3F1097-32, Application for Amend to License DPR-72,addressing Methodology for post-LOCA Boron Precipitation Prevention. Revised Framatome Technologies,Inc Document 51-5000519-02, Boron Dilution by RCS Hot Leg Injection, Encl
| ML20212F312 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/31/1997 |
| From: | Cowan J FLORIDA POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20212F319 | List: |
| References | |
| 3F1097-32, NUDOCS 9711040255 | |
| Download: ML20212F312 (37) | |
Text
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?AWpr Seefset he, to MS October 31~ 1997 3F1097 32 U.S. Nuclear Regu!atory Commission Attn: Document Control Desk
' Washington, DC 20555 C')01
Subject:
License Amendment Request #223, Revision 0 Post LOCA Boron Precipitation Prevention
References:
1.
FPC to NRC letter dated September 12,1997, " Post LOCA Boron Precipitation Mitigation Plan" 13F0997 28]
2.
FPC to NRC letter dated October 31,1997, " License Amendment Request
- 220, Revision 1; Revision of Operating License Condition 2.C.(5) (TAC No. 99128)" 13F1097-08) 3.
NRC to FPC letter dated August 23,1996, " Crystal River Unit 3 Integrated Performance Assessment Process (IPAP) Final Assessment Report (NRC Inspection Report No. 50 302/96 201)" 13N089612) f0
Dear Sir:
3 Florida Power Corporation (FPC) hereby submits a request for an an,endment to its Facility
( l Operating License No. DPR 72 for Crystal River Unit 3 (CR 3).
The attached License Amentirnent Request (LAR) #223, Revision 0, addresses the methodology for post-loss of coolant sccident (LOCA) boron precipitation prevention for CR-3. FPC concludes that the change in boron precipitation prevention methodology is safe; however, it does represent an unreviewed safety question (USO) in that it represents a change in the previously NRC-approved preventative methodologies; e.g., incorporating credit for hot leg nozzle gaps into EE-its design and licensing basis as a qualified passive method ior boron precipitation mitigation y
under certain unlikely scenarios.
Therefore, this action requires Nuclear Regulatory gjj' Commission (NRC) approval, g
E
' As discussed in more detail herein, this LAR addresses some post-LOCA boron precipitation g-prevention additional methodologies that have been previously approved by the NRC for g*
Babcock & Wilenx (B&W) type plants and methodolegies for which FPC is seeking approval E
through this LAR. The Post LOCA Boron Precipitation Prevention Plan provided in Attachment B of this submittal reflects additional consideration oy FPC of its boron precipitation mitigation options oreviousiv discuss 1d in Rafarance 1. Attachment B provides the current FPC plan on F# 38M Pd86
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U.S. Nucle:r R:gulatory Commission 3F1097-32 Page 2 of 5
- this matter and supersedes Reference 1. The information in Attachment B forms the basis for requesting this LAR as well as LAR #220 (Reference 2).
Additionally, this submittal responds to the restart issue identified in the IPAP report regarding boron precipitation following design basis LOCAs (Reference 3). The post LOCA boron precipitation prevention methodologies provided in Attachment B provide information to support closure of the NRC restart issue.
FPC has re evaluated previously relied upon boron prr,cipitation prevention methods and modified its methodology to address generic issues for B&W type plants that have been previously communicated to the NRC. Some of these modifications are not part of the original CR 3 licensing basis. This methodology,in part,'1 reflected in a decision matrix (Attachment B, Figure 1) that addresses boron dilution methods available, including reliance on flow through the hot leg nozzle gaps in some limited cases, and the use of compensatory measures as necessary. These systems and procedures,in combination, win effectively prevent post-LOCA boron precipitation. However, the change in the methodology revises the licensing basis for CR 3 and, therefore, requires approval by the NRC Although this LAR is necessary in response to the USO and in accordance with 10 CFR 50.59, no actual changes to the Technical Specifications are necessary or requested.
The methodology presented in this submittal will ensure that CR 3 can effectively preven +
boron precipitation in the long term cooling phase following a design basis LOCA.
Additior ally, FPC is working as part of the B&W Owners Group (B&WOG) to resolve the generic issues with the NRC.
A summary of the attachments to LAR #223 is listed hlow:
Attachment A - License Amendment Reauest #223. Revision O. Post-LOCA Boron Precioitation Prevention This attachment describes the proposed changes along with the 10 CFR 50.92(c) evaluation and conclusion that the proposed changes do not involve a significant hazard.
Attachment B - Post-LOCA Boron Precioitation Prevention Plan This attachment provides a discussion of the methodology for post-LOCA boron precipitation prevention for CR-3.
Attachment C - Acronyme and Abbreviations This attachment lists the acronyms and abbreviations used in the submittal.
Attachrnent D FTl Document 51-5000519-03 This attachment provides the current revision to Framatome Technologies incorporated (FTI) Document 51-5000519-03. The FTl document provides analyses to support the post-LOCA boron precipitation prevention plan.
In Reference 1, FPC discussed our plan to submit a request for approval of an additional active method that will provide defense in depth for post-LOCA boron precipitation by September 30,1997. This method would provide hot leg injection via reverse flow through portions of the low pressure injection system. Reliance on this method of boron dilution is not required
U.S. Nucl%r R:gulattry Commission 3F1097 32.
Page 3 of 5 prior to restart of CR 3 and is still under evaluation by FPC. Current plans are to provide this submittal by November 14,1997.
FPC requests that the NRC provide review and approval of this LAR to support restart of C8-
- 3. If you have any questions regarding this submittal, please contact Mr. David Kunsemiller, Manager, Nuclear Licensing at (352) 563 4566.
Sincerely,
$hO John Paul Cowan Vice President Nuclear Production JPC/dah Attachments cc: Regional Administrator, Region ll NRR Project Manager Senior Resident inspector
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'U.S; NucleCr RegulCtory_ Commission 3F1097 32 Page 4 of 5 List of Commitments Letter Reference Commitment Due Date i
Cover Letter, in Reference 1, FPC discussed our plan to November 14,1997 Page 2 submit a request for approval of an additional active method that will provide defense in depth for post-LOCA boron precipitation by September 30, 1997. This' method would provide hot leg injection via reverse flow through portions of the low pressure injection system. Reliance on this method of boron dilution is not required prior to restart of CR 3 and 13 still under evaluation by FPC. Current plans are to provide this submittal.
Attachment B, Operator initiated cooldown and November 30,1997 Page 7 depressurization would be used to decrease RCS pressure to the point where the drop line could be opened. This would lnclude utilizing the emergency feedwater pumps, diesel generator backed auxiliary feedwater pump (FWP-7), ADVs, TBVs, PORV, and - HPVs.
These multiple options provide significant defense in depth for assuring that the period that the drop line cannot be utilized for boron precipitation prevention is minimal. FPC will evaluate the ability of various secondary side cooldown and primary side depressurization methods to demonstrate the ability of the plant to rapidly reach an RCS pressure which will allow the drop line to be opened.
- U.S. Nuclear Regulatory Commis:'on
-3F1097 32 Page 5 of 5 -
- STATE OF FLORIDA.
COUNTY OF CITRUS John Paul.Cowan states that he is the Vice President, Nuclear Production for Florida Power Corporation; that he is authorized on the part of said company to sign and file with the Nuclear Regulatory Commission the inf ormation attached hereto; and that all such statements made and matters set forth therein are true and correct to the best of his knowledge, information, and belief.
kfM John Paul Cowan Vice President Nuclear Production Sworn to and subscribed before me this 34. day of _P clo 6e r
,1997,by John Paul Cowan.
Signature of Notary Public State of Florida S?
USA ANN MCBRIDE
, Notary PuNic.Stateof Florlds Wy Comm. Enp Oct. 25, l#9 Comm. No. CC 505458.
L/$x ksp p, h7) f,s'l.
q (Print, type, or stamp Commissioned Name of Notary Public)
Produced Personally [
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Known Identification
U. S. Nucinar Regulatory Commission Attachment A 3F1097 32 Page 1 of 4 ATTACHMENT A LICENSE AMENDMENT REQUEST (LAR) #223, REVIGION O POST LOCA BORON PRECIPITATION PREVENTION LICENSE DOCUMENT INVOLVED:
Crystal River Unit 3 (CR 3) License EUMMARY OF CHANGES:
The original licensing basis for CR 3 relied upon passive design features, systems, ano procedures which credited one passive and two active methods to prevent post-loss of coolant accident (LOCA) boron precipitation. Reactor vessel vent valve (RVVV) overflow (passive method) was credited for the first 40 days. FPC also implemented procedures that controlled two active boron precipitation prevention methods: 1) recirculation via the decay heat system drop line to the Reactor Building (RB) Emergency Core Cooling System (ECCS) sump IDL-RB Sumpl, and 2) hot tog injection via the Auxiliary Pressurizer Spray (APS).
