ML092920184

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Initial Exam 280,281/2009-301 Post Exam Comments
ML092920184
Person / Time
Site: Surry  Dominion icon.png
Issue date: 07/31/2009
From: Gerald Bichof
Virginia Electric & Power Co (VEPCO)
To:
Region 2 Administrator
References
09-510, 50-280/09-301, 50-281/09-301
Download: ML092920184 (25)
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Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 July 31 , 2009 Regional Administrator Serial No.09-510 U. S. Nuclear Regulatory Commission Docket Nos. 50-280 Region II 50-281 Atlanta Federal Center License Nos. DPR-32 61 Forsyth Street, S. W., Suite 23T85 DPR-37 Atlanta, Georgia 30303-8931

Dear Mr. Reyes:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 WRITTEN LICENSE EXAMINATION COMMENTS In accordance with NUREG-1021 , Section ES-402, the following comments are submitted concerning the Reactor Operator and Senior Reactor Operator written initial examinations administered at Surry on July 29,2009.

SRO QUESTION: #86 (KIA 015/017.AG2.2.22)

The question as given to the applicants follows:

Initial plant conditions on Unit 2 are as follows:

  • A power increase is in progress following reactor startup.
  • Reactor power is at 8%.
  • Pressurizer Spray valve 2-RC-PCV-2455A cannot be opened.
  • All three RCPs are operating.
  • Normal Charging has been tagged out at 1-CH-MOV-1289A and excess letdown is in-service due to required charging line piping repairs.

Current plant conditions on Unit 2 are as follows:

  • RCP 'C' trips on ground overcurrent.

Based on the above conditions, which ONE of the following describes whether action statements of the following LCOs are required to be performed:

Action statements within ...

a. Technical Specification Section 3.1.A.4 is required.

Technical Specification Section 3.1.A.5 is NOT required.

b. Technical Specification Section 3.1.A.4 is NOT required.

Technical Specification Section 3.1.A.5 is required.

c. both Technical Specification Section 3.1.A.4 and 3.1.A.5 are required.
d. neither Technical Specification Section 3.1.A.4 nor 3.1.A.5 are required.

Current ANSWER: (C)

COMMENTS:

Issue #1 Given the Unit conditions presented in the stem of the question, Technical Specification 3.1.A.4 is NOT required and Technical Specification 3.1.A.5 is applicable.

Technical Specification 3.1.A.4 (Reactor Coolant Loops) is applicable for Loop Stop Valves and the associated Reactor Coolant Loops, not Reactor Coolant Pumps.

Technical Specification 3.1.A.1 (Reactor Coolant Pumps) is the correct Technical Specification that would be applicable for the conditions given within the question (Le.,

loss of motive force/RCP in the 'C' Reactor Coolant Loop). As such, Technical Specification 3.1.A.4 is not applicable.

Additionally, although Technical Specification 3.1.A.4 states that, "POWER OPERATION with less than three loops in service is prohibited" this statement is made with regards to Reactor Coolant Loops, vice Reactor Coolant Pumps. This is clarified within the specification as the follow-on statements are related only to Reactor Coolant Loop Isolation Valves, as is the basis section of this Technical Specification. Additional clarification for returning a Reactor Coolant Loop to service is contained within Technical Specification 3.17 (Loop Stop Valve Operation). Technical Specification 3.17 states that in order to retum a loop to service (Le., have a loop in service) certain items must be met (see highlight areas in Attachment 1). When discussing returning a Reactor Coolant Loop to service in Technical Specification 3.1.A.4. and 3.17 (and their associated basis sections), there is no discussion on Reactor Coolant Pump Operation. Finally, statements within UFSAR Chapter 4 support this argument. contains the associated Technical Specifications and Basis information to support this conclusion.

Issue #2 The question stem asks the candidate to determine, "Based on the above conditions, which ONE of the following describes whether action statements of the following LeOs are required to be performed'. Technical Specification 3.1.A.4 has no applicable action statements to be performed. As such, action statements within Technical Specification Section 3.1.A.4 are NOT required.

