ML18139A732
| ML18139A732 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 07/24/1980 |
| From: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | Ferguson J VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| References | |
| NUDOCS 8009100056 | |
| Download: ML18139A732 (15) | |
Text
e UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 101 MARIETTA ST., N.W., SUITE 3100 ATLANTA, GEORGIA 30303 JUL 2 4 1980 Virginia Electric and Power Company ATTN: J. H. Ferguson Executive Vice President-Power P. 0. Box 26666 Richmond, VA 23261 Gentlemen:
cop*y The enclosed IE Bulletin No. 80-18, is forwarded for action.
A written response is required.
In order to assist the NRC in evaluating the value/impact of each Bulletin on licensees, it would be helpful if you would provide an estimate of the manpower expended in conduct of the review and preparation of the report(s) required by the Bulletin. Please estimate separately the manpower associated with corrective actions necessary following identification of problems through the Bulletin.
If you desire additional information regarding this matter, please contact this office.
Sincerely,
Enclosures:
- 1.
IE Bulletin No. 80-18 w/encls
- 2.
List of Bulletins Recently Issued cc w/encl:
W.R. Cartwright, Station Manager P. G. Perry, Senior Resident Engineer J. L. Wilson, Manager
,8 0.0.910 o o5 G
e-
- UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 July 24, 1980 SSINS No.: 6820 Accession No.:
8005050062 IE Bulletin No. 80-18 MAINTENANCE OF ADEQUATE MINIMUM FLOW THRU CENTRIFUGAL CHARGING PUMPS FOLLOWING SECONDARY SIDE HIGH ENERGY LINE RUPTURE Description of Circumstances:
Letters similar to the May 8, 1980 notification made pursuant to Title 10 CFR Part 21 (enclosure) were sent from Westinghouse to a number of operating plants and plants under construction (list, within enclosure) in early May, 1980.
The letters and the enclosed "Part 21" letter contain a complete description of the potential problem summarized below. The letters indicated that under certain conditions the centrifugal charging pumps (CCPs) could be damaged due to lack of minimum flow before presently applicable safety injection (SI) termination criteria are met. The particular circumstances that could result in damage vary somewhat from plant to plant, but involve unavail-ability of the pressurizer power operated relief valves (PORVs), with operation of one or more 'CCPs repressurizing the reactor during SI following a secondary system high energy line break. Since the SI signal automatically isolates the CCP mini-flow return line, the flow through the CCPs is determined by the individual pump characteristic head vs. flow curve, the pressurizer safety valve setpoint, and the flow resistances and pressure losses in the piping and in the reactor core. That minimum flow may not be adequate to insure pump cooling, and resulting pump damage could violate design criteria before current SI termination criteria are met.
Westinghouse recommends that plant specific calculations outlined in the letter (enclosure) be performed to determine if adequate minimum flow is assured under all conditi.ons. If adequate minimum flow is not assured, Westinghouse recommends specific equipment and procedure modifications which will result in adequate minimum flow. The recommended modifications assure availability of the necessary minimum flow by assuring that the mini-flow bypass line will be open when needed, but will be closed at lower pressures when the extra flow resulting from bypass line closure might be necessary for core cooling.
July 24, 1980 Page 2 of 3 Actions to be taken by PWR licensees listed in the enclosure as "operating plants," and those listed as "non-operating plants" which are nearing licensingi:
are listed below:
- 1.
Perform the calculations, outlined in the enclosure, for your plant.
- 2.
.If availability of minimum cooling flow for the CCPs is not assured for all conditions by the calculations in 1:
- a.
Make modifications to equipment and/or procedures, such as those suggested in the enclosure, to insure availability of adequate minimum flow under all conditions.
If modifications are 'made as described in the attachment for interim modification II, verify that the Volume Control Tank Relief Valve is operable and will actuate at its design setpoint.
- b.
Justify that any manual actions necessary to assure adequate minimum flow for any transient or accident requiring SI can and will be accomplished in the time necessary.
- c.
Verify that any manipulations required (valve opening or closing, along with the instrumentation necessary to indicate need for the action or accomplishment of the action, etc.) can be accomplished without offsite power available.
- d.
Justify that flow available from the CCPs with the modifications in place will be sufficient to justify continued applicability of any safety related analyses which take credit for flow from the CCPs (LOCA, HELB, etc.).
- e.
Justify that all Technical Specifications based on the Item 2.d analyses remain valid.
- 3.
