ML16340B721
| ML16340B721 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 05/08/1981 |
| From: | Cherny F Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML16340B719 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8105120017 | |
| Download: ML16340B721 (36) | |
Text
UNITED STATES OF AtiERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOf~aIC SAFETY AND LICENSING BOAPD 4
In the Yiatter of
)
5 6
PACIFIC GAS AND ELECTRIC COMPANY 7
8
{Diablo Canyon Nuclear Power Plant I 9
Unit Nos.
1 and 2)
)
Docket Nos.
50-275 O.L.
50-323 O.L.
10 11 12 TESTIHONY OF FRANK C.
CHERNY ON RELIEF, SAFETY AND BLOCK VALVES AT DIABLO CANYON NUCLEAR FACILITY 13 g.
Please state your full name.
14 A.
Frank C. Cherny.
15 g.
By whom are you employed?
16 A.
I am employed by the U.S. Nuclear Regulatory Commission.
I am a
17 18 Section Leader in the Hechanical Engineering Branch, Division of Engineering, Office of Nuclear Reactor Regulation.
19 g.
Describe the nature of your work with respect to the Diablo Canyon 20 operating license proceeding.
21 A.
Task Coordinator for NUREG-0737 Item II.d. 1 "Performance Testing of 22 Boiling Water Reactor and Pressurized Water Ractor Relief and Safety 24 25 Valve.
The Diablo Canyon SER input for qualification of safety and relief valves, and block valves, as required by THI Item I.D.1 was prepared under my supervision.
26 g.
Would you detail your professional qualifications?
27 A.
Attached is a copy of my professional qualifications.
y l
28 g.
Have you, in the course of your professional experience, had 29 occasion to review the design and qualification of relief, safety or 30 block valves at Nuclear Power Plants other than Diablo Canyon?
31 A.
- Yes, I have reviewed such designs or assisted in their review.
I 32 33 30 36 have also participated in many meetings, discussions, design reviews of a generic nature, for the last five years, related to the design of overpressure protection systems for nuclear power plants.
tetany such activities were related to direct participation in industry standards writing activities associated with overpressure protection 37 38 of nuclear reactor plant components and inservice testing of nucelar power plant pressure relief devices.
39 g.
In the course of your work on Diablo Canyon did you review the 40 design and qualification of Reactor Coolant system relief valves?
41 A.
Yes.
42 g.
In the course of your work on Diablo Canyon did you review the 43 design and qualification of Reactor Coolant system safety valves?
44 A.
Yes.
45 g.
In the course of your work on Diablo Canyon did you review the 46 design and qualification of Reactor Coolant system block valves?
47 A.
Yes.
48
-g.
Are there General Design Criteria (GDC) which must be met which 49 apply to relief and safety valves?
1 50 A.
- Yes, GDC 1, 14, 15 and 30 require the Applicant to assess their 51 RCPB, including safety and relief valves, to meet certain standards.
52 g.
What standards of review does the Staff use when reviewing for 53 compliance with GDC 1, 14, 15 and 30?
54 A.
55 56 57 58 59 60 61 In reviewing for compliance with GDC 1, 14, 15 and 30 the following standards are used:
(a)
- Standard Review Plan (SRP) 3.9.2, "Dynamic Testing and Analyses of Systems, Components, and Equipment.";
(b)
SRP 3.9.3, "ASME Code Class 1,
2 and 3 Components, Component
- Supports, and Core Support Structures.";
(c) Regulatory Guide 1.48 "Design limits and loading combinations for seismic Category 1 fluid systems components.";
and (d) Regulatory Guide 1.68 "Pre-operational and Initial Startup Test Programs for Water Cooled Power Reactors."
62 g.
Would you briefly describe what "Standard Review Plan (SRP) 3.9.2.,
63 "Dynamic Testing and Analyses of Systems, Components, and Equipment" 64 requires as regards reactor coolant system safety and relief valves?
65 A.
Standard Review Plan 3.9.2 requires a (1) piping vibration preop.
66 67 68 test program (2) seismic qualification of safety related mechanical equipment (3) dynamic system analysis to insure structural adequacy of piping loops for LOCA & SSE.
69 g.
Would you briefly describe what SRP 3.9.3 "ASME Code Class 1,
2 and 70 71 3 Components, Component Supports, and Core Support Structures" requires as regards reactor coolant system safety and relief valves?
V 1
72 A.
Standard Review Plan 3.9.3 (1) specifies what load combinations and 73 74 stress limits apply (2) operability assurance program for "active" pumps and valves (3) design of pressure relief valve supports and associated discharge piping supports.
