergency Depressurization of the Reactor [
required to be initiated? +
i a. Immediately since a General Emergency has been declare ,
b. The Noble Gas release rate reaches 1200% of the Tech Spec limi [
'
c. The measured dose rate at the site boundary is reported at' 110 mrem / hou .!
)
d. Dose projections estimate a,, Off-Site dose of 4.9 rem thyroid I (CDE). I
!
i
!
!
!
I
,
I i
i t
&
P l
l l
i PAGE 43
)
_ -
-, . _ _
. . _ , . . _ . . _ - - _ _ . - _ . . _ . . - _ . _ _ _ _ _ . . _ _ _ _ _ . . _ . . . , _ . _ _- . . _ . . . . _ . . . . . _ . _ _ . . . _ _ _ _ . . . - _
, l t ,
i l
<
J
I I
f .
l l ** . "NRC 97-1 SRO, Rev 0" EXAMINATION ** i l l l l t
l l
l QUESTION 80 POINT VALUE: 1.00 l
- A chemistry sample has been directed due to steadily rising SJAE rad
' '
monitor reading The following sequence of events occur at the times noted:
0900 chemistry reports coolant activity of 4.3 pei/ml based on
- . sample l 0905- the SS enters a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to Hot Shutdown LCO based on
!
coolant activity 0910 the SS declares an Unusual Event based on abnormal core conditions
- 0915 A plant shutdown is started to comply with Technical l Specifications The NRC must be notified of events in progress no later than
l l ! l' l l ~ 1015 :
'
i
I
!.
'
l
!
!
l
.
l l
[
d W
PAGE 44 l
- . , .
._ _ _ -.__ ..,.._ _ . _ _ _ _ . . _ . _ _. _ - _ _ _ _ _ _ . . . _ _ _ _ _ _ _ _ _ . _ _ _ . -
,
!
l l
!
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
QUESTION 81 POINT VALUE: 1.00 A General Emergency has been declared on July 4th due to an unisolable RCIC steam line break with indications of fuel failur The following '
weather data is availabl :
l Temperature 97'F ;
i- Upper wind speed 6.7 mph
! Lower wind speed 5.8. mph
! Upper wind direction 236.8*
,
Lower. wind direction 237.3*
l Stability class D
\ .
Maximum projected off-site dose per PEP-03.4.7 is 45 mrem TEDE and
'
maximum off-site field survey readings are 28 mrem / hour. The release is expected to drop rapidly due to the emergency depressurization of
~
l the Reactor in progres ~
What Protective Action Recommendation (PAR) should be made to Off-Site Agencies? Evacuate zones A, B, C, G, H, K and shelter zones D, E, Evacuate zones A, B, C and shelter zones D, E, F, G, H, K.
- Shelter all zones due to the release being below EPA Protective j l Action Guideline '
L j Shelter all zones since an evacuation is expected to take over l eight hours to complet l t
i
!
l I
!
!
l
.
1'
l l PAGE 45 l
,
!
~ . -. -
,
. ._ - ._ . ... . . _ _ . . _ . _ _ . _ . - _ _ _ _ _ . _ _ _ . . .
____.,
i .
i
- "NRC 97-1 SRO,_Rev 0" EXAMINATION ** I QUESTION 82- POINT VALUE: 1.00 A Unit Two (2) Reactor startup is in progress per GP-0 Heatup and pressurization of the Reactor is being performe The operating CRD Pump trip Attempts to restart CRD per OP-08 and AOP-02.0 are unsuccessfu AOP-02.0 requires the operator to insert a manual Reactor scram only if Reactor pressure.is below: psig psig psig psig QUESTION 83 POINT VALUE: 1.00 )
)
Unit One (1) is operating at 100% power when Recirculation Pump 1B !
trips, resulting in the following conditions:
'
Total Core Flow (P603) 39 Mlbm/ Hour l Total Core Flow (U1CPWTCF) . 35 Mlbm/ Hour ,
Indicated Core Plate DP 4.7 psid l APRMs 68% ;
LPRM Upscale /Downscale alarms None What region of the Thermal Power Limitations Map is the plant operating ;
in, and what operator action is required to be taken? ; Region B, raise total core flo Region B, insert control rods per ENP-2 Region A, immediately insert a manual scra % Buffer, increase monitoring of nuclear instrumentatio i PAGE 46 i
i (
- _ _ _. . . _._._____..~.__.__._._-._...___.-_.....___.m.m i l
,
!
- - "NRC 97-1 SRO, Rev 0" EXAMINATION **
-
i QUESTION 84 POINT VALUE: 1.00 I
,
'
Unit Two-(2) was' operating at power when a~ trip and lockout of BOP bus i 2B required the operator to insert a manual Reactor scra Shortly _
L following the scram,.the following indications are noted:
! Recirc pump A #1 seal pressure 1000.psig l Recirc pump A~#2 seal _ pressure ~ 1000 psig l_ Recirc pump B #1 seal pressure 100 psig Recirc pump B #2 seal pressure
'
50 psig Drywell pressure 1.4 psig, risin Average drywell temp 140'F, rising Average primary containment temp 126*F, rising
- The operator is required to enter:
f AOP-14.0 and isolate Recirc pump AOP-14.0 and isolate Recirc pump B.
,
- EOP-02-PCCP and isolate Recirc pump EOP-02-PCCP and isolate Recirc pump QUESTION 85 POINT VALUE
- 1.00
- A situation arises requiring immediate evacuation of the control room prior to completion of any immediate actions per AOP-32 0.. RPS is aligned:
i L '
RPS Lus A Powered from RPS MG Set A RPS Bus B Powered from RPS MG Set B l
l If the RPS EPA breakers are opened in the exact sequence specified by
'
AOP-32.0, opening which EPA breaker will result in a reactor scram? EPA Breaker 1.
,
l b EPA Breaker EPA Breaker 3.
!
, EPA Breaker i
!
l PAGE 47 t
s
, ,-, . , , . . . , . _ . . . , . ~ . _ _ - __ _ . _ ~ . _m
. .- , . _ . . . - _ . _ . . . . _ . . . - _ . _ - - _ . . _ _ - . . _ . . > _ . _ - . _ _ _ . _ _ _
- ' "NRC 97-1 SRO, Rev 0" EXAMINATION **
QUESTION 86 POINT VALUE: 1,00 FollowingLa-loss of feedwater on Unit Two (2) , HPCI and RCIC are being s used to restore Reactor water level to the normal band. The operator notes the following alarms and indications:
250 Batt B Under Voltage Alarm sealed in Battery Bus 2B-1 Voltage 0 volts (XU-2)
Battery Bus 2B-2 Voltage 0 volts (XU-2)
Battery' Bus 2B-1 Voltage 0 volts (ERFIS)
Battery Bus 2B-2 Voltage 0 volts (ERFIS)
.How is the operation of HPCI and RCIC affected by the power loss? HPCI continues to inject to the Reactor, RCIC isolates due to loss of isolation logic powe RCIC continues to inject to the Reactor, HPCI isolates due to loss of isolation logic powe HPCI continues to inject to the Reactor,'RCIC coasts down due to loss of flow controller powe RCIC continues to inject to the Reactor, HPCI coasts down due to loss of flow controller powe PAGE 48 y
. _ _ - , .. -
._ _. . _ - . _ . - _ . . - . ._ - . . . _ _ . _ . _ . _ _ - . . - . . _ . _ . _ . _ - _ . . _ _
E l
- "NRC 97-1_SRO, Rev 0" EXAMINATION **
!
i.
! QUESTION 87 POINT VALUE 1.00
- Foll'owing a Loss of Off-Site Power to Unit One -( 1 ) , the operator is [
l performing AOP-36.1. Plant conditions are:
i Diesel Generator 1 Running at 3575 KW load l Diesel Generator 2 Running at 3680 KW load l Reactor Building HVAC Isolated l The operator is directed to restart Reactor Building HVAC using three (3) supply! fans (75 KW each) and three (3) exhaust fans (45 KW each) .
! How will starting two supply and exhaust fans from MCC 1XG and oD2
,
supply and exhaust fan from MCC 1XH affect Diesel Generator (DG)
- maximum loading? i i
l DG1 only maximum load will be exceede DG2 only maximum load will be exceeded, DG1 and DG2 maximum load will be exceeded, i
i DG1 and DG2 will remain within maximum load limits.
t QUESTION 88 POINT VALUE: 1.00 l While operating at rated power, the following indications are noted:
Gen Bus Under Freq Relay Alarm sealed in Generator frequency 59.2 Hz If frequency remains at this value for five minutes, the turbine must'
be tripped to prevent damage due to excessive: I volts / hertz in the main generator winding volts / hertz in the main transformer winding )
'
i
, resonance vibrations in low pressure turbine bladin j resonance vibrations in high pressure turbine bladin i l I i
.! !
i
i
'
PAGE 49
.
-
_ _ _ _ _ . - - . . _ _ _ - . . - - _ . - - _ _ . _ _ . _ _ . . . _ _ . _ _ . _ . . . . . . . .
u
!
1
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
L l-QUESTION 89 POINT VALUE: 1.00 ;
Following a total loss of RBCCW, several control rods fail to fully l insert on a manual reactor scra How long can the CRD Pumps be operated without Cooling Water per the Abnormal Operating Procedure?
I l minute l
.
l minute i minutes.
l minute !
QUESTION 90 POINT VALUE: 1.00 I i
Unit Two (2) was at 20% power during plant startup when a sudden rise in off gas flow is accompanied by a lowering condenser vacuum. The reactor was manually scramme Plant conditions:
i Condenser vacuum 15" Hg, slowly lowering j Reactor pressure 921 psig, steady Bypass valves one partially open
.What is the minimum additional reduction in condenser vacuum that would result in the loss of automatic Reactor' pressure control?
- " H ;
b, 6" H " H " H l l.
{'
.
f i.
4 h
$
i i
'
PAGE 50
.
...y,. ,y . _ . ._. _ . , ,
., ._ . _ _ . _ _ , _ _ _ . _ . . _ _ _ _ _ _ . _ = _ _ = . _ _ _ . _ .. _ . . . . _ __ _
l r ** "NRC 97-1 SRO, Rev 0" EXAMINATION ** l l
.
'
l ;
QUESTION 91 POINT VALUE: 1.00 Unit One' (1) is performing. refueling operation A fuel handling I accident results the the following radiation alarms l
Area Rad Refuel Floor Hi (white alarm) i Area Rad Rx Bldg Hi (red alarm, blue bar) l I
Area Rad Control Room Hi (red alarm, red bar) I ( Rx Bldg Vent Rad Hi (red alarm, blue bar) !
l Rx Bldg Vent Rad Hi Hi (red alarm, blue bar)
i
!
'
How will the Reactor Building HVAC and Control Building Emergency Air i Filtration (CBEAF) systems respond to the above radiation alarms? l l Reactor Building HVAC isolates, CBEAF remains in standb J
,
l Reactor Building HVAC remains in operation, CBEAF initiate j
'
! Reactor Building HVAC remains in operation, CBEAF remains in standb Reactor Building HVAC isolates and CBEAF initiates, i l
!
l QUESTION 92 POINT VALUE: 1.00 j l During normal full power operation of Unit Two (2), the following j alarms and indications are noted:
'
Air Compressor D Trip Alarm sealed in Air Compressors A/B/C Running l Instrument Air Pressure low Alarm sealed in Instrument Air header pressure 100 psig The operator should verify that air compressors A, B and c'are loaded and that: Service air isolation valves, PV-706-1 and PV-706-2, have automatically closed.
, Interruptible air isolation valves, PV-722-1 and PV-722-2, have i automatically closed.
' Standby reactor building air compressors have automatically
,
started and loaded.
d l Backup nitrogen rack isolation valves, RNA-SV-5482.and SV-5481, l have automatically opene PAGE 51 l
!
. --. . _ . . _ -
_
-- -.._. . .
. - . - . - ~ . . . - . . - . . . - - - - . . --. . . . - .
i
, ** "NRC 97-1 SRO, Rev 0" EKAMINATION **
l- )
i QUESTION 93 POINT VALUE: 1.00
!
Following a loss'of-shutdown cooling, Alternate. Shutdown Cooling has
'
been established per AOP-15.0. -Plant conditions are:
,
RHR Loop A in suppression pool cooling i RHR Loop B injecting to the Reactor Vessel I
! Reactor pressure is 115 psig SRV G is open
! It becomes desired to make a slight adjustment to raise the cooldown
-
l l rat This may be accomplished by closing SRV G and opening: j i '
j SRV SRV '
l j SRV .!
l SRV I a- j
! !
QUESTION-94 POINT VALUE: 1.00 l I
'i
. Unit Two (2) is operating with the following; plant conditions: i i
! Reactor power 85% .
Core flow 51 Mlbm/hr !
Rod line 110%'
Recirc MG sets . Scoop tubes locked I
.
Which of the following conditions authorizes the operator to manually
' initiate Select' Rod Insert (SRI)?
l L
' Condenser vacuum lowers and approaches the turbine trip !
setpoint.
l Feedwater heating is partially lost and APRMs approach the
! scram setpoin ;
,
L A reactor feed pump trips and reactor level approaches the
L scram setpoin A recirculation pump trips placing the plant in Region A of the Thermal Power Limitations Ma ,
,
PAGE 52 I
i
.
.
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
QUESTION 95 POINT VALUE: 1.00 A clorine release of 25 lbs C1/sec is in progres The following i meteorological data is available: !
l Air temperature 46.8*F Lower wind direction 238.7*
Lower wind velocity 7.8 mph
,
Stability class E The concentration of clorine is expected to drop below 10,000 ppm at a downwind distance of between: .25 and 0.5 mile .5 and 0.75 mile ' .75 and 1.0 mile .0 and 1.25 mile QUESTION 96 POINT VALUE: 1.00 Unit One (1) is in a Station Blackout. E Buses are being cross-tie The Reactor is being cooled down at 100*F/ Hou SRVs are being used to reduce Reactor pressure and HPCI/RCIC are being used to maintain Reactor water leve The following Drywell Temperature readings are reported from the Remote Shutdown Panel:
CAC-TR-778, Point 1 314 F CAC-TR-778, Point 3 300 F CAC-TR-778, Point 4 297*F What action is required? Immediately open seven ADS valves, Raise the cooldown rate to >100*F/ hou Align LPCI for injection, then open seven ADS valves, Align fire water for injection, then open seven ADS valve PAGE 53
. -___--. ... _ __ - . _ _ . _ _ _ _ . . __ ~ _. . . _.. . _ _ . _ . . _ .
!
!
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
!
'
l
'
QUESTION.97 POINT VALUE: 1.00
!
During Station Blackout conditions, outside air temperature drops below 32* HPCI is being used to maintain Reactor Water Level above TAF.
i AOP-36.2 directs HPCI suction valve breakers to be turned OF This action will maintain HPCI suction from the: Suppression Pool to prevent exceeding minimum reactor vessel feedwater nozzle temperature requirements.
l l Suppression Pooi to prevent loss of HPCI due to excessive
! cooling of the pump and turbine lube oil, CST to prevent inadequate flow causing loss of CST suction capability due to suction line freezin CST to prevent a false low CST level suction transfer signal due to the level switches freezing.
.
i QUESTION 98 POINT VALUE: 1.00 The Control Room has been evacuated due to a fire in the Control Buildin Shutdown from outside the Control Room is in progress per l
ASSD Procedure Reactor cooldown is in progress in preparation to place shutdown cooling in service. The following Reactor pressures are recorded at j the indicated times: '
0000 750 psig I 0015 640 psig
, 0030 390 psig I 0045 340 psig 0100 310 psig
!
The Reactor Vessel cooldown rate is: less than 100*F/ Hour, maintain present cooldown rate.
l l greater than 100*F/ Hour, maintain present cooldown rate.
l less than 100*F/ Hour, increase cooldown rate to greater than 100*F/ Hour.
, greater than 100*F/ Hour, decrease cooldown rate to less than l 100'F/ Hour.
! PAGE 54
. _
-. .
T I
- "NRC 97-1 SRO, Rev 0" EXAMINATION **
- QUESTION 99 POINT VALUE: 1.00 i
Core defueling is in progres All control rods are fully inserted into the reactor core._ A fuel assembly has just been placed in the !
- fuel pool _and unlatched. 'The main hoist has been raised to a safe l elevation to pass through the cattle chute (HOI normal-up) with the bridge still over the fuel pool locatio The next step requires that a fuel assembly be removed from the reactor i core and placed in the fuel poo When will the ROD BLOCK INTERLOCK #1 light on the Interlock Status !
Display Panel first light as the next step is performed?
] As the bridge is moved near the reactor core (LS1 is actuated).
l When the bridge is over the reactor core (LS1 is actuated) ;
and the main hoist is lowered into the reactor vesse J When the fuel assembly is latched, with both grapple hooks close ' When the fuel assembly is being raised and the main hoist loaded signal is actuated.
l QUESTION 100 POINT VALUE: 1.00 .
Core Spray Pump 2A 4160 volt breaker is racked in per'OP-50, with 125V -
DC available at the switchgear. A LOCA results in a condition requiring the. auto start of the-pump.
Refer to LL-09113 sheet 15, Core Spray Pump 2A Control Wiring Diagra The pump breaker is closed by energizing the: X coil, when relay K12A is de-energized and relay K15A is
! energize Y coil, when relay K12A is de-energized and relay K15A is energized.
l
, X coil, when relay K12A is energized and relay K15A is
[ de-energize d.- Y coil, when relay K12A is energized and relay K15A is de-energized, f
PAGE 55 i
i i i
!
i
, 1 I
i i
i i ** "NRC 97-1 SRO, Rev 0" EXAMINATION **
i I
l
.
!
1 l
i
.
l l
l
\
>
l l l
I l
,
I
- END OF "NRC 97-1 SRO, Rev 0" EXAMINATION **
PAGE 56 OF 56
. , . . - . . . ... , . - _ n . - ,-. .
. . . . . .~ . . . . - . . .. .. .
1 j
,
NAME: ' MW& C DATE: / / SCORE:
CRADED BYr ALTERNATE GRADE:R (if required)
EXAM: NRC 97-1 SRO. Rev 0 CLASS: Et,C 96-1 COURSE CODE: ROA029 l
. ABCk 2 AhCD 4 AB@D 6 ABC@
l J A@CD 2 ABCh 4 ABC@ 6 hBCD A-B@D 2 ABC@ 4 ABC@ 6 A@CD
- ABC@ 2 AB@D 44. @BCD 6 ABC@
i AB@D 2 hBCD 4 ABC@ 6 ABC@ AB@D 2 ABhD 4 AhCD 6 A@CD ABC@ 2 A@CD 4 AB@D 6 hBCD @BCD 2 ABC@ 4 A'BC@ 6 AB@D AB@D 29. @BCD 4 AhCD 6 AhCD 1 AB@D 3 A@CD 5 ABCh 7 ABhD l 1 ABC@ 31. @B'CD 51. @BCD 7 A@ C D'
1 @BCD 3 ABC@ 5 A'B@D 7 AB@D 1 ABC@ 33. @BCD 5 ABC@ 7 A@CD 1 A@CD 3 ABC@ 5 A.BC@ 7 ABC@
15. @BCD 3 A@CD 5 A@CD 7 A@CD 1 A@CD 36. @BCD 56. @BCD 76. hBCD 1 A@CD 3 @BCD 5 @BCD 7 ABhD l 1 @BCD 3 AB@D 5 ABhD 78. @BCD 1 A@CD- 3 A@CD 5 AB@D 7 ABC@ <
2 AB@D 4 ABhD 6 ABC@ 8 ABhD
.
. . . - - . . . . . - . . . . . - . . . , . . - . - . . . . . , . + . _ . . ~ - _ . - - . . - . . ,- .
A I
i l
l
NAMES
'
NWCf - EV DATE:- / / SCORE,
!' /-
' GRADED SY . ALTERNATE GRADER: (if required)
L EXAM: NRC 97-1 SRO. Rev 0 CLASS: MLC 96-1 COURSE CODES ROA02B l l
,
j
, 81. @BCD
- . 8 ABC ;
.
8 A@CD j i' 8 A@CD !
i J
85. - ABCh
8 ABhD !
.
I 8 ABCh l
[
8 ABhD l .8 'AB@D , . @BCD
-
9 ABCh l
$-
.
92. @BCD i f
.
9 ABCh
j
,
.9 AhCD .
{ 9 hBCD
!
'
9 hBCD' ;
9 'A-BCh 9 ABhD 99. @BCD
. 10 ABhD l,
'!
!
l
. - . . _ _ , . - - - ..
.g i
k
(.
s
>
!
>J t t
d I
i
,
I I
% .]
- . . . -. _-, - . . - .
I i
4 ES-401 Site-Specific Written Examination Form ES-401-7 Cover Sheet
_
i
! U.S. Nuclear Regulatory Commission ,
'
Site-Specific ;
Written Examination
'
Applicant Information Name: Region: II Date: 04/25/97 Facility / Unit: Brunswick /1 & 2 ,
,
License Level: Reactor Operator Reactor Type: GE BWR-4 Start Time: Finish Time:
-
Instructions j Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. The passing grade requires a final grade of at least 80.00 percen Examination papers will be collected four hours after the examination start Applicant Certification All work done on this examination is my own. I have neither given nor received ai Applicant's Signature Results Examination Value Points Applicant's Score Points Applicant's Grade Percent
)
- v WRITTEN EXAMINATION GUIDELINES 'After you complete the examination, sign the statement on the cover sheet indicating that the work is your own and you have not received or. given assistance in completing the examinatio . To pass the examination, you must achieve a grade of 80.00 percent or greater.-
"-
Every question is worth one poin . For an initial examination, the time limit for completing the examination is four hour , You may bring pens and calculators into the examination room. Use only black ink to ensure legible copie . Print your name in the blank provided on the examination cover sheet and the answer sheet. You may be asked to provide the examiner with some form of positive identificatio . Mark your answers on the answer sheet provided and do not leave any question blank. Use only the paper provided and do not write on the back side of the page If you decide to change your original answer, draw a single line through the error,
. enter the desired answer, and initial the chang . If the intent of a question is unclear, ask questions of the NRC examiner or the designated facility instructor onl .~ Restroom trips are permitted, but only one applicant at a time will be allowed to leave. Avoid all contact with anyone outside the examination room to eliminate even the appearance or possibility of cheatin . When you complete the examination, assemble a package including the examination questions, examination aids, answer sheets, and scrap paper and give it to the NRC
'
examiner or proctor. Remember to sign the statement on the examination cover sheet indicating that the work is your own and that you have neither given nor received assistance in completing the examination. The scrap paper will be disposed of immediately after the examinatio ,
1 you have turned in your examination, leave the examination area as defined by t.~ proctor or NRC examiner. If you are found in this area while the examination is still in progress, your license may be denied or revoke . Do you have any questions?
<
'l
. . _ _ _ . __. _ _ . . . . . .
. .
_ . _ - _ . _ _ - _ _
.,
t t
f f
s
!
- "NRC 97-1 RO, Rev 0" EXAMINATION ** !
!
.!
I QUESTION 1 POINT VALUE: 1.00 [
t Unit One (1) is operating with the following plant conditions:
Core Thermal Power 2558 MWth l Reactor, pressure 1030 psig !
Core Flow' 77 Mlbm/hr i i
l-The SAFETY LIMIT for THERMAL POWFR is the MINIMUM CRITICAL POWER'
-
n RATIO (MCPR) shall not be less thans .07 l
g .08 j t .09 i
!
- .10 :
l m
QUESTION 2 POINT VALUE: 1.00
!
A Unit Two (2) startup is in. progress per GP-0 The Reactor is !
critical with Reactor power at the point of adding hea l
.
!
Coolant temperature is being raised to saturation conditions, with l Reactor steam dome pressure at 0 psig. RWCU is in servic i t
2PT-01.7, Heatup/Cooldown Monitoring is being performed. Which of the j following conditions would result in UNSATISFACTORY Acceptance !
Criteria of the PT? !
i G31-TI-R607, Pt 5 indicates 169*F '
C12-TR-R018, Ch 151 indicates 164*F L Reactor water level is 200" , G31-TI-R607, Pt 5 indicates 187*F ;
,
C12-TR-R018, Ch 151 indicates 183*F l i
Reactor water level is 210" l
- G31-TI-R607, Pt 5 indicates 174*F !
l C12-TR-R018, Ch 151 indicates 170*F l
,
Reactor water level is 200" ; G31-TI-R607, Pt 5 indicates 198"F C12-TR-R018, Ch 151 indicates 195*F l Reactor water level is 210"
,
PAGE 1 L
'
!
_ . _ . . 4
_ _ _ , _ . . - _. _ _ .- __ _ - . . _ . _ - _ ,_ . . . . - . m. - .
'
l t
!
l
- "NRC 97-1 RO, Rev 0" EXAMINATION **
l l
QUESTION 3 POINT VALUE: 1.00 Unit one (1) is operating with the following conditions:
Reactor Power 55%
Reactor Feed Pump 1A operating Reactor Feed Pump 1B idling Recirculation Pump Speeds 58%
Reactor Feed Pump (RFP) 1A trips and Reactor Level Hi/Lo alarm I Reactor Level drops to the scram setpoint and. continues to lower to I
+110 inches before the operator brings the idl!ng RFP on line (30
- seconds after the trip of RFP 1A) to restore Reactor Leve What should be the present status of the Recirculation Pumps?
.
! Running at 58% speed.
i
.
i Running on limiter #2.
l Running on limiter # Tripped on ATWS ARI/RPT.
l QUESTION 4 POINT VALUE: 1.00 l
Unit Two (2) power ascension is in progress per GP-0 The operator is performing OP-02, Section 5.3, Speed / Power Increase Using the Recirculation Pump A(B) Speed Control (an Information Use procedure) .
! While raising the Recirculation Pumps' speed, the operator is cautioned to maintain Recirculation Pumps A and B within:
l % when below 58 M1bm/hr core flow, and 10% when =58 M1bm/hr core flo ,
l % when below 58 M1bm/hr core flow, and 5% when =58 M1bm/hr l core flow.
t % when below 58 M1bm/hr core flow, and 20% when =58 Mlbm/hr core flow.
i
- % when below 58 Mlbm/hr core flow, and 10% when =58 Mlbm/hr
) core flow.
i I
j
!
PAGE 2 j i
, , , - , - -
- .
.- ....-. ..- - . .
i I
L
'
l l
- "NRC 97-1 RO, Rev 0" EXAMINATION **
l
. QUESTION 5 POINT VALUE: 1.00
-
.
.
!
Unit Two (2) has inserted a manual Reactor scram due to lowering
.
L l condenser vacuu Control rods failed to insert on'the scra Plant i
! conditions: 1 i
,_ Reactor power 31% n l Steam flow 3.2 Mlbm/hr j
}. Reactor pressure 960 psig, controlled by EHC Drywell pressure 0.6 psig
'
Mode Switch RUN
"
i Main Turbine Tripped on low vacuum l The operator is performing LEP-02, Section 3 to reset and scram the Reacto Jumpers to bypass RPS trip signals have been requested but HQI yet' installed. Which of the following would prevent the ,
operator from resetting RPS prior to jumper installation?
, Scram discharge volume Hi Hi level RPS trip sealed i :
' Turbine stop valves closed with reactor power above 30%. l
! Reactor water level is controlling at the.setdown setpoin l l IRMs upscale Hi Hi due to being inserted but not ranged u j l l
.
l i
.+
l l
!
'
l i
!. PAGE 3 l
..'
! -.
,
i
!
- "NRC 97-1-RO, Rev 0" EXAMINATION **
QUESTION 6 POINT VALUE: 1.00 Unit One (1) is operating at 27% power during Unit startup. The Turbine Generator has been synchronized to the grid. A total loss of L Division II DC Switchboard 1B results in a reactor scram. During this
! transient:
57 control rods fail to fully insert l Reactor pressure peaks at 1132 psig
'
Reactor water level lowers to +107 inches BOP Buses fail to transfer to the-SAT
- Diesel Generator 2 fails to start l What is the expected status of the Alternate Rod Injection (ARI)
system? ARI has:
i auto initiated on high reactor pressure.
l auto initiated'on low reactor water level, l
- not auto initiated but can be manually initiate I not auto initiated and cannot be manually initiate I
! l QUESTION FOINT VALUE: 1.00 Following a line break on Unit Two (2) plant conditions are: l i
Reactor pressure 450 psig )
l Reactor water' level +70 inches Drywell pressure 4.8 psig Drywell temp (average) 165 deg F !
All Drywell Cooler Fans will: l a, trip but can be restarted per SEP-1 j i trip and cannot be restarted per SEP-1 l i auto start but can be tripped at the RTGB.
I auto start and cannot be tripped at the RTGB.
t i
l r
PAG .
,-
- ._ , . , .
- ... - .. = . < ,
l l
I I
!
,
- "NRC 97-1 RO, Rev 0" EXAMINATION ** l
,
i QUESTION 8 POINT VALUE: 1.00, I i
Unit Two (2) is in OPERATIONAL CONDITION 5, with CORE ALTERATIONS in o progress and. SECONDARY CONTAINMENT INTEGRITY establishe l
!
I The Interruptible Instrument Air Header ruptures and the Interruptible Air Header isolation valves are closed. Non-Interruptible Air Header
'
pressure is norma How is SECONDARY CONTAINMENT INTEGRITY affected? Reactor Building Supply and Exhaust Fans remain in service, SBGT auto start ! Reactor Building Supply and Exhaust Fans remain in service, i SBGT' remains in standb Reactor Building Supply and Exhaust Fans trip and SBGT auto starts to maintain negative pressur Reactor Building Supply and Exhaust Fans trip but SBGT must be manually started to maintain negative pressur QUESTIO POINT VALUE: 1.00 Unit Two - (2) is operating at power with Diesel Geneator 3 (DG3) under clearance. A lockout of a BOP Bus initiates a transient resulting 3h a l
'
reactor scram signal and an ATWS. Plant conditions:
Reactor power 10%
Bus E3- De-energized SLC Switch PUMP B RUN position Bus E7-E8 Cross-tie Breakers racked in ;
How will the SLC~ system respond when the Bus E7-E8 Cross-tie Breakers are closed? SLC squib valve: A fires, no SLC pump start B fires, no SLC pump start A fires, one SLC pump start B fires, one SLC pump starts.
[
t PAGE 5 6-
_,
.__ ____. _ _ _ _ . . __ . _ _ _ _ . _ . _ _ _ . - _ _ _ . . ~ . _ .
,
. ;
- ** "NRC 97-1 RO, Rev 0" EXAMINATION **
s I QUESTION 10- POINT VALUE: 1.00
-
- The unit is at 10% power during reactor startup. The operator withdraws rod 22-19 to position 48. The following indications are ,
noted- 1 l
Rod Drift alarm seals in
-
-Rod Overtravel alarm seals in
<- Rod 22-19 four rod display is blank
What operator action is required? Enter substitute rod position data into the RWM.
< Insert rod 22-19 to position 46 to attempt recoupling.
4 Insert 22-19 to a position with an operable reed switc Fully insert rod 22-19 and disarm the HCU electrically or
hydraulicall !
QUESTION 11 POINT VALUE: 1.00 A Unit One (1) reactor scram occurs from 100% powe The operator completes the immediate scram actions and notes the following Rod Worth
! Minimizer (:RWM) display:
!. ALL RODS IN: NO SHUTDOWN: YES
[ RODS NOT FULL IN: 001 When the operator depresses the List Rods RWM key, rod 02-51 is
-
displayed. What can be the furthest withdrawn position of rod 02-51 !
to cause the above display, and the basis for that rod position? I , shutdown margin calculation , maximum suberitical bank withdrawal positio , maximum suberitical bank withdrawal positio , maximum subcritical bank withdrawal positio i i
l PAGE 6
_
.- . . - - _ - . _ _ . - _ . - _ . - - - - - . - . ~ . . _ _ . - . . . , _ _ - _~ - ~-
i
!
l l
l
- "NRC 97-1 RO, Rev 0" EXAMINATION **
l
)
QUESTION 12 POINT VALUE: 1.00 Which of the following valves must be cleaed during performance of PT-14.2.1, Rod Scram Time Testing, to avoid erroneous scram times? V105,'exhast header isolation valv V113, charging header isolation valv V103, drive water header isolation valv V104, cooling water header isolation valv QUESTION 13 POINT VALUE: 1.00 During a low water level condition, CRD Flow maximization is being implemented per SEP-09 with the Reactor Building accessibl The operator is directed to maintain Charging Water Header pressure ,
a950 psig while opening the Flow Control and Pressure Control valve i
This limitation will prevent pump: trip on overcurrent protectio ! trip due to low suction pressur flowrate in excess of runout capacit discharge pressure dropping below reactor pressur l l
!
l i
PAGE 7 l
._ . . _ . - . . . _ . . . .__ . ~ . . . - - _ . . _ . _. _ _ _ _._ . -.m .. . . . . . .m__. _ . _ __._ _ _ _ . _
'
t
,
i
- . >
,
'
t
,
'
- "NRC 97-1 RO,-Rev 0" EXAMINATION **'
i
)
!,
QUESTION 14 POINT VALUE .1.00 i The initial reactor startup is in progress following refueling per ;
i GP-0 Initial criticality has just been achieved and PT-14.3.1, ;
- .Insequence Critical Shutdown Margin Calculation, is being performed
~
.
- .
i
,
The operator ~ notes the following SRM readings:
-
'
I SRM. Channel A 8.0 E4- !
SRM Channel B 7.0 E4 '
'
SRM Channel D 5.6 E5 i
- - )
All IRMs are on Range What automatic protective functions (if any). i
[
should have occurred?
. ' Alarm onl L i Rod block onl i
s 1 Rod block and-1/2 scra *
t
!. Rod block and full scram.
i I
I PAGE 8 l
l
_. .. . - - -. . . .. .
-
.
. -
'
!
i
- "NRC 97-1 RO, Rev.0" EXAMINATION **
l t
'
QUESTION 15 POINT VALUE: 1.00 [
!
During a Reactor startup following refueling, the operator is ]
l performing PT-50.2, IRM Range 6/7 Overlap Determination. The following '
data is recorded:
l
Ranae 6 Rance 7
.
- IRM C 45 5 l IRM D 43 4 IRM E 37 3 IRM F 52 6 IRM G 44 4 IRM H 50 6 The Level 2 acceptance criteria of PT-50.2 is:
l satisfactory for all Division I and II IRM unsatisfactory for at least one Division I IRM onl unsatisfactory for at least one Divivision II IRM onl unsatisfactory for at least one Division I and one Division II IR l QUESTION 16 POINT VALUE: 1.00 APRM Channel C has all associated LPRM inputs Operable, with the LPRM function switches in Operate. During performance of a MST, the LPR function switches are placed to and left in Bypass, one at a tim What is the MINIMUM number of LPRM function switches that should have been placed in Bypasa when a Neutron Monitoring trip signal (1/2 scram) i
'
is received? l ! l l l
PAGE 9
- ., _ __ _. _ _. _ _ _ _ _ _ _ __. _- _ _ . . . .~._. _.._._ _ _ _ _ . _ . _ _ _ _ . _
s
).
a
- "NRC 97-1 RO, Rev 0" EXAMINATION **
!
'
l QUESTION 17 POINT VALUE: 1.00 Unit Two - (2) is operating at 55% powe Control Rod withdrawal.is in progress for plant startu The following indications and alarms occur:
Rod 46-11 is selected Rod Out Block alarm sealed in !
APRM Downscale/Inop alarm sealed in APRM Channel A indicates 0% power Rod Block Monitor (RBM) Channel A will:
1 be bypassed automaticall '
' remain in normal operation enforcing rod block i fail downscale resulting in a rod bloc ) transfer to APRM Channel E for reference powe !
!
QUESTION 18 POINT VALUE: 1.00 j
. Unit Two (2) is operating at 100% powe The Standby Gas Treatment i System (SBGT) is in the standb5- alignmen A Trip of both Reactor Feedwater Pumps results in the following plant conditions:
Reactor water level +80 inches ;
Reactor pressure 945 psig l Drywell pressure +0.3 psig l Rx Bldg pressure -0.4" WC HPCI and RCIC have initiated to restore Reactor water leve All systems respond as designed during the transien What operator action is reauired concerning the SBGT System? Secure'one SBGT fan per OP-1 Open Post LOCA Vent valves (SGT-V8 and V9). Open Primary Containment Suction valve (VA-2F-BFV-RB). Restart Rx Bldg HVAC and secure both SBGT trains per SEP-0 PAGE 10
. . _ .
..
. ,- -_ . . - . . - - _ - _ - . . . _ _ _ - . - . - = - _ - _ . -
!
- '"NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 19 POINT VALUE: 1.00 Following a LOCA, the Containment Hydrogen / Oxygen monitors have been placed in service using CAM overrides per the Hard Car After placing the mondtors in service, the following alarms and 1 indications occu ;
1 Reactor Building Vent Rad Hi alarm sealed in j Reactor Building Vent Rad Hi Hi alarm sealed in i Reactor Building Vent Radiation recorder channel A pegged high l Reactor Building Vent Radiation recorder channel B 0.5 mr/hr l How does' this failure affect Hydrogen / Oxygen (H2/02) monitors? Both Di'rision I and II H2/02 monitors remain in servic l
! Only Division I H2/02 monitor isolates, but can be placed back in servic Only Division I H2/02 monitor isolates, and cannot be placed back ir.. servic Both Di. vision I and II H2/02 monitors isolate, but can be placed back in servic QUESTION 20 POINT VALUE: 1.00 Unit one (1) i s operating at rated power, when the loss of an electrical power distribution system results in the following Group 1 PCIS status 1:eght indications on P601:
Inboard M3IV DC solenoid Out Inboard MSIV AC solenoid Out l Outboard MSIV DC solenoid Out Outboard MSIV AC solenoid Lit l
What power dt.stribution system has been lost?
' Division I AC.
' Division I D Division II AC.
i Division II D PAGE 11 i
i t
- "NRC 97-1 RO, Rev 0" EXAMINATION **
,
l
,
QUESTION 21 POINT VALUE: 1.00 During severe accident conditions, E&RC has been directed to take i samples at the Post Accident Sample Station to determine the extent of '
core damag .
They have requested that the control room open the RHR Heat Exchanger sample valves (E11-F079/F080). The valid Group 2 isolation signal for-
'these valves is overriden by placing: CAM Div I/Div II Isol Ovrd switches on XU51 to O l Heat exchanger sample control on XU-75/79 to LOCA .CAC Div I/Div II AC Isol Ovrd switches on XU51 to O I Control Power For PASS Isol Valves switches on XU75/XU79 to O QUESTION 22 POINT VALUE: 1.00 l
Unit Two (2) is operating at power when BOP Bus 2C trips on bus lockou The associated Diesel Generator auto starts and energizes I its E Bu One minute later, a Reactor scram occur Following the l scram:
117 control rods fail to fully insert SLC system is intiated with both SLC pumps Reactor water level is lowered to +125 inches, after SLC is started The operator notes that the RWCU system Inboard Isolation Valve (F001)
and the Outboard Isolation Valve (F004) are both close What closed the valves? F001 and F004 both closed when power was los l l F001 closed on low RPV level, F004 closed when power was los F001 closed on low RPV level, F004 closed when SLC was starte l F001 closed when power was lost, F004 closed when SLC was I starte l l
l i
PAGE 12
. . _ . - _ . - _ _ . . _ _ _ . _ . _ _ _ . - _ . _ _ . _ _ . _ _ _ _ _ _ . _ _ _ .--
. -
.
