ML20012F297
| ML20012F297 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 03/30/1990 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20012F295 | List: |
| References | |
| 51-1176431-02, 51-1176431-2, NUDOCS 9004110097 | |
| Download: ML20012F297 (67) | |
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L Attachment 2, FPC Letter to !!RC, Dated March 30, 1990 l t
51-1176431 02 I
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CRYSTAL RIVER 3 REACTOR VESSEL LOW TEMPERATURE OVERPRESSURE PROTECTION (LTOP) ;
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l Prepared for Florida Power Corporation by i l BW Nuclear Service Company '
Engineering and Plant Services Division P.O. Box 10935 l Lynchburg, Virginia 24506-0935 l
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Page 2 of 68 9004110097 900330 PDR ADOCN 05000302 P PDC
I 51-1176431 02 i
i EXECUTIVE
SUMMARY
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This report summarizer. the results of a program to revise the reactor vessel l Low Temperature Overpressure Protection (LTOP) system for Crystal River 3 (CR- -
3), in response to NRC gentric letter 8811 (Reference 1). This LTOP system revision incorporates a new approach, suggested by the NRC in Generic Letter !
88 11, which utilizes a non Appendix G LTOP fracture mechanics evaluation, based on demonstrating that LTOP events at CR 3 are expected to be infrequent events. .
A non Appendix G fracture mechanics evaluation results in higher allowable pressures for LTOP than those permissible for normal operation, while maintaining adequate safety margins from the standpoint of reactor vessel ,
i fracture prevention. An Appendix G approach would severely impact the i pressure / temperature (P/T) window available during plant heatup and cooldown.
l As summarized in section 2, this new approach maintains essentially the same .
LTOP limitations on plant operation through 21 EFPY as exist for the current, ,~
i.e. original 5 EFPY system, thus minimizing any LTOP impact on the currently l
available P/T window through 21 EFPY. The original PORV setpoint of 550 psig and enable temperature of 280 F at 5 EFPY becomes, using the new approach, a PORV setpoint of 555 psig with an enable temperature of 283 F at 21 EFPY. In addition, this program has defined an End of-Life (EOL), i.e. 32 EFPY, LTOP system for CR 3 which will not severely impact plant operation during heatup l and cooldown, thus minimizing the concern of LTOP system infringement on the l pressure / temperature window available during heatup and cooldown at EOL.
i Using the non Appendix G LTOP approach results, a new draft LTOP Technical l Specification and Backup Bases Document have been developed. ,
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51-1176431 02 CONTENTS Elat 1.0 II W X AL APPROACH 6 t 1.1 PROGRAM OBJECTIVES 6 1.2 DESIGN CRITERIA / REQUIREMENTS 6 1.3 THE EXISTING (ORIGINAL) CR 3 LTOP SYSTEM 7 1.4 ELEMENTS OF THE NEW APPROACH 7 2.0
SUMMARY
11 2.1 TWENTY ONE EFPY CR 3 LTOP SYSTEM 11 2.2 THIRTY TWO EFPY CR 3 LTOP SYSTEM 11 2.3 MARGINS AND INSTRUMENTATION ERROR 12 2.4 13 2.5 FUTURE APPENDIX G P/T LIMIT REVISIONS 16 2.6 COMPARISON VERSUS OBJECTIVES 16
.3.0 LTOP EVENTS DEFINITION 18 .
3.1 1977 BASIS 18 3.2 REVIEW OF INDUSTRY EXPERIENCE 18 3.3 NEW BASIS 19 4.0 LTOP EVENTS - FRE00ENCY CLASSIFICATION 20 4.1 DESIGN FEATURES AND OPERATING PRACTICES 20 4.2 NUREG/CR 5186 RESULTS 21 4.3 B&W REVIEW 21 4.4 EARLIER RANCHO SECO ASSESSMENT 22
4.5 CONCLUSION
S 22 5.0 LTOP EVENTS - RCS PRESSURE RESPONSE 26 5.1 TECHNICAL APPROACH 26 5.2 ANALYSES 28 5.3 RESULTS 39 6.0 FRACTURE MECHANICS EVALUATION 41 I
6.1INIRODUCTION 41 6.2 TECHNICAL APPROACH 44 6.3 BELTLINE ANALYSIS 50 i 6.4 CLOSURE HEAD ANALYSIS 50 6.5 N0ZZLE CORNER CRACK ANALYSIS 52
. 6.6
SUMMARY
AND CONCLUSIONS 54 l
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3 51 1176431 02 7.0 LTOP SYSTEM DEVELOPMENT 55 7.1 TWENTY.ONE EFPY CR 3 LTOP SYSTEM 55 7.2 THIRTY-TWO EFPY CR 3 LTOP SYSTEM 60 7.3 INSTRLMENTATION ERROR 62 8.0 REF2RENCES 63 s
APPENDIX A: A Summary Description of the Existing CR-3 65 '
LTOP System and a Comparison Versus Requirements .
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51 1176431 02 1.0 TECHNICAL APPROACH The objectives, design criteria, and key elements of the program are summarized as follows:
1.1 PROGRAM OBJECTIVES In developing this non Appendix G LTOP system, as suggested in NRC Generic Letter 88 11, the following objectives were identified:
1.) Develop a long term solution to the LTOP requirements, thereby eliminating ,
the cost and licensing interface involved with having to revise the system at ,
each P/T limit revision. In the long run, this should be very cost effective :
for both FPC and the NRC.
2.) Develop an LTOP system which will maximize the increase in the allowable ,
LTOP pressures to be gained by a non-Appendix G approach, with appropriate ;
- justification of adecuate safety from the standpoint of fracture prevention, i thus minimizing LTOP sased administrative restrictions during plant heatup and coolcown. This will minimize LTOP interference with the pressure-temperature window during heatup and cooldown of the plant. It will also minimize the i LTOP-related Technical Specification requirements and make the operator's job easier.
3.) Investigate alternatives to locking out HP! below the LTOP enable temperature and lower the enable temperature at which HPI lockout must be considered.
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- 4.) Review and update the set of CR-3 postulated LTOP events, based on a review ,
, of industry and B&W designed NSS experience since the initial (1976) evaluation.
5.) Confirm that the postulated LTOP events are. appropriately classified as
- infrequent events at CR-3.
i 6.) Develop an approach which could be applied generically, to the other B&WOG
- units, in accordance with B&WOG and NRC desires to standardize.
- The success of the program in meeting these objectives is summarized in section j 2.6.
l 1.2 DESIGN CRITERIA / REQUIREMENTS 1
1.2.1 Governina Criteria
. By letter dated October 1,1976 (Reference 2), the NRC staff requested Florida Power Corporation (FPC) to provide systems to protect against Low Temperature Overpressurization (LTOP) events. Per that letter, the governing criteria to Page 6 of 68 i
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51 1176431 02 apply in detemining the adequacy of LTOP protection was that no single i equipment failure or single operator error would result in exceeding 10CFR50, appendix G limitations.
NRC Generic Letter 88 11 (Reference 1), recently modified the governing criteria to permit consideration of alternatives to Appendix G LTOP setpoints with appropriate justification of adequate safety from the standpoint of fracture prevention.
1.2.2 additional criteria In the late 1970's agreement was reached on a CR-3 LTOP " System", initially developed based on the reactor vessel 5 EfPY Technical Specification P/T limits, which relied on operator action as the first line of defense and; as a backup, a low Pressure PORY Setpoint (550 psig) below 280 F. The following specific criteria for system performance were defined (Reference 4): ,
1.) Ooerator Action: No credit can be taken for operator action for ten minutes after the operator is aware of a transient.
2.) Sinole Failure: The system must be designed to relieve the pressure transient given a single failure in addition to the failure that initiated the ,
pressure trancient.
3.) Testab111tvr The system must be testable on a periodic basis consistent with the system's employment.
4.) Seismic and IE EE 279 Criteriat Ideally, the system should meet seismic Category 1 and IEE! 279 criteria. The basic objective is that the system should not be vulnerable to a common failure that would both initiate a l pressure transient and disable the overpressure mirtigating system. Such l
events as loss of instrument air and loss of offsite power must be considered.
1.3 THE EXISTING (ORIGINAL) CR-3 LTOP SYSTEM A brief description of the existing, i.e. original, CR 3 LTOP system and how it addresses each of the design criteria is provided in Appendix A.
