ML14066A479

Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event

ML14066A479
Person / Time
Site:	Byron, Braidwood
Issue date:	03/04/2014
From:	Kanavos M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BW140017, IR-13-001, IR-13-003
Download: ML14066A479 (77)

Text

March 4, 2014 BW140017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event

Reference:

Braidwood Station, Units 1 and 2, NRC Integrated Inspection Report 05505000456/2013003; 05000457/2013003 and 07200073/2013001, dated August 14, 2013 In the referenced NRC Inspection Report, Braidwood Station received a Green Finding with associated Non-Cited Violation (NCV)05000456/2013003-08; 05000457/2013003-08, "Failure to Account for PZR [Pressurizer] PORV [Power Operated Relief Valve] Accumulator Leakage During Hot Standby and Subsequent Cooldown Period Following a Postulated Earthquake."

Specifically, Region III Inspectors concluded that Braidwood Station failed to adequately evaluate PZR PORV air accumulator leakage during a natural circulation cooldown event. This performance deficiency was attributed to the inappropriate reliance on the nonsafety-related Chemical & Volume Control System (CVCS) Auxiliary Spray to provide the Reactor Coolant System (RCS) depressurization credited safety function.

the course of the a significant amount of discussion took regarding assumptions in the current basis the Station circulation cooldown event resulting from a seismic event with subsequent loss of offsite power.

The specific item of debate was whether or not CVCS Auxiliary Spray could be credited for RCS pressure control and depressurization, in lieu of the PZR PORVs, during a natural circulation cooldown. This depressurization is necessary to reduce RCS pressure to allow the Residual Heat Removal (RHR) System to be utilized for the subsequent RCS cooldown to cold shutdown conditions.

March 4, 2014 U. S. Nuclear Regulatory Commission Page 2 As noted in the inspection report, as part of the corrective action for this NCV, Braidwood Station would revise procedures to reflect the appropriate PZR PORV accumulator air leakage acceptance criteria and seek clarification and confirmation from the NRC's Office of Nuclear Reactor Regulation (NRR) regarding the acceptable analysis assumptions used in the natural circulation cooldown event. to this letter presents a discussion with associated documentation which demonstrates that the NRC approved crediting CVCS Auxiliary Spray for RCS pressure control during a natural circulation cooldown event for Braidwood and Byron Stations during the initial licensing process.

Exelon Generation Company, LLC (EGG) respectfully requests that the NRC provide a response to this letter by March 6, 2015, and either affirm the stated position in or describe why the stated position is not consistent with the Braidwood Station and Byron Station CLB for a natural circulation cooldown event. EGC appreciates the NRC's review of this issue.

If you have any questions regarding this letter, please contact Mr. Phillip Raush, Regulatory Assurance Manager, at (815) 417-2800.

Respectfully, Mark E. Kanavos Site Vice President Braidwood Station : Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event

Attachment 1 Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event

1.0 INTRODUCTION

In the referenced NRC Inspection Report (i.e., Reference 1), Braidwood Station received a Green Finding with associated Non-Cited Violation (NCV)05000456/2013003-08; 05000457/2013003-08, "Failure to Account for PZR [Pressurizer] PORV [Power Operated Relief Valve] Accumulator Leakage During Hot Standby and Subsequent Cooldown Period Following a Postulated Earthquake." Specifically, Region III Inspectors concluded that Braidwood Station failed to adequately evaluate PZR PORV air accumulator leakage during a natural circulation cooldown event. This performance deficiency was attributed to the inappropriate reliance on the nonsafety-related Chemical & Volume Control System (CVCS) Auxiliary Spray to provide the Reactor Coolant System (RCS) depressurization credited safety function.

During the course of the subject inspection, a significant amount of discussion took place regarding the assumptions in the current licensing basis (CLB) for the Braidwood Station natural circulation cooldown event resulting from a seismic event with subsequent loss of offsite power.

The specific item of debate was whether or not CVCS Auxiliary Spray is credited in the CLB for RCS pressure control, in lieu of the PZR PORVs, during a natural circulation cooldown. This depressurization is necessary to reduce RCS pressure to allow the Residual Heat Removal (RHR) System to be utilized for the subsequent RCS cooldown to cold shutdown conditions.

The following discussion reviews system design and licensing basis requirements for the Residual Heat Removal (RHR) System; and identifies the associated CLB historical documentation. Embedded in these requirements and documentation are discussions that address use of the CVCS Auxiliary Spray during the RCS depressurization in preparation for use of the RHR system.

The purpose of this review is to clearly establish the current licensing basis and analysis assumptions for the natural circulation cooldown event and gain NRC concurrence of the conclusions stated below; or have the NRC describe its interpretation of the CLB based on the attached documentation to eliminate future uncertainty.

2.0 REGULATORY GUIDANCE 2.1 SRP Section 5.4.7 and BTP RSB 5-1 (Enclosure A) regulatory system are

• ",,,,.\.A U"" Heat Removal Enclosure A contains SRP Section 5.4.7 and BTP RSB 5-1 in their entirety; however, the following excerpts from BTP RSB 5-1 are specifically noted:

Attachment 1 Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event Functional Requirement A.1 (on page 5.4.7-12) states:

The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-grade systems. These systems shall satisfy General Design Criteria 1 through 5.

Section H, "Implementation," defines three classes of plants, each with different implementation requirements for compliance with the BTP. It should be noted that both Byron Station and Braidwood Station are Class 2 plants.

Table 1, "Possible Solution for Full Compliance with BTP RSB 5-1 and Recommended Implementation for Class 2 Plants," outlines compliance options for the BTP requirements.

Column 1 of this table lists the "Design Requirements of BTP RSB 5-1." Item 1.a addresses "Capability Using Only Safety Grade Systems." Column 2 of the table, "Process and [System or Component]," (on page 5.4.7-16) specifically addresses "Depressurization (Pressurizer auxiliary spray or power-operated relief valves)." The associated "Recommended Implementation for Class 2 Plants" (Le., column 4) for this item states the following:

Compliance will not be required if a) dependence on manual actions inside containment after SSE or single failure or b) remaining at hot standby until manual actions or repairs are complete are found to be acceptable for the individual plant. 1/

3.0 INITIAL LICENSING BASIS CHRONOLOGY During the initial licensing discussions for Braidwood and Byron Stations, numerous pieces of docketed correspondence specifically addressing compliance with the requirements of BTP RSB 5-1 were exchanged between EGC (then Commonwealth Edison Company) and the NRC.

The following excerpts from docketed correspondence are specifically noted (note that some of the text has been underlined for emphasis):

3.1 May 1979, FSAR Question 212.6 (Enclosure B)

FSAR Question 2.6 documents the following NRC question:

following are noteworthy excerpts from the EGC response to this question:

Byron/Braidwood is subject to the technical requirements of RSB 5-1 as they apply to Class 2 plants. Onlv partial compliance with the technical position is required where manual actions or repairs can be demonstrated to be an acceptable alternative to strict compliance.

The safe shutdown design basis for Byron/Braidwood is hot standby. The functional requirements of RSB 5-1 impose the following assumptions on the system(s) used to go to

Attachment 1 Clarification of licensing Basis Assumptions for a Natural Circulation Cool down Event cold shutdown within a reasonable amount of time, provided that limited manual actions, as allowed by the recommended implementation for Class 2 plants, are performed.

Depressurization can be accomplished using portions of the chemical and volume control system. Either boric acid from the boric acid tanks or refueling water can be used as desired for depressurization with the flow path being via the centrifugal charging pumps and auxiliary spray valve to the pressurizer.

3.2 June 1980, FSAR Question 212.47 (Enclosure C)

FSAR Question 212.47 documents the following NRC question:

The NRC Regulatory Requirements Review Committee has recently approved a new staff position (BTP RSB 5-1) for the residual heat removal system. The technical requirements of this position for your plant are described below. Your response to these requirements should be in sufficient detail to enable the staff to review your compliance. System parameters assumed should be the most limited parameters allowed by Technical Specifications.

The NRC question further went on to note all Functional Requirements listed in BTP RSB 5-1.

EGC's response to this question simply referenced the responses to FSAR Questions 212.6 and 212.154.

3.3 September 1982, FSAR Question 212.154 (Enclosure D)

FSAR Question 212.154 documents the following NRC question:

The response to 212.47 references the response to 212.6. The response to 212.6 does not cover aI/ of the concerns of Q212.47. Provide a response to 212.47 that addresses each item of concern as listed in Q212.47.

The following are noteworthy excerpts from the EGC response to this question:

In the response to FSAR Question 212.1 Section 3, "Cold Shutdown Scenario, EGC reiterated the previous response noted in FSAR Question 212.6 above in Section 3.1).

The safe shutdown design basis for Byron/Braidwood is hot standby. The functional requirements of RSB 5-1 impose the following assumptions on the system(s) used to go to cold shutdown: a loss of offsite power, the most limiting single failure, and that only safety grade systems are available. Under these conditions, the plant is capable of being taken to cold shutdown within a reasonable amount of time, provided that limited manual actions, as allowed by the recommended implementation for Class 2 plants, are performed.

Attachment 1 Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event In the response to FSAR Question 212.154, Section 3.e, "Depressurization," EGC stated:

As stated above, there are two methods of depressurization available to the operator: either the CVCS auxiliary spray; or the PORVls. The centrifugal charging pumps in the CVCS are Seismic Category I pumps and are powered from the ESF buses. The auxiliary spray valve is an air-operated valve as are other valves in the flow path. In the event of a seismic event or loss of offsite power event where air is lost to the valves, every effort will be made to either open the valves with a portable gas cylinder or load the air compressors onto the emergency buses.

As an alternative, depressurization could be accomplished by discharging RCS inventory from the pressurizer to containment via the pressurizer power-operated relief valves. /I In the response to FSAR Question 212.154,Section III, "Depressurization," Items A and B, EGC stated:

A. Auxiliary spray valve 8145 - This normally closed valve fails closed on loss of air. In this case, 8145 can be opened by using a portable nitrogen bottle. If 8145 is stuck closed as a result of a single failure, the redundant pressurizer power operated relief valves can be used to depressurize the RCS by discharging the pressurizer inventory to the pressurizer relief tank.

B. Charging valves 8146 and 8147 - These valves fail open on loss of air. In this case, 8146 and 8147 can be closed by using portable nitrogen bottles. If either is stuck open, the redundant pressurizer power-operated relief valves can be used to depressurize the RCS by discharging the pressurizer inventory to the pressurizer relief tank.

3.4 February 1982, Byron Station Safety Evaluation Report (Enclosure E)

The NRC issued the Byron Station Safety Evaluation Report (SER) (NUREG-0876) in February 1982. The NRC subsequently issued the Braidwood Station SER (NUREG-1002) in November 1983. Much of the Braidwood SER is identical to the Byron SER; therefore, much of the Byron SER text was not repeated in the Braidwood SEA. SER Section 5.4.3 discusses the Residual Heat Removal System. The following excerpt found on page 24 discusses natural circulation conditions:

~e a either closing of a single valve in the supply line or opening of one of the several valves in lines connected to the supply line. If manual actions to correct for such failures were not successful, a backup method of depressurization would involve opening either of the seismic Category I PORVs of the pressurizer that discharge to the pressurizer relief tank. To reduce the potential for a higher than normal containment temperature and humidity, the applicant has stated that the procedures for depressurization using the PORVs will include precautions to ensure integrity of to rtQ,nr,;;:.c 4

Attachment 1 Clarification of licensing Basis Assumptions for a Natural Circulation Cooldown Event also requires confirmation that the PORVs will function in the environment that is expected in achieving cold shutdown.

The staff concludes that with satisfactory resolution of the confirmatory items described above, [i.e., see Attachment E discussion on referencing the Diablo Canyon tests on page 5-231 the Byron station meet[sl the requirements of BTP 5-1.

3.5 October 14, 1988, Response to NRC Natural Circulation Questions (Enclosure F)

In a letter from R. A Chrzanowski (Commonwealth Edison) to T. E. Murley (NRC), "Request for Additional Information Response," dated October 14,1988, Commonwealth Edison documented their response to NRC questions related to the natural circulation capability comparison between Byron and Braidwood Stations and the Diablo Canyon Station.

In Question 12, the NRC asked the following:

Page 9 of response [referring to the response provided on page 9 of FSAR question 212.154] - Are personnel trained to handle portable gas cylinders?

The following response was provided:

No specific training is required for operation of portable gas cylinders. However, this is an accepted maintenance and testing activity for non-routine operation of select valves.

3.6 November 4, 1988, NRC Natural Circulation Safety Evaluation (Enclosure G)

In a letter from S. P. Sands (NRC) to H. E. Bliss (Commonwealth Edison), "Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2, Natural Circulation Cooldown," dated November 4, 1988, the NRC issued a safety evaluation (SE) addressing natural circulation cooldown at Byron and Braidwood Stations. In Section 3.0, "Conclusion," the NRC stated the following:

The staff assessed the capability of Byron/Braidwood, to meet the requirements of RSB BTP 5-1. We have identified and evaluated the plant parameters that may affect application of the Diablo Canyon natural circulation test results to Byron/Braidwood.

• hnn.ff<:>ic and our C7VCIIUQ!lIVi demonstrated that t$V,rOf1!/ti/'aU:IW()Oa and that It should be noted that earlier in the SE, in the "Depressurization" section (on page 6), the NRC stated:

The Diablo Canyon test demonstrated that the RCS could be depressurized from cooldown conditions to the RHR initiation pressure under natural circulation conditions using the pressurizer auxiliary spray and/or pressurizer power operated relief valves.

5

Attachment 1 Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event At Byron/Braidwood, depressurization may be accomplished through the use of the pressurizer PORVs or the pressurizer auxiliary spray. However, the pressurizer auxiliary spray is not safety grade. and thus is not available for the RCS depressurization per BTP RSB 5-1.

It is EGCs position that the NRC erred when making this statement (i.e., the underlined statement above) as the statement suggests BTP RSB 5-1 does not permit use of the pressurizer auxiliary spray for RCS depressurization since the system is not safety grade. The above statement is true for Class 1 plants; however, Byron Station and Braidwood Station are Class 2 plants as acknowledged in November 4, 1988 cover letter transmitting the SE.

As noted in Section 2.1 above, BTP RSB 5-1, Table 1, "Possible Solution for Full Compliance with BTP RSB 5-1 and Recommended Implementation for Class 2 Plants," clearly allows an exception to full compliance with the BTP for Class 2 plants regarding the use of only safety grade systems; i.e.,

Compliance will not be required if a) dependence on manual actions inside containment after SSE or single failure or b) remaining at hot standby until manual actions or repairs are complete are found to be acceptable for the individual plant. 1/

It is clear from sequence of documentation outlined above; manual actions to support operation of pressurizer auxiliary spray were thoroughly discussed and vetted by the NRC during the initial licensing process for Byron and Braidwood Stations. Again, as stated in the response to FSAR Question 212.154:

In the event of a seismic event or loss of offsite power event where air is lost to the valves.

every effort will be made to either open the valves with a portable gas cv/inder or load the air compressors onto the emergency buses.

4.0 CONCLUSION

Based on the documentation presented above, the use of manual actions to ensure availability and use of the CVCS Auxiliary Spray for RCS depressurization during a natural circulation cooldown were specifically addressed and documented. It is also clear that BTP RSB 5-1, Table 1, allows use of non-safety systems, supported by manual actions, to achieve cold shutdown in lieu of "Using Only Safety Grade Systems." It appears that the NRC had an oversight when they stated: lithe pressurizer auxiliary spray is not safety grade, and thus is not available for the RCS depressurization per BTP RSB 5-1, in the November 4, 1988 SE.

II It is EGC's conclusion that the current licensing basis for Byron and Braidwood Stations assumes that it is acceptable to credit the use of the CVCS Auxiliary Spray for RCS depressurization during a natural circulation cooldown, following a seismic or loss of offsite power event.