L ' -
)
- "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 23 . POINT VALUE: 1.00
- Unit One (1) is performing Alternate Emergency Depressurizatio RCIC
'
' has been placed into pressure control to aid in Reactor pressure reductio Circuit alterations have been performed per EOP-01-RVCP and-EOP-01-SEP-10 for HPCI and RCI Which of the following conditions, by interlock, will remove the RCIC System from the pressure control mode of operation? Reactor pressure lowers to 50 psig, Drywell pressure rises to 2.5 psi ,
i
' Reactor water level lowers to +110 inche Suppression pool level rises to -23 inche )
i QUESTION 24 POINT VALUE: 1.00 Unit One (1) failed to scra RHR is placed in Suppression Pool cooling per the Hard Card without use of overrides. Reactor water level is then deliberately lowered to suppress power. Plant conditions are:
Reactor water level -45 inches (NO36/NO37)
Reactor pressure 800 psig Drywell pressure 0.6 psig Suppression pool temp 130*F Suppression Pool Cooling valves have close ~
Returning Suppression Pool Cooling to service requires:
( placing the Think Switch to Manual only.
l bypassing the 2/3 core height interlock onl !
l i l placing the Think Switch to Manual, then bypassing the 2/3 core l
'
height interloc l bypassing the 2/3 core height interlock, then placing the Think !'
) Switch to Manual.
.
PAGE 13 (
l
- _ . _r , . , _ , . . .- .-- y ,, e +
. . . _ . . . _ _, . _ - . _
- . ._. _ _ _ _ _ ~ - _ . _ _ . ..
,
l l
4 i
<
'
,
- "NRC 97-1 RO, Rev 0" EXAMINATION **
,
'
i
!
QUESTION 25 POINT VALUE: 1.00 !
Consider the following normal RHR suction valve interlocks for control
,
of Reactor water level:
.
<
l!
' Shutdown Cooling suction isolation valves (F008/F009) will j automatically isolate on low Reactor water leve I
< Shutdown Cooling pump suction valve (F006) cannot be opened
,
unless Torus common suction valve (F020) is closed.
j During Shutdown Cooling operation from Remote Shutdown stations per
<
'AOP-32.0, which of the above (if any) interlocks are functional?
. onl i J 1 onl , c; both 1 and a
'
$ neither 1 nor !
i 1 I
QUESTION 26 POINT VALUE: 1.60 l
'
\
!
Unit One (1) is operating at rated power when a Loss Of Off-site Power !
occurs simultaneously with a line break in the drywel Plant j conditions are: i
! i j Reactor water level -50"- (NO36/N037)
!
Reactor pressure 150 psig
- Drywell pressure 20 psig ,
Diesel Generator 2 Tripped '
i
. No E Bus cross-tie actions have been performe What is the expected status of RHR Loop 1B? Pump 1B running under dead head condition . Pump 1D running under dead head condition Pump 1B running with injection to the reacto Pump 1D running with injection to the reacto PAGE 14
. .- .
_
,
'
l
1
~l t !
i
,
l ;
L ** "NRC 97-1 RO, Rev 0" EXAMINATION ** l r
,
QUESTION 27 POINT VALUE: 1.00 l
l Following a amall line break, RHR Loop A is placed in Drywell and l Suppression Chamber Spray using required overrides per SEP-02 and l 'SEP-0 Plant-conditions are:
l
' Reactor Water Level +175 inches Reactor Pressure 900 psig Drywell Pressure 15.0 psig
.
I Reactor water level drops to -60 inche How will RHR Loop A and RHR L Service: Water (RHR SW)' Loop A respond?
!
l The RHR Loop A drywell/ suppression chamber spray valves:
l auto close, RHR SW Loop A pump (s) tri ({
l l auto close, RHR SW Loop A remains running.
I remain open, RHR SW Loop A pump (s) tri remain open, RHR SW Loop A remains running.
L
! l l QUESTION 28 POINT VALUE: 1.00 During normal power operation of Unit Two (2), an ECCS Division I Trip Cabinet Trouble alarm is received. Investigation shows that BQTH power supplies to the trip cabinet (XU- 63 ) have failed and all associated' trip unit meters indicate downscale with no trip lights li A DBA LOCA then occurs resulting in Reactor water level rapidly j
! dropping below the Top of Active Fuel and rapid Reactor ;
depressuriztion. How will Division I Low Pressure ECCS (Core Spray 2A j i and_RHR LPCI Loop 2A) respond? '
l
, Core Spray 2A initiates, LPCI 2A fails to initiate.
l l
< Core Spray 2A and LPCI 2A both fail to initiat l
[ Core Spray 2A fails to initiate, LPCI 2A initiates.
I I
j Core Spray 2A and LPCI 2A will both auto initiat :
4 !
!
I PAGE 15 I
--_ _
, _ _ ._ . . . - . . _ _ . -
. . ~ - . . - - - , - - _ _. - -. . . - - . . . . -..... .- ._-. - ..- - - _
l-
- "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 29 POINT VALUE: 1.00 Fbilowing a large line break on Unit Two (2), Core Spray Pump 2A is injecting to the reacto ~
Core Spray Pump 2B has been overriden of Plant conditions are: i I
Reactor water level 200" Reactor pressure 25 psig L Suppression pool leve inches Suppression pool temperature 210* !
Suppression chamber pressur .5 psi )
. Core Spray Pump 2A flow rate is currently 5500 gp The operator should reduce core Spray Pump 2A flow rate by throttling: Inboard Injection valve (E21-F005A) to approximately 3000 gpm.
i Inboard Injection ~ valve (E21-F005A) to approximately 4500 gp Outboard Injection valve (E21-F004A) to approximately'3000 gp Outboard Injection valve (E21-F004A) to approximately 4500 gpm.
I' QUESTION 30 POINT VALUE: 1.00 Unit Two (2) HPCI has automatically initiated on a valid initiation signa The operator observes the'following indications:
Steam Supply Pressure O psig Turbine Exhaust Pressure O psig Pump Discharge Pressure 0 psig Turbine Speed 1000 RPM, lowering HPCI TURB TRIP SOL ENERG HQT Alarming
, 1Which of the following would explain the above indications?
l
' Isolation due to ruptured exhaust diaphra j
! Overspeed trip from speed feedback signal failur Loss of oil pressure to the turbine control syste l
, 1
[ Loss.of 125 VDC input to the 24/52.5 VDC power supplie l
PAGE 16 r
'
. c, - , , ,
-. .. . . . - - _ . . - - . . , - , . . . . . - ~ . . , . - . - - . . .. . . . . . _ . . .
- "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 31 POINT VALUE: 1.00 A Reactor scram has occurred due to a Group 1 isolatio ADS /SRVs are required for Reactor pressure control.
During verification of Group Isolations, the operator inadvertantly closed the PNS supply to the.drywell. isolation valves, and failed to notice the Backup Nitrogen valves were closed.
No Group 10 isolation signal is received during the transien How !
will.the operation of ADS /SRVs be affected? Only ADS valves _can be opened manually until the accumulator supply is deplete Both ADS and SRV valves can be opened manually until the accumulator supply is deplete Only ADS valves will be supplied by Backup Nitrogen when drywell pneumatic header drops below 95 psi i Both ADS and SRV valves will be supplied by Backup Nitrogen j when drywell pneumatic header drops below 95 psi !
QUESTION 32 POINT VALUE: 1.00 Unit One (1) has experienced a high Reactor pressure transient following a Main Turbine trip. Current plant conditions:
All rods in j Reactor Pressure 950 psig controlled by EHC !
Eleven (11) SRV's green indicating lights lit Eight (8) amber memory lights illuminated Determine the extent of the pressure transien psig l 1 psig psig I psig .
PAGE 17
- -. . _ - . . . - . __ . - . _ _ . . - . - . . ~ . .- - . _ - . . . - . -. ..
- "NRC 97-1 RO, Rev 0" EXAMINATION **
-
QUESTION'33 POINT VALUE: 1.00 Unit-One (1) is operating at power with Core Spray Pump 1.B under clearance. A Loss of Off-site Power occurs to H9IH units. Diesel Generators (DGs) 2 and 4 auto start, DGs 1 and 3 trip and lockou A stuck open SRV and loss of high pressure injection causes Reactor water level to lower. All available low pressure ECCS pumps have been manually started. Plant conditions are:
Reactor water. level LPCI/ Core Spray initiation signal just received Reactor pressure 650 psig Drywell pressure 0.6 psig l ADS Inhibit Switches AUTO '
l Assuming NQ operator action, ADS will: Auto initiate in 1 minute 23 seconds.
, Auto initiate in 1 minute 45 seconds.
!
I Not auto initiate due to lack of high drywell pressure, Not auto initiate due to lack of ECCS pump permissive.
l
.i
!
li.
l l
'
,
4 PAGE 18
!
- .
- - _ . - . - -.-_____ - . - _ - . _ - _ - - - . . . . - . - . . . - .
i
- "NRC 97-1 RO, Rev 0" EXAMINATION ** !
t QUESTION'34 POINT VALUE: 1.00 Unit Two (2) has a Loss Of Off-Site Powe HPCI and RCIC both faile Reactor Water Level dropped to +30 inches before CRD reversed the level tren Current conditions are:
Reactor water level +60 inches, rising Reactor pressure 800-1000 psig using SRVs Drywell pressure 2.2 psig, rising Average.drywell temp 180*F, rising Generator primary lockout Tripped '
RBCCW Pumps are tripped, and Nuclear Service Water (NSW) cooling water valves (SW-V103/V106) are closed. What actions are required to restore
.RBCCW to control containment parameters? Reset the Primary Generator lockout, align RBCCW cooling to the conventional header.
l Reset the Primary Generator lockout, reopen NSW cooling water valves SW-V103/V106.
l Reset Core Spray initiation logic, align RBCCW cooling to the conventional heade Reset Core Spray initiation logic, reopen NSW cooling water valves SW-V103/V10 '
<
!
i l
!
,
!
!
PAGE 19
- - - .
.-. . - ... - _~ . = - - - . . ~ . - - . . . - - -
i
,
b
- ** "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 35 POINT VALUE: 1.00 During full power operation, an unisolable Instrument Air Header rupture occurs. .The Non-Interruptible Instrument Air header isolation
valves . (IAN-V50 and IAN-V51) are closed per the AOP. The operator notes the following:
BOTH Reactor Building Standby Air Compressors fail to star ~
Backup Nitogen automatically. aligns on low RNA header pressur PNS header pressure remains normal.
) Assume no Group 1 isolation signal How will the MSIVs respond to the loss of pneumatics?
I Inboard and outboard MSIVs remain ope Inboard and Outboard MSIVs drift close i l Inboard MSIVs drift closed, Outboard MSIVs remain open.
t l Inboard NSIVs remain open, Outboard MSIVs drift closed.
.
! QUESTION 36 POiY.' VALUE: 1.00 t 1 Following a Reac'or scram, the operator trips the main turbine and
<
notes the folle ine turbine valve positions:
, Turbir- stop valves are closed 3 Turf)ae control valves are closed
.
'
Ce..cined intermediate intercept and stop valves are open ,
one turbine bypass valve open 25% '
This condition could result in turbine damage caused by:
, necepeeding of the turbin ;
- excessie' axial thrust on the shaf i I
! overpressurization of the LP shell excessive moisture impingement on LP blades.
i'
!
i
)
-
i l
PAGE 20 ;
. .- ._ -_-
- . . _ _ _ _ , _ ,_ _ ._ _ _ _ . _ _ . - . _ . . . _ _ . . .. . _ . - _ . . . . _ _ _ . . _ _ . _ - . ,
i
)
!
<
- "NRC 97-1 RO, Rev 0" EXAMINATION **
i QUESTION 37 POINT VALUE: 1.00 Unit Two - (2) is operating steady state at rated power with the .
following Electro Hydraulic Control (EHC) conditions: !
Reactor pressure 1005 psig EHC Pressure Setpoint 920 psig PAM pressure 950 psig Pressure regulator A In control Pressure regulator B 5 psig biac
,
The Pressure Averaging Manifold (PAM) pressure input to Pressure i Regulator A fails low. The PAM pressure input to Pressure Regulator B '
is unaffected. How will the EHC System respond? Pressure regulator B takes control and stabilizes PAM pressure j at 945 psig, Pressure regulator B takes control and stabilizes PAM pressure !
at 955 psi J Control valves close, reactor pressure and neutron flux rise and the reactor scram Control / Bypass valves open and steam.line pressure lowers to the Group 1 isolation setpoint.
!
!
i
,
PAGE 21
!
l
,
_
. .. - - __ - -
. _ . . _ . . _ - . . - _ . _ _ . _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ . . . _ .
l
!
- "NRC 97-1 RO, Rev 0" EXAMINATION **
l -__
,
I QUESTION 38 POINT VALUE: 1.00 Unit One (1) is operating at 100% power with the following Condensate system alignment:
Condensate pumps A & B Running l Condensate pump C Standby / Auto l Condensate Booster pumps B & C Running Condensate Booster pump A Under clearance
,
I 4KV BOP Bus 1D trips and locks out due to a bus faul The transient l results in a Reactor scram. What is the availability of the Condensate l system to provide makeup to the Reactor vessel?
Condensate pump:
l A is available, no Condensate Booster pump is availabl , B is available, no Condensate Booster pump is availabl A is available and Condensate Booster pump C is availabl . B is available and Condensate Booster pump C is available.
l l
'
QUESTION 39 POINT VALUE: 1.00 l Unit Two (2)'is operating at 100% powe The Digital Feedwater Control l System (DFCS) is aligned as follows:
l Master Controller Auto, set at 187 inches l Level instrument N004A 187 inches
'
Level instrument N004B 187 inches Level instrument N004C failed downscale Mode Select switch 3 Element Level Select switch Level A Level instrument N004A fails downscal Assuming no operator action, Reactor water level will: rise and flood the main steam lines.
! drop to the low level scram setpoin rise resulting in a main turbine and feed pump trip.
i
' remain at 187" with level instrument N004B in control.
!
r PAGE 22 _ _ . .
- . _~ ~ _ . . . - - - . - . - -_-.- -.- - . = - ._. - . - .
<.
i
'/
,
- - "NRC 97-1 RO, Rev 0" EXAMINATION **
'
F
$
QUESTION 40 POINT VALUE: 1.00 Unit Two (2) is in operation at 100% powe The following annunciator
, . and conditions exist:
A-07, 1-2 RFP FW CONTROL SIGNAL FAILURE.
[ Amber light above A RFPT lockout switch out Amber light above B RFPT lockout switch lit
]
-
A Reactor Recirculation Pump trips. Reactor power lowers and Reactor water level begins to rise uncontrollably. How can the operator
.
restore Reactor water level to the normal band?
' Operate RFP A MSC control in the Lower directio Operate RFP B MSC control in the Lower direction.
'~ Place RFP A MGU in Manual and lower output demand.
, Place RFP B MGU in Manual and lower output deman *
- .-
,
j QUESTION 41 POINT VALUE: 1.00 l
l The control Building Ventilation system has initiated in the radiation
- protection mode. .The alignment of system controls is
'
Emergency. Filtration Fan A Pref f Emergency Filtration Fan B Stby
.
- Emergency Filtration Fan A has been running for 10 minutes when a high
'-
temperature is sensed in the Train A charcoal bed. Assuming the ,
f radiation initiation signal is still present, Emergency Filtration Fan l
i= A will:
l'
j immediately trip and Fan B will immediately auto start.
- - ; remain running since the high temperature trip is bypasse {
i immediately trip and Fan B will start after a 10 second delay, i- .
i
. remain running since the high temperature provides alarm onl F PAGE 23
4
. - - - - - , ., -- ,.
-- . - - ~ - = - . - . - - . - . - _ - - ~ _ - - - . - ~ -
,
i ;
,! 1 5:
'
i
- "NRC 97-1 RO, Rev 0" EXAMINATION **
,
QUESTION 42 POINT VALUE: 1.00
,
'
A Loss of Off-Site Power has occured. Secondary Containment isolate Reactor Building Ventilation was restarted by guidance of EOP-03-SCCP l ., using SEP-0 Plant conditions are:
i 1 1' Reactor Water Level is +150' inches, slowly rising
'
Drywell Pressure is 1.2 psig, slowly rising CAC Vent Purge Isol Ovrd (CAC-CS-5519) is in OVERRIDE Reactor Building Vent Rad Monitors have been reset I PCIS Isolation Reset push buttons on P601 have been depressed j Which of the following would cause the Reactor Building to re-isolate?
f Drywell pressure rises above 2.0 psig.
- ! Reactor level drops to the Top Of Active Fue I
- .
j Main Stack Radiation Monitor exceeds the Hi'-Hi setpoin ;
}
_ Reactor Building Vent Exhaust temperature exceeds 140 QUESTION 43 POINT VALUE: 1.00 Following a Loss of Off-Site Power, Diesel _ Generator #1 (DG1) is
- running in AUTO, tied to Bus El. DG1 parameters
'. Kilowatt load 3500 KW 1- Terminal Voltage 4160 Volts Reactive load 1300 KVAR Frequency 60 Hz
'
Off-Site Power has been restored and the BOP buses energized from the
{ ' '
The operator depresses the DG #1 CONTROL ROOM MANUAL push button on i the RTGB. DG #1 frequency will be approximately: Hz.
- H H :
j Hz.
.
PAGE 24
,--
. __ . . . . . - _ ~
_ .- .- ._
. - . .. . - - - . . . .-
l
.
'
.
- "NRC 97-1 RO, Rev 0" EXAMINATION **
i t QUESTION 44 POINT VALUE: 1.00 During performance of PT-12.2C, Diesel Generator (DG) 3 has TRIPPED on i reverse power due to misoperation of the governor control switc A I Loss of Off-Site Power then occurs.
.
'
What operator action is required to start DG3 following the Loss of Off-Site Power?
- Reset the engine lockout only.
4 Reset the generator lockout only, . Reset the engine lockout, then the generator lockou ! Reset the generator lockout, then the engine lockou ;. QUESTION 45 POINT VALUE: 1.00 The following sequence of events occurs on Unit One (1) :
Time = 0 seconds off-site. power is lost ;
,
Time = 5 seconds A LOCA signal is recieved !
{ Time = 10 seconds Diesel generators energize thier respective E
'
Buses l-l The Motor Driven Fire Pump normal feeder breaker from:
] Bus El closes at Time = 25 seconds.
. Bus E2 closes at Time = 25 second ! Bus El closes at-Time = 30 seconds, Bus E2 closes at Time = 30 seconds.
.
W I
.
<
1
-
PAGE 25
?
i i
- "NRC 97-1 RO, Rev 0" EXAMINATION **
{
QUESTION 46 POINT VALUE: 1.00 Unit One (1) is in an outage with the SAT energized and the UAT in backfeed alignment. All backfeed selector switches are in the BACKFEED positio Electrical system alignment:
BOP Buses 1C/1D Powered from UAT E Buses E1/E2 Powered from BOP Buses DGs 1, 2, 3, 4 Operable in standby alignment A sudden fault pressure occurs in the Main Power Transformer resulting in a Backup Main Generator lockou How will the electrical distribution system respond? BOP buses 1C/1D transfer to the SAT, four DGs receive an auto start signa BOP buses 1C/1D transfer to the SAT, no DGs receive an auto start signa BOP buses 1C/1D are de-energized, four DGs receive an auto start signa BOP buses 1C/1D are de-energized, DGs 1 and 2 only receive an auto start signal.
QUESTION 47 POINT VALUE: 1.00 A fire in Diesel Generator (DG) Cell #1 has required entry into ASSD-03 for safe shutdown of both units utilizing Safe Shutdown Train B.
Status of the electrical plant is:
DG 3 and 4 Running, tied to associated E Buses DG 1 and 2 Tripped, not available 4KV Bus E2 Energized from 4KV Bus El via E3 E3 to El to E2 Cross-tie breakers in FIRE What will trip the El to E2 cross-tie breakers? LOCA signal on Unit 1 or Unit DG3 high lube oil temperatur DG3 ground overcurren DG3 differential overcurren PAGE 26
- . . . .-. - - - . - .
_ .
. _ . _ ~ . - - . . . . . - -
b
- "NRC 97-1 RO, Rev 0" EXAMINATION **
. QUESTION 48 POINT VALUE: 1.00 A'"250' V BATT A GROUND" alarm has been received on Unit one (1). The following readings are' reported from the Battery Room:
P Bus milliamps N Bus milliamps 'l Charger 1A-1 135 volts, in float Charger 1A-2 135 volts, in float Per OP-51 and AI-115, the ground is on the: P Bus, action level 1 applie P Bus, action level 2 applie N Bus, action level 1 applie N Bus, action level 2 applie QUESTION 49 POINT VALUE: 1.00 Unit One (1) jus operating at 100% power. The UPS system is in its normal alignment for both units. A total loss of UPS occurs on Unit one (1), followed shortly by a spurious Reactor scram. Plant conditions:
Reactor water level lowers to +135 inches Reactor pressure 950 psig controlled by EHC APRM recorders on P603 indicate 100% power Digital Feedwater controller displays are blank How will the loss of UPS affect the plant during this transient? Reactor power cannot be determined to be less than 3% from P60 Reactor feed pumps will not respond to the reduced Reactor water leve EHC pressure control will be lost as the main turbine coasts down to zero speed.
! Reactor Building HVAC is lost until the main stack rad monitor i is transferred to Unit Tw !
,
PAGE 27 l.
l . - .
. _ _ _ _ - - - . . _ . _ _ _ . _ _ _ _ . _ _ _ . . . . . _ _ _ .
l l
,
l-
- "NRC 97-1 RO, Rev 0" EXAMINATION **
i QUESTION 50 POINT VALUE: 1.00 An Auxiliary. Operator has received 1.95 rem TEDE for the current year.
- The AO is needed to perform work in a 25 mrem /hr fiel The work is l expected to last 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> In accordance with NGGM-FM-0002, Radiation Control and Protection Manual,-the worker requires
- approval from the Manager - E&RC to exceed the annual
,
administrative dose limit.
- approval from the Plant General Manager to exceed the annual administrative dose limi , approval from the Site Vice President to exceed the annual administrative dose limi no special authorizations since the annual administrative limit should not be exceede ,
l QUESTION-51 POINT VALUE: 1.00 Both Units have lost off-site power. The only available Diesel Generator is DG2. Buses E2 and E4 cannot be cross-tied. Unit Two (2)
UPS has been de-energized for DC load stripping.
! What instruments are available to monitor Reactor Water Level it the l Unit Two (2) RTGB?
l Fuel Zone indicator NO36 onl Fuel Zone indicator NO37 onl Fuel Zone indicator NO36 and Narrow Range indicators N004A/B/ I Fuel Zone indicator N037 and Narrow Range indicators N004A/B/ l I
.
l J
PAGE 28
!
i l
_ _ _ . - _ . _ _ , _ _
_ . - ~ _ . - - - -_ - -. . - . - . - . - . . . - . - - - - . - . . . . . . -
.
l
r e
i
!
- "NRC,97-1 RO, Rev 0" EXAMINATION **
f
!
!
QUESTION 52 POINT VALUE: 1.00 :
.
.i You are escorting a visitor with a red badge in'the protected are It l becomes desired to temporarily give up your escort dutie !
l How may this be accomplished? i
! You and the visitor must exit the protected are You may turn over escort duties to a security guard onl You may turn over escort duties to any other qualified escort, :
and notify security at the access point you entere You may turn over escort duties to any other qualified escort, and notify security at the secondary alarm statio ;
QUESTION 53 POINT VALUE: 1.00 A Temporary Procedure Change is developed and designated as Revision To Follo This Temporary Change receives:
Interim approval on March 2nd Final approval on March 6th What is the last date this Temporary change may be used WITHOUT l
receiving any allowable extention(s)? May 1s May 5t I May 31s June 4th.
I
! i
!
'
l
,
PAGE 29
- _ _ . _ _, ._ ,
!
- "NRC 97-1 RO, Rev 0" EXAMINATION **
l QUESTION 54 POINT VALUE: 1.00 Which of the following describes the level of use of Emergency Operating Proceduree?
,
' ! Reference use procedure Continuous use procedure ! Information use procedure Exempt from level of use requirement ,
l I
!
I QUESTION 55 POINT VALUE: 1.00 I
Per PLP-21, which of the following would be an UNACCEPTABLE method of performing independent verification by use of RTGB indications?
I Independent verification of a Core Spray System: l pump breaker closure using red light indicatio ;
b, pump breaker rack in status using green light indicatio injection valve opening using system flow indication meter.
j injection valve standby position using green light indicatio l
.
i
!
I
l PAGE 30
!
- "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 56 POINT VALUE: 1.00 A RCIC valve lineup checksheet is being performed near the end of a refueling outage. The working copy has been verified in accordance with the requirements of AP-00 Per AP-008, the working copy is REOUIRED to be reverified: daily using a controlled copy, or every 3 days using NRC daily using a controlled copy, or every 3 days by contacting document service every 3 days using a controlled copy, or every 7 days using NRC every 3 days using a controlled copy, or every 7 days by contacting document service I QUESTION 57 POINT VALUE: 1.00 During RTGB walkdown in preparation for shift turnover, the oncoming Reactor Operator notes that annunciator A-5 2-2, Rod Out Block:
Alarm is sealed in Has a yellow dot affixed to the window If the signal causing this alarm clears, then comes back in without the alarm being reset, the subsequent alarm condition is indicated by the alarm window flashing: slowly with an audible alar rapidly with an audible alar slowly, then rapidly without an audible alar ' rapidly, then slowly without an audible alarm.
,
PAGE 31 l
i
._ . . ._ _- .-_ _..- _ . - . . m .. _ __ _ . ._ ,
!
t
!
. !
-
. -
- "NRC 97-1 RO, Rev 0" EKAMINATION ** '
!
l
<
QUESTION 58 POINT VALUES.l.00 l
!
l A motor-operated valve has-been manually backseate This valve is: {
t non safety related .
normally operated from the RTGB i i
~
.AP-013 requi res that the valve: ! control switch is caution tagged denoting the valve is i
backseate .
.e T handwheel is locked in position by use of an approved valve
- locking device.
- .
l must be manually operated off of the backseat prior to motor operatio motor breaker is placed under clearance in the off position.
d QUESTION 59 POINT VALUE: 1.00 i
A plant shutdown is in progress per GP-05 for refuelin Current plant 4 conditions are:
! Reactor power 20%
- Drywell oxygen 19.0% !
Drywell entry'may be made with the approval of the Manager-E&RC and the
.
authorization of the Shift Superintendent only if reactor power is
- reduced at least
- %, oxygen concentration is acceptable.
' %, oxygen concentration must be raised.
i-l %, oxygen concentration is acceptable.
I %, oxygen concentration must be raised.
'.
i l PAGE 32
_ _ .
. . , _ _
_. . . _ . _ _ _ _ . __ _ . _ .. .._ -.. . __ ._.
.
'
i
,
- "NRC 97-1 RO,'Rev 0" EXAMINATION **
,
! i i
l QUESTION 60 POINT VALUE: 1.00 '
Which of the following valve / actuator types may be used as a clearance
. boundary isolation _ component, provided the associated restrictions of
,4 AI-58 are satisfied?
, Solenoid operated ball valve marked as fail-closed on the print.
- Motor operated globe valve normally used as a flow control valve.
4 Pressure balanced diaphragm operated pinch valve not equipped ,
with a handwhee i
i Double acting cylinder operated butterfly valve not marked as fail-closed on the print.
.
. QUESTION 61 POINT VALUE: 1.00 A clearance request has been received for a fluid system with the ;
- following normal operating parameters
f Pressure 475 psig
' Temperature' 175*F Per AI-58, the clearance: May_use single valve boundary isolatio Must use dual valve boundary isolation due to pressure onl Must use dual valve boundary isolation due to temperature onl Must use dual valve boundary isolation due to pressure and temperatur i i
PAGE 33
- "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 62 POINT VALUR: 1.00 During accident conditions on Unit One (1), the following plant conditions exist:
Reactor water level -70 inches (Fuel Zone)
Reactor pressure 1100 psig Drywell average temp 190*F Drywell ref leg area temp 270*F Injection sources None available Under these conditions, peak fuel clad temperature will not exceed: *F, provided Reactor water level remains above -80 inches, *F, provided Reactor water level remains above -90 inche *F, provided Reactor water level remains above -80 inche *F, provided Reactor water level remains above -90 inches.
QUESTION 63 POINT VALUE: 1.00'
Following accident conditions, the crew is executing the Reactor Vessel Flooding Procedure, EOP-01-RXF Plant conditions are:
Control rods Fully inserted Reactor water level Unknown The operator is directed to control injection flow to the Reactor to maintain at least SRV/ ADS Valves open and Reactor pressure: ; above the Minimum Alternate Flooding Pressure, ; above the Minimum Alternate Flooding Pressur ; at least 50 psig above suppression chamber pressure, ; at least 50 psig above suppression chamber pressur PAGE 34
.- __ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - > . __ . _ _ , - . ._ m . . . . _ . . . . _ . _ _ _ . _ _ _ . . . _ _ _ _ _ _ . . _ . . . _ . .
- "NRC 97-1 RO, Rev 0" EXAMINATION **
! . QUESTION 64 POINT VALUE: 1.00 i
'
, A heavy influx of marsh grass on the Circulating Water Screens has caused a-loss of all Circulating Water pumps and a reactor scra Plant conditions are:
Group 1 isolated
,
Condenser vacuum is 0" Hg
! Turbine speed is 500 rpm, dropping EHC Electrical Malfunction in alarm due to loss of the PMG.
- 'The marsh grass is now cleared and the Circulating Water System has l been restarted. Is the Main Condenser available as a heat sink? No , the MSIVs are closed, No , the-EHC system is not availabl No, the condenser is not under vacuu Yes, all required systems are availabl QUESTION 65 POINT VALUE
- 1.00 l
Following a loss of all high pressure injection on Unit One ( 1 )' , seven ADS valves have been manually opened to restore' adequate core coolin Plant conditions are now:
Reactor water level +25", N026A/B Reactor water level +140", NO36/37 Reactor pressure 25 psig Drywell average temp 155*F Drywell ref leg area temp 280*F l
[ Reactor water level may be determined using:
I N026A/B onl NO36/37 only.
4 Both NO26A/B and NO36/37.
! Neither N026A/B nor NO36/37.
,
I i
!
j PAGE 35 l l
.
. .
.. ..
. . - - - - . -. .~ _ - -.. -.
l i !
'
!
i
l !
- i l * * - "NRC 97-1 RO, Rev 0" EXAMINATION ** !
! I
!
!
l QUESTION 66 POINT VALUE: 1.00 A Unit Two (2) reactor scram has occurred. Seven control rods failed ;
to fully insert and are between positions 08 and.1 Conditions are-
'
All APRM Downscale lights are LIT ;
MSIVs are open !
Total Steam Flow 3.6 E6 lbm/Hr, dropping
!
Reactor Pressure 900 psig, dropping :
Narrow Range Level Instruments (N004s) +155 inches, rising l Master Feedwater.setpoint at +170" '
Two Reactor Feed Pumps in operation l
With current plant conditions, the operator is required as an !
IMMEDIATE action to:
! i trip the Main Turbine.
! trip one Reactor Feed Pum place the Mode Switch to SHUTDOW enter Alternate Control Rod Insertio j i
? l l
!
QUESTION 67 POINT VALUE: 1.00 l
Following a group 1 isolation and a reactor scram, the operating crew is performing the Reactor Scram Procedure, EOP-01-RS Plant conditions are:
i Reactor water level 195 inches, slowly rising (N004s)
l Reactor pressure 800-1000 psig, controlled by SRVs L Drywell pressure 1.0 psig, slowly rising l Suppression pool temp 94*F, slowly rising l Suppression pool level -27.5", slowly rising l The operating crew is recuired to enter EOP-01-RVCP and execute l concurrently with the scram procedure if: drywell pressure rises to 1.5 psig.
) reactor water level rises to +230 inche I
1
' suppression pool temperature rises to 111* ! i
' suppression pocl level rises to -26.5 inches.
)
I l PAGE 36 i
_
. . _ . . _ - _ . . . - - ~ . . . . - . . . _ . . . _ . . . . - - ~ ~ - . , . - . . . . - - . - . - - . - - . - - . ,
,
t
i i
I' !
'
!
- "NRC 97-1 RO, Rev 0" EXAMINATION **
!
I
-
!
,
QUESTION 68 POINT VALUE: 1.00 Following-a Unit'One (1) Reactor scram, the crew has entered and~is ;
executing EOP-01-RSP, Reactor Scram Procedur Plant conditions are- '
!
Reactor water level 220", rising Reactor pressure 945 psig, stable MSIVs- Open ;
Per EOP-01-RSP, the MSIVs must be manuall'y closed if Reactor water level cannot be maintained below: l
- ~
230" l
l L "
>
r "
l l- "
l
'
.
l'
!. QUESTION 69- POINT VALUE: 1.00-l l The entry conditions for Unit One (1) - EOP-01-RVCP, React'or Vessel l Control Procedure:for Reactor pressure and water level are Reactor pressure is greater than:
- psig, or Reactor water level less than 153".
l
! psig, or Reactor water level less than 166". pstg, or Reactor water level less than 153".
l L psig, or Reactor water level less than 166".
l
!
l l
l l
'
,
i l
PAGE 37
, - - -
. . - . - _ _ - - - ---- , .-. -. --.
- "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 70 POINT VALUE: 1.00 During ATWS conditions on Unit Two (2) , injection to the Reactor has been terminated and prevented'to suppress Reactor power. Plant conditions are:
Reactor water-level -35 inches (Fuel Zone), lowering Reactor power 7%
Reactor pressure 950 psig MSIVs Closed SRVs One open CAC-TR-4426-1B, Point A 195*F CAC-TR-4426-1B, Point B 203*F CAC-TR-4426-2B, Point A 187'F CAC-TR-4426-2B, Point B 191*F Assuming APRMs are NOT downscale, what is the HIGHEST indicated Reactor water level injection may be re-established to the Reactor? inches inches inches inches QUESTION 71 POINT VALUE: 1.00 Following an incomplete Reactor scram, the operating crew is executing EOP-01-LPC, Level / Power Contro A decision step is reached asking
"Is The Reactor Shutdown?".
Which of the following conditions would satisfy the definition of
" SHUTDOWN" as it applies to the Reactor? All operable APRMs indicate downscal The Reactor is subcritical on range 6 of IRM The entire SLC Tank has been injected to the Reactor, Hot Shutdown Boron Weight has been injected to the Reacto i PAGE 38
--
. . . . - . - . . - . - . . - - - . . . . . . ~ . - - .- - . ~ . - . . - . - - . --
J
.
.
- "NRC 97-1 RO, Rev 0" EXAMINATION **
.
.,
QUESTION 72 POINT VALUE: 1.00 Following a reactor scram and a group 1 isolation, SRVs are being used to maintain reactor pressure 900-1000 psig.
.
Which of the following conditions requires ALL group 1 isolations to
- be defeated and the reactor vessel rapidly depressurized to tl.e main condenser?
3 Suppression Pool Level is +4' 6" Suppression Pool Temperature is 95*F Suppression Pool Level is -l' 6" Suppression Pool Temperature is 170*F Suppression Pool Level is -4' 3"
. Suppression Pool Temperature is 156*F 4 Suppression Pool Level is -8' 1" Suppression Pool Temperature is 105*F j
j QUESTION 73 POINT VALUE: 1.00 l
i During accident conditions, Reactor Water 3evel cannot be restored
! above the Top of Active Fuel. Service Water is injecting to the Reactor Vessel to raise Containment Leve Service Water injection MUST be secured, IRRESPECTIVE of Adequate Core '
Cooling when Primary containment level reaches: feet, to prevent covering the highest reactor vessel vent path capable of rejecting all decay heat, feet, to prevent covering the highest primary containment vent path capable of rejecting all decay hea ! .5 feet, to prevent covering the highest reactor vessel vent path capable of rejecting all decay hea .5 feet, to prevent covering the highest primary containment vent path capable of rejecting all decay hea !
l PAGE-39
.,_ . _ . . . _ _ . _ -._._ _ . . _ . . _ . _ _ _ _ _ _ - - . . . _ . _ _ _ _ _ _ _ . _ _ _ . _ _
r
I
!
l-t t
l
- "h'RC 97-1 RO, Rev 0" EXAMINATION ** ,
t
!
l ;
I -QUESTION 74 POINT VALUE: 1.00 !
L Following a large Recirculation line rupture, EOP-01-PCFP, Primary Containment Flooding Procedure, is being exectuted.. The following '
indications are available:
CAC-LI-2601-1 +5.9 feet '
CAC-PI-1257-2A 23 psig
.CAC-PI-1230 21 psig ,
CAC-PI-4176
~
25 psig I
CAC-PR-1257-1 22 psig ;
What is Primary containment water level? +14.5 feet +9.9 feet +7.6 feet +4.6 feet QUESTION 75 POINT VALUE: 1.00
.
A seismic event has occurred that has resulted in a Loss of Off-Site Power and high power ATWS condition !
!
The SLC Storage Tank outlet line completely severed at the tank during the earthquake. The SLC tank is EMPTY making the SLC pumps unavailable for boron injectio l Which system should be selected for alternate boron injection? CRD RCIC i RWCU Condensate
! ,
L 1
!
PAGE 40
,p
. . _ - - _ __ - _ _ _._.. ._ -._. - . . _ . . _ . . _ . _ . - . . . _ . _ - . _ _ . - _. _ .
!-
'
!
- "NRC 97-1 RO, Rev 0" EXAMINATION **
l
!. ' QUESTION 76 POINT VALUE: 1.00 A Condensate header rupture in the cable spread area of the Control
!
Building has resulted in a loss of all UPS and RPS powe _
Plant status is as follows:
Blue scram lights 137 illuminated '
IRM Indications 50 on Range 10 l l
l What method of EOP-01-LEP-02, Alternate Control Rod Insertion, would be l MOST effective in inserting the withdrawn rods? , Vent the scram air heade Vent the overpiston area of control rod ,
l Scram individual rods with the scram test switche Insert control rods with the Reactor Manual Control Syste i
. QUESTION 77 POINT VALUE: 1.00
.During a low reactor water level condition, Alternate Coolant Injection using demineralized water is being aligned using the HPCI system. A ;
valid HPCI isolation signal is present, resulting in an automatic ;
closure signal to the HPCI Injection valve (E41-F006). I How is the HPCI Injection Valve (E41-F006) opened to provide injection to the reactor?
E41-F006 is opened from the: RTGB after placing the HPCI ASSD Interlock Defeat Switch on the RTGB to BYPASS.
( MCC by placing the breaker's NORMAL / LOCAL switch to LOCAL to bypass valve interlocks.
,
i RTGB and the breaker at the MCC is opened by an AO when the valve indicates full ope i RTGB after jumpers are installed to bypass the valve auto
>
closure interlocks.
-
l
) PAGE 41
, - - - - - - - ,
. .. . , _ _ . ._.___.__.m.____..__.__._..._.___.m__ _ . _ , _
l l
!
l
- "NRC 97-1 RO , Rev 0" EXAMINATION **
l QUESTION 78 ' POINT VALUE: 1.00 Following a loss.of drywell cooling, a small steam leak in the drywell results in the following containment conditions: ;
Drywell pressure 9 psig, rising Suppression chamber pressure 8 psig, rising Suppression pool level +2 feet Average Drywell temp 270*F, rising The crew is directed to initiate drywell spray to control drywell temperatur Under current plant conditions, drywell spray may: l be initiated, all required conditions are me NOT be initiated, suppression pool level is too hig NOT be initiated, suppression chamber pressure is too lo , NOT be initiated, conditions are in the UNSAFE region of the !
Drywell Spray Initiation Limit.
,
i
'
QUESTION 79 POINT VALUE: 1.00 l The Suppression Chamber Spray Initiation Pressure is in the Suppression Chamber and is based on: .7 psig; intrusion of air into. primary containment due to Reactor Building-Torus vacuum breaker operation, .7 psig;. the lowest suppression chamber pressure that RHR-system logic will allow sprays to be initiated.
L psig; 95% of the noncondensibles in the drywell have been
!
transferred to the suppression chamber airspace.
i
! psig; the highest pressure that initiation of sprays will prevent exceeding the Pressure Suppression Pressure Limit.
i
<
s I
PAGE 42
_ __ _ _ _
. . ___... . _ _ . _ _ . _ _ _ . . _ _ _ _ _ _ _ . . . _ _ _ _ _ _ _ . - _ . . _ . . . _ . . - . . _ . _ . . . . . _ . . . _ _ _ .
