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Forwards Util Comments on Reactor Operator & Senior Reactor Operator Exams Administered on 880112.Ref Matl for Comments Also Encl
	ML17303A769
Person / Time
Site:	Palo Verde
Issue date:	01/19/1988
From:	Van Brunt E; ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:	; NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
	161-00751-EEVB, 161-751-EEVB, NUDOCS 8802010130
	Download: ML17303A769 (195)
	v • d • e

la Ai REGULAT i INFORMATION DISTRIBUTION YSTEM (RIDS)

~i ACCESSION NBR: 8802010130 DOC. DATE: 88/01/19 NOTARIZED: NO DOCKET 0 FACIL: STN-50-528 Palo Verde Nuclear Stationi Unit 1. Arizona Publi 05000528 STN-50-529 Palo Verde Nuclear Stationi Unit 2i Arizona Publi 05000529 STN-50-530 Palo Verde Nuclear Stations Unit 3i Arizona Publi 05000530 AUTH. NAME AUTHOR AFFILIATION VAN BRUNT'. E. Arizona Nuclear Power Prospect (formerly Arizona Public Serv REC'IP. NAME, RECIPIENT AFFILIATION Document Control Branch (Document Control Desk)

SUBJECT:

Forwards u'til comments on reactor operator & senior reactor ope'rator exams administered on 880112. Ref matl,for comments also encl.

DISTRIBUTION CODE: M003D COPIES RECEIVED: LTR ENCL SIZE:

TITLE: Operator Requalification Program NOTES: Standardized plant. 05000528 Standardized plant. 0500052'9 Standardized p lant. 05000530 I

REC P IENT COPIES REC IP IENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD5 LA 1 0 PD5 PD 1 LIC ITRAe E DAVIS' 1 1 lNTERNAL: 'ACRS 10 10 AEOD/DQA 1 1 AEOD/DSP/TPAB 1 NRR/DLPG/HFB 1 1 NRR/DLPG/OLB 1 01 1 1 RQN5 1 1 EXTERNAL: -'LPDR NRC PDR 1 1 NSIC NOTES:

TOTAL NUMBER OF COPIES REQUIRED: LTTR 24 ENCL 23

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Arizona Nuclear Power Project P.O. BOX 52034 ~ PHOENIX, ARIZONA 85072-2034 161-00751-EEVB/PGN Docket Numbers: STN 50-528/529/530 January'9, 1988 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 ATTN: Document Control Desk

Dear Sirs:

Sub)ect: Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2 and 3 Comments on Reactor Operator and Senior Reactor Operator Exams Administered on January 12, 1988 File: 88-001-762;I 88-056-026 ANPP has reviewed the Reactor Operator and Senior Reactor Operator exams admin-istered on January 12, 1988. In accordance with ES 201 in NUREG 1020, we are providing you with our comments in the attachments to this letter. Attachment A contains comments on the RO exam and Attachment B contains comments on the SRO exam. Reference material for the comments is also contained in the attachments.

Very truly yo s, E. E. Van Brunt, Jr.

Executive Vice President Project Director EEVB/PGN/cal Attachment cc: 0. M. De Michele A. C. Gehr J. B. Martin R. J. Pate G. W. Johnston (w/a)

J. R. Ball

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ATTACHMENT 'A COMMENTS ON RO EXAM PAGE SECTION I QUESTION 1.04 Part B answer is wrong. Should be selection 4 ~

See attached Reference from NLA06, Pages 7-3 thru 7-5.

QUESTION 1.12 Correct Reference is NLC55 12E016.

QUESTION 1.14 The way this question is worded, two options are correct. The Question should be deleted or both options accepted. CEA Insertion Limits are the subject of Tech. Spec. 3.1.3.6 (PDIL) which guarantees Option A and D'nd that adequate minimum'hutdown margin is available (which is very close to Option B). But Option B is worded to imply that "Reactivity available for shutdown is mazxmized". The subtle difference between "Maximized" an "Adequate Minimum" shutdown margin ma1ces Option B a correct answer. However, Option C refers to CEA misalignment effects and is not the subject of CEA Insertion Limits (Tech. Spec.

3.1.3.6). Instead, this, concept is the subject of Tech. Spec. 3.1.3.1 ~ See attached Tech. Specs.

QUESTION 1.16 The correct reference is NLC55 10E017 thru 20.

The Key is wrong. The correct Option is A. See attached reference.

I QUESTION 1.18, Correct reference is 12E05/12E06/12E017 in NLC55.

Part A Answer is wrong. The'core is always more responsive to reactivity additions if beta is smaller, unless the reactor is tripped and the stable 1/3 DPM SUR is created."' small negative reactivity insertion will not'reate a 1/3 DPM SUR.

QUESTION 1.21'orrect Reference is 11E07 of NLC55.

Both B and D are correct. See reference.

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ATTACHMENT 'A'OMMENTS ON RO EXAM PAGE 2 SECTION II QUESTION 2.02a It is not clear that the ques'tion is calling for a piece of equipment.. "HOW" denot'es a description of the method of performance this coupled with the fact that the "normal" power supply to PNB-D26 is the associated battery could lead the candidate to describe the evolution, of transferring from the regulator to the i'nverter. The inverter only, auto transfers in one direction, inverter to regulator'.

Ref. PN-15E-1 Section 1.3 PN-15E-2 Section 2.2 QUESTION 2.03 The answer appears to'e addressing both "HOW" and "WHAT" while'the question only asks for the WHAT WHAT "prevents thermal shock of the charging nozzle" HOW "Ensure adequate preheating of the charging flow" QUESTION 2.05 Part d. is the "best" answer, however the candidate could be confused since there is misleading information in d. the relief settings are actually 1250 psig to 1315 psig.

QUESTION 2.09b Tech. Spec..3.7.1.3 states 300 K is for essential feedwater.

QUESTION 2.10 The function of the L/D Backpressure valves is as r

stated in answer C1 answer 42 occurs because a component in the flowpath which contributes a it is headloss, it drop.

was not designed to attain that final'ressure Ref. Lesson Plan NLC21-00-XC-004 Page 11 Section G3 Another function of the letdown backpressure valves is to isolate at a high temperature out of the L/D Heat exchanger (148op) to pro'tect downstream components (IX, PRM, Boronometer).

Ref. CH Training Article Page CH-11

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ATTACHMENT 'A'OMMENTS ON RO EXAM PAGE 3 SECTION II Continued QUESTION 2.12c To ensure 75 F/Hr is not reached PVNGS imposes an administrative limit of 60 F/Hr. Some of the candidates may answer with 60 F/Hr since that is the number they use procedurally.

ref 410P-1ZZ10 QUESTION 2.13 Before the blackout, the valve position of a. and

c. will be OPEN per the given reference.

QUESTION 2.16 The stated condition of the question is that a LOW SERVICE AIR alarm is in alarm. This would lead the candidate to use the approach of a abnormal event with respect to the Instrument 9 Service Air System. (Note: The Tow SERVICE AIR alarm does not normally come in on a turbine trip/load rejection). The answer to part. a. is a design basis of the IA System and therefore not an abnormal event with respect to it. The candidate

'may answer more in line with events not planned to use excessive air:

T Resin fluffing Atomizing air on boilers Open relief in system High use of air powered tools by maintenance,.

Further two "types" of events are required, turbine trip and load rejection are the same "type" of event in'hat they actuate the, same valves in the secondary plant resulting in the massive air usages

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ATTACHMENT 'A COMMENTS ON RO EXAM'AGE 4 SECTION III QUESTION 3.02a CSAS should not be in answer because received until >8.5 psig Pcntmt it is not ref. SA-ESF-57. \

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a. MSIS should be part of answer because PS~G <919 ps la ~

ref. 'SA-ESF-57.

b. CIAS should be part of answer because Pres < 1837 psia.

ref. SA-ESF-57.

c ~ MSIS should be answer because PS~G < 919 psia.

ref. SA-ESF-57.

QUESTION 3.03 Some candidates may also include actions initiated by the BOP-ESFAS'eguencer when it receives a SIAS signal; the charging pumps block relay prevent auto-start for 40 seconds. This information should not be xegarded as incorrect.

trips, I

3.05 already to exclude emerg. mode QUESTION Comment so answer as it made is correct.

QUESTION 3.06a Some candidates may include combinations of signals for "SBCS demand", to include:

modulate and permissive quick open Z1 and Z2 guicl open Y1 and Y2 this signal is also interlocked with SBCS not connected (in test) or in emergency off. This information should not be counted wrong.

ref. SF-SBCS LP, Pg. 27, and SF-CEDMCS LP, Pg. 25.

a ~ "RC loops high" may be termed "high Tc.

b. High pressurizer pressure pre-trip will also cause CWP (2 of 4 chan.)

ref. SF-CEDMCS LP, Pg. 27.

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ATTACHMENT 'As COMMENTS ON RO EXAM PAGE 5 SECTXON X3I Continued

b. For Unit ¹1 ~onl , the following cycle 2 CPC condition will also cause a CWP output;

1) CEA deviation

2) Pre subgroup deviation penalty

3) Pre PLCEA insertion penalty

4) Pre out of sequence penalty It is recommended that this additional information not be required, but not counted wrong if supplied'.

ref. CPC LP NLC56-ZC-007 (new revision)

b. Answer should be "LPD pre-,trip", not "LPD".

ref. SF-CEDMCS LP, Pg. 27.

QUESTION 3.09 Question only asks "fail high, fail low, or stay the same". Candidates should not have to specify

, reason in answer for full credit.

'QUESTION 3.10 No comment on technical content, but point value for answer desired seems much too high (recommend

.5 per item for consistency with rest of section).

QUESTXON 3.12 Same comment as on 3.10 above; recommend .5 pt.

per item.

The question may be misleading because flow demand is a signal and not actually a control program.

Some candidates may describe operation or draw a diagram. These should be accepted 'as correct.

The refill demand (RTO Controller) should also be considered correct.

QUESTXON 3 '3 Comment already made on exam to clarify question; "error present for a while". This will sti,ll require candidate to assume "a while" is long enough for signal to decay off in order to arrive at correct answer.

QUESTION 3.14 Some candidates may answer as "high rate" insertion. This should not be counted, wrong.

ref. SF-RRS fig. SF-RRS-1

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ATTACHMENT 'A'OMMENTS ON RO EXAM PAGE 6 SECTION IV QUESTION 4.01 a ~ 5 15 minutes after a loss of forced flow before indications of natural circ. will be apparent.

Page 41 of 41EP-17Z01 .rev. 3.

b. May also say Tc stable or decreasing as per flow diagram on page 42 of 41EP-1ZZ01 rev. 3.

QUESTION 4.02 The questions ask for the systems that'an be used. Chemical and Volume Control System,(CH or CVCS) and Safety Injection System (SI) should be acceptable answers.

QUESTION 4.04 Any tolerance allowed?

a. < 467 psig

b. < 295oF QUESTION 4.05 a.&b. Question are stated in the same way but answer for

a. accounts for quarterly dose already received, where b. does not ~

The a. portion seems to be asking for the "rule" or limit, not the individuals allowed dose remaining.

REFERENCE:

NLA09: REACTOR THEORY TEXT: REV 003 CHAPTER 10: SECTION 10.2.4: PAGE 10-38 Thru 10-41 QUESTION:

If ASI is +.05 and core is at HFP, what is power difference between top and bottom?

How is ASI monitored?

How is ASI controlled?

What is ESI?

247

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NLC55: REACTOR THEORY REVIEW:

OBJECTIVE: 10-E018:

EXPLAIN WHY AXIAL POWER DENSITY OSCILLATIONS OCCUR AND WHY THE MAGNITUDE OF THESE OSCILLATIONS MUST BE LIMITED.

MAJOR POINTS:

~ Xenon inserts reactivity into local regions.

~ Consider partial CEA insertion at constant power.

~~ Power pushed down

~~ Xenon +Ap in bottom: Power 0 more

~~ Xenon -Ap in top: Power 4 more

~~ After 4-6 hours, switch

~~ 24-26 Hour cycle

~~ Divergent oscillation

~ Limit Local Axial Power Density

~ Normal power changes also initiate

~ Refer tq PVNGS plots May 1986

REFERENCE:

NLA09: REACTOR THEORY TEXT: REV g03 CHAPTER 10: SECTION 10.2.5: PAGE 10-42 QUESTION:

Can dilution initiate an oscillation' 248 W

NLC55: REACTOR THEORY REVIEW:

OBJECTIVE: 10-E019:

SPECIFY HOW AXIAL POWER DENSITY OSCILLATIONS ARE MONITORED AND CONTROLLED.