1.4 ELEMENTS OF THE NEW APPROACH The main steps involved in the Crystal River 3 LTOP revision summarized in this report are illustrated in Figure 1.4-1.
The LTOP system revision described herein is based to a great extent on the existing licensed CR-3 LTOP concept / system, as defined in Appendix A.
The approach used to develop the original CR-3 LTOP system consists of the following main elements:
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ELEMENTS OF THE ORIGINAL LTOP SYSTEM APPROACH i i
- Element 1: Consider postulated LTOP events as anticipated operational l occurrences.
Specift ation ppendixGP/T$
- Element 3: Implement the PORV/ Pressurizer bubble approach.
- Element 4: Define administrative controls to limit the severity of LTOP '
events or to preclude specific postulated events. !
- Element !' t Require the operator to maintain the Technical Specification Appendix G >/T limit during heatup and cooldown. ;
The new approach would replace elements 1 and 2 as follows:
ELEMENTS OF THE NEW LTOP APPROACH
- Element 1: Demonstrate that the frequency of an LTOP event that would exceed Appendix G limits is expected to be much less than one per reactor '
lifetime.
- Element 2: Therefore develop and protect a non appendix G frt.cture l l,
mechanics P/T limit.
- Element 3: same t
- Element 4: same
- Element 5: same -- note that the operator still maintains the Technical Specification Appendix G P/T limit during heatup and cooldown.
The following tasks were. required to develop the new system:
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Figure 1,4-1 Crystal River 3 ,
6 Non Annandix G LTOP Anoroach i
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51 1176431 02 2.0
SUMMARY
This section summarizes the most important aspects of the revised LTOP system ,
for CR 3. The system parameters at 21 and 32 EFPY are identified, instrumentation error is discussed, and HPI options are explained. In addition, the relevance of the 15 EFPY CR 3 Normal Heatup and Cooldown Technical Specification P/T limits is emphasized, and a comparison of the results versus objectives is provided.
l l 2.1 TWENTY-ONE EFPY CR 3 LTOP SYSTEM l l
The resulting CR 3 21 EFPY LTOP system consists of the following elements: ;
- An enable temperature of 283 F.
- A pressurizer level < 220 inches.
- Reactor coolant system (RCS) pressure and temperature must be maintained in accordance with normal heatup and cooldown Technical Specification P/T limits for 15 EFPY (Reference 5).
- RCS heatup and cooldown rates (F/hr) must be maintained in accordance with normal heatup and cooldown Technical Specification P/T limits for 15 EFPY (Reference 5).
- Inadvertent HPI injection and inadvertent core flood tank discharge must be administrative 1y precluded as credible events.
2.2 THIRfY-TWO EFPY CR-3 LTOP SYSTEM The resulting CR-3 32 EFPY LTOP system consists of the following elements:
- An enable temperature of F.
- A pressurizer level < '
inches.
- Reactor coolant system (RCS) pressure and temperature must be naintained :
in accordance with normal heatup and cooldown Technical Specification P/T limits for 15 EFPY (Reference 5).
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- RCS heatup and cooldown rates (F/hr) must be maintained in accordance with '
I normal heatup and cooldown Technical Specification P/T limits for 15 EFPY ,
(Reference 5).
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- Inadvertent HPI injection and inadvertent core flood tank discharge must be administrative 1y precluded as credible events.
2.3 HARGINS AND INSTRUMENTATION ERROR l
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51 1176431 02 Figure 2.4 1 (R.3 HPl System Schematic Page 15 of 68 j
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- 2.5 FUTURE APPEN0lX G P/T LIMIT REVISIONS l As noted in sections 2.1 and 2.2 above, the revised CR 3 LTOP system requires i that reactor coolant system (RCS) pressure and temperature, and the RCS neatup !
and cooldown rates (F/hr be maintained in accordance with normal heatup and
' t cooldown Technical Specif)ication P/T limits for 15 EFPY (Reference 5). f l
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.i f I The basis for maintaining the RCS heatup and cooldown rates within the 15 EFPY (
l P/T Ter,hnical Specification requirement is that these rates were used as a i l
basis to develop the LTOP fracture mechanics limits. The specific rates used I i for LTOP are provided in section 6.2.1.5. [
Therefore. FPC must assure that stbseauant P/T limit revisions maintair the !
initial P/T condition and heatun ano cooldown rate ass - tions used in LT04 l l
r 2.6 COMPARISON VERSUS OBJECTIVES !
I Following is a brief discussion of the revised CR 3 End of Life (EOL) LTOP l system, summarized in sections 2.1 and 2.2 above, as compared with the j technical objectives identified in section 1.1. ,
t Objective 1: Develop a long term solution to the LTOP requirements, thereby !
eliminating the cost and licensing interface involved with having to revise the i system at each P/T limit revision, in the long run, this should be very cost !
effective, i Result: An EOL LTOP system has been developed for CR 3 which will permit I heatup and cooldown at E0L i.e. it has minimal impact on the P/T window j available during heatup and cooldown.
Objective 2: Develop an LTOP system which will maximize the increase in the allowable LTOP pressures to be gained by a non Appendix G approach, thereby 4 minimizing LTOP based administrative restrictions during plant heatup and r cooldown. This will minimize LTOP interference with the pressure temperature j window during heatup and cooldown of the plant. It will also minimize the i LTOP related Technical Specification requirements and make the operators job ,
easier. I l
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51 1176431 02 l Result: The CR 3 E0L LTOP does not require any administrative requirements !
which further reduce the P/T window available during heatup and cooldown.
Only the PORY setpoint 555 psig below 283 F at 21 EffY, and psig belov l F at 32 EFPY wili restrict the available P/T window, and even this restriction can be eliminated if additional setpoints for the PORV are utilized. ,
Objective 3: Investigate alternatives to locking out HPI below the LTOP enable temperature.
l Result: Even with the higher LTOP allowable pressures resulting from a non- ;
appendix G approach, the analyses documented in section S below still !
demonstrate that inadvertent HPl actuation below the LTOP enable temperature !
must be precluded. However, this non Appendix G appro&ch did result in .
lowering the temperature at which HP! must be locked out compared to other existing approaches - i.e. it minimizes, compared to other existing '
approaches, the range of temperature over which HP! must be locked out. l Objective 4: Rigorously review and update the set of postulated events for l B&WOG LTOP, based on a review of industry and B&WOG experience over the last !
decade. l Result: This was completed, as documented in section 3.0. ;
Objective 5: Confirm that LTOP events are expected to be infrequent events at ;
B&WOG plants. j Result: This was completed, as documented in section 4.0. !
Objective 6: Develop a B&WOG generic approach, in accordance with B&WOG and NRC desires to standardize whenever possible. l t
l Result: The approach taken for this CR 3 LTOP is applicable to the other B&W l l operating units, as desired. l 6
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51 1176431 02 3.0 LTOP EVENTS DEFINITION 3.1 1977 BA$l$
1 In the 1976 77 time frame, evaluations were conducted to examine the CR 3 l system design and operation for susceptibility to overpressurization events j
during startup and shutdown and to determine the pressure response of the Reactor Coolant System (RCS) to potential events which cause pressure
) increases.
l At that time (pre 1977) there were 30 documented pressurized water reactor (PWR) events which resulted in RCS pressure exceeding technical specification Appendix G limits. Of these events,15 involved charging / letdown imbalance, 4 j
i involved reactor coolant pump (RCP) starts, 2 involved accumulator inadvertent i openings,1 involved inadvertent safety injection., 7 involved testing, and 1 l was unknown. Only one of these events involved a B&W Owners Group (BWOG) pl ant. This event was not a transient, but was a planned operation as part of the zero power physics testing procedure. The pressure limit included a two-year irradiation shift. Since the event occurred before initial criticality, J
the pressure limit with the two year irradiation limit removed was not exceeded.
Based on operating experience to that date, an evaluation of potential i scenarios, and B&W plant design and operating characteristics, the following seven events were chosen for further evaluation in the development of a CR 3 LTOP systemt l
j a. Erroneous actuation of the High Pressure injection (HPI) system
- b. Erroneous opening of the core flood tank discharge valve i c. Erroneous addition of nitrogen to the pressurizer
- d. Makeup control valve (makeup to the RCS) fails full open
- e. All pressurizer heaters erroneously energized
- f. Temporary loss of the Decay Heat Removal System's (DHRS) capability to remove decay heat from the RCS
- g. Thersal expansion of the RCS after starting an RC pump due to stored thermal energy in the steam generator These eve'nts, their evaluation. and the assumptions associated with them formed the basis for the original CR 3 LTOP system design, documented in reference 4.