.
- - "NRC 97-1 RO, Rev 0" EXANINATION' **
QUESTION 80 POINT VALUE: 1.00-During an accident on Unit One (1), the following primary containment and plant conditions exist:
l Reactor' pressure 798 psig Suppression pool level -42 inche Suppression pool temperature 171*F Suppression chamber pressure 17 psig l' Current conditions are in the: SAFE region of all Containment Limits, b. ' UNSAFE region of the Heat Capacity Level Limi ~
l UNSAFE region of the Heat Capacity Temperature Limi UNSAFE region of the Pressure Suppression Pressure Limit.
l
,
l QUESTION 81 POINT VALUE: 1.00
. A primary system discharging into Secondary Containment has resulted in
[ - one' area exceeding the Maximum. Safe Operating Radiation Level, but
!
'
within the EQ envelop. The radiation level in this area is subsequently reduced below the Maximum Safe value.
( ..
l A second area subsequently exceeds its Maximum Safe Operating Radiation u Level. 'What. action is required by the Secondary Containment Control'
!
Procedure?
a '. ' Shutdown the Reactor per GP-0 I 1 Scram the Reactor and initiate a cooldown s100'F/ Hour, l Scram the Reactor and initiate a cooldown >100*F/ Hour, Scram the Reactor and open seven ADS valves.
i'
1
,
PAGE 43 l
(
_ . . . . - ~. . _ -- . _ . , _ .. --
, .__ _ _ _ - _ _ . . _ _ . _ . _ _ . _ . _ . _ . . _ _ . . . . . . . _ . . . _ _ _ . . . . ~
-
i i l; 1 <
- "NRC 97-1 RO, Rev 0" EXAMINATION **-
l I I i !
, QUESTION 82 POINT VALUE: 1.00 t
i -While performing PT 10.1.1, RCIC OPERABILITY TEST, the RCIC steam
, supply line ruptured. RCIC failed to automatically isolate and ;
! attempts to manually isolate.RCIC are unsuccessfu j The following Steam Leak Detection NUMAC channels are in alarm: ;
. B21-XY-5949A, Channel Al-3, reading 298'F B21-XY-5949B, Channel Al-3, reading 294*F 3 B21-XY-5948B, Channel A5-4, reading 301oF B21-XY-5948A, Channel A5-4, reading 303*F
'
No other channels are in alar What action is recuired to be be
taken? f (
. i !
, Scram the reactor and emergency depressuriz l l Scram the reactor and commence a cooldown at normal rate I
!
i
i Scram the reactor and rapidly depressurize to the main
,
condense l
i Shutdown the reactor using GP-05 or scram the reactor as !
'
directed by the Shift Superviso :
l
!9
,
b
-
t e
4 '
i I
,
&
!
-
PAGE 44
---.-.. . - _ . . . -
- .- . - - - . - . . . . . - - . ... - -_ . - . . - . . - ~ - . . . - -
I l
!
l
- "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 83 POINT VALUE: 1.00 l Unit Two (2) is operating at power when a. rupture of RWCU piping downstream of the Non Regenerative Heat Exchangers occurs. RWCU
,
'
!
Inboard Isolation Valve (G31-F001).and Outboard Isolation Valve (G31-F004) BOTH fail in the open positio Plant conditions:
i Rx Bldg 50' temp 135'F Rx~ Bldg 20' temp 105*F
-
S Core Spray Room Flood Level Hi Hi alarm sealed in S RHR Room Flood Level Hi alarm sealed in
.The operating crew is required to enter EOP-03-SCCP ands continue attempts to isolate the leak, commence an immediate
, plant shutdown per GP-05.
L I
~ continue attempts to isolate the leak, scram the reactor when 50' temperature exceeds 140*F.
- immediately scram the reactor and consider anticipation of l emergency depressurizatio immediately scram the reactor and open seven ADS valves for j emergency depressurization, j l
l b
l l 1 3 l
!
l l
!
!
!
i-l PAGE 45 I
,. -
. - - . . - _ _ _ . - . - - . _ . - . . - - . - - - . . - - . . . . . . . _ . . -_
l
'
'
,
L ! ! i I -
I
>
- "NRC 97-1 RO, Rev 0" EXAMINATION **
l l
l
,
QUESTION 84 POINT VALUE: 1.00 Following core damage, an unisolable steam leak in the Turbine Building
, requires declaration of a General Emergency due to loss of three out of l three fission product barriers.
I j The crew is executing EOP-04-RRCP, Radiation Release Control Procedur Field surveys and Off-Site dose projections (PEP-03.4.7) are'being
,
l l
l performed.
!
l When, per EOP-04-RRCP, is Emergency Depressurization of the Reactor l required to be initiated?
a. Immediately since a General Emergency has been declare b. If the Off-site release rate exceeds the Emergency Action Level
.
for an Alert.
L l c. If the Off-Site release rate approaches or exceeds the Emergency Action Level for a Site Area Emergenc d. If the off-Site release rate approaches or exceeds the l' Emergency Action Level for a General Emergency.
t
!
QUESTION 85 POINT VALUE: 1.00 i
!
During emergency conditions, it has been determined that exposures in j excess of 10CFR20 limits may be required to protect equipment important j to reactor safety which is needed to protect the population of ,
'
Brunswick County from a large releas In accordance with PEP-03.7.6, the Emergency Worker Dose Limit is:
,
a. 10 Rem TED This dose may be authorized by the Site Emergency l Coordinato ,
b. 10 Rem TEDE. This dose shall be authorized by the Radiological Controls Manage l c. 25 Rem TED This dose may be authorized by the Site Emergency Coordinator.
i d. 25 Rem TED This dose shall be authorized by the Radiological
!
Controls Manager.
l l.
'
PAGE 46
,. . _
.
-- - .- . - . . . - . . . - - ..- - . . . - . - . - .-
!
l
'
,
!
,
- "NRC 97-1 RO, Rev 0" EKAMINATION ** .
QUESTION 86 POINT VALUE: 1.00 A Unit Two ' (2) Reactor startup is in progress per GP-0 Heatup and
,
pressurization of the Reactor is being performed.
!
The operating CRD Pump trip Attempts to restart CRD per OP-08 and-AOP-02.0 are unsuccessfu AOP-02.0 requires the operator to insert a manual Reactor scram only if Reactor pressure is below: psig > psig psig psig
,
QUESTION 87 POINT VALUE: 1.00 l Unit One (1) is operating at 100% power when Recirculation Pump 1B trips, resulting in the following conditions:
Total Core Flow (P603) 39 Mlbm/ Hour 4 Total Core Flow (U1CPWTCF) 35 Mlbm/ Hour i Indicated Core Plate DP 4.7 paid APRMs 68%
LPRM Upscale /Downscale alarms None What region of the Thermal Power Limitations Map is the plant operating in, and what operator action is required to be taken? Region B, raise total core flo Region B, insert control rods per ENP-2 Region A, immediately insert a manual scra % Buffer, increase monitoring of nuclear instrumentatio I'
L i l PAGE 47
-.
. ._ _ . _ _ . _ . . . . - . - _ _ _ _ . . . . ..___.__.__.__._._._...m _ _ . _ .
1
i
- "NRC 97-1 RO, Rev 0" EXAMINATION ** 4
- .
"
!
- ._ QUESTION 88 POINT VALUE
- 1.00
.. .
!
Unit Two (2) was operating at power when a trip and lockout of BOP bus 2B required the operator to insert a manual Reactor scra Shortly following the scram, the following indications are noted:
3 Recirc pump A #1 seal pressure 1000 psig
Recirc pump A #2 seal pressure 1000 psig Recirc pump B #1 seal pressure 100 psig
'
Recirc pump B #2 seal pressure 50 psig f
- Drywell pressure 1.4 psig, rising Average drywell temp- 140*F, rising
Average primary containment temp 126*F, rising The operator is required to enter: AOP-14.0 and isolate Recirc pump < AOP-14.0 and isolate Recirc pump B.
'
, EOP-02-PCCP and isolate Recirc pump EOP-02'-PCCP and isolate Recirc pump ~
!
i-
QUESTION 89 POINT.VALUE: 1.00 A situation arises requiring immediate evacuation of the control room 4 prior to completion of any immediate actions per AOP-3 RPS is
- aligned
RPS Bus A Powered from RPS MG Set A l
RPS Bus B Powered from RPS MG Set B If the RPS EPA breakers are opened in the exact sequence specified by
~AOP-32.0, opening which EPA breaker will result in a reactor scram?
[
j EPA Breaker 1.
! EPA Breaker 2.
4
-
PAGE 48
- . _ , , __
_ . . _ _ _ . . . . . _ ._ - _ . _ . -_ - _ . - _ . - - _ _ - . _ - - - ~ - - - - -
!
,
- "NRC 97-1 RO, Rev 0" EXAMINATION **
' QUESTION 90 POINT VAI.UE: 1.00
!
-Following a loss of feedwater on Unit Two (2), HPCI and RCIC are being -
used to' restore Reactor water level to the normal band. The operator !
notes the following alarms and indications: ,
250 Batt B Under Voltage Alarm sealed in Battery Bus 2B-1 Voltage 0 volts (XU-2) ,
' Battery Bus 2B-2 Voltage 0 volts (XU-2)
Battery Bus 2B-1 Voltage 0 volts (ERFIS)
Battery Bus 2B-2 Voltage 0 volts (ERFIS)
l How is the operation of HPCI and RCIC affected by the power loss?
l L HPCI continues to inject to the Reactor, RCIC isolates due to loss of isolation logic powe . RCIC continues to inject to the Reactor, HPCl isolates due to loss of isolation logic powe HPCI continues to inject to the Reactor, RCIC coasts down due i to loss of flow controller powe RCIC continues to inject to the Reactor, HPCI coasts down due l to. loss of flow controller powe I
!
l i
h
>
l
!
!
!
r I
l
!
PAGE 49 i
i !
o
- "NRC 97-1 RO, Rev 0" EXAMINATION **
I QUESTION 91 POINT VALUE: 1.00 Following a Loss of Off-Site Power to Unit One (1), the operator is performing AOP-3 Plant conditions are i
Diesel Generator 1 Running at 3575 KW load I Diesel Generator 2 Running at 3680 KW load '
Reactor Building HVAC Isolated j l
The operator is directed to restart Reactor Building HVAC using three j (3) supply fans (75 KW each) and three (3) exhaust fans (4 5 KW each) .
How will starting two supply and exhaust fans from MCC 1XG and one supply and exhaust fan from MCC 1XH affect Diesel Generator (DG)
maximum loading? l DG1 only maximum load will be exceede DG2 only maximum load will be exceeded, DG1 and DG2 maximum load will be exceede DG1 and DG2 will remain within maximum load limit QUESTION 92 POINT VALUE: 1.00 Following a Loss Of Off-Site Power on Unit One (1), Diesel Generators
- 1 and #2 are tied to their respective Emergency Buses. A rupture of the Unit One (1) Nuclear Service Water Header in the Service Water Building results in the following indications:
Nuclear Header Pressure (XU-2) O psig Nuclear Hdr Serv Wtr Press-Low Alarm sealed in Diesel Generators #1 and #2: have no available cooling and will tri have no available cooling but will continue to ru cooling water supply will automatically transfer to the Unit Two (2) Nuclear Service Water heade cooling water supply will automatically transfer to the Unit One (1) Conventional Service Water header.
I PAGE 50
!
. - . . - . . . - ~ - .. - - - . . . - - . - . - . - . _ _ - - ~ - - . . , . . .
.
' j I
5 l i I
'
i i !
j
- "NRC 97-1 RO, Rev 0" EXAMINATION **
I 2 ' QUESTION 93 POINT VALUE: 1.00
Unit Two (2)' was ' at 20% power during ' plant startup when a sudden rise
, in off gas flow is accompanied by a lowering condenser vacuum. The reactor was manually scramme Plant conditions:
"
Condenser vacuum 15" Hg, slowly lowering
, Reactor pressure 921 psig, steady
.
Bypass valves One partially open J
What is the minimum additional reduction in condenser vacuum that would
) result in the loss of automatic Reactor pressure control? " H " H " Hg.
i i " Hg.
! I i
d
. QUESTION 94 POINT VALUE: 1.00 )
i
.
Unit one (1) is performing refueling' operations. A fuel' handling- ;
j' accident results the the following radiation alarms:
,
[ Area _ Rad Refuel Floor Hi (white alarm)
j Area Rad Rx Bldg Hi (red alarm, blue bar)
2 Area Rad Control Room Hi (red alarm,. red bar)
-Rx Bldg Vent Rad Hi (red alarm, blue bar)
,
Rx Bldg Vent Rad Hi Hi (red alarm, blue bar)
,
i How will the Reactor Building HVAC and Control Building Emergency Air
[ Filtration (CBEAF) systems respond to the above radiation alarms?
f Reactor Building HVAC isolates, CBEAF remains in standby.
! Reactor Building HVAC remains in operation, CBEAF initiates.
'
' Reactor Building HVAC remains in operation, CBEAF remains in standby.
j Reactor Building HVAC isolates and CBEAF' initiate ,
i l PAGE 51 '
i
.. . - - --
- - . . . - . ~ . - _ . - = - . __ ..~.- . - - -., . ~ . - - _,
,
+
,
'
- "NRC 97-1 RO, Rev 0" EXAMINATION . **
?
!
QUESTION 95 POINT VALUE: 1.00 1 During normal full power- operation of Unit Two (2) , the following alarms and indications are noted:
Air Compressor D Trip Alarm sealed in Air Compressors A/B/ Running Instrument Air Pressure low Alarm sealed in Instrument Air header pressure 100_psig The operator should verify that air compressors A, B and C are loaded and that: Service air isolation valves, PV-706-1 and PV-706-2, have automatically close , Interruptible air isolation valves, PV-722-1 and PV-722-2, have
! automatically close Standby reactor building air compressors have automatically started and loade Backup nitrogen rack isolation valves, RNA-SV-5482 and SV-5481, have automatically opene i QUESTION 96 POINT VALUE: 1.00 Following a loss of shutdown cooling, Alternate Shutdown Cooling has ;
l been' established per AOP-1 Plant conditions are:
'
l RHR Loop A in suppression pool cooling
RHR Loop B injecting to the Reactor Vessel Reactor pressure is 115 psig SRV G is open It becomes desired to make a slight adjustment to raise the cooldown rate. .This may be accomplished by closing SRV G and opening: SRV H.
, SRV J.
r-
'** "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 97 POINT VALUE: 1.00 Unit Two (2) is operating with the following plant conditions:
Reactor power 85%
Core flow 51 Mlbm/hr Rod line 110%
Recirc MG sets Scoop tubes locked Which of the following conditions authorizes the operator to manually initiate Select Rod Insert (SRI)? Condenser vacuum lowers and approaches the turbine trip setpoin Feedwater heating is partially lost and APRMs approach the scram setpoin A reactor feed pump t.;ips and reactor level approaches the scram setpoin A recirculation pump trips placing the plant _in. Region A of the Thernal Power Limitations Ma .
PAGE 53
. - . - . . - . . . . . . . . . - . - - . ~ . ~ . - . - . . . . . - . . = . . - . . - . ~ . . ~ - . - ~.-..- --
!
,
- "NRC S'l-1 RO, Rev 0" EXAMINATION **
, 1
- QUESTION 98 POINT VALUE: 1.00 Following a loss of Off-Site Power to BOTH Units, the following conditions exist on Unit One (1) :
DG 1 Tripped DG 2 Tripped HPCI Unavailable RCIC Operating at rated flow !
1 #
The operator is required by AOP-36.,2Y to commence a Reactor cooldown. at ;
a rates ! greater than 100*F/ hour to between 50 and 150 psig Reactor pressur greater than 100*F/ hour to between 150 and 300 psig Reactor l
! pressure, j i
l as close as possible, but not to exceed 100*F/ hour to between 50 and 150 psig Reactor pressur as close as possible, but not to exceed 100*F/ hour to between j 150 and 300 psig Reactor pressur '
l l i
l
!
l
i
'
,
T i j PAGE 54 i
, _ --
_. . _ . . . _ - _ _ _ . _ . _ _ . _ . . . _ . _ . _ _ _ _ _ _ . _ _ _ _ . _ . . ._ _ . _ _ _ _ __ .
!
l
- "NRC 97-1 RO, Rev 0" EXAMINATION **
QUESTION 99 POINT VALUE: 1.00
Core defueling is in progress. All control rods are fully inserted into the reactor core. A fuel assembly has just been placed in the fuel pool and unlatched. The main hoist has been raised to a safe elevation to pass through the cattle chute (NOT normal-up) with the bridge still over the fuel pool location.
l The next step requires that a fuel assembly be removed from the reactor l core and placed in the fuel poo When will the ROD BLOCK INTERLOCK #1 light on the Interlock Status Display Panel first light as the next step is performed?
l l As the bridge is moved near the reactor core (LS1 is actuated). When the bridge is over the reactor core (LS1 is actuated)
and the main hoist is lowered into the reactor vesse When the fuel assembly is latched, with both grapple hooks closed.
,
l When the fuel assembly is being raised and the main hoist l loaded signal is actuated.
l l QUESTION 100 POINT VALUE: 1.00
!
Core Spray Pump 2A 4160 volt breaker is racked in per OP-50, with 125V DC available at the switchgear. A LOCA results in a condition l requiring the auto start of the pump.
!-
'
Refer to LL-09113 sheet 15, Core Spray Pump 2A Control Wiring Diagra The pump breaker is closed by energizing the: X coil, when relay K12A is de-energized and relay K15A is energize Y coil, when relay K12A is de-energized and relay K15A is
! energized.
,
! X coil, when relay K12A is energized and relay K15A is de-energized.
l Y coil, when relay K12A is energized and relay K15A 10 de-energized.
,
[
! PAGE 55 i
- . _
,
- "NRC 97-1 RO, Rev 0" EXAMINATION **
l l
l
,
,
l i
- END OF "NRC 97-1 RO, Rev 0" EXAMINATION **
PAGE 56 OF 56 l
~ ._ .. . ~ ,_ - _.. - ._ . . . - - - . . .. . ..=. . .~. _ . . _~ . . .
HAME: bNfO D CATE / / SCDRE: _
/ i CRADED BYs ALTERNATE GRADERS (if required)
EKAM: NRC 97-1 PO, Rev 0 CLASS: RLC 96-1 COURSE CODES POA019 ABCh 2 AB@D 4 hBCD 61. @BCD A@CD 2 @BCD 4 ABCh 6 ABhD ABhD 2 A@CD 4 hBCD 6 ABCh'
_ ABCh 2 hBCD 4 ABCh 6 ABC@ @BCD 2 AhCD 4 ABC@ 6 hBCD AB@D 2 ABhD 4 AhCD 6 @BCD
- :*; ::: : :*; ::: e::*::::::: l @BCD~ 2 @BCD 4 ABhD 6 ABCh 1 ABCh 3 ABC@ 5 AhCD 7 hBCD 1 ABhD 3 AhCD 5 AB@D 7 AhCD !
_
1 A@CD 3 ABhD 5 AB@D 7 ABCh 1 AB@D ' 33, hBCD 53. @BCD 7 ABCh 1 AhCD 3 ABhD 5 ABC@ 7 AhCD 1 hBCD 3 ABCh 5 AhCD 7 ABhD 1 ABCh 3 @BCD 5 ABhD 7 AhCD 1 @BCD' 3 A@CD 5 ABCh 7 ABhD 1 A@CD 3 ABC@ 5 hBCD 7 AhCD 1 hBCD 3 @BCD 5 AhCD 7 ABhD l 2 A@CD 4 A@CD e ABC@ 8 A@CD
j
-
. _ . _ .. _ _ _ _ . _ _ _ . _ _ . _ . _ . _ . . . . . . . _ _ _ _ __
-
RD Answer h) Dms , , Scm
'
ORADED BY ALTERNATE GRADERS (if required)
EXAMS NRC 97-1 RO. Rev 0 CLASS: HLC 96-1 CDURSE CODE: ROA019 8 AhCD- 1 82. hBCD 8 AB@D 8 ABC . ABhD 86. 'ABCh i
,
87. 'AhCD
.8 Ah'CD 8 9 '. ABCh 9 AB@D 9 ABCh 9 AB@D 93. $BCD 9 ABCh 9 hBC ,
9 ABCh 9 A$CD
.
9 AhCD 9 @BCD l 10 ABhD
- .
1 i
,
l
- i l i
'j !
.
-
l
'
l
!
'
!
I i
, -- . . , _ , - - - . , - . - . . _- - - -- _-=. -.
.
_ - - - -- .---.-_--..--- _-
I
!
i i
'
}
.
, REFERENCES i
i
FOR WRITTEN TEST I
.-
1
i l
i l '
i i
N.
!
.
<
1 l
,
k i
i e
l l
}
t
!
i
'
i
!
i I
i
}
<
l
!
f a
,
G
!
I_.___ -
, ' -
i
,.. 4
- )
EOP-01-U Attachm2nt 6 Page 1 of 17 I i
i i I
<
j l
!
-
EOP-01-UG Attachment 6 Reactor Water Level Caution 1 (Caution 1) i
~1 s-l
.
l
'
l
!
i l OEOP-01-UG l Rev. 24 l Page 83 of 131 l
_
, -
1 U EOP-01-UG Attachm:nt 6 Page 2 of 17 ATTACHMENT 6 REACTOR WATER LEVEL CAUTION
,
(Caution 1)
A reactor water level instrument may be used to determine reactor water level only when the conditions for use as listed in Table 1 are satisfied for that instrumen TABLE 1 CONDITIONS FOR USE OF REACTOR WATER LEVEL INSTRUMENTS E2TE Reference leg area drywell temperature is determined using Figure 13, ERFIS,
,
or Instructional Aid based on Figure 1 NOTE Immediate reference leg boiling is not expected to occur for short duration excursions into the unsafe region due to heating of the drywell. The thermal j time prohibit will constant associated with the mass of metal and water in the reference leg immediate boiling of the reference leg, Reference leg boiling is an obvious phenomeno Large scale oscillations of all water level instruments associated with the reference leg that is boiling will occu This occurrence will be obvious and readily observable by the operato Additionally, if the operator is not certain whether boiling has occurred, he can refer to plant history as provided on water level recorders or ERFI Reference pressure leg boiling is indicated by level oscillations without corresponding oscillation If no boiling of the reference legs occurs and drywell temperature and pressure are restored to the safe region, then the instrument in question should continue to be use Instrument Narrow Range Level Instruments Conditions for Use C32-LI-R606A, B, C (N004A, B, The reference leg area drywell C) temperature is in the SAFE region of C32-LPR-R608 (N004A, B)
Indicating Range 150-210 Inches the Reactor Saturation Limit (Figure 14)
Cold Reference Leg AND Unit 1 oniv: The indicated level is in the SAFE region of Figure 1 Unit 2 oniv: The indicated level is in the SAFE region of Figure 15 Shutdown Range Level Instruments B21-LI-R605A, B (N027A, B) The reference leg area drywell Indicating Range 150-550 Inches temperature is in the SAFE region of Cold Reference Leg the Reactor Saturation Limit (Figure 14)
AHD Unit 1 oniv: The indicated level is in the SAFE region of Figure 1 Unit 2 oniv: The indicated level is in the SAFE region of Figure 16A.
lOEOP-01-UG l Rev. 24 l Page 84 of 131 l
,
. ,
j
, . ;
EOP-01-UG Attechm:nt 6 Page 3 of 17 ATTACHMENT 6 (Cont'd)
_
TABLE 1 (Cont'd)
Instrument Conditions for Use Wide Range Level Instruments Temperature on the Reactor B21-LI-R604A, B (N026A, B) Building 50' below 140*F C32-PR-R609 (N026B) (B21-XY-5948A A2-4, Indicating Range 0-210 Inches B21-XY-5948B A2-4, ERFIS Cold Reference Leg Computer Point B21TA102, QB !
B21TA103) l
-
N.!2 E the reference leg area drywell temperature is in the UNSAFE region of the Reactor Saturation Limit (Figure 14),
THEi the indicated level is greater than 20 inches
H the reference leg area drywell temperature is in the SAFE region of the Reactor
' Saturation Limit (Figure 14),
THEN the indicated level is greater than 10 inche l l
l l i I l !
l 1 l
I l
I OEOP-01-UG Rev. 24 l l Page 85 of 131 l
!
. .
t A EOP-01-UG Attechmsnt 6 Page 4 of 17 l
ATTACHMENT 6 (Cont * d)
-
TABLE 1 (Cont'd)
l Instrument Conditions for Use Fuel Zone Level Instruments The reference leg area drywell j B21-LI-R610 (NO36) temperature is in the SAFE B21-LR-RG15 (NO37) region of the Reactor j t
Indicating Range -150 - +150 Inches Saturation Limit (Figure 14), l
! Cold Reference Leg !
AND II the reference leg area
'
drywell temperature is less than 440*F, ELEN the indicated level is greater than -150 inches 2 IE the reference leg area I drywell temperature is greater than or equal to 440*F, 'DigH the indicated level is greater than -130 inche i j
AND l Reactor Recirculation Pumps are shutdow F2.T To determine reactor water level at TAF, see Unit 1 only: F192re 17 and !
Unit 2 onlv: Figure 17A l
.
To determine reactor water level at j the minimum steam cooling level (LL-4), see Unit 1 only: Figure 18 and Unit 2 only: Figure 18A To determine reactor water level at i the minimum zero injection level (LL-5), see Unit 1 only: Figure 19 and Unit 2 only: Fd Ture 19A l
!
!
- Continued on next pag l OEOP-01-UG l Rev. 24 l Page 86 of 131 l
.. _
. .
., =
EOP-01-UG Attachmsnt 6 t
Page 5 of 17 ATTACHMENT 6 (Cont'd) i
. TABLE 1 (Cont'd)
!
Instrument Conditions for Use l FOTE I
Each figure has two curves:
The upper curve for reference leg area drywell temperature greater !
than 200*F. The lower curve for i reference leg area drywell <
- temperature less than or equal to 200*F. If containment conditions are such that reference leg area temperatures could not be controlled and maintained less than the 200*F !
requirement, then the upper lines on !
the graph should be utilize N.q These level instruments are valid for indication with RHR LPCI flo I l
l l
l l
i
,
I
!
l
,
l i
lOEOP-01-UG l Rev. 24 l Page 87 of 131 l
-- - _. .
_
,
'
. ,
l
. . 1 EOP-01-UG
,
i Attnchm2nt 6 Page 6 of 17
!
ATTACHMENT 6 (Cont'd)
FIGURE 13 l
LEVEL INSTRUMENT REFERENCE LEG AREA DRYWELL TEMPERATURE CALCULATIONS 1 For all Level Instrumente EXCEPT B21-LI-R605 A, B, (N027 A, B); the reference leg area drywell temperature is the highest of the following !
point ,
l I Recorder l
l
,
CAC-TR-4426-1B Point A
,
CAC-TR-4426-1B Point B i
CAC-TR-4426-2B Point A j CAC-TR-4426-2B Point B QE Microprocessor CAC-TY-4426-1 Point 5801 i
'
CAC-TY-4426-1 Point 5803 CAC-TY-4426-2 Point 5802 i
CAC-TY-4426-2 Point 5804 ' For Level Instruments B21-LI-R605A, B (N027A, B), the reference leg area !
drywell temperature is the highest of the following points:
Recorder CAC-TR-4426-1A Point D CAC-TR-4426-1B Point B CAC-TR-4426-2A Point C l
CAC-TR-4426-2A Point D
!
CAC-TR-4426-2B Point A CAC-TR-4426-2B Point B QB Microprocessor l CAC-TY-4426-1 Point 5822 CAC-TY-4426-1 Point 5803 l CAC-TY-4426-2 Point 5823 l CAC-TY-4426-2 Point 5824 CAC-TY-4426-2 Point 5802 CAC-TY-4426-2 Point 5804 I
l
\
l OEOP-01-UG l Rev. 24 l Page 88 of 131 l l'
__ __._..-__._._..___.._. . . . - . . _ . _ . _ _._ _._._ _ ._ __ __ _ . _ _ _ _ . _ _ _ _ . .
i .
c
- .
!
,
,
Attachment 6 i
i Page 7 of 17 ATTACHMENT 6 (Cont'd)
l
.
$ -
FIGURE 14 l
REACTOR SATURATION LIMIT i
i i
i
!
i i
1 AB '
j.
<
!
gg .nsats
.
. -* .mauseama i
w
& .msmmassemes
.espassamasuman i
A _maassenes
- SAB ..seammammans g .smaapWassassamasasses l F iiiI iiiiiiiiiiI i i i i i i i i i i
!
I
{
-
,,
$$M$MM II I I I II I I I I I I I I I I I I I I I i i I lIIi11IIIIIiIIIIIIIIIIIII l
- 6 iiiiiiiiiiil l i i i i i i i i i i i i 1IiilIiIIIIIiiIiIIIIIIIII k i i i i i i i i i i i i i i i i i i i i i i i i iIIIIIIIIIIIIIIII4IIIiiiI i
! @ s= M
m j (
()
-
+H+H+H+H+W+H-H 1 II I i i 1 i i i I I I i i i i i i i I
- gg I lf i l l l i l l i l i l l f i l l l t
,
I i i i i i i i i l i i 1 i i i i 1i# s 1 t i i i i i i i i i i i i i i i i i i i i i l EW Er i
l l l f i l t lill l i l l i l i l l i l il l i i l Il l 1 1 l l I I l i l l et 1 1 1 1 1 l l l 1 e 1 1 I sif f l i l l j % ( i i i t i l l i l lll i l l f i lli WI I I I I I II I I I I I i i1 i i I I I I I I I I I I i I i I i i I e I i I I I i i I El i I I I I I I II i i l I I iiI i i I i i i I i I II i I i i i i iiiiiiiiI El 1 1* 11l Ii 17 1 1 I I I I I I I I I El I I I I I I 1 i I i II i i ! I i I I i if f Ii iiiI I I I i i I l I I i i I i i i i i I I I i I I I I I II II II riii - 1 II i eiiiiiieiiiiI i i i i iiiiiiI i i iiiiiii'
II i 1 1 I I I I I I I I I I i i I i I I I I i i I i i I i 1 i Iil I (
,
i
. . . . . .
3 Y Y Y Y REACTOR PRESSURE (PSIG)
,
!
!
l e
i
!
$
!
i
i l OEOP-01-UG l Rev. 24 l Page 89 of 131 l
l l
'
. _ _ _ . . . . _ , _ -
. .
)
- . .
E0P-01-UG Attachment 6 l Page 8 of 17 i
)
ATTACHMENT 6 (Cont'd)
!
-
FIGURE 15 j ,
UNIT 1 NARROW RA! ICE LEVEL i l
INSTRUMENT (N004A, B, C) CAUTION i 170
.
m Z
O 165 g SAFE l w '
b d
166 O A W
p 4.-.
<C sG 9:.
O f >-
155 . Z >G
"
-A #$ UNSAFE *
jYW:' %
300 350 400 450 REFERENCE LEG AREA DRYWELL TEMP ( F)
I
l
.
l l OEOP-01-UG l Rev. 24 l Page 90 of 131 l
. _ _ _ . _ . . . _ _ . . . . _ _. . . _ _ . . . . _ . . _ _ . - _ _ _ _ _ _ _ _ m. _ ...-..__ _ _ _ ._...-.-_._...m___...-__
d . .
J
<
.. ,
EOP-01.UC Att chm 3nt 6
.
Page 9 of 17-
,
I
,
! ATTACHMENT 6 (Cont'd)
!
~
! FIGURE 15A UNIT 2 NARROW RANGE LEVEL !
,
j INSTRUMENT'(N004A, 5, C) CAUTION
,
j i
!.
.I 1 +
+
! .
- ._ - ;
( , 175 ,
!
>
1 ^
!
e z
- .
-
- v 165 g SAFE ,
1 w
!
L 3J 16 0
- O 5 w s F W A
.
l <C i c
o 15 5 p
,
a f.: . .
....
... .. .
.3
-
!, Z ## '
~ .
i j f, UNSAFE $ l
! .:. -.e e: e: -
i i 15 0
!
! . 350 400 450 1
!
I REFERENCE LEG AREA DRYWELL TEMP ( F) !
!
i I
e I OEOP-01-UG l Rev. 24 [ Page 91 of 131 s'
i
-
-. . _ _ _ _ ,
, i
. -.
=
- .
jT E0P-01-UC
'
Attachment 6 Page 10 of 17
-
ATTACHMENT 6 (Cont'd)
a
-
- FIGURE 16 UNIT 1 SHUTDOWN RANCE LEVEL
- INSTRUMENT (N027A, 8) CAUTION '
i
!
! . . .
300
-- -
e + 1 ,,,,, ,, .- -
,
n z
- SAFE U -
a 250 w ,
'
3a o
w A e 1 l<- (T 200 O
w .s p./- -
a
.. .
, en :
z e '
'
~-
, ,, gf5
.eff" UNSAFE I 150
jgg 150 gg 250 3gg 350 4gg 450 i
REFERENCE LEG AREA DRYWELL TEMP ( F)
i
i i
!
I
OEOP-01-UG l- Rev. 24 Page 92 of 131 l
__ .__ - _ _ _
,
. .
.
,
Attachment 6 Page 11 of 17
ATTACHMENT 6 (Cont'd) .
-
,
FIGURE 16A l UNIT 2 SHUTDOWN RANCE LEVEL
INSTRUMENT (N027A, B) CAUTION i
30s i
1 .
'
n z SAFE
,
C i
a 250 w
d
_a a 1 W -
e de,
% 2gg /yy a
~ : _
.:. c
. .. ..
- ..- .
-
9 47
- -
z #
'
H
,
,,gf,.'
"
.-- .-
- # UNSAFE 150 100 200 300 400 REFERENCE LEG AREA DRYWELL TEMP ( F)
OEOP-01-UG l Rev. 24 Page 93 of 131
.- .=..---.- - _ . . - . . . . _ - . - - . - . - . . . - . . . - . - - . . . - . . . - . . - . . - . . . . . . . . . . ,
. . ,
,. .
.l
- Attachment 6 ,
Page 12 of 17
\
ATTACHMENT 6 (Cont'd)
,
,
f
-
,
. FIGURE 17 e
<
UNIT 1 REACTOR WATER LEVEL AT TAF
,-
,
i
g _ .. i
!
-18 ABOVE 4 '
m~ ' Tv *?,;
,v -20 y sa
-
,
i
,
1 *
?i: 19
"g !
o -30 -
,
Z g N.;...
I h :
,
.. .
_
U ! = .
w'.
. .
.
-
_4g .
,
_a n , ,
w -
~
.
, , i
. .. -1
.
_J-50
"' '
i '
.
9 ,
,
..
. .
.
'
" "-- o - -
..'
," =
w -68 BELOW ai s ==
"
~ ~70 TAF o
Z
~
-80-90
- 18 0-1,150
)
100 300 500 700 900 1.10 0 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)
i
!
!
NOTE WHEN REACTOR PRESSURE IS LESS THAN 60 PSIC, USE INDICATED LEVE i TAF IS -7.5 INCHE '
l l
l OEOP-01-UG l Rev. 24 Page 94 of 131 l
. . _ . . . _ _ . _ _ _ _ _ _ _ . . _ . . . . _ _ .
.. . _ . _ . _ ,
.
,
-
.
.
!
E0P-01-UC l Attachment 6 Page 13 of 17 l ATTACIDENT 6 (Cont'd)
~
FIGURE 17A- i UNIT 2 REACTOR WATER LEVEL AT TAF ,
, . . . .
l
-10 ABOVE y
1 TAF
--
Ry -20 ,
V
.1 %h -
..
1 2
$* ' ' '
"
lh
%
.. .. .
o -30 m;P O
z ' m w% ;
i hy '
'
... .
7= ,
^ ,
'
-40
'
'-
a m. ;p m .,
-,
w d -50
-
- ,,
"'" ; ,,
--
.,,
.
' " gu-
_a ,
..4 -
'
..q
. .
gun
.
,
o -
= '-
6 .ao =
w -60 BELOW
' '
6*4 "
.H
o ~70 TAF Q
z
-80-90 :3
- 10 0 10 0 300 500 700 900 1,10 0 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)
mz WHEN REACTOR PRESSURE IS LESS THAN 60 PSIC, USE INDICATED-LEVE TAF IS -7.5 INCHE OEOP-01-UG Rev. 24 l Page 95 of 131
._ ... _ _ _ . _ _ . . _ _ _ .__ _ _ . . _ . _ . _ . . . _ _ . .
d
. -
'
<
Attcchment 6 -
Page 14 of 17
. i ATTACHMENT 6 (Cont'd) )
l
-
FIGURE 18 UNIT 1 REACTOR WATER LEVEL AT LL-4 (MINIMUM STEAM COOLING LEVEL)
!
t 0 i l .
!
-
-10 ,
i'
-
G w-28 ABOVE T LL-4
'
O -30 Z
s
'
v i
'
-40 %%
,
w '
,
1 b; -50 N; g
'
'
- a - ,, -
j g ,
,g' . ...
-- - ,
, ,
w -6z t , ---
,
q '. ,
~ .
,
.~ -.
. ,.
.
,,
o-70 * - 2
,,
" -- ~
' " -- y BELOW a b '
'"U-80 LL-4 k# 24"
, ,
l-90-10 0-1,150 10 0 300 500 700 900 1,100 60 200 400 600 800 1,000 REACTOR PRESSURE (PSIG)
i HQIE WHEN REACTOR PRESSURE IS LESS THAN 60 PSIC, USE INDICATED LEVE j LL-4 IS -35 INCHE j i
OEOP-01-UG l Rev. 24 l Page 96 of 131 l l
, . l
.
). i
.
E0P-01-UG l i
Attachment 6 3
{ Page 15 of 17 1 l i
ATTACHMENT 6 (Cont'd)
l-i
-
~
FIGURE 18A i
UNIT 2 REACTOR WATER LEVEL AT LL-4 1'
!
(MINIMUN STEAN COOLING LEVEL)
i
- .
i 8
4 4 -s
.
j -
S -28 ABOVE 1 w r LL-4 i o -3s
.
z
~
v ,
<
'
_a-40 * ---
%
- w '
[...
.
I
,
d a -50 $mISm,,,- l
.-
---
.
- .
.m y"
.
m -, ,' [...
..
!
w -6B
'
'-
-
,, , ,, ,
). r :
$.n:
..
"- ...
-, 6
- -78
<
%. m -
.. ,,
fE"
,, gg4
.o
. -
,
,
z
-
BELOW h =
,%
! *
-80
'
,
N
,. LL-4 m ar E
_9g i :c:
'
--10 0 l
.
?
10 0 300 500 700 900' 1,10 0 1 60 200 400 600 800 1,000 l l
REACTOR PRESSURE (PSIG)
l l
lfGIE {
WHEN REACTOR PRESSURE IS'LESS THAN '
60 PSIC, USE INDICATED LEVE {
LL-4 IS -35 INCHE '
.
l
.l l
i OEOP-01-UG l Rev. 24 Page 97 of 131
[
_- . .. __ .. . - .. .. - - - . - - __-__ - . . - ...
. .
. E0P-01-UG Att cha2nt 6 j
Page 16 of 17
.
AITACHMENT 6 (Cont'd) ,
l
-
FIGURE 19 i
UNIT 1 REACTOR WATER LEVEL AT LL-5
- (MINIMUM ZERO INJECTION LEVEL)
B-1s
!
, &
w-28 ABOVE I LL-5 0 -38 z
O-40 w
d J-5s (
o -se %_--,
l w .
,s , ,_
%W, '
l
!
S -78 " --
am
,
-
a .%. . ..
Z
_gg bh ~ --
.J
% .,,,
,
/*g"=rr
~ r'
. . . .
. .
. .
BELOW
-
t ... <:, g..w...
