ML19209A172
| ML19209A172 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 08/29/1979 |
| From: | Goodwin C PORTLAND GENERAL ELECTRIC CO. |
| To: | Engelken R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| References | |
| NUDOCS 7910030058 | |
| Download: ML19209A172 (11) | |
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\\Q-August 29, 1979 Trojan Nuclear Plant Docket 50-344 License NPF-1 Mr. E. H. Engelken, Director Nuclear Regulatory Cosmicsion Region V Suite 202, Walnut Creek Plaza 1990 N. California Blvd.
Walnut Creek, CA 94596
Dear Sir:
IE Bulletin 79-06C, dated July 26, 1979, was transmitted to Portland General Electric Company (PGE) for action concerning operation of reactor coolant pumps (RCPs) after a LOCA.
PCE has reviewed the content of this Eulletin and evaluated its schedules and implications on the Trojan Nuclear Plant.
Attached is PCE's response to the subject Bulletin which has been pre-pared in cooperation with the Westinghouse Owner's Group. As we indi-cated in our July 31, 1979 letter relating to IE Bulletin 79-06C, this response satisfies both IE Bulletins79-06A (Items 4 and 7.c) and 79-06C.
Sincerely,,..
t C. Goodwin, Jr.
Assistant Vice President Thermal Plant Operation and Maintenance CC/lo!/4sb4A3 Attachment c:
Mr. Lynn Frank, Director State of Oregon Department of Energy Mr.11. R. Denton, Director Office of Nuclear Reactor Regulation 79100300SF 1084 273 u
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PGE RESPONSE TO IE BULLETIN 79-06C Short-Term Action 1.
In the interim, until the design change required by the long-term action of this Bulletin have been incorporated, institute the fol-lowing actions at your facilities:
A.
Upon reactor trip and initiation of HPI caused by low reactor coolant system pressure, immediately trip all operating RCPs.
B.
Provide two licensed operators in the control room at all times during operation to accomplish thin action, other immediate and folicw-up actions required during such an occurrence. For facilities with dual control rooms, a total of three licensed operators in the dual control room at all times meets the requirements of this Bulletin.
PGE Response A.
Immediate Operator Actions in the existing Trojan Emergency Instruction EI-1, " Loss of Reactor Coolant" (copy attached) requires the operator to stop all reactor coolant pumps immediately upon verification of (1) reactor trip, (2) tur-bine trip, and (3) initiation of safety injection (including initiation of HP1 by low reactor coolant system pressure).
Therefore, this item is already being implemented at Trojan.
B.
Since the receipt of the subject Bulletin at the Trojan Nuclear Plant (August 1,1979), two licensed operators have been provided in the control room at all times during opera-tion to ensure the immediate trip of RCPs and associated follow-up actions required during such an occurrence.
Appropriate documentation for this change has been completed 2
at Trojan.
Independent of IE Bulletin 79-06C, Administrative Procedures at Trojan already require a second licensed operator to be in the control room following a reactor trip until stable conditions are reached in one of the operating modes.
This would include any reactor trips caused by safety injec-tion. Although we are complying with this short-term action, it is not considered necessary to have two licensed operators just to ensure tripping of RCPs. Therefore, we a
consider the assignment of two licensed operators to be an interim measure.
Short-Term Action 2.
Perform and submit a report of LOCA analyses for your plants for a range of small-break sizes and a range of time lapses between reactor trip and pump trip.
For each pal: of values of the parameters, 1084 276
determine the peak cladding temperature (PCT) which results. The range of values for each parameter must be wide enough to assure that the maximum PCT or, if appropriate, the region containing PCTs greater than 2200*F is identified.
s PCE Response 2.
A series of Loss-of-Coolant Accident (LOCA) analyses for a range of break sizes and a range of time lapses between initiation of break and penp trip applicable to the two,
three-and four-loop plants has been performed by Westing-house for the Westinghouse owner's Group. A proprietary report summarizing the results of these analyses will be submitted to Mr. D. F. Ross (NRC) by Fk. Cordell Reed (Chairman, Westinghouse Owner's Group) on August 31, 1979.
