ML050490484
| ML050490484 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 11/30/2004 |
| From: | Lanksbury R NRC/RGN-III/DRS/OLB |
| To: | |
| Shared Package | |
| ML050270138 | List: |
| References | |
| 50-440/04-301 50-440/04-301 | |
| Download: ML050490484 (92) | |
Text
FACILlTY POST-EXAMINAT10 N COMMENTS DATED DECEMBER 13,2004, AND JANUARY 6,2005 FOR THE PERRY INITIAL EXAMINATION - NOV/DEC 2004
FENOC FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant 10 Center Road RO. Box 97 Perry, Ohio 44081 December 13, 2004 PY-CEI/OI E-0628L United States Nuclear Regulatory Commission 2443 Warrenville Road, STE 210 Lisle, Illinois 60532-4352 Attention: Mr. Dell R. McNeil Division of Reactor Safety Perry Nuclear Power Plant Docket No. 50-440 NRC License Operator Exam for Class 03-01
Dear Mr. McNeil:
The Perry Nuclear Power Plant (PNPP) staff is respectfully requesting the Nuclear Regulatory Commission (NRC) staff to review Attachment 1 regarding specific issues that have been identified with select operator license exam questions and answers used in the most recent PNPP exam for Class 03-01.
Also, the PNPP staff requests that the operator license exam grades for Class 03-01 be officially released.
Please contact me at (440) 280-5056 if you have questions or require additional information.
Sincerely, J. Duffield Perry Training Section Manager
Attachment 1 PY-CEll0 IE-0628L Page 1 of 2 Question no. 1 -
2 correct answers. Answer A is also correct, the A pump could be downshifted to minimize loop flow mismatch.
Reference - Tech Spec 3.4.1 loop flow mismatch Helps Hayes, Lesiak, Pry, Weeks Question no. 2 -
2 correct answers. Answer A is also correct assuming all automatic actions occur. George Lesiak was not given the information that the bus EH12 stayed de-energized.
Helps Lesiak Question no.10-2 correct answers. Answer D is also correct because the radiation alarm requires entry into ONI-D17.
Reference ONI-D17 3.0 IMMEDIATE ACTIONS 3.1 EVACUATE the affected area Helps Hetrick, Pry, Slack, Jones, Rainey, Weeks Question no. 14 -
Toss out for Reactor Operators. This is beyond the scope of what is required knowledge.
Helps Jones, Rainey, Evans Question no.27-2 correct answers. Answer A is also correct per Tech Spec 3.6.3.2 Bases, which states one train can handle hydrogen generated from 75% of the fuel clad.
Reference TS 3.6.3.2 Helps Lesiak, Weeks, Rainey Question 34 2 correct answers. Answer D is also correct because both answers are parts of performing SPI 1.8
Reference:
PEI-SPI 1.8 Helps Evans Question 45 2 correct answers. Answer A is also correct because Drywell pressure has an effect on suppression pool level. This changes the pressure above SRV riser. Therefore, SRV temperature will be a range dependent upon Drywell pressure.
Helps Evans, Hayes Question no. 55 2 correct answers. Answer C is also correct because there will be a dp developed and the rod may possibly move due to the differences in surface areas of p under and p over.
Reference GEK - says possible cause can be inoperable directional control valve Helps Rainey Question no. 58 2 corrrect answers. Answer C is also correct because at a suppression pool temperature of 150" F, RHR A can be placed in suppression pool cooling.
Reference - SPI 3.2 directs the following:
3.7.3.1 OPEN RHR A(B) TEST VALVE TO SUPR POOL E l2-F024A(B).
Helps Weeks
Attachment 1 PY-CEI/OIE-0628L Page 2 of 2 Question no. 60 2 correct answers. Answer D could also be correct because it depends on the speed of the MSlV closure. If slow enough, Answer D would also be correct.
Helps Hayes, Jones, Kloosterman, Rainey, Lesiak, Jardine, Hetrick, Weeks Question no. 65 2 correct answers. Answer B is also correct because the surveillance covers 2 Tech Specs -
3.9.1 .Iand 3.9.2.2, making Answer B and C correct. The question states the surveillance fails not just the one rod out interlock.
Helps Jones, Weeks, Hetrick, Hayes, Pry Question no. 66 Throw out. These are certification requirements they receive prior to their first watch. No INPO ACAD requirements require this to be taught prior to receiving a license.
Helps Rainey, Jones, Evans, Weeks, Hayes, Pry, Lesiak Question 73 2 correct answers. Answer A is correct because the alarm response mode shall be given at the next brief. The answer does not state when the announcement should be made.
Helps Pry Question 74 2 correct answers. Answer C is correct because all positions on a shift crew can be fire brigade members. There is no reference that disqualifies any position.
Helps Jones Question 75 2 correct answers. Answer B is correct as stated in PAP-0528 as it directs Subsequent actions should be performed in order along with the site expectation that should is considered shall. The old procedure used to say the steps may be performed out of order. This statement has been removed.
Helps Lesiak, Pry, Slack, Jones, Rainey, Weeks Question 79 2 correct answers. Answer B is correct because minimum steam cooling water level with injection meets the definition of adequate core cooling.
Reference:
PEI Bases Definition of Adequate Core Cooling Helps Pry, Hetrick Question 86 2 correct answers. Answer D is correct because the pump may be required for adequate core cooling and should not be completely removed from service until it is known whether it is needed or not.
Helps Hayes Question 92 Throw out. The only correct answer is non-conservativeand would not be in the best interest of the health and safety of the public.
FENOC RrstEnergy Nuclear Operating Company Richard Anderson Vice President-Nuclear Perry Nuclear Power Plant 10 Center Road Perrx Ohio 44087 440-280-5579 Fax: 440-280-8029 January 6,2005 PY-CEI/OlE-O630L United States Nuclear Regulatory Commission 2443 Warrenville Road, STE 210 Lisle, Illinois 60532-4352 Attention: Mr. Dell R. McNeil Division of Reactor Safety Perry Nuclear Power Plant Docket No. 50-440 NRC License Operator Exam for Class 03-01
Dear Mr. McNeil:
By letter dated December 14, 2004, the Perry Nuclear Power Plant (PNPP) requested the Nuclear Regulatory Commission (NRC) staff review specific issues with select operator license exam questions and answers used in the most recent PNPP exam for Class 03-01.
In light of subsequent discussions concerning our letter, we are resubmitting our request with changes and additional supporting information.
As in our first letter, the PNPP staff also requests that the operator license exam grades for Class 03-01 be officially released.
If you have any questions or require additional information, please contact John Duffield, Manager Perry Training Section, at (440) 280-5056.
Sincerely, A
Attachment
- 1. Perry Initial License Exam Class 03-01 cc: Document Control Desk NRC Project Manager NRC Resident Inspector
Attachment 1 PY-CEI/OIE-O630L Page 1 of 87
, Perry Initial License,Exam Class 03-01 Note: NRC Answers are bolded. Proposed alternate answers are bolded and italicized.
QUESTION 001 Based upon further review by the utility, the appeal for this question is being withdrawn.
QUESTION 002 Given the following initial plant conditions:
- Mode 3 with a plant cooldown in progress following an extended high power run.
- RHR loop "B" is in Shutdown Cooling (SDC) mode
- Coolant temperature is 335°F
- RPV pressure is I10 psig Select the statement that describes the effect on the SDC Suction Isolation Inboard and Outboard Valves (1 E12-FO09 and IE12-FO08) if Bus EH12 (4.16 KV) trips:
a) 1EI2-FO08 and IE12-FO09 will shut.
b) lEl2-F008 and 1E12-Fob9 will NOT shut.
c) 1E12-FO08 will shut, lE12-FO09 will NOT shut.
d) IEI2-FO08 will NOT shut, 1 E12-FO09 will shut.
Comment:
2 correct answers - Mr. Lesiak indicated he did not receive information supplied by the proctor that bus EHI2 stayed de-energized and did not energize on the diesel supply. Answer A is also correct assuming the diesel re-energizes the bus and power is restored to the IE12F009 valve.
Reference:
ONI-R22-1, Loss of an essential 4160 volt bus Drawing 208-013 Sh 12, Nuclear Steam Shutoff System Licensee's Position:
The utility believes that there are two correct answers.
QUESTION 010 Based upon further review by the Utility, the appeal for this question is being withdrawn. Condition report 04-06852 was written to ensure the wording in the two procedures is made consistent.
Aiachment 1 PY-CEYOIE-0630L Page 2 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-R22-I
Title:
LOSS OF AN ESSENTIAL AND/OR A STUB Use Category:
In Field Reference Revision: Page I
4.16KV BUS Q2 e LOSS OF AN ESSENTIALAND/OR A STUB 4.16KV BUS Effective Date: 9-3-04 Preparer: Tracey L. Rose I 8-20-04 Date
Attachment 1 PY-CEI/OIE-OG3OT, Page 3 of 87
~ ~~ ~~ ~~ ~ ~ ~~~ ~-
Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-R22-1
Title:
Use Category:
LOSS OF AN ESSENTIAL AND/OR A STUB In Field Reference Revision: Page 4.16KV BUS 5 4of9 Auxiliary and Startup Power Section, Lonq Response Benchboard, IH I R-PR7n INTERBUS XFMR LH-1-A LOCKOUT RELAY I.2 Parameters The changes in plant parameters vary depending upon the bus lost, the reactor power level at the time of bus failure, and the components presently being supplied from the bus.
2.0 AUTOMATIC ACTIONS 2.1 Emergency Diesel Generator for the de-energized EH Bus receives an auto start signal.
2;2 Standby units start as operating equipment is lost.
2.3 IF EHI 1 AND EH12 are de-energized THEN the following startheposition when power is regained:
MCC, SWGR, and Misc. Elect Equip Area HVAC System, M23/M24 Control Room HVAC and Emergency Recirculation System, M25/M26 Makeup Water Pretreatment System, P20 Service Water System, P41 Emergency Closed Cooling System, P42 Nuclear Closed Cooling System, P43 Emergency Service Water System, P45 Control Complex Chilled Water System, P47 2.4 An NSSSS isolation signal is generated to Div 1 (Div 2) components due to de-energization of NSSSS relays when Bus EHI 1(EH12) is lost.
.4ttachment I PY-CELOIE-O63i\L Page 4 of S7 I PERRY NUCLEAR POWER PLANT Instruction Number:
ONI-R22-1 I I
Title:
LOSS OF AN ESSENTIAL AND/OR A STUB 4.1 6KV BUS Use Category:
In Field Reference Revision: Page 5 5 of 9 3.0 IMMEDIATE ACTIONS NA 3.1 IF the loss of an essential 4.16KV Bus was due to a loss of offsite power, 0
THEN REFER TO ONI-RIO, LOSSof AC Power.
4.0 SUPPLEMENTAL ACTIONS NOTE Protective relays and flags should be recorded and the cause of the trip should be determined prior to resetting tripped breakers.
4.1 REFER TO the following instructions concurrently with this instruction:
0 0 ONI-C11-1, inability to Move Control Rods ONI-E12-2, Loss of Decay Heat Removal.
0 ONI-P41, Loss of Service Water.
0 0 ONLP43, Loss of Nuclear Closed Cooling.
0 0 ONLB21-4, Isolation Restoration.
0 4.2 REFER TO EPI-AI and DETERMINE if an Emergency Action Levels have been exceeded.
Attachment 1 PY -CEI/OIE-O630L Page 5 of 87 instruction Number:
I
Title:
LOSS OF AN ESSENTIAL AND/OR A STUB 4.16KV BUS Use Category:
In Field Reference Revision: Page 5 6of9
' NOTE Divisional valves receiving an isolation signal from the NSSS System will reposition to the isolated position as soon as power is restored.
NA 4.3 An EH Bus is de-energized The Diesel Generator failed to start lnterbus Transformer LH-2-A is available LH-2-A is the preferred option to restore power to the bus quickly.
I THEN PERFORM the following:
0 4.3.1.a VERIFY the PREFERRED SOURCE EH1114 EH1212 EH1303 BRKR is open 0 4.3.1.b CLOSE the ALTN PREFERRED EH1115 EH1213 EH1302 SOURCE BRKR.
NA 4.4 An EH Bus is de-energized lnterbus Transformer LH-1-A is available LH-I-A is the preferred option to restore power to the bus quickly.
ITHEN PERFORM the following: I 0 4.4.1.a VERIFY the ALTN PREFERRED EH1115 EH1213 EH1302 SOURCE BRKR is open.
a 4.4.1.b CLOSE the PREFERRED SOURCE EH1114 EH1212 EH1303 BRKR.
Attaclment 1 PY-CFLOTE-06301, Page 6 of 87 s:
I
Attachment 1 PY-CEI/OIE-O63 OL Page 7 of 87 QUESTION 014 Based upon further review by the Utility, the appeal for this question is being withdrawn. The initial request was to remove this question based on being beyond the JTA requirements for RO's.
Feedback from the Lead Examiner indicates this was a 3.8 WA requirement for the RO. Utility review of the WA Catalog, assuming this was a knowledge under 230000 A4.14, makes this appeal basis invalid.
QUESTION 027 A LOCA has occurred resulting in significant Hydrogen generation. One division of Hydrogen Igniters is in operation and one Combustible Gas Mixing Compressor is operating. Both Hydrogen Recombiners are shutdown due to Hydrogen concentration exceeding 6% in containment. Hydrogen concentration is continuing to increase. Which one of the following statements best explains why Hydrogen concentration is continuing to increase?
a) Hydrogen generation has exceeded fhe operational capability of the one division of Hydrogen Igniters that are in service, b) A continuing increase in hydrogen concentration is indicative of a steam inert or Oxygen starved environment.
c) Hydrogen concentration will continue to increase until the Hydrogen Igniters reach their operating temperature which can take several hours.
d) The indicated increase must be due to a malfunction of the Hydrogen Analyzer since actual concentration cannot exceed 6% as long as the Hydrogen Igniters are in operation.
Commenf:
2 correct answers - answer A could also be correct, Tech Spec Bases 3.6.3.2 states that "the H2 igniters are installed to accommodate the amount of hydrogen equivalent to that generated from the reaction of 75% of the fuel cladding with water". The question stem provides the information that only one train of H2 igniters is in operation and that there is "significant" H2 generation. No specifics are provided as to the actual amount-of H2 generated, and % of the required igniters are not functioning.
The trainee could infer that the potential exists that it has exceeded the capability of the single H2 igniter train in operation, before an inert environment has occurred.
Reference TS 3.6.3.2.bases Licensee's Position:
The utility believes that there are two correct answers.