MAJOR POINTS:

Plot ASI vs Time using ESI as reference +.01 ASI Units When power's coming out of the core bottom, push zt down

~~ Use Rod Insertion

~~ Raise T-AVG by dilution

REFERENCE:

POWER OPERATION PROCEDURE QUESTION:

Why does changing T-AVG cause an ASI variation?

249

NLC55: REACTOR THEORY REVIEW:

OBJECTIVE: 10-E020:

SPECIFY HOV ASI CHANGES AS THE CORE POWER IS INCREASED FROM HZP TO HFP USING EITHER CEA MOTION OR DILUTION AS THE CONTROL MECHANISM.

MAJOR POINTS:

~ Dilution: .Power moves down ASI more +

~ CEA's : Power moves up ASI more-(Moving Out)

REFERENCE:

APPLICATION OF BASIC PRINCIPLES QUESTION:

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NLC55::REACTOR THEORY REVIEW:

OB JECTIVE:~ I~E>4:

DEFINE REACTOR PERIOD AND STARTUP RATE AND PERFORM CALCULATIONS USING P(t) =P 0 e~

~ ~

P(t) = P 10 T m 26 SUR

REFERENCE:

NLA09:REACTOR THEORY TEXT: REV 803 CHAPTER :SECTION :PAGE QUESTION-

h A

NLC55 -:REACTOR THEORY REVIEW:

B JECTIVE: /Z,-E 0/7:

%PLAIN WHAT IS 16XNT BY THE STAT12Q'NT "THE CORE BECOMES MORE RESPONSIVE AS IT AGES" AND IDENTIFY WHY THIS OCCURS.

MAJOR POINTS:

,ORE AGE VARIATIONS Exercises 13.1 and 13.2 show that "the core becomes more reactive as ages." Because the weighted average delayed neutron fraction (5) it decreases. The same insertion of reactivit creates a lar er SUR.

REPERENCE: NLA09:REACTOR THEORY TEXT: REV 803 CHAPTER :SECTION :PAGE QUESTION:

CP NLC55: REACTOR THEORY REVIEW:

OBJECTIVE: 11-E07:

PLOT POWER vs TIME FOR SEQUENTIAL POSITIVE REACTIVITY INSERTIONS INTO A SUBCRITICAL REACTOR. Qv MAJOR POINTS:

CR+

For each Ap:

CRg E

Time to stabiljze 0 When K ff =~

eff .99 starts taking'Long Time"

~ More generations

~ Larger effect'of delayed neutrons

REFERENCE:

REQUESTION: NLA09: REACTOR THEORY TEXT: REV 803 CHAPTER 11: SECTION 11.5 : PAGE 11"25 Thru 11-32 In the proce'ss of loading fuel you note that the countrate rises for approximately 60 seconds, before it stabilizes at a higher level after a fuel assembly is inserted. Briefly explain why this is, or is not an acceptable situation.

It is not acceptable. If the countrate requires 60 seconds to stabilize, it is an indication that the reactor is nearly critical, and Keff ff is much higher than that allowed by Tech Specs for fuel loading.

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.872 1000 .880 12.2 1.07 3.7

.880 1000 .888 13.1 1 ~ 07. 4;1

.888 1000 .896 14.1 1.08 4A

. 896 1000 . 904 15.3 1.08 4.8

.904 1000 .912 16.7 1.09 5.3

.912 1000 .920 18.4 1.10 5.9

.920 1000 .929 20.7 1.13 6.8

.929 1000 .938 23.7 1. 15 7.9

.938 1000 .947 27.7 1.17 9.4

.947 1000 .956 33.4 1.20 11.5 hk=p

.956 1000, .966 43.2 1.29 15. 0

.966 1000 .976 61.2 1.42 21.6

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.986 1000 .996 368 3.5 2Mi Change to 100 pcm

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.997 100 .998 737 1.50 269

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1.0 Introduction 1.1 Function The function of the 120UAC Class IE Power System is to distribute electrical power at 120VAC to all 120VAC Class IE loads. The 120VAC (Instrument AC) provides power for vital instrumentation and control loads in the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS).

The 120VAC Class IE Power System is designed to conform to the following bases:

a. The on-s'te x e 1 ect roc power system xs splat 1nto two independent load groups, each with its own off-site and on-site power supplies, buses, transformers, loads, and associated DC control power; either load group is capable of safely shutting down the unit.

b. The on-site power unit includes two redundant Class IE electric systems; the 120VAC system supplies power at 120VAC to plant safety-related systems.

No automatic transfers are provided between redundant: load groups.

d. The 120VAC Class IE Power System is designed to satisfy the single failure criterion For each protection and control channel, one 120VAC Class IE power source is provided Raceways are not shared by safety and non-safety related cables.

Special identification criteria apply for Class IE equipment cabling and raceways.

h. Safety related equipment is designed with the capacity to be tested periodically.

1.3 Basic S st:em Descri tion The 120VAC Class IE Power System is a single phase, 60 Hz ungrounded distribution system comprised of the folloying components:

a. Inverters

b. Distribution Bus

c. Voltage Regulator

d. Static Transfer Switch The system is designed with two different power supplies for reliabity:

a. Normal (Class IE 125VDC Control Center)

b. Abnormal (Class IE regulat:ed AC'power supply)

There is an automatic transfer bet:ween the buses, and no connections exist between independent: supplies. The system provides power for the four protective channels of the Reactor Protection System. Each channel is electrically and physically isolat:ed from the others.

The reason t:hat this voltage must be highly regulated is because it supplies vital and sensitive inst:rumentation which is highly susceptible to variations in line voltages. Such variations could result in possible protective actions taking place as a result of spurious conditions rather than actual abnormal conditions.

PN-15E-1

2.0 Com onent Descri tion.

The 120VAC Class IE Power System (see Figure PN-2) is made up of the following major components:

a. Inverters b . Transfer Switch

c. Distribution Bus

d. Line Voltage Regulator.

2.1 Inverters Static inverters supply regulated 120VAC power to vital instruments. The input to the inverter is a filtered DC which provides transient-free power. DC to AC conversion is accomplished by silicon controlled rectifiers (inverter) which produce a single phase, highly regulated (2 1/) AC signal.

2.2 Static Transfer Switch The Static Transfer Switch is an electronic solid state assembly which is used to automatically transfer from the inverter to the voltage regulator without interruption. Manually, the load can be switched to either source of power by pushing either the FORWARD (to the inverter) or Reverse (to the voltage Regulator) control buttons on the Static Switch Control Panel.

2.3 Distribution Bus There are four Class IE 120VAC distribution buses. They are two-wire, ungrounded distribution panels with thermal magnetic trip circuit breakers for main feeder, and fuse switches for branch circuits. The fuses are quick-operating to ensure that faults will be isolated before the inverter field collapses. Each panel, which supplies one channel of the RPS and ESFAS control systems, is physically separated and electrically isolated for reliability. If one fails, the other three can safely monitor vital system parameters and shutdown the reactor if necessary.

2.4 Line Volta e Re ulator Each 120VAC ungrounded distribution panel is supplied with a 480,3$

120VAC, 1$ voltage regulator as an alternate power source in case of inverter failure or scheduled maintenance. These are ferro-resonant, constant voltage transformers capable of supplying a highly regulated

(+ 2X), single phase, 60Hz signal.

PN-15E-2

CONTROLLED BY USER PLANT SYSTEMS CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with an indicated level of at least 25 feet (300,000 gallons).

APPLICABILITY: MODES 1, 2, 3,¹ and 4.*¹ ACTION:

Mith the condensate storage tank inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a. Restore the CST to OPERABLE status or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOW within the following 6 hours, or b, Demonstrate the OPERABILITY of the reactor makeup water tank as a backup" supply to the auxiliary feedwater pumps and restore the condensate storage tank to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOMN with a OPERABLE shutdown cooling loop in operation within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.7.1.3.1 The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the level (contained water volume) is within its limits when the tank is the supply source for the auxiliary feedwater pumps.

4.7.1.3.2 The reactor makeup water tank shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the reactor makeup water tank is the supply source for the auxiliary feedwater pumps by verifying:

a. That the reactor makeup water tank supply line to the auxiliary feed system isolation valve is open, and

b. That the reactor makeup water tank contains a water level of at least 26 feet (300,000 gallons).

"Until the steam generators are no longer required for heat removed.

Not applicable when cooldown is in progress.

PALO VERDE " UNIT 1 3/4 7-6 AMENDMENT NO. 20

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REV MBER A PAGE 11 LES uMBER NLC21-00-XC-004 COURSE TITLE NSSS LESSON TITLE CVCS STUDENT OBJECTIVE(S) SEQUENCE METHODS & ACTIVITIES

6. Nuclear Cooling Water (Shell side)

a. Design Pressure = ~

150 PSIG

b. Design Temperature ~ 250 'P

c. Design Plow 1500 GPM

d. Normal Plow ~ 680 GPM

e. Normal Inlet Temp. 105 'F

f. Normal Outlet Temp. = 128 'F G. Back Pressure Control Valve (PV-201 P and Q)

l. One inch, Pneumatic Operated, Globe Valves

2. Fail closed, modulating 2. Explain: Loss-Of-Air, Valves fail closed

3. Maintains sufficient back pressure to prevent Letdown Flashing in the letdown heat exchanger

4. Three position selector switch on B-03

a. PV-201P

b. PV-201Q

'c, Bggh

exchanger has letdown flow on the tube side and nuclear cooling water flow on the shell side. The nuclear cooling water flow is automatically regulated to maintain the letdown heat exchanger outlet temperature at approximately 120 F. The letdown heat exchanger tubes are constructed of stainless steel for corrosion prevention, and the shell is carbon steel.

2.1.7.1 Letdown Heat Exchan er Data Quantity 1 Designation CHN-E02 Type Shell-and-Tube Design Pressure: Tube Side 650 psig Shell Side 150 psig Design Temperature: Tube Side 550 F Shell Side 250 F Design Flow:..Tube Side 150gpm Shell Side 1500 gpm Normal Flow: Tube Side 72 gpm Shell Side 580 gpm materials: Tube Side Stainless Steel Shell Side Carbon Steel Hanufacturer Reco 2.1.8 Letdown Back ressure Valves (CH-PV-201P CH-PV-201 These air diagragm open, spring-closed globe valves control letdown backpressure at a nominal 460 psig, with high and low pressure alarms at 500 psig and 420 psig. This prevents boiling in the letdown heat exchanger, and the final pressure reduction:to a nominal 40-95 psig protects downstream components. Additionally, a high high temperature condition of 148 F, as measured at the outlet of the letdown heat exchanger, alarms and shuts the letdown backpressure valves.

2.1.9 Purification Filters Two parallel purification filters (CHN-F36 and F19) remove particulate materials from the letdown flow. Both filters are designed to remove 98 percent (by weight) of all particles 2 microns or larger. One filter is normally in service and the other is isolated. Both filters are removable cartridge type and are designed for efficient removal and replacement.

2.1.9.1 Purification Filter Data Filters CHN-F19, F36 Quantity 2 Type Removable Cartridge Design Flow 150 gpm Normal Flow 72 gpm Design Pressure 200 psig Normal Pressure 60 psig Design Temperature 250 F Normal Temperature 120 F Pressure Drop (at 30 gpm): Clean 3 psid Loaded 10 psid Efficiency 98/o CH-11 Rev. 02

PALO VERDE NUCLEAR GENERATING PROCEDURE NO.

'STATION MANUAL 410P-1ZZ10 HOT STANDBY TO COLD SHUTDOWN REVISION MODE 3 TO I'jODE 5 Page 15 of 39 NOTE Record data on 41ST-lRC01, each time the pressurizer spray valves or auxiliary spray valves are cycled.

4.3.20 SELECT the pressurizer spray valve:

RC-100E OR RC-100F 4.3.21 Notify Radiation Protection if the atmospheric dump valves are going to be used for cooldown to ensure proper monitoring is performed.

CAUTION A HIGH RCS COOLDOWN RATE MAY EXCEED THE CAPABILITY OF THE CHEMICAL AND VOLUME CONTROL SYSTEM. MONITOR PRESSURIZER LEVEL AND VOLUME CONTROL TANK LEVEL CLOSELY.

CAUTION MAINTAIN RCS TEMPERATURE, PRESSURE, AND COOLDOWN RATE WITHIN THE LIMITS OF T.S.