This report includes the maximum peak cladding tempera-tures (PCTs) for each break size and pump shutoff time considered.
It is concluded that if the reactor coolant pumps are tripped prior to the reactor coolant system pressure reaching 1250 psia, the resulting PCTs are less than or equal to those reported in the FSAR.
As we stated in the response to Item 1.A above, the current Emergency Instruction at Trojan requires immediate stopping of the RCPs upon llPI initiation, which is conservative relative to the 1250 psia criterion.
The Westinghouse analyses show that there is a finite range of break sizes and RCP trip times, in all cases 10 min or longer, which will result in PCTs in excess of 2200*F as calculated with conservative Appendix K models.
For these cases, the operator would have at least 10 min to trip the RCPs following the break, especially in light of the con-servatism in the calculations.
This is appropriate for manual rather than automatic action,.bpsed on the guidelines for terminatica of RCP operatioa pres 4nted in WCAP-9600, "Small Break Analysis for Westinghouse NSSS Systems".
Short-Term Action 3.
Based on the analyses done under Item 2 above, develop new guidelines for operator action, for both LOCA and non-LOCA transients, that take into account the impact of RCP trip requirements.
For B&W-designed reactors, such guidelines should include appropriate requirements to fill the steam generators to a high level, following RCP trip, to promote natural circulation flow.
PGE Response 3.
The guidelines developed by the Westinghouse Owner's Group were submitted to the NRC in Section 6 and Appendix A of WCAP-9600. The analyses provided as the response to Item 2 1084 277
above are consistent with these guidelines.
llence, no changes to the guidelines for operator action are needed for either LOCA or non-LOCA transients.
Short-Term Actions 4.
Revise emergency procedures and train all licensed reactor operators and senior reactor operators based on the guidelines developed under Itas 3 above.
PGE Response 4.
The action taken in response to Iten 1 above is sufficient as an interim measure and there is no immediate need for changing Trojan emergency procedures. Ilowever, the Westing-house Owner's Group has a continuing effort to review the necessity for revising emergency procedures, including such considerations as operation of the reactor coolant pumps.
The following is the expected schedule for revising the LOCA, steamline break and steam generator tube rupture emergency procedures if changes are decided to be necessary:
Mid-October 1979: Guidelines which have been reviewed by the NRC will be provided to each member utility. Utility personnel responsible for writing procedures will meet with the Owner's Group Sub-committee on Procedures and Westing-house to provide the background for revising specific plant emergency procedures, if appropriate.
Mid-December 1979:
Plant-specific pr,ocedures will be revised, if appropriate.
Mid-February 1980: Revised procedures will be implemented and operators trained, if appropriate.
Short-Term Action 5.
Provide analyses and develop guidelines and procedures telated to inadequate core cooling (as discussed in Section 2.1.9 of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations") and define the conditions under which a restart of the RCPs should be attempted.
PGE Response 5.
Analyses related to inadequate core cooling and definition of conditions under which a restart of the RCPs should be ~
1084 278
attempted will be performed by the Westinghouse Owner's Group. Resolution of the requirements for the analyses, an acceptable schedule for providing the analyses, and appropriate guidelines and proccdures resulting from the analyses will be arrived at between the Westinghouse Owner's Group and the NRC Staff.
Long-Term Action 1.
Propose and submit a design which will assure automatic tripping of the operating RCPs under all circumstances in which this action may be needed.
PCE Response 1.
As discussed in the response to Short-Term Item 2, we do not believe that automatic tripping of the RCPs is required based on the analyses that have been performed and the guidelines that have been developed for manual RCP tripping.
We prenose that further discussion of this item be delayed until the NRC Staff has completed its review of the Owner's Group submittal of August 31, 1979.
