ML19290E289
| ML19290E289 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/19/1980 |
| From: | Matthews P Office of Nuclear Reactor Regulation |
| To: | Capra R Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.E.1.1, TASK-TM TAC-44673, NUDOCS 8003100142 | |
| Download: ML19290E289 (20) | |
Text
,, ' p FEB 191980 DIST IBUTION:
CKET FILE (50-312)
NRR READING ASB READING MEMORANDUM FOR: 'R. Capra, Project Manager Light Water Reactors Branch 3. DPM FROM:
P. R. Matthews. Section Leader Auxiliary Systems Branch, DSS
SUBJECT:
RANCHO SECO AFW SYSTEH We have reviewed the Rancho Seco AFW System Reliability Analysis sutunitted by the licensee on December 17. 1979. Our positions and requests for additional information are attached. Inputs from ICSB, RSS and PAS are 1 included. Sections C and 0 of the attachment also address AFW system concerns applicable to Rancho Seco realted to SRP 10.4.9 and the design basis for the AFW system flow requirements. The inclusion of these latter two areas along with the reliability analysis review makes the scope of review of the Rancho Seco AFW system consistent with the W and CE operating plant AFW system reviews accomplished previously by the staff under the B&O Task Force.
This review is considered to be part of the long tenn follow-up to the B&W plant shutdown orders of 1979.
The licensee should be requested to respond and make cornitmentsdto resolve the issues discussed in the attachment.
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P. R. Matthews. Section Leader Auxiliary Systems Branch Division of Systems Safety cc: w/ attachment D. Garner V. Benaroya C. Liang W. LaFaye
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1.
HRC Staff Review of Rancho Seco AFW Svstem Reliability Analysis A.
Staff Comments on Attachment 1 1.
Sectim 1.5 defines the AR! system nission success criterion as attainment of flow from at least one AFW pump to at least one steam generator. We consider this definition to be incomplete.
The criterion should include the requirerent to deliver the AFW flow to the steam generator before the steam generator boils dry since that is the primary function of the AFW system. The success criterion should be revised accordingly and submitted in a report supplement.
However, we do accept the discussion that proposes supplemental criteria; namely, maintaining adequate core cooling by use of an HPI pump. This supplemental criteria would then make AFW system reliability based on 15 and 30 minute operator action time applicable.
However, it is not applicable for Case 3: LOAC, i.e. total ac power blackout since the HPI pumps require ac power. The AFW system reliability comparison for the 15 and 30 minute operator action time should be deleted for Case 3 in Figure 6 and in the figure in the Executive Sumary section and submitted in a report suople-ment. It is noted that the staff AFW system reliability review of W and CE operating plants did not rely on this supplemental criteria.
2.
Section 2.4.2 indicates AFW pumo suction and discharge pressure instrumentation is provided in the control room. The site visit
., of December 13,1979 by f1RR personnel indicates tl1at such instru-mentation does not exist. Verify that this discrepancy does not affect the reliability study results and that there are no AFW system procedures that are dependent on such ir.strumentation.
3.
Explain why Section 3.3.1, 3.3.2 and 3.3.4 do not identify the ICS-NNI power supply as a potential single failure source which can fail to ooen valves FV 20527 and 20528 upon ATW system opera-tion demand. In this regard, there appears to be a discrepancy between assumption (8) on page 8 which indicates ICS was considered as single control device with signals to both AFW trains and between the fault trees on pages A-7 and A-8 which appear to treat the ICS signal to each valve as channelized. The discussion of Item 2.1.7.a on page 1 of Attachment 2 recognizes the potential for this particular failure mode.
We recognize that you have implemented a crocedure for operating the AFW system it. dependent of the ICS and have committed to a long term installation of a safety grade AFW automatic initiation and control system which is independent of the ICS in order to correct this potential deficiency. Your reponse to this question should indicate whether any other conclusions or recormlendations affecting AFW system reliability are affected.
, B.
Staff Positions / Request for Additional Infonnation on AFW System Outstanding Items - Attachment 2 1.
AFW Automatic Initiation System (Item 2.1.7.a - flVREG 0578) a.
Safety Grade Design - Your letter of October 13, 1979 responding to flRC letter of September 13, 1979 on Lessons Learned - flUREG 0578 comitted to installing a safety grade AFW automati-initiation and control system independent of the ICS duri, g the 1981 refueling outage.
In accordance with NUREG 0578 and NRC letter of October 30, 1979, this modification should be implemented by January 1,1981. Your design and orocurement activities should be expedited accordingly. Also, the modification design should be sub-mitted for Staff review by August 1,1980 to support the January 1,1981 installation date. The revised design should add low steam generator water level signal as an AR4 system initiating event to the existing initiating signals.
b.
