ML20039F639
| ML20039F639 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 01/08/1982 |
| From: | Jensen W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20039F620 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8201130179 | |
| Download: ML20039F639 (18) | |
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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0fmISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of PACIFIC GAS AND ELECTRIC COMPANY Docket Nos. 50-275 0.L.
50-323 0.L.
(Diablo Canyon Nuclear Power Plant,
)
Unit Nos. 1 and 2)
)
TESTIMONY OF WALTON L. JENSEN, JR.
REACTOR SYSTEMS BRANCH DIVISION OF SYSTEMS INTEGRATION January 8, 1982 foRkoo$oNo$5 T
I UNITED STATES OF AMERICA NUCLEAR REGULATORY CDPMISSION 2
3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4
In the Matter of
)
5
)
PACIFIC GAS AND ELECTRIC COMPANY Docket Nos. 50-275 - OL 6
50-323 - OL (Diablo Canyon Nuclear Power 7
Plant, Unit Nos. 1 and 2)
)
8 NRC STAFF TESTIMONY OF W. JENSEN RELATIVE TO THE CLASSIFICATION OF PRESSURIZER HEATERS AS COMPONENTS 9
IMPORTANT TO SAFETY 10 (CONTENTION 10) 11 Q)
Please state your name and position with the NRC.
12 A)
My name is Walton L. Jensen, Jr.
I am an employee of the U. S. Nuclear 13 Regulatory Commission assigned to the Reactor Systems Branch, Division of 14 Syst' ems Integration, Office of Nuclear Reactor Regulation. In this position 15 I have lead responsibility for the Reactor Systems Branch review of Diablo 16 Canyon Nuclear Power Plant - Units 1 and 2.
17 Q)
Have you prepared a statement of professional qualifications?
18 A)
Yes. A copy of this statement is attached to this testimony.
19 Q)
In the course of your review, are you required to make detenninations of 20 whether systems and components are appropriately classified?
21 A) Yes.
In the review of analyses of design basis transients in Chapter 15 of 22 FSARs, I determine that those systems which are relied upon to mitigate the 23 event and achieve safe shutdown are appropriately classified as safety 24 grade.
In the review of chapter 5.2 of FSARs, I determine that the reactor 25 system pressure boundary is protected with safety grade equipment.
2 26 Q)
What is the purpose of this portion of your testimony?
27 A)
The purpose of my testimony is to respond to Contention Number 10 (as admitted by the Board in the September 30, 1981 Memorandum and I
I 28 Order) which states:
29 The Staff recognizes that pressurizer heaters and associated controls are necessary to maintain natural circulation at het stand-by conditions.
30 Therefore, this equipment should be classified as " components important to safety" and required to meet all applicable safety-grade design 31 criteria, including but not limited to diversity (GDC 22), seismic and environmental qualification (GDC 2 and 4), automatic initiation (GDC 20),
32 separation and independence (GDC 3 and 22), quality assurance (GDC 1),
adequate, reliable on-site power supplies (GDC 17) and the single failure 33 criterion. The Applicant's proposal to connect two out of four of the heater groups to the present on-site emergency power supplies does not 34 provide an equivalent or acceptable level of protection.
5 Q)
What is the function of the pressurizer heaters?
6 A)
The pressurizer heaters are part of the normal control system which regulates primary system pressure. When the pressurizer heaters are 38 activated, boiling occurs within the pressurizer producing steam which acts to increase reactor system pressure. The reactor system pressure 40 may be reduced by operation of the pressurizer sprays which condense the staam in the pressurizer.
(FSAR Chapter 5) 42 43 Q
What are hot standby conditions at Diablo Canyon?
44 A)
Hot standby is defined in NUREG-0452, Rev.2, " Standard Technical Specifi-45 cations for Westinghouse Pressurized Water P.eactors" as a condition for 46 which the core is subcritical by at least 1% in reactivity and the reactor 47 coolant temperature is above 350*F.
