ML20101C773
| ML20101C773 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde, 05000000, 05000470 |
| Issue date: | 12/18/1984 |
| From: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| ANPP-31502-WFQ, NUDOCS 8412210191 | |
| Download: ML20101C773 (16) | |
Text
{{#Wiki_filter:' r n I r Arizona Public Servicc Company. Director of Nuclear Reactor Regulation December 18, 1984 Mr. George W. Knighton, Chief ANPP-31502 WFQ/TFQ Licensing Branch No. 3 Division of Licensing U.S. Nuclear Regulatory Commission . Washington, D.C. 20555
Subject:
Palo Verde Nuclear Generating Station (PVNGS) g-Units 1, 2, and 3 Docket Nos. STN 50-528/529/530 Post-FDA Proposed CESSAR Changes File: 84-056-026; G.I.01.10 Refer mces: - (A) Letter from A. E. Scherer, CE, to D. G. Eisenhut, NRC, dated October 16, 1984 (LD-84-060);
Subject:
CESSAR Startup Testing. (B) Letter from A. E. Scherer, CE, to D. G. Eisenhut, NRC, dated December 5, 1984 (LD-84-070);
Subject:
CESSAR Amendment 10.
Dear Mr. Knighton:
As discussed with the NRC staff, it is APS' understanding that post-Final Design Approval (FDA) proposed CESSAR changes will not be reviewed and approved by the NRC staff, unless a change meets the criteria of the NRC's Standardization Policy. That is, a proposed change to CESSAR will be reviewed and approved if there is an increase in the safety and health of the public. With this staff position in mind, APS has reviewed all of the proposed CESSAR changes submitted to the staff, including CESSAR Amendment 9, issued on February 27, 1984 With regard to those changes which will not be reviewed on the CESSAR docket due to the Standardization Policy, APS hereby requests that the attached, previously submitted, proposed CESSAR changes be reviewed as proposed PVNGS FSAR changes. This request is justified because approval of the attached proposed FSAR changes will allow 1) incorporation of improvements into the initial test program, 2) use of the most recent CE secondary water chemistry guidelines, and 3) removal of unwarranted restrictions on the transfer of hydrogens. Your review and approval will also accept 1) the justification for using primary system blowdown and 2) separation analyses with respect to Regulatory Guide 1.75 for Plant Protection System Cabinets. h QO N -)
r ll: . Mr. G. W. Knighten: Po:t-FDA Proposed.CESSAR Ch:ngra ANPP-31502 - Page 2 + - Please contact Mr. William Quinn of my staff if you have any questions ~on this ~ matter. Very.truly ours 1 c'. CLUL-qil i E. E. Van Brunt,'Jr. ~ APS Vice-President: Nuclear Production - I. ANPP Project Director. - Attachments-EEVB/TFQ/mb cc: E. A..Licitra (w/a) A. C. Gehr (w/a) R. P. Zimmerman (w/a) i I l s f P 2 h + t .. -m.
'g ~ 7g. - ~. s p' - l w I' w g,9 --'. i-u m 1 y ~,Q. I l .c December 118, 1984 - ] -: ANPP-31502 A4 L' STATE OF ARIZONA ) ) ss. _ COUNTY OF MARICOPA) ~ I, Edwin E., Van ~ Brunt, Jr., represent that I an ' Vice President, Nuclear. Production of Arizona Public Service. Company,
- that the: foregoing document has been. signed by 'me on behalf = 'of Arizona Public Service Company with full authority to do so, that 'I have ' read 'such - document :and know Its contents,.and that' to the best of my knowledge and belief, the ' statements made therein are true.
5 g (g ( N( <\\ tJ L : + Edwin E.: Van Brunt, Jr. - ) ' Sworn t6 b fore me this' _ day of M o,.1984. .s-i b ffg ~ b -s, e ~ JJ l
- Notary Pubif6 n1 My.; Commission Expires
s 'hNjssica Expires April 6,1967 ,s f f , p x e y e + E' S t 9 rJ '\\ l.' f ; f 9 t q I g k 3 Y yi m.
r= EXCERPTS FROM CESSAR AMENDMENT 9 ISSUED FEBRUARY 27, 1984 l l l
e-5.4.13.4 Tests and Inspections { The valves are inspected during fabrication in accordance with ASME III Code requirements. 5.4.13.4.1 ' Pressurizer Safety Valves The inlet and outlet portions of the valves are hydrostatically tested with water at the appropriate pressures required by the applicable section of the ASME Code. Set pressure and seat leakage tests are performed with steam usin'g a pro-rated spring. Final set pressure tests are performed 9 ~ .with the final springs using either high pressure air-or low pressure steam with an hydraulic assist device. Final seat leakage tests are performed prior to shipment with the final springs using either hot air or hot nitrogen. Valve adjustment shall be made to a valve ring setting combination selected to providestablevgeoperationonthebasisoftheEPRISafetyValveTest Program results 5.4.13.4.2 Main Steam Safety Valves The inlet. portion of the valve is hydrostatically tested with water in accordance with the ASME Code. Set pressure and set leakage tests are performed using steam. Adjustment is made to provide a valve blowdown 9 meeting the requirement specified in Table 5.4.13-2.. (1) CEN-227 " Summary Report on the Operability of Pressurizer Safety Relief Valve in C-E Designed Plants", December 1982. 9 5.d-42 Amendment No. 9 February 27, 1984
ic ~ '; g ' TABLE 5.4.13-1 PRESSURIZER SAFETY VALVE PARAMETERS Property-Parameter Design pressure, Ib/in~.2a 2500 . Oesign._ temperature, *F 700 Fluid Saturated Steam, 4400 ppm, boron, 9 pH = 4.5 to 10.6 Set pressure, Ib/in.2a-2500 + 1% Min. capacity, lb/h at accumulation l9 pressure, each .'460,000 1 Type' Spring loaded safety-balanced bellows. Enclosed bonnet. Orifice area, in.2 4.34 ) Accumulation, % 3 k. Backpressure Max. ' buildup / max superimposed, Ib/in.29 700/340 9 Minimur.1 blowdown pressure, psia 2040 Typical materials Body ASME SA 182 GR. F316 ' Disc ASTM A637, GR. 688 Nozzle ASME SA 182, Gr. 347 -l9 i Amendment No. 9 i february 27. 1984 i L
. h ': .) TABLE 5.4.13-2 { , MAIN STEAM SAFETY VALVE PARAMETERS s Property Parameter Design' pressure,-lb/in. 9 -1375 Design temperature. *F 575 Fluid Saturated Steam Set pressure. lb/in.' 29 ~ 1255, 1290, 1315 l Min. capacity, Ib/h at accumulation 9 Pressure 6 19 x 10 Total (20 Valves) Type Spring loaded Orifice area,'in.2 16 Accumulatton, % 3 Backpressure Max. buildup / max superimposed..lb/in.29 125/0 lS Approx. dry weight, lbs. 1545 Minimum blowdown pressure, psig 1175 9 , Typical materials Body ASME SA 105 Otsc ASTM A565 GR. 616 Nozzle ASME SA 182. GR.-F316 i Amendment No. 9 february 27. 1084
{ l j i APPENDIX SA i l OVERPRESSURE PROTECTION FOR C0f180STION ENGINEERING SYSTEM 80 - PRESSURIZED WATER REACTORS TABLE OF CONTENTS SECTION TITLE PAGE NO. 1.0 INTROD_ULTl0N SA-1 2.0 ANAL _YSIX 5A-1 2.1 METHOD SA-1 2.2 ASSUMPTIONS SA-1 2.2.1 SECCNDARY SAFETY VALVE SIZING 5A-2 2.2.2 PRIMARY SAFETY VALVE SIZING SA-3 2.2.3 ACCEPTABILITY OF SAFETY VALVE BLOWDOWN SA-4 9 3.0 9,0ft,L(1510N1 SA-5 s i Amendmnt No. 9 february 27. 1984
? O tempera tur -3.5 x 10'g coefficient can vary between zero and for various phases of core life. Therefore, a coefficient of zero is chosen to maximize the power / pressure ( transient. Doppler coefficient of .8 x 10-5 AK/K/F is used in the c. loss-of-load analysis. Actual 02 grating coefficients can be exegeted to range from -1.4 x 10 at zero power to -1. x 10 aK/K/F at full power. By choosing a relatively small Doppler coefficient, the reduction in reactivity with increasing fuel temperature is minimized, thereby maximizing the rate of power rise, d. No credit is taken for letdown, charging, pressurizer spray, turbine bypass, or feedwater addition after turbine trip in the loss-of-load analysis. Letdown and pressurizer spray both act to reduce primary pressure. By not taking credit for these systems, the rate of pressurization is increased. By not taking credit for the addition of feedwater, the steam l9 generator secondary inventory will be depleted at a faster rate. This in turn reduces the capability of the steam generator to remove heat from the primary loop, thereby maximizing the rate of primary pressurization. The analysis reflects consideration of plant instrumentation e. error and safety valve setpoint errors. For example, all safety valves are assumed to open at their maximum popping This extends the period of tine before energy can pressure. I be removed from the system. The reactor trip setpoint errors are always assumed to act in such a manner that they delay reactor trip, again resulting in maximum pressurization. f. Pressurizer pressure at the onset of the incident is 2200 psi. By using the lower limit of the normal plant operating pressure, the time required to trip the plant on high pressure is increased. 2.2.1 SECONDARY SAFETY VALVE SIZING The discharge piping serving the secondary safety valves is designed to accommodate rated relief capacity without imposing unacceptable backpressure on the safety valves. The secondary safety valves are conservatively sized to pass excess steam flow. This limits steam generator pressure to less than 110% of steam generator design pressure during worst case transients. A plant's secondary safety valves consist of three banks of valves with staggered set pressures. The valves are spring-loaded type safety valves procured in accordance with ASME Boiler and Pressure Vessel Code, Section !!I. 9 SA-2 Amendment No. 9 februarv 27.19M
e Figure SA-2 depicts the steam generator pressure transient for this worst case loss-of-load incident. As can be seen in Figure SA-2, the steam generator pressure remains below 110 percent of design pressure during the incident. 2.2.2 PRIMARY SAFETY VALVE SIZING The reactor drain tank, inlet and discharge piping are sized to preclude unacceptable pressure drops and backpressure which would adversely affect valve operation. Primary safety valve backpressure is limited by the design pressure of the valve bellows. These bellows prevent any accumulated backpressure from being imposed on the valve spring, thus allowing valve operation at its design setpoint rather than at its setpoint plus backpressure. The design basis incident for sizing the primary safety valves is a loss of turbine-generator load in which the reactor is not immediately tripped. No credit is taken for any pressure-reducing devices except the primary and secondary safety valves. In reality, the incident would be terminated by a number of reactor trips. These include: g a. Steam generator low level trip; b. High pressurizer pressure trip; c. Manual trip. If the high primary pressure trip were to become inoperative, other reactor trips would proceed to shut the reactor down as their setpoints are exceeded. A series of loss-of-load studies are run with various sizes of primary safety valves. As can be seen in Figure SA-1, af ter the safety valve capacity increases to a certain size, additional increase in capacity has negligible effect in reducing the maximum system pressure experiences during the loss-of-load transient. C-E's primary safety valves are chosen so as to minimize the maximum pressure experienced during the loss-of-load transient. The minimum specified safety valve capacity is identified on Figure SA-l. Figures SA-2, SA-3 and SA-4 present curves of steam generator pressure, maximum Reactor Coolant System pressure and core power versus time for the worst case loss of turbine-generator load. As can be seen on Figures 5A-2 and SA-3, the maximum steam generator pressure and reactor coolant loop pressures remain below 110% of design during this worst case transient. The first, second, and third banks of secondary safety valves open at approximately 3.7, 5, and 6.2 seconds, respectively. The secondary safety valves remove energy from the Reactor Coolant System and thus mitigate the pressure surge. The primary safety valves are conservatively assumed to open at 1 percent above the normal Reactor Coolant System design pressure 5.7 seconds af ter the initiatien of the upwt condition. SA-3 Amendment No. 9 February 27, 1984