Although not credited in the originallicensing basis, CH 3 has another inherent design feature that provides a passive method of preventing post LOCA boron precipitation via natural circulation; i.e., through the gaps between the hot leg nozzles and the reactor vessel (RV) internals.
The effectiveness of the hot leg nozzle gap method for post-LCCA boron precipitation prevention has been previously accepted by the NRC as a generic backup method (Reference 16). For certain accident scenarios, this LAR recognizes the gaps as the primary prevention method.
Nonconforming conditions have been identified regarding the effectiveness of the RVVV overflow and the reliance on natural circulation through the hot leg nozzle gaps during certain small break LOCAs (SBLOCA). These conditions have been cominunicated to the NRC by Babcock & Wilcox (B&W) and the B&W Owners Group (B&WOG) as generic issues with B&W type plants.
Accordingly, FPC has re-evaluated its boron precipitation prevention methods and modified its methodology. Some of these methodology modifications are not part of the original licensing basis. The revised methodology relies upon a decision matrix to outline boron dilution methods that are available, reduce reliance on flow through the hot leg nozzle gaps es a passive design ieature, and direct the use of compensatcry measures as necessary.
These systems and prncedures, in combination, will effectively prevent post-LOCA boron precipitation.
FPC concludes that the change in boron precipitation prevention methodology is safe. An unreviewed safety question (USO) exists since FPC intends to incorporate credit for hot leg nozzle gaps into its design and licensing basis as a primary passive method for boron precipitation mitigation under certain unlikely scenarios. Therefore, this action requires Nuclear Regulatory Commission (NRC) approval. Although this LAR is necessary in response to the USO and in accordance with 10 CFR 50.59, no actual changes to the Technical Specilications are necessary or requested.
RGson for Reouest As a result of generic industry issues regarding post-LOCA boron precipitatior'. FPC has identified methods necessary to resolve these concerns.
These methods require an
U. S. Nuclxr R:gulatory Commission Attachment A 3F1097 32 Page 2 of 4 amendment to the FSAR to change the CR 3 licensing basis for post LOCA boron precipitation. This LAR requests approval for the use of the hot leg nozzle gap method for post LOCA boron precipitation, approval of the new methodology described herein, and supersedes the information provided in Reference 34.
Justification for Recuelt FPC has re-evaluated generic and CR-3 plant specific methods to prevent post LOCA boron precipitation and ha3 developed an enhanced methodology. The results of this re-evaluation and the results of supporting analyses are provided in Attachment 8. FPC concludes that the boron precipitation mi'gation methodology provided in Attachment B will offectively prevent post-LOCA boron ciecipitation. In summary, CR 3 has the following passive and active methods available to ensure the boron solubility limits in the RV are not reached following a design basis LOCA:
Passive Methods Active ifyds RVVV Overflow Drop Line to E Sump Hot Leg Nozzle Gaps Auxiliary Pressurizer Spray CR 3 Post-LOCA boron precipitation prevention methods will be coordinated in the Technical Support Center (TSC) using procedures that implement a decision matrix (see Figwe 1 of Attachment B). This matrix assesses the effectiveness of the passive design features (RVVV overflow and hot leg nozzle gaps) following a LOCA by monitoring the boron concentration in the RB Sump as anindication of boron concentrationin the RV. If conditions exist whereby boron concentration in the RV is a concern or RB Sump boron sampling is not available to assess the passive design features, the decision matrix will direct the initiation of available active methods for boron dilution.
Reliance on hot leg nozzle gaps for post-LOCA boron dilution has been accepted by the NRC as a generic backup method for boron precipitation provention. Althougli specific NRC approval of this application at CR 3 has not been requested nor granted, the NRC did recognize during the Integrated "orformance Assessment Process (IPAP) that the gap method (backup)is acceptable in the interim until FPC obthins such approvals. Attachment B provides the results of analyses that include CR 3 specific inputs regarding the offectiveness of the hot leg nozzle gap method. FPC requests NRC review and approval of this method as accepta' ole for primary use by CR 3 in combination with other passive and active methods listed above and the monitoring and decision making methodology described herein to prevent post-LOCA boron precipitation as part of this LAR.
Single failure protection for post-LOCA boron precipitation prevention methods is provided in most cases with the comt. nation of the passive design methods (RVVV overflow and hot leg nozzle gaps) with DL-RB Sump and/or APS. (DL-RB Sump and APS in combination are not single f ailure-proof since some valves in either flow path share a common motor control center for their power supply.) For most cases where RVVV overflow is not postulated to occur, the hot leg nozzle gap method with DL RB Sump and/or APS provide for single failure protection.
In a class of breaks of low probability (i.e., SBLOCA between 0.05 ft and 0.17 f t' located 8
below the pipe centerline on the reactor coolant pump discharge line - the probability of such
U. S. Nucl:ar R:gulatory Commission Attachment A
- 3F1097 32 Page 3 of 4 a break is less than 2.5 x 104 per year), the DL-RB Sump may not be available due to limitations on the RB Sump screens, in such cases RB Sump boron concentration monitoring is used as an input to the decision making process in the TSC. Other data and analyses described in Attachment B are used to determine the strategy and procedures to employ for specific conditions. Prior to the time that APS would be offectivo, hot leg nozzle gap flow will prevent post LOCA boron proc'pitation, in addition, CR-3 will utilize compensatory measures to reduco reactor coolant system (RCS) pressure to allow opening of the drop line. This will minimize the timo that an active method would be unavailable for this unlikely scenario.
FPC considers that those systems and procedures are adequate for ensuring that boron precipitation will not occur during post-LOCA long term cooling.
The information in Attachment B supports this conclusion.
NO SIGNIFICANT HAZARDS CONSIDERATION:
An evaluation of the proposed license amendment has boon performed in accordance with 10 CFR 50.91(a)(1) regarding significant hazards considerations, using the standards in 10 CFR 50.92(c).
1.
Does not involve a significant increase in the probability or consequences of an accident previously evaluated.
This LAR addresses the methodology that will be used following a design esis LOCA to ensure that the boron concentration in the reactor vessel does not reach the solubility limit during long term :ooling. This methodology utilizes systems and procedures that will be implemented following the previously evaluated accident (i.e., a LOCA). This proposed chango does not result in any modifications to the plant or change in a proceduro that is used prior to the postulated accident; thoiefore, those changes cannot result in an increase in the probability of an accident previously evaluated.
The methodology in this LAR will be implemented to ensure that boron precipitation, which may interfere with long term c0 Jing, w41 not occur following a design basis LOCA.
This methodology consists of systems and procedwes to provide additional defense in depth that for varying plant conditions will prevent the boron concentration in the RV from reaching the boron solubility limits. Evaluations are provided in this submittal that conclude that those methods are effectivo.
By ensuring that borca solubility limits are not reached in the RV, the analyses for the ECCS that ensure adequate core cooling following a design basis LOCA remain applicable.
Therefore, the consequences of accidents previously oveluated are not increased and offsite dose consequences remain a small fraction of 10 CFR Part 100 limits.
2.
Does not create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed changes reflect the methodology that will be used for CR 3 following a design basis accident to provent a boron precipitation event, which previously has been evaluated. Tha proposed LAR does not involvo any new accident initiators nor any modification to the plant nor a change in the oporation of the plant prior to the postulated
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U. S. Nucirr Regulatory Commission Attachm:nt A 3F1097 32 Page 4 of 4 design basis LOCA. Therefore, the possibility of a new or different kind of accident is not created.
3.
Does not involve a significant reduction in the margin of safety.
This change does not result in a reduction to the margin of safety for any accident. The proposed LAR ensures adequate defense in depth in that systems and procedures available following a design basis LOCA will prevent the precipitation of boron in the RV
- that could interfere with ECCS flow.
ENVIRONMENTAL IMPACT EVALUATION:
10 CFR 51.22(c)(9) provides critoria for and identification of lice sing and regulatory actions eligible for categorical exclusion from performing an environmental basessment. A proposed amendment to an operating license for a facility requires no environmental assessment if operation of the facility in accordance with the proposed amendment would not (1) involve a significant hazards consideration, (2) result in a significant change in the types or significant increase in the amounts of any effluents that may be released off site, or (3) result in a significant increase in individual or cumulative occupational radiation exposure. FPC has reviewed this license amendment and believes it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(c), no environmental impact statement or environmental assessment needs to ba prepared in connection with the issuance of the proposed license amendment.