3.4.8.1. MAlNTAIN PRESSURIZER COOLDOWN RATE WITHIN THE LIMITS OF T.S. 3.4.8.2 (LCTS I'/030124).

4.3.22 Establish a cooldown rate of approximately 60'F per hour using the SBCS or the main steam atmospheric dump valves.

PV2i6 000 AEV.8/82 584 A

II 4.3 Abnormal Condition The Instrument and Service Air System is designed to meet the maximum anticipated air demands without compromising proper operation of associated equipments The maximum demand for instrument air is expected to occur during a load rejection or turbine trip from full-load. It is estimated that 75 percent of all air-operated valves in the Condensate, Feedwater,,and Steam.

Systems will be actuated from full-closed to full-open (or vice versa) in about 60 seconds. The compressed air inventory in the receivers meets this requirement. The standby compressors will automatically start as air pressure decreases to 110 and 105 psig to restore normal system pressure. In the unlikely event syst: em pressure decreases to 90 psig, the priority control valve (PV-1) will automatically close to isolate plant service air loads. The maximum demand for plant service air is expected to occur during shutdown.

(The demand for instrument air should be minimal at this time.)

An orxfice

\

is provided in the instrument air line inside containment to limit flow to 50 cfm in the unlikely event of a branch line rupture.

Nitrogen gas automatically aligns itself to the Service/Instrument air header. At 85 psig and decreasing, PV-52 opens and supplys N 2 gas to the header During decreasing pressure from 125 psig, the following occurs: (see Figure IA-11)

a. 110 psig decreasing, first standby compressor starts

. b. 105 psig decreasing, second standby compressor starts (105 psig increasing, N2 gas PV-52 closes) c 95 psig decreasing, Instrument, Air Header Pressure Low

d. 90 psig decreasing, PV-1 closes (reset 90 psig manual) 85 psig decreasing, PV-52 opens (N supply) Low Service Air Pr~ssure 2

alarm On a decreasing service air pressure, the Nuclear Operator should verify Chat the air compressors are operating normally and determine the cause for excessive air use on the system. On a high delta P across filters, the Operator should shift to the standby train. If there is a problem with the instrument air dryers, the Operator should shift to the standby train.

Finally, on any failure of the air compressor, the Operator should insure the cause has been determined prior to resetting the air compressor.

IA"12 Rev. 2

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TABLE SA-ESF-IV ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM SETPOINTS ANO MARGINS TO ACTUATION Nomina I Norma I Nomina I Ful I Operation Actuation Margin to Actuation Si nal Power Ran e Set oint ~ac ua ion SIAS &. CIAS Low Pressurizer Pressure 2250 psia 2100-2350 psia 1837 psia(1) 263 psi High Containment Pressure 0 psig 0 psig 3.0 psig 30 psi CSAS High-High Containment Pressure 0 psig 0 psfg 8.5 psig 8.8 psl RAS Lou Refueling Water Tank Level 70-95) 7.4g 62.6g MSIS Low Steam Generator Pressure 1070 psia 1070-1170 psia 919 psia(1) 151 ps i High Conta inment Pressure 30O 0 psig 0 psig psig 3.0 psi High Steam Generator Level 55/o 30-74% 17$

EFAS Low Steam Generator Level and 82$ 72-90%%uo 25. 8%%d 46.2g Steam Generator Differential Pressure(2) 0 psid 0 psid 185 psid 185 psi

( 1) Setpoint can be manually decreased as pressure is reduced and Is automatically increased as pressure increases.

(2) This is a calculated, not sensed, variable.

SA-ESF-57 Rev. (1)

Cl

2) The IRS adds sodium hydrozide (Na01k) to the spray water to .enhance the removal by absorption of iodine from the containment atmosphere. This minimizes the iodine release which would othexwise occur if containment integrity were to be violated following an accident.

4. Recirculation Actuation Signal (RAS) (Figure SA-ESF-9)

a. A RAS will be generated on a low Refueling Mater Storage Tank (RWST) level of 7.4% following a SIAS.

b. Pretrip alarms are provided at 11% RMST level.

c .,

~ The'AS is provided to initiate recirculation of borated water from the containment sump for long term post-accident Emergency Core Cooling using the Safety Injection System.

1) Following SIAS actuation, the RWST serves as the source of borated water fox core injection and containment spray.

2) Mhen the RWST water inventory is depleted, a RAS will cause the ESF systems to automatically shift their suctions to the containment sump.

3) The sump, being full of water previously pumped from the RMST for"core injection and containment spray, can now serve as the borated water supply for long term core cooling.

5. Hain Steam Isolation Signal (HSIS) (Figure SA-ESF-10)

a. A HSIS will be generated by the ESFAS if any of the following conditions occur:

1) SG No. 1 pressure less than (variable setpoint-919 psia nominal)

2) SG No. 1 level greater than 91/ (NR)

3) SG No. 2 pressure less than (variable setpoint-919 psia nominal)

4) SG No. 2 level greater than 91/ (NR)

5) Containment pressure greater than (3.0) psig.

A13..inputs are s ared with the correspon ing reactor trips.

Pretrip alarms are provided for all inputs. Low SG Press Pre-Trip 960 psia. Low SG Level Pre-Trip 47.1/. High Containment Press Pre-Trip 2.5 psig.

The HSIS will isolate the Hain Steam, Hain Feedwater, sample and blowdown lines on both steam generators regardless of which SG (if either) was responsible for the actuation.

In the event of a HSLB, the HSIS will limit the energy release to containment and prevent excessive containment pressures. By isolating both SIFAH Generator, this allowing only one to release its energy. It also prevents complete blowdown of both SG's and the result.ing loss of heat sink for the reactor core.

SA-ESF-8 Rev. (1)

1 6.7 SEQUENCER 6.7.1 HODE 1 SAFETY INJECTION ACTUATION SIGNAL CONTAINMENT SPRAY ACTUATION SIGNAL Sequence Starts 0'iesel Time Seconds 0' Generators Start High Pressure Safety Injection Pumps A & B Charging Pump Permissive Actuation Relay Blocked Low Pressure Safety Injection Pumps A & B Diesel Generator Buildings Essential Exhaust Fans A &, B (1st of two necessary signals)

Control Room Essential Ventilation A

Jt

~

~~

Fuel Building Essential Ventilation 10 Auxiliary Feed Pump (Essential B) 15 Containment Spray Pumps A & B 20 Essential Cooling Mater Pumps A & B 25 Essential Spray Ponds Pumps A & B and DG Bldg Essential Exhaust Fans A & B Finally Start 30 Essential Chillers A & B 40 Charging Pump permissive Actuation Relay Fuel Building Essential Ventilation

1. Essential AFUs start (by sequencer) 2., Opens Dampers (by FBEVAS relays)

a. HFA-M06

b. HFB-M06 Closes Dampers (by FBEVAS relays)

a. HFA-H05

b. HFB-H05

4. M06 Dampers have priority over H05 Dampers if both SIAS and FBEVAS exist at the same time (by SIAS relays).

SA-BOP-24 Rev. (1) f

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l REV NUMBER: 01 COURSE TITLE: I Review PAGE.

ZESSON NUMBER: NLC56-01-ZC-003-01 ZESSON TITLE: Steam B ass Control S stem STUDENT OBJECTIVE(S) SEQUENCE METHODS & ACTIVITIES

3. AWP to CEDMCS is blocked if:

a. Emergency off.

b. "SBCS CONNECTED" ¹1 or ¹2 are not set.

B. Reactor Power Cutback Demand (RPC) signal sent to RPCB system.

E08 State the function of 1. Generated to match reactor SBCS signals sent to power to secondary power other systems. after large load reject.

2. Generated by Quick open Discussed already in Quick circuitry. Open logic; necessary..

use TP-4 if

3. RPC signal to RPCB blocked if.

a. Emergency off.

b. "SBCS CONNECTED" ¹1 or ¹2 are not set.

C. Automatic Motion Inhibit (AMI) signal sent to CEDMCS.

EO8 State the function of 1. Generated to allow rapid SBCS outputs sent to reloading of turbine other systems. following turbine trip or load reject of large magnitude.

REV NUMBER: 1 COURSE TITLE: I view PAGE: 25 LESSON NUMBER:

NLC56-01-ZC-11-01 LESSON TITLE: CEDMCS STUDENT OBJECTIVE(S) SEQUENCE METHODS 5 ACTIVITIES

c. Generated under following TP-16 conditions:

(1) Any of the Tc's that input to RRS 575oF ave 6 F greater than Tref.

(3) SBCS valid signal TP-17 to open valves:

(a), Quick Open Z1 and Z2 (b) Quick Open Z1 and Z2 (c) Modulation and Permissive

2. AMI

a. Prohibits any reg. CEA motion in AS.

I REV I NUMBER: 1 COURSE TITLE: I view PAGE: 27 LESSON NUMBER: NLC56-01-ZC-11-01 LESSON TITLE: CEDMCS STUDENT OBJECTIVE(S) SEQUENCE METHODS & ACTIVITIES

3. CWP a.= Inhibits any CEA withdrawal in any mode.

b. Two out of four logic from PPS.

(1) CPC CWP .output (2.

of 4).

(a) High LPD pretrip.

(b) Low DNBR pretrip.

(c) RPCB Flag set.

(2) Pressurizer 2409 PSIA pressure high pretrip.

4. =.

Upper Electrical Limit (UEL)..

a. Prevents further individual moti'on or subgroup motion when the first CEA in that group reaches this extreme limit.

li t

REVIS NUMBER: 02 COURSE TITLE: I & Review PAGE: 49 LESSON NUMBER: NLC56-01-ZC-007-02 LESSON TITLE: Core Protection Calculator STUDENT OBJECTIVE(S) SEQUENCE METHODS 5 ACTIVITIES

4. Associated safety channel N.I. trouble (NUC INST CH INOP annunciator on B04):

a. High voltage < + 800 VDC

b. Any of the 3 drawer calibrate switches not in OP~RAT

c. Any of 5 drawer trip test switches not in circuit card

'OPP'.

Any drawer removed or in the wrong position.

D. CPC channel CNP output trips Point out new DNBR pre-trip setpoint.

1. DNBR pre trip:

a. Quality margin in limiting noae < 0.05

b. DNBR < 1.39

2. LPD are-trip

a. LPD > 20 kw/ft

3. RPC (RZ Power Cutback) flag set

a. RPC flag received from CEACs; they sense group 5 only or groups 4 and 5 falling in

b. RPC flag set for 20 seconas

0 I

8 REVIS NUMBER: 02 COURSE TITLE: I 6 C Review PAGE: 50 LESSON NUMBER: NZC56-01-ZC-007-02 LESSON TITLE: Core Protection Calculator STUDENT OBJECTIVE(S) SEQUENCE METHODS & ACTIVITIES

4. CEAC CEA DEVIATION (CEADEV) Note that the four (4) new

a. Provides CWP or LPD penalty iffactor the DNBR causes of CWP listed here are part of the program to received from an operable de-sensitize the CEA CEAC is greater than 0 related penalty factors.

(penalty of > 1.0)

5. CP C pre-penalty (IPCMI) a ~ Provides CWP prior to receiving a CPC generated penalty factor CWP occurs at > 5 inches deviation between subgroups within a group c ~ CWP occurs at < 30 inches withdrawn on a PLCEA

d. CWP occurs when a Reg group is < 7.5 inches above the next-higher numbered Reg. group

1.0 INTRODUCTION

The Reactor Regulating System"-(RRS) is a non-safety related plant control system. It is designed as an automatic control system to help the primary plant respond to turbine load changes faster and more efficiently than a control room operator. It also allows the operator to attend to other plant conditions which might require his attention during transients. It is, nevertheless, necessary for the operator to be familiar with the actions initiated by the RRS to allow him to respond properly in conjunction with all other syst: em responses during transients.

In accordance with procedure, the operator must maintain the plant T avg within + 2 F of program temperature (T f) during power operation.

ref Even if RRS is not used to automatically control CEA's due to plant information not being available from RRS, the operator would have a difficult time trying to comply with the procedural temperature band constraints.

1.1 Functions

a. Program reactor coolant average temperature as a function of turbine load so as to satisfy the 100 percent power main steam pressure requirements.

b. Provide signals to the CEDMCS to reposition the regulating CEAs to maintain the reactor coolant average temperature within a deadband of the programmed reference temperature.

c. Provide automatic control of the reactor coolant average temperature (TA<G) and reactor power, in conjunction with the operation of the other control systems, for each of the following plant conditions during plant operation between 15 percent and 100 percent power:

Load rejection of any magnitude.

Loss of one out of two operating feedwater pumps.

