Text

Requalification Program Evaluation Rept 50-293/86-45OL on 861001-03.Evaluation Results:Four Senior Reactor Operators Passed Exam & One Reactor Operator Failed Written Exam. Several Areas Identified as in Need of Improvement
	ML20212A054
Person / Time
Site:	Pilgrim
Issue date:	02/20/1987
From:	Collins S, Cresendo F, Howe A, Keller R; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20212A003	List: IR 05000293/1986045; ML20211R080;
References
	50-293-86-45OL, NUDOCS 8703030273
	Download: ML20212A054 (115)
	v • d • e

{{#Wiki_filter:_ - _ - _ _ _ -.__. U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Requalification Evaluation Report EVALUATION REPORT NO.

50-293/86-30 (0L) DOCKET NO.

50-293 LICENSE NO. DPR-35 LICENSEE: Boston Edison Co.

800 Boylston St.

, Boston, MA 02199 FACILITY NAME: Pilgrim Nuclear Station DATES OF EXAMINATION: October 1-3, 1986 CHIEF EXAMINER: m 2 L(67 F.J. Cresc@zo, ReactoQngineer (Examiner) date REVIEWED BY: dd., A b A /f #7 A.G. Howe, R6 actor Engineer (Examiner) date REVIEWED BY: d.

c M. 8 8N. M /s.

,2 -/ 9 -8 7 R.M. KellerV Chief,@rofects Section No.1C date APPROVED BY: I3TYiLLt /[AlhW 2hhf87 ST J. Collind, Depu~ty 21 rector, DRP '~date SUMMARY: The administration of the facility's annual requalification i examinations was audited by the NRC. The requalification program was evaluated as satisfactory. All of the four Senior Reactor Operators passed the examinations and one of two Reactor Operators failed the written examination.

Several areas within the requalification evaluation process were identified as in need of improvement and the facility training staff committed to improving these areas.

8703030273 070224 PDH ' ADOCK 05000273 PDH i i -- , - - - m- -. -.. _ - .,_ ..-,._.,-,,-- - -.-_,_,_._ _.,-.__-,._ _... -,. - _,. ... -. - - - - - _ - - - - -

__ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ DETAILS 1.

EXAMINATION RESULTS: R0 SRO TOTAL Pass / Fail Pass / Fail Pass / Fail Written Examination 1/I 4/0 5/1 Oral Examination 2/0 4/0 6/0 Evaluation of Facility Written Examination Grading: Satisfactory Overall Program Evaluation: Satisfactory The inspectors also interviewed or observed other members of the licensee's training department and operations department.

2.

SCOPE The purpose of this inspection was to determine the effectiveness of the licensee's requalification training program used to identify needed areas of retraining for licensed operators. Within this scope, the following specific areas were inspected.

A.

Facility prepared written examinations were reviewed by Region I examiners prior to administration.

B.

Region I examiners observed the administration of the written examinations.

C.

The facility evaluators were observed by Region I examiners during administration of requalification oral examinations.

D.

Parallel grading of completed written and oral examinations was conducted by Region I examiners.

3.

WRITTEN EXAMINATION The facility prepared examinations were reviewed prior to administration.

This review revealed the examinations to be acceptable in content with the exceptions of Sections 4 and 7 (Procedures). With few exceptions, most questions in these sections did not test knowledge of procedures.

Several questions were system " purpose" oriented, others addressed physical operation of systems rather than procedural requirements of system operation.

In general, many of these questions would have been more appropriately placed in sections 2, 3, or 6 (system design and instrumentation). Due to this deficiency, the majority of facility prepared questions in these sections were replaced by NRC prepared questions. A selective few questions in other sections were replaced by NRC prepared questions, however, as noted above, the adequacy of

__ _ _- ._

l l other examination sections was acceptable. The specific substitutions were not made known to the facility prior to administration of the examinations.

) In addition to the deficiencies noted in Sections 4, and 7, the inspectors noted that the examinations, in entirety, required too many candidate responses to be completed in four hours. Consideration was given by the inspectors to reduce the length of the examinations, however, this was not done in order to minimize active NRC involvement in the evaluation ' process.

During post examination reviews, the problems noted above were discussed with the facility training staff.

The training supervisor agreed with the inspector's findings and has committed to improving training department procedures to eliminate these problems in the future.

The facility's proctoring of the examination administration was found to

be adequate. All necessary precautions were taken to eliminate possible examination compromise.

Copies of the completed written examinations were graded by Region I examiners and the NRC's results were compared to facility results. This comparison found that the facility's grading techniques were consistent with those of the NRC examiners. No significant deviations between NRC grading and facility grading results were noted. All deviations were e investigated by the NRC examiners on a question by question basis. No problem areas were identified and in fact, most deviations were due to

harsher facility grading. One Reactor Operator was identified as j needing additional retraining and all the Senior Reactor Operators passed the written examinations.

In accordance with Pilgrim training procedures, the one Reactor Operator will be removed from licensed duties, participate in accelerated training, and successfully complete another written exami- , nation prior to resumption of licensed duties.

4.

ORAL EXAMINATIONS t Two Reacter Operator and four Senior Reactor Operator requalification oral examinations were observed by Region I examiners.

Prior to administra-tion, the NRC examiners discussed the procedures and forms to be used during the exams with the facility evaluators.

Three specific problem l' areas were noted during these discussions.

First, the facility has no specific instructions to describe proper exam administration or correct form usage.

For example, although the form contains a sufficient number of areas to be covered, the examiners were not required to complete all ,' areas nor were they required to complete a minimum number or spectrum of areas. This could result in omission of significant subject areas during

examination administration. Additionally, this could contribute to inconsistencies in subject material covered between examiners or from year to year.

Second, the method of formulating overall pass / fail decisions does not adequately address weaknesses in significant areas. The procedure for determining pass / fail criteria is as follows: . . -..._____,.,.__,,.mr ..7 y_.,,,_._,.m.. ,,,, - - ._____.-,._._._..,,,,,,.._.e, _, _..,,, _ -, _ - - -,, _.,., -, - - m.c-e - ,, _ _

a.

The examiners ask questions of the examinees and subjectively grade the responses on a 1-5 scale.

b.

Upon completion of the exam the points are totaled and then divided by the number of questions asked. An average of 3 or greater is required to pass the exam.

This method results in an objective overall conclusion based on many individual subjective evaluations. The concern with this method is that a significant weakness in one or two areas could easily be masked by good performance in other areas. This is inconsistent with NRC examination practices whereby a candidate must achieve a minimum satisfactory grade in most, if not all, subject areas in order to pass an oral examination.

Third, the form was found to be very difficult to use in conducting an integrated examination.

The form is approximately 14 pages long and is divided into numerous sections.

These problems were discussed with the training staff.

The training supervisor agreed with the inspectors findings and has committed to restructuring the Forms and modifying the procedures to preclude future problems.

During observation of the examinations, the facility examiners were observed for adequacy of subjects covered and level of detail. The NRC examiners did not actively participate in questioning of candidates.

Generally, the facility examiners demonstrated adequate examination techniques within the specific areas covered.

The facility examiners were well prepared and the examinations were well integrated for the subjects covered.

It was noted that some major subject areas were either omitted from all discussion or were only briefly discussed.

Specifically, these areas included Emergency Operating Procedures, Emergency Plan Procadures and Technical Specifications. During follow-up discussions with the facility examiners various reasons were given for these omissions.

In one case the examiner simply forgot to discuss Emergency Operating Procedures.

The inspectors did not feel that the specific omissions were of great concern considering the individual candidates. They did, however, find the omissions to be indicative of a lack of procedural controls or guidelines for administration of oral examinations and incomplete l planning by the facility evaluators.

I The candidates performed well during the oral examinations with the exception of one Reactor Operator who was observed to be marginal.

Some minor generic weaknesses were noted by the NRC examiners.

These conclusions were consistent with the facility corclusions.

5.

FINDINGS The examiners are confident that the facility is capable of identifying licensed operator training deficiencies.

Several areas in need of improvement were identified and discussed with the facility training

-- - L' ) - Si .

,- .. . A ' staff. The most significant of these areas is the need for better quality assurance in written examination preparation, and for more specific instructions regarding administration of oral examinations.

' , Thepic,ilitytrainingstaffagreedwiththefindingsoftheinspectors with no dissenting, comments. Changes to the Pilgrim training department procedures will be heviewed by' Region I in order to monitor continusd; improvements in these areas.

. 6.

J. Exit Interview '-

.- ' NRCAttendees:( Frank Cresc nzo,' Reactor Engineer (Examiner) ' ' Allen Howe, Reactor Engineer (Pxaminer) . ,

Martin McBride { SRI) ' ' Jeff Lyash (RI) . r ,' Facility Attendees: H.

E~. Pedersen, Station Manager E.~Ziemanski, Management Services Manager , S. Hudson, Operations Man:ger P.-Hamilton, Senior Compliance Engineer ' H. Balfour, Staff Assistant ' ' E. Grahm, Compliance Group Leader T. Sullivan, Watch Engineer , D. Hughes, Senior Nuslear Training Specialist - ,

~

- Caramen t.s.

, The Examiners presented the scope and findings as discussed'above with the exception'of' parallel grading results.

, , . Attachments ,- 1.

R0 exam and answer key.

' ' 2.

SR0 exam and answer key.

, o s,, P ' s . ) .4--~,+~ -.-- .---~% r----s

ArrnehenGnY / PILGRIM NUCLEAR P0HER STATION .. "- REACTOR OPERATOR LICENSE REQUALIFICATION EXAMINATION FACILITY: PILGRIM REACTOR TYPE: BHR-GE3 EXAMINATION DATE: EXAMINER: D.H. HUGHES 1A g e",,Q " APPLICANT: Md wh INSTRUCTIONS TO APPLICANT: Ike separate paper for your answers.

Write answers on one side only.

Staple question sheets on top of the answer sheets.

Point values for each question are indicated in parenthesis after the question. A passing grade requires at least a 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination begins.

% OF CATEGORY APPLICANT'S CATEGORY VALUE IQI&t, SCORE VALUE CATEGORY 25.00 25.00 1. PRINCIPLES OF NUCLEAR POWER PLANT OPERATION, THERMODYNAMICS, HEAT TRANSFER, AND FLUID FLOH.

25.00 25.00 2. PLANT DESIGN INCLUDING SAFETY AND EMERGENCY SYSTEMS.

25.00 25.00 3. INSTRUMENTS AND CONTROLS.

25.00 25.00 4. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROLS.

100.00 100.00 TOTALS FINAL GRADE % All work done on this examination is my own.

I have neither given nor received aid.

APPLICANT'S SIGNATURE l ,, '

_ R0-4 SECTION ONE - ' QUESTION 1.01 <3.00)

For each of the following conditions, state whether the change will bring the system closer to, or have no effect on the power at which the onset of transition boiling would occur. Assume all other parameters remain constant for each case.

Briefly explain, 'I'0' a.

Increase in coolant flow rate.

b.

Decrease in coolant subcooling.

(1 0) c.

Decrease in coolant system pressure.

(1.0) ! QUESTION 1.02 (4.oo) The reactor is exactly critical near core end of life (EOL). Control rods are withdrawn to insert.0008 46K/K.

a.

What is the resulting period? (State your assumptions and show all work.)

(1.5o) b.

How long will it take for power to increase by a factor of 10? (Show all work.) (1.50) c.

What would the period be if.0058 4LK/K had been added to the " exactly critical" reactor? (Show your work.)

(1.0) QUESTION 1.03 (4.5) For each of the pairs of conditions listed below, state which condition would have the areater differential rod worth and why.

a.

Reactor moderator temperature of 150 Deg F or 500 Deg F.

(1.5) b.

A rod moved from position 00 to 20 or a rod moved from position 24 to 48 in a reactor performing a startup.

(1.5) c.

For a rod near the outer perimeter of the core or a rod near the center of the core at 100 percent power.

(1.5) QUESTION 1.04 (3.5) l a.

For the following transients, indicate which coefficient of I reactivity tends to change power first and in which direction - alpha T, alpha D or alpha V.

NOTE: No explanation required.

1.

Fast closure of one MSIV.

(.5) 2.

Control rod drop.

(.5) I 3.

SRV lifting and then resetting - consider both in your answer.

, (1.0) (*** CATEGORY 01 CONTINUED ON NEXT PAGE***) t ,

b.

The core void percentage is at a maximua with cini.Tum recirc ficw at

the 1007. rod pattern. Why does the void percentage decrease as power is raised to 1007. with recirc flow? What provides negative reactivity feedback as reduced voids add positive reactivity.

(1.5) . QUESTION 1.05 (2.00) Following a reactor scram, from 1007. power, explain what happens initially to the following parameters (increase, decrease, or remain the same) and why.

a.

Actual core flow.

(.5) b.

CRD flow (through the pumps).

(.5) c.

Control rod drive mechanism temperature (after accumulator has fully discharged).

(.5) d.

Steam line pressure drops (with MSIV's open).

(.5) QUESTION 1.06 (4.00) a.

Briefly describe the effect that transition boiling has upon a fuel bundle that makes it undesireable.

(2.0) b.

Briefly describe how we prevent the occurrence of transition boiling in the core.

(By monitoring what parameter?) (2.0) QUESTION 1.07 (2.00) The reactor is shutdown and cooled down after operation at 1 0 0*/. for a long time. About 24 hours after shutdown, K-eff was measured to be 0.97. Assuming that no core alterations are in progress, that temperature is constant, and that rod posi ti ons are constant, answer the fol1owing: a.

What two factors are affecting K-eff ? (0.50) b. Is positive or negative reactivity being added by each ? EXPLAIN. (1.00) c.

Will K-eff increase or decrease ? (0.50) ! l QUESTION 1.08 (2.00) i ) The plant is being cooled down from hot operating conditions. Given an l initial pressure of 935 psig, what would be the plant pressure two hours i after the start of the cooldown if the cooldown rate was maintained at the maximum allowed under normal conditions? (See attached section of Steam Tables.)

l l . (***END OF CATEGORY Ol***) l l . _ _ _ . _ _ -.. .

. - -- - -.. _ . Table 2: Saturated Steam: Pressure Table G . . $pecifet Volume [nthafpy [ntropy - m '. Abs Press Temp $8t.

$8t Sat.

1 81 $8t $81 Abs hess fl LD/$0 la.

(8hr .$ p

. b0uio - [vap V890r b004 [v8p Yapor b0use [ esp Yapor LD/50 in U1 V eg 'g hr h,g h s, 5,g s ' g g p S mell 32 018 0 016022 33024 33024 0 0003 075 5 075 5 0 0000 21872 2 1872 8 08485 e tt $9 323 0 016032 12355 1235 5 27382 060 1 087 4 0 0542 2 0425 2 0967 8 21 SM 79 % 6 0 016071 641 5 641 5 47623 048 6 096 3 O NS 9446 2 0370 AM

101 74 0 016136 33359 333 60 6973 036 1 05 8 0:16 84 % i 9781

la 16224 0 016407 23 515 73 532 (30 20 000 9 31 1 0.349 6094 4443 le 14 8 19321 0 016592 38404 38 420 lli 26 962 1 43 3 0 2836 5043 7879 10 8 m ' 14 M6 212 00 0 016719 26 782 26 799 18037 970 3 50 5 0 3121 4447 7%8 148e8 m ..es 18 8 213 03 0 016726 26 274 26290 18121 969 7 50 9 0 3137 4415 7552 15 8 30 0 227 % 0 016834 20 070 ' 70 08' ' ' 196 2 I 9601 ' 164 1 0 3682 ,3313

6995 NO

- ' 156 3 0 3358 13%2 1 7320 29 8 ' ". * * * * NO 250 34 0 017009 13 72M 13 7436 218 9 9452 ' de 0 . 26725 0 017151 10 4794 to 49% 236 I 933 6 169 8 0 3921 ,2844 56765 de 0 '. y to t ' 20 02 0 017274 8 4967 8 5140 250 2 923 9 174 1 04112 ,2474 1 6586 ut ?i III 292 71 0 017383 71%2 71736 262 2 915 4 1776 04273 .2167 1 6440 es t . 'f le 8 30293 - 0 017482 6 1875 6 20$0 272 7 907 8 Ito 6 0 4411 1905 l 6316 75 8 Jt a1 312 04 0 0!?S73 5 4536 5 4711 2B21 900 9 ul83 1 0 4534 .1675 1 6208 et 8 NO 320 28 0 0!?659 4 8719 4 8953 290 7 894 6 .,185.3 0 4643 1 1470 1 6113 Ot t 15 0 327 82 0 017740 4 4133 4 4310 298 5 888 6 ll872 0 4743 1284 16027 tes t lit 8 334 79 0 01782 4 0306 4 0484 305 8 883 I 1188 9 048 4 1115 l 595e tte a 120 0 34127 0 01789 3 7097 3 7275 312 6 8778 l190 4 0 4919 0%0 587, 120 0 '

IN G 34733 0 017 % 3 4364 3 4544 315 0 872 8 1917 0 4998 0815 5813 ING las t 353 04 0 01803 3 2010 32190 325 0 864 0 193 0 0 5011 .0681 5752 188 0 19s t 358 43 0 01809 2 9958 3 0139 330 6 863 4 194 1 0 5141 0554 5695 tut let t 36355 0 01815 2 8155 2 8336 336 1 859 0 .195 1 0 5206 0435 . %41 tes t 178 0 36842 0 01821 2 65 % 2 6738 341 2 854 8

l% 0 0 5269 0322 5591 179 1

lie s 373 08 0 01827 2 5129 2 5312 346 2 850 7 1% 9 0 5328

l 0215 5543 las t 19e I 377 53 0 01833 2 3847 2 4030 350 9 846 7 1897 6 0 5384

l 0113

, 5498 100 0 200 0 381 80 0 01839 2 2689 2 2873 355 5 842 8 198 3 0 5438 1 0016

15454 200 0 21s t 38591 0 01844 2 16373 2 18217 359 9 839 1 199 0 0 5490 0 9923 5413 215 6 273 0 38988 0 01850 2 06779 2 08629 364 2 835 4 199 6 0 5540 0 9834 5374 278 1 2N I 393 70 0 01855 l t 7991 1 99846 368 3 831 8

.200 1 0 5588 09748 5336 238 0 i 244 I 39739 0 0!*60 100009 19t F69 372 3 828 4 32008 05634 09665 5299 34s a 2te s E0 9 ' 0 0:p" 824!? 1 84317 376 1 825 0 201 1 0 5679 0 9585 5264 22 0 258 9 404 44 0 01870 75548 17741 F 3799 821 6 201 5 0 5722 0 9508 5230 260 0 270 0 40780 0 01875 69137 1 71013 3t3 6 414 3 201 9 0 5764 0 9433 5197 till 200 0 41107 0 01883 63169 1 65049 3871 815 1 202 3 0 58c5 09%1 SIM 298 9 298 8 41425 0 01885 1 57597 1 59482 390 6 812 0 202 6 0 5844 0 9291 5135 298 8 3es t 41735 0 01889 1 52384 1 54274 394 0 808 9 1202 9 0 5882 0 9223 1 5105 set t 3se 4 431 73 0 01912 1 30642 1J2554 409 8 794 2 1204 0 0 6059 0 8909 L 4968 854 8 des e 444 60 0 01934 1 14162 1 16095 424 2 780 4 1204 6 0 6217 0 8630 3 4847 ese e a s e 4HO 456 28 0 01954 1 01224 1 03179 4373 7675 l204 8 0 6360 0 8378 ' 4738 454 8 bes t 46701 0 01975 0 90787 0 92762 449 5 755 1

204 7 0 6490 08: 41 4639 tes t 860 1 476 94 0 01994 0 82183 0 84171 460 9 7433

.2043 O Mil 0 7936 ' 4461 see t 4547 lu t tot I 486 20 0 02013 0 74962 0 76975 471 7 732 0 l203 7 0 6723 0 7736 4H e 494 89 0 02032 0 68811 0 70843 481 9 720 9 2028 0 6828 0 7552 143si ass a fee 8 503 08 0 020 % 0 635c5 0 655 % 4916 710 2

2038 0 6928 0 7377

.I4304 Tse 8 798 8 510 84 0 0?069 0 58s80 0 60949 Sor 9 699 8 1200 7 0 7022 0 7210 1 4232 fM 8 188 8 51821 0 02087 6 54809 0 % #96 509 8 6896 1199 8 0 7111 0 705) 1 4163 000 0 4H 8 52524 0 0210" 051)97 05342 514 4 679 5 198 0 0 7197 0 6899 1 4096 558 I eIt t 531 95 0 02123 0 47968 0 50091 526 7 M97 'l% 4 0 7279 0 6753 the 8 538 3' O 02141 0 45064 0 47205 534 7 %Q0 194 7 0 7354 0 M12 .l8032 Bee 8 .3970 958 e les8 544 't 0 02159 0 42436 0 445 % 542 6 6504 192 9 0 7434 0 6476 .3910 less e leH t W,53 0 02177 0 40047 0 42224 550 1 6409 191 0 0 7507 0 6344 .3851 1958 8 1100 0 526 28 0 02195 0 37863 0 43058 557 5 631 5 189 1 0 7578 0 6216 . 3794 1188 0 11H g M182 0 02214 0 35859 0 38073 564 8 622 2 11870 0 7647 0 6091 l 3738 11H 8 12ee l 56719 0 02232 0 34013 0 36245 5719 613 0 1184 8 0 7714 05%9 13683 1200 0 1250 8 572 38 0 02250 0 32306 0 34556 578 8 603 3 1826 0 7780 0 5850 13630 125e i 1300 0 57742 0 02269 0 30722 0 32991 585 6 %46 180 2 0 7843 0 5733 l3577 1380 8 13H I 58232 0 02288 0 29250 03 537 592 3 5454 1778 0 7906 0 % 20 ( 3525 135s t 4408 8 587 07 0 02307 027871 0 30178 5988 576 5 175 3 0 7966 05507 l 34'4 lass t 14W O 591 70 0 02327 0 26584 0 28911 605 3 5674 172 8 0 8026 0 5397 13423 84u 8 1901 8 5 % 20 0 02346 0 25372 0 27719 611 7 558 4 ll101 0 8085 0 5288 3373 1988 8 f tM I 600 59 0 023 % 0 24235 0 ?M01 418 0 549 4 tl67 4 05142 0 5152 3324 lust 1888 8 604 87 0 02387 0 23159 0 25545 624 2 540 3 .364 5 0 8199 0 5076 3274 tese s Inte t 609 05 0 02407 0 22143 0 24551 630 4 5313 161 6 0 8254 04973 3225 1658 8 178B 8 613 13 0 02428 0 21174 02E07 436 5 522 2 .158 6 0 8309 04867 3176 17se t 17H I 61712 0 02450 9 20263 0 22713 642 5 513 1 155 6 0 8363 04765 31'? 1750 0 1808 8 621 02 0 02472 0 9390 0 ?!861 648 5 503 8 .52 3 0 8417 04662 3079 lage t 18H 8 624 83 0 02495 0 8558 0 21052 654 5 494 6 49 0 0 8470 0 4561 3030 isse s 1000 0 628 % C 02517 0 7761 0 20278 660 4 4852 45 6 0 8522 0 4459 298i Isse l i ItM e 632 22 0 02541 0 6999 0 9540 666 3 4758 42 0 0 8574 0 4358 2931 1960 0 3ESI 8 635 80 0 02 % 5 0 62 % 0 8831 6721 466 2 38 3 0 8625 0 42 % 2881 2ess e tiesI 642 76 0 02615 0 4885 0 7501 683 8 446 7-22 2 0 882 0 3644 2676 32M t 30 5 0 8727 0 4053 2780 fios a M 2700 0 64945 0 02669 0.3603

6272 495 5 426 7 2300 0 655 89 0 02727 0 2406 0 15133 707 2 406 0 113 2 0 8979 0 3640 , 2569 23e18 3e081 M211 0 02790 6.1287 0 14076 719 0 384 8 103 7 0 9031 0 3430 '2460 2440 0 flat 0 E811 0 02859 010?09 0 13068 731 7 3616 1093 3 0 9139 0 3206 12H5 250s 8 3081 0 67391 0 02938 0 09172 0 12110 744 5 1376 1082 0 0 9247 0 2977 1 2225 29e0 0 2701 8 67953 0 03029 0 08165 0 11194 757 3 312 3 1069 7 0 93 % 02741 1 2097 27eII p 3Est e 684 96 0 03134 0 07171 0 10305 770 7 2 51 10M 8 0 9468 0 2491 .1958 feels - flee t 690 22 0 03262 0 06158 0 09420 785 1 254 7 1039 8 0 9588 0 2215 1603 2004 8 3080 0 69533 0 03428 0 05073 0 08500 801 8 218 4 1020 3 0 9728 01891 .1619 asse t 3108 8 700 28 0 03641 0 03771 0 07452 824 0 169 3 993 3 0 9514 0 1460 i l373 3100 8 33EIS 705 08 0 04472 0 01191 005M3 8755 56 1 9316 1 0351 0 0482 1 0832 3200 1 82eB2' 70547 0 05078 0 00000 0 05078 906 0

906 0 1 0612 0 0000

0612 82e8 7*

g 'Cntal pressure '** - - -

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ R0-4

SECTION TWO QUESTION 2.01 (4.5) . With regard to the reactor core isolation cooling system minimum flow valve (M0-60): a.

What four conditions will cause the valve to close? (2.0) b.

State the specific conditions that must be satisfied to complete the opening logic for the valve.

(2.5) QUESTION 2.02 (6.00) Answer the following concerning the RHR system.

1.

State the RHR pump interlocks.

(1,0) 2.

State the RHR valve interlocks affecting containment cooling and spray valves.

(5.0) QUESTION 2.03 (2.00) Why was a modification (PDC 85-40) performed to install an Ultrasonic Flow Monitor on the Standby Liquid Control System ? DUESTION 2.04 (3.50) The reactor is operating at 85% power when 125 VDC panel D17 becomes and remains de-energized.

a.

What three power supplies would you check in an effort to re-energize the panel ? (1.50) b.

How will this transient affect reactor power ? EXPLAIN.

(1.00) c.

Which of the f ollowing components are NOT available as a result of this transient ? (1.00) 1) Diesel Generator A 2) RHR pump C 3) Core Spray pump B 4) RCIC 5) Primary protective relays on bus A2 (*** CATEGORY 02 CONTINUED ON NEXT PAGE***) _ _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ - _ _

- QUESTION 2.05 (3.00) . State the purpose of the ADS (Automatic Depressurization SYSTEM) and list the auto initiation signals (include setpoints).

QUESTION 2.06 (2.00) State two conditions identified by General Electric as possible operating conditions that may require adjustment of the TCV/TSV scram bypass setpoint (as a result of abnormal core thermal power, to turbine first stage pressure relationship).

QUESTION 2.07 (2.00) State the reason for the modification of PDC 85-53, "RHCU valve 1201-2 jog open feature".

QUESTION 2.08 (2.00) Regarding the Diesel Generators: a.