76 Q,
Would you briefly describe what Regulatory Guide 1.48 "Design limits 77 and loading combinations for seismic Category 1 fluid systems 78 components" identifies as a standard as regards reactor coolant 79 system safety and relief valves?
80 A.
Regulatory Guide 1.48 delineates acceptable design limits and load 81 82 combinations associated with normal operation and accident conditions.
83 Q.
Would you briefly describe what Regulatory Guide 1.68 84 "Pre-operational and Initial Startup Test Programs for Water Cooled 85 Power Reactors" identifies as a standard as regards reactor coolant 86 system safety and relief valves?
87 A.
Regulatory Guide 1.68 requires pre-operational testing of safety and 88 relief valves to demonstrate that they will operate as required.
89 Q.
What remains to be done by the Applicant to comply with GDC 1, 14, 90 15 and 30?
91 A.
The tests performed to date do not cover loadings which result from 92 transition flow from steam to water or solid fluid flow.
93 g.
How will the loadings which result from transition flow or solid 94 fluid flow be addressed with respect to reactor coolant system 95 safety and relief valves?
96 A.
A test program has been initiated by the Electric Power Research 97 98 99 100 101 102 103 104 105 106 107 108 109 110 112 113 114 115 116 117 118 Institute (EPRI) which will address safety and relief valve operability including loadings resulting from transition flow from steam to water and solid fluid flow.
PGEE has committed to participating in this program and has as one of its objectives to satisfy the long-term requirements on SRV testing as set forth in Section 2. 1.2 of NUREG-0578, "THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations" included as of October 31, 1978 as Item II.D. in NUREG-0737.
The Applicant has referenced the ongoing EPRI/NSAC safety and relief valve testing program.
A description of the EPRI program was provided to NRC by EPRI in 1979 and an updated revision in July 1980.
As noted in Supplement 14, the Staff is generally in agreement that the NUREG-0737 technical requirements for safety and relief valves can be met subject to receipt of additional information which was requested by letter of November 26, 1980 to Russel l C.
Youngdahl.
By letter of December 15, 1980 EPRI responded to both the Staff'.s November 26, 1980 letter and NUREG-0737.
PGSE has referenced the EPRI December 15 response in their February 26, 1981 letter.
The Staff has not completed its review of the December 15, 1980 letter primarily as regards EPRI proposed documentation submittal dates for safety and relief valves and comments made in the letter regarding resolution of a block valve test program after
119 120 121 122 123 124 125 126 127 128 129 130 131 132 133 134 July 1, 1981, the scheduled completion date for safety and relief valve testing.
Based on the Staff review of the EPRI program and PG&E's assurance that the EPRI program is applicable to the Diablo Canyon safety and relief valve designs, I believe there is adequate assurance that the NUREG-0737 requirement regarding performance verification of the RCS relief and safety valves will be met satisfactorily for Diablo Canyon.
Should this program demonstrate that these valves are not qualified for the above-stated loadings the Staff will require the licensee to take corrective actions.
Present schedules indicate that this testing will be completed by July 1, 1981.
With regard to the safety valves, there is presently no evidence that these valves will not operate properly during the anticipated transients which produce transition flow and solid fluid flow.
135 g.
Other than with respect transition flow and solid fluid flow, has 136 the Applicant demonstrated compliance with the standards you 137 identified as necessary to meet GDC 1, 14, 15 and 30?
138 A.
Yes, compliance with Standard Review Plan (SPR) 3.9.2, "Dynamic 139 140 141 142 Testing and Analyses of Systems, Components, and Equipment," at Diable Canyon which includes relief and safety valves is demonstrated in the Safety Evaluation Report on Diablo Canyon (SER)
Section 3.9. 1.
and SER Supplements 7, 8, and 9.
I 143 Compliance with SRP 3.9.3, "ASME Code Class 1, 2, and 3
144 145 146 147 150 151 152 153 154 155 156 157 158 159 Components, Component
- Supports, and Core Support Structures" at Diablo Canyon, which includes relief and safety valves, is demonstrated in SER Sections 3.9.2 and 5.2. 1.
Compliance with Regulatory Guide 1.48 "Design limits and loading combinations for seismic Category 1 fluid systems components" at Diablo Canyon, which includes relief and safety valves, is demonstrated in SER Section 5.2.1.
8 14).
Compliance with Regulatory Guide 1.68 "Pre-operational and Initial Startup Test Programs for Water Cooled Power Reactors" at Diablo Canyon, includes testing of relief and safety valves, is demonstrated in SER Section 14.