Ten percent step change in NSSS load. The final load after a step increase will not exceed 100 percent of rated NSSS power.

4) Five percent per minute changes in NSSS load. The final load after an increase will not exceed 100 percent, of rated NSSS power.

5) Maintain steady-state conditions.

d. Provide indication to the operator to allow verification of system operation and to facilitate manual control of the reactor coolant average temperature.

e. Provide reactor coolant average temperature, reactor power, CEA withdrawal demand, and turbine load signals to the SBCS.

f. Provide a reactor coolant average temperature signal to the pressurizer level control syst: em (PLCS) and Feed Mater Control System (FWCS).

(NONE)

SF-RRS-1 Rev. 0

FOB 'I FORMATlON OiXLY

'ALO VERDE NUCLEAR GENERATING PROCEDURE NO. APPEi%)W E STATION MANUAL 41EP-IZZOl Page 1 ef S REV(SION E(KRGENCY OPERATIONS Page 41 of 179 NATt RAL CIRCULATION VERIFICATIOiN ~

OBJECTIVE This Appendix is used to verity that natural circulation flov is establisned, and is maintaining the plant in a stable configuration.

This Appendix also addresses void reduction/elimination.

CAUTION 7

IF RCS SUBCOQLED .'MARGIN DECRFASES TO LESS TH>b 28 F, INITIATE SIAS.

ihOTE Natural circulation flow vill normally be established by the performance or critical safety zunctions. It will be approximately 5"15 minutes after loss of forced

flow'before indications of natural circulation >>ill be apparent-. Any operator inauced change vill reauire approximately 10 minutes for the effect. to become apparent, due co increased loop transport time. Once natural circulation zlov is establisned, loop c should

~

be approximately equal to SG saturation cemperature.

This provides the operator with a good indication of RCS and SG thermal coupling.

ROTE Natural circulation flow requires at, least 1/3 steam generator tube coverage. iiaintaining steam generator wide range level above coverage.

0,'ill ensure at least 1/3 cube

],'.P VERIFY NATURAL CIRCULATION FLOV IS ESTABLISHED BY PERFORMING FLOV CHART FIGURE l.

Ij the flowchart confirms natural ci.rculation flow. monitor flow indications and adeauate core cooling by maintaining the data sheer., page 7 of 7, updated every 15 to 30 minutes.

PVti6400 REV &f62 (9bS>l

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FOR I(. FORMATION ONLY KHKRGENC4'PERATZGNS ( APoEHOZX E Vt lPAGE nF 8 Re~iSIOn 3 P~ge ~2 Of 17a

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HEAT REMOVAL ECOVERZH RCS HEAT REHQVAI GQ TQ VKS CUNCTZONAI RKCOVKRV PROCEOURE RCS l I tl RCS )28-F SVBCOOt.EO SUBCOQLEO MARGIN AT LEAST 1 SG

)28"F $ 3%X CWR)

H/'F W YES THRQTTl K St Fl OH Ta MAINTAIN IHvENTQRY RV AHO SLRC QL!HG OUTLET NQ PLENUM FULL YES RCS HO RKS'TQRK FQRCEO FORCKO CONT!LIVE ZQ nQNLTGR VQtorNG CIRCULATZON CIRCULA IOH NATURAL CIRCUt ATION SUSPEC.ED ~

B'Y PER IOOI CALLY 5'TEP + 4 RESTQREO USIHG >15 FIGURE

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YES VES coHFIRn Rcs vo !DING pKR CONTINUE RECOVERY PROCEDURE STEP 3, 0 CONTROL VOZOING WITw ~QoCEO CIRCUt ATION PKR STEP 4 0

t CONTROLLED Br USER REACTIVITY CONTROL SYSTEMS 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 BOR'ATION"CONTROL SHUTDOWN MARGIN - ALL CEAs" FULLY"INSERTED LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.0X delta k/k.

APPLICABILITY: MODES 3, 4", and 5" with all full-length CEAs fully inserted.

ACTION:

With the SHUTDOWN MARGIN less than 1.0X delta k/k, immediately initiate and continue boration at greater than or equal to 26 gpm to reactor coolant system of a solution containing greater than or equal to 4000 ppm boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0X delta k/k at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of at least the following factors:

1. Reactor Coolant System boron concentration,

2. CEA position,

3. Reactor Coolant System average temperature, Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration.

4.1.1.1.2 The overall core reactivity. balance shall be compared to predicted values to demonstrate agreement within + 1.0X delta k/k at least once per 31 Effective Full Power Days (EFPD}. This comparison shall consider at least those factors stated in Specification 4. 1.1.1.1, above. The predicted reactivity values shall be adjusted l,'normalized} to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD after each fuel loading.

See Special Test Exception 3. 10. 9.

PALO VERDE " UNIT 1 AMENDMENT NO. 23

0, CONTROLLED BY USER REACTIVITY..CONTROL.5YSTEHS..

SHUTDOWN MARGIN, -..KN1. -.. ANY CEA, WITHDRAWN LIMITING CONDITION FOR OPERATION 3.1.1.2

~ h ~ ~

a~ The SHUTDOWN MARGIN shall be greater than or equal to that shown in Figure 3.1-1A, and

b. For T ld less than or equal to 500 F, KN 1 shall be less than 0.99.

APPLICABILITY: MODES 1, 2~, 3", 4"; and 5~ with any full-length CEA fully or par ti ally wi thdd ann.

ACTION:

a. With the SHUTDOWN MARGIN less than that in Figure 3.1-1A, immediately initiate and continue boration at greater than or equal to 26 gpm t(c>he reactor coolant system of a solution, containing greater than "or-equa'l to'4000 ppm boron or equivalent until th'e required SHUTDOWN MARGIN is restored, and

b. With T l less than or equal to 500 F and K greater than or equal to 0.9$ , Immediately vary'CEA positions and/oI initiate and continue boration at greater than or equal to 26 gpm to the reactor coolant system of a solution containing greater than or equal to 4000 ppm boron or equivalent until.'the required KN 1 is restored.

~ ~

I SURVEILLANCE REOUIREMENTS C

4.1.1.2.1 With any full-length CEA fully or partially withdrawn, the SHUTDOWN MARGIN shall be determined to be greater than or equal to that in Figure 3.1.1A:

a ~ Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable CEA(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable as a result of excessive friction or mechanical interference or known to be untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).

See Special Test Exceptions 3. 10. 1 and 3. 10 9 PALO VERDE - UNIT 1 3/4 1-2 AMENDMENT NO. 23

'I l

CONTROLLED BY USER

-- (509,6.5) 6' t I t 4

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FlGURE 3.l- lA SHUTDOWN MARG'lN VERSUS COLD LEG TEMPERATURE PALO VERDE

" UNIT l 3/4 1"2a AMENDMENT NO. 24

1 CGNTRGLLED BY USER REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREHENTS'Continued}

b. When in MODE 1 or MODE 2 with k ff greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that CEA group withdrawal is within'he Transient Insertion Limits of Specification 3. 1.3.6.

C. When in MODE 2 with k ff less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior'o achieving reactor criticality by verifying that the predicted critical. CEA position is within the limits of Specification 3.1. 3. 6.

d. Prior to initial operation above 5X RATED THERMAL POWER after each fuel loadi'ng, by consideration of the factors of e. below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.

e. When in MODE 3, 4, or 5, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of at l'east the following factors:

1. Reactor Coolant System boron concentration,

2. CEA position,

3. Reactor Coolant System average temperature,

4. Fuel burnup based on gross thermal energy generation, 5; Xenon concentration, and

6. Samarium concentration.

4.1.1.2.2 When in MODE 3, 4, or 5, with any full-length CEA fully or partially withdrawn, and T o) less than or equal to 500'F, KN shall be determined to be less than 0.9$ M least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consi3eration of at least the following factors:

l. Reactor Coolant System boron concentration,

2. CEA position,

3. Reactor Coolant System average temperature,

4. Fuel burnup based on gross thermal energy generation,

5. Xenon concentration, and

6. Samarium concentration t

4.1.1.2.3 The overall core reactivity balance shall be compared to predicted values to demonstr ate agreement wi thi n + l. OX delta k/k at l east once per 31 Effective Full Power Days (EFPD}. This 'compar)son shall consider at least those factors stated in Specification 4.1.1.2.1.e or 4.1.1-2.2. The predicted reactivity values shall be adjusted (normalized} to correspond to the actual core conditions prior to exceeding a fuel 'burnup of 60 EFPD after each fuel loading.

PALO VERDE " UNIT 1 3/4 1-3 'AHENOHENT NO. 23

CONTROLS ED BY USER

'I

. 3/4. 9 REFUELING OPERATIONS 3/4. 9. 1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the. following reactivity conditions is met:

~ t

a. Either a Keff of 0.95 or less, or

b. A boron concentration of greater than or equal to 2150 ppm.

APPLICABILITY: MODE 6".

ACTION:

With the require'ments of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration a4 greater than or equal to 26 gpm of a solution containing > 4000 ppm boron or its equivalent until K

ff i s reduced to 1 ess than or equal to 0 95 or the boron concentrati on i s

~

restored to greater than or equal to 2150 ppm, whichever is the more restrictive.

SURYEILLAHCE RE UIREMENTS 4.9.1.1 The. more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full-length CEA in excess of 3 feet from its fully inserted position within the reactor pressure vessel.

4.9.3..2 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours.

The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tensioned or with the head removed.

PALO VERDE - UNIT 1 3/4 9-1 AMEHDt1EHT HO. 21

H FOR INFORMATION ONl Y PROGEOURE PALO YERDE NUCLEAR GENERATING NO. APPED)IX P STATION MANUAL 41EP"1ZZ01 Page 1 of 14 REVISION EMERGENCY OPERATIONS Page 92 of 179 RESETTING SAFETY INJECTION ACTUATION SIGNAL (SIAS)

AihD P COKTAD~KNT ISOLATION ACTUATION SIGNAL CIAS CAUTION RESET CRITERIA ."HJST BE >KT BEFORE AT r.~.'PTING TO RESET SIAS/CIAS.

1.0 VERIFY PLANT CONDITIONS i'KET THE FOLLOWING CRITERIA BEFORE RESETTING SIAS/CIAS:

1.1 SISS reset criteria

'.1.1 RCS is at least ZS subcooled.

1.1.2 Pressurizer level is greater than 33",~ and under control.

1. 1.3 At least one S/G is removing heat from the RCS, and indicated level is greater than 35",o 4'.R. and recovering.

1;1.4 Safety injection flow is not required to maintain PZR level.

NOTE IF the SIAS signal can't be reset due to the actual pressures being less than the actuation setpoint, at the CRS's discretion, the actuation setpoint may be lowered by use of test signals at the PPS cabinets 1,1.S RCS pressure is stable and above the SIAS setpoint.

1.1.6 Containment pressure is below SIAS setooint and not increasing.

PV216400 REV LIE 1965Pl

FOR INFORMATlON ONLY PROCEDVRE PALO VERDE NUCLEAR GENERATING NO. APPENDIX P STATION MANUAL 41EP-1ZZ01 Page 2 o= 14 .

REVISION E"iERGENCY OPERATIONS Page 93 of 179 "RESETTING SAFETY INJECTION ACTUATION SIGNAL (SIAS AaW CONTAIHNE'Vs 'ISOLATION ACTUATION SIGNAL ('CIASI 'ont'd 1.1.7 RVLis indicates void restricted to upper head.

1.2 CIAS reset criteria 1.2.1 Containment pressure is belov CIAS setooint and not, increasing.

1.2.2 RCS pressure is stable and aoove SIAS/CIAS setpoint.

2.O +AN THE CONDITIONS Or 1.O ARE:1ET, RESET Am RESTORE SIAS/CIAS TO STANDBY.

CAUTION IF ANY OF T/K FOLLOWING CRITERIA CANNOT BE

.'ET,. TsKY, FULL SI FLOti'!ST BE R:--INITIATED.

1) RCS SUBCOOLED GREATER TELE 28 F

2) RVL'is VESSEL LEVEL INDICATES VOID RESTRICTED To UPPER 1KAD

3) PZR LEVEL GREATER THKX 33'.~ Ai%)

CONTROLLABLE.

4) ONE S/G CAPABLE OF ~1ADTAINING HEAT RE.'10VAL 2.1 At the Plant Protection System modules on BOS reset all four safety channels by pressing the reset button on each module.

Obtain the Actuation Trip Path Reset. and Auxiliary Relay Cabinet. door i;eys.

2.3 At the PPS cabinets reset all four channels'by:

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FOR iNFORMATION ONLY PROCEDURE PALO VERDE NUGLEAR GENERATING NO.