3.2 REVIEW 0F INDUSTRY EXPERIENCE Page 18 of 68 l
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in developing a revised LTOP system for CR.3, it was felt necessary to review l the original event definitions to determine if any changes would be required as !
. a result of post.1977 industry experience. l NUREG/CR 5186, 'Value/ impact Analysis of Generic issue 94, ' Additional Low !
Temperature Overpressure Protection for Light Water Reactors," NovemDer 1988 !
! (Reference 6), was prepared by Pacific Northwest Laboratory for the USNRC, !
This document contained a Licensin Event Report (LER) data base of 30 !
overpressure Mitigation System (OMS)gchallenge events which occurred between !
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1980 and 1986. The event descriptions of this data base were reviewed against
, the pre 1977 events and the CR 3 .1977 event basis for the purpose of j l
identifying any new or different event scenarios. !
( AEOD Case Study C401 " Low Temperature Overpressure Events at Turkey Point Unit f 1 4." USNRC, March 1984 (Reference 7), was also reviewed for a better .
I understanding of the event details and system design details as they affect the !
pressure transients at the different PWRs. ,
Of the 30 events between 1980 and 1986 where the setpoint of the protection ,
system was reached 21 were of the mass addition type and 9 were of the energy :
addition type. Fifteen of the mass additions involved a charging / letdown i imbalance and 6 involved inadvertent safety injection. Seven of the energy >
additions involved pressure surge on RCP starting and 2 involved excessive i delta temperature between the RCS and the steam generator. None of these +
events involved a $1WOG plant. .
The review did not identify any new events, equipment failures or operator l errors that would not be enveloped by the 1977 CR 3 basis events.
3.3 NEW 8ASl$ :
As the result of the industry experience review, it was concluded that the same i event definitions developed in 1977 would be approsriate for developing a revised LTOP system, and the same events should be evaluated.
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I 4.0 LTOP EVENTS FREQUENCY CLAS$1FICAT10N Low Temperature Overpressurization (LTOP) events that exceed Appendix G P/T !
limits are not anticipated operational occurrences at CR 3. One reason is that !
the CR 3 hettup, operation, and cooldown is performed with a gas bubble in the i pressurizer. Reviews of operating experience both by Pacific Northwest Lab i l (PNL) (Reference 6) and B&W indicate that LTOP events at CR 3 are expected to i i have a frequency of occurence much less than once per reactor lifetime. l 1 ;
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4.1 DESIGN FEATURES AND OPERATING PRACTICES !
With respect to low temperature overpressurization considerations, the B&W.
designed CR 3 plant has two design features or operating practices which are !
- fundamentally different than that of most other pressurized water reactor (PWR) ;
! designs. l
! First, CR 3 is designed, except for system hydrotest, to always have either a .
l gas or steam bubble in the pressurizer. This is true even during vent and !
I fill, startup, and shutdown operations. Other PWR designs conduct these i
! in a water solid condition. :'
) operations The " waterwith solid' thesystem Reactorhas Coolant System for the potential (RCS) a lmost instantaneous pressure .
increase for mass addition transients and, for energy addition transients, the '
pressure increase is significantly faster than it will be for a system with a steam or gis space. Operating experience has shown that most all low I temperature overpressure events occurred during startup or cooldown during a i
! water solid cordition. The practice of conducting these operations at Crystal i River 3 (CR 3) with a pressurizer gas or steam volume allows the operator i sufficient time to identify, access, and terminate the transient should 4 ,
precursor to an overpressurization event occur. Operating with a steam or gas e space in the pressurizer is the most significant difference between the CR 3 ;
plant and other PWR designs and, undoubtedly, is one of the major reasons that no low temperature ovepressurization events have occurred during the commercial l operation of any of the BWOG plants, including CR 3. j Second, at low pressures other PWR designs are such that an inadvertent $
isolation of the Residual Heat Removal System (RHRS) will terminate letdown flow from the RCS. In this non CR 3 design, because the normal letdown line orifices restrict the flow, a line to the Chemical and Volume Control System (CVCS) is taken from the discharge of the RHR$ heat exchangers. A positive displacement charging pump is normally operating in this mode supplying makeup :
and RC pump seal injection cooling. An inacvertent isolation of the RHR$ ;
suction line effectively bottles up the RCS and the system is exposed to the '
discharge head of the charging pump. The most susceptible RCS condition for ['
pressure transients is when the RCS is water solid or nearly so such as at the completion of the filling and venting process when the volume of remaining air is extremely small. The loss of letdown flow with continued charging flow has been one of the major causes of low temperature overpressure events at PWRs.
The CR 3 design does not route the letdown flow through the Decay Heat Removal Page 20 of 68 i
51 1176431 02 System (DHR$): and, equally as important, does not operate in a solid or near solid system condition.
4.2 NUREG/CR.5186 RESULTS The recent work by PE for the NRC, NUREG/CR 5186 (Reference 6), demonstrates that LTOP is not an anticipated operational occurrence at CR 3. The PNL team analyzed the Licensee Event Report (LER) data base for all operating Pressurized Water Reactors (PWRs), investigating overpressure mitigation system (OMS challenges which occurred between 1980 and 1986. The LER data base was used) to determine the frequency of challenges to the overpressure protection systems, the fraction of OMS challenges for which the OMS failed, and the overpressurization. PNL resulting evaluatedfrequency of Reactor three classes of LTOPCoolant System mitigation (RCS) systems: pi lot operated relief valve (PORV) category, Residual Heat Removal (RHR) relief valve category, and PORV plus pressurizer bubble category, the latter being comprised of B&WOG plants and the former two representing the other PWRs. Table 4.21, reproduced from the NUREG, sumarizes some results from the PNL review.
At B&WOG plants during 1980 to 1986, there were no events that challenged the LTOP overpressure mitigation system hardware (PORV). By assuming a challenge, FNL bounded the B&WOG LTOP challenge frequency (less than 0.01s per reactor-year). A prediction of PORV reliability 0.087 failures / demand) was made based on LTOP challenges that occurred at th(e other PWRs. The product of LTOP challenge frequency and PORV reliability yields the NUREG/CR 5186 prediction of less than 0.0016 overpressurizations per reactor year for B&WOG plants. This is less than one overpressurization event every 625 reactor yearst in other words, much less than once per reactor lifetime.
The B&WOG plants had much better results than the other PWR owners groups because all B&WOG plants fall into the 'PORY plus nitrogen bubble" category of overpressurization mitigation system. Because a bubble is maintained in the pressurizer, B&WOG plants have a much slower pressurization than the other PWRs that bestup and cooldown with a water solid pressurizer and depend on PORVs or RHR relief valves for protection. (NUREG/CR 5186 points out that essentially all of the recorded LTOP events occurred during water solid operation at Westinghouse Owners Group and Combustion Engineering owners Group plants.)
Consequently, the B&WOG operator has time to prevent pressurization and avoid challenges to the overpressure mitigation system (PORV).
4.3 B&W REVIEW In addition, a review has been performed by B&W of B&WOG operating history, from commercial operation through February 198g. The review supports the conclusion that LTOP events exceeding Appendix G P/T limits are not anticipated operational occurrences for CR 3.
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l The final result was that, in every event, the operators acted to terminate the event well before the PORY or P/T limits would have been challenged, in addition, there was a search to identify any instances where the PORV function may have been inoperable, if challenged. Two LERs were found involving unavailability of the PORV for LTOP. These occurrences are i
consistent with expected PORY unavailability (e.g. NUREG/CR 5186 reports PORV l
unavailability of 0.087 failures / demand).
4.4 EARLIER RANCH 0 SEC0 ASSESSMENT The NRC Safety Evaluation Report (SER) that approved a 1983/84 Rancho Seco LTOP submittal addresses the frequency of LTOP events. The SER concluded that "a low Temperature Overpressure event, which would violate the Appendix G pressure / temperature limits, should not be considered an anticipated operational occurrence at Rancho Seco Nuclear Generating Station.' Although not prepared specifically to address Crystal River, it supports the conclusion that'the LTOP event is not an anticipated operational occurrence.
