IR 05000293/1997006: Difference between revisions

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{{Adams
{{Adams
| number = ML20149K131
| number = ML20216H215
| issue date = 07/18/1997
| issue date = 09/10/1997
| title = NRC Operator Licensing Exam Rept 50-293/97-06 (Including Completed & Graded Tests) for Tests Administered on 970505-09
| title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-293/97-06
| author name =  
| author name = Meyer G
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name = Boulette E
| addressee affiliation =  
| addressee affiliation = BOSTON EDISON CO.
| docket = 05000293
| docket = 05000293
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-293-97-06, 50-293-97-6, NUDOCS 9707290264
| document report number = 50-293-97-06, 50-293-97-6, NUDOCS 9709160178
| package number = ML20149K125
| title reference date = 08-18-1997
| document type = EXAMINATION REPORT, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| page count = 103
| page count = 3
}}
}}


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September 10, 1997 l
Enclosure U.S. NUCLEAR REGULATORY COMMISSION
P i
 
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==REGION I==
I E. Thomas Boulette, PhD      i Senior Vice Presideat Nucleer      !
. License N DPR-35 -
Boston Edison Company     i Pil9tim Nuclear Power Station 600 Rocky Hill Road Plymouth, Massachusetts -02360 5599    l SUBJECT: NAC INSPECTION No._50 293/97-06 (REPLY)
; Report N Jocket No.- 50-293  .
        '
Licensee:  Boston Edison Company 800 Boylston Street Boston, Massachusetts 02199 Facility:~ Pilgrim Nuclear Power Station
Dear Dr. Boulettei
.
: This. letter refers to your August 18,1997 correspondence, in response to our July 18,1997 lette i'
Exemination Period: May. 5 - 9,1997 Examiners: D. Florek, Seninr Operations Engineer .
Thank you for informing us of the corrective and preventive actions documented in your letter to assure that applicants for initial operator examinations will have properly  ?
C. Sisco, 0,serations Engineer ,
performed the required five significant control manipulations. These actions will be ;
S. Dernh., Examiner in Training l
i examined during your next licensed operator examinatio Your cooperation with un is appreciate
      '
S. Willoughby, Contract Examiner
, Approved by: G. Meyer, Chief Operations and Human Performance Branch ;
Division of Reactor Safety j
 
i 9707290264 97071g w
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PDR ADOCK 05000293 V  pg ,
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EXAMINATION SUMMARY Examination Report 50-293/97-006 (OL)
Initial examinations were administered to six senior reactor operator (SRO) instant applicants during the period of May 5 -9,1997, at the Pilgrim Nuclear Power Statio OPERATIONS Five of six applicants passed the examination. One SRO instant applicant failed the written and operating portion of the examination. The five applicants that passed were well prepared for the examinations. The applicants consistently understood and implemented the emergency operating procedures well. Some weak areas of understanding were identified during the written exam and operating tes Two of the applications were found to be deficient in that the applicants had not performed the five significant control manipulations on the plant as required by 10 CFR 55.31(a)(5).
 
The applicants' qualification records did not support performance of five significant control-manipulations. The root ceuse for this problem appeared to be that the BECO program guidance inappropriately permitted multiple significant control manipulation credit for a single, extended power chang l l
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Details 05.1 Operator initial Examinations Scope
      ]
The examiners administered initial examinations to six instant SRO applicants in accordance with NUREG-1021, " Examiner Standards," Revision I Observations and Findinas The results of the initial examinations are summarized below:
i SRO PASS / Fall
      '
j Written  5/1 Operating  5/1 Overall  5/1 The Boston E-jison Company (BECO) staff reviewed the written examination and assisted in the validation of the operating examination during the week of April 21,1996. The BECO staff provided comments on the examination that  )
significantly improved the examination. The BECO staff, who were involved with the examination review, signed security agreements to ensure that the initial examinations were not compromise In a letter, dated May 16,1997 (see Attachment 2), BECO provided six comments on the written examination. The NRC accepted two of the six comments. As a result, one question was deleted from the examination and two correct answers were accepted in one question. The NRC resolution of facility comments is summarized in Attachment 3.'
The following summarizes the written examination questions that were missed by at least three applicants, indicating a weakness in the understanding of the subjec Ques 3 Knowledge of the normal indication for the core spray line break detection monito Ques 33 Knowledge of the method to move an MSIV by use of the MSIV test push-butto T Ques 36 Knowledge of the air ejector off gas radiation rnonitor signals that willinitiate the 13-minute timer, i
 
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2 Ques 38 ' Ability to use technical specifications related to inoperable IRMs.-
Ques 43 Ability to determine procedure' entry to a given set o ,
        ;
condition I Ques 61 Knowledge of the number of drifting rods in a nine-rod array that require placing the mode switch in shutdow .
- Ques 76 Ability to determine the method and reason for depressurizing the reactor to a given set of condition !
Ques 85 Knowledge of the method to track the duration of surveillance During the operating test, at least two applicants performed poorly in each of the ,
'following areas:      !
Refueling operations Recognizing a loss of control room annunciators
- The above test items represent areas of weak understanding or performance and are provided to enable improvement of the training progra During.the dynamic simulator test, the following item was significant and a consistent positive observatio j l
Knowledge and understanding of the emergency operating procedures (EOPs).
 
During the development and administration of the examination, the examiners noted the following item for further BECO consideration of possible procedure improvement Emergency Operating Procedure 5.3.21 page 34 of 58 indicated that the installation of the jumper in panel C915 from jumper location DD-24 to DD-25 defeated the high drywell pressure and low RPV levelisolation signals for MO-47, Shutdown Cooling Outboard Isolation Valve. This jumper also affected isolation signals for MO-29B LPCIinjection valve. The procedure did not provide a note that this valve was also affected by installation of the jumpe : Conclusions      j Five_of six of the applicants were well prepared for the examination, and as a result, j five applicants passed the examination. One SRO instant applicant failed the l
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examinatio ;
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Sincerely, i
S Glenn W. Meyer, Chief  ,
  *
Operator Licensing and Human Performance Branch  ,
Division of Reactor Safety Docket No. 50 293      ,
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,yggpm,m-  - IE55%5ER    -
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05.2 Reactivity Manipulations Scope The inspector reviewed the BECO records to determine how the applicants complied with 10 CFR 55.31(a)(5). This section of 10 CFR requires that applicants must perform five significant control manipulations on the plant that affect reactivity or power leve Findinos BECO Document 0-RO-04 "NRC Licensed Nuclear Plant RO/SRO initial Qualification," dated August 1996, required a minimum of five significant reactivity manipulations with amplification that effort should be made to diversify the reactivity manipulations. Ten examples were identified for meeting the requiremen Four of the examples related to 10% power changes with control reds or recirculation flow. The inspector considered each of the ten examples as an appropriate significant control manipulatio Based on review of the individual applicant qualification records, four of the six applicants had performed five significant cont'ol r manipulations on the plant, although one of these four applicants did not have diverse manipulation Based on review of the individual applicant qualification records, the NRC examiner identified on May 5,1997, that two of the six applicants had not performed five significant control manipulations on the plant. Although the BECO guidance specified the minimum conditions for a manipulation, the minimum conditions had inappropriately been used to credit more than one manipulation when a single, extended power change occurred. For example, one applicant reduced power with recirculation flow from 100% to 68% over 56 minutes. BECO considered this to be three of the five significant control manipulations. The NRC staff disagreed with BECO and considered this to be one significant control manipulation. Another applicant reduced power from 100% to 50% initially with recirculation flow and then later with control rods over 77 minutes. BECO considered this to be all five of the required five significant control manipulations. The NRC staff disagreed with BECO and considered this to be two significant control manipulations. BECO was informed of the examiner's conclusion and informed that this would not impact the administration of the remainder of the examination. The resolution of this issue was pursued after the examination was administere The final applications submitted on April 18,1997, indicated that these two applicants had performed their five required significant control manipulations. After the NRC staff review of the supporting data for the application, the .NRC staff concluded that one applicant had performed two of the five significant control manipulations and the other applicant had performed three of the five significant control manipulations. These two applicants did not meet the requirements of 10 CFR 55.31(a)(5).
m
4 In discussions with BECO and the NRC on June 4,1997, the NRC reiterated the NRC position and informed BECO that these two applicants passed the examination but would not be issued licenses until the applicants and BECO submitted revised Form NRC-398s after five significant control manipulations were performed on the plant. BECO acknowledged the NRC staff finding and indicated that they had initiated actions to have the applicants perform additional significant control manipulations on the plant after the examination was administered and would submit revised Form NRC-358 BECO submitted revised Form NRC-398s in a letter dated June 1,1997. BECO also provided the details of how the applicants satisfied the 10 CFR requirement for significant control manipulations. Based on the revised applications and supporting data, the NRC subsequently issued licenses for these individual Conclusion The BECO guidance and examples of how to meet the requirements of 10 CFR 55.31(a)(5) were acceptable. However, the BECO practice of giving multiple significant control manipulation credit for a single, eendad power change was not acceptable. The examiner concluded that BECO had violated 10 CFR 55.31(a)(5),
which requires that applicants for operator licenses must have performed five significant control manipulations on the plant that affects affect reactivity or power l level. With the multiple manipulations removed, one SRO applicant had performed I two significant control manipulations, and another SRO applicant had performed three significant control manipulations. (VIO 97-06-01)
E8  Review of UFSAR Commitments A recent discovery of a licensee operating their facility in a manner contrary to the updated final safety analysis report (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 examination activities discussed in this report, the examiners reviewed portions of the UFSAR that related to the selected examination activities, questions or topic areas. The particular section reviewed was Table 5.2.4. The specific question reviewed was consistent with the UFSA . _ _ _ - _ .
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V. Manaaement Meetinas X1 Exit Meeting Summary At the conclusion of the examination, the examiners discussed their observations of the examination process with members of BECO management. BECO acknowledged the  !
examiners' observations. The BECO personnel present at the exit included the following:
      !
J. Alexander, Training Manager    .
      '
M. Briggs, Principal Instructor K. DiCroce, Sr. Regulatory Affairs Engineer L. Olivier, Vice President Nuclear    l M. Santiago, Operations Training Manager    1 (T. Sullivan, Plant Manager    .
.T, Trepanier, Operations Department Manager    !
T. Venkataraman, QA Group Manager
      ,
NRC Personnel-S. Dennis, Operations Engineer    l D. Florek, Sr. Operations Engineer    ]
      '
R. Laura, Senior Resident inspector C. Sisco, Operations Engineer
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Attachments:      !
      ! SRO Examination and Answer Key Facility Comments on Written Examinations    ; NRC Resolution of Facility Comments    i l Simulation Facility Report    I
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ATTACHMENT 1 SRO Examination and Answer Key
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t U. S. NUCLEAR REGULATORY COMMISSION SITE SPECIFIC EXAMINATION SENIOR OPERATOR LICENSE REGION 1 APPLICANT'S NAME:
FACILITY:  Pilarim 1 REACTOR TYPE: BWR-GE3 DATE ADMINISTERED: May 5,1997 INSTRUCTIONS TO APPLICANT:
Use the answer sheets provided to document your answers. Staple this cover sheet on top of the answer sheets. Points for each question are indicated in parentheses after the question. The passing grade requires a final grade of at least 80%.
Examination papers will be picked up four (4) hours after the examination start TEST VALUE APPLICANT'S SCORE FINAL GRADE %
100.00 All work done on this examination is my own. I have neither given nor received ai l l
l Applicant's Signature
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i SENIOR REACTOR OPERATOR    Prga 2 ANSWER SHEET Multiple Choice (Circle or X your choice)
  . lf you change your answer, write your selection in the blan MULTIPLE CHOICE  O23 a .b cd __
001 a b c d  024 a b c d 002 a b'c d _
O25 a b c d 003 ' a b c d  026 a b c d 004 a bc d  027 a b c d 005 a b c d  028 a b cd 006 a. b c d  029 a b c d  ;
007 a b c d  030 a b c d 008 a b c d  031 a b c d 009 a b c d  032 a b c d 010 a b c d  033 a b c d
  '011 a b c d  034 a b c d
, . 012 a b c d  035 a b c d 013 a b c d  036 a b c d 1  014 a b c d  037 a b c d 015 a b c d  038 a b c d 016 a b c d  039 a b c d 017 a b c d  040 a b c d 018 a b c d-  041 a b c d 019 - a b c d  042 a b c d 020 a b cd  043 a b c d 021 a b c d  044 a b c d 022 a. b c d  045 a b c d f
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SENIOR REACTOR OPERATOR -  Pzgn'3 ANSWER SHEET-Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blan ,
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046'.a b c d  069 a. b c d  l i
i 047 a'b c d_  070 a b c-d-  )
048 ab c d  071 a b c d 049 a bc d  072 a b c d  )
050 a b c -d  073 ab c d i
074 a b c d  l 051 a b-c'd
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052' a b c d  075 a b c d I
053 a b c d  076 a b c d  i
054 a b c d  077 a b c d 055 a b c d  078 a b c d 056. a b c d  079 a b c d  1 057 a b c d  080 a b c d 058 a b c d  081 a b c d
- 059' a b c d  082 a b c d
, 060 a b c d  083 a b c d 061 a b c d  084 a b c d 062 a b c d  085 a b c d  :
063 a b c d  086 a b c d 064 a b c d  087 a.b c d 065 a b c d  088 a b c d 066 a b~ c d  089 a b c d 067 a-b c d  090 a b cd 068 -a b c d  091 a b c d I
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SENIOR REACTOR OPERATOR    Prg3 4 ANSWER SHEET Multiple Choice (Circle or X your choice)
If you change your answer, write your selection in the blan .
092 a - b c. d 093 a b c-d _
094 a b c d _
095 a b c d _
096 a b c d _
097 a b c d _
098 a b c d _
099 a b c d _
100 a b c d _
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  ( * * * * * * * * * * END OF EX AMIN ATION * * * * * * * * * * )
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NRC RULES AND GUIDELINES FOR LICENSE EXAMINATIONS During the administration of this examination the following rules apply:
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        ; Cheating on the examination means an automatic denial of your application and could result in more severe penalties.
. . After the examination has been completed, you must sign the statement on '
the cover sheet indicating that the work is your own and 'you have not received of given assistance in completing the examination. This must be done after you complete the examinatio , Restroom trips are to be limited and only one applicant at a time may leav You must avoid all contacts with anyone outside the examination room to avoid even the appearance or possibility of. cheatin ,
! Use black ink or dark pencil ONLY to facilitate legible reproduction . Print your name in the blank provided in the upper right-hand corner of the examination cover sheet and each answer sheet.
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  ~ Mark your answers on the answer sheet provided. USE ONLY THE PAPER
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PROVIDED AND DO NOT WRITE ON THE BACK SIDE OF THE PAGE.
      -.--o ,_w.., .r.-,,


' The point value for each' question is indicated in parentheses after the ;
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question.
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t If the intent of a question is unclear, ask questions of the examiner only.
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: When turning in your examination, assemble the completed examination with examination questions, examination aids and answer sheets in addition,
,  turn in all scrap paper.
 
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1 Ensure allinformation you wish to have evaluated as part of your answer is on your answer sheet. Scrap paper will be disposed of immediately
;  following the examination.
 
;
i 1 To pass the examination, you must achieve a grade of 80% or greate . There is a time limit of four (4) hours for completion of the examinatio . When you are done and have turned in your examination, leave the
;  examination area (EXAMINER WILL DEFINE THE AREA). If you are found in this area while the examination is stillin progress, your license may be i  denied 'or revoked.
 