STATION MANUAL 410P-1SI01 REVISION SHUTDOWN COOLING INITIATION Page 5 of 40 I

3.0 LIMITATIONS AND PRECAUTIONS 3.1 Do not .exceed the following shutdown cooling loop limitations:

The shutdown cooling loops temperature shall not exceed 350 F.

Pressurizer pressure shall not exceed 410 psia when a shutdown cooling loop is open to the RCS.

3.2 Do not exceed the RCS cooldown rates of Tech. Spec. 3.4.8.1-.

3.3 Do not exceed a 19 F per minute heatup rate on the shutdown cooling loop.

3.4 The Containment Spray (CS) system shall be operable in Modes 1 thru 3, and in Mode 4 whenever shutdown cooling is not in operation. (per Tech.'pec. 3 '.2) 3.5 When using a SDCHX for RCS cooldown, its outlet valve to the containment spray header, SIA-HV687 Train A, or SIB-HV695 Train B, shall be shut to prevent spraying RCS coolant into containment in the event of a CSAS.

3.6 The LPSI pump mini-flow recirc., 100 gpm, provides sufficient heat removal for only one hour of pump operation. Continuous operation of the pump requires, at 'a minimum, 1300 gpm flow to cool the pump.

3.7 The CS pump mini-flow recirc., 150gpm, provides sufficient heat removal for only one hour of pump operation.

I 3.8 Take care to ensure that there is always a flow path for the LPSI and CS pumps when they are operating with their mini-flow recirc.

valves shut.

3.9 Do not exceed a LPSI pump flow of 5000 gpm, or a motor run current of 60 amps.

3. 10 Do not exceed a CS pu'mp flow of 4550 gpm, or a motor run current of 95 amps.

I ATTACHMENT 'B'OMMENTS ON SRO EXAM PAGE 1 SECTION V QUESTION 5.19 The keyed answer (B) is the correct Theory

Response

this if the +0.25 ASI must be corrected time. However, the Procedure Appendix at (410F-1ZZ05"B) says to wait unit ASI crosses ESI heading negative. If an operator candidate writes an answer, instead of selecting an option from the multiple choice, the answer should be graded accordingly.

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COMMENTS ON SRO EZAM PAGE 2 SECTION VI QUESTION 6.01a The question asks for a specific manipulation.

The answer has two manipulations that must be accomplished. A candidate may respond in the singular which would be "go to setpoint override" rather than detail the switch changes necessary to go to setpoint override.

ref. PPCS Training Article 3.5.1.

QUESTION 6.02 The answers are reversed. Answer (a) should be (5) and answer (b) should be (3).

ref. NLC56-01-ZC 11 01 Pg. 51.

QUESTION 6.05 This question is extremely open to individual interpretation based on the understanding of the term "core damage". Core damage could be interpreted as anything that exceeds the acceptance criteria of 10 CFR 50.46 especially since the question reference (NLC21-00-RC-06 E01) addresses those acceptance criteria. Core damage could also mean any change to the'ore which is a degradation from what was installed, such that allowable fuel pin leakage would be considered "core damage". Thus, if the former assumption is made, the answer is TRUE while the latter would be FAT SE.

QUESTION 6.06 Of the choices given, choice (b) is the most correct but the candida'tes may explain that the actual basis is 2 HPSI and 3 charging pumps which would be both (a) and (b).

ref. T.S. basis 3/4.4.8

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ATTACHMENT 'B'OMMENTS ON SRO EXAM PAGE 3 SECTION VI Continued QUESTION 6.07 There is, in reality, only one principle flow path besides controlled bleedoff. That path is to the RCS (NLC21-00-RC-02, Pg. 15). However, in an attempt to answer the question which requires two flow paths, some candidates will probably answer "vapor seal (3rd stage) leakage to the RDT" as is on the answer key. Others may believe the question is asking how the seal injection flow is accounted for in two principle flow paths'hey would answer 1)RCS and 2) VCT from controlled bleedoff. The vapor seal flow rate is normally 0 gpm so it would not be considered as a principle flow path, by most people, during normal RCP operation.

QUESTION 6.11d The containment'spray system is a subsystem of Safety Injection, (SI) so an acceptable answer should be SX to EN to SP.

QUESTION 6.13 a ~ The question may be misleading since the diesel generator needs to be overridden for any automatic start that is not reset. Therefore, in addition to LOP, any signal that causes an automatic start should be accepted as correct. These are:

SIAS/CSAS or APAS1 (2). Also, LOP essentially automatically resets as soon as the DG output breaker closes. It does, however, have a "memory" circuit that necessitates the use of override to perform a normal shutdown.

If for some reason,, the automatic reset did not it occur, the DG could not be shut down normally, would require an Emergency stop. The sequence of events presented is also somewhat misleading since the DG had to be placed in override in order to perform the parallel operation stated.

b. A correct answer should not require the candidate to say "then to STOP" since the DG will be in override as soon as START 'has been selected.

ref. NLC22-00-RC-033 Pgs. 48, 49 41AO-1ZZ52.

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'I QUESTION 6.16c While the new CWP's in Cycle 2 are part of the pxogram to de-sensitize CEA penalty factors, their purpose is not to help de-sensitize the penalty factors; they are more of a result of the program.

Their purpose is to alert the operator and stop rod withdrawal prior to a penalty being inserted thus preventing unnecessary reactor trips. Also, the original point breaL".down for part c. has allotted (0.5) pts. while the question allowed (1.0).

ref. NK,C56-01-ZC-007-02 CPC functional design spec (proprietary).

QUESTION 6.18 The candidate can also respond that internal to the high range monitor will shift to alternate filters as activity increases on the filters.

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ATTACHMENT sB COMMENTS ON SRO EXAM PAGE 5, SECTION VII QUESTION 7.03 Answer key calls for'2 answers, 1: Containment Sump Level Indication; 2 is AFN-P01 Low Suction Pressure Trip. 'esson Plant NLC32-03-RC-001 refers to 41RO-1ZZ01 Step 5.16.1 (reenergize M19 for Containment Sump Level Indication only), thus enabling AFN-P01 Low Suction Pressure Trip enabled by locally monitoring Pump Suction Pressure.

QUESTION 7.04 App. E of 41AO-1ZZ13 does not delineate an "upper detector string" as such, it refers to detector numbers 1 to 4 for determination of upper head void size QUESTION 7.13 Comment on question,,by examiner, directed candidates to not use "RWT" only-CST level 41AO-1ZZ13 addresses "RMWT"vice "RWT", this could lead candidate to assume that "RMWT" is available as a water source and would definitely result in an answer exceeding that given'n the answer key.

QUESTION 7.14 This question does not tell what mode the plant is in, while the answer key implies Mode 1 (41AO-1ZZ17 rev. 2 Step '1-1.2, states "Xf the plant is in Mode 1"). However, or 3, a trip may not be ifnecessary.

the plant was in Mode 2 Additionally, NLC32-030RC-025 student objective EO3 states: reason for reactor trip A. "MSIV's and FWIV's fail shut". It does not state "Economizer FWIV" as the answer key does. I Additionally the question requires devices "in,two SEPARATE plant systems" the answers given are both in the same plant system (SG) this could cause the candidate to" answer with a device in some other system which may cause a trip.

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SECTION VII Continued QUESTION 7.16

b. Wants the immediate action and the answer key states "Close the Affected RCP Bleedoff Valve" while the reference (NLC32-03-RC-27 E04) states "within 1 minute shut the seal bleedoff valve".

This is to an "immediate" action. Procedure 41AO-1ZZ29, Step 8.3 states that the seal bleedoff valve should be shut immediately, once the, um has been shut down.

QUESTION 7.18 While the examiner gave instructions to candidates to specify page number for App. "B" of 720P-9RZ01, the answer key states number "7".

QUESTION 7.19 Reference NLC31-03-RC-006 "E03" does not address ASI. It is addressed in 410P-1ZZ04, Rev. 3, App.

"G" which is not listed as a reference for this question.

QUESTION 7.20 Reference NLC31-03-RC-006 does not'pecifically address which plant parameter changes significantly during feed valve- swapover. It is addressed in 410P-1ZZ04, Rev. 3, App. "D".

QUESTION 7.24 This answer could include "and RCS Pressure/inventory control is lost". Reference 41A0-1ZZ14, Rev. 1.

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ATTACHMENT 'B "COMMENTS ON SRO EXAM PAGE 7 SECTION VIII QUESTION 8.03d Part d of the question may lead to confusion since 40AC-'ZZ02, Conduct of Shift Operation directs the Unit 1 Shift Supervisor to assume the EC position in the event of Security Violation involving the entire site.

ref. 40AC-9ZZ02 Step 4.1.2.

QUESTION 8.04b Per conduct of Shift, 40AC-9ZZ02, 5.2.2.2 the Pire Team Advisor is responsible to "assess the potential safety consequences of the fire and advise control room personnel." This would imply all safety concerns not just safety related plant system.

QUESTION 8.13a The question states the fuel bundle is located over the top of the core. If the assumption is made that the bundle is above the= flange level than it is not longer a core alteration, per T/S

'defn. 1 9g an/ the answer, would be "No."

QUESTION 8.20c If assumption is made that Az. Tilt remains at .11 than per T/S 3.2 ' action b.2 power should be reduced to 50'.

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PROCEDURE PALO VERDE NUCLEAR GENERATING NO. APPENDIX B STATION MANUAL 41OP-1ZZ05 Page 1 of J

6 REVISION POWER OPERATION Page 24 of 35 ASI CONTROL DURING POWER OPERATIONS' 1.0'BJECTIVE-To provide the operator with guidelines that will aid in maintaining ASI within the limits specified by the CE Core Operating Guidelines to ensure fuel conditioning requirements and to maintain Tech. Spec limits.

. 2.0 PERSONNEL INDOCTRINATION Due to the long active fuel length in System 80 plants it is anticipated that divergent axial xenon oscillations will occur over a wide range of power levels 'and cycle burnups. This means that if uncontrolled, power will peak with larger and largei values in the top and bottom of the core as burnup progresses (see Figure 2).

Because this can cause damage to fuel, a means of controlling these oscillations becomes necessary.

The general method for controlling oscillations will be with either CEA (reg groups of PLCEA's) movement dilution, or power transients, dampen the power as it begins to swing to the top (negative ASI) of the core. Dampening is intended to prevent the ASI from falling outside limits specified by the CE Core Operating Guidelines. However, dampening operations should be initiated when ASI is expected to exceed +.02 of the ESI being used.

will go through a complete cycle about. every 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . If

~P ASI dilution method is used, as ASI crosses ESI toward negative, dilution will be initiated to dampen .the cycle. The amount of dilution will be restrained by the ability to maintain temperature.

If CEA's are being used to dampen the cycle, reg groups or PLCEA's will be partially inserted as ASl crosses ESI toward negative.

When the axial power begins to shift to a more positive ASI, the rods are slowly withdrawn to complete the dampening of the oscillation. Consideration should be given for use of the PLCEA's for ASI control since 'these do not have requirements placed upon them by the PDIL's.

As the core approaches end of cycle (EOC), the oscillations are expected to become more and more unstable. This will entail performing dilutions at -more frequent intervals or driving the CEA's in further and at more frequent intervals.

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41 FOR lNFORMATlON ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO. APPENDIX B STATION MANUAL 41OP-1ZZ05 Page 2 of 6 REVISION POMER OPERATION Page 25 of 35 ASI CONTROL DURING POWER OPERATIONS (Cont'd If the xenon oscillation is excessive and the ability to use dilution or CEA movements does not exist, then a power transient.

may b'e used. A power transient should not be used unless absolutely necessary.

If the fuel is not conditioned for full power operation (i.e., if t'e fuel has not, been operated at full power for more than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months within the last 60 days), then a power ascension should be performed unrodded. The unrodded power ascension will subject the fuel uniformly to the progressively higher linear heat rates and avoid the localized power ramps caused by CEA withdrawal. Since the core is xenon free in these. situations, the only factor affecting ASI is the change'n RCS temperature.

If the "fuel is already conditioned for full power and the coze is not xenon free, the power ascension should be performed rodded.

Vith xenon in the core, an additional factor is present which may aggravate the change in ASI during the power increase, causing a significant overshoot of the target value and a substantial xenon oscillation upon the return to full power. Consequently, to counteract the force of the xenon, the ASI must be driven to its equilibrium value by insertion of CEA's.