Primary Containment and Drywell Hydrogen Igniters B 3.6.3.2 Attachment 1 PY-CEVOE-O63OL B 3.6 CONTAINMENT SYSTEMS Page 8 of 87 B 3.6.3.2 Primary' Containment and Drywell Hydrogen Igniters BAS ES BACKGROUND The primary containment and drywell hydrogen igniters are a part of the combustible gas control required by 1 0 CFR 50.44 (Ref. 1) and GDC 41. "Containment Atmosphere Cleanu (Ref. 21, t o reduce the hydrogen concentration i n t e rimary containment following a degraded core accident. The f: 'I 1ydrogen igniters ensure the combustion of hydrogen i n a manner such t h a t containment overpressure fai 1ure is prevented as a result of a postulated degraded core accident.
10 CFR 50.44 (Ref. 1 ) requires boiling water reactor units w i t h Mark 111 containments t o install suitable hydrogen control systems. The hydrogen igniters are installed t o Q 27 accommodate an amount of hydrogen equivalent t o t h a t generated from the reaction of 75% of the fuel claddins w i t h water. This requirement was placed on reactor units w i t h Mark 111 containments because they were not designed for inerting and because of their low design pressure.
Calculations indicate t h a t i f hydrogen equivalent t o t h a t generated from the reaction o f 75% of the fuel cladding w i t h water were t o collect i n primary containment, the resulting hydrogen concentration would be far above the lower fl ammabi 1i t y 1imi t such t h a t , without the hydrogen igniters ,
i f the hydrogen were ignited from a random ignition source, the resulting hydrogen burn would seriously challenge the primary containment.
The hydrogen igniters are based on the conce t of controlled E
ignition using thermal igniters designed t o e capable of functioning i n a post accident environment, seismically supported and capable of actuation from the control room.
Hydrogen igniters are distributed throughout the drywell and primary containment i n which hydrogen could be released or t o which i t could flow i n siqnificant auantities. The hydrogen igniters are arranged i n two i ndependent d i vi si ons such t h a t each containment region has two igniters, one from each division, controlled and powered redundantly so t h a t ignition would occur i n each region even i f one division failed t o energize.
(cont i nued)
PERRY - UNIT 1 B 3.6-95 Revision No. 1
Primary Containment and Drywell Hydrogen Igniters B 3.6.3.2 Attachment 1 PY-CEI/OIE-O63OL BASES Page 9 of 87 BACKGROUND When the hydrogen igniters are energized they heat up t o a (continued) surface temperature 2 1700°F. A t this temperature, they ignite the hydrogen gas t h a t is present i n the airspace i n the vicinity of the igniter. The hydrogen igniters de end on the dispersed location of the igniters so that loca ockets of hydrogen a t increased concentrations would burn 7
1 efore reaching a hydrogen concentration significantly higher than the 1ower f l ammabi 1i t y 1imi t .
AP PL I CABLE The hydrogen igniters cause hydrogen i n containment t o burn SAFETY ANALYSES i n a controlled manner as i t accumulates following a degraded core accident (Ref. 3 ) . Burning occurs a t the lower flammability concentration. where the resulting temperatures and pressures are relatively benign. Without the hydrogen igniters, hydrogen could build up t o higher concentrations t h a t could result i n a violent reaction i f ignited by a random ignition source after such a buildup.
The hydrogen igniters are not included for mitigation of a Design Basis Accident (DBA) because an amount of hydrogen e equivalent t o that generated from the reaction o f 75% of the fuel cladding with water is far i n excess of the hydrogen calculated for the limiting DBA loss of coolant accident (LOCAL The hydrogen concentration resulting from a DBA can The hydrogen igniters are considered t o be risk significant in accordance with the NRC Po7 icy Statement.
LCO Two di vi si ons of primary containment and drywell hydrogen igniters must be OPERABLE, each w i t h 90% or more o f the igniters OPERABLE (i .e., no more t h a n five igniters i noperabl e. 1 (conti wed)
PERRY - UNIT 1 B 3.6-96 Revision No. 1
Primary Containment and Drywell Hydrogen Igniters B 3.6.3.2 Attachment 1 BASES PY-CEI/OIE-O630L Page 10 of 87 LCO This ensures operation of a t least one hydrogen igniter (continued) division, w i t h adequate coverage o f the primary containment and drywell. i n the event of a worst case single active failure. This wi 11 ensure t h a t the hydrogen concentration remains near 4.0 v/o.
APPL ICABI LIN In MODES 1 and 2 , the hydrogen igniter is required t o control hydrogen concentration t o near the fl ammabi 1i t y limit o f 4.0 v/o following a degraded core event t h a t would generate hydrogen i n amounts equivalent t o a metal water reaction of 75% o f the core cladding. The control o f hydrogen concentration revents overpressuri zati on of the primary contai nrnent . Tie event t h a t could generate hydrogen i n quantities sufficiently high enough t o exceed the flammability limit is limited t o MODES 1 and 2.
In MODE 3, both the hydrogen production rate and the total hydrogen produced after a degraded core accident would be less t h a n t h a t calculated for the DBA LOCA. Also, because o f the limited time i n this MODE. the probability of an accident requiring the hydrogen igniter is low. Therefore, the hydrogen igniter is not required i n MODE 3.
In MODES 4 and 5 . the probability and consequences of a degraded core accident are reduced due t o the pressure and temperature 1imitations . Therefore, the hydrogen igniters are not required t o be OPERABLE i n MODES 4 and 5 t o control hydrogen.
ACTIONS -
A. 1 With one hydrogen igniter division inoperable, the ino erable division must be restored to OPERABLE status R
wit i n 30 days. In this Condition, the remaining OPERABLE hydrosen iciniter division is adequate t o Derform the hydrogen burn function. However, the overall re1 i a b i 1i t y is reduced because _a sincrle failure i n the OPERABLE subsystem l y . The 30 day Completion Time is based on the low probability o f the occurrence o f a degraded core event t h a t would generate hydrogen i n amounts equivalent t o a metal water reaction o f 75% of the core cladding. the amount of time available after the event for operator action t o prevent hydrogen (cont i nued 1 PERRY - UNIT 1 B 3.6-97 Revision No. 1
Attachment 1 PY-CEUOIE-0630L Page 11 of 87 QUESTION 034 The plant was operating at 100% reactor power when the plant experienced an earthquake. A medium break LOCA occurred and RPV'Level2 was reached. All ECCS systems responded correctly. RPV level is currently 180 inches and slowly increasing. The reactor failed to scram and all efforts to manually insert control rods have failed. Standby liquid control has failed to correctly initiate and shut down the reactor {failed SQUIBB valves). The Unit Supervisor has decided to initiate Alternate Boron Injection (ABI) in accordance with PEI-SPI I.8. What needs to be done in order to successfully initiate Alternate Boron Injection?
a) Secure HPCS.
b) Close E22-FO04 (HPCS Injection Valve) c) Secure both SLC pumps d) Connect a low pressure hose from the SLC storage tank to fhe sucfion of the AB1 pump; start the A51 pump, open the A51 pump discharge valve.
Comment:
2 correct answers - answer D is also correct. Both answers are parts of performing SPI 1.8 Verbal discussions with the lead examiner indicates that there is no correct answer and that both of the above are partial answers. The question will be removed from the test. The utility concurs with this decision.
Reference:
PEI-SPI 1.8 Licensee's Position:
The utility agrees with the NRC position on no correct answers.
QUESTION 045 Based upon further review by the Utility, the appeal for this question is being withdrawn.
Attachment 1 PY-CEI/OIE-063OL Page 12 of 87 QUESTION 055 While attempting to insert a control rod, the operator depresses the INSERT pushbutton and observes the following:
- No rod motion
- CRD DRIVE WATER HEADER FLOW at 0 gpm
- CRD COOLING WATER FLOW at 60 gpm Which ONE of the following is the possible cause of these indications?
a) CRD Flow Control Valve failed closed.
b) Associated drive water stabilizing valves failed closed.
c) Associafed Insert Exhaust Directional Confro! Valve (DCV 927) failed closed.
d) Associated Insert Drive Directional Control Valve (DCV 123) failed closed.
Comment:
The two correct answers was based upon the fact that there may be no delta p developed without the exhaust valve being open. Although the rod may move due to the differences in surface areas of p under and p over and leakage through the seals, the potential exists that no flow will be developed with the exhaust valve closed, as there would be no flow path. This would make answer C also correct.
Reference:
GEK 755988 Licensee's Position:
The utility believes there are two correct answers.
~~ ~~
Attachment 1 PY-CEUOIE-0630L 45 - r Page 13 of87 GEK-75598B A- In F i g u r e 4-17, a waveform i s i l l u s t r a t e d o f notch-out from p o s i t i o n "24" w i t h flow t h r o u g h v a l v e 120 r e s t r i c t e d t o 0.5 gpm (1.89 !L/min) less t h a n normal. Note t h a t :
- 1. CRD notch-out operation is successful.
- 2. An i n c r e a s e i n drive-down d P o c c u r r e d .
- 3. The s e t t l e p e r i o d is e x t e n d e d .
- 4. A s l i g h t loss o f r e s p o n s e o c c u r r e d a t t h e commencement of t h e CRD settle period.
BO I n F i g u r e 4-18, a waveform i s i l l u s t r a t e d of notch-out from p o s i t i o n "24" with f a i l u r e of v a l v e 120 t o a c t u a t e . Note t h a t :
- 1. No notch-out CRD o p e r a t i o n o c c u r r e d .
- 2. Drive-down'dP i s i n c r e a s e d .
- 3. A J O S S of r e s p o n s e o c c u r r e d a t t h e commencement of t h e s e t t l e period.
- 4. No s e t t l e p e r i o d i s i n d i c a t e d .
4l4-76 O s c i l l o s c o p e waveform t r a c e s s h o w i n g e a k a g e a n d f l o w r e s t r i c t i o n a t d i r e c t i o n a l c o n t r o l v a l v e 1 2 1 a r e o b t a i n e d by applyjng n o t c h - i n and notch-out s i g n a l s t o t h e CRD from p o s i t i o n "24" and o b t a i n i n g a waveform trace p h o t o g r a p h of t h e dP a c r o s s t h e CRD p o s i t j o n .
4-77 Leakage. Leakage o f 1.5 gpm (5.7 l?,/min) a c r o s s valve 1 2 1 d u r i n g a notch-out o p e r a t i o n r e s u l t s i n a drive-down d P d e c r e a s e . Leakage i n e x c e s s of 1.5 gprn (5.7 R/min) may r e s u l t i n f a i l u r e of the CRD t o notch-out b e c a u s e of l o s s e s i n drive-down p r e s s u r e and p r e s s u r e t o unlock t h e CRD c o l l e t . Refer t o F i g u r e s 4-19, 4-20, and 4-21. Compare t h e s e and t h e p h o t o g r a p h o b t a i n e d w i t h r e f e r e n c e F i g u r e 4-6.
A. I n F i g u r e 4-19, a waveform i s i l l u s t r a t e d of notch-out from p o s i t i o n "24" w i t h 1.5 gpm (5.7 J?/min) l e a k a g e a c r o s s v a l v e 121. Note t h a t :
- 1. The CRD notch-out is successful.
- 2. A d e c r e a s e o c c u r s i n drive-down dP.
- 3. A s l i g h t g a i n o c c u r s i n s e t t l e dP.
- 4. S e t t l e time i s i n c r e a s e d .
4-43
Attachment 1 PY-CEI/OIE-O630L Page 14 of 87 GEK-7 5 5 98B
/ Figure 4-19. Waveform Trace: Notch-Out from P o s i t i o n 24 with i
1.5 gpm (5.7 R/min> Leakage Across Valve 121 Figure 4-20. Waveform Trace: Notch-Out from Position 24 with 3 gpm (11.4 R/min) Leakage Across Valve 121 4-44
Attachment 1 PY-CEIlOLE-0630L Page 15 of 87 GEK-75598B I
PO> PU PO> PU PU> PO TIME ). .
(0.5 SEC/DIV)
Figure 4-21. Vaveform Trace: Notch-Out from P o s i t i o n 24 w i t h Valve 1 2 1 F a i l e d Full-Open B. I n Figure 4-20, a waveform i s i l l u s t r a t e d o f notch-out from p o s i t i o n "24" with 3 gpm (11.4 R/min) leakage a c r o s s valve 121. Note t h a t :
- 1. The CRD notch-out is successful.
- 2. A loss in drive-down dP occurs.
- 3. S e t t l e t i m e i s extended.
C. I n Figure 4-21, a waveform is i l l u t r a t e d of notch-out from p o s i t i o n "24" w i t h valve 1 2 1 f a i l e d i n the f u l l - o p e n p o s i t i o n . Note t h a t :
- 1. The CRD f a i l e d t o notch-out.
- 2. A l o s s i n drive-down dP occurs.
- 3. No s e t t l e period i s i n d i c a t e d .
4-4 5
Attachment 1 PY-CEI/OIE-0 63OL Page 16 of 87 GEK-75598B G1 A. In Figure 4-22, a waveform i s i l l u s t r a t e d of notch-in p o s i t i o n "24" w i t h 1.4 gpm (5.3 &/mid f l o w through valve 121. Note that:
- 1. CRD notch-in i s s u c c e s s f u l .
- 2. A s l i g h t loss i n drive-in dP occurs.
- 3. Overall t h e waveform trace a p p e a r s normal (see reference Figure 4-51.
B. In F i g u r e 4-23, a waveform i s i l l u s t r a t e d of notch-in with 0.7 gpm (2.65 k/min) flow through valve 121. Note t h a t :
1.. The CRD notch-in i s s u c c e s s f u l .
- 2. Drive-in dP i s low.
c 3. Drive-in dP e q u a l s s e t t l e dP.
C. In F i g u r e 4-24, a waveform is i l l u s t r a t e d of notch-in from p o s i t i o n "24" w i t h valve 1 2 1 failed-closed. Note t h a t :
- 1. CRD f a i l e d t o notch-in.
- 2. Drive-in dP is erratic and reversed (i.e., PO g r e a t e r than PU).
- 3. The s e t t l e p e r i o d is extended.
4-79 DIRECTIONAL CONTROL VALVE123. Measurement of dP a t t h e HCU is n o t r e q u i r e d f o r t h e d e t e c t i o n of leakage through valve 123. O s c i l l o s c o p e waveform traces showing f l o w r e s t r i c t i o n a t valve 123 are o b t a i n e d by applying notch-in signals t o t h e CRD.
4-80 Leakage. Leakage through v a l v e 123 can result i n CRD d r i f t - i n when the CRD i s a t a l a t c h e d p o s i t i o n and n o movement s i g n a l i s a p p l i e d . The CRD d r i f t r a t e is dependent upon t h e r a t e of l e a k a g e through v a l v e 123 ( i . e . , t h e g r e a t e r t h e l e a k a g e through valve 123, t h e h i g h e r t h e CRD d r i f t rate). Should valve 123 f a i l i n t h e full-open p o s i t i o n , t h e CRD w i l l d r i f t i n a t approximately i t s normal speed of 3 i n . / s e c (76 mm/sec).