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SENIOR REACTOR OPERATOR  Pcgs 7
. QUESTION: 001 (1.00)
The HPCI system has automatically initiated due to a low reactor water level. Drywell pressure remains within normal limits. The HPCI turbine slowly lowers reactor pressure. Reactor pressure continues to decrease and reaches 80 psi Which ONE of the following is the expected automatic response?
a. Group IV isolation, but no HPCI turbine trip and no Group Vil isolation b. Group IV isolation and HPCI turbine trip, but no Group Vil isolation c. Group Vilisolation and HPCI turbine trip, but no Group IV isolation
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d. Group IV solation, Group Vil isolation, and HPCI turbine trip l
I OUESTION: 002 (1.00)
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Which ONE of the following signals will NOT require resetting the Trip and Throttle Valve to restart the RCIC turbine?
! a. RCIC Turbine Mechanical Overspeed b. Reactor high water level c. Manual trip pushbutton on 904 panel
; d. High Steam Supply line differential pressure l
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- SENIOR REACTOR OPERATOR _  P ga 8
 
QUESTION: 003 (1.00)
With the plant operating at 100% power, the 'A' Core Spray Line Break Detection Monitor is reading approximately -3.0 psi This reading is:
,
a. Indicative of an 'A' Core Spray Line break inside the shrou b. indicative of an 'A' Core Spray Line break outside the shroud.
: c. normal due to the differential pressure across the dryers and separators being approximately -3.0 psid at 100% powe d. normal due to changes in water density after the instrument was calibrated to read zero under cold condition QUESTION: 004 (1.00)
The following conditions exist:
- SBLC Tank Temperature 45 Degrees F
-
SBLC Tank Volume 4000 gallons
- SBLC Tank Concentration 9.1% weight %
- B-10 Isotope Enrichment 53 %
What is(are) the MINIMUM required action (s) that you as the NWE should immediately initiate?
a. Perform a SBLC flow tes b. Determine whether the sodium pentaborate solution meets the original design criteri c. Perform a SBLC flow test and determine whether the sodium pentaborate solution meets the original design criteri d. Immediately commence a plant shutdown such that the plant can reach cold shutdown within 24 hour ., _ -.  - .-
SENIOR REACTOR OPERATOR  Paga 9 QUESTION: 005 (1.00)-
Given the following conditions:
- The plant is in cold shutdow *
- An RHR system test is in progres The LPCI Override Control Switch (S178) has been taken to MANUAL OVERRID Drywell pressure is at O psig and stead As part of the test, reactor water level is simulated at -60 inche The next step in the procedure is to take the control switch for the Torus Spray Valve MO-1001-37B to open.
 
Which ONE of the following explains why MO 1001-37B will NOT open when the control switch is taken to open?
a. The 15 minute time delay is not timed ou b. Drywell pressure is at atmospheri c. The RPV Level Override Keylock Switch (S188) is not in MANUAL j OVERRID d. The 5 minute time delay has not timed out and the MO-1001-28B i
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is not close l l
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SENIOR REACTOR OPERATOR  Pegs 10 l
QUESTION: 006 (1.00)
,
The following conditions exist:
  - The plant is operating at 100% powe The "A" SBGT Fan is in AUT The "B" SBGT Fan is in STB A valid SBGT Initiation signal occur The "A" SBGT Fan initially starts, runs for 10 seconds, then trips for an unknown caus Which ONE of the following describes the expected automatic response of the "B" SBGT Fan?
        'l The "B" SBGT Fan will:
a. start when the initiation signal is received, run for 65
  . seconds and then stop, then restar b. start immediately after the "A" SBGT Fan trips and continue
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running uninterrupted c. start after 65 seconds and continue to run uninterrupte d. start when the initiation signal is received and continue running uninterrupte j
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SENIOR REACTOR OPERATOR    Prgs 11
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QUESTION: 007 (1.00).      ,
The following conditions exist:
  ;
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The plant is at 100% powe it is determined that the 3A SRV will NOT open under ANY
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conditio Which ONE 'of the following states the MINIMUM action REQUIRED by
              >
E. Thomas Boulette    -2 l
cc w/encli            ,
L. Olivier, Vice President . Nuclear and Station Director          +
T. Sullivan, Plant Department Manager          :
N. Desmond, Regulatory Relations          >
D. Tarantino, Nu: lear Information Manager          i
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Technical Specifications?
R. Hallisey,- Department of Public Health, Commonwealth of Massachusetts        j i  The Honorable Therese Murray            !
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a. Place the plant in Cold Shutdown within 24 hour b. Reduce reactor coolant pressure below 104 psig within 24 hour c. Provided HPCI is operable, enter 14 day LCO. When this LCO is expired, place the plant in Cold Shutdown within 24 hour d. Provided HPCI is operable, enter 14 day LCO. When this LCO is expired, reduce reactor coolant pressure below 104 psig within
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24 hours.
 
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QUESTION: 008 (1.00)
The Honorable Joseph Gallitano          i B. Abbenet, Departmont of Public Utilities
During a high drywell pressure condition, a valid ADS signal exists and
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. the ADS system has initiated. With the initiation signal still present
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both initiation Signal Timer Reset Pushbuttons are depresse Which ONE of the following describes the expected automatic response of the ADS system?
j All ADS valves will:
a. remain ope b. close and remain closed indefinitel c. close and remain closed for 105 seconds then reope d. close and remain closed for 11 minutes then reopen.
 
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SENIOR REACTOR ' OPERATOR  P:ga 12
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Chairman, Plymouth Board of Selectmen
QUESTION: OO9 (1.00)
The following conditions exist:
  - The "A" and "B" Reactor Feed Pumps are in servic Both Reactor Recirculation Speed demands are at 60%. -
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  - An instrument failure causes the Feedwater Regulating Valves to reduce feedwater flow to 3 Mlbm/hr with reactor water level reaching a minimum of + 17 inches.
 
  - The operator takes manual control of the Feedwater Regulating Valves and is returning water level to normal with a current
    '
level of + 18 inche '
  - No operator action is taken on the Reactor Recirculation Syste Which ONE of the following describes the expected response of the  l
. Recirculation Flow Controllers?    '
'The Recirculation Flow controllers will demand lowering speed to a. 44% without a rate limitation signa !
  .b. 44% at a rate of 1.5% per secon c. 26% without a rate limitation signa d. 26% at a rate of 1.5% per second.
 
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' QUESTION: 010 (1.00)
An Emergency Diesel Generator (EDG) has started due to a LOCA signa Which ONE of the following signals will cause an EDG trip?
a. Engine Overspeed i
b. Engine Low Lube Oil Pressure c. Engine High Lube Oil Temperature
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d. Engine Crankcase High Vacuum
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SENIOR REACTOR OPERATOR:  ' Piga 13 )
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QUESTION: 011 (1.00)    l l
The following conditions exist:
- HPCI is injecting water from the CST to the RP l
- The HPCI Suction Valves From Suppression Chamber MO-2301-35 and-MO-2301-36 Control Switches are in Aut The CST Low Level Alarm comes i Which ONE of the following describes the expected response of the HPCI system?
The HPCI Suction From CST MO-23016 will receive a close signal-
      )
a. as soon as both the MO-2301-35 and the MO-2301-36 valves reach
. full ope b. as soon as both the MO-2301-35 and MO-2301-36 valves come off their closed seats, c. as soon as either the MO-2301-35 or the MO-2301-36 valve reaches full ope l l
d. at the same time the MO-2301-35 and MO 2301-36 valves receive ;
an open signa )
I QUESTION: 012 (1.00)
Which ONE of the following states where the RCIC turbine receives stearn and where the RCIC pump discharges?
a. Steam from "C" Main Steam Line and Discharge to "A" Feedwater
: Line b. Steam from "D" Main Steam Line and Discharge to "B" Feedwater
: Line c. Steam from "D" Main Steam Line and Discharge to "A" Feedwater Line
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d. Steam from *C" Main Steam Line and Discharge to "B" Feedwater Line i
 
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Pegs 14 l SENIOR REACTOR OPERATOR I
OUESTION: 013 (1.00)
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Which of the,following SRM rod block (s) is(are) bypassed by moving IRM range switches from Range 2 tc Range 3?  l a. SRM Downscale Rod Block only  -
b. SRM Inoperable Rod Block only  ;
I c. SRM Downscale Rod Block and Detector Retract Not Permitted Rod Block d. SRM inoperable Rod Block and SRM Downscale Rod Block u
QUESTION: 014 (1.00)    ,
With Reactor Power at'100%, an SRV spuriously lifts. Action to close !
the valve are successful. Immediately after valve closure, the  j downstream temperature is checke Which ONE of the following is an expected approximate downstream temperature?
a. 212 degrees F b. 295 degrees c. 375 degrees F  .l d. 525 degrees F i
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SENIOR REACTOR OPERATOR  P gs 15-QUESTION: 015 (1.00)
The following conditions exist:
- A half scram exists on RPS "A" due to APRM testin A fire caused a loss of RPS Bus "B" and a full scra '
- The half scram testing was stopped and APRMs were returned to normal.~
- The SCRAM DISCHARGE INSTRUMENT VOLUME HI LEVEL SCRAM BYPASS switch is then taken to bypas Which ONE of the following describes when the RPS "A" half scram may be reset?
a. immediatel b. after the air dump test switch is placed in isolat c. after the SDIV vent and drain valves come fully ope d. after RPS "B" is energize !
l QU' - ION: 016 (1.00)
The mode switch is in RUN. Which ONE of the following scram signals is automatically bypassed 2 seconds after taking the mode switch to SHUTDOWN?
      , i a. Mode switch in shutdown b. Main steam isolation valve closura c. Turbine stop valve closure d. Scram discharge volume high level
 
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SENIOR REACTOR OPERATO Pags 16
:
QUESTION: 017 (1.00).
 
. The ATWS logic system has automatically initiated due to low reactor water leve ~
Which ONE of the following actuations will be delayed by 9 seconds?
a. Rod insertion b. Reactor Recirc Pump Field Breaker Trip
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c. Reactor Recirc Pump Drive Motor Breaker Trip d. Reactor Feed Pump Trip
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QUESTION: 018.(1.00)
:
'
. During a reactor shutdown, the control rod selected on the Rod Select Matrix is NOT in the rod group of the latched step. As reactor power
' decreases, at what point will this condition cause an insert and withdraw block?
a. Steam Flow drops below 35%
b. All APRM readings drop below 20%
c. Steam Flow or Feed Flow drops below 20%
d. Steam Flow and Feed Flow drop below 35%
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SENIOR REACTOR OPERATOR  P:gs 17 QUESTION: 019 (1.00)
The following conditions exist:
- The plant is operating at 100% powe APRM "C".is bypassed for maintenanc APRM "E" then fails giving a constant reading of 95% regardless of inpu A half scram already exists on RPS "B" Which ONE of the following meets the action REQUIRED?
a. Initiate insertion of operable rods and complete insertion of all operable rods within sixteen hour b. Reduce power level to IRM range and place mode switch in the startup/ hot stand'y position within eight hour c. Reduce turbine load and close main steam isolation valves within eight hours, d. Reduce power to less than 45% of design.
 
QUESTION: 020 (1.00)
A TIP trace is being performed when a high drywell pressuro signal occurs. Select the expected automatic actio a. The shear valve fires with the detector stillin the cor b. The ball valve closes with the detector stillin the cor c. The detector withdraws into its shield and the ball valve closes, d. The detector withdraws into its shield and the shear valve fire ,
 
e- A , asa _ -ir e 4 # ww .e4.1 - ,a 4 4 +4 -
4- beM iJ---'a-
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SENIOR REACTOR OPERATOR  Pi:gs 18
      .
QUESTION: 021 (1.00)
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      '
While operating at 80% power, an instrument failure causes the throttle pressure sensed by the EPR to fail high. No operator action is take Which ONE of the following is the expected result?  i a. The reactor would scram on a high pressure scram signa t b. The MPR would take control and pressure would increase by approximately 10 ps c. The reactor would scram on a low pressure scram signal.
 
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d. The reactor would scram on a MSIV closure scram signa QUESTION: 022 (1.00)
At 500 psig during a reactor startup and heatup, the #1 Bypass Valve (BPV) comes partially ope i
'
Which ONE of the following errors is the cause?
Failure to maintain the:
,
l a. EPR 40-80 psig below reactor pressure  l l
b. EPR 40-80 psig above reactor pressure  ;
I c. MPR 40-80 psig below reactor pressure d. MPR 40-80 psig above reactor pressure
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SENIOR REACTOR OPERATOR  P:ga 19
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QUESTION: 023 (1.00)'
With the plant operating at 20% power both Reactor Recirculation Pumps irip. The operator manually scram the reactor. Post scram, fuel zone level indicators read:
a. falsely high since less flow exists through the jet pumps than existed during calibration conditions.
, b. falsely low since less flow exists through the jet pumps than  ;
existed during calibration condition c. falsely low due to decreased density of the water in the vessel against calibrated conditions, d. falsely high due to decreased density of the water in the vessel against calibrated condition QUESTION: 024 (1.00)
With the "A" Loop of RHR in Lo  oling, RPV level decreased to 12 l inches. The Shutdown Cooling Outtsumo I:,ulation Valve MO-1001-47  l stopped in mid-stroke. All other valves have responded as expecte i
      !
Which ONE of the following is REQUIRED in order to open the "A" Loop  !
LPCI Injection Valve #2 MO-1001-29A?
a. The MO-1001-47 valve must be closed, b. The MO-1001-29A must be manually reset, c. Reactor coolant pressure must be greater than 76 psig, d. The Group ll isolation signal must clear and the Group 11 logic  l must be rese I l
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SENIOR REACTOR OPERATOR  PJgn 20 QUESTION: 025 (1.00)
A reactor scram has occurred. Electrical busses A-5 and A-6 have transferred to the Start-up Transformer. Which ONE of the following-describes the drywell cooler response?
a. The running drywell coolers will trip and start after a 45 second time delay. The drywell coolers in standby remain in standby, b. The running drywell coolers will trip. The drywell coolers in standby will start after a 45 second time dela c. The running coolers will stay in service. The drywell coolers in standby willimmediately start when A-5 and A-6 are reenergize d. The running coolers will stay in service. The drywell coolers in standby will start after a 45 second time dela !
QUESTION: 026 (1.00)    l Primary Coolant Temperature is 245 degrees F when Shutdown Cooling is placed in service, immediately thereafter, a fire disables the Shutdown Cooling Outboard Isolation Valve MO-1001-47 motor operator. The valve 1 is in the open positio I Which ONE of the following meets the MINIMUM REQUIRED action?
a. Verify the ability to manually close the MO-1001-47 valve, then reestablish shutdown coolin )
      )
b. Verify the ability to close the MO-1001-50 valve, then reestablish shutdown cooling, c. Close either the MO-1001-47 or MO-1001-50 valve and open the respective breake d. Station an operator to manually close the MO-1001-50 valve if required and continue in shutdown coolin ,._
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SENIOR REACTOR OPERATOR  Paga 21 QUESTION: 027 (1.00)
l When valving in a CRD hydraulic control accumulator, the 305-102
'(5Nithdraw Riser Isolation Valve) and the 305-112 (Scram Discharge Riser isolation Valve) are required to be open prior to opening the 305-101 1 l
(Insert Riser isolation Valve). This prevents-a. .a single rod scram when opening the 305-101 valv I b. excessive scram time of that rod in the event of a reactor scra l
      !
c. damage to the accumulator in the event of a reactor scra I d. damage to the drive mechanism in the event of a reactor scra )
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QUESTION: 028 (1.00)    l l
While operating at 100% power, a control rod is determined to be i uncoupled. Attempts to couple the rod have been unsuccessfu l l
Which ONE of the following states the MINIMUM REQUIRED actions?
a. Verify that the control rod can be moved with drive pressure and maintain the control rod at the target positio b. Fully insert the control rod and hydraulically disarm the CR l
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c. Fully insert the control rod and electrically disarm the l directional control valve l l
d. Fully insert the control rod, electrically disarm the  )
directional control valves and then declare the rod inoperabl i
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, SENIOR REACTOR OPERATOR  Pzga 22 i QUESTION: 029-(1.00)
With the plant at pnwer, it is determined that the MO-1001-37B (B Loop Torus Spray) and MO-1400-25A (A Loop Core Spray Inboard injection)
valves have failed their operability test. Both volves are currently close The maximum time allowed before the plant must be in COLD SHUTDOWN is:
a 24 hours (1 day).
96 hours (4 days),
c. ~ 168 hours (7 days).-
d.192 hours (8 days).
QUESTION: 030 (1.00)
A tagout, which has been in effect on the "A" Reactor Recirculation Pump for 7 days, has just been cleared. The "A" Reactor Recirculation Pump is started and immediately manually tripped. On the second start p attempt, the pump starts and runs for 10 minutes and then is manually
,
trippe When is the SOONEST that another start of the "A" Reactor Recirculation
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Pump may be attempted?
Chairman, Duxbury Board of Selectmen'          :
a. Immediately b.15 minutes after the second trip c. 45 minutes after the second trip d. 4 hours after the second trip i
Chairman, Nuclear Matters Committee          !
      !
Plymouth Civil Defense Director _ _          -
 