4.5 CONCLUSION
S LTOP events that exceed Appendix G P/T limits are not anticipated operational occurrences for CR 3. In fact, this review indicates that there has never been an overpressurization event that violated Appendix G P/T limits or challenged Page 22 of 68
51 1176431 02 the low temperature PORV setpoint in over 100 years of B&WOG commercial operating experience. The 86WOG design, that is, operation with a bubble in the pressurizer, is the most important factor in the exceptional B&WOG record for LTOP events. This conclusion is supported by B&W review of potential overpressurization events and by NRC sponsored analysis reported in NUREG/CR.
5186.
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51 1176431 02 i TABLE 4.2 1 Results from NUR[UCR.$186 t
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Overpressure Mitigation OMS Challenges Overpressurizations i,
$vitem (OMS) Catenori car._ Reactor Year Der Reactor Year !
PORV 0.094 0.0082 j I
RHR Relief Valve 0.125 0.0179 PORY + N2 Bubble * <0.018 <0.0016 j
- CR 3 is in this category, f t
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51 1176431 02 TABLE 4.3 1 Postulated LTOP [ vents
- 1. Makeup control valve f ails open.
- 2. Inadvertent Core Flood Tank discharge.
Inadvertent HPl actuation.
3.
- 4. Nitrogen addition to pressurizer.
- 5. Inadvertent pressurizer heater energitation.
- 7. Thermal RCS expansion after start of Reactor Coolant Pump due to stored energy in OTSG.
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51 1176431 02 5.0 LTOP EVENTS RCS PRES $URE RESPONSE 5.1 TECHNICAL APPROACH 5.1.1 Backaround This section sumarizes the results of the pressure response analysis of the Reactor Coolant System (RCS) to the following CR 3 LTOP design basis transients, identified in section 3.0:
- a. Erroneous actuation of the High Pressure injection (HPI) system
- b. Erroneous opening of the core flood tank discharge valve
- c. Erroneous addition of nitrogen to the pressurizer
- d. Makeup control valve (makeup to the RCS) fails full open
- e. All pressurizer heaters erroneously energized
- f. Temporary loss of the Decay Heat Removal System's (OHRS) capability to remove decay heat from the RCS
- g. Thermal expansion of the RCS after starting an RC pump due to stored thermal energy in the steam generator The RCS pressure response due to the postulated CR 3 LTOP events was determined for the initial CR 3 LTOP system in 1977 (Reference 4). A reanalysis is required because the initial and limiting (final) P/T values have changed, i.e.:
the initial pressure and temperature conditions at which an LTOP event could initiate has changed due to changing P/T Tech Spec requirements, as sumarized below.
the limiting LTOP fracture mechanics limit for which 10 minute operator action is required - has changed, based on the analysis summarized in section 6.
The results are summarized in section 5.3 and interpreted, i.e. are applied to define an acceptable LTOP system, in section 7.
5.1.2 lett Pressurizar Model A B&W 177 stand alone pressurizer model with RCS makeup and letdown flow capabilities, developed from the Modular Modelling System (MMS) code, was utilized in these pressurization studies. The HMS pressurizer model provides pressure feedback to the RCS system which modifies the surge rate into the pressurizer as pressure increases, a feature which acts to decrease makeup flow in accordance with the makeup pump characteristics and increase letdown flow Page 26 of 68
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51 1176431 02 4
L with increasing RCS pressure, thereby providing a realistic model of RCS system i l
. behavior, j 5.1.3 'RCS Pressure Leon A* and 'RCS TemneraturA*, f i
! Throughout this document, all analytical results are reported as 'RCS Pressure !
Loop A". l l
I I
L !
?
l ,
! Plant parameter values used in the analysis are actual values. Intrumentation !
error is discussed in Section 2.3. !
l l
5.1.4 Initial RCS Pressure and Temnerature !
4 l f i
I l
F 5.1.5 Alarms and Indications ,
Although the time available for operator action is based on the time at which j the first alarm occurs, and not necessarily on initiation of the event, all of !
the analyzed transients start with at least one signal which alerts the j operator.
The MV transient is alarmed on initiation of the event by a high makeup flow >
al arm. In addition, numerous pressurizer level and makeup tank level alarms '
are sounded within the first ten minutes to alert the operator.
Loss of decay heat removal capability could only be caused by loss of flow in ;
the Decay Heat Removal System or in the cooling water system serving the Decay Heat Removal System. Loss of flow in either system would immediately actuate l low flow alarm (s), thus alerting the operator. l The operator would be alerted to an inadvertent pressurizer heater actuation by ;
the pressurizer heater status indications on the control console, increasing ;
Page 27 of 68 t
P d
l 5
1 ;
i 51 1176431 02 !
RCS pressure, and higher than normal latdown line and makeup line flow rates. l among others. l t
i As described in the following sections, the other credible LTOP events are {
i either precluded or self limiting, with operator action to terminate the ;
l transient not required, j 5.1.6 Pressurizar Levels i l
As reported in the following sections, the mass addition transients were !
analysed at various RCS temperatures and pressures for several different ,
i initial pressurizer levels. Pressurizer level alarms and typical operating !
j ranges for CR 3 are presented below: l
- I l I I 5.1.7 Makeuo Tank Levels [
I l
- For the makeup tank, which is the normal suction source for the makeup /HP! i pump, water level at the start of an LTOP transient is conservatively assumed -
l to be just below the high alam level at inches. The relationship of this
. level to the other makeup tank level setpoints is:
\
l r The initial makeup tank level is not used in the analysis except to determine ,
when low level alarms would be actuated. ,
5.2 ANALYSES Details of the pressure response evaluations performed for each of the LTOP ,
events are reported in this section and summarized in Section 5.3.
5.2.1 Erroneous actuation of the Hioh Pressure Iniection (HPI) System t The results of the erroneous HPI actuation pressure response evaluation are reported for various initial RCS pressures, RCS temperatures, and pressurizer levels in Table 5.21, and illustrated in Figure 5.21.
Page 28 of 68 l
l
_ _ _ _ ._ _ - - - _,__ - - -._.i
I !
i l
l 51 1176431 02 }
The MMS pressurizer model includes heaters, spray and relief valve l capabilities. The following facts apply to the HP! analysis: ;
l !
l l l
i i
- As discussed in detail in section 7.0, because the results demonstrate that 10 l
- minutes is not available for operator action, the CR 3 LTOP system will require i l
that erroneous actuation of the HP! system be precluded below the LTOP enable ,
temperature. ;
, 5.2.2 Erroneous onanina of The core Flood Tank Discharae Valve f
( i As discussed in section 7.0, this event is not credible because the core flood :
tank motor operated block valves, CFV 5 and CFV 6, will be closed and their .
j breakers locked out and red tagged whenever the core flood tank pressure is i greater than or equal to the maximum allowable RCS pressure for the existing -
RCS temperature. This allows testing and draining of the Core Flood System, ;
l yet precludes the LTOP event. !
l i l 5.2.3 Erroneous Addition of Nitronen to the Pressurizar It is not credible that this event can overpressurize the RCS, Nitrogen is !
added to the pressurizer during plant cooldown at an RCS pressure of 50 psig or !
less. Nitrogen addition is controlled by a 50 psig regulator and a 100 psig regulator, arranged in series. A relief valve located downstream of the 100 '
psig regulator provides protection in the event of regulator failure. This system is shown on FSAR Figure 6 28.
5.2.4 Makeuo control Valve (Makeuo to the RCS) Fails Full Doen ,
i The results of the makeup control valve fails open pressure r'sponse
- evaluation are reported for various initial RCS pressures, RCS temperatures, and pressurizer levels in Table 5.2 2, and illustrated in Figure 5.2 2. In addition. Figure 5.2 3 shows the pressure versus time results for the event at the limiting initial RCS temperature and pressure. F and psig, where !
the margin between the normal heatup and cooldown P/T limit at 15 EFPY and the LTOP P/T limit from section 6 is minimum. :
The MMS pressurizer model includes heaters, spray and relief valve capabilities. The following facts apply to the makeup control valve fails open i analysis:
l Page 29 of 68
}
I 1
1 l 1 51 1176431 02 1 l
1 3
l l
- i 1 :
As discussed in detail in section 7.0, because of these results, the CR 3 L10P )
I system will require that pressurizer level be limited below the LTOP enable l j temperature to less than 220 inches at 21 [FPY and to less than inches at ;
32 EFPY. I i
)
l Two other significant results of the makeup control valve failing open analysis are the following:
1 1 1.) With Makeup Tank level, i.e. inventory, limited to 70 inches or less, j and the automatic switchover to BWST suction prevented, the Makeup Control Valve failing open event will be self limiting i.e. will terminate prior to exceeding the LTOP P/T limit.
