ML20136H632
| ML20136H632 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/15/1985 |
| From: | Carey J DUQUESNE LIGHT CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM TAC-44562, NUDOCS 8511250153 | |
| Download: ML20136H632 (9) | |
Text
r
'Af Telephone (412) 393-6000 Nuctea, Group
$,,i,,,d November 15, 1985 ort PA15077-0004 Director of Nuclear Reactor Regulation
/ United States Nuclear Regulatory Commission Attn:
Mr. Steven A. Varga, Chief Operating Reactors Branch No. 1 Division of Licensing Washington, DC 20555
Reference:
Beaver Valley Power Station, Unit No. 1 Docket No. 50-334, License No. DPR-66 NUREG-0737; Item II.D.1 - Response to Request for Additional Information Gentlemen:
In accordance with our submittal dated October 14, 1985, we are providing responses to questions 1 through 6 of your request for additional information dated August 22, 1985 on the above referenced subject.
The responses to your requests for information are included in the attachment to this submittal.
If you have any questions regarding the information contained herein, please contact me or members of my staff.
Very tr ly yours, g112 g h h oo 4
. Carey p
Vice President, Nuclear Attachment cc:
Mr. W. M. Troskoski, Resident Inspector U. S. Nuclear Regulatory Commission Beaver Valley Power Station Shippingport, PA 15077 U. S. Nuclear Regulatory Commission c/o Document Management Branch Washington, DC 20555 Director, Safety Evaluation & Control Virginia Electric & Power Company gfh P.O. Box 26666
(,
One James River Plaza Richmond, VA 23261 e
O ATTACHMENT Response to NRC Request For Additional Information Dated August 22, 1985 NUREG-0737 Item II.D.1 Performance Testing of Relief and Safety Valves Beaver Valley 1 QUESTION 1 Some information received indicates Beaver Valley Unit 1 uses Limitorque SMB-000-15 actuators with the PORV block valves.
he EPRI test report submitted June 1, 1982 shows the actuators are Limitorque SMB-00-15.
Clarify which actuator is used at Beaver Valley Unit 1.
RESPONSE
Beaver Valley Unit 1 uses Limitorque SMB-00-15 actuators with PORV block valves. hese motor operators were replaced during the 4th refueling outage with fully environmentally qualified operators.
QUESTION _2 In response to question 8 of our request for additional information, it was stated the PORV air supply pressure was increased to 60 psig and the air supply tubing size was increased, but the response did not state what new tubing size was used.
Provide the size of the new tubing installed at the plant.
Also the response stated that, after the modifications were made, testing of the PORVs resulted in acceptable stroke times being achieved. he answer did not indicate what these stroke times were.
Provide the results of the stroke time testing for our review.
RESPONSE
Approximately 10 feet of 1/2 inch copper tubing was installed in the 1
air supply to PORVs PCV-E-455C and PCV-RC-455D. his modification was performed under Design Change Package 137 and was part of the modifications performed for cold over pressure protection.
%e air I
supply pressure to PCV-E-455D was increased to 60 psig to obtain acceptable stroke times after modification to the air supply solenoids performed under Design Change Package 579.
i l
Test BVT 1.2-2.6.1 was performed for PCV-E-455C on 8/20/83 and for PCV-E-455D on 9/6/83 to verify acceptable stroke times after the DCP 579 modification. %e results of these tests are shown in Table 2.1.
During these tests timing signals for recording PORV (PCV-RC-455C and D) opening and closing times were obtained from the valve "ONN" status light (red),
valve "CIDSE" status light (green), and air solenoid valve. hose
. signals were recorded and timed using an oscillograph operating at a chart speed of 1 iysec. PORV opening time is defined as time from air solenoid valve changing state (de-energized to energized) until valve I
"Ct4SE" status light (green) extinguishes.
PORV closing time is defined as time from air solenoid valve changing state (energized to de-energized) until valve "OR N" status light (red) extinguishes.
PORV cycle time is the sum of open and close times for each individual valve operation.
Se acceptance criteria for avr 1.2-2.6.1 are:
I
- 1) Maximum opening time less than or equal to 2.5 seconds, j
- 2) Maximum closing time less than or equal to 8.0 seconds, and
- 3) Minimum cycles time greater than or equal to 7.0 seconds.
