3F1297-24, Provides Supplemental Info to Support Util Response to NRC Request for Addl Info,Dtd 970128.Clarification to Util RAI Responses,Discussed During NRC Audit USI A-46 & Calculations Encl also.W/10 Oversize Drawings

Provides Supplemental Info to Support Util Response to NRC Request for Addl Info,Dtd 970128.Clarification to Util RAI Responses,Discussed During NRC Audit USI A-46 & Calculations Encl also.W/10 Oversize Drawings

ML20203J282
Person / Time
Site:	Crystal River
Issue date:	12/16/1997
From:	Rencheck M FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20203J289	List: 3F1297-24, Provides Supplemental Info to Support Util Response to NRC Request for Addl Info,Dtd 970128.Clarification to Util RAI Responses,Discussed During NRC Audit USI A-46 & Calculations Encl also.W/10 Oversize Drawings ML20203J303 ML20203J315 ML20203J370 ML20203J385 ML20203J400 ML20203J417 ML20266D724 ML20266D729 ML20266D734 ML20266D740 ML20266D744 ML20266D748 ML20266D752 ML20266D758
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR 3F1297-24, GL-87-02, GL-87-2, TAC-M69440, NUDOCS 9712190180
Download: ML20203J282 (48)

Text

AM I

Florida l

Power 8 222a2T ""J o O %i*.'. m .n

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December 16,1997 -.

3F1297-24 _

1 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 l

Subjwt: Supplemental Response for the Resolution of Unresolved Safety issue (USI) A-46 ,

(Generic letter 874)2)(TAC No. M69440) ,

}

References:

1.

2.

FPC to MRC letter,3F1295-18, dated December 31,1995 NRC to FPC letter,3N0596-04, dated May 2,1996 l[ h

[

i

3. NRC to FPC letter,3N0197-20, dad January 28,1997 I

4. FPC to NRC letter,3F0397-28, dated March 27,1997 t al

5. FPC to NRC letter,3F089/-01, dated August 1,1997 j

6. FPC to NRC letter,3F1297-33, dated I,ecember 8,1997 ri i i  !

I J

Dear Sir:

N i

The purpose of this letter is to provide supplemental information to suppor, Florida Power r jJ Corporation's (FPC) responses to the NRC's Request for Additional Information (RAI), dated h January 28,1997 (Reference 3). In addition, this letter provides clarification to FPC's RAI l responses which were discussed during the NRC audit of USl A-46 (References 4 and 5).  ;

fl

{ d On November 4 - 7, 1997. an NRC USI A-46 audit was performed by Messrs. P.Y. Chen and Mj i

S.B. Kim of the NRC/NRR, and Mr. K. Bandyopadhyay of Brookhaven National laboratory, s W

The purpose of this audit was to verify that FPC's USI A-46 program was effectively being f, j l implemented. The NRC audit team recognized that FPC had implemented additional actions which b .Q j were not speci0cally addressed in FPC's RAI responses. The team also requested clarification to 4 some of the RAI responses. both FPC and the NRC audit team agreed that a supplemental l response was needed to address these remaining issues. gg, E

Responses to RAI question numbers 1, 2, 4, 10, 11, 12, 14b, 14c 16,18g, 20b, and 23, as E /  ;

submitted in References 4 and 5, hve been clarificd or amended, and are provided in Enclesure 2. R -

As discussed with the NRC, the responses to the remaining RAI questions are either acceptable as Q,.

originally submitted, or were found to be acceptable during the audit, and do not require any a;, , .3 further clarification. E J g ;ggg ,

y

['

i, in addition, as discussed during the audit, the seismic adequacy of the Emergency Feedwater (EFW) tank has been confirmed. The evaluation is prcvided as Attachment B to this submittal as k

o Calculation S974)316. Also, as requested, FPC is providing copies of the original design basis 4 H calculations for selected tanks in Attachment F to this submittal. .O CRY 8TAL RIVER ENERGY COMPLEX: 187a0 w power une street . Crystal River, Florida 34428 4708 . (352)795 4486 9712190100 971216 . progress camp y PDR ADOCK 05000302 -