P. Gromer, Massachusetts Secretary of Energy Resources          !
SENIOR REACTOR OPERATOR  Pags 23 I
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QUESTION: 031 (1.00)
;_  ~J. Milier, Senior issues Manager
l The plant is operating at 100% power when the "B" Reactor Recirculation i Pump trips. No operator action is taken.
              ;
 
Which ONE of the following describes the initial steady state to final steady state change in the "A" Reactor Recirculation Loop Jet pump flow and the reason for the change?
The "A" Reactor Recirculation Loop Jet pump flow will: l l
a. increase due to lower core pressure dro I b. increase due to decreased core voidin l l
c. decrease due to higher core pressure dro d. decrease due to increased core voidin QUESTION: 03 (1.00)
While operating at 0% power, it is determined that th Main Steam Line High Flow switches o the "B" Main Steam Line will N trip under a high flow condition.
 
Which ONE of the following the MINIMUM RE IRED action? r a. Direct l&C personnel to m ually trip t inop blejpi Eiles. f b. Direct l&C personnel to manu ly i ert a ha f isolation on the "B" Group 1 Cha (j p s c. Initiate an orderly shutdown db in Cold Shutdown Condition within a MAXIMUM of 30 h urs afte he instrument failur d. Initiate an orderly shutd n and have th Main Steam Lines isolated within a MAXI UM of 10 hours a r the instrument failur ;
 
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SENIOR REACTOR OPERATO P:ge 24 QUESTION: 033-(1.00)
Depressing a.n outboard MSIV test pushbutton will:
a. energize the AC test valve and vent air from the underside of the pisto b. energize the AC test valve and admit air to the underside of the pisto ,
c. deenergize the AC test valve and vent air from the underside of the pisto ' d. deenergize the AC test valve and admit air to the underside of r the pisto QUESTION: 034 (1.00)
Which ONE of the following conditions requires Rod Block Monitor Operability?
a. MCPR is 1.35 and Reactor Power is 25%.
b. MCPR is 1.45 and Reactor Power is 75%.
c. MCPR is 1.55 and Reactor Power is 95%.
d. MCPR is 1.60 and Reactor Power is 100%.
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. SENIOR REACTOR OPERATOR  P g) 25 QUESTION: 035 (1.00)
A loss of 120V Bus A (Y-3) occur Which ONE of the following describes the effect on the RWCU system?
    ;
a. Half of the logic for closing the MO-2 and MO-5 valves is made u b. MO-2 goes closed. As soon as MO-2 comes off the open seat, the operating RWCU pump (s) will trip. MO-5 remains ope c. MO 2 goes closed. As soon as MO 2 comes off the open seat, the ,
operating RWCU pump (s) will trip and MO-5 will go close j d. MO-5 goes closed. As soon as MO-5 comes off the open seat, the operating RWCU pump (s) will trip. MO-2 remains ope QUESTION: 036 (1.00)
The OFF GAS ISOL CH PRM SEL switch is in position 2. Which ONE of the following conditions of the Air Ejector Off Gas Radiation Monitors will 4 cause tiie 13 minute timer to initiate?
a. Hi radiation signal on both channels b. Hi Hi radiation signal on one channel c. Hi radiation signal on one channel and Downscale Trip on the !
other channel
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d. Downscale trip on one channel and inop trip on the other channel 1 . - _ . . _
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l l- SENIOR REACTOR OPERATOR   P ga 26
   - J. Fleming            -.
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A. Nogee,- MASSPIRG            >
QUESTION: 037 (1.00)
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l A reactor startup is in progress with reactor power in the intermediate range. IRM "A" then starts to intermittently swing upscale and then downscal Which ONE of the following conditions on IRM "A" will cause a Rod Block but NOT cause a Half Scram?
Office of the Commissioner, Massachusetts Department of Environmental Quality Engineering            i Office of the Attorney General, Commonwealth of Massachusetts
The IRM reads:
"
a.1 (on the 0-40 scale) while on range *
T. Rapone, Massachusetts Executive Office of Public Safety          i Chairman, Citizens Urging Responsible Energy          l 1 Commonwealth of Massachusetts, SLO Designee          ,
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b. 3 (on the 0-40 scale) while on range i
 
I c. 36 (on the 0-40 scale) while on range d. 39 (on the 0-40 scale) while on range !
QUESTION: 038 (1.00)
With the Mode Switch in Startup, at 1200 on 5/5/97, the Downscale Trips for IRM Channels "A", "B", and "E" are made inoperabl ]
Which ONE of the following is the LATEST that one of these channels must be placed in a tripped condition?
a. 1300 on 5/5/97    i b.1200 on 5/6/97 c.1200 on 5/12/97 d. 1300 on 5/12/97
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l SENIOR REACTOR OPERATOR  PIga 27 '
OUESTION: 039 (1.00)
The plant was. operating at 100% power with the "B" CRD pump in service.
 
Subsequently, a valid LOCA signal generated a scram. The plant responded as expected except, the startup transformer feeder breaker to bus A 5 failed to close. A-5 has been automatically energized from the shutdown transformer.
 
Which ONE of the following describes the status / availability of the CRD pumps?
a. "B" CRD pump is runnin "A" CRD pump can be started since no load shed signal was generate b. "B" CRD pump is not runnin "A" and "B" CRD pumps cannot be started due to load shed signa c. "B" CRD pump is not runnin "A" and "B" CRD pumps can be started since no load shed signal was generate ,
d. "B" CRD pump is runnin "A" CRD pump cannot be started due to load shed signa l
 