(SER).
Compliance with the appropriate sections of Appendix 8 to 10 C.F.R. Part 50 at Diablo Canyon, including relief and safety valves, is demonstrated in SER Section 17.4.
(SER).
160 g.
Is there anything, in your opinion, other than compliance with GDC 161 1, 14, 15 and 30, which demonstrates the reliability of the Reactor 162 coolant system relief and safety valves at Diablo Canyon?
163 A.
Yes.
In addition to complying with the requirements of GDC 1, 14, 164 165 166 167 15 and 30 as discussed above.
The reactor coolant system safety valves were originally designed and tested for operation on saturated steam in accordance with the applicable edition and addenda of Section III of ASME Boiler and Pressure Vessel Code.
f
158 169 170 171 172 173 Verification of this testing appears in the FSAR for Diablo Canyon, Table 5.2-1.
As required by Article 9 of the ASME Boiler and Pressure Vessel
- Code, the safety valve relieving capacity has been provided so that the pressure limitation specified in the Code will be maintained under all of the system transients and accidents postulated to occur.
174 g.
Why do you believe that this further demonstrates the qualification 175 of the reactor coolant system and safety valves?
176 A.
Section III of the ASME Code provides specific valve functional 177 178 179 180 181 requirements and installation requirements for the reactor coolant system safety valves.
Additionally, it places restrictions on the types of pressure relief valves that can be used for such application and provides testing requirements for certifying the relieving capacity of the safety valves.
182 g.
Is there anything else you believe demonstrates the qualification of 183 the reactor coolant system and safety valves?
184 A.
Both safety valves and two of the three relief valves have been 185 186 187 188 189 seismically qualified to be functional after exposure to loads resulting from the maximum hypothetical earthquake for Diablo Canyon as documented in Amendment 50 to FSAR Table 7-7.
Also Safety and Relief valves will be operationally tested during the pre-op test program performed in accordance with Regulatory Guide 1.68.
190 g.
Why do you believe that these facts further demonstrate the 191 qualification of the reactor coolant system relief and safety 192 valves?
193 A.
The pre-op testing of safety and relief valves demonstrates the 194 195 196 197 198 operational readiness of the valve to lift within the prescribed set pressure range.
Seismic qualification assures that the valves will function should they be subjected to the maximum hypothetical earthquake which was postulated for Diablo Canyon.
199 g.
Are you familiar with I and E Bulletin 81-2?
200 A.
Yes.
201 g.
Would you briefly summarize Bulletin 81-2?
202 A.
I and E Bulletin 81-2 is entitled "Failure of Gate Valves to Close 203 204 205 206 207 208 209 210 211 212 Against Differential Pressure".
The Bulletin discusses valve closure tests recently performed by EPRI at the Marshall Test Facility on seven gate valves of the type commonly used as PWR PORV Block Valves.
The testing included closing the valve against full flow steam differential pressure conditions selected as being representative of those that a
PORV might be expected to close against.
The Bulletin discusses the fact that three of the seven tested gate valves failed to fully close when subjected to the test conditions.
It notes that valves of the type that failed are also supplied for utilization in a number of safety related applications
213 214 215 216 217 218 219 in addition to the PORV block valve application where closure with a differential pressure across the valve is a requirement.
It requests that Licensees and Construction Permit Holders determine whether any of the "failed" valves are installed in such applications or planned to be installed.
If no valves in this category are found this is to be reported to NRC.
If one or more valves of this type are identified in such an application or 220 221 222 223 224 225 226 intended for such an application, it imposes requirements for specific actions to be taken by both Licensees and Construction Permit Holders.
Action to be taken include an evaluation of the significance of the valve failure to close on system operability in accordance with the plant technical specifications, modification of valves so they are qualified for the intended service or obtaining of qualified replacements.
227 g.
What type of valves are present at Diablo Canyon?
228 A.
The Diablo Canyon plant has three Crosby HB-BP-86 (6M6) safety 229 valves, three Masoneilan 20,000 Series (2
NPS)
Power Operated Relief 230 Valves and three Velan ¹B10-30548013M Motor Operated Block Valves.
231 g.
Have any of these valves undergone the EPRI testing program?
232 A.
As of May 5, 1981 the following testing had been performed by EPRI 233 234 235 on valves of the type installed at Diablo Canyon:
Power Operated Relief Valve - A Masoneilan 20,000 Series PORV-full flow steam test.
I 236 237 238 kB10-3054B013H Velan Hotor Operated Block Valve - full flow steam test.