If the fuel is already conditioned for full power operation and the core is xenon free, the power ascension may be performed either zodded or unrodded. If the power ascension is expected to be relatively 'smooth and uninterrupted, then an unrodded ascension may be more desirable because the xenon concentration will not have had time to become significant. If the power ascension will.be interrupted by temporary holds at intermediate power levels, then a zodded ascension may be more desirable since a significant xenon

'concentration may build in before full power is attained.

Maintaining ASI to within +.05 of the full power ESI during transient conditions below 50;o power is desirable. Also, maintaining AS1 to within +.01 of the full power ESI during transient conditions above 50";. power is desirable. ASI should be maintained within the +.05 full power ESI band at all power levels. However, ASI shall be within +.05 of rhe full'power ESI pxior to exceeding 50io power. The only exception to the +.05 value above 50/ power is in extreme cases when there exists the inability to control ASI to +.05. In these cases, actions shall be taken to return ASI to +.05 ESI as soon as possible. Reactor Engineering shall be contacted to evaluate the effects of exceeding the +.05 value.

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FOR INFORMATION ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO. APPENDIX 8 STATION MANUAL 410P-1ZZ05 Page 3 of 6 REVISION POWER OPERATION Page 26 of 35 ASI CONTROL DURING POMER OPERATIONS (Cont'd During a reduction in power, ASI should be maintained within +.05 of the ~fu?1 ower ESI if the power reduction is less than 8 days.

CEA's will be inserted as necessary to maintain ASI during the power reduction. If the power reduction is expected to be greater than 8 days, Reactor Engineering shall be contacted to provide a new ESI value for ARO at the reduced power level. This will ensure that fuel conditioning is maintained uniform throughout the coze.

For cycle 1, below 75% power with COLSS in service (55% power, COLSS out of service), the normal overlap between groups 4 and 5 may be modified for ASI control." To prevent an out of sequence trip, group 4 should be positioned at least 10 inches higher than group 5 and can be as high as the normal overlap value.

When normal overlap is not maintained between groups 4 and 5, both groups of rods will need to be evaluated separately in LCO 3.1.3.6 to insure compliance with the transient insertion limit. The maximum insertion limit for group 4 is 60 inches withdrawn when the normal overlap is not maintained. when groups 4 and 5 are positioned closer together than-normal overlap, group 4 becomes the limiting group. For example, to allow a group 4 insertion to 85 inches withdrawn (group 5 at, 75" withdrawn), the power limit is approximately 47% (cycle 1, COLSS inservice).

Late in core life at low power level where the core is heavily top peaked, the most effective CEA configuration for forcing power to the bottom of the core is expected to be part lengths at 75 inches, group 5 at 75 inches and group 4 at 85 inches withdrawn.

Anytime the normal CEA overlap is not maintained and power is greater than 20%, the timing requirements of LCO 3.1.3.6 action b.

shall be observed.

ASI can be read from the ASI meter on 804 SEN-JI-9, or from core monitoring Computer Point I.D. 8 RJANA04 (preferred point), or from CPC Point I.D. 8268.

3.0 PRECAUTIONS 3.1 Reactor power and RCS temperature shall be maintained as constant as possible during all ASI dampening operations, 3,'2 The dampening action, either dilution or CEA'ovement, shall be in small frequent steps. Allow several minutes to pass before initiating additional dampening actions so that the impact of the move upon ASI can be determined.

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FOR INFORMATiON ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO. APPENDIX B STATION MANUAI 410P-1ZZ05 Page 3a of 6 REVISION POMER OPERATION Page 26a of 35 ASI CONTROL DURING POWER OPERATIONS (Cont'd 3.3 If any indicated ASI reaches + 0.25 or a CPC DNBR, or LPD pretrip setpoint is reached action shall be taken to bring the ASI, LPD or DNBR back to within allowable limits. An acceptable means of complying with this precaution would be to reduce power by boration when bottom peaked or inserting regulating CEA's or PLCEAs when severely top peaked.

3.4 If the dilution method of ASI control is being used and RCS temperature is not being adequately maintained, then control of ASI using CEA's may be necessary.

4.0 INITIAL CONDITIONS 4.1 A trend chart, recorder should be used to monitor ASI to know when I

to initiate dampening, actions.

5.0 INSTRUCTIONS 5.1 Honitor ASI and determine if the next negative peak or the following positive peak will exceed +.02 of the full power ESI.

The peaks can be projected using the ASI program on the Operator Assistance Calculation System or the method discussed in the worksheet in Appendix B.

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PROCEOURE PALO VERDE NUCLEAR GENERATING NO. APPENDIX B STATION MANUAL 410P-12205 Page 4 of 6 REVISION

" POlKR OPERATION Page 27 of 35 ASI CONTROL DURING POMER OPERATIONS (Cont'd 5.2 If the projected next negative peak or the projected following positive peak will exceed +.02 of the"full power ESI, then dampening of the oscillation should begin halfway down the backside of the present positive peak (see Figure 2) using dilution or CEA insertion.

5.3 Dampen the oscillation using dilution or CEA insertion. Naintain RCS temperature and power as necessary.

5.4 After the dampening action has been performed, ASI will be seen to drift towards the lower limit. The dampening action will be completed when this no longer occurs.

5.5 Secure the dampening action. Allow ASI to drift to the positive side of the band.

5.6 Vhen ASI has crossed ESI toward positive, begin CEA withdrawal, if used.

5.7 Secure the dampening action.--

5.8 "Repeat steps 5.1 to 5.7 when necessary.

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II 3.4.2 Pressurizer Hi h Pressure Heater Cutout In the same manner as the low-level interlock, at 2350 psia increasing, the selected high pressure alarm/control bistable (PSH-100X, PSH-100Y) trips and blocks closure of all heater breakers. The interlock remains in effect until pressure decreases to below the 2330 psia bistable reset, point.

3.5 Overrides 3.5.1 Normal Pressure Set oint Override The Normal'ressure Setpoint Override allows the backup heaters to be energized while pressure is above the 2225 psia control bistable (PC-100X, PC-100Y) r'eset point. This feature is initiated by taking the heater bank control switch (HS-100-4, 5, 6, 7, 8, 9) first to OFF then to ON. A white OVERRIDE light illuminates to indicate this mode. After a ten second time delay, the override can be cancelled by taking the control switch to OFF. 10 seconds must then elapse before the override can be re-initiated. In the override mode, heater operation is still subject to the low-level and high pressure interlocks.

3.5.2 SIAS Override As required by Technical Specification 3/4.4.3.1 the two Class 1E powered backup heater banks trip on SIAS. The SIAS Override allows these heater banks to be reenergized with a SIAS present. The override is initiated in the same manner as the Normal Setpoint Override, by taking the control switch (HS-100-4, 5) to OFF then ON. A separate white OVERRIDE light illuminates to indicate this mode. Once initiated, the heater banks can be deenergized and reenergized as desired, but the override permissive remains in effect until SIAS is reset. In the override mode, heater operation is still subject to the low-level and high-pressure interlocks.

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interface between CEDMCS and other A. The following provide input to plant systems. CEDMCS:

1. RSPT's

a. UEL for interlock and ops module indication.

b. LEL for interlock and ops module indication.

c. DRC for indication to mimic and reset to PC.

2. Plant Computer Inputs.

a. UGS

b. LGS

c. Sequential overlap permissives.

3. PPS

a. CWP

4. SBCS

a. AMI if 15K or if logic net following load reject.

b. ANP for SBCS demand.

1 CONTROLLED BY USER REACTOR COOLANT SYSTEH BASES SPECIFIC ACTIYITY (Continued Reducing T ld to less than 500 F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4. 8 PRESSURE/TEMPERATURE LIHITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Chapters 3 and 5 of the FSAR. Ouring startup and shutdown, the rates of temperature and pressure changes are limited so as not to exceed the limit lines of figure 3.4-2. This ensures that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

Ouring heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at.

the outer wall. These thermal induced compressive stresses at the inner wall tend to alleviate the tensile stresses induced by the internal pressure.

At the outer wall of the vessel, these thermal stresses are additive to the pressure induced tensile stresses. The magnitude of the thermal stresses at either location is dependent on the rate of heatup. Consequently, each heatup rate of interest must be analyzed on an individual basis for both the inner and 4 outer wall.

The heatup and cooldown limit curve (Figure 3.4-2) is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup or cooldown rates of up to 100 f per hour. The heatup and cooldown curve was prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figure 3.4-2.

The reactor vessel materials have been tested to determine their initial RTNT, the results of these test are shown in Table 8 3/4.4-1. Reactor oper a-tion and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RTNpT Therefore, an adjusted reference temperature, based PALO VERPE - UNIT 1 B 3/4 4-6 ~l

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GONTROLLED BY USER REACTOR COOLANT SYSTEM BASES PRESSURE/TEMPERATURE LIMITS Continued) upon the fluence and residual element content, can be predicted using.

Figure B 3/4.4-1 and the recommendations of Regulatory Guide 1.99, Re'vision 1, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curve Figure 3.4-2 includes pre-dicted adjustments for this shift in RT at the end of the applicable service period, as well as adjustments for possNTe errors in the pressure and temperature sensing instruments.

The actual shift in RTN of the vessel material will= be established periodically during operatioII Tby removing and evaluating, in accordance with ASTH E185-73 and Appendix H of 10 CFR 50, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the delta RTN T determined from the surveillance capsule is different from the calculateII delta RTNDT for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50. The reactor vessel material irradiation surveillance specimens are removed and examined to determine changes in material properties. The results of these examinations shall be used to update Figure 3.4-2 based on the greater of the following:

(1) the actual shift in reference for plates H-6701-2 and H-4311-1 and weld 101-142 as determined by impact testing, or (2) the predicted shift in reference temperature for the limiting weld and plate as determined by RG 1.99, "Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Haterials."

The maximum RT T

for all Reactor Coolant System pressure-retaining materials, with the fxception of the reactor pressure vessel, has been deter-mined to be 40'F. The Lowest Service Temperature limit is based upon this RT<nT since Article NB-2332 (Summer Addenda of 1972) of Section III of the ASHE Bo>Thr and Pressure Vessel Code requires the Lowest Service'emperature to be RTNn~ + 100 F for piping, pumps, and valves. Below this temperature, the system preKCure must be limited .to a maximum of 20Ã of the system's hydrostatic test, pressure of 3125 psia. However, based upon the 10 CFR Part 50 Appendix G analysis, the isothermal condition for the reactor vessel is mo~e restrictive than the Lowest Service Temperature line. Therefore, only the isothermal line is shown on Figure 3.4-2.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

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1 PALO VERDE NUCLEAR GENERATING PROCEDURE NO. Pcs Kl STATION MANUAL 41AO-1ZZ52 PclV P>

REVISION DIESEL GENERATOR OPERATIONS AFTER ESFAS ACTUATIONS Page 6 of 12 3.0 TRANSFERRING 4.16 KV BUS S03 (S04 FROM DIESEL GENERATOR A (B TO NORMAL SUPPLY AND SUBSE UENT DIESEL SHUTDOWN AFTER A LOP ONLY START SIGNAL OR DIESEL START DUE TO A LOP SIGNAL WITH AFAS AND OR SIAS SIGNALS REGARDLESS OF WHETHER THE AFAS AND OR SIAS SIGNALS HAVE BEEN RESET NOTE Only the high-resistance grounded Diesel Generator supply will operate with a sustained ground fault on the system when in "Emergency Mode."

Therefore, the following alarms are checked prior to synchronizing with offsite power because the Diesel Generator breaker will open when placing the Diesel Generator in "Override" if any of these conditions exist.

3.1 Verify "Ground Overvoltage," "Overcurrent," and "Load Unbalance Trip" alarms are not present at the annunciator on DGA(B) B01.

3.2 Record and reset targets on undervoltage relays 727-1UV, 727-2UV, 727-5UV and 727-6UV on PBA-S03A (PBB-S04A); and 727-3UV, 727-4UV, 727-7UV, and 727-8UV on PBA-S03R (PBB-S04R).

NOTE Complete Steps 3.3 thru 3.14 for only one Diesel Generator at a time to prevent paralleling both Diesel Genexators to the Offsite Power Source simultaneously. (Reference FSAR 8.3.1.1.3.5) 3.3 Take the "Start/Stop" switch DGA-HS-1 (DGB"HS-2) on B01 to the "Start" position. Observe the white "Override" light comes on.

3.4 Place Diesel Generator A(B) "Speed Mode Select Switch in the "Droop" position. PEA-SS-GOlD (PEB-SS-G02D). Output frequency may decrease depending on Diesel Generator load.