P PDR

LU.S. Nuclear Regul: tory Commission

' 3F1297-24 Page 2 of 3-

, ' On December 2,1997, FPC met with the NRC Staff to discuss the plans for the resolution of USI A-46. At.that meeting, FPC stated the schedule for the resolution of USI A-46 will be improved by completing the resolution of all outliers prior to startup from Refueling Outage 11 instead of Refueling Outage 12.- Seventy (70) outliers whien were part of our level I and level 11 systems (safety-related), as defined by the System Readiness Review Plan, are being resolved prior to startup from the current outage. Forty-three (43) outliers in Ixvel 11 and Level 111 systems (non-safety related) are labeled as post-restart and are being resolved prior to startup from Refueling Outage 11. Ilowever, of these forty three (43) post-restart items, FPC has completed seventeen (17) items during the current outage. The remaining twenty-six (26) post-restart _ outliers will receive an operability assessment prior to restar* from the current outage to confirm that the failure of any of the outstanding non-safety related outliers will not affect the ability to achieve safe shutdown of the plant. This assessment will also include a review of potential seismic interaction concerns.

As part of the closure to USI A-46, FPC will perform a confirmatory self-assessment of USI A-46 activities for CR-3. This effort will assess the completeness of the resolution of USI A-46 issues.

This action will be completed prior to startup from Refueling Outage 11.

Enclosure 1 provides additional clarification and information that was discussed during 'the NF C audit- Enclosure 2 provides clarifications and supplemental information for selected RAI responses, as discussed above. Enclosure 3 provides the restart outlier resolution schedule and status. Enclosure 4 provides the post-restart outlier resolution schedule and status. Enclosure 5 p ovides a list of FPC commitments, if you have any questions regarding this letter, please contact Mr. David F. Kunsemiller, Manager, ,

Nuclear Licensing at (352) 563 4566.

Sincerely,

.W. nc k, Director

/-

Nuclear Engineering and Projects MWR/cgp

~

cc: Regional Administrator, Region II (w/ encl. & att.)

_ Senior Resident inspector (w/ encl. & att.)

NRR Project Manager (w/ encl. & att.)

P.Y. Chen (w/ encl. & att.) -

t _ . . _. __

J

i

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. .. . , . u .

LU.S? Nuclear Regulatory Commission?

3F1297 24 .

? ._

L Page 3 of 3i '

'inclosures:

l. . c Response to .NRC Audit issues 21 Supplemental Response to NRC RAIi

' 3. - Restart Outlier Resolution Schedule and Status -

4. . Post-Restart Outlier Resolution Schedule and Status .

. 5.c : List of Regulatory Commitments:

Attachments:

0 A. 1.1 Introduction

12. Caveat and Anchorage Review (Stevenson & Associates' Letter dated 12/9/97)

3. " Extent-of-Condition" Review for Thin Base Metal Anchorage - Class 20

4. Calculation S97-0541, Rev 0, " Confirmatory Study of Selected Safe :

Shutdown Equipment Anchorage for USI A-46"

. B. Calculation S97-0316, Rev. O, " Seismic Evaluation of Emergency Feedwater Tank,

.  : EFT-2"-

g

' C. ~' Calculation S-96-0013, Rev,1,- " Qualification of Tanks per U.S.I. A-46" .

D. - Calculation S97-0542, Rev. O, "CR3 Structural Margin Evaluation (Study)" l E. . Screening Evaluation Work Sheets (SEWS) for RCPM-3A and RCPM-3B F. Design Basis Tank Calculations for:

The Pressurizer Quench Tank (Tag Number WDT-5)

Reactor Coolant Bleed Tanks (Tag Numbers WDT-3A, 3B, 3C)

Dedicated Emergency Feedwater Tank (Tag Number EFT-2)

. G.- Letter on Grounding Resistor  !

. H. . Memo on " Resolution _of Questions Regarding Possible Mis-classification SSEL Equipment" l.. Discussion on Specific Cabinet Internals and Weight Concerns 7  ? J. ;Information Copies of Referenced Drawings E

d $

d f

4 y, .-A--..,. *

.x I

I 1

l ENCLOSURE 1 i

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKE'I NUMBER 50-302/ LICENSE NUMBER DPR-72 1

l RESPONSE TO NRC AUDIT ISSUES l

4

- . . . - , - ~ .