RECOMMENDATION:

Issue 1 and 2 resolution:

Based on the above information the correct answer for this question should be (B), as Technical Specification 3.1.A.4 is not applicable and 3.1.A.5 is applicable. Additionally, Technical Specification 3.1.A.4 contains no action statements to be performed.

REFERENCES:

  • Excerpts from UFSAR Chapter 4

SRO QUESTION: #93 (KIA 079.G2.2.22)

Given the following plant conditions:

  • Unit 1 is at 100%
  • A loss of Containment Instrument Air has occurred
  • 1D-C6, PRZR PWR RELIEF W LO AIR PRESS, annunciates
  • Containment Instrument Air was crosstied with Instrument Air
  • Containment Instrument Air Pressure = 85 psig and increasing
  • All Pressurizer PORV air bottles are properly aligned with air pressures of 1050 psig Which ONE of the following correctly states (1) the status of LCO 3.1.A.6 (PORV Operability) and (2) the Technical Specification required operator actions, if any?
a. (1) The LCO is met.

(2) No further action associated with the Pressurizer PORVs is required.

b. (1) The LCO is met.

(2) Verify Pressurizer PORV operability by closing Pressurizer PORV Block Valves, manually cycle the Pressurizer PORVs, and then re-open the Pressurizer PORV Block Valves.

c. (1) The LCO is NOT met.

(2) Restore the Pressurizer PORV backup air supply within 14 days OR be in HSD within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. (1) The LCO is NOT met.

(2) Close and remove power from both Pressurizer PORV block valves within one hour AND be in HSD within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Current ANSWER: (A)

COMMENTS:

Given the Unit conditions presented in the stem of the question, Technical Specification 3.1.A.6 is NOT met.

The conditions given in the question state that annunciator 10-C6 (PRZR PWR RELIEF W LO AIR PRESS) has actuated and that Pressurizer PORV air bottles are pressurized to 1050 psig. Since annunciator 1O-C6 will only annunciate when Pressurizer PORV air bottle pressure is less than 1000 psig or when Pressurizer PORV Back-up Air System (Le., air pressure downstream of the pressure regulator) is less than 80 psig, it can logically be concluded that the cause of the alarm was due to low air pressure downstream of the pressure regulator.

Since air pressure downstream of the pressure regulator is less than 80 psig, the PORVs must be declared inoperable in accordance with ARP 1O-C6 Step 4 and Technical Specification 3.1.A.6. This means that the LCO for 3.1.A.6 is NOT met and that the required operator actions are to restore the Pressurizer PORV backup air supply within 14 days OR be in HSO within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Attachment 2 contains the associated Technical Specification and Procedure to support this conclusion.

RECOMMENDATION:

Based on the above information the correct answer for this question should be (C), the ARP and plant conditions show the Technical Specification is NOT met.

REFERENCES:

If you have any questions or require additional information, please contact us.

Attachments

  • Attachment 1 - Supporting Reference material for Question 86
  • Attachment 2 - Supporting Reference material for Question 93

copy:

Mr. Malcolm Widmann United States Nuclear Regulatory Commission Region" Sam Nunn Atlanta Federal Center 61 Forsyth Street, S. W., Suite 23T85 Atlanta, Georgia 30303-8931 Document Control Desk United States Nuclear Regulatory Commission Washington, D.C. 20555-0001 Senior Resident Inspector Surry Power Station

ATTACHMENT 1 WRITTEN LICENSE EXAMINATION COMMENTS REFERENCE MATERIAL TO SUPPORT COMMENTS SRO QUESTION: #86 Surry Power Station - Units 1 & 2 VIRGINIA ELECTRIC AND POWER COMPANY

Attachment 1 - Reference Material for Question #86 TS 3.1-1 08-03-95 3.1 REACTOR COOLANT SYSTEM Applicability Applies to the operating status of the Reactor Coolant System.

Objectives To specify those limiting conditions for operation of the Reactor Coolant System which must be met to ensure safe REACTOR OPERATION.

These conditions relate to: operational components. heatup and cooldown. leakage, reactor coolant activity, oxygen and chloride concentrations, minimum temperature for criticality. and Reactor Coolant System overpressure mitigation.