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- SENIOR REACTOR OPERATOR  Pign 28
      ,
QUESTION: 040-(1.00)-
Which ONE of the following administrative precautions related to valves are required when lining up RHR "A" loop for shutdown cooling?
a. MO-1001-7A "RHR PUMP A TORUS SUCTION" red tag close MO-1001-7C "RHR PUMP C TORUS SUCTION" red tag close ,
MO-1001-43A "RHR PUMP A SHUTDOWN COOLING SUCTION yellow tag close MO-1001-43C "RHR PUMP C SHUTDOWN COOLING SUCTION yellow tag close b. MO-1001-7A "RHR PUMP A TORUS SUCTION" red tag close MO-1001-7C."RHR PUMP C TORUS SUCTION" red tag close MO-1001-438 "RHR PUMP B SHUTDOWN COOLING SUCTION red tag close MO-1001-43D "RHR PUMP D SHUTDOWN COOLING SUCTION red tag close c.1001-6A "RHR PUMP C SUCTION VALVE FROM THE TORUS" red tag close A "RHR PUMP A SUCTION VALVE FROM THE TORUS" red tag close MO-1001-43B "RHR PUMP B SHUTDOWN COOLING SUCTION red tag close MO-100143D "RHR PUMP D SHUTDOWN COOLING SUCTION red tag close d.1001-6A "RHR PUMP C SUCTION VALVE FROM THE TORUS" yellow tag close A "RHR PUMP A SUCTION VALVE FROM THE TORUS" yellow tag close MO-1001-43A "RHR PUMP A SHUTDOWN COOLING SUCTION yellow tag close ;
MO 1001-43C "RHR PUMP C SHUTDOWN COOLING SUCTION yellow tag close _ _
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SENIOR REACTOR OPERATOR  Pags 29 1 QUESTION: 041 (1.00)
_
Given the following conditions:
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The plant is in cold shutdown No recirculation pumps are in service RHR pump "A"is in shutdown cooling RWCU is in service Reactor shutdown level instrument indicates 40 inches Which ONE of the fo:iowing describes reactor coolant temperature indication if the "A" RHR pump trips. Assume no operator action, a. Recirc loop "A" temperature indicator is representative of reactor coolant temperatur b. Recirc loop "B" temperature indicator is representative of reactor coolant tempestur c. RWCU bottom head drain temperature indicator is representative of reactor coolant temperatur d. No temperature indicator is representative of reactor coolant temperature,
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SENIOR REACTOR OPERATOR  Prgs 30
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- QUESTION: 042 (1.00)
.The following conditions exist:
!' - EOP-02 is being executed .
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The Mode Switch is in Shutdown and ARI has been initia'te '
- The MSIVs are close Reactor power is 2.5% and no boron has been injecte ~ Alternate _Depressurization is required by EOP-0 Four SRVs can be opene Which ONE of the following actions should be taken to control reactor '
I water level?
a. Secure all sources of injection. When pressure decreases below 200 psig, slowly inject with LPC b. Secure all sources of injection. When pressure decreases below
:. 400 psig, slowly inject with the Condensate Pump c. Secure all sources of injection except CRD and RCIC. When
, pressure decreases below 200 psig, continue injection flow rate with RCIC and CR d. Secure all sources of injection except CRD and RCIC. When pressure decreases below 270 psig, slowly inject with the
.
Condensate Pumps.
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i SENIOR REACTOR OPERATOR    Pcgs.31  i
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QUESTION: 043 (1.00)
The following conditions exist:      l
- A manual' scram was inserted from 20% powe l
- No other scram signals exis !
l
- Reactor power is on intermediate range 6 and decreasin Three control rods are at position 06. All other rods are  ,
fully inserte Which ONE of the following is the required action?
a.' ' enter PNPS 2.1.6. No EOP entry is require b. ' enter EOP-01, then exit EOP-01 and enter EOP-02 at R- ,
        ;
c. enter PNPS 2.1.6, " Reactor Scram", then exit PNPS 2.1.6 and
  .
enter EOP-02 at R d. enter PNPS 2.1.6, " Reactor Scram", then enter EOP-02 and execute concurrently with PNPS 2. i l
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< SENIOR REACTOR OPERATOR  pig 3 32-s QUESTION: 044 (1.00)
The following conditions exist:
- EOP-02 is being execute Boron is being injected with the SBLC syste '
- Initial SBLC tank level was 4100 gallon ;
Reactor Water Levelis being lowered to reduce reactor powe ;
- Current SBLC tank level is 3000 gallon Torus water temperature is 112 degrees Reactor water level is'-100 inche Which ONE of the following actions is REQUIRED 7 a. Reise reactor water level to the +12 to +45 inch band and perform Alternate Depressurizatio '
b. Raise reactor water level to the + 12 to +45 inch band. Do not perform Alternate Depressurizatio !
      !
c. Maintain reactor water level at its current value and perform
  . Alternate Depressurizatio d. Maintain reactor water level at its current value. Do not perform Alternate Depressurizatio QUESTION: 045 (1.00)    {
l While operating at 100% reactor power, reactor pressure starts to oscillate approximately 10 psi peak to peak and pressure control is shifting alternately from the EPR to the MPR and back to the EP Which ONE of the following actions are REQUIRED?
a. DJace the EPR control switch to off, b.' Reduce reactor power to approximately 75%.
. c. Raise the MPR setpoint to prevent pressure control from
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swapping between regulators.
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d. Lower.the MPR setpoint to allow the MPR to take control of pressur . .  . -. . .-  .
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SENIOR REACTOR OPERATOR  Pcga 33
      ,
t QUESTION: 046 (1.00)
With the plant in Cold Shutdown, some solvent that is improperly stored in a Control Room locker ignites. The Nuclear Watch Engineer makes the decision to evacuate the Control Room and to call for off-site assistance to put out the fire. Control is established at remote -
shutdown stations 20 minutes after the Control Room evacuatio What is the MINIMUM event level classification?
a. Unusual Event b. Alert
      '
c. Site Area Emergency d. General Emergency
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QUESTION: 047 (1.00)
A LOCA has occurred. Which ONE of the following REQUIRES exiting the RPV level control leg of EOP-017 a. Reactor water level is -165 inches and increasing with Reactor Pressure at 200 psi b. Reactor water level is -125 inches and decreasing with Reactor Pressure at 175 psi c. Reactor water level is -125 inches and increasing with Reactor Pressure at 100 psi d. Reactor water level is -125 inches and decreasing vdth Reactor Pressure at 75 psig.
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- SENIOR REACTOR OPERATOR  P gs 34
      ;
QUESTION: 048 (1.00)
~ Which ONE of the following conditions REQUIRES Alternate Reactor-  ,
Pressure Vessel Depressurization assuming a primary system is '  ,
dischaiging into secondary containment?
      .
a. RCIC torus piping area temperature is 300 degrees F and RCIC
  '
      ,
turbine area temperature is 195 degrees b. HPCI compartment water levelis 8 inches and HPCI turbine area  ,
temperature is 195 degrees .:
c. RHR "B" and "D" pump area temperature is 300 degrees F and RHR
  ."A" and "C" pump area temperature is 195 degrees , d. Main Steam Tunnel area temperature is 300 degrees F and RHR "A" and "C" pump area temperature is 220 degrees ,
QUESTION: 049 (1.00)
Following a Nitrogen Line leak in the drywell, AO-4356 (Nitrogen / Air Isolation Valve to the Drywell) was closed. By calculation, how many times over the next eight hours can each SRV be actuated?
NOTE: COUNT EACH OPEN AND CLOSE CYCLE AS ONE ACTUATION a. 5      ,
b. 10 c. 20
d. 40
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SENIOR REACTOR OPERATOR  P ga 35 QUESTION: 050 (1.00)
in the event that torus water level cannot be maintained above 95 inches, HPCl is secured in order to prevent:
a. exceeding the Primary Containment Pressure Limi '
b. exceeding the Pressure Suppression Pressur c. exceeding the Heat Capacity Temperature Limit, d. isolating HPCI on high exhaust pressur ' QUESTION: 051 (1.00)
The following conditions exist:
- Reactor pressure is 10 psi Drywell pressure is 4 psi ' Torus bottom pressure is 15.2 psi Torus water level is 303 inche Select the correct action and its reason.
.
Under these conditions:
a. Alternate RPV Depressurization is required to prevent SRV Tail Pipe failure, b. Suppression Chamber Spray initiation is required using enly those RHR pumps not required to provide adequate co coolin c. Suppression Chamber Spray initiation is not allowed sinc. the Torus Spray Sparger is covere d. Suppression Chamber Spray initiation is not ai. owed since the Torus-Drywell Vacuum breakers are covered.
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' SENIOR REAC f0R OPERATOR  P;gs 36 QUESTION: 052 (1.00)
A trip of the "A" Reactor Recirculation Pump has occurred. The plant is operating in Region ll of the Power / Flow Map after the immediate actions
of 2.4.17 have been complete Which ONE of the following is REQUIRED 7 Exit Region 11 by:
a. manually scramming the reacto b. restarting the "A" Reactor Recirculation Pum c. increasing the speed of the "B" Reactor Recirculation Pum d. inserting control rods in reverse order of the pull shee l QUESTION: 053 (1.00)    !
l The following conditions exist:    l l
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Torus water level is 105 inche Torus water temperature is 180 degrees Reactor pressure is 700 psi Which ONE of the following states whether Alternate RPV Depressurization is required, not required, or prohibited and the reaso Under these conditions, Alternate RPV Depressurization is a. not required since primary containment limits are not exceeded, b. required to ensure the energy released during an RPV blowdown I can be accepted.
.
. c. required since the downcomers are now exhausting to the torus free air spac d prohibited since the SRV Tail Pipes are now exhausting to the torus free air spac !
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SENIOR REACTOR OPERATOR  Paga 37 OUESTION: 054 (1.00)
To initiate a reactor scram when the control room has been evacuated, it is undesirable to deenergize the RPS busses as the means of scramming because:
    ,
a. ' nuclear instrumentation needed to monitor reactor power will become denergize b. pressure control using turbine bypass valves will be lost after the scra c. RPV level control will unnecessarily transfer from feedwater to HPC I d. groups I, ll, Ill, and VI isolations will be defeate j QUESTION: 055 (1.00)
The following conditions exist:
- The plant is operating at 75% powe At 0800 one Safety Relief Valve opene At 0802 EOP-03 has been entered due to torus water temperature reaching 80 degrees F.
At what point should a Reactor Recirculation pump speed reduction and manual reactor scram be performed?
a. Immediately when it is determined that the SRV cannot be reclose b. At 081 c. When torus temperature reaches 120 degrees d. When the " unsafe"_ region of the Heat Capacity Temperature Limit curve is entere ...-. . . .. .. .- - . - . - - . - - .. - . . . .
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SENIOR REACTOR OPERATOR    P:go 38 '
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QUESTION: 056 (1.00)
Drywell spray was initiated in accordance with EOP-03. As drywell  1 temperature and pressure are decreasing, the unacceptable region on the Drywell Spray initiation Limit curve is entered at a Drywell temperature of 250 degrees ~ Which ONE of the following is the REQUIRED action?
a. Secure drywell spray when drywell pressure drops below psig, b. Secure drywell spray when torus bottom pressure drops below psi c. Adjust drywell spray as necessary to maintain operation within the Drywell Spray Initiation limit curv ;
l d. Immediately secure drywell spra l l
QUESTION: 057 (1.00)
A loss of feedwater heating has occurred. Which ONE of the following is the REQUIRED immediate operator action?
Run back Reactor Recirculation flow until:
a. reactor power has been reduced 25% below its pretransient level without regard to current total core flow, b. total core flow has been reduced to 36 Mlb/hr without regard to the current reactor power level, c. reactor power has been reduced to at least 25% below its pretransient level AND total core flow has been reduced to at least 36 Mlb/h d. reactor power has been reduced 25% below its pretransient level OR total core flow has been reduced to 36 Mlb/h _ , . , , . .. . .
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- SENIOR REACTOR OPERATOR    Paga 39 -
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QUESTION: 058 (1.00)'
A startup is in progress with reactor pressure at 900 psig when the "A" CRD pumps trips and the "B" CRD pump cannot be started. .Two accumulator alarms, both in the same nine rod array, illuminat D Which ONE of the following is the required. action?
,
s. Manually scram the reacto b. Determine the cause of the alarms. If both alarms are due to low gas pressure then manually scram the reactor, c. Fully insert one of the rods with an accumulator alarm and    1
        -
disarm its directional control valve , d. Enter LCO to be in Cold Shutdown within 24 hours.'
QUESTION: 059 (1.00)-
While operating at 100% power, a recirculation pump seal failure causes EOP-03 entry on high drywell pressure and high drywell temperatur Following initiation of suppression chamber spray, drywell pressure
- stabilizes at 4 psig, torus bottom pressure stabilizes at 8 psig, and drywell temperature stabilizes at 175 degrees Which ONE of the following actions is REQUIRED?
a. Declare an Unusual Event and initiate drywell spray in
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i accordance with the Primary Containment Pressure leg of EOP-0 b. Declare an Alert and in:tlate drywell spray in accordance with the Drywell Temperature leg of EOP-0 . c. - Declare an Unusual Event. Do not initiate drywell spra d. Declare an Alert. Do not initiate drywell spray.
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: SENIOR REACTOR OPERATOR    Prgs 40
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- QUESTION: 060 (1.00)        I
With the plant operating on the 65% load line, condenser vacuum starts to decrease. Reactor Recirculation Flow is reduced in accordance with    -i plant procedure !
After the Reactor Recirculation Flow reductions the plant will be operating, a. in the scram regio b. In the exit region.-
I c. In the caution zon d. above the MELLA line.
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QUESTION: 061 (1.00)
l The plant is operating at 100% power when control rods start to drift.
i The MAXIMUM number of control rods in a nine rod array that are allowed to drift WITHOUT REQUIRING tho mode switch to be placed in shutdown is:
a. one rod without regard to whether the rods are drifting in or ou b. two rods if rods are drat;ng in and one rod if rods are
- drif ting ou ,
l c. two rods without regard to whether the rods arr liiting in or    I out.
[ d.~ three rods if rods are drifting in and two rods if rods are drifting out, t
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SENIOR REACTOR OPERATOR  Pcgs 41 OUESTION: 062 (1.00)-
The plant is operating at power when a total loss of TBCCW occurs, immediate actions are complete in accordance with plant procedure Which ONE of the following describes RPV pressure and level control?
a. ' RCIC is being used in the' level control mode and HPCI is being
.used in the pressure control mod b. HPCIis being used in the level control mode and RCIC is being used in the pressure control mode, c. HPCI is being used in the level control mode and SRVs are being used to control pressure. RCIC remains shutdow d. RCIC is being used in the level control mode and SRVs are being used to control pressure. HPCI remains shutdow QUESTION: 063 (1.00)
The plant is operating at 100% power when a loss of Bus A5 occur Which ONE of the following action (s) is(are) required?  ,
a. Reduce reactor power to maintain steam tunnel temperature below
'
160 degrees b. If steam tunnel temperature exceeds 160 degrees F scram the reactor and close the MSIV c. If steam tunnel temperature exceeds 160 degrees F scram the reactor. Maintain the MSIVs ope d. If steam tunnel temperature exceeds 160 degrees F commence a normal plant shutdow I SENIOR REACTOR OPERATOR  Pcga 42 I
      ,
QUESTION: 064 (1.00).
:
With the plant at 100% power on the 100% load line, reactor water level j
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starts to decrease due to unknown causes. . Level is currently + 25 inches and is trending down at 1/2 inch per minut I l
Which ONE of the following is the required action assuming water level continues to fall?    :
      :
a. Insert rods using the RPR rods until below 70% load line, then l reduce core flow to 36-40 Mlb/h .)
I b. Insert rods using the RPR rods until below 70% load line, then
      '
reduce recirculation pump speed to minimu c. Reduce recirculation pump speed to minimum, then insert rods as necessary to exit the caution zone, d. Reduce core flow to 36-40 Mlb/nr, then insert rods using the RPR rods until below 70% load line, then reduce recirculation pump speed to minimu .
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QUESTION: 065 (1.00)
The plant is operating at 100% power with the "B" Reactor Recirculation Pump scoop tube locked when a reactor scram occur Which ONE of the following actions are REQUIRED?
a. Direct a licensed operator to manually position the "B" Reactor Recirculation MG set scoop tube to minimum spee b. Direct any member of the operating crew to manually position the "B" Reactor Recirculation MG set scoop tube to minimum spee c. Unlock the scoop tube, if possible, then run the "B" Reactor Recirculation pump to minimum spee d. Trip the "B" Reactor Recirculation Motor Generator Set.
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SENIOR REACTOR OPERATOR  Pigs 43 QUESTION: 066 (1.00)
Given the following conditions:
- A fuel leak occurs and as a result the reactor is manually scramme Due to the fuelleak, the CRD HCU east and west areas radiation levels reach 1200 mR/hr and 1250 mR/hr respectivel The west Scram Discharge Volume vent and drain valves have failed ope Under these conditions, Alternate RPV depressurization is:
a. not required since the CRD HCU east and west areas are considered the same area, b. required in order to protect secondary containment from failin c. required to allow the scram to be reset and the primary system leak isolated, d. not required since there is no primary system discharging into secondary containment, i
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' SENIOR REACTOR OPERATOR  Prga 44
OUESTION: 067 (1.00)
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The following conditions exist:
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A seismic event has caused the torus suction lines to both Core Spray loops to crack downstream of the Core Spray Suction (MO-1401-3) valve These cracks result in the water level in the SE and NW
;  Quadrants to reach 8 inches and 10 inches above the floor respectivel Efforts to lower the water level are only able to maintain leve There is no primary system discharge into secondary containmen !
Which ONE of the following is required by EOP-047  ;
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a. Isolate the Core Spray suction from the toru b. Maintain the Core Spray suction aligned to the torus, c. Perform Alternate RPV Depressurizatio d. Transfer the Core Spray suction for both loops to the CST.
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OUESTION: 068 (1.00)    i Which ONE of the following conditions violates secondary containment integrity?
a. Both drywell personnel access doors are ope b. Reactor water cleanup MO-1201-2 (RWCU Suction) valvo is failed '
open.
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c. Reactor building ventilation is secured due to dampers failing close d. One refuel floor exhaust isolation damper is failed open with the other refuel floor exhaust isolation damper open and fully operable.
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: SENIOR REACTOR OPERATOR  Pega 45 OUESTIONi 069 (1.00)
The following conditions exist:
- A reactor startup is in progress
- The Reactor Mode Switch is in "Startup/ Hot Standby"
- Reactor pressure is 850 psig The main turbine is tripped
-- A valid Group Iisolation has occurred
- All systems operated as designed Which ONE of the following conditions caused the Group Iisolation?
a. Low main steam line pressure b. Two main steam lines isolating c. High main steam tunnel temperature d. High reactor water level
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l QUESTION: 070 (1.00)    i l
A steam leak in the drywell has occurred and the control room crew has entered EOP-01 and EOP-03. TI-9019 and TRU-9044 on panel C903 are l
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broken. In accordance with the data contained in the attached 2.1.27, which ONE of the following is the instrument run temperature for the "A" channel instruments?
a. 208 degrees F b. 210 degrees F c. 216 degrees F d, 220 degrees F
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SENIOR REACTOR OPERATOR    Pag 3 46  ,
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i OUESTION: 071 (1.00)      ,
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initiating suppression chamber spray prior to torus bottom pressure  l
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reaching 16 psig prevents fatigue failure of l
a. SRV Tail Pipe b. Torus Drywell vacuum breakers.
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c. downcomer l
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d. the Reactor Building-Torus vacuum breaker !
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QUESTION: 072 (1.00)
a l  The following conditions exist:
  - A core off-load is in progres The Refuel Bellows Seal Rupture alarm is received followed 2 minutes later by the Fuel Pool Low Level alar Currently an irradiated bundle has been removed from the core but is still above the reactor vesse Which ONE of the following is the REQUIRED action?
a. Immediately evacuate the refuel floor and leave the bundle hoisted above the reactor vessel, b. Return the bundle to the in-core position that it came from,  j c. Place the bundle in the nearest open in-core positio d. Place the bundle in the nearest open fuel pool positio i
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SENIOR REACTOR OPERATOR  pig 3 47 OUESTION: 073 (1.00)
The following conditions exist:
-
The reactor is shutdow At 1600 all RPV water level indication was lost due to electrical problems and EOP-16 was entere At 1630 conditions to flood the RPV were established with A, B, and D SRVs open and RPV pressure 52 psig above torus pressur At 1640 electrical power was restored and water level can be determine Which ONE of the following actions are REQUIRED?
a. Immediately exit EOP-16 and enter EOP-01 at L-1 and P- b. Continue vessel flooding until 1819 then immediately exit EOP-16 and enter EOP-01 at R-1, c. Concurrently execute EOP-16 and EOP-01 at L-1 and P- d. Continue vessel flooding until 1819 then stop all injectio Verify that RPV level decreases before the MCUTL is reached, i then exit EOP-16 and enter EOP-01 at R- I I
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SENIOR REACTOR OPERATOR-    P gs 48 QUESTION: 074 (1.00)
"
The following conditions exist:
A failure to scram has occurre '
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- No boron has been injecte '
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Reactor power is 30%.
; - The Main Turbine is trippe The Main Condenser is availabl ,
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  .orus water level is norma Due to difficulty in establishing suppression pool cooling, the
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. Heat Capacity Temperature Limit (HCTL) was exceeded.
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Which ONE of the following states the proper method of controlling reactor pressure?    ,
j a. Reactor pressure should be reduced using the main turbine  1 bypass valves to stay below the HCTL curv b. Reactor pressure should be reduced using the SRVs to stay below l
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the HCTL curve.
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c. Alternately depressurize using the main turbine bypass valve i
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d. Altemately depressurize using the SRVs.
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OUESTION: 076 (1.00)
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Which ONE of the following actions allow the operator to disregard NPSH limits?
a. After a successful reactor scram, Core Spray is being used to  l maintain level between -125 to +45 inche b. After a successful reactor scram, LPCI is being used to maintain level between + 12 to +45 inche c. Durin0 an ATWS, LPCI is being used to maintain level between-155 to -140 inches after level was lowered until reactor power dropped below 3%.
d. During an ATWS with Alternate RPV Depressurization required and all SRVs INOPERABLE, LPCI is being used for injection.
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SENIOR REACTOR OPERATOR    pig 3 49 QUESTION: 076 (1.00)
The following conditions exist:
- A steam leak occurs just upstream of the Main Turbine Stop Valves with ooth MSIVs in the "A" main steam line failing to clos A reactor scram is successful in inserting all rods full Both Main Stack Process Radiation Monitors have been reading 2.5E+4 for the last 25 minute Off-site release rate projections are 2 R/ hour Whole Body at the site boundar Select the correct action and its reaso Under these conditions the preferred method of depressurizing the RPV is using:
a. SRVs because of the scrubbing potential of the torus water, b. SRVs because the heat removal capability is greater than the Main Turbine Bypass Valves, c. Main Turbine Bypass Valves because the hotwell is the preferred heat sin d. Main Turbine Bypass Valves because the heat removal capability is greater than the SRV s
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i SENIOR REACTOR OPERATOR  Pcgs 50
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QUESTION: 077;(1.00)
- With a Reactor Building Vent Radiation Hi-Hi Alarm present, EOP-04 ,
directs the operator to verify secondary containment isolation of 1 Reactor Building Heating and Ventilation and the initiation of Standby
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. Gas Treatment Syste j
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'This verification will ensure that:
      !
a, the Reactor Building atmosphere is contained at a positive pressure until it can be treated and release I b. a trected and controlled ground release of the activity is j provide j i c. a treated and controlled elevated release of the activity is provide l d. both the primary and secondary containments are maintained at a
;
slightly negative pressure, i
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.      1 i OUESTION: 078 (1.00)
The following conditions exist:    i
1
- A reactor startup is in progress with RPV pressure at 500 psig.
" - It is determined the "A" Channel of Group i PCIS has one reactor high water level switch (16A-K105A) that will NOT tri The MINIMUM time allowed to place 16A-K105A in the tripped condition is:
a. one hour, b. two hour c. six hours, d. twelve hours.
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- SENIOR REACTOR OPERATOR  Pcge 51
- QUESTION: 079 (1.00)    ,
~ The following conditions exist:
- ' A successful automatic reactor scram occurred on high reactor ;
pressur The main condenser is ava'ilable but not currently in servic The operator is attempting to stabilize pressure between 900-1060 psig using SRV l Re-establishing the main condenser as a heat sint::
a.- is not allowe i b. is preferred but is allowed only if no valid MSIV isolations exis c. is required immediately after valid MSIV isolation signals are overridde d. is only allowed if the SRVs become unavailabl ;
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QUESTION: 080 (1.00)    l With the plant operating at 100% power, the control room becomes uninhabitable because of toxic gas. Evacuation is ordered and only the immediate Actions of Pf4PS Procedure 2.4.143 were carried ou At this point reactor water level is being maintained by:
a. Reactor Feed Pumps and CR b. RCIC and CR :
c. HPCI and CR d. CRD onl >
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SENIOR REACTOR OPERATOR  Prgs 52 OUESTION: 081 (1.00)-    *
With the plant at 100% power, an MPR and EPR f ailure caused the turbine stop valves to close and the turbine bypass valves to remain close Reactor pressure peaked at 1330 psig at which time the reactor scrammed on high flu Select the statement below that correctly describes the transien a. No safety limit violation occurred. The Stop Valve closure scram was the only RPS trip failur J b. No safety limit violation occurred. The Stop Valve closure scram was not the only RPS trip failure, c. A safety limit violation occurred. The Stop Valve closure scram was the only RPS trip failur d. A safety limit violation occurred. The Stop Valve closure scram was not the only RPS trip failur l
QUESTION: 082 (1.00)
Following a reactor scram, the Mode Switch should be taken to Shutdown as soon as possible in order to:
a. disable the low steam pressure isolatio b. enable the high reactor water level isolatio c. insert another scram signal for 2 seconds, d. allow MSIV closure without generating a scram signa _
SENIOR REACTOR OPERATOR  P:ga 53 '
OUESTION: 083 (1.00)
The following conditions exist:
- The reactor was shutdown at 023 Due to loss of level indications, EOP-16 was entered at 103 At 1100 flooding conditions were established with 3 SRVs ope Flooding was stopped as soon as Flooding Cornpletion Time was reached, t
Assuming RPV levelinstruments do not respond, which ONE of the 16116 wing is the LATEST time at which injection must be reinitiated?
- a. 1214 b. 1217 c. 1254
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d. 1257 OUESTION: 084 (1.00)
A worker in the Emergency Response Organization had 100 mrem TEDE for the current year and 2.5 Rem TEDE lifetime prior to the declaration of an emergency. Which ONE of the following is the MAXIMUM TEDE this worker can receive over the course of the emergency without special authorization?
a. 2.4 Rem b. 2.5 Rem    I c. 4.9 Rem d. 5.0 Rem    i
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e-SENIOR REACTOR OPERATOR  Pago 54  ;
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QUESTION: 085 (1.00)    j
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A surveillance on the Reactor Water Cleanup High Flow Isolation is du l
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Which ONE of the following describes how the duration of the surveillance is tracked and when the inoperability clock begins and ends?
a. The surveillance is tracked in the NOS Logboo The clock starts when the system is removed from service and ends when the system is returned to normal lineu b. The surveillance is tracked in the NOS Logboo The clock starts when the system is removed from service and ends when the NWE signs off the surveillanc l c. The surveillance is tracked using an LCO Maintenance Planning i
Checklist. The clock starts when the system is removed from I
i  service and ends when the system is returned to normal lineu d. The surveillance is tracked using an LCO-Maintenance Planning Checklist. The clock starts when the syt. tem is removed from  :
-,  service and ends when the surveillance is signed off by the  I l
  ' work group.