Provide the results of calculations performed under Item 1, and describe any modifications made as a result of Item 2 (include the justifications requested).
Actions to be taken by PWR licensees not listed in the enclosure are listed below:
- 1.
In a quantitative manner similar to 1 above, determine whether or not minimum cooling is provided to centrifugal pumps used for high pressure injection, for all conditions requiring SI, prior to satisfying SI
- Those listed in the enclosure considered to be "nearing licensing" are:
North Anna 2, Diablo Canyon 1, McGuire 1, Salem 2, and Sequoyah.
These plants must respond in writing within the specified time.
Other non-licensed plants whether or not listed in the enclosure, are not required to submit a written response at this time.
IE Bulletin No. 80-18 July 24, 1980 Page 3 of 3 termination criteria. If a "minimum flow bypass" line is present which remains open during high pressure injection, and if that line guarantees that minimum cooling flow will be provided to the pumps under such condi-tions, then no further calculations are required if al-1 safety relat_ed analyses (Item 2.d above) assumed presence of the open line.
- 2.
Same as 2 above.
- 3.
Same as 3 above.
Licensees of all operating PWR power reactor facilities and those nearing licensing* shall submit the information requested within 60 days of the date of this letter. Include in your response to this Bulletin, (a) your schedule for any changes proposed, (b) if reactor operation is to continue prior to completion of the proposed changes, include your justification for continued operation.
Reports shall be submitted to the Director of the appropirate NRC Regional Office and a copy forwarded to the Director, NRC Office of Inspection and Enforcement, Division of Reactor Operations Inspection, Washington, D. C. 20555.
Approved by_GAO, B280225 (R0072); clearance expires 7-31-80. Approval was given under a blanket clearance specifically for identified generic problems.
Enclosure:
Ltr from T. M.. Anderson,~
to V. Stello, IE dtd 5/8/80
- Those considered to be "nearing licensing" are: North Anna 2, Diablo Canyon 1, McGuire, Salem 2, and Sequoyah.
.* Electrlc Corporation e
Water Reactor DIYlsions Mr. V. Stello, Director
- office of Inspection and Enforcement U. S. Nuclear Regulatory.Corrmiss;on 1717 H Street Washington, D. C.
20555
. IU:a Techna(02)' DfY!sloo ICl355 Plnwin P!nns)1YIWa 1523:l
.* *May 8, 1980 NS-TMA-2245 8'0-oi.. *1-0() a
Subject:
Centrifugal Charging Pump Operation Following Secondary Side High Energy Line Rupture
Dear Mr. Stello:
This letter is to confinn the telephone conversation of May 8, 1980 betweP.n Westinghouse and Mr. Ed Blackwood of D1v1s1on of Reactor Operations Inspection, Office of Inspection and Enforcement, regarding notification made pursuant to Title 10 CFR Part 21.
A review of the Westinghouse Safety Injection (SI) Termination Criteria following a secon9ary side high eneriY line rupture (feedline or steaml1ne rupture at high.1n1tial power levels} has revealed a potentia1 for conse*
quential damage of one or more centrifugal charging ~umps (CCPs) before the SI tennination criteria are satisfied and CCP operation tenninated.
Such consequential damage may adversely impact long-tenn recovery operations for the initfatfng event and 1s not pennitted by design criteria. This concern exists for plants which utilize the CCPs as Emergency Core Cooling System (ECCS) pumps. where the CCPs are automatically started, and where the CCP mfniflow 1solatfon valves are automatically isolated upon safety injection initiation. Attachment A identifies plants potentially subject to this concern.
A surrmary of the concern and recomnendations follow.
- Following a secondary side high energy line rupture and associated reactor triP.. Reactor Coolant System (RCS) pressure and temperature initially decrease.
Safety injection is actuated and the CCPs start to increase RCS inventory.
Reactor Coolant System pressure and temperatur~ subsequently increase due to the loss of secondary inventory, steamline and feedline isolation, RCS inventory addition and reactor ccre decay heat generation. The accident scenario may vary with rupture size and specific plant design, but it will develop into a RCS heatup transient with accompanying increase in RCS pressure.,
As RCS pressure increases, the pressurizer power-operated relief valves (PORVs) are designed to limit RCS pressure to 2350 ps1a *. Although these valves are nonnally available, they are not designed as safety-related equip~
ment. It can be postulated that, due to either loss of offsite power,
o')
- Mt. e May 8. 1980 NS-TMA-2245 adverse environment inside containment, the pressurizer PORV in manual mode, or the PORV block valve in a closed position, due to PORV leakage, the pressurizer PORVs may not be operable.