4-46
Attachment 1 PY-CEUOIE-0630L Page 17 of 87 GEK-75598 B F i g u r e 4-22. Waveform Trace: Notch-In from Position 24 with f l o w of 1.4 gpm (5.3 R/min) through Valve 122 rg>W pD>w
- nqcd 10.5 7.0 1.5 0 UP 1.s 7.0 10.5 F i . g u r e 4-23. kaveform Trace: Notch-In from Position 24 w i t h 0.7 gpm (2.65 R/min> f l o w T h r o u g h Valve 1 2 1 I
4-4 7
' . Attachment 1 PY-CEUOIE-063 OL Page 18 of 87 GEK-75 59 8B PO> Pu PO> PU PU>PO PU> PO TIME *
(0.5 SEC/DIV)
Figure 4-24. Waveform Trace: Notch-In from P o s i t i o n 24 w i t h Valve 1 2 1 Failed-Closed 4-81 Flow R e s t r i c t i o n . With flow through v a l v e 1 2 3 r e s t r i c t e d t o 3 gpm ( 1 1 . 4
&/min), t h e CRD f a i l s t o s a t i s f a c t o r i l y complete t h e notch-in o p e r a t i o n .
Simiiarly, no CRD movement is apparent w i t h f l o w through v a l v e 123 r e s t r i c t e d t o 0.5 gpm (1.89 R/min). Refer t o F i g u r e s 4-25 and 4 - 2 6 . Compare these and t h e photograph o b t a i n e d w i t h r e f e r e n c e F i g u r e 4 - 6 .
A. In F i g u r e 4 - 2 5 , a waveform is i l l u s t r a t e d of notch-in from p o s i t i o n "24" w i t h flow through valve 1 2 3 r e s t r i c t e d t o a maximum of 3 . 2 gpm (12.1 R/min). Note t h a t :
- 1. The CRD f a i l e d t o notch-in.
- 2. Drive-in dP is normal.
3- The s e t t l e period is extended.
4-48
Attachment 1 PY-CEVOIE-0630L Page 19 of 87 should maintain drive temperatures below 250°F. The cooling water, see Figure 3 1, is supplied to the drive via the insert header and the insert port. The flow of the cooling water is upward through the strainer between the outer tube and thermal sleeve via a set screw plug orifice (23) located in the main flange. The flow continues upward through the outer screen and into the reactor.
Normal cooling water header pressure is equal to reactor pressure plus 20 psi, and the normal cooling water flow rate through each drive mechanism is approximately .20 - .34 gpm. Utilize Figure 35 for a simplified flow path representation of this section.
- b. Insert Function On a rod insert signal, Figure 32, water from the drive header enters the insert port and is routed to the underside of the drive piston. Simultaneously, water from above the drive piston is exhausted through the flow ports (69) in the buffer shaft, down between the piston tube and indicator tube, and out the withdraw port.
Unlocking the collet fingers is not required for CRDM insertion.
The collet fingers are forced out of the locking notch as the index tube moves upward. The fmgers grip the outside wall of the index tube and snap into the next lower locking notch for single notch insertion to hold the index tube in position. Utilize Figure 35 for a simplified flow path representation of this section.
C. Withdraw Functions The collet locking mechanism requires a hydraulic pressure greater than reactor vessel pressure to unlock the fingers for CRDM C1 l(CRDM) - Rev. 5 34
Attachment 1 PY-CEI/OIE-O63OL Page 20 of 87 Hi t h d P
FIGURE Cll(CRDM)-32 ROD INSERTION FLOW PATH C1 l(CRDM) - Rev. 5 85
Attachment 1 PY-CEI/OIE-O630L Page 21 of 87 CHARGING 1 I i I SCRAM DISCHARGE I I EXHAUST I I I I l I COOLING 1
DRIVE I I 1 I
I I
I 1 i i 105 112 i
114 if- OUTLET VALVE 7 VALVE TYPICAL OF 177 DRAIN FIGURE C11(CRDH)-6 HYDRAULIC CONTROL UNIT PIPING DIAGRAM C11(CRDH), Rev. 7 47
Attachment 1 PY-CEI/OLE-O63OL Page 22 of 87 QUESTION 058 A Main Steam line break (18 minutes ago) has resulted in the following plant conditions:
m lines with LPCS due to exceeding RPV the Containment Spray Flow Valve (F064A) did NOT open.
toff Valve (F028A)
Open the LPCl A Injection Valve (F042A)
Open the RHR A Test Valve to Supp. Pool (F024A)
Shutdown RHR Pump A Comment:
rs. Due to, suppression p perature of150 degrees F, if PEI-SPI 3& you may go ppression poolcooling on the -- - ---
p available, the SRO rrect, as opening the erating without minimum flow protection. .-
Reference:
PEl-Bases SPI 3.1 Licensee's Position:
The utility believes' there are two correct answers QUESTION 060 Based upon further review by the Utility, the appeal for this question is being withdrawn.
Procedure Number:
PERRY NUCLEAR POWER PLANT PEI Bases
Title:
Use Category:
PEI Bases Document Reference PELT23 Containment Control Revision: Page Suppression Pool Temperature Control 6 314 of 392 STEP:
required for adequate core cooling, operate a available Suppression Pool cooling DISCUSSION When suppression pool temperature cannot be maintained below the most limiting suppression pool temperature LCO value (95"F), explicit instructions are given to operate all available methods of suppression pool cooling.
Maintaining adequate core cooling takes precedence over maintaining suppression pool temperature below the LCO value since catastrophic failure of the containment is not expected to occur at this temperature. In addition, further action is still available for reversing the increasing suppression pool temperature trend. Therefore, only if the operation of a RHR pump is not required to assure adequate core cooling is it permissible to use that pump for suppression pool cooling. This step however, does permit alternating the use of RHR pumps between the RPV injection mode and suppression pool cooling modes, as the need for each occurs, and so long as adequate core cooling can be maintained.
Attachment 1 PY-CEI/OIE-O630L Page 24 of87 PEI-SPI 3.1 Page: i Rev.: 0 The Cleveland Electric Illuminating Company PERRY OPERATIONS MANUAL Plant Emergency Instruction 4 s e TITLE: SPECIAL PLANT INSTRUCTION 3.1 CONTAINMENT SPRAY OPERATION REVISION: 0 EFFECTIVE DATE : 8-19-94 PREPARED: PEI Improvement Team 8-16-94
/ Date EFFECTIVE PICS
Attachment 1 PY-CEUOIE-0630L Page 25 of 87 PEI-SPI 3 .1 Page: 1 of 9 Rev.: 0 PEI-SPI 3.1 Containment Spray Operation ENTRY CONDITIONS This instruction is entered when Containment pressure or hydrogen concentration necessitate spraying of Containment, or when Containment sprays are to be terminated.
SCOPE This instruction provides the necessary actions to manually initiate Containment sprays both before and after the RHR Containment Spray Loop has been used to vent Containment. The high Drywell pressure interlock is bypassed, if necessary, to allow spray initiation. Steps are also provided to realign Containment Spray to RPV injection or Suppression Pool Cooling. RPV injection may be routed inside or outside the shroud.
NECESSARY EOUIPMENT Control Room PEI-SPI File Cabinet:
- four PEI-SPI keys CC 599' D/01, OSC PEI File Cabinet:
- one green locking tab IB 599' K / 0 5 :
- one 12 ft ladder LOCATION OF REQUIRED LOCAL ACTIONS AX 599,, above the RHR A(B) HX Room door:
- C / O 7 (C/03), RHR A (B) FPCC Supplement Cooling Discharge Vlv lE12-F099A(B)
(CONTINUED ON NEXT PAGE)
Attachment 1 PY-CEI/OIE-O630L Page 26 of87 PEI-SPI 3.1 Page: 2 of 9 Rev.: 0 PEI-SPI 3.1 Containment Spray Operation (Continued) 1.0 IF RHR Containment Spray Loop A(B) is NOT lined up to vent Containment,
-THEN INITIATE Containment Spray Loop A(B) as follows:
1.1 -IF a high Drywell pressure LOCA signal is present, THEN PROCEED TO Step 1.3 of this instruction.
1.2 AT H13-P629 (P618),
PLACE CNTMT SPRAY A(B) HI DW PRESS BYP E12A-S75(S76) keylock switch in BYPASS.
NOTE RHR Loop B Containment Spray manual initiation pushbutton must be depressed for at least 35 seconds to 'allow the signal to seal in.
I I 1.3 ARM and DEPRESS CNTMT SPRAY A(B) MANUAL INITIATION E12A-S63A (B) pushbutton.
1.4 V E R I M RHR PUMP A(B) E12-C002A(B) is running.
1.5 VERIFY ESW PUMP A(B) P45-C001A(B) is running.
1.6 VERIFY ECC PUMP A(B) P42-C001A(B) is running.
1.7 VERIFY the following valves are open:
CNTMT SPRAY A(B) FIRST SHUTOFF E12-F028A(B)
CNTMT SPRAY A(B) SECOND SHUTOFF E12-F537A(B)
(CONTINUED ON NEXT PAGE)
Attachment 1 PY-CEYOIE-0630L Page 27 of 87 PEI-SPI 3 -1 Page: 3 of 9 Rev. : 0 PEI-SPI 3.1 Containment Spray Operation (Continued)
NOTE RHR A ( B ) HX'S BYPASS VALVE E 1 2 - F 0 4 8 A ( B ) will not close within ten minutes after a LOCA initiation signal.
I I NOTE RHR A ( B ) H X ' S BYPASS VALVE E 1 2 - F 0 4 8 A ' ( B ) will only automatically close if RHR A ( B ) HX'S INLET VALVE E 1 2 - F 0 4 7 A ( B ) and RHR A ( B ) HX'S OUTLET VALVE E12-FO03A ( B ) are open.
1.8 VERIFY the following valves are closed:
L P C I A ( B ) I N J E C T I O N VALVE E l Z - F 0 4 2 A ( B )
RHR A ( B ) TEST VALVE TO SUPR POOL E l Z - F 0 2 4 A [ B )
RHR A ( B ) HX'S BYPASS VALVE E 1 2 - F 0 4 8 A ( B )
SHUTDOWN COOLING A ( B ) TO FDW SHUTOFF E 1 2 - F 0 5 3 A ( B )
(CONTINUED ON NEXT PAGE)
Attachment 1 PY-CEI/OIE-O63OL Page 28 of 87 PEI-SPI 3 .1 Page: 4 of 9 Rev.: 0 PEI-SPI 3.1 Containment Spray Operation (Continued)
- CAUTION *
- System realignment other than as directed in the following step *
- may result in water hammer severe enough to cause pressure *
- boundary failure and subsequent uncontrolled release to the *
- environment.
- 2.0 -
I F RHR Containment Spray Loop A(B) is lined up to vent Containment OR Containment is currently being vented via RHR Containment Spray LOOP A(B) r THEN COMMENCE Containment Spray with RHR Loop A(B) as follows:
2.1 IF Containment is currently being vented via Containment Spray Loop A(B)
THEN CLOSE CNTMT SPRAY A(B) FIRST SHUTOFF ElZ-F028A(B) to secure venting.
2.2 FILL the drained RHR A(B) piping as follows:
NOTE RHR A(B) HX'S OUTLET VALVE E12-F003A(B) is throttled open to allow the water leg pump to fill up the drained portion of the RHR header.
2.2.1 THROTTLE open RHR A(B) HX'S OUTLET QALVE E12-F003A(B) to obtain 4-6% open.
2.2.2 AT AX 599' C/07(C/03), above RHR A(B) HX Room door, CLOSE and LOCK RHR A(B) FPCC Supplement Cooling Discharge Vlv 1E12-F099A(B).
2.2.3 VERIFY the following valves are open:
CNTMT SPRAY A(B) FIRST SHUTOFF ElZ-F028A(B)
CNTMT SPRAY A(B) SECOND SHUTOFF E12-F537A(B)
(CONTINUED ON NEXT PAGE)
Attachment 1 PY-CEIlOIE-0630L Page 29 of 87 PEI-SPI 3.1 Page: 5 of 9 Rev.: 0 PEI-SPI 3.1 Containment Spray Operation (Continued) 2.2.4 VERIFY t h e f o l l o w i n g v a l v e s a r e c l o s e d :
LPCI A ( B ) INJECTION VALVE E 1 2 - F 0 4 2 A ( B )
RHR A ( B ) HX'S BYPASS VALVE E 1 2 - F 0 4 8 A ( B )
RHR A ( B ) HX'S INLET VALVE E 1 2 - F 0 4 7 A ( B )
2.2.5 VERIFY ESW PUMP A ( B ) P45-C001A(B) is running 2.2.6 VERIFY ECC PUMP A ( B ) P42-C001A(B) is running.
2.2.7 START RHR PUMP A ( B ) E 1 2 - C 0 0 2 A ( B ) .
2.2.8 OPEN RHR A(B) H X ' S INLET VALVE E 1 2 - F 0 4 7 A ( B ) .
2.2.9 WHEN f l o w i s p r e s e n t a s i n d i c a t e d on RHR A ( B ) PUMP FLOW E12-R603A ( B ) ,
-THEN THROTTLE RHR A ( B ) HX'S OUTLET VALVE E l 2 - F 0 0 3 A ( B ) s l o w l y t o o b t a i n a f l o w r a t e of 1 0 0 0 - 1 5 0 0 gpm.
2.2.10 WHEN a p p r o x i m a t e l y f o u r m i n u t e s have passed, THEN OPEN RHR A ( B ) HX'S OUTLET VALVE E 1 2 - F 0 0 3 A ( B ) .
(CONTINUED ON NEXT PAGE)
Attachment 1 PY-CEUOIE-0630L Page 30 of 87 PEI-SPI 3 - 1 Page: 6 of 9 Rev.: 0
@ee PEI-SPI 3.1 Containment Spray O p e r a t i o n (Continued) 3.0 TERMINATE RHR Containment S p r a y Loop A ( B ) a s f o l l o w s :
NOTE L P C I A ( B ) I N J E C T I O N VALVE E12-FO42A ( B ) c o n t r o l s w i t c h i s o v e r r i d d e n c l o s e d b e f o r e r e s e t t i n g Containment S p r a y l o g i c w i t h a L P C I i n i t i a t i o n s i g n a l p r e s e n t t o prevent uncontrolled LPCI i n j e c t i o n i n t o t h e RPV.
3.1 -
IF L P C I A ( B ) i n i t i a t i o n s i g n a l i s p r e s e n t ,
THEN TAKE LPCI A ( B ) INJECTION VALVE E12-F042A(B) control s w i t c h t o CLOSE t o o b t a i n t h e amber o v e r r i d e l i g h t .