If the reactor temperature were 48 reduced below 350*F the state of the reactor would be described as " hot 49 shutdown" and below 200*F as " cold shutdown". During operation at " hot 50 standby" reactor decay heat is removed using the steam generators and for hot and cold shutdown operation the reactor heat is removed utilizing i
51 the Residual Heat Removal (RHR) system.
52 53 Q ) Are the Pressurizer Heaters necessary to maintain the reactor system t
54 at hot standby?
l 1
55 A) Yes, operation of the pressurizer heaters is required to maintsin primary 56 system pressure at hot standby. Tests at the Sequoyah Nucler.i Power Plant 57 have demonstrated that the effect of deenergizing the pressurizer heaters 58 would be gradual depressurization of the primary system (100 psig/ hour).
59 The Sequoyah tests demonstrated that the reactor system pressure could 60 be controlled using the safety grade coolant charging pumps. However, con-61 tinued maintenance of an elevated reactor system pressure in this manner at 62 hot standby would cause the pressurizer level to increase and cause an eventual 53 loss of the pressurizer steam bubble. Operation of the reactor at hot 64 standby is not permitted at Diablo Canyon without a pressurizer steam bubble 65 and the reactor would be cooled to a hot or cold shutdown condition and placed 66 on the RHR system.
67 Q)
Can the reactor be brought to a cold shutdown condition without the pressurizer heaters?
69 A) Yes, operation of the pressurizer heaters is not required to bring the plant to cold shutdown which is a safe and stable condition. One of the first 71 actions in operating procedure L-5 for bringing the plant from hot standby to cold shutdown is to turn off the pressurizer heaters. Although the procedure requires that the reactor system pressure be temporary held at approximately 400 psig until the RHR system can be activated (for which the pressurizer heaters may be utilized) this function could also be accomplished by control 76 of flow from the charging system. Cooling of the reactor system from hot 77 standby to cold shutdown causes a decrease in the pressurizer level. The 70 operator could control pressurizer level by adding water from the safety 79 grade coolant charging pumps. The plant may be maintained at cold shutdown 80 without pressurizer heaters.
81 Q)
Is it nocessary from a safety standpoint to maintain the plant at a hot 82 standby condition?
83 A) No. No particular safety function is served by maintaining the plant in 84 a hot standby condition.
85 Q) Could natural circulation be maintained at Diablo Canyon following a loss of 86 Pressurizer heaters?
87 A) Yes. The tests at the Sequoyah Nuclear Power Plant demonstrated that the 88 natural circulation could be maintained without operation of the pressurizer 89 heaters while the reactor coolant was in a subcooled condition (below the 90 boiling temperature). Operators at Diablo Canyon are instructed to maintain 91 subcooled conditions within the primary system.
(See Emergency Operating 92 Procedure OP-22 Emergency Shutdown and Emergency Operating Procedura OP-23 93 Natural Circulation) 94 95 Q)
If the reactor system reached the boiling temperature could natural circulation 96 be maintained?
97 A) Yes.
For plants with U-tube steam generators, such as Diablo Canyon, 98 the high points of the coolant loops are the U-bends of the stean generator 99 tubes which are continually covered with secondary coolant supplied by the 100 main or auxiliary feedwater system. Steam formed in the coolant loops of 101 a plant of the Diablo Canyon design would be condensed by the steam generators 102 with no interruption of natural circulation.
If sufficient steam were 103 present, the mode of natural circulation would change from single-phase natural 104 convection to two-phase boiling-condensation. Tests at the LOFT and 105 Semiscale facilities have demonstrated that continuous natural circulation 106 will occur at plants equipped with U-tube steam generators in the presence 107 of steam in the coolant loops as long as steam generator cooling remains 108 available. Semiscale results are documented in Report No. EGG-SEMI-5507, 109
" Quick Look Report for Semiscale Mod-2A Test S-NC-2," July 1981. LOFT 110 results are documented in report No. NUREG CR-1570 " Experimental Data 111 Report for LOFT Nuclear Small Break Experiments L3-7," August 1980.
5 112 g) The pressurizer heaters are utilized in a number of emergency operating procedures.