As a result of the RCS heatup and ;nventory increase. the RCS pressure could rise to the pressurizer
-. 1round 11 can occur between 1800 and 4200 seconds depending on operator ection
- and available equipment.
During the 1n1t1al portion of this transient, the SI termination criteria may not be satisfied. Consequently, the RCS pressure can reach the pressurizer safety valve relief pressure before CCP operation is terminated. During this period, the minimum flow required for CCP opera-tion must be satisfied by flow to the RCS since the CCP min1f1ow isolation valves are automatically closed on safety injection initiation:* This -requires that the CCPs be able to deliver their minimum required flow to the RCS at the safety valve sctpoint pressure.
To evaluate th1s concern, Westinghouse has developed a calculational method and has reviewed typical CCP head versus flow perfonnance curves and other representative plant parameters. The cijlculational method considers the effects of safety valve relief setpoint accuracy, RCS piping resistance, ECCS piping resistance, number of CCPs operating, technical specification allowable CCP head degradation, and uncertainties associated with in-plant verification testing. The analyses for two CCP operationJ the best estimate condition, is similar to the analysis for one CCP operation except that the flowrate used to detenn1ne ECCS piping line loss must ensure the minimum flow through each pump.
For example, at a specific required h~ad, the pump with the higher developed head may be required to deliver greater than the minimum flow in order to pennit the lower head pump to meet the minimum flow requirement.
This generic evaluat;on indicates that sufficient flow to satisfy CCP minimum flow requirements to avoid pump degradation may not be ensured for a secondary system high energy line rupture under the conditions described above.
Based on the generic evaluation, Westinghouse recorrmends that operating plants perfonn a plant specific evaluation to assess this concern.
Attachment B provides the Westinghouse calculational method and a sample ca~culat1on which can be used in this evaluation. Based on Westinghouse generic review, satis-factpry results may not be obtained. Should a plant specific concern be identified, the following recomnendations have been developed and can be tailored to specific plant applications for the interim until necessary design modif;catfons can be implemented.
The interim modifications consist of system alignment and operating procedure changes to provide backup to the pressurizer PORVs 1n ensuring that CCP minimum flow requirements are satisfied. In conjunc-tion with the interim modifications. it is recommended that p1ant5, (a) review the pressurizer PORV operations to maximize the availability of these valves to 1fmit challenges to the pressurizer safety valves, and {b) review the maintenance operations and technical specifications for the backup (1.e., third) charging pump to maximize its availability for long-term recovery from a secondary side rupture. These recorrmendations. in combination with the inter1m
e **-****
- Mr. ~y 8, 1980 NS-TMA-2245
.. modifications described below. are considered sufficient to address this con-cern in the interim until necessary design modifications ca~_.be implemented.
... Interim Modification I
. Jhis interim modification 1s preferred and requires that component cooling
- - -water be supplied to the seal w~ter heat exchanger following safety injection initiation 1n order to provide cooling for CCP min1f1ow.
- 1. Verify that CCP miniflow return is aligned directly to the CCP suction during nonnal operation with the alternate return path to the.. volume control tank isolated (1ock closed).
2, Remove the safety injection initiation automatic closure signal from the CCP miniflow isolation valves *
. 3. Modify plant emergency operating procedures to instruct the operator to:
- a. Close the CCP m1niflow 1so-lation valves when the actual RCS pressure drops to the calculated pressure for manual reactor coolant pump trip.
- b. Reopen the CCP miniflow isolation valves should the wide range RCS pressure subsequently rise to greater* than 2000 psig.
Interim Modification II This modification is an alternative for plants in which component cooling water is not supplied to the seal water heat exchanger following safety injection initiation. Since miniflow cooling is not provided, this alterna-tive directs miniflow to the volume control tank to pennit the CCP minimum flow requirements to be satisfied with cool uncirculated water. The volume control tank acts as a surge tank to collect miniflow following safety injection initiation with excess flow directed to a holdup tartk via the
- volume control tank relief valve.
- 1. Align the CCP miniflow to the volume control tank during nonnal oper~-
tion with the mini"flow return path direct to the CCP suction isolated (lock closed). Verify that the volume control tank relief valve and discharge line capacity exceeds the miniflow requirements of all CCPs plus the reactor coolant pump seal return f1ow.
- 2. Saine as Interim Modif1cat1on I, Item 2.