State the automatic starting signals (include setpoints) for DG's? (1.4) b.

In the event of a LOCA start, which of the diesel enaine protection functions are still capable of shutting down the engine? (.6) l l (***END OF CATEGORY 02***) . -

R0-4 ~ SECTION THREE , QUESTION 3.01 (3.00) . A ground overcurrent situation causes a lockout of 4160 vac bus A-5.

Assuming the A-5 bus was initially powered from the Unit Auxiliary Transformer, answer the following questions True or False.

a.

The Unit Auxiliary Transformer bkr to A-5 will trip.

(.5) b.

The Startup Transformer bkr to A-5 will shut if the Auto Transfer switch is in the ON position.

(.5) c.

A direct result of the A-5 bus lockout is auto start of Diesel Generator "A".

(.5) d.

If Diesel Generator "A" is started or is running when the lockout occurs, its output bkr will automatically shut to supply A-5 with power.

(.5) e.

A lockout signal on bus A-5 will trip any supply bkr connected to it and prevent all power supply bkrs to A-5 from shutting.

(.5) f.

The Shutdown Transformer is the only source available to A-5 under lockout conditions.

(.5) QUESTION 3.02 (4.00) Describe the operating relationships between core spray injection valves (M0-25A and MO-25B) and the presence of a core spray initiation signal.

(NOTE: there are two parts to this answer.)

QUESTION 3.03 (6.00) Concerning Diesel Generator load shedding... a.

What is the purpose of DG load shedding? (2.0) b.

What conditions must be met in order to cause the Diesel Generator load shedding logic to actuate? (List each individual signal in the logic.)

(4.0) (*** CATEGORY 03 CONTINUED ON NEXT PAGE***) t .

.. -- __ - .. - - -. QUESTION 3.04 (4.00) . Concerning the Recirculation Pump Control System: a.

What is the purpose of limiting recirculation pump speed to 28% when - less than 20% total feedwater flow exists? (1.0) b.

What is the reason we limit recirculation pump speed to 65% when operating with less than three feed pumps and a low reactor water level alarm? (1.0)

c.

What is the function of the recirculation runback reset pushbutton and indicating light on panel 904? What precaution is associated with using this reset button? (2.0) QUESTION 3.05 (2.00) , Hill the Core Spray pumps start under each of the following i circumstances? If not, why? (NOTE: Answer part a and b independently of each other.)

) a.

High drywell pressure (2.5 psig).

(.6) b.

Low Rx water level (-49"). (1.4)

QUESTION 3.06 (2.00)

Concerning the Rod Block Monitor: a.

What is the function of the Rod Block Monitoring System? (1.0) b.

What are the minimum number of inputs required for each RBM, to prevent an inoperative alarm? (1.0) QUESTION 3.07 (2.00) , Explain how and why indicated Rx level would respond to the following, (assume all plant parameters -- actual level, steam flow, pressure remain unchanged): a.

Small leak in level transmitter reference leg isolation valve packing gland.

(1.0) b.

Equalizing valve for level transmitter leaks by.

(1.0) QUESTION 3.08 (2.00) Regarding the Turbine Control System (I.P.R.)

Explain why the back-up pressure regulator is normally set at a slightly higher pressure than the controlling pressure regulator.

(***END OF CATEGORY 03***)

e .- - . - - - ,_..._., __ - .. .. . _ . . . - - - -. _ _ _ _ _ _ _ _.. - - -

R0-4 SECTION FOUR - QUESTION 4.01 (3.00) Procedure 1.3.11 " Reset of Lockout Relays and Relay Targets" states that lockout relays and protective relay targets shall not be reset until authorized by one of three supervisors.

a. Who are the three (3) supervisors ? ( 1. 50.' b. What are the exceptions to this rule ? (1.50' QUESTION 4.02 (4.00) a.

Explain the automatic actions that will occur if Fire Suppression System Header pressure is decreasing.

State the normal fire suppression system pressure band.

If no water, is available on site, where can the fire system receive a water supply? '3'0U' QUESTION 4.03 The reactor is at 100% power wher. the three (3) inch instrument air / nitrogen line in the drywell ruptures. Based on procedure 2.4.21 " Double Ended Break of the 3" Instrument Air / Nitrogen Line in the Drywel1": a.

What are three (3) symptoms or automatic actions ? (1.5) b.

What immediate actions should be taken ? (1.5) QUESTION 4 04 (3.oo) Concerning the P.A.M.

panels: a.

What are six of the nine plant parameters which can be monitored at the P.A.M.

panels ? (2.00) b.

What is the one instrument that is not to be used while the recirculation pumps are running and briefly explain why ? (1.Oc.

(*** CATEGORY 04 CONTINUED ON NEXT PAGE***)

.. -. . - - -. .

QUESTION 4.05 (3.00) You are on watch and notice that suppression pool bulk temperature has risen to 88 deg F and is increasing slowly. A HPCI full flow tomt is in progress, a. At what suppression pool temperature should suppression pool cooling be initiated ? (0.75) b.

At what suppression pool temperature should HPCI testing be secured ? (0.75) , c.

At what temperature must the reactor be scrammed ? (0.75) d.

State any emergency procedures you would enter as a result of the initial conditions given above.

(O.75) t QUESTION 4.06 (1.00) Concerning EOP-3 "RPV Level Restoration", step III.F(3) directs the operator to determine if at least two injection subsystems are lined up for injection with pumps running. What is an injection subsystem ? (1.00) , QUESTION 4.07 (4.00) E0P-06 (Primary Containment Control Level) establishes both high level and

low level in the suppression pool as entry conditions.

What are the high and low level setpoints for entry into E0P-06? a.

(1.0) b.

What are the bases for establishing high and low level limits? (3.0) QUESTION 4.08 (4.00) Considering the E0Ps, under what two conditions is Boron injection required? (***** END OF CATEGORY 04

- - - - )

(************* END OF EXAMINATION ***************) - .. . _ __.

, . _ _ _. -. ~. _ _ _ _ _ _ _ _ _ _ _ _,. _ ., _ _ __ ... .

- , TYPICAL N.R.C.

EOUATION SHEET . F e ma v = D/t Cycle officisney a V = TIr28. (cyl) s = V t + 1/2(at2) o " . A = AN A = A e-A t g W = ag a = (Vf - Vo)/t E = ac2 A= En2/tif2 = 0.693/ tift KE = 1/2(mv2) ATg = ( A T. ) - ( A T1 ) PE = agh L n( A T. ) ( AT, ) , 't * Vo + at , , Q + m,hin = W/J + m,hout ', A E = 931 A m ha u + PV I = I e-JAX I = I e-E X c c = dCp 47 her5Ah I = I lo-x/TVL o k = UA ( A T)u Ah = Cp A T TVL = 2.303/#. P = P 10sur(t) h = hg + Xhg H E = 0.693/g o g P = P et/T Q = r*n v 4f/,7 CR (1 - Keff 1) CR (1 - Keff 2)

2 = o ... ..... , - -. - - - -. - - - - - - - - - - - - - - - - - - - - - SUR = 26.06/T WL-pAv sCR = s /1 - Keff c 7 = (f.*/p ) + ((4 p )/Ep) H = 1/(1 - Keff) = CR /CRo

p= A6* + e M = (1 - K,ff o)/(1 - K,gy 1) AT+1.

SDM = (1 - K,ff) (* = 10-4 seconds P " (Kett-1)/Keff - A Keff/Keff , 5 = 0.1 seconds-1 l P = (E$V)/(3x1010) Igdi=Id22 T,= F N Idf=Idh l

. . R/hr = (0.5 CE)/d2(meters) ._ _.. .. -- , WATCR PARAMETERS (atmospheric conditions) MISCELLANEOUS COUVERCIONS , 1 gal. = 8.345 lbm. = 3.70 liters I curie = 3.7 x 1010dps 1ft3 = 7.48 gal.

1 kg = 2.21 lbm Den:ity = 62.4 lbm/fL3 = 1 Em/cm3 1 hp = 2.54x103 Btu /hr = 550 ft-lbf/sec ! Heat of Vaporization = 970 BTU /lbm 1 tr.J = 3.413 x 106 Btu /hr I Heat of Fusion = 144 DTU/lbm 1 in. = 2.54 cm 1 ATtt - 14.1 psi = 29.92 in. Hg.

oF = 9/5 oC + 32

, CC = 5/9 (oF - 32) - ' l . ,

- . D . .,. - .. -, -, -. -, , ,, &

- SECTION ONE QUESTION 1.01 (3.00) ' For each of the following conditions, state whether the change will bring the system closer to, or have no effect on the power at which the onset of transition boiling would occur. Assume all other parameters remain constant for each case.

Briefly explain.

a.

Increase in coolant flow rate.

(1.0) b.

Decrease in coolant subcooling. (1.02 c.

Decrease in coolant system pressure.

<1.o) ANSWER a.

Farther from (.25); each lbm of fluid scends less time in contact with the clad surface (0.75) b.

Closer to (.25): 21h reauired to achieve critical cuality decreases as the inlet enthalov increases (0.75) c.

Farther from (.25); from a lower oressure. hfa is hiaher (o.75) QUESTION 1.02 (4.00) The reactor is exactly critical near core end of life (E0L). Control rods are withdrawn to insert.0008 A K/K.

a.

What is the resulting period? (State your assumptions and show all work.) < 1. 50 3 b.

How long will it take for power to increase by a factor of 10? (Show all work.)

(1.50) c.

What would the period be if.0058 AK/K had been added to the " exactly critical" reactor? (Show your work.)

(1.0) ANSHER Notes to examination grader... Accept.1 sec-1 as 4 (icating a normal up power transient) for 1.

from~ formula sheet) or any value from.08 sec-3 to.1 sec-I (ind part a. of this question.

2.

Accept.1 sec-l as f (from formula sheet) or any large value for J( that is given by the student to represent a prompt critical J (3.01 sec-I is given by the General Electric . academic series as prompt critical f ). This applies to part c.

3.

Accept.0056 as B core for E0L (or a number close to it, obviously representing B core E0L).

(*** CATEGORY 01 CONTINUED ON NEXT PAGE***)

ANSWER: - a.

Assumptions: B core EOL .0056 (.25 pt) ' /* - 10-I4se .1 sec-' T= + 7 - 10-4 +.0056 .0008 (.25 pt) .0008 .1(.0008) 'T .0048 - 60 seconds (1 pt) .00008 , b.

P__ - et/rf (.25 pt) Po 10 - etby 10 - et/60 (.25 pt) in10 - t/60 t - 2.3 (60) (.25 pt) t - 138 seconds (o.75) T - /* + B - f c.

f fR T - 10-4 +.0056 .0058 .0058 .1(.0058) T = 10-4 .0172 seconds .0058 QR if Q added exceeds B core, the reactor is prompt critical (1 pt).

, (*** CATEGORY 01 CONTINUED ON NEXT PAGE***)

QUESTION 1.03 (4.5)

For each of the pairs of conditions listed below, state which conattion would have the areater differential rod worth and why.

a.

Reactor moderator temperature of 150 Deg F or 500 Deg F.

(1.5) . b.

A rod moved from position 00 to 20 or a rod moved from position 24 to 48 in a reactor performing a startup.

(1.5) c.

For a rod near the outer perimeter of the core or a rod near the center of the core at 100 percent power.

(1.5) ANSHER a.

Reactor moderator temperature at 500 deo F (.75), since thermal diffusion lenath will be much creater (.75).

b.

A rod moved from oosition 00-20 (.75) due to the hiaher neutron flux igen (.75) and tlee lower control rod density in this region of the Core.

c.

A rod near the center of the core will have the greatest differential worth (.75) since the local flux is hiahest with resoect to averaae flux in the core (.75).

QUESTION 1.04 (3.5) a.

For the following transients, indicate which coefficient of reactivity tends to thange power first ind in which direction - Elpha T, alpha D or alpha V.

NOTE: No explanation required.

1.

Fast closure of one HSIV.

(.5) 2.

Control rod drop.

(.5) 3.

SRV lifting and then resetting - consider both in your answer.

(1.0) b.

The core void percentage is at a maximum with minimum recirc flow at the 100% rod pattern. Why does the void percentage decrease as power is raised to 100% with recirc flow? What provides negative reactivity feedback as reduced voids add positive reactivity.

(1.5) ANSHER a.

CONDITION COEFFICENT POWER 1.

Fast closure of one MSIV voids (.25) up (.25) 2.

Control rod drop Doppler (.25) down (.25) 3.

SRV l a.

lifting voids (.25) down (.25) b.

reseating voids (.25) up (.25) . t ! l (*** CATEGORY 01 CONTINUED ON NEXT PAGE***) _ __

b.

Power is increased by raising recirculation pump speed, thereby, increasing flow through the core.

Increased flow crovides a shorter ' boilina lenath (.3) which in turn reduces the cercent voids (.3).

Lowerina the cercent voids adds oositive reactivity (.3) from o< v, which will cause a rise in nower until a sufficient increase in oower has caused once aaain a hiaher cercentaae of voids (.3).

However, - durina this oower increase. the fuel temoerature is risina. causina neaative reactivity to be added via e4 D (.3).

Therefore, to achieve a new higher steadystate power where reactivity - 0, the initial positive reactivity addition caused by the removal of voids must be matched by an equal amount of negative reactivity to be added via o< v and erd. Since the fuel temperature increase ( e4 D) contributes some negative reactivity, the void fraction need not contribute as much.

Therefore, at a higher power, void fraction will be lower.

QUESTION 1.05 (2.00) Following a reactor scram, from 100% power, explain what happens initially to the following parameters (increase, decrease, or remain the same) and why.

a.

Actual core flow.

(.5) b.

CRD flow (through the pumps).

(.5) c.

Control rod drive mechanism temperature (after accumulator has fully discharged).

(.5) d.

Steam line pressure drops (with MSIV's open).

(.5) ANSHER Ref. PNPS Heat Transfer Fluid Flow, Thermal Limits Handbook and Recent PNPS SR0 NRC Exam (12/11/84).

a.

Increase (.25) due to the reduction in head loss due to the steamina and void formation reduction (.25).

b.

Flow increases (.25) as pumps attempt to recharce accumulator (.25).

c.

Increase (.25) due to FCV shuttina because of "b" above which reduces coolina water flow (.125) AND Hi temoerature reactor water introduced throuch ball check (.125).

d.

Decrease (.25) due to reduction in head loss due to reduction in steam flow rates (.25).

(*** CATEGORY 01 CONTINUED ON NEXT PAGE***) l l l

QUESTION 1.06 (4.00) . a.

Briefly describe the effect that transition boiling has upon a fuel bundle.that makes it undesireable.

(2.0) ' b.

Briefly describe how we prevent the occurrence of transition boiling in the core.

(By monitoring what parameter?) (2.0) ANSHER a.

Transition boilina involves partial film blanketina and subseauent co11anse of the steam alona the surface of the fuel rod. (.5) The boilina condition is unstable at this time. and steam blanketino occurs random 1v (.5).

The problem is that the clad surface heats uo when steam blankets it. and cools when replaced with liauid (.5).

Constant thermal cyclina leads to material failure (.5).

b.

We monitor Minimum Critical Power Ratio (MCPR) (.67) which compares actual bundle oower to a bundle power at which transition boilina is oredicted to occur (.66) (for a given set of conditions).

By stavina safelv away from " critical oower". transition boilina is avoided (.67).

QUESTION 1.07 (2.00) The reactor is shutdown and cooled down after operation at 1 0 0*/. for a long time. About 24 hours after shutdown, K-eff was measured to be O.97. Assuming that no core alterations are in progress, that temperature is constant, and that rod positions are constant, answer the f ollowing: a.

What two factors are affecting K-eff ? (0.50) b. Is positive or negative reactivity being added by each ? EXPLAIN. (1.00) c.

Will K-eff increase or decrease ? (0.50) ANSWER (2.00) (0.50) a.

Xe and Sm b.

Xe decays so positive reactivity is added.

(0.50) Sm is stable and added to by the decay of Nd so negative reactivity is added.

(0.50) c.

The net effect is that K-eff will increase.

(0.50) REFERENCE PNPS Reactor Theory Handbook p.

24-30.

(*** CATEGORY 01 CONTINUED ON NEXT PAGE***) . ! _ - __ .

_ . . . QUESTION 1.08 (2.00) The plant is being cooled down from hot operating conditions. Given an initial pressure of 935 psig, what would be the plant pressure two hours after the start of the cooldown if the cooldown rate was maintained at the maximum allowed under normal conditions? (See attached section of Steam Tables.)

ANSHER Assume: maximum cooldown rate - 100*F/ hour.

(.4) Thus, 200*F temperature change has occurred.

935 osia - 950 osia (.4) then TSAT - 538.4*F (.4) 538.4*F - 200*F = 338.4*F (.4) If TSAT (t - 2 hours) = 338.4*F, then PSAT - 115.5 psia or 100.5 psig (.4) (Grader: accept answer in psia QI in psig) (***END OF CATEGORY Ol***)

. __ Table 2: Saturated Steam: Pressure Table C , m $pecit c volume fathalpy Entropy Abs Press Temp $81 Sat $8t $8t.

$st $8t Abs Press , Q lbl5e in.

78hr bemd Ev8p V8por bemd Evap Vapor bemd Evap Vapor LD/50 in p t vg v,, g hg h,g h S v , g g s,g sg p i e teell 32018 0 016022 33024 3302 4 0 0003 l075 5 ,075 5 0 0000 2 1872 21572 9 88855 02S 59 323 0 016032 1235 5 1235 5 27 382 060 1 .087 4 0 0542 2 0425 2 0967 825 LM 19 546 0 016071 641 5 641 5 47623 048 6 096 3 0 0925 9446 2 0'70 eM I4 101 74 0 016136 333 59 333 60 6973 036 1 05 0 0 1326 44 % 8'81

le 16224 0 016407 73 515 23 532 30 20 000 9 31 1 0 2349 6094 8443 la le 8 193 ?! 0 016592 38404 38420 61 26 982 1 43 3 0 2836 5043 78 79 le 8 84 80s 212 00 0 016719 26 742 26 799 8017 970 3 50 5 0 3121 4447 7%8 14 805 E 18 4 213 03 0 016726 26274 26290 .8121 969 7 1 509 0 3137 4415 7552 184 Me 227 % 0 016434 20 070 20 087 19627 960 1 1156 3 0 3358 I 3%2 1 7320 Me Me 250 34 0 017009 33 7266 1374% 218 9 Se5 2

164 1 0 3642 3313 6995 Me p/ ;

40 8 26725 0 017151 10 4794 104%5 236 1 933 6 169 8 0 392) 2844 6765 es e Me 281 02 0 017274 84%7 8 5140 250 2 923 9 174 1 0 4112 2474 6586 Me hy.

ss 4 312 04 0 017573 5 4536 5 4711 282 I 900 9 1831 0 4534 1 1675 6208 se e MB 292 71 0 017383 71%2 71736 262 2 915 4 177 6 0 4273 l.2167 6440 es e Me 302 93 0 017482 61875 6 2050 272 7 907 8 .180 6 0 4411 11905 6316 Mt si b Ms 32028 0 017659 4 8779 4 8953 290 ? 894 6 1853 0 4643 1 1470 ,6113 et 8 b IM e 32782 0 017740 4 4133 4 4310 298 5 868 6 (187.2 0 4743 1284 I 6027 its 4

918 8 334 79 0 01782 4 0306 4 0444 305 8 8831 (188 9 0 4434 1115 1 5950 tit t IMS 34127 0 01789 3 7097 3 7275 312 6 477 8 .190 4 0 4919 0960 1 5879 tM8 ' e lat 4 34733 0 017 % 3 4364 3 4544 319 0 872 8 191 7 0 4998

0815 1 5813 IM8

' 540 0 353 04 0 01803 3 2010 3 2190 325 0 a68 0 193 0 0 5071 .0681 1 5752 14s 0

lie s 35843 (L01809 2 9958 3 0139 330 6 8634 194 1 0 5141 0%4 3 % 95 iMt

100 0 363 55 0 01815 2 8155 2 8336 336 1 859 0 1951 0 5206 10435 , %41 188 0 170 0 368 42 0 01821 2 65 % 2 6738 341 2 854 8 1%0 0 5269 l 0322 .5591 tie s , IM t 373 08 0 01827 2 5129 2 5312 346 2 050 7 .196 9 0 5328

0215

'5543 tes t les e 37753 0 01833 2 3847 2 4030 350 9 846 7 ,197 6 0 5384

0113 L 5498 IM 8 fec e 381 80 0 01839 2 2689 2 2873 355 5 842 8 l198 3 0 5438 1 0016 L 5454 290 0 218 e 385 91 0 01844 2 16373 2 18217 359 9 8391

1990 0 5490 0 9%23 5413 31s t 225 0 389 88 0 01850 2 06779 2 08629 3642 835 4

199 6 0 5543 0 9834 5374 22e s 238 s 393 70 0 01855 9799:

99846 368 3 831 8 .200 1 0 5588 0 9748 5336 230 t Mes 39739 0 01860 89909 91769 372 3 428 4 200 6 0 % 34 0 9665 !?99 248 s !st e 40097 0 01865 82452 84317 376 1 825 0 2011 0 5679 0 9585 !?64 25s e 284 8 40444 0 01870 75548 77410 379 9 82 t 6 201 5 0 5722 09%8 5230 Met tre a 40780 0 01875 69137 71013 383 6 818 3 201 9 0 5764 0 9433 !!97 278 8 2ee s 41107 0 01880 63169 65049 3871 015 1 202 3 0 5805 0 9361 5166 fee t 2st e 41425 0 01885 57597 59482 390 6 812 0 .202 6 0 5844 0 9291 5135 tes t 3es e 41735 0 01889 1 57384 1 54274 394 0 808 9 1202 9 0 5882 09223 1 5:05 8es e su 8 43173 0 01912 1 30642 132554 409 8 794 2 12040 0 6059 0 8909 14%8 nee 4es e 444 60 0 01934 1 14162 1 16095 424 2 780 4 12046 0 6217 0 8630 1 4847 40s 8 .

= 4H 8 4M 28 0 01954 1 01224 1 03179 4373 7675 204 8 0 6360 0 8378 .4738 ele e 888 8 46701 0 01975 0 90.87 0 92762 449 5 7551 204 7 06490 0 8148 4639 see s SM 8 476 94 0 01994 0 82183 0 84177 460 9 7413 204 3 0 M11 0 7936 4547 Stel est e 486 20 0 02013 0 74 % 2 0 76975 4717 732 0 203 7 0 6723 0 7730 4461 8ee t EH 6 49489 0 02032 0 68811 0 70843 481 9 720 9 202 8 0 68?8 0 7552 4381 650 0 70s 8 503 08 0 020$0 0 63505 0 655 % 491 6 710 2 201 8 0 6928 0 7377 .4304 ist t 7M e 510 84 0 02069 0 58880 0 60949 500 9 699 8 1200 7 0 7022 0 7210 l4232 75s e Met 518 21 0 02087 0 54829 0 % 896 509 8 6896 l1994 0 7111 0 7051 i4163 sell BH S 525 24 0 02105 0 51197 0 53302 Sle 4 679 5 198 0 0 7197 0 6899 14096 858 0 set e 531 95 0 02123 047%8 0 50091 526 7 669 7 ' 1%4 0 7279 0 6753 4032 ess e tes e 538 39 0 02141 0 45064 0 47205 534 7 660 0 L194 7 0 7358 0 % 12 3970 354 3 tM68 544 58 0 02159 0 42436 0 445 % 542 6 6504 192 9 0 7434 0 6476 3910 tese s leH 8 550 53 0 02177 0 40047 0 42224 550 1 640 9 l1910 0 7507 06344 .3851 tale s lies e 556 28 0 02195 0 37863 0 40058 5575 631 5 .189 1 0 7578 0 6216 .3794 1its e liu s 56182 0 02214 0 35859 03eM3 564 8 622 2 L187 0 0 7647 0 6091 1 3738 1158 6 1200 0 M719 0 02232 0 34013 0 36245 571 9 613 0 ll84 8 0 7734 05%9 I 3683 1200 8 1258 9 572 38 0 02250 0 32306 0 345 % 578 8 603 8

182 6 0 7780 0 5850 3630 inst l

138s 8 57742 0 02269 0 30722 0 3?991 585 6 594 6

1802 0 7843 0 5733

.3577 13ee 3 t 11M e 542 32 0 02268 0 29250 0 31537 592 3 585 4 177 4 0 7906 0 5620 .3525 use e i 14ee s 58707 0 02307 0 27871 0 30178 598 8 576 5 1753 0 7966 05507 3474 lest e l 14M e 591 70 0 02327 0 26584 0 28911 605 3 M74 172 8 0 8026 05397 3423 1458 8 1900 8 5 % 20 0 02346 0 25372 ON719 411 7 558 4 170 1 0 8085 0 5288-3373 1540 4 18M 8 600 59 0 023 % 024235 0 1 % 01 618 0 5494 167 4 0 8142 0 5182 . 3324 tute 1 Met 604 87 0 02387 0 13159 0 25545 624 2 540 3 164 5 0 8199 0 5076 IJ274 toes t ten 8 609 05 0 02407 0 22143 0 24551 630 4 531 3 Ill 6 0 8254 04971 1 3225 1sSe 8 list e 61313 0 02428 021178 0 23607 636 5 522 2 158 6 0 8309 04867 1 3176 trete 1790 0 41712 0 02450 0 20263 0 22713 642 5 Sill 55 6 0 8363 0 4765 1 3128 1750 s toes t $2102 0 02472 0 9390 0 21861 648 5 503 8 52 3 0 8417 04662 3079 tese s 18M 8 624 83 0 02495 0 8558 0 21052 654 5 494 6 490 0 8470 04%I 3030 IIW s less e 628 % 0 02517 0 7761 0 20278 660 4 485 2 45 6 0 8522 04459 2981 teos e I 3e088 tele e $32 22 0 02541 0 6999 0 9540 666 3 475 : 42 0 0 8574 0 4356 2931 195s t fees e 635 80 0 02565 0 62 % 0 8831 672 3 466 2 38 3 0 8625 0 42 % 1.2881 fees e flee t 642 76 0 02615 0 4885

7501 683 s 446 7 30 5 0 8727 04053 2780 21e0 0 W 27e8 8 64945 0 02669 0. 3603 0 6272 6955 426 7 22 2 0 8828 0 3848 2676 27ts e 23ee e 655 89 0 02727 0 12406

5133 707 2 406 0 .13 2 0 8929 0 3640 2%9 23es a

662 13 0 02790 0 11287 0 14076 219 0 364 8 103 7 0 9031 0 3430 2460 sees s 290e s 66811 0 02859 0 10209 0 13068 731 7 3616 l093 3 0 9139 0 3206 1 2345 25es e less e 673 91 0 02938 0 09172 0 12110 744 5 3376 '082 0 0 9247 02977 1 2225 2ess e 3705 8 679 53 0 03029 0 08165 0 11194 757 3 312 3 .069 7 0 93 % 0 2741 1 2097 liss e

Mat t 684% 0 03134 0 07171 0 10305 7707 2851 .LOM S 0 9468 0 2491 1958 fees t a foss e 69022 0 03262 0 06158 0 09420 785 1 254 7 l039 0 0 9588 0 2215 1803 fees s test e 695 33 0 03428 0 05073 0 08500 801 8 218 4 1020 3 0 9728 01891 1619 seen e . 3108 8 700 28 0 03641 0 03771 0 07452 424 0 169 3 993 3 0 9914 0 1460 1373 3108 8 83es e 705 04 0 04472 0 0I191 0 0 % 63 8755 56 1 933 6 1 0351 0 0442 De32 SNes 82MJ' 70547 0 05078 0 00000 0 05078 906 0

906 0 1 0612 0 0000 . 0612 32e: 2* g

Critcal pretsure

'# _ . _.. (

R0-4 . SECTION TH0 QUESTION 2.01 (4.5) Hith regard to the reactor core isolation cooling system minimum flow valve (H0-60): a.