...
g, go
-
,
,
,,
-98 LL-5 =-
saa, an EcunL 70 aser
- 10 0-1,150 16 0 300 568 700 900 1,10 0 66 200 408 606 806 1,000 REACTOR PRESSURE (PSIG)
mz WHEN REACTOR PRESSURE IS LESS THAN 60 PSIC, USE INDICATED LEVE LL-5 IS -50 INCHE OEOP-01-UG l Rev. 24 Page 98 of 131 l
- . _ . _
. _ . _ . . . ~ . _ _ _ _ . _ . . _ . . . _ . . . _ . . - _ ..._..____._.-._..______..._..__.._...__4
. .
. -
<
- Attochment 6
Page 17 of 17
,
ATTACHNENT 6 (Cont'd)
i ~
FIGURE 19A l; . UNIT 2 REACIDR WATER LEVEL AT LL-5 (NININUM ZERO INJECTION LEVEL) ;'
i
'. l
' B ,
-1B i'
r
i (
-
9w-28 ABOVE i
!. I LL-5 l
<
o -3g :
t z
s 1 v
..
g -40 i W
- - dJ -50 i
---.-
j 1 .
!'
-Q w -6B l h -
,, ,
'
!
N
'
, %..;.. -.
-
e !
>
j -Q F<--
- -70
..
~
p.,;;;
..%
. .. -, ,
'
-* = , ,'
O
,
p* ..
.,,,*
'
Q '-
-
-
- ~ ?=,,,,
! .Z c=
~ =,;*~
-
F= .,** uo i i H
'
_gg =sc- == :_
'
j
-
'
f BEL W % hg, o j
1.
'
-98 - LL-5 , ..: ..
,
$E 1-
-100 i
.
l 1 t
100 300 500. 700 900
'
.
1,10 0 ,000
[ REACTOR PRESSURE (PSIG) !
.
j i !
!
e HQH
. WHEN REACTOR PRESSURE IS LESS THAN
_60 PSIC, USE INDICATED LEVE j i LL-5 IS -50 INCHE '
4 . !
i l
e OEOP-01-UC l Rev. 24 l Page 99 of 131
!- ... ._ . - . _ . . . , , _ ,, - , _ _ _ _ _ . _ , . . - - _ _ - .
. . - . , ~ . . . ~ . - . . . . . = . . . - - . -
- - - = - - . . - - . . ~ - . . .
~....-...~..--..w_--..
. . .it f-i
.
Attachmsnt 10 Page 1 of 4.- )
! .I
)
, _ l
,
i J
EOP-01-UG Attachment 10
- Secondary Containment Temperature And Radiation Limits
-
,
!
,
l-
,
!
l l
,
L 1
! !
? l
!
i l
l.
l l
l' , , I (' I' l i-
t l OEOP-01-UG l Rev. 24 l Page 128 of 131 l l-l'
I I - . m _ . .- - - , . . . . , . - - , . , ,.-..&_ -- . - , . , . . . .
_ .- .- . . - - . . - . . - - - _ . - . _ , --
4.~ . .-
. - S
,
,
EOP-01-UG Attachm2nt 10 ;
'
Page 2 of 4 ATTACHMENT 10 SECONDARY CONTAINMENT TEMPERATURE AND RADIATION LIMITS i
-
FIGURE 20
,
SECONDARY CONTAINMENT AREA TEMPERATURE TABLE 1 I AREA TEMPERATURE LIMITS !
PLANT PLANT STEAM LEAK l INSTRUMENT MAX NORM MAX SAFE AREA AUTO
' LOCATION DETECTION NUMBER / OPERATING OPERATING GROUP DESCRIPTION CHANNEL / LOCATION WINDOW )
VALUE ('F) VALUE ISOL l (NOTE 1) (*F)
N CORE N CORE PANEL XU-3 VA-TI-1603 120 175 N/A SPRAY SPRAY ROOM S CORE S CORE PANEL XU-3 VA-TI 1604 120 175 N/A SPRAY SPRAY ROOM RWCU PUMP B21-XY-5949A G31 TE N016A l ROOM A B21 XY 59498 G31-TE-N0168 CH. Al-1 RWCU PUMP 821-XY-5949A G31-TE N016C
, ROOM B B21 XY 5949B 140 225 3
'
RWCU G31 TE-N0160 CH. A2-1 RWCU HX B21-XY-5949A G31-TE-N016E ROOM B21-XY-5949B G31-TE-N016F CH A3-1 N RHR 821-XY-5948A E11-TE-N009A N RHR EQUIP ROOM CH. AS-4
' 175 295 N/A PANEL XU 3 i VA-TI-1601-
, S RHR 821-XY-5948B E11 TE-N0098
EQUIP ROCN CH. A5 4 175 295
- S RHR N/A PANEL XU-3- VA-TI 1602
<
RCIC EQUIP B21-XY-5949A E51-TE N023A ROOM B21-XY-5949B ESI-TE-N0238 165 295 CH. Al-3 HPCI HPCI EQUIP B21-XY-5948A
-
E41-TE N030A ROOM B21-XY-59488 E41 TE-N030B 165 4 CH. Al-1 l
RCIC STM B21-XY-5949A E51-TE-N025A
- l TUNNEL B21-XY 59498 E51-TE-N0258 190 295 5
, TUNNEL CH. A5-1 1
4 20 FT NORTH B21-XY-5948A B21-TE-5761A CH. Al-4 i i
- 20 FT 140 20 FT SOUTH B21-XY-594BB B21-TE-57638 200 N/A j CH. Al-4 {
l 50 FT NW B21-XY-5948A B21-TE-5762A 50 FT CH. A2 4 140 200 N/A 50 FT SE B21 XY-5948B B21 TE 5764B CH. A2-4 REACTOR MULTIPLE ANNUNCIATOR WINDOW ALARM N/A 3,4, AND/0R 5 :
BLDG AREAS PANEL A-02 5-7 SETPOINT REACTOR MSIV ANNUNCIATOR WINDOW ALARM N/A 1 BLOG PIT PANEL A 06 6-7 SETPOINT i
NOTE 1 MAX NORM OPERATING VALUE IS THE ANNUNCIATOR / GROUP ISOLATION SETPOINT WHERE APPLICABLE l OEOP-01-UG l Rev- 24 l Page 129 of 131 l I
i . . (
EOP-01-UG Attechmtnt 10 Page 3 of 4 ATTACHMENT 10 (Cont'd) '
FIGURE 21 SECONDARY CONTAINMENT AREA DIFFERENTIAL TEMPERATURE TABLE 2 AREA DIFFERENTIAL TEMPERATURE LIMITS PLANT AREA PLANT STEAM LEAX MAX NORM AUTO LOCATION DETECTION OPERATING GROUP DESCRIPTION CHANNEL VALUE (*F) ISOL ,
(NOTE 1) '
'
RWCU PUMP B21-XY-5949A ROOM A B21-XY-5949B CH. A4-1 RWCU PUMP B21-XY-5949A RWCU ROOM B B21-XY-5949B 45 3 CH. A5-1 RWCU HX B21-XY-5949A ROOM B21-XY-5949B CH. A6-1 N RHR N RHR B21-XY-5948A 50 N/A EQUIP ROOM CH. A6-4 S RHR B21-XY-5948B 50 N/A EQUIP ROOM CH. A6-4 S RHR RCIC B21-XY-5949A EQUIP ROOM B21-XY-5949B 45 5 CH. A2 3 HPCI HPCI B21-XY-5948A EQUIP ROOM B21-XY-5948B 45 N/A CH. A3-1 RCIC STM B21-XY-5949A 45 5 STEAM TUNNEL B21-XY-5949B TUNNEL CH. A4-3 HPCI STM B21-XY-5948A TUNNEL B21-XY-5948B 45 4 CH. A6-1 REACTOR MULTIPLE ANNUNCIATOR ALARM 3, 4, AND/OR 5 BLDG AREAS A-02 6-7 SETPOINT NOTE 1:
MAX NORM OPERATING VALUE 15 THE ANNUNCIATOR / GROUP ISOLATION SETPOINT WHERE APPLICABLE lOEOP-01-UG l Rev. 24 l Page 130 of 131 l
- . . - - - - - , . ~ - - . _ _ .... ~. . . . - . . . . - - . - - . - . . . . ~ . = . - . . - . . - . - . . . -
, e
_, . 'I : )
(- .;
EOP-01-UG l J- <
Attachment 10 Page 4 of 4
'
\
j ATTACHMENT 10 (Cont'd)
,,
l
[ b
- -
. FIGURE 22 )
SECONDARY CONTAINMENT AREA RADIATION
] .:
\
=
TABLE 3 ;
AREA RADIATION LIMITS
j PLANT PLANT LOCATION ARM MAX NORM MAX SAFE -
AREA DESCRIPTION CHANNE . OPERATING OPERATING VALUE (mR/HR) VALUE (mR/HR)
$ N CORE N CORE SPRAY 15- 200 * 7000 e , SPRAY ROOM s_
ii
- ' S CORE S CORE SPRA * 7000 l SPRAY ROOM
.
N RNR N RH * 7000 ROOM
.
,
S RHR S RER 18 200 * 3000'
ROOM
[
l HPCI .HPCI ROOM- .N/A N/A * 3000
! {
k' N ACROSS 19
- FROM TIP ROOM i-RX DRYWELL 20 f BLDG ENTRANCE 80 * 2000 20 W DECON ROOM 22
,
ELEV
.
- - RAILROAD 23 i DOORS 7 RX BLDG SAMPLE 24
-
50 FT STATION
<'
ELEV 80 * 2000 RX BLDG j AIR LOCK 25 --
4 l
'-
RX N OF FUEL 27 80 * 7000 l l~ BLDG STORAGE POOL l 117 FT ELEV 'BETWEEN RX 28 1000 7000
& FUEL POOL
! CASK WASH 29 90 * 7000 !
t-
AREA l
j RX BLDG SPENT FUEL 30 90 * 3000 l l 80 FT ELEV COOLING SYSTEM !
i 1
,
!
l'
I *
' CONTACT E&RC TO DETERMINE IF MAX SAFE OPERATING VALUE IS EXCEEDED 4 a i I ' l OEOP-01-UG l Rev. 24 l Page 131 of 131 l j i
a
, ,%,, - , - . - . , , , - - , , , , - - , . . - , , . - -. , ,-, , , _ _ - .e,_,
... . \
.
R
-
CP&L CAROLINA POWER & LIGHT COMPANY BRUNSWICK NUCLEAR PLANT Reference Use
-
'
DATE COMPLETED FREQUENCY: )
UNIT _ % PWR GMWE A. Post Refuehng Outage SUPERVISOR j 8. As determined by Reactor Engineering REASON FOR TEST (check one or more):
Routine surveillance OWP#
WR/JO # l
Other (explain) ,
l
.
.
.
PLANT OPERATING MANUAL I
VOLUME X
'
PERIODIC TEST ;
RECEIVED BY BNP ENPE _1 v OCT 2 21996 UNIT %g 0 ;ocs 7 m.q ,,
NUCLEAR 0000,'EtNT CONiROL OPT-5 SRM SIGNAL-TO-NOISE RATIO AND IRM RANGE 6/7 OVERLAP DETERMINA TION l REVISION 11 l
EFFECTIVE DATE QA 10 -g-jG Sponsor b 9/27/9(e
{/ Date Approval lokW ManagW Engineering Date
.
OPT-5 Rev.11 Page 1 of 6
)
-
-__..._____
., ..
I .
l-l p . PURPOSE The purpose offthis procedure is to demonstrate that adequate SRM Signal-To-NoiseLRatio exist and to verify correct overlap of the IRM Ranges 6 and .0 REFERENCES f: Startup Test Instruction 6 - SRM Performance and Control Rod l: Sequence, Document No. 22A2229 AJ.
i ~Startup' Test Instruction 10-IRM Performance, Document No. 22A2229 A .3 0PT-14.3.1, In-Sequence' Critical Shutdown Margin Calculation i
i
'
.2 . 4' 0FH-11, Refueling
\. .
!
( OPT-50.1, LPRM and APRM Initial Sensitivities
]j ' PREREOUISITES The prerequisites'for this PT are listed with each'section.
l PRECAUTIONS'AND LIMITATIONS j . The SRM and IRM shorting links are removed to activate the SRM noncoincident scram until OPT-14'3.1, In-Sequence Critical 1 Shutdown
.
l
- Margin Calculation, has been satisfactorily completed. Therefore, control rods should be inserted in reverse order when the power level l of 105 cps.is reached to prevent a scram at 5 x 105 cps prior to
!- installing the shorting link .2 Caution should be used when SRM shorting links are removed or l j
,
installed so that inadvertent scrams are avoided.
! Independent verification is required in this P .0 SPECIAL TOOLS AND EOUIPMENT
!
I Electronic calculator l
l ACCEPTANCE CRITERIA
!
L Level 1:
6. SRMs have a count rate of 3 cps or more when fully inserted in the cor . The signal to noise ratio for the SRMs is two or more, i Level 2:
- 6. IRM readings in Range 7 are within 8 to 12 percent of the
{ readings;in Range 6.
l',
"
Jq i
l 0PT-50.2 Rev. 10C Page 2 of 6 )
! '
l
< __ _ - . _ . , _ . __ .~ _ _ _ . .
- ..
- * ANNUNCIATIONS EXPECTED None PROCEDURE Initials SRM Initial Data Check NOTE
- This section should be performed within one week prior to initial criticalit . Prerequisites
. I 8.1. FH-ll fuel loading is complete and all control '
rods are fully inserte '
8.1. SRMs are operable and fully inserted in the cor l Procedure 8. Record date and tim Date Time l
Counts Fully Counts Fully Signal to Inserted Withdrawn Noise Ratio l (A) (B) (A-B)/B
!
SRM C SRM D 6. Record the count rate for SRM A when fully inserted in the cor .
!
l 8. Verify that SRM A has a count rate of a 3 cps when '
l fully inserted.
!
l 8. Move SRM A to its fully withdrawn position and record
- the fully withdrawn count rat . Return SRM A to its fully inserted position. . Calculate and record the signal to noise ratio for SRM A as shown.
'
'
8. Verify that SRM A has a signal to noise ratio 2 . Repeat Steps 8.2.2 through 8.2.7 for SRM B.
.
-
OPT-50.2 Rev. 10C Page 3 of 6
. _
,
,
. PROCEDURE Initials 8. Repeat Steps 8.2.2 through 8.2.7 for SRM .2.10 Lepeat Steps 8.2.2 through 8.2.7 for SRM .2.11 Independently verify Signal-To-Noise Ratio /
calculations in the above tabl Ind.Ve .2.12 Independently verify SRMs have a count rate 2 3 cps /
when fully inserted into the cor Ind.Ve .2.13 Inform the Unit SRO, that the Signal-To-Noise Ratio portion of the OPT-50.2 is complete and Sat /Unsa ,,
(circle) '
. IRM Ranges 6 and 7 Continuity Check 8. Prerequisites 8.3. Reactor is critical in the Intermediate (IRM)
Rang . Procedure Steps 8.3. Record date and tim Date Time 8.3. For each IRM, establish a power level high in __
i IRM Range Record reading belo (IRM Range 6 uses the black 0-125 scale.)
8.3. For each IRM, switch the IRM to Range 7 and ,
record readin (IRM Range 7 uses the red 0-40 l scale.) Calculate percentage of Range 7 to '
Range 6 readings by the formula:
Percent - ((Range 7 / Range 6) * 100) !
l IRMs A B C D E F G H Range 6 Range 7
l Percent .
'
i 8.3. Verify that the IRM readings in Range 7 are within 8 to 12 percent of the readings in Range l
l
-
l OPT 50.2 Rev. 10C Page 4 of 6 i
I
.
. PROCEDURE Initials 8.3. Independently verify calculations in /
Step 8.3.2.3 and that readings in Range 7 are Ind.Ve within 8 to 12 percent of readings in Range If readings are outside the 8 to 12 percent band, a calibration of the IRM amplifiers should be performe .3. If adjustments were required per Step 8.3.2.5, record data following the final adjustment, for the applicable IRM(s) in the table belo .
IRM A B C D E F G H Range 6 Range 7 Percent 8.3. Independently verify calculations performed in /
Step 8.3.2.6 and that readings in Range 7 are Ind.Ve within 8 to 12 percent of readings in Range .4 Inform Unit SRO of PT status (SAT /UNSAT) . Inform the Neutron System Engineer if calibrations were required (Yes/No) and of the PT status (Sat /Unsat).
l
I
!
I l
l l
'
l
l l
i i
i
-
l OPT-50.2 Rev. 10C Page 5 of 6
_. - - . - - -_=_ ,- - _ . _ _ _ _ . . . . . ~ .
. ,
j d i
!
! Ceneral Comments / Recommendation i
l l
l Initials Name (Print)
PT performed by l
.
L l PT has been satisfactorily completed l
Responsible Engineering Manager l {
(Signature)
'
PT has NOT been satisfactorily completed Responsible Engineering Manager (Signature)
ll
!
Reviewed by !
(Signature)
i
!
I Corrective Action Required:
'
a. ( ) None
- b. ( ) WR/JO No.
,.
c. ( ) Other (explain)
l
~
,
OPT-50.2 Rev. 10C Page 6 of 6
. . . _ . . . - - . - - .
-
o
}
ATTACHMENT 2 ,L j- _
Page l' of 1 c
.
Data Sheet for Battery Ground Detection l BATTERY (1 A)
! DATE: TIME: PERFORMED BY:
l CURRENT: "P" mA "PN" mA "N" mA'
~
_
BATTERY BUS VOLTAGE: "1 A-1 " V DC "1 A-2" V DC
.
i
,
"1 A" RESISTANCE = y DC + V DC - 50 =
(P)mA + (N)mA BATTERY (1B)
DATE: TIME: PERFORMED BY:
CURRENT: "P" mA "PN" . mA "N" m BATTERY BUS VOLTAGE: " 18-1 " ..
"1B" RESISTANCE = V DC + V DC - 50 =
'
(P)mA + (N)mA EXAMPLE:
BATTERY (1B)
DATE: 6/5/91 TIME: 1302 PERFORMED BY: XXXXXXXXXXX CURRENT: "P" 2.15 mA "PN" 0.08 mA "N" 1.75 mA BATTERY BUS VOLTAGE: "1 B-1 " 135 V DC "1 B-2" 140 V DC
"1B" RESISTANCE = 135V DC + 140V DC - 50 . =
2.15 mA + 1.75 mA The "- 50" factor in the equation is to account for the presence of a 50 KO resistor in series with the milliamp moto Since the overall resistance (20.5 KO) is below the setpoint (25 KO), this would suggest there is a ground on the system which will first need to be verified. Once '
the ground has been verified, ground hunting activities should commenc According to Attachment 2, the ground is located on the N Bus. Al-115 should then be referenced for monitoring activities which will show that Action Level 1 should be entere P-51 Rev.30 Page 117 of 117 i
. .
c I
~
@hh
-
CAROLINA POV!ER & LIGHT COMPANY BRUNSWICK NUCLEAR PLANT Inronnat lon i
l Use PLANT OPERATING MANUAL i VOLUME l l BOOK 2 ADMINISTRATIVE INSTRUCTION ca
- -i c.-
- . . w UNIT $fC ,
O - . ' -R'
...
.
0Al-115 125/250 VDC SYSTEM GROUND CORRECTION GUIDELINES REVISION 2 REEIVED B1 EMP
, EFFECTIVE DATE JUN 0 81995 (p-8- 95
^
.
NUCLEAR 000UllefiCONIFOL Sponsor 91 bmmw- f-9- T dll ,,
Date Approval fY! k< 1 15
Plant General Manager (U1/U2)/ Date'
Manager - Technical Support 0Al-115 Re Page 1 of 8
.
1
' 4' ,1 i
LIST OF EFFECTIVE PACES e OAl-115
,
, Pane (s) Revision *
1-2 2 i .
3 0 l l
4-8 2
- >
i
!
i i
l i
J
.
l t
!
.
t
.
. .
l
i l
,
J
l l l
,
I i g i
i i:
!
l
1 i !
I
- -
0-AI-115 Rev, 2- Page 2 of 8
.. . _-. .. . . . .
. .
.
1.0 DC GROUND POLICY STATDiENT It is the policy of Carolina Power & Light Company to engineer, construct, and operate nuclear power plants without jeopardy to the health and safety of the public and of its employee The function of the this procedure is to establish a set of criteria for actions which should be taken based upon the magnitude of grounds on the 125/250 VDC distribution system. Completion of the various actions will provide a reasonable assurance that the 125/250 VDC distribution system will centinue to satisfy the CP&L Polic It shall be the responsibility of all organitations at the plant and of all corporate support groups to provide support needed to ensure that the
DC Cround program is successful, 2.0 PURPOSE AND SCOPE
The 125/250 VDC system at BNP is an ungrounded syste The nystem was designed as an ungrounded system so that the system could sustain one solid ground fault and still be able to meet all requirements for continued safe operatio On the ungrounded system when the first ground fault occurs the system is de5raded but it is not degraded below an acceptable grounded leve The de ground detectors provide a means of monitoring the de distribution system for significant grounds. If one ground occurs the ground detectors will alert plant cperations so that the ground can be corrected before a second more significant ground occur The purpose of this document is to provide a set of guidelines for plant i operations to use to determine the urgency of action and the actions to be tsken to correct a known de ground based upon the severity of the i groun The initial values selected for this document should provide a l reasonable assurance that the operation of the de distribution system l will not be adversely affecte I The effectiveness of this document will depend on the full support of all levels of planc and corporate managemen This includes the provision of adequate resource 3 and a dedication to correcting grounds expeditiously as required by th. action statement Changes to these guidelines will be made as the detailed evaluation of the de distribution system continues and as refinements and improvements in the de ground detection system dictat l 0-AI-115 Rev. O Page 3 of 8
- - . . - .__ - - . .. -- . - . . - -- . ..
. .
- ,
!
!
3.0 BASIS FOR ACTION LEVELS Electrical Evaluation BNP-E-6.116 was issued to provide a basis for the , ;
setpoint of the battery ground detectors. The value established by this 1
!
evaluation is 25K ohms. The ground detectors' operate with a 15% band, which corresponds to 21.3K to 28.8K ohm *
l
- 4 Electrical devices at BNP were researched and the most sensitive device j
'
of concern'was determined to be GE HFA relays. These relays have a ; ;
nominal dropout current of 3.75 m This value was utilized to derive an I appropriate setpoint. Therefore, at ground levels just below 25K ohms, I situations (involving two very strategically located grounds) could l !
develop that might hold-in a normally energized rela : :
,
! J At ground levels below 15K ohms, sufficient currents are allowed to l i develop which under the right conditions (involving two strategically i !
located grounds) could result in picking up a de-energized relay. Since ! i most relays in the system are normally de-energized relays, this level is j ]
considered the most urgent to correc Above the ground detector setpoint of 25K ohms, there are no required actions to be taken. Action Level 1 is between 25K and 15K ohms and l 1 plant procedures should be entered to locate and correct the ground
' '
'
conditio Grounds below 15K ohms shall be considered the most severe condition (Action Level 2) and shall require the greatest degree of f effort towards' resolutio This procedure provides guidance to plant operations as to prudent 4 l actions to be taken when DC System grounds are encountered. These guidelines do not necessarily ensure that a system operating problem will not occur if the guidelines are followed. However, due to the fact that l!
- 'very specific, isolated, and independent events must occur with the l DC grounds on the system to create the potential for operational i l problems, the guidelines provide a relative degree of confidence that !
j l corrective actions can be completed before these system operating I problems occur.
'
l RESPONSIBILITIES
!
l Plant General Manager is responsible for establishing and l maintaining the de ground program and shall assign specific responsibilities for the implementation of the progra .2 The Manager of Operations is responsible for ensuring that j appropriate corrective actions are followed as outlined in this l
-
guidelin .3 The Manager of Maintenance is responsible for assuring prompt and
- effective repair of plant eqdpment with de ground .4 The Manager of Technical Support is responsible for any support functions towards monitoring and/or repair of de grounds.
.
,
0-AI-115 Rev. 2 Page 4 of 8
-
. , ,
.
!
i l
l 5.0 DEFINITIONS i l
5 .1 - Action Levels i
!
Operation at or below the action-level 1 or 2 values presented in *
this document may not. result in immediate misoperation of de electrical equipment, however, the de electrical equipment will be operaging in a degraded conditio '
(-
, 5. . Action Level 0 (Ground Resistance 2 25kD) !
l i
I
'
This is the achievable value which should be maintained by ;
applying good operating and maintenance practices. .Every l
..
effort should be made to operate with the de grounds at or {
above this valu I
5. Action Level 1 (Ground Resistance 15kn to 25ka) !
This level represents the level of operation of the de '
system below the ground detector alarm setpoint of 25k ,
Operation with one ground in this range will not result in i the inadvertent pickup of a deenergized relay, but could result in the inadvertent hold in of an energized relay if '
a second hard ground were to occu Prompt actions should be taken to restore the ground resistance to Action Level 0 values to minimize the possibility of continued i degradation of the de distribution syste I 5. Action Level 2 (Ground Resistance 5 15kO)
This level represents the level of operation that could result in the inadvertent pickup of a deenergized relay if j a second ground $25KO occur Immediate actions shall be <
taken to restore the ground resistance to a minimum of j Action Level 1 values prior to encountering a second i ground s25KO. Continuous investigation of possible ;
grounds should be' conducted until the ground is located i and correcte I l
i
.
i l
l 0-AI-115 Rev. 2 Page 5 of S
!
. . .. - =
l
>
. .
'
)1
,
i
. i l 6.0 ACTION REQUIREMENTS
!
l 6.1 Action Level 0 (h25kD)
!
l l At Action Level O the de grounds are above the de ground detector ,
l alarm point of 25kD. No actions are required providing the dc '
ground detector alarms are operabl If the' de ground detector alarms are determined to be inoperable, then the ground current shall be measured once per shift per OP-51, 1- j to verify that the grounds are above the alarm point.
, l 6. l If the DC ground annunciator alarms or ground current !
, measurements demonstrate a ground <25kD, perform the following: ;
I Note ground in Shif t Supervisor's log along with any
.
6.1. l {
possibly related activity / evolutio ,
i l 6.1. If ground measurement determines that ground is actually below 25kD, Operations should perform ground hunting per OP-51 and appropriate action leve .2 Action Level 1 (<25ka co >15kD)
At Action Level 1 the de grounds are above the point where ;
deenergized relays may pickup if a hard ground occurs on the '
opposite bu This action level shall be entered if ground current measurements demonstrate that a ground between 25kD and 15kD actually exist Action steps for Action Level 1 are as follows: ;
i 6. ,
The ground current shall be measured once per shift
-
i l per OP-51, to ensure that the ground resistance has not i i degraded to the Action Level 2 poin ,
i l
6. Efforts shall be undertaken to isolate all circuits per !
OP-51 which can be isolated without affecting plant l operation in an effort to locate the ground. The ground 2 hunting effort will be cot. ducted on an eight hour per day ,
five days per week basi .!,
l-l l
l l I e 0-AI-ll5 Rev.. 2 Page 6 of 8
I.
'
.
!
.
!
L .
[ 6.0 ACTION REQUIREMENTS (Continued) ,
! Action Level 1 (<25ka to >15kO) (Continued)
6. If the ground cannot be located within 14 days then the actions Steps 6.3.2 through 6.3.4 of Action Level 2 shall be implemente . 2.4 ' Circuits identified with grounds should be corrected expeditiously, and if possible, placed under clearance while waiting correctio .3 Action Level 2 (515kD)
'
At Action Level 2 the de grounds are at the point where deenergized relays may pickup or energized relays may dropout if another
$25kn ground occurs on the opposite bus. This action level shall be entered if measurement of the ground current demonstrates that the ground is less than or equal to 15kD. Action steps for Action Level 2 are as follows:
6. The ground current shall be measured twice per shift to ensure that the ground resistance has not degraded or to determine if any worse grounds develo '
6. Efforts shall be immediately undertaken to isolate all dc i circuits per OP-51 which have not been isolate The ground hunting effort will be conducted on an 24 heur per day seven day per week basi . If af ter seven days the ground has not been located and no more circuits can be isolated per OP-51 due to the existing plant operating conditions, PNSC is to approve subsequent ground hunting action . Circuits identified with grounds should be corrected expeditiously, and if possible .placed under clearance while waiting correctio I 7.0 REFERENCES FP-84882, Battery Ground Detector 7 -. 2 IEN 88-86, Operation with Multiple Grounds in Direct Current Distribution Systems, October 21, 1988 IEN 88 86, Supplement 1, Operation with Multiple Grounds in Direct Current Distribution Systems, March 31, 1989 0-AI-115 Rev. 2 '
Page 7 of 8
- . __ . . . -
.. __ _ _ _ . . _ . _ _ _ _ _ . _ _ _ _ .
. .
, .
I-
I i :
!-
- 7.-0 REFERENCES (Continued)
. l SDCD 51, System Design Criteria Document for the DC Electrical
- System, April 9,1988 L L l Commonwealth Edison Company, Ground Task Force Final Report, '
May 25, 1989 t- ' Calcul'ation BNP-E-6.116, Ground Detection Setpoint Basis
,
for 125/250V DC System )
l ! j
}
l
.
!
!
l !
"
l
l
! 1 l '
l
- i l
l
'
),
i i
l l
l l
,
i
!
!
!
i I
I i
i !
l !
r 1 l
l
0-AI-115 Rev. 2 Page 8 of 8
<
N,- , -
, ,. -
l
~.
. ,
'
l
.
ATTACHMENT L (Cont'd)
i Table 1 l Tagout Devices l
DISCUSSION Table 1 lists various devices that may be used as isolation boundaries for clearances. The suggested tagout method and known restrictions on use for clearances are given. If it is desired to use a particular ,
device and actuator not shown, or to use a restricted device shown, a ]
j special evaluation must be performed by a person knowledgeable of the device, and the evaluation logged or otherwise documente Clearance Boundary Isolation Components TAGOUT DEVICE: GATE VALVE P%NUAL SOLENOID MOTOR CYLINDER
- @ B X C)<J W C)<
l TAGOUT METHOD Tag on manual Tag on motor Tag on air supply . l operator breaker ("BKR isolation Locked Off or (" Closed"). !
Removed').
Tag on gag device Tag on u nusi (" Gagged Closed-).
operato Tag on actuator ("Do not remove actuator").
RESTRICTIONS Consider leakage Do not use as If valve is marked history and use two tagout bouncary if fail closed on valves in series if not listed on print. no gag is leakage drawing as a fail- needed unless unacceptabl closed valv system pressure is expected to exceed valve spring
.
, pressure.
l
!
I
!
OAI-58 Rev. 50 Page 79 of 97
!
. *
O ATTACHMENT L (Cont'd)
Table 1 Tagout Devices Clearance Boundary Isolation Components TAGOUT DEVICE: GATE VALVE SPRING-OPPOSED PRESSURE BAUWCED PNElf% TIC .
HYDPMLIC DIAPHRAGM DIAP)fMGM
'
S m-
>< >< ><
TAGOUT METHOD Tag on air supply isolation (" Closed").
Tag on gag device (" Gagged Closed").
RESRICTIONS Must use gag devic DO NOT USE AS Use as clearance Prone to leakag CLEARANCE Consider leakage boundary is BOUNDAR dependent on history & use two valve valve model, isolation if necessary, spring closur and hydraulic i
!
supply. Evaluate case-by-case for us .)
l I
-.
!
l
l I
!
!
i OAI-58 Rev. 50 Page 80 of 97 l
. - _ .
. _ _ . . . - . . . . _ _ . . _ _ . _ . - . - - . _ . _ ~__,_~____.._-m.._~.
.- i
! !
,1 :
!
P i
-'
l
!
,
ATTACHMENT L '(Cont 'd) I i
. Table 1 -)
i Tagout Devices '
i TAGOUT DEVICE: GLOBE VALVE !
I
~l MANUAL i
SOLEN 0ID i MOTOR CYLINDER i
)
-
00 1
@ g- ;
!- -
-
'
l
!
I
-
,
TAGOUT E TH00 Tag on manual Tag on solenoid l Tag on motor Tag on atr supply operato power supply breaker ("BKR isolation (" Closed"),
breake Locked Off or Removed"). Tag on gag device NOTE: Other (" Gagged Closed").
.
components may Tag on manua '
l receive feed from operato Gaggtng device l' same breake .
neede ,
l-
- RESTRICTIONS.
'
Use as tagout Do not use as
.
boundary only Wien Do not use as l isolation isolation boundary if marked on crawings boundary if valve is normally as valve is used as a flow normally used as control valve, a flow control'
valv Gagging device neede .l
.
l !
l' ,
!
l
-
,
! I i
b
!'
t l- I OAI-58 Rev. 50 Page 81 of 97 i
l'
. . - - . .. _ _ .
~ , . . . , , j
. .
l
!
- I l
i l
ATTACID4ENT L (Cont'd)
l Table 1 Tagout Devices TAGOUT DEVICE: GLOBE VALVE SPRING-OPPOSED PNEUMATIC DIAPHRAGM I
l O
s s -
'
/ 's, TAGOUT METHOD Tag on air supply '
i isolation (" Closed").
i Tag on gag device (" Gagged Closed").
Gagging device neede RESTR.lCTIONS Do not use as isolation boundary if valve is normally used as a flow control valv Sagging device neede .
.
i l
l i
!
,
OAI-58 Rev. 50 Page 82 of 97
, _ _ .
. . ,
I-4 I
l ATTACID4EWT L (Cont'd)
Table 1 f
'
Tagout Devices TAGOUT DEVICE: BUTTERFLY VALVE
-_
MANUAL MOTOR CYLINDER CYLINDER ($1NGLE-ACTING) (DOUBLE ACTING) I
.
@ ET E3:
-
y M l-d M i TAGOUT Tag manual Tag motor breaker METHOD operato Tag air supply Tag air supply (*BKR Locked Off or isolation valve Removed"). isolation valve (" Closed"). (* Closed").
Tag manual Tag gag device operato Tag gag device (" Gagged Closed"). (" Gagged Closed").
RESTRICTIONS Valves are prone to Valves are prone to Valves are prone to leak by. Co.isider leak by. Consider Valves are prone leak by. Consider to leak by, leakage history, leakage history, leakage histor Use redundant Consider leakage i
Use redundant Use redundant history. Use valves if leakage valves if leakage valves if leatage i redundant valves unacceptable. Some unacceptable. Some unacceptable. Some valves can leak by if leakage valves can leak by valves can leak by unaccept-due to being closed
,
I due to being closed due to being closed past the sea past the seat, able. Some past the sea valves can leak Gagging device by due to being needed if closed past the positioning seat. Gagging oppos1te of fail device needed, ocsition.
!
,
l l
l i
0AI-58 Rev. 50 Page 83 of 97
. . . _ - - . . .. ._ . - . ., ..
. =
'
t
.
Y
!
ATTACHMENT L-(Cont'd)
Table 1 Tagout Devices TAGOUT DEVICE: BALL VALVE
,
PANUAL SOLEN 0ID CYLINDER CYLINDER (SINGLE-ACTING)
(DOUBLE ACTING)
- *
Ef E3:
..
XK D@G D@G N
TAGOUT METHOD Tag manual operato Tag on air supply Tag on. air '
Isolation valse supply isolation '
(" Closed"). valve Tag on gag device (" Closed").
(' Gagged Cicsed'), Tag on gag device (* Gagged Closed"). !
1 RESTRICTIONS )
00 NOT USE AS If valve is marked t Gagging device TAGOUT BOUNDAR fail-closed on print, neede no gag 15 neede I
1 i
!
!
l
, .
l l l
,
i l
OAI-58 Rev. 50
!
Page 84 of 97 l
i
!
i - -
..
. .
'
,
{
P t
l-ATTACHMENT L (Cont'd)
! Table 1 Tagout Devices TAGOUT DEVICE: PLUG VALVE l
l l MANUAL CYLINDER CYLINDER (SINGLE ACTING) (DOUBLE ACTING) !
<^> O
'
K>i i
l l
TAGOUT METHOD Tag manual Tag on air supply operato Tag on air supply Isolation valve isolation valve (" Closed'). j (" Closed").
!
Tag on gdg devic Tag on gag
!
devic l
I
!
RESTRICTIONS Prone to leak b Prone to leak b Prone to leak b Consider valve Consider valve history and apply Consider valve history and apply history and apply redundant redundant isolation redundant isolation if if leakage not isolation if leakage not acceptabl leakage not acceptable, acceptabl No gagging device needed in fail Position valve in positio desired position prior to isolatino air.
!
i I
L i
!
OAI-58 Rev. 50 Page 85 of 97
'
,
.
ATTACHMENT L (Cont'd)
Table 1 >
Tagout Devices TAGOUT DEVICE: ANGLE VALVE MANUAL MOTOR i
@ i
- &
!
_
TAGOUT METHOD Tag on manual Tag on motor operator, breaker ("BKR Locked Off or ,
Removed"). I
'
Tag on manual operato )
RESTRICTIONS
...
DAI-58 Rev. 50 Page 86 of 97
.
I
l
!
ATTACHMENT L (Cont'd)
)
Table 1 !
!
Tagout Devices l
TAGOUT DEVICE: DIAPHRAGM (PINCH) VALVE
!
W UAL PRESSURE BALANCED PRESSURE BALANCED DIAPHRAGM (WITH I DIAPHRAGM (N0 '
HANDWHEEL) HANDWHEEL)
.
w -- M
~
TAGOUT METHOD Tag on handwhee Tag on Hancwhee RESTRICTIONS Record throttle 00 NOT USE AS position of valve TAGOUT BOUNDAR and restore to this position on return to ,
servic *
0AI-58 Rev. 50 Page 87 of 97
... . . . . . - - _ . - . . _ . -
a s
.
-
EOP-01-UG Attechment 5 '
,
Page 15 of 31 ATTACHMENT 5 (Cont'd)
-
FIGURE 1 DRYWELL SPRAY INITIATION LIMIT 450 - - .: .\ y -c ,- ,
, ,- m.a - , , .
,,,,
?h5 '
%d ?:,W:i;/ << fyk, is t , .
'i s
% Vi , ,, ? - , i,
'
400 "
,' *
NSAFh ', ' '
,
,
,,
w / -
y > < >, ,
,
-
w g
~,
- ss;; .v y- , :::q:> . ,
Lcne
U *
W 350 ,,
4 M
., g y
-
4 '
3 300
&
Z '
T 2' SAFE 250 W
o
" h w
,
g200
_J
._J w ,.
'
[ 15 0
- x ,
.
a
4 10 0
50 ,
- -
5 15 . 25 35 45 55 65 75 0 10 20 30 40 50 60 70 DRYWELL PRESSURE (PSIG)
.
E ,
DRYWELL AVERAGE AIR TEMPERATURE IS DETERMINED USING ATTACHMENT 4 OF THE " USER'S GUIDE" i
,
lOEOP-01-UG l Rev. 24 l Page 66 of 131 l
_. . . . - . . - - . ._ - . - . -.~ . . . . . . . . - - . . . . . . - . - . - - . . _ . . . . . . . - . - - - . - . . . _ - _ . - . . .
t 8
.
-
i
,
EOP-01-UG Attachmsnt 5
Page 16 of 31
,
i ATIACHMENT 5 (Cont'd)
i
.
-.
FIGURE 2 UNIT 1 HEAT CAPACITY LEVEL LIMIT
!
-
- meron mzeune summa m umus
meew = uur m * "m " a"w n aus atm(v)
N *==
um m ne me me -
.
======. e -
,
,
em m em ma wn -
t
=====. e
.
.
.
.m m m me e -
.
am mem me an -
! -
,
= mam me .w -
-
m. m sw -
i- .
= m am me m -
~
~
.
im m us mo my -
.
arm us m so, _
[ uns m= = me Ar,em
.
err
-
.
t a-v
.
_J-1
'
W d -2 i- - J -3 t
J FE l O -4 O
!, a -5 i
z -6 O
i G -7 l- m 1 W -8 UNSAFE (Y.