)
his assures that the opening time is within the cold overpressure l
protection analysis assumptions of 2.0 seconds stroke time, 0.4 second disk pressurization time and 0.3 second signal delay time. ne cycle time also assures that there is sufficient capacity of the backup nitrogen system to allow for cycling the PORV's for 10 minutes.
QUESTION 3 Nhat is the torque setting of the Lim' r ;m PORV block valve operator in 4
the plant and what torque does the or utor produce at this setting?
RESPONSE
%e torque setting of the Limitorque operators on PORV block valves MOV-RC-535, MOV-RC-536 and MOV-RC-537 are set at 1.75 within the range of 0-5.
ne torque produced by the operator at this setting is 86 ft-lbs.
l 2
i I
r.
QUESTION 4 In response to stion 3 of our request for additional information, reference was to a Westinghouse Owners Group analysis which showed valve blowdown up to 10% did not result in the pressuriser filling nor were there any adverse effects on plant safety. Provide a copy of this report for our review.
RESPONBE he concern addressed in question 3 of your original request for additional information is to assure adequate cooling for decay heat removal in the event of a safety valve lifting and exceeding the design blowdown of 5%. Se original question 3 from your July 2, 1984 letter was:
"Results from the EPRI tests on the Target Rock 69C safety valve indicate that blowdowns may exceed the design blowdown of 5%. he consequences of potentially higher blowdowns were not addressed in the Beaver Valley Unit 1 submittal.
Blowdowns in excess of the design blowdown of 5% could cause the pressure to be sufficiently decreased that adequate cooling might not be achieved for decay heat removal. Discuss these consequences of higher blowdowns if blowdowns in excess 5% are expected."
h e maximum blowdown would occur if the safety valve stuck open. his would be equivalent to, and be bounded by small break IDCA analysis.
Se Beaver Valley Unit 1 UFSAR section 14.3.1 presents the analysis of small break IDCAs using the WFIASH code.
he analysis for a 6 inch diameter break, which is the limiting size in terms of highest peak clad temperature, results in a maximum hot spot clad temperature of 1729 1
degrees F.
In addition small break IDCA analyses are being performed in response to NUREG-0737 items II.K.3.30 and II.K.3.31 using the NCFFRUMP code to verify the original licensing calculations. Se NRC has reviewed l
and issued SER's on the NCFFRUMP code (NCAP-10079) and the small break l
ECCS evaluation model using the NCFFRUMP code (NCAP-10054).
A plant specific response to NUREG-0737 item II.K.3.31 will be provided in I
accordance with the schedule for this ites, and is expected to reconfirm adequate cooling for decay heat removal.
3
QUESTION 5.
In the piping analysis report submitted June 24, 1983, it was stated the pr r functioning of the insulated boxes enclosing the SRV loop seals d be verified by testing. Provide the results of the testing for our review to confirm the SRV loop seal temperatures are now in the range of 310 degree F.
RESPCNSE After installation of the insulated boxes enclosing the SRV loop seals performed under Design Change 606, BVr 1.3-2.6.5 was performed to verify the SRV loop seal temperature.
Se results of the test performsd on 9/23/83 after a 20 hour2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> soak time are shown in Table 5.1.
Se observed temperatures fall within a range of 300 to 390 degrees F.
his is acceptable based on the following reasons:
1.
Tw of the three valves exceed the minimum temperature criteria of 310 degrees F.
2.
h e coldest valve has a temperature approximately 96 degrees F above the saturation temperature applicable to a back pressure of 15 7BfK-15 PSIA is the sum of the expected gage operation pressure of the relief tank and the absolute pressure of the containment building during power generation.
He temperature elevation above the back pressure saturation temperature observed by EPRI during the valve testing program was approximately 96 degrees F.
EPRI has not determined if the same support reaction measured in the hot loop seal test could be obtained at lower temperatures. Since the phenomena involved is flashing, the same support load should be observed at L
lower temperatures as long as the temperature is above saturation.
l 3.
Se temperature profile expected is such that the balance of the loop seal water is hotter than the section measured in DCP 606.
4.
Operation of safety valves in nuclear plants is a very rare occurrence.
5.
All the piping and supports meet seaver valley Unit I stress criteria.
6.
In the piping analysis it was assumed that all valves open simultaneously.
In actuality this is not the case.