EGC respectfully requests that the NRC provide a response to this letter by March 6, 2015, and either affirm the stated position stated above or describe why the stated position is not a

6

Attachment 1 Clarification of Licensing Basis Assumptions for a Natural Circulation Cooldown Event

5.0 REFERENCES

1 Braidwood Station, Units 1 and 2, NRC Integrated Inspection Report 05505000456/2013003; 05000457/2013003 and 07200073/2013001, dated August 14, 2013 6.0 ENCLOSURES A. Standard Review Plan (SRP) (i.e., NUREG-0800) Section 5.4.7, "Residual Heat Removal (RHR) System," and associated Branch Technical Position (BTP) RSB 5-1, "Design Requirements of the Residual Heat Removal System" B. FSAR Question 212.6 C. FSAR Question 212.47 D. FSAR Question 212.154 E. Byron Station Safety Evaluation Report (NUREG-0876), Section 5.4.3, "Residual Heat Removal System" F. Letter from R. A. Chrzanowski (Commonwealth Edison) to T. E. Murley (NRC), "Request for Additional Information Response," dated October 14,1988 G. Letter from S. P. Sands (NRC) to H. E. Bliss (Commonwealth Edison), "Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2, Natural Circulation Cooldown," dated November 4, 1988

Enclosure A Standard Review Plan (SRP) (Le., NUREG-0800) Section 5.4.7, "Residual Heat Removal (RHR) System," and associated Branch Technical Position (BTP) RSB 5-1, "Design Requirements of the Residual Heat Removal System"

. NUREG-OSOO (Form~rly NUREG-75/087l U.S. NUCLEAR REGULATORY COMMISSION STANDARD REVIEW PLAN OFfiCE OF NUCLEAR REACTOR REGULATION

• 5.4.7 RESIDUAL HEAT REMOVAL (RHR) SYSTEM REVIEW RESPONSIBILITIES Primary - Reactor Systems Branch (RSB)

Secondary - None I. AREAS OF REVIEW Rev. 2 - July 1981 USNRC STANDARD REVIEW PLAN StBndard review plans me prepared for the guidance of the Office of Nuclear Aeector Regulation staff respo.lsible for the raview of

_3PlllicatioRs 10 construct and operate nuclellr power plants. Thesa documents are made available to the public a:s part of the Commission's to inform the flIJGlear imfustrv and the general public of regulatory proc~diJree and pOlicies. Standard review i:J!ans are not for or Commission's regulations and with them i:s not required. The standard review the Content Safety tor .Nuo:::!ear Plants_

Published standard review De revised peric as approprillte, to accommodate comments andto ref!ecf new informs-tlon and experience.

COry'lmetlts and sU9ge$tiOn!l for l~provemer:'t will be considered: and should be sent to the. U.S.Nuclear. Regulatory.Coll)misJiion, Office of Nuclear Reactor* Regulatto,n, WashIngton. D_C, 20555. ... . "

1'.- t removed in the heat exchangers is transpm'ted to the ultimate heat sink by the component coolingwateY' or service wate.r system. In PWRs, the RHR system is I

also used to fi11, drain, and remove heat from the refueling canal during refueling operations, to circulate coolant through the core during plant startup prior to ReS pump operation, and in some to provide an auxil iary pressurizer spray.

The RHR system in BWRs is typically composed of four sUbsystems. The contain-ment heat removal and low pressure emergency core cooling subsystems are discussed in SRP Sections 6.2.2 and 6.3. The shutdown cooling and stearn condensing (via ReIe) subsystems are covered by this SRP section. These subsystems maKe use of the same hardware, consisting of pumps, piping, heat exchang~rs, valves, monitors, and* controls. In the shutdown cooling mode; the BWR RHR system can also be used to supplement spent fuel pool co01;ng. As in the PWR, the low pressure RHR piping is protected from high RCS pressure by isolation valves_

The steam condensing mode of RCIC operation in BWRs (when included in the plant design) provides an alternative to the main c:ondenseror normal RCIC mode of operation during the initial cooldown. Steam from the reactor is transferred to the RHR heat exchangers where it is condensed. The condensate is piped to the suction side of the RCIC pump. The RCIC pump returns the condensate to the reactor vessel. The heat removed in the heat exchangers is transported to the ultimate heat sink by the service water *system.

Other means of removing decay heat in the event that the RHR system is inoper-able have been proposed for some BWRs .. These approaches Use some of the piping that is used for the stearn condensing mode of RCIC. These approaches are also covered by th1SSRP section.

The reactor coolant temperatures and pressure must be decreased before the low pressure RHR system can be placed in operation; therefore, the review of the decay heat removal fUnction must consider all conditions from shutdown at normal

.reactor operating pressure and temperature to the cold depressurized condition.

RSB revi eVls the requ; rements for re 1i abil ity and capabil ity of removi ng decay heat identified in NUREG-0660 (II.E.3.2 and II.E.3.3), NUREG-0718 (II.B.7), and NUREG-0737 (111.0.1.1). With respect to the staff review for compliance with Branch Technical Position RSB 5-1 (Ref. 5), the Auxiliary Systems Branch (ASB), I

. t Chemical Engineering 8ranch (CMEB), and RSB effort is divided as follows:

1.. For BWRs, the RSB reviews the processes and systems used in the cooldown of the reactor for the entire spectrum of potentia) reactor coolant syst8m pressures and temperatures during decay heat removal.

2. For PWRs, the RSB reviews the approa~h used to meet the functional require-ments of BTP RSB 5-1 with respect to cooldown to the conditions permitting operation of the RHR system, Since an alternate approach to that normally used for cooldown may be specified, the reviewers identify a11 components and systems used. The CMEB has primary review responsibility for the review of the pertinent portions of the eves (SRP Section 9.3.4). The ASB, as part of its primary review Y'esponsibility for SRP Sections 10.3 and 10.4.9 reviews the atmospheric dump valves.and the source for auxiliary feedwater, respectively, for conformance to STP RSB 5-1. The RSB reviews the pressurizer relief valve and ECCS. if u~ed.In addition, the RSB*

"reviews the tests and support; ana1ysi sconcerni ng mixi ng of borated down rcuhiti as red 5.4.

'3. For both PWRs and BWRs, the ASS reviews the component cooling or service water systems that transfer decay heat from the RHR system to the ulti*

mate heat sink as part of its primary review responsibility for SRP Sections 9.2.1 and 9.2.2.

4. The RSB reviews the design and operating characteristics of the RHR system with respect to its shutdown and long-term cooling function.

Where the RHR system interfaces with ,other systems (e-9., ReIe system, component cooling water system) the effect of these systems on the RNR system is reviewed. Overpressure protection provided by the va1ving ,

, between the ReS and RHR system is arso reviewed.

In addition: the Reactor Systems Branch will coordina+-.e eVd'luations of other branches that interface with th~ overall review of th~ RHR system as follows:

The Containment Systems Branch,verifies that portions of the RHR system pene-trating the containment barrier are designed with acceptable isolation features to maintain containment integrity for all operating conditions including acci-dents as part of its prima'-'y review responsibility for SRP Section 6.2.4; The structural Engineering Branch (SEE) determines the aCi.eptability of the design analysis, procedures and criteria used to establish the ability of seismic Category I structures housing the system and supporting systems to withstand the effects of natural :->henomena such as safe shutdown earthquake (SSE); the probable maximum flood (PMF), and tornado missiles as part of its primary review responsibility for SRP Sections 3.3.1,3.3.2, 3.5.3, 3.7.1 thru 3.7.4, 3.8.4 and 3.8.5. The Materials Engineering Branch (MTEB) verifies that inser-vice inspection requirements are met for system components as part of its I primary review responsibility for SRP Section 6.6 ,and, upon request, verifies ,

the compatibility of the materials of construction with service conditions as I part of its primary )'eview responsibility fat SRP Section 6.1. The Mechanical Engin~ering B~anch (MEB) determines that the components, piping and structures ,!i are de'signed and tested in accordance with applicable codes and standards as part of its primary review responsibility fOl' SRP Section's 3.9.:!. through*

3.9.3. The MEB also determines the acceptability of the seismic and quality I

group classifications for system components as part of its primary review responsibi 1 ity fot' SRP Sections 3.2.1 and 3.2.2 The effects of pipe breaks inside and outsid2 of containment, such as pipe whip and jet impingement, are reviewed by MEB and ASB as part of their primary review responsibilities for SRP Sections 3.6.2 and 3.6.1, res~ectively. The MEB also reviews adequacy of the inservice testing program of pumps and valves as part of its primary review responsibility for SRP Section 3.9.6. The Procedure and Test Review Branch (PTRB) reviews the proposed preoperational and stal'tup te51; programs to confirm that they are in conformance with the intent of Regulatory Guide 1.68 as part. of its primary review responsibility for SRP Section 14.2. The PTRB alSO has primary review responsibility for Task Action Plan items lI.K.l (C.1.10) of NUREG-0737 (OLs only) and I. C. 6 of NUREG-0718 (CPs only) regarding procedures to ensure that system operability status is known. The ASB reviews flood protection as part of its primary review responsibility for SRP Section 3.4.1. The ASS identifies the structures systems and components to be protected against externally generated missiles and reviews the adequacy of protection against such missiles as part of its primary review responsbility for SRP section 3.5:1.4 and 3.5.2. -The ASB also reviews protection against internal'lygenerated mis511es both inside and outside of to~tainment as part of its primary review responsibi 1 ity for SRP Sections 3.5. L 1 and 3.5.1. 2.

3 Rev. 2 July 1981

Th~ Power Systems Branch (PSB) identifies the safety-related electrical loads and determines that power systems supplying motive or control power for ,the RHR system meet acceptable criteria and will perform these intended functions during all plant operating and accident conditions as part of its primary review responsibility for SRP Sections 8.1, 8.2, 8.3.1, and 8.3.2. The Instru-mentation and Control Systems Branch (leSB)} as part of its primary review responsibility forSRP Sections 7.1 and 7.4 reviews the instrumentation and control systems for the RHR system to determine that it will perform its design function as required and conform to all applicable acceptance criteria.

The ress also reviews the provisions taken to meet GOC 19 with respect to, equipment outside of the control room for hot and cold shutdown. The RadlO-logical Assessment Branch (RAB) has primary review responsibility for SRP Section 12.1 through 12.5 including Task Action P1an items I1.B.2 of NUREG-0737 and NUREG-0718 which involve a radiation and shielding design review and corrective actions taken to ensure adequate aCcess to vital areas and protec-tion of safety equipment (CPs and OLs). The review for Fire Protection, Technical Specifications, and Quality Asurance are coordinated and performed by the CMEB, LicenSing Guidance Branch (lGB) and Quality Assurance Branch (QAB) as part of their primary review responsibility for sRp Sections 9.5,1, 16-0 and 17.0, respectively. .

For those areas of review identified above as being reviewed as part of the primar,Y review respons ibi 1ity Jf other branches, the acceptance ct'iteri a necessdry for the review and their methods of application are contained in the refererlced SRP Sect ion of the corresponding primary branch.

II. . ACCEPTANCE CRITERIA The Reactor Systems Branch acceptance criteria are based on meeting the require-ments of the fol1owing regulations:

A. General Design Criterion 2 with respect to the seismic design of sy~tEm~,

structures and components' whose fai.lure could cause an unacceptable reduction in the capability of the residual heat removal system. Accept-ability is based on meeting position C-2 of RegUlatory Guide 1.29 or its equivalent. .

8.- General Design Criterion 5 which requires that any sharing among nuclear' power units of strllctu~'es, systems ahd components important to safety will not significantly impair their safety function.

C. General Design Criterion 19 with respect to control room requirements for normal operations and shutdown, and; D. General Design Criterion 34 which specifies requirements for a residual heat removal system.

Specific criteria necessary to meet the requirements of General Design Criteria 2. 5, 19, and 34 are as follows:

1. The system or systems are to satisfy the functional, isolation, pressure relief, pump protectiori and test reqUirements specified in Branch Technical Position RSB 5-1. . *

2. Interfaces between the RHR system and RCIC and component or' systems s d

with; and provides proper support (where required) for, the other. In relation to these and other shared systems (e.g., emergency core cooling and containment heat removal systems), the RHR system must conform to GDC5.

3. The requirements for the reliability and capability of removing decay heat under the following Task Action Plan items must also be satisfied:

a.Meeting Task Action Plan item II.E.3.2 of NURfG-0660 which involves systems reliability. NRR will conduct a generic study to assess the capability and reliability of shutdown heat removal systems under various transients and degraded plant conditions including comp1ete loss of all feedwater. Deterministic and probabilistic methods will be used to identify design weaknesses and possible system modifica-tions that could be made to improve the capabiJity and reliability of these systems under all shutdown conditions. (CPs and Ols).

Specific requirements wi! 1 be based on the results of this study.

b. Meeting Task Action Plan item II.E.3.3 ~f NUREG~0660 which involves a coot'dinated study of shutdown heat removal requirements, An effort to evaluate shutdown heat removal requirements in a comprehen-sive manner is required, thereby permitting a judgment of adequacy in terms of overall system requirements. As part of this project j .

NRR will conduct a study to assess the desirability of and possible requirement for a diverse heat-removal path, such as feed and bleed~

particularly if all secondar'y-side cooling is unavai1able. The NRC staff will work with the recently established ACRS Ad Hoc Subcommit-tee on this matter to develop a mutually acceptable overal1 study program. (CPs and OLs). Specific requirements will be based on the results of this study.

c.Meeting Task Action Plan item II.B.B of NUREG-0718 (Ref. 7) which involves description by the applicants of the degree to which the designs conform to the proposed interim rule on degraded core accidents. (CPs only)

d. Meeting Action Plan item 111.0.1.1 of NUREG-0737 (Ref. 8) and

. NUREG-0718 (Ref. 7) which involves primary coolant sources outside of containment (CPs and OLs).

4. When the RHR system is used to control or mitigate the consequences of an accident, it must meet the design requirements of an engineered safety featUre system. This includes meeting the guidelines of Regulatory Guide 1.1 regarding net positive suction head.

III. REVIEW PROCEDURES The procedures below are used during the construction permit (CP) review to assure that the design criteria and bases' and the pre1iminary design as set forth in the PY'eliminary Safety Analysis R~port meet the acceptance criteria given in subsection II. . ..

FOfoperating license (OL) reviews, the procedures are utilized toverffy that the initial design criteria and bases have been appropriately implemerited ih the final design as forth in the Final Safety Analysis Report. The OL 5 .. 4. Rf:HI. 2 - *1981

review also includes the proposed technical specifications, to assure, that they are adequate in regard to limiting conditions of operation and periodic surveillance testing. .

As noted in sUbsections I and II, the R5B review for PWRs is 1imited to the row pressure - low t!;mperature RHR system. Fot' BWRs, the review is to include all of the systems used to transfer residual heat from the reactor over the entire range of potential reactor Goo1ant temperatures and pressures. The fonowing steps are to be applied by the reviewer for the appropriate systems, depending on whether a PWR or BWR is being reviewed. These steps should be adapted to CP or OL reviews as appropriate. .

1. Usin9 the description given in the applicant's Safety Analysis Report (SARl. including component lists and p~rfDrmance specifications, the reviewer determines that the system(s) piping and instrumentation are such to allow the system(s) to operate as intended, with or without offsite power and given any single active component failure. This is accomplished by reviewing the piping and instrumentation diagrams (P&IDs) to confirm that piping arrangements permit the required flow paths to be achieved and that sufficient process sensors are available to measure and transmit required information. A failure modes and effects ana1ysis (or.

similar system safety analysis) provided in the SAR is used to determine conformance to the single failure criterion. .

2. Using the comparison tables of SAR Section 1.3, the RHR system is compared to designs and capacities of such systems in similar plants to see that there are no unexplained departures from previously reviewed plants.

Where possible, comparisons should be made with actual performance data' from similar systems in operating plants. "

3. . From the system description Clod P&IDs. the reviewer determines that the*

isolation requirem&nts of Branch Technical Position RSB 5-1 (Ref. 5) are satisfied~ , I

4. The reviewer determines that the RHR system design, has provisions to prevent damage to the RHR pumps in accordance with Branch Technical Position RSB 5-1 (Ref. 5). The reviewer checks the isolation valves in the suction line for potential closure, NPSH requirements, pump t'unout, and potential loss of mini flow line during pump testing. If operator action is required to protect the pumps, the reviewer evaluates the instrumentation required to alert the operator and the adequacy of the time frame for operator action.