A. Operational Components Specifications I. Reactor Coolant Pumps

a. A reactor shall not be brought critical with less than three pumps. in non-isolated loops. in operation.

Amendment Nos. 203 and 203 Page 1 of 8

Attachment 1 - Reference Material for Question #86 TS 3.1-4 05-31-95

4. Reactor Coolant Loops
a. Loop stop valves shall not be closed in more than one loop unless the Reactor Coolant System is connected to the Residual Heat Removal System and the Residual Heat Removal System is OPERABLE.
b. POWER OPERATION with less than three loops in service is prohibited The following loop isolation valves shall have AC power removed and their breakers locked. sealed or otherwise secured in the open position during POWER OPERATION:

Unit No. I Unit No.2 MOY 1590 MOY 2590 MOY 1591 MOY2591 MOY 1592 MOY 2592 MOY 1593 MOY 2593 MOY 1594 MOY 2594 MOY 1595 MOY 2595

5. Pressurizer
a. The reactor shall be maintained subcritical by at least I % until the steam bubble is established and the necessary sprays and at least 125 KW of heaters are operable.
b. With the pressurizer inoperable due to inoperable pressurizer heaters. restore the inoperable heaters within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the Reactor Coolant System temperature and pressure les'>

than 350°F and 450 psig, respectively, within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

c. With the pressurizer otherwise inoperable. be in at lea"t HOT SHUTDOWN with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the Reactor Coolant Sy'>tem temperature and pressure less than 350°F and 450 psig. respectively.

within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Amendment Nos. 199 and 199 Page 2 of 8

Attachment 1 - Reference Material for Question #86 The following paragraph contains the basis section for Technical Specification 3.1.A.1:

Ba.,is Specification 3.I.A-l requires that a sutlicient number of reactor coolant pumps be operating to provide coastdown core cooling now in the event of a loss of reactor coolant flow accident. This provided flow will maintain the DNBR above the applicable design Iimit.(l) Heat transfer analyses also show that reactor heat equivalent to approximately 10% of rated power can be removed with natural circulation; however. the plant is not designed for critical operation with natural circulation or one loop operation and will not be operated under these conditions.

The following paragraph contains the basis section for Technical Specification 3.1.A.4:

The limitation specified in item 4 above on reactor coolant loop isolation will prevent an accidental isolation of all the loops which would eliminate the capability of dissipating core decay heat when the Reactor Coolant System is not connected to the Residual Heat Removal System.

Page 3 of 8

Attachment 1 - Reference Material for Question #86 TS3.17-1 04-22-93 3.17 LOOP STOP VALVE OPERATION Applicability Applies to the operation of the loop stop valves.

Objective To specify those limiting conditions for operation of the loop stop valves which must be met to ensure safe reactor operation.

Specifications I. The loop stop valves shall be maintained open unless the reactor is in COLD SHUTDOWN or REFUELING SHUTDOWN.

2. A hot or cold leg stop valve in a reactor coolant loop may be closed in COLD SHUTDOWN or REFUELING SHUTDOWN for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for valve maintenance or testing. If the stop valve is not opened within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. the loop shall be isolated.
3. Whenever a reactor coolant loop is isolated. the stop valves of the isolated loop shall have their AC power removed and their breakers locked open. *
4. Whenever an isolated and fi11ed reactor coolant loop is returned to service. the following conditions shall be met:
a. A source range nuclear instrumentation channel shall be operable and continuously monitored with audible indication in the control room during opening of the hot leg loop stop valve, during relief line flow. and when opening the cold leg stop valve in the isolated loop. Should the count rate increase by more than a factor of two over the initial count rate, the hot and cold leg stop valves shaH be re-closed and no attempt made to open the stop valves until the reason for the count rate increase hac'> been determined.
  • Power may be restored to a hot or cold leg loop stop valve in an isolated and filled loop provided the requirements of Specifications 4.b or 4.c are met. respectively. Power may be restored to a loop stop valve in an isolated and drained loop provided the requirements of Specifications 5.a and b are met.