What four conditions will cause the valve to close? (2.0) b.

State the specific conditions that must be satisfied to complete the opening logic for the valve.

(2.5) ANSHER NOTE TO GRADER: For the opening logic (part b), the correct conditions with incorrect and/or qualifiers, constitutes half credit for each response stated as such.

a.

CLOSE MO-60

~

1.

Turbine trip (.5) 2.

Control switch (.5) 3.

High RCIC flow > 80 gpm (.5) 4.

High Rx level +48" (.5) B.

OPEN HO-60 1.

Turbine not tripped (.5) 2.

AND Rx water level 5 +48" (.5) 3.

AND a.) Control switch to open (.5) b.) QR low low reactor level (RCIC Initiation Signal, -49") (.5) AND RCIC low flow < 80 gpm (.5) QUESTION 2.02 (6.00) Answer the following concerning the RHR system.

1.

State the RHR pump interlocks.

(1.0) 2.

State the RHR valve interlocks affecting containment cooling and spray valves.

(5.0) ANSHER ['must be open for shutdown cooling; HO-7 must be open for allP ~ 1.

a.

other modes.)

(.5) . (***CAGEGORY 02 CONTINUED ON NEXT PAGE***)

._... . _ _ -_ - - - _. -

b.

White lights for RHR pumps indicate manual cperation used to shutdown RHR pump while initiation signal present.

Pumps secured manually with initiation sianal oresent won't resoond to - any subseauent initiation sianals.until oriainal sianal reset.

(Can be operated manually.)

(.5) ' 2.

a.

Drvwell sorav (M0-23, 26) and torus coolina and sorav (M0-34, 36, and 37) valves isolate. if onen. on receiot of LPCI auto initiation sianal.

(.5) b.

Torus coolina valves (M0-34 and M0-36) with initiation sianal oresent. recuire the followina before they can be opened: (.5) 1.

' manual' selected (.5) 2.

valve "onen" selected on control switch (.5) 3.

2/3 core coveraae satisfied (.5) c.

1 osi drvwell oressure sianal reauired to oermit drvwell/ torus sorav ooeration if an initiation sianal is cresent in addition to "b" above.

(.5) d.

If 2/3 core cc>eraae loaic is not satisfied. must have the followina to ocen containment coolina valves: (.5) 1.

manual kevlock override of 2/3 core coverace interlock (.5) 2.

b.1. fron1 above (.5) 3.

b.2. from above (.5) GRADER NOTES: A.

Part 2 may appear in order other that given here.

Check to see that all answers are given, regardless of how stated.

B.

Give credit for valves identified by name nr by number.

C.

Underlined sections required. Other parts of answer not required for full credit.

(*** CATEGORY 02 CONTINUED ON NEXT PAGE***) . r - -__m.,_

QUESTION 2.03 (2.00) Why was a modification (PDC 85-44) performed to install en Ultrasonic Flow Monitor on the Standby Liquid Control System ?

ANSWER This modification will provide the ability to periodicly test the SLC system flow at increased (newly required) flow rates without altering existing SLC hardware.

(2.00) REFERENCE PNPS Requal Training Program LP O-RO-08-01-01 pg. 23.

QUESTION 2.04 (3 50) Th3 reactor is operating at 85% power when 125 VDC panel D17 bscomes and remains de-energiz ed.

a.

What three power supplies would you check in an effort to re-energize the panel ? (1.50) b.

How will this transient affect reactor ;0wer ? EXPLAIN.

(1.00) c.

Which of the following components are 'lO T available as a result of this transient ? (1.00) 1) Diesel Generator A 2) RHR pump C 3) Core Spray pump B 4) RCIC 5) Primary protective relays on bus A2 ANSWER (3.50) a.

MCC B-14 via charger (0.50) MCC B-10 via charger (0.50) Battery B (0.50) b.

Reactor power will decrease because recirc MG B has tripped.

(1.00) c.

3), 5) (0.50 each) REFERENCE PNPS Procd. # 5.3.12, Loss of Essential Bus D-5 p.2 125 VDC LP p.

10, Figure 1.

Enabiling Objectives 3, 6.

I (***LATEGORY 02 CONTINUED ON NEXT PAGE***) . _ _ _ _ _

( QUESTION 2.05 (3.00) State the purpose of the ADS (Automatic Depressurization SYSTEM) and list the auto initiation signals (include setpoints).

' ANSHER 1.

The ADS services as a backuo to HPCI in case HPCI cannot or does not resoond to LOCA (.5).

If HPCI ineffective AND AT LEAST ONE LOW PRESSURE pump (RHR or CS) is running, ADS will blow down to allow LPCI/CS to keen the core cool and covered (.5).

. - - . __ 2.

a.

Hiah drvwell oressure (.2) (2.5#) (.2) AND (.1) b.

Rx lo lo water level (.2) (-49") (.2) AND (.1) c.

At least one LPCI/CS numo runnina (.2) (150 osi) (.2) AHQ (.1) d.

120 second timer aoes to comoletion (.5) QUESTION 2.06 (2.00) State two conditions identified by General Electric as possible operating conditions that may require adjustment of the TCV/TSV scram bypass setpoint (as a result of abnormal core thermal power, to turbine first stage pressure relationship).

ANSWER 1.

Operating with less than full capability of FH heater trains.

(1.0) 2.

Operating with a turbine bypass valve open.

(1.0) QUESTION 2.07 (2.00) State the reason for the modification of PDC 85-53, "RHCU valve 1201-2 jog open feature".

ANSHER Formerly,1201-2 would travel full open if an open signal was applied.

It was impossible to stop 1201-2 at any mid-position.

This characteristic of 1201-2 caused a rapid water inrush through the RHCU system into the filter demins, which caused damage to the ion exchange resin.

The modification allows "iocaina" of the 1201-2 valve (1.0).

It may be stocoed at any interim oosition to allow control of water inrush (.5).

After th.e flow transient ends. 1201-2 may be fully oDened by the coerator (.5).

. (*** CATEGORY 02 CONTINUED ON NEXT PAGE***)

( . . QUESTION 2.08 (2.00) Regarding the Diesel Generators: a.

State the automatic starting signals (include setpoints) for DG's? (1.4) b.

In the event of a LOCA start, which of the diesel enaine protection functions are still capable of shutting down the engine? (.6) ANSWER a.

1.

Unit auxiliary (.2) and startuo transformer (.2) breakers to the bus open and undervoltage on the secondary of the startuo transformer (.2); or 2.

Hiah D.H. Press (.2) (2.5#) (.2); gr 3.

10 lo vessel level (.2) -49" (.2) b.

Overspeed (.6) (***END OF CATEGORY 02***)

! I

r

R0-4 ) SECTION THREE , , QUESTION 3.01 (3.00) l A ground overcurrent situation causes a lockout of 4160 vac bus A-5.

- Assuming the A-5 bus was initially powered from the Unit Auxiliary ' Transformer, answer the following questions True or False.

a.

The Unit Auxiliary Transformer bkr to A-5 will trip.

(.5)

b.

The Startup Transformer bkr to A-5 will shut if the Auto Transfer switch is in the ON position.

(.5) c.

A direct result of the A-5 bus lockout is auto start of Diesel Generator "A".

(.5) d.

If Diesel Generator "A" is started or is running when the lockout occurs, its output bkr will automatically shut to supply A-5 with power.

(.5) e.

A lockout signal on bus A-5 will trip any supply bkr connected to it and prevent all power supply bkrs to A-5 from shutting.

(.5) f.

The Shutdown Transformer is the only source available to A-5 under lockout conditions.

(.5) ANSHER a.

True (.5) b.

False (.5) c.

False (.5) d.

True (.5) e.

False (.5) f.

False (.5) .

!

i (*** CATEGORY 03 CONTINUED ON NEXT PAGE***)

! < .. -.. - _ -. -.. - _ - _, _.. -. - _. _ - , _ _ _ _,. _ _ _ _ _.,, _,..._ - _ _ _. _. _ . ., _. _.. _ _.. .. ~. .

I QUESTION 3.02 (4.00) - Describe the operating relationships between core spray injection valves (M0-25A and M0-25B) and the presence of a core spray initiation signal.

(NOTE: there are two parts to this answer.)

" ANSWER 1.

Core spray injection valves get an ooen sianal when an initiation sianal is oresent (1.0) and reactor oressure falls below 400 osia (1.0).

2.

The 25 valves (one oer CS loon) may be throttled or shut manually.

fLv_en with an initiation sianal oresent (.75).

However, they will not resoond to any auto sianals until the criainal sianal clears (.75).

(May be re-ooened manually.)

(.5) White light on, near valve control, indicates that a manual override of an initiation signal was employed.

QUESTION 3.03 (6.00) Concerning Diesel Generator load shedding... a.

What is the purpose of DG load shedding? (2.0) b.

What conditions must be met in order to cause the Diesel Generator load shedding logic to actuate? (List each individual signal in the logic.)

(4.0) ANSWER a.

Strios the associated emeraency bus loads in orecaration for secuential startino of maior loads onto the DG.

(1.0) This orotects DG from hiah currents (due to multiple motor starts) and orevents stallina the diesel.

(1.0) b.

LOCA sianal and loss of offsite power sianal simultaneous 1v.

(.5) As indicated by: 1.

LOCA initiation signals from CS or RHR a.

Hiah drvwell oressure (.5); or b.

Low low Rx wtr. level AND low Rx oressure (.5) (*** CATEGORY 03 CONTINUED ON NEXT PAGE***) _ _ _ _ _. .. _. .. - -

I Either LOCA - Signal 2.

OR low oressure oumos runnina (all three) per diesel (.5) a.

RHR A & C AND CS A (for DG A) b.

RHR B * D AND CS B (for DG B) ' AND 3.

Loss of offsite power as indicated by: a.

DG A(8) bkr closed (.5) AND startup XFMR bkr open (.5) Either Loss Offsite Power Signal 4.

OR loss of offsite power as indicated by: a.

Aux XFMR bkr ooen (.5) b.

Startuo XFMR bkr open OR Startuo XFMR undervoltaae (.5) (i.e., 1 or 2 and 3 or 4 above) QUESTION 3.04 (4.00) Concerning the Recirculation Pump Control System: a.

What is the purpose of limiting recirculation pump speed to 28% when less than 20% total feedwater flow exists? (1.0) b.

What is the reason we limit recirculation pump speed to 65% when operating with less than three feed pumps and a low reactor water level alarm? (1.0) c.

What is the function of the recirculation runback reset pushbutton and indicating light on panel 904? What precaution is associated with using this reset button? (2.0) ANSWER a.

At feedwater flow less than 20% there is insufficient subcoolina of the downcomer water (.2) for high speed operations, which may lead to recirculation oumo cavitation.

(.8) b.

With less than three feedwater pumps and vessel level low, reactor power is limited by a recirculation run back to ensure that the ! steamia, rate is within the makeuo caoacity of the feed system.

(1.0) (*** CATEGORY 03 CONTINUED ON NEXT PAGE***) . - -

I c.

The indicatina liaht informs the coerator that #2 recirculation numo soeed limiter is in control.

(.5) The initiatino conditions must be - corrected and the reset nushbutton utilized to break the seal in.

(.5) The coerator must be sure to reduce the settina of the master (if in master manual) recirculation oumo controller until it is in i control and the deviation between it and the individual controllers . is zero.

(.5) If this is not accomolished and the runback is reset.

the recirculation cumo will ramo un to match the controller outout t (.5) QUESTION 3.05 (2.00) Hill the Core Spray pumps start under each of the following circumstances? If not, why? (NOTE: Answer part a and b independently of each other.)

a.

High drywell pressure (2.5 psig).

(.6) b.

Low Rx water level (-49"). (1.4) ANSHER a.

Yes (.6) b.

No (.6).

Also need low Rx pressure, less than 400 psig.

(.8) QUESTION 3.06 (2.00) Concerning the Rod Block Monitor: a.

What is the function of the Rod Block Monitoring System? (1.0) b.

What are the minimum number of inputs required for each RBM, to prevent an inoperative alarm? (1.0) ANSWER a.

The function of the Rod Block Monitor is to monitor the local neutron flux levels durina the withdrawal of a selected rod (.5) and aenerate trio sianals to actuate rod inhibit and annunciator circuits when the monitored neutron flux levels exceed oreset limits in relation to the reactor recirculation flow.

(.5) b.

Must have. at a minimum. half of the LPRMS sucolvina each RBM channel.

If less than half are available, the associated RBM channel is inoperable.

(1.0) (*** CATEGORY 03 CONTINUED ON NEXT PAGE***) . . -

f ' QljESTION 3.07 (2.00)

Explain how and ehy indicated Rx level eould respond to the following, (&ssume all plant parameters -- actual level, steam flow, pressure remain unchanged): . a.

Small leak in level transmitter reference leg isolation valve packing ',, - gland.

(1.0)

b.

Equalizing valve for leval transmitter cleaks by.

(1.0) ' AN3HER ' a.

Ref. leg pressure will decrease, the delta-P across the cell will decrease and indicated level will increase.

(1.0) b.

The delta-P across the cell will decrease and indicated level will increase.

(1.0) QUESTION 3.08 _ (2.00) Regarding the Tur,b..ine Control System (I.P.R.)

, Explain why the back-up pressure regulator is normally set at a slightly higher pressure than the controlling pressure regulator.

ANSWER In the event that it:e initial oressure.reculator fails in such a manner as to close turbine control valves (as designed), the back-uo oressure reculator will take over at a slictttiv'hiaher_pnsS.yyra.

(1.5) Ihis limits the Dressure excursion on th6 vessel.

(.5).

- % a , / ' '~ . t s. - , s \\ ' -

, - , . (***END OF CATEGORY 03***) , \\ r _ _ _ _ - - _ _ _ _ - _ -.. _. - _ - - - -.

I R0-4 SECTION FOUR . QUESTION 4.01 (3.00) ' Procedure 1.3.11 " Reset of Lockout Relays and Relay Targets" states that lockout relays and protective relay targets shall not b2 reset until authorized by one of three supervisors, a.

Who are the three (3) supervisors ? (1.50) b.

What are the exceptions to this rule ? (1.50) t ANSWER (3.00) c. Chief Operating Engineer, Chief Maintenance Engineer, . Electrical Engineer (0.50 each) b.

Exceptions are the turbine trip lockout relay action and the reactor recirculation system (0.5) when the cause us known not to be of an electrical failure and the trip condition has been cleared. (0.5) These lockouts and targets may be reset as authorized by the Watch Engineer. (0.5) REFERENCE PNPS proc. 1.3.11 , J QUESTION 4.02 (4.00) a.

Explain the automatic actions that will occur if Fire Suppression System Header pressure is decreasing. State the normal fire suppression system pressure band.

If no water, is available on site, where can the fire system receive a water supply? ANSHER >- a.

1).

Normal system oressure is 110-125 osia (.5), the Electric Jockey P_ ump (.5) (50 aom) starts at 110 osia: stoos at 125 osia if in y altto (.5).

(In manual, continuous run.)

2).

If system oressure falls to 95 osia (.5), the Electric Fire Pumo t Starts (.5).

(2000 GPM - Runs for eight minutes after pressure returns to 125 psig if in auto start / auto stop.

In auto start / man stop, must be shutdown manually.)

(*** CATEGORY 04 CONTINUED ON NEXT PAGE***) ? . t - -.-. . _ -. -

I I 3).

If System oressure falls to 85 osia (.5), the diesel enaine driven oumo starts (.5).

(It must be shutdown manually.)

Diesel also starts on less of AC p::wer.

4).

If no on site Dumos available. cross connect oublic fire system with PNPS system with a municioal numoer.

(.5) . ,+ Ref: 2.4.54 QUESTION 4.03 (3 00) The reactor is at 100'/. power when the three (3) inch instrument air / nitrogen line in the drywell ruptures. Based on procedure 2.4.21 " Double Ended Break of the,3" Instrument Air / Nitrogen Line in the Drywel1": a.

What are three (3) symptoms or automatic actions ? (1.5) b.

What immediate actions should be taken ? (1.5) ANSWER (3.00) a.

1.

inst. air header low pressure 2.

drywell pressure increase 3.

drywell 02 concentration increases 4.

cryogenic tank pressure low 5.

drywell to torus d/p alarm .- 6.

unit scram 7.

MSIV closure 8.

N2 supply to drywell hi/ low pressure (3 req'd. at 0.5 each)

b.

1.

scram the unit if not allready scrammed 2.

close AO-4356 (N2/ air isolation to drywell) 3.

closely monitor plant parameters (3 req'd. at 0.5 each REFERENCE PNPS Proc. 2.4.21 pg.2 (*** CATEGORY 04 CONTINUED ON NEXT CAGE ***) I

~ QUESTION 4.04 (3.00) . Concerning the P.A.M.

panels: -a.

What are six of the nine plant parameters which can be monitored - at the P.A.M.

panels ? (2.00) b.

What is the one instrument that is not to be used while the recirculation pumps are running and briefly explain why ? (1.00) ANSWER (3.00) a.

1.

Drywell pressure 2.

Torus level 3.

Torus temperature 4.

Drywell/ torus temperatures 5.

Nobl e gas radiation levels at exhausts 6. Containment H2 and 02 concentrations 7. Safety / relief valve open/close monitoring 8.

Reactor water level 9.

Position indication of sixteen 1" combustibl e gas control valves ( six at O.33 each) b. Reactor water level fuel zone i ndi cati on (0.5) due to false reading from recirc pump d/p.(0,5) REFERENCE PNPS Proc. 2.2.119, 2.2.120.

QUESTION 4.05 (3.00) You are on watch and notice that suppression pool bulk temperature has risen to 88 deg F and is increasing slowly. A HPCI full flow ( test is in progress.

a.

At what suppression pool temperature should suppression pool cooling be i n i t i,t t ed ? (0.75) , l b.

At what suppression pool temperature should HPCI testing be secured ? (0.75) l ' c.

At what temperature must the reactor be scrammed ? (O.75) d.

State any emergency procedures you would enter as a l rt 421 t of the initial conditions given above.

(O.75) l l ANSWER (3.00) a.

80 deg. F (O.75) b.

If 90 deg. F is exceeded, HPCI testing must be stopped.(0.75) c.

110 deg. F (0.75) d.

EOP-4 (0.75) REFERENCE l PNPS T/S 3.7.A.1, EOP-4.

l l (*** CATEGORY 04 CONTINUED ON NEXT PAGE***) _

r . QUESTION 4.06 (1.00) , Concerning EOP-3 "RPV Level Restoration", step III.F(3) directs the operator to determine if at least two injection subsystems , ere lined up for injection with pumps running. What is an injection subsystem ? (1.00) . ANSWER (1.00) An independant means of makeup, that is one with physical separation of flowpaths, components and injection points.

(1.00) REFERENCE PNPS EOP-3 pg. 56 QUESTION 4.07 (4.00) E0P-06 (Primary Containment Control Level) establishes both high level and low level in the suppression pool as entry conditions.

a.

What are the high and low level setpoints for entry into E0P-06? (1.0) b.

What are the bases for establishing high and low level limits? (3.0) ANSWER Ref.

E0P-06 BWR EPGs App 8 pg. B.7-4 (*** CATEGORY 04 CONTINUED ON NEXT PAGE***)

, _ _ _ _ _ - - - - - - - - - - - - - -. - - - - - a.

Hi level +6" or 140" (.5) Lo level -6" or 128" (.5) - b.

Discussion: Both high and low suppression pool water levels require entry to the Containment Control Guideline.

Low levels reduce the , suooression cool heat canacity (.5) and may result in exoosure of drvwell vents and numo suction strainers (.5).

Hiah levels reduce the suooression chamber volume (.5), thereby, increasina the sunoression chamber oressure (.5), increasina static and dynamic leadi (.5) in the containment, and may submerce drvwell-to-sunoresssion chamber vacuum breakers. (.5) QUESTION 4.08 (4.00) ' ~ ConsideringtheE0Ps,underwhat(d conditions is Boron injection t required? ANSHER 4.08 Per E0P-02 under two conditions: 1.

If RX. Power is above 3% (or indeterminant) (1.0) and Sucoression cool temoerature is above 110'F (or will be above 110*F before the Reactor can be shutdown) (1.0).

Inject Boron.

2.

If RX. Power is below 3% but all rods are not inserted cast 04 (1.0) position and SDM is not verified for CR oattern existina (1.0).

Inject Boron (***END OF CATEGORY 04***) - - - - - _ - - - - - - - - _ _ _ - - - - - - - - - - - - - - - - - J

e PILGRIM NUCLEAR POWER STATION

rreshmen b8-F ~-

SENIOR REACTOR OPERATOR LICENSE REQUALIFICATION EXAMINATION i i e FACILITY: PILGRIM REACTOR TYPE: BHR-GE3 EXAMINATION DATE: gjgQTro EXAMINER: D.H. HUGHES APPLICANT: U LH INSTRUCTIONS TO APPLICANT: Use separate paper for your answers. Write answers on one side only.

Staple question sheets on top of the answer sheets.

Point values for each question are indicated in parenthesis after the question. A passing grade requires at least a 70% in each category and a final grade of at least 80%. Examination papers will be picked up four (4) hours after the examination begins.

% OF CATEGORY % OF APPLICANT'S CATEGORY VALUE TOTAL SCORE VALUE CATEGORY 25.00 25.00 5. THEORY OF NUCLEAR POWER PLANT OPERATION, FLUIDS, AND THERMODYNAMICS.

25.00 25.00 6. PLANT SYSTEMS DESIGN, CONTROL, AND INSTRU-MENTATION.

25.00 25.00 7. PROCEDURES - NORMAL, ABNORMAL, EMERGENCY AND RADIOLOGICAL CONTROL.

25.00 25.00 8. ADMINISTRATIVE PROCEDURES, CONDITIONS, AND LIMITATIONS.

100.00 100.00 TOTALS FINAL GRADE % All work done on this examination is my own.

I have neither given nor received aid.

APPLICANT'S SIGNATURE . . -_ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _.

SRO-4 . SECTION FIVE QUESTION 5.01 (1.5) . For each of the following conditions, state whether the change will bring the system closer to, or have no effect on the power at which the onset of transition boiling would occur. Assume all other parameters remain constant for each case.

Briefly explain.

a.

Increase in coolant flow rate b.

Decrease in coolant subcooling c.

Decrease in coolant system pressure QUESTION 5.02 (4.5) The reactor is exactly critical near the end of life (E0L), Control rods are withdrawn to insert.0008 46K/K.

a.

What is the resulting period? b.

How long will it take for power to increase by a factor of 10? c.

What would the period be if an additional.005 46K/K were added initially? QUESTION 5.03 (4.5) For each of the pairs of conditions listed below, state HHICH condition would have the GREATER DIFFERENTIAL R0D HORTH and briefly, WHY.

a.

Reactor moderator temperature of 150 deg F or 500 deg F.

b.

For a rod next to a withdrawn control rod or next to an inserted control rod.

c.

For a rod at position 10 or position 40 in a core operating at 100% power.

l (*** CATEGORY 05 CONTINUED ON NEXT PAGE***) l , ,-. - - -

QUESTION 5.04 (3.5) . a.

For the following transients, indicate which coefficient of reactivity tends to change power first and in which direction - alpha-e T, alpha D or alpha V.

NOTE: No explanation required.

1.

Fast closure of one MSIV 2.

Control rod drop 3.

SRV lifting and then resetting - consider both in your answer b.

The core void percentage is at a maximum with minimum recirc flow on the 100% rod line. Why does the void percentage decrease as power is raised from this point to 100% power? QUESTION 5.05 (2.00) Following a reactor scram, from 100% power, explain what happens initially to the following parameters (increase, decrease, or remain the same) and why, a.

core flow b.

CRD flow (through the pumps) c.

CRD temperature d.

steam line pressure drops QUESTION 5.06 (4.00) Hith regard to Delayed Neutrons a.

Define Beta.

(1.0) b.

Exolain how and why the value of Beta changes from beginning of core life to the end of core life.

(3.0) ^ . (*** CATEGORY 05 CONTINUED ON NEXT PAGE***)

QUESTION 5.07 (3.00) ,, 1.

With regard to thermal limits: ~ a.

State the three PNPS thermal limits, b.

Discuss the purpose for the establishment of the three thermal limits.

(What adverse situations are avoided by observance of the thermal limits?) c.

State the acronyms used by the process computer in presenting the thermal limits to the operator.

QUESTION 5.08 (2.00) The plant is being cooled down from hot operating conditions. Given an initial pressure of 935 psig, what would be the plant pressure two hours after the start of the cooldown if the cooldown rate was maintained at the maximum allowed under normal conditions? (See attached section of Steam Tables.)

(***END OF CATEGORY 05***)

, _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ _ _ l . x . Table 2: Saturated Steam: Pressure Table C . P $pecific Volume (athalpy (ntropy I Abs Press Temp Sat.

$8t Set $8t sat $at Abs Press , ,l Lb!$q in.