! 1
' _g D
i m-s B 5 16 15 26 25 36 35 48 45 58 55 60 ATHC ( F)
um SUPPRESSION POOL WATER TEMPERATURE IS DETERMINED BY:
POINT A ON CAC-TR-4426-1A QR POINT A ON CAC-TR-4426-2A QB COMPUTER POINT G050 QB COMPUTER POINT G051 QB CAC-TY-4426-1 QB CAC-TY-4426-2 lOEOP-01-UG l Rev. 24 l Page 67 of 131 l
. _ . - .. . . - . . . . _ . . . . . . - . . . - . - . - . . . . - . . - . _ . - . . - - - - . - - - . - . . .
- . .
. .
,
!
'
,
Attachmsnt 5 !
Page 17 of 31 f
.
,
ATTACHMENT 5 (Cont' d)
!
) -
FIGURE 2A UNIT 2 HEAT CAPACITY LEVEL LIMIT
!
,
menn meam === Pom-n eewuu taar unas = coa. ===== m n=u=m=a Atec v
! == = == ess - -
-
.= = s- n .
i am = , , -
-
! = = = = , = . m - -
- .m = = , m = -
.
t an = ,mo err -
-
,
-
.
= = m ese e . .
i an = = ,mo y - .
t
= = = a s pies , - -
= = e .ser -
-
= = m := , -
.
um == = res Ar,.y arr
- ^
,
& s __ -- i
!
v i
_1 J
..
.
i w i- 3J -2 -3 l
l
'
SAFE
_a
' o -4 o
- a -5
4 Z -6 ,
o l
!
M -7 m '
- w -s UNSAFE
!
m a _g J a 1 m -1
<
- ,
l s 5 18 15 25 25 30 35 46 45 58 55 60 i
1 ATHC( f) i
!
l j
j i
SUPPRESSION POOL WATER TEMPERATURE IS DETERMINED BY:
q POINT A ON CAC-TR-4426-1A-j QB POINT A ON CAC-TR-4426-2A l
.
QR COMPUTER POINT G050
, QE COMPUTER POINT G051
'
l QB CAC-TY-4426-1 j QR CAC-TY-4426-2 s
a q l OEOP-01-UG l Rev. 24 l Page 68 of 131 l i
$
.
. .
.-
.
.-.-. .
. _-, - -..
. . ,
,
-
EOP-01-UG j Attechm2nt 5 '
Page 18 of 31 -
4 ATIACHMENT 5 (Cont'd) I i i -
-
FIGURE 3 UNIT 1 HEAT CAPACITY TEMPERATURE LIMIT *
n
.,i _ g. 228 .
. = 232 .
v : r xxx i
g 210 x xx:=
%
i D 200 UNSAFE
- .
..
g 190- -
- -
o_
2 180 W .
-
-
k
- 17 0 g
W - , ,
l 160 1
.- SAFE a 150
- O O 14g CL l
- I i Z 130
<
s O
w M 12 0 l
.
(n !
! W i
% 110 CL
!
i
'
Q- i
- .
D to 100 '
l:
.
l J l-1,15 0
,
10 0 300 500 700 900 1,10 0
- 0 200 400 600 800 1,000 !
[ REACTOR PRESSURE (PSIG)
i
.
i
E
.
SUPPRESSION POOL WATER TEMPERATURE IS DETERMINED BY:
'
POINT A ON CAC-TR-4426-1A
. 9E POINT A ON CAC-TR-4426-2A l j QE COMPUTER POINT G050
,
QE COMPUTER POINT G051 QB CAC-TY-4426-1 QB CAC-TY-4426-2
'
4 * VALID FOR SUPPRESSION POOL LEVELS ABOVE -31 INCHES
OEOP-01-UG
l Rev. 24 l Page 69 of 131 l
,
. .
.-.. . . . . - . . - . . . . . . . . . . ~ . . . ~ . . .
-.
.
,. q E0P-01-UC Attechnent'5
,
l -i'
,
Page 19 of 31
- ATTACHMENT 5 (Cont'd)
,
'
k
~.
FIGURE 3A-UNIT 2 NEAT CAPACITY TEMPERATURE LIMIT * !
. - -
i
, C o
220 m
'
- -
.
210
,
W .
'
'. '
$ 200 -
.
UNSAFE
!
-
,
y 190 -
.
1 - a, ,
,
!
2 18 0 w
'
,
l
!. ,
H- '
- .g 170 .
I w .
'
! l
! 160 -
-
SAFE
i a 150 ! !
O
,
'
} O 14g '
.
!
Z
130 l
.
m i
}' W 120 M
'
,
W i, i
% 110
. _ .
D 100 M
10 0 300 500 700 900 1,10 0 0 200 400 '600 800 1,000
'
-i REACTOR PRESSURE (PSIG)
l l
l HQIE SUPPRESSION POOL WATER TEMPERATURE IS DETERMINED BY: !
POINT A ON CAC-TR-4426-1A-98 POINT A ON CAC-TR-4426-2A '
gg COMPUTER POINT C050 gg COMPUTER POINT C051 98 CAC-TY-4426-1 '
98 CAC-TY-4426-2 i
- VALID FOR SUPPRESSION POOL LEVELS ABOVE -31 INCHES i
I OEOP-01-UC l Rev. 24 Page 70 of 131 l
- . . - . - . - -
.. _ . _ _ _ ._ _ _ . . . _ _ . . . _ _ - . .- _ _ __. . ,
EOP-01.UG Atte.chment 5 Page 20 of 31 ATTACHMENT 5 (Cont'd)
-
1 FICURE 4
UNIT 1 MAXIMUM CORE UNCOVERY TIME LIMIT '
a I '
g i 15,BBB, ,
'
1 (n u
'
r s3 : - , ,
. - i
, Z s / ! ,
. H 2 ' / l r
/ l
> I 1,BBB / j i -
! Z f . .
- 3 I / i
- O , / j
- O / +
i H : '
i
'
D
'
/
i I /
'
i V) '
10 0{
'
!
x .
/ :
, o f--
- r
! -
) i
! O '
- <
l
.
d m
- I
, x 16l .
La .
'
t
< :
'
- s d
l
>
}
m
'
H I '
s i
B 5 16 15 2B '
i
'
l I
- MAXIMUM CORE UNCOVERY TIME - MINUTES i
.
l i
'
.
!
I
- OEOP-01-UG l Rev. 24 Page 71 of 131 l
. _ . _ . _ . _ _ _ _ _ _ _ . . . _ _ _ . _ _. - - _ . _ - . . _ _ _ _ _ _ . . ___m-m . __ _ __ .. _._ _ - . . . _ _
.
i
,
"
l E0P-01 UG Att chment 5 Page 21 of 31 ATTACHMENT 5 (Cont'd) i
. !
-
FIGURE 4A UNIT 2 MAXIMUM CORE UNCOVERY TIME LIMIT !
O I
,
i,
,
(n 12.585l
.
,
a
!
l
W o
.
'
/
-
! - Z * # ;
- * l l / l l 1,280 l
'
Z /
3 $ /
O , /
O <
H 3 J l
D / l I / . i (n 18 8 i .
& *
l O l !
O * /
l'
!
[
E E
w Il .
tr
<
, .
LLI a
5 1 .
B 5 10 15 20 .
i !
MAXIMUM CORE UNCOVERY TIME - MINUTES i
)
OEOP-01-UG l Rev. 24 Page 72 of 131 l
__ . . . _ . . , . _ _ _ . - _ _ . - , _ . _ _ . _ _ _ . _ _ . . . . _ . . _ _ _ . . _ . _ _ _ _ _ - - . _ . _ _ _ . _ . _ . _ _ . _ _ ,
j 4 *
.
- .
.
- +
EOP-01-UC Attechnent 5
..
i
. Page 22 of 31 t
!
5-AITACHMENT 5 (Cont'd)
!
FIGURE 5
-
CORE SPRAY NPSH LIMIT l .
I
!- )
'
i I
C298
- l I V
! w 285 1
!. E
.
- :-
Q'275 E
i i
@ 2W l 2 258
! N !
g 248 235 .
3E
!
a 225 l
~ 218
2 258 O
'
- 3 3 M 19 8 m I i
W i'
z '
o. 18 5 Q
-
3 175 M
.
585 1,*.25 2,58W 3,585 4,588 5,585 6,588 '
S 1.895 2.988 3,588 4.895 5,885 6,868 7.886 CORE SPRAY FLOW (GPM)
NOTE i
SUBTRACT 0.5 PSIC F1t0M INDICATED SUPPRESSION CHAMBER PRESSURE FOR EACH :
FOOT OF WATER LEVEL BEIhW A SUPPRESSION POOL WATER LEVEL OF -31 INCHES {
(-2.6 FEET). I OEOP-01-UG Rev. 24 i l l Page 73 of 131 l J
,
l
_
.
- - , - - - - - ~ ~ ~ ~ "
-_ _ -. . - _ _ _ _ _ . _
- -. - - - - . . -. - - . - . .. .- - .. . . --
. .
E0P-01.UC l Att:chment 5 i Page 23 of 31 ATTACHMENT 5 (Cont'd)
- -
,
- '
C o
290
- W280
,-
,
g ___
~
D 270 ** NI1 *
'
'
<C
-
yCr 260 I I i
i I i '
,
2 250
.
% ---
_
~
g 240 ,
- 2H PSIG j ;
i W ~
Q 230 ___
l
3 '-
,.
i a 220 '
m
!
o '
- 10 3SIG t O '
'w
,
a. 210 ~
!
m .
1 ' .
'
Z 2g0 '
'
P --
i
.
- '5 PSIG "" ' !
.-
,_ 1 :
M 19 0 (n N '
m y ,.
.
-
%
a. 180 ,
m
'
0- '
- O PSIG -
,
D 170 m
!'
\ \
.i 0 5,000 10.000 15,000 20,000 J
RHR PUMP FLOW (GPM)
,!
E SUBTRACT 0.5 PSIC FROM INDICATED SUPPRESSION CHAMBER PRESSURE FOR EACH FOOT OF WATER LEVEL BEIDW A SUPPRESSION POOL WATER LEVEL OF -31 INCHES (-2,6 FEET).
OEOP-01-UC l Rev. 24 Page 74 of 131 l
.
._. _ _ _ _ _ _ . . . . . . . . _ . . . . _ . _ . - . _ . _ . _ _ _ _ . ~ _ _ _ . . . . _ _ , . _ _ _ . . . _ _ _ . . _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ . _ _
. .
,
'*
EOP-01-UG f Attechnen: 5 !
,
-
Page 24 of 31 r t
ATTACHMENT 5 (Cont'd) I
'
!
-
FIGURE 7 !
j UNIT 1 PRESSURE SUPPRESSION PRESSURE ~ !
.
i
+6 '
-
+5
- n v
[ +4 r g +3 i
- !
h ++2 1 SAFE s
o B o
l !
Z -2 O
! w l @ -3 NSAFE
?
w K
_4 .
- a. -5
,
- i l
.'")
l m -6 l
i -7 l
- -8 l
l 5 15 25 35 45 55 E 16 20 30 40 50 t0 SUPPRESSION CHAMBER PRESSURE (PSIG)
,
I
! ..
.
OEOP-01-UG l Rev. 24 l Page 75 of 131 i
. _ . _ _ _ _ . . . _ _ _ _ _ _ . _ _ _ _ . _ . . _ _ _ _ . _ , . . . _ . . . . . . _ _ . _ _ _ . . - _ _ _ . _ _ . _ _ _ _ _ . _ _ _ _ _ . .
< - ,
-
_
!. EOP-01-UC
!
Attachment 5
'
i Page 25 of 31 ATTACHMENT 5 (Cont'd)
j -
I FIGURE 7A
' UNIT.2 PRESSURE SUPPRESSION PRESSURE
- i l
t
,
+6 -
m
+5
'
+4 J
+3 y +2 !
l a ,, SAFE l
.J O 8 O { j
-
2 -2 ;
O '
E -3 vi !
W -4 i
% i ! '
D i u) -6 '
-7 l i
!
-8 i
5 15 25 35 45 55 B is 28 38 48 58 68 ,
SUPPRESSION CHAMBER PRESSURE (PSIG)
l
1 OEOP-01-UC Rev. 24 l l Page 76 of 131
. - . . .- -- - - . . - . - . - . . . _ - . . _ - - - . . .~. ...... - - - ... ... ..-..-..- -.- .-- ~~,
< . *
< ;
~
1 !
-
Attcchesnt 5 .i Page 26 of 31
,
l
ATTACHMENT 5 (Cont'd) i i- L !
'
-
[ FIGURE 8 U -
,
UNIT 1 SRV TAIL PIPE LEVEL LIMIT
'
A Y
I
!
I-
i
'
. +6 'm. . - s s p .3.'m. w.<me .s .s'.gi.3 .m ;31".1
-
r
.n! S i
1-m +5 N UNSAFE- ,m ,
. ,
-
-
- ., .,
L ,..
.
v yg .
, e
.
.. . .
.
' 6:
, , .
!-.
<
+3 ' '
.' .-
l.
,g
+2 '
'
J o SAFE
!.
" o +1 Z E
-
O w
i M -1
.
(n
,
y ;
Q:' -2 i
i i
O !
CL i 3 -3 ;
i cn i t- <
.4-1,158 :
i
.
"
i-18 6 388 588 758~ 988 1,18 8 j
\
B 286 -400 600 BBB 1,000
!
! REACTOR PRESSURE (PSIG)
.
f4 i i
l i :
a-A .
i
i
i
.!
a
.
OEOP-01-UG l Rev. 24 Page 77 of 131 l
i j.
i
_ . . . . . , _ _
_ _ _ _ _ . . . _ _ . _ _ _ _ _ . _ . _ . . . . _ . _ . _ _ _ . . . _ _ . _ _ _ _ . - _ _ . . . . _ _ . . . - . . _ . _ . . _ . _ _ _ . . . _ . _ . _ - . . . _ . . _ _ _ - . .
'
f.
, -
.
,e
$ EOP-01-UG Attechnent 5
!. Page 27 of 31
,
ATTACHKENT 5 (Cont'd)
'
! - I t
> FIGURE 8A
! !
4 UNIT 2 SRV TAIL PIPE LEVEL LIMIT
..
'
t i-
,
!
- +6 - .
- .. . , .1.:
. ,. ,.r.;,
.
- .
!. 7e . ' ,
.. .'
! -
^
F +5
% UNSAFE
, .., .
L'
,
u,, u
.
v- #4 ,
...
, c . ,
,
.
W m g ,.
- > +3 '
< w %
<
_a i +2
,
y I o SAFE
!-
i o +1 i
, CL i .Z 8
.
e O
w I m -1 i M
..
w I m -2 i 1
.
CL
- D -3
i m
!
-4
i
.i
). 18 9 388 500 700 900 118 0 l
0 250 400 688 880 1800 i
(
,
,
.
REACTOR PRESSURE (PSIG)
i
?
l
1
,
}
!,
- OEOP-01-UG l Rev. 24 l Page 78 of 131 i.
I'
I
!-
. _ _ _ _ _ . _ _ _ _ , . . _ . -- _ , _ _ _ . . _ _ _ _ .
. .
,
.
E0P.-01-UG Attachment 5~
Paga 28 of 31 ATTACHMENT 5 (Cont'd)
-
FIGURE 9 UNIT 1 CORE SPRAY VORTEX I.IMIT i
l l
!.
!
+6
~ i
+5
~ ' f
. +4 l h +3 ,
+2 1 -
%+8 !
J -l !
2 -2 i-3 z !
O -4 .
$ -5 .'
gr -6 k -7' '
b -6 I-9
- 18 588 1,588 2,588 3,588 4,588 5,588 6,588 8 1.888 2.888 3,888 4,888 5,888 6,888 7,888 CORE SPRAY FLOW (GPM)
OEOP-01-UC l Rev. 24 Page 79 of 131 l
. . . . . - . . . . . - - . . . . - . . - . . . . - - . . - . - . - . - - . . - - - . - . . . . . - _ _ - _ . _ . . - . - . .
' *
j .
!. '
,
~
i Attechmant 5 Page 29 of 31
.
e
! ;
i ATTACHMENT 5 (Cont'd) i
! !
'
-
FIGURE 10 UNIT 2 CORE SPRAY VORTEK LIMIT
,
,
i i
i i
!
i
! +6 l- +5
', '
.m 44 . -
l l--
! h- +3 t
'
v
+2
. d 1
3J +8
J -1 O
O -2 S_-5
$
w g -6 a, -7
"3 u) -5-9
-le Set 1,565 2,588 3,585 4,588 5,50s 6,588 s 1,855 2A88 3,888 4A88 5,588 6A88 7,885 CORE SPRAY FLOW (GPM)
OEOP-01-UG l Rev. 24 Page 80 of 131 l
, _ , _ _ .._. _
, __.___..-......._.__._...____.-._..._._m.._.__~~.. .-....__.______-__...._m__
! -.
.
<-
1 Attachment'S
,
Page 30 of 31
L ATTACHNENT 5 (Cont'd) .
.
I
' ' l
-
FIGURE 11-
,
UNIT 1 RHR VORTEK LIMIT
- . .
- .~ +5
.. e i- ' +4
!~
I i m +3
" vb +2
.
-
.
d SAFE
'
+1 f
. .
.
'
.
D a 0 ,
r--
'
- .
a -1 o -
-
'
o -2 .
-
.
a_
.
--
-3
-
-
Z ,f
o _,, ; 1 s
H .
u, ,
'
U) -5 w -
a- -6 ,
.
,, ct ,
a_
> -7 --
7 '
v)
-8 UNSAFE ***
-9-10 B 5,000 10,gg0 15.000 20,000
'
s OEOP-01-UC- Rev. 24 I- p ,, gg ,g 131
,
.,r-,,c- -y- ,-c, ,,-y ,,v, -,-,--+yw- , , -.- - , , - - - - --mv y.w.r. m - - . - - -.
. . . . _ . _
_ . . . . ~ _ _ . . . _ _ _ . . . . . _ _ . - , _ _ _ _ . . _ _ . - . _ . . _ _ - . . . . - . _ . . _ . _ _ . _ - _ . _ _ - _ _ . . _
-
,. )
'
i E0P-01-UG Att:chment 5
Page 31 of 31
.l
.
ATTACIDENT 5 (cont'd)
I \
l- -
FIGURE 12 '
i i UNIT 2 RHR VORTEK LIMIT
,
i i e .
,
,
+5 , !
).
,'-
p
.
- 4
+3 !
4 . '
v +2 ^
..
a w ,, SAFE
. ...
'
T 'c j
J , . . .
O -2 ,.'
,
'
a .
.-
~3 z
O '
5 '
w _4 n
-
.
m : i w -5 " -
.
w -
.
Cr "I
'
a -6 - y I i
a .
-) -7 m-8 UNSAFE ###
-9-10 0 5,000 10,000 15,000 20,000 LPCI (RHR) FLOW (GPM)
,
!
!
!
l
!
!
!
OEOP-01-UG l Rev. 24 l Page 82 of 131 l
. . _ . . _ _ _ _ _ _ _ _ . _ - . . . . _ _ _ _ . _ _ _ __ _ _ __ _ ___..___ _ _ _ ___- _ . _. ._ _ _ ____ - -
__
- . s j %
!-
/
t
)
l
,
4 FIGURE 1 l
Thermal Power Limitations Map
!
i i
l
l
! 90 '
l
! 85 i
!
j
-
l
i W 70
$ h
!
i o
e '
l W 55 -
i U <
g i i
!
W
60
,
!
- l #
45 -
! l
! 40 i
} 35 - '
!
l l 30 .
I 23.1 27 30.8 34.7 38.5 42.4 46.2 50.1
!
i MLBS/HR po) 95) (40) es)
!' (so) (ss) (ao) (es)
% CORE FLOW l
'
,
Region A - Manual SCRAM upon entry E Region B -Immediate Exit upon enny
!
g 5% buffer (Rod line & % Flow)-In:reased Macitating of Nuclear u ' mWon Required i
!
!
t
)
i i
i
v
i
1 1AOP-04,0 Rev. 3 Page 9 of 11 l
i
- _ _ - . - - - . -
. _- - - . - -
, ___-__.-----n, , , _ ... -.,. .a ww n .m an u mw as.na
. .n ,am asaan a:s.a n .u . man.n.n~s.La..w e b
/
i FIGURE 2
Rated Power vs. Core AP a
B1C11 POWER VERSUS CORE SUPPORT PLATE DELTA-P i 82
,
x .. m,,
x ~>ry
.
/
.'y';kk.h,'$.;;'T-'
-
'*'9';.g
! l
-, _ l l ....-
- ^ ' #1 - i
! .
?*& '
g,,s. g j 74 -
eiy
'
-en .
' " N =4
- 70 *#
a wgg W
,
-yggp?2l" z ' \
f%.N!N
66 #4'E~~
$98 satta p&#
'
19W
)
& MW i i
i
{ W Ik <
l
! 3 59
- o e w - I j
t 1 s$p[yQ N
' H wm
~
! Z $we
>
w wa aj !,
Q ,Y 1
'
x 51 .m u -
paew (- W i (1, s !!:t:c 9 !!188 a
! 47 .
---
i l = -
,3 '
!
43 ---
!
( h. -
Tl-if- M:'9.
i
" 7N '-*
l 39 f
- w# ;
l -
t '
- .
.<
i 35 -
!
m# !
'
I
,
31 r
,
i 27 '
j 1 .5 3 .5 5 5.5 i
CORE SUPPORT PLATE DELTA-P (PSID)
)
i i
!
>
!
t
!
i
1AOP-04.0 Rev. 3 Page 10 of 11
i
,-,.-.,r - - - - -e--- ------r --3- -- -----r- =r -.-- -- -- r - - - -- + - * - - - - - - - - - - - -
_ __
O FIGURE 3 Estimated Total Core Flow vs. Core Support Plate Delta-P for BlCll 21 _ i i i i i i , i i i i i 20 g: Percent Power Lines Top to Bottom:,90,80,70,60,50,40,30 100 <
gg / /
16 15 '
-
R'
E i
- HH
/HH/H
?' E H/Wl{/
10 '
9 g I Yb /Y ._
8 '
E AWM
5~ I' '
4 /
E AW E
AW 1: V -
0 lill ll'I till lill lill lill lill lill Illi IIII Illi lill I'll 20 25 30 35 40 45 50 55 60 65 70 75 80 85 Core Flow (Mlb/hr)
1AOP-04.0 Rev. 3 Page 11 of 11
! ! t' ; ,< i >!- ,
=
,
-
-
-
-
I
.
- v
- e R
.
-
.
.
-
-
'
-
_~
- _~
iI -
-.
-
-
-
- .
=. .
..
.
.. =. .
.*. . .
.* :
- . . - m
.
'
. .n
._ -
===r i
= . I,
=. . :. n ga....n
.
. .
- .
=
-
.
.m:
.=
=
=
..
.
.
-
"......
..
_- _
-
1 .
.n-
, i
-
. ..
.
.
a a
_
-
O@ =.' .
'
'
'
'
..
. " , ,
.
r. + l . O :
.
,.1 .
iI
.
. = .=,.
^
.
. . .
. .
.
. :
,. ,
f ,
.
- = .
M
_ . : .
A
],~
. .
R n
.u
-
n H i
. .
1,
. n .
, "
" p%
'- '.- .
.
s.,N
.
.
..
.. G A
r
.
s~ . -H
-
- -
-
.
i1
-
,
e m
%
.
. . o/*z s
.
.
I D
K
- ' , =
m =" C I
= I O
y 2* . 1l EL
-
k J
(.(n .
/.s
,V RB UL
f o
q Q
Qi H
t y
c
=
. .. ,
_
,
_
J' r=
a = .I GA I
FC
6
-
m s1 pI M
.
.
=..
==.
I R
_-
w
..
. ., a *.'."" ===
T C
N~ H _ge t
I
== E L
.
$.
p f.,
. .
. ,
, s
.
o\ n
.
E C
.. e 1T
.
.
l
.,1 ,
' H M A.= .
. *
E t
s ee . e r ,a ,> gt
- .s 1T m
z ,, ,
,
A.=.
.
= _- @E a
>
-
I Jy
-
. ..=- .* m
,gm A=E ,.=-
e >
, ,
. .
- .s. L.T "
i -
.
,
,
,
-
";- .-
'
",
" r'
, T_
y g
_ A=g1_,,t , -
-
=" .
. :
.
" ..='= "
.
." .
f
- .
C. . .*.
. "*'
. .
" ..
=
. ,""
.
.
.
.
.*. .
D-
- 6
0-M S -
S L
C-S P
O
. - -. _
, ,
-
i I
@&[
-
CAROLINA POWER & LIGHT COMPANY BRUNSWICK NUCLEAR PLANT PLANT OPERATING MANUAL VOLUME XXI ABNORMAL OPERATING PROCEDURE UNIT RECEIVED BY BNP
!
h Q\S \N
\0b JUN 2 51996 nucuAR DOCUENT C0m0L OAOP-1 LOSS OF SHUTDOWN COOLING REVISION 8 EFFECTIVE DATE (D cTl 9Co i
5'M -
'
Sponsor s 7 Date Approval /, , , , ' (_/ b-22-96 n'ager'- Operations mb)2s $2123 Date OAOP-1 Re Page 1 of 15 l
l
. . . . _ _ _ . _ .. _ . _ .. _ _ _ _ - . _ . _ . . _ . . _ . . .
. .
.
I r
i l
REVISION SUMMARY ;
i
' Added caution in various locations describing the potential for a false low RPV l
water level signal under specific plant conditions which incorporate OPS Al 95-01883 Task 9. Deleted cautions, notes, and steps which referenced actions 1 required for monitoring reactor coolant heatup/cooldown rates as this inforrnation is wholly covered in 1(2)PT-01.7 which is referenced in this procedur i LIST OF EFFECTIVE PAGES
,
,
Paae(s) Revision 1-15 8 i
l
..
)
.
l OAOP-1 Re Page 2 of 15
. . ~ . _ _ . _ - _ . . _ . _ _ _ . - _ . _. _ _ _ _ __ ___
. .
I
-
!
I SYMPTOMS RHR SWPUMP 1A(2A) TRIP (A-01 1-9) or RHR SWPUMP 1C(2C) TRIP l (A-013-9) annunciator in alar l RHR SW PUMP 1B(2B) TRIP (A-031-8) or RHR SWPUMP 1D(2D) TRIP (A-03 3-8) annunciator in alar .3 l RHR PUMP 1A(2A) TRIP (A-013-8) or RHR PUMP 1C(2C) TRIP (A-015-8) annunciator in alar l RHR PUMP 1B(2B) TRIP (A-03 3-7) or RHR PUMP 1D(2D) TRIP
,
(A-03 5-7) annunciator in alar t 1.5 1 RHR HX A/B DISCH CLG WTR TEMP HI (A-03 2-9) annunciator in alar . i RHR A/B DISCH & SUCT HDR PRESS HI (A-03 3-9) annunciator - !
,
in alarm.
i
. REACTOR VESS LO LEVEL TRIP (A-05 2-6) annunciator in alarm.
- Group 8 Isolation Valves clos l Increasing. Reactor Coolant Temperature and/or Pressur R16 1.10 High NSW or CSW header pressure approaching pump shutoff head
- (approximately 90 psig). 1
.
R16' 1.11 Unexplained changes in running RHRSW loop flow or pump discharge -
i
, pressure, ,
i
!
,
- AUTOMATIC ACTIONS I e
t
. IF a Group 8 Isolation Signal exists (Low Level One or High Steam Dome Pressure), THEN the following will occur:
,
-
RHR SHUTDOWN COOLING OUTBOARD ISOLA TION VAL VE, ;
E11-FOOB, will clos "
-
RHR SHUTDOWN COOLING INBOARD ISOLA TION VAL VE, ;
E11-FOOS, will close.
g i
! <
-J
^
,
,
OAOP-1 Rev 8 Page 3 of 15
'
,
__ _ . ._ .
. .
. AUTOMATIC ACTIONS
-
Loop A(B) INBOARD INJECTION VAL VE, E11-F015A(B)', will clos The RHR Pump in service for Shutdown Cooling will trip on a loss of suction pat .0 OPERATOR ACTIONS ^ Immediate Actions "
, None applicable i
- Supplementary Actions CAUTION
!
R20 IF reactor coolant temperature is greater than 212 F, AND reactor water level has been raised to greater than 218 inches for 10 minutes or more, THEN a '
false RPV low level signal could result when the reference leg condensing pot N12A(B) nozzle is uncovered as levelis subsequently lowered below 218 inche . IF Shutdown Cooling has been lost due to a tripped RHR Pump, THEN START an PHR Pump in the loop being used for Shutdown Coolin . IF forced circulation has been lost, AND natural circulation has NOT been established, THEN RESTORE AND MAINTAIN reactor vessel water level between 200" and 220" as read on B21-LI-R605A/B), or as directed by Shift Superintendent based on plant conditions, until forced circulation is restore * Low Level One Only OAOP-1 Re Page 4 of 15
/ _ __ _ __ _
. .
. Supplementary Actions NOTE: A Group 10 Isolation Signal will isolate the air supply to Reactor Head Vent Valves, B21-F003 and B21-F004. These valves fail closed on a loss of air supply or powe . DETERMINE the position of Reactor Head Vent Valves, B21-FOO3 and B21-FOO . IF vessel coolant-temperature is greater than 212 F, or is
'
indeterminate, THEN ENSURE Reactor Head Vent Valves,
,
B21-FOO3 and B21-FOO4 are CLOSE CAUTION Natural circulation cannot be depended on to provide adequate flow through the bottom head region or the recirculation loops. The recirculation loop l suction temperatures and bottom head temperatures therefore cannot be '
utilized for vessel coolant temperature monitoring for indication of boiling.
Under natural circulation conditions, reactor vessel pressure must be monitored for coolant temperature determination. If coolant temperature was initially less ,
i than 212 F, pressure must be closely monitored for indications of a trend of increasing pressure. If this trend is established, it must be assumed that 212 F has been exceeded, boiling is occurring, and a mode change has taken place.
,
3. DIRECT an operator to perform the following: MONITOR reactor coolant heatup/cooldown'in accordance with 1(2)PT-01.7 so that any unexpected trends can be promptly reporte !
NOTE: If the time to boiling in the reactor vessel CANNOT be determined, - ;
then it must be assumed that 212 F will be exceede l OBTAIN the approximate time to boiling in the reactor vessel based on current plant conditions (value should be in Daily Schedule Report).
0AOP-1 Re Page 5 of 15
_ . . ._ -. _ _ _ . - _ . . _ . _ . _ _ ._
,
'
l I
. Supplementary Actions l
l NOTE: Secondary Containment Pressure Seal Work Permits are tracked in - '
, accordance with OENP-5 )
3. IF it becomes apparent that Shutdown Cooling CANNOT be !
reestablished OR it has been determined that 212 F will be :
exceeded,THEN: DIRECT Engineering to restore Secondary Containment prior to l exceeding 212 i
'
I RESTORE Primary Containment prior to exceeding 212 !
I CLOSE Reactor Head Vent Valves, B21-FOO3 and B21-F00 ]
!
3. IF the operating RHRSW loop has been lost, THEN PERFORM the )
following:
l
,
R16l IF unexplained changes in flow or pump discharge pressure are l
observed in the running RHRSW loop, AND NSW or CSW header pressure approaches pump shutoff head (approximately 90 psig),
THEN ENSURE UNIT 1(2) SERVICE WA TER DISCHARGE OUTLET VAL VE,1(2)-SW-V442, is ope ; IF available, THEN START the idle RHRSW Booster Pump in the RHRSW loop being used for Shutdown Coolin ' IF the NSW or CSW Service Water Header has been lost, THEN l PLACE the RHRSW loop in operation using the other Service Water header in accordance with 1(2)OP-4 . IF NO RHRSW Booster Pumps can be placed in operation, THEN i
PLACE RHRSW in operation with NO RHRSW Booster Pumps i available in accordance with 1(2)OP-4 . IF Service Water is unavailable, THEN PLACE Fire Protection Water in operation to the Service Water Header in accordance with OAOP-1 OAOP-1 Re Page 6 of 15 i
. . . _ _ .
. .
.
. Supplementary Actions CAUTION R20 IF reactor coolant temperature is greater than 212 F, AND reactor water level has been raised to greater than 218 inches for 10. minutes or more, THEN a ;
" false RPV low level signal could result when the reference leg condensing pot ;
,
N12A(B) nozzle is uncovered as level is' subsequently lowered below 218 inche :
'
l
. ;
'
3. IF the operating RHR loop in Shutdown Cooling has been lost, AND a Group 8 isolation Signal does NOT exist, AND NEITHER l
RHR Pump in the RHR loop being used for Shutdown Cooling can be started, THEN SHIFT Shutdown Cooling loops in accordance
.
with 1(2)OP-1 . IF the operating RHR loop in Shutdown Cooling has been lost,
!
AND a Group 8 isolation Signal exists, THEN PERFORM the following: RESTORE AND MAINTAIN reactor water levelin the previously established band in accordance with Shift Superintendent directio . REDUCE reactor pressure below 125 psig in accordance
.
with OGP-0 . ENSURE RPS is energize ;
, WHEN ALL Group 8 lsolation signals have cleared, THEN RESET
,
the Group 8 Isolatio .
-
PLACE the RHR loop that was operating in Shutdown Cooling back in service in accordance with the following: IF piping cool down or drain down are a concern, THEN GO TO Step 3.2. IF piping cool down or drain down are NOT a concern, THEN CONTINUE with Step 3.2.8. OAOP-1 Re Page 7 of 15
,
. _ _ . ___ __ . _ . - - -
,
. Supplementary Actions
, CLOSE Loop A(B) OUTBOARD INJECTION VAL VE, E11-F017A(B).
t
' OPEN Loop A(B) INBOARD INJECTION VAL VE, ,
E11-F015A(B). 1
! OPEN RHR SHUTDOWN COOLING OUTBOARD ISOLA TION \
VAL VE, E11-FOO8.
,
l OPEN RHR SHUTDOWN COOLING INBOARD ISOLA TION VAL VE, E11-FOO CAUTION i
Failure to minimize RHR Pump operation while deadheaded may cause pump I damag l 1 START an RHR Pump in the loop being used for Shutdown ;
Coolin ' SLOWLY THROTTLE OPEN Loop A(B) OUTBOARD INJECTION VALVE, E11-F017A(B) to re-establish RHR loop conditions prior to the even WHEN RHR loop conditions have stabilized, FULLY OPEN Loop A(B) OUTBOARD INJECTION VALVE, E11-F017A(B). IF the reactor coolant temperature is less than 212 F, THEN ENSURE Reactor Head Vent Valves, B21-FOO3 and B21-FOO4 are OPE MAINTAIN RHR in Shutdown Cooling in accordance with 1(2)OP-1 OAOP-1 Re Page 8 of 15
.=
1 Supplementary Actions NOTE: Filling and venting may be required if the loop was idle for an extended period of tim '
CAUTION l
Piping may contain hot water at greater than 212 F due to previous Shutdown Cooling operatio . IF RHR has NOT been restored in Shutdown Cooling in
-
accordance with the above steps, THEN PLACE the RHR loop that
{
was operating in Shutdown Cooling back in service in accordance '
with 1(2)OP-17 as soon as conditions permi . IF necessary to minimize reactor coolant temperature rise, THEN PERFORM one of the following feed and bleed combination l i
Eff_Q BLEED COND/FW in accordance with RWCU Reject in accordance 1(2)OP-32 with 1(2)OP-14 CRD in accordance with Main Steam Line Drains in 1(2)OP-08 accordance with 1(2)OP-32 Core Spray in accordance with 1(2)OP-18 LPCI in accordance with 1(2)OP-17 3.2.10 IF NEITHER RHR loop can be placed in Shutdown Cooling, THEN PLACE the Condensate System in Condenser Cooling in accordance with 1(2)OF-3 .2.11 IF ALL of the above methods CANNOT maintain vessel coolant l temperature below 212 F, THEN INITIATE alternate Shutdown Cooling with the SRVs as follows: ENSURE ALL control rods are fully inserte _
OAOP-1 Re Page 9 of 15
. .
+
. Supplementary Actions IF the Reactor Recirculation Pumps are running, THEN PERFORM the following: RAISE AND MAINTAIN reactor water level between 200" and 220" as read on B21-L/-R605A/B), or as directed by Shift Superintendent based on plant condition STOP the running Reactor Recirculation Pumps in accordance with 1(2)OP-0 ' SHUT DOWN the RHR loop that was operating in Shutdown Cooling in accordance with 1(2)OP-1 . PLACE one RHR loop in the Suppression Pool Cooling mode in accordance with 1(2)OP-1 . IF Suppression Pool temperature rises above 95 F, THEN GO TO OEOP-02-PCCP, Primary Containment Control Procedure AND PERFORM CONCURRENTLY with this procedur . CLOSE the following valves: Inboard and Outboard MSIVs (B21-F022A-D and 821-F028A-D). HPCI inboard and Outboard Steam Supply Valves (E41-FOO2 and E41-FOO3). RCIC Inboard and Outboard Steam Supply Valves (E51-F007 and E51-FOO8).
, Inboard and Outboard Reactor Head Vent Valves (B21-F003 and B21-FOO4). Inboard and Outboard Main Steam Line Drain Valves (B21-F016 and B21-F019).
,
OAOP-1 Re Page 10 of 15
'
.
.
! Supplementary Actions SELECT one SRV based upon the desired cool down rate using the following table: '
HIGHEST B21-F013F B21-F013A B21-F013K B21-F013E COOLDOWN 821-F013H B21-F013B B21-F013L '
B21-F013G . B21-F013C B21-F013G B21-F013C B21-F013J B21-F013D B21-F013J B21-F013D !
-
821-F013A B21-F013E B21-F013E B21-F013A B21-F013B B21-F013F B21-F013F B21-F0138 l B21-F013K B21-F013H B21-F013H B21-F013K B21-F013L {
B21-F013L l 821-F013C B21-F013G B21-F013C B21-F013G B21-F013D B21-F013J B21-F013D B21-F013J i
LOWEST B21-F013E B21-F013K B21-F013A B21-F013F COOLDOWN 821-F013L B21-F013B B21.F013H PLACE the control switch for the desired SRV to OPE l
! RAISE AND MAINTAIN reactor water. level greater than 254 inche !
,
!
0AOP-1 Re Page 11 of 15
'
.
1 Supplementary Actions NOTE: The RHR pumps are preferred for injectio NOTE: Monitoring Tsar per 1(2)PT-01.7 may NOT be valid under these special conditions due to reactor pressure NOT necessarily relating to TSAT. Therefore, SRV tailpipe temperature recorder B21-TR-R614 on Panel H12-P614, and/or ERFIS trending should be utilized for monitoring reactor coolant cool down rat .
1 START one RHR or Core Spray Pum '
1 THROTTLE OPEN the injection valve on the affected pump until the SRV open . IF reactor pressura CANNOT be maintained less than 164 psig above Suppression Chi:mber pressure, THEN PLACE another SRV control switch to OPE .
1 PERFORM the following actions as necessary to maintain cool down rate less than 100 F per hour: THROTTLE CLOSE the injection valve on the affected pump until the desired SRV close RECORD reactor pressure at which the SRV close l psig THROTTLE OPEN the injectiors valve on the affected pump until the SRV reopen THROTTLE CLOSE the injection valve on the affected pump until reactor pressure is 10 to 20 psig greater than the pressure at which the SRV closed in Step 3. 2.1 1.1 . IF it is desired to ADJUST the cool down rate, THEN CLOSE the open SRV AND OPEN the next SRV that will ADJUST the cool down rate in the desired directio . REPEAT Step 3.2.11.14 until vessel coolant and Suppression Pool temperature are within 100 . CONTROL Suppression Pool temperature as necessary to maintain vessel coolant temperature above 75 AOP-1 Re Page 12 of 15
- . .
= .- -
, .