An RCS vent of 0.75 square inches, with a backpressure less than or l 2.) l equa to 100 psig, is adequate to maintain RCS pressure below the 32 EFPY
- 1. TOP P/T limit of psig, with a maximum makeup flow rate of gpm.
5.2.5 All Pressurizar Heaters Erroneousiv Enareired The results demonstrate that this is a very slow moving ' transient" about to psf / minute (during the initial 10 minute period) in the range of 1 interest. Table 5.31 sumarites the results of a typical case.
The following facts apply to the analysis:
l l
l Page 30 of 68 l
I i
51 1176431 02 !
As the results indicate, this is a very slow moving ' transient" and therefore i is not a serious LTOP threat, j 5.2.6 Temocrary loss oL the Decav Heat Removal System The analysis of the loss of decay heat system was made with conservative !
assumptions conceived. to produce the maximum rate of pressure increase af ter the loss of cooling capability. The results demonstrate that this is a very l slow moving " transient' about psi / minute (during the initial 10 minute [
period) in the range of interest. Table 5.3 1 summarizes the results. ;
1 The following facts apply to the analysis:
i 1
l l
Page 31 of 68
$1 1176431 02 As the results indicate, this is a very slow moving
- and therefore is not a serious (TOP threat.
5.2.7 Thermal Ernantion of the RCS Inventory Af ter RC Pume Start Several means of inadvertant overpressurization following start of an RC pump with stored thermal energy in the steam generators' secondary fluid inventory were examined in the 1976/1977 analyses (Reference 4). Although not a credible condition. the worst case (greatest change in RCS pressure) which was examined is that of starting an RC pump with the steam generators filled to % with F feedwater.
Since all other cases analysed in the references are bounded by this case, an analysis of the surge volume and pressurizer pressure change before and after start of the R C pump was made using the present operating conditions and limits at the Crystal River plant.
As in the loss of decay heat removal analysis, initial conditions and assumptions for this transient are unlikely but serve to illustrate that the effect of energy absorption from hot OTSG secondary water on the pressurizer pressure after start of an R C pump ist
- 2. is of the same magnitude as previously analyzed using present operating conditions and limits at the Crystal River plant.
The following facts apply to the analysis:
l l
l Page 32 of 68 l l
. .._ - - - - - - - - - - - d
$1 1176431 02 With the assumptions and initial conditions as stated above, the increase in '
primary pressure was calculated from the real gas law using appropriate compressibility factors.
The not change in RC pressure is from psig to psig. Therefore, this event is self limiting below the 32 [FPY LTOP P/T limit of psig.
Page 33 of 68 l
I
t 51 1176431 02 I
Table 5.2 1 Inadvertent HPl Actuation Results '
?
Initial Initial Initial Final Number RCS RCS Pressurizer RCS l of HP! Temp. Pressure Level Pressure (3)
Trains (degF) (psig) (inches) (psig) l
, ....... l a I i
i l
j 4
i i
t
[
i r
l Note: (1 !
(2 Corresponds to the 21 EFPY PORV Setpoint from Section 7.
i (3 Pressure 10 minutes after event initiation. !
I t
I I
i t
f l
i i
?
Page 34 of 68 i i
t i
t
51 1176431 02 Figure 5.2 1 l
1 Page 35 of 68
i 51 1176431 02 I
\
Table 5.2 2 [
i Makeup Contro) Valve Fails Open Results l l
I Initial Initial Initial Final l RCS RCS Pressuriter RC$
Temp. Pressure (1) Level Pressure (2) l (degF) (psig) (inches) (psig)
I i
l I
i f
l I Note. (1) 1 (2) Pressure 10 minutes after event initiation. ,
I i t l
l
[
f i
i r
i I
l Page 36 of 68 i l !
l 6
e k
51 1176431 02 Figure 5.2 2 Page 37 of 68 i
i $1-1176431-92 !
i I I
!! N_ ..
I ,
i a
i, n# L d N g
= ' i) fa
- g i
i
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$a '
~ t i
o 5 p l-a ;
M !
1 $
./ -
-n .
l i
1 n - ,
a
- 4 I I I , i i i I ' - ' :. ;
i e -
St ed. esneseg ,
Page 38 of ( i
. . . _ _ . . - . . . _ - . . - - _ . . . . . - .. . - . _ - - - - - - . . _ . , - . . -- . . . , _ _ - - . - - _ - _ ~ . - - . . . - . - . . ~ . . - - . . -
51 1176431 02 5.3 RE$ TILTS The RCS pressure response for each cf the events is sumarized in Table 5.31.
Table 5.31 illustrates that once HPI is admihistratively precluded the makeup control valve failing open becomes the limiting event. with approximately a psi / min pressurization rate. Pressurizer heaters failing on and loss of decay heat removal are lesser LTOP threats, with pressurization rates of approximately and psi / min respectively. All other events are either precluded or self limiting below the 32 EFPY LTOP P/T limit of psig, described in section 6.
Page 39 of 68 l
i i
51.!!76431 02 l TABLE 5.3 1 LTOP Event Pressure Response (3),(4)
Initial RC Initial 10 min. Pressurization i level Temp. Pressure Pressure Rate ;
Transient (inches) (F) (psig) (psig) (pst/ min) :
...... ..... ........ ......... .............. i HP! actuation.jpump (1) (1)
Makeup V1v Open !
Makeup Viv Open )
i DHS Loss i R.C Pump Start (2) (2) i PZR Heaters On l
, Note: 1 This event will be adminstratively precluded. l 2 This peak pressure is less than the 32 EFPY LTOP limit of '
psig.
(3) Core flood tank discharge and nitrogen addition to the o pressurizer are precluded events.
(4) The 21 EFPY LTOP P/T limit is $55 psig. The 32 EFPY LTOP :
P/T limit is psig (see section 6). l l
l
\
i t,
l Page 40 of 68 l r I >
51 1176431 02 6.0 FRAttuitt MECHANICS EULUATION
6.1 INTRODUCTION
LTOP P/T limits have been developed for CR.3 based on a fracture mechanics evaluation at 21 [FPY and 32 1.8PY.
The resulting (TOP limits for 21 [FPY and 32 [FPY, are provided in Table 6.1 1 and illustrated in Figure 6.1 1.
)
Page 41 of 68 I
l
51 1176431 02 Table 6.1 1 CR-3 LTOP P/T Limitt at 21 and 32 EFPY 1
i i
1 c
{
Page 42 of 68 i
i
I n.u?6431-02 1
I Figure 4 01-1 CR.) 21 and 32 EFPY LTOP P/T Lleits :
i s
1 :
1 1
l i :
i I i .
i
- CR 3 21 EFPY LTOP P/T Limit i i
CR.3 32 EFPY LT0P P/T Limit m ;
i =
l c a
t -
I 2 .
C S -
w ;
k- . ,
l CR-315 EFPY Nomal g P/T Limit .
l' i
W e
- i i i i i l i
a 3 T. P.retur.. r ,,,, 3 ,, ,,
t
~e.-evv -,w,r-n-_-., ,,,-e--e-, , s . , , - ,m--~,,,se,-,--w,,-- .----em-e n- -n- - - em,,,
I !
i i
i 51 1176431 02 {
l l I 6.2 TECHNICAL APPROACH 6.2.1 Methodoloov - Linear Elastic Fracture Mechanics (LEFM) i The analysis which determined the CR 3 LTOP fracture mechanics I pressure / temperature limits at 21 and 32 EFPY was performed using linear !
l elastic fracture mechanics methods, with the following governing equation: ;
Allowable Pressure = (K K!T
- K!res)/(KIPunit x S.F.)
where, K = material toughness
! K = stress intensity factor due to thermal stress, IT l K = stress intensity factor due to residual stress.
!res KIPunit = stress intensity factor due to unit pressure.
S.F. = safety factor Allowable pressures versus reactor coolant system . temperature in the reactor vessel downcomer region were obtained and adjusted to RCS pressure at the hot leg tap The major elements in the governing equation -- the safety factor, the tem erature transient assumed to determine KIT, the method of calculating residua stresses (Ki res), and the use of K -- are discussed in the sections '
which follow.