Testability - AR4 valves FV 20527 and 20528 should b3 actuated using-the existing design control signals during the current refueling outage. Also, until the safety grade system per item a. above is installed, we require that you implement the following periodic testing procedure:
., Esublish and implement a procedure for performing channel functional tests
- of the automatic initiation circuitry of the AFJ pump using '>e loss of main feedwater signal at least every 31 days.
- CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions, b.
Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
c.
Item 2.1.7.a.6 - Automatic Loading of Pump P-319 on to the Nuclear Service Bus (Diesel Generator) Upon Loss of Offsite Power - Before implementing the action proposed for this item in Attachment 2, you should provide for Staff review within 30 days of receipt of this letter the following information:
Provide results of tests and analyses that demonstrate that the automatic sequencing of this load does not adversely affect the previous diesel generator perfonnance and associated safety System actuations and functions. Regulatory Guide 1.9 should be tc,ilowed; however, exceptions will be considered if adequately justified.
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. 2.
AP4 Flow Indication (Item 2.1.7.b - NUREG 0578) a.
Safety Grade Design - The commitment in your October 19, 1979 letter to install a safety grade system during the 1981 refueling should be scheduled for January 1,1981. Your design should be submitted for Staff review by September 1, 1980 to support the January 1,1981 installation date.
b.
Testability - Until the safety grade desisa per item a. above is installed, we require that you cons.11t to the follcwing periodic testing procedure:
Establish and implement a procedure for perfonni'19 channel functional tests of the AR4 flow and channel checks
- of the steam generator level indicator channels at least every 31 days.
- CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indi-cation and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
3.
System 11odification for Periodic AFW Pump Testing We have reviewed your proposal (P-3 Attachment 2) to imorove reliability associated with periodic AR4 pump testing. Existing
. manual valve PAS-055 will be replaced by a motor-operated valve operable from the control room and the test flow path will be changed to prevent the testing of one train from affecting the availability of the other train. We conclude the proposed modification will improve APd system reliability. However, we require:
a) that the modification be implemented by January 1,1981 rather than during the 1981 refueling outage, b) confimation that the motor operated valve position will be indicated in the control room, c) confimation that upon completion of the modification, the pump full flow test procedure will be chan92d to accommodate the modified system, d) that you submit for Staff infomation prior to the modification a revised P&ID reflecting the final design since the exact configuration has not been detemined, and e) that until the modification is installed, you continue to implement your existing surveillance procedure which requires stationing an operator at valve PdS-055 during pump surveillance testing.
4.
Emergency Procedure for Alternate Water Sources (GS-4)
We accept your cormlitment (P-5 Attachment 2) to reference standard operating procedures in your emergency procedures to
. establish how to obtain water for the AFW system from sources other than the condensate storage tank. Your existing procedures do not appear adequate for obtaining water from the Folsum South Canal, particularly with respect to starting the pump (s). Review your procedures and verify that they are adeauate for supplying water from both the Canal and the plant reservoir.
5.
AFW System Flow Path Verification (GS-6)
Item 2 on P-5 of Attachment 2 commits to Technical Specification surveillance testing per ASME Code Section XI. The Code is not explicitly clear that the pump or valve flow test will assure a flow test from the condensate storage tank to the OTSG. We require that you propose the following modification for the AFW system Technical Srscification:
" Prior to start-up following a refueling shutdown or any cold shutdown of longer than 30 days duration, conduct a test to demonstrate that the motor driven AFW pumps can pump water from the CST to the steam generators."
6.
Condensate Storage Tank Level Indication and Alarm We accept your commitment (Item 3 on Page 5 of Attachment 2) to install safety grade CST level indication and alarms. However, we recuire:
a) that the safety grade system be installed by January 1,1981 rather than during the 1981 refueling
, b) confirmation that the safety grade design will include:
- 1) redundant sensors, detectors readouts, and alarms all the way from the CST to control roem, including power supplies
- 2) use of Class lE circuitry equipment, and power supplies.
7.
AFW Pump Endurance Test (Item 4, page 5, Attachment 5)
We have reviewed the results of your endurance test and find that we need additional infonnation to complete our review. The attachment titled " Revision to Recommendation No. 2 of ' Additional Short Term Recommendations' Regarding Auxiliary Feedwater Pump Endurance Test" (previously informally transmitted) requests further information regarding the conditions and results of your tests. You should respond to this revision to the extent practicable with the data available from your tests.
In particular, the information should also justify the acceptability of the rising temperatures for the east and west turbine bearings of pump P-318 which occurred during the latter part of the test run.
8.
Periodic Testing of AFW Motor Operated Valves We accept your position (Item 2, page 6 Attachment 2) to perform periodic testing of these valves on a quarterly frequency in accordance with AS"E Code Section XI rather than monthly.
. 9.