e In the event that a complete loss of load occurs without a simultaneous reactor trip, the protection provided by the high pressurizer pressure trip, primary / safety valves and secondary safety valves is sufficient to assure that the t integrity of the RCS and main steam system is maintained and that the minimum Df48 ratio is not less than 1.19. 2.2.3 ACCEPTABILITY OF SAFETY VALVE BLOWD0Wft 2.2.3.1
Background
Full scale, full pressure prototgcal testing of pressurixer safety valves was perfonned by EPRI in 1981. The blowdown settings required to insure stable valve operation during the blowdown from the set pressure were above the 5% setting specified in the ASME Code. In order to insure that the extended blowdown would not adversely affect overpressure protection or plant operation, analyses were performed to evaluate the f4SSS response. The analyses described below demonstrate that a blowdown setting, including associated uncertainties, of 18.5% is acceptable. 2.2.3.2 Results of Evaluation An extended blowdown of the safety valves could result in swelling of the pressurizer liquid level due to flashing and possible liquid carryover through the safety valves. Since the safety valve design specification specifies dry saturated steam flow conditions, it is desirable to show that these conditions are maintained during the extended blowdown. It is also desirable to verify that the RCS remains in a subcooled condition in order that the steam bubble formation in the RCS is precluded. 9( A computer analysis was performed of the Loss-of-load event with delayed reactor trip, similar to that used in safety valve sizing, except that a conservative 20% safety valve blowdown and initial conditions biased to maximize pressurizer liquid level were assumed. The purpose of this analysis was to determine the pressurizer liquid level response and the RCS subcooling under these conservative conditions. For additional conservatism, an additive adjustment was made to the computer-calculated pressurizer levels on the basis of a very conservative pressurizer model. This model assumed that the initial saturated pressurizer liquid did not mix with the cooler insurge liquid, that the initial liquid remained in equilibrium with the pressurixer steam space, and that the steam which flashed during blowdown remained dispersed.in the water level vs time curve showed a maximun of 98%(2)The adjusged pressurizer liquid phase and caused the liquid level to swell (1730 ft ), below the safety valve nozzle elevation of 107%, eo that dry saturated steam flow to the safety valves is assured throughout the blowdown. The computer analysis also showed that adequate subcooling was nuintained in the RCS during the blowdown, so that steam bubble formation is precluded. (1) CEft-227, " Summary Report on the Operability or Pressurizer Safety Relief Valves in C-E Designed Plants" Dacember 1982. (2) Water level expressed as the percentage of the distance from the lower Level nozzle to the upper level nozzle. SA-4 Amendment flo 9 February 27, 1984
] In addition, the System 80 safety analyses of pressurization events were re-evaluated to determine the impact of assuming an 18.5% blowdown below noninal set pressure (to 2040 psia) for the pressurizer safety valves in lieu of the 5% specified by the ASME Code. The evaluation indicated that, for the FWLB event analysis, which produces the greatest increase in pressurizer level, the increased blowdown would not result in the pressurizer liquid level reaching the safety valve nozzle elevation and thus normal safety valve operation would be assured. maintained during the blowdown. Further, subcooling in the RCS was 9 In summary, analyses show that adequate plant overpressure protection and RCS subcooling are ensured during a blowdown of 18.5% below nominal pressurizer safety valve set pressure. 3.0 CONCLUSIpRS, C-E's System 80 pressurized water reactor steam generators, and Reactor Coolant System are protected from overpres,surization in accordance with the guidelines set forth in the ASME Boiler and Pressure Vessel Code, Section III. Peak Reactor Coolant System and Secondary System pressures are limited to 110% of design pressures during worst case loss of turbine-generator loac. Overpressure protection is afforded by primary safety valves, secondary safety valves, and the Reactor Protection System. 5A-5 Amendment No. 9 February 27, 1984 m_
I e ~ .o. 14.2.5 REVIEW, EVALUATION, AND APPROVAL OF TEST RESULTS ( The development of administrative procedures for review, evaluation and approval ~of test results is the responsibility ofsthe Applicant. Advice and consultation will be provided by Combustion Engineering as appropriate. Test results shall be recorded as permanent plant records. 14.2.6 TEST RECORDS .a An official copy of each completed test procedure, including all required supplemental data, exceptions, conclusions and approval signatures shall be maintained in accordance with the Applicant's administrative-controls. 14.2.7 CONFORMANCE OF INITIAL. TEST PROGRAMS WITH REGULATORY G INDUSTRY STANDARDS The intent of the following Regulatory Guides will be followed with the noted differences. 14.2.7.1 Reg. Guide 1.68 Initial Test Programs for Water-Cooled Reactor Power Plants (Revision 0,11/73). The following exceptions and/or clarifications address only significant differences between the System 80 test program and the applicable regulatory 6 position. Minor terminology differences, testing not applicable to the plant design, and testing that'is part of required surveillance tests will not be addressed. I Guide 1.68 (Revision 0, 11/73). Reference is made to the applicable portion of Regula J 14.2.7.1.1 Reference Appendir. A, Section 8.1.c. l9 This section suggests that rod drop times be measured for all control element assemblies (CEAs) at hot and cold full-flow and no-flow conditions. The CESSAR CEA drop-time testin the regulatory guide; however, g is consistent with the recommendations of tests which do not provide meaningful data will be deleted. As outlined in test summary 14.2.12.3.4, the CEA drop 6 time testing will consist of: a.) One drop of each CEA at cold, maximum permissible flow conditions (2 or 3 reactor coolant pumps) and at hot, full-flow conditions.* b.) Those CEAs falling outside the two-sigma limit for similar CEAs will be dropped three additional times, c.) Hot no flow scram insertion rod drops will not be performed for System 80 reactors. C-E has demonstrated that rod drop times under full-flow conditions are more limiting than the drop times under conditions of 9 no-flow. Only "first-of-a-kind" plants are allowed to be critical at reduced temperature. Thus, if plant operating procedures preclude pulling CEA's prior to achieving HOTZERO POWER conditions, the cold drops will be 6 expected. 14,2-5 Amendment No. 9 Februarv 77. MM
1 ? 14.2.12.5 Power Ascension Tests 14.2.12.5.1 Natural Circulation Test
- 1.0 OBJECTIVE 1.1 To evaluate natural circulation flow conditions.
- 2. 0' PREREQUISITES 2.1 The reactor is at > 80% power. Previous power history is_such that the loop differential temperature (T T ) under natural circulation 9
isnotexpectedtodropbelow10FdUringtheperformanceofthe C test. 3.0 TEST METHOD 3.1 The reactor coolant pumps are tripped. 3.2 The plant is tripped. 3.3 The natural circulation power to flow ratio is determined. 4.0 DATA REQUIRED 4.1 Conditions of the measurement. 4.1.1 Power. 4.1.2 Previous power history. 4.1.3 RCS temperatures. 4.2 Time dependent information. 4.2.1 RCS temperatures. 4.2.2 Steam generator levels and pressures. 4.2.3 Pressurizer pressure and level. 5.0 ACCEPTANCE CRITERIA ~5.1 The natural circulation power to flow ratio is less than 1.0. 14.2.12.5.2 Variable Tavg (Isothermal Temperature Coefficient and Power Coefficient) Test 1.0 OBJECTIVE 1.1 To measure the Isothermal Temperature Coefficient and Power Coefficient of reactivity at selected power levels. Mhis test wilThe perf"ordied for the "first-of-a-kind" unit only. 14.2-05 Amendment No. 9 February 27. 1984 i
3-w-- Ji O' p i g e t L2.2 ~The RRS, FWCS,' SBCS, RPCS, and the pressurizer level and pressure { 1p T control, systems are in automatic operation.
- f(i
?. y ') , 3.0. TEST METHOD- .u ' Load -increases and 2ecreases -(steps and ramps) in acco'rdan,ce with
- 3.1 T
the C-E Fuel Pre-cohditioning Guidelines will be perfo'rmed at power 6-
- i levels in the 90 to 100% range and with swings in the 50 to 25 to 50% power level.
93 ,4.0-DATA' REQUIRED 9 ' 4 al. . Time dependent data. 4 d'.1 Pressurizer level and pressure. s '4.1.2 RCS~ temperatures. R 4.1.3 CEA position. '4.1.4 ' Power level and demand. 4.1. 5 - Steam ge'nerator levels and-pressures. '4.1.6-Feedwater and steam flow. o 4.1.7 FeedwiteE temperature. 9 I~ 5.0 ACClEPTANCECRkTERIA 5.1 The st'ep and ramp transients demonstrate that the plant performs load changes allowed by C-E's Fuel Pre-conditioning Guidelines and data has been taken that will demonstrate the plant's ability to 6_ 4 meet unit load swing design transients. P. 14.2.12.5.4 ' Control Systems Checkout Test 1.0 OBJECTIVE ..s 1.1. To ' demonstrate that the automatic control systems operate satis-g factorily during steady-state and transient conditions. g -2.0 PREREQUISITES . 2,1 The reactor is operating at the desired conditions. 2.2 The RRS, FHCS,- SBCS, RPCS, and the pressurizer level and pressure controls are in automatic operation. 4 -3.0 TEST METHOD-3.1 The performance of. the control systems during normal operations, transients and trips will be monitored to demonstrate that the . systems are operating satisfactorily. 14.2-87 Amendment No. 9 February 27, 1984 L
4: g ;4. '; V 'a.' '2.0E PREREQUISITES -2.1-The' SBCS, FHCS,- RRS, RPCS, and pressurizer pressure a~nd[ level ~ '7 controls are operable in either manual or automatic modes. + 23.0 TEST > METHOD ~ = 3.1 Performance of the feedwater systems will be_ monitored during. normal. operation, transients, and trips..Specifically, the l downcomer' to economizer transfer will be monitored for noise or 9 ~~ ibration due to Water Hammer. v . 4.~ 0 - DATA REQUIRED 4kl.l' P,eactor power .4.1.2 RCS temperatures _;4.1.3 . Pressurizer pressure 4.1.4 - Steam generator levels and pressures 4.1.5 Steam and feedwater flows-4.1.6-Feedwater_ temperature-4.1.7 CEA Position '5.0 ACCEPTANCE CRITERIA ( 5.1 The main and emergency feedwater systems perform as designated P by..the system description. ] 9- ^~ 114.2.12.5'.18 ~CPC Verification 1.0. OBJECTIVE To verify DNBR and Local Power Density (LPD) calculations of'the ~ CPCs. -2.0 -PREREQUISITES '2.1 . The reactor is at the desired power. level _ and CEA configuration.- .with_equilbrium Xe. ~
- 2.2 The CPCs are operational.