6.2.1.1 Fracture -- Toughness As summarized in section 4.0, LTOP events at CR-3 are expected to be infrequent events, i.e. category C, or emergency events, therefore a non Appendix G frnture mechanics approach is justified.
6.2.1.2 Postulated Flaw Size The LTOP analysis was based on a conservatively assumed flaw, size, shape and orientation.
6.2.1.3 Determination of RT Values Page 44 of 68 E
i 51 1176431 02 l Tables 6.21 and 6.2 2 summarize the development of the CR 3 beltline RTNDT values for the controlling materials at 21 and 32 EFPY respectively, using Regulatory Guide 1.99, Revision 2 methodology. An RTNOT of degF was used for the nozzle and closure head evaluations, consistent with approved B&W methodology. ,
i 6.2.1.4 Safety Factor was utilized.
As defined in the governing equation, a safety factor of t
6.2.1.5 Heatup and Cooldown Temperature Transient i The heatup and cooldown temperature transient assumption used to develop the ,
CR-3 LTOP P/T limits is based on the maximum rates allowed per the CR 3 15 EFPY Technical Specification P/T limits for normal heatup and cooldown (Reference ,
5).
In addition to the analysis for heatup and cooldown transients, a fracture mechanics analysis is also made for the isothermal conditions over the full range of operating temperatures.
6.2.1.6 Residual Stresses Welding processes generate residual stresses within the welded zone of reactor vessel materials. Subsequent heat treatment reduces the severity of the residual stress levels although complete stress relief is not feasible. Since residual stresses may contribute to the overall stress state, their effects are sometimes included in fracture mechanics analysis. However, residual stresses are self equilibrating across the thickness, i.e. the introduction of cracks or flaws in a residual stress region tends to redistribute stresses and relieve the severity of the residual stresses.
Page 45 of 68 1
I
$1 1176431 02 :
Residual stresses were considered in the development of the CR 3 LTOP fracture !
mechanics P/T limit.
6.2.1.7 Cladding Effects Since the reactor vessel cladding material is stainless steel whereas the base metal material is carbon steel, differential thermal expansion between the two <
materials during heatup and cooldown operations may create additional stresses t at the clad / base metal interface. To evaluate this effect
! models were constructed as follows:
i i
I The results of the analyses are summarized in Table 6.2-3.
The effects of cladding in the development of the final LTOP limit curves are evaluated based on the results of this analysis.
Page 46 of 68 l
l
l 51 1176431 02 Table 6.2-1 CR 3 RTNOT Values at 21 EFPY, using RG 1,99, Rev 2 ;
l t
l 4
l l
l '
l 1
\
Page 47 of 68
i i
51 1176431 02 Table 6.2 2 9 CR 3 RTNOT Values at 32 EFPY, using RG 1.99, Rev 2 E
t 1
t f
1 l
l Page 48 of 68
51-1176431 02 Table 6.2 3: Results of Analysis 4
I i
Page 49 of 68 i
P i
51 1176431 02 i 6.3 BELTLINE ANALYSIS Using the methodology described in section 6.2, the allowable reactor vessel beltline pressures as a function of reactor coolant system (RCS) inlet temperature were determined for. the normal heatup, normal cooldown, and ;
isothermal conditions.
The composite results at 21 EFPY and 32 EFPY are provided in Table 6.1-1 and illustrated in Figure 6.1-1. The minimum allowable pressure results from the analysis, and occurs at a fluid temperature of degF.
6.4 CLOSURE HEAD ANALYSIS ,
The normal heatup and cooldown closure head Technical Specification P/T limits i were modified for this analysis to account for the following: ,
a) this LTOP analysis uses the K toughness curve instead of KIR curve used for Tech. Spec P/T curves.
b) this LTOP analysis uses a safety factor of instead of 2.0 used for Tech. Spec. P/T curves.
l The resulting LTOP closure head limit is given in Table 6.4-1.
From these results the influence of the closure head limits in the development of the final LTOP limit curves is determined.
1 l
l l
l Page 50 of 68
-,---m ----m e - .-...-w e-.. , ,, ,-e--- m e s v ,
i
. l t
I 51 1176431 02 L
Table 6.4 1 [
Closure Head LTOP P/T Limit I f
i t
RCS RCS Temperature Pressure (degF) (psig)
C l
l l
t Page 51 of 68
t 51 1176431 02 6.5 N0ZZLE CORNER CRACK ANALYSIS ,
A postulated nozzle crack was analyzed to see whether it is controlling during any part of the normal heatup and cooldown or during steady state conditions.
The results are summarized in Table 6.51. A comparison of these results with the beltline results of Table 6.1-1 shows that the nozzle is less limiting than ,
the beltline From these results the influence of the nozzle corner crack in the development of the final LTOP limit curves is determined.
2 i
1 i
l l
l i
i l
Page 52 of 68 I
l l
51-1176431 02 Table 6.51 RV Nozzle LTOP P/T Limit RCS RCS Temperature Pressure (degF) (psig) l
, f l 1 l
l l
l l
l I
1 l
( l Page 53 of 68 ;
i
.I
l 51-1176431 02 ,
6.6
SUMMARY
AND CONCLUSIONS From the above discussion, the CR 3 LTOP P/T limits are as illustrated by Figure 6.1-1.
Pagi 54 of 68
___ _____ _ ~ . .
= _ ___
i 51 1176431 02 7.0 LTOP SYSTEM DEVELOPMENT 7.1 TWENTY-ONE EFPY CR 3 LTOP SYSTEM i Section 2.1 summarizes the CR 3 21 EFPY LTOP system. This section provides i additional detail.
7.1,1 Ten Minute Ooerator Action Protection 7.1.1.1 LTOP initial conditions (P/T)
In addition, the fracture mechanict analysis which developed the CR 3 21 EFPY LTOP P/T Limit used the maximum heatup and cooldown rates as limited by the CR-315 EFPY Technical Specification, ar described in section 6. .
In addition, RCS heatup and cooldown rates (F/hr) must be maintained in accordance with normal heatup and cooldown Technical Specification P/T limits-for 15 EFPY (Reference 5).
7.1.1.2 The CR 3 LTOP Pressure / Temperature Limit Table 6.1-1 lists and Figures 6.1-1 and 7.1-1 illustrate the CR 3 21 EFPY '
reactor vessel LTOP P/T fracture mechanics limit which must be protected by the LTOP system, including ten minute operator action criteria. This limit,
' developed as described in Section 6, defines the maximum allowable pressure as l a function of RCS temperature allowed during an LTOP event based on the LTOP t
fracture mechanics criteria.
7.1.1.3 Enable Temperature The CR-3 RCS code safety valves are set at 2500 psig and are fully open at 2575 psig, providing full relief capacity for any postulated LTOP event. The CR-3 Technical Specifications require that the code safeties be operable at RCS temperatures greater than 280 F, therefore, the code safeties are available to provide LTOP protection above 280 F.
System pressure overshoot, that is, increase of primary coolant pressure after pressure reaches the setpoint value, does not occur due to the rapid action of the code safeties and the relatively slow rates of pressure increase due to the steam bubble in the pressurizer l The CR 3 21 EFPY LTOP P/T fracture mechanics limit prohibits RCS pressures of Page 55 of 68
i !
l l 51 1176431 02 ,
l l 2575 or greater below an RCS temperature of 283 F, thus, the temperature below ,
which the LTOP system must be enabled, i.e. operable, is 283 F at 21 EFPY.
7.1.1.4 Inadvertent HP1 Actuation i Inadvertent HPI actuation must be precluded below 283 F at 21 EFPY.
Based on the HPI pressurization analyses documented in section 5, summarized in Table 5.21 and Figure 5.21, it is concluded that inadvertent HPI actuation must be precluded at all temperatures below the enable temperature because even with a inch initial pressurizer level and psig initial pressure, a 10 minute operator response time is not available for two train actuation and, to j
obtain a 10 minute operator response time for inadvertent single train i actuation, for all temperatures below 283 F. very restrictive pressurizer
- levels would be required.