3.2 PLACE CNTMT SPRAY A ( B ) MANUAL I N I T I A T I O N E12A-S63A(B) p u s h b u t t o n c o l l a r i n DISARM.
3.3 DEPRESS CNTMT SPRAY A ( B ) SEAL I N RESET E12A-S64A(B) p u s h b u t t o n t o reset t h e C o n t a i n m e n t S p r a y i n i t i a t i o n logic.
3.4 -
I F C o m b u s t i b l e Gas Mixing S y s t e m A ( B ) i s E- running, THEN CLOSE CNTMT SPRAY A ( B ) FIRST SHUTOFF E12-F028A(B).
3.5 CLOSE CNTMT SPRAY A ( B ) SECOND SHUTOFF E12-F537A(B).
3.6 -
IF d i r e c t e d t o i n j e c t i n t o t h e RPV, THEN COMMENCE i n j e c t i o n w i t h RHR A ( B ) Pump a s f o l l o w s : ,
3.6.1. IF d i r e c t e d t o i n j e c t o u t s i d e t h e shroud, THEN INJECT a s follows:
3.6.1.1 AT H13-P629 (P618) ,
PLACE RHR ISOL BYPASS E12-F053A(B) keylock s w i t c h i n BYPASS.
3.6.1.2 OPEN SHUTDOWN COOLING A ( B ) TO FDW SHUTOFF E12-FO53A ( B ) .
3.6.2 IF d i r e c t e d t o i n j e c t i n s i d e t h e shroud, THEN OPEN LPCI A ( B ) I N J E C T I O N VALVE E12-F042A(B).
( C O N T I N U E D ON NEXT PAGE)
Attachment 1 PY-CEUGIE-OS30L Page 31 of 87 PEI-SPI 3.1 Page: 7 of 9 Rev.: 0 3 (Continued) 3.7 -IF directed to place RHR A ( B ) in Suppression P o o l Cooling,
-THEN COMMENCE Suppression P o o l Cooling as follows:
NOTE RHR A ( B ) H X ' S BYPASS VALVE E l Z - F 0 4 8 A ( B ) and RHR A ( B ) HX'S OUTLET VALVE E 1 2 - F 0 0 3 A ( B ) will not close within ten minutes after a LOCA initiation signal.
3.7.1 THROTTLE RHR A ( B ) H X ' S OUTLET VALVE E l Z - F 0 0 3 A [ B ) to obtain 60-65% open.
3.7.2 VERIFY RHR A ( B ) HX'S BYPASS VALVE E l Z - F 0 4 8 A ( B ) is closed.
(CONTINUED ON NEXT PAGE)
Attachment 1 PY-CEI/OE-O63OL Page 32 of 87 PEI-SPI 3.1 Page: 8 of 9 Rev.: 0 PEI-SPI 3.1 Containment Spray Operation (Continued)
- CAUTION *
- flow may result in loss of flow for the LPCS Pump.
- 3.7.3 Combustible Gas Mixing System A(B) is NOT running, THEN THROTTLE RHR flow as follows:
3.7.3.1 OPEN RHR A(B) TEST VALVE TO SUPR POOL E12-FO24A ( B ) .
3.7.3.2 THROTTLE open RHR A(B) HX'S OUTLET VALVE E12-FO03A (B) to obtain 7100-7300 gpm.
3.7.4 IF Combustible Gas Mixing System A(B) is running, THEN THROTTLE RHR flow as follows:
3.7.4.1 OPEN RHR A(B) HX'S OUTLET VALVE E12-FO03A (B).
3.7.4.2 TRROTTLE open RHR A ( B ) TEST VALVE TO SUPR POOL E12-FO24A (B) t o obtain 7100-7200 gpm.
Attachment 1 PY-CEVOIE-063OL Page 33 of87 P E I - S P I 3.1 Page: 9 of 9 Rev.: 0 PEI-SPI 3.1 Containment Spray Operation (Continued)
Control Room B a c k P a n e l Locations P632 1 P623 I P691 I P669 I I P651 P653 I P694 1 P672 I PS85 1 P845 1 P619 1 P63d t I I i t 1 P614 I PS03 I p804 I P604 I P600 1 I
P870 1P969 I P907 I P811 PS09 P8lO P807 P808 P883 P800 P842 P823
- P637 P821 PS22 P866 P872 P868 P864 P612 P840 PS65 P869
Attachment 1 PY-CEUOIE-0630L Page 34 of 87 QUESTION 065 Given the following:
- The plant is in Mode 5 with refueling operations in progress.
- The refuel position one-rod-out interlock surveillance was last completed satisfactory at 0800.
- Then, when performed again at 2130 by operations, the one-rod-out interlock surveillance failed.
WHAT actions are required in accordance with PNPP Technical Specifications?
a) Immediately suspend loading of irradiated fuel into the RPV; initiate action to restore Secondary Containment to operable.
b) Immediately suspend in-vessel fuel movement with equipment associated with the inoperable interlock and insert all insertable control rods.
c) lmmediatefy suspend control rod withdrawal.and initiate actions to fully insert all insertable control rods in cells containing one or more fuel assemblies.
d) Immediately initiate action to insert all insertable control rods and place the mode switch in the SHUTDOWN position in Ihour.
Comment:
The question states the "surveillance" fails and both specifications are in the surveillance, not just the one rod out interlock. The surveillance SVI-C71-T0427, covers SR 3.9.1.1and 3.9.2.2 making B and C correct.
Reference:
SVI-C71-TO427 SR 3.9.1 .I SR 3.9.2.2 Licensee's Position:
The utility believes that there are two correct answers QUESTION 066 Based upon further review by the Utility, the appeal for this question is being withdrawn.
Attachment 1 PY-CEI/OIE-O63OL Page 35 of 87 SVI - C7 1-TO42 7 Page: i Rev.: 5 PERRY OPERATIONS MANLTAL Surveillance Instruction e TITLE: RX MODE SWITCH REFUEL MODE CHANNEL FUNCTIONAL REV1 SION: 5 EFFECTIVE DATE : 11-21-02 PREPARED: G a r y Kirsch 10-2-02
/ Date
Attachment 1 PY-CEUOIE-063OL Page 36 of 87 SVI-C71-TO427 Page: ii Rev.: 5 SCOPE OF REVISION:
Rev. 5 - 1. Added reference to Surveillance Work Orders ( S W O ) .
- 2. Changed MPL to Asset.
- 4. Minor format and administrative changes.
- 5. Revised inoperability notification format per CR 02-00241.
- 6. Added four step removal sequence for disconnecting cable.
- 7. Removed requirement to use Word Simulators for first performance . (OMCR 01- 0091)
- 8. P I C s from previous revision evaluated for incorporation - PIC-1, 2, 3 , 4 , 5 , 6 , 7, 8, 9, and 11.
Attachment 1 PY-CEUOIE-0630L Page 37 of 87 SVI-C71-TO427 Page: 1 Rev.: 5 R x Mode Switch Refuel Mode Channel Functional
1.0 DESCRIPTION
1.1 Scope
Instruction demonstrates operability of Reactor Mode Switch REFUEL Position interlocks by performance of two channel functional tests.
Q" eI Instruction fully satisfies functional surveillance requirements of Technical Specifications SR 3.9.1.1 and SR 3.9.2.2.
Instruction verifies operability for Reactor Mode Switch REFUEL Position interlocks:
- 1. Refuel Position One-Rod-Out Interlock
- 2. Refueling Equipment Interlocks
- a. All-rods-in
- b. Refuel platform position
- c. Refuel platform main hoist, fuel loaded
1.2 Frequency
At least once per 7 days.
1.3 Technical Specification Applicable MODES:
During in-vessel fuel movement with equipment associated with the interlocks.
MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn.
2.0 PRECAUTIONS AND LIMITATIONS
- 1. Steps marked with dollar sign ($1 immediately to left are required by Technical Specifications. Such items, if found to exceed Allowable Values or found to be inoperable, may be NRC reportable and shall be brought to immediate attention of US.
- 2. Steps designated with initial block/lines are to be initialed as values entered or step completed. These steps may require additional initials or signatures to be entered on attachments or Data Package Cover Sheet (DPCS)/Surveillance Work Order (SWO).
- 3. Instruction should be read completely before proceeding with performance.
- 4. Steps shall be performed in sequence and instruction carried through to completion unless directed otherwise.
Attachment 1 PY-CEUOIE-063OL Page 38 of 87 SVI-C71-TO427 Page: 2 Rev.: 5
- 5. US shall be notified immediately if instruction step cannot be completed as stated or if problems develop during instruction performance.
- 6. Actions taken in this instruction cause Reactor Mode Switch REFUEL Position interlocks to be inoperable.
- 7. Channels made inoperable in Section 5.0 for:
Refueling Equipment Interlock Refuel Position One-Rod-Out Interlocks Single Control Rod Withdrawal-Cold Shutdown Single Control Rod Drive (CRD) Removal-Refueling Multiple Control Rod Withdrawal-Refueling
- 8. Refueling platform and equipment shall be operated in accordance with SOI-F11/15, Fuel Handling, Refueling and Auxiliary Platforms (Unit 1).
- 9. During instruction performance, the following annunciator and status lights may come on:
- a. Annunciator ROD WITHDRAWAL BLOCK (UNIT CONTROL CONSOLE 1H13- P680- 05A-El0).
3.0 MANPOWER AND EQUIPMENT 3.1 Manpower/Location/Communication
- 1. Five individuals are recommended:
- a. Control Room Operator to perform switch manipulations and collect data.
- b. Two fuel handling personnel to manipulate refuel platform and limit switches.
- c. Two technicians to disconnect cables, connect probe word simulators and simulate rod withdrawal on 620' level in containment.
- 2. Establish communications among personnel.
3.2 Required Measuring and Test Equipment (MEcTE)
None
Attachment 1 PY-CEI/OIE-O63OL Page 39 of 87 SVI-C71-T0427 Page: 3 Rev.: 5 3.3 Additional Tools and Equipment
- 1. SOI-F11/15, Fuel Handling, Refueling and Auxiliary Platforms (Unit 1).
- 2. 2 RCIS Probe Word Simulators, if required.
- 3. Test Weight, 1L70-M0002EI Block #1 (372 lbs. dry) or Dummy Fuel Bundle (bottom of Dryer Pool).
4.0 PREREQUISITES Initials
- 1. Obtain US'S signature for Work Start on DPCS/SWO.
- 2. Instruction may be performed provided following conditions verified :
- a. Plant in MODE 3 , 4 , or 5 .
- b. Refueling Platform ready for operation in accordance with SOI-F11/15 with Dryer Storage Pool/Reactor Well Gate removed per 101-9. If Section 5.1.3 will not be performed, N/A this step's initial line.
- 3. Request RO verify all control rods inserted with exception of control rod(s) removed per Technical Specifications I
3.10.4, 3.10.5, or 3.10.6.
- 4. At P680, if REACTOR MODE SWITCH lC71A-Sl (P680-llE2)in SHUTDOWN, perform the following. If REACTOR MODE SWITCH in REFUEL, N/A this step's initial lines.
- a. Request RO have a second licensed operator or other technically qualified member of the unit technical staff verify all control rods remain fully inserted with the exception of control rod(s) removed per Technical Specifications 3.10.4, 3.10.5, or 3.10.6.
Obtain individual's signature.
Signature
- b. Request RO place REACTOR MODE SWITCH lC71A-Sl I (P680-llE2) in REEWEL.
- 5. Verify the following:
- a. Annunciator ROD WITHDRAWAL,BLOCK (P680-05A-E10)reset.
Attachment 1 PY-CElIOIE-0630L Page 40 of 87 SVI-C71-T0427 Page: 4 Rev.: 5 5.0 SURVEILLANCE INSTRUCTION Initials NOTE: Instruction divided into the following:
5.1.1 Test Preparation 5.1.2 Refuel Position One Rod Out Interlock 5.1.3 Refueling Equipment Interlocks 5.1.4 Test Restoration 5.1 Surveillance Test 5.1.1 Test Preparation
- 1. Print LEAD TEST PERFORMER'S name on Attachment 3.
- 2. Obtain RO's Authorization for Test Start on DPCS/SWO.
Detach Attachment 3 and give to RO.
- 3. Fill in START TIME/DATE, TEST NUMBER, TEST TITLE, and LEAD PERFORMER in Test Tracking Index.
- 4. Inform US of channels inoperability:
Technical Specification 3.9.1 Actions required during in-vessel fuel movement with equipment associated with the interlocks for:
Refueling Equipment Interlocks 3.9.1 Technical Specification 3.9.2 Actions required in MODE 5 with the reactor mode switch in the refuel position and any control rod withdrawn for:
Refuel Position One-Rod-Out Interlocks 3.9.2 Technical Specification 3.10.4 Actions required in MODE 4 with the reactor mode switch in the refuel position for:
Single Control Rod Withdrawal-Cold Shutdown 3.10.4 Technical Specification 3.10.5 Actions required in MODE 5 with LCO 3.9.5 not met for:
Single Control Rod Drive (CRD) Removal-Refueling 3.10.5 Technical Specification 3.10.6 Actions required in MODE 5 with LCO 3.9.3, LCO 3.9.4, or LCO 3.9.5 not met for:
Multiple Control Rod Withdrawal-Refueling 3.10.6 Record time and date. Obtain US'S signature.
/
Time Date US signature
Attachment 1 PY-CEVOIE-0630L Page 41 of 87 SVI-C71-T0427 Page: 5 Rev.: 5 SECTION 5.1.2 Initials 5.1.2 Refuel Position One Rod Out Interlock NOTE Controls and indicators used in this section are located on P680.
- 1. Request RO confirm SRM operability per Technical Specification 3.3.1.2.
- 2. Perform the following. If a control rod cannot be withdrawn and RCIS Probe Word Simulators must be installed, N/A this step's initial lines.
a.
b.
Request RO withdraw one control rod to position 02 in Individual Drive.
Record number of partially withdrawn control rod.
I Control Rod # -
- 3. Request RO perform the following. If a control rod can I be withdrawn and simulators not installed, N/A this step's initial lines.
- a. Verify SEQUENCE "A" selected.
- b. Select rod 30-59.
- 4. On 620' level in containment, perform the following.
If simulators not: installed, N/A this step's initial lines.
- a. In lH22-P071-A3,locate cable 1CllR96A on rod 30-59 connector.
- b. Independently verify cable 1CllR96A on rod 30-59 connector located. Independent Verifier:
- c. Disconnect cable 1CllR96A from rod 30-59 connector.