2.3: ~ The incore detector-system is operational. l 3.0 71-
- TEST METil0D -
3.1: Specified values _are recorded from the CPCs. '3.2 ~ The' values for LPD and DNSR obtained from the CPCs are compared with' the values calculated for the same conditions using the ~ CPC FORTRAN Sireulator. 14.2-100 Amendment No. 9 February 27, 1984
.a t o C-E Power Systems Tel 203/6881911 Combustion Engineering. Inc. Telex: 99297 1000 Prospect Hill Road Windsor. Connecticut 06095 M SYSTEMS POWER STN 50-470F October 16, 1984 LD-84-060 Mr. Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
CESSAR Startup Testing
Dear Mr. Eisenhut:
During preparation for startup of the first System 80'" plant, C-E has noted minor modifications which could be made to CESSAR Chapter 14 to facilitate an improved testing procedure. These changes affect only tests performed after fuel loading and do not in any way affect CESSAR's compliance with NRC requirements. These changes, along with a description of each change, are provided in the attachment for NRC review. These changes will be incorporated in a subsequent amendment to CESSAR. If you have any questions or comments, feel free to call me or Mr. T. J. Collier of my staff at (203) 285-5215. Very truly yours, COMBUSTION ENGINEERING, INC. 11 = A. E. T herer Director l Nuclear Licensing AES:las
- Attach, cc: K. Eccleston (USNRC Project Manager) f
- N ic.: q l
i. DESCRIPTION.0F' e LO-84-060 MODIFICATIONS TO Attachment CHAPTER 14'0F CESSAR-F Page'l ~ ~ ' ~ s 14.2.7.1.1: RG 1.68, Appendix A, Section B.1.c (page 14.2-5) Deleting the cold (260*F) partial flow CEA drops.is consistent with experience in previous startups, which shnwed that the hot, full-flow drops were more limiting. Low temperature criticality is allowed only on first-of-a-kind plants, and then only for short periods of time under close supervision. CEA insertability at cold conditions is still demonstrated during post-core hot functional testing, providing assurance that the CEAs can be tripped, if necessa ry. 14.2.10.1: Initial Fuel Loading (page 14.2-7a) The containment evacuation alarm described in the deleted material will not be provided, nor is there any requirement for such a device. Should a situation i exist requiring evacuation, the operator could (in the case of Palo Verde is required to) utilize the site public address system and the plant evacuation alarm. As stated in this section, audible count rate indicators will be provided in containment. 14.2.12.3.1: PCHFT Controlling Document' (page 14.2-69) Item 2.1 reflects the possibility that some of the pre-core tests may be postponed to, or rerun during, the post-core hot functional testing. The other changes reflect how thednstrumentation is to be calibrated. YU) 14.2.12.3.4: Post-CoreWfM Performance (page 14.2-72) The test method is up' dated to reflect the fact that no cold drops are to be performed. Verification of position indication and alarms is not temperature or schedule dependent and can be accomplished at 'any time, as the change reflects. The change to the required data reflects the fact that the RCS conditions are only a concern'for this test during the rod drops. 14.2.12.4.2: CEA Symetry' and Coupling Test (pages 14.2-77a,14.2-78) The CEA coupling test is deleted because of a difference in the System 80 design and previous C-E designs. In previous C-E. designs, CEAs and extension shafts were uncoupled during each refueling outege (CEAs remained in the l core)..In 'the System 80 design, CEAs and extension shafts are not uncoupled (CEAs are withdrawn into the-upper guide structure). . -N
ei o OESCRIPTION-OF f LD-84-060 MODIFICATIONS TO Attachment CHAPTER 14 0F CESSAR-F 'Page 2 ~ '~~ '14.2.12.4'.'4: Shutdown and Regulating CEA' Group Worth Tests (page 14.2-79) Testing on the first-of-a-kind plant has been modified ta perform the net shutdown measurement at low temperature (approximately 320*F) rather than at HZP conditions. This change is advantageous for the following reasons: (1) The measurement provides direct verification 'of the net shutdown worth at relatively cold conditions (shutdown margin following a cooldown). This measurement can be readily done on the first-of-a-kind unit where a low temperature test program is performed. (2) Less RCS boron dilution is required. (3) Potential cooldown event's during testing while in a highly rodded configuration would have less of a consequence. The remaining information, e.g., CEA group worths, obtained with the revised measurement approach is essentially equivalent to that obtained with the original test approach. For follow-on plants, the net shutdown measurement -is to be performed at 565 F, since low temperature testing is not performed. 14.2.12.4.7: Pseudo Oropped and Ejected CEA Worth Test (page 14.2-83) The wording is changed to provide flexibility in the testing methodology. The measurements of CEA worths via dilution (CEA insertion), boration (CEA withdrawal) or CEA compensation provide equivalent information. The-test method is reworded to clarify the conditions at which testing will be performed. 14.2.12.5.3: Unit Load Transient Test (page 14.2-87) i The test method is updated to reflect the conditions under which the test will be conducted, including the limiting factors. 14'.2.12.5.4: Control -Systems Checkout Test (pages 14.2-87,14.2-88) The test method is reworded to clarify the con'ditions at whitch testing will be peyfai med. Feedwater-temperature is added to the list of monitored parameters (4.1.7). The acceptance criteria are reworded to clarify the critaria to be used for evaluating steady state performance and transient responses. 14.2.12.5.6: Turbine T' rip (pages 14.2-89, 14.2-89a ) l L The Turbine Trip Test and the Unit Load Rejection Test lead to essentially-the L same plant response. Rather than perform redundant tests, the turbine trip test will be performed with the Reactor Power Cutback System ~ (RPCS) not in-
- service while the unit load rejection test is performed with the RPCS in-c
(.i
fa ~ OES'CRIPTION!0F e. tn_g4-060' -MODIFICATIONS TO Attachment CHAPTER'14'0F CESSAR-F ~ ~ Paglii'3 ~ .; servi ce. The Data Required section is reworded to specify the parameters to be .e' valuated against -acceptance criteria. Additional key parameters are to be
- monitored to provide supplemental information but are not evaluated against specific-acceptance criteria. The Acceptance Criteria section is reworded to specify the method to be used for evaluating the parameters against which acceptance criteria are applied. Since non-safety parameters monitored during the test.are not1 evaluated against specific acceptance criteria, the second '
sentence is eliminated. '14.2.12.5.7: Unit load Rejection Test (pages 14.2-90, 14.2-90a,-14.2-91) This test will be performed with the RPCS in operation. The sumary is ~ reworded in a manner similar to the Turbine Trip Test. 14.2.12.5.11: Xenon Oscillation Control (PLCEA) Test (pages 14.2-93,--14.2-94)- .The initial. conditions for the test are revised to allow this test to be performed at or above 50". power. The prerequisite.that testing at.the.80% plateau be completed is not required. The ' acceptance criteria'is reworded to - liminate the phrase "throughout core life", since-this requirement cannot' be. e 1-Edemonstrated directly from the test results. The~ test data, in conjunction wi.th design analyses, demonstrates that xenon oscillations are readily controllable throughout 11 fe..
- 14.2.12.5.12: " Ejected" CEA Test (page 14.2-95) 14.2.15.5.13: " Dropped" CEA Test -(page 14.2-96)
The rewording of the acceptance criteria clarifies the procedure to be used for
- evaluating-the test results.- The rewording does not change the intended Tacceptance criteria.
14.2.12.5.14: Steady' State Core Performance Test - (pages -14.2-96,14.2-97) ' The objective-is reworded to coincide with the primary reason for. performing - the test. 0bjective 1.1 is deleted since specific acceptance criteria are not applied for this purpose.. The Test Method and Data Required sections are - reworded to more clearly specify the way the' test will be performed. . Acceptance.Criteriai-5.1 is not required as. the COLSS and CPC systens adequately . monitor DNBR-and LPD limits during power escalation. Acceptance Criteria 5.2 'iscreworded to specify that enre peaking factors are also evaluated against-specific acceptance criteria. 1 'l l _a
T - ed t8 DESCRIPTION OF E' LO-84-060 MODIFICATIONS-TO. Attachment CHAPTER 14'0F CESSAR-F ~ Page 4 14.2.12.5.15: Intercomparison of PPS, CPC and PMS ' Inputs (page '14'.2-99) The. rewording reflects the proper terminology. for.this parameter, i.e., the temperature shadowing factor. This measurement is planned for both first-of-a-kind and follow-on units, so the asterisked footnote is deleted. 14.2.12.5.17: Main and Emergency Feedwater Systems Test (page 14.2-99a ) This page was missing from Amendment 9, February 27, 1984 14.2.12.5.18: CPC Verification (page 14.2-101) Incore detector maps (Section 4.5) are not required for this test so this -requirement is deleted. The Acceptance Criteria section is reworded to clarify the procedure for applying the criteria. 14.2.12.5.19: Steam Bypass Valve Capacity Test (pages 14.2-101,14.2-102) The test description is altered to reflect capacity testing of each ADV and: -SBCS valve individually. Individual valve capacities are required to show that. the valve capacities assumed in Chapter 15 (Safety Analysis) are conservative. F Prerequisites are changed to delete the requirement for automatic SBCS operation, because individual valve modulation (open and close) is not possible ~ in automatic control. Table 14.2-1: Low Power Physics Tests The table is modified to be consistent with the revised test summaries (described above). -Table 14.2-2: Power Ascension Test 'The table is modified to be consistent with the test summaries and to reflect - the planned testing approach. The footnote'cn the coefficient measurements'is added to clarify that the test is performed with CEA movement and must be performed at a power level which allows the required CEA motion based on margin . considerations. Ta bl e ' 14.2-7 : Physics (Steady State) Test Acceptance C iteria The acceptance criteria for net shutdown worth, and dropped and ejected CEA worths, were inadvertently omitted. These are added in this amendment. -Dropped and ejected CEA worths, and power distribution comparisons, are not required on follow-on units. This change nakes Table 14.2-7 consistent with -Tables-14.2-1 and.14.2-2. Other addtions are added for clarification and do not affect the established acceptance criteria.
- e. G f, )
MODIFICATIONS TO CHAPTER 14 0F CESSAR-F i e i l l I I t f l l l L
14.2.5 REVIEW, EVALUATION, AND APPROVAL OF TEST RESULTS The development of administrative procedures for review, evaluation and ~ approval of test resuTts is the responsibility of the Applicant. Advice and consultation will be provided by Combustion Engineering as appropri_a_te. Test results shall be recorded as permanent blant records. 14.2.6 TEST RECORDS An official copy of each completed test procedure, including all required supplemental data, exceptions, conclusions and approval signatures shall be l maintained in accordance with the Applicant's administrative controls. l 14.2.7 CONFORMANCE OF INITIAL TEST PROGRAMS WITH REGULATORY GUIDES AND INDUSTRY STANDARDS The intent of the following Regulatory Guides will be followed with the noted differences. 14.2.7.1 Reg. Guide 1.68 Initial Test Programs for Water-Cooled Reactor Power Plants (Revision 0, 11/73). The following exceptions and/or clarifications address only significant l differences between the System 80 test program and the applicable regulatory position. Minor terminology differences, testing not applicable to the c_, l plant design, and testing that is part of required surveillance tests will not be addressed. Reference is made to the applicable portion of Regulatory Guide 1.68 (Revision 0, 11/73). 14.2.7.1.1 Reference Appendix A, Section B.1.c. This section suggests that rod drop times be measured for all control element assemblies (CEAs) at hot and cold full-flow and no-flow conditions. The CESSAR CEA drop-time _trLting is consistent with the reconsnendations of the regulatory guide; ho' wever,' tests which do not provide meaningful data will be deleted. As outlined in test suninary 14.2.12.3.4, the CEA drop time testing will consist of: a.) One drop of each CEA et ::'f. ' : : P ' " n : 2 : ^...... ' ^ - ' :::::: :::' c.; 7.., '...J at hot, full-flow conditions.X l b.) Those CEAs falling outside the two-sigma limit for similar CEAs will be dropped three additional times. L c.) Hot no flow scram insertion rod drops will not be perfonned for System 80 l reactors. C-E has demonstrated that rod drop times under full-flow I conditions are more limiting than the drop times under conditions of no-flow. t l T) T rAf drop Ge fut d 2fo f /4 km was e RminM/ sin e e d. 01 "is o in " la a 11 we be ri ic a r d fE/'s p ,0if p ig o dur s rec d pu e ea E o it n,te ld ro e efi pr o t hc2', {u//-flo< mi a nsoM N'"'l
- AiN ?'
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- 2*
yl &
- s QS
- Y"*4 clua; Nwfn,w p y Ces U**,f.