==
Conclusions:==
(
1.
The proposed license amendment does not involve a significant hazard as described previously in the evaluation.
2.
The proposed LAR does not involve any modification to the plant nc. a change in the operation n.f the plant prior to the postulated design basis LOCA. The proposed license amendment does not result in a significant change in the types or a significant increase in the amounts of any effluonts that may be released off site and does not involve irreversible environmental consequences beyond those already associated with the Final Environmental Statement.
I 3.
The proposed license amendment doos not change the operation of the plant nor the i
l consequences of a design basis accident and, therefore, does not result in a significant increase to the individual or cumulative occupational radiation exposure.
1.
i
U. S. Nuclear Regulatory Commission Attachm:nt B 3F1097 32 Page 1 of 22 ATTACHMENT B POST-LOCA BORON PRECIPITATION PREVENTION PLAN 1.
Purpose This attachment provides a discuss ~on of the systems and procedures that will be used for preventing post-loss of coolant accident (LOCA) boron precipitation in the reactor vessel (RV) for Crystal River Unit 3 (CR 3). This attachment provides additional detail and the results of analyses to support License Amendment Request (LAR) #220 and LAR #223. This submittal supersedes the informatior, provided in Reference 34. Applicable correspondence related to post LOCA boren precipitation is listed in the Reference section.
11.
CR 3 LICENSING BASIS /BACKGilOUND Original Licensing Basis The originallicensing basis for CR 3 regarding post LOCA boron precipitation control consisted of design features and procedures to prevent boron precipitation. The following boron dilution methods were credited:
Reactor vessel vent valves (RVVV) overflow, which would be effective for greater i
than 40 days After the RVVV overflow is no longer effective, Recirculation via the decay heat system drop line to the reactor building (RB) emergency core cooling system (ECCS) sump [DL-RB Sump]
Hot leg injection via the auxiliary pressurizer spray (APS)
The ability to prevent post-LOCA boron precipitation following a single failure was addressed.
The methods available, in combination, would be able to effectively prevent boron precipitation, in the event that a single active component failure does not allow operation of the DL-RB Sump method, the operator would have an attemative of selecting the APS method to provide boron dilution.
Licensing Basis - Background The original licensing basis is comprised of generic evaluations performed by Babcock &
Wilcox (B&W), specific analyses presented by FPC, and evaluations performed by the NRC.
As a result of these evaluations, CR-3 depended on RVVV overflow until such time as it would become ineffective (in excess of 40 days). Af ter this time, one of two active methods would be credited (DL-RB Sump and APS).
In Reference 6, FPC provided the methods and procedures that would be used at CR-3 for post LOCA boron precipitation prevention. Single failure was addressed in this submittal by the combination of the DL-RB Sump and APS methods, in Reference 9, the NRC approved
U. S. Nuclear R:gulatory Commissien Att chment B 3F1097 32-Page 2 of 22 i
the procedures and methods (DL RS Sump & APS) for CR 3 and use of these methods in combination to meet single failure criteria, 111.
ASSESAMhNLQElBEY10181t10NDIT10NS The following discussion summarizes generic issues regarding boron precip;tation that have been discussed by the B&W Owners Group (B&WOG) with the NRC.
RVVV Overflow Analysis 1
Subsequent to the originallicensing basis, B&W analyses in 1991 determined that the RVVV overflow would not be effective for breaks below the center line of the cold leg reactor coolant pump dhcharge piping. The analyses indicated that the post LOCA height of the two i
phase mixture in the RV may not be sufficient to provide flow of liquid through the RWVs.
The analysis also indicated that the APS flow of 40 gpm may not be adequate in and of itself j
to prevent boron concentration from reaching the solubility limit if the core boil of t rate was greater than APS flow, in response to this problem, the B&WOG provided analyses to the NRC that credited natural circulation in the RV through the hot leg no;zle gaps. As decided in a December 3,1992 meeting between the B&WOG and the NRC, and approved by the March 9,1933 NRC letter l
(Reference 16), the hot leg nozzle gap flow would provide an adequate backup to the primary l
recirculation flow path. At that time this issue was considered to be resolved.
)
SBLOCA Spectrum At the December 3,1992 meeting between B&WOG and the NRC, small break LOCAs (SBLOCA) were also discussed. Engineering judgment was used to separate the SBLOCA break spectrum into three groups (larger SBLOCAs, smallest SBLOCAs, and intermediate cold leg pemp di.= charge SRI OCAO ta r%menerMe tW the S8l.OCA boron concentration was bounded by the large break LOCA (LBLOCA) value without gap flow. The validity of these engineering judgments was questioned by the NRC in late June 1996 (Reference 19). The j
B&WOG Regulatory Responso Group (RRG) was activated to effectively address the NRC questions. An initial written response defining the issues and quantifying the safety significance was provided by letter from the RRG to the NRC on July 3,1996 (Reference 20).
Also, a RRG _and NRC meeting was held on July 9,1996 to provide new technical l
assessments adequate for justifying continued operation of the plants (Reference 21). As L
stated in the July 3,1996 letter and presented at the July 9,1996 meeting, it was identified l
that reliance on the hot leg nozzle gap flow may be necessary until the reactor coolant system (RCS) conditions allow another means of boron concentration control to be established.
Single Fallure Analyals Following the 1991 B&W determination that the RVVV overflow would not be effective for certain breaks and that the APS method may not be effective by itself for boron dilution, the I'
hot leg nozzle gap method was credited as an acceptable backup method. The combination of the DL RB Sump and hot leg nozzle gap method would provide protection for single f ailures.
This position was stated in Reference 22.
U. S. NucleCr Regulat@ry Commission Attachment B 3F1097 32 Page 3 of 22 Recently, a~ single failure has been identified that would prevent the use of both the DL-RB
' Sump and APS methods. Loss of the 3AB ES motor control center (MCC) would result in the loss of power to valves in the flow path of the drop line and APS lino. This would invalidate the originallicensing basis for single failure which relied upon the APS method should a singic 4
active component failure prevent operation of the DL-RB Sump method. Evaluation of compliance with the single failure criterion is presented in Section V of this attachet.
IPAP Report During the integrated Performance Assessment Process (IPAP) inspection (Reference 24), the NRC team identified a concern that the methods described in the Crystal River Unit 3 (CR-3)
Finel Safety Analysis Report (FSAR) for post-LOCA boron dilution were different from those approved by the NRC. The IPAP report stated that the passive method involving gaps in the reactor internals was accepted as a backup to active boron dilution methods in an NRC letter dated March 9,1993 to the B&WOG (Reference 16); however, a piant specific evaluation was never submitted for NRC review regarding the acceptcbility of the gaps at preventing boron precipitation. The IPAP report also stated that a preliminary safety evaluation performed by the NRC determined that CR 3 was safe to allow continued operation until this issue can be resolved.
The methods utilized at CR 3 to prevent post-LOCA boron precipitation are provided in this attachment (see Section VI). The methodology will ensure that CR 3 can effectively prevent L
boron precipitation in the long term cooling phase following a design basis LOCA.
Additionally, a discussion of the plant specific results for the evaluations of the effectiveness of the hot leg nozzle gap flow method is included in Section VI of this attachment. This information alsc supports closure of the NRC restart issuo.
IV. CR 3 POST-LOCA BORON PRECIPITATION PREVENTION PLAN h
The following discussion provides an overview of the post-LOCA boron precipitation mechanism as it applies to CR 3 and describes each of the methods for prevention of boron precipitation after a LOCA.
l Desedption of Post LOCA Boron Precipitation Boric acid, which is added tc the RCS to provide core reactivity control during steady state 1
operation and during accioent condidons, is dissolved in and freely circulatesin the RCS liquid.
The core flood tanks (CFT) and borated water storage tank (BWST) also contain boric acid for additional reactivity control during accidents. During LOCA transients, postulated conditions may result in this boron becoming concentrated in the core region due to boiling in the RV.
it-is possible that under certain temperature and concentration cc,oditions, boron could precipitate out of solution, potentially reducing core flow, and challenging long term core cooling.
l The possibility of boron precipitation is a function of the solubility of boron in the RCS. At L
-high RV temperatures, the solubility of boric acid is suf ficiently high so that boron precipitation cannot occur. _ Boron solubility decreases as RCS temperature decreases. At saturation temperatures below 305" F (corresponding to 72 psia), the core boron concentration could approach the solubility limit if given adequate time to concentrate in the core following a loss L
U. S. Nucl:ar R:gu!atory Commission Attachm:nt B 3F1097 32 Page 4 of 22 of subcooling. This temperature was conservatively determined assuming that all the boron initially in the RCS, plus the boron in the CFTs and all the volume of the BWST, is concentrated in the core region.