This valve is the same model as that used at Diablo Canyon.
239 g.
Were any of the block valves which failed the EPRI tests discussed 240 in I and E Bulletin 81-2 of the type to be used at Diablo Canyon?
241 A.
No.
242 g.
What testing remains go be completed on the valves at Diablo Canyon?
243 A.
Additional testing of the Hasoneilan Power Operated Relief Valve to 244 245 246 247 248 249 250 251 252 253 254 include additional fluid effects that the valves could be exposed to under design basis transient or accident events such as the effects of loop seals and subcooled and saturated liquid is scheduled to be completed by July 1, 1981.
Complete testing of the Crosby HB-BP-86 (6M6) Safety Valve is scheduled to be performed during June of 1981 with a scheduled completion date of July 1, 1981.
The need for additional qualification testing of the Velan PORV Block Valve, as of Hay 5, 1981, is under discussion between PWR utilities and the NRC Staff.
If additional testing is needed it must be completed by July 1, 1982 as specified in NUREG-0737.
255 g.
Will the remaining testing be completed prior to fuel load?
256 A.
The qualification testing of Safety Valves and PORV's of the type 257 installed on Diablo Canyon is scheduled to be completed by July 1, 258 198l, which will be well prior to fuel load.
I
0 259 g.
Have any Diablo Canyon valves failed during the testing program?
260 A.
As of tray 5,
- 1981, no valves of the type used in Diablo Canyon have 261 failed any of the EPRI tests.
262 g.
What is done if a valve fails a test acceptance criterion?
263 A.
EPRI has established a procedure so that all utilities participating 264 265 266 267 268 269 270 271 272 273 274
, 275 276 277 278 279 280 281 282 283 in the program, the NSSS vendors, the valve manufacturers, and the NRC are all notified within a few days of any instances where a
valve fails a test acceptance criterion.
The NSSS vendors, with assistance from EPRI, assist the individual utilities with plant specific evaluations of the safety significance of any such failures.
Depending on the results of these evaluations, actions are taken by the utilities in accordance with the regulations as regards reporting,to NRC, possibly declaring equipment inoperable, if installed on an operating plant, and modifications or replacements of affected components for both operating plants and plants like Diablo Canyon that have a Construction Permit.
The regulations also require the NSSS vendors and valve manufacturers to report safety related equipment anomalies.
Additionally, the Office of Nuclear Reactor Regulation at the NRC independently reviews the details of all reported failures on a case by case basis and a
decision is made as to what appropriate action should be taken.
For safety and relief valves of the type installed at Diablo
- Canyon, as noted above, testing is scheduled to be completed by July 1, 1981.
If any failures of valves of this type occur'n the EPRI testing
- program, NRC will require the effects of the specific
)
284 285 286 287 288 289 290 291 292 293 294 295 296 297 298 299 failure on safe operation of the Diablo Canyon to be expeditiously evaluated by PG&E on a schedule such that any necessary modifications or replacements of safety or relief valves can be made prior to initial fuel loading.
As noted above, it is not clear at thi s time whether additional testing is required to confirm the capability of the type of PORV Block Valve installed at Diablo Canyon to open and close against all fluid conditions that could result from design basis transients and accidents.
If more testing is required, as specified in NUREG-0737, it must be completed by July I, 1982.
If it is determined that additional testing of block valves is required to confirm their performance capability, a procedure will be established for expeditious handling of adverse test results.
Valve modifications or replacements, if any are warranted, will be made to the Diablo Canyon PORV Block Valves on a schedule consistent with the safety significance of any observed anomalies.
300 g.
In view of the above testimony, do you have an opinion as to whether 301 fuel loading and low power testing can commence at Diablo Canyon 302 while PORV Block Valves remain to be tested?
303 A.
As noted in this testimony, it has not been determined that 304 additional testing will be required to confirm the opening and 305 306 307 308 closing capability of the Diablo Canyon type block valves.
Based on the fact that the Diablo Canyon safety and relief valves will be fully qualified prior to fuel loading for service conditions far in excess of those conditions valves could be exposed
1 309 310 311 312 to during low power testing, the other factors discussed
- aboe, and on the testimony of Norman Lauben, it 'is my opinion that fuel loading and low power testing can commence at Diablo Canyon with no adverse affect on the health and safety of the public.
l
UNITED STATES OF AMERICA HUCLEAR REGULATORY CONHISSIOH BEFORE THE ATOliIC SAFETY AHD LICEHSIHG BOARD In the Hatter of PACIFIC GAS AND ELECTRIC COlRPAHY (Diablo Canyon Nuclear Power Plant
)
Unit Hos.