A
one temporary channel and one permanent channel will be equipped with audible count rate indicators in two locatinne +==nn m y in the containment and permanent in the main control room. @ c?-*'ia--* c m rtica %i s c - h d t; it e.. u.;.t aid; ---- 1;;;; c :-- & in : ::t--i-t :t i fic: t. -, Jm LL., y ;; ;.,.- m.c.t --t: t: pr;.id: _t: it e i uiroinn _ _ _ M " p a r t 7;t; M., ;., tL '. ' 1;;di c.i; :;:-" ina. __ _ _ _ ___ s Continuous area radiation monitoring will be provided during fuel handling and fuel loading operations. Permanently installed radiation monitors display radiation levels in the main control room and will be monitored by licensed operators. Fuel assemblies, together with inserted components, will be placed in the reactor vessel one at a time according to a previously established and approved sequence which was developed to provide reliable core monitoring with minimum possibility of core mechanical damage. The initial fuel loading procedure will include detailed instructions which will prescribe successive movements of each fuel assembly from its initial position in the storage racks to its final position in the core. The procedures will establish a system and a requirement for verification of each fuel assembly movement prior to proceeding with the next assembly. Multiple checks will be made for fuel assembly and inserted component serial numbers at successive transfer points to guard against possible inadvertent exchanges or substitutio l At least two fuel assemblies containing neutron sources will be placed into the core at appropriate specified points in the initial fuel loading procedure to ensure a neutron population large enough for adequate monitoring of the core. As each fuel assembly is loaded, at least two separcte inverse count rate plots will be maintained to ensure that the extrapolated inverse count rate ratio behaves as would be expected. In addition, nuclear instrumentation will be monitored to ensure that the "just loaded" fuel assembly does not excessively increase the count rate. The results of each loading step will be reviewed and evaluated before the next prescribed step is started. 14.2.10.1.1 Safe Loading Criteria Criteria for the safe loading of fuel require that loading operations stop immediately if: a.) The neutron count rate from either temporary nuclear channel unexpectedly doubles during any single loading step, excluding anticipated change due to detector and/or source movement or spatial effects (i.e., fuel assembly coupling source with a detector), or b.) The neutron count rate on any indivic'ual nuclear channel increases by a factor of five during any single loading step, excluding anticipated changes due to detector and/or source movement or spatial effects (i.e., fuel assembly coupling source with a detector). 14.2-7a
14.2.12.3 Postcore Hot Functional Tests ~ g 14.2.12.3.1 ePostcore Hot Functional Test-Controlling Document -1.0 OBJECTIVE-s To demonstrate the proper integrated operation of plant primary, secondary, and auxiliary systems with fuel loaded in the core. ~ 2.0 PREREQUISITES All precore hot functional testing has been completed,4r ufudu/. 2.1 2.2 Fuel loading has been completed, 2.3 All permanently installed instrumentation on systeas to be tested is available and calibrated,in ace r/necs. vid nchmc.34 y,c,~fu. As j c and test pecu Ans. 2.4 All necessary test instrumentation is available and calibrated /n Acm/ence w/M wean 64 spdfica% aJtot frwuhaur. i 2.5 All cabling between the CEDM's and the CEDM control system is connected. 2.6 Steam generators are in wet layup in accordance with the.NSSS chemistry manual. 2.7 RCS has been borated to the proper concentration. 3.0 TEST METHOD 3.1 Specify plant conditions and coordinate the execution of the related postcore hot functional test appendices. 4.0 DATA REQUIRED 4.1 As specified by the individual postcore hot functional test l-appendices. I 5.0 ACCEPTANCE CRITERIA 5.1 Integrated operation of the primary, secondary, and related l auxiliary systems is in accordance with the CESSAR descriptions. 5.2 .As specified by the individual postcore hot functional test appendices. L 14.2-69
-14.2.12.3.4 Postcore Control Element Drive Mechanism Performance 1.0 OBJECTIVE C 1.1 To demonstrate the proper operation of the CEDM's and CEA's under __ _.._.._. HOT SHUTDOWN and Hot,- Zero - Power-conditions.-- - - - - s 1.2 To verify proper operation of the CEA position indicating system and alarms. 1.3 To measure CEA drop times. 2.0 PREREQUISITES l 2.1 The CEDMCS precore performance test has been completed. 2.2 All test instrumentation is available and calibrated. 2.3 Plant Monitoring System is operational. 2.4 The CEDM cooling system is operational. 2.5 CEDM coil resistances have been measured. 3.0 TEST METHOD 3.1 Perform the following at HOT SHUTDOWN conditions: k,' y:w:rggry;< r-
- yuk of CEbM.
3.1.1 Withdraw and insert each CEA _ _ -. ;:_i: 'ndi::ti: : .a 1 0.'.0 "---~ : : ' ::r-d 1 :; t'- ':r
- ' E^r 7e: 'e= three edditiens'
- _:_. ::nts ef d=e-;- t' e f:
e+:' -l
- A '., '. 1 '...,. ^. ; de t i, ^. _. ;..... ' : : i ' _. :i : C E' ' :.
.~. 3.2 Perform the following at hot, zero power conditions: W 9"l&
- ctbM,
---::_.n;p;;;t;;c;g:;et.e..., k ' W' y: l 3.2.1 Withdraw and insert each CEA c,d l 3.2.2 Measure and record drop time for each CEA. 3.2.3 Perform three measurements of drop time for each of those CEA's falling outside the two-sigma limit for similar CEA's. 33 6 rfen f4< p /h ws.,al 9 %: 4.0 DATA REQUIRED 3,3,7 y;gt,,,ag age,gea.,f yg Jo*h mer,t9 pr*s. indiedr% a,./ alan,..r. 4.1 CEA drop time. /c 84m Ml enuM d 4.2 RCStemeraturebndp ssure .tMA WA. N -[ -)*. 4.3 CEA position and alarm indications. 14.2-72
a 14.2.12.4.2 CEA Symetry -d c~a'i~; Test ~ l 1.0 OBJECTIVE r 1.1 To demonstrate that no loading or fabrication errors that result in measurable CEA worth asymmetries have occurred. s 4.+ ....._....s., T. 2__ .2 ___L er a s.
- n.
2_;.. _._L._;,. y..,.. i w. n. 1:._. _1 .1 n.. c., c. FVa
- wii, a
i gw ya y5 g m 1
- V Ji6 I b bA3%
w ini W p 4 4 s ay % 4 6Eb 5 J M 4 i E W% pi b t I V' '. w ww . m s. ,...,~ ,,m ___;2.: ___ ._2 .m_ 2,, ___1_ __2 VI w wr r. u s b ivrig w w w.rg gj - ~, s i I V e VV rui u y gwa j w.ag,mg rv s u ww pw r v- -g, w w 1, .u_ _m _t . ec ee <,
- m. < _ a m,-
sus.. >u, er, 14.2-77a
2.0 PREREQUISITES e 2.1 The reactivity computer is in operation. 2.2 The reactor is critical at the desired conditions with the con-trolling CEA group partially inserted and in manual control. 3.0 TEST METHOD M CE^ C::;1'm; Ch::h (MOT C""T00"" "r:t :f :dind"} p calnetaA NA 4e 4 m e g - *_9
- 4,4 61 s
11 A e7 3 7;j ;; ; ; nanstiun emmetivitu incertinn sad than .e wi t harya, uma i s sut i Wa bim b( A 7. 3.1. 2 $$Ch ). 3 esh6At6d [w tirs ^ 3.Ed CEA Symmetry Test (hot, zero power conditions - 565'F, 2250 psia) 3. 1 The first CEA of a symmetric group is fully inserted with all remaining CEAs withdrawn except the controlling group, which is positioned for zero reactivity. 1 3./. 2 The inserted CEA is withdrawn while another CEA in the symmetric group is inserted and the differences in worth (net reactivity) of the CEAs is determined from the reactivity computer. 3. 3 The remainder of tne CEAs in the symmetric group are sequentially swapped until the relative worths of each CEA in the symmetric group has been determined. c 2 1 1 3./.4 Repeat steps 3.2.1 - 3.2.3 for the remainder of the groups. 4.0 OATA REOUIRED 4.1 Conditions of.the-measurement. 4.1.1 RCS temperature. 4.1.2 Pressu:*izer pressure. 4.1.3 Boron concentration. 4.2 Time dependent data. 4.2.1 CEA position. 4.2.2 Reactivity computer traces. 5.0 ACCEPTANCE CRITERIA 5.1 The relative worth of symmetric CEAs are within the acceptance criteria specified in Table 14.2-7. .4.4 Ali CCA; er; d;;;m trat;d t: 5: ::;;1:. l 14.2-78 1
F 14.2.12.4.4 Shutdown and Regulating CEA Group Worth Test l 1.0 OBJECTIVE i 1.1 To determine regulating-{iv - :=e =-Ti-J and shutdown CEA group worths necessary to demonstrate shutdown margin (i.e., worth of all CEA's less the highest worth CEA). 1.2 To demonstrate that the shutdown margin is adequate. 2.0 PREREQUISITES 2.1 The reactor is critical. 2.2 The reactivity computer is operating. 3.0 TEST METHOD 4 320*F Ava Swrtows 3.1 MOT =$N9790 W=99NDitt9MS measurement of regulating 4CEA groups down to the musum=pumre i g-57--i im ;--idh-, ; i d-- ^--i -. _, ( f o r " fi rs t-of-a-kind" plant only).* M *T ** *** C* d l"W4 3.1.1 The CEA group worths will be measured by dilution /boration of the RCS. 3.2 Hot, zero power measurement of regulating - CEA groups! l 3.2.1 The CEA group worths will be measured by dilution /boration of the RCS. 3.2.2 Where dilution /boration is not feasibit, worths may be determined by CEA drop and/or by use of alternate CEA configurations. 4.0 DATA REQUIRED l 4.1 Conditions of the measurement. 4.1.1 RCS temperature. l l 4.1.2 Pressurizer pressure. 4.1.3 CEA configuration. 4.1.4 Boron concentration. l 4.2 Time dependant information. 4.2.1 Reactivity variation (strip chart). 4.2.2 CEA positions. 4 09
- Fotto u-ow o w sTs m s.
mv s H Wb wo asAsos.uwT ti 6 A DE. er E 65.
- F i
I J 14.2-80 L
14.2.12.4.7 Pseudo Dropped and Ejected CEA Worth Test
- k 1.0 OBJECTIVE e
L 1.1 To measure the worth of the " dropped" CEA. + 1.2 To measure the worth of the " ejected" CE from tne zero power dependent insertion limit (ZPOIL). 2.0 PREREQUISITES 2.1 Reactor critical at hot, zero power conditions witn appropriate CEA configurations. 2.2 The reactivity computer is in operation. 3.0 TEST METHOD 3.1 Pseudo worst " dropped" CFA measu:ement 3.1.1 The pseudo worst and next worst " dropped" CEA worths are established on the basis of predictions and verified during the 0% metry check. drvffi) beoea dilution /6erafim aad/or (E4c[yf,.u h. 3.1.2 The worths of the worst and next worst EAs are then measured by ( 3.2 Pseudo worst " dropped" PLCEA and worst " dropped" PLCEA subgroup measurement. 3.2.1 The pseudo worst " dropped" PLCEA and worst " dropped" PLCEA subgroups are established by prediction. 3.2.2 The worths of the worst single PLCEA and PLCEA subgroup are measured by boron dilution /boration and/or CEA compensation. 3.3 Pseudo worst " ejected" CEA measurement 3.3.1 The worth of the pseudo worst " ejected" CEA is established by means of a prediction. ,,gmj u The worths of the worst and next worst [CEAs are measured by 3.3.2 boration ^g g g'g. .} m 2m ca j 4.0 DATTEYUI$I 4.1 Conditions of the measurement 4.1.1 RCS temperature 1 (
- This test will be performed only on the "first-of-a-kind" plant.
14.2 83 t
2.2 The RRS, FWCS, SBCS, RPCS, and the pressurizer level and pressure control systems are in automatic operation. ,g,f/g,1 g & 883 3.0 TEST METHOD 3.1 Load inc'reases and decreases (steps and ramps) in accordance with the C-E Fuel Pre-conditioning Guidelines will be performed at power levels in the 96 to Wet rangh:-d% 25 & 6:7. /xMr Pa ") *
- e 5 th nin-i-th; 50 t; M-te-s
- ~~,..., T..;.\\
f as 4.0 DATA REQUIRED 4.1 Time dependent data. 4.1.1 Pressurizer level and pressure. 4.1.2 RCS temperatures. 4.1.3 CEA position. 4.1.4 Power level and demand. 4.1.5 Steam generator levels and pressures. 4.1.6 Feedwater and steam flow. 4.1.7 Feedwater temperature. 5.0 ACCEPTANCE CRITERIA 5.1 The step and ramp transients demonstrate that the plant-performs load changes allowed by C-E's Fuel Pre-conditioning Guidelines and data has been taken that will demonstrate the plant's ability to meet unit load swing design transients. 14.2.12.5.4 Control Systems Checkout Test 1.0 OBJECTIVE i 1.1 To demonstrate that the automatic control systems operate satis-factorily during steady-state and transient conditions. 2.0 PREREQUISITES l 2.1 The reactor is operating at the desired conditions. l 2.2 The RRS, FWCS, SBCS, RPCS, and the pressurizer level and pressure controls are in automatic operation. 3.0 TEST METHOD The performance of the control systems during._% sh/e &M44 sf l 3.1 c,,s/As 4eeeo4enee :nf t. ';: will be monitored to demonstrate that the ( systems are operating satisfactorily. 14.2-87
4.0 DATA REQU MED 4.1 Tim'e dependent data. s 4.1.1 Pressurizer level and pressure. 4.1.2 RCS temperatures. 4.1.3 CEA position. 4.1.4 Power level and demand. 4.1.5 Steam generator levels and pressures. 4.1.6 Feedwater and steam flow.