113 Doesn't this indicate that they should be safety grade?
114 A) Although the pressurizer heaters would be expected to be nomally available 115 in a number of anticipated transients and accidents (as they are during 116 routine daily operations), they are not required to protect the reactor.
117 As discussed in Chapter 15 of the FSAR, reliance is placed on the reactor 118 protection system and the engineered safety features which are designed to 119 safety grade criteria.
It must be recognized that. emergency procedures are 120 written to present guidance on all options available to the operator for 121 coping with a plant transient or accident. To do otherwise would be an 122 unwise limitation of design and operator capability during each event. FSAR 123 Chapter 15 analyses, in which conservative assumptions on the unavailability 124 of non-safety grade components are typically imposed should not be confused 125 with an actual transient or accident during which the oparator would be 126 expected to make maximum use of ALL available systems, whether or not 127 they are safety grade.
128 Q)
Then why does NUREG-0578 state that "... there is a need to consider the upgrading of those pressurizer heaters and associated controls D0 required to maintain natural circulation at hot standby conditions to a safety-grade classification..
."?
A)
Failure of the pressurizer heaters to operate would allow the reactor 33 system to gradually depressurize which, in the absence of any corrective peradon acdon, woM evenhaHy cause anomade achadon of W m.
135 t
i w
-w y-
6 136 Section 2.2.2, Page A-2 of NtREG-0578 states the need to consider the upgrade 137 of these systems is "to achieve greater heater reliability and to decrease the 138 number of demands for operation of the Emergency Core Cooling System." The 139 repeated unnecessary actuation of the Emergency Core Cooling System is 140 undesirabl e.
The automatic actuation of ECCS for a loss of pressurizer neaters 141 would be an unlikely event at Diablo Canyon since adequate means is provided 142 to the operator to control system pressure utilizing the charging system and 143 by controlling the cooldown rate of the steam generators.
144 The capability to provide emergency power to the pressurizer heaters is g
available at Diablo Canyon to reduce the number of demands for ECCS to 16 yg operate in accordance with Item II.E.3.1 of NUREG-0737 and Item 2.1.1 of NUREG-0578.
SER Supplement 14.
148 149 Q)
Does the NRC consider the operation of the pressurizer heaters at Diablo 150 Canyon to be important to safety?
A) Yes. The pressurizer heaters are considered " components important to safety" with 151 152 respect to their pressure-control function.
Q) Should the pressurizer heaters then be required to be safety grade?
153 A) No. This pressure control function (of the pressurizer heaters) does not 154 mean it is necessary to meet safety grade criteria for the reasons
-155 sumarized below:
156 (1) The term "important to safety" applies generally to the broad class 157 of structures, systems, and components addressed in the General Design 158 Criteria.
159 (2) " Safety-grade" structures, systems and components are a sub-class of 160 all those "important to safety.'
161 a
i 162 (3)
All structures, systems and components encompassed by the tenn "important 163 to safety" (including the " safety-grade" sub-class) are necessary to 164 meet the broad safety goal articulated in Appendix A to 10 CFR Part 50
{
l l
165 of the regulations (i.e., provide reasonable assurance that a facility j
166 can be operated without undue risk to the health and safety of the 167 public).
168 (4)
Only " safety-grade" structures, systems and compnents are required 169 for the critical ac<:ident prevention, safe shutdown, and accident 170 consequences mitigation safety functions identified in Section III.C l
171 of Appendix A to 10 CFR Part 100.
172 173 Q) What are the critical safety functions which must be provided by safety 174 grade systems?
175 A) The critical safety functions identified in 10 CFR 100 which must be pro-vided with safety grade equipment are as follows:
177 (1) The integrity of the reactor coolant pressure boundary.
178 (2) The capability to shut down the reactor and maintain it in a safe 179 shutdown condition.
180 (3) The capability to prevent or mitigate the consequences of accidents 181 which could result in potential offsite exposure comparable to the 182 guideline exposures of this part.