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SENIOR REACTOR OPERATOR  P:gs 55
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QUESTION: 086 (1.00)
With the plant at 5% power, a closed motor operated valve located in the drywell must be tagged in the closed positio Which ONE of the following is the proper method for determining that the valve is in the closed position?
a. The position should be first verified by the indirect method before power is isolated. The isolation of the power supply may then be performed. Independent verification of the power supply is not require b. The position should be first verified and independently verified by the indirect method before power is isolated. The isolation of the power supply may then be performed and independently verifie c. The first verifier should enter the drywell for verification of valve position. The independent verifier may perform an indirect verification of remote valve positio d. The first verifier and the independent verifier should make separate drywell entries for verification of valve positio QUESTION: 087 (1.00)
Which ONE of the following may enter the Controls Area without receiving permission from the NWE/NOS or Control Room Operator?
a. Operations Department Manager
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b. NRC Resident inspector ej. Station Director d. The Outside Nuclear Plant Reactor Operator (NPRO)
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SENIOR REACTOR OPERATOR  P:gs 55 QUESTION: 086 (1.00)    .
          *-T9FuW'4 '
With the plant at 5% power, a closed motor operated valve located in the drywell must be tagged in the closed position.
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Which ONE of the following is the proper method for determining that the valve is in the closed position?
a. The position should be first verified by the indirect method before power is isolated. The isolation of the power suppl may then be performed independent verification of the power supply is not required, b. The position should be first verified and independently verified by the indirect method before power is isolated. The isolation of the power supply may then be performed and independently verifie c. The first verifier should enter the drywell for verification of valve position. The independent verifier may perform an indirect verification of remote valve positio d. The first verifier and the independent verifier should make separate drywell entries for verification of valve positio l l
QUESTION: 087 (1.00)
Which ONE of the following may enter the Controls Area without receiving permission from the NWE/NOS or Control Room Operator?
a. Operations Department Manager b. NRC Resident inspector c. Station Director d. The Outside Nuclear Plant Reactor Operator (NPRO)
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SENIOR REACTOR OPERATOR  Pigs 56
- QUESTION: 088 (1.00)
Absent a basis to assign a longer duration, Which ONE of the following is the normal duration of a temporary modification?
a. Installation until the end of the shift b.- 6 weeks following installation c. 6 months following installation d. installation until the end of the refueling outage QUESTION: 089 (1.00)
The MINIMUM amount of parallel watchstanding REQUIRED in order to reactivate an NRC reactor operator license is:
a. 8 hour i b. 40 hour )
i c. seven 8 hour shift l d. five 12 hour shift l l
QUESTION: 090 (1.00)
 
Which ONE of the following is the MINIMUM REQUIRED protective equipment  l for handling Sodium Hypochlorite outdoors?  1 Safety Goggles and:
a. Rubber Gloves b. Rubber Gloves and Apron, Rubber Safety Boots, Forced Air Respirator    i c. Rubber Gloves and Apron, Rubber Safety Boots d. Rubber Gloves and Apron, Rubber Safety Boots, Respirator i
 
SENIOR REACTOR OPERATOR  Prgs 57 QUESTION: 091 (1.00)
Which ONE of the following conditions would allow a fail open air operated valve to be DANGER tagged in the closed position?
      '
a. The valve is gagged in the closed position with a device to ensure it does not change stat '
b. The DANGER tag is only for equipment protection and no maintenance will be performed under this tagou c. The air supply to the valve is also DANGER tagged in the open positio d. A " Human Red Tag" is assigned to monitor the status of air to the valve.


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QUESTION: 092 (1.00)
Which ONE of the following conditions PROHIBIT the use of a " Human Red Tag"?
a. The only qualified tagger available to be a " Human Red Tag" is
- a member of the work grou b. Two isolation points are required to provide isolation.
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c. The work is expected to take 2 hours to complet d. The work is expected to take 1 hour to complete with only 1/2 hour left in the current shif l l
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SENIOR REACTOR OPERATOR  Prgs 5 '
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l QUESTION: 093 (1.00)    1
An offsite fire department is responding to the site during a fire in /
  - vital are )
  ' Which ONE of the following describes the security reouirements in order !
  - to allow access to the protected area / vital area?  l i
a.' The fire truck and firemen must be searched prior to entering ;
  ~ the protected area. No additional search is required prior to entering the vital area provided security escorts the' tea b. No search is required of the fire truck or firemen prior to entering the protected area provided security escorts the team, ,
however both the truck and firemen must be searched prior to >
entering the vital are :
l c. No search is required of the f;c - truck or firemen prior to l entering the protected area or vital area provided security escorts the tea l l
d. No search is required of the fire truck or firemen prior to entering the protected area or vital area provided security and operations department escort the tea )
QUESTION: 094 (1.00)
You are working in a Hot Particle Control Zone (HPCZ) in a double set of protective clothing. Which ONE of the following is the proper method of removing the protective clothing when exiting the area?
a. Remove both sets of protective clothing at the step off pad at the exit of the HPC b. Remove both sets of protective clothing at the step off pad at
  . the exit of the buffer zon c. Remove the outer set of protective clothing at the step off pad at the exit of the HPCZ and the' inner set of protective clothing at the step off pad at the exit of the buffer zone, d. Remove the outer set of protective clothing at the step off pad at the exit of the buffer zone and the inner set of protective clothing at the step off pad at the exit of the HPCZ zon ;...  .- ,
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SENIOR REACTOR OPERATOR  Pzg3 59 l l
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QUESTION: 095 (1.00)    l An ALERT has been declared. Which ONE of the following describes the l REQUIRED emergency notification?  I l
l a. The NRC must be notified within 15 minutes after the declaration of the ALERT. State and local agencies must be 1
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notified immediately thereafter, not to exceed 1 hou b. State and local agencies must be notified within 15 minutes after the declaration of the ALERT. The NRC must be notifind immediately thereafter, not to exceed 1 hou c. The NRC must be notified within 1 hour after the declaration of the ALERT. State and local agencies must be notified immediately thereafter, not to exceed 1 hour and 15 minutes, d. State and local agencies must be notified within 1 hour after the declaration of the ALERT. The NRC must be notified immediately thereafter, not to exceed 1 hour and 15 minutes QUESTION: 096 (1.00)
Which ONE of the following describes the required manning of the Fire Brigade?
The Fire Brigade shall consist of five members:
a. including the Brigade Leader. Two of these persons may also be part of the crew required for safe shutdown of the plan b. including the Brigade Leader. These persons may not be part of
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the crew required for safe shutdown of the plan c. excluding the Brigade Leader. Two of these persons may also be part of the crew required for safe shutdown of the plan d. excluding the Brigade Leader. These persons may not be part of the crew required for safe shutdown of the plan . . . . - ..
SENIOR REACTOR OPERATOR  P:ga 60 l
OUESTION: 097 (1.00)
l During an emergency, a reasonable action that departs from Technical  l Specifications must be taken immediatel in accordance with PNPS procedures, which ONE of the following MUST  ;
approve taking this action?    1 a. An on shift licensed Reactor Operator and on shift licensed Senior Reactor Operator    i i
b. A licensed Senior Reactor Operator only c. A licansed Senior Reactor Operator and the Operations Department Manager d. A licensed Senior Reactor Operator and the Operations Department Manager and the Plant Manager  j i
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l OUESTION: 098 (1.00)
You have worked the foDowing schedule:
- Thursday 1st scheduled day off    l
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Friday 2nd 7 am to 7 pm    I
- Saturday 3rd 7 am to 7 pm    )
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Sunday 4th 7 am to 3 pm
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Monday 5th 7 am tn 3 pm
- Tuesday 6th 7 am to 9 pm
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Wednesday 7th 7 am to 3 pm
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Thursday 8th 7 am to ?
Which ONE of the following represents the LATEST you can be required to work on Thursday the 8th, without special approval being granted?
(Assume turnover time is NOT included)
a.3pm b. 5pm c. 7 pm d. 9 pm i
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SENIOR REACTOR OPERATOR    Pzga 61 QUESTION: 099 (1.00)
The only individual available for a call-in for TSC staffing informed the Nuclear Watch Engineer (NWE) on the phone that he has consumed alcohol within the previous 5 hour Which ONE of the following describes the individuals ability to work in the TSC?    :
a. not permitted to work in the TS b. permitted to work in the TSC provided the individual informs the NWE that alcohol has not impaired his ability to work in the TSC. A blood alcohol concentration test is at the NWE discretion, based on the NWE phone discussions with the  ;
individua c. permitted to work in the TSC only if a blood alcohol concentration test is performed upon arrival on site and the concentration is less than 0.0 d. permitted to work in the TSC only if a breathalyzer  l test is performed upon arrival on site and the blood to alcohol ratio is greater than or equal to !
l QUESTION: 100 (1.00)
A procedure is currently being performed which requires the installation l of a jumper. It is discovered that the procedure directs the jumper I placement in a position that would cause an unexpected ESF actuation. A change to the jumoer position is require Which ONE of the following is the required method to revise this procedure to change the jumper position?  l a. Editorial correction b. SRO Change c. Minor Revision d. Major Revision    ]
  ( * * * * * * * * * * END OF EXAMIN ATION * * * * * * * * * *)
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  ' BOSTON EDIS0N  RTYPE H6.02
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PILGRIM NUCLEAR POWER ST. TION Procedure No. 2.1.27 DRYWELL TEMPERATURE INDICATION i
REQUIRED REVIEWS  REVIEWERS AND APPROVERS hE. A$ oves /NNW 6b7M
    "" Writ *"  '**'''
Thi"k  '*d""*'
STAR Act 8)
    'wcynical Reviewer
      'gg)q,V Date Review  *# 9 # ^ * '' u' 9/#6/
Validator '
Date' W SAFETY REVIEW E0"!9ED/
    /i f Procedure A' ner
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e/d
      /Dgte NOT REQUIRED N/+
QAD Man'ager ORC REVIEW REQUIRED /    Date MT REQ'J:"C0 AA L  lo /k 194
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ORCC{plirman '
Date 0m Ibb smlk  fobt9/94 sporpible anager  / 1Dhte Effective Date: /0/d8 9%
020095    2.I.27 Rev. 3
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REVISION LOG REVISION 3 Date originated 5/94 Paaes Affected Descriotion 4 Add PDC 92-58 to Reference ~5,7,10,12,14,15 Revise Kaye nomenclature and delete channel points from old Kaye recorder in accordance with PDC 92-5 Editorial 2C Date Originated 3/93 Paaes Affected Descriotion-7,9,11,13 Delete references to Station Honeywell Computer System as it is obsolet Editorial 2B Date Originated
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Paoes Affected pescription      i
 
4,5,7 Editorial corrections to reflect new E0P numbers and entry  l conditions and to add new Editorial Correction rev bar  l identification I Editorial 2A Date Originated Paaes Affected Qgscriotion 4,5,7 Incorporated editorial corrections to Main Control Room Panel Labels per PDC 87-78 REVISION 2 Date Originated Paaes Affected Descriotion All Reformat to comply with PNPS 1.3.4- .
2.1.27  Rev. 3 Page 2 of 15
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,    .IABLE OF CONTENTS
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EA9A PURPOSE AND SC0PE................................................. 4 REFERENCES........................................................ 4 3.01 . DEFINITIONS....................................................... 4 DISCUSSION........................................................ 4 PRECAUTIONS AND LIMITATIONS....................................... 6 PREREQUISITES.....................................................    -6' PROCEDURE......................................................... 7 ATTAC HME N T S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8
' ATTACHMENT 1 - TE-5050 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE ESTIMATE..................................... 9 ATTACHMENT 2 - TE-8125 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE DETERMINATION............................... 10 ATTACHMENT 3 - TE-5050 TEMPERATURE ELEMENTS - INSTRUMENT RUN T EM PE RATUR E E ST I MAT E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 ATTACHMENT 4 - TE-8125 TEMPERATURE ELEMENTS . INSTRUMENT RUN TEMPERATURE DETERMINATION............................... 12 ATTACHMENT 5 - TE-5050 TEMPERATURE ELEMENTS - LOCATION INFORMATION..... 13 ATTACHMENT 6 .TE-8125 TEMPERATURE ELEMENTS - LOCATION INFORMATION..... 14
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2.1.27 Rev. 3 Page 3 of 15
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.  .  . . _ .0 - PURPOSE AND SCOPE    -
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This Procedure provides instructions for determining Drywell bulk temperature when
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the Emergency Operating Procedures _(EOPs) require measurement of this parameter.
 