AFd System Operation During Loss of All AC Power We agree with the conclusion stated in Item 3, page 6 of Attach-ment 2 that the AFW system will start upon loss of all ac power (turbine pump starts - flow control valves open). Mcwe /m,
further operation of the flow control valves is dependent on air supplied from an ac powered air compressor. Consequently, we require you to:
a.
Verify that your procedure A.51 (1) requires an operator to be stationed at the flow control valves following loss of all ac power and (2) provides for adequate lighting at the local valve station and communication with the control room which are independent of ac power for two hours.
b.
modify your design by January 1,1981 to enable the control room operator to control AFW system operation and steam generator level from the control room for two hours without dependence on ac power.
- 10. Technical Specification AFd System Outage LC0 lie have reviewed your response to item 7 of IE Bulletin 79-05A and your revised Technical Specification Limiting Condition for Or9 ration proposed by Amendment 63. We conclude that it is unacceptable. We require that you propose a modification to your
T echnical Specifications to provide for the following:
(1)When two independent 100" capacity flow paths are not available, the capacity shall be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or the plant shall be placed in a cooling mode which does not rely on steam generators for cooling within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and (2) When at least one 100% capacity flow path is not available, the reactor shall be made subcritical within one hour and the facility placed in a shutdown cooling mode which does not rely on steam generators for cooling within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
C.
AFirl System Standard Review Plan Section 10.4.9 A review of the Rancho Seco arf system using the guidelines of Standard Review Plan Section 10.4.9 and associated Branch Technical Position 10-1 indicates the following areas need further evaluation. The results of these evaluations and recorrended corrective actions should be submitted for staff review. Any necessary corrective actions should be imple-mented by January 1,1981.
1.
Postulated High Energy Pipe Breaks a.
In the event of a postulated break in the main steam or main feed system inside or outside containment, the Rancho Seco plant does not have redundant instrumentation and controls to automatically limit or terminate AFW system flow to a depres-surized steam generator and to direct the minimum required flow to the U tact steam generator.
, b.
The Rancho Seco AFW system does not meet the high energy line break criteria in SRP 10.4.9 and Branch Technical Positon 10-1; namely, that the AFH system should maintain the capability to supply the required flow to the steam generator (s) assuming a pipe break anywhere in the AFW pump discharge lines concurrent with a single active failure.
The licensee should evaluate the postulatti pioe breaks stated above and (1) determine any AFW system design changes or procedures neces-sary to detect and isolate the break and direct the required feed-water flow to the intact steam generator (s) before they boil dry or (2) describe how the plant can be brought to a safe shutdown condition by use of other systems which would be available following such postulated avents. No operator action in less than 10 minutes should be assumed.
D.
Design Basis for AFW System Flow Requirements f
In a letter from Babcock & Wilcox (B&W) to SMUD dated Mav 16, 1979, B&W summarized results of an analysis of the required AFW flow rate for the Rancho Seco Plant which shows that reactor decay heat of 4.5 percent of rated power plus RC pump power requires an AFW flow of approxi-mately 760 gpm. This required AFW flow is based on a required start time of 36 seconds for the AFW pumps following a reactor trip simul-taneous with loss of main feedwater flow.
. We conclude that the design basis provided above for establishing AFW system flow requirements is insufficient in that it does not identify the limiting plant transients or accidents that were considered in establishing the bounding AFW system flow requirements.
We require that you provide the AFWS flow design basis infomation required in Enclosure 2 for the Rancho Seco design basis transients and accident conditions.
Your response should include the following:
1.
List of all events needing AFW to mitigate the consequences.
2.
Justification that the bounding non-LOCA calculation will serve as a conservative basis for sizing the AFW system for non-LOCA core cooling considerations.
In other words, show that the calculation will bound all of the non-LOCA events requiring AFW.
3.
The non-LOCA analysis should include a loss of feedwater event using FSAR type assumptions to maximize heat removal requirements (1.2 ANS decay heat, 2" power level measurement uncertainty, RCP
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heatinput). The calculation should not take credit for "antici-patory reactor trip" since it will not occur under all conditions.
Lifting of the PORV is not precluded; however, credit for pressure reifef through the valve should not be assumed.
. 4.
For the small LOCA events, reference may be made to the B&W Report,
" Evaluation of Transient Behavior and Small Reactor Coolant System Breaks in the 177 Fuel Assembly Plants dated May 7,1979.
The acceptance criteria for the event will be:
1.
Reactor Coolant System pressure remains less than 110% of design pressure (2750 psig).
2.