U.S. Nuclear Regulatory Commission Enclosure 1-3F1297-24 Page 1 of 6 -

I NRC Audit Question 1:

The licensee willprovide clarifications to its earlier response or additional trl formation in response to the remaining RAIitems (i.e., RAIItem Nos. 1, 2, 4,10,11,12,14t, c,18g, and 23 [ sic]). The response will address the SE open issues directly on an item-by-item basis.

FPC Response:

- Supplemental information and clarifications for each of the remaining RAI items listed acd open issues from the SER are provided in Enclosure 2. In addition, supplemental

~

information for RAI items 16 and 20b are also provided.

NRC Audit Question 2:

The licensee will corfirm that all equipment i.; anchored and anchorage has been inspected following the GIP 2 guidelines. In addition. Ihe licensee willprepare bounding calculations in accordance with GIP-2 for verification of anchorage. As a minimum, all poorly or improperly anchored equipment will be identiped and checkedfor anchorage adequacy.

FPC Response:

FPC's response to the specific information requested above is discussed in FPC's response to NRC RAI Request Number 12 in Enclosure 2 and in Attachment A of this submittal. l The bounding calculations are contained in Calculation S97-0541, which is provided in Attachment A of this submittal.

U.S. Nuclear Regulatory Commission Enclosure 1 3F1297-24 Page 2 of 6 NRC Audit Question 3:

The licensee will cortfinn that all GIP-2 caveats have been ven)Ld during seismic

• walkdown " and satisfiedfor all equipment items.

FPC Response:

Supplemental infoimation to address this issue is provided in response to NRC RAI Request Number 11 in Enclosure 2. In addition, an independent review was performed by Stevenson and Associates. The assessment report is included in Attachment A of thi.c submittal.

NRC Audit Question 4:

The licensee will develop a top level procedure for operator action, in general, and relay chatter, in particular, and submit itfor staf review.

FPC Response:

FPC procedure AP-961, Earthquake, is being revised to provide guidance to operators on <

how to cope with relay chatter subsequent to a seismic event and identificanon of bad actor relays. FPC will submit the revised AP-961 to the NRC for infonnation.

FPC will revise Section 6 of the CR-3 Plant Specific Procedure (PSP) to add a description of the relay functionality review that was performed as part of the relay walkdown. This is discussed further in Enclosure 2, SER Concern 2.

_a

. .. -- ._ . - -- _ = - _--- - -

U.S. Nuclear Regulatory Commission Enclosure 1 3F1297-24 Page 3 of 6 NRC Audit thestion 5:

The licensee will develop a complete list of equipment in the USI A-46 safe shutdown paths and cor$nn that no additional equipment items need to be included in the current list.

FPC Response:

The current Safe Shutdown Equipment List (SSEL) is considered complete. The list has been prepared and thoroughly reviewed by a multi-disciplined team which included the

- Operations Department.

FPC plans to perform a confirmatory audit to address this SSE! 8" sue prior to the end of Refueling Outage 11. This is related to commitment 3F1297-24v ; t Enclosure 5.

4 f

NRC Audit Question 6:

The licensee will venfy that all equipment items are appropriately classified and re-evaluate inappropriately classified equipment.

FPC Response:

FPC has completed a review of the specific components discussed during the Audit (Transfer Switches VBXS-1 A, IB, IC, ID, IE, 3A, 3B, 3C, 3D, DPXS-1). In addition, FPC has completed a review of all other equipment classifications. This review indicates the classification of the identified components is appropriate as shown on the FPC PSP Screening Evaluation Work Sleets (SEWS). This is discussed further in Attachment H of this submittal.

f i

iU.S. Nuclear Regulatory Commission Enclosure-l -

J 3F1297-24 . Page 4 of 6  :

a. i NRC Audit Question 7: -

The licensee will venfy whether the Diesel Generator grounding resistors 'need to be

- included in the safe shutdown equipment list. If so, the seismic ad quacy of the ceramic  :

insulators in the loadpath should be investigated. l

-i

- FPC Response: - l FPC has completed a review of the ; diesel generator grounding resistor. The Lreview

~ determined that the ground resistor is not required for the safety function of the diesel  ;

generator or for safe shutdown capabilities. This is discussed fmdier in Attachment G of this submittal.