. Supplementary Actions 1 WHEN a normal method of Shutdown Cooling can be established, THEN SHUT DOWN alternate Shutdown Cooling as follows: STOP the ECCS pump (s) used for vesselinjectio WHEN the SRV(s) that were opened have closed, THEN PLACE the control switch for the SRV(s) to CLOSE OR AUT . IF the reactor coolant temperature is less than 212*F, THEN
,
OPEN Reactor Head Vent Valves, B21-FOO3 and B21-FOO CAUTION R20 IF reactor coolant temperature is greater than 212*F, AND reactor water level
.
'
has been raised to greater than 218 inches for 10 minutes or more, THEN a false RPV low level signal could result when the reference leg condensing pot N12A(B) nozzle is uncovered as level is subsequently lowered below 218 .
inche ' RESTORE AND MAINTAIN reactor water level between 200" and 220", or as directed by the Shift Superintendent, based on plant condition WHEN directed by the Shift Superintendent, THEN SHUT DOWN the RHR loop used for Suppression Pool Cooling in accordance with 1(2)OP-1 .0 GENERAL DISCUSSION An extended loss of the decay heat removal function can lead to elevated vessel coolant temperatures, localized or bl.lk coolant boiling and potentially result in a depletion of reactor coolant and eventual uncovering of the cor If no forced circulation exists during a Loss of Shutdown Cooling event, natural circulation must be established. Natural circulation however, cannot be depended on to provide adequate flow through the bottom head region or the recirculation loops. The recirculation loop suction temperatures and bottom head temperatures therefore cannot be utilized for vessel coolant temperature monitoring for heatup rate determination or indication of boilin In addition, if RWCU is not in service with suction from the bottom head, vessel bottom head drain temperature cannot be used for verification of Tech Spec 3.4.6.1 (Pressure / Temperature Curves) compliance for cooldown AOP-1 Re Page 13 of 15
- .. . . - -
.
. GENERAL DISCUSSION l
'
Under natural circulation conditions, reactor vessel pressure must be monitored for vessel coolant temperature determination if vessel coolant temperature is less than 212 F, pressure must be closely monitored for indications of a trend of increasing pressure. If this trend is established,it j must be assumed that 212*F has been exceeded, boiling is occurring, and a
,
mode change has taken plac !
R2O !
,
If reactor coolant temperature is greater than 212 F, and reactor water level i has been raised to greater.than 218 inches for 10 minutes or more, then a false RPV low level signal could result when the reference leg condensing l
>
!
' ^ pot N12A(B) nozzle is uncovered as level is subsequently lowered below 218 inches. This false signalis the result of water exiting the nozzle and condensing pots at the same time steam is re-entering the reference le This counter flow condition sets up the conditions conducive to steam bubble creation and collapse. This causes a momentary upward pressure spike in the reference leg, which gives a momentary indicated false signal to
,
the transmitters involved.
While irradiated fuel remains in the reactor vessel during an outage, maintaining the decay heat removal function remains a key to shutdown
' safety. The risk associated with a loss of decay heat removal event is dependent on a number of factors, including the decay heat load present and the existing plant configuration. Outage risk assessment will ensure that adequate contingency plans are in place prior to reducing decay heat removal capabilit This procedure addresses a loss of normal decay heat removal capability during shutdown conditions. The procedure provides contingencies for the following methods of decay heat removal: !
'
- RHRSW Loop Failure
- RHR Loop Failure
- Condenser Cooling Failure
- Feed and Bleed Combinations
- Alternate Shutdown Cooling with SRVs This procedure also provides contingencies for restoring primary and secondary containment, and initial emergency actions for a loss of Shutdown Coolin Industry events have occurred which demonstrate that use of the SRVs to steam to the Suppression Poolis a viable method of decay heat remova Makeup requirements can be supplied by a single CRD Pump in this mode of cooling. Outage risk assessment should prevent the need for ever using this mode of coolin AOP-1 Re Page 14 of 15
,. .
_ _ _ _ _ _ ._ ..__ _ _ . _ _ _ _ . _ _ . . - . _ . . _ _ _ . _ . .
'
.
i a
i
_5.0' REFERENCES 5. '1 Regulatory Guide 1.33, Quality Assurance Program Requirements-(Operations) (November 1972), Appendix A, item .2 ANSI Standard N18.7-1976, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants, Section 5.3.9.2, item (3) Technical Specification 3.4.6.1, Pressure / Temperature Limits Technical Specification 3.6.1.1, Primary Containment Integrity
- Technical Specification 3.6.5.1, Secondary Containment integrity
, FSAR Section 5.4.7, 7.4.3, 15. .7 NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management OAOP-18.0, Nuclear Service Water System Failure (2)OP-02, Reactor Recirculation System Operating Procedure 5.10 1(2)OP-08, Control Rod Drive Hydraulic System Operating Procedure 5.11 1(2)OP-14, RWCU System Operating Procedure 5.12 1(2)OP-17, RHR System Operating Procedure 5.13 1(2)OP-18, Core Spray System Operating Procedure 5.14 1(2)OP-32, Condensate and Feedwater System Operating Procedure 5.15 1(2)OP-43, Service Water System Operating Procedure R16 5.16- IER #92-21-03 (IFI); FACTS #93B9034 5.17 OEOP-02-PCCP, Primary Containment Control Procedure l -5.18 OENP-54, Building Ventilation Pressure Control Program i
L 5.19 1(2)PT-01.7, Heatup/Cooldown Monitoring
R2O 5.20 CR 95-01883, False RPV Water Level Low Level 1 Signals
>
i l
- OAOP-1 Re Page 15 of 15 i
{
l
'
_ _
_ . - _
- . . .-, . . _ - - - -- . .. ~ . . . . - . . . - - - .
e -
T
<
Section 1 - Primary Containment Water Level Calculation Suppression Pool water level instruments may be used for levels up to
+2 fee CAC-LI-2601-1 (XU-51)
r CAC-LR-2602 (XU-51)
! IE Suppression Chamber pressure is less than 75 psig, THEN CALCULAT ,
!
Primary Containment water icvel as follows: l Pi -
Suppression Chamber pressure plus head of water i
-
CAC-PI-1257-2A (XU-51)
CAC-PI-1257-2B-(P601)
P2 -
Drywell pressure at greater than 85 ft elevation CAC-PI-1230 (P601)
PCu - 2,3 feet (P, - P2 ) + 5.3 feet psi
>
i l TIME 5 '
P,(ps ig)
l T P (psig)
l Pi - P 2 i
L
! x x x x x 2.3
! x 2.3
!
+ + + + 5,3- + +5.3
'
PC,(ft)
.,
e l OEOP-01-PCFP l Rev. 10 l Page 11 of 22 l
.
__ __ _ . . .. . _. _ . . . . . _ _ _ , __ _
! . -
t
.
I
'
l Section 1 (Continued) 1 IE Suppression Chamber pressure is greater than 75 psig, THEN CALCULATE
- Primary Containment water level as follows
i {
i P, - Primary Containment pressure plus head of water CAC-PI-4176 l (XU-51) j CAC-PR-1257-1 (XU-51)
P, - Drywell pressure at greater than 85 ft elevation CAC-PI-1230 (P601)
-
i P, measured using CAC-PI 4176 PC , - 2. 3 feet (Pi - P ) + 28.5 feet
psi P measured using CAC-PR-1257-1
P,Cu - 2.3 feet (P, - P ) + 30. 5 feet
psi Using PI-4176 TIME Pi (psig) i P2 (psig) l-l Pi -
P3
l I
x x x x x )
x i I
I
+ 2 + 2 + 2 + 2 + 2 + 28.5 i
!
l P C.,
i l
[ l OEOP-01-PCFP l Rev. 10 l Page 12 of 22 l l
l
_ _ . . . . __ _ .. _ . .
,
.
Sestion 1 (Continued)
Using PR-1257-1 i
I I
TIME I i
-
P,(psig)
~
l l
,
'
P:(psig)
P, - P 2 !
!
x x x x x x i
"
+ 3 + 3 + 3 + 3 + 3 + 3 i l
!
PC.,( f t )
1 I H Drywell water level is greater than 60.5 feet. THEN USE CAC DRWELL
'
VATER LEVEL INDIC, CAC-LI-1216 on P60 l
!
i I
'
. l I
!
,
!
.
l OEOP-01-PCFP l Rev. 10 l Page 13 of 22 l
'
. .
,
.
C CP&L
-
CAROLINA POWER & LIGHT COMPANY continuous BRUNSWICK NUCLEAR PLANT use DATE COMPLETED FREQUENCY; UNIT % PWR GMWE
"**"'"**'*'*"**V'*h*"P'* 'd *~-
SUPERVISOR REASON FOR TEST (check one or more):
Routine surveillance OWP#
WR/JO #
Other (explain)
PLANT OPERATING MANUAL VOLUME X PERIODIC TEST UNIT ~
B N P RECIPIENT ID 2 m: ebb JONTROLL E I -
2PT-0 HEA TUP /COOLDOWN MONITORING REVISION o RECEIVED BY BNP EFFECTIVE DATE NAY 131996
/ NUCLEAR DOCUMENT CONEL Sponsor D, i
_Sk. 9k %
Datel
'
Approval -
Ov'T _
Date Menggerations 2PT-0 Re Page 1 cf 1 f l
l
. . . . . _ . . ._ . . _ _ . - . . - - _ . _ - . .
..
. _ _ _ ___ _ . . _ . . _ . _ . _ . . .
l'
- REVISION SUMMARY i This procedure was issued to provide a method to determine compliance with the
,
requirements of Technical Specifications 3.4.6.1, items a and b, except during inservice hydrostatic or leak testing. This surveillance satisfies Technical j
Specification 4.4.6.1.1 Surveillanco Requirement during Reactor Coolant System
'heatup and cooldow '.lST OF EFFECTIVE PAGES e,
I r
- Pace (s) Revision
f 1-11 0
.
.
$
l
! !
.
.
.
. .
f l
.
'
,
.
l i
, ,
f
- !
-
i .
4 i
!
2PT-0 Re Page 2 of 11 i
~ . , - .-
. --
. .
,
1.0 PURPOSE i
1.1 This surveillance is performed to determine compliance with the I l
requirements of Technical Specifications 3.4.6.1, items a and b, except ;
during inservice hydrostatic or leak testing. This surveillance satisfies !
Technical Specification 4.4.6.1.1 Surveillance Requirement during Reactor Coolant System heatup and cooldow .2 This test involves data collection of certain plant parameters, and 1 i
confirmation that reactor coolant system pressure and temperature are maintained within limit '
2.0 REFERENCES 2.1 Technical Specifications i I GE SIL No. 430, Reactor Pressure Vessel Temparature Monitoring 2.3 GE SIL No. 251, Control of RPV Bottom Head Temperatures 2.4 GE SIL No. 251 Supplement 1, BWR Vessel Bottom Head Coolant Temperature !
i 2.5 OGP-02, Approach to Criticality arW Pressurization of the Reactor j
2.6 OGP-05, Unit Shutdown l
2.7 20P-17, Residual Heat Removal System Operating Procedure 2.8 OAOP-15.0, Loss of Shutdown Cooling 2.9 OEOP 01-RSP, Reactor Scram Procedure I 3.0 PRECAUTIONS AND LIMITATIONS 3.1 IF PPC Display 860, RPV HEATUP/COOLDOWN MONITOR, is available, THEN the value of DT/HR to be recorded on Attachment 2 is the as-displayed numerical valu .2 IF Tm is above 212*F, OR in natural circulation, and PPC Display 860, RPV HEATUP/COOLDOWN MONITOR is NOT available, THEN coolant
~ temperature is to be determined from the Steam Tables utilizing reactor steam dome pressure as read on RTGB instrument <
2PT-0 Re Page 3 of 11
-
'
. .
.
3.0 PRECAUTIONS AND LIMITATIONS WHEN Tsn is less than 212 F, OR when RPV pressure monitoring instrumentation is NOT valid, THEN coolant temperature is to be determined by one of the following methods:
3. Reactor Recirc Pump running AND loop is NOT isolated from the Reactor, THEN use R6 circulation Suction Temperatures read on B3bTR-R65 . RHR Pump is running in Shutdown Cooling mode with the Heat Exchanger aligned as follows:
~
3.3. RHR HX in service: Use RHR HX 2A(B) Inlet Temperature as read on E41-TR-R605 Point 1(2), on Panel H12-P61 .3. RHR HX NOT in service: Use RHR HX 2A(B) Outlet Temperature as read on E41-TR-R605 Point 3(4), on Panel H12-P61 .4 Bottom Head temperature during heatup and cooldown may be determined in a number of ways depending on the status of the RWCU System and the Reactor Recirc Pump . During heatup, C12-TR-RO18, Channel 151, Bottom Head metal temperature on Panel H12-P007, is the preferred source due to the metal temperature response lagging the coolant temperature response. IF Channel 151 is unavailable, THEN alternate methods can be used depending on. RWCU status as follows:
-
RWCU in service: Use Vessel Bottom Drain Temperature as read on G31-TI-R607 on Panel P603, OR C12-TR-RO18 Channel 153, on Panel H12-P00 RWCU NOT in service AND Reactor Recire Pumps 2A AND 2B running at 2: 20% speed: Use Recire Suction Temperature as read on B32-TR-R65 _
2PT-0 Re Page 4 of 11
- - . - . - . - . - ~ . - - .. .-
. .
.
.
., i
-
!
,
.;
3.0 PRECAUTIONS AND LIMITATIONS i
!
! 3. During cooldown, Bottom Head coolant temperature is the '
,
preferred source due to coolant temperature response leading !
vessel metal temperature respons !
--
IF RWCU is in service, THEN Vessel Bottom Drain coolant ,
temperature can be determined from G31-TI-R607, OR C12-TR-RO18, Channel 15 ;
i l
-
IF RWCU is NOT in service, THEN Recirc suction temperature !
as read on B32-TR-R650 may be_ utilized, BUT ONLY IF both '
~ Recire Pumps are running at greater than or equal to 20% !'
spee IF RWCU AND Recire are NOT in service, THEN use C12 TR RO18, Channel 151,
4.0 PREREQUISITES Non'e -
5.0 SPECIAL TOOLS AND EQUIPMENT Steam Tables 6.0 ACCEPTANCE CRITERIA WHEN the following criteria are met, THEN this' procedure may M considered satisfactory:
6. Calculated heatup or cooldown in any one-hour period does NOT exceed 100"F. This is the value of Differential Temperature / Hour (DT/HR) recorded on Attachment 2 or .1.2' Operation is to the right and/or below limiting lines of Technical Specification Figures 3.4.6.1-1 or 3.4.6.1-2, as applicable. This is determined by verification that the BOTTOM HEAD value recorded on Attachment 2 or 3 is to the right and/or Delow limiting lines for vessel pressure of the appropriate Technical Specification Figures (Attachment 4 or 5).
I i
e I
2PT-0 Re Page 5 of 11
- :
. .
-
i s
7.0 PROCEDURAL STEPS
'
NOTE: Data thould be recorded at an increased frequency of at least once per 10-15 minutes during transient conditions such as scram, loss of heat sink, uncontrolled depressurization, et .1 COMPLETE the applicable portions of the Data Sheet of Attachment 2 or 3, at least once per 30 minute O'
7.2 INITIAL on the appropriate Data Sheet to indicate that both acceptance criteria are me ;
'
7.3 IF either acceptance criterion is NOT met, OR limits are being approached, THEN IMMEDIATELY NOTIFY the Unit SC .4 CONTINUE monitoring until determined unnecessary by the Unit SC Q )
'
l
!
i
.
2PT-0 Re Page 6 of 11 l _
.. . .
.
. . . - . - . - . . - - _ - . . - . - . . - - . - . . - . . . - _ - . - . . - , . . - -
. .
.
'
{ ATTACHMENT 1
- - !
Page 1 o! I
'
Certification and Review Form '
!
'
2 PT-01.7 - I
- !,
General Comments and Recommendations.7 i
'
' .
i initials Name (Print)
Test procedure performedliyi'
4 !
!
'
>
'
!
"
t
'
l ;
,
'
,
Exceptions to satisfactory performance
.
l
!
!
'
t Corrective action required
,
Test procedure hn been satisfactorily completed Unit SCO '
Signature Date Test procedure has not been satisfactorily completed Unit SCO Signature Date Test has been reviewed by
.
Shift Superintendent Signature Date 2PT-0 Re Page 7 of 11
'
.
.
l
l
,
ATTACHMENT 2 Page 1 of 1 2PT-0 Data Sheet (Tsar > 212 F or Natural Circulation)
DT/HR' l T,y 'F + /- BOTTOM HEAD'
T.,, ' F MAX / MIN 100*F / Initials 'F / Initials DATE/ TIME CURRENT in hour (cnteria met) (criteria met)
/ /
/ /
/ /
/ /
/ /
/ /
/ /
/ /
/ / I
,
/ / l
/ /
/ /
l
/ /
/ /
/ /
/ /
/ /
/ /
/ / I
/ /
This Attachment may be duplicated as needed for periodic monitorin *
IF PPC DISPLAY 860, RPV HEATUP/COOLDOWN MONITOR is available, THEN RECORD the displayed value of DT/HR and NA the T.,r columns. IF PPC DISPLAY 860 is not available, THEN DT/HR is calculated as the larger of T.,1 CURRENT minus the maximum or minimum T.,, in the previous hour. The value may be positive or negative dependent on plant heatup or cooldow **
BOTTOM HEAD temperature during a plant cooldown is determined from G31-TI-R607, Pt. 5 or C12-TR-RO18, Ch.153, IF RWCU is in service. BOTTOM HEAD temperature during a plant heatup is determined from C12 TR-RO18, Ch.151 (preferred), or either C12-TR-RO18, Ch.153, or G31-TI-R607, Pt. 5, IF RWCU is in service. IF RWCU is NOT in service, BOTTOM HEAD temperature monitoring should be performed using C12-TR-RO18, Ch.151, OR Recirc suction temperature IF both Recire Pumps are running at greater than or equal to 20% spee '
2PT-0 Re Page 8 of 11
.
'
. .
ATTACHMENT 3 Page 1 of 1 Data Sheet (Tsu < 212'F or invalid Pressure Instruments)
2PT-0 '
DT/HR *
RECIRC RECIRC RHR + /- BOTTOM HEAD" A B SDC 100*F / Initials DATE/ TIME "F
F / Initials
'F *F teriteria met) (criteria met)
/ /
/ /
/ / j
/ / !
/ /
/ /
/ /
/ /
i /
/ / {
/ /
/ /
/ /
/ /
!
/
/
/ /
/ /
/ /
/ /
/ /
This Attachment may be duplicated as needed for periodic monitorin *
DT/HR is calculated as the largest change in either valid RECIRC A or B temperatures in the previous hour, or if in SDC, the largest change in the appropriate RHR HX temperature in the previous hour. The value may be positive or negative dependent on plant heatup or cooldow **
BOTTOM HEAD temperature during a plant cooldown is determined from G31 TI-R607, Pt. 5 or C12-TR-RO18, Ch.153, IF RWCU is in service. BOTTOM HEAD temperature during a plant heatup is determined from C12-TR-RO18, Ch.151 (preferred), or either C12-TR-RO18, Ch.153, or G31 TI-R607, Pt. 5, IF RWCU is in service. IF RWCU is NOT in service, BOTTOM HEAD temperature monitoring.should be performed using C12-TR-RO18, Ch.151, OR Recirc suction temperature IF both Recirc Pumps are running at greater than or equal to 20% spee PT-0 Re Page 9 of 11
-
.
.
ATTACHMENT 4 Page 1 Of 1 Tech Spec Figure 3.4.6.1-1 Amendment 172 Pressure-Temperature Limits Reactor Vessel 1200 NORMAL OPERATION WITH CORE NOT CRITICAL I
n'
1100 '
I e
1000 X f I f
.Es 1!
-
0,
$
800 f)f f
- J
.y 700 S l '[
tu r i
CC 600 '
o .-
0) (550) ,/
en
>
$ 500 l o-
. I an Luli r
A *
100 .
l J
l 200
/
r,f ~
100 ;;;;; @'
- -g
-____ u O
l 100 l200 300 400 500 600 (70) (170)
TEMPERATURE (*F)
NOTES:
1. OPERATE TO RIGHT AND/OR BELOW UMmNG UNES 2. * INDICATES BOTH HEATUP AND COOLDOWH RATE 3. PRESSURE AND TEMPERATURE INTERSECTIONS NOTEC BY PARENTHESES s
2PT-0 Re Page 10 Of 11
__ .
_ ._ . _ _ . _ _ _ _ .
,
. , ;
.' l
)
ATTACHMENT 5 i Page 1 of 1
<
Tech Spec Figure 3.4.6.1-2 Amendment 172 Pressure-Temperature Limits Reactor Vessel NORMAL OPERATION WITH CORE CRITICAL -
120 , ,
l
-
,
,
, 1100 l
I tooo 1 1
..
$j l ,
< ;
ooo -i ,
800 [ '
,
e) !
^ I . l
. Too [ ,
l
>
S / :
tu . r .
Q' 600 ' '
a f *
l m ;
e i .
E-s*
a
!, ,
,
,
,
i i !
400
=, [
.
l
, /,/_
!
'
l 200 /
JPt": -
,
r'
'
s f * ~
'
100 ,
.r : 2 .;:: = t .
i
' 3 l k1
'
, .._.: l o too [foo soo 400 soo soo (193)
(210)
TEMPERATURE (*F)
1. OPERATE TO RIGHT AND/OR BELOW UMmNG UNES 2. * INDICATES BOTH HEATUP AND COOLDOWN RATE 3. PRESSURE AND TEMPERATURE WTERSECTIONS NOTED BY PARENTHESES 4. OPERATION IN SHADED AREA PERMtTTED ONLY WHEN WATER LEVEL IS WHITHIN NORMAL RANGE FOR POWER OPERATIO ,
2PT-0 Re Page 11 of 11 I
!
- -- r6 MGo VOLTS [\,m- 354 o
- s ly z . ' -
.,
A " di io' \
igrp i
- 6 5"J3"}.m g g 2 '
s a g qct
'
@C viu- .
G t...m. -14 i y
'n :. ccz s tu
,s "-
a
<'
3 q$J g L_ ts gg
.
t (.- (~ CTS I (_. ( too/% U-((3
-
U *
-
I 05
-
p
-
~
U o
- V OJ el d 4 -
uJ e5( P Q
N: m O
$$% Ng f 3Se O zo u OC dew 7 s~ S~ ~l ~ ',
$E" ,Q)p cx 5@ , w e2 zge 9)
s r d oiy
~~E sV gid '1D k
6g o
my 5 e 8 5 Yd 1750 ,M a f l MI -- w p t.sb== L%
@. l Et at M T L'
_e
' e :: a la -}rst.\-COO \h
.
I
'k v Q *
g 35A
- ;S S A n7 ^t 0 zN 3G #
Cn'
. .
-
_ _ __ _ _ _ _ _ . -. ..
. .. . .. . -
" ' '
/' "G A-so/ 5) cc-so/si i io ' clo D '
1 ACS 115 I~~~~~
A C. 5 7 % i O -
(O' N '
I de
m"a o' <,L- - .
si
.s s1
NB > @r @ ,,-
1 stuho:
<
S
-
i m
.
i
- ar,@.. e
. Eo st s1 9 m cc so
,-
51 s11 i
.
!
g-oo y =i ..
== == == g=eo !
._
s do y o
,,
y- 's "s *s *zAi4S"EC 4 A u,y "" a APERVU RE i
.
, ,. cgs ..
.
9 _
.
CAN !
,q
_
. 87 Also Avallisb!o on !
- "
eg =c/gg Aperturt Ca rd i
1 1Acc ;
S_75 07 .5.), i i
[- T 5()qg t_ g 51 i r. - a - s e j
'
g "- g g' .4 ; REVISED PERI !
otz s9 "
1~ < (:' PDC-88-COM
.' ,
'
i 21A
-
f5\ 7S- i gg.,w l
'
71 2 ,b 7
" 4 2. 1 O h, .
AS BUIL7 PER :
5 \ 5o h ' "
. SC ric%
l t
j PM 77-13 8 i B . ;
!
g ~150cA A 50MTTMit! '
!
. 7OW A S E 0l'T PEA G
k '+ g og 4 4 F"E
-
C P' \ \ 2.~7 l'
<
2.-# c-4 AF S Q
'
EE V') PER n :
??j u 3 uBX 74Cl _
-
3 tl.75 MM
== ??S 9-19-74 '
%nN r<E.raneuca ows D 1 }ELETEDgyh jg s2ulTot e5 7.pp.sese Q 4 pea to :i] Hi2 Psci & - t'I. FG27
,~ 4~ > oi ~
3 Rgve sec-
,;'
i. , } av8 PER e, . 7- n M SY/C4Y/C)h _ /y7[ _o WtsT \%us
'
REV DESCRIPTION fb .A
l l
znc N:sa 49/hhw D AT E 10 G-11 STATS a sG.40RT4 CARO \.lu A, N. 6 o "S" (c} l CAROLINA POWER A LIGHT CollPAN7 BRUNSWICX STEAM ILECTRI:: PLAMT EV. 57 l
~
~4160 V O LT S WGR E 3 .
., .,
i
-
_ . _ _ - . __COM P T AI6 109)
_.._C O R E__SE R A ( PUMP 2A
.C.O NT RO L W_lRI N.G _DJ A G RA M
C LL-09I13 C% RD I5 -
l
.. - - . - - _ . . - . . .
.
I h
i REFERENCES
.
FOR WRITTEN TEST
- SRO ONLY
-
!
l i
e
!
>
!
!
i
!
l
i
$
)
)
!
.
,,
>
.
1/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS s
'
1/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION !
1. f.imiting Conditions for Operation and ACTION requirements shall be appitcable during the OPERATIONAL CONDITIONS or other states specified for i l
each specificatio ;
1. Adherence to the requirements of the Limiting Condition for Operation I and associated ACTION within the specified time interval shall constitute compliance with the specification. In the event the Limiting Condition for !
' operation is restored prior tc expiration of the specified time interval, completion of the ACTION statemeat is not require !
1.0.1 In the event a Limiting Condition for Operation and/or associated ACTION requirements cannot ,be sat:isfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in at least HOT SIIUTn0WN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that permit operation under the permissible ACTION statements for the specified time interval as measured, from initial discovery or until the reactor is placed in an OPERATIONAL CONDITION in which the specification is n'ot applicable. Exceptions to these requirements shall he stated in the individual specification . Entry into an OPERATIONAL CONDITION or other specified applicability state shall not he made unless the conditions of the Limiting Condition for
_
6- Operation are met without reliance on provisions contained in the ACTION statements unless otherwise excepted. This pr'ovision shall not prevent passage through OPERATIONAL CONDITIONS required to comply with ACTION requirement .0.5 When a system, subsystem, train, component, or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERARLE for the purpose of satisf ying the requirements of its applicable Limiting Condition for Operation, provided: (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system (s),
subsystems (s), train (s), component (s), and device (s) are OPERABLE, or likewise satisf y the requirements of this specification. Unless both conditions (1)
and (2) are satisfied, the unit shall be placed in at least HUr SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in at least COLD SHUTDOWN within the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> This specification is not applicable in Conditions 4 or 5 I
C7 BRilNSWICK - UNIT 2 3/4 0-1
- - -
. .
- RETYPED TECH. SPEC Updated Thru. Amend.
-
'
s
'
-s -
- .
, ~[ APPLICABILITY a _
-
SURVEILLANCE REQUIREMENTS
- -
4. Surveillance Requirements shall be applicable during the OPERATIONAL CONDITIONS or other states specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requiremen . Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interva '
.
4. Performance of a Surveillance Requirement within the specified time !
interval shall constitute compliance with OPERABILITY requirements for a Limiting Condition for. Operation and associated ACTION statements unleas otherwise required by the specification. Surveillance requirements do not have to be performed on inoperable equipmen . Entry into an OPERATIONAL CONDITION or other specified applicable state i shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specifie .0.5 Surveillance Requirements for inservice inspection and testing of ASME
,
Code Class 1, 2, and 3 components shall be applicable as follows:
4 Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g) (6)
(i).
.
l l
,
!
l
BRUNSWICK - UNIT 2 3/4 0-2 Amendment No.180
i*
}
l rm i 6 3/4.0 APPLICABILITY i
-
SURVEILLANCE REQUIREMENTS (Continued)
. -
b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
ASME Boiler and Pressure Vessel Required frequencies l Code and applicable Addenda for performing inservice l
>
terminology for inservice inspection and testing l inspection and testing activities activities Weekl y At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days j Yearly or annually At least once per 366 days ' The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activitie ,
- Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirement Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specificatio .
f. The Inservice Inspection Program for piping identified in NRC Ceneric Letter 88-01 shall be performed in accordance with the staff positions on schedule, methods & personnel & sample expansion inicuded in this lette .
La-BRUNSWICK - UNIT 2 3/4 0-3 Amendment No. 180
.
,
. !
i a
SODIUM PENTABORATE SOLUTION VOLUME CONCENTRATION REQUIRE ;
FIGURE 3.1.5-1
,
.
,
i
_
. ;
I Sodium Pentaborate Solution Volume
, Concentration Requirements I 2 I l l 2 '
'
' l i
- I 2 Repon of Requred Vadume Concentrabon
^
1 \^
~
'
1 ~~ --
A -
The vdimemourermanrmos Lt i me msar m - Wtw wb e _
_ onermoonetaless esom y 1 .
-
1 ~
1 ~
.b
' ;
i h* 1 *
I i
1 j
-
1 ~
1 ' ' ' ' ' 'I' ' ' '
2000 2500 3000 3500 4000 4500 5000 5500 i NetVolume d Solubonin Tank (gads)
!
.
BRUNSWICK - UNIT 2 3/4 1-20 Amendment No. 212
U
,.
j .
,
, ,
e
'$
e O
140 t s *
130 g ,20
,
-
" 110 g
/
'
/ -
h100 -
v
'
"N /
~
80 /
60 /
/
0 5 10 15 20 25 30 35 40
.' PERCENT SODI W PENTABORATE BY WEIGHT OF SOLUTION
,'O SATURATION TEPG'ERATURE OF SODitN
-
PENTABORATE SOLUT10N
,f FIGURE 3.1.5-2 '
. .
,
R
-
.
- - _ _- ._ ___ ______ __ - _ ___- - _ __ _ _ _ _ _ . _ _ _ . _ _ . . .. - . . . - .
- - - . .
- .
,
'
!
INSTRUMENTAT10N
!
3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION l
LIMITING CONDITION _FOR OPERATION 3.3.2 The isolation actuation instrumentation channels shown in Table 3.3.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.2- i
{
i APPLICABILITY: As shown in Table 3.3.2- ;
ACTION: With an isolation ointtrip lessset $umn conservative than the actuation value showninstrumentation channel in the Allowable Values co
.
Table 3.3.2-2, declare the channel inoperable until the channel is !
restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint valu For any isolation actuation Trip Function with less than the Minimum Number of OPERABLE Channels per Trip System required by Table 3.3.2-1: Within one hour, verify sufficient channels remain OPERABLE or are !
placed in the tripped condition * to maintain automatic isolation i actuation capability for the Trip Function, and Place the inoperable channel (s) in the tripped condition * within:
a) 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for trip functions common to RPS Instrumentatio l and b) 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for trip functions not common to RPS Instrumentation Otherwise. take the ACTION required by Table 3.3.2- Delete I l The provisions of Specification 3.0.3 are not applicable in OPERATIONAL CONDITION ,
l SURVEILLANCE REQUIREMENTS 4.3.2.1 Each isolation actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK. CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.2- .3.2.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip syste *
An inoperable channel need not be placed in the tripped condition where this '
would cause the Trip Function to occur. In these cases. if the inoperable channel is not restored to OPERABLE status within the required time, the ACTION required by Table 3.3.2-1 for the Trip Function shall be take l
.
BRUNSWICK - UNIT 2 3/4 3-10 Amendment No. 206
_ _-
. .
,
..
INSTRUMENTATf0N t
SURVEILLANCE REQUIREMENTS (Continued) !
4.3. }
demonstrated to b5 within its limit at least once perEach 18test months.The shall include at least one logic train such that both logic trains are tested at I i least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months. where N is the total number of 1 redundant channels in a specific isolation functio l i
i l
i i
!
' Radiation monitors are exempt from response, time testin !
,.
!
I l
!
l l
l l
l BRUNSWICK - UNIT 2 3/4 3-11 Amendment No. 206
. - _ - _ _ __
_ _ _ _ _ _ _ _ _ . . ___ ___ . _ _ _ _ . _ _ _
,.
.
.
.
-E
{
g
,
n TABLE 3.3.2-1
.
ISOLATION ACTUATION INSTRUMENTATION
H VALVE GROUPS MINIMUM NUMBER APPLICABLE N OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b)(c) CONDITION ACTION, 1. PRIMARY CONTAINMENT ISOLATION Reactor Vessel Water Level - Low. Level 1 2. 6 2 1. 2. 3 20 8 2 1. 2. 3 27 Lod. Level 3 1 2 1. 2. 3 20
$ Drywell Pressure _- High 2. 6 2 1. 2, 3 20 Y Main Steam Line iG (Deleted)
1 Pressure - Low I 2 1 22 Flow - High I 2/line 1 22 Flow - High l 2 2. 3 21 Main Steam Line Tunnel Temperature - High l ' 2'* 1. 2. 3 21 Condenser Vacuum - Low 1 2 21 1
& Turbine Building Area
@ Temperature - High l 4'"
o 1. 2. 3 21 g Main Stack Radiation - High (h) 1 1. 2. 3 28 m Reactor Building Exhaust S Radiation - High 6 1 1. 2. 3 20
- - _ _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ _ _ _ - - _ - _ _ - .. . - .
_
- _ _ _ _ _
,
.
..
,
~,
'
.
'$n TABLE 3.3.2-1 (Continued)
-
ISOLATION ACTUATION INSTRUMENTATION <
C N
- VALVE GROUPS MINIMUM NUMBER APPLICABLE
" OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b)(c) CONDITION _ ACTION' '
2. SECONDARY CONTAINMENT ISOLATION Reactor Building Exhaust (1) 1 1,2,3,5, 23 Radiation - High and *
6 1 1,2,3 20 t w Drywell Pressure - High (1) 2 1,2,3 23 g 2, 6 2 1,2,3 20
'
" Reactor Vessel Water Level - (1) 2 1,2,3 23 w Low, Level 2 3 2 1,2,3 24 3. REACTOR WATER CLEANUP SYSTEM ISOLATION A Flow - High 3 1 1,2,3 24 : Area Temperature - High 3 2 1,.2, 3 24 r Area Ventilation A Temperature - High 3 2 1 2,3 24 SLCS Initiation 3 '" NA 1, 2 24 kg Reactor Vessel Water Level -
Low, Level 2
3 2 1,2,3 24
.EI A Flow - High - Time Delay NA 1 1,2,3 24 l
@ Piping Outside RWCU Rooms Area 3 , 2, 3 '
Temperature - High 24
+
.
'
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._________.____._____.____..m__1 - - _ . _. _ _ _
' t; - e *
- * <
.
..* (g *
- _
g TABLE 3.3.2-1 (Continued)
E
.g ISOLATION ACTUATION INSTRUMENTATION O VALVE CROUPS MINIHUM NUMBER APPLICABLE e
OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SICNAL(a) PER TRIP SYSTEM (b)(c) CONDITION ACTION U s u CORE STANDBY COOLING SYSTEMS ISOLATION
' , High Pressure Coolant Injection System Isolation HPCI Steam Line Flow - High 4 1 1, 2, 3 25 i
' HPCI Steam Line Flow - High l
Time Delay Relay NA 1 1, 2, 3 25 ,
- HPCI Steam Supply Pressure - Low 4 2 1, 2, 3 25 w 7 1 1, 2, 3 25 l
w HPCI Steam Line Tunnel Temperature - High 4 2 1, 2, 3 1
@
f Bus Power Monitor NA IE} 1/ bus 1, 2, 3 26 ' HPCI Turbine Exhaust - t Diaphragm Pressure - High 4 2 1,2,3 25 ,
, HPCI Steam Line Ambient Temperature - High 4 1 1, 2, 3 25 t HPCI Steam Line Area a Temperature - High 4 1 1, 2, 3 25 l , . HPCI Equipment Area I Temperature - High 4 1 1, 2, 3 25 D
z 1 Drywell Pressure - liigh 7(k) 3 3, y, 3 73 l
?
O
.
_ _ . _ . _ _ . _ _ _ . _ _ . _ _ _ _ _ _ _ . _ . _ _ . _ _ . _ . _ _ . . _ _ _ . _ _ . _ . _ _ _ _ _ _ . _ . . . _ _ _ _ _ _ . . _ _ . _ _ _ _ _ . _ _ _ _ _ . . _ _ _ _ . _ _ _ . _ ._ - . _ _ _ - . - . _ . _ __ . _ _ _ _ - _ _ _ _ , . , _ _ . . _ . . _ . . . _
$' 9 -
'
V h
. . y .
._
s
. . .
TABLE 3.3.2-1 (Continued)
'
E g ISOLATION ACTUATION INSTRUMENTATION h VALVE CROUPS OPERATED BY HININUM NUMBER OPERABLE CHANNELS APPLICABLE OPERATIONAL i
' *
SICNAL(a)
TRIP FUNCTION PER TRIP SYSTEM (b)(c) ' CONDITION ACTION <
.E f 4 CORE STANDBY COOLINC SYSTEMS ISOLATION'(Continued)
u " Reactor Core Isolation Cooling System Isolation e
, .
i RCIC Steam Line Flow - High 5 1 1, 2, 3 25 !
, RCIC Steam Line Flow - High l .
Time Delay Relay NA 1 1,2,3 2$ f
. RCIC Steam Supply Pressure - Low
g) 2
1, 2, 3 1, 2, 3
25 l
' RCIC Steam Line Tunnel
" Temperature - High '5 2 1, 2, 3 25
>
s~
w Bus Power Monitor NA IE) 1/ bus 1,.2, 3 '
.'. ,
i k'i 6 !
'
- Pressure - High 5 2 1, 2, 3 25
'
0 RCIC Steam Line Ambient ;
Temperature - High 5 1 1, 2, 3 25 l !
!
' RCIC Steam Line Area , l
'
A Temperature - High 5 1 1,2,3 25 l r t
' RCIC Equipment Room Ambient Temperature 'High 5- 1 1, 2, 3 25 l l
RCIC Equipment Room
{
,
1 A Temperature - High 5 1 1, 2, 3 25 l
-
11. RCIC Steam Line Tunne i NA 1 1,2,3 - 2 5-W Temperature - High j
'
, Time Delay Relay [
12. Drywell Pressura - High 9(k) 1 1, 2, 3 25
- :
Y ,
.
- , , - _ . . . ~ . - , _ . . -
. . . - ,-. . . . . . ._ .._,-.-_.-._m, . . _ _ _ _ ~ , . . , , . . . - - . . . . . . . . - , + _ . . _ . . . - . _ . , .
.
ge ej- .,
. . ,,- .
'
.
5 TABLE 3.3.2-1 (Continued)
E
. ISOLATION ACTUATION INSTRUMENTATION ,
M M
.
'
VALVE CROUPS HINIMUM NUMBER APPLICABLE s
OPERATED BY OPERABLE CHANNELS OPERATIONAL E -TRif FUNCTION SICNAL(a) PER TRIP SYSTEM (b)(c) CONDITION ACTION
.
i w SHUTDOWN COOLINC SYSTEM ISOLATION !
t Reactor Vessel Water Level - 2, 6 2 1, 2, 3 20 ;
8 1,2,3
'
Low, Level 1 2 27 Reactor Steam Dome Pressure - High 8 III I 1, 2, 3 27 l t
!
.
>
i U* ,
hs .
!
h I ' i
. .
I i
I
t
'
?
i
-
- .
I i A
.
. s
_- - - - . _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ . _ _ . . _ _ _ .
- - _ . _ _ _ . . . . _ _ _ _
-
_
-_,________m.-.____ _
-
_ ___ _.__. __ . -...... -.._.. .-_ _. __ .... _ - _ _._. __,__..__ .._ ._-._-.._ . .
.
- .
, ,
TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION i ACTIONS j ACTION 20 - Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> i ACTION 21 - Be in at legst STARTUP with the main steam line isolation valves closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT SHUTDOWN within 6 i hours and in COLD SHUTDOWN within the next 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> l l
' ACTION 22 - Be in at least STARTUP within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> l I
ACTION 23 - In OP{ RATIONAL CONDITIONS 1, 2, or 3, establish SECONDARY CONTA~NMENT INTECRITY with the standby gas treatment system opera'.ing within one hou In OPERATIONAL CONDITION 5 or when handling irradiated fuel in ;
the secondary containment:
1) Establish SECONDARY CONTAINMENT INTECRITY with the standby gas treatment system operating within one hour; I
2) Otherwise, suspend handling of irradiated fuel in the l secondary containment, CORE ALTERATIONS, or activities that could reduce the SHUTDOWN MARCI ;
,, -
ACTION 24 - Isolate the reactor water cleanup syste ACTION 25 - Close the affected system isolation valves and declare the affected system inoperabl j ACTION 26 - Verify power availability to the bus at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ACTION 27 - Deactivate the shutdown cooling supply and reactor vessel head spray isolation valves in the closed position until the reactor steam dome pressure is within the specified limit ACTION 28 - Close the affected isolation valves within 14 days or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . ,.
BRUNSWICK - UNIT 2 3/4 3-17 Amendment No. 178
-
_ . . . . . . . _
.
f
. , . ';
s .. .
,
TABLE 3.3.2-1-(Continued) '
'
ISOLATION ACTUATION INSTRU'AENTATION
- '
NOTES
When handling irradiated fuel in the secondary containmen !
(a) See Specification 3.6.3.1. Table 3.6.3-1 for valves in each valve grou i
!
(b) When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated ACTIONS may be delayed as follows:
(1) For up only on.e to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> channel perfor tripTrip Functions with a design that provides syste I i!
-(2)- For up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for all Trip Functions, provided the Trip Function maintains isolation actuation capabilit .(c) Delete l )
-(d) A channel.is OPERABLE if 2 of 4 instruments in the channel are OPERABL (e)' With reactor. steam pressure a 500 psi (f) - Closes only RWCU outlet isolation valv )
(g) Alarm onl (h) Isolates containment purge and vent valve (i) Does not isolate E11-F015 (j) Does not isolate B32-F019 or B32-F02 (k) Valve isolation depends upon low steam supply pressure coincident with
-
-;
high drywell pressur (1) Secondary containment isolation dampers as listed in Table 3.6.5.2-1,
,
i
i !
l l
'
l BRUNSWICK - UNIT 2 3/4 3-17a Amendment No. 206
. . _ . _ _ . _ , _ - - _ _ .
-
. .
,
?
INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATIQM I LIMITING CONDITION,FOR OPERATION 3.3.3 The Emergency Core Cooling System (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints I set consistent with the values shown in the Trip Setpoint column of Table 3.3.3- !
APPLICABILITY: As shown in Table 3.3.3- {
ACTION: With an ECCS actuation instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of .
Table 3.3.3-2. declare the channel inoperable until the channel is l )
restored to OPERABLE status with its trip setpoint adjusted consistent '
with the Trip Setpoint value.
' With one or more ECCS actuation instrumentation channels inoperable, take the ACTION required by Table 3.3.3- i The provisions of S3ecification 3.0.3 are not applicable in OPERATIONAL CONDITI1N SURVEILLANCE REQUIREMENTS 4.3. Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK. CHANNEL CALIBRATION and I CHANNEL FUNCTIONAL TEST o !
thefrequenciesshowningerationsduringtheOPERATIONALCONDITIONSandat able 4.3.3-1 4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months and shall include calibration of time delay relays and timers necessary for proper functioning of the trip syste .3.3.3 The ECCS RESPONSE TIME of each ECCS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months, where N is the total number of redundant channels in a specific ECCS functio BRUNSWICK - UNIT 2 3/4 3-33 Amendment No. 206
.
. .
-
- <
.,1
_
.
P g E
o p; TABLE 3.3.3-1 n
.
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION E .
M MINIMUM APPLICABLE m-OPERABLE CHANNELg OPERATIONAL- '
TRIP FUNCTION PER TRIP FUNCTION ' CONDITIONS ACTION
'
1. CORE SPRAY SYSTEM Reactor Vessel Water Level - Low. Level 3 4 1.2.3. i Reactor Steam Dome Pressure - Low (Injection Permissive) 4 1.2.3. ' I Drywell Pressure - High 4 1. 2. 3 ca .30 i .
,
2 Time Delay Relay 1/ pump 1.2.3.4.5- 31 1 Bus Power Monitor'* 1/ bus 1.2.3. . LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM Drywell- Pressure - High 4 1. 2, 3 30 I Reactor Vessel Water Level - Low. Level 3 4 1. 2. 3. 4*'. S'*) 30 'l Reactor Vessel Shroud Level (Drywell Spray Permissive) 1/ valve 1. 2. 3. 4"' 5*' 31 I t Reactor Steam Dome Pressure - Low (Injection Permissive) !
1. RHR Pump Start and LPCI Injection Valve Actuation 4 1. 2. 3. 4' 5*' '
2. Recirculation Loop Pump Discharge Valve Actuation 30 l
{
- s 4 1. 2. 3. 46'.. 5*' 30 l ,
g RHR Pump Start - Time Delay Relay- 1/ pump 1. 2. 3. 4*'. 5"' 31 I A l Bus Power Monitor'* 1/ bus 1. 2. 3. 4*' . 5*' 32 !
O i
[
a
___ _ -_. _. - -_ _ _ - _ - _ - - _ - _ - _ - - _ _ _ _ _ _ _ _ _ _ _ - _ - _ - - _ _ _ _ _ _ - _ - _ _ _ - _ - _ _ _ - - - _ _ _ - _ _ - _ _ _ _ _ _
______ ___________ __ _ _______ _ _ _ _ _ _ . _ _ _ _ _ _ _ .
_
.
.
%,
E p; TABLE 3.3.3-1 (Continued)
'
. EMERGENCY' CORE COOLING SYSTEM ACTUATION-INSTRUMENTATIOM
$.
--4 MINIMUM APPLICABLE su ' TRIP FUNCTION OPERABLECHANNELp OPERATIONAL PER TRIP FUNCTION * CONDITIONS ' ACTION 3.11LGH PRESSURE COOLANT INJECTION SYSTEM - Reactor Vessel' Water Level - Low. Level 2 4 1. 2. 3 30 l Drywell Pressure - High 4 1. 2. 3 30 I
. Condensate Storage Tank Level - Low 2 1. 2. 3 33 Suppression Chamber Water Level - High 2'c3 1, 2. 3 33 ' Bus Power Monitor'* 1/ bus 1. 2. 3 32
- ADS Inhibit Switch 2 1, 2. 3 36 i F Reactor Vessel Water Level - Low. Level 3 4 1. 2. 3 36 1
- Reactor Vessel Water Level - Low. Level 1 2 1. 2. 3 36 1 AD$ Timer 2 1. 2. 3 36 I ( Core Spray Pump Discharge Pressure - High (Permissive) 4 1. 2. 3 36 .I i
f A RHR (LPCI MODE) Pump Discharge Pressure - High (Permissive)
Bus Power Monitor'*
2/ pump 1. 2. 3 36 I
'
1/ bus 1, .
Os
{
i
_ . _ - . _ . . . _ . . . . . _ . _ . _ _ . . _ . . _ . . _ . . . - _ . - - . . _ _ . . . _ . - . . _ , . _ _ - . _ . _ . -
._ .__ _ _ . . . ___. ._ . - _ _ , . _ _ _ _ . _ . _ . . _ . . . _ . . _ _ _ _ _ . . . . _ _
. _ ,
v .,
ty /
.
. .
,
"
.
-
s
.
TABLE 3.3.3-1 (Continued)
'
l EMERCENCY 008E COOLING SYSTEM ACTUATION INSTRUMENTATIDW g
I
. APFLICABLE
'I OPERATIONAL TOTAL NLRIBER CHAM ELS MINIMIM CHAISIELS {
OF CHANNELS TO TalIP OPERABLE CONDITIONS ACTION l j i FUNCTIONAL UNIT
_
-
., .
I -
, IDSS OF POWER 4.16 kw Emergency Bus 1/ bus 1/ bus 1/ bus 1,2,3,4I *),5I *) 34 i
, i Undervoltage (Loss of !
Voltage)
' 4.16 kw Emergency Bus 3/ bus 2/ bus 2/ bus 1,2,3,4I *I,5I *I 35 i Undervoltage (Degraded Voltage) i
- .
- i j*
. l
.
I I
- l
. .
,
!
- !
I *
siL .
'
. I, .
,
p i i
$ ;
3 e
'
,
. - . _- . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - __ _ _ _ _ _ - _ _ _ _ _ _- __-_-_ _ ____________ - __ _ _ _ _ _ _ _ _ _ _ - . .-_ -___ -
. -- - . . - - .- - - _~ - - - . - - - . ~- -_ -
-
e .
D r .
TABLE 3.3.3-1 (Continued) '
!
EMERGENCY CORE COOLING SYSTEP ACTUATION INSTRUMENTATION i
. EE ACTION 30 - With the number of OPERABLE channels less than required by the '
Minimum OPERABLE Channels per Trip Function requircment: I Within I hour, verify sufficient channels remain OPERABLE or *
are placed in the trioped condition to maintain automatic !
ECCS actuation capability .for the Trip Function, and
.
l Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, place all inoperable channels that do not cause the Trip Function to occur in the tripped conditio '
Otherwise.. declare the associated ECCS inoperabl :
I ' ACTION 31 - With the number of OPERABLE channels less than required by the
Minimum OPERABLE Channels per Trip Function requirement, declare l l the associated ECCS inoperable.
i
! ACTION 32 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, verify- I bus power availability at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or declare the associatec ECCS inoperable.
l
!~ ACTION 33 - With the nuraber of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at I least one inoperable channel in the tripoed condition within one hour or declare the HPCI system inoperabl l ACTION 34 - With the numbe of OPERABLE channels less than the Total Number i
!
of Channels, declare the associated emergency diesel generator inoperable and take the ACTION required by Specification 3.8.1.1 l or 3.8.1.2. as appropriate.
t ACTION 35 - With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; operation may then continue until performance of the next required CHANNEL FUNCTIONAL TES ACTION 36 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, verify within one hour that a sufficient number of channels remain OPERABLE to maintain actuation capability of either ADS Trip System A or ADS Trip System B and restore the inoperable channels to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Otherwise, declare ADS inoperable.
l
,
l
,
(-
BRUNSWICK - UNIT 2 3/4 3-37 Amendment No. 206 i
I
. - -, _ , , , . - - . . ,
._ .- _ _ . _ _ . _ . . _ _ _ . _ . . . - _ _ _ _ . . _ _ . _ . _ _ _ _ _ _ . ~ _ . . _ _ _ _ _ _ . _ . . . . _ _ _ _ _
,
!
.
.
L
.
TABLE 3.3.3-1 (Continued)
i ,
'
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION
-
HQT (a) When a channel is placed in.an inoperable status solely for performance of
. required Surveillances, entry into associated ACTIONS may be delayed for-
,
up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the Trip Function or the redundant Trip Function l maintains ECCS actuation capability.
! .
(b) Not applicable when two core spray system subsystems are OPERABLE per l Specification 3.5.3.1.
l-
!- ,
. (c) Provides signal to HPCI pump suction valves onl '
'l (d) Alarm onl '
'
(e) Required when ESF equipment is required to be OPERABLE.
!- ,
I l
l L
l r
I L
l i
,
?
BRUNSWICK - UNIT 2 3/4 3-38 Amendment No. 206 l r . . , . . - ., . , - - . - . - -
r . - . , , , , - . - - , , . , . .,
_ . _ , _ . . _ - . _ - . _ _ _ . . . _ _ _ _ _ _ _ - _ _ . _ . . . __. _ _ _ . _ _ ._ ___ _ .m __ .
.
.
. .
e t
. ,
- FIGURE 3.4.6.1-1- I PRESSURE-TEMPERATURE LIMITS ;
, .s REACTOR VESSEL :
.
- 1 J l i -
NORMAL OPERATION WITH CORE NOT CRITICAL I J ,
l
!' 1200 , , ,. , , ., . . . . . . .. , , , 4 , , , . . . , . , , , , , , ,
l'
i i
e i i ,4 . . + . . . .. i i , 4 . i i , , . , , . , . , , . , j
,
. , , i i ...+ . i ,4 : # i . . , 4 . . . . , . , , , . . . . . . 3 i .
4 ei i , . ., si,,. , i . , , e i , , , , . . ; . . . :
i
, .t . , , ii 6 , . ., Is 4 . 4 i . i . . .
l 1100 , , , , . i . ,. , . ., , , , , , .
- .
, , , ,
l
,
i
, , ,
i
, , ,
, .
j I
, . . i 6 . . 6 . . . ii , . . , i . 6 . . . . . . .
- i i + ,. . 1 . , . . . . . . . . . . . .
.. 4 e ii + .i
,
, , , , . , . . . . . . .
' ' ' '
- 1000 . i e ,' ' .' . . . .' .' .' ,' ,c y
' ' ' ' ' ' ' ' ' ' '
-
, , , . -,
4 . .
. i .
. .
.
. . .,
+ ..i
. . .i 7.... , . ,
i . i
.
. .
.
. . . .
, , . 4 i . , i
.
.
. .
.
.
.
.
.
.
,
.
, , .
. . . . . .
i
.
! . . . . . . .
, , , . #4 , i ,i ! , r Ei . . . . , . . . . . . 4 4 . . . . . . .
S 4 900
' ' '
,' ,'
'
' ' ,' ' ' ' Y' .'
' ' ' ' ' ' ' ' ' ' ' ' '
. . . .' i .
.'
,-
t i i s i e is
, y i i
.
,
. .
i
.
, , .
, . .
.
, , ,
. . ..
i ,
.
.
, i i . ia i . # i , e i i i . , , . ,
t ii . , . ,
g e i
. t 4
.i
.
900 , ,
. , , , . , . ,,
i , , , .
t . .
, .
.
,
,
, , , ,
, i
, , ,
, , .
e i i i . . 6 f ., i , ,
, e ,
, , I . i , , .
, , 8 . i . i 6 4== i 3 . 4 . i 6 , , i , . , .
f 6 e i
. 4 f 1 a f 6 i .. . f f . . . 4 4 . . , . . . I p , i . 6 i 4 e i . 3 . . t , i , . . . . . , . 4 i . . . i
- . . , ,,6 .
,
....i . . ... , , . . i . ,,,, . . . . , ,
.p
.
8 . i , i . . e t i 6 i e i 4 l 6 . .,i4 . . . . , i i .
[ . 4 4 , .i, J , i i , , 6 . . . , . .
- e
'
i e , i . , is o , .
'
, . . . , , i . . . . . . .
,i wr 600 , , .' .' ' ' . .' . ,' ;
'
e'" , ,'
' ' ' '
. . . .
' ' '
,' i .
'
,
,
G i , 4 . ., .i . ., i . , . . , , , , , ,
! 5 (550) ' *6 e ii ' , , ... . . ' i . ' . i .
',
'
,_ m . , i . . .
'
i , , ,e 4 4 i i , . i . . , .
!
!, . *
500 ,' , .' '
,' ,' ' ', , ,' .' .' .' .'
' ' ' ' ' ' ' ' ' '
'
.
, , , , , . . .' , , . . . . . , ,
. . i . i ..., .i . ,.. . , i . , , ..,. .
--
3- . i ,iii i, ., . . , , . , , , , .
, , i,, i, ., ,, i . . , , .... . i
. .
! 400 , , , . ,w
, ;
r
,,,
.. . .i
, ,
i i i i . . .
, ,
..
,,
.
,
.
. ,
I (
j i 6 4g i i it . . . .
.
, , i , . . . . .
, .; i , i . , . . i i . ,
j . , . . . . .. .
300 ! i 4 .
- . , f i , . , i . , . . . . . .
<
/l .f 6 , i t i 6 4 , . .
.'
i i e i s i i , , ,
{ , g I , . } , e . . . l
!
t,
-
200 /, 7
'
,
,f . i . .
- . .
. . . .^
, i . .
l i
l r i
, 4 . !
, . .
j 100 . . ..
, . . .. ;
,
.. 3 . 6 4 i l
, , . , , . . !
! .
i , , , . . . I l 0 l
100 1 200 300 400 500 6
. (70) (l70) l
. TEMPERATU84: (* F)
.
6-d j BAEEE i FUEL IN REACTS
4 ,
< .1 16 XEFFg 10 N/CH" > 1 IEY j
j-4 .- RTg = 33 (1/4 T) rs! ZESTRIBENT LOCATION CCEIRECTION IIICLUDED ,
- REG. OtTIDE 1.99 *
REY. 2 i $38 ,
1 OPERATE TO RIGET AND/OR RELOh! LIMITING LINES
, - tT/ * IWICATES BOTE EEATUP AND rm vm RATE , '" '
j PRES 8URE AND TDWERATURE INTERSECTION 5 NOTED BT PARENTEESEE
d BRUNSWICK - UNIT 2 3/4 4-15 Amendment No. 172
-
, _ . _ . - _ _ , . . .
. . . - _ . . . _ _ . _
.
< :
.
'. .
,
FIGURE 3.4.6.1-2 PRESSURE-TEMPERATURE LIMITS REACTOR VESSEL f NORMAL OPERATION VITH CORE CRITICAL ]
~
.
]
i i
1200 i
' '
,.
i i ,
l .
1100 . '
i e , , i
.. .a !
1 i , .
-
'
l
,
s I
I g ,
,
i .
,
i +
1 ,
.
$ ;
i . .
90C ,'
.t
' '
,
l i o i ,
I C , 6 .
- '
. ,
l 000 ' ' '
r g
,
4 .
, , , .
, .
A
.~ r , e .
-
e 700
'
' ' '
-
, i !1 ,
-
e i ,i . ,
w J l ' 3 , l' i I i - e i ei ~ / f f f n 6 0 -
100 7 00 300 400 500 60C i M (2 01 T M M M (*P')
.
,
.. .
. . i
"
/. . I assE ! FUEL 3 REACTIlR l . s7.1isXq10,N/Ot2 , g ,gy 4 37 = S3' (1/4 T) . SSI IusTRWENT LOCATION CCRRECTION DCLUDED
. . GUIDE 1.90 REY. 2 I
- -# - 3 33 l OPERATE TO E2tMT AfD/CE BELOW LIMITUG LINES * IEICATES BOTE EEATUF AND fvrwarm EATE PREBREE AND TEMPERATURE INTERSECTItlME NOTED ST PARDTEESIS
orEnATIce In caoss-sArcusD AaEA PEnnIrrED esLT uuEn mTEm utvEL u wITuIs sammL aAsoE ran scmEn orERATIO BRUNSWICK : UNIT 2 3/4 4-16 Amendment No.172-
, ..._. __ _ . . _ _ ._ . . , .. ._ _ - _ - - ..
. - . - . - . .-.- _-.-- -~--._.-_ _ ._ - -
- - - .
. . . .,
i l f
- g *
l
4
. 1 i
I i .
1 .
- FIGURE 3.4.6.1-3a j PRESSURE-TEMPERATURE LIMITS ly REACTOR VESSEL
~
{ HYDROSTATIC AND LEAK TESTS
,
'
,
1200 . , . , , . . . . 6 . , . . . . ,
e e i i . , e e i . . , . , e e ,
, i t . e i i e i e . , e e i a e * , i e , e 6 1 4 , .. . ,
a ! # . i e s I i t , . i , i 6 e i , e , , . . . . i .
i
i t e a e e i . , i , , e e t e t i i . I i 6 i .a . , .
j . . . 4 . . ..i i , e . . 4 . .
6 . . i . . . . . . . . .
,g+[r f ,.no ,,
,
i . . , . . , i . . e , , , i . . e , , , , . . . . . .
{ e a i e i ! , 6 6 , 4 i e l e 6 . e i 6 e 6 . i , e . , %f fi
, s , e i + 6 e a e i e . e i a e i . 6 i , e * t r.
- , , , , 6 , , , , , , , . . . . . . . . . . I t ,. a j i , 6 i + i e 6 e ee i e 6 e 4 a e (10U31 7 , , .
- i . t e 6 . 4 . , . . . , . , , , n i e , i , I i e i 6 # , . 6 , e . N
.' 4 6 , 6 . 6 e i , , . , e , . d""T"' "
9 .
. . i i 6 . 6e 4 e i 4 I . , , i . . .
- .
4 i
. . e i a e 6 . 4 . . . , e ta . . . e e 6 e i et . gw ,
.,
>
900 , . . . . . , , . . , i 6 . . . . . , , . . . , . . . r- ,
a e' , t . 6 i i t i . .
e e i s , i e 6 6 e e d; ,
2r l
.
1 4 6 , , e ,
i , . . e i e i ;" . m 6 l i e + 6 i e t i 6 I f 6 , 4 6 I i e ,
'
"I'" .
-
6 6 i e a i e i 6 i 6 . e . . 6 , 6 i g i ygt g ii t
' 900 e ,
- . , . . , . , . . e i . . . . . , . . . . . , , , . , n .f g e l e i . e . .i e i i s , 6 i , , , 4 i s , e 6 t e e e e , H i f . . i .
.
. e i 6 ep e , e i i . 4 , ,i, i 6 . . e . . . . J . * r, / . . i
! ,
- 6 i . . 4 a e i i . , t ., 6 i + , , . . . , 4 , it f j g }i f .
, e i t e e i 6 4 i a e e i 6 .4 4 6 s 6 6 e a e i , , , e t if e a e .
{.
, , . . . i i i ! i t e , , i . . t . . i i f e .
A e i i s i e e e i ! , e ,! i i t ej i i , e e# j.,,
d
,0. ' ! e e i e . e i . t .6 e t i e i ! , , . . a ## se a e . . .
d
. . ! e e a t e * i ,# # e . 4 e i 6 , . . i y . . I i . . . .
)
! g g
,00 o
6 i . .
i
,
- 6
. +i e e i e
, ,
e6
, ,
e I i 6
.
,
,
..
,
,
,
i 6
,,
, ,
l g
,
i
,D,.
.f yS i
I e
,
,
6 i , .
, , , ,
4 . . e
. . .
,
l I 6
} .
-.
g' . , a 4 .
4 e + t . , 6 #p
't . . . . . . . . .
g . ## i e ii f i .6 . + 6 . . . e . . .
j g
.
g
='
. . .
. t . .
i e
,
e I
. e
. . . .
i
.
a 6 i
4 . s e#
mp g ~:
l t
' e . . .
. . . . , e
. . e
.
e
.
.
.
+
.
...
.
j i . i . . . 4 6 e i i 4 , ;,,.sFi _
e e i i , . e . . .i f*. *
. . . . . 6 , , (4= , -m - 6 , i i 6 t . . . .. . .
l q t e i e e e i 6
, . I e . 6 l , e . i . . . . ) , . . . , e rel ! 6 4 6 i e 6 . . i e i e . .
i i 6 . t t . i e 2 . ! , . 6 , * , e i ' '
t . i l # + 4 4 . 4 , 6 . e 6 e i s . r
, i 6 6 . i s . . . 4 i , i . . .' . .
- i , . . . e, * i e , # , a . . . 6 4 .
1 . . . . . 6 6 i e i e .
4 l 1 e
"
300 '
(299) eMi' l l l . ,!l'l .
, . . . . , , , ii , . .
6 . . 6 ,, . 6 , , . . .
I . . .
i i i , 6 i , a 6 i 6 l- 200 , , i . . . . . . .
. . . .
i . . .
< ' . .
100 d , ,
,
,
e
. . I
'
, , i 6 . . .
, . , , .
i , i i
. , , , . . i . . tto 120 130 140 150 1G0 170 100 l
TEMPERATURE U F1 !
B&EE ! FUEL IN REAC1tR REAC1tR NOT CRITICAL . ,
3.* REG. GUIDE 1.90 REY. 2 6, < 8 EFFT I I.SX10 N/Q [ > 1 M V I RT = 77' (1/6 T) l ISSI INSTRWWif 14 CATION CORRECTION INCLUDED EEL
, OPERATE TO RIGET AND/QR BE14W LDtITfMG LINES
'V " * INDICATES BOTE EIATUP AND r=rrw nrmar garg PREsstRE AND TEMPERATURE INTERSECTIONS NOTED SY PARENTEESES 6 oFERATINs LIMIT INDICATES TDfERATURE REQUIRED IF TEST FRESSURE WAS EECEEDE BRUNSWICK - UNIT 2 3/4 4-17 Amendment No. 172 ;
.
' _ _ _ _ _ _ - _ _ . - . . _ _ _ - _ -_ _ _ _ . ~ - . _ _ . _ _ _ _ _ _ _ . _ . _ _ . . _ .
. .
FIGURE 3.4.6.1-3b PRESSURE-TEMPERATURE LIMITS e REACTOR VESSEL
,
,
l
.
HYDROSTATIC AND LEAK TESTS ,
s t
1200 . .., . . . , . . . ... . , . . , , . . . , , . . . . . ., . .
4 i i i . . i t . . t , 6 . . i . ! , i . . , , , . . .
- , , t . , . . . , t i i e I . i I e i , . . 4 , . . . . .
i . . , e ea
. , , . i . ie ,
s i i + t . . . . . .i . . . . -
- -
i i , . , , . . . . . . 4 . ..ie 4 4 . . . .. sg ft-~
,V, 8. f
,
+
1100 , , , , , , , , , 6 i . ,i, , i . . . . .
j i i t .
I , i . .i6 i , t ! e i If (1G-U)
- . , 6 i i i 1 i . . i OY / . .
j , t . , . . e i e i( . O G5 s i # . . . . i ;
s
. 4 i i . e i e i . . e . . . . ,ri .
! 1000 , , , , , i i , , , , . . . , , , e .. . , , , .
- i -
,
,
, , , i i i
e i i . . . . i i , 4
,
. .
j . . e, . .
, , . , ,ie , , . . 6 . . ,, . cs .
- . i i . ; 9 . . . i 6 , i . . . . i . . .9
,
' ' ' ' '
' ' ' '
. .
'
q
'
.
j N..a
'
900 l , . ; , l . l'l i , . . rJ "
,
.
i i
s i i i ,
i
!
.
,
i
, ,
i
, .
. .
a . , a
.
, i
. .
- i , , , e , i e i . i . . - . .
! . . 6 . . 4 6 6 . 6 i . . e . . . 1 . . ;. G n
! S00 . , ,, i , ... , , , , , , , , ,,.. , , i , i , i i , n ,,.f . J
,
i , . ..i 6 . i i i e i 6 6# t . i a e i i i . A 4 if i . .
i i 6 i , 6 ,,4 i , + .. . 64 . . .- f . .
.
6 , , . i i 6 i i i i . . . .. 6 . i , 6 trsu) . . > +
) e . i e . . . 1 i i . . . . . i . . . . i . p i- -. 4 e i i . . , ... . . .i,, , , , . .
. . .
j e i . +,e e i i i i . . . i 6 i . . , , . . ;i,. . . i*. .
. . . . . .. , ,,. i , . . 6 e i a . i . f . . . . . . . .
.
3 i 4 6 . 4 . 6 6 . . . . .... .6 , 4 i i , # . .; ri . . i i . . . . .
!
w
E00
' ' '
' , ',
.
.6 e
'
.
'
i
,
'
,
.
'
i
.
' ' '
.
. ,
4 .
' ' ', 'i
,.
,.6 .
' ' ' '
i , . i i , i
' ' '
e i
. . .
'
ef sn i .
' ' ','
i '
,
. . .
.
'
.
. .
. . .
.' .
'
.
. .
, g . ..
6 6 ,
6 i , .
i e i i i i
. . .
4 . . 6
, i , !
i ,
i .
,
. #.
if i /g ! , i i i i ,
, ,
. .
i
.
. . .
, . . . . .
.
-
W It 500
.
, ,
t
. . .
, , ,
. 6 .
. 4 . I
, ,
+4
.
i . . i
, , . .
4 . . ._ . .
, , .f
. .
t i .
f
.
, , g [i ,
e i ,i q #
,
.6 e i
, . , ,
i . i 6 i i . .
, , , .
6 i , .
. . .
. . .
. . .
. . . .
. .
'>
y e
i,
.,
. i i i i i i i e i , i
, p i #
i . . .
- ,
. gG . !
i
.
, ,
i 6
.
i
. . . i i i , e
. 6 6 ,
. . . .
. . .
. . . . .
i i . . i i . i , i . i i . .
ii e 4 i x e i , 6 . . . . .
400 , 6 , , , w i . , , , , , , , , , , , . , , , , , . .
. . ,
i 4 6 i i 6 i . . i , i . , 6 . i . . . i 4
, , , 1 i i , 6i . 4 i + . . . . . . . , , . , . . .
i i . , i 6 i i t 6 . i , e i i , 6 . . 6 . , . . . .
i e i i i . i , . . . . i i i . . . . r .
4 6 ) i
, iii . . . . , , t , , , . , . . . . .
g gg) i i ipi m#, , i . 4 i e e i . , . , . . , . , .
. . , , i i . 6 , , . . . .
i i i +, . , i i . . . . .
-
' ' ' ' ' ' ' ' ' ' ' ' ' ' '
200 H
_ ! , i i . , i . . , , . .
_$ , . . , i . t , i . . ,
i i 6 . . . . . 6 . .
, i . . , i i , .
_L e i i , i . , 6 .
100 . . . # . i ,
. i , , . .
i i . . ..
g , i . t .
l l
! ' t , t , l , iI . i ' '
70 80 90 100 110 120 130 140 -150 160 170 180 1:
TEMPERATu'tE (*F)
IME . FUEL IN REACT m , 10 % 2 , g ,gy .4 X 10 N o /Q1 4 RT = 82 (1/4 T) SI INST 173ENT LOCATItal CORRECTION INCLUDED RES. GUIDE 1.99 REY. 2 REACTM NOT CRITICAL EE C.# OPERATE TO RIGBT AND/CE BELOW LIMITING LINES * INDICATES BOTE BEATUP AND l'M M8 RATE
's . FRESSURE AND TEMPERATURE INTERSECTIONS NOTED 31 FARENTEESES j 4 - CFERATING LIMIT INDICATES TEMPERATURE REQUIRED IF TEST FRESSURE WAS EXCEEDE i BRUNSWICK - UNIT 2 3/4 4-18 Amendwnt No.172 L
._ _ _ , - - ,__ _ _ _ _ . , r
-
.
.
y .
{-
. ,
1
FIGURE 3.4.6.1 3c i
PRESSURE-TEMPERATURE LIMITS i
. . REACTOR VESSEL '
1
- -
HYDROSTATIC AND LEAK TESTS 1200 , , , , , , , , ... , . , ,. . ... . . . c . . .
i i . 4 i . ii ,i, i e .. , .. i,.. . , , , , ,
, . , , , , .. . , . . . . . . . . ;
, . , . . . * *i i ii ,i ...i . . , , ; . . ,
'
1100.-
l' ' '
,' *
. .
' '
l
,
ll '
ll' l,'l . ll' lll,sp,@ _ _
j e i i i.,i ...i
]. i ve i
i # .
ii
, . . . . i.,i gf .
-
, .
i i i . + ovacer sr > .
. i i i .
1000 ,
,
e , . ,. . , , .
j , , , i , , . .... , , . . , ,.
,
,, , ,
i i , ...
i e- i i
. , , , , ,
'
- . , m i . . . . . . , e i . . i ;
j , e i 6 , . 6 . e . . , n .
' ' , , , , . . t i.
4 -
900 * i ,
i . i . 6 e
. . .
'
J
"
P
=
i- i
, , , , i . , i i . .' , , .e i , e i . , 6 , % .
}' # . , i , 4 m , . , .
' '
800 '
'
' ' ' ' '
l
-
,' ' ' '
, e
.
. . . ,- , e i e . . . . .' . w .' .;
} i . , , . i .. e , 6 . . i . J jr
. . i , 6 , , i ii 6 e , . . e , , . js' ( ,y
} .. . * * , * 6 4 i ,i , , i . i (/Uaj jr .
j ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' ' '
0 ' .'
e
{
. -
e i ,
-. i
,
i
.
e
e i .
, .
i .
,
i i
i
.
i
. .
. . .
,
' . , . ' '.', '.'
. . .
. yi
' f ' '
. , , ,
.
,
{ j i ,
. , ,
. . .e i . i i , , , i . . , , . . . .
. . . . . . , . , , i 4 . .. . . . . ,r . . ,
.
, .. . .
. . , , 6 . 4 . 6 . . . i s . 4 4 6 i .. , , . . . ,
u, 600 , , , , , , , ,, , , ,,, , , , , ,,
. . .sr
,, , , , ,,
.
,
,
i
! E i i , i . i 6 . i , , i 6 i 6 ..e ,.i . . . , , , .
- *
3 . . i 4 i i i i i , , i# , i ,.. . ..
a
- ,g'
. , , , 4 i , .. . i i i ,r1 . e . . . ..
! * ' ' ' ' ' ' ' ' '
- , 500 ', ,' . '.
, . . .
- i!
. 6 , .
.!
.
i
. .
e i if i 6 0,' / ,,, , , i ..
.
4 * *
. . . 6 .. . ,
. .
, i m ,
.
i i i 3 i 4 . .
i , .
i . .
. . .
....
....
. .
.
,
e 6 i . . i ,m.- i ,
T gor 4 '
(410) ' ' '
'
i .
'
,
' ' '
i , i i , .... . . .
400 .
', , i
~
,
'
',!' g' CM i i .
., ' .' '
. . '. ... .
' ' '
. , , 2 e i i i , . .ie i6 i , . . i
- i , , e i, . . . . . . . . , . .
f 6 i 4 , i s , 4 i , , . +, i ... . .
' ' ' ~ ' ' ' ' ' '
'
l ,
', , . . , . . ,
,', i ', i . . .. ,
,. g gg) i , i . i i ri_ ma i4 . . . .
- _ e., i , , .. i , i . . .
- ;e i i , 4 i,, i ... . .
,, i i... . .
i 200 -j , 6 i ' ' ' '
s . .' . i .' i' .' ' i . .
i ,, . 6 .. . . .
'
, .i 6 . .
,t ii . 4 .
100
'
, ., , .
i e , .
. t . , .
'
, 6 l , , ,
i , , , . . . . .
go' f40 150 160 170 180 19C i
.
70 e6 100 110 120 130
.
TEMPE:RATURE (* F)
! E0EEI.L j FUEL IN REACTCR l- I12Egy 4- S.3 X 10 N/Of > 1 DEV i
'
4 RT = SS' (1/4 T) I INSTRUPENT LOCATION CORRECTION INCLUDED
1 REG. GITIDE 1.De REY. 2 REACTCE NOT (2ITICAL
m f p.,, OPERATE TO RIGET AND/CR RELOW LDfITING LINES j 'iy - * IEICATES BOTE EEATUF AND cmmsa! RAIE
- PRESSURE AIS TEMPERATURE INTERSICTIONS NOTED ST PARENTIESEE
- 6 oFIRATING Lact INDICArES TInrERATURE REQIUM IF TEST FRES$URE WAS EECEEDID.
- BRUNSWICK - UNIT 2 3/4 4-19 Amendment No. 172 l - _ - . . -_ _ _ _ _ _ ., _ - . _ _
_ _
>
_
l 3/ EMERGENCY CO'RE COOLIN'3 SYSTEMS l
3/4. HIGH PRESSURE COOLANT INJECTION SYSTEM l
LIMITING CONDITION FOR OPERATION
__
3. The High Pressure Coolant Injection (HPCI) system shall be OPERABLE with:
l One OPERABLE high pressure coolant injection pump, and ' An OPERABLE flow path capable of taking suction from the suppression I pool and transferring the water to the pressure vesse l i
APPLICABILITY: CONDITIONS 1, 2, and 3 with reactor vessel steam dome pressure '
greater than 150 psi l ACTION:
l With the HPCI system inoperable, POWER OPERATION may continue provided l the ADS, CSS, and LPCI systems are OPERABLE; restore the inoperable HPCI system to OPERABLE status within 14 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> i i With the surveillance requirements of Specification 4.5.1 not perfomed at the required frequencies due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applicable provided the
'
appropriate surveillance is perfomed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after reactor steam pressure is adequate to perfom the test SURVEILLANCE REOUIREMENTS 4l5.1 The HPCI shkil be demonstrated OPERABLE: At least once per 31 days by: Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with wate BRUNSWICK - UNIT 2 3/4 5-1 Amendment No.188
.
k
q "O '
L:
ENERCENCY CORE COOLING SYSTEMS .
SURVEILLANCE RE0UIREMENTS-(Continued) Verifying that each valve (manual, power-operated, or automatic)
in the flow path that is not locked, sealed, or otherwise secured in position, is -in its correct positio h. At least once per 92 days, by verifying that the system develops a flow of at least 4250 gym for a system head' corresponding to a i reactor pressure 11000 peig when steam is being supplied. to the turbine at 1000, +20. -18, pei , At least once per 18-months byt t'
.
' performing a system functional test which includes . simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each 'automatie valve in the flow path actuates to its correct position. Actual in',ection of coolant into the reactor vessel is excluded from - I this tes .- Verifying that the system develops a flow of at least 4250 gpa for a system head corresponding to a reactor pressure of > 165 ,
psig when steam is being supplied to the turbine at' 165,3,15, l pei l
- Verifying that the suction for the 'HPCI systes'is automatically transferred from the condensate storage tank to the suppression l
,
pool on a condensate storage tank low water level signal or '
suppression pool high water level signa i
.
.
i l
5 BRUNSWICK - UNIT 2 3/4 5-2 RETYPED TECH. SPEC ,, a . .a .n. . ... a . 7, i- -
~- - -- - _ __ _
, _. ,__ _ - -, _ _ _ _ _. _ . _ , ._ _._ __. _ _._ _ _ _ _ __, _ .._ _______ ___ __
,/'4* ,
i
<
l EMERCENCY CORE COOLING SYSTEMS <
t
%. ,
-
3/4.5.2- AUTOMATIC DEPRESSURIZATION SYSTEM -
,
,
LIMITING CONDITION FOR OPERATION ,
.
352 The Automatic Depressurization System (ADS) shall be OPERABLE with at
. least seven OPERABLE ADS valve APPLICABILITY: CONDITIONS 1, 2, and 3 with reactor vessel steam done pressure j
.) 113 pei ,
ACTIONt 'With one ' ADS valve inoperable, POWER OPERATION may continue provided the NPCI; CSS, and LPCI systems are OPERABLE; restore the inoperable ADS valve to OPERABLE status within 14 days or be in at least HOT SNUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . With two or more ADS valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SH'JrDOWN within the next 24-hour ' With the Surveillance Requirement of Specification 4 5.2.b not performed at the required interval 'due to low reactor steam pressur the provisions of Specification 4.0.4 are not applicable provided the ,
appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter reactor
'
,,
vessel steam pressure is adequate to perform the test j SURVEILLANCE REOUIREMENTS 4. The ADS shall he demonstrated OPERABLE at least once per 18 months by:
- Performing a system functional test which includes ' simulated ,
automatic actuation of the' system throughout . its emergency operating sequence, but excluding actual valve actuatio Manus 11y opening each ADS valve when the reactor steata dome pressure is 1100 psig and observing that either; The control valve or bypass valve. position responds accordlur,1) .
or Thete is a corresponding change in the measured steam flo j l
a BRUNSWICK - UNIT 2 3/4 5-3 RETYPED TECH. SPECS. _ ,
-
-.~,.-.x_s_---w..-- ~ ~ . - -
.l S2 ' ' i
? !
j
> ,
1 l 4 i
$
' 5'-
,
EMERGENCY CCRE C00LINC SYSTEMS i
d j_ '. 3/4.5.3 LOW PRESSilRE- C00LINC SYSTEMS .