No fuel failure (DNBR >l.30).
ir. ::.: r i Revision to Recomendation No. 2 of " Additional Short Term Reco=mendations" Regarding Auxiliary Feedwater pump Endurance Test s,.sv-The licensee should perform an endurance test on all AFW system pumps. The test should continue for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after achieving the following test conditions:
- pump / driver operating at rated speed and
- Pump developing rated dis' charge pressure and flow or some higher pressure at a reduced flow but not exceeding the pump vendor's maticum permitted discharge pressure value for a 48-hour test
- For turbine drivers, steam temperature should be as close to normal operating steam temperature as practicable but in no case should the temperature be less than 400'F..
Following the 48-hour pump run, the pumps should be shut down and allowed to cool down until pump temperatures reduce to within 20*F of their values at the start of the 48-hour test and at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. have elapsed.
Following the cool down, the pumps should be restarted and run for one hour. Test acceptance criteria should include demnstrating that the pumps remain within design limits with respect to bearing / bearing oil tempera-tures and vibration and that ambient pump room conditions (temperature, humidity) do not exceed environmental qualification limits for safety-related equipment in the room.
The ifcensee should provide a sumary of the conditions and results of the tests. The sumary should include the following: 1) A brief description of the test method (including flow schematic diagram) and how the test
was instrumented (i.e., where and how bearing temperatures were measured).
- 2) A discussion of how the test conditions (pump flow, head, speed and steam temperature) compare to design operating conditions. 3) Plots of hearing / bearing oil temperature vs. time for each bearing of each AR pump / driver demonstrating that tempera ure design limits were not exceeded. 4) A plot of pump room ambient temperature and humidity vs.
tic:e demonstrating that the pump room achient conditions do not exceed environmental qualification limits for safety-related equipment in the room. 5) A statement confirming that the pump vibration did not exceed allowable limits during tests.-
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Basis for Auxiliary Fee 6 water System Flow Reouirements As a result of recent staff revie:s of operatin; plant Auxiliary Feed-water Systens (AFWS), the staff concludes that the design bases and criteria provided by licensees for establishino APd5 requirements for flow to the steam generator (s) to assure adequate ramoval of reactor decay heat are not well defined or documented.
We require that you provide the following AFWS flow design basis infor-mation as appitcable to the design basis transients and accident con-ditions for your plant.
1.
a.
Identify the plant transient and accident conditions considered in establishing AFWS flow requirements, including the following events:
s
- 1) Loss of Main Feed (LMFW)
- 2) LMFW w/ loss of offsite AC power
- 3) LMFW w/ loss of onsite and offsite AC power
- 4) Plant cooldown
- 5) Turbine trip with and without bypass
- 6) Main steam isolation valve closure
- 7) Main feed line break
- 8) Main steam line break
- 9) Small break LOCA
- 10) Other transient or accident conditions not listed above b.
Describe the plant protection acceptance criteria and corres-ponding technical bases used for each initiating event identi-fied above. The acceptance criteria should address plant limits such as:
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- Maximum RCS pressure (PORY or safety valve actuation)
- Fuel temperature or damage Jimits (DNB, PCT, maximum fuel central temperature)
- RCS cooling rate limit to avoid excessive coolant shrinkage
- Minimum steam generator level to assure sufficient steam generator heat transfer surface to remove decay heat and/or cool down the primary system.
2.
Describe the analyses and assumptions and corfe'sponding technical justification used with plant condition considered in 1.a. above including:
a.
Maximum reactor power (including instrument error allowance) at the time of the initiating transientmr accident.
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b.
Time delay from initiating event to reactor trip.
c.
Plant parameter (s) which initiates AFWS ficw and time delay between initiating event and intmduction of AFV5 flow into steam generator (s).
d.
Minimum steam generator water level when initiating event occurs.
e.
Initial steam generator water inventory ahl depletion rate before
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and after AFWS flow cocmences - identify reactor decay heat rate used.
3-f.
Maximum pressure at which steam is released frce steam generator (s) and against which the AFW pump must develop sufficient head.
g.
Minimum number of steam generators that must receive AFW flow; e.g.1 out of 27. 2 out of 47 h.
RC flow condition - continued operation of RC pumps or natural circulation.
- i. Maximum AFV inlet temperature.
J. Following a postulated steam or feed line break, time delay assumed to isolate break and direct AFV flow to intact steam generator (s). AFV pump flow capacity allowance to accomodate the time delay and maintain m'infr:um steam generator water level.
Also identify credit taken for primary system heat removal due to blowdown.
k.
Volume and maximum temperature of water in main feed ifnes between s. team generator (s) and AFWS connection to main feed line.
1.
Operating condition of steam generator nonnal blowdown following initiating everit.
m.
Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.
n.
Time at hot standby and time to cooldown RCS to RHR systera cut-in temperature to size AFW water source inventory.
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3.
Verify that the AFd pumps in your plant will supply the necessary flow to the steem generator (s) as, determined by iteas 1 and 2 above considering a single failure. Identify the margin in si:ing the pump f. low to allow for pump recirculation flow, seal leakage and pump wear.
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