4 5

i NRC Audit Question 8:

The licensee will cortfirm that the questionable cabinet internals (cantilever box and ceramic ,

1 insulators) are, in fact, acceptable (or otherwise have been modified). The licensee will also cortfirm that similar situations do not exist in other equipment.

4 4

FPC Response:

FPC has'compkted a review of the above listed components. It is FPC's understanding that the canulever box issue dealt with the Anticipaid Transient Without Scram Logic Cabinet (tag number ATCP-1), The ~ ceramic insulator issue dealt with 480V- Turbine-Auxiliary Bus A (tag number MTSW-3A) and the 480V Reactor Auxiliary Bus A (tag

number MTSW-3C). The issue'with 4160V Engineered Safeguards (ES) Bus 3B - Morth (tag number MTSW-2E) and the 4160V ES Bus 3B - South (tag number MTSW-2F) dealt with a potential transformer mounted on top of the cabinets. This is' discussed further in -

Attachment I of this submittal.-

4 4

4 y ,.4y= -y g g-se 4 - - -op +-. me w -_eg- - -w-+e ,mq e.y-e.- - - ,w n4.-.ew.4 eeei-e 46 rwa +~ e, 4w 4.--m' *-- eA44' *- r @4"T

U.S. Nuclear. Regulatory Commission Enclosure 1. a 3F1297 24 = Page 5 of 6:

NRC Audit Question 9:- .

- 1he licensee will submit sample calculationsfor cable and conduit supports.

FPC Response:

' Subsequent to the USI A-46 audit, FPC provided sataple design basis calculations for cable and conduit supports to the NRC. FPC has received feedback from the NRC that indicates .

there _ are no open issues with FPC raceways. - This subjer is also~ discussed 'in the-supplemental response to NRC RAI Request Number 16 in Enclosure 2. FPC has '

performed a Limited Analytical. Review (LAR) of select cable and conduit supports to Lverify our earlier conclusion that CR-3's configuration is acceptable. = The LARt are '

included in calculation S97-0542 (Attachment D of this submittal).

NRC Audit Question 10:

The licensee will complete and then submitfor staf review calculationsfor outlier tanks.

FPC Response:

p In Reference 5, FPC stated in the original response to NRC RAI Request Number 23 that minor programmatic deviations were discovered. These deviations consisted of having '

several tanks and heat exchangers not having completed calculations as required by PSP.

These tanks and heat exchangers were declared outliers originally due to the lack of formal calculations. The following items were listed as post-restart outliers. However, as committed-in Reference 5, FPC has completed the calculations. The specific tanks in c - question for this concern are: Reactor Coolant (RC) Drain Tank 3A (tag number WDT- .

- 3A), RC. Drain Tank 3B (tag number WDT-3B), RC Drain Tank 3C (tag m.mber WDT-3C), RC Drain Tank (tag number WDT-5), Main Condenser A _(tag number CDHE-4A),

l and Main Condenser B (tag number CDHE-4B). The calculation, S-96-0013, Revision 1~,-

has been completed. This calculation is included in Attachment C of this submittal.

~

The Condensate Storage Tank (tag number CDT-1)-has been recently identified as an

- outlier. An error in the existing calculation was found during a review by Dr. Robert Kennedy.' His review was part of. a third party assessment of our structural extent of condition review. A Precursor Card was generated to document this error (PC 97-7423).

=This issue was' discussed with the NRC during the audit. CDT-1 is included in the group of -

y --- 9- s+-

- , m e ry vyr.- - w -p,9p, f

U.S. Nuclear Regulatory Commission Enclosure 1 3F1297-24 Page 6 of 5 post restart outliers. FPC has performed a seismic margin calculation as part of the operability assessment. This calculation is contained in Attachment D of this submittal.

FPC has made an attempt to locate original design basis calculations for this tank and has been unsuccessful to date. FPC has initiated another Precursor Card to document our inability to locate this calculatiori in a timely manner (PC 97-8523).

NRC Audit Question 11:

The licensee will cortfirm that all outliers have been satisfactorily resolved through design documentation andfield rnodifications.

