05000323/LER-2005-001, TS 3.4.10 Not Met During Pressurizer Safety Valve Surveillance Testing Due to Random Lift Spread
| ML050940202 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 03/28/2005 |
| From: | Oatley D Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| DCL-05-029, OL-DPR-82 LER 05-001-00 | |
| Download: ML050940202 (6) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(i) 10 CFR 50.73(a)(2)(vii), Common Cause Inoperability 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded 10 CFR 50.73(a)(2)(viii)(A) 10 CFR 50.73(a)(2)(ii)(B), Unanalyzed Condition 10 CFR 50.73(a)(2)(viii)(B) 10 CFR 50.73(a)(2)(iii) 10 CFR 50.73(a)(2)(ix)(A) 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(x) 10 CFR 50.73(a)(2)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.73(a)(2)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown 10 CFR 50.73(a)(2)(v), Loss of Safety Function |
| 3232005001R00 - NRC Website | |
text
Pacific Gas and Electric Company' David H. Oatley Vice President and General Manager Diablo Canyon Power Plant P. 0. Box 56 Avila Beach, CA 93424 805.545.4350 Fax: 805.545.4884 March 28, 2005 PG&E Letter DCL-05-029 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-323, OL-DPR-82 Diablo Canyon Unit 2 Licensee Event Report 2-2005-001-00 Technical Specification 3.4.10 Not Met During Pressurizer Safety Valve Surveillance Testing Due to Random Lift Spread
Dear Commissioners and Staff:
In accordance with 10 CFR 50.73(a)(2)(i)(B), PG&E is submitting the enclosed licensee event report regarding the pressurizer code safety valves being outside the Technical Specification 3.4.10 set pressure due to random lift spread.
This event was not considered risk significant and did not adversely affect the health and safety of the public.
Sincerely, David H. Oatley ddm/2246/A0630775 Enclosure cc/enc:
Bruce S. Mallett David L. ProuIx Girija S. Shukla INPO Diablo Distribution cc:
A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Caltaway
- Comanche Peak.
Diablo Canyon.
Palo Verde
- South Texas Project
- WolfCreek
NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB: NO. 3150-0104 EXPIRES: 06/30/2007
, the NRC may digits/characters for each block) not conduct or sponsor, and a person Is not required to respond to, the
- 3. PAGE Diablo Canyon Unit 205000323 1 OF 5
- 4. TITLE TS 3.4.10 Not Met During Pressurizer Safety Valve Surveillance Testing Due to Random Lift Spread
- 5. EVENT DATE
- 6. LER NUMBER
- 7. REPORT DATE
- 8. OTHER FACILITIES INVOLVED SEUNILRVFACILrTY NAME DOCKET NUMBER MONTH DAY YEAR YEAR NUMBER RNEOV MONTH DAY YEAR 05000 FACILnY NAME DOCKET NUMBER 01 27 2005 2005 -
001 -
00 03 28 2005 05000
- 9. OPERATING MODE 1.THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR§: (Check all that apply)
O 20.2201(b)
Cl 20.2203(a)(3)(i) 0 50.73(a)(2)(i)(C)
El 50.73(a)(2)(vii) 1 E 20.2201(d)
El 20.2203(a)(3)(ii)
El 50.73(a)(2)(ii)(A)
El 50.73(a)(2)(viii)(A)
El 20.2203(a)(1)
E: 20.2203(a)(4)
El 50.73(a)(2)(ii)(B) al 50.73(a)(2)(viii)(B) 0 20.2203(a)(2)(i) a 50.36(c)(1)(i)(A) a 50.73(a)(2)(iii)
El 50.73(a)(2)(ix)(A)
- 10. POWER LEVEL E 20.2203(a)(2)(ii)
El 50.36(c)(1)(ii)(A) al 50.73(a)(2)(iv)(A)
El 50.73(a)(2)(x) o 20.2203(a)(2)(iii)
El 50.36(c)(2) 0 50.73(a)(2)(v)(A)
E 73.71(a)(4) o 20.2203(a)(2)(iv)
El 50.46(a)(3)(ii) 0 50.73(a)(2)(v)(B) a3 73.71(a)(5) 100 El 20.2203(a)(2)(v)
El 50.73(a)(2)(i)(A)
El 50.73(a)(2)(v)(C)
El OTHER 0
20.2203(a)(2)(vi) 0 50.73(a)(2)(i)(B) a 50.73(a)(2)(v)(D)
Specify in Abstract below or in NRC Form 366A
- 12. LICENSEE CONTACT FOR THIS LER FACILILY NAME TELEPHONE NUMBER (Incdude Area Code)
Lawrence M. Parker - Senior Regulatory Services Engineer (805) 545-3386CAUSE SYSTEM COMPONENT l MANUJ-REPORTABLE l
CAUSE
SYSTEM COMPONENT MANU-REPORTABLE I
FACTURER TO EPIX FACTURER TO EPIX X
AB I RV l_C710 Yes ll____
- 14. SUPPLEMENTAL REPORT EXPECTED
- 15. EXPECTED MONTH DAY YEAR SUBMISSION El YES (If yes, complete 15. EXPECTED SUBMISSION DATE) 0 NO DATE ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)
On January 27, 2005, during scheduled testing in accordance with Surveillance Test Procedure M-77, "Safety and Relief Valve Testing," PG&E identified two of three pressurizer safety valves (PSVs) outside the Technical Specification (TS) 3.4.10, "Pressurizer Safety Valves,... lift setting of >2460 and <2510 psig." The as-found lifts were 4.4 and 3.6 percent low, and within analyzed safety limits; thus, this condition did not adversely affect the health and safety of the public.