- 4. I.1 Fedm1%- te ACCEPTANCE CR b so s M.
5.0 RIA 5.1 The control systems maintain the reactor power, RCS temperature, pressurizer pressure and level, and steam generator levels and pressures within their control bands during 4eWP steadN, state and a g yJf. 4,3 s._. -- ~ g& A M & M x e p r 4 % n W
- ..-+
14.2.12.5.5 Reactor Coolant and Secondary Chemistry and Radiochemistry Test 1.0 OBJECTIVE 1.1 To conduct chemistry tests at various power levels with the intent of gathering corrosion data and determining activity buildup. ~1.2 To verify proper operation of the process radiation monitor. 1.3 To verify the adequacy of sampling and analysis procedures. 2.0 PREREQUISITES 2.1 The reactor is stable at the desired power level. 2.2 Sampling systems for the RCS and CVCS are operable. 3.0 TEST METHOD 3.1 Samples will be collected from the RCS and secondary system at various power levels and analyzed in the laboratory using applicable sampling and analysis procedures. 3.2 Samples will be collected at the process radiation monitor at various power levels, analyzed in the laboratory, and compared with the process radiation monitor to verify proper operation. 14.2-88
F 4.0 DATA REQUIRED 4.1 Conditions of the measurement. 4.1.1 Power. = 4.1.2 RCS temperature. 4.1.3 Boron concentration. 4.1.4 Core average burnup. 4.2 Samples for measurement of gross activities and/or isotopic activities. 5.0 ACCEPTANCE CRITERIA 5.1 Measured activity levels are within their limits. 5.2 The process radiation monitors agree with the laboratory analyses within measurement uncertainties. 5.3 Procedures for sample collection and analysis are verified. 14.2.12.5.6 Turbine Trip Test ( \\. 1.0 OBJECTIVE 1.1 To demonstrate that the plant responds and is controlled as designed following a 100% turbine trip e'd,.4 A W S A f*
- 4 -
l 2.0 PREREQUISITES 2.1 The reactor is operating above 95% power. 2.2 The SBCS, FWCS, RRS, % and pressurizer pressure and level control systems are in automatic operation. 2.5 Tha RkT i ** M AM 0*def S*%
- 3.0 TEST METHOD 3.1 The turbine is tripped.
3.2 The plant behavior is monitored to assure that the RRS, SBCS, 3 FWCS,fC!I,andpressurizerpressureandlevelcontrolsystems i maintain the NSSS within operating limits. 4.0 DATA REQUIRED 4.1 Power level prior to trip. uesp uMA fMd The following4 parameters are monitored throughout the transient. 4.2 A 14.2-89
4.2.1 Pressurizer pressure, level,x d :; n., :n 1 %s Lag 4.2.2 RCS temperature (--d -- :: u :. l 4.2.3 SG pressures.nd ':: '. s J q A C D f* ,6m--
- 1 mo
.A kmsAm n.,,.... 7 2* ,- L. s h.. _f -* * ~..., - - ' .._A y. smm, -.4y _ A A - f*.-- C 7-- -. e _3 pm t. - - - ; -... A. - uwwwu.,1 t a-a E a q e uw r. a n,- ___1_______ __; rca k '" W 7, m,w rii w y%u 1 ww ..w..=w w-- -=r 7-M % M y ;J-t ( ~ ~ b r A, ,usawa b g,s l i i 14.2-89a
= 't 5.0 ACCEPTANCE CRITERIA 5.1 The test will be evaluated against single valued acceptance limits for those safety parameters which approach a safety limit. 4e gf}i., {_ t{n fm _- l_-.1 y t ,_._1_.__ _.?,.___-. _._:'_ 14.2.12.5.7 Unit Load Rejection Test 1.0 OBJECTIVE plantHsjvn/s andn ude dedn </n)oe/ To demonstrate that the 7." ^^~ '
- ..j........
foHew;,,y. a rev$ feuejde......... :... '. 1.1 7 n m.eu s se nu e - 2.0 PREREQUISITES 2.1 The reactor is operating above 95% power.
- CCbMCS, 2.2 The SBCS, FWCS, RRS, RPCS, and pressurizer pressure and level control are in automatic operation.
3.0 TEST METHOD 3.1 A breaker (s) is tripped so as to subject the turbine to the maximum credible overspeed condition. CEB/fcS, 3.2 The plant behavior is monitored to assure that the RRS, SBCS, RPCS, FWCS, and pressurizer pressure and level control systems maintain the monitored parameters. t l 4.0 DATA REQUIRED i 4.1 Plant condition prior to trip. fg g narf yce crbr,a b l 4.2 Thefollowing} parameter temonitored[throughoutthetransient. Pressurizerpressure,hevel:..f;;.;,r. 4.2.1 RCShempe& Wet l l 4.2.2
- r. ture5;nd p:;;.cc.,
s 4.2.3 SGpressurefrf'..;'. l h.5 5N S h 55/. ]e) p/A N h A lrdn N W h & N & ant N Y * ~~ l l l 14.2-90
(
- ==&=;=c
.__.. :...,,,,,;;g ..= 4 r--, P-M.... w -. s.., -..-... -. -. ~ w -- p; r... ( I I l I 14.2-90a
{ r s 5.0 ACCEPTANCE CRITERIA 5.1 The test will be evaluated against single valued acceptance limits for those safety parameters which approach a safety limit. 4+r
- ddition, the tim; d
- p ndent "C5 ::p;r;tur: :nd prc::;r :: a:
- EC ':~:': :-f pr:::;r:: ri 5: :: ;: :
i: ::p::t:d ;;';::, 14.2.12.5.8 Shutdown from Outside the Control Room Test 1.0 OBJECTIVE 1.1 To demonstrate that the plant can be maintained in HOT STANDBY from outside the control room following a reactor trip. 2.0 PREREQUISITES 2.1 The reactor is operating at > 10% of rated power. l 2.2 The capability to cooldown on the shutdown cooling systems has been demonstrated during pre and post core hot functional tests. 2.3 The remote shutdown panel instrumentation is operating properly. 2.4 The communication systems between the control room and remote shutdown location has been demonstrated to be operational. 2.5 The remote shutdown instrumentation controls and systems have been preoperationally tested. 3.0 TEST METHOD 3.1 The operating crew evacuates the contral room (standby crew l remains in the control room). l, 3.2 The reactor is tripped from outside the control room, 3.3 The reactor is brought to HOT STANDBY by the operating crew from outside the control room and is maintained in this condition for at least 30 minutes. 4.0 DATA REQUIRED. 4.1 Time dependent data. l I 14.2-91
2.2 Results of the radiation surveys performed at zero power conditions ( are available. 3.0 TEST METHOD s 3.1 Measure gamma and neutron dose rates at 20, 50, 80 and 100% power levels. 4.0 DATA REQUIRED 4.1 Power level. 4.2 Gamma dose rates in the accessible locations. 4.3 Neutron dose rates in the accessible locations. 5.0 ACCEPTANCE CRITERIA 5.1 Accessible areas and occupancy times during power operation have been defined. 14.2.12.5.11 Xenon Oscillation Control (PLCEA) Test" 1.0 OBJECTIVE 1.1 To demonstrate a technique for damping xenon oscillations. 2.0 PREREQUISITES thi-T=. , _ _ 2 0t, _ - - _ : ':: 5::: :: ;!:ted.
- 1 --
cy"" :~t. Sc! f w. 4Lw
- 2. I I The reactor is 7
't' : _t _r m. n cr._
- w. w.
- 2. 7 2 The COLSS and the incore detector system are in operation.
3.0 TEST METH00 3.1 A free oscillation is establised. 3.2 The PLCEA's/or CEA's are used to dampen the oscillation. 4.0 DATA REQUIRED 4.1 Reactor conditions. 4.1.1 Power level. 4.1.2 Boron concentration, k.
- This test will be performed only on the "first-of-a-kind" olant.
14.2-93
4.1.3 RCS temperatures. 4.1.4 Burnup. ~ r 4.1.5 CEA, position. 4.2 Time dependent data. 4.2.1 Incore detector maps. 4.2.2 Excore detector information. 4.2.3 PLCEA's and CEA position. 5.0 ACCEPTANCE CRITERIA 5.1 The technique necessary to damp xenon oscillations t' : ;' tut -- e_ &++e using the PLCEAs and/or CEA's has been demonstrated. I 14.2.12.5.12 " Ejected" CEA Test
- 1.0 OBJECTIVE 1.1 To determine the power distribution associated with the pseudo CEA ejection from the full power dependent insertion limit (FPOIL)
CEA configuration. 2.0 PREREQUISITES + 2.1 Testing at 80% power has been completed. 2.2 The reactor is at approximately 50% power.with equilibrium conditions and with the CEAs at the FPOIL. 2.3 The incore detector system is in operation. 3.0 TEST METHOD 3.1 The " worst" case CEA (selected by calculation) is fully withdrawn. 3.2 Incore detector maps are taken before and after withdrawal of the static " ejected" CEA. 3.3 The next worst " ejected" CEA is withdrawn while inserting the previous CEA. 3.4 An incore detector map is taken. 3.5 The CEAs are returned to normal configuration.
- This test will be performed only on the "first-of-a-kind" plant.
14.2-94
4.0 DATA REQUIRE 0 4.1 Conditions of the measurement. 4.1.1 Boron coecentration. 4.1.2 Burnup. 4.2 Time dependent data. 4.2.1 Power. 4.2.2 Incore and excore detector readings. 4.2.3 RCS temperature. 4.2.4 CEA position. P. ? 5.1 r... ... within the acceptance band specified in Table 14.2-7.cf 't.., f..
- ic.-
- :c.,1, c.....': _
Fcdic;c;.e?.w. 14.2.12.5.13 Dropped CEA Testa 1.0 OBJECTIVE I 1.1 To determine the power distribution resulting from a " dropped" CEA. 2.0 PREREQUISITES 2.1 Testing at 80% power has been completed. 2.2 The reactor is at approximately 50% power with equilibrium conditions for the desired CEA configuration. 2.3 The incore detector system is in operation. 3.0 TEST METHOD 3.1 A full length CEA is selected, based on calculations, which will best verify the dropped rod assumptions used in the, safety analyses. 3.1.1 TheselectedCEAisrapidlyinsertedtotheful1[po tion. 3.1.2 The CEA remains inserted for a preselected time. 3.1.3 Excore and incore instrument signals are recorded before and after the CEA insertion.
- This test will be performed only on the "first-of-a-kind" plant.