183 184 Q)Then does the NRC require that the pressurizer heaters he safety grade?
185 A) No. Operation of the pressurizer heaters and associated controls are not 186 required to provide any of these critical safety functions and the NRC 187 therefore does not require them to be safety grade.
_g.
188 189 NRC STAFF TESTIMONY OF W. JENSEN RELATIVE TO 190 PRIMARY SYSTEM RELIEF AND BLOCK VALVES 191 (CONTENTION 12) 192 193 'Q) Please state the purpose of this portion of your testimony.
'"194 -A) The purpose of this testimony is to respond to Contention 12 (as 195 admitted by the Board in the September 30, 1981 Memorandum and 196 Order) which reads:
197 Proper operation of power operator relief valves, associated block valves and the instruments and controls for these valves is essential to miti-198 gate the consequences of accidents. In addition, their failure can cause or aggravate a LOCA. Therefore, these valves must be classified 199 as components important to safety and required to meet all safety-grade design criteria.
200 201 202 203 204 205 206 207 208 209 210 211 212 213 214
215 Q)
What are the functions of the PORVs and Block Wlves?
216 A) The function of the PORVs at Diablo Canyon is to open at the set pressure i
217 of 2350 psig and relieve pressurizer steam so as to preclude the necessity 218 of the safety valves from being opened for mild transients. The set 219 pressure for the safety valves is 2485 psig. The PORVs may also be 220 manually opened by the operator at any pressure below their setpoint to 221 provide a backup means of pressure control in accordance with the operating 222 procedures. Manual opening of the PORVs would cause them to relieve steam 4
- 223 at the pressure of the reactor system.
Reliance is placed on the Reactor 224 Protection Systems and the Engineered Safety Features to mitigate design 225 basis events rather than the PORV. The function of the block valve is to 226 provide the operator a means to isolate a leaking or failed open PORV.
227 (FSAR Chapter 5) 228 The PORVs and associated block valves are not required to provide low 4
229 temperature protection at Diablo Canyon for the first fuel cycle (see 230 Diablo Canyon SER Supplement 6 pages 5-3 to 5-7).
231 232 Q)
Is proper operation of the PORVs or block valves essential to mitigate 233 the consequences of accidents?
234 A) No, proper operation of the PORV, associated block valve, and instruments 235 and controls is not required to mitigate the consequences of any design 236 basis accident. Analyses of design basis accidents are contained in 237 Chapter 15 of the FSAR. As discussed in Chapter 15 of the FSAR reliance 238 is instead on the engineered safety features to mitigate the consequences 239 of anticipated events. The engineered safety features include the Emergency 240 Core Cooling System (ECCS), primary and secondary system safety valves, the 241 Auxiliary Feedwater System and the Residual Heat Removal System (RHR). all 242 of which are safety grade.
10 243 A failure of the PORV or associated instruments and controls which results 244 in inability to isolate the flow path through the valve causes the equivalent 245 of a small-break loss-of-coolant accident. The accident would be terminated 246 by closure of the block valve.which is an immediate action to be taken by 247 the operator in the event of a small-break LOCA. Even if the block valve 248 were not isolated, the capability of the Erergency Core Cooling System is 249 sufficient to permit safe shutdown of the reactor with no core uncovery 250 or core damage. This is demonstrated by analyses contained in Chapter 15 251 of the Diablo Canyon FSAR and in Section 3.3 of Volume III of WCAP-9600.
252 Q) What offsite doses would result from a stuck open PORV that was not isolated by the operator?
A) Since, a stuck open PORV which is not isolated will not result in damage to the fuel element cladding, the fission products contained in the fuel elements would not escape from the core. The only releases to the public 257 would be from radioactive materials already contained in the primary
< 258 coolant. This material would include activated corrosion products contained 259 in the primary coolant and fission products which might have leaked into 260 the coolant during operation.
261 262 The offsite doses to the public for a stuck open unisolated PORV which did 263 not cause fuel failure would be much less than the guidelines of 10 C.F.R.100.