The PSVs were disassembled, inspected, and reset within the TS 3.4.10 lift setpoint requirements. PG&E believes the cause of the PSVs being outside the TS allowance is random lift spread.
PSV lift setting repeatability has been recognized as an industry-wide problem. PG&E has participated in extensive investigative test programs, both jointly with the nuclear steam supply system vendor, Westinghouse Owners Group, and independently. The results of the industry investigations are documented in WCAP-12910, "Pressurizer Safety Valve Set Pressure."
PG&E has previously enhanced the PSV maintenance activities and testing procedures resulting in improved performance. No further corrective actions are required.
NRC FORM 366(6-2004)
PRINTED ON RECYCLED PAPER NRC FORM 366 (6-2004)
PRINTED ON RECYCLED PAPER
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL NUMBER REVISION NUMBER Diablo Canyon Unit 2_
0 5
0 l 3 1 2 3 2005 0 0 1 l 1 02 1 OF 5
TEXT I.
Plant Conditions
Unit 2 has operated in various plant modes with the described condition.
II.
Description of Problem A.
Background
TS 3.4.10, "Pressurizer Safety Valves," requires that three PSVs [AB][RV]
shall be operable with a lift setting greater than 2460 psig and less than 2510 psig corresponding to ambient conditions of the valve at nominal operating temperature and pressure. This upper and lower pressure limit is based on a nominal pressure of 2485 psig, with an upper and lower tolerance limit of plus or minus one percent.
Surveillance Test Procedure (STP) M-77, "Safety and Relief Valve Testing," verifies the PSV's lift setting in accordance with the requirements of the ASME Boiler and Pressure Vessel Code, Section Xl. The initial (as-found) lift setting is evaluated for TS compliance. STP M-77 requires that the valves lift within the required tolerance in order to declare them operable.
STP M-77 test methodology obtains the as-found lift setting by placing the PSVs in an environmentally controlled enclosure and heating the ambient air to the temperature conditions typical at Diablo Canyon Power Plant (DCPP). The loop seal is also heated to simulate the piping temperature conditions at DCPP. Testing is accomplished by the addition of steam at a defined ramp rate. Steam is added until physical evidence of stem movement is visible on the remote data acquisition display screen. The data is then reviewed to ascertain 'first discernible stem movement" and the pressure at which it took place.
B.
Event Description
Following the Unit 1 eleventh refueling outage (1 R1 1) in May 2002, the subject PSVs lift settings were verified to be within the range required by TS 3.4.10. The PSVs were then returned to warehouse stock. During the Unit 2 eleventh refueling (2R1 1) outage in February 2003, these three PSVs were placed in service in Unit 2 without any additional adjustment of the lift settings. The valves were replaced during 2R12 in November 2004 and tested offsite in January 2005.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL NUMBER REVISION l
~NUMBER Diablo Canyon Unit2 0
5 0 1 3 l2 l 3 2005 0 0 1 l-0 3
OF 5
TEXT On January 27, 2005, during scheduled offsite testing in accordance with STP M-77, 'Safety and Relief Valve Testing," PG&E identified two of three PSVs outside the TS 3.4.10, "Pressurizer Safety Valves,... lift setting of
>2460 and <2510 psig." The initial lift settings were 4.4 and 3.6 percent low.
The second valve lifts of these same two PSVs were found to be 3.1 percent low and 0.3 percent low, respectively.