- d. Independently verify cable 1CllR96A disconnected from rod 30-59 connector. Independent Verifier:
- e. Connect RCIS Probe Word Simulator to Mux cabinet on connector where cable 1CllR96A disconnected and simulate position 02.
- f. In lH22-P072-A3, locate cable 1CllR524B on rod 30-59 connector.
- g. Independently verify cable 1CllR524B on rod 30-59 connector located. Independent Verifier :
- h. Disconnect cable 1CllR524B from rod 30-59 connector.
- i. Independently verify cable 1CllR524B disconnected from rod 30-59 connector. Independent Verifier:
- j. Connect RCIS Probe Word Simulator to M u cabinet on connector where cable 1CllR524B disconnected and simulate position 02.
Attachment 1 PY-CEVOIE-0630L Page 42 of 87 SVI - C71- TO427 Page: 6 Rev.: 5 SECTION 5.1.2 Initials
- 6. Request RO engage ROD SELECT CLEAR pushbutton.
- 7. Confirm selected rod does not clear.
- 8. Request RO disengage ROD SELECT CLEAR pushbutton.
- 9. Request RO attempt to select another rod.
I
$ 10. Confirm another rod cannot be selected.
- 11. Perform the following:
- a. Request RO insert partially withdrawn control rod to position 00 and N/A initial line for Step 5.1.2.11.c. If simulators installed, N/A this steps initial line.
- b. Confirm rod recorded in Step 5.1.2.2.b indicates 00.
If simulators installed, N/A this steps initial line.
- c. Simulate rod insertion to position 00 with both simulators.
- 12. Confirm the following status lights off:
- 13. Request RO press DRIVE MODE pushbutton and confirm GANG DRIVE mode selected.
I
- 14. Confirm the following:
- b. Annunciator ROD WITHDRAWAL BLOCK (P680-05A-ElO) on.
- 15. Request RO press DRIVE MODE pushbutton and confirm INDIVIDUAL DRIVE selected.
I
- 16. Confirm the following:
- b. Annunciator ROD WITHDRAWAL BLOCK (P680-05A-ElO) reset.
Attachment 1 PY-CEVOIE-0630L Page 43 of 87 SVI- C71- TO427 Page: 7 Rev.: 5 SECTION 5.1.2 Initials
- 17. Request RO engage ROD SELECT CLEAR pushbutton.
- 18. Confirm selected rod clears.
- 19. Request RO disengage ROD SELECT CLEAR pushbutton.
- 20. On 620' level in containment, perform the following.
If Section 5.1.3 to be performed or if simulators not installed in Step 5.1.2.4, N/A this step's initial lines.
- a. In lH22-P071-A3, disconnect RCIS Probe Word Simulator from M u x cabinet.
- b. Connect cable 1CllR96A to rod 30-59 connector.
- c. Independently verify cable 1CllR96A connected to rod 30-59 connector. Independent Verifier:
- d. In lH22-P072-A3,disconnect RCIS Probe Word Simulator from M u cabinet.
- e. Connect cable 1CllR524B to rod 30-59 connector.
- f. Independently verify cable lCllR524B connected to rod 30-59 connector. Tndependent Verifier:
- g. Verify rod 30-59 indicates correct position.
- 21. If only performing Refuel Position One Rod Out Interlock check, N/A initial lines for Section 5.1.3 and proceed to Section 5.1.4. If not, N/A this step's initial line and proceed to Section 5.1.3.
5.1.3 Refueling Equipment Interlocks NOTE: Overvessel switches S1 and 52 are tested assuming an approach to the vessel from the south.
- 2. On Refuel Bridge Operator Status Console, confirm the following status lights off:
- a. OVER VESSEL S1.
- b. OVER VESSEL S2.
- 4. Confirm status light OVER VESSEL S2 on.
- 5. Request RO move Refuel. Bridge away from vessel until I status light OVER VESSEL S2 off.
- 6. Confirm status light OVER VESSEL S1 off.
Attachment 1 PY-CEVOIE-063OL Page 44 of 87 SVI-C71-TO427 Page: 8 Rev.: 5 SECTION 5.1.3 Initials
- 7. Request RO move Refuel Bridge Main Hoist one foot in I downward direction.
8, Confirm Refuel Bridge Main Hoist moved in downward direction.
- 9. Request RO move Refuel Bridge Main Hoist one foot in 1 upward direction.
- 10. Confirm Refuel Bridge Main Hoist moved in upward direction.
- 11. On Refuel Bridge Operator Left Console, confirm status light HOIST LOADED off.
- 12. In 1H13-P651 (CONTROL ROD POSITION PANEL), confirm Refuel Platform Fuel Loaded (PG) LED off.
- 13. In lH13-P652, confirm Refuel Platform Fuel Loaded (PG)
LED off.
- 14. Request RO lift Test Weight Block #1 (or Dummy Fuel I Bundle) using Refuel Bridge Main Hoist, above 15 IN and out of weight test area to clear RED ZONE light.
- 15. On Refuel Bridge Operator Left Console, confirm status light HOIST LOADED on.
- 16. In P651, confirm Refuel Platform Fuel Loaded (PG) LED on.
- 17. In P652, confirm Refuel Platform Fuel Loaded (PG) LED on.
- 18. On Refuel Bridge Operator Status Console, confirm status light ROD BLOCK 2 INTERLOCK off.
- 19. On P680, confirm the following:
- a. Annunciator ROD WITHDRAWAL BLOCK (P680-05A-E101 reset.
- 20. In P652, confirm Refuel Platform Overcore (PC) LED off.
- 21. Request RO hold Limit Switch S2 in actuated (up) I position until Step 5.1.3.25. See Attachment 3 for location. (Simulates Refuel Bridge being located over reactor vessel.)
Attachment 1 PY-CEVOIE-063OL Page 45 of 87 SVI - C7 1- TO42 7 Page: 9 Rev.: 5 SECTION 5.1.3 Initials
$ 22. On Refuel Bridge Operator Status Console, confirm status light ROD BLOCK 2 INTERLOCK on.
- 23. At P680, confirm the following:
- a. Annunciator ROD WITHDRAWAL BLOCK (P680-05A-ElO)on.
$ b. Status light WITHDRAWAL BLOCK (P680-05C) CH 2 on.
- 24. In P652, confirm Refuel Platform Overcore (PC) LED on.
- 25. Request RO release Limit Switch S2. I
- 26. On Refuel Bridge Operator Status Console, confirm status light ROD BLOCK 2 INTERLOCK off.
- 27. At P680, confirm the following:
- a. Annunciator ROD WITHDRAWAL BLOCK (P680-05A-E10) reset.
- 28. In P652, confirm Refuel Platform Overcore (PC) LED off.
- 29. Perform the following. If a control rod cannot be withdrawn and simulators are installed, N/A this step's initial lines.
- a. Request RO withdraw one control rod to position I 02 in Individual Drive.
- b. Record number of partially withdrawn control rod.
Control Rod # -
- 30. Request RO perform the following. If a control rod can be withdrawn and simulators not installed, N/A I
this step's initial lines.
- a. Verify SEQUENCE "A" selected.
- b. Select rod 30-59.
- 31. On 620' level, simulate rod position 02 on both simulators. If simulators not installed, N/A this step's initial line.
- 32. On Refuel Bridge Operator Status Console, confirm the following status lights off:
- a. ROD BLOCK 1 INTERLOCK.
- b. REFUEL INTERLOCK.
C. BRIDGE REV. STOP 1.
Attachment I PY-CEI/OIE-O63OL Page 46 of 87 S V I C71-TO42 7 Page: 10 Rev.: 5 SECTION 5.1.3 Initials 33, In P651, confirm Refuel Platform Overcore (PC) LED off.
- 34. Request RO hold Limit Switch S1 in actuated (up) position until Step 5.1.3.47. See Attachment 3 for I
location. (Simulates Refuel Bridge being located over reactor vessel.)
- 35. On Refuel Bridge Operator Status Console, confirm the following status lights on:
- a. ROD BLOCK 1 INTERLOCK.
- b. REFUEL INTERLOCK.
C. BRIDGE REV. STOP 1,
- 36. At P680, confirm the following:
- a. Annunciator ROD WITHDRAWAL BLOCK (P680-05A-ElO) on.
$ b. Status light WITHDRAWAL BLOCK (P680-05C) CH 1 on.
- 37. In P651, confirm Refuel Platform Overcore (PC) LED on.
- 38. Request RO turn RAISE/LOWER GRAPPLE rheostat on I Refuel Bridge Operator Right Console toward RAISE direction and hold until Step 5.1.3.40.
$ 39. Confirm hoist does not move.
- 40. Request RO release RAISE/LOWER GRAPPLE rheostat. I
- 41. Request RO depress JOG DOWN pushbutton on Refuel 1 Bridge Operator Right Console and hold until Step 5.1.3.43.
$ 42. Confirm hoist does not move.
- 44. Request RO rotate Bridge Speed Control on Refuel I Bridge Operator Right Console to REVERSE and hold until Step 5.1.3.46.
$ 4.5. Confirm bridge does not move,
- 46. Request RO release Bridge Speed Control.
- 47. Request RO release Limit Switch S1.
Attachment 1 PY-CEVOIE-0630L Page 47 of 87 SVI- C71-TO427 Page: 11 Rev.: 5 SECTION 5.1.3 Initials
- 48. On Refuel Bridge Operator Status Console, confirm the following status lights off:
- a. REFUEL INTERLOCK.
- b. ROD BLOCK 1 INTERLOCK.
- c. ROD BLOCK 2 INTERLOCK.
- d. BRIDGE REV. STOP 1.
- e. OVER VESSEL SI.
- f. OVER VESSEL S2.
- 49. At P680, confirm the following:
- a. Annunciator ROD WITHDRAWAL BLOCK (P680-05A-E101 reset.
- 50. Request RO place Test Weight Block #1 (or Dummy Fuel I Bundle) in safe condition and ungrapple.
- 51. On Refuel Bridge Left Console, confirm status light HOIST LOADED off.
- 52. Perform the following:
- a. Request RO insert partially withdrawn control rod to position 00 and N/A initial line for Step 5.1.3.52.c. If simulators installed, N/A this step's initial line.
- b. Confirm rod recorded in Step 5.1.3.29.b indicates 00. If simulators installed, N/A this step's initial line.
- c. Simulate rod insertion to position 00 with both simulators.
- 53. On 620' level in containment, perform the following.
If simulators not installed in Step 5.1.2.4, N/A this step's initial lines.
- a. In lH22-P071-A3,disconnect RCIS Probe Word Simulator from Mux cabinet.
- b. Connect cable 1CllR96A to rod 30-59 connector.
- c. Independently verify cable 1CllR96A connected to rod 30-59 connector. Independent Verifier:
- d. In 1H22-P072-A3,disconnect RCIS Probe Word Simulator from Mux cabinet.
- e. Connect cable 1CllR524B to rod 30-59 connector.
- f. Independently verify cable 1CllR524B connected to rod 30-59 connector. Independent Verifier:
- g. Verify rod 30-59 indicates correct position.
Attachment 1 PY-CEJYOIE-06301, Page 48 of 87 SVI- C71- TO427 Page: 12 Rev.: 5 Section 5.1.4 Initials 5.1.4 Test Restoration
- 1. Inform US of channel operability. Record time and date.
Obtain USS signature.
/
Time Date US Signature 5.2 Plant/System Restoration
- 1. Inform RO of system restoration.
- 2. Fill in STOP TIME/DATE in Test Tracking Index.
5.3 Acceptance Criteria E: Satisfactory instruction completion based on Technical Specification items (marked with dollar sign).
- 1. All Technical Specification required items as indicated by dollar signs ($1 performed satisfactorily.
I I YES [ I NO, US notified
- 2. All other items performed satisfactorily.
[ 1 YES [ 1 NO, I&C Supervisor notified
- 3. Check blocks on DPCS/SWO to indicate acceptable or unacceptable test results.
5.4 Records The following records are generated by this instruction:
Quality Assurance Records Data Package Cover Sheet/Surveillance Work Order SVI-C71-T0427, pages 3 through 12, and:
Attachment 1, M&TE/Comment/Signature Sheet Non-Quality Records None
Attachment 1 PY-CEUOIE-063OL Page 49 of 87 SVI-C71-TO427 Page: 13 Rev.: 5
6.0 REFERENCES
6.1 Technical Specifications 6.2 Drawings B-208-020 B-208-086 6.3 Vendor/Technical Manuals GEK 75559 Perry 1/2 Operations and Maintenance Instruction REFUELING PLATFORM EQUIPMENT ASSEMBLY 767345768, G14, E.A. Number 2, 4/10/84 (File
- 147G) 6.4 Commitments The following commitments are either partially or fully satisfied by this instruction:
None 7.0 ATTACHMENTS 7.1 Attachment 1 - M&TE,/Comment/SignatureSheet.
7.2 Attachment 2 - Reference Drawing.
7.3 Attachment 3 - Alarm, Status/Indication Light, and General Information List.
Attachment 1 PY-CEI/OIE-O63OL Page 50 of 87 Attachment 1 S V I -C71-TO427 Sheet 1 of 1 Page: 14 Rev. : 5 Rx Mode Switch Refuel Mode Channel Functional M&TE/Comment/Signature Sheet Comments :
Performed By: / /
/ /
Independent Verifier: / /
/ /
/ /
/ 1 Signature Initials Date
Attachment 1 PY-CEI/OIE-O63OL Page 51 of 87 SVI -C71-TO427 Sheet 1 of 1 Page: 1 5 Rev.: 5 Reference Drawinq Refueling Platform W A T E D ON FAR SIDE OF TRUCK
Attachment 1 PY -CEI/OXE-O630L Page 52 of 87 Attachment 3 S V I - C71- TO427 Sheet 1 of 1 Page: 16 - LAST Rev.: 5 Alarm, Status/Indication Liqht, and General Information List S V I
Title:
R x Mode Switch Refuel Mode Channel Functional NOTE: This attachment is for information only and is provided as an aid to the RO. Additional alarms and/or plant impact may occur due to the specific plant conditions at the time of performance. Desired changes to this attachment should be documented in the Comments section of the DPCS/SWO.
The following alarm may be received intermittently:
LOCATION Annunciator ROD WITHDRAWAL BLOCK P680- 05A-E10 The following status lights may come on or go off intermittently:
STATUS LIGHTS LOCATION Status light INSERT REQUIRED CH 1 and CH 2 P680-05C Status light WITHDRAWAL BLOCK CH 1 and CH 2 P680-05C PLANT IMPACT:
At least ( 3 ) three Rod B l o c k signals will be generated during performance of this instruction. Refuel Bridge will be unavailable f o r use if Section 5.1.3 performed .