264 265 Q) Are you aware of any study of the probability of PORV failure 266 caus'ng a LOCA?
267 268 A) The probability of a stuck open PORY is evaluated in WCAP-9804 "Probabilistic 269 Analysis and Operational Data in Response to NtREG-0737 Item II.K.3.2 for 270 Westinghouse NSSS Plants," February 1981. This report concludes that the
_ 11 271 probability of a stuck open PORV produced LOCA is 2.1 x 10-6 per reactor l
272 year for plants similar to Diablo Canyon. The NRC review of this report 273 is scheduled to be completed in April 1982. According to WASH-1400, the 274 median probability of a small break LOCA with a break diameter between 275 0.5 in, and 2.0 in. is 10-3 per reactor-year with a variation ranging from 276 10-2 to 10-4 per reactor-year. WCAP-9804 concludes that the probability 277 of a stuck open PORV is a very small contributor to the overall probability 278 of a small break LOCA.
279 Q) What additional requirements will be imposed on Diablo Canyon if the NRC 280 does not agree with the conclusions of WCAP-9804?
281 A)
If the NRC concludes that the probability of a small break LOCA caused by 282 a stuck open PORV is significant contributor to small break LOCA occurrence i
283 contrary to the conclusions of WCAP-9804, the applicant will be required to 284 install an automatic block valve closure system. The requirements for such 285 a system are discussed in Item II.K.3.1 of NUREG-0737.
I
(
l286G)
What would be the effect of a stuck open PORY in conjunction with a LOCA produced 287 from some other cause?
288 A) A stuck open PORV in conjunction with a LOCA would be unlikely since the LOCA 289 would act to depressurize the plant and therefore would not challenge the PORVs.
290 In the event that the PORV was to open inadvertently following a small break in 291 the primary system piping, the effect on the raactor system would be equivalent 292 to increasing the break size. The effect of an increase in break size would 293 fall within the spectrum of small-break sizes already analyzed for Diablo 294 Ca nyon. The small break spectrum is described in Chapter 15 of the FSAR.
295 296 Y
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297-Analyses of open PORVs in conjunction with a small break loss of coclant 298 accident for which all feedwater was lost are contained in Section 4.2 301 of Volume III of WCAP-9600. These analyses demonstrate that opening PORVs 302 improved core cooling.
303 The inadvertent opening of a PORV following a large break in the primary 304 system piping would not significantly affect the consequences of a large 305 break LOCA since the size of the flow path provided by an open PORV would 306 not be significant in comparison with that' of the break.
307 a) The PORVs and block valves are utilized in a number of plant emergency 308 procedures. Doesn't this mean that they have to be safety grade?
309 A) No. Although operation of the PORVs and block valves would be expected to be 310 normally available in a number of transients and accidents, these components 311 are not required to protect the reactor. As discussed in Chapter 15 of the FSAR, 312 reliance is placed on the Reactor Protection System and the Engineered Safety 313 Features, which are designed to safety grade criteria.
It must be recognized 314 that emergent.y procedures are written to present guidance on all options 315 avcilable to the operator for coping with a plant transient or accident. To 316 do otherwise would be an unwise limitation of design and operator capability 317 during such events.
FSAR Chapter 15 analyses in which conservative assumptions 318 on the unavailability of non-safety grade components are typically imposed should 319 not be confused with an actual transient or accident during which the operator 320 would be expected to make maximum use of ALL available systems, whether or 321 not they are safety grade.
322 Q) Does the NRC consider operation of the PORVs and block valves to be important 323 to safety?
324 A) Yes. The PORVs and block valves are considered " components important to safety" 325 with respect to their pressure control function.
_13 _
326 Q) Then should the FORVs and block valves be required to meet safety grade criteria 327 with respect to their proper operation?
328 A) No. The PORVs and block valves are not required to meet safety grade criteria 329 for the reasons summarized below:
330 (1) The term "important to safety" applies generally to the broad class 331 of structures, systems, and components addressed in the General Design 332 Criter'ia.