4 FPC Response:

Enclosures 3 and 4 provide oudier resolution schedule and status lists. FPC will provide confinnation of completion of resolution of these outliers. The schedule for resolution of A-46 has been improved from-FPC's earlier conunitment of Refueling Outage 12. FPC will resolve the seventy (70) identified restart outlier issues prior to the unit restart from the current outage and complete the resolution of post-restart outliers by Refueling Outage 11.

. - Currently, FPC has completed seventeen (17) of the forty-three (43) post-restart outliers. A status of outlier resolutions was discussed with the NRC in a meeting held on December 2, 1997. The remaining twenty-six (26) post-restart outliers will receive an operability assessment prior to restart from the current outage to confirm that the failure of any of the outstanding non-safety related outliers will not affect the ability to achieve safe shutdown of the plant.

As part of the closure to USI A-4(,, FPC will perform a confirmatory self-assessment of USI A-46 activities for CR-3. This effort will assess the completeness of the resolution of USI A-46 issues. Upon completion of this confirmatory review of the A-46 program, FPC will infonn the NRC of resolution of USI A-46 for CR-3. This action will be completed prior to startup from Refueling Outage 11,

l ENCLOSURE 2 i

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DP.R-72 a

SUPPLEMENTAL RFsSPONSE TO NRC RAI i

VERIFICATION OF SEISMIC ADEQUACY OF EQUIPMENT IN OPERATING REACTORS UNRESOLVED SAFETY ISSUE (USI) A-46, GENERIC lei FER 87-02

U.S. Nuclear Regulatory Commission Enclospre 2 3F1297-24 Page 1 of 26 CRYSTAL RIVER UNIT 3 .

l SUPPLEMENTAL RESPONSE TO NRC RAI UNRESOLVED SAFETY ISSUE (USI) A-46, GENERIC LETTER 87-02 By letter dated December 31,1995 (3F129518), Florida Power Corporation (FPC) provided the documentation of the seismic evaluation (The Report) performed to address USI A-46 at Crystal River Unit 3 (CR-3). The evaluation was performed using FPC's Plant Specific Procedure (PSP) for resciving USI A-46.

By NRC letter dated January 28,1997 (3N0197-20), the NRC determined that additional inforn ation was necessary to complete their review of the CR-3 Report and provided a Request for Additional Information (RAI). FPC completed responses to the NRC's RAI and submitted them by FPC letters dated March 27,1997 (3F0397-28), and dated August 1,1997 (3F0897 01). In addition FPC submitted a letter to the NRC dated December 8,1997 (3F1297-33) providing " Plans for Resolution of USI A-46 and Large Bore Piping and Piping Supports."

The NRC conducted an Audit of the CR-3 USI A-46 Program during November 4 - 7, 1997, to verify that the program was effectively being implemented. Responses to RAI questions 1, 2, 4,10,11,12,14b,14c,16,18g, 20b, and 23 have teen clarified and amended. No additional information for the other RAI questions is incluid with this submittal, based on our understanding that those issues were satisfactory and closed by the NRC audit team. The following provides FPC's responses to the above RAI questions.

U.S. Nuclear Regulatory Commission Enclosure 2

. 3F1297-24 Page 2 of 26

- NRC RAI REQUEST NUMBER 1 In the Safety Evaluation (SE) '(Reference 1), the staff has taken several exceptions and identified specific issues related to your A-46 implementation procedures (References 2 and 3). Since you performed :he equipment verification (called walkdown) before receiving the SE, your walkdown report *(Reference 4) does not completely address the staff concerns.

Moreover, since the walkdown report basically contains a summary of the data, it is not clear from the report whether and how many of the staff concerns have been addressed through the walkdown. Therefore, please provide the necessary information to show that the open issues identified in the SE (Reference 1) have been addressed during the walkdown.

FPC SUPPLEMENTAL RESPONSE:

The supplemental responses are included herein, the positive Ondings of the NRC A-46 audit team, and this response to the SER will address the Staff's concerns. The following information is provided to address each of the SER concerns.

SER Concern 1:

The licensee's approach to achieve and maintain hot standbyfor 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />sfollonin an SSE is acceptable provided that the licensee confirms that the equipment necessary to assure core decay heat removalfor 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in both of the safe shutdoun paths are seismically adequate.

FPC Response:

The following information is provided to summarize FPC's efforts to address the above issue.