F8hr Liquid (vap V4por Lsquid Evap Vapor Liquid Ev80 Vapor LD/So in

D I vg v,g v hg h,g h 8, 5,g S ' g g g p 0 00065 32 018 0 016022 33024 33024 0 0003 1075 5 075 5 0 0000 21872 21872 0 04865 8 25 59323 0 016032 1235 5 1235 5 27382 1060 1 087 4 0 0542 2 0425 2 0967 SM 8 54 19 586 0 016078 641 5 641 5 47623 1048 6 0% 3 0 0925 9446 2 0370 8 58 IS 101 74 0 016136 333 59 333 60 6973 1036 1 05 8 0 1#26 8455 9781

59 16224 0 016437 73515 73532 30 20 1000 9 31 1 0 2349 6094 6443

10 8 19321 0 016597 38404 38 420 61 26 982 1 43 3 0 28 % 5043 7879 la s g 14 696 212 00 0 016719 26 782 26 799 8017 970 3 50 5 0 3121 4447 7%s 14 696 li t 213 03 0 016726 26274 26 290 3111 %97 50 9 0 3137 4415 7552 18 8 M8 227 % 0 016834 20070 20087 196 27 E01 1156 3 0 3358 13 % 2 1 7320 29 9 3e e 250 34 0 017009

3 1766 13 7436 21B 9 945 2 1864 I O M82 13313 16995 34 8 44 0 26725 0 017151 10 4794 104%5 2M1 933 6 1169 4 0 3921

.2844 l 6765 44 s ~3 He 281 02 0 017274 8 4967 8 5140 250 2 923 9 1174 1 0 4112 2474

6586 Se e s

Se t 29? ?l 0 017383 71%2 71736 262 2 915 4 11776 04273 '.2167 6440 6e s . 70 0 3C2 93 0 017482 4 1875 6 2050 272 7 9078 1180 6 0 4411

1905 6316 is t

.r se g 312 04 0 017573 5 4536 5 4711 282 1 900 9 1183 1 0 4534 1675 6208 se t se 8 32028 0 017659 4 8779 4 0953 290 7 894 6 1185 3 0 4643

1470 16113 se e tas s 327 82 0 017740 4 4133 4 4310 298 5 888 6 1157 2 0 4743 1 1264 1 6027 tot e 110 8 334 19 0 01782 4 0306 4 0444 305 0 Bal l 184 9 0 4834 1 1115 1 5950 lit a 17e 8 34127 0 01789 3 7097 37275 312 6 877 8 190 4 0 4919 1 0960 1 5879 120 0

= lu e 34733 0 017 % 3 4364 3 4544 319 0 872 8 191 7 0 4998 1 0815 1 5813 138 8 148 3 353 04 0 01803 3 2010 3 2190 325 0 868 0 .193 0 05071 10t81 1 5752 148 8 IHe 35843 0 01809 2 9958 3 0139 330 6 863 4 11941 0 5141 1 0554 1 5695 Ins t 16s s 36355 0 01815 2 8'56 2 8336 3M1 859 0 .195 1 0 5206 1 0435 1 5641 168 0 lis t 364 42 0 01821 2 65 % 2 6738 341 2 854 8 .1% 0 0 5269 1 0322 1 5591 178 8

18e a 373 08 0 01827 2 5129 2 5312 346 2 850 7 1% 9 0 5328 1 0216 1 5543 Its e Ise e 377 53 0 01833 23847 2 4030 3509 846 7

.1976 0 5384 1 0113 1 5498 iMe 200 0 381 80 0 01819 2 2689 2 2873 355 5 8424 1198 3 0 5438 1 0016

'1 5454 fee t 210 0 38591 0 01844 2 16373 218217 359 9 819 1 1199 0 0 5490 0 9923 15413 fte t ??s e 389 88 0 01850 2 06779 2 08629 364 2 835 4 1199 6 0 5540 0 9534 .5374 22e e 23t s 39370 0 01855 i97991 99846 M83 B318

200 1 0 $588 09740

$3M 238 0 248 s 39739 0 01860 09909 91769 372 3 828 4 200 6 0 % 34 09%S 5299 244 0 25s e 40097 0 01855 82452 84317 376 1 825 0 201 1 0 5619 0 9585 5264 25a 0 264 9 4S4 44 0 018'O 75548 77418 379 9 821 6 201 5 0 5722 0 9508 $230 26s e 27s t 40783 0 010'5 69137 71013 383 6 Bl8 3 201 9 0 5764 0 9433 5197 278 8 Fes t 41107 0 0:080 63169 65049 3871 SIS I 20? 3 0 5805 09MI SIM 2st e fee t 414 25 0 01885 57597 59482 390 6 412 0 2026 0 5844 0 9291 .5135 290 0 300 6 417 35 0 01889 1 52384 1 54274 394 0 008 9 1202 9 0 5882 0 9223 1 5!05 3ee t 354 0 431 73 0 01912 1 30647 132554 409 8 794 2 1204 0 0 6059 0 8909 1 4968 35s 8 ese e 444 60 0 01934 1 14162 1 16095 424 2 780 4 1204 6 0 6217 0 8630 4 4847 488 e a s e 458 0 4%28 00199 1 01224 1 03179 4373 767 5 I?04 8 0 6360 0 8378 4738 450 0 las t 46701 0 01975 0 90787 0 92762 449 5 ?$$ 1 .204 7 0 6490 0 8148 8639 Ses e SW I 476 94 0 01994 0 82183 0 84177 460 9 7433 .204 3 O Mil 0 7936 EM7 558 4 Seel 486 20 0 02013 0 74 % 2 0 76975 4717 732 0 .203 7 06??3 0 7738 4461 saa e 65s 8 494 49 0 02032 0 68811 0 70843 481 9 ??O 9 202 0 0 6828 0 75S2 4381 g5 g 708 0 503 08 0 02050 0 63505 0 655 % 491 6 710 2 .201 8 0 6928 0 7377 18304 790 0 IN I S10 to 0 02069 0 58880 0 609a9 500 9 699 8 1200 7 0 70?2 0 7210 1 4232 750 0 See t SIS 21 0 02087 0 54809 0 %8% 509 8 649 6 1199 4 0 1111 0 7051 14163 ees e tu s $25 24 0 02105 0 51197 0 53302 518 4 679 5 1199 0 0 7191 0 6899 1 4096 854 4 tot 8 531 95 0 02123 0 47968 0 50091 526 7 669 7 196 4 0 1279 0 6753 1 4032 ses 8 9M e 538 39 0 02141 0 45064 0 47205 534 7 M00 194 7 0 7358 0 6612 . 3970 tus stee g S44 58 0 02159 0 424 % 0 445 % $42 6 650 4 192 9 0 7434 0 6476 .3910 tese s leH e 55053 0 02177 C e0047 0 42224 550 1 640 9 1910 0 7507 0 6344 .3851 tale s 1100 0 5 % 28 0 02195 0 37863 0 80058 5575 631 5 189 1 0 7578 0 6216

3794 lite t flW 8 561 82 0 02214 0 35859 0 38073 564 8 622 2 1870 0 7647 0 6091

.3738 115* 4 1298 I % 719 0 02232 0 34013 0 36245 571 9 613 0

184 8 0 1714 0 5969

.3683 ties 8 1254 0 572 38 0 02250 0 32306 0 34556 578 8 603 8 1182 6 0 ??80 0 5850 3630 1254 e Ilse t $7742 0 C2269 0 30722 0 32991 585 6 594 6 1180 2 0 7841 0 5733 3577 13st 8 1350 0 582 32 0 02288 0 29250 0 31537 592 3 5854 1177 8 0 7906 0 $620 .35?5 13u t 1888 8 58107 0 02307 0 27871 0 30178 598 8 5765 1175 3 079% 05507 34 74 1400 0 1454 8 591 70 0 02327 0 26584 0 28911 605 3 5674 1172 8 0 8026 0 5397 3423 14W O IHee 596 20 0 02346 0 25377 0 27719 611 7 558 4 .I70 1 0 8085 0 5288 3371 1200 15W 8 600 59 0 023 % 0 24235 0 26601 618 0 5494 167 4 08:42 O bit? 3324 1554 e 16es e 404 87 0 0218' 0 23159 025545 624 2 S40 3 164 5 0 8199 0 5076 3274 18et t IGN O 609 05 0 02407 0 22143 0 24551 6304 531 3 161 6 0 8254 0 4971 1 3225 teses s 17M 8 613 13 0 02420 0 21178 023607 636 5 522 2 158 6 0 8309 04867 l 3176 17se s 1754 1 61712 0 02450 0 20263 0 22713 642 5 S131 55 6 0 8363 04765 3128 1758 9 18e0 0 623 J2 0 024?2 0 19190 0 21861 648 5 S038 52 3 0 8417 04%2 3079 ises e 18 88 624 83 0 02495 0 18558 0 21052 654 5 494 6 49 0 0 8470 04%I .30 30 isss 4 teelt 628 56 0 02517 0 17761 0 ?0278 660 4 485 2 45 6 0 8522 0 4459 ?981 tsee s .' 1950 0 432 22 0 02541 0 16999

9540 6%3 475 8 42 0 0 8574 04358 2931 tou t . 2000 0 635 80 0 02 % S 0 162 %

8831 672 1 4% 2 38 3 0 8625 0 42 % 2881 fees e '3 fiM 8 64? ?6 0 02615 0 14885

1503 6833 446 7 30 5 0 8127 0 4053 2780 fise s

??ss e 64945 0 C2669 0136c3 0 6272 695 5 426 7 . 22 2 0 8828 0 3848 26 4 2res 8 y 23M O 65589 0 02727 0 12406 0.5133 7072 406 0 183 2 0 8929 0 3640 2569 23es t 2400 0 M218 0 02790 O !!287 0 14016 719 0 384 8 103 7 0 9031 0 3430 2460 feas t .. flee g 66811 0 02859 0 10209 013068 731 7 361 6 1093 3 0 9139 0 3206 .2345 25es e Inse l 673 91 0 02938 0 09172 0 12110 744 5 3376 1082 0 0 9247 02977 2225 2640 1 2798 0 67953 0 03029 0 08165 0 11194 757 3 312 3 1069 7 0 9356 0 2741 2097 2700 8

4 fees t 684%

0 03154 0 07171 0 10305 770 7 2851 1055 8 0 9468 02491 1958 flee s

flee t 69022 0 03262 0 06158 0 09420 185 1 254 7 1039 8 0 9588 0 2215 1833 29es e seest 69533 0 03428 0 05073 0 64500 BOI S 218 4 1020 3 0 9728 0 1891 1619 asse t 8107 8 700 28 00M81 0 03771 0 07452 824 0 1693 993 3 0 9914 0 1460 1 1373 31e4 8 3290 8 705 08 0 04472 0 01191 0 05663 8755 56 1 933 6 8 0351. 0 0482 1 0832 87se t 8288 2* 70547 0 05074 0 00000 0 05028 906 0

906 0 1 0612 0 0000 1 0612 32ss l' ' Critic 81 pressure 'd . _-. _ _ _ _ _ _ _. _ _ _ _ _ _ _ _.. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. . , . TYPICAL-N.R.C. EOUATION SHEET s F = ma v = s/t Cycle efficiency = V = TI"r26 (cyl) s = V t + 1/2(at2) o A = AN A = A e-A t a W = mg a = (Vf - Vo)/t E = mc2 A= En2/t1/2 = 0.693/ tift KE = 1/2(mv2) A T.M = ( 6 T, ) - ( A T1 )

PE = mgh E n( A T. ) ( AT ) g ,. vf = va + at , , Q + mhin = W/J + m,hout , A E = 9314 m h=u+PV I = I e-#X I = I e-Z X c c Q = dCp AT ksemAh I = I 10-x/ M o Q = UA (A T)LM Ah = Cp AT TVL = 2.303/g P = P 10sur(t) h = h; t xhg HVL = 0.693/g o P = P et/T h : r*nv6f/y CR (1 - Keff 1) = CR (1 - Keff 2)

2 o ... . ... .. - -. -rh=pAv.. - - -. _ --.-- -..

.- - - - - - - - - - . -. SUR = 26.06/T scR = c /l - Keff c ' T = (&*/p ) + ((4 p )/Ep] H = 1/(1 - Keff) ' = CR /CRo

' p= AO* + 8 M = (1 - Keff o)/(1 - Keff 1) AT+1 SDM = (1 - Keff) C* = 10-4 seconds p = (Kef f-1)/K rf = A K rf/Keff e e 5 = 0.1 seconds-1 i P = (EQV)/(3x1010) Igdi=Id22 E = FH 1df=Id)

2 R/hr = (0.5 CE)/d2(meters) .. _ - _ _ WATER PARAMETERS (atmospheric ccnditions) MISCELLANEOUS CONVERCIO!!S , 1 gal. = 8.345 lbc. = 3.70 liters 1 curie = 3.7 x 1010dps

- 1fL3 = 7.48 gal.

1 kg = 2.21 lbm Density = 62.4 lbm/ft3 = 1 gm/cm3 1 hp = 2.54x103 Blu/hr = 550 ft-lbf/sec Ileat of Vaporization = 970 BTU /lba 1 K.J = 3. 413 x 106 Blu/hr i Heat of Fusion = 144 DTU/lbm 1 in.

2.54 cm l 1 ATM - 14.1 psi - 29.92 in. Hg.

OF = 9/5 oc + 32 ~ oc = 5/9 (oF - 32) - . e L

WoEZJ

SECTION SIX QUESTION 6.01 (4.5) , Hith regard to the minimum flow valve (MO-60) for the RCIC System: a.

What four conditions will cause it to close? (2.0) b.

State what specific conditions must be met to satisfy the opening logic for the valve.

(2.5) QUESTION 6.02 (2.5) State the three auto isolation signals to RCIC (include setpoints) and identify the two valves affected directiv by auto isolation.

QUESTION 6.03 (3.00) A ground overcurrent situation causes a lockout of 4160 VAC bus A-5.

Assuming the A-5 bus was initially powered from the Unit Auxiliary Transformer, answer the following questions True or False.

The Unit Auxiliary Transformer breaker to A-5 will trip.

(.5) a.

b.

The Startup Transformer breaker to A-5 will shut if the Auto Transfer switch is in the ON position.

(.5) c.

A direct result of the A-5 bus lockout is auto start of Diesel Generator "A".

(.5) d.

If Diesel Generator "A" is started or is running when the lockout occurs, its autout breaker will automatically shut to supply A-5 with power.

(.5) A lockout signal on bus A-5 will trip any supply breaker connected to e.

it and prevent all power supply breakers to A-5 from shutting.

(.5) f.

The shutdown transformer is the only source available to A-5 under lockout conditions.

(.5) l ! (*** CATEGORY 06 CONTINUED ON NEXT PAGE ***) t .,. i

_ _ _ _ _ _ _ _ _ _ Concerning the Recirculation Pump Control System: ' a.

What is the purpose of limiting recirc. pump speed to 25% when less than 20% total feedwater flow exists? (1.0) b.

What is the reason we limit recirc. pump speed to 65% when operating with less than three feed pumps and a low reactor water level alarm? (1.0) c.

What is the function of the recirc. runback reset pushbutton and indicating light on panel 904? What precaution is associated with using this reset button? (2.0) QUESTION 6.05 (4.00) Describe the operating relationships between core spray injection valves (M0-25A and M0-25B) and the presence of a core spray initiation signal.

(There are two parts to the answer.)

QUESTION 6.06 (2.00) Regarding the Diesel Generators: a.

State the automatic starting signals (include setpoints) for DG's.

(1.4) b.

In the event of a LOCA start, which of the diesel enaine protection functions are still capable of shutting down the engine? (.6) QUESTION 6.07 (2.00) Concerning the Rod Block Monitor: a.

What is the function of the Rod Block Monitoring System? (1.0) b.

What are the minimum number of inputs required for each RBM to prevent an inoperative alarm? (1.0) , (*** CATEGORY 06 CONTINUED ON NEXT PAGE ***)

. 6.__P68Ml_@Y@IENS_DEgl@N1_CgNIBg61_8NQ_lNSIBUNEN18IlgN . QUESTION 6.0;8 (3.00) With the plant operating at 100% power, recirc in Master Manual, en operator inadvertently decreases the MHC pressure setpoint by 5 psi. What will be the initial response and final status of the following due to this action? Briefly explain for initial response only. Refer to the attached MHC logic diagram if necessary. Assume load limit set at 100% and flow limit set at 105%. (3.00) c. TCV position.

b.

BPV position.

c.

Reactor power.

d.

Reactor pressure.

> I l l !

l ' (***** END OF CATEGORY 06

- - - - )

l - _ - - - _ -, _ _ - __ _. __ _ __ __ _ _ _ _ _ _ _ _ _ _ _ _ _____ __ -

._ . . . - _ -__ .-- -_ - . .. . 7.

PRQQEDQBES _NQB_ MAL 1_ADNQBMBL _gMgBQgNQY_AND

BBD196901986_QQNIBQL

. QUESTION 7.01 (2.00) i m.

In accordance with procedure 2.1.1 "Startup From Cold Shutdown," under what conditions shall secondary containment integrity be maintained? (1.50) b.

You are allowed a certain period of time after placing the reactor

in the RUN mode before the primary containment atmosphere oxygen ' is required to be less than 4% by weight. How long is this period?(0.50) . DUESTION 7.02 (3.00) a.

During an ATWS condition if RPV water level should become undeterminable, EDP-08, "RPV POWER CONTROL BY LEVEL", directs the

i operator to maintain power above 8% by controlling injection to the RPV.

What is the significance of 8% and why must power be maintained above it? (2.00) ' b.

If, in the above situation, the operator is unable to maintain i reactor power above 8%, he will be directed to depressurize ' to below the " MINIMUM ALTERNATE FLOODING PRESSURE". Explain the basis

for this pressure.

(1.00)

QUESTION 7.03 (3.00) In reference to procedure 2.4.143 - Shutdown From Dutside Control Room Due To Inhabitability of Control Rooms a.

WHERE are the Operators AND Operating Supervisor directed to

go following assembly in the 23' 4kv switchgear area? (1.00) l ' b.

WHAT is the " preferred" method to scram the reactor AND WHY is it the preferred method? (1.00) c.

WHEN should the reactor feedwater pumps be tripped? (1.00) < ! ! l { (***** CATEGORY 07 CONTINUED ON NEXT PAGE

- - - - )

i . -.. - - -,.. - ,.. . - . -. - - - - . -., -. - -.. -. - - -. - -.. - -. - -. - - -. -. - - - -. -.

. Z:.__P_EQQE_QQEES__- _ NQEMALt_AQNQRMALt_EMERGENQY_ANQ $ E8919 LOG 1G86_QQNIEQL DUESTION 7.04 (2.00) Entry into procedure EDP-03 section III.F3 " Steam Cooling" will occur only if all possible injection to the RPV becomes unavailable. If, during steam cooling, RPV pressure drops below 700 psig, the procedure directs tha operator to emergency depressurize. What is significant about 700 psig cnd why is RPV depressurization required when pressure drops below 700 psig? (2.00) QUESTION 7.05 (4.00) The Automatic Actions of Procedure 2.4.19 "RECIRC Pump M-G Set Scoop Tube Lock Up", describe the conditions that will cause such a lockout. State these conditions and setpoints (if applicable).

QUESTION 7.06 (1.00) Concerning the P.A.M. panels; List the one instrument that is not to be used while the recirculation pumps are running.

i , l l ' (*** CATEGORY 07 CONTINUED ON NEXT PAGE***) {

r QUESTION 7.07 (4.00) h.

j E0P-06 (Primary Ccntainment Control Level) establishes both high level and low level in the suppression pool as entry conditiens.

a.

What are the high and low level setpoints for entry into E0P-06? b.

What is the basis for establishing high and low level limits? QUESTION 7.08 (4.00) Considering the E0Ps, under what conditions is Boron injection requiredi OUESTION 7.09 (2.00) If both stack dilution fans are lost, the associated Off-Normal procedure requires that both standby gas treatment units be placed in service. Why is this necessary? . (***END OF CATEGORY 07***) ' , .'

. - SRO-4 . SECTION EIGHT QUESTION 8.01 (4.00) . a.

State the manning requirements for minimum shift crew compliment as required by Technical Specification Table 6.2-1 (restated in Procedure 1.3.34, " Conduct of Operations") for the following conditions: 1). Operating (2.0) 2).

Cold Shutdown or Refueling (2.0) QUESTION 8.02 (3.00) a.

Procedure 1.3.34, " Conduct of Operations", delineates conditions when the control room operator may leave the area of the controls without obtaining a qualified relief. State those conditions.

(1.0) b.

1.3.34 also delineates conditions under which a formal relief turnover is not required for the control room operator. State those conditions and describe the turnover that agM be given under these circumstances.

(1.0) c.

1.3.34 defines " Controls".

Briefly describe controls as defined by this procedure.

(1.0) QUESTION 8.03 (2.00) Define Primary Containment Integrity.

QUESTION 8.04 (5.00) , Classify the following as an Unusual Event (U), Alert (A), a Site < Emergency (S), a General Emergency (G), or not applicable (N/A).

(Use the Attached PNPS 5.7.1.1).

1.

Failure of the reactor to shutdown on manual initiation within 15 minutes with MSIVs open.

(.5) 2.

Loss of physical control of the reactor building by operations or security personnel.

(.5) 3.

Loss of physical control of the main control room by BECo operators.

(.5) 4.

In HOT STANDBY, a fire occurs in the Admin. Bldg. which is still out of control 25 minutes after fire fighting efforts have begun.

(.5) (***** CATEGORY 08 CONTINUED ON NEXT PAGE *****) l,..._ - . . - -. - - - _ _ _ _ . _. - _ -. . _ _ _ _ -. - - _.. _ -.. - _.

I 5.

The reactor in HOT STANDBY with coolant system leakage in excess of . 50 gpm as indicated by drywell indicators.

(.5) 6.

During operation at 100% power, an MSIV isolation occurs on a high = radiation signal.

Drywell pressure increases above 10 psig and main stack rad monitor is at 3 rem and increasing.

(.5) 7.

A hurricane with sustained winds greater than 80 mph (onsite measurements).

(.5) 8.

Failure of an SRV to reset following a scram-related actuation.

(.5) 9.

Greater than 200 uci/ml of total iodine in the reactor water.

(.5) 10.

Loss of all control room annunciators for 15 minutes or more.

(.5) QUESTION 8.05 (2.00) Define core alteration, as defined in Technical Specifications.

QUESTION 8.06 (3.00) Hith fuel in the vessel, midway through the refueling outage with Rx.

cavity flooded, the "C" RHR pump is tagged out for electrical repairs to the motor windings. The "B" RHR Loop has been tagged out for maintenance to repair the "B" heat exchanger bypass valve, M0-1001-168 (leaks excessively past seat).

The NPO has just reported that the "A" RHR pump, currently in shutdown cooling, is leaking excessively (10 gpm) and recommends securing it. All other CSCS systems are operable. The Hatch Engineer decides to secure the "A" pump for repair.

List all applicable LCOs and state any actions that are required.

NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS T0*

ANSWER.

FULLY REFERENCE ALL APPLICABLE SECTIONS OF THE T.S. YOU

USE TO DEVELOP YOUR ANSWER.

QUESTION 8.07 (6.00) a.

A core reload has recently been completed.

Due to a suspected problem with the Graphitar Seals, Control Rod Drive Mechanisms 18-11 and 38-39 are being removed for maintenance.

Describe the conditions that must be satisfied (as stated by Tech Specs) prior to and during removal of the mechanisms.

(4.0) b.

Under what conditions does an inoperable hydraulic accumulater (HCU) constitute an inoperable control rod? (Reference 2.1.15; OPER-09.)

(1.0) c.

Under what conditions (short of repair) can an inoperable accumulator be cleared of its inoperable status? (Reference Tech Specs 3.3.D.)

(1.0) (*** END OF EXAMINATION ***) . . _ _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

. LDitT*!G C'";01TIO t ear OPEF.iTIO!; S'JRVRIL!.'.!$ R'.nti ers' ?;T . 3.3 'FE/.C!'.':TY C0!'TTO:. , ! 4.3 P.EACTIVIT/ C0!; 70L '% I Anp1!c.bility: \\ge A niicability:

Applies to the opera:ienal status of the control rod system, Applies to the surveillnnce require-r.ents of the control rod systee.

Objective: Objective: To assure the cbility cf the con;- ' trol rod systen to control reac-To verify the' ability of the control tivity.

rod system to control reactivity.

Specificati:n: Soccification: A.

Resetivitv Limitations A.

Reactivity Limitations 1.

Reactivity tare-in - core loac:nq 1.

Reactivity car.7in - core loading

~ ~

^ The core loading thsil be 11 cited to that which can Sufficient control rods shall { be cade subcritic.1 in the be withdrawn follonn; a re-cost reactive ccndition fueling outage when core during the cperating cycle alterations were pyrforms-d with the str ngest eperable to deconstrate with a ::ar- % control rod in its full-gin of 0.25 percent t.k that the core can be cade sub-ti out Eosition and a.11 other L ' operable rods fully in-crit'icci at any time in the - sorted.

y subsequent fuel cycIt-with the strengest operabic con-

trol rod fully witdrawn } and all other operable rods

fully inserted.

, t 2.

Roactivity car-in - inecer-

.

2.

Reactivity car-in - inoner-able c-r1 : c.

d able control rods L a., Control rod drivas ,j Each partially or fully with-which cannot be noved

drawn operable control rod with control rod drive dl shall be exercised one notch . pressure shall be con-fj at 1 cast once cach week.

sidered inoperablo.

If This test shall be perfo : ed a partially or fully a withdrawn control red I,' at 1 cast once per 24 hours in drivo ennnot be coved with? the event power operation in continuing with three or core drive or scrc.a pressure Il v inoperabic centrol rods or in the reactor shall be -

'

the event power operation is brought to a shutdown

{ conditionwithin48hoursy partially withdrawn rod which continuing with one fully er u unless investigation il

o

der.:nstrates that the .# cannot be coved and for which e cause of the failure is .1 control rod drive cechanism y

n:t due to a failed con- !j danage has not been ruled out.

, '-( trol red drive cc:har. ism The surveillanco need nor he ,1 ' c:llet housing.

completed within 21 hours a f the" { nurber of inoperable rods has - . .

~ i l Ar.codsUNCONTROLLEl . . -- = -. = -.....-.... _ c9py.

.. i - -------. - -. .

- t.u... a :. 2 t~:,tc it5 iOR EI:P5T105 l Stl.v!!'.I./.?:Cli i!FE!!M!L'!Prr 3.3.A ACTIVITY CONT.'0L 4.3 Rr?,CTIV ITY - CD.'CROL I . b.

Tho control rod dir:c. l been reduced to 1 css T I-tionni control valves than three and if it for insparab13 control has boon de:onstrated reds sha*1 he disarmed that control rod l , electrically and the drive cc-hanisu colle: - . control rods th:11 be housinr. failure is not in such positic.u that cause of an ienovable Specification.$.!,.A.1 control rod.

is ::et.

c.

Control rod drives I

which are fully in-rcrted and electrically disamed shall not be considered inoperable.

$ - < d.

Control rods with serra l times greater than I those permitted by l . ' . _ Specification 3.3.C.3 l -- are inoperable, but t - if they can be naved with control rod drive pressure they need not I i be disarted electri-l cally.

! (^ l - M; e.

During reactor pcNer operatien, the number , of incperable control ! . U*

- 't!"1 E5'i2 rods shall not.::cced i

eight.

Specificatica

3. 3. A.1 =us t ba met [ 1.

The coupling integrity an.cl at all times.

[ be verified for each with-drawn control red ca follovs: ' . a.

b' hen the red is withdrovn B.