333 (2) " Safety-grade" structures, systems and components are a sub-class of 334 all those "important to safety."
335 (3) All structures, systems and components encompassed by the term "important 336 to safety" (including the " safety-grade" sub-class) are necessary to 337 meet the broad safety goal articulated in Appendix A to 10 C. F. R. Part 50 338 of the regulations (i.e., provide reasonable assurance that a facility 339 can be operated without undue risk to the health and safety of the public).
340 (4) Only " safety-grade" structures, systems and components are required for the 341 critical accident prevention, safe shutdown, and accident consequences b42 mitigation safety function identified in Section III.C of Appendix A to 343 10 C.
F.' R. Part 100.
344 345 Q) What are the critical safety functions which must be provided by safety grade 346 equipment?
347 A) The Critical Safety functions identified in 10 CFR 100 which must be provided 348 with safety grade equipment are as follows:
349 (1) The integrity of the reactor coolant pressure boundary.
350 (2) The capability to shut down the reactor and maintain it in a safe 351 shutdown condition, or 352 (3) The capability to prevent or mitigate the consequences of accidents which
_ l4 353 could result in potential offsite exposure comparable to the 354 guideline exposures cf this part.
355 Q) Then should tne PORVs, associated block valves be required to be safety grade?
356 A) No. Proper operation of the power operated relief valves and associated block 357 valves is not required to provide any of these critical safety functions and the 358 NRC therefore does not require this equipment to be safety grade.
359 Q) Then why has the NRC required that the PORVs and block valves be supplied with 360 emergency pcwer?
361 A) Emergency power has been provided to two of the three PORVs and to the three 362 block valves to reduce the number of challenges to safety valves and ECCS 363 during operation in accordance with Item II.G.1 of NUREG-0737 ar.d Item 2.1.1 364 of NUREG-0578 (see Diablo Canyon SER Supplement 10).
365 366 367 368 369 370 371 372 373 374 1
375
)
1 1
, ALTON L. JENSEN, JR.
W 376 377 PROFESSIONAL QUALIFICATIONS 378 379 380 I am a Senior Nuclear Engineer in the Reactor Systems Branch of the Nuclear 381 Regulatory Commission.
In this po'sition I am responsible for the technical 382 analysis and evaluation of the public health and safety aspects of reactor 383 systems.
384 385 From June 1979 to December 1979, I was assigned to the Bulletins and Orders 386 Task Force of the Nuclear Regulatory Commission.
I participated in the 387 preparation of NUREG-0565, " Generic Evaluation of Small Break Loss-of-Coolant 388 Accident Eehavior in Babcock & Wilcox Designed 177-FA Operating Plants."
389 390 From 1972 to 1976, I was assigned to the Containment Systems Branch of the 391 NRC/AEC, and from 1976 to 1979, I was assigned to the Analysis Branch of the 392 NRC.
In these positions I was responsible for the development,and evaluation 393 of computer prog ams and techniques to calculate the reactor system and 394 containment system response to postulated loss-of-coolant accidents.
395 396 From 1967 to 1972, I was employed by the Babcock and Wilcox Company at Lynchburg, 397 j Virginia. There I was lead engineer for the development of loss-of-coolant 398 computer programs and the qualification of these programs by comparison with 399 experimental data.
400 M
- 401 From 1963 to 1967, I was employed by the Atomic Energy Commission in the 402 Division of Reactor Licensing.
I assisted in the safety reviews of large 403 power reactors, and I led the reviews of several small research reactors.
404 I received an M.S. degree in Nuclear Engineering at the Catholic University of 405 America in 1968 and a B.S. degree in Nuclear Engineering at liississippi State 405 407 University in 1963.
408 I am a graduate of the Oak Ridge School for Reactor Technology, 409 1963-1964.
410 411 I am a member of the anerican Nuclear Society.
412 I am the author of three scientific papers dealing with the response of B&W 413 414 reactors to Loss-of-Coolant Accidents and have authored one scientific paper 415 dealing with containment analysis.
416 417 418 419 420 421 422 423 424 425 2
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