3.5 Verify the Diesel Generator is actually in "Droop" by raising and/or lowering the speed with PEA(B)-SC-G01(2) and noting a corresponding change in frequency on PEN-SI-G01(2).

3.6 Place the Synchronizing Switch for 4.16 Kv bus PBA-S03 (PBB-S04) normal supply PBA-SS-S03L (PBB-SS-S04K) in the "ON" position.

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STAT:lON lNANUAL ' 41AO"1ZZ52 REV/SION DIESEL GENERATOR OPERATIONS AFTER ESFAS ACTUATIONS Page 7 of 12 3.7 Using Diesel Generator..A(9). Speed Switch PEA-SC-G01 .(PEB-SC-G02) adjust Diesel Generator speed to cause the synchroscope to move

. slowly. in the "Fast" direction.

3.8 Match the Generator Voltage with Bus Voltage using Diesel

, Generator A(B) Voltage Raise/Lower switch PEA-.EC-G01 (PEB-EC-G02) located. on 801, .Running. Voltmeter.MAN-EI-002R represents, Bus voltage.. Incoming,voltmeter MAN>>EI-0021'epresents ESF Transformer voltage.

CAUTION

'IF' SIAS OR AFAS SIGNAL IS RECEIVED WHILE THE DIESEL GENERATOR IS IN OVERRIDE AND OPERATING IN PARALLEL WITH OFFSITE POWER, THE DG OUTPUT BREAKER WILL NOT TRIP. IMMEDIATE OPERATOR ACTION IS:.REQUIRED TO TRIP THE DG OUTPUT BREAKER.

~ CAUTION THERE IS NO SYNC"CHECK PROTECTION ON THE NORMAL SUPPLY BREAKER.

3.9 When the Syncroscope needle is at the 12 o'lock position close 4.16 Kv bus S03 (S04) normal supply breaker PBA-HS-S03L (PBB-HS-S04K).

3.10 Place synchronizing switch PBA-SS-S03L (PBB"SS-S04K) in "OFF",

PV216 000 ABI, 8/82

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FOa iMFOaMATiOw ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO.

STATION MANUAL- 41AO-1ZZ52 REVISION DIESEL GENERATOR OPERATIONS AFTER ESFAS ACTUATIONS Page 9 of 12 NOTE The white "Override" light will remain ON after performing step 3.16 are not reset.

if any AFAS or SIAS signals 3.16- Turn Diesel Generator A (B) "Start/Stop" Switch DGA-HS-1 (DGB-HS-2) on B01 to the "Stop" position.

3.17 Ensure Lube Oil Circulating Pump DGA-P04 (DGB-P04) starts after the Engine stops.

NOTE If Diesel Generator A(B) has run more than one hour while loaded at less than 2.5 MW; Then operation for the purpose of carbon burnout is required.

If carbon burnout operation is required and an AFAS or SIAS signal is present and the white "Override" light is on; Then proceed to step 3.18.

OR If carbon burnout operation is required and all AFAS and SIAS signals are reset and the white "Override" light is off; Then proceed to step