Codtrel ! cds the first tir.e subse tuent

1,. Each control rod shall bo' to each reibelir.;, outr se ' couple:i to its drive or or after raintennnce, ob erve discernible r:s-completely inserted.nd the control rod direc-pon:e of the nu: lent ins trueen tation. Hovent, tier.r.1 or control valves for ir.itittl rods when dinnr.ed <:lectrically.

rc:panse is not discern-This require ent does not ap;,1y in the refuel con-ible, subsequent exerci in6 of these rod after the ditic.. When the rer.etor reactor is critical : hall is "*. ed.

P.so centrol be yarforr.ed to vertfy rc:.1-r.es ray be renoved ca '_.; e-3;ecification inctrucentation rc:;cnse.

.

J.A.1 is OJt.

. ! Amendment No. 14

-

.UNCONTROLLE COPY.

'- , ...

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . CONDITION FOR CPERATION SURVEILLANCE REQUIREMENTS . . _ 3.3.B Control Rods 4. 3. B Control Rods 2.

The control rod drive hous-b.

When the rod is fully ing support system shall be withdrawn the first in place during reactor power time subsequent to each ~ operation and when the reac-refueling outage or af-tor coolant system is pres-ter maintenance, observe surized above atomspheric that the drive does not pressure with fuel in the .go to the overtravel reactor vessel, unless all position.

control rods are fully in-serted and Specification 2.

The control rod drive housing 3.3.A.1 is met.

support system shall be in-spected after reassembly and 3.

a.

No control rods shall the results of the inspec-be moved when the reac-tion recorded.

tor is below 20% rated power, except to shut-3.

Prior to control rod with-down the reactor, unless drawal for startup or inser- - the Rod Worth Minimizer tion to reduce power below (RWM) is operable.

A 20% the operability of the maximum of two rods may Rod Worth Minimizer (RWM) be moved below 20% de-shall be verified by: sign power when the RWM , is inoperable if all a.

verifying the correct-other rods except those ness of the control which cannot be moved with rod withdrawal sequence ' control rod drive pressure input to the RWM com-are fully inserted.

puter.

b.

Control rod patterns and b.

performing the RWM the sequence of withdrawal computer diagnostic or insertion shall be es-test.

tablished such that: c.

verifying the annunci-1) when the reactor is ation of the selection critical and below errors of at least one 20% design power the out-of-sequence control maximum worth of any rod in each distinct insequence control RWM group rod which is not elec-trically disarmed is d.

verifying the rod block less than 0.010 delta function of an out-of-k.

sequence control rod which is withdrawn no 2) and when the reactor more than three notches, is above 20% design power the maximum worth of any control rod, in-cluding allowance for a single operator error, is less than 0.020 delta ( k.

Amendment No. 39 -

.UNCONTROLLE

(Qfy.

__

z -.: :. - - ~ ' '3.3.8. Control Rods 4.3.5, Control Rods

4.

Control rods shall not be with- . 4.

Prior to control rod with-drawn for startup or refueling drawal for startup or ,/.. unless at least two source range during refueling, verify channels have an observed count - that at least two source rate equal to or greater, than range channels have an

.- three counts per second.

observed count rate of at least three counts per 5.

During operation with limiting second.

control rod patterns, as deter- ' mined by the Reactor Engineer, 5.

When a limiting control rod . either: pattern exists, an instru-ment functional test of the ' a.

Both RBM channels shall be RBM shall be perfomed operable: or prior to withdrawal of the , . designated rod (s) and b.

Control rod withdrawal shall daily thereafter.

be blocked: or

c.

The operating power level shall be limited so that the MCPR will remain above - the Safety Limit MPCR

assining a single error that results in complete withdrawal of any single operable control rod.

! C.

Scram Insertion Times C.

Scram Insertion Times i 1.

The average scram insertion 1.

Following each refueling time, based on the deener-outage, each operable , gization of the scram pilot control rod shall be sub-valve solenoids as time zero, jected to scram time tests . - .. of all operable control rods from the fully withdrawn ' in the reactor power opera-position.

If testing is tion condition shall be no not accomplished with the greater than: nuclear system pressure ' above 950 psig, the % Inserted Average Scram measured scram insertion From Fully Insertion time shall be extrapolated Withdrawn Times (set) to reactor pressures above 950 psig using previously .

.55 detemined correlations.

1.275 Testing of all operable

2.00 control rods shall be com-i

3.50 plated prior to exceeding 40% rated themal power.

. - . k Amendment No. J5

l ' 83 - 10NCONTROLLE - - '-- COPY.

, -. ._ - . __

___. _ _ _ _. _ _ _ __ _ .. _ _ _ _ _ _ _ _ _ .- . ! - . ! UV.IMN3 C0WTMCN FtR 0?EMMONS SUPVETIMN"E REOUIRDOT f . 3 3.C Berar Insertion Time 4 3.C Seras Insertion Time i ' ' W.. The average of the serna 3.

At 16 week intervals. 50% . insertion times for the of the control rod drives -

three fastest control rods shen be tested as in le.3.C.1 of au grcmps of four ese-so that every 32 weeks a u . trol rods in a two by two of the control rods sha n - array sha n be no greater have been tested. Vbenever

' than: 50f, of the control rod drives ' have bee: s:ra: tested, an $ Inserted Avg. Scram evaluation shan be ande to ' ! From Fun y Insertion provide reasonable assurance Withdrawn fire Sec.

that proper control rod drive performance is being

.58 maintained.

' . '

1 35 - $0 3.12 -

5 30 _

The maximum scram inser-tion time for 905 inser- . tion of any operable con-

trol rod shan not szeeed

7 00 seconds., i . D.

Control Rod Accumulaters .D.

Contrel Red Accurulaters . At au reactors operating pres.

Once a shift, ch"eck the status sures, a rod accumulator any of the pressure and level alarms # , be inoperable provided that no for each accumulator.

j j other control rod in'the nine- !' rod square array aroung this . rod has a . ' l ' 1.,Tnoperable accumalster.

. . , 3.

Directisest centret valve electriestly dise M d . I while in a non. fully ta-serted posities.

3.

Scres (neertion time

greater than the marin e i permissible insertisa . l time.

. l

. , , If*a control red with sa

inoperable secumulater is , , , inserted " full-in" and its

- directions 1 eentrol volves ( are electrically disarmed.

At shall not be eensidered .. have an io.,aras1.

s4 - .ceu oi.t.r.

- . . . , 4 ne n n o. es ~ ..UNCONTROLLE ' ~ ~ ~ ~ ~ copy.

-. . _- . - . _ - .

- LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT - E.

R activity Anomalics E.

Reactivity Anomalies . The reactivity equivalent of During the startup test program . ( the difference between the and startups following refuel-actual critical rod configur-ing outages, the critical rod ation and the expected con-configurations will be compared figuration during power to the expected configurations operation shall not exceed 1% at selected operating conditions.

A K.

If this limit is exceed-These comparisons will be used ed, the reactor will be shut as base data for reactivity down until the cause has been monitoring during subsequent - detennined and corrective ac-power operation throughout the tions have been taken if such fuel cycle. At specific power actions are appropriate.

operating conditions, the ' critical rod configuration will be compared to the configuration F.

If Specifications 3.3.A expected based upon appropriately through D above cannot be met, corrected past data. This com-an orderly shutdown shall be parison will be made at least initiated and the reactor every full power month.

shall be in the Cold Shutdown - condition within 24 hours.

Specifications 3.3.A through D above do not apply when there is no fuel in the reactor vessel.

- G.

Scram Discharge Volume G.

Scram Discharge Volume 1.

The scram discharge 1.

The scram discharge volume volume drain & vent drain and vent valves shall valves shall be operable be verified open at least once whenever more than one per month. Each valve shall operable control rod is be cycled quarterly. These wi thdrawn.

valves may be closed intennit-tently for testing under 2.

If any of the scram dis- . administrative control charge volume drain or vent valves are made or 2.

During each refueling outage found iniperable an verify the scram discharge orderly shutdown shall be volume drain and vent valves; initiated and.the reactor shall be in Cold Shutdown a) Close within 30 seconds within 24 hours.

after receipt of a reactor scram signal and b) Open when the scram is reset.

. ' ( - . . Amendment 65 ..UNCONTROLLE , '- COPY.

. . 1*MITI!:0 CC*:3 TIC! S FOR CTZMTICN SWlVEft.tA11CE 3:U T O "TliT

_ / 35.CC1E AC CO?!TAIf. WIT C001.f MC' 36 5 ' con! MID Cn!!TAIM6?.7 C00LD3

',

SYS;ES SYSTn*3 A:*11esbility Avliesbility AFP11cs to the c; rati=nal status of Applies to the Sune111anze Res ire =ent:

u the core and su;;ressica pool coolir.g of the core azed suppression pool cooling subsysta=s.

subsy:te=s which are required when the corresponding 11=1:12:3 Condition for op.

., eration is in effect.

Ctfective C53ective ' To assure the operability of the core To verify the operability of t'ha core and and suppression pool cooling subsysta=s su;;ression pool cooling subsysten: unter under all esaditicas ter Wich this all conditions for Wich this cooling ca-cooling capatility is as essential re-pability is as essential res;=nse to sta-sponse to stati:n abn=r-*'* ties.

tien abnor=alities.

l - Sseeif3 cation Seecifiention ,i A.

Co., St-tv p.nd 7,7:: Setsvete-s A.

Ce -+ ? --s * P - 4 T *>-* ?"M"* * *- - 1.

Both core s;rt.y,:bsy:t'e=s sha.11 1.

Core 3; ray subsysta= restisg.

to c;rra. tis c eeve-irrat.iated Ita= N e.ev fuel is is the vessel act prior g to re::t:r startup frs= a Cold .a. Sir.:1sted once/Operatir.g \\ Conditi:n, ex :;t as spe:ifiet Au* m tic cycle in 3.$.A.2 below.

Actuation tast.

. b.

Pump operability' On=e/ month

i

i ! . c.

Motor Operated once/= cath and Valve operability on=e/ cycle from the Alternata Shutdown Fanel , . ' . d.

Pu=p flow rate Each pu=p shall deliver at laast 3600 g;n assi=st - a system head corresponding to a I reactor vessel . ' pressure of 104 ps' e.

Core Sp;sy Header A p Instru=antation Ane=d=ent No. M. 62 - (

. 103

e - - - - - - -,,, - -. -. + - -, - - - - - - .. - _..____ .,, _ _ _ _ -, _. _,.. - _ _ --,u , -.,,, -, _. - - _ _ _ _ _ _ - _ _ - _ _,. _ _ _ _.. _ _, - -.. - - - -. - -. - - -..

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

. . . .. (- . tre:7:NC COS'3ITICNS TCR C?EMTION St'RVIIttANCE EOU1F". INT _ 3.3.A Core Spray and L?ct subsyste=s 4.5.A Core $ stay a.nd I?CI Sub syste.s (c cat' d) (cont' d) - . , Check Onca/ day ' , , , Calibrata Once/3 e.:nthe

.- . Tast On=a/3 months 2.

Tscs and af:a the data that osa 2.

When it is detar= bad that ena c::a of the cera spray subsyste=s is spray subsysta= is inoparable, r.ade or four.d to be isope:able the operable cc:a spray subsystas, for any reason, continued reactor the L?cI subsyste.= and the disse.1 opera:ics is permissible during ganarators shall be de=sastrated to the sue:endbg seve= days, pro-be operable i==edia: sly. L a ope:- vided tha: during such seven days abla esta spray subsysta= shall be a11' active ce=ponents of the otha tamanstrated to ha operabia daily

c::a spray subsysta: a=d active thereaftar.

cc=pene=:s of the L?C1 subsysta= and the diesel ganarators arm op-arable.

. 3.

The L7C1 Satsysta=s shall be oper-3.

L7CI Subsysta= Testing shal.1 be as , able whenever 1::adia:ed fuel is follows: . , ( it tha rea::c: vessel, a=d prior . ' to saaet=: startup f:c= a Cold a.

Si=ulated Au::=a-Once/Operathg Ccuditics, ax:ept as specified tic Actuation Tast Cycle in 3.5 A.4, 3.5.A.5 and 3.5.7.5.

. On'a/nonth b.

Pu=p Operabili:y c . - c.

Me:cr Opera:ed once/Y.:::h and valve operability Onta/ cycle frew the Alta:.ata

. Shutdown Panal

d.

pu=p Flow Once/3 sanths , Three LPCI pt._ps shall deliver

14,t.00 sp= against a sys:ss bead corresponding to a . - - vessel pressure of 20 psig . O e . e A and=ent No. N, 62

g . 104 .

r = - ,

. i .A .. LIMITING CONDITIONS FOR OPERA $0N SURVEILIANhEREQUIREMENT ) -., .1 . ... 3.5.A Core Epray and_LTCI Subsystems 4.5.A , Core Spray and LPCI Subsystsas ~ (Cont'd) (Cont'd) , 4.

From and af ter tha'dets that one 4.

Whos it is determised that 'one of of the Elf '(LPCI) ptacps is made the RER (LPCI) pumps is inoperable or found to be inoperable for any at a time W en it is required to be operable, the conta4nment cooling ' , reason, continued ructor opers-suttystes, the remaining. active com-P tion is permissible only during , the succouding thirty days pro-ponents of the LPCI Subsystem, both.

' vided that during such thirty days core spray systems and the diesel

* * * * * ' * * * * *

8* I'***' 3****** ** * *

** ***** * *

the remaining active components of be operable immediately and-the op-the LPCI Subsystem, and all active etable LPCI purops daily thereafter.

, . components of both core spray sub-systems and the diesel generators t _ are operable.

5.

When it is determined that the' 5.

From and after the date that tho' LPCI subsystem is, incperablef both LPCI subsystem is made or found to core; spray subsystems, the'contain , Ni t , s be inoperable for any reason, ment ccoling subsystem and the f continued reactor operation is per-diesel generators required for missible only during the succeed-operationiof such components if ' ing seven days unless it is sooner no external, source of pvist were made operable, provided that during available shall be demonstrated

' ^ such seven days the containment to be operable immediately and , k' cooling subsystem (includingit LFCI daily thereafter.

psamps) and active componeats of - both core spray subsystems, and ' the diesel generators required for operation of such components if no external source of power were avail-able shall be o prable.

\\ ! 6.

If the requircaeats' of 3.5.A can-not be met, an crderly, shutdown ' of the reactor shall be initiated and the reactor shall be in the - Cold Shutdown Condition within 24 hours.

, . v M e t i.

/ e, s

tes* ( . , 105 $ - - - _ _ _ _ _ _ f ______j _

- ' .,..... . .... ... .. .. ... . .a. . . , ,

. '[" LIMITING CONDITION FI)R OPERATION ' SURVEILLANCE REQUIREMENT w ,

3.5.5 Containment Coolina Subsystem 4. 5. B Containment Coolina subsystem 1.

Except as specified in 3.5.B.2, 1.

Containment Cooling subsystem Testing 3.5.B.3, and 3.5.F.3 below, shall be as follows: , ' - both contaitznent cooling subsystem loops shall be operable whenever Item Frequency s

irradiated fuel is in the reactor l vessel and reactor coolant temper-a.

Pump & Valve Operability Once/3 months ature is greater than 212 F, and and Once/cycit

. prior to reactor startup from a from the Cold Condition.

Alternate Shutdown i Station - b.

Pump Capacity Test After pump Each RBCCW pump shall maintenanca deliver 1700 gym at and every 70 ft. 'IDE.

Each '3 months SSWS pump shall de-liver 2700 gym at 55 i f t.

TDM.

c.

Air test on dryvell Once/5 years ,( and torus headers and nozzles .. i ) i i \\ , + .

Conditional relief granted from this LCO for the period October 31, 1980 through November 7, 1980.

Amendment No. 44 ,n.

. 106 _ _ - - _. _ _ _ _ _ _ _. _ _. - -. _ -.. _ _,, _____ _... _ . . . . - .

. ._ ._ -- - . - _. _ . _ - - __ ,... ~....... - --- . -

. .

. ,7' ~ LIMITING CONDITION FOR OPERATION ' SURVEILLANCE REQUIRDENT l s

3.5.B containment Coolina Subsystem 4.5.B Containment Coolina subsystem (Cont' d)

(Cont' d) 2.

From and after the date that one 2.

When one containment cooling subsystem containment cooling subsystem loop loop becomes inoperable, the operable is made or found to be inoperable subsystem loop and its associated s for any reason, continued reactor diesel generator shall be demonstrated . operation is permissible only during to be operable immediately and the the succeeding seven days unless operable containment cooling subsystem

I such subsystem loop is sooner made loop daily thereafter.

operable, provided that the other containment cooling subsystem loop, . i including its associated diesel generator, is operable.

3.

If the requirements of 3.5.B can-l not be met, an orderly shutdown shall be initiated and the reac- ' tor shall be in a Cold Shutdown I Condition within 24 hours.

C.

HPCI Subsystem C.

HPCI Subsystem F 1.

The RPCI Subsystem shall be oper-1.

HPCI Subsystem testing shall be per-able whenever there is irradiated formed as follows:

, fuel in the reactor vessel, reactor pressure is greater than 104 psig, s.

Simulated Auto-Once/operatin and prior to reactor startup from antic Actuation cycle a Cold Condition, except sa speci-Test fisd in 3.5.C.2 and 3.5.C.3 below.

b.

Pump Operability Once/nonth and once/ cycle from the - Alternate

Shutdown Station , c.

Motor Operated Once/ month Valve Operability and - Once/ cycle - from the Alternate f Shutdown Station ! d.

Flow Rate at once/3 months

Conditional relief granted from this 1000 psig l r LCO for the period October 31, 1980

V through November 7,1980.

e.

Flow Rate at Once/operatin l 150 psig cycle ' Amendment No. 44 . . 107 l l .. -. - . - . . - - _ , . ~ .. . .- -. _ _ _ _... _. - -..

.__ _ . _ - _ _ _. _

, 3.5.C HPCI Subsystem (Cont'd) ' 'le.5.C HPCI subsysts t (Cont'd) - ,e S e HPCI pu=p shall deliver at least 4250 sps for a sys- , .,. f tem head corresponding to a , , reactor pressure of 1000 to ' 150 psis.

- 2.

From and after the date that the 2.

When it is determined that the HPCI ~ HPCI Subsystem is made or found to . Subsystem is inoperable the RCIC, the - ' be inoperable for any reason, con-LPCI subsystem, both core spray sub- , ,, - , 'tinued, reactor operation is per.

systems, and the ADS subsyster actus- ' missible only during the succeed-tion logic shall be demonstrated to ing seven days unless such subsys-be operable i==ediately. Se RCIC tem is sooner made operable, pro-system and ADS subsystem logic shall ( viding that during such seven be demonstratei to be operable daily l days all active components of the thereafter.

- l ADS subsystem, the 3CIC system,

the LPCI subsystem and both core - , i spray subsystems are operable.

! 3.

If the requirements of 3.5.C can-not be met, an orderly shutdown '

shall be initiated and the reac- - tor prassure shall be reduced to or , below 104 psig within 24 hours.

- e ' s( I . 3 5.D Reactor Core Isolation Coeling 4.5.D Reseter Core Isolatten Ceoli..g - (RCIC) Subsystem (RCIC) Subsystem ,, , ' 1.

De RCIC Subsystem shall be oper-1.

RCic Subsystem testir.g shall be per-able whenever there is 1. radiated for=ed as follows: fuel in the reactor vessel, the re-actor pressure is greater than 104 a.

Simulated Auto-Once/ operating psig, and prior to reactor startup matic Actuation cycle from a Cold Condition, except as Test specified in 3 5.D.2 below.

b.

Pump Operability once/ month and Once/ cycle from , , ! the Alternate Shutdown Statien ~ c.

Motor Operated once/ month and Valve Operability once/ cycle fres - ,the Alternate .; Shutdown Station; - r-10S Q , Amendment No.

- - - -. - - - , _ ... -.. ,, _,,, _,,, ,,, --_ ~ _ _ _ _ -. . - -. -. . .

.-. - - . .. . -. - _ , , ... _ _.. 3.5.D React *or Core Isolation Cooline .k.5 3 neneter en a 1:aiss en emit., - (RCIC)Subsvstem (Cont'c) (ac!c) sub:yttes (cant's)

*

- - . d.

Flow Rate at Cnew/3ser.ths . c, 1000 7sig { . s.

a.

Flow Rate at Cace/=perssia; - . 130 ;sig cycle . - . .. .

The RC*C ya=p iha11 deliver at . . aca.st kCO g;= for a sy:tc= h:2

. . sorrespondit: to a reactor ;,res.

.' - sure of leco t.15e,sig . , , . 2.

Tres and after the date that the 2.

When it is deter ined that the Ac.c . RC:cs is made or found to be isoP-subsystem is ino;erable, the EPC : ' erstle for any reason, senti =ued shall be decor.strated to be o;ersLle reactor power operati n is per=is-i==st.istely and veshly thereafter.

sible only during the succeeding , , seven days ;rsvided that dring . . such seven days the XPC*S is cP*r* . able.

. . .

,3.

If the requirements of 3.3.3 cannot . ha =st, an ordarly shutdown shall

. , be ini:14:ad and the raset== pres-eure shall be reduced to or below , 104 peig with1= 24 hours.

. , ( 3 3.Z Aute=atie Dewesrei=stien 4.5.E [uta=atie flewarre-i:stie. '

Systes ( A 5) Systes (ADS) . . 1.

The Automatic Depressurizatics Sub-1.

During each operati=g cycle the

' . systen shall be operahla whenever fallowi=g tests shall be ;ert=r=si , there is irradiated fuel in the on the ADS: reactor vessel and the reactar pres.

, l sure is gres*.ar than 1016 psig and s.

A sis = lated au.o=stic cc.2:tio= , i prist ta a startup fres a Cold Can-tast shall be ;erfor=ed prirr ts dition, 'stept as specified in ,startup after each refaeli=4 oui.- , e ,

3 3 2.2 below.

age.

, . , b.

With the reactor at pressure, . . each relief valve phall be man-

us11y opened until a c rresponding

- . - change in rasctor pressure or main turbine bypass valve pc 41:imns

. indicate that steam is flowine frsa the valve - . . c.

Perform a test from the alter:ste Amendmen: No. 57 shutdova anal to verify that he f(. relief vafve solenoids actuata.

. ' - Test shall be perfomed after each . - refueling outage prior to startuo.

i 109

. ' e..... . = =.. . * * = = . -..... ..m.. .... . .. - ..y.-

=

. .. _ -, . - - - - - - - - - - -, , -.,.,,., - , - - ,.,, ~.... - -,, - - - - - - -.

' ~ T.

.. ~ ~ ~ ~ ~~ ............ _....... _ _....... .....____v.. . , .

3 5,E Auto.atic Drore::uri:stien . u.5.E Auto =atie De -essurization . System (ADS) (Cont'd) ' System ( ADS ) (Cont'd)

I 2.

From and after the date.that one 2.

When it is determined that one val-valve in the automatic depressur-of the ADS is inoperable, the ADS ization sub=ystem is cade or found subsystem actuation logic for the other ADS valves and the NPCI sub-

to be inoperable for any reason, "I**** *

**
- - 1. 1. d ** *

contiriued reactor operation is per- ' - operable immediately and at least missible only during the succeeding weakly thereafter until the valve . , ' seven days unless such valve is is repaired.

' sooner mada operable, provided that , during such seven, days the HPCI subsystem is operable.

<- . . . 3.

If the requirements of'3.5.I can-

- not be met, an ordarly shutdown - . . shall be initiated and the reec-tor pressure shall be reduced to at least 104 psig within 24 hours.

' ' .

'

a , 3 5.F Mini =u= Lov Prescure Cooline 4.5 7 Mini- - lov Pressu-e Cooling and Diesel Generator Avail-and Diesel Generator. Avail-ability ability n 1.

During any period when one diesel 1.

When it.is deter-hed that one die generator is inoperable, continued generator is W ahle, all low p reactor operation is per=issible sure core cooling and con-h-r.t

l only during the succeeding seven, h subsystecs shall be de=en- , l days unless such diesel generator - stated u be opeM1e " eda aly ' is sooner made operable, provided [ that all of the low pressure core and %- thereafter.

In addition ' and containment cooling subsystems .the operable diesel generator shs.1

and the remaining diesel generator be de=onstrated to be operable i=- . shall be operable.

If this re-mediately and daily thereafter u=: quirement cannot be met, an order-the inoperable diesel is repairsi.

lyshutdownshallheinitiated - and the reactor shall be placed ' in the Cold Shutdown Condition within 24 hours.

. . . , . 2.

Any co=binatien of inoperable com-ponents in the core and conta.in-ment cooling systens shall not de- .[ feat the capability of the re-nS-U( ing operable co=ponents to fulf4" ! the cooling functions.

! 110

l Amendment No. 15 i l ,w- - .-,- _

.- '. .

. . LIMITING CONDITION FOR OPERATION SURVEII1ANCE REQUIRD(ENT g of 3.5.F Minimum Low Pressure Cooling ' and Diesel Generator Avail-ability (Cont'd) 3.

When irradiated fuel is in the re-actor vessel and the reactor is in

the Cold Shutdown Condition, both core spray systems, the LPCI and - ' containment cooling subsystems may be inoperable, provided no work is being done which has t.he potential . for draining the reactor vessel.

. 4.

During a Aefueling outage, for a period of 30 days, refueling oper-ation may continue provided that one core spray system or the LPCI system is operable or Specification 3.5.F.5 is met.

5.

When irradiated fuel is in the reactor vessel and the reactor is in the Refueling Condition with the torus drained, a single e-control rod drive mechanism may , s( be removed, if both of the fol-loving conditions are satisfied: . I a) No work on the reactor ves-sel, in addition to CRD re-moval, will be performed which ! has the potential for exceed- . ing the maximum leak rate from a single control blade seal if it became unseated.

b) i) the core spray systems are operable and aligned with a , suction path from the conden- ' sate storage tanks.

ii) the condensate storage tanks shall contain at least 200,000 gal-lons of usable water and the - refueling cavity and dryer / separator pool shall be flooded to at least elevation 114'-0".

l 3.5.G , (Intentionally lef t blank) ,,

1

l 111 l 1-.-A-er Nc. 10 l ' ,-.

I.

, . LIMITINC CONDITION FOR OPERATION SURVEILIANCE REQUIPJMENT g.

' 3.5.H Maintenance of Filled Dis-3.5.H Maintenance of Filled Discharme charme Pipe Pipe Whenever care spray subsystems, LPCI The following surveillance requirements subsystem, HPCI, or RCIC are required shall be adhered to to assure that the to be operable, the discharge piping discharge piping of the core spray sub- . from the pump discharge of these sys-systems, LPCI subsystem, HPCI sad RCIC .'tems to the last block valve shall be are filled: filled.

1.

Every month prior to the testing of the LPCI subsystem and. core spray subsystem, the discharge piping of these systems shall be vented from the high point and water flow ob- - served.

2.

Tc' lowing any period where the LPCI subsystem or core spray subsystems have not been required to be oper-able, the discharge piping of the inoperable system shall be vented from the high point prior to the return of the system to service.

P 3.