PSV lift setting repeatability has been recognized as an industry-wide problem. PG&E has participated in extensive investigative test programs, both jointly with the nuclear steam supply system vendor, Westinghouse Owners Group, and independently. The results of the industry investigations are documented in WCAP - 12910, "Pressurizer Safety Valve Set Pressure."
C.
Inoperable Structures, Systems, or Components that Contributed to the Event None.
D.
Other Systems or Secondary Functions Affected
None.
E.
Method of Discovery
This condition was identified during routine scheduled testing performed in accordance with STP M-77 at the offsite testing facility.
F.
Operator Actions
None.
G.
Safety System Responses None.
Ill.
Cause of the Problem A.
Immediate Cause Two of three PSVs did not lift within the TS 3.4.10 lift-setting tolerance.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6)
PAGE (3)
YEAR SEQUENTIAL NUMBER REVISION NUMBERI DiabloCanyonUnit2 05 1
0 1 3 2 3 2005
- - I 0 l0 1
0 0 4 1 OF l 5 TEXT B.
Root Cause The cause of the as-found lift setting has been determined to be random lift spread.
C.
Contributory Cause None.
IV.
Assessment of Safety Consequences
The limiting event for evaluating the lift setting is the loss of load analysis that requires the maximum reactor coolant system (RCS) pressure of 2750 psia not be exceeded. The RETRAN computer model was run to determine if, with the as-found PSV lift setpoint, the RCS pressure would exceed 110 percent of ASME design acceptance criteria, or 2750 psia. The analysis confirmed that the as-found set points would have maintained adequate RCS pressure relief capacity, such that the plant remained bounded by the limiting loss of load analysis provided in Final Safety Analysis Report Update, Section 15.2.7, "Loss of External Electrical Load and/or Turbine Trip." Also, the as-found lift setting was reviewed for potential interaction with the pressurizer power-operated relief valves (PORVs) and the potential for inadvertent low pressure lifting, and were found acceptable.
Therefore, this event was of very low risk significance, was not a Safety System Functional Failure, and did not adversely affect the health and safety of the public.
V.
Corrective Actions
A.
Immediate Corrective Actions
The valves were disassembled, inspected, reset within tolerance, and returned to warehouse stock.
B.
Corrective Actions to Prevent Recurrence No corrective action to prevent recurrence is required because the inherent characteristics of the valve are within the safety analysis basis of DCPP.
LICENSEE EVENT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1)
DOCKET NUMBER (2)
LER NUMBER (6) l PAGE (3)
YEAR SEQUENTIAL NUMBER REVISION NUMBER DiabloCanyonUnit2 0 5 10 0 01 3 12 3 2005 0
0 1
0 5 1 OF l 5 TEXT VI.
Additional Information
A.
Failed Components None.
B.
Previous Similar Events
LER 1-94-009, Revision 2, submitted in PG&E Letter DCL-95-248, dated November 7, 1995, regarding PSVs found outside TS limits during the Unit 1 sixth refueling outage. The root cause of this event was determined to be random lift-setting spread. No corrective action to prevent recurrence was required because this inherent characteristic of the valve was within the analysis basis of DCPP. However, a prudent action to replace the PSV upper spring washer was recommended. The implementation of this prudent action has been deferred until NRC concerns regarding valve performance can be acceptably resolved.
LER 1-95-016, Revision 2, submitted in PG&E Letter DCL-98-077, dated May 28,1998, regarding PSVs found outside TS limits during the Unit 1 seventh refueling outage. The root cause of this event was determined to be random lift-setting spread. No corrective action to prevent recurrence was required because this inherent characteristic of the valve was within the analysis basis of DCPP. However, a prudent action to replace the PSV upper spring washer was recommended. The implementation of this prudent action has been deferred until NRC concerns regarding valve performance can be acceptably resolved.
LER 2-2001-004, submitted in PG&E Letter DCL-01-090, dated August 27, 2001, regarding one PSV found 3.4 percent low and one PSV found 2.8 percent high during offsite testing. The root cause of this event was determined to be random lift-setting spread. No corrective action to prevent recurrence was required because this inherent characteristic of the valve was within the analysis basis of DCPP.
LER 1-2002-005-00, submitted in PG&E Letter DCL-02-091, dated August 9, 2002, regarding one PSV found 1.9 percent low and one PSV found 2.6 percent high during offsite testing. The root cause of this event was determined to be random lift-setting spread. No corrective action to prevent recurrence was required because this inherent characteristic of the valve was within the analysis basis of DCPP.