3. 1'9.
OR If carbon burnout operation is NOT required; Then proceed to step 3.20.
3.18 Reset any AFAS or SIAS signals that are present when allowed and in accordance with 41EP-1ZZ01, EMERGENCY OPERATIONS.
3.19 Verify the Diesel Generator A(B) white Override" light is off, then restart and load the Diesel in the "Test" mode per 410P-1DG01 (410P-1DG02) for the purpose of Carbon burnout.
3.20 If Diesel Generator A(B) has run more than one hour, perform the Appendix J test in 410P-1DG01 (410P-lDG02).
3.21 When the Essential Spray Pond Train A(B) is no longer required for other equipment operation and when Diesel Generator lube oil temperature is <160'F the Essential Spray Pond Train A(B) may be shutdown per 410P"1SP01 (410P-1SP02).
PV2N 000 AEV, 8182
REVI NUMBER: 02 COURSE TITLE: I Review PAGE: 49 LESSON NUMBER: NZC56-01-ZC-007-02 LESSON TITZ,E: Core Protection Calculator STUDENT OBJECTIVE(S) SEQUENCE METHODS & ACTlYlTlES
4. Associated safety channel N.I. trouble (NUC INST CH INOP annunciator on B04):
, a. High voltage < + 800 VDC
b. Any of the 3 drawer calibrate switches not in trip
'OPERATE'.
Any of 5 drawer not in test'witches circuit
'OPP'.
Any drawer card removed or in the wrong position.
D. CPC channel CNP output trips Point out new DNBR pre-trip setpoint.
DNBR pre trip:
a. Quality margin in limiting node < 0.05
b. DNBR < 1.39
2. ZPD pre-trip
a. LPD > 20 kw/ft
3. RPC (RZ Power Cutback) flag set
a. RPC flag received from CEACs; they sense group 5 only or groups 4 and 5 falling in
b. RPC flag set for 20 seconds
I REVI NUMBER: 02 COURSE TITLE: I 6 Review PAGE: 50 LESSON, NUMBER: NLC56-03.-ZC-007-02 LESSON TITLE: Core Protection Calculator STUDENT OBJECTIVE(S) SEQUENCE METHODS 5 ACTIVITIES CEAC CEA DEVIATION (CEADEV) Note that the four (4) new
a. Provides or CWP if the DNBR LPD penalty factor
. causes of CWP listed here are part of the program to received from an operable de-sensitize the CEA CEAC is greater than 0 related penalty factors.
(penalty of > 1.0)
5. CPC Dre-aenalty (IPCMI)
a. Provides CWP prior to receiving a CPC generated penalty factor
b. CWP occurs at > 5 inches deviation between subgroups within a group c.: CWP occurs at < 30 inches withdrawn on a PLCEA
d. CWP occurs when a Reg group is < 7.5 inches above tne next-higher numbered Reg. group
(
FOR mNFORMATlQN ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO.
STATION MANUAL 41RO-1ZZ01 REVISION REACTOR TRIP Page 13 of 30 5.10 Reactor Power Decreasing or Stable YES/NO 5.11 Tank Levels 5.11.1 Refueling Mater Tank 5.11.2 Reactor Hakeup Mater Tank feet, 5.11.3 Uolume Control Tank F 11,4 Condensate Storage Tank feet 5.12 Steam Generator No. 1 Level Stable or Recovering YES/NO 5.13 Steam Generator No. 2 Level Stable or Recovering YES/NO 5.14 SG Feedwater Source 5.15 Condenser Vacuum Available for SBCS Operation YES/NO 5.16 Containment Parameters Normal or Recovering YES/NO (Temp, Pressure, Humidity, Isolation).
5.16.1 IF SIAS actuated, THEN direct an auxiliary operator to override and reenergize NHN-H19 to provide GNAT Radwaste Sump Level indication to the Control Room.
5.16.2 Containment Sump Level 5.16,3 List abnormal parameter(s)
PV21640D AFV. 6/bt (96SPl
FOR lNFORMATION ONLY PROCEDURE PALO VERDE NUCLEAR GENERATlNG NO. APPENDIX E STATlON MANUAL 41AO"1ZZ13 Page 1 of 2 REVISION NATURAL CIRCULATION COOLDOMi Page 43 of 47 RVUH LEVEL VS. -PZR LEVEL, FOR RCP RESTART 1.0 RVUH level can be read on page 212 of the QSPDS. Level indication is delineated either as a pictorial representation which indicates which sensors are uncovered (voided) or as a percentage.
The following chart. should be used by the Shift Supervisor as a guide to determine the desirability of RCP Restart with a void either indicated or suspected in the Rx Vessel Upper Head. The following guide lines will preclude emcpying the Pressurizer on RCP Restart.
Elevation Minimum Indicted Maximum PZR Level Heter -
of Detector Void Void Void for RCP Reading Dectector Above FAP- Size a Size b Size c Restart d
(%) Number (inch) (ft ) (ft') (<<) (%)
67 1 223 230 500 810 48 41 2 169 810 1082 1354 78 16 3 115 1354 1627 1898 100 0 4 ., 61 1898 1967 2039 100 73 46 2039 2100 2161 100 47 6 32 2161 2222 2284 100 21 7 18 2284 2345 2406 100 0 8 4 2406 2441 100
"-FAP = Fuel Alignment Place
a. This void size would exist if the detector were just barely uncovered.
b. These values correspond to the percent liquid volumes indicated by the QSPDS.
c. This void size would exist, if the next lower detector were just barely .
covered.
d. Based on the maximum void size.
l 1[
r 0 I=.=~CIQR V=SS=I VOID VCI Uhl=
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1 FOR INFORMATION ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO.
STATION MANUAL 41AO-1ZZ17'EVISION LOSS OF 125 VDC CLASS IE ELECTRICAL POWER Page 4 of 49 1.0 VERIFICATION CHECKS 1.1 Loss of DC Bus PKA-H41 1.1.1 Indications
1) Voltmeter PKA-EI-M41 indicating zero volts.
2) "125V IE CC M41 CHGR A/AC PNL D21 TRBL" ALARH (window number 1A04A)
CAUTION ON A LOSS OF PKA-H41 THE HSIV'S AND FEEDWATER ECONOMIZER ISOLATION VALVES FAIL CLOSED.
1.1.2 If the plant is in Hode 1, TRIP the REACTOR manually and perform the actions of 41EP-1ZZ01 concurrently.
1.1.3 Proceed to step 2.0 on a loss of PKA-M41 1~2 Loss of DC Distribution Panel PKA-D21 1.2.1 Indications
1) Plant computer alarm
2) "125V IE CC M41 CHGR A/AC PNL D21 TRBL" ALARH (window number 1A04A)
CAUTION ON A LOSS OF PKA-D21 TllE HSIV'S AND FEEDWATER ECONOHIZER ISOLATION VALVES FAIL CLOSED.
1.2.2 If the plant is in Mode 1, TRIP the REACTOR manually and perform the actions of 41EP-1ZZ01 concurrently.
1.2.3 Proceed to step 4.0 on a loss of PKA-D21.
PV216 000 REV. 8I82 684 A
tor 4
.REV ON NUMBER: 0 COURSE TITLE: Si 0 erator Course 'B'E AGE: 5 NUMBER: NLC32-03-RC-027-00 LESSON TITLE: RCP and Motor r enc STUDENT OBJBQTIVE S) SEQUENCE METHODS & ACTIVITIES E04. With an RCP running, Operat5ng 1imits with loss oi seal gV. Roger to 8,0.
state the limits with injection and/or loss of NCH. RCP is a loss of NC and/or running. Emphasize shutting the seal injection. bleed-off valve.
A. No NC 10 minutes.
B. No injection indefinite.
C. No NC, no injection 1 minute.
Close bleed-off valve.
V." limits with loss of seal C
E05. Wi'th an RCP not Operating V. Refer to 9.'2.
running, state the injection and/or loss of NCW. RCP is limits with a loss of shutdown. This is for an RCP in NC and/or seal hot standby, emphasize injection. A. No NC, no injection 20 minutes. that seal bleed-off Insure that the bleed-off valve must be shut within, is shut within one minute. one minute when seal injection is lost.
B. No NC insure bleed-off valve is This is to prevent open. seal damage.
C. No injection insure bleed-off valve is shut.
BOG. State when seal VI. When should seal injection be VZ. Refer to 9.2.
injection should be maintained.
in operation.
A. RCS temperature >250op.
B. RCS pressure >150 psia.
C. RCS level higher than top of, cold legs.
PVT62.03ll AEV 1 66
FOR INFORMATlON ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO.
STATION MANUAL 41AO-1ZZ29 REVISION REACTOR COOLANT PUMP AND MOTOR EMERGENCY Page 19 of 30 NOTE The Reactor Coolant Pumps may be operated for an indefinite period of time without seal injection provided adequate NCM flow is maintained to the RCP's. This is detrimental however due to increased wear and contamination of the RCP seals and the water lubricated carbon journal bearing.
NOTE If the seal injection flow to the affected RCP(s) is low or lost the cyclone filter on the pump may have clogged. Consideration should be given to shutting down the affected RCP and flushing the cyclone filter per section 12-15 of 410P-1RC01 "RCP Operations".
S.2 VERIFY adequate seal injection flow (normal 6.6-7.5 gpm) on 803.
If seal injection flow is inadequate or lost, proceed to 410P-1CH03 "RCP Seal Injection System".
Seal Flow Indication (803)
RCP-1A CHN-FIC-241 RCP"18 CHN-FIC-242 RCP-2A CHN-FIC-243 RCP-28 " CHN-FIC-244 8.3 If seal injection and NCl( flow are lost simultaneously, immediately CLOSE the controlled bleedoff valve on the affected R.C.P., once the pump has been manually shutdown.
Bleedoff Valve H. S. Number (803)
RCP"1A RCN"HS-430 RCP-18 RCN-HS-431 RCP-2A RCN-HS-432 RCP-28 RCN"HS-433
~16 OOD AEY. 6I62 o~p~
1 I
lt
FOR INFORMATION ONLV PALO VERDE NUCLEAR GENERATING PROCEDURE NO. Appendix G STATION MANUAL 410P-1ZZ04 Page 1 of 3 I REVISlON PLANT STARTUP MODE 2 TO MODE 1 Page 27b of 27 LOW POWER ASI CONTROL 1.0 DISCUSSION At low reactor power the CPC's use a pre-programmed ASI value. The CPC's begin calculating-ASI using actual neutron power at approximately 6% to 7X reactor power. It is possible for the actual CPC ASI value to-exceed the CPC Aux Trip value of +0.5 and cause a reactor trip when the CPC's begin using actual readings.
The COLSS calculation of ASI generally becomes accurate at approximately 4X reactor power. Therefore, it is necessary to determine ASI using COLSS to predict the value of CPC ASI prior to 6% 7X reactor power.
A COLSS ASI value with +0.35 to -0.35 gives reasonable assurance that the CPC calculated ASI will not exceed the CPC Aux Trip value when the CFC ASI calculation becomes valid t
(6%-7% Reactor Power).
If COLSS ASI is outside the limits of +0.35 to -0.35, a hand calculation should be performed to provide an estimate of CPC ASI.
If this hand calculation is within the range of +0.45 to -0.45, the CPC ASI should be within the CPC trip limits of +0.5 when Reactor Power reaches approximately 6X and the power ascension may continue.
With COLSS out of service, the only means to reliably monitor ASI is by using the hand calculation. However, due to the uncertainties in the estimation of ASI, the results of the hand calculation of ASI should be more restrictive. If the calculation is within +0.40 to -0.40, the CPC ASI should be within the CPC trip limits and the power ascension may continue.
Action should be taken to control ASI to within these administrative li,mits. These actions may include insertion/
withdrawal of Reg. group or Part-length CEA's or RCS boration/dilution.
To control ASI near EOL, variance from normal regulating group overlap may be required.
For cycle 1, below 75% power with COI SS in service l'55 power, COLSS out of service), the normal overlap between groups 4 and 5 may be modified for ASI control. To pre-vent an out of sequence trip, group 4 should be positioned at least 10 inches higher than group '5 and can be as hig as the normal overlap value.
~2lb400 ABr.e!~ Mg*
FOR lNFORMATlON ONLY PROCEDUR<
GENERATING NO. Appendix G STAT1ON MANUAL 410P-1ZZ04 Page 2...of 3 REVlSION PLANT STARTUP NODE 2 TO NODE 1 3 Page 27c of 27 LOW POWER ASI CONTROL When normal overlap is not maintained between groups 4 and 5, both groups of rods will need to be evaluated separately in LCO 3.1.3.6 to insure compliance with the transient insertion limit. The maximum insertion limit 'for group 4 is 60 inches withdrawn when the normal overlap is not maintained. When groups 4 and 5 are positioned closer together than normal overlap, group 4 becomes the'limiting group. For example, to allow a group 4 insertion to 85 inches withdrawn (group 5 and. 75" withdrawn), the power limit is approximately 47% (cycle 1, COLSS inservice).
Late'in core life. at low power level where the core is heavily top peaked, the most effective CEA configuration for forcing power to the bottom of the core is expected to be part lengths at 75 inches, group 5 at 75 inches and group 4 at 85 inches withdrawn.
Anytime the normal CEA overlap is not maintained and power is greater than 20%, the timing requirements of LCO 3.1.3.6 action b. shall be observed.
PV2lOCOO RKV.II% 5&A
~ 4 E ~
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t FOR INFORMATION ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO. Appendix G
,STATION MANUAL 410P-lZZ04 Page 3.of 3 REVISION A
PLA.'iT STARTUP .':ODE 2 TO NODE 1 Page 27d of 27 ..
\
2.0 If COLSS ASI exceeds the range of +0.35 to -0.35, perform the following to evaluate and correct ASI:
2.1 If COLSS ASI is outside the limits of +0.35 to -0.35, the following calculation should be performed to provide an estimate of CPC ASI:
CPC ASI = (CPC ID 012 - (CPC ID 010)
(CPC ID 012) + (CPC ID 010) 2.1.1 If the calculation is within the range of +0.45 to -0.45, the CPC ASI should be within the CPC trip limits of +0.5 when Reactor Power reaches approximately 6"...
2.1.2 If the calculation is outside the range of +0.45 to -0.45, actions should be taken to bring ASI to within this range.
3.0 1f COLSS ASI is unavailable, perform the following to evaluate and correct ASI:
F 3.1 Perform the calculation qj,2.+
3.1.1 I f the calculation is within the range of +0.40 to -0.40, the CPC ASI should be within the CPC trip limits when Reactor Power reaches approximately 6",o.
3.1.2 I f the calculation is outside the range of +0.40 to -0.40, actions should be taken to bring, ASI to within this range.
SIMBA 0'.5 OCO PEV libl
I l
" E "U~"EAR GENmmING PROCEDURE NO. APPENDIX D J
STATION 41OP-1ZZor Page'1 of 1 REVISION MANUAL'LANT STARTUP MODE 2 TO MODE 1 Page 25 of 27 r
Guidelines for Downcomer to Economizer Chan cover in Auto The downcomer to economizer swapover occurs at approx. 16.5 power on the contxol channels. then the transfer occurs Tcoldd will initially decrease approx. 3.5 F. It is desirable to .step thru the swapover in a smooth continuous manner in order to minimize the possibility of the Feedwatex control system hunting back and forth between the downcomer and economizer. It is also desirable to have sufficient reactivity available in the rods to stabilize temperature after the swapover occurs. Until fuxther modifications are accomp1ished to the FiCS, feed pump. speed should be in manual for the swapover.
Instructions 1.0 With power slightly below the swapover point and RCS temperature stabilized, smoothly increase turbine load while adjusting the rods or diluting to maintain xeactor power and temperature.
2.0 When the swapover occurs-continue-to increase turbine load slightly to ensure both steam generators swapover and remain on the economizer.
3.0 Verify steam bypass valves remain shut.
4.0 Allow sufficient time for RCS parametexs to stabilize.
5.0 Adjust Tcold ld as necessary.
~ '4 ~ ~
0 FOR INFORMATION ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO.
STATION MANUAL 41AO-1ZZ14 REVISlON EXCESSIVE RCS LEAKRATE Page 7 of 14 NOTE If. the leakage is greater than limits of the Tech Spec 3.4.5.2, it must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
F 11 Calculate RCS leakrate in accordance with Appendix B.
2.12 Place the VCT makeup mode selector switch in Auto.
2.13 If the RCS leakrate is greater than the capacity of the charging pumps and RCS pressure/inventory control is lost. TRIP the Reactor and proceed to 41EP-1ZZ01, Emergency Operations.
3.0 IDENTIFYING THE SOURCE OF THE RCS LEAKAGE NOTE If the leakage is greater than limits of the Tech Spec 3.4.5.2. It must be reduced to within limits within 4 hours.
CAUTION ENSURE ANY OPERATOR WHICH IS SENT TO LOCATE THE SOURCE OF RCS LEAKAGE IS WEARING A SCBA AND HAS A SURVEY METER IN POSSESSION..
3.1 Dispatch operators to do a walkdown of the charging and letdown piping to determine if leakage can be isolated.
3.2 In the control room monitor the Control Boards to assist in determination of leakage. Appendix C may be of assistance in determining abnormal parameters.
3.3 If RCS leakage cannot be isolated, perform Plant Shutdown per 410P-1ZZ05, Power Operations.
3.4 If RCS leakage has been isolated, stabilize RCS pressure/temperature and Pressurizer level and perform Leak Rate determination in accordance with Appendix A, and Steps 2.1 to 2.6.
pV216.000 AEV, 8/82 (485P)
I'l FOR INFORMATION ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO.
STATION MANUAL 40AC-9ZZ02 REVISION CONDUCT OF SHIFT OPERATIONS Page 8 of 48 4.1 ~ 2 The Unit 1 Shift Supervisor has the following additional responsibilities:
~ Function as the primary contact between PVNGS and Transmission Control Center (TCC) concerning operation of the 525 KV switchyard.
Keeping the other Units informed of operations concerning the 525 KV switchyard.
Coordinating activities concerning the Startup Transformer Yard.
~ Function as the primary contact between the Units and WRF concerning services supplied by the WRF.
~ Responsible for station security on backshifts, weekends, and holidays.
~ Make reports or notifications as detailed in this procedure for occurrences in- the common areas of the siCe.
4.1.3 The following administrative duties have been assigned to the Shift Supervisor, however, they may be delegated to another designated operator with the approval of the Unit Superintendent. The SS is to be kept informed of plant status and activities in progress'dditional administrative duties which do not detract from operational safety shall only be assigned with the approval of the Unit Superintendent.
Review and approve work orders Authorize and cancel clearances Authorize installation and removal of temporary modifications.
Review and concur with radiation exposure and radioactive effluent release permits.
Control of keys to vital areas.
Authorize hot work permits.
Review and concur with chemical control instructions.'uthorize Surveillance Tests per 73AC-9ZZ04.
>'N>6 COO REV. 8/82
FOR INFORMATlON ONLY PROCEDURE PALO VERDE NUCLEAR GENERATING NO.
STATION MANUAL 40AC-9ZZ02 BEY(SION CONDUCT OF SHIFT OPERATIONS Page 12 of 48 5.1.2 The Unit Superintendent may augment the normal shift complement as needed to support plant operating schedules.
NOTE Technical Specification 6.2 shall always take precedence.
5.2 Minimum Operation Shift Composition 5.2.1 The shift complement shall not be less than as specified in technical specifications table 6.2.1.
5.2.2 In addition to the requirements of Technical Specification Table 6.2-1 the following personnel requirements of specification 6.2 shall be met:
5.2.2.1 ALL CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
NOTE Fire Team composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to accommodate unexpected absence, provided immediate action is taken to required positions.
fill the 5.2.2.2 A site Fire Team of at least five members shall be maintained onsite at all times. The Site Fire Team will be manned by the Fire Protection Group and will respond as the need arises for fires inside the protected area. Operations will operate fire protection equipment only during response to an emergency situation or to protect equipment and/or area damage.
e PV2 f8 000 REV, 8/82
FOR INFORMATION ONLY PROCEDURE PALO YERDE NUCLEAR GENERATING NO.
STATION MANUAL 40AC-9ZZ02 REVISION CONDUCT OF SHIFT OPERATIONS Page 13 of 48 The Fire Team shall not include the Shift Supervisor, the STA, nor the (5) other members of the minimum
.shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency. Operations shall provide an advisor to the site fire team. The fire team advisor shall be competent to assess the potential safety consequences of the fire and advise Control Room personnel. 'The fire team advisor, as a minimum, shall be a ANSI 3.1 qualified NO'II certified in all areas.
5.3 'Minimum Control Room Manning
,NOTE The Control Room and the controls area are shown on Appendix G.
5.3.1 At, least one RO or SRO shall be in the controls area when fuel is in the vessel. The RO/SRO shall not leave the controls area without proper relief except during an emergency effecting plant safety which requires entering other areas within the CR to verify receipt of an annunciator or initiate corrective action, or when directed by the Shift Supervisor to evacuate to the Remote Shutdown Panel.
5.3.2 In addition to the operator at the controls an SRO shall be in the Control Room during modes 1 through 4 The SS
~
shall normally be in the Control Room and maintain the Control Room command function, however, during any absence of the SS from the Control Room during modes 1 through 4, a qualified SRO per T.S. Table 6.2-1 shall be designated to assume the Control Room command function. In modes 5 and 6 the individual designated to 'assume the Control Room command may be a qualified SRO or RO.
5.4 Nuclear Operators I and II Assigned Areas.
5.4.1 NO I and II areas of responsibility are assigned by the Unit Superintendent.
5.5 Overtime shall be limited in accordance with 10AC-OZZ07, Overtime Limitations and Technical specification 6.2.2.2.
5.6 Around-the-clock coverage is provided by six crews working rotating shifts.
PV218 000 REV. 8I82
I'

ML17303A769: Difference between revisions

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1.0 INTRODUCTION

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