7.1.1.5 MVCV Fails Open As documented in section 5, because inadvertent HPI actuation and inadvertent core flood tank discharge will be precluded as events below 283 F, the event:
makeup control valve fails open becomes the limiting event for determining the I
administrative requirements to provide a ten minute operator response time, with approximately a >si/ min pressurization rate. Pressurizer heaters failing on and loss of cecay heat removal are lesser LTOP events, with pressurization rates of approximately and psi / min respectively (see section 5.3). All other events are either incredible or self limiting below the 21 EFPY LTOP P/T limit of 555 psig, described in section 6.
Based on the MUCV pressurization analyses documented in section 5, summarized in Table 5.2-2 and Figure 5.2-2, it is concluded that limiting pressurizer level to less than or equal to 220 inches will provide 10 minute operator response time below the LTOP enable temperature at 21 EFPY.
i 7.1.2 PORV Protection 7.1.2.1 Setpoint Table 6.1-1 lists and Figures 6.1-1 and 7.1-1 illustrates the CR 3 21 EFPY reactor vessel LTOP P/T fracture mechanics limit which must be protected by the LTOP system, including the PORV. This limit, developed as described in Section 6.0, defines the maximum allowable pressure as a function of RCS temperature l allowed during an LTOP event. A dual setpoint PORV would have to have its lower setpoint based on protecting the fracture mechanics limit over its entire range. Since the minimum value of the LTOP fracture mechanics limit for CR-3 at 21 EFPY is 555 psig, the PORV setpoint required at 21 EFPY is 555 psig.
System pressure overshoot, that is, increase of primary coolant pressure after pressure reaches the setpoint value, does not occur due to the rapid action of the PORY and the relatively slow rates of pressure increase due to the steam bubble in the pressurizer.
Page 56 of 68
51 1176431 02 7.1.2.2 Enable Temperature ,
The basis for the CR 3 21 EFPY LTOP enable temperature of 283 F is discussed in ,
section 7.1.1.3. .
7.1.2.3 PORV Capacity As discussed in section 7.1.1.4, inadvertent HPI actuation inadvertent core flood tank discharge will be precluded as LTOP events below 283 F at 21 EFPY. '
Therefore, the limiting LTOP event with respect to required PORV capacity becomes the MUCV failing open event, which results in pressurizer insurge rates of less than gpm. ,
As documented in the CR 3 FSAR, Table 4-1, the steam capacity of the PORV at 550 psig is 25,985 lb/hr, equivalent to a liquid insurge volume rate into the pressurizer of 2650 gpm. The PORY capacity increases as the RCS pressure increases.
Therefore, the PORV capacity significantly exceeds the capacity required for the limiting LTOP event the MUCV failing open event. '
7.1.3 Ooerability 1
There are no o >erability problems with the 21 EFPY CR-3 LTOP system, i.e.
heatup and coolcown operations will not be adversely impacted.
r l
l Page 57 of 68
i i
51-1176431 02
( ,
l Table 7.1-1 i
i CR 3 15 EFPY APPENDIX G P/T LIMIT 1 i
! MAX 1 MUM ALLOWA8LE PRESSURE VS TEMPERATURE DURING HEATUP AND C00LDOWN ,
i l
(Reference 5) ,
Maximum Maximum Maximum
. Pressure Pressure Composite l RCS During During Pressure Temperature Heatup Cooldown Heatup and Cooldown l (degF) (psig) (psig) (psig) 303 I 60 303 285 t 126 303 285 303 150 319 285 319 l
175 354 337 354 200 409 412 412 225 492 516 516 L
250 611 674 674 280 825 977 977 l 320 1242 1486 1486 375 2275 <2275 2275 l
1 l
l l
l l
l i
Page 58 of 68
/
---w, , ., , , ,---------w - - - - - - - - - - - - - - -
u -n76' u-02 4 I Figure 7.1 1 CR.3 21 and 32 EFPY LTOP P/T Limits 1 ,
I l =
+
. CR.3 21 EFPY LT0P P/T Limit ,
CR.3 32 EFPY LT0P P/T Limit m a
g -
a
? -
i G -
C 8 -
=
E i 6
CR 3 15 EFPY Normal Q P/T Limit l
1 1
W 1 i i i i I RCS Temperature. 'F Page 59 of 68
. . , . . , , , . - . .,.w-.Wr.- ,.,,-.,www.-,ew,,.,w-* yp,ey,m.y -wwe--
4 51 1176431 02 7.2 THIRTY TWO EFPY CR-3 LTOP SYSTEM Section 2.2 summarizes the CR 3 32 CFPY LTOP system. This section provides additional detail.
7.2.1 Ten Minute Ooerator Action Protection 7.2.1.1 LTOP initial conditions (P/T)
In addition, the fracture mechanics analysis which developed the CR 3 32 EFPY LTOP P/T Limit used the maximum heatup and cooldown rates as limited by the CR-315 EFPY Technical Specification, as described in section 6.
In addition, RCS heatup and cooldown rates F/hr must be maintained in accordance with normal heatup and cooldown Tech (nical) Specification P/T limits for 15 EFPY (Reference 5).
7.2.1.2 The CR-3 LTOP Pressure / Temperature Limit Table 6.1 1 lists and Figures 6.1-1 and 7.1-1 illustrate the CR-3 32 EFPY reactor vessel LTOP P/T fracture mechanics limit which must be protected by the LTOP system, including ten minute operator action criteria. This limit, developed as described in Section 6, defines the maximum allowable pressure as a function of RCS temperature allowed during an LTOP event based on the LTOP fracture mechanics criteria.
7.2.1.3 Enable Temperature j The CR-3 RCS code safety valves are set at 2500 psig and are fully open at 2575 psig, providing full relief capacity for any postulated LTOP event. The CR 3 Technical Specifications require that the code safeties be operable at RCS temperatures greater than 280 F, therefore, the code safeties are available to I provide LTOP protection above 280 F. i System pressure overshoot, that is, increase of primary coolant pressure after ,
pressure reaches the setpoint value, does not occur due to the rapid action of I the code safeties and the relatively slow rates of pressure increase due to the steam bubble in the pressurizer The CR 3 32 EFPY LTOP P/T fracture mechanics limit prohibits RCS pressures of 2575 or greater below an RCS temperature of F, thus, the temperature below which the LTOP system must be enabled, i.e. operable, is F at 32 EFPY.
1 Page 60 of 68 l
i l
1 I
51 1176431-02 l 7.2.1.4 Inadvertent HPI Actuation Inadvertent HPI actuation must be precluded below 310 F at 32 EFPY, Based on the HPI pressurization analyses documented in section 5, summarized in Table 5.21 and Figure 5.21, it is concluded that inadvertent HPl actuation must be precluded at all temperatures below the enable temperature because even ,
with a inch initial pressurizer level, a 10 minute operator response time is '
not available for two train actuation and, to obtain a 10 minute operator response time for inadvertent single train actuation, for all temperatures below F, very restrictive pressurizer levels would be required.
7.2.1.5 MUCV Fatis Open i
As documented in section 5, because inadvertent HP! actuation and inadvertent I core flood tank discharge will be precluded as events below F, the event: '
makeup control valve fails open becomes the limiting event for determining the i administrative requirements to provide a ten minute operator response time, ,
with approximately a asi/ min pressurization rate. Pressuriter heaters '
failing on and loss of cecay heat removal are lesser LTOP events, with pressurization rates of approximately and psi / min respectively see section 5.3). All other events are either incredible or self limiting b(elow the 32 EFPY LTOP P/T limit of psig, described in section 6.
l Based on the MUCV pressurization analyses documented in section 5, summarized l in Table 5.2-2 and Figure 5.2-2, it is concluded that limiting pressurizer inches will provide 10 minute operator level to less than or equal to >
response time below the LTOP enable temperature at 32 EFPY.
7.2.2 PORY Protection 7.2.2.1 Setpoint Table 6.1-1 lists and Figures 6.1-1 and 7.1-1 illustrates the CR-3 32 EFPY reactor vessel LTOP P/T fracture mechanics limit which must be protected by the LTOP system, including the PORV. This limit, developed as described in Section
- 6.0, defines the maximum allowable pressure as a function of RCS temperature allowed during an LTOP event. A dual setpoint PORV would have to have its lower setpoint based on protecting the fracture mechanics limit over its entire range.
Since the minimum value of the LTOP fracture mechanics limit for CR 3 at 32 EFPY is psig, the PORV setpoint required at 32 EFPY is psig.