LEAD TEST PERFORMER:
(PRINT)
Refueling Equipment Interlocks 3.9.1 Attachment 1 3.9 REFUELING OPERATIONS PY-CETIOIE-OB3OL Page 53 of 87 3.9.1 Refueling Equipment Interlocks LCO 3.9.1 The refueling equipment interlocks shall be OPERABLE.
APPLICABILITY: During in-vessel fuel movement with equipment associated with the interlocks.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A. 1 Sus end in-vessel Immediately refueli ng equi pment interlocks inoperable.
?
fue movement with equipment associated with the inoperable i nterl ock (s 1.
OR A.2.1 Insert'a control rod Immediately
.withdrawal block.
AND A. 2.2 Verify all control I mmedi ate1y rods are ful'ly inserted.
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Perform CHANNEL FUNCTIONAL TEST on each o f 7 days the following required refueling equipment Q L s interlock inputs:
e a. A7 1 -rods-in,
- b. Refuel platform position, and
- c. Refuel platform main hoist, fuel 1oaded .
PERRY - UNIT 1 3.9-1 Amendment No. 116
Refuel Position One-Rod-Out Interlock 3.9.2 Attachment 1 PY-CEUOIE-063OL 3.9 REFUELING OPERATIONS Page 54 of 87 3.9.2 Refuel Position One-Rod-Out Interlock LCO 3.9.2 The refuel position one-rod-out interlock shall be OPERABLE.
APPLICABILITY: MODE 5 w i t h the reactor mode switch i n the refuel position and any control rod withdrawn.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refuel position one- A. 1 Suspend control rod Immedi a t e 1y rod-out i nterl ock withdrawal .
A. 2 I n i t i a t e action t o Immediately f u l l y insert a l l insertable control rods i n core cells containing one o r more fuel assembl ies.
SURVEI LLANCE REQUIREMENTS SURVE I LLANCE I FREQUENCY SR 3.9.2.1 Verify reactor mode switch locked i n refuel 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> position.
I (continued)
PERRY - UNIT 1 3.9-2 Amendment No. 69
SURVEILLANCE FREQUENCY SR 3.9.2.2 ----_-_------------NOTE--------------------
(2b . 5 Not required t o be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> e after any control rod i s withdrawn.
Perform CHANNEL FUNCTIONAL TEST. 7 days PERRY - UNIT 1 3.9-3 Amendment No. 69
Attachment 1 PY-CEUOIE-0630L Page 56 of 87 QUESTION 073 Select the statement below that reflects an Operations Section expectation for TRANSIENT ALARM RESPONSE during implementationof Plant Emergency Instructions (PEI).
a) Entry in to the TRANSIENT ALARM RESPONSE mode shall be announced by the Unif Supervisor.
b) Locked in alarms that are abnormal for the present plant status should be communicated to the Unit Supervisor.
c) Recurring alarms that annunciate ON1 or PEI entry conditions do NOT need to be re*
announced.
d) The TRANSIENT ALARM RESPONSE mode will remain in effect until the PEis are exited.
Comment:
Reference for transient response (PYBP-POS-2-3) to this question. Answer A is correct because the alarm response mode shall be given at the next brief. This would make answer A correct. Answer B is also correct during transient response, as it is the RO's responsibility to report alarms that are not normal for the situation.
Reference:
NOBP-POS-2-3 Licensee's Position:
The utility believes that there are two correct answers QUESTION 074 Based upon further review by the Utility, the appeal for this question is being withdrawn.
QUESTION 075 Based upon further review by the Utility, the appeal for this question is being withdrawn.
PLANT OPERATIONS SECTION BUSINESS PRACTICE Number:
PYBP-POS-2-3
Title:
Revision: Page Transient Response Guidelines 2 8 of 14 5.4 Alarm Response/ARl Usage During Emergency/ Off-Normal conditions, critical alarms may indicate degradation in the level of plant or personnel safety, potential for equipment damage, or direct entry and actions using emergency or off-normal procedures. Identified critical alarms shall be communicated to the Unit Supervisor. During Emergency/ Off-Normal conditions, annuciators that are not critical or are expected based upon Q73 plant conditions, do not need to be communicated. The Unit Supervisor shall announce entry into transient annuciator response during the first transient crew briefing. As time permits, annunciator reviews are performed for unexpected conditions and Alarm Response Instructions are reviewed for plant response.
Following Emergency/ Off-Normal conditions and when the plant has been stabilized, the Unit Supervisor shall announce the resumption of normal annunciator alarm instruction response.
The reactor operator or balance of plant operator shall own alarm response instructions. This performance will be in conjunction with the execution of the ONIs and PEls. The expectation is that the operators own the ARls and will take appropriate action without interfering with execution of the ONls or PEls.
This allows the Unit Supervisor to maintain a control and oversight role during the transient.
Various annunciator windows are color coded to assist the operator in responding to multiple alarm events. Color-coding provides a visual indicator as a means to aid the operator in prioritizing alarm response. The color-coding is intended to be a tool to aid the operating crew set priorities based upon potential significance of the degrading plant conditions or equipment failures. All annuciators are important, however during the initial phases of a transient the red and amber windows require immediate evaluation.
5.5 ON1 Usage This section provides guidance for multiple ON1 transient type events with the intention of relieving the Unit Supervisor of the need to direct each action step to the operators. The ability to assign owners for ON1 actions is required to ensure that the oversight and control role of the Unit Supervisor is not compromised.
The Unit Supervisor is responsible for ensuring all ON1 immediate and subsequent actions are addressed. The Unit Supervisor may assign all or part of the subsequent actions to the reactor operators. Attachment Iprovides a reference for suggested responsibilities for ON1 actions.
Attachment 1 PY-CEI/OIE-063OL Page 58 of 87 QUESTION 079 The plant was operating at 100% power when a LOCA occurred. All control rods are fully inserted. LPCS and LPCl 'A' are both injecting into the RPV. NO other ECCS pumps are available. As long as both pumps are injecting, RPV water level can be maintained above TAF.
Suppression Pool temperature is 130°F and rising. Select the statement below that correctly describes the use of LPCl 'A' for Suppression Pool cooling.
a) LPCl 'A' must be diverted to Suppression Pool Cooling to ensure that Suppression Pool temperature is maintained below the Heat Capacity Limit, since LPCS can maintain adequate core cooling through spray cooling alone.
b) LPCI 'A' may be diverted fo Suppression Pool cooling as long as LPCS is able fo maintain RPV water level above -25 inches (the Minimum Steam Cooling RPV wafer level). .
c) LPCl 'A' must be diverted to Suppression Pool Cooling, irrespective of adequate core cooling, when neither Suppression Pool temperature nor Reactor pressure can be maintained below the Heat Capacity Limit (HCL) d) LPCl 'A' may be diverted to Suppression Pool Cooling only if additional injection sources become available to be used with LPCS to maintain RPV water level above 0 inches.
Comment:
PEI bases defines adequate core cooling as level being above the Minimum Zero Injection Water Level. As Answer B states that LPCl may be diverted to Suppression Pool Cooling as long as LPCS can maintain RPV water level above -25 inches, (consistent with the non ANVS flow chart not requiring ED until level cannot be restored and maintained above -25 ") answer B is also correct.
Answer D is correct because maintaining level above zero inches ensures the core is cooled.
Reference:
PEI Bases Definition of Adequate Core Cooling PEI B13 non ATWS flowchart Licensee's Position:
The utility believes that there are two correct answers QUESTION 086 Based upon further review by the Utility, the appeal for this question is being withdrawn.
Attachment 1 PY-CEIlOIE-0630L Page 59 of 87 Procedure Number:
PERRY NUCLEAR PIOWER PLANT PEI Bases
Title:
Use Category:
PEI Bases Document Reference Plant Emergency Instruction (PEI) Revision: Page Definitions and Usage of Key Words 6 25 of 392 The meaning of the following terms is discussed in the context of their use within the PEls. This information is provided in order to facilitate a consistent and technically accurate understanding of the entry conditions, operator actions, cautions, and execution of the PEls.
Adequate Core Coolinq Heat removal from the reactor sufficient to prevent rupturing the fuel clad. Within the EPGs, three viable mechanisms for establishing adequate core cooling are defined-core submergence, spray cooling, and steam cooling.
Submergence is the preferred method for cooling the core. The core is adequately cooled by submergence when it can be determined that RPV water level is at or above the top of the active fuel. All fuel nodes are then assumed to be covered with water and heat is removed by boiling heat transfer.
Adequate spray cooling is provided in BWW3 through BWW6 designs, assuming a bounding axial power shape, when design spray flow requirements are satisfied and RPV water level is at or above the elevation of the jet pump suctions. The covered portion of the core is then cooled by submergence while the uncovered portion is cooled by the spray flow. Currently this method is not used and is under design review to ensure that it is acceptable for use at Perry.
Steam cooling is relied upon only if RPV water level cannot be restored and maintained above the top of the active fuel, cannot be determined, or must be intentionally lowered below the top of the active fuel. The core is adequately cooled by steam if the steam flow across the uncovered length of each fuel bundle is sufficient to maintain the hottest peak clad temperature below the appropriate limiting value-1500°F if makeup can be injected, 1800°F if makeup cannot be injected. The covered portion of the core remains cooled by boiling heat transfer and generates the steam, which cools the uncovered portion.
Steam cooling with makeup capability if RPV water level cannot be restored and maintained above the top of the active fuel; during RPV flooding when the reactor may not be shutdown; if RPV water level is intentionally lowered below the top of the active fuel to reduce reactor power or if emergency RPV depressurization is required. When RPV water level cannot be restored and maintained above the top of the active fuel and when RPV water level is intentionally lowered below the top of the active fuel, adequate steam flow is established by maintaining RPV water level above the Minimum Steam Cooling RPV Water Level. When the reactor is not shutdown under all conditions without boron during RPV flooding and when emergency RPV depressurization is required under failure-to-scram conditions, adequate steam flow exists as long as RPV pressure is above the Minimum Steam Cooling Pressure. In all cases, the peak-clad temperature is limited to 1500°F, the threshold for fuel rod perforation.
Steam cooling without makeup capability is employed during steam cooling. With no makeup to the RPV, adequate steam flow exists as long as RPV water level remains above the Minimum Zero-Injection RPV Water Level. When RPV water level drops below this elevation, emergency depressurization must be performed. The peak clad temperature is permitted to rise to 18OO"F, the threshold for significant metal-water reaction, to maximize the heat transfer to steam and to delay the depressurization as long as possible.
Attachment 1 PY-CEI/OIE-O63OL Page 60 of 87 Procedure Number:
PERRY NUCLEAR POWER PLANT PEI Bases
Title:
Use Category:
PEI Bases Document Reference Plant Emergency Instruction (PEI) Revision: Page Definitions and Usage of Key Words 6 26 of 392 The minimum RPV water level at which adequate steam flow exists is higher when makeup capability exists because:
The limiting fuel temperature is lower (1500°F). The higher limit of 1800°F is used only when cladding perforation cannot be avoided.
With injection, water at the core inlet is subcooled. Some of the energy produced by the core must then be expended in raising the temperature of the liquid to saturation and less steam will be produced to cool the uncovered portions of the core.
The "adequate core cooling" state can be defined only within the context of the EPG guidelines and contingencies. Once conditions requiring entry of the SAGs exist, a normal core configuration can no longer be assumed and the same criteria cannot be applied. While one of the objectives of the SAGs is to submerge the core or core debris, restoring RPV water level to above the top of the active fuel in accordance with the RPV and Primary Containment Flooding guideline does not necessarily reestablish adequate core cooling.
Assure Make certain that a specified state or condition is established and will be maintained.
Encompasses an implied action to operate appropriate systems, as available, to accomplish the stated objective. Both direct and indirect indications may be used to determine that the specified state or condition has been achieved and will be maintained (refer to the discussion of "adequate core cooling").
Available The state or condition of being ready and able to be used (placed into operation) to accomplish the stated (or implied) action or function. As applied to a system, this requires the operability of all necessary support systems (electrical power supplies, cooling water, lubrication, etc.) for the systemlcomponents to work as designed.
Boron Injection Initiation Temperature Defined to be the greater of either:
The suppression pool temperature at which initiation of a reactor scram is required by Technical Specifications, or, The highest suppression pool temperature at which initiation of boron injection usinb SLC will result in injection of the Hot Shutdown Boron Weight of boron before suppression pool temperature exceeds the Heat Capacity Limit.
Perry uses the first criteria as the limit.
Bwassing Temporarily disabling the functioning of an automatic protection feature. As used in the PEls, this term is generally limited to conditions where a bypass feature has been included in the system (e.g., bypassing a high drywell pressure interlock).
Attachment 1 PY-CEUOIE-063OL Page 61 of 87 Procedure Number:
PERRY NUCLEAR POWER PLANT PEI Bases
Title:
Use Category:
PEI Bases Document Reference Plant Emergency Instruction (PEI) Revision: Page Definitions and Usage of Key Words 6 29 of 392 I lniect Slowly The phrase Inject slowly...in RPV Control (ATWS) - Level means if indicated level goes below the bottom of active fuel (BAF), injection into the vessel should be promptly increased with the available systems to restore level indication to just on-scale on the fuel zone instruments. Once on-scale indication is obtained, injection should then be reduced and controlled to slowly increase RPV level to the desired band. The injection rate should be controlled such that RPV power oscillations and spiking is minimized.
Line UD for iniection Establish the initial conditions necessary for system operation including positioning of valves and breakers, installation of spool pieces, etc. as directed in the SPls or, if not specified in the SPls then per the respective system Sol.
Maintain below/above Take the action necessary to prevent the value of the parameter from exceeding identified limits.
Momentarily exceeding a parameter value due to instrumentation perturbation is excluded from this determination.
The Minimum Number of SRVs Rewired for Decay Heat Removal (MNSDHR) 121 The least number of SRVs which, if opened, will remove all decay heat from the core at a pressure sufficiently low that the ECCS with the lowest head will be capable of making up the SRV steam flow.
The Minimum Steam Coolina Pressure [MSCP)
The lowest RPV pressure at which steam flow through open SRVs is sufficient to preclude any clad temperature from exceeding 1500°F even if the reactor core is not completely covered. The MSCP is a function of the number of open SRVs.
Minimum Number of SRVs Rewired for Emersencv DeDressurkation [MNSRED) m The least number of SRVs which corresponds to a Minimum Steam Cooling Pressure (MSCP) sufficiently low that the ECCS with the lowest head (LPCI) will be capable of making up the SRV steam flow at the corresponding MSCP. The MNSRED is utilized to ensure the RPV will depressurize and remain depressurized when emergency depressurization is required.
Minimum Steam Coolinn RPV Water Level (MSCRWL)
(-25)
The lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered part of the core from exceeding 1500°F. This water level is utilized to preclude fuel damage when RPV water level is below the top of active fuel.