During the cooldown following a SBLOCA,if the core remains saturated, with core pressures above 72 psia, boron precipitation cannot occur even if all the available boron is placed within the core mixing volume. Prevention of boron precipitation at temperatures lower than 305' F with these conservative assumptions must rely upon boron dilution methods.
For LBLOCAs, the RCS depressurizes quickly to the containment pressure, which results in reduced boron solubility limits. For these cases, the time required for the core boron concentration to increase to the solubility limit is in excess of five hours. Provention of boron precipitation beyond five hours after a LBLOCA must rely upon boron dilution methods as described in the following section. SBLOCAs require more time to reach a condition in which boron may precipitate.
Current Pnst LOCA Boron Precipitation Prevention Methods Reactor coolant system circulation and boron concentration in the post-LOCA, long term, cooling environment have been analyzed for the potential of unacceptably high boron concentrations in the core region. Both large and small breaks of the reactor inlet (cold leg) and outlet (hot leg) are considered. The methods available for preventing post LOCA boron precipitation consist of passive design features and active methods as described below. The licensing basis for these methods is discussed above. The LOCA scenarios in which each l
method is available are indicated in Table 2 and discussed Inter in this attachmen:.
PASSIVE METHODS:
RVVV Overflow This passive design feature results in natural circulation through the RVVVs and will automatically occur for cold leg breaks which result in sufficient refill of the RV.
Overflow through the RVVVs to the vessel downcomer draws dilute ECCS water from the lower plenum into the core region, which limits the maximum boron concentration to less than the solubility limit. RVVV overflow depends on the location and size of the break bemg above the center line of the reactor coolant pump cold leg discharge piping.
Hot Leo Nozzle Gao Flow (Fiaures 5 and 61 This passive design feature automatically utilizes natural circulation from the upper core region through the gaps between the hot leg outlet nozzles and core support shield to the l
downcomer. The flow path for the hot leg nozzle gaps is as follows. The steam being produced by the core flows through the RVVVs, through the broken cold leg, and out of the break. ECCS flow replaces the boil off from the vessel with the excess flow above the boil-off rate being discharged out the break. The flow of water through the hot leg nozzle gaps recirculates through the downcomer and mixing occurs. The volume of water passed by the gaps is replaced with water of reduced coron concentration. resulting in a reduction in core boron concentration. Evaluations of the effectiveness of the hot leg
U. S. Nuclear Regul tory Commission Attachment B 3F1097 32 Page 5 of 22 nortle gap method for preventing post-LOCA boron precipitation at CR-3 is discussed in this attachment, i
ACTIVE METHODS:
l ECCS Flow Path Throuch the Reactor Vessel This method applies to those breaks on the hot leg side of the RV. The ECCS flow path of RCS from the cold leg, into the downcomer, through the core, and out the hot leg (to the break) will supply a continuous flow of dilute ECCS water through the RV which will limit the boron concentration to less than the solubility limit. Breaks on the cold leg side 6f the RV do not provide this flow path.
Droo Line to the RB Sumo (DL-RB Sumo) (Fiaures 2 and 3)
In this method, aligning the decay heat drop line to the RB Sump will allow gravity feed of reactor coolant from the hot leg to the RB Sump which would then be recirculated back to the RV with inc ECCS system. This will result in injection flow through the core at a rate eLaal to the drop line flow rate. This method requires precautions to be exercised when RCS and RB pressures are not equal to ensure that an excessive differential pressure (6P) does not adversely affect the integrity of the RB Sump screens.
In order to determine the time at which the drop lino should be aligned to the RB Sump, analyses were performed that did not credit RVVV entrainment or hot leg nozzle gap flow.
Based on these conditions, the decay heat drop line should be opened within five hours following a LBLOCA initiated from 2772 MWt. Some additional time would be available for CR-3 due to its lower rated thermal power of 2544 MWt.
In the case of a LBLOCA, RCS pressure and RB pressure reach equilibrium in a short time period after the initiation of the break, in these cases, use of the DL-RB Sump method can be accomplished within five hours without jeopardizing the integrity of the RB Sump screens. For SBLOCAs that do not result in RCS refill, preventing boron precipitation using the drop line method would require alignment of the drop line to the RB Sump prior to the RCS pressure decreasing below 72 psia. However, this would resultin a AP across the RB Sump screens that could compromise their integrity. This case is provided in
" Limited SBLOCA" discussicn below.
Hot Lea iniection with APS (Fiaure 4) lo this method, diluts water from the RB Sump is injected into the hot leg to force a reverse flow through the RV. This will provide a constant flushing of boron in the RV.
The flow path is through the APS line, via low pressure injection (LPI) system train A, out of the pressurizer through the surge line into the hot leg, and then into the RV. APS would be ineffective until the flow rate exceeds the boil-off rate. At this time, the excess flow (above the boil-off rate) would provide boron dilution. Using conservative analysis assumptions, APS would not be effective for the first 34 days after the accident.
Additionally, flow from APS could disturb the boron dilution flow path that exists from the hot leg nozzle gap method. Therofore, initiation of APS reouires precautions to ensure
U. S. Nucl:ar Regulatory Commission Attachm:nt B 3F1097 32 Page 6 of 22 that this flow does not hinder the effectiveness of the gap flow. This method would only be used upon indication that the hot leg nozzle gap method is not effective.
Increased flow rates could also be available with the APS method of boron dilution; Hydraulic analyses show that greater than 40 gpm could be available for hot leg injection as RCS pressure decreases. Additionally, APS using the high pressure injection (HPI) system could provide rnuch greater than 40 gpm of flow. Increasing the flow rate available would reduce the length of time until APS would be effective following the LOCA.
Hot Leg injection via Reverse Flow through LPI (HLI-RF)
FPC is developing an additional active method that will provide defense in depth for the current methods of post-LOCA boron dilution. This method, which provides hot leg injection via reverse flow through a non-operating LPI pump, wi'.I require NRC approval prior to implementation. Reliance on this method of boron dilution is not required prior to restart of CR 3 and is not credited in this submittal. FPC will submit a request for NRC review and approval of this method.
LOCA Scenarios Post LOCA precipitation control methodologies for CR-3 are consistent with the NRC approved methods except for a small class of breaks where refill of the RV is insufficient for RVVV ovorflow and the drop line to the RB Sump cannot be upened (due to excessive RCS versus RB AP which could damage the screens) prior to reaching the analyzed boron solubility limits in the RV. During this LOCA scenario, other methods (including the hot leg nozzle gap flow method) must be relied upon until the drop line can be opened. The probability of a break which relies on hot 10g nozzle gap flow to prevent boron precipitat:on is extremely remote (less than 2.5 x 10 per year - see Limited SBLOCA discussion below).
4 Each break scenario, along with the methods used for post-LOCA boron precipitation control, is illustrated in Table 2 and discussed below. References to the break ;cenarios shown on Table 2 are included below in brackets (e.g.,1811, [F3), etc.).
To evaluate core boron concentration control, LOCAs (Al are divided into hot leg and cold leg breaks. For hot leg breaks [B1), ECCS injection results in a continuous flow of liquid through the core equivalent to the ECCS injection flow rate. This flow will be adequate to ensure core boron concentration remains well below solubility limits and no additional actions are necessary to prevent boric acid precipitation.
- Cold leg breaks which occur in the reactor coolant pump (RCP) suction piping [C1) will result in the vossel filling to an adequate level to allow the internal RVVVs to open and overflow to the downcomer annulus. This height of water in the vessel results from the center line of the RCPs being several feet above the bottom of the RVVVs. Adequate recirculation will be assured by the additional vessel liquid level associated with the froth mixture due to core boiling.
U. S. Nuclear R::gulatory Commission Attachment B 3F1097 32 Page 7 of 22 Cold leg breaks on the RCP discharge piping above the piping centerline (D1] will result in a j
RV level which allows the RVVVs to open with overflow to the downcomer region. This is an effective means of preventing boron concentration in the vessel.