14.2-95
3.2 PLCEA 3.2.1 The PLCEA, selected as prescribed in 3.1.1, is rapidly inserted to the fulhPin position. 3.2.2 The PLCEA remains inserted f,or a preselected time. ~ 3.2.3 Excore and incore instrument signals are recored before and after the CEA insertion. 4.0 DATA REQUIRED 4.1 Conditions of the measurement. 4.1.1 Boron concentration. 4.1.2 Burnup. 4.2 Time dependent data. 4.2.1 Power. 4.2.2 Incore and excore detector readings. 4.2.3 RCS temperatures. 4.2.4 CEA position. 'I kW e ms<2 5.1 ..,.~..........-..u.. r.-. .....__.._..._,..._._~r..___._.._ within the acceptance ; cit;ri specified in Table 14.2-7. M 14.2.12.5.14 Steady State Core Performance Test 1.0 OBJECTIVE T triter WSSS :.9d cv;r:l' ? 2nt perfor 2nce 2nd 0;t blish e -data base f r futur: u::. 1.X'l To determine core power distributions using incore instrumentation. 2.0 PREREQUISITES 2.1 The reactor is operating at the desired power level and CEA configuration with equilibrium Xe. 2.2 The incore instrumentation system is in operation. 14.2-96 1
s 3.0 TEST METHOD \\ m r u c.' k l. 3.1 Selected plant computer outputsp CPC outputs --d----I'---' --_n___ 3.2 Reactor power is determined by performing a heat balance. 3.3 The core power distribution is obtained using the incore detectors. 4.0 DATA REQUIRED 4.1 Conditions of the test. 4.1.1 Reactor power. 4.1.2 CEA positions. 4.1.3 Boron concentration. 4.1.4 Core average burnup. 4.1.5 Selected plant computer outputs and CPC outputs. L '..'. J..... '. ^ ..J...., 4.1. Y4 Incore detector maps. l 5.0 ACCEPTANCE CRITERIA N I d:l:
- l rid ' thr-j cl Ir I r 9 -- --
r.I---I 'A i r::: ::: r :: p d ^er=' + t u t C "" _"_ - d ' ^"_ _ ' _i. i t. ! ' ',; t i,. ;.,,;,. ; d; d d.. '.; 3.. u 5.11 Agreement between the predicted and measured power distributions !~ within the acceptarce criteria specified in Table 14.2-7. andccM fiAAAg &% Art. l 14.2.12.5.15-Intercomparison of PPS, Core Protection Calculator (CPC), and PMS Inputs 1.0 OBJECTIVE l.1 - To verify that process variable inputs / outputs of the PPS, the CPCs, the PMS, and the console instruments are consistent. 2.0 PREREQUISITES 2.1 The plant is operating at the desired conditions. l l 2.2 All CPCs and CEACs, and the PMS are operable. l l 14.2-97
y t ,, s,
- o Y
3.0 TEST METHOD Planir radial peaking factors are ' verified for various CEA configura-3.1 , tions by comparison of the CPC values with values measured with the inc, ore detector _ system. 3.2 The CEA shadowing factors are verified by comparing excore detector responses for various CEA configurations with the unrodded excore s responses.
- 3. 3 -
The shape annealing factors are measured by comparing incore s pouer distributions and excore detector responses during a free Xezoscillation. r4asa. $3.4 The temperature
y-factors are verified by comparing core.
y power and excore detector responses for various RCS temperatures. 4.0 DATA REQUIRED (. 4.1 Conditions of the measurement. 4.1.' l - P' owe r. 4.1.2 Burnup. 4.2 Time dependent data. 4.2.1 Incore and excore detector readings. 4.2.2 CEA position. 4.2.3 RCS temperatures.
- 5. 0 ACCEPTANCE CRITERIA Measured W ial peaking factors determined from incore flux maps 5.1 are no higher than the corresponding values used in-the CPCs.
Shv0w:'~ l 5.2 The CEA shadowing, factors, and temperature ---- 3, factors used h' in the CPCs agree within the acceptance criteria specified in the l CPC test requirements." l 5.3 The shape annealing matrix have been measured and the boundary point power correlation constants used in the CPCs are within ) the limits specified by the test requirements.** W c s k. f %\\ ( l Ch.ig w* r. a w [M _ ^ - D 0 " th0 '*"'"& Of
- kind" %
a r1... vs czy-- us(v)W. !
- As specified in the appropriate revisions or supplenents of M l
r 14.2-99 5'
t' - e. 14.2.12.5.17 ' Main and Emergency Feedwater Systems Test -1.0 OBJECTIVE 1.1 To der.ionstrate that the operation of the main feedwater and emergency feedwater systems during Hot Standby. Startup and other normal 7 operations, tranlients, and plant trips is satisfactory. = e d . - - ~ I i Amendment No. 7 March 31, 1982 ( un*s rf y yn Q h s J J 3, F % 2.%:stv) 14.2-99a ene,. ,nn.,, ,.m.-- . - - -,.. -... ~. -,
.o +e (. 4.0 DATA REQUIRED 4.1 Reactor : power. 4.2 CEA positions. s 4.3 Boron concentration. 4.4 Specified CPC inputs, outputs, and constants. l 44 P:::: f:t::t:: xp. 5.0 ACCEPTANCE CRITERIA-c M by. cc: A + @ 5.1 The values of DNBR and LPD ^'_'.._'2 - the CPCs are 'th'- th: 2.3,_
- _ eco e,
__2 1._ Z'5%' e'Li.-siy"K"Oi pK J, ' ' ~ -' ~ 14.2.12.5.19 Steam Bypass Valve Capacity Test 1.0 OBJECTIVE 1.1 To demonstrate that the maximum steam flow capacity of c' ;'- atmospheric steam dump valve upstream of the main steam isolation valves is less than that assumed for the safety analysis. a x s ~. uur s To measure the capacity ofath; b :q : p^ valve --d := 3. th: 1.2 Yh" 'f ~& h & S H YL AC.4
- 9*
2.0 PREREQUISITES C 2.1 The reactor power is > 15% full power. 6 ": 5""5 -?5 ;= - * --+t. k up=.. G e n. +1 J : =u =-# -- ^ l sondenseen 2.)[
- Control systems are in automatic where applicable.
o/va-2.K 3 The operation of the atmospheric steam de=p=, turbine by pass and shutdown cooling system have been demonstrated as part of the HOT FUNCTIONAL testing. 3.0 TEST METHOD 3.1 The individual steam flows through each of the atmospheric dump valves upsteam of the MSIVs are measured.
- d h
The capacits.3ees of n!..^. : steam bypass valveW we-measured. 3.2 .4.0 DATA REQUIRED 4.1 Reactor power. 4.2 RCS temperatures. 14.2-101
.s... - 4.3 Pressurizer pressure. ? '4.4 Steam generatpr levels and pressure. ) 4.5 Steam dump and bypass valve positions. s 5.0 ACCEPTANCE CRITERIA 5.1 The capacities of the individual steam dump valves are less than the values use "PM N[ne"d in the safety analysisM +<A A A vdw f ' %- 'g't:d steamgas 5.2 The ca aciti s v lveJL hav.re been of n L. _ _ % s 4 ju % %. v~gsured <td A<auf.A L c 1 14.2.12.5.20 Incore Detector Test 1.0 OBJECTIVE To verify conversion of the fixed incore detector signals to 1.1 voltages for input to the plant computer. 1.2 To collect baseline performance data for the movable incore detector system.
- 2. 0 PREREQUISITES 2.1 The reactor is at the specified power level and conditions.
2.2 The plant computer is operable. 2.3 The incore detector system is operable. 3.0 TEST METHOD 3.1 Fixed incore detector signal verification. 3.1.1 Amplifier output signals are measured based on test input signals. 3.1.2 Group _ symmetric instrument signals are measured. 3.2 Data is recorded from the movable incore detectors during core j traverses. -4.0 DATA REQUIRED l 4.1 Reactor power. 4.2 CEA position. l l 14.2-102 f l }
j i. ) -( r TABLE 14.2-1 ~ d LOW DOWER PHYSICS TESTS Test Title F i rs t-o f-a-kind
- Follow-On Units d l a
Low Power Biological Shield 320*F/565'F 565*F Survey Test
- '*CEA 4emphrgtSymme try 99 mew 565'F 565*F Test 4
Isothermal Temperature Coefficient 320*F-565'F 565*F Test Regulating CEA Group Worth Test 320*F & 565*F 565'F 3se ; Shutdown CEA Group Worth Test SW5 565 F 0 Differential Boron Worth Test 320*F & 565*F 565'F Critical Boron Concentration Test 320'F-565* 565'F Pseudo Dropped and Ejected CEA 565*F N/A 1. Worth Test
- An expanded test program is conducted for the "first-of-a-kind" in order to validate the design, the design methods, and the safety analysis assumptions.
" Or, th; "first of ;- kir.d" plc.c.t th; CCA cc.pli.; ch;;k i; p;rfer;;d et 200 I, rd tM CE? cy
- try t;;t i; p;rfe,n,;d at 505"I
- A Reduced testing is contingent upon the demonstration that " Follow-On" plants behave in an identical manner conformancewiththeAcceptanceCrite%riagiveninTable14.2-7.the First-Of-A-Kind M
D I
w... -- TABLE 14.2-2 (Sheet 1 of 2) POWER ASCENSION TEST Test Title First of-a-Kinda Follow-On Units ** 'l Natural Circulation Test
- > 80%
N/A Variable Tavg (Isothermal Temperature " l Coefficient & Power Coefficient) Test 20, 50, 80, 100% 50 & 100% S0 1roh Unit Load Transient Test i,t '^^"- 50,1o s 1 Control Systems Checkout Test io,50,809,1097. 50, 80% l RCS and Secondary Chemistry and Radiochemistry Test 20, 50, 80, 100% 20, 50, 80, 100% Turbine Trip Test 100% 100% Unit Load Rejection Test tuuD100% M 100% l Shutdown from Outside the Control Room Test > 10% > 10% Loss of Offsite Power Test > 10% _ 10% Biological Shield Survey Test 20, 50, 80, 100% 20, 50, 80, 100% Xenon Oscillation Control Test WhSo7. N/A I Dropped CEA TEST 9 Post leS% N/A l " Ejected" CEA Test es Post J80% N/A l i Steady-State Core Performance i Test i 20, 50, 80, 100% 20, 50, 80, 100% Intercomparison of PPS, CPC and Process Computer Inputs 20, 50, 80, 100% 20, 50, 80, 100% I Verification of CPC Power 20,50)et,tsW 20, 50)E W l Distribution Related Constants medmemmesdd48t*
- An Expanded test program is conducted for the "first-of-a-kind" in order to validate the design, the design methods, and the safety analysis assumptions.