1 and 2)
)
Docket Nos.
50-275 O.L.
50-323 O.L.
FRANK C.
CHERNY PROFESSIONAL QUALIFICATIONS tlECHAHICAL ENGINEERING BRANCH DIVISION OF ENGINEERIHG I am a fiechanical Engineer in the tdechanical Engineering Branch responsible for the review and evaluation of design crit ria of mechanical components, of methods of dynamic analysis and testing of safety related systems and components and of criteria for protection against dynamic effects associated with postulated failures of fluid system components for nuclear service.
I graduated from Marquette University with a B.S.
degree in h1echanical Engineering in 1965.
From July 1965 to November 1968 I was employed by the Babcock 5
Wilcox Co. at offices in both Barberton and Akron, Ohio.
During the majority of this period I was engaged in materials engineering
- work, primarily writing technical ordering requirements for primary pressure boundary materials to be used for reactor vessels, steam generators, and presurrizers for both commercial and U.
S.
Navy nuclear systems.
In addition I had assignments of several months duration each in quality
I
control engineering and nuclear stean supply system performance engineering.
From November 1968 to !!ay 1974 I was employed in the Pressurized Mater Reactors Division of Westinghouse Nuclear Energy Systems.
!iy work experience during this period includes the following:
From November 1968 to Hay 1970 and September 1970 to April 1971 as a Reactor Vessel Project Engineer based in Honroeville, Pa.:
( 1)
I had overall project engineer responsibility for design and construction of reactor vessels for several Westinghouse nuclear power plants in the U.S.
tiy responsibilities included preparation of Design Specifications and review of vendor decumentation for compliance with Westinghouse, Utility, ASIDE, Architect Engineer and AEC requirements.
I was personally responsible for coordination of the technical aspects of the transfer of two partially completed reactor vessels from a U.S.
manufacturer's shop to a European manufacturer for completion when schedular problems developed at the U.S. manufacturer.
(2)
After the U.S.-Europe transfer of these components, I assumed responsibility for technical coordination between the primary Westinghouse nuclear engineering office in the U.S.
and an overseas office established in Brussels, Belgium to do project engineering work for mechanical components used in Westinghouse nuclear plants both in the U.S.
and in Europe.
From June 1970 to September 1970 and from April 1971 to December 1972 I was employed by Westinghouse Nuclear Energy Systems in Europe based in Brussels, Belgium.
!1y responsibilities included:
l
~
(1)
During the June-September, 1970 period and from April 1971 to about April 1972 I had project engineer responsibility for several reactor pressure vessels and a pressurizer.
I also acted as Westinghouse engineering representative for U.S.
AEC guality Assurance audits of European vessel manufacturers.
(2)
From April 1972 to December 1972 I served as a lead engineer with a broader scope of responsibility.
I was responsible for reactor pressure
- vessels, pressurizers and reactor vessel supports fabricated in Europe for Westinghouse Nuclear Plants.
Several engineers and a technician reported directly to me during this period.
The work included preparation of Design Specifications, review and approval of vendor design and manufacturing documentation, and coordination with both U.S.
and European utility and regulatory representatives.
From December 1972 to May 1974 I was again based in Honroevi lie, Pa., this time as Senior Reactor Yessel Project Engineer.
I was responsible for the technical adequacy of several reactor pressure vessels being manufactured in the U.S. for use in Westinghouse Nuclear Plants in Europe.
I was also responsible, during the majority of this period, for the training of a Westinghouse Nuclear - Europe engineer temporarily based in the U.S.
In Hay of 1974 I star ted work for the Regulatory Division of the U.S. Atomic Energy Commission and have remained through the transition into the U.S. Nuclear Regulatory Commission.
In 1977 I was appointed as a Section Leader in the Mechanical Engineering 8ranch.
As a branch technical reviewer and later as a Section Leader I have been
participating in the review of construction permit and Operating License s
applications.
Since July of 1974 I have served as a member of tne ASl1E Section III Subgroup On Pressure Relief whicn is responsible for writing industry standards for the overpressure protection of light water reactor plant components.
Additionally, since 1977 I have been a member of the AS)~!E Working Group on Safety and Relief Valves.
The Working Group has recently completed work on a proposed industry standard entitled "Require!vents for Inservice Performance Testing of Nuclear Power Plant Pressure Relief Devices."
The proposed standard was issued for public comment by AS!1E early in 1981.