Amendment Nos. t 77 and 176 Page 4 of 8

Attachment 1 - Reference Material for Question #86 Basis Section of Technical Specification 3.17 Basis The Reactor Coolant System may be operated with isolated loops in COLD SHUTDOWN or REFUELING SHUTDOWN in order to perform maintenance. A loop stop valve in any loop can be closed for up to two hours without restriction for testing or maintenance in these operating conditions. While operating with a loop isolated. AC power is removed from the loop stop valves and their breakers locked opened to prevent inadvertent opening. When the isolated loop is returned to service. the coolant in the isolated loop TS 3.17-5 05-22-01 mixes with the coolant in the active loops. This situation has the potential of causing a positive reactivity addition with a corresponding reduction of shutdown margin if:

a. The temperature in the isolated loop is lower than the temperature in the active loops (cold water accident). or
b. The boron concentration in the isolated loop is insufficient to maintain the required shutdown margin (boron dilution accident).

The return to service of an isolated and filled loop is done in a controlled manner that precludes the possibility of an uncontrolled positive reactivity addition from cold water or boron dilution. A tlow path to mix the isolated loop with the active loops is established through the relief line by opening the hot leg stop valve in the isolated loop and starting the reactor coolant pump. The relief line tlow is low enough to limit the rate of any reactivity addition due to differences in temperature and boron concentration between the isolated loop and the active loops. In addition. a source range instrument channel is required to be operable and continuously monitored to detect any change in core reactivity .

Page 5 of 8

Attachment 1 - Reference Material for Question #86 Basis Section of Technical Specification 3.17 (continued)

The limiting conditions for returning an isolated and filled loop to service are as follows:

a. A hot leg loop stop valve may not be opened unless the boron concentration in the isolated loop is greater than or equal to the boron concentration corresponding to the shutdown margin requirements for the active portion of the Reactor Coolant System.
b. A cold leg loop stop valve can not be opened unless the hot leg loop stop valve is open with relief line flow established for at least 90 minutes at greater than or equal to 125 gpm. In addition. the cold leg temperature of the isolated loop must be at least 70°F and within 20 0 P of the highest cold leg temperature of the active loops. The boron concentration in the isolated loop must be verified to be greater than or equal to the boron concentration corresponding to the shutdown margin requirements for the active portion of the Reactor Coolant System.
c. A source range nuclear instrument channel is required to be monitored to detect any unexpected positive reactivity addition during hot or cold leg stop valve opening and during relief line now.

Amendment Nos. 226 and 226 Page 6 of 8

Attachment 1 - Reference Material for Question #86 UFSAR Chapter 4:

Revision 4O.S-Updated Online 06/30109 SPS UFSAR 4.2-13 Thermal sleeves are installed at the following locations where high thermal stresses could otherwise develop due to rapid changes in tluid temperature during normal operational transients:

I. Return line from the residual heat removal loop.

2. Both ends of the pressurizer surge line.
3. Pressurizer spray line connection to the pressurizer.
4. Charging line connection.
5. Loop fill header connections to each loop.

4.2.2.7 Small Valves All valve surfaces in contact with reactor coolant are austenitic stainless steel or equivalent corrosion-resistant materials. Connections to stainless steel piping are welded Valves that perform a modulating function are equipped with two sets of packing and an intermediate leakoff connection.

4.2.2.S Loop Stop Valves The reactor coolant loop stop valves. one of which is shown on Figure 4.2-S. are remotely controlled. motor-operated gate valves that permit any loop to be isolated from the reactor vessel during cold or refueling shutdowns. A stop valve is installed on each hot leg and in each cold leg.

During return to service of an isolated filled loop. coolant is circulated through a bypass line.

which contains a remotely controlled. motor-operated stop valve. This bypass valve is closed during normal loop operation. A valve pump interlock circuit prevents the starting of the reactor coolant pump in a given loop unless either (a) both hot leg and cold leg loop stop valves are open or (b) the cold leg loop stop valve is closed and the bypass valve is open. The interlock also prevents pump operation if the bypass valve and either of the stop valves are closed.