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101/sb/4kk4A6 ~
1084 279
PORTLAND GENERAL ELECTRIC C05'PANY
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- List of Effective j
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Pages:
July 3, 1979 Page 1 of 6 - Rev. 7 Page 2 of 6 - Rev. 7 Revision 7*
SAFETY-RELATED Page 3 of 6 - Rev. 7 Page 4 of 6 - Rev. 6 Page 5 of 6 - Rev. 6 Page 6 of 6 - Rev. 6 EMERGENCY INSTRUCTION EI-1 S OF REACTOR COOLANT 7/$ 79 APPROVED BY DATE
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A.
SYMPTOMS Listed below are the symptoms which may indicate a large Icak'?n the reactor coolant system which will result in a loss of reactor coolent:
1.
Pressurizer low pressure.
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Pre jurizer low level.
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3.
High containment pressure.
4.
liigh containment humidity.
5.
liigh containment recirculation sump level.
6.
High containment radiation alarm.
B.
AUTOMATIC ACTIONS 1.
2.
3.
Safety injection is initiated.
4.
Containment spray may be initiated.
C.
Do!EDIATE OPERATOR ACTIONS 1.
Verify.cactor trip, turbine trip, and safety injection has occurred.
CAUTION:
If pressuri:cr pressure drops to 1765 psig and there is no automatic safety injection, manually start safety injection.
2.
STOP reactor coolant pumps.
3.
Verify al? engineered safeguards valves and equipment are aligned and operating with status lamp panel.
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4.
Verify safety injection flow when pressure is below pump's shutoff head.
EI-1 Page 1 of 6 Revision 7 1084 280
5.
Verify spray initiated if the containment pressure reaches the high-high set point.
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6.
Verify the ECCS is keeping the RCS pressure above saturation.
The incore thermocouples may provide better indication of core temperatures than the RTD bypass loops after the RCP's ar.e turned off.
If it is not possible to keep RCS pressure above saturation, see subsequent action for response.
7.
When 50 F subcooling can be maintained, AND if continued operation of the charging pumps would result in unsafe plant conditions, stop the charging pumps.
CADTION:
A.
Do not stop the safety injection or residusi heat removal pumps for at least 20 minutes unless continued operation is likely to result in unsafe plant conditions.
B.
If it becomes necessary to reset containment isolation the attached list must be verified before resetting.
C.
Maintaining pressurizer level alone may not prevent ex-cessive boiling in the RCS and resultant voids that may compromise the core cooling capability and natural circula-tion.
Keeping pressure within the limits of the pressure tenperature curves of figure 3.2 of the CROCTRM ensures saturation is not reached.
8.
If both RHR pumps are running, manually isolate trains by closing RHR cross-connect valves M0-8716A/B.
D.
SUBSEQUENT OPERATOR ACTIONS 1.
If there is an increasing pressurizer relief tank level, pressure and/or temperature, along with a high relief line temperature and the pressurizer pressure is below 1765 psig, isolate the POR's to see if one is stuck open.
2.
If pressurizer pressure and/or level are decreasing and Tave is remaining -
constant, a loss of coolant accident is indicated.
It may further be distinguished from a loss of secondary coolant or S/G tube rupture as follows:
a.
An increase in containment pressure, a containment high radiation alarm, and rising sump water level indicates a loss of coolant accident.
b.
An increasing pressurizer relief tank level, pressure, and/or temperature with possibly a high relief line temperature after both POR's are isolated indicates a loss of coolant accident due to a stuck open safety valve.
c.
A condenser air removal equipment radiation alarm or a steam generator blowdown radiation alarm indicates a steam generator tube rupture.
d.
Abnormally low pressure in one or more steam generators, coincident s
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with low pressurizer pressure and level and decreasing Tave indicate a main steam line break or feed line break.