Safe Shutdown Equipment List:

FPC ht, developed a comprehensive Safe Shutdown Equipment List (SSEL). Not only did this list include the equipment required to safely shut down the plant following an

. earthquake but also included electrical equipment that might contain control devices (refer to NRC RAI Question Number 4 for further information). The list was prepared using the guidelines of the GIP. The list was prepared by individuals trained by Seismic Qualification Users' Group (SQUG). This list was reviewed and accepted by FPC's Operations Department.

J

• Unless othensise mentioned, "the report" means Reference 4, and all subsequent page number and section umber citations arefrom this report.

U.S. Nuclear Regulatory Commission Enclosure 2 3F1297-24 Page 3 of 26 Seismic Walkdown:

FPC has completed all the seismic review walkdowns. These walkdowns were done by experienced and well trained Seismic Capability Engineers familiar with the GIP and FPC's PSP. The walkdowns were done using the PSP. liowever, FPC maintains the GlP caveats were implicitly included due to the experience and knowledge of the Seismic Review Team (SRT). The SSEL and walkdown effort had a third party review as documented in the report (Reference 1). FPC has recently had another review of our program to compare GIP caveats against the PSP SEWS evaluation. This review did not find any concerns. This review also performed an anchorage study that shows anchorage acceptable (except for those identified as outliers).

Outliers:

The seismic screening resulted in one-hundred-thirteen (113) outliers out of approximately eight hundred (800) items of equipment reviewed. FPC performed a ranking of the outliers and made a commitment to resolve the more safety significant outliers (seventy) prior to restart from the current outage. On December 2,1997, FPC met with the NRC Staff to discuss the plans for the resolution of USI A-46. At that meeting, FPC stated the schedule for the resolution of USI A-46 will be improved by completing the resolution of all outliers prior to startup from Refueling Outage 11 instead of Refueling Outage 12. Seventy (70) outliers which were part of our Ixvcl I and Level 11 systems (safety-related), as defined by the System Readiness Review Pir.n, are being resolved prior to startup from the current outage. Forty-three (43) outliers in Level 11 and level 111 systems (non-safety related) are labeled as post-restart and are being resolved prior to startup from Refueling Outage 11. Of those forty-three (43) post-restart items, FPC has completed seventeen (17) items during the current outage. The remaining twenty-six (26) post-restart outliers will rcceive an operability assessment prior to restart from the current outage to confirm that the failure of any of the outstanding non-safety related outliers will not affect the ability to achieve safe shutdown of the plant. De assessment will also include a review of any seismic interaction.

. SER Concern 2:

The licensee's approach to resolve relay issues is acceptable provided that the licensee revises Section 6 of CR-3 PSP, Revision I, to include its commitments to the staf positions as delineated in Reference 8.

FPC Response:

FPC will revise Section 6 of the CR-3 PSP to add a description of the relay functionality review that was performed as part of the relay walkdown (Commitment l

I

U.S. Nuclear Regulatory Commission Enclosure 2 3F1297-24 Page 4 of 26 3F1297-24-06). In addition, FPC will revise procedure AP-961, Earthquake, to include identi0 cation of the bad actor relays and guidance for appropriate operator action to cape with the malfunction of the relays. There is a commitment to provide the appropriate operator procedure for dealing with relays (Commitment 3F1297 04).

SER Concern 3:

The licensee's approach to evaluate the adequacy of equipment anchorage is wt completely acceptable. The staf will venfy the licensee's implementation of Section 4.4 of the GIP, Revision 2, and the summary report's documentation of the results of anchorage emluation and how any outliers are handled.

FPC nesponse:

FPC acknowledges there are three important attributes to verify adequate anchorage:

verifying actual anchorage is installed, verifying proper installation of the anchorage, and verifying there is adequate capacity of the anchorage.

Verifying Anchorage Installation Exists:

The primary action to verify the existence of anchorage has been satisned. FPC performed a 100% inspection of accessible anchorage and hands-on check of all non-energized equipment and energized equipment with external anchorage. There is a specific caveat an the PSP SEWS to verify the existence of anchorage. This has been documented on all applicable SEWS (the exception is line mounted equipment). Where the anchorage was missing, or considered poor, then that equipment was declared an outlier. Out of the one-hundred-thirteen (113) outliers, approximately twenty-seven (27) are because of missing or poor anchorage. At the conclusion of the outlier resolution effort, there will be no unanchored SSEL equipment. The PSP also required the Seismic Review Team (SRT) to verify anchorage load path.