To ensure against an accidental start-up of an un borated andlor cold isolated loop. an additional valve interlock system is provided that meets the IEEE-279 Criteria/or Nuclear Power Plant Protection Systems, August 1968. This is shown on Reference Drawing 1. which indicates a relief line and bypass around the cold-leg stop valve. These additional valve temperature and tlow interlocks require that a controlled tlow of reactor coolant is circulated through the relief line of the inactive loop insuring that boron concentration and temperature of the isolated loop are brought to equilibrium with the remainder of the reactor coolant system. prior to opening the cold leg loop stop valve. This controlled t10w will minimize the possibility of a sudden reactivity addition from cold water or boron dilution.

The valve-temperature and valve-t1ow relief line interlocks are provided to:

I. Prevent opening of a hot-leg loop stop valve unless the cold-leg loop stop valve is closed.

TN. page was publleI1ed 8lectrtt1ically fOr use n DoctnIenllJm. The nformatlOn contained In

!he DoctJmentlll1 ~ 01 the UFSAR trey be _ 1 from !he InfoonatlOn fOUnd In the hardcopy version 0/ the UFSAR SUCh dIt!8r8nces 1ft ntantlOnoJ am oro the resUlt oIoppn:r.red changes 10 the UFSAR thel heve not)'81 been submitted to !he NRC.

Page 7 of 8

Attachment 1 - Reference Material for Question #86 UFSAR Chapter 4 (continued):

Revision 40.8-Updated Online 06/30/09 SPS UFSAR 4.2-14

2. Prevent opening of a cold-leg loop stop valve unless:
a. The hot-leg loop stop valve has been opened a specified time.
b. The loop bypass valve has been opened a specified time.
c. Flow has existed through the relief line for a specified time.
d. The cold-leg temperature is within 20 0 P of the highest cold-leg temperature in other loops and the hot-leg temperature is within 20 0 P of the highest hot-leg temperature in the other loops.

Returning an isolated loop to service requires that the above interlocks be satisfied. a minimum temperature exists in the loop. and that core reactivity be monitored using a source range nuclear instrument channel.

If a loop was initially drained. the above interlocks can be bypassed. The initially isolated and drained loop may be returned to service by partially opening a loop stop valve and filling the loop in a controlled manner from the reactor coolant system. If using the Volume Control Tank (VCT) as the makeup source. the charging now from the VCT is periodically sampled during the backtill evolution to ensure its boron concentration meet" the minimum refueling water boron concentration requirement established by Technical Specification 3.IO.A.9. Makeup to the Reactor Coolant System solely through auxiliary spray during the backfill evolution is prohibited to ensure that a sufficient fraction of makeup now is mixed with coolant in the active Reactor Coolant System volume and nows through the core. where the source range instrumentation is available to provide secondary indication of improperly blended makeup flow. The vacuum-assisted backfill evolution involves initiation of reactor coolant pump seal injection in the isolated and drained loop to allow establishment of a partial vacuum prior to partially opening the cold leg loop stop valve. The following controls are required to assure that no sudden positive reactivity addition or loss of reactor coolant system inventory occurs during the backfill evolution:

I. Only one loop should be tllled at a time.

2. The isolated loop must be verified to be drained.
3. Adequate reactor coolant inventory exists to assure that. during the till operation. decay heat removal is maintained. This minimum inventory level should not be violated during the fill operation.

If this method is used to till a loop. then the loop is no longer considered to be isolated and the requirements for returning the isolated loop to service are not applicable as long as the loop stop valves are opened within a specified time.

The parameters of each reactor coolant loop stop valve are shown in Table 4.1-7.

This paoo .... ~ oloottr1lcally lor use " Docoo>onttJm. The Worrnotloo oon1alned In

!he Document"" _ at Ihe UFSAR may be dlflerent from tho InIomMltlon found In the hardCopy wrslon 01 the UFSAR. S u c l 1 _ . "'" "lBntlonoJ and are tho reaull 01 ~od changH to 1M UFSAR that he"" not)'Ot boon IUbmItlod to tho NRC.