- REFERENCES
      , DEVELOPMENTAL
[1] PNPS Technical Specifications Table 3. [2] PNPS Technical Specifications Tables 3.2.H and 4. [3] PDC 87-78C, Improvements to Labels, Nameplates on Main Control Room Panels i
[4] PDC 92-58, Kaye Recorder Replacement  ! IMPLEMENTING
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[l] PNPS 2.2.49, " Primary Containment Cooling System"
[2] PNPS 8.7.1.4.2, ' Primary Containment Integrated Leak Rate Test" DEFINITIONS None DISCUSSION
[1] The following sections of the Emergency Operating Procedures require measurement of Drywell temperature:
, (a) E0P-1, RPV Control: RPV Water Level Instrument Run temperatures associated with the RPV Saturation Temperature Figure of Caution (b) E0P-2, Failure to Scram: RPV Water Level Instrument Run temperatures
:  associated with the RPV Saturation Temperature Figure of Caution (c). E0P-3, Primary Containment Control:
    .
  (1)' Entry condition _(150"F)
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  (2) Drywell temperature path (3) RPV Water Level Instrument Run temperatures associated with the RPV
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Saturation Temperature Figure of Caution (4)- Figure 5: (SPDS 031) DSIL (Orywell Spray Initiation Limit)
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2.1.27 Rev. 3 Page 4 of 15
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- -  .- . .- . - . . . . - .- - .. GUSSION (Continued)
(d) E0P-4, Secondary Containment Control: RPV Water Level Instrument Run  ,
temperatures associated with the RPV Saturation Temperature Figure of
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Caution (e) E0P-16, RPV Flooding E0P-26, RPV Flooding, Failure To Scram:
  (1) Temperatures near the RPV Water Level Instrument Reference Leg vertical run [2] Drywell temperature is normally monitored in the Control Room by using TRU-9044, DRYWELL TEMP / PRESS Recorder, and TI-9019, DW TEMP Indicator, on Panel C903. TRU-9044 receives its input from a single temperature element located at a relatively low elevation in the Drywell. TI-9019 receives its input from a single temperature element located just below the neck of the Drywell. Both of these temperature elements measure ambient Drywell air space temperatur [3] The TE-5050A through P temperature elements are used to evaluate Drywell tem >erature with respect to Technical Specifications limits (refer to Technical Specifications Table 3.2.H). The Drywell locations of these elements are listed in Attachment 5. These elements are used to monitor  ,
Drywell temperature for Technical Specifications requirements because of their I reliability, location, and their redundancy (dual-element RTDs). In addition,  I these temperature elements are the primary elements used for the Primary Containment Integrated Leak Rate Tes [4] Local Drywell air temperature indication is supplied by the TE-8125 series  I temperature elements. The TE-8125 series temperature indication consists of 20 RTDs located throughout the Drywell which provide input to the Kaye Tem Computer (refer to Attachment 6).
 
[5] When TRU-9044 and TI-9019 are not available, selected Drywell temperature elements are used to estimate an average temperature near the RPV water level instrument runs and an average bulk Drywell temperature. Temperatures near the RPV water level instrument runs are monitored by those thermal elements which are located in the upper elevations of the Drywell since mest of the instrument runs are found in this region of the Drywell. Bulk average Drywell  '
temperature is a weighted average temperature based on the volume of the Drywell. By averaging more readings from the lower region of the Drywell (which contains most of the Drywell air space) than from the upper region of the Drywell, a representative average Drywell temperature is obtained. More sophisticated methods to calculate a_ weighted average Drywell temperature are available, as part of the ILRT Procedure, PNPS 8.7.1.4.2. The method outlined
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in this Procedure, however, attempts to balance the complexity and time
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consuming aspects of the sophisticated approach against the requirement to rapidly obtain a value for Drywell temperature suitable for use in the E0Ps.
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  -. PRECAUTIONS AfC LIMITATIONS
[1] The Drywell temperature shall be maintained within the following limits when the reactor coolant temperature is above 212 (a) Above elevation 40': $ 194*F  ;
(b) Equal to or below elevation 40': s 150*F Upon determination that the Drywell temperature at any elevation has exceeded the above limits, the Drywell temperature at each elevation shall be logged every 30 minutes. The Drywell temperature shall be reduced to within the above limits within 24 hours; otherwise corrective action shall be as specified in Technical Specifications Sections 3.2.H.2 and 3.2.H.3.
 
i (Tech Spec 3.2 H.1)
[2] If the Drywell temperature has exceeded either limit of Technical Specifications Section 3.2.H.1 for greater than 24 hours, an engineering
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evaluation shall immediately be initiated to assess potential damage and render a determination of ability of safety related equipment to perform its intended functio If either limit of Technical Specifications Section 3.2.H.1 has been exceeded for greater than 24 hours, ferther action to justify continued operation shall be determined by an engineering evaluation which must be completed within one week. (Tech Spec 3.2.H.2)
[3] If the requirements of Technical Specifications Section 3.2.H.2 have not been I met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours. (Tech Spec 3.2.H.4)
[4] If the Drywell temperature at any elevation exceeds 215*F and the temperature cannot be reduced to below 215 F within 30 minutes, a reactor shutdown shall be initiated and the reactor shall be in cold shutdown condition within 24 hours. (Tech Spec 3.2.H.4)
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[5] When reactor coolant temperature is above 212*F, the Drywell air temperature '
limits will be determined by reading the instruments listed in Techaical Specifications Table 3.2.H. These instruments shall be logged once per shift, and each reading compared to the limits of Technical Specifications Section 3.2.H.1. (Tech Spec 4.2.H.1)
J PREREQUISIT_E.ji
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None
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- CEDURE
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[1] DETERMINE bulk Drywell temperature using one of the following methods (listed in order of preference):
(a) SELECT the higher of the valves indicated on TI-9019, DW TEMP Indicator, and TRU-9044, DRYWELL TEMP /F 'SS Recorder (Panel C903).
 
CAUTION  ,
i j The instruments listed below are not environmentally qualified for use in a harsh '
environment. Under accident conditions, they should only be used if either  j TI-9019 or TRU-9044 is not available for us (b) Highest probable Drywell temperature from EPIC points DRY 002 or DRY 00 (c) For a more representative bulk teaperature, AVERAGE the TE-5050 series RTDs using the computer points in accordance with Attachment (d) For a more representative bulk temperature, AVERAGE the TE-8125 series RTDs using the Kaye Temp. Computer in accordance with Attachment . (e) All of the TE-5050A through P series RTDs can be read locally at Panel C85, '
Reactor Building El. 23' East, for Attachment I dat [2] DETERMINE RPV water level instrument run temperature using one of the  !
      '
following methods (listed in order of preference):
(a) SELECT the higher of the values indicated on TI-9019 and TRV-9044  !
  (Panel C903). l (b) AVERAGE the TE-5050 series RTDs in ecordance with Attachment (c) AVERAL .he TE-8125 series RTDs using the Kaye Temp. Computer in accordance with " '.achment (d) The TE-5050A through P RTDs can be read locally at Panel C85, Reactor l Building El. 23' East, for Attachment 3 dat [3] Additional information on Drywell temperature elements and location is contained in PNPS 2.2.49, " Primary Containment Cooling System". 1
 
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2.1.27 Rev. 3 Page 7 of 15
 
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, ATTACMENTS ATTACHMENT 1 - TE-5050 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE ESTIMATE ATTACHMENT 2 - TE-8125 TEMPERATURE ELEMENTS - BULK DRYWELL TEMPERATURE. DETERMINATION ATTACHMENT 3 - TE-5050 TEMPERATURE ELEMENTS - INSTRUMENT RUN TEMPERATURE ESTIMATE ATTACHMENT 4 - TE-8125 TEMPERATURE ELEMENTS - INSTRUMENT RUN TEMPERATURE DETERMINATION ATTACHMENT 5 - TE-5050 TEMPERATURE ELEMENTS - LOCATION INFORMATION ATTACHMENT 6 - TE-8125 TEMPERATURE ELEMENTS - LOCATION INFORMATION
 
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-    - ._  . _ - - _ ATTACHMENT 1 Sheet 1 of 1 TE-5050 TEMPERATURE ELEMENTS BULK DRYWELL TEMPERATURE ESTIMATE
E. Thomas Boulette    3 Distribution w/ encl:
[1] SELECT one temperature element in each group of temperature elements Alg!
Region I Docket Room (with concurrences)
RECORD its temperatur [2] COMPUTE the average temperature as follows:
PUBLIC Nuclear Safety Information Center (NSIC)
(a) Average - (A + B + C + D + E + F)/6 T'.ME TE-5050 EPIC GROUP  COMPUTER POINT ELEMENT #    TEMPERATURE ( F)
NRC Resident inspector        ;
A DRY 002 A -----OR---- -------------- ------- ------- ------- ------- ------- -------
R. Conte, DRP M. Conner, DRP C. O'Daniell, DRP P. Milano, NRR A. Wang, NRR W. Dean, OEDO c
B DRY 004 i
DOCDESK Inspection Program Branch, NRR (IPAS)
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E DRY 010 B -----OR---- ---- - --- ------- ------- ------- ------- -- ---- -------
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G DRY 014 C -----OR---- -------------- ------- ------- ------- ------- ------- -------
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H DRY 116 L        l DRY 122      '
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D -----0R---- -------------- ------- ------- ------- ------- ------- -------
DOCUMENT NAME: A: PILG9706. REP
M DRY 124 K
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DRY 120 E -----0R---- -------------- ------- ------- ------- ------- ------- -------
To reeche a copy or this dorpment lascate la the bott 'C' = Copy without attachment / enclosure T = Copy whh attachmenvenclosure 'N' = No copy OFFICE - @lgRjM  l Al/DRS l Rl/  Rl/  l Ri/
J DRY 118 N
NAME DM /  *
DRY 126 F -----OR---- -------------- ------- ------- ------- ------- ------- -------
GMeyer (, ,
p DRY 130
DATE QF/27/97  09/8)/97 "j" 08/ /97 08/ /97 08/ /97
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AVERAGE Performed By  Date  Reviewed By  Date 2.1.27 Rev. 3 Page 9 of 15
OFFICIAL RECORD COPY i
 
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ATTACHMENT 2 Sheet 1 of 1 TE-8125 TEMPERATURE ELEMENTS BULK DRYWELL TEMPERATURE DETERMINATION
[1] SELECT one temperature element in each group of temperature elements 88Q
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RECORD the temperature indicated on the Kaye Temp. Compute [2] COMPUTE the average temperature a: follows:
  (a) Average - (A + B + C + 0 + E + F)/6 TlME  .
TE-8125 GROUP ELEMENT #  TEMPERATURE (*F)
 
A ----0R----- -------- --------- --------- --------- --------- -------- l l
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8 .... 0R.... ........ ......... ......... ......... ......... ........ 1
 
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C -----OR---- -------- --------- --------- --------- --------- --------
 
13 D -----OR---- -------- --------- --------- --------- --------- --------
 
15 E -----OR---- ----- -- --------- --------- --------- --------- --------
 
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F -----OR---- -------- ------- *- --------- --------- --------- --------
 
AVE: GE Performed By' Date Reviewed By Date 2.1.27 Rev. 3 Page 10 of 15
 
ATTACHMENT 3 Sheet 1 of 1 TE-5050 TEMPERATURE ELEMENTS INSTRUMENT RUN TEMPERATURE ESTIMA1E
[1]. DETERMINE.the rack (s) of concern 8tlQ RECORD the indicated temperature for each element in that group.


[2] COMPUTE the average temperatur Instrument Runs for Rack 2205 A Channel Instruments TLME
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!TE-5050  EPIC ELEMENT COMPUTER POINT
#    TEMPERATURE ( F)
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A
........... ......
DRY 002
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Zw . . . ....... ....... ....... ....... .......
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........... ........ . ..... . .. .. ....... ....... ....... ....... .......
AVERAGE Instrument Runs for Rack 2206 B Channel Instruments TLME TE-5050  EPIC ELEMENT COMPUTER POINT
#    TEMPERATURE (*F)
..... ..... ...... . ..... . $ ....... ....... ....... ....... .......
D  DRY 008 2.Ir
........... ..................  ....... ....... ....... ....... .......
4.....
E  DRY 010 LIO .
AVERAGE Performed By  Date  Reviewed By  Date 2.1.27 Rev. 3 Page 11 of 15
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ATTACHMENT 4 Sheet 1 of 1
 
TE-8125 TEMPERATURE ELEMENTS INSTRUMENT RUN TEMPERATURE DETERMINATION
[1] DETERMINE the rack (s) of concern A151 RECORD the indicated temperature for each  l element in that grou !
[2] COMPUTE the average temperatur Instrument Run for Rack ?205 A Channel Instrument T;:ME TE-8125 ELEMENT #  TEMPERATURE (*F)
 
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  [b
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10 -,g AVERAGE
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Instrument Runs for Rack 2206 B Channel Instrument T::ME TE-8125 ELEMENT #  TEMPERATURE (*F)
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Pdyrn Nuc a P w Staten Plymouth. Massachusett 02360 L J. 00 Met vu Ps wwa Nuom ope'etens   August 18, 1997
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  "'8 8''' *" D""'"    BECo Ltr. 2.97. 084 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 License DPR 35 Docket 50-293
2.[h
  - SUBJECT:- REPLY TO NOTICE OF VIOLATION (REFERENCE NRC INSPECTION REPORT N /97 06)
  ........ ......... ......... ......... ......... ........
Enclosed is Boston Edison Company's reply to the Notice of Violation contained in the subject inspection repor The following commitments are made in this letter:
 
. Prior to the present class of license candidates beginning their on shift training phase, additional guidance will be incorporated into the license candidate shift qualification book. This guidance will
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. specify that continuous, significant control manipulation by a common mechanism (e.g.,
AVERAGE Performed By  Date  Reviewed By  Date 2.1.27 Rev. 3 Page 12 of 15
recirculation flow or control rod movement) will count as a single significant manipulation, regardless of the size of the overall plant power change. This action will be completed by January 1,199 . A requirement for specific leview of each license candidate's significant control manipulations by training department personnel will also be added to the candidate shift qualification book prior to the present class of !icense candidates beginning the on shift phase of their training. This action will be completed by January 1,199 Please do not hesitate to contact me if there are any questions regarding the enclosed repl ~
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1 J. Olivier RLC/dmc/vio97 06 Enclosure 1: _ Reply to Notice of Violation-
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ATTACHMENT 5 Sheet 1 of 1
W KO3 wl "Ytf
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LTE-5050 TEMPERATURE ELEMENTS LOCATION INFORMATION
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Temperature EPIC Elevation Azimuth Area Element Point ID (feet) (degrees) Monitored
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TE-5050A DRY 002 86 0 2' out from vessel below an exh. register TE-50508- DRY 004 89 180 l' out from vesse TE-5050C DRY 006 86 50 4' out from vessel above supply register
- TE-50E00 DRY 008 90 330 2' below head exh. hol TE-5050E DRY 010 60 270 2' out from bio-shield.
 