5. Using the system process diagrams, P&1Ds, failure modes and effects analysis; and component performance specifications, the reviewer deter-mines that the system(s) has the capacity to bring the reactor to conditions permitting operation of the RHRsystem in a reasonable period of time. assuming a single failure of an active component with only either onsite or offsite electric power available. For the purposes of this review, 36 hou'tsis considered a reasonable time period. The ASS is responsible 'for the review of the initia1cooldown phase for PWRs.

Therefore, this )~e\liew effort is to be coordinated.with that branch. For the purposes of the review of both PWRs and BWRs, only the operation of

. safety grade equipment is to he. assumed." .

6. The cool down function is to be reviewed to determine if it can be per-formed from the contro1 room assuming a s'ing1e failure of an active component. with only either onsite or offsite electric power available.

Any operation required outside of the control room is to be justified by the applicant. Like Item 5, the initial cooldown for PWRs is to be reviewed by ASB.

7. By reviewing the system description and the P&IOs, the reviewer confirms the RHR system satisfies the pressure relief requirements af Branch Technical Position RSS 5-1 (Ref. 5).

8. By reviewing the piping arrangement and system description of the RHR system, the reviewer confirms that the RHR system meets the requirements of GDC 5 (Ref. 2) concerning shared systems.

9. The RSB reviewer contacts the ASB reviewer in conjunction with his review of the RHR system heat sink and refueling system interaction to inter-

. change info~mation and assure that the reviews are consistent with regard

. to the interfacing parameters. For example, the ASS review determin~s the maximum service or component cooling water temperature. The RSB, reviewer then reviews. the RHR system description to determine that this maximum temperature has been allowed for in the RHR system design.

10. The RSB reviewer contacts his counterpart in the leSB to obtain any needed information from their review. Specifically, leSS confjrms that automatic actuation and ~emote-manual valve controls are capable of performing the functions required, and that sensor and monitoring pro-visions are adequate. The instrume~tation and controls of the RHR system are to have sufficient redundancy to satisfy the single failure criterian.

11. TheRSB reviewer contacts his counterpart in CSB ~o that the information needed concerni ng the; r revi ews wi 11 be interchanged.

12. The RS8 l'eviewer contacts his counterpart in PTRB to discuss any special test requirements and to confirm that the proposed preoperational test program for the RHR system is in conformance with the intent of Regulatory Guide 1.68. .

13. The proposed plant technical specification~ are reviewed to:

a. Confirm the suitability of the limiting conditions of operation, including the proposed time limits and reactor operating restric-tions for periods when system equipment is inoperable due to repairs and mafntenance~

b. Verify that the frequency and scope of periodic surveillance testing is adequate.

14. The reviewer' contacts the SEB reviewer to confirm that the systems employed to i'emove residual heat are housed in a structure whose design and d~sign criteria provide adequate protection against wind) tornadoes, floods., and missiles, as appropriate.

15. FOI' PWRs, the reviewer, confirms. that theauxi 1i ary Jeedwater supply

~at1sfies the requirements of Br~nch Technical Position RSB 5~1.

.4 Rev. 2 ly

16. The RSB reviewer provides information to other branches in those areas where the RSB has a review responsibility that is not explicitly covered in steps 1-15 above. These additional areas 'of review responsibility include:

a. Identification of engineered safety features (ESF) and safe shutdown electrical loads, and verification that the minimum time intervals for the connection of th ESF to the standby power systems are satisfactory.

b. Identification of vital allxiliary systems associated with the RHR system and determination of cooling load functional requirements and minimum time interval$.

c. Identification of essential components associated with the main steam supply and the auxiliary feedwater system that are required to operate during and following shutdown.

17. The RSB review evaluates the applicant responses to the following Task Action Plan items:

a. JI.E.3.2 of NUREG-0660 (CPs and Ols)

b. JI.E.3.3 of NUREG-0660 (CPs and OLs)

c. II. B. B of NUREG-0718 (CPs only)

d. 11.0.1.1 of NUREG-0737 and NUREG-0718 (CP~ and OLs)

IV. EVALUATION fINDINGS The reviewer verifies that the SAR contains sufficient information and his review supports the following kinds of statements and conclusions, which should be included in the staffls Safety Evaluation:

For PWRs The residual heat removal function is accomplished in two phases:

the initial cooldown phase and the residual heat removal (RHR system) operation phase. In the event of 10s5 of offsitepower, the initial phase of cooldown is accomplished by use of the auxiliary feedwater system and the atmospheric dump valves. This equipment is used to reduce the reactor coolant system temperature and pressure to values that permit operation of the RHR system. The review of the initial cooldown phase is discussed in Section_ of the SER.

The review of the RHR system operational phase is discussed below.

, The residual heat removal (RHR) system removes core decay heat and provides long-term core cooling following the initial phase of t'eactor cooldown. The scope of review of the RHR system for the plant included piping and instrumentation diagrams, equipment layout drawings) failure modes and effects analysis, and design performance specifications for ~ssential components. The review has included the applicant's proposed design criteria and design bases for the RHR system,and.h,is analysis of the adequacy of those ct'iteria

~nd.bases and the conformance of, the design to these crite~ia and Rev. 2

The staff conc1udes that the design of -t:he Residual Heat Re.moval Syst~m is acceptable and meets the requirements of General Design Criteria 2, 5, 19, and 34. This conclusion is based on the following:

(1) The app1icant has met the General Design Criterion 2 with

~espect to position C-2 of Regulatory Guide 1.29 concerning the seismic design of systems, structures and components whose fai 1ure coul d cause an unacceptable reduction in the capabil ity of the residual heat removal system; .

(2) The applicant has met the requirements of General Design Criterion 5 with respect to sharing of structure! systems and components by demonstrating that such sharing does not signifi-cantly impair' the ability of the Residual Heat Removal System to perform it safety function including in the event of an accident to one unit, an orderly shutdown and cooldown of the remaining units.

(3) The applicant has met General Design Criterion 19 with respect to the main control room requirements for normal operations and shutdown and General Design Criterion 34 which specifies i'equ1re-ments for the residual heat removal system by meeting the regulatory position in Branch Technical PosHian RSB 5-l.

In addition, the applicant has met the requirements of the following Task Action Plan Items:

(1) Task Action Plan item I1.E.3.2 of NUREG-0660 (Ref. 10) as it fa 1ates to systems capabil ity and re 1i ability of shutdown heat removal systems under various transients.

(2) Task Action Plan item Ir.E.3.3 of NUREG-0660 (Ref. 10) as it relates to a coordinated study of shutdown heat removal .

requirements. .

(3) Task Action Plan item II.B.8 of NUREG-071B (Ref. 7) as it relates to description by the applicants of the degree to which the designs conform to the proposed interim rule on degraded core accidents (CPs on1y). .

(4) Task Action Plan item 111.0.1.1 of NUREG-0737 (Ref. 8) and NUREG-0718 (Ref. 7) as they relate to primary coolant sources outside of containment (CPs and Ols).

For BWRs The residual heat removal function is accomplished in two phases:

the initial cooldown phase and a low pressure-temperature operation phase. In the event of loss of offsite electrical power, the initial coo 1down phase is accompli shed us i ng the reactor' core i so 1at ion cooling (Rete) system and the safety/ relief valves. The low pressure-temperature mode of operation is usually accomplished by .

the residual heat removal (RHR). system .. However, certain sing;le failures can render the RHR system inoperative. In that event; two of the and are

The scope of review of these systems for the plant included PJP; ng and i nstrumentat ion diagrams l_ gqiJjpment layout drawi ngs, .

failure modes and effects analysis, and design performance specifica~

tions for essential components. The review has included the applicant's proposed design criteria and design bases for these systems and his analysis of the adequacy of those criteria and bases and of the conformance of the design to these criteria and bases .

. The staff concludes that the design of the Residual Heat Removal System is acceptable and meets the requirements of General Design Criteria 2. 5, 19, and 34. This conclusion 15 based on the following:

(1) The applicant has met General DestgnCriterion 2 with rGspect to position C-2 of Regulator,)! Guide 1.29 concerning the seismic design of systems, structures and components whose failure coula cause an unacceptable reduction in the capability of the residual heat removal system.

(2) The applicant has met the requirements of General Design Criterion 5 with respect to sharing of structures, systems, and components by demonstrating that such sharing does not signifi-cantly impair the ability of the Residual Heat Removal System to perform .its safety function including in the event of an accident to one unit, an oroerly shutdown and cool down of the remaining units.

(3) The applicant has met General D~sign Criterion 19 with respect to the main control room requirements for normal operations and shutdown and General Design Criterion 34 which specifies require-ments for the residual heat removal system by meeting the regulatory position in Branch Technical Position RSB 5-1.

~.. .

In addition, the applicant has met the requirements of the following Task Action Plan Items:

(1) Task Action Plan item II.E.3.2 of NUREG-0660 as it relates to systems capability and reliability of shutdown heat removal systems under various transients.

(2) Task Action Plan item II.E.3.3 of NUREG-0660 as it relates toa coordinated study of shutdown heat removal requirements.

(3) Task Action Plan item II.B.8 of NUREG-0718 (Ref. 7) as it relates to description by the applicants of the degree to which the designs conform to the proposed interim rule on degraded core acddents (CPs only).

(4) Task Action Plan item 111.0.1.1 of NUREG-0737 (Ref. 8) and NUREG-0/18 (Ref. 7) as they relate to primary coolant sources outside of containment (CPs and OLs). .

In addition to the above criteria, the acceptability of the RHR .

,system may be based on the degree of design similarity with prev.iously approved plants. Deviations. from these criteria from other types of, RHR systems (e. g., systems that are designed to withstand reactor ant system operating or systems located entirely i de nmnt) 1 be an indi

V. IMPLEMENTATION The fo11owing is intended to provide guidance to applicants and licensees regarding the NRC staff*s plans for using this SRP section.

Except in those cases 1n which the applicant proposes an acceptab1e alter-native method for complying with specified portions of the Commission's regulations, the method described herein will be used by the staff in its eva1uation of conformance with Commission regUlations.

Imp'Iementation schedules for conformance to parts of the method discussed herein are contained in the l'eferenced BTP RSB 5-1, regulatory guides, and NUREGs. '

VI. REFERENCES L 10 CFR Part 50, Appendix A" General Design Criterion 2, IIDesign Bases for Protection Against Natural Phenomena." ,

2. 10 CFR Part 5.O~ Appendix A, General Design Criterion 5, "Sharing of Structures, Systems and Components."

3. 10 CFR Part 50, Appendix A, General,Design Criterion 19, "Control Room. 1I

4. 10 CtR Part 50, Appendix A, General Design Criterion 34, IIResidual Heat Removal. II

,5. Branch Technical Position RS85~1, "Design ReqUirements of thf! Residual Heat Removal System,fI attached to SRP Section 5.4.7. '

6. Reglatory Guide 1.29" uSeismic Design Classification. 1I

7. NUREG-071B, IILicensing Requirements for Pending Applications for Construc-tion Permits and Manufacturing License. 1I

8. NUREG-0737, lIClarification of TMI Action Plan~ Requirements. II

9. Regulatory Guide 1.1, "Net Positive Suction Head for Emergency Core Coolin9 and Containmf;ntHeat Remova1 Systems.1l

10. NUREG-0660, llNRC Action Plan Developed asa Resu1t of the TMI-2 Accident.1I

,l~

1 -I

BRANCH TECHNICAL POSITION RSB 5~1 DESIGN REQUIREMENTS OF THE RESIDUAL HEAT REMOVAL SYSTEM BACKGROUND GDC 19 states' tha.t, "A control room shal1 be provided from which actions can be taken to operate the nuclear power unit under normal conditions. If Norma1 operating conditions including the shutting down of a reactor; therefore, since the residual heat removal (RHR) system is one of several systems involved in the norm ql shutdown of all reactors, this system must be operable from the contro 1 room.

GDC 34 states that IIS uitable redundance . . . shall be provided t.o assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available), the system safety function can be accomp1ished, assumi ng ,a si ngl e fai1 ure.I I , ..

In most current plant desiyns the RHR system has a lower desi~n pressure than the reactor coolant system (ReS), is 10cated outside of contalnment and is part of the emergency core cooling system (EeeS). However, it is possible for the RHR system to have different 'design characteristics. For example, the RHR system might have the same design pressure as the ReS, or be located inside of containment. Plants which may have RHR systems that deviate from current designs will be reviewed on a case-by-case basis. The functiona1, isolation, pressure relief, pump protection, and test requirements for the RHR system are in~luded in this position.

BRANCH POSITION A.. Functional Requirements The system(s) which can be used to take the reactor from normal operating conditions to cold shutdown* shall satisfy the functional requirements listed below.

1. The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-grade systems.

These systems shall satisfy General Design Criteria 1 through 5.

2. The system(s) shall have suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities to assure that for onsite electrical power system operation (assuming offsite power is not available) and for offsite electrical power system operation (assuming onsite power is not available) the system function can be accomplished assuming a single failure.

l1!'

Processes involved in cool down are heat removal, depressuriz~tion, flow circulation, and reactivity control. The cold shutdown condition, as .

described in the Standard Technical Specifications, ref~rs to a 5ub~

critical reactor with a reactorcoo1ant temperature no greater than 200 Cl F for a PWR and 212°F for a BWR.

Rev, 2 - 1981

3. The system(s) shall be capab1e of being operated from the control room with either onlY onsUe or on1y offsite power available. In demonstrating that the system can perform its function assuming a 51ng1e failure) limited operator action outside of thp control room would be considered acceptable if suitably justified.

4. The system(s) shall be capable of bringing the reactor to a cold shutdown condition, with only offsite or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.

B. RHR System Iso1ation Requirements The RHR system shall satisfy the isolation requirements listed below.

1. The following shall be provided in the suction side of the RHR system to isolate it from the ReS.

(a) Isolation shall be provided by at least two power-operated va1ves in series. The valve positions shall be indicated in the control room.

(b) The valves shall have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR system design pressure. Failure of a power supply shall not cause any va1ve to change posit1Jn.

(c) . The valves shall have independent diverse interlocks to protect against one or both valves being open during an ReS increase

. above the design pressure of the RHR system.

2. One of the following shall be provided on the discharge side of the RHR sy~~em to isolate it from the ReS:

(a) The va1ves~ position indicators, and interlocks described in item l(a) thru l(c) above, . .

(b) One or more check valves in series with a normally closed power-operated valve. The power-operated valve position shall be indicated in the control room. If the RHR system discharge line is used for an ECCS function, the power-operated valve is to be opened upon receipt of a safety injection signal once the reactor coolant pressure has decreased below the ECCS design pressure.

(e) Three check valves in series, or (d) Two check valves in series, provided that there are design provisions to permit peri~odic testing of the check valves. for leak tightness and the testing is performed at least annua'ly~

Rev. 2 ~ Ju1y1981

C. Pressure Relief Requirements The RHR system shall satisfy the pressure reHef requirements listed below.

1. To protect the RHR system against accidental overpressurizatior. when it is in operation (not isolated from the ReS), pressure relief in the RHR system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code. The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the ReS shall be considered when selecting the pressure relieving capacity of the RHR system. For example, during shutdown cooling in a PWR with no.steam bubble in the pres-surizer, inadvertent operation of an additional charging pump or inadvertent opening of an EeeS accumulator valve should be considered in selection of the design bases.

2. Fluid discharged through the RHR system pressure re1ief valves must.

b~ collected and contained such that a stuck open relief valve will not:

(a) Result in flooding of any safety-related equipment.

(b) Reduce the capability of the Eees below that needed to mitigate the consequences of a postulated LOCA.

(c) Resu1t in a non-isolatab1e situation in which the water provided to the ReS to maintain the core in a safe condition is discharged nutside of* the containment . .