Of all the potential cold leg break sizes, only those breaks on the RCP discharge piping below the piping centerline between approximately 0.05 ft' and 0.17 ft'IF21 have the potential for reliance on hot leg nozzle gap flow to prevent boron precipitation in the absence of a single failure.
Breaks smaller than about 0.05 ft [F3] will either remain at pressures and 8
temperatures above the solubility limit or will result in refill of the RCS and RV sufficient to result in RVVV overflow, which will promote the circulation of diluta ECCS water througn the core and prevent boron precipitation. Breaks larger than about 0.17 f t'lF11 will result in RCS depressurization which will allow the drop line to be operod without damaging *he RB Sump screens prior to the core boron concentration reaching the solubility limit. This w!Il also promote circulation of dilute ECCS water through the core and prevent boron precipitation.
8 Of a!! the potential cold log breaks between approximately 0.05 ft' arid 0.17 f t, only treaks in this range which extend below the cold leg nozzle centerline would require additional measures to prevent boron precipitation if passive methods are shown not to be offective [F2]
(this is discussed below). Cold leg breaks above the centerline on the Cold Leg Pump Discharge (CLPD) piping, as well as Cold Leg Pump Suction (CLPS) breaks, will result in RV water levels which establish RVVV overflow, which will promote the circulation of dilute ECCS water through the core and prevent boron precipitation.
Limited SBLOCA - Area F2 on Table 2 As discussed above, the only break which would rely on hot leg nozzle gap flow to prevent boron precipitation (in the absence of a single failure)is a cold leg break in the range of 0.05 ft to 0.17 ft' v!hich occurs on the CLPD piping below the cold leg nozzle centerline. The 8
probability of such a break is less then 2.5 x 104 per year. This is approximately 1000 times less likely than the occurrence of a LBLOCA.
l In this small area of concern, FPC will utilize the design features of the plant along with j
procedures to prevent post-LOCA boron precipitation. The decision to implement an active boron dilution method will be based on the acceptubility (and avail bility) of the boron concentration in the RB Sump. In the limited cases where the DL-RB Sump method could not be used due to the RB Sump screen concern, other methods would be relied upon.
1 Additionally, other defense in depth methods would be used to minimize the time needed to reach conditions where the drop line could be opened. Preliminary analytical results indicate that the drop line could be opened at a RCS temperature of 2800 F. As stated earlier, at saturation temperatures above 305* F, boron concentration cannot reach the solubility limit in the RV. Operator initiated cooldown and depressurization would be used to decrease RCS pressure to the point where the drop line could be opened. This would include utilizing the emergency feedwater pumps, diesel generator backed auxiliary feedwater pump (FWP-7),
I atmospheric dump valves (ADVs), turbine bypass valves (TBVs), power operated relief valve (PORV; and the high point vents (HPVs). These multiple options provide significant defense j
h depth for assuring that the period that the drop line cannot be utilized for boron precipitation prevention is minimal. FPC will evaluate the ability of various secondary side
U. S. Nuclzr R:guttory Commissi:n Attachm:nt B 3F1097 32 Page 8 of 22 cooldown and primary side depressurization methods to demonstrate the ability of the plant to rapidly reach an RCS pressure which will allow the drop line to be opened.
The passive method of the hot leg nozzle gar flow would be available during this time. Hot leg nozzle gap flow has been evaluated fx CR 3 as effective for prevention of post-LOCA boron precipitation.
Other methods would only be necessary if the RB Sump boron concentration indicated that RV concentrations were approaching the solubility li' nit. As discussed earlier, the APS method would be available in this situation. Increased flew rates could also be available with APS via the LPI system or the HPl system to make this method effective ettlier than 34 days post accident.
Detection of Boron Concentration Boron concentration in the RB Sump, which will be sampled post LOCA, will be used as an indication of boron concentration in the core region and, therefore, as a measurement of the effectiveness of boron dilution in the core region. The sampling method using the post accident sampling system (PASS) has been successfully tested. A flow path from the RB Sump to the PASS boronometer and gamma spectroscopy was established to ensure that RB Sump boron concentration could be determined. Qualification of the PASS system is addressed in LAR #220 (Reference 35).
i As the RV boron concentration increases due to boiling of the water recirculated from the RB i
Sump, the boron concentration in the RB Sump will decrease. FTl document 51 5000519-03 (Attachment D) includes an evaluation of the expected change in the RB Sump beron concentration as the core approaches the boron solubility limit, where boron precipitation may occur (see Section 4 of FTl docuraent 51 5000519-03). Several SBLOCA cases were l
analyzed assuming the limiting 0.05 ft' break with different initial BWST, CFT, and RCS boron concentrations. Plots of boron solubility limit versus temperaturo, with the core region boron concentration translated into an equivalent BWST/RB Sump concentration difference, is included in Attachment D as figures 4A,48,5A,58, and SC.
l As the core region boron concentration increases, the RB Sump boron concentration decreases. The degree of boron concentration in the RV is, therefore, monitored by the difforence between the initial BWST concentration and tha RS Sump boron concentration.
l This relationship is derived analytically based on the assumption that the BWST volume is pumped into the RB Sump via ECCS and RB spray. Using the boron solubility temperature l
curves and the RCS boron concentration versus BWST/RB Sump concentration difference, a l
relationship between a given BWST/RB Sump concentration difference and allowable incore l
temperature was developed and plotted as a curve. Therefore, core temperature must remain below the curve to ensure that boron precipitation does not occw. Similarly, for a given temperature, the difference between the initial BWST boron concentration and the sampled RB Sump boron concentration (which is associated with the core boron concentration) must remain below the curve to preclude boron precipitation.
Decision Matrix Utilization of post-LOCA boron dilution methods at CR-3 will be coordinated in the Technical Support Center (TSC) using a decision matrix (Figure 1). The decisinn matrix shows
- 9 logic for determining the actions to be taken if subcooled conditions cannot be achieved following
U, S. Nuclear Regul: tory Commission Attachm:nt B 3F1097 32 Page 9 of 22 i
l a LOCA. The decision process for controlling post-LOCA boron precipitation requires the monitoring of the boron concentration in the RB Sump as an indication of boron concentration -
in the RV, RCS incore temperature, RCS pressure, RB pressuref and time after subcooling
-i margin is lost in the RCS. -
l The passive design features for CR 3 (RVW overflow and hot leg nonle gaps) will perform their function without the reliance on any operator action. These features are a part of the CR 3 RV design, and their effectiveness for post LOCA boron precipitation control has been evaluated. - Although these. methods are expected to _ effectively. limit the RV boron concentration from reaching the solubility limit, 'the RB Sump boron concentration will be monitored to ensure the effectiveness'of these methods post LOCA is known by the
~ emergency response team, if conditions exist whereby boron concentration in the RV is a concern or RB Sump sempling cannot be performed, the decision matrix will direct the initiation of the available active methods in accordance with procedures (described below). If the OL RB Sump method cannot be initiated due-to the AP effects on the RD Surnp screens, additional defense in depth measures will be initiated as discussed in this attachment.
Procedures Procedures will be revised to reflect the post LOCA baron precipitation prevention plan as described in this attschment. Tho use of the de hion matrix will be identified m EM-225,
" Duties of the Technical Support Center Accident Assessment Team." Post-LOCA boron precipitation prevention will be coordinated by personnel in the TSC. Guidance will describe the following:
Discussion of the inechanics of post-LOCA boron precipitation, conditions under which it could occur, and a discussion of the passive design features and active methods available for CR-3.
Specific guidance on the monitoring and implementation of post LOCA boron precipitation prevention methods as required by the decision matrix.
~
A description of the critical parameters which must be monitored as cart of the boron mitigation strategy. Parameters to monitor include:
RCS Temperature (T..)
RCS Pressure RB Sump Boron Concentration
-RB Pressure
- Time at Saturated Conditions Guidance on the monitoring of the offectiveness of the passive design features based on HB Sump boron concentration.
Guidance on the initiation of an active method of post-LOCA botors oilution. This guidance willinclude information as described below.
U. S. Nucl:ar Regulatory Commission Attachment B 3F1097 32 Page 10 of 22 4
Appropriate plant conditions for initiation of an active method will be described. The appropriate active method for use in the post-LOCA environment will be defined by the previously addressed plant parameters. Active methods of boron precipitation prevention will not be initiated unless RB Sump boron concentration indicates that the RV boron solubility limit is being approached. Periodic monitoring of the RB Sump boron concentration willindicate a trend that will be used in determining actions that will be taken. The absence of a decreasing RB Sump boron concentration r.1akes any increase in vessel concentration impossible, and an active boron di!ution method unnecessary. RB Sump boron concentration would remain essentially constant unless the passive methods of boron cfilution are ineffective.