Page 8 of 8

A ITACHMENT 2 WRITTEN LICENSE EXAMINATION COMMENTS REFERENCE MATERIAL TO SUPPORT COMMENTS SRO QUESTION: #93 Surry Power Station - Units 1 & 2 VIRGINIA ELECTRIC AND POWER COMPANY

Attachment 2 - Reference Material for Question #93 SURRY POWER STATION ANNUNCIATOR RESPONSE PROCEDURE NUMBER PROCEDURE TITLE REVISION 9

1D-C6 PRZR PWR RELIEF W LO AIR PRESS PAGE 1of6 REFERENCES 1D-22

1) UFSAR 4.0
2) 11448-ESK-10C, 10AN
3) Tech Spec3.1.G, 3.1.A
4) 1-DRP-005, Instrument Setpoints
5) DCP 93-054-3, Reassessed CRDR - Outage Related (Relocation)
6) DCP 93-037-3, PORV Backup Air System Modifications
7) Tech Spec Amendment 198
8) 11448-FM-75E, Sheet 2
9) Tech Spec Amendment 231 10)TRM 3.7.18 11 )PI S-2005-3320, PORV Air Bottle Pressure 12)DCP 05-043, PRZR PORV Air Bottle Cameras 13)PI S-2006-1699, PORV Air Bottle Pressure PROBABLE CAUSE
1) Alarm actuates when PS-IA-104A or PS-IA-1 04B senses backup air bottle pressure less than or equal to 1000 psig.
2) Alarm actuates when PS-IA-103A or PS-IA-103B senses PORV Backup Air System pressure less than or equal to 80 psig downstream of air bottle regulator.
3) Instrumentation failure has occurred.

CONTINUOUS USE Page 1 of 9

Attachment 2 - Reference Material for Question #93 NUMBER PROCEDURE TITLE REVISION 9

1D-C6 PRZR PWR RELIEF W LO AIR PRESS PAGE 20f6 ACTIONI EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE:

  • During normal operation. a PORV Is inoperable If less than two air bottles are in service or bottle pressure is less than 1000 pslg. If two air bottles (each with greater than 1000 psig) cannot be maintained aligned to the associated PORV. the applicable Tech Spec clock must be entered.
  • When OPMS is required. a PORV is inoperable if two bottles are in service with less than 1269 psig or three bottles are in service with less than 1000 psig. If two air bottles (each with greater than 1269 psig) OR three bottles (each with greater than 1000 psig) cannot be maintained aligned to the associated PORV. the applicable Tech Spec clock must be entered.
  • The PORV remote monitors should be used to verify alarm before declaring PORVs inoperable or making CTMT entry. If bottle pressure and regulator pressure unavailable on monitors, this procedure should be initiated starting with Step 2.
1. CHECK PORV AIR BOTTLE AND Do the following:

REGULATOR PRESSURE ON REMOTE MONITORS o a) Increase surveillance of local monitors.

D

  • Either in-service air bottle pressure - D b) Submit Condition Report to investigate LESS THAN OR EQUAL TO 1000 PSIG annunciator.

OR c) !E air bottle pressure greater than 1000 psig. but less than required D

  • Either regulator downstream pressure -

based on notes above, THEN do the LESS THAN OR EQUAL TO 80 PSIG following:

D 1) !E RCS temperature less than 350°F.

THEN GO TO Step 2 RNO.

Air Bottle pressure given as 1050 D 2) !E RCS temperature greater than or psig. Therefore regulator equal to 350°F. THEN GO TO Step 3.

downstream pressure must be less D d)!E air bottle pressure adequate, than or equal to 80 psig. THEN GO TO Step 14.

Page 2 of 9

Attachment 2 - Reference Material for Question #93 NUMBER PROCEDURE TITLE REVISION 9

10-C6 PRZR PWR RELIEF W LO AIR PRESS PAGE 30f6 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: The PRZR PORVs are required for the mitigation of a steam Generator tube rupture when the RCS is greater than 350°F.

2. CHECK RCS TEMPERATURE - GREATER i.E the vessel head is bolted AND the THAN OR EQUAL TO 350°F PORVs are providing OVerpressure Protection, THEN do the following:

Plant conditions were given as o a) Declare PORVs inoperable.

100% power. o b) Review Tech Spec3.1.G o c) GO TO step 5.