3. If inter10cks are provided to automatica11y close the isolation valves when the ReS pressure exceeds the RHR system de~ign pressure, adequate re 1i e f capaci ty 5 ha 11 be prov i ded du ri ng the time pe ri ad while the valves are ~1osing, D, Pump Protection Requirements The design and operating procedures of any RHR system shall have provisions to

.prevent damage to the RHR system due to overheating, cavitation or loss of adequate pump suction fluid.

E. Test Requirements The isolation valve operability and interlock circuits must be designed so as to permit on line testing when operating in .the RHR mode. Testabil ity shall meet the requirements of IEEE Standard 338 and Regulatory Guide 1. 22.

The preoperational and initial startup test program shan be in conformance with Re~u1atory Guide 1.68. The programs for PWRs shall include tests with, supporbng ahalysis to (a) confirm that adequate mixing of borated water added prior to or during coo1down can be achieved under natural circulation conditions anct permit estimation of the times required to achieve ,such mixing, and .

(b) confirm that the cooldown under .natural circulation'cond1tion~ can be achieved within the 1imits specified in the. emergenc~1 operating procedures.

Comparison with perform~nce of previoClsly tested plants; of similar, design may be substituted for these tests. '

F. Operational Procedures The operational pl'ocedures for bringing the p-lant from normal operating power to cold shutdown shall be in conformance with Regulatory Guide 1.33. For pressurized water reactors, the operational procedures shall include specific procedures and information required for cooldown under natural circulation conditions.

G. Auxi 1iary F.eedwater' Supply The seismic Category I water supply for the auxiliary feedwater system for a PWR shall have sUfficient inVentory to permit operation at hot shutdown for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by cooldown to the conditions permitting operation of the RHR system. The inventory needed for cooldown shall be based on the lonQest cool down time needed with either only onsite or only offsite power aval1able with an assumed single failure.

H. Imp1ementation For the purposes of implementing the requirements for plant heat removal capabilitity for compliance with this position, plants are divided into the fo11owing three classes:

Class 1 Full compliance with this position for all plants (custom or standard) for which CP or PDA applications are'docketec on O~

after January 1, 1978. See Table 1 for possible solutions for full compliance. ' '

Class' 2. Partial implementai:;ion of this position for a11 plants (custom or standard) for which CP or PDA applications are docketed before January 1, 1978, and for which an Ol ,issuance is expected' on or after January 1, 1979. See Table 1 for reco,mmended implementation for -Class 2 plants.

Class 3 The extent to ~hich the implementation guidance in Table 1 will bebackfitted for all operating reactors and all other plants (custom or standard) for which issuance of the OL is expected before January 1, 1979, wi 11 be based on the cambi ned I&E and DOR review of related plant features for operating reactors. '

19SJ

TABLE 1. POSS1BLE SOLUTION FOR FULL COMPLIANCE WITH BT? RSB 5-1 AND RECOMMENDED IMPLEMENTATION FOR CLASS 2 PLANTS Process and [System Possible Solution for Recommended Imp lementatfon for or Compone~~ Full Compl,iance__ .lliss 2}'1ants (see Note 1)

Functional R~qufrement for Long-term cool i ng [IlHR dl'OP Provide double drop line (or valves Camp 1lance will not be l'squi l'ed if no to Cold Shutdown 1ine] in para1lel) to prevent s1ngle valve it can be shown that correction ror, failure from stopping RHR cooling single failure by manua1 actions, Capab n ity Us 1I1g Safety function. (Note: This I'equirement inside Ot' outside of containment or Grade Systems in conjunction with meeting effects' return to hot'standby until manual of'single failure for long-term actions (ot' repairs) are found to Capabfiity ,with either cooling and isolation requirements be acceptable fol' the f ndi-...:i dual onsite or only offsite power involve increased number of plilnt.

and with single failure independent power supplies ann (Hmitedactiol1 outside CR to meet SF) possibly more than four valves), I Reasonable timefol' cooldo'vln assuming most limiting SF and ofrsite or only onsit~

pow!.!)',

Heat removal and ReS circulation Provide safety-grade dump valves, Compliance required.

during cooldown to co)~ shutdown operator's, and power supp1y, etc. so (Note: Need SG cooling to mafn- that manual action shou1d not. be taln ReS circulatlon even after l"equ'i!'ed after SSE except to meet RHR in operation when Linder single failure.

natul'al circulation [steam dump valves].)

Depressurization (Press0rizer Provide upgrading and additional Comp 1 j ance wi 11 not be requ i red if duxi 1i ary spray ot' power- , valves to ensure operation of aux- a) dependence on manual actions operated, t'el i ef va hes), iliary pressurizer spray lIsing only 'inside containment after SSE or safety-grade SUbsystem meeting single single failure,or b) remaining dt failure. Possible aHernative mil.'! hot standby until manual actions involve using pressurizer powet'- or repairs are complete are found operated relief valves which have to be acceptable for the individual 1.0 been upgraded. Meet SSE and single p Jant.,

00 failure without manual operation inside ~ontainment,

TABLE 1. POSSJBLE SOLUTIDrI FOR FUll COMPLIANCE WITH BTP RSB 5-1 AND RECOMMENDED H1P'LEMENTATION FoR CLASS 2 PLANTS Process and [System Possible SoJutian for Recommended Implementation f;)I' or Compon~~ Full Compliance Class 2 Plants (~~~ Note l~_

Boratian for cold shutdown Provide procedure and uPQrading where Same .15 above,

[eVeS and boron sampling], necessal')' such that boration to cold shutdown concentration meets the requirements of I. Solution could range from (1) upgrading dnd adding valVes to have both 'letdown and chat'g-ing paths safety grade and meet sfngle failure to (2) use of br..r::kllP procedures involving less cost. For example, bor--

atjoll without letdoliln may be acceptable and'eiiminate need for upgrading let-down path. Use of EeCS for injectIon of borated water' maya 1so be accept-ab le. Need surveill ance of DI)"On

n co'ncentrat i on (boronometer' and! or sampling). limited operatol' action inside or outside -of containment if justified.

RHRIsolation RHR System Comply with one of allc....able Conlp I laBee t'llquil'ed, (Pl ants arrangemellts given, normally meet the requirement undel' eXisting SRP Section 5.4,7).

Pressure Re1ief Collect and contain relief disch'arge RHR System Determine piping, etc" needed to meet I'equirement to provide in Cumpliance wil' not be j'equiN:d, if it 1s shown that ;:1equate .

I.

des i911. a1 ternate methods of dispoFing of

,'" discharge are available, c..

c:

~

TABLE 1. POSSIBLE SOLUTION FOR FULL COMPLIANCE WITH aTP RSB 5-1 AND RECOMHENOEfl I~;PLEMENT~TION FOR C~ASS 2 PLANTS Process and [System Possible S016tion for Recommended Imp 1ementat i on fOI' or Component] _---LulL~ 1i anc~_ ,_~JEE~_ ,t,t:jAntLls.!.!? lIg!~n-,-

Requi rement A.G. 1:68. For PWR5, Run tests confirming analysis to Comp 1i ilnCe l'squi red.

plus analysis for coal down meet requirement.

und"r ilaturalcirculatiol1 to confirm.adequate mixing and coo'JdownwHhfn limits

~peGifiedfn tOP.

,Procedure A,G. 1.33. For PWRs, Develop procedures and lnformation Comp 1ianee ft'Qui red.

ludespec Hie procedures and from tests and analysis, fnforlllatlnn for corildown I:nde)'

ci rcul<'ltion,

-?'

11 i (!cry Feedwater Category 1 supply for auxiliary FW for at least four at hot shutdown'plus Emergency feedwater Supply From tests and analysis obtain

'conservative estimate of auxiliary FW 5 upp Iy to mest requi rement and Compliance will not be required, if it is shown that an adequate a1 ternate seismic Categlll'Y I I

down to RHR cut~in based provide seismic Category I supply, source is avai1ab1e.

longest time for only.

te or only offsite power assumed sing'le failure.

The implementation for Class 2 plants does not result in a major impact whne p)'oviding additional capabi lityto go to cold shutdaY/n. The major

'impact results from the requirement fO)' safety-grade steam dllmp valves.

Enclosure B FSAR Question 212.6

B/B-FSAR AMENDMENT 20 MAY 1979 QUESTION 212.6 "DISCUSS Byron/Braidwood compliance with the technical position (RSB 5-1) requirements contained in Regulatory Guide 1.139 regarding capability of cooling the plant down to RHR conditions using only safety-grade equipment and assuming loss of offsite power."

RESPONSE

Byron/Braidwood is subject to the technical requirements of RSB 5 1 as they apply to Class 2 plants. Only partial com-pliance with the technical position is required where manual actions or repairs can be demonstrated to be an acceptable alternative to strict compliance. The safe shutdown design basis for Byron/Braidwood is hot standby. The functional re-quirements of RSB 5 1 impose the following assumptions on the system(s) used to go to cold shutdown: a loss of offsite power, the most limiting single failure, and that only safety grade systems are available. Under these conditions, the plant is capable of being taken to cold shutdown within a reasonable amount of time, provided that limited manual actions, as allowed by the recommended implementation for Class 2 plants, are per-formed. Residual heat removal system operation conditions

(- 350 F, 400 psi) can be achieved in approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, including the time required to perform any necessary actions while at hot standby.

Four key functions are required to achieve and maintain cold shutdown. Means for performing these functions are described below.

Circulation of the reactor coolant can be provided by natural circulation, with the reactor core as the heat source and the steam generators as the first heat sink, and then by the re-sidual beat removal pumps.

Removal of residual heat can be accomplished first by venting steam to the atmosphere while maintaining steam generator level with the auxiliary feedwater system; and then via the residual heat exchanger. Hot standby can be maintained by releasing steam via the safety grade steam generator safety valves.

cooldown to 350 F can be accomplished by releasing steam via local operation of the steam generator power-operated relief valves. Then cooldown to cold shutdown conditions can be achieved with the residual heat removal system. A sufficient Seismic Category I supply of auxiliary feedwater is provided to permit 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> operation at hot standby plus cooldown to residual heat removal system initiation conditions.

Boration can be accomplished using portions of the chemical and volume control system. Boric acid from the boric acid

B/B-FSAR AMENDMENT 20 MAY 1979 pumps by the boric acid transfer pumps. The centrifugal charging pumps can inject the boric acid into the reactor coolant system via the safety ection flow path or the normal charging and reactor coolant pump seal injection flow paths. Makeup in excess of that needed for boration can be provided from the refueling water storage tank.

Depressurization can be accomplished using portions of the chemical and volume control system. Either boric acid from the boric acid tanks or refueling water can be used as desired for depressurization with the flow path being via the centri-fugal charging pumps and auxiliary spray valve to the pressurizer.

The plant design provides the capability for conducting natural circulation cooldown tests if required. However, because of the great similarity in design between all Westinghouse pres-surized water reactors, Byron/Braidwood can reference those tests conducted at other units rather than conducting such tests on the Byron/Braidwood project. However, specific pro-cedures for cooling down using natural circulation will be prepared prior to startup of the project.

Q212.6-2

Enclosure C FSAR Question 212.47

FSAR AMENDMENT 26 JUNE 1980 QUESTION 212.47 "The NRC Regulatory Requirements Review Committee has recently approved a new staff position (BTP RSB 5-1) for the residual heat removal system. The technical require-ments of this position for your plant are described below.

Your response to these requirements should be in sufficient detail to enable the staff to review your compliance.

System parameters assumed should be the most limited para-meters allowed by Technical Specifications:

BRANCH POSITION (A) Functional Requirements The system(s) which can be used to take the reactor from normal operating conditions to cold shutdown* shall satisfy the functional requirements listed below.

(1) The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown* using only safety-grade systems.

These systems shall satisfy General Design Cri teria 1 through 5.

(2) The system(s) shall have suitable redundancy in components and features, and suitable inter-connections, leak detection, and isolation capa-bilities to assure that for onsite electrical power system operation (assuming offsite power is not available) and for off site electrical power system operation (assuming onsite power is not available) the system function can be accomplished, assuming a single failure.

(3) The system(s) shall be capable of being operated from the control room with either only onsite or only offsite power available. In demonstrating that the system can perform its function assuming a single failure, limited operator action outside of the control room would be considered acceptable if suitably justified.

(4) The system(s) shall be capable of bringing the reactor to a cold shutdown condition, with offsite or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.

• Processes involved in cooldown are heat removal, depressurization, flow circulation, and control. The cold shutdown the Standard f Q2

FSAR AMENDMENT 26 JUNE 1980 (B) RHR System Isolation Requirements The RHR system shall satisfy the isolation requirements listed below.

(1) The following shall be provided in the suction side of the RHR system to isolate it from the RCS.

(a) Isolation shall be provided by at least two power-operated valves in series. The valve positions shall be indicated in the control room.

(b) The valves shall have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR system design pressure. Failure of a power supply shall not cause any valve to change position.

(c) The valves shall have independent diverse interlocks to protect against one or both valves being open during an RCS increase above the design pressure of the RHR system.

(2) One of the following shall be provided on the discharge side of the RHR system to isolate it from the RCS:

(a) The valves, position indicators, and interlocks described in item l(a) - (c),

(b) one or more check valves in series with a normally closed power-operated valve.

The power operated valve position shall be indicated in the control room. If the RHR system discharge line is used for an ECCS function, the power-operated valve is to be opened upon receipt of a safety ection signal once the reactor coolant pressure has decreased below the ECCS design pressure.

(c Three check valves in series, or (d) Two check valves in series, provided that there are design provisions to permit periodic testing of the check valves for leak tightness and the testing is performed at least annually.

(C) Pressure Relief Requirements

FSAR AMENDMENT 26 JUNE 1980 (1) To protect the RHR system against accidental overpressurization when it is in operation (not isolated from the RCS), pressure relief in the RHR system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code. The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS shall be considered when selecting the pressure relieving capacity of the RHR system.

For example, during shutdown cooling in a PWR with no steam bubble in the pressurizer, inadvertent operation of an additional charging pump or inad-vertent opening of an ECCS accumulator valve should be considered in selection of the design bases.

(2) Fluid discharged through the RHR system pressure relief valves must be collected and contained such that a stuck open relief valve will not:

(a) Result in flooding of any safety-related equipment.

(b) Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA.

(c) Result in a non-isolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside of the containment.

(3) If interlocks are provided to automatically close the isolation valves when the RCS pressure exceeds the RHR system design pressure, adequate relief capacity shall be provided during the time period while the valves are closing.

(D) Pump Protection Requirements The design and operating procedures of any RHR system shall have provisions to prevent damage to the RHR system pumps due to overheating, cavitation or loss of adequate pump suction fluid.

(E) Test The isolation valve operability and interlock circuits must be designed so as to permit on line testing when operating in the RHR mode. Testability shall meet the requirements of IEEE Standard 338 and Regulatory Guide 1.22.

4

B/B-FSAR AMENDMENT 34 NOVEMBER 1981 The preoperational and initial startup test program shall be in conformance with Regulatory Guide 1.68.

The programs for PWRS shall include tests with supporting analysis to (a) confirm that adequate mixing of borated water added prior to or during cooldown can be achieved under natural circulation conditions and permit estimation of the times required to achieve such mixing, and (b) confirm that the cooldown under natural circulation conditions can be achieved within the limits specified in the emer-gency operating procedures. Comparison with performance of previously tested plants of similar design may be substituted for these tests.

(F) Operational Procedures The operational procedures for bringing the plant from normal operating power to cold shutdown shall be in conformance with Regulatory Guide 1.33. For pressurized water reactors, the operational procedures shall include specific procedures and information required for cooldown under natural circulation conditions.

(G) Auxiliary Feedwater Supply The seismic Category I water supply for the auxiliary feedwater system for a PWR shall have sufficient inventory to permit operation at hot shutdown for at least four hours, followed by cooldown to the conditions permitting operation of the RHR system.

The inventory needed for cooldown shall be based on the longest cooldown time needed with either only onsite or only offsite power available with an assumed single failure."

RESPONSE

Branch Technical Position RSB 5-1 is addressed by the response to acceptance review Question 212.6 (see page Q212.6 1).

For additional information on these items, see the response to 212.154.