- 3. Same as Interim Modification I. Item 3.
e*
e
- Mr. V. Stel 1o May s. 1980 NS-TMA-2245
. Based on the generic evaluation, Westinghouse has initiated efforts to perfonn additional plant specific analyses for non-operating plants and to develop design modifications to resolve any identified concerns. *The mod1f1cations will be designed to safety-related standards and will be compatible with
...
- Westinghouse SI tennination criteria and standardiied technical specifications.
-: ~f you require further 1nformat1on. please call Ray Sero (412-373-4189} of my
- -staff.
TMA/jaw Attachments
~r~a.:~~.,..,~.,..------
~n~erson. Manager Nuclear Safety Department
~
I
- . ~*..
. ~
3-Loop
- Beaver Va 11 ey.1
- i Far*ley 1
.*. :..,~rry 1 & 2 North Anna 1 I 2 e
PfERATING PLANTS NON-OPERATING PLANTS Be~ver Valley 2 far1ey 2
- Shearon Harris 1,2 1 3 & 4 Vfrg11 SUTJner e
ATTAC~ENT A 4-Loop
......Cook 1 & 2
... Salem l & 2 Trojan
- Zion 1 & 2 5equoyah 1 Braidwood 1 & 2 Byron 1 &.2 Calloway 1 I 2 Catawba 1 & 2 Camanche PeAk 1 & 2 D1ab1o Canyon 1 & 2 Jamesport 1 & 2 Haven Marble H111 1 & 2 McGuire 1 & 2 Millstone 3 Seabrook l & 2 Sequoyah 2 Ster1 ing Vogtle 1 & 2 Watts Bar 1 & 2 Tyrone Wolf Creek
f I
~:. -
... anACHMENT B MINIMUM CENTRIFUGAL CHARGING PUMP FLOW DURING TWO PUMP PARALLEL SAFETY INJECTION OPERATION In order to ensure that minimum pump flow is maintained during parallel safety injection operation of two centrifugal charging pumps (CCPs)
- Westinghouse provides below a sample c:alc:ulatfon utiHzing actual plant data and determinas what actual CCP developed head at the miniflow flowrate
- ._.-ust be available.
Step 1: Individually detennine the developed head of each CCP at the m1ni-f1ow*flowrate of 60 gprn from field test data~
(two pumps for 4-loop plants and three pumps for 3*loop plants)
Sample:
Maximum developed head pump 257-1.4 psi d a;; 5940 ft. ll 60 gpm Minimum developed head pump 2554.1 psid = 5900* ft. 8 60 gpm Step 2: Correct the pump head for testing error. Add the appropr1ate error in detenn1n1ng the ftbove measured-developed head, i.e~.
- fnstrument error plus reading error. to the 11axfmum developed head and s_ubtract this _error.from the minimum developed head
- Sample: Pressure 1ns~rument accuracy of _:!:0.5 percent x span of measuring ~nstrument of 3000 psig
- 15 _ps1 (35 ft. of head). p1us 10 psi (23 ft.) read,ng 1ccuracy
- 58 ft.
The resultant CCP developed. heads at m1niflow which can be supported are a maximum developed head of 5ggs ft. for the ~aximum head pump. and a minimum developed head of 5842 ft. for the minimum head pump.
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-2*
Step 3: Detennine total CCP flow.
Construct a pump curve fo: the maxi-mum head pump that 1s parallel to the actual *as~bu11t* vendor pump curve and passes through the above detennined developed head at the min;flow flowrate which is the measured developed
. head pl us the detenni ned measurement accuracy. A See attach-anent F; gure l * )
Use this head versus flow curve to detennine the flow delivered by the maximum head pump (strong pump) at the developed head of the minimum head pump (weak pump) at the min1flow flowrate (1.e., 5842 ft. as determined in Step l).
Sample:
As 11lustrAted in Figure 1, the delivered flow of the strong pump at 5842 ft. 1s 150 gpm.
Therefore. the total flow from _both CCPs which guarantees that the weak CCP will be delivering at least 60 gpm is 210 gpm (150 gprn + 60 gpm).
Step 4: Dete,:mfne Injection Piping Head Loss.
The head loss due to fr1c:tion* 1n the safety injection/RCP seal injection pf ping is detenn1ned as follows:
- ~e 6hf is equal to the strong CCP developed head at runout flow. This resistance is established during the CCP flew balance testing which limits CCP flow to the r-unout limit.