.
CORE SPRAY SYSTEM i,
. .
.
j j LIMITING CONDITION FOR OPERATION :
i i
3.5.3.1- Two independent Core Spray System (CSS) subsystems shall be OPERABLE with each subsystem comprised of: One pump, and I
,
i An OPERABLE flow path capable of taking suction from at least one of j the following OPERABLE sources and transferring the water through the
- spray sparger to the reactor vessels i i
! In OPERATIONAL CONDITION 1, 2, or 3, from the suppression pool.
t l l In OPERATIONAL CONDITION 4 or 5*:
l
a) From the suppression pool, or ;
.
'
b) When the suppression pool is inoperable, from the
!- condensate-storage tank cocc. tining at least 150,000 gallons j of water.
i l- APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4, and 5*. '
ACTION:
i l In OPERATIONAL CONDITION 1, 2, or 3:
,
i With one CSS subsystem inoperable, POWER OPERATION may continue !
- provided both LPCI subsystems are OPERA 8LEl restore the j inoperable CSS subsystem to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD j SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
.
i With both CSS subsystems inoperable, be in at least HOT SHUTDOWN i
i within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> l
! l
-
%. .
l
'
- l
- The core spray system is not required to be OPERABLE provided that the l' l i reactor vessel head is removed and the cavity is flooded, the spent fuel '
,
pool gates are removed, and the water level is maintained within the i limits of Specifications 3.9.8 and 3. i
ss
)
i BRUNSWICK - UNIT 2 3/4 5-4 Amendment No. 127
! i
_ , ._
. _ _ _ . _ . ,. . _ . . _
_ . . . . . . . . . .
-.-._.-. - . - - - . - _ . . - - - _ - ~ . - - . - - . . ~ . - - - - . - - . .
i
u) .
a, 1- .
i jeN EMERCENCY CORE COOLING SYSTEMS
! LIMITING CONDITION *FOR OPERATION (Continued)
i ACTION (Continued)
!
{- In the event:the CSS is actuated and injects water into the
}
reactor coolant system, a Special Report shall be prepared and i submitted to the Commission pursuant to Specification 6.9.2
{ within 90 days describing the circumstances of the actuation and ,
j the total accumulated actuation cycles to dat '
I
" In OPERATIONAL CONDITION 4 or 5*:
i.
} With one CSS subsystem inoperable, operation may continue provided that at least ces LPCI subsystem is OPERABLE within 4
'
,
i houest otherwise, suspend all operations that have a potential for draining the reactor vessel.
{
j With both CSS subsystems inoperable, operation may continue provided that at least one LPCI subsystem is OPERABLE and both *
LPCI subsystems are OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Otherwise, suspend j' all operations that have a potential for draining the reactor vessel and verify that at least one LPCI subsystem is OPERABLE within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i f*
j :.. The provisions of Specification 3.0.3 are not applicabl _
! SURVEILLANCE REQUIREMENTS I .5. Each CSS subsystem shall be demonstrated OPERABLE: 4 l
l At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the condensate storage tank
, minimum required volume when the condensate storage tank is required i to be OPERASL ,
r ;
I
' At least once per 31 days by:
]
i
{ Verifying that the system piping from the pump discharge valve
- to the system isolation valve is filled with water.
i *
i j * The core spray. system is-not required to.be OPERABLE provided that the 4 reactor vessel head is removed and the cavity is flooded, the spent fuel l
'
pool gates are removed, and the water level is maintained within the !
,
. limits of Specifications 3.9.8 and 3. .
/
\;;p
'
BRUNSWICK - UNIT 2 3/4 5-5 Amendment No. 127
. .. .- _ . . . - - - . - _ . . _ . _ - - . - . ~ . __ - ... , - ; U -
.
. _ _. . . .. . . . . - . . . . - . . . . -
l
,
i l I
< + . ..
l EMERCENCY CORE CDOLINC SYSTEMS !
- I
SURVEILLANCE REQUIREMENTS (Continued) !
, Verifying that each valve (manual, power-operated, or automatic)
in the flow path that is not locked, sealed, or otherwise
secured in position, is in its correct positio At least once per 92 days by l
i
- Verifying that each CSS pump can be started from the control
!
,
room and develops a flow of at least 4625 spe on -ecirculation flow against a system head corresponding to a reactor vessel
- pressure of > 113 psi ,,
l Performing a CHANNEL CALIBRATION'of the core spray header AP
instrumentation and verifying,the setpoint to be 5, +1.5, paid l
) greater than the normal indicated A '
' At least once per 18 months by performing a system functional test l vhich includes simulated automatic actuati'on of the system throughout {
its emergency operating sequence and verifying that each automatic I valve in the flow path actuates to its correct position. Actual ;
injection of coolant into the reactor vessel is excluded from this J
'
' test.
i
.
.
.
.
i
.
.
!
i ,:7 BRUNSWICK - UNIT 2 3/4 5-6 A==ad==nt No. 160
_ .__ _ _. . _ _ . __ _ _ _ _ - - - . - _ . _ . . _ _ _ _ . . _ m _ .
,. cp .
'
o EMERGENCY CORE COOLING SYSTEMS LOW PRESSURE COOLANT INJECTION SYSTEM LIMITING CONDITION POR OPERATION
-_ _ _ .
3.5.1.2 Two independent Low Pressure Coolant Injection (LPCI) subsystems of ;
the residual heat removal system shall be OPERABLE with each subsystem l comprised of: '
i TVo pump I ,An OPERABLE flow path capable of taking suction from the suppression pool and transferring the water to the reactor pressure vesse j APFLICABILITY: CONDITIONS 1, 2, 3, 4*, and 5*.
ACTION: In CONDITION 1, 2, or 3:
~ With one LPCI subsystem or one LPCI pump inoperable, POWER l OPERATION may continue provided both CSS subsystems are OPERABLE; restore the inoperable LPCI subsystem or pump to ;
OPERABLE status within 7 days or be in at least HOT SHUTDOWN j within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> , With both LPCI subsystems inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> ! With the LPCI system cross-tie valve open or power not removed from the valve operator, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6. within 90 days describing the circumstances of the actuation and the total accumulated actuation cycles to dat ; In CONDITION 4* or 5* with one or more LPCI subsystems inoperable, take the ACTION required by Specification 3.5. The provisions of Specification 3.0.3 are not applicabl l
!
- Not applicable when two CSS subsystems are OPERABLE per Specification
!-
1.5. .
BRUNSWICK - UNIT 2 3/4 5-7
RETYPED TECH. SPEC Updated Thru. Amend. 78
.
. ~ . - - -- . -. - --_ .- . . . -. . - - - - -- . - . - - . . -
. Cp .
- .
,
'
'
,
l
!
EMERCENCY CORE COOLING SYSTEMS
-
-
)
SURVEILLANCE REQUIREMENTS
.
- -
l l 4.5. Each LPCT absystem shall be demonstrated OPERABLE: - ;
l l At least once per 31 days by:
l Verifying that the system piping from the pump discharge valva to the system isolation valve is filled with water,
' Verifying that each valve (manual, power-operated, or automatic)
in the flow path that is not locked, sealed, or otherwise F secured in position, is in its correct position, and
,
5 Verifying that the subsystem cross-tie valve is closed with I power removed from the valve operato l
,' At least once per 92 days by verifying each pair of LPCI pumps discharging to a comanon header can be started from the control room and develops a total flow of at least 17,000 gym against a system-head corresponding to a reactor vessel pressure of > 20 psi At least once per 18 months by performing a system functional test ;
'
which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position. Actual injection of coolant into the reactor vessel is excluded from this tes ,
,.
em
[
l
'
l l
I RRUNSWICK - UNIT 2 3/4 5-8
'
RETYPED TECH. SPEC l i
__ __
- 'O 3/4.8 ELECTRICAL POWER SYSTEMS . .
l I /4.8.1 A.C. SOURCES ,
! ,
i j ,OPERATINC~
l 1.IMITING CONDITION- FOR OPERATION i
'
herh + -es a minimum,'the following A.C. electrical power sources shall be j- OPERABLE: Two physically independent circuits , per unit , between the offsite l
'
transmission network and the onsite Class 1E distribution system, and
!
! Four separate and independent diesel generators, each with:
!
l A separate engine-mounted fuel tank containing a minimum of 100 1
,
- gallons of fuel,
- A separate day fuel tank containing a minimum of 22,650 gallons
- of fuel, and
$ A separate fuel transfer pum A plant fuel storage tank containing a minimum of.74,000 gallons of fue APPLICABILITY: OPERATIONAL CONDITIONS 1,'2, and ACTION With one offsite circuit of the above required A.C. electrical powg
' sources not capable of supplying the Class IE distribution system:- Demonstrate the OPERABILITY of the remaining A.C. offsite source by performing Surveillance Requirement 4.8.1.1.1.a within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter; Demonstrate the OPERABILITY of the diesel generators by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafteri
) Restore the inoperable offsite circuit to OPERABLE status within !
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> !
i
" With Unit 1 in OPERATIONAL CONDITION 4 or 5 and one of the required Unit 1 offsite power circuits not capable of supplying the Unit 1 Class 1E distribution system, either restore the inoperable Unit 1 offsite circuit i to OPERABLE status within 45 days or place Unit 2 in at least HOT SHUTDOWN within the nex,t 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The provisions of ACTIONS 3.8.1.1.a.1, 3.8.1.1.a.2, and 3.8.1.1.a.3 are not applicabl BRUNSWICK - UNIT 2 3/4 8-1 Amendment No. 176 '
i
- - . , - - - - . . - . . -..- -
.. , .-
.,
_ . . . _ . . _ . . . _ . _ . _ . _ . - _ . . . . _. _.- _.... . .. _ __ . ~ . . _ . _ . . . . _ _ _ . _ _ _ _ . _
'
,
1 .
>
h...l'
, ELECTRICAL POWER SYSTEMS ,
'
i . LIMITING CONDITION FOR OPERATION (Continued)
ACTION (Continued):
, With a diesel gengrator of the above required A.C. electrical power source inoperable : ,
i j ' I
!
,
i i
5 P i
I
.
Ir.
- v .
'
lt
!
,
i i
'
.
f
.
.
a
1 4 !
j
.
' A diesel generator shall be considered to be inoperable f rom the time of il failure until it satisfies the requirements of Surveillance Requirements- l
6 4.8.1.1.2.a.4 and 4.8.1.1.2. i
! BRUNSWICK - UNIT 2 3/4 8-la Amendment No. 176 I4 . . . . __ _ . . . . _ _ . _ _ __ _ .
g Cr
'
-
.
.
. .
.
% ELECTRICAL POWER SYSTEMS '
LIMITING CONDITI'ON FOR OPERATION (Continued)
.
ACTION (Continued) Demonstrate.the OPERABILITY of the A.C. offsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter; 4 Demonstrate the OPERASILITY of the remaining diesel generators by performing Surveillance Requirements 4.8.1.1.2.a.4 and
-
4.8.1.1.2.a.5 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> thereafter; Restore the inoperable diesel generator to OPERABLE reatus within 7 days or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
. With one offsite circuit and one diesel generator cf the above
- required A.C. electrical power sources inoperable
- Demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirements 4.8.1.1.1.a, 4.8.1.1.2.a.4, i and 4.8.1.1.2.a.5 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> lj thereafter;
f"- Restore at least one of the inoperable sources to OPERABLE
status withip 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or be in at least HOT SHUTDOWN'vithin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 l hours; i
l
' With the inoperable offsite A.C. power source restored, demonstrate the OPERASILITY of the remaining A.C. power sources at required by ACTION b; restore four diesel generators to
', OPERABLE status within 7 days from time of initial loss or be in f
at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD
- SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; l With the inoperable diesel generator restored, demonstrate the OPERABILITY of the remaining A.C. power sources as required by ACTION ai restore two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> from cima of initial loss or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> . With two of the above required offsite A.C. tircuits inoperable
- Demonstrate the OPERABILITY of four diesel generators by performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 within two hours and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
', thereafter, unless the diesel generators are already operating; v
BRUNSWICK - UNIT 2 3/4 8-2 Amendment N . _ ___._ _ _._.-._ _ _.-.___.___. _ . . . _ _ - _ . _ . . _.
.i* .
N
. -
. .
I ch'
'
ELECTRICAL POWER SYSTEMS
'
LIMITINC CONDITION FOR OPERATION (Continued)
l ACTION (Continued)
<
l Restore at least one of the inoperable offsite sources to.
OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT SHUTDOWN i within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;.
- . With one offsite source restored, demonstrate the OPERASILITY of
the remaining A.C. power sources as required by ACTION a;
'
restore two offsite circuits to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />
"
from time of initial loss or be in at least HOT SHUTDOWN within
,
the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24
, hours.
'
t
!. With two of the above required diesel generators inoperable:
!
i Demonstrate the OPERABILITY of the remaining A.C. power sources by performing Surveillance Requirements 4.8.1.1.1.a,.
j 4.8.1.1.2.a.4, and 4.8.1.1.2.a.5 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least
- once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter; !
i i < Restore at least three diesel generators.co 0PERABLE status i
!
' <.. within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least NOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />;
-
- 1 l With one diesel generator restored, demonstrate the OPERASILITY l i of the remaining A.C. power sources as required by ACTION b; ,
i restore at least 4 diesel generators to OPERABLE status within 7 l days from time of initial loss or be in at least HOT SKUTDOWN j
, within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the ;
'
following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
i t
i i SURVEILLANCE REQUIREMENTS i
4.8.1.1.1 Each of the above required independent circuits between the offsite i transmissica network and the onsite Class 1E distribution system shall be:
1,
! Determined OPERABLE at least once per 7 days by verifying correct j breaker alignments and indicated power availability, and i
j Demonstrated OPERABLE at least once per 18. months during shutdown by
manually transferring unit power supply from the normal circuit to
- the alternate circui $
ig
. La-I
- BRUNSWICK - UNI /4 8-3 Amendment No. 54, 13. '
,, _ _ __ _
- _ . _ _ - _ _ _ - _ _ _ _ . . - . _ _ - . . . _ . . -_-__.__.-._.m.mm_-_ -_
3 . 4, .
II.
a
- '
', i ELECTRICAL POWER SYSTEMS SURVEILLANCE gervIpmmTS (Continued) 1 i
i
~
,
4 . 8 ' 1' .1. 2 Each' diesel generator shall be demonstrated OPERABLE:
.
! i At least once per 31 days on a STAG 3ERED TEST BASIS by:
{
. Verifying the fuel level in the engine-mounted fuel tank, . !
2 Verifying the fuel level in the day fuel tank,
$ Verifying the fuel transfer. pump can be started and transfers j fuel.from the day tank to the engine mounted tank, , i
, -
,
Verifying the diesel, starts and accelerates to at least 514 rpm
~
l 4.
- -
in less than or equal to 10 seconds,* Verifying the generator is synchronized,' loaded to greater than or equal to 1750 kw, and operates for greater than or equal to 15 j minutes, and l) Verifying the diesel generator is aligned to provide standby
- power to the' associated emergency buses, s
j; At least once per 31 days by verifying the fuel level in the plant
[
fuel storage tank.
k' At least once per 92 days by verifying that a sample of diesel fuel i from the fuel storage tank, obtained in accordance with AS2W-D270-65,
is within the acceptable limits specified in Table 1 of ASTM-D975-74 g when checked for viscosity, water and sediment,
!
' At least once per 18 months during shutdown by:
i
- Subjecting the diesel to an inspection in accordance with
procedures prepared in conjunction with its manufagpurer's i
)
recommendations for this class of standby service, l.
I'
' Verifying the generator capability to reject a load equal to one core spray pusy without tripping, i
!
,
i
!
- The diesel generator start (10 seconds) from ambient conditions shall be I performed at least once per 184 days in these surveillance tests. All other l engine starts for the purpose of this surveillance testing may be preceded by
- a manually initiated engine prelube period and/or other warmup procedures I recomunended by the manufacturer so that mechanical stress and wear on the l diesel engine is minimize i-I ** For Cycle 9 only, . the surveillance interval for Technical Specification 1 4.8.1.1.2.d.1 may be extended until November 21, 1991.
i BRUNSWICK - UNIT 2 3/4 8'-4 Amendment No. 185
)
. .
.- -. -. . - , - . - , , , - - - , , -
.
.
,
,
- ** . 1 e' \
-
\
- \
.
-
l
.
.
,
." . .
rs ELECTRICAL POWER SYSTEMS
.
^
SURVEILLANCE REQUIREMENTS (Continued)
I -
- Simulating a loss of offsite power in conjunction with an emergency core cooling' system test signal, and
a) Verifying de-seergization of the emergsacy buses and load shedding from the emergency buse b) Verifying the diesel starts from ambient condition on the
!
,
auto-start signal, energises the emergency buses with permanently connected loads, emergizes the auto-connected l loads through the load sequence relays and operates for greater. than or equal to 5 minutes while its generator is loaded with the emergency loads.
i Verifying that on the emergency co e cooling system test signal, all. diesel generator trips except engine overspeed, generator differential, low lube oil pressure, revsrse power, loss of field and phase overcurrent with voltage restraint, are automatically bypasse . Verifying the diesel generator operates for greater than or
,
equal to 60 minutes while loaded to greater than or equal to ( ,
3500 k . Verifying that the auto-connected loads to each diesel generator do not exceed the 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> rating of 3850 k . Verifying that the automatic load sequence relays are OPERABLE i
{
with each load sequence time within 10% of the required valu .
I l
l l l
.
.
l
- -
~
_ , , _ _ , _ __,
- -
- "
- "I
.-- -. = . - ..
!' ,
l, r
!
l
l l
ATTACRMENT 1 Spatial Area Estimation To estimate the potential spatial area for concern during a hazardous material release, it will be necessary to obtain basic meteorological data and complete Form 1 which is attached to this procedur Obtain Meteorological Data:
1, Obtain the following metebrological data from the BNP on-site meteorological monitoring system and complete Form 1:
, Lower wind direction: (deg.) this is the direction from which the wind is blowin Lower wind velocity: (mph)
If velocity is 0 to 5 mph . . Light If velocity is 5 to 15 mph . . Moderate If velocity is 15 or more . . Strong i Stability type: A B C D E F G (circle one)
If A, B, or C . . . Unstable If D . . . Neutral If E, F, or G , Stable l If the BNP Control Room cannot provide the required information, contact
!
the National Weather Service (763-8331) at the Wilmington International Airpor Identify yourself and briefly describe the emergency which exists. Ask to speak to the Shift Forecaster On Duty, and request the following information from the Wilmington Airport Hourly Observation:
i Wind direction: (deg.) this is the direction from which the wind is blowin Wind velocity: (mph)
!
If velocity is 0 to 5 mph . . Light !
, If velocity is 5 to 15 mph . Moderate
!
If velocity is 15 or more Strong Stability Type: Ask the Forecaster if the atmospheric stability from the surface to 500 feet would be classified as:
(circle one) Unstable l Neutral Stable
'
,
l OAOP-34.0 Rev. 8 Page 10 of 19 l
. .
.
ATTACHMENT 1 (Cont'd) If neither the BNP Control Room or the National Weather Service can be contacted for weather information, use the following to estimate conditions which exist at the accident site:
Wind Direction: Estimate the direction where north should b Face that direction and determine from which direction the wind is COMING FROM (to your right will be East. .
your left will be West.. from the front will be North..' from the back will be South).
(circle one) North Northeast
- East Southeast South Southwest West Northwest Wind Velocity: From the following, estimate the category which best describes the wind velocity:
(circle one)
Light Wind felt on the face; leaves rustle; ordinary wind vanes are move Moderate Leaves and small twigs on trees in constant motion; wind will extend a light fla Strong Raises dust and loose paper; small branches on trees are moved; small trees in leaf begin to swa Stability Class: Use the following to estimate the atmospheric stability affecting the release of gaseous material (circle one) Stable Usually occur at night: Clear skies with calm winds and cool surface temperature. Most likely to occur in early morning (2 a.m. until sunrise); also if you observe rising smoke which is suddenly stopped in vertical ris Neutral Occur during night or day: If skies are cloudy or if there is rain (or rainshowers) then the atmospheric stability is neutral. During evenings from the period of sunset to about 2am; conditions are most likely neutra Unstable Usually during the daylight hours: Skies must be clear or mostly clear, usually during a warm day with winds moderate or strong. Winds tend to be gusty so that smoke dissipates rapidl A0P-34.0 Rev. 8 Page 11 of 19
_
__- . - _ _ . - . _ - __ .
. . __ . .
'
,
W ATTACRMENT 1 (Cont'd)
,
NOTE:
Tables 1 through 3 are used for 25 lbs C1/sec release Assistance from the TSC or engineering may be required for accurate plume spatial extents for other release rate .
With the basic. meteorological information, an estimate of the spatial extent of a plume can now be mad The following should be done:
Unstable ~
If the atmospheric stability is UNSTABLE, use Table 1 of this I
procedur Under the appropriate wind velocity group and the !
'
concentration of concern, find the crosswind distance (in feet) and the downwind distance (in feet or miles) where that area of equal concentration may be foun Example: Moderate wind velocity and you want to find the area j of 10 ppm. Use Table 1, look in the middle group under j Moderate and under the 10 ppm column. At 0.25 miles downwind, the width of the plume of material is 276 feet to the right of the centerline. . .a total of 552 feet wide at 0.25 miles downwind. The plume is a total of 718 feet wide at 1.25 miles. The area of 10 ppm ended between 1.25 and 1.5 miles downwind of the releas Neutral If the atmospheric stability is NEUTRAL, use Table 2 of this procedure. Under the appropriate wind velocity group and the concentration of concern, find the crosswind distence (in feet)
and the downwind distance (in feet or miles) where that area of equal concentration may be foun Example: Strong wind velocity and you want to find the area of 10 pp Use Table 2, look in the right-hand g'toup under Strong and under the 10 ppm column. At 0.25 miles downwind, the width c.f the plume of material is 284 feet to the right of the centerlin .a total of 568 feet wide at 0.25 miles downwind. The plume is a total of 968 feet wide at 1.25 miles. The area of 10 ppm ended between 1.75 and 2.0 miles downwind of the releas Stable If the atmospheric stability is STABLE, use Table 3 of this procedure. Under the appropriate wind velocity group and the concentration of concern, find the crosswind distance (in feet)
and the downwind distance (t. feet or miles) where that area of equal concentration may be foun Example: Light wind eelocity and you want to find the area of 1000 pym. Use Table 3, look in the left hand group under Light and under the 1000 ppm column. At the widett point of the plume, 1.50 to 1.75 miles from the release point, the plume is only 98 feet to the right of the centerline.. . . a total of 196 feet wide. However, the
,
1000 ppm concentration ends between 2.75 and 3.00
!
OAOP-34.0 Rev. 8 Page 12 of 19
- - .
.-
. - ~. . . . .
.
.
?
-ATTACHMENT 1 (Cont'd)
miles from the release point. Under Stable conditions and light winds very small concentrations of gaseous material will travel very far downwind provided no other factors '
act upon the dispersion of the plum *** Critical Notes ***
.
1. Under stable conditions and. light wind speeds, gaseous plumes of released material will travel in concentrated form for great distances, provided
,
' j that no other. influences act upon the plum i
\
- If the wind dots not remain steady in any direction, the plume will not !
remain concentraced at extended distances, but will disperse the material J within a much smaller area. Observation of the wind direction is 1
.
important not only to estimate the downwind areas of concern, but to assist in the determination of potential concentrations of released ;
materia *If there are any close-by structures, large bodies of water or wooded areas to where the release point is located, these influences could locally affect the. wind direction and velocities observe ]
Large objects will direct the wind flow and reduce the velocity. Within a downwind distance ten times the height of a large object, wind direction and velocities will be influenced, thus caution should be exercised in concentration determinatio l The wind direction reported by meteorologists and the Brunswick Control Room personnel is the wind direction FROM where the wind is coming!
Gaseous material released from a source will travel 180' opposite from the wind direction being reported by observation, therefore, in order to have the correct direction for possible actions to be taken, the opposite direction from which the wind is being reported must be use .
t Meteorological data obtained from the on-site meteorological monitoring system or from the National Weather Service will use compass direction " Plant North" is rotated 45' to the east of compass nort !
i l
.
0AOP-34.0 Rev. 8 Page 13 of 19 i
_
.
.
L l1
.
CONTINUOUS GAS RELEASE @ 25 LBS CL/SECOND (3.79 M5/SEC) FOR UNSTABLE ATMOSPilERIC CONDITIONS CROSSWIND DISTANCE FROM CENTERLINE TO EDCE OF ISOPLETli (FEET) FOR VARIOUS CONCENTRATIONS OF CL (ppm) AT SEVERA1. WIND VELOCITIES DOWNWINO LIGHT H0DERATE DISTANCE STRONG Mitf5 FEET 10 1.000 5.000 10.000 10 l1.000 5.000 10.000 10 1.000 5.000 10.000 0.25 1320 335 0 0 0 276 0 0 0 241 0 0 0 0.50 2640 543 407 317 0.75 3960 701 455 282 1.00 5280 821 457 1.25 6600 911 359
.
1.50 7920 973 '
l.75 9240 1010 2.00 10560 1019 2.25 11880 1001 2.50 13200 949 -
2.75 14520 857 3.00 15840 706 3.50 18480 4.00 21120 4.50 23760 5.00 26400 6.00 31680 7.00 36960 8.00 42240 9.00 47250 10 0 52800 1 .0 105600 OAOP-34.0 Rev. 8 Page 14 of 19
_ _ _ _ _ _ . _ . _ _ _ _ _ . - _ _ _ _ _ _ - . _ _ - - _ - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ - .__ _--
lI I l I
"
0 1
0 0 f
.
o
1
_
..
.
S 0 e
.
_
N 0 g
.
.
.
.
O 0 0 a
I '
P
.
T G N
I D O R
N T OS S 0 CU 0 0 0 O
C I 1 *
I R R A E V l
i P R 0 4 4 5 8 4 8 5 S O 8 1 8 0 8 9 6 bM FS 1 2 4 4 5 4 3 1 E
T) I A EI TT L EC 0 A 0 R (F O L 0 0
"U l iV E 1 E T N ED LN
_ R PI 0
_ O F S OW 0 0 0 I L E 5 _
) T A A CFR R E
S OE E D
V D
/EE H
0 2 'M DGS 0
0 E
9 ETA
L .O B 3 T)
A ( n i
I
~ E p 1 1 5 3 2 2 5 6 8 2 2 0 9 D N p 1
4 8 4 5
6
7
7
7
6
6
4
( 1 NI O LRL C
E EC l S T
/ NF 0
L EO 3 0 C C S 0 S MN 1 B OO LRI FT 5 A 0
2 CT ER 0 0
@NN AE T
H E TC G S SN I O I
L A 0 E DC 0 0 0 L
E D 1 R N I
SW A S C O S 0 7 1 0 9 7 6 1 2 0 8 4 0 2 3 1 0 8 6 9 7 4 5 5 3 4 5 0 4 8 8 4 5 5 S R 1 3 5 7
9 0 1 2 3 4 4 5 5 5 5 5 4 0 U C 1 1 1 1 1 1 1 1 1 1 1 1 1 O
U 8 N
I T 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 .
N T
E
2
4 0 0
0
0
0 4 6 8 0 2 4 8 2 6 0 8 6 4 5 0
0 0 v O DE E 3 6 % 2
6 9 2 5
8
2
5
8
4
1
7
4
6 9
2
2
8
2
6 5 e C NC F 1 2 3 6 7 9 1 1 1 1 1 1 2 2 2
3 3 4 4 5 7 0 R
N IWA NT
.
WS OI S
DD E 5 0 5 0 5 0 5 0 5 0 5 0 0 0 0 0 0 0 0 0 0 0 0 3 L 2 5 7 0 2 5 7 0 2 5 7 0 5 5 0 0 0 0 0 -
I O 0 O 1 2 2 2 0_ 0 5 0 P M 1 1 1 2 3 3 4 5 6 7 8 9 1 1 2 O
,
g 4_ A O
( ( ( I l g l _
- _ _ _ _ _ _ _ _ _ . . _ _ _ . - _ - _ _ - - _
- e Tl 3
- *
CONTINUOUS CAS RELEASE @ 25 LBS CL/SECOND (3.79 M'/SEC) FOR STABLE ATMOSPHERIC CONDITIONS CROSSWIND DISTANCE FROM CENTERLINE TO EDGE OF ISOPLETil (FEET) FOR VARIOUS CONCENTRATIONS OF CL (ppm) AT SEVERAL WIND VELOCITIES DOWNWIND LIGHT HDDERATE STRONG DISTANCE mites FEFT 10 1.000 5.000 10.000 10 1.000 5.000 10.000 10 1.000 5.000 13.000 0.25 1320 56 38 29 24 51 30 18 8 49 26 8 0 2640 97 60 38 24 88 42 82 30
_ 0 50 0.75 3960 133 75 37 119 44 111 12 1.00 5280 166 86 18 147 36 136 1.25 6600 197 93 173 159 1.50 1920 226 98 197 180 1.75 9240 254 98 219 199 2.00 10560 280 95 240 217 .
2 25 11880 305 88 260 234 2.50 13200 330 75 279 250 2.75 14520 354 52 297 265 3.00 15840 377 315 278 3 50 18480 420 347 304 4 00 21120 462 377 326 4 50 23760 502 404 345 5.00 26400 540 429 362 6 00 31680 610 474 388 7.00 36960 676 512 406 _
8.00 42240 736 543 414 9.00 47250 792 569 404 1 .0 79700 1060 607 20 0 105600 1212 456 OAOP-34.0 Rev. 8 Page 16 of 19
_ _ . .__ _-. _ - - - _ _ _ - _ _ _ _ . _ _ _ _ _ _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ - - _ _ _ - _ - _ _ - - - _ - . - - _ _ _ _ _ _ - - _ _ _ _ - _ _ - _ _ - - _ _ - _ _ - _ - _ - _ _ - - _ - - _ _ _ _ - _ - _ _ _ _ -
, . . . ~ . . - - .
-
~
.- j
,.
.
.1
<
l ATTACHMENT 1 Page 1 Of 1 PAR Flowchart GENERAL-i EMERGENCY )
( DECLARED l
\
1 f
EVACUATE ZONES A, B C AND 10 MILES DOWNWIND, SHELTER
, REMAINING ZONES 1P CONTINUE ASSESSMENT BASED ON DOSE PROJECTION ACTUAL RELEASE DATA, AND WEATHER CONDITIONS 1 P COMPARE DOSE ASSESSMENT OR MEASURED DOSE AGAINST EPA PAGs TO DETERMINE ADDITIONAL ACTIONS ACTIONS FLOW CHART PROJECTED DOSE (Rem) RECOMMENDED ACTIONS COMMENTS WHOLE BODY (TEDE) < NO ADDITIONAL PROTECTIVE * STATE / COUNTY MAY RECONSIDER PREVICUS AND ACTION RECOMMENDATIONS RECOMMENDED PROTECTIVE ACTION THYROID (CDE) <5 0 REQUIRED * CONTINUE MONITORING / ASSESSMENT OF ENVIRONMENTAL RADIATION LEVELS l
e STATE / COUNTY WILL DETERMINE ACTUAL l PROTECTIVE ACTIONS TO BE TAKEN FOR ]
WHOLE BODY (TEDE)1 OR ABOVE RECOMMENDED MANDATORY THE GENERAL PUBLIC. TAKING ANY EXISTING !
OR EVACUATION OF GENERAL - CONSTRAINTS OR SPECIAL CONS!DERATIONS I THYROID (CDE) $ OR ABOVE PUBLIC FROM THE INTO ACCOUNT. ALTERNATIVE ACTIONS MAY BE l AFFECTED AREA TAKEN BY STATE AND LOCAL AUTHORITIE '
- CONTINUE MONITORING / ASSESSMENT OF ENVIRONMENTAL RADIATION LEVELS l
l
'
!
l
!
l f i i
'
OPEP-02.6.28 Re Page 7 Of 8 t
l L _ , . . _ . . _ _. _ _ _ _ . _ _ . . _ . ..._ _ _ . . _ .- . . . ._ ._ _ . _
. .
s J
l
"
ATTACHMENT 2 i
'
Page 1 of 1 '
Evacuation Zones and Time Estimates /10 Mile EPZ Map
<
I s
WIND FROM . EVACUATE SHELTER ACUATjNS
- ZONES ZONES yp (
< WINTER / SUMMER l I
,
NORTH (338 022) A.B. E.F.G. TO 480 NORTHEAST (C23 067) A,B, E.F G. TO 480 i EAST (068 - 112) A,B,C. F.G. TO 490
$
S A B.C.D.E.F.G,H K
_ C'*(HEAST (113- 157) . 185 TO 500
, '.cUTH (158 - 202) A B.C.E.F G. D 185 TO 685
,
- .. SOUTHWEST (203 - 247) A B.C.G D.E,F 175 TO 685 WEST (248 292) A.8,C. D.E. TO 685 '
NORTHWEST (293 337) A.B. D.E.F. TO 685 ALL ZONES IN 10 MILE EPZ A.B.C.D.E.F,G. TO 695 f
QPinsievel
!: # hn
WP .
\
i se - '% IR '== l
, mw e ;
-
'
e r'
.
'='
( b@j
, ::T i f ,m l 3 g )
" .
!
,
j f ::::
w
.
s
,
~ l m
+
,
N d -
i s- ;=
- W !
< m
/ e
"
l2Y. l=
, , E .
i? @
W -
421
- anacu r+ =i ,@iA d
<t0" ;hlk' fi 4 'Ihj 17?!!!-
-
'
>
I .
%
-
~H #m* !@#h
- n Te e
p g
GTsiiW@W um, u l Sme 1 I a., MDfj$.. ,
() .
n e D e ,s
.s
,
1 1 - :-
..s
.
-
t g N 3 ga s a1 Mw2stm et * tie
<
3 4% %
J
. . ? gMei m m#
.r.ccs.a. s .1
. M.W[
ymy feir #4.
. Shwp M$
'**= s xmams 4gtl1ig 4 gjig ,
i gans.n any -
+ - >
- smggpwn
-
wwe M se ~* e" 4 h is p c! m ,,1 e==JEMJii/MWs4Gesustr T m
l 4_EJi46fGiseTI?Cs
- Yih!TINFj$F$iNjik1 x "
Asi5ive!%iO
4 AA$!2franLM h $$$g$l5F WN
,
if ti nihlbirss %
'Miiks0EM35%;SMed mhi.hkipiis
iTM NNN'"
+
' MEFAFSEFF% h .JA W t
%s&NWesemiGI6~4F.d" ****15iF WT4EMK.yitpiW4gpF2FV
... .
!
9 i
!
I
-
OPEP-02.6.28 Re Page 8 of 8
i i
.
.. . . . - _ - .. . - _ . . . . . . . . - ~ . . .
- . . _ . . . - . . . . - .
, .
1 (
TABLE 1 t
Reactor Pressure Vs Saturation Temperature i
l
'
.
'I c
600 d
-
-
^ ._
, 550 ~r f
o - d "
3 v mus8P'
"
w se-2 9
x 5 0 0 ' _ ._
a , ,
e - -- ' '
x J~
m _ .. ... #
'
Y 450 , l
<
r p w r 'i
.
k a
Z o 900 I
- .r i F-
'
i
.
E g
-
x r i i
.o 350 , !
i
,
x 1 m J l f
w ; r 3 z 300 , 1 C r 1 J r i t
o ,
O ! i
u 250 I b
_ __
-
,
I I I l
- l O se sae 300 600 900 1200
.
!
t REACTOR PRESSURE (PSIG )
!
4
1
BSEP-0/ASSD-02 Page 38 of 144 Rev. 25
, _ _
-
B S
E P
- 2W,8t OOJq2s bW%aw%<ReW .
I
.s =g E w~ T
- / I E A O 2 3
4 s s M S o 0 0 o o S
D Ok G 0 c o o P
-
2
"
_
-
s
,
-
-
.
,'
%
, s
, s
'
'
,'
T I R e
,
P M ' a a
g E t c
e I
~- o N r 3 % _
T 9 C A H w o B o o L
-
f O l E
_ U ' d o
4 R ,
w
S
-
-
s m
!
n P
l o
-
2 ' t
'
- s
_ ,
_ .
~
-
~l
.
s /
'
' r
,' Y .
,
,5
'_
c oI sl oO R l d O*
e
.v pooF e n/
w h
5
.r
.
. .
. ._ _ _ _ _ _ . . . _ _ _ . .- -_ .
,
!
ATTACHMENT 7 Page 1 of 1
.Drywell Te_mperature Calculation Using Remote Shutdown Panel Recorder inputs
!
. Values are obtained from Recorder CAC-TR-778 Above 70' Elevation
,. PT 1_ x0.141 = 'F
-
Between 28' and 45' Elevation PT 3 x 0.404 = F i
!
Between 10' and 23' Elevation PT 4 x 0.455 = F l
Average Drywell Temperature F (Sum of 3 Regional Weighted Areas)
l l- ;
! !
l :
l l
! !
. I i
i-l OAOP-3 Rev.12 Page 152 of 152 I 1:
'
e .-
_ _ - - .
.
..
.
,
FIGURE 4 CORE mal' SHOWING LOCATIONS OF SRM/IU!S
<
.
..: . ,,,
- 1 : . .. . . .,
(_
.
.
.
\ .. . . '
.
53 . , . . , ,-
. . . . . .2 .
.
.. -
a .
a _. . . . . _ .
- .
+.O ++.O+, +9
- .
--
'n'
.
. +e +.
. +.e+. .
, ,
-: :
u
'gi ' , , ' . .'Es, ,- 'Oi '*'- -
..
, ,
_ _ . _. . . _ . .__ _ _. _._ .
__
,
.
n - r . . . .e -
.2.e., : .:.-
L s. e :_ _:. . _:_ .:
.
. _ :. _:. .:. .a,_ .
_
. . . . ,
n- ,m . ,
O , .e ,
, as. . ., .
, ,
r r r T rvrT
,
r .
,e .e:._:.e. e o.:. r .:.e--
--
- .
-
- r - : . _:. ,
,
.,: _ _ . .
- 1 - . ,
<
,
, @r..* _:. R.._:_ t,' .:
. lin c ~
---
~tt, ,ett
'
. 9_ s .
'
. ._
e. . e .4
.
- _ __ _., e ,
.
. ,
'
rrrT
-
7 T- T r T g .
n ., e.-.e.,- rT
.
,, ., .,
. .. _._ __ _
. . _. _ _._ ..
-
,
.
n l -
.e., .6, 0 , ,, - - ,9 ,, 1 I
, . -
rT r c )
ce l
r 7@ ,n r .;, T. , ., .,r r@c
_ _ _ _ . -
, ., ..
.:
.- .'r s
l
,
,
_ _._ __. . ... -
s 5
,
l l- . . . .
es -
-
. e 1_ ._ :.
=_: 0 , .. -. .. c, . y y :
-l l
,; L :.e .:...
.
_
' '
'. Th S d i' .'
0: -
4 : .l l -+
-=
.-. :.n 3 I-
-
,. . . .
_
-., .............~,,.,c, . i i
- 38 40 44 ...44
.
00 04 C8 :: '16 * "O .'" 4 ~13 i
.. *._5..d '
'. .
..
.;.. ~ s g..hs.w... ..Y ..
. .. ,
,
,.
.
._
.
.
.
.%'Q ..
l
'N.? .
.f CONTROL RCCS (13*) : B SRM DETICTORS (4) .
. .!
.
O IRM DETIcTORS tB) *
A
a _.SCL*RCI LOCAT:CNS (4) (Ramond)
e IN coRI DITIcTo A STE:NCS (31) 4 LPPJJ CITICTORS FIR ST3.!NG
,
!
0FH-17 20 Rev. 5 l J