System pr. essure overshoot, that is, increase of primary coolant pressure after pressure reaches the setpoint value, does not occur due to the rapid action of the PORV and the relatively slow rates of pressure increase due to the steam bubble in the pressurizer.
7.2.2.2 Enable Temperature l
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The basis for the CR 3 32 EFPY LTOP enable temperature of F is discussed in section 7.2.1.3. ;
7.2.2.3 PORV Capacity As discussed in section 7.2.1.4, inadvertent HPI actuation and inadvertent core I flood tank discharge will be precluded as LTOP events below F at 32 EFPY. .
Therefore, the limiting LTOP event with respect to required PORY capacity becomes the MUCV failing open event, which results in pressurizer insurge rates of less than gpm.
As documented in the CR 3 FSAR, Table 41, the steam capacity of the PORV at I 550 psig is 25,985 lb/hr, equivalent to a liquid insurge volume rate into the pressurizer of 2650 gpm.
Therefore, the PORV capacity significantly exceeds the capacity required for i the limiting LTOP event - the MUCV failing open event. !
7.2.3 comrability There should be no operability problems with the 32 EFPY CR-3 LTOP system, i.e.
heatup and cooldown operations will not be adversely impacted. If the limitation on pressure and temperature which results from reducing the PORV setpoint to psig at F is a problem then additional setpoints for the <
PORY would be one possible solution.
7.3 INSTRUMENTATION ERROR The CR 3 LTOP system limits sumarized in 7.1 and 7.2 above, e.g. enable
! temperature, PORV setpoint, and pressurizer level, do not directly take into I account instrumentation errors. This is consistent with current LTOP practice, 'i and is disrusted in detail in section 2.3.
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8.0 REFERENCES
1.) NRC Generic Letter 8811. "NRC Position on Radiation Embrittlement of ,
Reactor Vessel Materials and its Impact on Plant Operations", July,1988. I 1
2.) NRC letter: J.F. Stolz to FPC, " Verification for compliance with Appendix G ,
Pressure Temperature Limits During Startup and Shutdown", October 1, 1976.
3.) NRC, Reaulatory Guide 1.99. Revision 2. " Radiation Damage to Reactor Vessel ,
Material", May 1988.
4.) Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No. 21 to License No. OPR 72, Florida Power Corporation Et A1, '
Crystal River 3 Nuclear Generating Plant, Docket No. 50 302, Dated July 3, 1979.
5.) B&W Document: BAW 2091 (77-2091-00), " Pressure Temperature Limits for 15 ,
EFPY For the Crystal River 3 Nuclear Plant", August 1989. '
6.) NUREG/CR 5186, "Value/ Impact Analysis of Generic Issue 94, ' Additional Low Temperature Overpressure Protection for Light Water Reactors,'" November 1988.
7.) AE00 Case Study C401, " Low Temperature Overpressure Events at Turkey Point Unit 4," USNRC, March 1984.
References 2 and 4 are controlled and retrievable through the FPC document control system and can be used as acceptable references.
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51 1176431 02 APPENDIX A A SUMARY DESCRIPTION OF THE EXISTING CR-3 LTOP SYSTEM AND A COMPARISON VERSUS DESIGN CRITERIA 1
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In the late 1970's agreement was reached on a CR 3 LTOP "Sistem', initially developed based on the reactor vessel 5 EFPY Technical Specification P/T limits, which relied on operator action as the first line of defense tnd; as a backup, a Low Pressure PORY Setpoint (550 psig) below 280 F. Following is a summary of how each of the design criteria were met by the system from an NRC perspective. The text in quotation marks is from the NRC SER (Reference 4):
1.) Ooerator Action: No credit can be taken for operator action for ten minutes after the operator is aware of a transient.
"FPC has provided an evaluation of:
- a. Erroneous actuation of the High Pressure Injection (HPI) system,
- b. Erroneous open%( of the core flood tank discharge valve,
- c. Erroneous addittu of nitrogen to the pressurizer,
- d. Makeup control valve makeup to the Reactor Coolant System (RCS) fails full open,
- e. All pressurizer heaters erroneously energized. I
- f. Temporary loss of the Decay Heat Removfal (DHR) System's capability to remove decay heat from the RCS.
- g. Thermal expansion of RCS after starting a reactor coolant pump (RCP) I due to stored thermal energy in the steam generator.
We accept these analyses. These analyses show that, in the event of a i postulated mass addition, actuation of the relief valve will limit RCS pressures to the relief valve set)oint and hence below Appendix G limits.
Should the relief valve fail closec, or actuation circuitry fail, the system pressure would continue to increase. With the exception of postulated high pressure safety injection, the nitrogen bubble in the pressurizer will provide at least ten minutes, and in some cases substantially longer time, for operator action."
"For all postulated heat addition transients and for all mass additions other thas- inadvertent high pressure safety injection, the CR 3 OMS meets single failure and operator action criteria." -
"To preclude HPI in the temperature range of 280 F to 150 F, the licensee i must " rack out" the HPI isolation valve circuit breakers with the valves in their normally closed position. HPI pump operation within this range is still necessary for makeup and RCP seal flow."
" Credit for administrative controls is consistent with past NRC staff actions. We have permitted a manually enabled system, credit for blocking l Page 66 of 68 l
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51 1176431 02 safety injection actuation signal, credit for successfully blocking one of two high pressure safety injection trains, and credit for blocking acculmulator injection."
" System pressure overshoot, that is, increue of primary coolant pressure after pressure reaches the low setpoint value, does not occur on B&W NSSS due to the rapid action of the electrical PORV and the relatively slow rates of pressure increase due to the nitrogen blanket in the pressurizer" 2.) Sinole Failure: The system must be designed to relieve the pressure transient given a single failure in addition to the failure that initiated the pressure transient.
"The OMS consists of active and passive subsystems."
"The active subsystem is simply the modification of the actuation circuitry of the existing electrical PORV to provide dual setpoints, a normal operation setpoint of 2450 psig and a low pressure setpoint of 550 psig.
The low pressure setpoint is employed when the RCS is below 280 F. This system is manually enabled. An alars will function should the operator fail 1 to enable the system. An Alarm has also been installed to monitor the position of the pressurizer relief valve, RC V2."
"The passive subsystem consists of the introduction of a nitrogen blanket at the top of the pressurizer. The reactor is operated during heatup and cooldown with a steam or nitrogen bubble. The bubble functions as a mechanical damper. This subsystem is part of the original B&W design."
"The CR-3 OMS is both redundant and functionally diverse " (Note that !
diversity although desirable was never an NRC staff design criteria.)"
"For all postulated heat addition transients and for all mass additions other than inadvertent high pressure safety injection, the CR-3 OMS meets single failure and operator action criteria."
"To preclude HPI in the temperature range of 280 F to 150 F, the licensee must " rack out" the HPI isolation valve circuit breakers with the valves in their normally closed position. HPI pump operation within this range is stt11 necessary for makeup and RCP seal flow."
" Credit for administrative controls is consistent with past NRC staff i actions. We have permitted a manually enabled system, credit for blocking safety injection actuation signal, credit for successfully blocking one of two high pressure safety injection trains, and credit for blocking acculmu' ator injection."
3 3.) Testability: The system must be testable on a periodic basis consistent with the system's employment.
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"The system is testable and is to be tested prior to use. The PORV is to be tested each shutdown."
4.) Seismic and IME 279 Criteria! Ideally, the system should meet seismic Category I and IEJE 279 criteria. The basic objective is that the systen should not be vulnerable to a common failure that would both initiate a pressure transient and disable the overpressure mitigating system. Such events as loss of instrument air and loss of offisite power must be considered.
"The CR-3 OMS is tolerant of seismic events. FPC has performed analyses for the pilot assembly connection pipe assuming seismic motion of 3.0 horizontal 1 and 3.0 vertical. The actual valve meets Class I requirements. Testing '
with simulated seismic loadings has not been performed. This was not a requirement at the time the plant was designed and constructed. Even if it is assumed that the valve, connection pipe, or actuation circuitry, failed due to a seismic event, the nitrogen blanket in the pressurizer would-provide protection for postulated low temperature overpressure events."
"The system does not strictly meet IEEE 279 criteria. The basic objective i of this criterion, prevention of common mode failure, is met by virtue of the subsystem diversity."
"The CR-3 OMS is both redundant and functionally diverse." (Note that diversity although desirable was never an NRC staff design criteria.)
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