Q212.47-4

Enclosure D FSAR Question 212.154

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 QUESTION 212.154 "The response to 212.47 references the response to 212.6.

The response to 212.6 does not cover all of the concerns of Q212.47. Provide a response to 212.47 that addresses each item of concern as listed in Q212.47."

RESPONSE

The information which follows supplements the responses previously provided for Questions 212.47 and 212.6.

1. Operating Procedures General operating procedures are presently being reviewed and will be revised.

Procedures to be revised will include at least the following:

a. Reactor Trip EO

b. Reactor Trip Recovery ESO.1 C. Natural Circulation Cooldown ESO.2

d. Post-LOCA Cooldown and Depressurization ES 1.2

e. Loss of AL ECA.2.

2. Natural Circulation - Comparison of Byron/Braidwood and Diablo Canyon.

Byron/Braidwood and Diablo Canyon Unit 1 have been compared in detail to ascertain any differences between the two plants that could potentially affect natural circulation flow and attendant boron mixing. Because of the similarity between the plants, it was concluded that the natural circulation capabilities would be similar, and therefore, the results of prototypical natural circulation cool down tests being conducted at Diablo Canyon will be representative of the capability at Byron/Braidwood.

The general configuration of the piping and components in each reactor coolant loop is the same in both Byron/Braidwood and Diablo Canyon. The elevation head by these components and the system piping is similar in both plants.

To compare the natural circulation capabilities of Byron/Braidwood and Diablo Canyon, the hydraulic resistance coefficients were compared. The coefficients were generated on a per loop basis. The hydraulic resistance coefficients applicable to normal flow conditions are shown in Table Q212.154 1.

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 The general arrangement of the reactor core and internals is the same in Byron/Braidwood and Diablo Canyon. The coef ficients indicated represent the resistance seen by the flow in one loop.

The reactor vessel outlet nozzle configuration for both plants is the same. The radius of curvature between the vessel inlet nozzle and downcomer section of the vessel on the two plants is different. Based on 1/7 scale model testing performed by Westinghouse and other literature, the radius on the vessel nozzle/vessel downcomer juncture influences the hydraulic resistance of the flow turning from the nozzle to the downcomer. The Diablo Canyon vessel inlet nozzle radius is significantly smaller than that of Byron/Braidwood, as reflected by the higher coefficient for Diablo Canyon.

The resistance coefficient for the RCS piping between the plants differs slightly due to the loop isolation valves on Byron/Braidwood. This difference in flow resistance has been taken into account in the loop resistance calculation.

Steam generator units were also compared to ascertain any variation that could affect natural circulation capability by changing the effective elevation of the heat sink or the hydraulic resistance seen by the primary coolant. it was concluded that there are no differences in the original design of the steam generators in the two plants that would adversely affect the natural circulation characteristics.

As indicated, the difference between the total resistance coefficients for the two plants is insignificant. It is expected that the relative effect of the coefficients would be the same under natural circulation conditions such that the natural circulation loop flow rate for Byron/Braidwood would be within 2% of that for Diablo Canyon.

The coefficients provided reflect the flow rate and associated heat removal capability of an individual loop in the plant.

The comparison, therefore, does not take into consideration the number of loops available nor the core heat to be removed.

An evaluation of the Byron/Braidwood steam relief and auxiliary feedwater systems has been performed to demonstrate that cooling can be via two steam generators the most 1 active failure, i.e" the failure of an atmospheric valve.

Loop circulation flow is dependent on reactor core decay heat which is a function of time based on core power operating history. Under natural circulation flow conditions, flow

AMENDMENT 49 B/B-FSAR JULY 1987 into the upper head area will constitute only a small percentage of the total core natural circulation flow, and therefore, will not result in an unacceptable thermal/hydraulic impedance to the natural circulation flow required to cool the core.

For typical 4 loop plants (including Byron/Braidwood), there are two potential flow paths by which flow crosses the upper head region boundary in a reactor. These paths are the head cooling spray nozzles, and the guide tubes. The head cooling spray nozzle is a flow path between the downcomer region and the upper head region. The temperature of the flow which enters the head via this path corresponds to the cold leg value (i.e., TCOLD). Fluid may also be exchanged between the upper plenum region (i.e., the portion of the reactor between the upper core plate and the upper support plate) and the upper head region via the guide tubes. Guide tubes are dispersed in the upper plenum region from the.

center to the periphery. Because of the nonuniform pressure distribution at the upper core plate elevation and the flow distribution in the upper plenum region, the pressure in the guide tube varies from location to location. These guide tube pressure variations create the potential for flow to either enter or exit the upper head region via the guide tubes.

To ascertain any difference between the upper head cooling capabilities between Diablo Canyon and Byron/Braidwood, a comparison of the hydraulic resistance of the upper head regions was made. These flow paths were considered in paral leI to obtain the results as shown in Table Q212.154-2.

As indicated in Table Q212.154 2, the effective hydraulic resistance to flow in Byron/Braidwood is slightly less than Diablo Canyon. Assuming that the same pressure differential existed in both plants, the Byron/Braidwood head flow rate would be 112% of the Diablo Canyon flow.

It can, therefore, be concluded that the results of the natural circulation cool down tests performed at Diablo Canyon will be representative of the natural circulation and boron mixing capability of Byron/Braidwood. The results of these tests will be reviewed for applicability. A natural circu-lation cooldown test will be performed at Byron Unit 1 prior to startup following the second refueling if the Diablo Canyon prototype test does not sat results.

3. Cold Shutdown Scenario Byron/Braidwood is subject to the technical requirements of RSB 5-1 as they apply to Class 2 plants. Only partial Q2 .154 3

B/B-FSAR AMENDMENT 42 MAY 1983 compliance with the technical position is required where manual actions or repairs can be demonstrated to be an accep-table alternative to strict compliance. The safe shutdown design basis for Byron/Braidwood is hot standby. The func-tional requirements of RSB 5-1 impose the following assumptions on the system(s) used to go to cold shutdown; a loss of offsite power, the most limiting single failure, and that only safety grade systems are available. Under these con-ditions, the plant is capable of being taken to cold shutdown within a reasonable amount of time, provided that limited manual actions, as allowed by the recommended implementation for Class 2 plants, are performed. Residual heat removal system operation conditions (350 F, 400 psi) can be achieved in approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, including the time required to perform any necessary actions while at hot standby. Cold shutdown conditions (T < 200 F) can subsequently be achieved within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Refer to Subsection 5.4.7.1.

Four key functions are required to achieve and maintain cold shutdown. Means for performing these functions are described below.

a. Coolant Circulation Circulation of the reactor coolant can be provided by natural circulation, with the reactor core as the heat source and the stearn generators as the first heat sink, and the residual heat removal pumps as second heat sink.

b. Residual Heat Removal The function of residual heat removal is performed in two stages in accomplishing the cooldown from hot standby to cold shutdown.

The first stage is from hot standby to 350 F. During this stage, circulation of the reactor coolant is provided by natural circulation with the reactor core as the heat source and the stearn generators as the heat sink. Stearn is initially released via the stearn generator safety valves to maintain hot standby. This occurs automatical as a result of turbine and reactor . Stearn release for cooldown continues via the stearn generator atmospheric relief valves. As the cooldown proceeds, the operator adjusts these valves to increase the amount of stearn dump, to a reasonable cooldown rate. Feedwater makeup is provided by the auxil feedwater system.

The stearn generator safety valves are Seismic Category I spring-loaded valves that can automatically maintain the plant in a safe hot standby condition for an extended period of time. The stearn generator atmospheric relief valves

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 are also seismically qualified. Should a single failure render one of the atmospheric dump valves inoperable, the plant could be cooled down to the RHRS initiation temperature via the three active loops. Additionally, the 8 inch manual valve upstream of the failed relief valve could be closed while the failed valve was repaired or replaced. Communi cations for any local operations would be made by the use of hand held two-way radios.

The auxiliary feedwater system has sufficient alignment capability and flow capacity to ensure that feedwater can always be provided to all steam generators. A motor-driven pump is provided which feeds all four steam generators.

A separate system incorporates a diesel driven pump which can also supply feedwater to all four steam generators.

The auxiliary feedwater system is capable of providing feed-water for an extended period of time. The primary source of feedwater is the condensate storage tank which has a minimum available volume of 276,000 gallons. Backup is provided from the Seismic Category I service water system.

In the unlikely event that sufficient auxiliary feedwater was not available in the condensate storage tank, the pump suction is automatical switched to the backup source of essential service water.

The status of each steam generator can be monitored using safety-related instrumentation located in the control room.

Separate indication channels for both steam generator pressure and water level are available.

The second stage is from 350 F to cold shutdown. During this stage, the RHRS is brought into operation. Circulation of the reactor coolant is provided by the RHR pumps and the heat exchangers in the RHRS act as the means of heat removal from the RCS. In the RHR heat exchangers, the residual heat is transferred to the component cooling water system which ultimately transfers the heat to the essential service water system.

The RHRS is a fully redundant system. The RHRS includes two RHR pumps and two RHR heat exchangers. Each RHR pump is powered from a different emergency power train and each RHR heat exchanger is cooled by a different component cool water . The component cooling water and the essential service water systems are both des to Seismic Category I. if any component in one of the RHR subsystems were rendered inoperable as the result of a single failure, cooldown of the plant would not be compromised; however, the time for cooldown would be extended. The operation of the RHRS can be monitored using instrumentation in the

FSAR AMENDMENT 39 SEPTEMBER 1982 control room. There is indication of the pump discharge flow, the pump operating status and the component cooling flow from the discharge of the RHR heat exchangers.

C. Boration and Inventory Control Boration is accomplished using portions of the chemical and volume control system (CVCS). The boric acid transfer pumps supply four weight percent boric acid from the boric acid tanks to the suction of the centrifugal charging pumps which inject the borated water into the reactor coolant system (RCS) via the normal charging and/or reactor coolant pump seal injection flow paths. Makeup in excess of that required for boration can be provided from the refueling water storage tank (RWST) using the centrifugal charging pumps and the same injection flow paths as described for boration. Two motor-operated valves, each powered from different emergency diesels and connected in parallel, transfer the suction of the charging pumps to the RWST.

The two boric acid tanks, two boric acid transfer pumps, two centrifugal charging pumps and the associated piping are of Seismic Category I design. There is sufficient boric acid capacity in the boric acid tanks to provide for a cold shutdown with the most reactive rod withdrawn. The boric acid transfer pumps and centrifugal charging pumps are train oriented and can be loaded on the emergency diesels. Should a common valve make both the normal and alternate charging lines unavailable, the reactor coolant pump seal injection flow would be sufficient for boration. The RCS can be borated to the cold shutdown concentration by accommodating the boration flow in the steam space of the pressurizer and in the space made available as the RCS shrinks due to cooling. Boration to cold shutdown without letdown is possible as discussed below.

Boration and makeup can be monitored using instrumentation in the control room. indications available include boric acid transfer and centrifugal charging pump operating status and boric acid tank and RWST water level. This water level instrumentation is safety-related. Sampling can be done continuously or intermittently from several sampling connec-tions in the normal letdown path, if it is available, or from two separate RCS hot legs. In the worst case situation, the amount of boron ected can be calculated by monitoring the in the boric acid tanks.

d. Boration for Cold Shutdown Without Letdown The plant is maintained in a hot standby condition while the operator evaluates the initial plant conditions and the avail ability of equipment and systems (including non-safety grade

FSAR AMENDMENT 39 SEPTEMBER 1982 equipment) that can be used in shutdown. Prior to initiating cooldown, the operator will determine the boration requirements and the method by which the plant can be taken to cold shutdown.

In performing the cooldown, the operator integrates the functions of heat removal, boration and makeup, and depres-surization, attempting to accomplish these functions without letdown from the RCS. Once the plant is cooled to 3500 F and depressurized to 425 psia, RHRS operation is initiated and the RCS is taken to cold shutdown conditions.

A calculation was performed examining the feasibility of reaching cold shutdown (defined to be 2000 F and atmospheric pressure) without letdown for Byron Unit 1. The calculated results (Tables Q212.154-4 and 5) show that the resulting boron concentration as a function of time and temperature is sufficient to maintain the core subcritical with the required shutdown margin. The basis of the calculation is as follows.

The plant is presumed to be operating at power with boron concentration ci and xenon concentration Xi. The reactor is tripped and brought to zero load hot standby conditions (defined at Tavg of 557 F, 2250 psia, and pressurizer liquid level at 25% of span). The most reactive control rod is presumed to stick during the trip. For some unexplained reason after reaching zero load hot standby conditions, the normal and excess letdown lines are lost. The operator must then proceed to cold shutdown via either of two methods.

In the first method, all RCS depressurizations are accom-plished by spraying into the pressurizer from the CVCS auxil-iary spray connection. In the second method, depressurization is accomplished by opening the pressurizer power-operated relief valves. Normal pressurizer spray is presumed to be unavailable because the reactor coolant pumps may not be operating.

Initially, independent of the method used to reach cold shutdown, 4 wt.% boric acid from the boric acid storage tanks will be used to fill the pressurizer liquid level from 25% of span to 95% of span while the plant is held at zero load hot standby conditions. This will provide sufficient boron to compensate for xenon decay at hot standby.

The 95% pressurizer level indication will ensure that the is not water solid. The will be held constant at 2250 psia us of zer heaters and the reactor coolant temperature will be reduced from 557 F to 350 F by relying on natural circulation through the steam generators. At 350 F, the pressure will be reduced, using either of the two depres-surization methods, from 2250 psia to 415 psia. The RHR Q2 7

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 system will then be used to reduce the temperature from 350 F to 200 F while maintaining a constant pressure of 415 psia. At 200 F and 415 psia, the pres sure will be reduced to atmospheric using the same method as used during the previous depressurization.

As stated previously, the operator must select one of two methods for depressurization either via eves auxiliary spray or the PORVls. If the eves auxiliary spray is to be used, the pressurizer liquid level at the beginning of the depressurization cycle (i.e., at 350 F and again at 200 F) should be as low as possible while still keeping the heaters covered. This will minimize the mass of saturated liquid in the pressurizer which will flash to steam as the pressurizer pressure is reduced. It will also provide space in the pressurizer for accepting the spray water without overranging the pressurizer level indicator. when cooling from 557 F to 350 F and from 350 to 200 F, the operator first allows the coolant to contract to 25% of pressurizer level indication span, and thereafter adds 4 wt.% boric acid to maintain the 25% of span indication and increase boron concentration. In this way, the pressurizer heaters will be kept covered and the operator will avoid a situation wherein there is too much liquid in the pressurizer without the ability to let down. Depressurization from 2250 psia to 415 psia at 350 F, and from 415 psia to atmospheric pressure at 200 F, will be accomplished by spraying until pressurizer level indication increases to 95% of span.

If the PORVls are to be used for depressurization, the pres-surizer liquid level at the beginning of each depressurization cycle (i.e., at 350 F and again at 200 F) should be as high as possible to maximize the mass of steam which can be discharged without uncovering the pressurizer heaters.

When cooling from 557 F to 350 F, the operator will maintain the pressurizer level at 95% of level indication span by adding 4 wt.% boric acid, via some other method than spraying, to makeup for coolant contractions in the pressurizer.

Depressurization from 2250 psia to 415 psia at 350 F, and from 415 psia to atmospheric pressure at 200 F, will be accomplished by opening one PORV, discharging steam and allowing the pressurizer liquid level to decrease.

e.

As stated above, there are two methods of available to the operator: either the eves auxiliary spray; or the PORVls.

Q212.154-8

FSAR AMENDMENT 39 SEPTEMBER 1982 The centrifugal charging pumps tn the CVCS are Seismic Category I pumps and are powered from the ESF buses. The auxiliary spray valve is an air-operated valve as are other valves in the flow path. In the event of a seismic event or loss of offsite power event where air is lost to the valves, every effort will be made to either open the valves with a portable gas cylinder or load the air compressors onto the emergency buses.

As an alternative, depressurization could be accomplished by discharging RCS inventory from the pressurizer to contain-ment via the pressurizer power-operated relief valves.