RB Cump 'coron concentration trending would also reflect other brcak scenarios that do not require the initiation of active methcds of post LOCA boror; precipitation prevention. Hot leg breaks inherently provide ECCS flow through the core region, resulting in no concentrationincrease. Also,if adequate subcooling is present, boron cannot concentrate in the RV since no boiling would be occurring. In these cases, the RB Sump boron concentration would not indicate an approach to RV boron solubility limits.
Guidance on the use of the active methods will be provided.
DL RB Sump This method can be implemented once RV and RB pressure are approximately equal. Criteria that will be used to ensure that the AP between l
the RV and RB is acceptable will be included based on the analysis of the drop line and RB Sump screens. This method requires that one train of the LPI system be secured to accomplish the flow path from the hot leg to the RB Sump.
APS - This method can be implemented once the flow rate will exceed the boil-off rate. An evaluation of the effectiveness of the passive design features will be performed to ensure that APS flow does not interfere with the natural circulation via the hot leg nozzle gaps. Additionally, higher flow rates via the APS would result in the APS method being effective earlior than 34 days. This could be accomplished with higher flow rates from the LPI system at lower RCS pressures and by higher flow rates from the HPl system.
Guidance will be provided to use defense-in-depth methods if needed. These l
methods could be used for those small prcbability break scenarios where the DL-RB l
Sump method cannot be used due to the limitation on the RB Sump screens and RV solubility limits could be reached. Cooldown will be accomplished by several methods already available to the operator including the use of emergency feedwater (EFP-1, EFP-2) or auxiliary feedwater (FWP-7) in conjunction with the TBVs or ADVs.
Depressurization will inherently occur during cooldown, but can es augmented through the use of the PORV and the HPVs. Guidance will also be provided on the strategy to restart a second LPI pump, if previously secured. This action could increase the RCS pressure (and increase RCS tempercture) for a limited spectrum af breaks to values above the boron solubility temperature limit.
U. S. NucleOr Regulatory Commission Attachment B 3F1097 32 Page 11 of 22 Figures and tables will be included to provide the information needed to support the decision ".iatrix. This information willinclude the minimum time needed for the RCS to rp%n the solubility limit ut. der saturated conditions and allowable RB Sump boron concentration decrease (both parameters are a function of RCS temperature), as well as APS effectiveness as a function of RCS pressure and time after the accident.
Appropriate revisions will also be raade to the applicable Emergency Oporating Procedures (EOP). EOP-08, "LOCA Cooldown" will be revised to instruct the Chemistry department to begin sampling the RB Sump for boron concentration. Chemistry instructions are currently contained in EM-307, " Sampling and Analysis of the Reactor Coolant System, the Reactor Building Sump, and the Miscellaneous Waste Storage Tank Under Accident Conditions," but are in the process of being transferred to procedure CH 632D, " Post Accident Sampling and Analysis of Reactor Building Sump," as part of procedure upgrades.
EOP 14, " Emergency Operating Proceduro Enclosures," Enclosure 20, " Boron Control," will contain the verified and validated plant procedure for operation of the DL-RB Sump and APS methods, initiation of these methods will be coordinated through the TSC as directed by EM 225.
V.
SINGLE FAILURE ANALYSIS The combination of methods available for post-LOCA boron precipitation preventic.m at CR-3 are single failure-proof ss described below. The varioue methods, in and of themtebros, are not sing l0 failure-proof; however, when reviewed in combination witn the methods available in occh instance (including defense in depth in area F2 on Table 2) are adequate to perform the boron precipitation prevention function following a single failure (the effectiveness of the defense in depth measures is being evaluated as stated above). The methods that are relied upon for each break size and location are listed in Table 2.
The single failure review of the post LOCA boron prevention methods presented to the NRC in Reference 6 (January 13, 1976) credited the combination of the DL-RB Sump and APS methods. When it was identified that the RVVV overflow may not be effective in all cases, it was also idontified that the APS method when limited to 40 gpm flow would not be effective by itself until approximately a month af ter the accident. Since the APS would not provide a backup to the DL-RB Sump method in this case, the hot leg nozzle gap method was credited as an acceptable backup method. This resulted in the combination of methods avanabie meeting single failure criteria, The ability of the post-LOCA boron precipitation prevention methods to perform their intended function with a single f ailure was restated in the response to Request 5 of the 10 CFR 50.54(f) request for information (Reference 22).
Recently, a single failure has been identified that would prevent the use of both the DL-RB Sump and APS methods. Loss of the 3AB ES MCC would result in the loss of power to valves on both the drop line and APS line. This is not consistent with the original licensing basis, but does not affect the current single failure compliance as stated above. Since the establishment of the hot leg nozzle gap method as an acceptablo backup method (APS would not be offective until > 34 days at a flow rate of 40 gpm), the APS method is not relied upon as the backup method for the DL-RB Sump method for any break size / location. For those cases when APS would be ef fective, the likelihood of a LOCA. in combination with this single failure, would be extremely remote,
U. S. Nucl:ar R:gulatory Commission AttacnmInt B 3F1097 32 Page 12 of 22 VI. EVALVAMON OF THE HOT LEG NOZZLE GAP METHOD F,QR CR 3 The of fectiveness of the hot leg nozzle gap method to prevent boron precipitation in the post-LOCA long term cooling period has been evaluated by FT!. Table 1 contains a listing of correspondence with the NRC providing ir. formation on hot leg nozzle gaps.
The current analysis that supports the of factiveness of the hot leg nozzl6 gap method for CR 3 is contained in FTl document 51-1266113 00.
This docuinent is a proprietary report submitted to the NRC, by the B&WOG in Referer.co 27. This evaluation was performed utilizing conservative assumptions that would bound all of the B&W plants. The full spectrum of LOCAs was analyzed for initial core power levels of 2568 anti 2772 MWt with limiting gap sizes for all plants, in each core power category, without credit for DL-RB Sump or RVVV liquid overflow or entrainment. The most limiting solubility margins were produced by the highest power lovel plant with ths smallest measured gap size. The boron concer? ration calculations using the nominalisothermal limiting gap size for all plants with an initial core power of 2772 MWt demonstrated that the hot leg nozzle gap flow is adequate to prevent the core boron concentrations from reaching the solubility limit.
The analysis determined that hot leg nozzle gaps will be open and provide sufficient dilution flow, it was also determined that a reduction of the smallest as built gap area by ninety porcent during the long term cooling phase does not preclude the gap flow from providing adequate core boron dilution. The gap flow mechanism was also shown to provido adequate flw for a small range of non-isothermal temperatures. For the smallest gap sizes, the shell temperature can be up to 25 F cooler than the RV internals without totally closing the gap and inhibiting the dilution flow.
This analysis was conservative for CR 3 based on hot leg nozzio gap size and thermal power level. The hot leg nozzle gap sizes for CR 3 are the largest of the B&W plants. These sizes are based upon actual as built measurements. Gap sizes for CR 3 are listed in Table 1 of FTl Documsnt 51-126113-00. Additionally, the maximum power rating of CR 3 is 2544 MWt.
Using the gap size for CR-3 with an initial core power level of 2568 MWt results in adequate core boron dilution with the shell temperaturu up to 75 F cooler than the internals.
The conclusions of this analysis were also submitted under the docket for CR 3 by FTl Document 51 5000519-00 in Reference 34. The current revision of FTl Document 51-5000519 03 is included as Attachment D in this submittal.
Vil. HEFERENCES 1.
BAW-10091, Topical Report, August 1974, "B&W's ECCS Evaluation Model Report with Specific Application to 177 FA Class Plants with Lowered Loop Arrangement" 2.
BAW 10091, Supplement 1, Topical Report, December 1974, " Supplementary and Supporting Documentation for B&W's ECCS Evaluation Model Report with Specific Application to 177-FA Class Plants with Lowered-Loop Arrangement" 3.
NRC to FPC letter dated July 7,1975, B&W Topical Report BAW-10103, f 3N0775-03) 4.
FPC to NRC letter dated September 19,1975, Additional Information on Emergency Core Cooling System,13F0775-02]
U._5. Nucle;r RIgudtory Commission.
Attachment B j
3F1097-32 Page 13 of 22 5.
NRC to FPC. letter dated December 8,1975, Request for Additional Information on ECCS Analysis, [3N1275 021-6, FPC to NRC letter dated January 13,1976, Response to 12/8/75 Request for Additional Information on ECCS Anelysis,13F0176 03) j 4
7.