IF the vessel head is NOT bolted AND the PORVs are NOT providing Overpressure Protection, THEN do the following:

o a) Locally check PORV air bottle and regulator pressure.

o b) GO TO step 14.

Page 3 of 9

Attachment 2 - Reference Material for Question #93 NUMBER PROCEDURE TITLE REVISION 9

PRZR PWR RELIEF W LO AIR PRESS 1D-C6 PAGE 40fS ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE: A PORV is capable of being manually cycled if CTMT IA pressure is greater than 80 psig and power is available to the PORV.

  • 3. CHECK BOTH PORVS - CAPABLE OF Do the following:

BEING MANUALLY CYCLED a) Within one hour, close and deenergize o . Normal CTMT IA pressure greater the PORV Block valve for PO RV(s ) which than 80 psig - AVAILABLE (PI-IA-1 01) can NOT be manually cycled:

AND o

  • Power to PORVs - AVAILABLE o

o

the PORV power supplies, so they are available. o d) GO TO Step 5.

4. DECLARE PRZR PORVS INOPERABLE DUE

- TO INOPERABLE AIR SUPPLY AND START 14 DAY CLOCK lAW TECH SPEC 3.1.A.S.f

5. REVIEW TRM 3.7.18 Page 4 of 9

Attachment 2 - Reference Material for Question #93 NUMBER PROCEDURE TITLE REVISION 9

1D-C6 PRZR PWR RELIEF W LO AIR PRESS PAGE Sof6 ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED NOTE:

  • There are four backup air bottles for each PORV. Two are normally valved in and two are normally isolated.
  • Backup air bottle pressure Is needed for long-term ability to cycle PORVs. If PORV air bottle pressure is less than 1000 psig or downstream regulator pressure is less than 80 psig and the RCS Is greater than 350°F, a 14 day clock to restore the PORV(s) backup air supply(ies) must be started. This clock may be stopped when the low pressure condition is corrected.
6. LOCALLY OR REMOTELY CHECK 0 !E air bottle pressure greater IN-SERVICE AIR BOTTLE PRESSURE ON than 1000 psig, but less than required, BOTH PORVs - EITHER LESS THAN 1000 THEN GO TO step 7.

PSIG

!E pressure downstream of either PORV bottle regulator less than 80 psig, THEN do the following:

0 a) Direct 1& C to adjust regulator(s).!E greater than 80 psig obtained, THEN GO TO step 12.

b) IF pressure can NOT be increased to greater than 80 psig, THEN do the following:

0 1) Initiate Condition Report to repair failed regulator(s).

0 2) GO TO step 14.

0 !E pressure downstream of both regulators greater than 80 pSig, THEN initiate a Condition Report AND GO TO step 12.

7. LOCALLY INVESTIGATE AND IDENTIFY AIR LEAKS AS NECESSARY
8. SWAP PORV AIR BOTTLES lAW 1-0P-IA-007, SWAPPING PRZR PORV AIR BOTTLES Page 5 of 9

Attachment 2 - Reference Material for Question #93 NUMBER PROCEDURE TITLE REVISION 9

PRZR PWR RELIEF W LO AIR PRESS 10-C6 PAGE 60f6 ACTION! EXPECTED RESPONSE RESPONSE NOT OBTAINED

9. CHECK AIR BOTTLE PRESSURE ON BOTH o GO TO Step 14.

PORVs - GREATER THAN REQUIRED PRESSURE

10. CHECK PRESSURE DOWNSTREAM OF o Direct I & C to adjust regulator( s). !E greater BOTH BOTTLE REGULATORS - GREATER than 80 psig obtained, THEN THAN 80 PSIG GO TO Step 11.

!.E pressure can NOT be increased to greater than 80 psig, THEN do the following:

o a) Initiate Condition Report to repair failed regulator( s).

o b) GO TO Step 14.