Procedure Number:
PERRY NUCLEAR POWER PLANT PEI Bases
Title:
Use Category:
PEI Bases Document Refe rence Plant Emergency Instruction (PEI) Revision: Page Operational Limits and Setpoints 6 38 of 392 RPV Level Setpoints 177.7 =I78 in. Low level scram setpoint.
100.75=100 in. Two feet below the feedwater spargers 16.5 in. ADS initiation setpoint.
0 in. Top of Active Fuel (TAF).
w 79 e -25=-25 in. Minimum Steam Cooling RPV Water Level.
-43.76 = -42.5 in. Minimum Zero Injection RPV Water Level.
-45 in Jet Pump Suction RPV Pressure Setpoints 1064.7 = 1065 psig High RPV pressure scram setpoint.
900 psig Pressure at which all turbine bypass valves are fully open.
133 = 130 psig Highest RPV pressure at which the shutoff head of a low-water-qualify alternate injection subsystem (excluding SLC) is reached.
53.5=60 psig Decay Heat Removal Pressure (above containment)
RPV Temperature Setpoints Between 70°F RPV water temperature for cold shutdown conditions.
and 2OOOF 1OOOFlhr RPV cooldown rate LCO.
Containment/DrvwellTemperature Setpoints 95OF Most limiting suppression pool temperature LCO and containment temperature LCO.
145OF Drywell temperature LCO.
185OF Containment design temperature and environmental qualification temperature for safety related electrical equipment in the containment.
33OOF Maximum temperature at which ADS is qualified, drywell design temperature.
Procedure Number:
PERRY NUCLEAR POWER PLANT PEI Bases
Title:
Use Category:
PEI Bases Document Reference PEI-B13 RPV Control (Non-ATWS) Revision: Page 6 95 of 392 STEP:
Injectionsystem.
subsystem or alternate injecfon subsystem lined up with the pump YES WHEN g of the foiimng condaions exist:
Emergemy Depressurization is required
- Emergency Depressurization is required
- Rpv level be restored R W level decreases to -42.5 in and maintainedgreater then
-25 in.
THEN proceed I
DISCUSSION This override step applies throughout the performance of the remainder of RPV Level Control.
Emergency RPV depressurization permits injection from low head systems, maximizes the total injection flow, and minimizes the flow through any primary system break.
Attachment 1 PY-CEIlOIE-0630L Page 64 of 87 Procedure Number:
PERRY NUCLEAR POWER PLANT I PEI Bases Use Category:
Reference Revision:
97 of 392 DISCUSSION (Continued)
If an injection source is not available then Steam Cooling is entered. If an RPV injection source becomes available during this time and RPV water level cannot be restored and maintained above the Minimum Steam Cooling RPV Water Level (MSCRWL)(-25), then emergency RPV depressurization is required.
If the injection source can restore and maintain RPV water level above the MSCRWL, Steam Cooling is exited when RPV water level is increasing. Emergency RPV Depressurizationis not immediately required for the following reasons:
While peak cladding temperature could exceed 1800°F while RPV water level is increasing to the MSCRWL, the length of time required for the increase is expected to be short since the MZIRWL and MSCRWL are only approximately one foot apart. Provided the system continues to operate, the core will ultimately be cooled by submergence.
Cooling by submergence is usually preferable to blowing down since submerging the core not only quenches the uncovered, heated portion of the core, but also adds inventory for long-term cooling.
A blowdown would deplete the remaining inventory of water in the RPV. Since the available injection source must make up this lost inventory, the time required to raise RPV water level above the MSCRWL and cool the core by submergence could be lengthened.
If RPV water level is increasing, the system is providing sufficient injection to overcome break flow.
If RClC is restored, depressurizing could result in a low RPV pressure isolation. Even if the system can be operated at low pressure by defeating isolation interlocks, continued operation without bypassing the interlocks is preferable since the system is designed for operation above the low RPV pressure isolation setpoint and normal operation imposes less workload upon the operating crew.
If attempts are unsuccessful in starting pumps in one or more systems, injection subsystems, or alternate injection subsystems, the operators will utilize Steam Cooling. Steam Cooling is performed by allowing RPV water level to decrease through boil off until it drops to the Minimum Zero-Injection RPV Water Level (-42.5 in.). During this period the fuel temperature in the uncovered portion of the core increases and heat is transferred from the fuel rods to the steam.
Minimum Zero-Injection RPV Water Level is defined as the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to prevent any clad temperature in the uncovered portion of the core from exceeding 1800°F.
Attachment 1 PY-CEVOIE-0630L Page 65 of 87
Title:
PERRY NUCLEAR POWER PLANT Procedure Number:
Use Category:
PEI Bases J PEI Bases Document Reference PEI-B13 RPV Control (Non-ATWS) Revision:
96 of 392 DISCUSSION (Continued)
If an injection source is available, emergency depressurization is delayed until RPV water level reaches the top of the active fuel, but may be performed anytime RPV water level is between the top of the active fuel and the Minimum Steam Cooling RPV Water Level.
If it is expected that operating injection systems will reverse the level trend before RPV water level drops to the Minimum Steam Cooling RPV Water Level, the blowdown may be delayed.
If it is believed that available injection systems are capable of restoring and maintaining RPV water level above the Minimum Steam Cooling RPV Water Level following RPV depressurization, the blowdown may be performed as soon as RPV water level reaches the top of the active fuel.
If it is not expected that available injection systems will restore and maintain RPV water level above the Minimum Steam Cooling RPV Water Level following RPV depressurization, the blowdown should be delayed as long as possible.
The emergency depressurization should be performed even if RClC is the only source of injection to the RPV. RClC operation can continue even after the RPV has been depressurized, since the low pressure isolation may be defeated. The system can sustain some flow as long as RPV pressure is above the value at which the turbine stalls.
Emergency depressurization is not performed while RPV water level is above the top of the active fuel because:
The core will remain adequately cooled as long as RPV water level remains above the Minimum Steam Cooling RPV Water Level.
The time before RPV water decreases to the top of the active fuel can best be used to line up additional injection sources. If the decreasing RPV water level trend can be reversed, emergency depressurization may not be required.
The emergency depressurization requirement in this step is predicated upon three conditions:
9 RPV water level has dropped at least to the top of the active fuel.
At least one injection source is available.
The available injection sources cannot restore and maintain RPV water level above the MSCRWL if a blowdown is not performed.
Immediate emergency RPV depressurization is thus not required, even with RPV water below the MSCRWL, if available injection sources can restore and maintain RPV water level above the MSCRWL.
The MSCRWL is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F assuming the most limiting top-peaked power shape prior to reactor shutdown.
Attachment 1 PY-CEUOIE-0630L Page 66 of 87 QUESTION 092 Given the following conditions:
- Reactor Plant at 100% RTP
- The following annunciators are in alarm:
- HOT SURGE TANK LEVEL HI
- HTR 4 ISOL HOT SRG TK LEVEL HI
- The Extraction Steam supply and Steam Seal Evaporator drains to Heater 4 have automatically isolated
- N21-F220, Hot Surge Tank Level Control Bypass Valve indicates closed
- N21-F230, Hot Surge Tank Level Control Valve is partially open and is unresponsive to the Hot Surge Tank Level Controller signals (in either AUTO or MANUAL)
- Local manual control of N21-F230, Hot Surge Tank Level Control Valve was unsuccessful
- Hot Surge Tank level is 150" and increasing slowly Which one of the following actions should you direct the ATC Operator to perform while maintaining current power level?
a) Shutdown one of the Condensate Booster Pumps.
b) Perform the "SecuringFlow to the Hot Surge Tank" section of SOI-N21 c) Throttle open Condensate Minimum Flow Recirculation Valve (N21+245, Short Cycle Clean-up Valve) d) Manually trip all Hofwell and Condensate Booster Pumps Comment:
There is no correct answer listed. With a Heater 4 isolation, the ATC is 'required to take the immediate actions of ONLN36, which require the operator to reduce reactor power to less than 95%.
The answer listed as correct is in the subsequent actions of the ARI, that would not be done until after the immediate actions are completed and power is decreased. This is asking the SRO to direct a supplemental action, prior to the required immediate action. If he did order supplemental actions, answers B (ARI H I3-P680-0002-E2, action 4.7.3), C (ARI H13-P680-0002-E2, action 4 3 , and D (ARI Hl3-P680-0002-E2, action 4.7.4) are all subsequent actions in the ARI that could be used for mitigating the Hot Surge Tank High level. In this situation, there would be three potential actions to mitigate the problem.
References:
ARI-H 13-P680-0002-E2 ONI-N36 SOLN21 Licensee's Position:
The Utility believes the question should be deleted as the immediate actions of ONI-N36 require the operator to lower reactor power.
Amchment 1 PY-CEUOIE-0630L Page 67 of 87
~~
Instruction Number:
@t2 PERRY NUCLEAR POWER PLANT ARI-H 13-P680-0002-E2 e
Title:
HOT SURGE TANK LEVEL HI Use Category:
In Field Reference Revision: Page Computer Point ID - None 5 51 of 57 1.O CAUSE OF ALARM 1.1 Hot Surge Tank level greater than 131 inches as sensed by 1N21-N336.
1.2 Hot Surge Tank temperature below 300", causing inaccurate remote indicationlalarrn of Hot Surge Tank level.
I.3 Failure of HOT SURGE TANK LEVEL CONTROL IN21-F230, IN21-R475, controller.
2.0 AUTOMATIC ACTION Heater 4 will isolate on further increase to 134 inches.
3.0 IMMEDIATE OPERATOR ACTION None 4.0' SUBSEQUENT OPERATOR ACTION NOTE At Hot Surge Tank temperatures below 300°F, HOT SURGE LEVEL & CNDS TO HTR 4 FLOW, will indicate up to 15 inches high. The tank level sight glass should be used if high level data is needed.
0 4.1 TAKE manual control of HOT SURGE lN21- 1N21-TANK LEVEL CONTROL. F230 R475 0 4.2 REDUCE Hot Surge Tank level to 1N21- 1N21-between 105-130 inches. F230 R475 NA 4.3 IF necessary, 0 THEN REFER T O SOI-N21 and PERFORM Alternate Hot Surge Tank Level Control.
Instruction Number:
PERRY NUCLEAR POWER PLANT ARI-H 13-P680-0002-E2
Title:
Use Category:
HOT SURGE TANK LEVEL HI In Field Reference Revision: Page Computer Point ID - None 5 52 of 57 NA 4.4 I F necessary, IN21-0 THEN MANUALLY THROTTLE Hot F230 Surge Tank Level Control locally.
(HB 600 E/2)
NA 4.5 IF necessary, lN21-THEN MANUALLY THROTTLE Hot F220 Surge Tank Level Control Bypass, locally. (HB 600 E12)
R THROTTLE OPEN CNDS MIN 1N21- fN21-RCIRC FLOW CONTROL. F245 R247 0 MAINTAIN motor current for the operating CBPs e 353 amps.
R 4.6 MAINTAIN motor current for the operating Hotwell pumps e 195 amps.
NA 4.7 1 Hot Surge Tank level rises I 1 Hot Surge Tank goes off scale high I THEN PERFORM the following:
NOTE HOT SURGE LEVEL & CNDS TO HTR 4 FLOW, (blue pen) or local (sightglass) level indication may be used to determine Hot Surge Tank level. HEATER SHELL PRESS 4, local pressure indicator ERlS point N21EA040, or Process Computer point N36BAOl3 may be used to determine Hot Surge Tank pressure.
R 4.7.1 VERIFY HST LVL CV MANUAL CONTROL in OFF.
0 4.7.2 VERIFY HOT SURGE TANK LEVEL 1N21- lN21-CONTROL in MANUAL and minimum. F230 R475 NA 4 - 7 3 IF necessary, R THEN REFER TO SOI-N21 and PERFORM Securing Flow to the Hot Surge Tank.
Attachment 1 PY-CEYOIE-063OL Page 69 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ARI-H 13-P680-0002-E2
Title:
HOT SURGE TANK LEVEL HI Revision: Page Comwter Point ID - None 5 53 of 57 NA 4.7.4 Unable to stop Condensate flow to the Hot Surge Tank THEN PERFORM the following:
0 MANUALLY TRIP all Hotwell Pumps 0 MANUALLY TRIP all Condensate Booster Pumps END OF SECTION
Attachment 1 PY-CEVOIE-0630L Page 70 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: I Page I 8 Iof 17 Q92 E ) LOSS OF FEEDWATER HEATING Effective Date: 8-30-04 Preparer: Dan Roniger I 8-10-04 Date
Attachment 1 PY-CEVOIE-0630L Page 71 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 2 o f 17 TABLE OF CONTENTS PAGE 1.o ENTRY CONDITIONS 3 2.0 AUTOMATIC ACTIONS 3 3.0 IMMEDIATE ACTIONS 5 4.0 SUPPLEMENTAL ACTIONS 5
5.0 REFERENCES
12 6.0 RECORDS 12 7.0 SCOPE OF REVISION 12 8.0 ATTACHMENTS 13
Attachment 1 PY-CEIIOIE-063OL Page 72 of 87
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Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference I Revision: I Page I.O ENTRY CONDITIONS 1.I Alarms 0 HTR 6A (6B)EXST & INLET DRNS ISOL LEVEL HIGH HTR 5A (5B) EXST & INLET DRNS ISOL LEVEL HIGH HTR 4 ISOL HOT SRG TK LEVEL HI HEATER 3A (3B) EXST ISOL LEVEL HIGH HEATER 2A (ZB, 2C) LEVEL HIGH HEATER 1A (IB, I C ) LEVEL HIGH 1.2 Parameters 1-2.1 Decreasing feedwater temperatures OR differential temperatures as indicated on CONDENSATE SYSTEM TEMPERATURE recorder, 1NZI-R216 and FEEDWATER TEMPERATURE recorder, IN27-RO66 on IH13-P842.
1.2.2 Feedwater heater levels outside the normal operating range.
1.2.3 Feedwater heater pressures outside the normal operating range.
I.2.4 Feedwater heater temperatures outside the normal operating range.
2.0 AUTOMATIC ACTIONS 2.1 Possible rod block AND/OR reactor scram due to high neutron flux.
Attachment 1 PY-CEUOIE-063OL Page 73 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 4 of 17 NOTE The most probable cause of a loss of feedwater heating is the automatic isolation of a heater or heaters on high or high-high level. Refer to Attachment 2 for applicable isolation valves.
2.2 APRM flux levels increase with NO change in:
0 Control rod position Recirculation flow 2.3 The following will occur on a heater isolation:
0 The extraction steam block valve, positive assist check valve, and associated valves listed in Attachment 2 will close.