- NRC to B&W letter dated February 4,1976, Topical Report Evaluation of BAW 10103 l
8.
FPC to NRC letter dated December 10,1976, " Crystal River Unit #3 Operating License DPR 72," 13F1276 06) 9.
" Supplement No. 3 to the Safety Evaluction Report by the Office of Nuclear Reactor Regulation in the Matter of Florida Power Corporation, Et Al, Crystal River Unit No. 3, Docket No. 50 302," December 30,197613N1276101 10.
NRC to B&W letter dated February 18,1977, " Evaluation of BAW-10103" 1 1. - BAW 10103A, Revision 3, Topical Roport, July 1977, "ECCS Analysis of B&W's 177-FA Lowered Loop NSS" 12.
B&W to NRC letter dated November 7,1991, JHT/91-186, " Post LOCA Boron Precipitation" 13.
FPC to NRC letter dated December 4,1991, " Licensee Event Report 91 011 00,"
13F1291-04) 14.
B&W to NRC letter dated January 5,1993, ESC 005, As-built Gap Measuren,ents 15.
B&WOG to NRC letter dated February 4,1993, OG 1136, " Post LOCA Reactor Vessel Recirculation to Avoid Boron Precipitation" 16.
NRC (A. C. Thadani) to B&WOG letter dated March 9,1993, " Post LOCA Reactor Vossel Racirculation to Avoid Boron Precipitation," [3NO393-18) 17.
NRC to FPC letter dated June 17,1996,"NRC Integrated Inspection Report 50-302/96-04 Notice of Violation," [3N0696-08) 18.
NRC to FPC lettor dated June 26,199C, " Crystal River Nuclear Generating Plant Unit 3
Boron Precipitation Following Design Basis Accident - Request for Information Pursuant to 10 CFR 50.54(f)," [3N069617) 19.
NRC letter dated June 28,1996, "NRC Conference Call with Babcock and Wilcox Owners Group Regarding Boron Precipitation in B&W Designed Reactors Post SBLOCA"
-20.
B&WOG to NRC letter dated July 3,' 1996, " Post LOCA Boron Precipitation" 21.-
NRC letter dated July 16,1996, " Summary of Public Meeting with the BWOG Hegulatory Response Group to Discuss Boron Precipitation Concerns During Long-Term Core Cooling Following a Small Break Loss-of-Coolant Accident"
U. S. Nucl::r 3:gul: tory Commission Attcchment B 3F1997 32 Page 14 of 22 22.
FPC to NRC letter dated July 26,1906, " Response to Borcn Precipitation Issue,"
[3F0796-191 23.
B&WOG to NRC letter dated July 31,1996, OG 1601, "B&WOG Post-LOCA Core Boron Dilution Management" 24.
NRC to FPC letter dated August 23,1996," Crystal River Unit 3 Integrated Performance Assessment Process (IPAP) Final Assessment Report (NRC Inspection Report No. 50-302/96 201)," [3N0896121 25.
NRC to FPC letter dated January 7,1997,"NRC Inspection Report No. 50 302/9619,"
(3N0197 04) 26.
NRC to FPC letter dated March 12,1997, " Notice of Violation and Exercise of Enforcement Discretion (NRC Inspection Report Nos. 50 302/96-12 and 50-302/96-19)," [3NO397 09) 27.
B&WOG to NRC letter dated March 27,1997, OG-1644, "' Post LOCA Boron Concentration Management,' FTl Document No. 51-1266113-00 (Proprietary), March 1997" 28.
FTl Document 51-1266113-00, March 1997, " Post LOCA Boron Concentration Management" 29.
NRC to FPC letter dated April 11,1997, " Notice of Violation and Exercise of Enf orcement Discretion (NRC Inspection Report Nos. 50-302/96-12 and 50 302/96-19)
NRC to FPC letter,3NO397-09, dated March 12,1997," [3F0497 34) 30.
FPC to NRC letter dated June 16,1997,"NRC Notice of Violation, Integrated Inspection Raport No. 50 302/96 19, NRC to FPC letter, 3N0597-13, dated May 16, 1997,"
l 31.
FFC to NRC letter dated June 26,1997, " License Condition 2.C.(5) Requiring Installation and Testing of Flow Indicators," 13F0697 08]
32.
FPC to NRC letter dated August 4,1997, " Drop Lino Valve Fosition Indication,"
l 13F0897 241 33.
FTl Document 51 5000519-00, August 1997, " Boron Dilution by Hot Leg injection" 34.
FPC to NRC letter dated September 12,1997, " Post LOCA Boron Precipitation Mitigation Plan," 13F0997-281 l
35.
FPC to NRC letter dated October 30,1997, " License Amendment Request #220, Revision 1; Revision of Operating License Condition 2.C.(5) (TAC No. 99128)"
13F1097-08) 36.
CR 3 FSAR, Revision 23, Section 4.3.10.1, " Boron Dilution"
U. S. Nucle:r Regulatory Commission Attachm:nt B 3F1097-32 Page 15 of 22 TABLE 1 HOT LEG NOZZLE GAP FLOW Submittal Description BAW 10091, Evaluation crediting hot log nozzle gaps determined that the Supplement 1 boron remains dilute throughout the post accident period.
(12/74)
NRC and B&WOG Presented the results of analyses supporting gap flow.
moeting 12/3/92 Conservative calculations were performed to establish the flow areas through the gaps. Using these conservative flow areas, thermal hydraulic calculations were performed to demonstrate that adequate flow will occur though the gaps to prevent precipitation. Conservatisms were also incorporated into the thermal hydraulic calculations.
B&W to NRC lettter As built measurements for gap prwided to the NRC.
dated 1/5/93 NRC to B&WOG NRC approval of gaps as an adequate backup to the primary letter dated 3/9/93 recirculation flow path.
FPC to NRC, Provided as-built gap measurements, gap size and flow rate as a Response to 10 function of RCS temperature, and determination that boron CFR 50.54(f) concentrations remain significantly below the solubility limits for Roquest, 7/26/96 reductions in flow of 50 and 75 perce11 using the smallest gap sizes for the must limiting break.
B&WOG to NRC Analyses on gap flow for B&W plants. Concludes gap flow is FTl Document 51-aciequate boron dilution method for both SBLOCAs and 1266113-00 LBLOCAs. As built gap measurements provided for each plant, (3/27/97) and analyses performed based on the smallest gap size (CR-3 ht.s largest gaps). Confirmed that reduction of the smallest as-built gap area by 90 percent does not preclude adequate gap flow for boron dilution.
FPC to NRC Re-affirms for CR-3 that hot leg nozzle gap flow is adequate FTl Document 51-boron dilution method.
5000519-00 l
(9/12/97)
U. S. Nuclear Regulatory Commission Attachment B 3F1097 32 Page 16 of 22 TABLE 2 - Post-LOCA BORON PRECIPITATION CO!!(RC' LOCA Clasemcation Boron Precipitation Prevention Methods Hot Constant flow of ECCS through teactor vessel Bree s (B1)
All RCS Breake (A)
RVVV overflow, DL RB Sump. Hot Leg Nozzle
$y') "
Gaps,APS l
Cold leg Breake (B2)
Abovs piping RVVV overflow, DL-RB Sump, Hot Leg Nozzle centerline Geps,APS (D1)
RCP l
Otcharge LB 0,CA DL RB Sump, Hot Leg Nozzle Gaps, AP*
(C2)
X > 0,17 ft' DL R2 Surrp Hot Leg Nozzle Geps, APS ng SBLOCA 0.05 ft8 < X < 0.17 ft' DL-RB Sump (Lknited ecreene), Defense in hne (D2)
(E2).
(F21 Dacth_ Hot Leo Nonle Cana APS X < 0.0S ft2 RVVV overflow DL RS Sump, Hot Leg Nozzle (F3)
Gaps. APS 1
U. S. Nucbar Regulatory Commision Attachment B 3F1097-32.
Page 17 of 22 --
Figure 1 - Decision Matrix lS RCS Saturate'd?
Yes.
IF Concentration < Allowable Value-OR Unavailable-
. No THEN Establish Active Method Is RCS Pressure ~ RB Pressure?
Yes No Continue:Cooldown; v
- u t
Align Drop Line Establish Appropriate.
to RB Sump Mitigation Pian as.
Defined by the TSC If Active Method can not be Established, Then Rely on Passive Method (Gap Flow).