Whenever the HPCI or RCIC system is e lined up to take suction from the torus, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system and water flow observed on a monthly basis.

4.

The pressure switches which monitor - the discharge lines to ensure that they are full shall be functionally tested every month and calibrated every three months.

. s . - d Amendment No. 39

- 112 i _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

- _- . . . ' EASES: " . , , 3 5.A Core Strav a=d IJCI Subsystem

. M i.

' This specificatien assures that adequate emergency cooling capability is . available whenever irradiated fuel is in the reactor versal.

, Based on the loss of coolant s.nalysis performed by General Electric in accordance with Section 50.46 and Appendix K of 10CFR50, the Pilgrim I Emergency Core Cooling Systems are adequate to provide sufficient c= cling to the core to dissipate the energy associated with the ' loss of coola'nt accident, to li=it calculated fuel clad temperature to less than 22000F, ' to limit calculated local =etal water reaction to less than or aqual to 170 and 'to li=it calculated core wide metal water reaction to less than or equal to 1%. Se li51 ting conditions of operation in Specifications 3 5.A.1 through 3 5.A.6 specify the co=binatiens of operable subsystems to assure the availability of the minic::m cooling systems noted above. No single fail-ure of CSCS equip =ent occurring duri:s a loss-of-coola:t accident under these li=iting conditions of operation will result in inadequate ecoli=g { of the reactor core.

. Core spray distributic. has been shevn, in full-scale tests of syste=s - s1=ilar in design to that of Pilgri=, to exceed the =1-d - require =e=ts by at least 25%. In addition, coc' r effective =ess has been decc=strated at less than half the rated flow in si=ulated fuel asse=blies with heater rods to duplicate the decay heat characteristics of irradiated fuel. The e accident analysis is additic ally ec se.rvative in that r.o credit is take= b fer spray 'ecolant enter 1=g the react:r before the inter a1 pressure has fallen to ir4 psig.

.

< - . . , . The IJCI subsystem is designed to provide emergency cooling to the ccre by fl W g in the eve t of a loss-of-coolant accident. This system functions.in co=hi=ation with the core spray system to prevent excessive fuel clad temperature. The LPCI subsystem and the core spray subsystem provide adequate ecoli=g for break areas of approxima.tely 0.2 square feet up to and includi=g the double-ended recirculatien li=e break without assistance from the high pressure emergency ccre coo 14 g subsyste=s.

The allowable repair times are established so that the average risk rate for repair would be no greater than the basic risk rate. The method and concept are described in reference (1).

Using the results developed in- . . . . my ~ . Amendment No. 15 113 , _ _ _ _ _ _

I.. - , . BASES:

., D* 3 5.A core spray snd hPCI Subeyrte.s (cont'd) v this reference, the repair period is found to bc less than 1/2 the test in-terval. Bis accumes that the core spray cubsystems and LPCI constitute a . 1 out of 2 system; however, the combined effect of the two systems to li=it , excessive clad te: pcraturce cust also be considered. The tcst interval sper- ,

ified in Specificatica 14 5 was 3 months. Werefore, an allovable repair
j period which mair.tnine t!.e bahic risk considering sincie failures should

' ., be loss than 30 days and this specification.is within this period. For - multiple failures, a shorter interval is specified and to ic; rove the as-

surance that the remaining sy tems will function, a daily test in called for. Although it is recognized that the information given in reference (1) . provides a quantitative method to estimate. allowable repair times, the lack

of operating data to supy. ort the analytical approsch prevents complete ac- , ceptance of this method at this time. Derefore, the tir..es stated in the specific items were established with due regard to judseent.

Should one core sprcy subsystem become inoperable, the remaining core spray and the LPCI system are availabic should the need for cera cocling ari:e.

To assure that the re=aining core spray and LIC subsystems end tne dic el - { generators are availt.ble, they are demonstrated to be opera,ble ir:ediately.

This de::nnstra. ion include: a manual initiation of the ;ur.ps and associated .

ve.1ves and diesel generators. 3ased on judgments of the'rcliability of the @M remaininc systc=s: 1.e., the core spray and LPCI, a seven-day repair pe.viod was obtained.

~ Should the loss " f one LPCI pucp occur, s. nearly fdll complenent of core und j o ! containment coolin6 equipment is available. Three LPCI pumps in conjunction f with the core spray subsyste= will perform the core coolin; function.

Ba-cause of the ave,ilt.bility of the =njority of the core coclirq equip = cat, e

!

which vill be de::onstrated to be operable, a 30-day rep.ir period is justi-i tied. If t'he LMI subsystem is not available, at least 2 LMI pumps must be available to fulfill the containment coolinE function. The 7-day repair . ' period is set on this basis.

, i The LPCI Subsyr. tem is not considered inoperabic unen the RilF. Symitem is operating in the shu:down cooling mode.

' , . (1) Jacobs, I.M., " Guidelines for Determining Safe Test Intervals and - Repair Times for Enginecren Safeguards", General Electric Co.

A.P.E.D., April,1969 (APED 5736) > . . . . h ' . ' 114 . . . - - _ _ _ - _ _ _.. _ _ _. _ _. _ _

- . .. .

. . . ., %. ' f BASES: 3.5.3 containment coolina Subsystem ' . The containment cooling subsysten for Pilgrim I consists of two independent loops each of which to be an operable loop requires one LCPI pump, two RBCCW pumps, and two SSW pumps to be operable.

There are installed spares for mar-

sin above the design conditions. Each system has the capability to perform '

its function; i.e., removing 64 x 106 Beu/hr (Ref. Amendment 18), even with some system degradation.

If one loop is out-of-service, reactor operation is permitted for seven days with daily testing of the operable loop and the ap-propriate diesel generator.

With components or subsystems out-of-service, overall core and containment cooling reliability is maintained by demonstrating the operability of the remaining cooling equipment. The degree of operability to be demonstrated depends on the nature of the reason for the out-of-service equipment. For routine out-of-service periods caused by preventative maintenance, etc., the pump and valve operability checks will be performed to demonstrate opn-- ability of the remaining components.

However, if a failure, design defi-ciency,. etc., caused the out-of-service period, then the demonstration of operability should be thorough enough to assure that a similar problem does not exist on the remaining components.

For example, if an out-of-service period were caused by f ailure of a pump to deliver rated capacity, the other pumps of this type might be subjected to a capacity test.

In any event, surveillance procedures, as required by Section 6 of these specifications, - detail the required extent of testing.

Since some of the SW and RBCCW pumps are required for normal operation, capacity testing of individual pumps by direct flow measurement is imprac-tical. The pump capacity test is a comparison of measured pump performance parameters to shop performance tests combined with a comparison to the per-formance of the previously tested pump. These pumps are rotated during op-eration and performance testing will be integrated with this or performed during refueling when pumps can be flow tested individually. Tests during normal operation vill be performed by measuring the shutoff head.

Then the l pump under test will be placed in service and one of the previously opera-l ting pumps secured. Total flow indication for the system will be compared for the two cases. Where this is not feasible due to changing system con-ditions, the pump discharge pressure will be measured and its power require-ment will be used to establish flow at that pressure.

e ' , o l 115 n.

t .- l , ' . f

._ .-. _ - .

e .

, - -m BASES _: , 'f * 3 5.C HrCI , ' The limit'.ng conditions for operating the HPCI System are derive.1 from the

Station Nuc1 car Safety Operational Analysis (Appendix C) and a detailed ibnctional analysis of the liPCI System (Section 6).

. . ' . The HFCIS is provided to assure that the reactor coi c is adequately cooled To limit fuel clad temperature in the event of,a small break in the nuclear - system and loss-of-coolant which does not result in rap,id depressurization ,

of the reactor vessel. The HPCIS permits the reactor to be shut down while maintaining sufficient reactor vessel water level inven ory until the ves-sel is depressurized. The llPCIS continues to operate until reactor vessel pressure is below the pressure at which LPCI operation or, Core Spray System operation maintains core ecoling.

The capacity of. the system is selected to provide this required core cool-ing. The 11pCI pump is designed to pump 4250 spm at reactor pressures between 1100 and 150 psig. Two sources of water are available. Initially, decinerali:cd water from the cendensate stora5e tank is used instead of inject-ing water from the suppression pool into the reactor.

' When the HPCI System begins operatien, the reactor depressurizes more rapicy than vould occur if HPCI was not initiated due to the condensation of steam e As the y' by the cold fluid p ped into the reacter vessel by the HPCI System.

reactor vessel pressure continues to decrease, the }*FCI flow momentarily reaches equilibrium with the flov through the break. Continued depressuri-zation causcs the break flov to decrease below the HPCI eov and the liquid inventory bc51ns to rise. This type of response is typical of the sman breaks. The core never uncov'ers and is continuously cooled throughout the transient so that no core damage of any kind occurs for breaks that lie with-in the capacity range of the HPCI.

The analysis in the niAR, Appendix G, shows that the ADS provides a sin 61e failure proof path for depressurization for postulated transients and acci-dents. The RCIC is required as an alternate source of makeup to the HPCI only in the case of loss of all offsite A-C power. Considering the JIPCI and ' the ADS plus RCIC as redundant paths, reference (1) methods would give an es-timated allowable repair time of 10 days based on the one month testing fre ' Considering this and the judgments of the reliability of the ADS and quency.

! RCIC systems, a 7-day period is specified.

The HPCI and RCIC as well as all other Core Standby Cooling Cystems must be - operable when starting up from a Cold Condition.

It is realized that the HPCI is not designed to operate until react'or pressure exceeds 150 psig and is cuto-matically isolated before the reactor pressure decreases below 50 psig. It is . the intent of this specification to assure that when the reactor is being . -( - started up from a Cold Condition, the HPCI is not knowr. to be inoperable.

% . 116 - . . .,, _,,. - _. _ _ _ _ - - -...... .... m .._ -..,. _ _. _ _., - , - - - -

i

.

. BASES: ' 3.5.D RCIC f,vstem );! . ,a - . The RCIC is designed to provide makeup to the nucicar system as part of the planned operation for periods when the norr.a1 heat sink is unavailable. The , nuclear safety analysis, ISAR Appendix G, shows that RCIC also serves as re-dundant makeup system on total loss of all offsite power in the event that HPCI is unavailable. In all other postulated accidents and transiente, the . ADS provide: redundancy for the HPCI. Based on this and judgments on the I - . -, reliability of the hPCI system, an allowabic repair time of ceven days is ., , specified. Immediate and weekly demonstrations of HfC1 operability during - RCIC outage is considered adequate based, on judcment and practicality. More frequent testin6 would cause undesirable steam flow interruption and thermal cycling transients.

. . . . . .. . - ( . .

' . . . ~ l . . . . . . . , I ~ _ l ' l . - . l l - . l C 117 ps ' ( . - . . . .

O e , _ -.,

. .. . 3ASES: - , , 3 5.E Aute=atie Deerass=1catie= Sy=:a= (A:5)

l 7.,,C.

2 e li=iting conditions for operating the AIS are derived frc= the Statics > Nuclear Operational Analysis (Appendix G) a=d a detailed functional a.aly-sis of the ADS (Section 6).

. This specification ensures the operability of the ADS under all conditions for which the aut==atic or ~""=1 depressurizatics of the nuclear systes is , .. an essential response to station abnor=alities.

. The nuclear system pressure relief system provides ante =stic nuclear cyste= - depressuricatics for s-*" breaks is the nuclea.: syste= so tha-the icv pres-sure coolant injectics (LPCI) and the core spray subsyste=s cas c;erate to protect the fuel barrier.

Because ce Autc=stic Depressurication Systes does not provide takeup to te reacter pr -= y vessel,== credit is taken for the stes= cooli=g cf the core d caused by the systa= actuatics to gevide f.'.r.her conservatis= to.he CSCS.

Perfor=asce a-*'ysis of the Autc=stic Depressuricatics Syste= is ce=sidered caly with respect to its depressuri-d g effect in ec j"-a-d-- with 1pC: c; Core Spray.

There are fcur valves provided and each has a capacity of

800,0C0 lb/hr as a reactor pressurerof 1125 psig.

. De alloa.ble cut of service ti=e for one ADS valve is deter-d ei as seven l days because of the redunda:cy and because 'the F.pCIS is de=c=strated to be w[ operable during this period. Therefore, redu=das: protectics for the cere with a s=all break in tha nuclear systa= ic st available.

y The ADS test circuit per=its cc=tinued s=ved* ce es the operable relief valves to assure that they vdD be available if re gi. red.

. l

, t

. . . . - . l ' . . . - =.t.

i t J .

118 . Ameninent No. 15 . _ _ _. -_ -. -. .- . . . - -

_ _ _ _ _.

. . . [ BASES: - -. 3.5.T Minimum Low Pressure Cooling and Diesel Generator Availability The purpose of Specification F is to assure that adequate core cooling equip- ' ment is available at all times.

If, for example, one core spray were out of service and the diesel which powered the opposite core spray were cut of service, only 2 LCPI pumps would be available.

It is during refueling outages Nthat major maintenance is performed and during such time that all low pres-sure core cooling systems may be out of service. This specification provides that should this occur, no work will be performed on the primary system which could lead to draining the vessel.

This, work would include work on certain control rod drive components and recirculation system. Specification F al-lows removal of one CRD mechanism while the torus is in a drained condition without compromising core coolink,empability. The available core cooling capability for a potential draining of the reactor vessel while this work is perfomed is based on an estimated drain rate of 300 gpm if the control rod blade seal is unseated. Flooding the refuel cavity and dryer / separator pool to elevation 114'-0" corresponds to spproximately 350,000 gallons of water and will provide core cooling capability in the event leakage from the control rod drive does occur. A potential draining of the reactor vessel (via control rod blade leakage) would allow this water to enter into the torus and after approximately 140,000 gallons have accumulated (needed to meet minimum NPSH requirements for the LPCI and/or core spray pumps), the torus would be able to serve as a common suction header. This would al-y low a closed loop operation of the LPCI system and the core spray system l (once re-aligned) to the torus.

In addition, the other core spray system is lined up to the condensate storage tanks which can supplement the re- ~ fuel cavity and dryer / separator pool water to provide core flooding, if required.

Specification 3.9 must also be consulted to determine other requirements for the diesel generators.

l . . l l - l , l . ( 119 ' Amendment No. 39 .- .- -, _. - ..... - .- - _.

- .

. . . BASES: %.f "f-3 5.G , .. I

. i ~

. . (THIS PAGE BrunC: TALLY LEFT BLAmt) , w.

p - . ? . l e

l i @ l Y 2, 120

. I .

. . BASTS: ' ~ E m.

. .. i 3. 5 11 Maintenance of Tilled Dischare.c Pipe i a_ . - \\ If the discharce Pipinc of the core spray, LTCI subsystem, IITCI, and ECIC , are not filled, a vnter ha::.cr can develop in this pipinc when the pump and/orpu=psarestarted. An analysis has been done which shows that if . a vnter hammer were to occur at the time at which the system vere required, ! the syster. would still perform its desicn function. lioucver, to minimize dacace to the discharge piping and to ensure added parcin in the operation ' of these cy:tems, this Technical Specification requires the discharge lines to be filled whenever the system is in an operable condition.

, ,

- . . . . - . O

. . . y- _ . e G / e e f

. . e e ' ', . . - p_ . . - . . ~ 121 . . . _..

.. .- - ' . ,, . . B ASE3: . . ) h 4.5 core end containmnt coolina Svstems Survet11ance trenuencies - %(~. The testing intervs1 for the core and containment cooling systems is based on induntry practice, quantitative reliability analysis, judgnent and prac-ticality. The enre cooling ystem: have not been designed to be fully test- . able during operation. For example, in the case of the IIPCI, automatic initiation durinr. power operation would result in pu:sping cold varer into the recctor vessel which is not desirsble. Complete ADS testing during power operation csuncs an undesirabic locs-of-coolant inventory. To in-screace the availsbility of the core and containment cooling cystems, the ~ components which eshe up the cystem; i.e., inscrementation, pu=ps, valves, - , etc., are tested frequently. The pu:eps and tastor operated injection valves are also testerl cach month to assure their opersbility. A sisadsted auto-antic actustion test once each cycle combined with centhly tests of the pumps and injection valves is deemed to be adequate testing of these systems.

- When cot'ponents and subsystems are out-of-service, overall core end contain- , ment cooling reliability is msintained by demonstrating the operability of the remaining equipment. The degree of operability to be dc=onstrated de-pends on the nature of the reason for the out-of-service equipment. For routine out-of-service periods caused by preventative maintenance, etc., the pump and valve operability chcchs will be perfert:cd to dc=snstrate operability of the remaining ce=ponents. Ilowever, if a failure, de:;ign deficirney, esuced the outage, then the demonstratien of operability should be thorough encur.h to assure that a generic prnbica does not exist. For exampic. if"an out-of-cervice period were caused by failure of a pu:hp to deliver rated especity due to a de-sign deficiency, the other pu=ps of this type might be cubjected to a flow rste h test in addition to the operability checks.

k Redundant operabic components are cubjected to increesed testing during equipi::ent out-of-service times. This adds further conservatics and in-creases assurance that adequate cooling is availabic should the need arice.

. . O e . e t . . . %. q' 122

. . .

3,10 _ Care Alterations ,4,10 Core Alterations i

fyylienhil i ty Applienbility ~ Applies to the fuel handling and Applica to the periodic tecting of

(~ core reactivity limitations during thooc interlocho and instrumenta-

refueling and core alterationn.

tion ur.cd during refueling and core

' alterationo.

Ohicetive_ Ohicetive_ j To ensure that core resetivity in To verify the operability of in-within the capability of the control strum:ntation and interloc!:3 used j _- rodc and to prevent critienlity in refueling and core alterations.

' during refueling.

Speciffention Specificction - .. . A.

P.cruelina Interlocks A.

Re fu eli ne,. In t er1_ocho During core citerations when fuel is Prior to cny fuel handling with the in the vessel the reactor r. ode suitch heed off the rencter vessc1, the re-

shall _be locked in the " Refuel" fueling interloch: ch:11 Le func-position and the refuelling inter-tionally tested. They ch:11 be tented locks shall be opernble, at vechly intervalc thercaiter until no lonr,cr ree,uired. They :,n 11 al o - be tected follouing any repr.ir vork ascocir.ted uith the interl'ochc.

P B.

Core I!nnitorine D.

Core l'on,itorin9, g.

During core bltera'tions when fuel is hrior t.o t:nhing cay alterations to the B in the vessel two Sp.M's shall be core the SIS 'n ch:ll be function:.11y opercble, one in the core quadrant tested and checked for neutron res-where fuel or control rods'are being ponse. Thercnf tec, uhile required to raoved and one in an adjacent quadrant.

be oport.bic, the SRl!'c vill be checked For an SRM to be considered operahic, dcily for recponce.

the following conditions chcIl be satisfied: . 1.

The Stu! chc11 be in crted to the

nore:1 operating,IcVel.

(Uce of cpecial coveabic, dunhing type detcetore during initici fuci 1cnding , and us,ior core citerstienc in pince ' of non:31 detcetcrc is pen:ionibic as long ne the detector io connected 'to the normal SM! circuit.)

. O . . \\ 202 - . . ' ..UNCONTROLLE . _= _ _._ COPY.

~ - .. .

' .

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS ' ', - - 2.

The SRM shall have a minimum of 3 Spiral Reload cps except as specified in 3 and 4 below.~ During spiral reload, SRM operability will be verified by using a portable 3.

Prior to spiral unloading, the external source every 12 hours until SRM's shall have an initial count the required amount of fuel is loaded rate of 13 cps.

During spiral un-to maintain 3 cps. As an alternative loading, the count rate on the to the above, up t'o two fuel assemblies SRM's may drop below 3 cps.

will be loaded in different cells containing control blades around each 4.

During spiral reload, each SRM to obtain the required 3 cps.

control cell shall have at least Until these assemblies have loaded, one assembly with a minimum ex-the cps requirement is not necessary.

posure of 1000 MWD /t.

C.

Spent Fuel Pool Water Level C.

Spent Fuel Pool Water Level Whenever irradiated fuel is stored Whenever irradiated fuel is stored in in the spent fuel pool, the pool the spent fuel pool, the water level water level shall be maintained shall be recorded daily.

at or above 33 feet.

' D.

Multiple Control Rod Removal D.

Multiple Control Rod Removal

Any number of control rods and/ Within 4 hours prior to the start of or control rod drive mechanisms removal of control rods and/or control , may be removed f rom the reactor rod drive mechanisms from the core and/ pressure vessel provided that at or reactor pressure vessel and at least least the following requirements once per 24 hours thereafter until all , ' are satisfied until all control control rods and control rod drive rods and control rod drive mech-mechanisms are reinstalled and all

anisms are reinstalled and all control rods are fully inserted in the control rods are fully inserted core, verify that: in the core.

a.

The reactor mode switch is op-a.

The reactor mode switch is operable etable and locked in the Re-and locked in the Refuel position Afl fuel position per Specifica-per Specification 3.10.A.

tion 3.10.A, except that the Refuel position "one rod out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanisms to be removed, after the fuel assemblies have been removed as specified below.

b.

The SRM channels are operable per i l Specification 3.3.B.4.

b.

The source range monitors (SRM) are operable per Specification 3.3.B.4.

c.

The Reactivity Margin requirements I( c.

The Reactivity Margin require-of Specification 3.3.A.1 are satisfied.

ments of Specification 3.3.A.1 are satisfied.. 203 , ljNCONTROLLE ' ' COPY.

. . -_ -- - -- - - - - - - - -. --

- - --- . . LIMITING CONDITIONS FOR OPERATIONS SURVEILLANCE REQUIREMENTS of - d.

All control rods in a 3x3 array d.

All control rods in 3x3 array centered on each of the control centered on each of the control rods being removed are fully rods removed or being removed inserted and electrically or are fully inserted and electric-hydraulically disarmed, or have ally or hydraulically disarmed, the surrounding four fuel assem-or have the surrounding four blies renoved from the core cell.

fuel assemblies removed.

e.

All other control rods are fully e.

All other control rods are fully ' inserted.

inserted.

f.

The four fuel assemblies are f.

The four fuel assemblies surround-removed from the core cell ing each control rod and/or control surrounding each control rod rod drive mechanism that is to be or control rod drive mechanism removed from the reactor vessel at to be removed from the core the same time are removed from the and/or reactor vessel.

core and/or reactor vessel.

. . . Amendment No. 41 203a P . .* ..UNCONTROLLE '- COPY.

- - _. .- - -

. ' B0510N . . EDISON

r l

\\_ l . I l u - ( - -s l NUCLEAR OPERATIONS DEPARTMENT PILGRIM NUCLEAR POWER STATION Procedure No. 5.7.1.1 EMERGENCY CATEGORIES AND ASSOCIATED EMERGENCY ACTION LEVELS

. rc- . \\- List of Effective Pages - .. 5. 7.1.1 -1 5. 7.1.1 -2 _

5.7.1.1-3 Attachments i-5.7.1.1 A-1 5.7.1.18-1 5.7.1.1 B-2 5. 7.1.1 C-1 i

5.7.1.10-2 5.7.1.10-1 5.7.1.10-2 5.7.1.1E-1 5.7.1.1E-2 l 5. 7.1.1 F-1 Approved 09[h/ y ORC Ch'dirman T,lNfif f Date . , 5. 7.1.1 -1 Rev. 4 ' . - -

. . m.a - . , 'l 1.

PURPOSE The purpose of this procedure is to define each emergency action level C for declaring an emergency in each category.

II.

ACTION - A.

Declaration of Emergency by the Watch Engineer ~ 1.

If emergency conditions are suspected, refer to Attaehment A - through E to determine if an EAL has been reached or exceeded ' and if so, the emergency class.

, 2.

Af ter taking all necessary imediate response actions, declare the emergency and proceed to the next step.

3.

Follow the instructions given in the proper procedure for that emergency class declared.

. ' Personnel Emergency - Procedure No. 5.6.1 - Unusual Event - Procedure No. 5.7.1.2 - Alert - Procedure No. 5.7.1.3 , - f Site Area Emergency - Procedure No. 5.7.1.4 - - General Emergency - Procedure No. 5.7.1.5 - r 4.

As conditions change, the emergency classification may have to \\._ be upgraded or downgraded as stated in Attachments A through . t %

- E.

. - < y , h.

5.

Request the Security Supervisor to call-in additional - W operations personnel as necessary. N_QT[: Make every "ife practicable effort to minimize personnel radiation exposure.

Do not exceed normal radiation exposure restrictions without the specific approval of the Emergency Director (Watch

! Engineer). The urgency and potential risk of the emergency 'l situation will determine the degree of precautions to be taken (i.e., surveys, respiratory protection, protective clothing).

III.

RESPONSIBILITIES , .i e.

It is the responsibility of the Watch Engineer to ensure any emergency situation is promptly identified and properly classified and the ! l appropriate class procedure is implemented. It is also his responsibility to initially function as the Emergency Director until , ,i properly relieved of this duty.

- ' It is the responsibility of the Station Manager to relieve the Watch Engineer as Emergency Director as soon as possible after being notified of an emergency. It is also his responsibility to ensure the proper response, assessment and corrective action is implemented in accordance with the appropriate procedures.

\\( , 5. 7.1.1 -2 Rev. 4 ' . I , _ _. _ . _ -. _ _ _. _ _. _ _ _, .. --

v

- ..; w ec . , ' ". IV.

DISCUSSION , . Emergency situations are classified to cover the entire spectrum of I( possible radiological and non-radiological emergencies that may be - .c ' encountered at PNPS. The classifications are: k 1.

Personnel Emergency _ 2.

Unusual Event . '. ?- . , .

3.

-Alert % l' 4.

Site Area Emergency j g L 5.

General Emergency li Accidents may be initially classified in a particular category and later changed to another classification if conditions warrant.

/ ' - . - . Each of the emergency classes are characterized by Emergency Action . Levels (EAL's). Except for the Personnel Emergency Class, the EAL's c , ' l consist of specific sets of plant parameters (i.e., instrument ' indications, system status, etc.) to be used to initiate emergency class designation, notifications, and mobilization of emergency organizations.

o These EAL's are sumarized in Attachment F. "EAL CHART." A controlled . copy of the EAL CHART'is located in the following Emergency Response ll Facilities: Control Room, Technical Support Center (TSC) Emergency Operations Facility (EOF), and the Recovery Center. If an EAL is revised. Document Control must be notified in order to ensure that the > , [(,. , controlled copies of the EAL CHART are appropriately modified.

, _ , f., ' 'r ' #b, p.,.. +4 A , ' The on-duty Watch Enaineer, initially acting 'as' the Emergency' Director '- l 7 . ^ 3 -pJg.&. has the ultimate responsibility to initially classify and declare ' w; @s = ., ' -. - emergency conditions based on the EAL's. He may also declare an '-

- emergency condition based on any event that may affect the safe operation

of the plant or the health and safety of plant personnel and the general ' l public.

. V.

ATTACHMENTS A.

Personnel Emergency - Emergency Action Levels B.