- Reduced testing is contingent upon the demonstration that " Follow-On" plants behave in an identical manner as the "First of a-Kind" plant i
through conformance with the acceptance criteria given in Table 14.2-7. l
- Initial Power Level co n [' W. & ' W' A *-. wnM X% ~ M m t
g,g J wm J, M k c e $ ; ~ w_ _ t -- l -- - - - -- -
m T CESSAR Table 14.2-7 PHYSICS (STEADY STATE) TEST ACCEPTANCE CRITERIA TOLErgucEy Parameter First of-a-kind _LPTT Follow on Plant Symmetry Test i 1 1/2 ( CEA Group Worths 1 1 1/2 ( 1 15% or.1% Ap 4 ~1 4 L VorK (Abd Shm:idw) 10 1lo%or.05%ap.k'a Eh'Wk'Ap/*F Temperature Coefficient te E* 2 5 x 10 l Critical Baron Concentration 13 10 1 100 ppa Boron Worth 1 50 ppe Dg WEpcM W wWs 1 15 pps/% ap 1 10 pps/% op .g.25* *N.I % y. A PAPT ~ g Power Distribution em 7 M (Radial and Axial) RMSg5% RM5g3%* I ee Peaking Factors (Fxy,FR,Fzl,Fq) 1 10% 1 7.5%" Temperature Coefficient 1 5 x 10-4 @/*F !.3 x 10'4 ap/*F Power Coefficient 1 2 x 10'4 @ /% 1 2 x l',' # ' M p' w W Pseudo Ejected CEA A Ap* (2D Power Density Comparison) 1 20% g A, '# g Oropped CEA , p/ITA 12" gg[g l i (20 Power DensitygComparison) at 50% power and above e ...,....m na - H =-> '2 "__^' aa 2 ~ h ens' = fcuo'***-neb ~)' N N * 's *%o 3 - - - Ar (7HM MundM oc ro<a e 4 sss+1cstes tu cose an ns m een w e 4 xs n t M 4 y C4Ga es Asspot'c/ art). i f
'_ j '- ', - t & _ .?"d ;f .~. - 3.- l- 'CESSAR AMENDMENT 10 I EXCERPTS FROM'DECEtEER 5, 1984 f LETTER FR0f1 A. E. SCHERER, CE, TO D.'G. EISEllHUT, NRC (LD-84-070T. 'f 9 e i m - 1 l i _3
level, resulting in a transient pressure below normal operating pressure. To minimize the extent of this transient, the backup heaters are energized, contributing more heat to the water. Backup heaters are deenergized in the event of concurrent high-level error and high pressurizer pressure signals. ' ~ ~ "~ ,A low-low pressurizer water level ~ signal deenergizes all heaters before - - they are uncovered to prevent heater damage. The pressure control program is shown in Figure 5.4.10-5. 5.4.10.3 Evaluation It is demonstrated by analysis in accordance with requirements for ASME Code, Section III, Class 1 vessels that the pressurizer is adequate for all normal operating and transient conditions expected during the life of the facility. Following completion of fabrication, the pressurizer is subjected to the required ASME Code, Section III hydrostatic test and post-hydrostatic test non-destructive testing. During hot functional testing, the transient performance of the pressurizer is checked by determining its normal heat losses and maximum ;r:::urf:stice l aosdepressurizationrate/. This information is used in setting the pressure I controllers. Further assurance of the structural integrity of the pressurizer during plant life will be obtained from the inservice inspections performed in accordance with ASME Code, Section XI, and described in Section 5.2. Overpressure protection of the Reactor Coolant System is provided by four ASME Code spring-loaded safety valves. Refer to Section 5.4.12 and 5.4.13. 5.4.10.4 Tests and Inspections Prior to and during fabrication of the pressurizer, non-destructive testing is performed in accordance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code. Table 5.4.10-2 summarizes the pressurizer inspection progr&m, which also includes tests not required by the Code. Refer to Section 5.2.1 for inservice inspections of the pressurizer. 5.4-j9
~ t 7.16.4.2 Provisions shall be made to preclude the introduction of air into the SCST er :M;;i ;; :::t:! : : during fill operations. l 7.16.4.3 If : ?nt; i: ::d te tr:::fer h;;dr::f : t: th; SCST, ci' ' th; _.. _._3 NM8.T_ _@ meta' : ? crt: Of th: ?.u ? :heu!d 5: ty; 20t :r J U ;t;j-i- : _. _ 1r e--i. . u... o u. ---u.- - s w...- m r _ c u c-
- m ', ~; '. C',;.1 ' "_...',"T 4 2' C 'll"_'_ _. 5 i. 7 " _ ' Z.. ". '.'.,. "'. :7 ".:.
____m _m_2 ) 7.16.5 FIRE PROTECTION l A fire protection system shall be provided to protect the Iodine Removal System and shall include, as a minimum, the following features: Facilities for fire detection and alanning. a. b. Facilities or methods to minimize the probability of fire and its associated effects. c. Facilities for fire extinguishment. d. Methods of fire prevention such as use of fire resistant and non-combustible materials whenever practical, and minimizing exposure of combustible materials to fire hazards. Assurance that fire protection systems do not adversely e. affect the functional and structural integrity of 7-safety related structures, systems, and components. f. Care should be exercised to ensure fire protection systems are designed to assure that their rupture or inadvertent operation does not significantly impair the capability of safety related structures, systems, and components. i 7.17 ENVIRONMENTAL l l See Section 7.7 ard CESSAR Section 3.11 for environmental interfaces. l 7.18 MECH /.NICAL INTERACTION l 7.18.1 IRS components shall be properly supported such that pipe stresses i and support reactions are within allowable limits, as defined in CESSAR Section 3.9.2. CE provides the Applicant the loads at the supports / structures interface locations for components that CE supplies, under normal, upset, emergency, faulted, and test conditions, as described in CESSAR Section 3.8.5. 7.18.2 IRS piping and fittings shall be Seismic Category I. l f 68-26
4 M. Twst AT,W - -- ~ ~ - 4 7.16.4.3 All transfer lines and pump components in contact with the hydrazine - solution should be clean and hydrazine compatable as recommended by the' chemical manufacturers. i. 4 2 4 i. 1 2 I i l l a 1 - i i I t
- --e,
+w. ,,.,.,-y.4-,,,-4.m,., ,,__,._,,_._,,,e.m. ,,,_m,.,_.,n,,,,._. .n., .-.,n. p.,w,..
1.75, " Physical Independence of Electric Systems". A discussion of the physical independence is provided below which describes the compliance with Section 4.6 of IEEE 279-1971 and General Design Criteria 3 and 21. General Design Criterion 17 is di_s_c.ussed in the. Applicant's. Safety Analysis Report. The PPS cabinet is divided into four bays which are separated by mechanical and thermal barriers. Each bay contains one of the four redundant channels of the RPS and ESFAS. This provides the separation and independence necessary to meet the requirements of Section 4.6 of IEEE 279-1971. M @. iii u. = =: 1 : = m u..: = u,z___a. c. 2_____.....;=i; = =:_ : = = :. = c : m i. _2 .u.. __ :__i_ ___
- Z C -.. Z i.. ;. 1 ii L_" J C f;;'a'.C..'.;';.'.'i "." ' Z '"'.; '.';;
c_ "CI,.'; '"'X;;.1...:Z ;;1' ~i' Cii!.iZ C";E 7 Zi Z_iV'..... feed IE[ E'. 5 " U E'E N IEE-U c-EEU E' M er tE"I--thii-U E '" ' r The ESFAS Auxiliary Relay Cabinets provide separation and independence for the selective two-out-of-four actuation logics and actuation relays of the two redundant ESF Systems' Trains. Each train's logic and relays are contained in a separate cabinet with all of the train A actuation circuits in one cabinet and all of the train B actuation circuits in the other cabinet. There are mechanical and thermal barriers within the cabinets to protect different portions of the selective two-out-of-four logic from spurious actuation. The two cabinets are physically separated from each other. The RTSS consists 'of four RTSG. Each RTSG and its associated switches, T contacts, relays, etc. is contained in a separate cabinet. Each cabinet is physically separated from the other cabinets. This method of construction ensures that a single credible failure in one RTSG cannot cause malfunction or failure in another cabinet. The separation and independence of the power supplies for each of the above systems is discussed in Chapter 8.0. The interface requirements appear in Section 7.1.3 while tha implementation will appear in the Applicant's Safety Analysis Report. Protection system analog signals, sent to the Plant Monitoring System (PMS), are isolated from the protection system. Digital signals are also isolated for the associated signals coming from the protection system. All of these isolation techniques ensure that no credible failures on the output side of the isolation device will effect the PPS side and that the independence of the PPS is not jeopardized. The test results reports on the isolation devices (within CESSAR Licensing scope) will be submitted for review prior to installation of the devices in the first Applicant's facility. 7.1.2.11 Conformance to IEEE 387-1972 Conformance to IEEE 387.1972, "IEEE Trial-Use Standard: Criteria for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations", as criteria in the design of these systems is discussed in the Applicant's Safety Analysis Report. 7.1-8
CESSAR 7.1.2.10 _ _ Replace paragraph 3 with the following paragrap't. plished through 6 inch separation or barriers or conduit.Sep mation of the logic matrices (A8, AC, BC, AD, 80. CD), initiation circuitsHowever actuation. circuits, 6 inch separation is not maintained, nor can barriers or , and 4 e.onduit be utilized. tion achieved is acceptable.An analysis has been performed to show that the separa-Tests and analyses have also been completed to demonstrate that no single credible egent in one PPS bay can prevent the circu try in any other bay from performing its safety function. i l l l I l l l l I l l t
TABLE 9.2-1 no,m mu.um ,. r,.m m m. m m 5% Tem btMiNEt hi M A.. E F F i.u tmT --MAKEUP WATERALIMIT,S pH" 6.0 to 8.0 Conductivity less than 4 pmhos Chloride O.OCK Less than 4rM> ppa C1 Fluoride Less than pm F e..____2_2 e_i:2.. ,___u__ m -...-.,..(7-, u_.. n _ _ _ _ _ A _ J In _ _ _ _ A _; Silica '"^g, Less than 0.01 ppa S odiu m Leu h 6.oon nm
- If water contains C0, the pH specification may be lowered to 5.8 to compensate for CO absorption.
2
- 0:::r:ti:r ;';:: : n::r :t!:r i: -d: ; ::t:r :y;t:: d::';.. but it f:
et :::'dcr:d ::::::r3
- " rt ! : t: :::--d: y :'.: ;.:t: :n';.
l l l I e i {
~, time when exposed to normal reactor coolant chemistry conditions, approaching low steady state talues within approximately 200 days. The high pH condition produced by high ammonia concentration (to 50 ppm) minimizes corrosion product release and assists in the rapid development of the passive oxide , film. Most of the flim is established within 7 days at hot, high pH conditions. To'ald in maintaining the pH during this passivation period, lithium in the form of lithium hydroxide, is added to the coolant and maintained within a 1-2 ppm 11thium-7 range. At power, oxygen concentration is limited by maintaining excess dissolved hydrogen gas in the coolant. The excess hydrogen forces the water decomposition / synthesis reaction in the reactor core to water rather than hydrogen and oxygen. Oxygen in the makeup water is removed in the same way. In order to minimize the effect of crud deg.osition on the reactor core heat transfer surfaces, lithium-7 hydroxide additions to the reactor coolant are made. The lithium-7 hydroxide produces pH conditions within the reactor coolant at operating temperature which reduces the corrosion product solu-bility and, hence, the dissolved crud inventory in the circulating reactor coolant. The elevated pH promotes condition; within the coolant for selective deposition of corrosion products on cooler surfaces (SG) rather than hotter surfaces (core). An additional advantage is the formation of a more stable and tenacious passive oxide layer on out-of-core system surfaces. lithium concentration is maintained within a 0.1 '.0 ppm 11thium-7 rangeThe _g,o.2 during operation. d': '
- r: ' _' ' :,
'u:r 5:r:: :: ::nt--t' ': ' ^^ :, __,_.,_,__,,.u__ .m_ u _t _2 u__2 g y:,13 e g g ,1 o.._ e_ .o. .m_ m - _ 1_ _ Il _ ; ::: : _ l'ii:1 "; _ _ _ T _ 1. :_ ; ' 7 's ' _ _ ~21_ 1 I 'Z:'i 'ZL iP_ ' 1_ :::1 Efii_ L' 2:" 3._j'_;i.!i_L 'i.u'i T M _:1Titul' l'i k 7 " i:' C ~ ] g' ' E[,5EEE 'E' : U '- A E-E:-5h i'En b u '- U - bt'I-- y.......... 9.3.4.1.3.3 Reactivity Control. Boron concentration is normally controlled by feed and bleed. To change concentration, the makeup system supplies either reactor makeup water or boric acid to the Volume Control Tank, and the letdown stream is diverted to the holdup tanks via the preholdup ton exchanger and the gas stripper. Toward the end of a fuel cycle, with. Iow boric acid concentration in the coolant, feed and bleed becomes ineffi-cient, and the deborating fon exchanger is used to reduce the RCS boron concentration. The ion exchanger contains an anion resin initially in the hydroxyl form, which is converted to a borate form as boron is removed from the reactor coolant. 9.3.4.2
System Description
9.3.4.2.1 System The normal reactor coolant flow path through the CVCS is indicated by the heavy ifnes on the Piping and Instrumentation Diagrams, Figures 9.3-1 through 9.3-4. 9.3-5 J
TABLE NO. 9.3-1 (Sheet 1 of 2) OPERATING LIMITS REAC.Tod. CootANT' l 1.0 MAKEUP WATER Analysis Normal ^ ' - - - -
- Chloride (C1)
<0.15 ppm S!'!:: (Si0 ) '0.0' ;;r '. 02 ; ;;. g 0: d eti"y a 0 p-'e r i - 3 y'::,/ - c pH 6.0 - 8.0 (1) 1 Fluoride (F) 44)- <0,1 ppm '0 ' ;;; Suspended Solids <o.5 ppm 2.0 PRIMARY WATER n'.he.\\ h Qon. Core Load and Analysis Hot Functionals fM(2) Piti:' Criticality "nn: Operation pH (77') 9.0 - 10.4 4.5 - 10.2N 4.5 - 10. N Conduetivity -feb(3) W (3) feb(3) Hydrazine 30 - 50 ppm (4) 30 - 50 ppm (4) 1.5 x 0xygen ppm (5) (max. 20 ppm) Ammonia <50 ppm <50 ppm (0.5 ppm Ofssolved Gas (6)
- 1. 0 - 2.. o 4
- n. 0 - 2.0 g
Lithium 1-2 ppe ?.2 '.0 ppm 0.2
- .0 ppm
--- (?) N-50cc(STP)/kg Hydrogen (H O) (6) 2 Oxygen .d0.1 ppm .g0.1 ppm M 50) 50.1 ppm (10) Suspended Solids <0.5 ppm, 2 ppm max <0.5 ppm,2ppmmax@) <0.5 ppm,2ppmma$h Chloride 4015 ppm 40.15 ppm 40.15 ppm Fluoride 40.1 ppm .g0.1 ppm .d0.1 ppm Boron < Refueling '"00 ;;-- Concentration g g4 gg c.~..a A.-
TABLE 9.3-1 (Cont'd)(Sheet 2of2) Notes for Table No. 9.3-1 t s (1) May be as low as S.8 ff proven due to CO2 absorption. (?) " ur ' d: '* 't :;;!!: $!: t
- ter 3-t -d:d 'cr pr' : y -d: ;.