This operation can be integrated with the cooldown function near the end of the cooldown to 350 F. As RCS inventory is relieved to the containment the pressurizer temperature and pressure is reduced, thus, reducing the pressure in the RCS. Makeup is provided as necessary to maintain a minimum level in the pressurizer. RCS pressure and temper-ature and pressurizer level can be monitored using safety-related instrumentation in the control room.

Because the pressurizer PORV's are not qualified for operation in an adverse environment, it is necessary to ensure that the environmental conditions in the containment do not com-promise the operation of the PORV's during depressurization.

This can be done either by ensuring the integrity of the PRT or ensuring PORV operation is not impaired due to the containment environment following the rupture of the PRT.

The PRT, as described in Subsection 5.4.11, is designed to absorb a discharge of steam equivalent to 110% of the full power pressurizer steam volume without exceeding pressure and temperature design values.

The volume of steam vented from the pressurizer to depres-surize the plant from a hot standby to a cold shutdown con-dition is not necessarily less than the volume of steam at 110% power. However, the rate of release is significantly lower and can be controlled to ensure that the integrity of the PRT is maintained. The depressurization operation can be halted at any time to cool the PRT which allows a volume of steam to be discharged without compromising of the PRT. Hence we have concluded that zation via the PORV's will not cause an environment in the containment that is adverse to the operation of the PORV'S. A will be added to the procedures to state that the integrity of the PRT must be maintained during this mode of depressurization. The ability to depressurize using the PORV's without rupturing the PRT will be demonstrated on the Byron/Braidwood simulator.

Q2 9

B!B-FSAR AMENDMENT 39 SEPTEMBER 1982 If such abil cannot be demonstrated, confirmation will be provided that the paRV's will function in the containment environment that is expected in achieving cold shutdown.

f. Instrumentation Safety-related instrumentation is available in the control room to monitor the key functions associated with achieving cold shutdown. This instrumentation is discussed in Section 7.5 (Safety Related Display Instrumentation) and includes the following:

a. RCS wide range temperature

b. RCS wide range pressure C. Pressurizer water level

d. Steam generator water level (per steam generator)

e. Steamline pressure (per steamline)

f. RWST level

g. Boric acid tank level (per boric acid tank)

h. Containment pressure.

This instrumentation is sufficient to monitor the key functions associated with cold shutdown and to maintain the RCS within the desired pressure, temperature, and inventory relationships.

Operation of the auxiliary systems which service the RCS can be monitored by the control room operator, if desired, via remote communication with an operator in the plant.

Q212.154 10

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 SINGLE FAILURE EVALUATION Residual Heat Removal A. From Hot Standby to 350 F

1. Reactor coolant loops and steam generator Four reactor coolant loops and steam generators are provided, anyone of which can provide natural circulation flow for adequate core cooling.

Even with the most limiting single failure (a steam generator power-operated relief valve),

three of the reactor coolant loops and steam generators remain available.

2. Steam generator atmospheric relief valves -

Four valves are provided (one per generator),

any two of which is sufficient for residual heat removal. In the event of a single failure, three power-operated relief valves remain available.

3. Condensate storage tank (non Category I) Upon depletion of the primary source of auxiliary feedwater in the condensate storage tank, a backup source of auxiliary feedwater can be provided to the suction of the auxiliary feedwater pumps from either train of the Seismic Category I essential service water system.

B. From 350 F to Cold Shutdown

1. RHR pumps A and B - Two RHR pumps are provided, either one of which can provide adequate circu-lation of the reactor coolant. Each pump is powered from a different emergency power train.

In the event of a single failure, either pump can provide sufficient RHR flow.

2. RHR suction isolation valve 8701A and 8701B (to RHR pump A) and 8702A and 8702B (to RHR pump B) The two valves in each RHR subsystem are each powered from different emergency power trains. Failure of either power train can prevent initiation of RHR cooling in the normal manner from the control room. In the event of such a failure, the affected valve(s) can be deenergized and opened with its handwheel by operator action outside of the control room. Other single Q212.154-11

BjB-FSAR AMENDMENT 39 SEPTEMBER 1982 failures that would only affect one of the RHR subsystems can be tolerated and adequate cooling can be provided by the redundant subsystem.

3. RHR heat exchanger A and B - If either heat exchanger is unavailable for any reason, the remaining heat exchanger can provide sufficient heat removal capability.

4. RHR flow control valves HCV-606 and HCV-607 - if either of these normally open, fail open valves closes spuriously, sufficient RHR cooling can be provided by the unaffected RHR subsystem.

5. RHRjSafety Injection System cold leg isolation valves 8809 A and B - If either of these normally open, motor-operated valves, which are powered from different emergency power trains, closes spuriously, sufficient RHR cooling can be provided by the unaffected RHR subsystem. The affected valve can be deenergized and opened with its handwheel.

6. Component cooling water system - Two redundant subsystems are provided for safety-related loads.

Either subsystem can provide sufficient heat removal via one of the RHR heat exchangers.

7. Essential service water system Two redundant subsystems are provided for safety-related loads.

Either subsystem can provide sufficient heat removal via one of the component cooling water system heat exchangers.

Boration and Inventory Control A. Boric acid tanks 1 and 2 - Two boric acid tanks are provided. Each tank contains sufficient four weight percent boric acid to borate the RCS for cold shutdown.

B. Boric acid transfer pumps 1 and 2 Each pump can be powered from a different emergency power train.

In the event of a single failure, either pump can sufficient boric acid flow.

C. Isolation valve 8104 If valve 8104, which is supplied from emergency power and is normally closed, cannot be opened due to power train or operator failure, 0212.154-12

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 it can be opened locally with its handwheel. If valve 8104 cannot be opened with its handwheel, an alternate flow path is available via air-operated, fail open valve FCV-110A and normally closed manual valve 8439.

D. Refueling water storage tank isolation valves LCV-112D and LCV-112E - Each valve is powered from a different emergency power train; only one of these normally closed motor-operated valves needs to be opened to provide a makeup flow path from the RWST to the centrifugal charging pumps.

E. Centrifugal charging pumps I and 2 - Pumps 1 and 2 are powered from a different emergency power train.

In the event of a single failure, anyone pump can provide sufficient boration or makeup flow.

F. Flow control valve (FCV-121 - This valve fails open on loss of air.) If FCV-121 closes spuriously, the centrifugal charging pumps can safely operate on their miniflow circuits. Efforts would be made to open it. Boration can be accomplished by starting the positive displacement pump or by using the cold leg injection flow path.

G. Normal charging flow control valve HCV-182 - This normally open valve fails open on loss of air or power. If HCV-182 closes spuriously, the charging pumps can operate on their miniflow circuits until operator action can open bypass valve 8403.

H. Normal charging isolation valves 8105 and 8106 -

If either of these normally open, motor-operated valves, each of which is powered from a different emergency power train, closes spuriously, operator action can be used to deenergize the valve operator and reopen the valve with its handwheel.

I. Normal charging isolation valve 8146 If this normally open valve closes spuriously, alternate charging valve 8147, which fails open, can be used.

J. Reactor coolant pump seal ection valves 8355 A, B, C t and D - If any of these normally open, motor-operated valves closes spuriously, operator action can be used to deenergize the valve operator and reopen the valve with its handwheel.

Q212.154-13

FSAR AMENDMENT 39 SEPTEMBER 1982 III. Depressurization A. Auxiliary spray valve 8145 - This normally closed valve fails closed on loss of air. In this case, 8145 can be opened by using a portable nitrogen bottle. If 8145 is stuck closed as a result of a single failure, the redundant pressurizer power-operated relief valves can be used to depressurize the RCS by discharging the pressurizer inventory to the pressurizer relief tank.

B. Charging valves 8146 and 8147 - These valves fail open on loss of air. In this case, 8146 and 8147 can be closed by using portable nitrogen bottles.

If either is stuck open, the redundant pressurizer power-operated relief valves can be used to depres-surize the RCS by discharging the pressurizer inven-tory to the pressurizer relief tank.

C. RHR suction isolation valve 8701 A and Band 8702 A and B - The RHR suction isolation valves are qual ified for the steamline break environment. Therefore, they are qualified for the less severe environment that would result if, as described in the above A and B, the RCS is depressurized by discharging the pressurizer inventory to the pressurizer relief tank.

IV. Instrumentation Sufficient instrumentation is provided to monitor from the control room the key functions associated with cold shutdown.

All necessary indications are redundant. Thus, in the event of a single failure, the operator can make comparisons between duplicate information channels or between functionally related channels in order to identify the particular malfunction.

Refer to Section 7.5 (Safety-Related Display Instrumentation) for applicable details.

Q212.154 14

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 TABLE Q212.1S4-1 COMPARISON OF HYDRAULIC RESISTANCE COEFFICIENTS UNITS: Ft/(Loop gpm) 2 x 10-10 DIABLO CANYON UNIT I BYRON/BRAIDWOOD Reactor Core and Internals 7.6 7.14 Reactor Nozzles 36.8 27.55 RCS Piping 24.0 30.0 Steam Generator 114.4 116.9 Total 182.8 181. 6 Flow Ratio: DIABLO CANYON/BYRON/BRAIDWOOD (182.8/181.6)1/2 1.0033 Q212.154 15

FSAR AMENDMENT 39 SEPTEMBER 1982 TABLE Q212.154-2 COMPARISON OF UPPER HEAD REGION HYDRAULIC RESISTANCE DIABLO CANYON UNIT 1 BYRON/BRAIDWOOD Flow area (ft2) 0.77 0.844 Loss coefficient 1. 51 1. 45 Overall hydraulic resistance (ft 4) 2.57 2.038 Relative head region flowrate 1. 00 1.12 (Based on hydraulic resistance)

Q212.154 16

BIB FSAR AMENDMENT 39 SEPTEMBER 1982 TABLE Q212.154-3

SUMMARY

OF SYSTEMS AND EQUIPMENT REQUIRED FOR COLD SHUTDOWN BORATION WITHOUT LETDOWN Boric Acid Tank Boric Acid Transfer Pump Centrifugal Charging Pump Charging Line Pressurizer Level Indication Pressurizer Heater CVCS Auxiliary Spray Pressurizer Relief Valve Residual Heat Removal Loop Q212.154-17

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 TABLE Q212.154-4 COLD SHUTDOWN VIA CVCS AUXILIARY SPRAY TIME RCS TAVERAGE BORON CONCENTRATION HOURS F PPM 0 557 Ci 3 557 536 + .923 Ci 4 500 543 + .935 ci 450 842 + .89 Ci 400 1072 + .857 Ci 9 350 1266 + .83 Ci 12 350 1581 + .77 Ci 18*/37** 200 1583 + .77 Ci 24*/43** 200 2125 + .693 Ci

• 2-Train RHR Cool down

• • Single Train Cooldown Q212.154 18

B/B-FSAR AMENDMENT 39 SEPTEMBER 1982 TABLE Q212.154-5 COLD SHUTDOWN VIA PORVIS TIME RCS TAVERAGE BORON CONCENTRATION HOURS F PPM 0 557 Ci 3.0 557 536 + Ci 5.5 500 927 + Ci 7.5 450 1185 + .9 Ci 9.0 400 1389 + .87 Ci 10.5 350 1159 + .84 Ci

11. 0 350 1617 + .87 Ci 15.0*/36.0** 200 1996 + 81 Ci 16.0*/37.0** 200 2029 + .824 Ci

• 2 -Train RHR Cooldown

• • Single Train RHR Cooldown Q212.154-19

Enclosure E Byron Station Safety Evaluation Report (NUREG-0876)

Section 5.4.3, "Residual Heat Removal System"

with Model 0 steam generators. and at foreign planti to evaluate it~ impact on the other domestic plants with similar ~team genelator design. lhe slaff has not completed thf' I?valuat ion of this pt'oblrm and its impact on the inl(>gr'ity of Byron ~tation ~~odel 0 steam generctors. lhc- staff will address this issue in a supplement to this SER.

5.4.2.2.3 Conclusions Conformance with RC'gulatory Guide 1.83. NURfG-04~2, and the in~-pection requlre-ments of Section XJ of the AS~R Code constitutrs an acceptable basis for m(1l't ina.

in part, the requirements of GOC 32. for the rec~ntly identified flow-induced -

tube vibT'atioll problem in the \o.'eslinghouse t'lodpl 0 qeam generators, because the staff has not completC'd its evaluation of this problem, final conclusions regarding the Byron steam gene~'ator tube in~el'vic.e in~pect ion wi 11 he documented in a supplement to this SER.

5.4,3 Residual Heat Removal Sy!:>tem The res.idual heat t'emoval (RHR) system is de!.'i{1fH?d to remove hh"1l from the reactor coolant ~yst(>m aftC'r the ~.yst('m tem~)(,l'atllt'(' .3nd pressun* h<ne been reduced to appl'o~imately 3&OOr and 400 psi!}. t'I'!:-JlE?ctiVl'ly. lhe RHR system ;s capat)le of reducing the reactor coolant to the cold shLJtdll\>.n tf'mJH:rature and maintain this tempel"ature urt i 1 the plant lS started lip again.

The RHRS op('rates in the following manes:

Junclio;1S 1n conjunction with the high fH.'ad portion 01 th(> fmH9t'ncy (on.'

.1 cooling system (leeS) to provide i:lject 10n of ~HlratN1 wal{'1' from UH~

refut'l ing water ~,t('rage tank (RWS1) into the f~C:, (0111 h'!j~ ,1urinq th{)

inject ion phase folll)wirHl (l LO(l\,

PrOV1<1es h 1 ng*t('r11 cnre coo)in~J durinq t/lp r(>(irC:ul.~tion phas~ of a lOU,_

Ihi!> function is accomplishe;J hy aliqning th(> RHH ~yst(>m to t<l~e fluid from the containment sump, cool H by l:i!'CIJlation tlll'OIH]h til!:' RHR IH't1t

('X(.hanyers, tlnd ~l!pply it to tile l'I',">CtOI' (ort',

(4)

C;)lHl~cl(*d hi thf: C\'C~ \"la Uw )o~'pr('~~ufl' i(lto(;wn ~!':I.' t.o (onlloi "(oMtnr coolant pr(>s~un' Byron R

Des lyn data fot' the I{HR~ are Pressure 000 p.,iq I (>mpera tun:> 400°1 Pump capac ity 3(IOn qpm Number of independf'nt lr,ijn~ .~

fhe I{CS cooldown tilll(' with Ollf' RJtR l"i11n from initi.ll 1,11(11111:' \,y tlOU p~iphf"'ir prp~':j\lI" (11)<1 /OO()I it, 2~.

0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

The two RHR trains are independent In act ion ;Hld po .... p'*!-d by '('p;jf;d I' IHlWPI' slJppl i(>5 to P"o\'l(j(, r'£'dundancy.

Ihe applicant ha!> ddlh'l~ss(1d tllf- rilpahility tlf thf' Rim ,*v~t{,Jll ju ~'I('t't BfP RSB !l-l, "Of!S;f~n R('Quirf'm(>nts of the Re~iliual 1l(~l1t I-;PHwvo1 ~>\'~, tf'~ "

Four JH'O((>SS(,~ ar(~ lllvolvNi in t:,hn9 the p](1/1t from t',flt.tandhy to (old shutdown conditions. lhl:~(, al'(>: (1) i'('mOVa' 01 f('siduai heat ,1M ~t(ll'(>(j ef1C'ryy; (2) circulatiol of ltll~ rp3ctul' cool.:wt; en tHH';1tinn of the rt,Htor coolant to the cold shutdown LOfon concl:ntT'ation dnd L()ulalll makpup; (jlld (4) depre~:;urllation. With lo~s of off~itl' pow(>T'. the n~i1<.tOI tooLlfIt PWIlPS, main COnc1t~IISf>1' and ttl (I main f(:,pd'nd\.f*" pump::. are ,maV,jj lahle. Heat. !'Plnoval alld coolant circulation un(1er' natural cHcu).,tiol! (oJ)ditiull':> i~ UHf!) corrtrollpo by u~E" of the ~tcam qenel'3tor Jtmo~phN'ic dump valvl's cin(l ttH! f1uxil1dry t('('<iwatt'r' sytl?r.l.

ltw water 5lJpply for thE' ,'lUxilii1l'y f(,f'dwatpr ~,y~ti'1l1 i~, plovidpd 'nit 1<11 iy t)'om the non.,eismic condensate ~torr1YI:' li.lll~.. fhf' tlperdtoi' Llill switttl Ill' ~,ource lit auxilial'y feE-dwale)' to the sei ... mic Category i serviCl., .... otr'l' :,v~\.I*nl, llJe foul' s.team generator' almo!>ptwl'ic l'el1(lf valvl's ';till be .iLtuatl'd hy Ch~<. il qUtdili ,<1. t solenoid-actuated, hydraldic operators IlOwered frnffl (>rnl'rq~flcy hll;'l'~. rwu valvps an' powered from olle ('merlJency bLJS iHld two from the ()Ul('" ernp.rqt'IlCY bus.