0 The normal AND alternate drain valves for the affected heater modulate to return heater level to normal.
0 Normal drains to the affected heater will be directed to the alternate drain path.
A&chment 1 PY-CEUOIE-063OL Page 74 of 87 I
Title:
PERRY NUCLEAR POWER PLANT' Instruction Number:
Use Category:
ONI-N36 I LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 5of 17 3.0 IMMEDIATE ACTIONS Operating with a reduced feedwater temperature (e.g., without all feedwater heaters in NOTE Shifting Recirculation Pumps to slow speed from minimum FCV position will not appreciably increase the margin to thermal limits. Therefore, to minimize the transient to Recirculation PumDs, do not transfer them to slow speed.
THEN REDUCE reactor power using Y?- Reactor Recirculation Flow Control Q Valves to meet all the following criteria:
n 0 I95% reactor power Less than or equal to the power level prior to the loss of feedwater heating.
4.0 SUPPLEMENTAL ACTIONS Cl 4.1 REFER TO ONI-C51 and EXECUTE concurrently with this instruction.
0 4.2 MONITOR Feedwater temperature
Attachment I PY-CEI/OIE-O63OL Page 75 of 87
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Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: I Page NA 4.3 rThe minimum required number of OPRMs to maintain trip capability are NOT OPERABLE.
Feedwater temperature is more than 5°F PDB-below the 4255°F curve of the A001 1 FEEDWATER TEMPERATURE VERSUS CORE THERMAL POWER graph.
THEN USE the Backup Stability PDB-Protection Regions Two Loop Power - A0006 Flow Reduced Feedwater Power To Flow Map I I 0 4.4 MONITOR the following for proper operation of individual heaters and heater strings:
Level Temperature 0 Pressure NA 4.5 IF the high level alarm is received for 1.4 1B IC feedwater heater 1, CI THEN REFER TO Attachment Ito VERIFY the normal and alternate drain valves open.
NA 4.6 IF the high level alarm is received for 2A 2B 2c feedwater heater 2, a THEN REFER TO Attachment 1 to VERIFY the normal and alternate drain valves open.
Attachment 1 PY -CEI/OE-O630L Page 76 of 87 I
Title:
PERRY NUCLEAR POWER PLANT Instruction Number:
Use Category:
ONI-N36 I LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 7of 17 NA 4.7 IF the high level alarm is received for 3A 3B feedwater heater 3, cf THEN REFER TO Attachment 2 to VERIFY the following valves close:
Extraction steam block valve Positive assist check valve Associated valves NA 4.8 IF the high level alarm is received for 5A 5B feedwater heater 5, n THEN REFER TO Attachment 2 to VERIFY the following valves close:
Extraction steam block valve Positive assist check valve Associated valves NA 4.9 IF the high level alarm is received for 6A 6B feedwater heater 6, 0
THEN REFER TO Attachment 2 to VEREFY the following valves close:
0 Extraction steam block valve Positive assist check valve Associated valves NA 4.10 IF feedwater heater 1 level rises to the ?A 1B IC high-high setpoint, n THEN REFER TO Attachment 2 to VERIFY the associated valves close.
Attachment 1 PY-CEI/OIE-O63OL Page 77 of 87 Instruction Number PERRY NUCLEAR POWER PLANT ONLN35
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 8of 17 NA 4.1 I IF level continues to increase in 2A 2B 2C feedwater heater 2 level rises to the 0 high-high setpoint, THEN REFER TO Attachment 2 to VERIFY the associated valves close.
NA 4.12 IF the hot surge tank level rises to the high-high setpoint, 0
THEN REFER TO Attachment 2 to VERIFY the associated valves close.
NOTE Recovery actions for feedwater heater #4 isolation are located in ARI-HI 3-P680-2-E1.
NA 4.13 IF Heater 1 AND 2 isolates on a A B C high-high level in either heater, 0
THEN MONITOR level in the isolated heaters.
NA 4.14 IF necessary, THEN OPEN the following shell side maintenance drain valves to prevent the isolated heater from flooding into the turbine blading.
0 Heater 1A Maintenance Drain to 1N26-Condenser F538A 0 0 ' Heater 1B Maintenance Drain to 1N26-Condenser F538B n 0 Heater I C Maintenance Drain to 1 N26-Condenser F538C 0 Heater 2A Maintenance Drain to lN26-Condenser F532A 0 0 Heater 2B Maintenance Drain to 1N26-Condenser F532B
Attachment 1 PY-CEUOIE-0630L Page 78 of 87 instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36 Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 9 of 17 0 Heater 2C Maintenance Drain to 9 N26-Condenser F532C NA 4.15 THEN PERFORM the following:
0 4.1 5-1 REFER TO Attachment 2 to VERIFY the extraction steam block valve is open.
0 4.15-2 CONFIRM NO extraction steam leak by investigating the following:
Alarms Sump levels Radiation levels NA 4.15.3 IF a leak has occurred, THEN REFER TO ONI-Nl1, Pipe a Break Outside Containment.
5 4.16 REDUCE main generator loading to within the allowable limits listed below:
Heater Number of Trains Lost Side of Heater Lost RFP Steam Supplv Main Extraction Basis I& 2 I Condensate 1125 1188MWe 1 MWe 5 2 Extraction 938 1000Mwe 2 MWe
Attachment 1 PY-CEI/OIE-O80L Page 79 of 87 I
Title:
PERRY NUCLEAR POWER PLANT Instruction Number:
Use Category:
ONI-N36 I LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 10 of 17 0 4.17 REDUCE reactor power to within the allowable limits listed below:
Heater Number of Trains Lost Side of Heater Lost -
Limit -
Basis 1842 1 Condensate 18,000 gpm 4 Condensate flow 1& 2 2 Condensate 9,000 gpm 4 Condensate flow 3 2 Condensate 11,900 gpm 3 Condensate flow 5, 6 2 Trains of the same Feedwater 18,400 gpm Feedwater 3 heater flow Bases for limits:
1 To prevent undue loading and overstressing of any turbine part.
(Isolation of extraction steam changes Main Turbine Stage pressures, stage pressure drops, and steam flow through the turbine so that bucket, diaphragm, and thrust loads are affected.)
2 To prevent the overloading of Feedwater Heaters 6A and 6B 3 Bypass line flow limitations 4 To minimize flow induced vibrations in the feedwater heater tubes 4.18 REFER TO SOI-N27 AND/OR SOI-N21 to LIME UP Isolated heaters.
Attachment 1 PY-CEUOIE-0630L Page 80 of 87 PERRY NUCLEAR POWER PLANT
Title:
LOSS OF FEEDWATER HEATING Revision: Page
~~
8 11 of 17 NA 4.19 IF desired, THEN REFER TO the following to 0
RETURN isolated heaters to service as applicable:
SOI-N27 SOI-N21 ARI-H13-P680-2 window E l NA 4.20 In single recirculation loop operation Final feedwater temperature is below the 425°F curve in Feedwater Temperature Versus Core Thermal Power, PDB-A0011 0 THEN REFER TO 101-3 to REDUCE reactor power to 5 23.8% (894 MWt) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
NOTE For planned reductions in feedwater temperature, FTI-BO010 must be performed prior to feedwater temperature reduction.
NA 4.21 IF feedwater heaters have been returned to service, 0
THEN REFER TO 101-3 to RESTORE reactor power level.
N A 4.22 IF the isolated heaters are NOT returned to service, 0
THEN REFER TO FTI-B0010, Preparation for Final Feedwater Temperature Reduction Operation.
Attachment I PY-CEUOIE-0630L Page 81 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 12 of 17 NA 4.23 IF the appropriate actions of FTI-BO010 have been completed, R
THEN REFER TO 101-3 to RESTORE reactor power level.
5.0 REFERENCES
Commitment BO0509 - Steps 4.2,4.16 Commitment FOI 554 - Steps 3.1,4.2 Commitment BO0916 - Caution for Step 3.1, Step 3.1 6.0 RECORDS The following records are completed@eneratedby this document:
Quality Assurance Records None Non-Quality Assurance Records None 7.0 SCOPE OF REVISION Rev. 8 1. Deleted immediate action to scram when in the increased awareness region.
- 2. Deleted actions to insert cram rods as a duplicate action of ONI-C51.
- 3. Deleted actions for verification of power limits and moved them to ON I-SPI-G4.
- 4. Reversed the order of Steps 4.21 and 4.22.
Attachment 1 PY-CEUOIE-0630L Page 82 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 13 of 17 8.0 ATTACHMENTS ATTACHMENT 1 - FEEDWATER HEATER NORMAL AND ALTERNATE DRAINS ATTACHMENT 2 - FEEDWATER HEATER ISOLATION ATTACHMENT 3 - TECHNICAL SPECIFICATION DISCUSSION
Attachment 1 PY-CEYOIE-063OL Page 83 of 87 PERRY NUCLEAR POWER PLANT Instruction Number:
ONI-N36 1 I
Title:
LOSS OF FEEDWATER HEATING 1 Use Category:
In Field Reference Revision: Page I
8 14 of 17 ATTACHMENT 1 - FEEDWATER HEATER NORMAL AND ALTERNATE DRAINS Page 1 of 1 Heater Normal drain valve Alternate drain valve 6A 1N25-F290A IN25-F280A 6B 1N25-F290B 1N25-F280B 5A 1N25-F340A 1N25-F330A 5B 1N25-F340B 1N25-F330B 3A IN26-Fl20A 1N26-Fl80A 1N26-Fl30A 3B 1N26-Fl20B 1N26-Fl80B IN26-FI308 2A 1N26-FO50A 1N26-FO70A 2B 1N26-FO50B 1N26-FO70B 2c 1N26-FO50C 1N26-FO70C 1A 1N26-FOIOA 1N26-FO30A 1B 1N26-FOIOB 1N26-FO30B IC 1N26-FOIOC 1N26-FO30C
Attachment 1 PY-CEVOIE-063OL Page 84 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
LOSS OF FEEDWATER HEATING Revision: Page 8 15 of 17 ATTACHMENT 2 - FEEDWATER HEATER ISOLATION Page 1 of 2 Extraction steam Positive assist Associated valves block valve check valve 1N36-FI20A I 1N36-F40A I IN25-Fl40A 1N25-FI70A MSR 1A
~ ~ _ _ _ _ __
6B 1N36-Fl20B 1N36-FI40B IN25-FI40B 1N25-FI70B MSR 1B IN25-F250B
~~ ~~~
1N25-FI45B 1N25-FI75B MSR 2B IN25-F255B 1 5A I IN36-F435A IN36-F455A I 1N25-F290A I HTR 6A I I 5B I IN36-F435B 1N36-F455B I IN25-F290B 1 HTR6B [
1N36-F260 I IN36-F250A 1N36-F250B 1N33-F160 I 1N36-F380A 1N36-F390A None 1N36-F380B 1N36-F390B None None None 1N26-FI20A HTR 3A 1N21-FI45A Condensate IN2 1-F170A None None 1N26-FI30A HTR 3A IN26-FI30B HTR 3B 1N21-FI45B Condensate lN21-Fl70B
Attachment 1 PY-CEVOIE-0630L Page 85 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36 I
Title:
Use Category:
LOSS OF FEEDWATER HEATING In Field Reference Revision: Page 8 16 of 17 ATTACHMENT 2 - FEEDWATER HEATER ISOLATION Page 2 of 2 Positive assist Associated valves block valve check valve 2c None None 1N26-Fl20B HTR 38 1N21-Fl45C Condensate 1N21-F170C 1A None None 1N26-FO50A HTR 2A 1N21-F145A Condensate IN21-F170A
~~
IB None None IN26-FO50B HTR 2B 1N21-F145B Condensate 1N21-F170B IC None None 1N26-FO50C HTR 2C
~ _ _
IN21-F145C Condensate 1N21-F170C
Attachment 1 PY-CEI/OIE-O63OL Page 86 of87
~~
Instruction Number:
PERRY NUCLEAR POWER PLANT ONI-N36
Title:
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LOSS OF FEEDWATER HEATING In Field Reference Revision: Page a 17 of 17 ATTACHMENT 3 - TECHNICAL SPECIFICATION DISCUSSION Page 1 of 1 1.O For unplanned reductions in feedwater temperature, comply with T.S.3.3.2.1 Action A for the Control Rod Block Instrumentation.
2.0 For unplanned reductions in feedwater temperature comply with one of the following four options for the RPS and EOC-RPT Instrumentation:
Restore the feedwater temperature within the Tech Spec Action Time@)
Disable the bypass (i.e., arm the Function) in accordance with the PDB-I0010 for RPS and EOC-RPT Implement the setpoint changes in accordance with FTI-BO010 within the Tech Spec Action Time(s)
Reduce power to 5 38% within the Tech Spec Action Time(s) 3.0 T.S. affected when feedwater heaters are removed from service causing a reduction in feedwater temperature T.S. 3.3.1 .I,RPS Instrumentation Channels A, B, C, D, E, F, G, and H for Turbine Stop Valve Closure, Table 3.3.1.1-1, Function 9.
0 T.S. 3.3.1 .I,RPS Instrumentation Channels A, B, C, and D for Turbine Control Valve Fast Closure, Trip Oil Pressure-Low. Table 3.3.1.1-1, Function 10.
T.S. 3.3.2.1, Control Rod Block Instrumentationfor Rod Withdrawal Limiter (RWL) Table 3.3.2.1-1, Function 1.a. Control Rod movement is prohibited with a RWL channel inoperable.
0 T.S. 3.3.4.1, End of Cycle Recirculation Pump Trip (EOC-RPT)
Instrumentation.
Attachment 1 PY-CEYOIE-0630L Page 87 of 87 Instruction Number:
PERRY NUCLEAR POWER PLANT SOI-N21 Section 6.2
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CONDENSATE SYSTEM In Field Reference Revision: Page a 20 of 60 0 6.2.1 CONFIRM the Reactor Feedwater Booster pumps are NOT in operation.
a 6.2.2 CONFIRM the Alternate Hot Surge Tank Level Control is NOT in effect.
R 6.2.3 CONFIRM the lN21-FZ20 VLV POSIT 1N21-indicates closed. R709 0 6.2.4 PLACE the HOT SURGE TANK LEVEL CONTROL 1N21-F230 in MANUAL on 1N21-1H13-P680. R475 a 6.2.5 CLOSE HOT SURGE TANK LEVEL CONTROL valve IN21-F230. 1N21-R475 Cl 6.2.6 CLOSE the Hot Surge Tank LCV Outlet 1N21-Is01Valve. F54I R 6.2.7 REFER TO SOLN27 and VERIFY RFP Seal injection Pumps Shutdown.
0 6.2.8 CLOSE the RFBPs Seal Wtr Supp 1N27-Press Reg Inlet Isol. F800 0 6.2.9 CLOSE the RFBPs Seal Wtr Supp 1N27-Press Reg Disch Bypass. F802 END OF SECTION