EI-1 Page 2 of 6 Revision 7 1084 281
CAUTION: Do not override automatic actions of engineered safety features without careful review of plant conditions and C~m only then if continued ESF operation will result in unsafe plant conditions.
Do not make operat_...a1 decisions based on a single plant paramater or indication when a confirmatory indication is availabic, for example, pressurizer level without confirm-ing with pressurizer pressure.
3.
If it is determined by the above descriptions that the accident is a loss of reactor coolant, proceed to step 4.
If the accident is not a loss of reactor coolant, proceed to the appropriate Emergency Instruction.
4.
If plant conditions require a planned evolution, i.e. stopping unneeded RHR p' naps, the safety injection signal may be reset after 20 minutes.
l CAUTION:
In the event of a loss of off-site power following manual blocking an automatic SI, the only loads that will re-sequence onto the diesel generator are those initiated by the shut-down sequencer.
All other ESF loads required to be in operation as a result of the initial safety injection, must be manually re-started by the operator.
5.
Implement the Emergency Plan.
6.
If the RCS has spent a period of time below saturation or RCS sampics
-x 7
show cladding damage or a buildup of hydrogen gas in the RCS, start the containment hydrogen recombiners and mixing fans and periodically vent the pressurizer to the pressurizer relief tank.
If WGDT's are full, it may be necessary to allow the PRT rupture disc to blow, venting gases to the containment.
If RCP's are available, maximize pressurizer sprays to aid in degassing RCS and dissolving any voids which may now exist in the vessel head area.
a' 7.
When the RWST LO LEVEL annunciator actd' tes, start aligning the safety injection system to take suction from the containment recirculation su=p as follows:
1084 282 NOTE:
Ensure the residual heat removal (RHR) pumps tripped automatically on RWST LO LEVEL signal.
Open RHR heat exchanger (Hx) component cooling water (CCW) inlet a.
valves M0-3210A and MO-3210B.
b.
Close RHR pump suction valves MO-8700A, MG-8700B and M0-8812 from the RWST.
Open RHR pump suction valves MO-8811A and MO-8811B from recir-c.
culation sump.
NOTE: These valves are interlocked such that MO-8700A/B must be
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closed before M0-8711A/B can be opened.
d.
Verify the RHR Hx outlet cross-connect valves MO-8716A and MO-8716B are closed to provide train separation if both RHR pumps are operable.
If both pumps are not operable Icave the valves open.
EI.1 Page 3 of 6
START west / cast RHR pumps A/B.
c.
'~j f.
Flow to the vessel through the two cold leg injection lines can be f
checked by using west RilR Hx "A" outlet flow FI-971A and FI-971B, s
and east RHR Hx "B" outlet flow FI-970A and FI-970B.
g.
Close safety injection pump miniflow line block valves M0-8813 and M0- 8bl4.
h.
Open RHR pump discharge valve isolation valve MO-8804B to the safety injection pump suction.
NOTE:
Valve M0-8804B is interlocked such that the reactor coolant system to PllR system isolation valvi Ma 731 or MO-8702 must be closed, safety injection pump miniflow block valves MO-8813 or MO-8814 must be closed, and recirculation sump isolation valve MO-8811B must be open before M0-8804B can be opened.
i.
Ope:. RHR pump discharge isolation valve MO-8804A to the charging pump suction.
NOTE:
Valve MO-8804A is interlocked suc's that the reactor coolant system to RHR system isolation valves M0-8701 or M0-8702 must be closed, safety injection pump miniflow block. valves M0-8813 or MO-8814 must be closed, and recirculation sump isolation valve MO-8811A must be open before MO-8804A can
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be opened.
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j.
Verify that east RHR pump "B" is supplying the safety injection pumps (increased safety injection pumps discharge pressure, PI-919, PI-923).
k.
Open RHR discharge to safety injection pump suction valves MO-8807A
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and MO-8807B.
1.
Close safety injection suction val e M0-8806 from RWST.