1n the evpnt that ,Hl ernE'rgency hus is lo!?t, the plant :n.,y c,t ill 1,1;:' c()ol.,.d down, but at J slow,"'.' ,';.JtP., B1P RSB !j-l feqliiY'(~s that ,) l\otUl,,1 chCll1at:on U*~tl with supporting analy~is, be conductpd to dpmon~lr(1te thp ;,hiLly to cool IjOV.:1 and depressurit'll t.tI£' plant and to ti('mons.trate that bot'oll rn;;":iIlQ ir, :*,utf:cient under such cl,'cumstanc..('s. Comparison with the performallu' ot IH'('viollsly t£'stNj j.llants of similar design may Of> substituted for ttJe~,e t(,!lots, iT jw,t/fi('(L The dppiicdnt plans to r'p1en'nce ~I!sb to he conductt.'d at Oiiddo Canyon to JlI('Pt.

this n:>Quircment. fhe applicant hi1!:> provided a pn~liIl11nal'V dsses~m(>!\t of the difl('renc('s hl'tween Oi.lblo Canyon ,]Ild Byron which r:dqht i111(,(1 buron 011;>:1119 under natural circulation. This. aSSf'ssment indicatp<, that tilt" rl>~ult<; uf till" Diablo Canyon tests and supporting analysis would sat isty til(> HII' R~1l :1-\

requin'ments for Byron How(1vpr, til(> ~taf1 do('s not pliHl tv 1'('i:Hh d (Onelll:'l"!/l on this maUl'" until the Diablo Canyon "(,Slllt~ havE' 11('('0 ('('viewed a/HI their applicabl1ily lo Ol'yon evaluated. "If the Diablo Callyon tests at'P not complf!tpd Ot* do not j)t'ovidl' sati!>fctctory results. the applic.ant /las l;omnlltt('d to perfol'flI such tests befoJ'l' S.tdfl.up attl"r t/l(:> first refllelin\j, This testing i~. not nec('ssat'Y fOl' fir~t-ly{le f

ron R

/)peded to ta~(:> thr plant trom hot ~.tandhy to the clJt~ill point of thr I<HI{ ~\stprn under condition'.> such .1!> ,-)'{(,(l(j('(1 \057} of of1sitf' power W-ht>rl the RC~ pump<, ,in- ....

not dvailaIJl(*. It would he' !))'eff>r'1bl .., to "IHl U,i', tp<,t .-tlt!:I' 01(> fir,;.t )!!lo;ld ,.,

when till' decay !If:at i'~ !*plativt*ly liHW" lhi*, II>llllld fl'SUlt !Il 'Hurl' mf'ilftlnqtili te!:.t data MId te~,tinq ulltkJ l'onditiofl':> IIlI)T'e f'('prt*t.,tlntalivp of 'flO~P ,)(cUl'lill()

ove:' thf.' 4u~yr plallt life, Bor-aUon to (olli "hutdowll (ondi1iunc, i~ 'lOrmally ;lnor1lpli~lH'd ttnollqh Uw ma~t'lJP line anti r'pi1ct\lJ' cool.1nt pump ~f'al inj('cti(,n flo*.... path.h,,* l'fi(~ct *t I lo!>s, of air ~.lJPply tf'fl[lel'l, (/1(> normal ltll10wn afld m,deup l)n('~) illopeltlhlf' because several valvl.>s dl'P ,lir np('ratHL Boralion without letduwll i~, .1CCorn-pl1~hed with 4 100'/% bOlic acid from the r'('dund()nt ~,-,i<,mic:.: CJtegory I boric arid tanH) and boric acid transfer pumps tllf'ouyh the charyio!l pumps to thc' ~.af(~ty injectitH1 flow path. Ihe dpplicant.. has l:>tated lhat there l~ ~utfit,i()nt volump from coolant contl'action .'itld prl'ssuri/pr' 5.team '>pi:iCO to accommodate the ','ollime of horic acid nef:'deiJ to r(lach cold <,hutdown, file '~t.."1tf has ~'Pqu~*~l('d (U!llirmi1' tory analysh to ~llPpOlt thi<, statpment; lncll/dinq thtl PTtl'(t-. .. It >,PIIOfl df*Cil)'.

LInder natural til'lulatioll (onditions, the normal ('UPP!y lor lIH~ pr£'!:-~,lH'i7('r

~,pray from til(' -:uid legs uf two c:001lll1t loops is lOSt. In this CllSP, th{'

P"t'SSUrlH'l' spray Cdn h~ supplied by flow from the ,:ent"ifugal chat'uirHJ pumps tt1l'ougl, a lil1f! branching off f)'om ttl(' ctldl'9illlJ pump!> of the eves. Thi~*.:oupply could helo5l by .:\ siflqlp. tai*lun.' involvinq eithe!' closing uf d ~-;ngll' vall,'11 ill thE' supply lifH"' or operlin~l of nne of s(>vt'l'al vaiv('s in Ilnes cunnecled to thf!

supply lill(', It manual actions to CO\T{!ct for such Llilun's wf.:'I'p flOt <,llcC(*ss1111.

a banup method of depl'('ssuri7ation would involvp npenlng t.>ith(,l' of tile two s('ismic Catf.'!Jot'y 1 PORVs of the Pl'(>S~,urilPr' that ;Jh;ch;1\'()p to tlie ....

pressuriz(,f 1'(>1 ief taM,. 10 1'f'l1uce the potent i~'; for ,'j hi9hpr' than ;iOl'mal -

conL:linmpnt tt'mpel'aturp and humidity, the applicant ha:> ~,t<lted !tltlt thp procp-dure~ for dc.'f)l'I.'!:-sLJri7ation llsing lhe PORVs will inciude pl'H.ClutioIlS to t'n~u!'e inteqrity of Hll> preSSlJrill11' I'eli(,t tank (PIn), In addition, UWibi I ity to depn,~,slJri7l' 'l'<ithllUl rupturiny the PRl rupture di.,k will :'(' "(Jlilil'nwd 'Ill tIll' HVl'oJ) sirnuid\.rn. ihl:' stall also r qlJjrp~; conlirmati()fl 1!);,t lh,' PORV'o .. i!:

ttJl)ction in til' ,-nvironmellt ttlat i~ E>Jt.III'Ctp(l in 'lChil*vinq ,(lId :hutrlown.

The sta1t rorl(ludes tllt;t with ~ati~.filrl()I'Y ,'('solution of ttl!' con1irmatory :if'IHS (jl'~trih('d abuve, the Byl'vf1 ::.tatioll meld lilt! ,'pquin,>rnenb il! !HP R~B trJ.

!J.4.3,l runctional  !~'quin'ment~

ltlE' RHR sy~.t(!m for Byron "IU5l me(*t (lDe 1 'hrough ~), GDe 1 through'~ j'pgardinq quality ~tandor(js and rH',ros. design bast:s for protection against fl.llural phenomena, fire protection, and t'nviron.!.entaI illld missil(~ design bast's dt"(>

covered in Sections 3 ,1nd 9.'1,1 of this I'('port, rpspectivr-ly, l;De '" II<,/lal'if1(j of Structures, System~, and Cumponents," is met tor UH~ Bytclfl i

---

lor (!ach unit. led~, detection tor tlw Rim ~,y5tem is discLJs~ed in Section 'i.2.'J of this repol'L Isolation valve .'H:d pOw('Y' supply }'edundancy are discu:,!,(>d lJlltl£>r' separat(,> topics in this sPction. 'he staff has r£'vl(>wed the 'll'SCl'iptiull ,)t 'Ill' RHR system and the pi(.Jing and instrumentation dianrams to v~rify that tilt, a c t l ff tl' 11m i 1 u' ,.,

Byron S[R ~-24

Enclosure F Letter from R. A. Chrzanowski (Commonwealth Edison) to T. E. Murley (NRC), "Request for Additional Information Response," dated October 14, 1988 (Commonwealth Edison response to NRC questions related to the natural circulation capability comparison between Byron and Braidwood Stations and the Diablo Canyon Station)

e Com onealth Edison One First National Plaza, Chicago, Illinois Address Reply to: Post Office Chicago, Illinois 60690 Bo~ 767 October 14, 1988 Mr. Thomas E. Murley, Director Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC. 20555

Subject:

Byron Stations Units 1 and 2 Braidwood Stations Units land 2 Request for Additional Information Response NRC Docket Nos. 50-454/455 and 50-456/457

Dear Mr. Murley:

The NRC staff telephonically requested additional information on September 2, 1988, concerning the natural circulation capability comparison between the Byron and Braidwood Stations and the Diablo Canyon Station. Further clarification of the requested information took place telephonically on September 28, 1988.

Enclosed with this letter is Commonwealth Edison's response to the NRC questions.

Please direct any further questions regarding this matter to this office.

Very truly yours,

~~:~

Nuclear Licensing Administrator RAC/k1j cc: Byron Resident Inspector Braidwood Resident Inspector L.N. Olshan - NRR S. Sands - NRR Region III Office encl.

5237K

\

ATTACHMENT J3y.r.QILan~L~rai_d.wQQd St.aUQJ1. ..R.f!JiPQnSeS t.2....thsLReques.t fQr AdditiQnal InfQrmatiQn Regarding Natural CirculatiQn CQQ1dQwn

1. What is the effect Qf RCP and mQtQr inertia?

In cQmparisQn tQ DiablQ CanYQn, effect is similar.' Diab1Q has a 6000 *HP mQtQr cQmpared tQ ByrQn/BraidwQQd's 7000 HP.

MQtQr inertia WQuld therefQre be similar, if nQt slightly higher due tQ the larger size mQtQr. Effect Qf pump WQu1d be similar since MQdel 93A is used.

2. What are quantitative values fQr thermal driving head?

An analysis WQuld need tQ be perfQrmed tQ determine these values. HQwever, with respect tQ DiablQ CanYQn, thermal driving head WQu1d be similar per cQmparisQn presented in questiQns 4 and 5 listed be1Qw.

3. Are these units with a T-cQ1d upper-head?

Yes. AlthQugh nQt specifically stated, this is inferred in the first paragraph Qf Q212.l54-3. It shQuld be nQted that in cQmparisQn, DiablQ CanYQn is a T-hQt upper head.

4. Is resistance per lQQP fQr DiablQ CanYQn cQnsistent with Qther plants?

Table Q2l2.l54-1 "CompariSQn of Hydraulic Resistance CQefficients" presents a cQmparison between Diablo CanYQn and ByrQn/BraidwoQd fQr nQrmal flQW cQnditions. The flow ratiQ between them is equal tQ apprQximately 1.0033 indicating that the total resistance cQefficients between the plants is insignificant. Per request Qf the NRC, the values repQrted for DiablQ CanYQn were verified with WestinghQuse as being accurate.

5. Flow resistance with Diablo is based on NORMAL flQw. At 1Qwer flQws-coefficient Qf frictiQn may increase. A statement is needed to justify cQmparison at nQrmal f1QW when NC flows are much lQwer.

At lQwer flows, coefficient of frictiQn may in fact increase, both fQr Diablo Canyon and Byron/BraidwQod. Page 2 of Q2l2.l54 states " ... It is expected that the relative effect Qf the (hydraulic) coefficients (Table Q2l2.154-1) would be the same under natural circulatiQQ cQnditions such that the natural circulatiQn loop flow rate for ByrQn/BraidwQQd WQuld be within 2% Qf Diablo CanYQn.

6. Page 2 of response states flow is dependent on decay heat.

Is this consistent with other Westinghouse statements i.e.,

independent of decay heat?

Per Q.212.154-2, loop flow circulation ~ dependent on decay heat. More specifics are necessary as to the source of the other Westinghouse statements to the contrary.

7. What is cooldown rate with 3 and 4 steam generators?

Using Natural Cooldown Circulation Procedure 1/2 BEP ES-O.2, a cooldown rate monitored at the RCS cold legs of less than 50F/hr. is required to assure no void formation in the upper head, to the point where the RHR system could be employed for further cooldown without void formations.

As indicated on Q212.154-2, an evaluation has been performed at Byron/Braidwood to demonstrate cooling can be provided via two steam generators. Therefore, the capacity of the atmospheric relief valves is determined to be adequate.

8. What time to RHR cut in temperature?

RHR cut in temperature (350 0 F, 400 psi) can be achieved in approximately 9 to 10.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. (Table Q212.154-4/5)

9. How much aux feed is needed for cooldown with 3 or 4 steam generators?

For Braidwood, a back-up source of water is available from the Essential Service (SX) Water System System with it's source being the Essential Service Cooling Pond. For Byron, the SX system source is the essential service cooling tower. Both are Safety Category 1.

In addition the refueling water storage tanks are available for a 'bleed and feed' type of cooldown. These tanks are approximately 500,000 gallons and are safety category 1.

Per telecon with NRC, B&L study of Diablo Canyon indicates 360,000 gallon of feedwater would be required there. The capacity of Byron/Braidwood Safety Category I systems exceeds this amount. Other non-safety systems could also be made available. By Tech Spec requirements, a minimum amount of 200,000 gallons of make-up water is required to be available for aux feed in the 500,000 gallon capacity condensate storage tanks

10. How long for boration?

Calculated results for the feasibility of reaching cold shutdown without letdown are given in Table Q2l2.154-4/S.

This shows the resulting boron concentration as a function of time and temperature. The CVCS aux spray is reliant on non-safety instrument air. Assuming loss of this air, and if efforts are unsuccessful in operating valves using portable bottles, the pressurizer PORVls could be used.

Per FSAR 9.:3.4.1.1b. the amount of boron available exceeds that amount required to borate the RCS to cold shutdown concentration.

11. How is it determined that upper head is sub-cooled prior to depressurization?

Byron Procedure 1/2 BEP ES 0.2 page 19 of 19 (Braidwood similar) discusses determination of sub-cooling using RCS pressure. Additionally, it should be noted that upper head sub-cooling can be accomplished without the use of CRDM fans.

12. Page 9 of response - Are personnel trained to handle portable gas cylinders?

No specific training is required for operation of portable gas cylinders. However, this is an accepted maintenance and testing activity for non-routine operation of select valves.

13. What is the assurance that PRT will not rupture?

The PRT is designed to absorb a discharge of steam equivalent to 110% of the full power pressurizer steam volume without rupture. Although the volume of steam released during depressurization from a hot standby to a cold shutdown condition during natural circulation is not necessarily less than this. the rate of release is significantly lower and can be controlled to ensure that the integrity of the PRT is maintained.

14. What is the volume of the upper head?

821.9 cu. ft.

• . 15. Is there a reactor vessel spray nozzle between downcomer and upper head? If so, will it's spray area provide better cooling effect in upper head?

With RCP's running, 2/3% of flow from the cold legs is diverted up through holes drilled in the upper (head) flange assemblies. The temperature of the flow which enters the head via this path corresponds to the cold leg value. Table Q.2l2.l54-2 indicates that the effective hydraulic resistance to flow in Byron/Braidwood is slightly less than Diablo Canyon. Assuming that a pressure differential was similar for both plants, Byron/Braidwood head flow rate would be 112% of Diablo Canyon. It's also assumed this flow rate comparison would be similar at natural circulation consistent with response to Question

5. Since Byron and Braidwood are T-cold upper-head, upper head will be cooler at initiation of natural circulation since Diablo is a T-Hot Plant.