(g)-@.). Special hot conditioning ifmits: Temperature >350*F_for 7-10 days (') Dur' ; r : ' nd rd
- 'uc.!* ;, ;"' ry 5:
!! !r 2: ?.a gy; te 65;g tr:n nrrtr:tir "I.us a(gr3) Consistent with additive concentration. ) Prior to a depressurization shutdown, reduce total gas to <10cc(STP HO g T.ustAT h (r,2 ) to limit the possibility for explosive mixtures. 7 twe ;--<< m t<7 t = tt' u tu--ti r i: <-- re -- f : 5:f ; u tu-- :e a th tt:-'t=, l
- '-t:f- '-2 ;;r L
. c: = h:d (trei n t:d by L bru ithrr;h), thr rn:rt t; th: 0.2-? ;;; r=;;. "f; r=;; :st i: '- ' rt: ;rfer t: :.rf tin! f t;. Th: 'ithi= r=;; d:n =t ;;;', er ;== !nei ;. { (9) "= r rd O' :, tr th; d:tr:tta; 'r =:h=;n f: ;?:n d '- g -cer"! : (~?e ;;-e) th: 'itht = -it
- e.2-0.5 ;;:.
'9) " t :;; 91: dur' ; rn !rd. 4) 4 Jr s;4e. is .wat4.6.1 .b 30-50ype .43 3 4;~e 1 j N Rc5
- s
\\ e ss R.,, \\50 *F, (,5) Prior +o . wee.Ai$ l50*F Audn be.br or k \\.a 400*F Aue:$ cooWoun, 3 Iasov @ l (y) Aa3 4 b.esiki.a Fr. -
- p. F c.<<
b are' d ir s 3 kg r.g., sk. J J be .w ;*F iaeA 14 +k t5 h25'cc.(sT6/w.,(9 J re i ds in c se na c. e h m,ote. Aeg ssin3skdA*gu r'~e piad vwatf be
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re. a. e L @ 9 4,.3., su._o u. <s a a A u,a u... ,_,m3 3 3 A e e c.h - e..t.d f u-rs. (*)) T k. ab a.r ) c.~ Alb e n oC o.s % 3.o e r i pc<.a:HeA l C. - up b 14 L.,me.i b .it.a er.) L.,a...r,b es. (lO N oE sff '* C AN\\ L h + f* ^ \\ boe.h. turL 1
Secondary water chemistry is based on the zero solids treatment method. This method employs the use of volatile additives to maintain system pH and to scavenga dissolved oxygen present in the feedwater. A neutralizing amine is added to establish and maintain alkaline conditions in the feedtrain. Neutralizing amines which can be used for pH control are j amoniz, morpholine, and cyclohexylamine. Ammenia should be used in plants employing condensate polishing to avoid resin foul 4ng. Although the amines are volatile and will not concentrate in the steam generator, they will j reach an equilibrium level which will establish an alkaline condition in 1 4 the steam generator. Hydrazine is added to scavenge dissolved oxygen present in the feedwater. Hydrazine also tends to promote the formation of a protective oxide layer on metal surfaces by keeping these layers in a reduced chemical state. Both the pH agent and hydrazine can be injected continuously at the discharge headers of the condensate pumps or condensate demineralizer, if installed. These chemicals are added as necessary for chemistry control, and can also be added to the upper steam generator feed line when necessary. Operating chemistry limits for #:: rter rd secondary steam ggneratqr A<- -wetee are give in Tables 10.3.4-1 and 10.3.4-2. "c"enJ >e"sh e. d vW w The limits stated are divided into two groups; normal and abnormal. The limits provide high quality chemistry control and yet permit operating flexibility. The normal chemistry conditions can be maintained by any plant operating with little or no condenser leakage. The abnormal steam generator limits are suggested to permit operations with minor system fault [ conditions until the affected component can be isolated and/or repaired. The following procedures are recomended to the applicant: I# Ihr ::.Ii"C C: riI r er II I^27 ^^^ r I r bl;;;d;'.J- :Z;I ) i r.;; d;t;;t ;r; th:r frur rt::/ r :;rci'i: :=du:th ity, th: :t::r ;;;; = t:r .::ter thtu!d 5: '-.cd'Ite!y : ;!:d rd !y::d f:r ch!:r'd: ::n;:nt=ti;r- ?.. !? chi:rfdc : = : = t= tt:r '- ith'^ 5= c!'9:* !:! ::, 0;0=tir: 27 ^r.ti nu:. 'h: ; = : = t:r: :h;;id b; ;rpi:d f;r ;FI; rid = ;t I;=0 i ^ c^ ;;r e ght ' cur CS*#t ~!: : = ducti';fty =c::d: 'Our.-h :/ - S. If ;bi: rid: : =^^-t= tter f: f =^^= "f b= l'9:' / !;:0 Or 2;;- M5' g 0.1 ;;;,: :=d=::r !=h f: 9df= ted rd != E ft !: tier precedern 4 ch u?d 5: 'citituted. i = rC [1.22 are : "dI I er de ce" IIrII " " CI ^ th: 200:r ;=:r t:r.Jt:r :t :t :dy :t t; p;r; tin :=diti;r; ith;ut
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r led M4.T When the normal range is exceeded. immediate investigation of the problem should be initiated, sampling frequency i increased to the abnormal level (at least twice per 8 hour shift) and blowdown increased to one (1) percent of the main steaming rate. The problem should be corrected and the parameter (s) returned to the normal range within one week. If this cannot be done, and the parameter has a listed abnormal range, power should be reduced to 25% as if the abnormal range had been exceeded, . When the abnormal range.is exceeded, power should be reduced to the i lowest.value (maximum of 25%) consistent with automatic operation of the feed system. Continued plant operation is then possible while corrective action is taken. Power reduction should be initiated within four hours of exceeding the abnormal range. The problem should i be corrected and the parameter (s) returned to the normal range within one hundred (100) hours. be shutdown. If this cannot be done, the unit should a R
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Of 0.? ;;. : ;i::: ;;;;r r:d : tic. ch: !d 5: t-it':t:d 'nz:d':t:!; : d then!d be~-ed :rd t: 25! "fth'- 'cu-Neurt. 6:r: :5:14 ::t :::::d 255
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4 th: ::::t :::d::::r 10:h2;: :::::: In; 0; r tf9; :;;;i"::t'::: i-T:ti: 10.0.' 1 0: : ::! Or :::::d :br:rr:! -it:, : thutd:n :h: ld i: ;;;;i t:d a'th'- ': r 5:;r:- Draining or flushing of the steam generators will be necessary to reduce the impurity concentration. 10.3.4.2 Corrosion Control Effectiveness Alkaline conditions in the feedtrain and the steam generator reduce general corrosion at elevated temperatures and tend to decrease the release of soluble corrosion products from metal surfaces. These conditions promote the formation of a protective metal oxide film and thus reduce the corrosion products released into the steam generator. Hydrazine also promotes the formation of a metal oxide film by the reduction of ferric oxide to magnetite. Ferric oxide may be loosened from the metal surfaces and be transported by the feedwater. Magnetite however, provides' i an adherent protective layer on carbon steel surfaces. Hydrazine also promotes the formation of protective metal oxide layers on copper surfaces. The removal of oxygen from the secondary waters is also essential in reducing corrosion. Oxygen dissolved in water causes general corrosion that can [ result in pitting of ferrous metals, particularly carbon steel. Oxygen is removed from the steam cycle condensate in the main condenser deaerating section. Additional oxygen protection is obtained by chemical injection of hydrazine into the condensate stream. Maintaining a residual level of hydrazine in the feedwater ensures that any dissolved oxygen not removed by the main condenser is scavenged before it can enter the steam generator. The presence of free hydroxide (OH-) can cause rapid corrosion (caustic stress corrosion) if it is allowed to concentrate in a local area. Free hydroxide is avoided by maintaining proper pH control, and by minimizing impurity ingress in the steam generator. l l l l l l l l 10.3-2(a) 1
TABLE 10.3.4-1 OPERATING CHEMISTRY LIMITS FoR SECONDARY STEAM GENERATOR WATER II) Abnormal N Normal Variable Specifications Limits pH ( min,.4 sgskam - 9.2 '.5 ".h (.,,,, c,..) 9.o - 9.r. se, C* kien Conductivity (3) < A umhos/cmN 0.4 - 2.0 08 !;nt'f: umhos/cm e... ._2 2 em u u --r-- ,, n.. in __ rr-Sf1tca ?!h: - r r -- Chloride RO,4;;- d (4) 2,0 - 100 ppL 5. 45 - 42O ppk 20 - 100 pek Su\\S.ka 4 15 ppg 1 5 - 10 o,,6 NOTES: (1) Normal specifications are those which should be maintained by continuous steam generator blowdown during proper operation of secondary systems. (2} "ic.;. ;l -it: '-d'u n : ':;! t r-d' t'er er f:t: rd ?!=t th tf t:- ch=!e 5: r r re " 15 : :' "-' t: 1: crr :I':
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. 'n /:- (2) Min.4 sgsh-is .43 .... A.<3 3sv.m e. A., o ;,,3 at.3 co-c. .Js, pe.e (3) 7.9-W,. 1- ~. A i i e. sLJJ ou,, \\;,3 .9 7.o /mL.s/ y is.....).A +Le ua:4 sk..\\A La LJA.an w;4L.o Q.ur L ucs. (f) 19. L. io.. Aida. skJA.w n ):-;k .t 500 pk is l
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r. TABt.E 10.3.4-2
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/ariable Specifications -it; pH mia. A F=i::t:r system unt:f ' ; a.
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8.8 - 9.2 b. Copper-free ' :i::t:r g,g j,g system Conductivity (Intensified cation) (F.edw.ker) < M.2 o umhos/cm O.S '.S.r'::/:- Hydrazine ( F.ad wak*') 10 - 50 ppb h s.\\ved A 0xy en 3 Feed) <-+ p p b 5 'O ;;i Condensate) < 10 ppb (33 !O ;;i <hppb Sodium 10 ;;t Copper (,Ya* AWE **) -f64-4 2,p% 4 tron (Fe.Jwdee) < N ppb 10 20 ;;t i : !: j,.itr.n;;r) ' 0 ;; - c,:' tS:;t =;;r)- -+-0. 0 ;;- dit'n 10 ;;t pH C. ate.\\ A A4;&iv. (5) NOTES: (1) flormal specifications are those which should be maintained during proper operation of secondary systems. (2)
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