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4 out from bio-shiel ~
  -TE~-5050F DRY 012 60 90 TE-5050G DRY 014 40 270 10' out from bio-shield under Main Steam Lin TE-5050H DRY 116 40 90 10' out from bio-shield under Main Steam Lin TE-5050J DRY 118 35 0 l' from CRD area inside wal TE-5050K DRY 120 35 180 l' from CRD area inside wal l l
TE-5050L DRY 122 22 205 13' out from CRD area outside !
wal TE-5050M DRY 124 22 45 13' out from CRD area outside ,
wal l TE-5050N DRY 126 15 270 8' out trom CRD rea nutside wal TE-50500 DRY 128 15 0 On CRD area outside wall.
 
N TE-5050P DRY 130 12 125 10' out from CRD area outside wal ~
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Y ATTACHMENT 6 Sheet 1 of 2
 
TE-8125 TEMPERATURE ELEMENTS LOCATION INFORMATION
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Temperature Elevation Azimuth Area Element (feet) (dearees) Monitored TE-8125-1 90 285 Head Exhaust TE-8125-2 90 210 Head Exhaust
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TE-8125-3 85 180 3' out from vessel
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TE-8125-4 85  0 4' out from vessel above ductwork TE-8125-5 82 300 In exhaust duct TE-8125-6 82  45 In exhaust duct TE-8125-7 80 270 In annulus
: TE-8125-8 80  90 In annulus TE-8125-9 54 270 6' out from  !
bio-shield l
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TE-8125-10 54 90 6' out from  '
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bio-shield TE-8125-ll 40 270 10' out from
,    bio-shield under Main
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Steam Line TE-8125-12 40  90 10' out from bio-shield under Main Steam Line  ,


l TE-8125-13 25 315 On CRD area outside wall
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TE-8125-14 '25 135 On CR0 area outside wall TE-8125-15 19 225- 14' out from CRD area outside wall TE-8125-16 19  45 14' out from CRD m es outside wal .1.27 Rev. 3 Page 14 of 15
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ATTACHMENT 6 Sheet 2 of 2 TE-8125 TEMPERATURE ELEMENTS LOCATION INFORMATION Temperature- Elevation  Azimuth Area Element  (feet)  idearees) Monitored TE 8125-17~ 14  265 On CR0 area outside wall TE-8125-1 On CRD area outside wall TE-8125-19  29  180 l' from CRD area
inside wall TE-8125-105 29  0 l' from CR0 area inside wall
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,' Pilgrim Nuclear Power Station cc: Mr. Alan Wang, Project Manager Project Directorate 13 Office of Nuclear Reactor Regulation Mall Stop: OWF 1402 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockvillo Pike Rockvillo, MD 20852 U. S. Nuclear Regulatory Commission Region 475 Allendalo Road King of Prussla, PA 19400 Senior NRC Resident inspector Pilgrim Nuclear Power Station
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2.1.27 Rev. 3 Page 15 of 15
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SENIOR REACTOR OPERATOR  Pegs 1 ANSWER KEY MULTIPLE CHOICE O23 c 001 b  024 b 002 b  025 d 003 d  026 c 004 c  027 d 005 b  028 c l
006 a  029 6 at A 007 b  030 c    !
 
008 a    }
s  4 000 a  M    -
010 a  033 a 011 a  034 b 012 a  035 b 013 c  036 d    I I
014 b  337 c 015 d  038 d l
016 a  039 d 017 b  040 b 018 c  041 d 019 b  042 c 020 c  043 d 021 d  044 b-    ,
- 022 d  045 d
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SENIOR REACTOR OPERATOR Pags 2 ANSWER KEY
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  .046 c 069 c
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  ,' Boston Edison     Docket N3. 50 293 Pilgrim      License No. DPR 35 Enclosure 1 Reply to Notice of Violation During the Initiallicensing examination conducted by the MC the week of May 5 9.1997, the lead cxaminer identified some of the reactivity manipulations performed by our applicants may not have satisfied the NRC's expectations. This resulted in the following violation of NRC requirements (VIO 50-293/97 06-01):
047 d 070)('C'
NOTICE OF VIOLATION
048 ' d  071 c-t
  *10 CFR 55.31(a)(5) requires, in part, that applicants for operator licenses must have performed five significant control manipulations on the plant that affects reactivity or power level, Contrary to the above, as identified on May 5,1997, two senior reactor operator applications, dated April 18,1997, and submitted to the NRC, had documented five manipulations based on credit taken for multiple manipulations due to the extent of the power change when one manipulation should have been credited. When multiple manipulations due to extended power changes were removed, one applicant had two significant control manipulations, and the other appilcant had three significant control manipulations, This lo a severity level IV violation (Supplement Vil)"
    '
REASON FOR THE VIOLATION An incorrect assessment of conditions substantiating a license candidate's required reactivity manipulations resulted in this violation. The factors that led to this incorrect assessment are discussed in detail in the following paragrap An interpretation of what constituted a "significant control manipulation" was based on reviews of NUREG 1202, " Answers to Questions at Public Meeting Regarding implementation of Title 10, Code of Federal Regulations, Part 55 on Operator Licenses", Regulatory Guide 1.8, " Qualification and Training of Personnel for Nuclear Power Plants",10 CFR 55.59, "Requalification", and 10 CFR 55.31(a)(5).
049 c  072 c 050 a 073' a 051 c  074 d 052 d  075 d A-053 b  076 b 054 b  077 c


055 a  078 a  l l
Following this review, we established an "in-house' listing of examples in the candidate shift qualification book. The listing would provide examples that would satisfy the requirement for a single reactivity manipulation for licensing purposes; this list of examples does not, however, provide limiting conditions describing how the manipulations are to be conducted or counte CORRECTIVE ACTIONS TAKEN AND RESULTS ACHIEVED To ensure the license applicants in question had met the NRC's minimum requirement for reactivity manipulations, those two candidates performed additional reactivity manipulations while the NRC was cvaluating the validity of the challenged manipulations. Following the satisfactory performance of two additional reactivity manipulations for one of the license candidates and three additional reactivity manipulations for the other license candidate, updated licenso applications for those two candidates i
056 a  079 b  l 057 d  080 a g 058 a  081 d 059 d  082 a  I
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061 a  084 d 062 a  085 a 063 c  086 b 064 d-  087 a
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065 d  088 c 066 b  089 b
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067 a 090 e
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  ' SENIOR REACTOR OPERATOR
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wera submitted to the NRC on June 16,1997 (r:f:r:nce BECo lett:r 2.97.062). The id:ntift:d viol: tion
ANSWER KEY
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was entered into our corrective action process by the generation of Problem Report (PR) 97.255 CORRECTIVE STEPS THAT WILL BE TAKEN TO AVolD FURTHER VIOLATIONS Prior to the present class of license candidates beginning their on shift training phase, additional j guidance will be incorporated into the license candidate shift qualification book. This guidance will specify that continuous, significant control manipulation by a cominon mechanism (e.g., recirculation
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ATTACHMENT 2 Facility Comments on Written Examination l
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e      10CFR50.55 h
Boeton Edinon b Pilgrim Nudear Power Station Rocky Hdi Road
    - Plymouth, Massachusetts 02360 L J. Olivier
,  Vice President Nuclear Operations
;. and Station Director May 16,1997 BECo Ltr. 2.97-054
 
U.S. Nuclear Regulatory Commission Region I '
: . 475 Allendale Ruad King of Prussia, PA 19406 Docket No. 50-293 ,
License No. DPR-35 Pilarim's 1997 NRC Written Examination Comments a
1  The written examination administered on May 5,1997, was considered to be an in-depth examination, which fairly tested the six (6) SRO candidate's knowledge in the appropriate areas. After thorough
, . analysis of the content of the examination, it is clear that the use of misleading information, use of the double negative context, and the asking of subjects not important to public health and safety were avoide However, specific requests on several written exam questions are submitted for your consideration in Enclosure 1. Enclosure 2 contains the reference documentation associated with each of the requests.
 
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- Your consideration of these requests is greatly appreciate .
e L. J. Olivier
- PMKINRCEXC Enclosure '
- cc:. See ne'xt page
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cc- Mr. Don Florek Region Allendale Road .
flow or control rod movement) will count as a single significant manipulation, regardless of the size of i  the overall plant power chang ,
King of Prussia, PA .19406 Mr. Alan Wang, Project Manager Project Directorate 1-3
A requirement for specific review of each licenso candidate's significant control manipulations by l
: Division of Reactor Projects - 1/11 Mail Stop: 14B2 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockville Pike Rockville, MD 2085 '
l training department personnel will also be added to the candidate shift qualification book prior to the present class of license candidains beginning the on shift phase of their training, i
' U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555
' Senior Resident inspector  .,
Pilgrim Nuclear Power Station
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ENCLOSURE 1
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l ENCLOSURE 1
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1 Question # 32
;  While operating at 100% power, it is determined that the Main Steam Lin'e High Flow switches on the "B" Main Steam Line will t10T trip under a high flow conditio Which ONE of the following is the MINIMUM REQUIRED action? Direct I&C personnel to manually trip the inoperable switche I
, Direct l&C personnel to manually insert a half Group i isolation on the "B" Group I -
Channe Initiate an orderly shutdown and be in Cold Shutdown Condition within a i  MAXIMUM of 30 hours after the instrument failure.
 
' Initiate an orderly shutdown and have Main Steam Lines isolated within a MAXIMUM of 10 hours after the instrument failure.
 
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ANSWER: d.
 
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p        4 DISCUSSION:      l The stem of this question states,"... it is determined that the Main Steam Line High Flow
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switches on the "B" Main Steam Line will NOT trip under a high flow condition..."
There are two trip systems associated with Group I PCIS, designated "A" and "B". Trip System
  "A" has inputs from MSL High Flow switches comprising two instrument channels, and Trip System "B" has eight inputs from MSL High Flow switches comprising two instrument channels, each steam line is equipped with four switches each, one for each instrument channel (Enclosure 2, Attachment 1, page 1).
 
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; . The stem of the question states that all flow switches on the "B" Main Steam Une are
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inoperable. Since this is the case, there are less than two operable instrument channels for 9 --
both PCIS logic trip systems. (See Enclosure 2, Attachment 1, Page 2)
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;  Since there are less than the minimum operable instrument channels for both trip systems,.
1  Attachment 1, page 3 states, "If the minimum number of operable instrument channels cannot
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be met for both trip systems, place at least one trip system (with the most inoperable channels)
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in the tripped condition within one hour or initiate thc appropriate action required by Table 3.2.A listed below for the affected trip function."
 
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Table 3.2.A requires action "B", which states, " Initiate an orderly load reduction and have Main
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  . Steam Lines isolated within 8 hours".
 
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Since there is no grace period (of one hour) for the two trip system inoperability (vice the one trip system inoperability), there is no obvious correct respons Page 1 of 10
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REQUEST (Question # 321:
Since the correct response is not offered as a choice in the responses, we request that this ,
question be deleted from the examinatio REFERENCE:
PNPS Technical Specifications, Table 3.2.A and associated notes (Enclosure 2, Attachment 1).
 
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Those actions will be completed by January 1,199 DATE WHEN FULL COMPLIANCE WAC ACHIEVED
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:  Full compliance was achloved on June 7,1997, when all required significant reactivity manipulations i h9d been completed for both candidates Following completion of the last significant reactivity manipulation, updated license applications for those two candidates were submitted to the NRC.
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! Question # 76      l The following conditions exist:
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- A steam leak occurs just upstream of the Main Turbine Stop Valves with both  )'
j  MSIV's in the "A" main steam line failing to clos A reactor scram is successful in inserting all rods fully.
 
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- . Both Main Stack Process Radiation Monitors have been reading 2.5+E4 for the last
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25 minute Off-site release rate projections are 2 R/ hour Whole Body at the site boundary.
 
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, Select the correct action and its raaso Under these conditions, the preferred method of depressurizing the RPV is using: SRVs because of the scrubbing potential of the torus wate 'l l
; SRVs because the heat removal capability is greater than the Main Turbine Bypass  i
;  Valve l Main Turbine Bypass Valves because the hotwell is the preferred heat sin .
1 Main Turbine Bypass Valves because the heat removal capability is greater than the SRVs.
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j ANSWER: b.
 
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!        I DISCUSSION:
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Because both answer "a" and "b" select the SRVs as the correct mechanism of depressurizing,  !
l the question then becomes discriminatory as to the basis for doing so. Appendix B of the Emergency Procedure Guidelines states that Contingency #2, Emergency RPV Depressurization may be required to:
Minimize radioactivity release from the RPV to the primary containment and secondary containment, or to areas extemal to the primary containment and secondary  '
: containmen Additionally, Appendix G states that the purpose of the Radioactivity Release guideline is to limit radioactivity release into areas cutside the primary and secondary containment Since distracter "a" implies that SRV's are used because they discharge to the primary containment, "a" can be construed as the correct answer. That is, given the situation provided, the fact that the SRVs tiischarge to the containment via the torus is more significant than the fact that the SRV's heat removal capability is greatedhan the bypass valve Since Appendix B also provides generic guidance that SRV's are used because of their heat
. removal capability, "b" is also correc .
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REQUEST: (Question # 76)
Because both answers "a" and "b" are correct per the EPGs, we request that answers "a" and
"a" both be accepted as correct, and the question be retained in the examinatio REFERENCEi Emergency Procedure Guidelines Appendix B, Section 11, Contingency #2 (OEl  l Document 8390-4B, [ Enclosure 2, Attachment 2]). l l
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4,            1 l Qitestion #'28        )
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,  While operating at 100% power, a control rod is determined to be uncoupled. Attempts
;  to couple the rod have been unsuccessful.
 
. Which ONE of the following states the MINIMUM REQUIRED action o Verify the control rod can be moved with drive pressure and maintain the control rod  ;
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at the target position.
 
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~ 'Fu!!y insert the control rod and hydraulically disarm the CR ; Fully insert the control rod and electrically disarm the directional control valves.-  j Fully hsert the control rod, electrically disarm the directional control valves and then declare the rod inoperable.
 
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ANSWER: J I
i, DISCUSSION:
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The only differentiation between distracter "d" and the correct answer "c" is whether the rod is declared inoperabl ]
If a control rod was uncoupled, it would be declared inoperable by Technical Specifications when the inoperability was discovered. The control rod would then be inserted and electrically disarmed to ensure control rod movement was preclude Taking this action does not eliminate the fact that the control rod was inoperable but does
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allow relief from the requirements of the associated Technical Specification actions for an  l uncoupled control rod. The control rod that was uncoupled would still be administratively    I controlled as an inoperable control rod, even though the action statement of Technical Specification 3.3.F does not have to be applied. At PNPS, if an action has to be taken
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on the part of Technical Specifications, the equipment inoperability is traced through the application of an " Active LCO"in the LCO lo i from a Tech Spec consideration only, the rod is not inoperable. However from an administrative and practical standpoint, the rod is indeed inoperable, and the Active LCO
:  is maintained to control the status of the rod. Therefore, if the cand:date approached the
  ' question from this perspective, distracter "d" can also be considered as an acceptable
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answer,
;,  While Procedure 2.2.87, 5.2.1[3] does state that a rod fully inserted and electrically L disarmed is not inoperable, it references Tech Spec 3.3.A.2 that concems rods that
;  cannot be moved with drive pressure. This statement does not apply to the conditions
  ' identified in the question.
 