The injection p1p.1ng resistance (k) is equal to the* developed head of the stronj CCP at its runout flow divided ~Y the (runout flowrate).
e.g.
t = developed head 2 = 6h2 c 1500 ft. 2 (runout flowrate)
Q tsso gpm)
e
~TTAC~ENT B The resistance of the injection piping (~hf)* at the tot~l CCP flow required to_ maintain 60 gpm through the weak CCP 1s:
4hf m kQ2 or Ahf ~ (4.96 X 10*3 :~2) (210 gpm)~
- Z19 ft.
- *
- Ste,e 5: Detennine head loss through the Reactor* Coolant System.
tons1der that the reactor coolant pumps are operating. therefore *
. the pressure drop from the CCP cold. leg injection nozzles through the reactor vessel to the pressurizer surge line off.the hot leg*
at full RCS flow are to be included. This pressure drop js appr"Oxfmately ~O psid (116 ft.) for 4-loop plants an~ 48 *psid (111 ft.) for 3-loop plants. This pressure drop must be ov~rcome by the CCPs 1n order to deliver flow to the *Res at the hot leg/
pressurizer pressure.
Step 6: Detennine the elevaticnal head between the RWST and the pressurizer safety.valves.
e.g.
RWST elevation 160 ft.
CCP suction elevation 100 ft.
RCS cold leg injection nozzle elev~tion -
126 ft.
Pressurizer safety valve elevation RWST to CCP suction minus CCP suction to RCS minus RCS to pressurizer safety valves (61 ft. assuming a full pressurizer}
corrected for density difference 187 ft.
60 ft.
- (-26 ft.)
- (-44 ft.)
-10 ft.
Thus. in this examp1e the CCPs must provide an additional 10 ft.
of elevatfonal head.
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- . ATTAC.HMENT B Step 7: Calculate the pressurizer safety valve relief pressure.
e.g.
relief pressure m safety valve. nominal relief pressure
+ lS setting tolerance relief pressure= 2485 psig + 25 psig
- 2510 psig (5798 ft.)
Step 8: Detennine the maximum RCS pressurizer pressure at which 60 gpm minimum flow is maintained through the weak CCP.
Maximum RCS pressure= (CCP developed head at total CCP flowrate) -
(1njeet1on piping head loss) - (head loss through RCS) - (elevc-t1on head loss)
Max1mt,,?n RCS pressure* 5842 ft. - 219 ft. - 116 ft. - 10 'ft.=
5497 ft. C 2380 psfg Comp~r~n51 this pressure ta the pressur*izer safety vnlve rel 1ef pressure (Step 7) of 2~10 psig. it is evident that the 60 gpm flow requfred for the weak CCP will not be maintained
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e IE Bulletin No. 80-18 July 24, 1980 Bulletin No.
80-18 Supplement 2 to 80-17 Supplement 1 to 80-17 80-17 80-16 80-15 80-14 80-13 80-12 80-11 80-10 RECENTLY ISSUED IE BULLETINS Subject Date Issued Maintenance of Adequate 7/24/80 Minimum Flow Thru Centri-fugal Charging Pumps Fol-lowing Secondary Side High Energy Line Rupture Failures Revealed by 7/22/80 Testing Subsequent to Failure of Control Rods to Insert During a Scram at a BWR Failure of Control Rods to Insert During a Scram at a BWR Failure of Control Rods to Insert During a Scram at a BWR 7/18/80 7/3/80 Potential Misapplication of 6/27/80 Rosemount Inc., Models 1151 and 1152 Pressure Transmitters with Either "A" or "D" Output Codes Possible Loss Of Hotline 6/18/80 With Loss Of Off-Site Power Degradation of Scram 6/12/80 Discharge Volume Capability Cracking In Core Spray 5/12/80 Spargers Decay Heat Removal System 5/9/80 Operability Masonry Wall Design 5/8/80 Contamination of 5/6/80 Nonradioactive System and Resulting Potential for Unmonitored, Uncontrolled Release to Environment Enclosure Issued To All PWRs reactor facil-ities with OLs and CPs and to PWRs nearing licensing.
All BWR power reactor facilities holding OLs All BWR power reactor facilities holding OLs All BWR power reactor facilities holding OLs All Power Reactor Facilities with an OL or a CP All nuclear facilities holding OLs All BWR's with an OL All BWR's with an OL Each PWR with an OL All power reactor facilities with an OL, except Trojan All power reactor facilities with an OL or CP