Close charging pump suction valves MO-112D and MO-112E from RWST.
m.
8.
Shift the suction on both spray pumps one at a time as follows:
a.
STOP west / east containment spray pump A/B.
b.
Close containment spray pump suction valve MO-2050 A/B from the RWST.
Open containment spray pump suction valve MO-2052 A/B from the c.
recirculation sump, d.
START west / east containment spray pump,A/B.
e.
Repeat steps 1 through 5 and shift the other pump.
i
(,j EI-1 Page 4 of 6 Revision 6 1084 283
f.
h' hen the NaOH TANK LO LO LEVEL annunciator actuates, close spray additive valves MO-2056A and 2056B to prevent air binding of the
" -' T spray pumps.
9.
If the hydrogen recombiners and mixing fans have not aircady been started l per 6 above, start them now.
10.
Approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> after the accident, depending on the recirculation sump boron concentration, align the safety injection systen for hot leg / cold leg recirculation as follows:
Open RHR to RCS hot legs isolation valve MO-8703.
a.
b.
Open cross tie isolation valve M0-8716A.
Close RHR Cold Leg Injection valves M0-8809A/B.
c.
d.
Verify hot leg recirculation flow on FI-600.
c.
Open hot leg isolation valve MO-8802A. -
f.
Verify flow to the reactor coolant system through the hot leg header on FI-918.
g.
Open hot leg isolation valve M0-8802B.
h.
Verify flow to the reactor coolant system through the hot leg header on FI-922.
i.
Close Cold Leg Safety injection valves M0-8821 A/B and M0-8835, 11.
Sample the recirculated coolant to determine boron concentration as follows:
Every 15 minutes for the first hour].-
a.
'b.
Every hour for the next 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
d.
Maintain boron concentration greater than 2,000 ppm B.
Use emergency borate mode to increase the boron concentration, as e.
required.
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EJC EI-1 Page 5 of 6
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Revision 6 1084 284
EMERGENCY INSTRUCTION EI-1 LOSS OF REACTOR COOLANT
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VALVE VERIFICATION BEFORE RESETTING CIS Before resetting Containment Isolation, you must verify the following valves are in the indicated positions:
Valve Description Panel Position Verification MO-4180 Containment Sump Discharge C19 Pull to Lock CV-4181 Containment Su=p Discharge C19 Pull to Lock CV-5652 Accum Sample Isol C17 Auto After Close CV-5661 Reactor Coolant Drain Tk Sarple C17 Auto After Close CV-4000 Reactor Coolant Drain Tk N2 Jupply C17 Auto After Close CV-4006 Reactor Coolant Drain Tk Outlet C17 Auto After Close CV-4301 Gas Collection Header Valve C17 Auto After Close CV-4471 Instrument Air to Containment C17 Auto After Close CV-4470 Service Air to Containment C17 Auto After Close CV-10001 Containment Purge Supply C17 Auto After Clo,se CV-10004 Containment Exhaust C17 Auto After Close M0-10002 Containment Purge Isol C17 Auto After Close MO-10003 Containment Exhaust Isol C17 Auto After Close CV-10014 Chilled Water Return C17 Auto After Close CV-10015 Chilled Water Supply C17 Auto After Close
( _
MO-2810 A Steam Generator Blowdown Isol CIS Auto After Close M0-2813 B Steam Generator Blowdown Isol C15 Auto After Close MO-2812 C Steam Generator Blowdown Isol C15 Auto Aftc2 Close M0-2808 D Steam Generator Blowdown Isol C15 Auto After Close CV-2811 A Steam Generator Blowdown Sample CIS,_ Auto After Close CV-2880 B Steam Generator Blowdown Sample C151_ Auto After Close CV-2814 C Steam Generator Blowdcen Sample C15 Auto After Close CV-2809 D Steam Generator Blowdcwn Sample C15 Auto After Close i
1084 285 i
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