Unusual Event Emergency Action Levels ,, , l5 C.

Alert Emergency Action Levels Ig.

0.

Site Area Emergency Action Levels ( E.

General Emergency Action Levels j i F.

EAL CHART

. o ( 5.7.1.1-3 Rev. 4 ' ' . i ' I - .- . .- _ _ _ ._ _ __

. . . . ATTACHMENT A - Personnel Emeraency - Emeroency Action Levels 1.

Onsite emergency medical treatment required for any individual with g ' without evidence of internal or external contamination.

2.

Offsite emergency medical treatment required for any individual without evidence of internal or external contamination.

"

~ . - . %. i _ '

s.

- . .; ~ ,,,. . h ' . t I li , r- [ ' ei T i: l _ 5.7.1.1A-1 Rev.,4 - - -. - - -. - -

. , _ ___ . . . . - ATTACHMENT 8 - Unusual Event Emeroency Action levels 1.

Release rate of airborne or liquid radioactive effluent technical specification for 15 minutes or more as indicated by the Main Stack Reactor Building Vent or Liquid Waste Discharge Radioactive Effluent Monitor.

-

- 2.

Greater than 500,000 uCi/sec at air ejector for

,"

15 minutes or more or an increase of - , - 100,000 uCi/sec within a 30 minute period using , '- the most recently calculated conversion constant ' 3A - from ar/hr to uti/sec 3.

Greater than 20 uCi/ml of total iodine in reactor water (confirmed).

g 5' 4.

Irradiated fuel in vessel, reactor coolant

,

temperature greater than 212*F and any of the i t following conditions: a.

Reactor coolant system unidentified leakage greater than 5 GPM and total leakage greater than 25 GPM when averaged over a 24 hour - period as indicated by drywell sump flow integrators.

- b.

Reactor coolant pressure in excess of tt hnical specification safety limits.

( c.

' Failure of a safety / relief valve to properly close after reduction of applicable pressure as-indicated -- s~ by thermocouple or acoustic monitors.

- ' IE d.

' Primary containment isolation valves inoperable l' requiring shutdown according to technical specifications.

Loss of ECCS or fire protection systems requiring e.

l plant shutdown according to technical specifications.

f.

Loss of all onsite AC power capability.

, g.

Loss of indication and annunciation on any safety related system requiring plant shutdown

-

,'~ according to technical specifications.

h.

A fire onsite that is not controlled within 20 minutes af ter fire fighting ef forts have begun.

. 5.7.1.18-1 Rev. 4 _ - - - - - - - - - - - .. . - - - . . -

_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _

, , . d ATTACHMENT 8 - Unusual Event Emeraency Action levels (Continued) ' 5.

An earthquake that causes observed station damage.

6.

Onsite airplane crash.

7.

Unplanned explosion onsite.

".

' - 8.

Notification of release of toxic gases within 1 mile of site boundary.

9.

Confirmation of a tornado touching ground onsite.

10.

A hurricane with sustained ( > 15 minutes) wind speed greater than 90 mph at the site.

11.

Transportation of an injured person offsite for - emergency medical treatment with evidence of internci or external contamination.

12.

Attempted entry onsite with evidence of intent to sabatage.

- O . % , . , . - - . , hg w ( 5.7.1.18-2 Rev. 4 . . _ _ _ _ _ _ _ _ _ _ _ _ _ _

.

. . c 4'e , I ATTACHMENT C - Alert Emeroency Action levels , 1.

Release rate of airborne or liquid radioactive effluents in excess of ten times technical specifications for 15 minutes or more as indicated '- by the Main Stack, Reactor Building Vent or Liquid Waste Discharge Radioactive Effluent Monitor.

2.

Greater than 5 Ci/sec at air ejector for 15 minutes

, or more using the most recently calculated - . - conversion constant from ar/hr to uti/sec.

s_ 3.

Greater than 200 uti/ml of total iodine in, reactor water (confirmed).

4.

Irradiated fuel in vessel, reactor coolant temperature greater than 212 degrees F and any of the following conditions: a.

Reactor coolant system total leakage in a excess of 50 gpm average over one hour as indicated by the drywell sump flow integrators.

I b.

Loss of all offsite power coincident . with loss of both emergency diesel generators.

' - Loss or potential loss of habitability of c.

Control Room as witnessed by physical /'~ indicators (eg. smoke, fire).

i ~ / d.

Any natural phenomena that could potentially -.,, - impair ECCS capability.

e.

Failure of the reactor protection system to ' l properly shutdown the reactor on valid trip signals after immediate and reasonable corrective actions.

. f.

Loss of all onsite D.C. power (24,125 and 250 VDC) for 15 minutes or less (greater than 15 minutes is a Site Area Emergency.

l g.

A fire within any process building that _ ,, ! has the potential to affect safety systems within one hour or less if not controlled.

_h. Loss of all control room annunciators for 15 minutes or more.

5. 7.1. l C-1 Rev. 4 k l - - -. .- - - .. . . _ . . . - - - - - - -. .... _. - _ - -

... .i

. . .-- - .w 2 9xsp.

.- . , . . ATTACHMENT C - Alert Emercenev Action Levels (Continued) ' 5.

Airplane crash onsite that causes substan-tial observed damage to process buildings or switchyards.

6.

An earthquake that causes substantial ) observed process building damage.

-., -,.

- s... 1.' 7.

Direct radiation levels observed in the plant - c.

.. - ' '*' which increase by a factor of 1000 over a period ' of 1 hour or less. (Excluding the results of controlled processes such as in-core neutron detector withdrawal into the TIP room.)

8.

An ongoing security compromise requiring , ' assistance by an offsite security force.

9.

Complete loss of safety related systems ~~ .. required by Technical Specifications necessary ' ' "O ~te ^ to neintain the reactor in a cold shutdown condition.

10.

Any other plant condition which in the judgment of the Watch Engineer warrants increased awareness - by offsite agencies.

! l - l ,-

( . ~ s ' = . ' \\ ,. <. - _g.

%. '"'h ll % * ?.'N

hb**,,

' % .. . , ,e ~

'Y-4[*

y,.

,. .. y . - -;q.p.[,l..g fh," g ..c 7 4.g.g,.9 ;;. ', '- . , f.

'j7ep : + , .. .- a . -, ,, , - - -. -_ __ ., S.q-- t, , ~

, .

( 5. 7.1.1 C-2 Rev. 4 l l ! -

, . . r . <. . I - , . ATTACHMEN,.T 0 - Site Area Emergency Action Levels k ~ 1.

Irradiated fuel in vessel, reactor coolant temperature greater than 212 degrees F and any of the following conditions: a.

Reactor vessel water level below the top of core ($ inches on 903 panel) and decreasing.

Y b.

Main steam line rupture outside primary

. ' containment that is not isolatable.

, * . c.

Loss of both of fsite power and onsite AC power for more than 15 minutes, ie. loss of Line 342 and 355 and shutdown transformer and both diesel generators.

d.

Loss of all onsite DC power (24,124 and 250 VDC) for 15 minutes or more.

i e.

Failure of the reactor to shutdown on manual initiation within 15 minutes with MSIV's open (with MSIV's closed this is a General Emergency) f.

Loss of all annunciators coincident with any g Alert Emergency Action Level.

l - ! -g.

Evacuation of the main Control Room without /-. establishing the ability within 15 minutes ( 'to begin shetdown procedures.

Anairborneradioactiveeffluentreleasewhich)esults . 2.

in a calcuated dose rate at any offsite location greater ,. 2 than 500 mres/hr for 2 minutes to the whole body or '""2 - . 2500 n.i.tm/hr to the thyroid under adverse meter' logy - o as indicated by a level of 5 x 105 cps on the Main Stack or Reactor Building Vent Radioactive Effluent Manitor.

3.

An airborne radioactive effluent release which results in a calculated projected dose at any effsite location in excess of 1 Rem whole body or 5 Rem thyroid using actual meteorology.

'4.

Drywell high range monitor reading in excess of 105 R/hr .. (equivalent to 1 Rem whole body at the site boundary in 10 hocrs).

l S.

Results of environmental measurements tf.at indicate l orojected doses in excess of the EPA Protective Action Guides of 1 Rem whole body and/or 5 Rem thyroid for the ' expected duration of the event.

6.

Loss of physical control of the main control room by l the Boston Edison Operators.

5.7.1.10-1 Rev. 4 - _. _. _ _. _ - .. . ... .. _ . . .

' . ... x -n ;+:. ' . , . . l ATTACHMENT D - Site Area Emeraency Action Levels (Continued) , 4 l 7.

High alarm on at least two refueling floor process radiation monitors for 10 minutes or more caused by an actual release of radioactive material from a fuel bundle (s).

, , . s-w .. .% . .

. - , .

- . . ~ .... p - , . =:.e: n,- , l . '? . ,' . ., . . . -y;,, ~ r

,. <

. ~ - w.

.

.- < 4 49,.

l . . . e.: . . . .

_ ' 5.7.1.10-2 Rev. 4 '.

i - - - - - - . - - - - .. . - -. -

__ _ _ _ . _ _ . "s - - . .,. + a..,..u,. . , . ATTACHMENT E - General Emeroency Action Levels .

./ 1.

A sustained (one hour) airborne radioactive

( effluent release which results in a calculated - dose rate at any location offsite in excess of 1 rem /hr whole body and/or 5 rem /hr thyroid under actual meteor clogical conditions.

. ~ 2.

-Drywell'high range monitor reading in excess of ^

*

,, . y m.2y \\ ~'~ site boundary in 10 hours) ~n .; .100 R/hr (equivalent to 5 Rem whole body at the . ~ ' %. , ' ',." " " ' - _" , 3.

Results of environmental measurements that indicate projected doses in excess of 5 Rem ' whole body and 25 Rem thyroid.

4.

The' occurrence of any two of the three events listed . . . below (A, B and C) with the potential for the third - j.'o .. ' ' event to occur as indicated by water level, pressure ! s - 'j A;, or radiation level trends while there is irradiated , fuel in the vessel and reactor coolant temperature ' is in excess of 212 degrees F.

a.

Loss of Fuel Cladding as defined below:

(first fission product barrier) ' I -- 1.

Greater than 200 uCi/ml of I-131 (confirmed) in reactor water or, . . 2.

Greater than 5 Ci/sec (corrected for a 30 minute - . s.

a..,t.. decay) at the air ejector using the most recently - z .'

.

m :, q q q ; i s; g i s.r alculated conversion constant from ar/hr to uti/sec or,1 - , N] y g, - .4 - 3 2.. .. _ _ _ , l@f?dM,%#E.#lOp?@3gHighradiationleveltriponmainstreamline. monitors (2 out of 4) caused by - rc e b.

ME .- .

,._,, j- - - .- \\ 4.

' Reactor vessel water level below the top of the

core (# inches indicated on 903 panel) for 15

~ minutes or more.

b.

Loss of Reactor Coolant Pressure Boundary as indicated below: (second fission product barrier) '! i ds ~ waM;. g. y g, 2 41.

Drywell atmosphere monitor reading in excess of ' v - c.

, 100 R/hr or, 2.

Drywell pressure greater than 10 PSIG and increasing ~c.

Loss of Primary Containment as defined below : (third fission product barrier) , 1.

Failure of any inboard and outboard isolation valve 5.7.1.1E-1 Rev. 4 ( . .,--. , ,, - - _. . ~. _ - - - - - _.... _. _ -,. - - '

. ..n,.: - . . . . - _ . ATTACHMENT E - General Emergency Action Levels (Continued)

. in series to close when an isolation signal is present or, .

' 2.

Drywell pressure in excess of 56 PSIG and increasing or, , 3.

Radiation level on Main Stack high range monitor in excess of 5 R/hr.

.. -~ . .- . ... ~ h. ',g S. ' operations or security personnel.

'~ 5.., Loss of physical control of the Reactor Building by Boston Edisgn . ". ' ',..% t-6.

Failure of the reactor to shutdown on manual initiation within

~

15 minutes with MSIV's closed.

7.

Irradiated fuel in vessel, reactor coolant temperature - greater than 212 degress F and any of the following conditions: a.

Reactor vessel water level below 2/3 core coverage (-48 inches indicated on 903 panel) and decreasing or, b.

Any combination of events or failure that are likely to result in the loss of any of the following functions within 24 hours af ter reactor shutdown: . . 1) Ability to reliably maintain water over the reactor core . ii) Ability to reliably remove decay heat from the reactor core .. .iii) Ability to reliably expell decay heat to the ultimate heat sink.

- .. -. - , ~ . , _ .. - py n , _ ;. g - :r.7

'

,. '

_, . g' p.<; : _ _;.' < , ., , .. ..lV, 5j5 -{{;.Q _ h'.pQs.

, ., bfh, .$fq*{ N',:&' O.* * %;, ~ % '. k a.

.' e*-

a , , , .~ .-, . w <

,jf n-

.. ,

q kj:..

,.;,, .4

c

_ . 5. 7.1.1 E-2 Rev. 4 ' . . .---________________._m_

.. _ _ _ - - . .n'. < , j . . . ATTACHMENT F EAL CHART

. . 47 USUAL t.yt.NT 3.T.E.3 f. met C.Y SEI. TRAL t t.h.et.N*T ALERT PwP.. ppd P.

.. . F .e P e. O,T.O.s kJ.1J LP.1J L7.9 4 L7,1.8 e A .- .w... . ,. , .n ...=.. .. .w .. .e.

, . . - ,..,..,..n.

.... =. . * . . =, .. ., . . . . ... . m i.m... o i.,... =.=, .. .a.-. w.

-... - .- , r.

.h.b g gg g i .n.r.i r ..

- "'""'s",

D.,., '1 a.c. , e.,ti.r =,. s -.,. w.

a = - . ,.. .r . . = .. ,.. - .. .. , . / h ft 1_ w us wgay,. I.r r.t

_ _ w a m.,w . . n u , .m.,"In . ... , . , , ....s . ,. n . ., . , . sma ~ , A. s ,.n = ...,=== A .., W Da r g W .s . .c,., m.s =. e, ,.; w .. . . r. mm.c. .n . t wic .,..c.. ., aa .- . .. .. g,.,,.,,g -.c .~. m.~. . = - mm =i . -... . i===- ...c ,,,,,,,c,,,,,,,,,,,,,, e r. g Lgir.: c.c,,,,,,,,

.c c w :r.

..c

.e m.ci.. ..-. . .-. - e m.o.. . """" ~ ' - esp.a .,. s 0.c r . ., $, e n n - .., c m -. ,.,.,.u, m...s.

. m%. e.. w.. =., = m.,., "r.L.. =w i i m e asuni .= .a .a .,a...-..=. . -. = - - . . "'"'$ , - - . e, s..-,m e . . 4. . W.t t.s oi,--= = pw = = rs: i ,-n

e . R .f W..Wr.e I . ,.a , - M ,i4.

. ..v . .. m.= .a . . s.

u. tor.th .r , n,.<., -. .a .. .-. ,.n,.5 .u.u.n.. ,,c . . .a.

. uni e.wr= . ...2 .., .. n.

-.==.1 eam.in . . u.. ., r, m, -. . m.es a w _ t,., m. = = - = .. . , .- -. -=-a mi.-. ...u.a .

- " " " *

) .,=.o.e...,.,..,- .f z C ,R , ig nN. c c.

.

u i.sr..,e ., . .. = a - . . u... .. . . . . .. s,rr <

* * = * *

=a

, -.. r = =. u., ' . ., = ,,,,,,,,,.,,,,,,,,,,,,.,,,,,,,,,,,

== C".a a % . .s -. . ..

"==

.k . .,= - - t.

'.g ? pf-. ,e,a .. .. - . . ..:,, ,. . rir

"-

v

. i ca t .. n .~- e i W e,er ', r.,ga.;si3rr i ~...n-insiw oeu u... - e - ,m u... n -oseamoa.e i.

u.w = .. = .- w. m. . . m -.a u, 6-e.c3..n - = = = =. u,.,,.,,., x - o w a veo.. s,w , 'g . ors m , .m m

, ... .iii . . =... -. nu .a .=i ...... ._- i+ I . ~,. . '. .iumas m.

. . -., w ! .

i C e., e. r =m ' " " " ' * . a.c y.ti> 6 e p ih, E *. r ". 1 , v ..,~ . , -a ,,., . .. ... .n.. cria x.

r.- ..,,, ,..-, .a . t a.6 mp r.1c.n

, ww ~ .b pag te t.t.fsteri g

.it.., p I

& ..

==,....o.~, ~== a .1 m yser*.a .,.ir.. r, .a,,=..= c,c tea ic.n..ie.n.iegi.a..' a..r:a .ex i.r .rr ,...x., .. ~. l - ,ni . - . w, l / s. .m.i-o . ls \\ l 5.7.1.1F-1 Rev. 4 l' - -- -- - - ,- - - - __ .... _. _ _. _ _, _ _ _ _ , _, _ _ __ . _

. d,.

_g % . MHC SYSTEM ., - TURBINE SPEED SET POINT ADJUST LOAD i - + $* f~ LIMIT

tvo t y \\ , . l SET g o . \\ ' y PolNT h ( ._g_,_,_ n SPE ED

- 2.5 % F LOW / 1 * ERROR ' toAo + ' . j

t

' ~ - - - - - - _.

n , cHAwarn aoy ! 4 7] ,80 g i l MSL i coNTaot

' , ' A c c e t e n 4 Tio n t_ _ _ _ _"__cH,' PRESSURE. j.

i5

' , g nELAY - ' . o l} RUNBACK L_________________, EPR Y Y N TROL 8. 0. J. M.

c c c _ VALVES - SET POINT , ADJUST Y SE T ! PolNT

j , - , - BYPASS . __7 z

__

._ yg gyg g

JA S L Y \\ d l g g PRESSURE I 3.j D I y ' 2.5% FLO W /1 e E R ROR , t _ _ _ _ _ _ _ _ _ _ _ _ __i MPR

LVG = LOW VALUE GATE

~ HVG = HIGH VALUE GATE . Figure 1

Rev.0 8/85 MHC Lesson Plan

U ~$.*

m __ l

MHC SYSTEM ., - TURBINE SPEED SET POINT ADJUST LOAD s - + e LIMIT 4,o f{ POINT h r SPEED- - - - - -, , ,' CHANGERLoAo + j l 2.5 % F LOW / 1 * ERROR -

,~~~~~~~~~~*1 i coy , , ! MSL I Y[] l C$7cH "

g

8 ' " ' PRESSURE,' \\ ' I I-RELAY

s FLOW LIMIT-r-

O l} RUNBACK L _ _ _ _ _ _ _ _ _ _ _ _ _ _ a, ! NN B ' O' J' M ~ C C C _ V VALVES - SET POINT l ADJUST Y

, ) . - - BYPASS p [- 1,------ -_1 V A LVE S i - t.S l g MSL PRESSURE I \\,j i

-}- l ' 2.5% F LO W / 1 e ERROR t______________i MPR

LVG = LOW VALUE GATE

' HVG = HIGH VALUE GATE Figure 1

Rev.0 8/85 MHC Lesson Plan I1 JOI \\1 A~

SRO-4 o SECTION FIVE QUESTION 5.01 (1.5) - For each of the following conditions, state whether the change will bring the system closer to, or have no effect on the power at which the onset of transition boiling would occur. Assume all other parameters remain constant for each case.

Briefly explain.

a.

Increase in coolant flow rate.

(.5) b.

Decrease in coolant subcooling.

(.5) c.

Decrease in coolant system pressure.

(.5) ANSWER a.

Farther from (.25); each lbm of fluid soends less time in contact with the clad surface (.25).

b.

Closer to (.25): A h reauired to achieve critical auality decreases as the inlet enthalov increases (.25).

c.

Farther from (.25); from a lower oressure. hfo is hiaher (.25) QUESTION 5.02 (4.5) The reactor is exactly critical near core end of life (E0L). Control rods are withdrawn to insert.0008 4K/K.

a.

What is the resulting period? (State your assumptions and show all work.)

(1.75) b.

How long will it take for power to increase by a factor of 10? (Show all work.)

(1.75) c.

What would the period be if.0058 A K/K had been added to the " exactly critical" reactor? (Show your work.)

(1.0) ANSWER Notes to examination grader... 1.

Accept.1 sec-l as A (from formula sheet) or any value from.08 sec-' to.1 sec-I (indicating a normal up power transient) for part a. of this question.

2.

Accept.1 sec-I as 4 (from formula sheet) or any large value for A that is given by the student to represent a prompt critical f (3.01 sec-I is given by the General Electric . academic series as prompt c itical 4 ). This applies to part c.

3.

Accept.0056 as B core for E0L (or a number close to it, obviously representing B core E0L).

(*** CATEGORY 05 CONTINUED ON NEXT PAGE***) ,

- _. . ANSWER: . a.

Assumptions: B core EOL .0056 (.25 pt) /* - 10-14sec .1 sec-I [-1+Bf e le T - 10-4 +.0056 .0008 (.25 pt) .0008 .1(.0008) f .0048 - 60 seconds (1 pt) .00008 b.

E__ et/7 (.25 pt) . Po 10 - et/g 10 - et/60 (.25 pt) in10 - t/60 t - 2.3 (60) (.25 pt) t - 138 seconds (1 pt) T =l^+B f C.

_ f jC y - 10-4,.0056 .0058 .0058 .l(.0058) T = 10~4 .0172 seconds .0058 QR if f added exceeds B core, the reactor is prompt critical (1 pt).

(*** CATEGORY 05 CONTINUED ON NEXT PAGE***)

[ QUESTION 5.03 (4.5) . For each of the pairs of conditicns listed below, state which condition would have the areater differential rod worth and why.

. a.

Reactor moderator temperature of 150 Deg F or 500 Deg F.

(1.5) b.

A rod moved from position 00 to 20 or a rod moved from position 24 to 48 in a reactor performing a startup.

(1.5) c.

For a rod near the outer perimeter of the core or a rod near the center of the core at 100 percent power.

(1.5) ANSWER a.

Reactor moderator temperature at 500 deo F (.75), since thermal diffusion lenath will be much areater (.75).

b.

A rod moved from cosition 00-20 (.75) due to the hiaher neutron flux lean (.75) and the lower control rod density in this region of the core.

c.

A r_QLDear the center of the core will have the greatest differential worth (.75) since the local flux is hiahest with resoect to averace flux in the core (.75).

QUESTION 5.04 (3.5) a.

For the following transients, indicate which coefficient of reactivity tends to change power first and in which direction - alpha T, alpha D or alpha V.

NOTE: No explanation required.

1.

Fast closure of one MSIV.

(.5) 2.

Control rod drop.

(.5) 3.

SRV lifting and then resetting - consider both in your answer.

(1.0) ANSWER a.

CONDITION COEFFICENT POWER 1.

Fast closure of one MSIV voids (.25) up (.25) 2.

Control rod drop Doppler (.25) down (.25) 3.

SRV a.

lifting voids (.25) down (.25) b.

reseating voids (.25) up (.25) (*** CATEGORY 05 CONTINUED ON NEXT PAGE***) , .-

~ 'b.

Power is increased by raising recirculation pump speed, thereby, . increasing flow through the core.

Increased flow orovides a shorter boilina length (.3) which in turn reduces the cercent voids (.3).

Lowerina the cercent voids adds nositive reactivity (.3) from d v, . which will cause a rise in oower until a sufficient increase in oower has caused once aaain a hiaher cercentaae of voids (.3).

However, durina this oower increase. the fuel temoerature is risina. causina neaative reactivity to be added via em D (.3).

Therefore, to a:hieve a new higher steadystate power where reactivity - 0, the initial positive reactivity addition caused by the removal of voids must be matched by an equal amount of negative reactivity to be added via osv and av d.

Since the fuel temperature increase ( o< D) contributes some negative reactivity, the void fraction need not contribute as much.

Therefore, at a higher power, void fraction will be lower.

QUESTION 5.05 (2.00) Following a reactor scram, from 100% power, explain what happens initially to the following parameters (increase, decrease, or remain the same) and why.

a.

Actual core flow.

(.5) b.

CRD flow (through the pumps).

(.5) c.

Control rod drive mechanism temperature (after accumulator has fully discharged).

(.5) d.

Steam line pressure drops (with MSIV's open).

(.5) ANSWER Ref. PNPS Heat Transfer Fluid Flow, Thermal Limits Handbook and Recent PNPS SR0 NRC Exam (12/11/84), a.

Increase (.25) due to the reduction in head loss due to the steaming and void formation reduction (.25).

. b.

Flow increases (.25) as pumps attempt to recharae accumulator (.25).

c.

Increase (.25) due to FCV shuttina because of "b" above which reduces coolina water flow (.125) AND Hi temoerature reactor water introduced throuah ball check (.125).

d.

Decrease (.25) due to reduction in head loss due to reduction in steam flow rates (.25).

(*** CATEGORY 05 CONTINUED ON NEXT PAGE***) _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _

QUESTION 5.06 (4.00) . Hith regard to Delayed Neutrons . a.

Define Beta.

(1.0) b.

Exolain how and why the value of Beta changes from beginning of core life to the end of core life.

(3.0) ANSHER a.

Beta is the fraction of all neutrons produced in a generation that are delayed.

(1.0) b.

The value of Beta varies with fuel type. At any time in core life.

the value of Beta in the core (B core) is the averaae of B for the different fuel tvoes. weiahted by the fraction of fissions in each fuel tvoe (1.0).

At BOL with no Pu-239. Beta core - 0.007 (1.0).

el EOL. Beta core - 0.0054. because a large fraction of fission is Pu-239 (1.0).

The same weighted averaging technique will give Beta core for any distribution of fuel types and fission fractions. The above example demonstrates that Beta core causes the reactor to respond more quickly to reactivity changes.

QUESTION 5.07 (3.00) 1.

With regard to thermal limits: a.

State the three PNPS thermal limits.

(.6) b.

Discuss the purpose for the establishment of the three thermal limits.

(What adverse situations are avoided by observance of > the thermal limits?) (1.5) c.

State the acronyms used by the process computer in presenting the thermal limits to the operator.

(.9) ANSHER a.

Minimum Critical Power Ratio (MCPR) (.2) Average Planar Linear Heat Generation Rate (APLHGR) (.2) Linear Heat Generation Rate (LHGR) (.2) b.

MCPR (.2) - by not allowing MCPR to go below 1.07, we orevent transition boilina in 99.9% of the fuel bundles under transient conditions.

(.3) APLHGR (.2) - maintaining APLHGR per Tech Specs ensures clad won't heat un greater than 2200*F if coolant is removed (LOCA) based on stored heat in a node. (.3) (*** CATEGORY 05 CONTINUED ON NEXT PAGE***) - - - - . - - . .

LHG8 (.2) by not exceeding limit of 13.4 kw/ft, will not oroduce . - 1% olastic deformation of fuel nins under steadv state conditions (.3) . c.