11. INITIATE A CONDITION REPORT TO HAVE EMPTY AIR BOTTLES REPLACED
12. VERIFY COMPLIANCE WITH APPLICABLE TECH SPEC
13. DECLARE PORVs OPERABLE
14. PROVIDE NOTIFICATIONS AS NECESSARY:

o . OMOC o . STA o . Shift Supervision

- END-Page 6 of 9

Attachment 2 - Reference Material for Question #93 TS 3.1-4a 09-17-08

6. Relief Valves Two power operated relief valves (PORVs) and their associated block valves shall be OPERABLE* whenever the Reactor Coolant System average temperature is ~ 350°F
a. With one or both PORVs inoperable but capable of being manually cycled, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve(s) and maintain power to the associated block valve(s).

Otherwise. be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce Reactor Coolant System average temperature to < 350°F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With one PORV inoperable and not capable of being manually cycled. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or capable of being manually cycled or close the a<>sociated block valve and remove power from the block valve. In addition. restore the PORV to OPERABLE status or capable of being manually cycled within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Otherwise. be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce Reactor Coolant System average temperature to < 350°F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
c. With both PORVs inoperable and not capable of being manually cycled. within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore at least 1 PORV to OPERABLE status or capable of being manually cycled. Otherwise. close the associated block valves and remove power from the block valves. In addition. be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce Reactor Coolant System average temperature to

< 350°F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Amendment Nos. 261 and 261 Page 7 of 9

Attachment 2 - Reference Material for Question #93 TS3.1-5 05-31-02

d. With one block valve inoperable. within I hour either restore the block valve to OPERABLE status or place the associated PORV in manual. In addition, restore the block valve to OPERABLE status in the next 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or. be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce reactor coolant average temperature to <350°F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
e. With both block valves inoperable. within I hour either restore the block valves to OPERABLE status or place the associated PORVs in manual. Restore at least I block valve to OPERABLE status within the next hour or, be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce reactor coolant average temperature to <350°F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
f. With one or both PORV(s) inoperable (but capable of being manually cycled) because of an inoperable backup air supply, within 14 days either restore the PORV(s) backup air supplyOes) to OPERABLE status or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce Reactor Coolant System average temperature to < 350°F within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
7. Reactor Vessel Head Vents
a. At least two Reactor Vessel Head vent paths consisting of two isolation valves in series powered from emergency buses shall be OPERABLE and closed whenever RCS temperature and pressure are >350°F and 450 psig.

Amendment Nos. 231 and 231 Page 8 of 9

Attachment 2 - Reference Material for Question #93 TS 3.1-5c 05-31-02 The power operated relief valves (PORVs) operate to relieve Reactor Coolant System pressure below the setting of the pressurizer code safety valves. The PORVs and their associated block valves may be used by the unit operators to depressurize the Reactor Coolant System to recover from certain transients if normal pressurizer spray is not available. Specifically. cycling of the PORVs is required to mitigate the consequences of a design basis steam generator tube rupture accident. Therefore. whenever a PORV is inoperable. but capable of being manually cycled. the associated block valve will be closed with its power maintained. The capability to cycle the PORVs is verified during each refueling outage (and is not required during power operations).

These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve leak excessively. The electrical power for both the relief valves and the block valves is supplied from an emergency power source to ensure the ability to seal this possible Reactor Coolant System leakage path.

With one or both PORVs inoperable (but capable of being manually cycled) due to an inoperable backup air supply, continued operation for 14 days is allowed provided the normal motive force for the PORVs. i.e .. the instrument air system. continues to be available. Instrument air ha<; a high system reliability. and the likelihood of it being unavailable during a demand for POR V operation is low enough to justify a reasonable length of time (i.e .. 14 days) to repair the backup air system.

The accumulation of non-condensable gases in the Reactor Coolant System may result from sudden depressurization. accumulator discharges and/or inadequate core cooling conditions. The function of the Reactor Vessel Head Vent is to remove non-condensable ga~s from the reactor vessel head. The Reactor Vessel Head Vent is designed with redundant safety grade vent paths.

Venting of non-condensable gases from the pressurizer steam space is provided primarily through the Pressurizer PORVs. The pressurizer is. however. equipped with a steam space vent designed with redundant safety grade vent paths.

References (I) UFSAR Section 14.2.9 (2) UFSAR Section 14.2.10 Page 9 of 9

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