5237K

Enclosure G Letter from S. P. Sands (NRC) to H. E. Bliss (Commonwealth Edison), "Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2, Natural Circulation Cooldown," dated November 4,1988

... I<. flo ' NOV 0 9 REC'O I ,-

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 November 4, 1988 Docket Nos. 50-454, 50-455 and 50-456, 50-457 Mr. Henry E. Bliss Nuclear Licensing Manager Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

SUBJECT:

BYRON STATION UNITS 1 AND 2 AND BRAIDWOOD STATION UNITS 1 AND 2, NATURAL CIRCULATION COOLDOWN (TAC NOS. 56199, 63239, 64018, 64045)

Dear Mr. Bliss:

The NRC staff has reviewed the results of the natural circulation cool down tests and boron mixing capability performed at Diablo Canyon and the comparison of these results with the Byron/Braidwood design. The NRC staff has completed its evaluation of Byron/Braidwood and has determined that the technical requirements of RSB 5-1 have been satisfied as they apply to Class 2 plants.

Therefore, Byron/Braidwood have satisfied the commitment set forth in the Byron/Braidwood FSAR, Amendment 48, dated July 1987 to either demonstrate compatibility with the Diablo Canyon prototype test results, or perform a natural circulation coo1down test if the test results could not be shown to be representative. The NRC staff has determined that the Byron/Braidwood analysis for this event is representative of the prototype test run at Diablo Canyon and is acceptable. The Safety Evaluation is enclosed.

Sincerely J Stephe Q.~

P. Sands, Project Manager Project Directorate 111-2 Division of Reactor Projects - III, IV, V and Special Projects

Enclosure:

as stated cc: see next page

Mr. Henry E. Bltss Byron Statton Commonwealth Edtson Company Untts 1 and 2 cc:

Mr. Wt11iam Kortter Regional Admtn1strator, Region III Atomic Power Distribution U. S. Nuclear Regulatory Westinghouse Electric Corporation Commtsston Post Office Box 355 799 Roosevelt Road. Bldg_ 14 Pittsburgh, Pennsylvanta 15230 Glen Ellyn. 1111n01s 60137 Michael Miller. Esq. Mr. M1chael C. Parker, Chief Sidley and Austin Dtviston of Engtneering One First National Plaza Illinois Department of Chicago, Illinois 60603 Nuclear Safety 1035 Outer Park Drtve Springfteld. Illinots, 62704 Mrs. Phtlltp B. Johnson 1907 Stratford Lane Joseph Gallo, Esq.

Rockford, I11tnots 61107 Hopkins and Sutter Sutte 1250 Ms. Lorratne Creek 1050 Connecttcut Avenue, N.W.

Rt. I, Box 182 Washington. D. C. 20036 Manteno, Illinois 60950 Douglass Cassel. Esq.

Dr. Bruce von Zellen 109 N. Dearborn Street Department of Bio10gtca1 Sctences Sutte 1300 Northern Illinois University Chicago. Illinois 60602 DeKa1b, Ill1nots 61107 Ms. Pat Morrtson Mr. Edward R. Crass 5568 Thunder1dge Drive Nuclear Safeguards &Licensing Rockford, I111n01s 61107 Sargent &Lundy Engineers 55 East Monroe Street Attorney General Chicago, Illinois 60603 500 South 2nd Street Springfield. Illinois 62701 U. S. Nuclear Regulatory Comm1ss10n Byron/Resident Inspectors Offices Chairman. Ogle County Board 4448 North German Church Road Post Offtce Box 357 Byron. Illinois 61010 Oregon, Illinois 61061 ESI Review Coordinator Environmental Protection Agency Region V 230 S. Dearborn Street Chicago, Il1in01s 60604 Commonwealth Ed1son Company Byron Station Manager 4450 North German Church Road Byron, 1111nots 61010

Mr. Henry E. Bliss Braidwood Station Commonwealth Edison Company Units 1 and 2 cc:

Mr. William Kort1er Ms. Lorraine Creek Atomic Power Distribution Route I, Box 182 Westinghouse Electric Corporation Manteno, Illinois 60950 Post Office Box 355 Pittsburgh, Pennsylvania 15230 Douglass Cassel, Esq.

109 N. Dearborn Street Joseph Gallo, Esq. Chicago, Illinois 60602 Hopkins and Sutter 1050 Connecticut Ave., N. W. Mr. Charles D. Jones. Director Suite 1250 Illinois Emergency Services Washington, D. C. 20036 and Disaster Agency 110 East Adams Street C. Allen Bock. Esq. Springfield, Illinois 62706 Post Offices Box 342 Urbana, Illinois 61801 Michael Miller, Esq.

Sidley and Austin Ms. Bridget Little Rorem One First National Plaza Appleseed Coordinator Chicago, Illinois 60603 117 North Linden Street Essex, Illinois 60935 George L. Edgar Mr. Edward R. Crass Newman &Holtzinger, P.C.

Nuclear Safeguards and 1615 L Street, N.W.

Licensing Division Washington, D.C. 20036 Sargent & Lundy Engineers 55 East Monroe Street Attorney General Chicago, Illinois 60603 500 South 2nd Street Springfield, Illinois 62701 U. S. Nuclear Regulatory Commission Resident Inspectors Office EIS Review Coordinator RR#l, Box 79 EPA Region V Braceville, Illinois 60407 230 S. Dearborn Street Chicago, Illinois 60604 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road, Bldg. 14 Glen Ellyn. Illinois 60137 Chairman Will County Board of Supervisors Will County Board Courthouse Joliet, Illinois 60434

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 ENCLOSURE 1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATING TO NATURAL CIRCULATION COOL DOWN BYRON UNITS 1 AND 2 AND BRAIDWOOD UNITS 1 AND 2 COMMONWEALTH EDISON DOCKET NOS. 50-454/455/456/457

1.0 INTRODUCTION

Branch Technical Position (BTP) RSB 5-1, "Design Requirements of the Residual Heat Removal (RHR) System," requires that test programs for pressurized water reactors (PWRs) include tests with supporting analysis to (1) confirm that adequate mixing of borated water added prior to or during cooldown can be achieved under natural circulation conditions and permit estimation of the tfmes requfred to achfeve such mfxfng, and (2) conffrm that the cooldown under natural circulation conditions can be achieved within the limits specified in the emergency operating procedures. In addition, the plant is to be designed so that the reactor can be taken from normal operating conditions to cold shutdown using only safety-grade systems. A comparison of performance to that of previously tested plants of similar design may be substituted for these tests.

Byron Units 1 and 2 and Braidwood Units 1 and 2 (Byron/Bra1dwood) are each classified as a Class 2 plant with regard to the implementatfon of the above BTP.

A natural circulation/boron mix1ng/cooldown test was performed at Dfablo Canyon Unit 1 on March 28-29. 1985. By memorandum dated February 4, 1987, the staff determined on the basis of the Diablo Canyon Unit 1 tests and submittals and the Brookhaven National Laboratory (BNL) technical evaluation report (TER),

2 that the Diablo Canyon Unit 1 systems meet the intent of BTP RSB 5-1 for a class 2 plant. In response to FSAR question 212.154 the licensee provided a comparison of natural circulation cool down for Byron/Braidwood with Diablo Canyon Unit 1. This included the following:

1) A comparison of the Diablo Canyon plant and Byron/Braidwood to demonstrate their similarity.

2) A cold shutdown scenario for Byron/Braidwood using the constraints of BTP RSB 5-1.

By letters dated October 7 and November 2, 1988, in response to NRC staff questions, the licensee provided additional information.

The staff safety evaluation for Diablo Canyon Unit 1, with the attached BNL TER entitled "Technical Evaluation Report for Diablo Canyon Natural C1rculation, Boron Mixing, and Cooldown Test," identified the plant parameters that may affect application of the test results to other plants.

These parameters are the basis for our evaluation and are discussed in the following sections.

2.0 EVALUATION Natural Circulation Diablo Canyon Unit 1 is rated at 3338 Mwt and has four loops in its reactor coolant system (RCS). Byron/Braidwood are of the four-loop RCS design and each rated at 3411 Mwt. The licensee has stated that the general conf1guration of the piping and components in each reactor coolant loop is the same in Byron/Braidwood as in D1ablo Canyon Unit 1. Significant parameters governing natural circulation are hydraulic flow resistance and thermal driving head. To demonstrate similar1ty in design for natural circulation, these two parameters were compared.

3 Data from Table Q212.154-1 show that the Byron/Braidwood hydraulic resistance coefficients at normal flow conditions are slightly lower than Diablo Canyon's.

A comparison of steam generator design between that of Diablo Canyon (Model 51) and Byron/Braidwood (Models 04/05) was made. The licensee concluded that the thermal driving head during natural circulation conditions would be greater at Byron/Braidwood than that of Diablo Canyon. The expected natural circulation flow at Byron/Braidwood considering these effects would be within 2% of Diablo Canyon. The thermal ratings of Byron/Bra1dwood and Diablo Canyon are close and differences in reactor power and decay heat levels between the plants are not expected to alter this conclusion. We find this explanation acceptable.

RCS Cooldown The plant's ability to cool the RCS at a specified cooldown rate, assuming a sufficient supply of auxiliary feedwater and subcooled RCS, is determined by the capacity of the atmospheric steam dump (ASD) valves. Steam flow through these valves removes the sensible heat and decay heat throughout the cool down period. The end of the cool down period, when the steam generator pressure is low, provides the most limiting conditions for valve capacity. The energy to be removed is determined by the water inventory and the amount of structural material in the RCS, the level of decay heat, and the cooldown rate.

The ASD valves for Byron/Braidwood are safety grade. Four valves are provided, one per steam generator. Each valve is hydraulically operated.

In the event of a single failure, three steam generators would be available for cooldown. The licensee stated, that as a result of an analysis based on a steam generator tube rupture, coinc1dent with natural c1rculation conditions, the plants could be cooled down with two steam generators. Therefore we find that there is reasonable assurance that the ASDs have sufficient capacity to perform an RCS cool down to the RHR in1tiation temperature in a reasonable time and the ASD capacfty is therefore acceptable.

4 Bypass Flow and Upper-Head Cooling A potential exists for void formation in the upper-head of the reactor vessel during the cool down/depressurization under natural circulation cond1t1ons if the upper head 1s relatively isolated from the rest of the ReS and its fluid temperature remains higher than the coolant temperature in the main flow paths of the RCS. Upper-head cooling under natural circulation cond1t1ons is influenced by core bypass flow and m1xing in the upper head.

Westinghouse plants may be divided into two groups according to the magnitude of the bypass flow: Thot and Tcold plants. For the Tcold plants, such as Byron/Braidwood sufficient bypass flow exists to make the temperature of the upper head fluid essentially equal to the cold-leg temperature. On the other hand, for the Thot plants, which 1ncludes D1ablo Canyon, the bypass flow 1s much smaller. For Thot plants this circumstance results in upper head temperature ranging between the cold-leg and the hot-leg temperatures and ra1ses a possibility of void formation in the upper-head region.

Contr1buting to the Tcold des1gn at Byron/Braidwood are spray nozzles located between the downcomer and upper head area. These spray nozzles allow better flow communication and mix1ng in the upper head during natural c1rculat1on.

The upper head volume for Byron/Braidwood 1s larger than that of Diablo Canyon. The NRC staff considers the upper head volume effect on coo11ng of the upper head to be small compared to the contr1bution of flow through the spray nozzles. We would therefore expect a shorter coo11ng time for a Tcold plant compared with that of a Thot plant of s1milar thermal rat1ng.

Boron Mix1ng The D1ablo Canyon boron mix1ng test evaluation demonstrated adequate boron mixing under natural c1rculation conditions when highly borated water was 1njected into the RCS. Contributing to the d1ffusion of the boron is the mixing effect created as the flow passes through the reactor coolant pumps

5 and the steam generator tubes. The plant's ability to achieve the proper shutdown margin, however, depends mainly on the injection rate of boron relative to the total inventory of water in the RCS. The required concentration change of about 300 ppm for the test was aChfeved fn less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The source of boron for Byron/Braidwood are the boric acfd tanks (BATs) and refueling water storage tanks (RWSTs) which have a boron concentration significantly less than that of Diablo Canyon. Thus, addition of a larger quantity of borated water over a longer time will be required to reach the desired concentration change. The licensee provided calculated results that indicate that less than two hours is needed for Byron/Braidwood to achieve the same concentration change demonstrated in the test. On the basis of those results the staff finds that there is reasonable assurance that sufficient time exists for boron injection and mixing to achieve the required shutdown margin.

Under normal operation at Byron/Braidwood, the borfc acid solution is injected into the RCS via the charging and reactor coolant pump seal injection lines.

Upon loss of instrument air, charging and boration injection flow is available through the high head safety injection flow path. The volume supplied via this path would be controlled through on-off operation of the charging pumps.

A concern during the boron mixing period of the natural circulation would be the RCS pressure increase due to the injection of additional mass into the system without letdown. The licensee has stated that during natural circulation conditions, sufficient boron can be added to the ReS, without letdown, to reach a cold shutdown boron concentration condition without going solid in the pressurizer. In addition, if the ReS pressure increases, the pressurizer power operated relief valves would be utilized for ReS pressure control. We find there is reasonable assurance that adequate ReS pressure and inventory control is available during boron addition.

6 Depressurization The Diablo Canyon test demonstrated that the RCS could be depressurized from cooldown conditions to the RHR initiation pressure under natural circulation conditions using the pressurizer auxiliary spray and/or pressurizer power operated relief valves.

At Byron/Braidwood, depressurization may be accomplished through the use of the pressurizer PORVs or the pressurizer auxiliary spray. However, the pressurizer auxiliary spray is not safety grade, and thus is not available for the RCS depressurization per BTP RSB 5-1.

The licensee stated that either of the two pressurizer power operated relief valves (PORVs), discharging to a pressurizer relief tank (PRT), is capable of providing the depressurization function. The PORVs are environmentally qualified and therefore will not be adversely affected by a harsh environment in the event of a PRT rupture disk failure. In the event that a PORV fails open, the PORV block valves, which are safety-grade, may be used to block the affected PORV flow path. Each PORV is designed with a safety related air accumulator to allow 50 cycles of valve operation. The licensee further stated that this capability is more than adequate to permit cooldown and depressurization during natural circulation conditions.

At the end of the depressurization, the ReS is approximately at 400 psig.

The RHR system may now be placed in service and the cooldown to cold shutdown condition continued.

We find there is reasonable assurance that adequate means are available for depressurization and reaching cold shut down conditions.

7 Cooling Water The primary auxiliary feedwater supply to the steam generators is provided by the condensate storage tank (CST) at Diablo Canyon and Byron/Braidwood. An additional source of auxiliary feedwater for Byron/Braidwood is the seismic category I essential service water (ESW) system. In the event of an insufficient water supply from the CST, the auxiliary feed pump suction is automatically switched to the backup source.

The Byron/Braidwood CSTs each has a capacity of 500,000 gallons with a technical specification (TS) minimum capacity of 200,000 gallons. The alternate or backup water source is the essential service water (ESW) system for both plants. For Byron, the ESW cooling basins have a 500,000 gallon capacity under normal operdtion (290,OOO gallon required by Technical Specifications). At Braidwood the ESW is provided by a cooling lake which provides an essentially unlimited back up auxiliary feedwater supply. Thus, a large backup supply of cooling water 1s ava1lable for Byron/Braidwood. We conclude that there is reasonable assurance that sufficient cooling water inventory exists to meet the proposed plant cooldown method *

3.0 CONCLUSION

The staff assessed the capability of Byron/Braidwood, to meet the requirements of RSB BTP 5-1. We have identified and evaluated the plant parameters that may affect application of the Diablo Canyon natural circulation test results to Byron/Braidwood.

On the basis of the licensee's submittals, and our evaluation as previously discussed. we conclude that the licensee has demonstrated that the D1ablo Canyon natural c1rculation tests are applicable to Byron/Braidwood and that they comply with the requ1rements of BTP RSB 5-1.

Principal Contributor: D Katze Dated: November 4. 1988