MCPR presented as MFLCPR (or FLCPR) (.3) APLHGR presented as MAPRAT (.3) LHGR presented as MFLPD (.3) QUESTION 5.08 (2.00) The plant is being cooled down from hot operating conditions. Given an initial pressure of 935 psig, what would be the plant pressure two hours after the start of the cooldown if the cooldown rate was maintained at the maximum allowed under normal conditions? (See attached section of Steam Tables.)

ANSWER Assume: maximum cooldown rate - 100*F/ hour.

(.4) Thus, 200*F temperature change has occurred.

935 osia - 950 osia (.4) then TSAI - 538.4 F (.4) 538.4*F - 200*F - 338.4*F (.4) If TSAT (t - 2 hours) - 338.4*F, then PSAT - 115.5 psia or 100.5 psig (.4) (Grader: accept answer in psia or in psig) (***END OF CATEGORY 05***)

- . g , TYPICAL N.R-C.

EOUATION SHEET . F = ma y= s/t Cycle efficiency = V = Tfr2( (cyl) s = V t + 1/2(at )

o A = AN A = A e-at o W = ag a = (Vf - Vo)/t E = mc2 A= En2/t1/2 = 0.693/tft KE = 1/2(mv2) ATg = Q T. ) - ( A Ti ) PE = agh En( A T. ) (AT) a ,,, vg = v + at h + Ehin = h/J + hhout ' A E = 931 d m h=u+PV I = I e-#X I = I e-EX c c 6 = ECp A T k = A Ah I = I 10-x/TVL o h = UA ( A T)Ln Ah = Cp AT Tvt = 2.303/g P = P 10sur(t) h = h, + X ht, HVL = 0.693/g o

P = P et/T @ = invaf/g CR (1 - Kegg 1) = CR (1 - Keff 2)

2 o - - - ... - - -. - - - - - - - - - - - - - -. - - - - - ,.. _ _ _ _.. _.

, SUR = 26.06/T in=pAv sCR = s /1 - Keff c

T = (f.*/p ) + (($ p )/Ep) H = 1/(1 - K gg) = CR /CRo e

P= A6* + 8 M = (1 - Kegg o)/(1 - K gg 1) e %T+1 SDM = (1 - K,gg) /K rg (* = 10-4 seconds p=(Keff-1)/Keff = A Keff e 5 = 0.1 seconds-1 P = (EQV)/(3x1010) Idii=Id22 E = FN Idf=Id l

. . R/hr = (0.5 CE)/d2(meters) ..7..._ _.., - WATCR PARAMETERS (atmospheric conditions) MISCELLANEOUS C0!!VERCIO!!S , 1 gal. = 8.345 lbm. = 3.78 liters 1 curie = 3.7 x 1010dps 1 fL3 = 7.48 gal.

1 kg = 2.21 lbm _ Den:lty = M.4 lbm/f t3 = 1 gm/cm3 1 hp = 2.54x103 Btu /hr = 550 ft-lbf/ cec Ileat of Vaporization = 970 BTU /lbm 1 M'.J = 3.413 x 100 Blu/hr Heat of Fu: ion = 144 DTU/lba 1 in.

2.54 cm

1 ATM - 14.7 psi = 29.92 in. lig, of = 9/5 oc + 32 oc = 5/9 (oF - 32) - . se . ,f , -. . +. - -, - - - - -.

SRO-4 . SECTION SIX-QUESTION 6.01 (4.5) - Hith regard _to the minimum flow valve (M0-60) for the RCIC System: a.

What four conditions will cause it to close? (2.0) b.

State what specific conditions must be met to satisfy-the opening logic for the valve.

(2.5) ANSHER a.

Close M0-60 1.

Closes on turbine trip (.5) 2.

Closes with control switch (.5) 3.

Closes ~on high RCIC flow (> 80 GPM) (.5) 4.

Closes on high Rx water Lvl (+48") (.5) b.

Open M0-60 (logic) 1.

Turbine not tripped (.5) 2.

AND Rx water lvl 1 +48" (.5) 3.

AND a.

control switch to 'open' (.5) b.

DEReactor water level lo lo (initiation signal) (.5) A.ND Low RCIC flow (< 80 GPM) (.5) NOTE TO GRADER: For the opening logic, the correct conditions with improper AND and OR conditions is sufficient for half credit.

QUESTION 6.02 (2.5) State the three auto isolation signals to RCIC (include setpoints) and identify the two valves affected directiv by auto isolation.

ANSWER 1.

ISOLATIONS a.

High steam line flow (300% w/TD) (.5 each) b.

Steam space temp high (150* - 200*F) (.5 each) c.

Rx pressure low (100 psi) (.5 each) (*** CATEGORY 06 CONTINUED ON NEXT PAGE***)

.. _ _ _ _ _ _ _ _ _ _ _ _ a.

Inboard steam line isolation, M0-16, shuts (.5) b.

Outboard steam line isolation, M0-17, shuts (.5) - NOTE: Valve numbers not required for full credit QUESTION 6.03 (3.00) A ground overcurrent situation causes a lockout of 4160 VAC bus A-5.

Assuming the A-5 bus was initially powered from the Unit Auxiliary Transformer, answer the following questions True or False.

a.

The Unit Auxiliary Transformer breaker to A-5 will trip.

(.5) b.

The Startup Transformer breaker to A-5 will shut if the Auto Transfer switch is in the ON position.

(.5) c.

A direct result of the A-5 bus lockout is auto start of Diesel Generator "A".

(.5) d.

If Diesel Generator "A" is started or is running when the lockout occurs, its output breaker will automatically shut to supply A-5 with power.

(.5) e.

A lockout signal on bus A-5 will trip any supply breaker connected to it and prevent all power supply breakers to A-5 from shu-ting.

(.5) f.

The shutdown transformer is the only source available to A-5 under lockout conditions.

(.5) ANSHER a.

True (.5) b.

False (.5) c.

False (.5) d.

True (.5) e.

False (.5) f.

False (.5) (*** CATEGORY 06 CONTINUED ON NEXT PAGE***) .. . __. ... _

QUESTION 6.04 (4.00) . C ncerning the Recirculation Pump Control System: -. a.

What is the purpose of limiting recirc. pump speed to 25% when less than 20% total feedwater flow exists? (1.0) b.

What is the reason we limit recirc. pump speed to 65% when operating with less than three feed pumps and a low reactor water level alarm? (1.0) c.

What is the function of the recirc. runback reset pushbutton and indicating light on panel 9047 What precaution is associated with using this reset button? (2.0) ANSHER a.

At feedwater flow less than 20%, there is insufficient subcoolina of the downcomer water (.2) for high speed operations, which mav lead to recirculation oumo cavitation (.8).

b.

With less than three feedwater pumps and vessel level low, reactor power is limited by a recirculation run back to ensure that the steamina rate is within the makeuo caoacity of the feed system (1.0).

c.

The indicatina liaht informs the coerator that recirculation oumo soeed limiter is in control (.5).

The initiatina conditions must be corrected and the reset oushbutton utilized to break the seal in (.5).

The coerator must be sure to reduce the settino of the master (if in master manual) recirculation oumo controller until it is in control and the deviation between it and the individual controllers is zero (.5).

If this is not accomolished and the runback is reset.

the recirculation oumo will ramo un to match the master controller outout (.5).

QUESTION 6.05 (4.00) Describe the operating relationships between core spray injection valves (H0-25A and M0-258) and the presence of a core spray initiation signal.

(There are two parts to the answer.)

ANSWER 1.

Core spray injection valves get an open sianal when an initiation sianal is oresent (1.0) and reactor oressure falls below 400 osia (1.0).

(*** CATEGORY 06 CONTINUED ON NEXT PAGE***)

2.

The 25 valves (one per CS 1000) may be throttled or shut manually.

. even with an initiation sianal present (.75).

However, they will not respond to any auto sianals until the criainal sianal clears (.75).

LMay be re-ooened manually.) (.5).

- White light "on" near valve control indicates that a manual override of an initiation signal was employed.

. I QUESTION 6.06 (2.00) Regarding the Diesel Generators: a.

State the automatic starting signals (include setpoints) for DG's.

(1.4) b.

In the event of a LOCA stort, which of the diesel enaine protection functions are still capable of shutting down the engine? (.6) ANSWER a.

1.

Unit auxiliary (.2) and startuo transformer (.2) breakers to the l bus open and undervoltaae on the secondarv of the startuo transformer (.2), E ' 2.

High D.W. Press (.2) (2.5#) (.2), E 3.

Lo Lo vessel level (.2) -49" (.2) b.

Overspeed (.6) (*** CATEGORY 06 CONTINUED ON NEXT PAGE***) !

_.

-. - .

I QUESTION 6.07 (2.00) . Ctncerning the Rod Block Monitor: . a.

What is the function of the Rod Block Monitoring System? (1.0) b.

What are the minimum number of inputs required for each RBM to prevent an inoperative alarm? (1.0) ANSHER a.

The function of the Rod Block Monitor is to monitor the local neutron flux levels durina the withdrawal of a selected rod (.5) and cenerate trio sianals to actuate rod inhibit and annunciator circuits when the monitored neutron flux levels exceed oreset limits in relation to the reactor recirculation flow (.5).

b.

Must have. at a minimum. half the LPRMs sucolvino each RBM channel.

If less than half are available, the associated RBH channel is inoperable (1.0).

QUESTION 6.08 With the plant operating at 100% power, recirc in Master Manual, cn operator inadvertently decreases the MHC pressure setpoint by 5 psi. What wi,ll be the initial response and final status of the following due to this action? Briefly explain for initial response only. Refer to the attached MHC logic diagram if necessary. Assume load limit set at 100% and flow limit set at 105%. (3.00) c. TCV position, b.

BPV position, c.

Reactor power.

d.

Reactor pressure.

ANSWER o. TCV remain at 100% due to load limit (.50) b.

BPV open 5% due to flow limit (.50) c.

Power decreases due to lower pressure (.50) , d.

Pressure decreases due to BPV (.50) FINAL o. TCV at 100% position (.25) b.

BPV shut (.25) c. power lower (.25) d.

pressure slightly lower (.25) REFERENCE Hope Creek LP I.0.

3,4,10 FJC 286 Pilgrim LP MHC figure 1

- Nine Mile Point 2 EHC L.P.

objective EO-6 (***END OF CATEGORY 06***) I _ - - - - - _ - - - - - - - - - - - - _ - - - - -

--

-- - -- - - a

_ _________ - _ __________ _______________________ ______ .

t QUESTION 7.01 a.

In accordance with procedure 2.1.1 "Startup From Cold Shutdown," under what conditions shall secondary containment integrity be maintained? (1.50) b.

You are allowed a certain period of time after placing the reactor in the RUN mode before the primary containment atmosphere oxygen is required to be lese than 4% by weight. How long is this period?(0.50) ANSWER a.

Whenever the reactor is critical or when the Rx water temperature is above 212 degrees and the head vent closed. (1.5) 6.

24 hours. (0.5) REFERENCE Procedure 2.1.1 Rev. 44 pg. 4 FJC 8 QUESTION 7.02 a.

During an ATWS condition if RPV water level should become undetermi nabl e, EOP-08, "RPV POWER CONTROL BY LEVEL", directs the operator to maintain power above 8% by controlling injection to the RPV.

What is the significance of 8% and why must power be maintained above it? (2.00) b.

If, in the above situation, the operator is unable to maintain reactor power above 8%, he will be directed to depressurize to below the " MINIMUM ALTERNATE FLOODING PRESSURE". Explain the basis for this pressure.

(1.00) ANSWER a.

8% represents the " Reactor Flow Stagnation Power" above which water level is known to be high enough to permit natural circulation.

Controlling injection flow to maintain power above 8% thus automatically ensures that RPV water level is above TAF.

(2.00) b.

Lowest RPV pressure at which steam flow through the open SRV's will most ef f ectively remove heat.

(1.00) REFERENCE Pilgrim EOP-08 pg. 28, G.E.

EPG FJC 295 QUESTION 7.03 In reference to procedure 2.4.143 - Shutdown From Outside Control Room Due To Inhabitability of Control Rooms a.

WHERE are the Operators AND Operating Supervisor directed to go following assembly in the 23' 4kv switchgear area? (1.00) b.

WHAT is the " preferred" method to scram the reactor AND WHY is it the preferred method? (1.00) c.

WHEN should the reactor feedwater pumps be tripped? (1.00) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ -

. . i ANSWER a.

One operator - 23' 4kv switchgear area (.25) One operator - 37' 4kv switchgear area (.25) CRO to RPS MG set room (.25) Oparating Supervisor to Inst Rack 2205 or 2206 (.25) b.

Open the breaker to the APRM's (at the RPS power panels) (0.5) Preferred because tripping the RPS supply infers a loss of power and causes undesirable events such as early closure of MSIV's to take place.

(0.5) , c.

Two RFP's tripped when unit trips (0.5) One RFP tripped when level starts to increase (0.5) REFERENCE Boston Edison Procedure 2.4.143, Shutdown From Outside Control Room, pg.6-8, Rev.5 FJC 297 QUESTION 7.04 Entry into procedure EOP-03 section III.F3 " Steam Cooling" will occur only if all possible injection to the RPV becomes unavailable. If, during steam cooling, RPV pressure drops below 700 psig, the procedure directs the operator to emergency depressurize. What is significant about 700 psig and why is RPV depressurization required when pressure drops below 700 psig? (2.00) ANSWER The minimum RPV pressure producing steam flow through one SRV sufficient to limit peak clad temperature to 2200 deg. F has been determined to be 700 psig. The surge of steam flow resulting from rapid depressurization will again lower fuel temperatures providing additional time to establish a source of coolant imjection.

(2.00) REFERENCE Pilgrim EOP-03 pg. attachement H pg 3 of 4 FJC 301

QUESTION 7.05 (4,00) The Automatic Actions of Procedure 2.4.19, "RECIRC Pump M-G Set Scoop Tube Lock Up", describe the conditions that will cause such a lockout.

State those conditions and setpoints (if applicable).

ANSWER 1.

M-G Drive Motor 801. of rated voltage (.5)

Low Voltage (.5) (or 3328 volts) 2.

Loss of Speed (.5) Control Signal 3.

Low 011 Pressure (.5) 30 psig (with 6 see TD) (.5) 4.

High 011 Temperature (.5) 210*F (.5) 5.

Generator Lockout (.5)

Ref: 2.4.19. . _,_._.___ _ _ _ _ ____ _ _ _ _

QUESTION 7.06 (1000) - Concerning the P.A.H. panels; O List the one instrument that is not to be used while the recirculation pumps are running.

ANSHER Reactor water level fuel-zone indication (1.0) due to false reading from recirculation pump DP. - Ref. proc. 2.2.120 QUESTION 7.07 (4.00) E0P-06 (Primary Containment Control Level) establishes both high level and low level in the suppression pool as entry conditions.

a.

What are the high and low level setpoints for entry into E0P-06? b.

What is the basis for establishing high and low level limits? ANSWER Ref.

E0P-06 BHR EPGs App B pg. B.7-4 a.

Hi level +6", +140" (.5) Lo level-6", +128" (.5), b.

Discussion: Both high and low suppression pool water levels require entry to the Containment Control Guideline.

Low levels reduce the sucoression cool heat capacity (.5) and may result in exoosure of drvwell vents and oumo suction strainers (.5).

Hiah levels reduce the suonression chamber volume (.5), thereby, increasina the sucoression chamber oressure (.5), increasina static and dynamic

loads (.5) in the containment, and may submerge drvwell-to-suooression chamber vacuum breakers (.5).

l l l (*** CATEGORY 07 CONTINUED ON NEXT PAGE***) l l

QUESTION 7.08 (4.00) .. Considering the E0Ps, under what conditions is Boron injection required? . ANSWER Ref. E0P-02 Per E0P-02 under two conditions: 1.

If RX. Power is above 3% (or indeterminant) (1.0) and Sunoression Dool temperature is above 110*F (or will be above 110*F before the Reactor can be shutdown) (1.0).

Inject Boron.

' 2.

If RX. Power is below 3% but all rods are not inserted cast 04 (1.0) position and SDM is not verified for CR nattern existino (1.0).

Inject Boron.

t

!

I (***END OF CATEGORY 07***)

O o QUESTION 7.09 If both stack dilution fans are lost, the associated Off-Normal procedure requires that both standby gas treatment units be placed in service. Why is this necessary? (2.00) -ANSWER Both standby gas treatment units are placed in service at rated flow to provide dilution air to reduce the hydrogen concentration in the stack and maintain suitable exhaust velocities at the top of the stack.

(2.00) REFERENCE Procedure 2.4.45 Rev.

5, Pg. 2

SRO-4 . SECTION EIGHT QUESTION 8.01 (4.00) i-a.

State the manning requirements for minimum shift crew compliment as required by Technical Specification Table 6.2-1 (restated in Procedure 1.3.34, " Conduct of Operations") for the following conditions: 1). Operating (2.0) 2).

Cold Shutdown or Refueling (2.0) ANSWER a.

1).

Operating Lic. SR0 - 1* (.5) Lic. R0 - 2 (.5) NLNP0 - 2 (.5) STA - 1 (.5)

An SR0 licensed STA may serve dual roles as STA and one of the SR0's.

(Not required for credit.)

2).

Cold Shutdown or Refueling Lic SR0 - 1 (.5) Lic. R0 - 1 (.5) NLNPO - 1 (.5) STA - none required (.5) Higher licenses may take place of lower grade licenses / personnel.

(Not required for credit.)

QUESTION 8.02 (3.00) a.

Procedure 1.3.34, " Conduct of Operations", delineates conditions when the control room operator may leave the area of the controls without obtaining a qualified relief.

State those conditions.

(1.0) b.

1.3.34 also delineates conditions under which a formal relief turnover is not required for the control room operator. State those conditions and describe the turnover that muit be given under these circumstances.

(1.0) c.

1.3.34 defines " Controls". Briefly describe controls as defined by this procedure.

(1.0) (*** CATEGORY 08 CONTINUED ON NEXT PAGE***)

.-w- --- . _....,. > - - -. -,,,. .% ,.-. ._,..n- , -,, ,,.,. -,, -m,,.. ,, - -.. - -.

t

_.-,--,--..,4.- . -

AN$HER , a.

...In the event of an emeraencv. the coerator may bri absent from " the area monentarily in order to verify the receiot of an annunciator , alarm or to initiate corrective actions. orovided he remains within the confines of the control room".

(1.0) b.

"Short duration reliefs durina steady state ooerations do not reauire formal relief turnover in accordance with Section V.E. of this procedure; however. verbal discussion of olant status and off-normal conditions must be conducted.

(1.0) c.

Controls are defined as aooaratus and mechanism 3. the manioulation of which directiv affect the reactivity or oower level of the reactor.

NOTE: Underlined sections are the essence of the statements. Answers may be worded different.

Ensure points are made that are equivalent to the answer key.

QUESTION 8.03 (2.00) j ~ Define Primary Containment Integrity.

ANSWER , Ref: T.S. pg. 3 Primary Containment Intearity - Primary containment integrity means that the drywell and pressure suppression chamber are intact and all of the following conditions are satisfied: (.4) 1.

All manual containment isolation valves on lines connected to the reactor coolant system or containment which are not required to be open during accident conditions are closed.

(.4) 2.

At least one door in each airlock is closed and sealed.

(.4) 3.

All automatic containment isolation valves are operable or de-activated in the isolated position.

(.4) . 4.

All blind flanges and manuays are closed.

(.4) (*** CATEGORY 08 CONTINUED ON NEXT PAGE***)

QUESTION 8.04 (5.00) o Classify the following as an Unusual Event (U), Alert (A), a Site Emergency (S), a General Emergency (G), or not applicable (N/A).

(Use the . Attached PNPS 5.7.1.1).

1.

Failure of the reactor to shutdown on manual initiation within 15 minutes with MSIVs open.

(.5) 2.

Loss of physical control of the reactor building by operations or security personnel.

(.5) 3.

Loss of physical control of the main control room by BECo operators.

(.5) 4.

In HOT STANDBY, a fire occurs in the Admin. Bldg. which is still out of control 25 minutes after fire fighting efforts have begun.

(.5) 5.

The reactor in HOT STANDBY with coolant system leakage in excess of 50 gpm as indicated by drywell indicators.

(.5) 6.

During operation at 100% power, an MSIV isolation occurs on a high radiation signal. Drywell pressure increases above 10 psig and main stack rad monitor is at 3 rem and increasing.

(.5) 7.

A hurricane with sustained winds greater than 80 mph (onsite measurements).

(.5) 8.

Failure of an SRV to reset following a scram-related actuation.

(.5) 9.

Greater than 200 uci/ml of total iodine in the reactor water.

(.5) 10.

Loss of all control room annunciators for 15 minutes or more.

(.5) ANSWER 1.

S (.5) 2.

G (.5) 3.

S (.5) 4.

U (.5) S.

A (.5) 6.

G (.5) 7.

N/A (.5) < 8.

U (.5) 9.

A (.5) 10. A (.5) (*** CATEGORY 08 CONTINUED ON NEXT PAGE***) i ! I

QUESTION 8.04-(5.00) o.

Classify the following as an Unusual Event (U), Alert (A), a Site Emergency (S), a General Emergency (G), or not applicable (N/A).

(Use the . Attached PNPS 5.7.1.1).

1.

Failure of the reactor to shutdown on manual initiation within 15 minutes with MSIVs open.

(.5) 2.

Loss of physical control of the reactor building by operations or security personnel.

(.5) 3.

Loss of physical control of the main control room by BECo operators.

(.5) 4.

In HOT STANDBY, a fire occurs in the Admin. Bldg. which is still cut of control 25 minutes after fire fighting efforts have begun.

(.5) 5.

The reactor in HOT STANDBY with coolant system leakage in excess of 50 gpm as indicated by drywell indicators.

(.5) 6.

During operation at 100% power, an MSIV isolation occurs on a high radiation signal. Drywell pressure increases above 10 psig and main stack rad monitor is at 3 rem and increasing.

(.5) 7.

A hurricane with sustained winds greater than 80 mph (onsite measurements).

(.5) 8.

Failure of an SRV to reset following a scram-related actuation.

(.5) 9.

Greater than 200 uci/ml of total iodine in the reactor water.

(.5) 10.

Loss of all control room annunciators for 15 minutes or more.

(.5) ANSHER 1.

S (.5) 2.

G (.5) 3.

S (.5) 4.

U (.5) 5.

A (.5) 6.

G (.5) 7.

N/A (.5) 8.

U (.5) 9.

A (.5) 10. A (.5) , I !

(*** CATEGORY 08 CONTINUED ON NEXT PAGE***) .---.-. -. -.. .--_ - . . -. - - - - - - - --

. QUESTION 8.05 (2.00) o . Define core alteration, as defined in Technical Specifications.

ANSHER T.S. Pg. 4 Core alteration: The act of movina any comoonent in the reaion above the core succort olate (.4), below the unoer arid (.4) and within the shroud (.4).

Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration (.4).

Normal movement of in-core instrumentation is not defined as a core alteration (.4).

' QUESTION 8.06 (3.00) Hith fuel in the vessel, midway through the refueling outage with Rx.

cavity flooded, the "C" RHR pump is tagged out for electrical repairs to the motor windings.

The "B" RHR Loop has been tagged out for maintenance to repair the "B" heat exchanger bypass valve, H0-1001-16B (leaks excessively past seat).

The NP0 has just reported that the "A" RHR pump, currently in shutdown cooling, is leaking excessively (10 gpm) and recommends securing it. All other CSCS systems are operable.

The Hatch Engineer decides to secure the "A" pump for repair.

List all applicable LCOs and state any actions that are required.

- - - - ******************************************************

NOTE: USE THE ATTACHED SECTIONS OF THE TECHNICAL SPECIFICATIONS T0*

ANSHER.

FULLY REFERENCE ALL APPLICABLE SECTIONS OF THE T.S. YOU

USE TO DEVELOP YOUR ANSHER.

ANSHER Ref: Technical Specifications 3.5.F.3 - Allows CS, LPCI and Containment Cooling systems may be inop if no vessel drain potential work is being done.

(1.5) 3.5.F.4 - Refueling may continue (30 days) provided that one CS system or LPCI system is operable or 3.5.F.5 is met (CR removal).

(1.5) NOTE: Attach complete TS Section 3.5 I l (*** CATEGORY 08 CONTINUED ON NEXT PAGE***)

QUESTION 8.07 (6.00) o ^ a.

A core relcad has recently been completed. Due to a suspected problem with the Graphitar Seals, Control Rod Drive Mechanisms 18-11 - and 38-39 are being removed for maintenance.

Describe the conditions that must be satisfied (as stated by Tech Specs) prior to and during removal of the mechanisms.

(4.0) b.

Under what conditions does an inoperable hydraulic accumulator (HCU) constitute an inoperable control rod? (Reference 2.1.15; OPER-09.)

(1.0) c.

Under what conditions (short of repair) can an inoperable accumulator be cleared of its inoperable status? (Reference Tech Specs 3.3.D.)

(1.0) ANSWER a.

3.10 Core Alterations D.

Multiole Control Rod Removal Any number of control rods and/or control rod drive mechanisms may be removed from the reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control rod drive mechanisms are reinstalled and all control rods are fully inserted in the core.

(1.0) a.

The reactor mode switch is operable and locked in the Refuel position per Specification 3.10.A, except that the Refuel position "one rod out" interlock may be bypassed, as required, for those control rods and/or control rod drive mechanistas to be removed, after the fuel assemblies have been removed as specified below.

(.5) b.

The source range monitors (SRM) are operable per Specification 3.3.B.4.

(.5) c.

The Reactivity Margin requirements of Specification 3.3.A.1 are satisfied.

(.5) d.

All control rods in a 3x3 array centered on each of the control rods being removed are fully inserted and electrically or hydraulically disarmed, or have the surrounding four fuel assemblies removed from the core cell.

(.5) e.

All other control rods are fully inserted.

(.5) (*** CATEGORY 08 CONTINUED ON NEXT PAGE***)

s.

f.

The fcur fuel assemblies are removed from the core cell H.

surrounding from the core cell surrounding each control rod or control rod drive mechanism to be removed from the core PO*' and/or reactor vessel.

(.5) b.

With reactor pressure less than 950 psig, declare a control rod inoperable if its hydraulic accumulator is inoperable.

(1.0) , c.

If an inoperable accumulator has its control rod fully inserted and disarmed electrically, it shall not be considered to have an inoperable accumulator.

(1.0) (***END OF EXAMINATION ***) ... }}

v t e Inspection Report - Pilgrim - 1986045
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \| 25 \|
1986 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 10(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 17(0) 19(0) 21(0) 22(0) 24(0) 25(0) 27(0) 28(0) 29(0) 30(0) 31(0) 32(0) 33(0) 34(0) 35(0) 36(0) 37(0) 38(0) 39(0) 40(0) 41(0) 42(0) 43(0) 44(0) 45(0) 99(0) Security Operator Exams Emergency Planning Other

IR 05000293/1986045

Text

REGION I

=

"==

Navigation menu

Search