4 i
l l
7-QUESTION 6 he piping analysis also indicated modifications would be made to one snubber and some anchor bolts would be proof-tested to higher loads than used during their original installation.
Were the snubber modification and the results of the new anchor bolt proof testing included in the piping analysis performed after EPRI testing?
In addition, after the anchor bolt testing is completed, provide a comparison of the allowable and calculated loads for those bolts tested.
RESPONSE
All of the pressurizer safety and relief valve (PSARV) piping system supports were evaluated to insure structural adequacy and compliance with the ASME Boiler and Pressure Vessel Code,Section III, subsection NF, 1977 edition with adenda to and including sunner of 1979 addenda. he results of this evaluation is presented in WCAP 10447 and shows that all of the supports meet the criteria without structural modification.
Although the calculated load on the snubber for support H-127 was less than the allowable it was noted that the swing angle exceeded the maximum allowed swing angle of 5 degrees as specified by the vendor. This excess swing angle would potentially bind the snubber and was due to upward movement of the pipe of almost 2" during normal operation. To reduce the swing angle the pipe clamp was lowered 0.5 inch so that the final angle in the normal operating position was decreased to a melmum of 4.56 degrees. This slight repositioning of the pipe clamp was evaluated as having no impact on the structural adequacy of tie support.
We evaluation of the loads on the anchor bolts was performed before the proof test program and therefore the results of the proof testing were not directly used in the evaluation.
%e proof tests however were used as confirmation of the adequacy of the bolts.
All of the anchor bolts analyzed in WCAP 10447 on the PSARV piping system supports were shown to be within the allowable load / stress limits.
A comparison of the allowable and calculated loads for th?se bolts tested are shown in Table 6.1.
5
r.
TABLE 2.1 PCV-RC-455C - STROKE TIMES (SEC.) 8/20/83 ORN CI M E CYCLE 1
2.47 6.25 8.72 2
2.38 6.22 8.60 3
2.38 6.22 8.60 4
2.42 6.08 8.50
'5 2.41 6.22 8.63 i
6 2.41 6.14 8.55 7
2.44 6.17 8.61 8
2.44 6.22 8.66 9
2.44 6.19 8.63 10 2.44 6.22 8.61 Avg.
2.42 6.19 8.61 i
PCV-RC-455D - STOKE TIMES (SEC) 9-6-83 onN CIME CYCLE i
1 2.34 7.16 9.50 2
2.34 7.09 9.43 3
2.34 6.94 9.28 4
2.34 7.00 9.34 5
2.34 6.97 9.31 6
2.33 7.27 9.60 7
2.36 7.16 9.52 8
2.34 7.45 9.79 9
2.34 7.41 9.75 10 2.34 7.38 7.72 Avg.
2.34 7.18 9.52 i
h e
r
(
6 l
7 I
r, TABLE 5.1 SW Loop Seal Temperatures TIME W-RC-551A RV-RC-551B W-RC-551C MIN.
Temp Gegree F Temp degree F Temp degree F INITIAL 308 388 339 2
308 388 341 4
308 388 342 6
308 388 342 8
308 388 341 10 305 388 343 12 308 387 342 14 308 387 342 16 308 388 343 18 308 388 344 20 308 388 344 22 308 388 343 24 308 388 341 26 305 388 342 28 308 388 344 Final 308 388 345 Average 308 387.875 342.375 l
l f
t l
l l
7 l
k
TABLE 6.1 Summary Table of calculated and Allowable Tension Loads for the various Anchor Bolts Either Tension or W rque Tested Maximum Anchor Calcu' nted Allowable Bolt Dia hre "
Load Tension Load Hanger No.
(In) sXI?S)
(KIPS)
H-31 1
4.15 5.86 H-33 3/4 2.54 3.12 H-115 1-1/4 2.40 4.13 1-1/4 3.12 4.13 1-1/4 4.92 6.26 1-1/4 2.62 4.13 1-1/4 3.34 4.13 1-1/4 4.68 6.26 N-116 1-1/4 6.71 6.76 H-120 1-1/4 6.36 6.98 H-128 1
4.35 5.58 H-129 1
4.64 5.58 H-303A 1/2 1.27 1.65 5/8 1.27 1.65 H-303a 3/4 1.18 3.54 H-304 1
2.78 5.29 H-305 1
3.17 5.29 l
l 8
\\