1 r
U. S. Nuc1:cr Regul: tory Commission Attichm:nt B 3F1097-32 Page 18 of 22 Figure 2 - DL RB Sump (Reactor Vessel Flow Path) nnnngno
-J lt/-
s7M)u 3
T T
_ II T R a
\\v
,a s
p
.I 4 L
C f Hot leg (Intact) 1 Coldleg(Broken
=k
}
i..
,/ :
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U. S. Nuclear Regulatory Commisslon Att: chm:nt C 3F1097 32
. Page 1 cf 1 ATTACHMENT C AC10NYMS AND ABBREVIATIONS ADV......... Atmospheric Dump Valves APS......... Auxillary Pressurizer Spray B&W........ Babcock and Wilcox B&WOG...... B&W Owners Group BWST........ Borated Water Storage Tank CFR......... Code of Federal Regulations CFT......... Core Flood Tank CLPD........ Cold Leg Pump Discharge CLPS........ Cold Leg Pump Suction CR 3........ Crystal River Unit 3 DL RB Sump... Drop Line to Reactor Building ECCS Sump AP.......... dif fetential pressure ECCS........ Emergency Core Cooling System EOP,........ Emergency Operating Procedure F........... Fahrenheit FPC......... Florida Power Corporation FSAR........ Final Safety Analysis Report FTl.......... Framatome Technologies incorporated gpm......... gallons per minute HLl RF....... Hot Leg injection via Reverse Flow Through Portions of LPI HPl.......... High Pressure lnjection HPV......... High Point Vent IPAP,.......
Integrated Performance Assessment Process LAR......... License Amendment Request LBLOCA...... Large Break LOCA LOCA........ Loss of Coolant Accident LPI.......... Low Pressure injection MCC........ Motor Control Center MWt......... Megawatt Thermal NRC,....... Nuclear Rogulatory Commission PASS........ Post Accident Sampling System PORV..,,,. Power Operated Relief Valve psia......... punds per square inch absolute RB.......... Reactor Building RCS......... Reactor Coolant System RCP......... Reactor Coolant Pump RRG.......... Regulatory Response Group RV.......... Reactor Vessel RVVV........ Reactor Vessel Vent Valve SBLOCA...... Small Break Loss of Cociant Accident TBV......... Turbine Bypass Valves TSC......... Technical Support Center
U. 3. Nucle:r R:gulatory Commission Attachm:nt D 3F1097 32 1
ATTACHMENT D t
FTl DOCUMENT 51 4000519 03
(
(ATTACHED) i Pages of the FTl Document are noted with the revision number of the last changs for that page. See Record of
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Rovisions for affected pages for each revision.
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INTEROFFICE CORRESPONDENCE i
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Nuclear Engineering NR3B 240 3560 Office MAC Telephone sveXCT! Cryt,tal River Unit 3 L
Quality Dveument Transmittal. Analysis / Calculation To: Recofds Management NR2A a.
J The following analysg/ calculation package is submitted as the OA Record copy:
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DOCNo if PC DOCUMENT IDtNteFicatioN NuMetmi mte
- systtuts, 10 tat PAtt$ tRANSMitttD M97 0098 3
RC,DH 102 TITLI l
BORON DILUTION BY HOT LEG INJECTION I
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.l VIND (VtNDon NAMil VINDoR DOCUMENT NUM8tR (DXREFI l SuPthSEDio DOCUMENT $ (DAntPI FTl 51 6000519 03 l 515000519 00 I
DHV 3 DHV 8 ll DHV 4 l DH 38 Fi l
DHV 41 l
l DHV-7 l
l l
NA ll ll l
l CowMENta (USAGE histnictioNs. Pnoen!st Anv. tTC.I REVISON #1 AND 2 OF THIS CALCULATION WERE NOT ENTERED INTO THE FPC SYSTEM j
NOTE-Use Tag number only for valid tag numbers (i.e., RCV 8, StW 34, DCH-99), otherwise; use Part number field (i.e., CSC14599, AC1459), if more space is required, wnte "See Attachment" and list on separate sheet.
I Dt $1C GA t AT VERiFICAT60N ENolNET DAtt SVF'445 i
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I cc: Nuclear Projects (if MAR /CGWR/PEERE Calculation Review form Part it' actions required XYes O No l
Return to Service Related)
O ve. x No (if Yes, send copy of the forrn to Nucitar Regulatory Assurance and a Supervisor, Config, Mgt. Info.
copy of the Calculation to the Responsible Organization (s) identified in Mc. Nucl. Operations Eng. (Original) w/ attach Part til on the Calculation F4 view form.
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r N CALCULATION REVIEW catewsw Page 1 of 2 cALCWAfload be0/h V, M97 0098 REV. 3 j
PART I -
DESIGN ASSUMPTION / INPUT REVIEW: APPLICABLE 6 Y.: O No The following organizations have reviewed and concur with the design assumptions and inputs identified for this calculation:
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LO/L9 9Y
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Nuclear Plant Technical Support w
System Engr 5+*****
M /3[97 Nuclear Plant Operations
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RESULTS REVIEW: APPLICABLE [Yes O No The following organizations have reviewed and concur with the results of this calculation and understand the actions which the organizations must take to implement the results. /h7 10 Nuclear Plant Technical Support
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System Engr N
Nuclear Plant Operations s,v.i..t_
y Nuclear Plant Maintenance O Yes N/A Nuclear Licensed Operator Training O Yu 7NM Manager, Site Nuclear Services,
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M97 0098 REV. 3
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PART lil. CONFIGURATION CONTROL: APPLICABLE T Yes O No The fellowing is a list of Plant procedures / lesson plans /other documents and Nuclear Engineering calculations which require updating based on calculation results review:
)
Document Date Reauired Responsible Oraantration EM 22b 11/15/97 EMERGENCY PLANNING (STEPHENSON)
Upon completion, forward a copy to the Manager, Nuclear Regulatory Assurance Group for tracking of actions if any items arc identified in Part Ill. If calculations are listed, a copy shall be sent to the original file and the calculation log updated to reflect this impact.
PART IV. NUCLEAR ENGINEERING DOCUMENTATION REVIEW The responsible Design Engineer must thoroughly review the below listed documents to assess if the calculation requires revision to these documents. If 'Yes,' the change authorizations must be listed below and issued concurrently with the calculation.
Enhanced Design Basis Document - O Yes [No l'C'8 Vendor Qualification Package (vo)
FSAR O Yes n'a *>
Topicai Design Basis Doc. O Yes #
"c'i improved Tech. Specification O Yes No n '"* 's E/SOPM O Yes o f'C'i improved Tech. Spec Bases C Yes No n*85 Othet gocuments reviewed:
Config. Mgmt. Info, Systera C Yes [No ICO'8' NIY O Yes O No ecmo poc. maim =:n Analysis Basis Document O Yes No "C*l O Yes O No scow ooc. mum.cu Dssign Basis Document O Yes d o "c'i O Yes O No A,3penden R Fwe Study 0YesINol'C88 O Yes O No icoua poc num co Fue Hazardous Analysis OYesUNo"ce>
0 Yes O No icum ooc mum cn rdPA Code Conformance Docurtwra O Yes No <'ce' O Yes O No ecm a ooc unmuce PART V. PLANT REVIEWS / APPROVALS FOR INSTRUMENT SETPOINT CHANGE PRC/DNPO approvalis required if a setpoint is to be physically changed in the plant through the NEP 213 process.
PRC Review Required O Yes YNo PRC Chairman
/Date DNPO Review Required C Yes No g
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/Date c4KN ewo, e i
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ENGINEERING INFORMATION RECORD M [,Rg,A,T,0 g E Documentidentifier 51 5000519 03 Title
. BORON DtLUTION SY HOT LEG INJECTION PREPARED BY:
REVIEWED BY:
Name J.D. CARLTON Name rG.J. WISSINGER Signatur i
(4d D*te 10/14/97 Signature w)
^ vADate 10/14/97 1
XL-Technical M ager Statement: Initials 1
V Reviewer is Independent.
Remarks:
THE.51. DOCUMENT DESCRIBES THE ECCS REQUIREMENTS FOR MANAGEMENT OF CORE BORON CONCENTRATION POST LOCAS. HOT LEG INJECTION BY REVERSE FLOW THROUGH THE DECAY HEAT DROP LINE IS EVALUATED AS ARE DUMP TO-SUMP AND NOZZLE GAP LEAKAGE.
PACES 4,32 AND 33 ARE THE ONLY PAGES THAT CHANGE FOR THIS REVISION.
PLEASE SEE ATTACHED CHANGE PAGES.
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