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REQUEST: (Question # 28)
We request that distracter "d" also be accepted as correc REFERENCE:
      , PNPS 1.3.34.2 (See Enclosure 2, Attachment # 3)
. - 3.0[1]" Active LCO" Definition
. ' 4.0 " Discussion" Operations Department Manager (Tom Trepanier, (508) 830-8364)
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+ Question # 50 in the event that torus water level cannot be maintained above 95 inches, HPCI is secured in order to prevent: exceeding the Primary Containment Pressure Limi exceeding the Pressure Suppression Pressure, exceeding the Heat Capacity Temperature Limi isolating HPCI on high exhaust pressur ANSWER: DISCUSSION:
The operators at PNPS are provided a " supplemental" approved handout for the study of the EOP procedures (Enclosure 2, Attachment 4). In this handout, the basis for the 95 inches torus level securing of HPCIis not stressed as the PCPL. The fact the exhaust will become uncovered is stressed, and HPCI will then directly pressurize the containment. The wording for the PCPL statement is "may exceed the PCPL", and not
"the basis for the uncovery is the PCPL". Wnen this question is considered, the fact that the primary containment would pressurize is a valid line of thought. From this direction, scrutinization of the choices through the use of the supplied EOPs would lead a candidaa to choose the most limiting curve between the PCPL and the PSP. This would of course be the PSP curve. Based on this line of reasoning, response "b"is considered also to be a valid respons REQUEST: (Question # 50)
We request that distracter"b" also be considered as correc REFERENCE:
EOP-03 Supplemental Training Materials / Flow Charts (Enclosure 2, Attachment 4)
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. Question # 29
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With the plant at power, it is determined that the MO-1001-37 (B loop Torus Spray) and MO-1400-25A (A Loop Core Spray Inboard Injection) valves have failed their operability test. Both valves are currently close The maximum time allowed before the plant must be in COLD SHUTDOWN is: hours ( 1 day) hours (4 days) hours (7 days)
t hours (8 days)
ANSWER: DISCUSSION:
PNPS 2.2.125, " Containment isolation System" lists the valves that are considered to be primary containment isolation valves. An identical listing is contained within the FSA Included in this listing are both the MO-1400-25A and the MO-1001-378 (see Enclosure 2, Attachment 5). As containment isolation valves, the administrative requirements require at least one valve in the line to be deactivated in the isolated position, unless the valve receives any signals other than the isolation signal. Whether the valve receives any other signals (other than the isolation signal) determines whether the valve has to be deactivated electrically or otherwise administratively controlled. If the requirements of this procedure are not met (and the questinn does not provide this information), an orderly shutdown shall be initiated and the reactor shall be in Cold Shutdown within 24 hour ,
This question was designed to test the applicants' ability to determine:
1) the impact of an Inoperable 378 valve on "LPCl* operability; 2) the impact of an inoperable 378 valve on the Containment Cooling Loop's Operability and, 3) the overall effect of 1 and 2 when coupled with an inoperable Core Spray syste At least one applicant, (during a followup interview), interpreted this question as a test of !
his ability to recognize that:
1) Both valves are PCIS valves 2) That at a minimum, the 25A would need to be deactivated since it receives an Auto
  ' Open signal and, 3) Determine the corrective actions for failed PCIS valve Since the questions asks for the maximum time allowed before the phnt must be in COLD SHUTDOWN, if a candidate were to assume that the question is testing his Page 6 f 10 g
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knowledgs of PCIS, thin it is rs: sort:bla that ths candidits would choso "c" es tha correct response, given that no other actions are take REQUEST: (Question # 29)
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Due to the two different ways that this question can be Interpreted, we request that both "a" and "b" be accepted as correc REFERENCE:
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PNPS Procedure 2.2.125 (Enclosure 2, Attachment 5)
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l Question # 27 l
'When' valving in a CRD hydraulic control accumulator, the 305-102 (Withdraw Riser . 1 Isolatio'n Valve) and the 305-112 (Scram Discharge Riser Isolation Valve) are required to be open prior to opening the 305-101 (insert Riser Isolation Valve). This prevents:
af a single rod scram when opening the 305-101 valv excessive scram time of that rod in the event of a reactor scram, damage to the accumu5 tor in the event of a reactor scra damage to the drive mechanism in the event of a reactor scra ANSWER: .
DISCUSSION:
While it is stated in PNPS 2.2.87 that valve misoperation dunng the isolation or restoration of a HCU can cause " severe damage to the mechanism", the isolation of the 102 (by itself) can also delay control rod insertion following a scram signal. As seen in Enclosure 2, Attachment 6, with the 102 valve shut, the exhaust path from the
- mechanism is isolated. Since the question does not state the position of the associated rod for the HCU being restored, the candidate could reasonably assume that the rod is in a position other than fully inserted. If the exhaust path is isolated, any scram signal will not permit the mechanism to scram at " normal" rates, if the control rod inserts at al REQUEST: (Question # 27)
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Due to the fact that response "b" contain the phrase " excessive scram time of the rod in the event of a reactor scram", we request that response "b" be also accepted as a correct answe ,
REFERENCE:
Figure 4 from PNPS Training Material (Enclosure 2, Attachment 6)
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ATTACHMENT 3 NRC Resolution of Facility Comments Ques 28 Disagree with BECO comment. The question stem requested " MINIMUM l REQUIRED actions" and the applicants had Technical specifications. The question clearly related to interpretation of technical specification-required actions. As specified in Technical Specification 3.3.A.d, control rod drives that are fully inserted and electrically disarmeo shall not be considered inoperable. Therefore answer d is incorrect. There was no change to the answer ke Ques 29 Agree with BECO comment. There was insufficient information provided in the question to rule out consideration of containment isolation system technical specifications. The answer key was revised to accept a or b as correct answer Ques 32 Agree with BECO comment that there is no correct answer to the question as written. There was no comment provided to this question during the preexam review. The question was deleted from the examinatio l Ques 50 Disagree with BECO. The question asks for the reason HPClis secured at a decreasing torus level of 95 inches. Enabling objective 10 required the applicant to " state the significance of torus levelless than 95 inches as regards the HPCI system." The significance, stated in O-RO-03-04-05, Rev
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1, IG 3 is to prevent exceeding the primary containment pressure limit ?
d (PCPL). The BECO response to the question is attempting to reword the-question to determine the first EOP-03 curve limit reached if HPCI exhaust is not secured at a decreasing torus level of 95 inches. This was not the
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question asked. There is only one correct answer to the question aske While the pressure suppression pressure (PSP) will be exceeded, it has a l
- relatively small consequence. The PCPL will be exceeded, which has a large j consequence, primary containment failure, and is the stated reason in the l reference material for securing HPCI at a torus level of 95 '"ches. There was no change to the answer ke Ques 7 Disagree with BECO comment. As described in 0-RO-03 04-07, Rev 1, IG 9, the purpose of performing alternate depressurization under the conditions of
' the question is to reduce the driving head and flow of any primary leak by rapidly reducing the pressure. In 0-RO-03-04-09, Rev 1, IG 18 the SRVs are used because the heat removal capability (40% power) is greater than the main turbine bypass valves and the RPV will be depressurized sooner. Tha basis for venting containment, when required, using the torus vents considers the scrubbing potential of the torus water to support the torus method as the preferred method. Venting of the primary containment was not required based on the conditions given in the question. Therefore, the only correct answer to this question was answer b. There was no change in the answer ke +
 
I ATTACHMENT 4 SIMULATION FACILITY REPORT Facility License: DPR 35 Facility Docket No: 50 293 Operating Test Administration: May 6-9,1997
 
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 f of a 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 IIfM  DESCRIPTION
 
None
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Latest revision as of 06:17, 18 December 2021

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-293/97-06
ML20216H215
Person / Time
Site: Pilgrim
Issue date: 09/10/1997
From: Meyer G
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To: Boulette E
BOSTON EDISON CO.
References
50-293-97-06, 50-293-97-6, NUDOCS 9709160178
Download: ML20216H215 (3)
v • d • e


Text

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September 10, 1997 l

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I E. Thomas Boulette, PhD i Senior Vice Presideat Nucleer  !

Boston Edison Company i Pil9tim Nuclear Power Station 600 Rocky Hill Road Plymouth, Massachusetts -02360 5599 l SUBJECT: NAC INSPECTION No._50 293/97-06 (REPLY)

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Dear Dr. Boulettei

This. letter refers to your August 18,1997 correspondence, in response to our July 18,1997 lette i'

Thank you for informing us of the corrective and preventive actions documented in your letter to assure that applicants for initial operator examinations will have properly  ?

performed the required five significant control manipulations. These actions will be  ;

i examined during your next licensed operator examinatio Your cooperation with un is appreciate

Sincerely, i

S Glenn W. Meyer, Chief ,

Operator Licensing and Human Performance Branch ,

Division of Reactor Safety Docket No. 50 293 ,

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E. Thomas Boulette -2 l

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L. Olivier, Vice President . Nuclear and Station Director +

T. Sullivan, Plant Department Manager  :

N. Desmond, Regulatory Relations >

D. Tarantino, Nu: lear Information Manager i

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R. Hallisey,- Department of Public Health, Commonwealth of Massachusetts j i The Honorable Therese Murray  !

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The Honorable Joseph Gallitano i B. Abbenet, Departmont of Public Utilities

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Chairman, Plymouth Board of Selectmen

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Chairman, Duxbury Board of Selectmen'  :

Chairman, Nuclear Matters Committee  !

Plymouth Civil Defense Director _ _ -

P. Gromer, Massachusetts Secretary of Energy Resources  !

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_ ~J. Milier, Senior issues Manager

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- J. Fleming -.

A. Nogee,- MASSPIRG >

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Office of the Commissioner, Massachusetts Department of Environmental Quality Engineering i Office of the Attorney General, Commonwealth of Massachusetts

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T. Rapone, Massachusetts Executive Office of Public Safety i Chairman, Citizens Urging Responsible Energy l 1 Commonwealth of Massachusetts, SLO Designee ,

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E. Thomas Boulette 3 Distribution w/ encl:

Region I Docket Room (with concurrences)

PUBLIC Nuclear Safety Information Center (NSIC)

NRC Resident inspector  ;

R. Conte, DRP M. Conner, DRP C. O'Daniell, DRP P. Milano, NRR A. Wang, NRR W. Dean, OEDO c

DOCDESK Inspection Program Branch, NRR (IPAS)

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DOCUMENT NAME: A: PILG9706. REP

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To reeche a copy or this dorpment lascate la the bott 'C' = Copy without attachment / enclosure T = Copy whh attachmenvenclosure 'N' = No copy OFFICE - @lgRjM l Al/DRS l Rl/ Rl/ l Ri/

NAME DM / *

GMeyer (, ,

DATE QF/27/97 09/8)/97 "j" 08/ /97 08/ /97 08/ /97

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OFFICIAL RECORD COPY i

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Pdyrn Nuc a P w Staten Plymouth. Massachusett 02360 L J. 00 Met vu Ps wwa Nuom ope'etens August 18, 1997

"'8 8 *" D""'" BECo Ltr. 2.97. 084 U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 License DPR 35 Docket 50-293

- SUBJECT:- REPLY TO NOTICE OF VIOLATION (REFERENCE NRC INSPECTION REPORT N /97 06)

Enclosed is Boston Edison Company's reply to the Notice of Violation contained in the subject inspection repor The following commitments are made in this letter:

. Prior to the present class of license candidates beginning their on shift training phase, additional guidance will be incorporated into the license candidate shift qualification book. This guidance will

. specify that continuous, significant control manipulation by a common mechanism (e.g.,

recirculation flow or control rod movement) will count as a single significant manipulation, regardless of the size of the overall plant power change. This action will be completed by January 1,199 . A requirement for specific leview of each license candidate's significant control manipulations by training department personnel will also be added to the candidate shift qualification book prior to the present class of !icense candidates beginning the on shift phase of their training. This action will be completed by January 1,199 Please do not hesitate to contact me if there are any questions regarding the enclosed repl ~

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1 J. Olivier RLC/dmc/vio97 06 Enclosure 1: _ Reply to Notice of Violation-

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,' Pilgrim Nuclear Power Station cc: Mr. Alan Wang, Project Manager Project Directorate 13 Office of Nuclear Reactor Regulation Mall Stop: OWF 1402 U. S. Nuclear Regulatory Commission 1 White Flint North 11555 Rockvillo Pike Rockvillo, MD 20852 U. S. Nuclear Regulatory Commission Region 475 Allendalo Road King of Prussla, PA 19400 Senior NRC Resident inspector Pilgrim Nuclear Power Station

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,' Boston Edison Docket N3. 50 293 Pilgrim License No. DPR 35 Enclosure 1 Reply to Notice of Violation During the Initiallicensing examination conducted by the MC the week of May 5 9.1997, the lead cxaminer identified some of the reactivity manipulations performed by our applicants may not have satisfied the NRC's expectations. This resulted in the following violation of NRC requirements (VIO 50-293/97 06-01):

NOTICE OF VIOLATION

  • 10 CFR 55.31(a)(5) requires, in part, that applicants for operator licenses must have performed five significant control manipulations on the plant that affects reactivity or power level, Contrary to the above, as identified on May 5,1997, two senior reactor operator applications, dated April 18,1997, and submitted to the NRC, had documented five manipulations based on credit taken for multiple manipulations due to the extent of the power change when one manipulation should have been credited. When multiple manipulations due to extended power changes were removed, one applicant had two significant control manipulations, and the other appilcant had three significant control manipulations, This lo a severity level IV violation (Supplement Vil)"

REASON FOR THE VIOLATION An incorrect assessment of conditions substantiating a license candidate's required reactivity manipulations resulted in this violation. The factors that led to this incorrect assessment are discussed in detail in the following paragrap An interpretation of what constituted a "significant control manipulation" was based on reviews of NUREG 1202, " Answers to Questions at Public Meeting Regarding implementation of Title 10, Code of Federal Regulations, Part 55 on Operator Licenses", Regulatory Guide 1.8, " Qualification and Training of Personnel for Nuclear Power Plants",10 CFR 55.59, "Requalification", and 10 CFR 55.31(a)(5).

Following this review, we established an "in-house' listing of examples in the candidate shift qualification book. The listing would provide examples that would satisfy the requirement for a single reactivity manipulation for licensing purposes; this list of examples does not, however, provide limiting conditions describing how the manipulations are to be conducted or counte CORRECTIVE ACTIONS TAKEN AND RESULTS ACHIEVED To ensure the license applicants in question had met the NRC's minimum requirement for reactivity manipulations, those two candidates performed additional reactivity manipulations while the NRC was cvaluating the validity of the challenged manipulations. Following the satisfactory performance of two additional reactivity manipulations for one of the license candidates and three additional reactivity manipulations for the other license candidate, updated licenso applications for those two candidates i

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wera submitted to the NRC on June 16,1997 (r:f:r:nce BECo lett:r 2.97.062). The id:ntift:d viol: tion

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was entered into our corrective action process by the generation of Problem Report (PR) 97.255 CORRECTIVE STEPS THAT WILL BE TAKEN TO AVolD FURTHER VIOLATIONS Prior to the present class of license candidates beginning their on shift training phase, additional j guidance will be incorporated into the license candidate shift qualification book. This guidance will specify that continuous, significant control manipulation by a cominon mechanism (e.g., recirculation

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flow or control rod movement) will count as a single significant manipulation, regardless of the size of i the overall plant power chang ,

A requirement for specific review of each licenso candidate's significant control manipulations by l

l training department personnel will also be added to the candidate shift qualification book prior to the present class of license candidains beginning the on shift phase of their training, i

<

Those actions will be completed by January 1,199 DATE WHEN FULL COMPLIANCE WAC ACHIEVED

Full compliance was achloved on June 7,1997, when all required significant reactivity manipulations i h9d been completed for both candidates Following completion of the last significant reactivity manipulation, updated license applications for those two candidates were submitted to the NRC.

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