In addition to the USI A-46 effort, there have been many inspections and other programs at FPC that help verify the existence, or other concerns with anchorage. The Structural Maintenance Rule Inspections made hands-on inspections of virtually every accessible anchorage in the plant. The System Readiness Review Plan performed walkdowns of plant equipment.

Verifying Proper Anchorage Installations:

The second area regarding proper installation hac also been satisfied. All anchor bolts were hand checked and visually inspected where accessible during the A-46 walkdowns. Again, any concerns noted were required to be documented on the SEWS.

U.S. Nuclear Regulatory Commission Enclosure 2 1 3F1297 24 Page 5 of 26 in addition to the A-46 walkdowns, the Structural Maintenance Rule inspection performed a visual inspection of anchorage in the plant (including non-SSEL and SSEL equipment). The Maintenance Rule walkdowns also included a wrench tightness test that was performed for one-hundred-ninety-three (193) expansion anchors using-methods based on the GIP. The one hundred-ninety-three (193) expansion anchors tested were a subset of the three-hundred-fifty-four (354) expansion bolts included in the scope of the Maintenance Rule. - Anchors tested represented approximately sixty

. (60) components; all but six (6) are SSEL items. Out of the one-hundred-ninety-three (193) bolts tested, three bolts failed the acceptance criteria. A follow up analysis conclude <i the loose bolta did not adversely affect the ;dequacy of the equipment. This informatiol is contained in the Structural Maintenance Rule inspection Report.

Verifying Anchorage Capacity:

The third area deals with the capacity of the anchorage. The program at FPC relied on the judgment and experience of well trained Seismic Capability Engineers to make the determination if the capacity of the anchorage is adequate. FPC understands the Staff's concern about making an engineering judgment about the adequacy of an anchorage capacity. To address this issue, FPC has performed a bounding anchorage calculation based on the requirements of the GIP.

This bounding calculation took a subset of the electrical equipment contained in the SSEL and performs a bounding calculation. The intent of this bounding calculation is to envelope the equipment at CR-3 and to ensure equipment anchorage adequacy.

These calculations are done to the requirements of the GlP. FPC maintains that mechanical equipment is inherently more rugged and generically better anchored.

Therefore, no further review of mechanical equipment is warranted. Thit .tudy included the following equipment classes:

Equipment Class 1 Motor Control Centers Equipment Class 2 Low Voltage Switchgear Equipment Class 3 Medium Voltage Switchgear Equipment Class 4 Transformers Equipment Class 14 Distribution Panels Equipment Class 16 Battery Chargers and inverters Equipment Class 20 Instrument and Control panels and cabinets A summary of the results of the calculation is included in Attachment A of this submittal. Calculations to support this study are included in Attachment A (FPC Calculation S97-0541).

The NRC has noted several items of equipment that they felt necced further review.

These items are "noted" in Attachment A. The specific items of equipment and the

U.S. Nuclear Regul tory Commission Enclosure 2 3F1297 Page 6 of 26 findings of the FPC review are documented in Attachment A of this enclosure. Other than the item discussed next, all equipment anchorages were found to be acceptable. .

In addition, the NRC has identined an issue with an electrical enclosure that was anchored through the base sheet metal without stiffened elements. Specifically, this is

- the SCR Cabinet for AilllE-4A and AlliiE-4B (tag number AllCP-4). To add.ess this specinc concern, FPC has generated a Precursor Card to document the new anchorage outlier FPC also has generated an " extent of-condition" review to verify there are no other cabinets with this concern. The " extent-of-condition" review found no additional

. items (see Attachment A of this enclosure for more information).

As part of our confirmatory self-assessment, FPC will also perform additional anchorage calculations for approximately one hundred (100) electrical components.

. This is approximately 50% of the total scope of SSEL electrical components. Based on the results of the evaluation, FPC may expand the scope of the calculations utilizing statistical methods.

This concern is also being addressed by FPC response to RAI Question 12, and information contained in Attachment A of this submittal.

SER Concern 4:

The licensers' approach to emluate cable and conduit racermy is not acceptable. The licensee should submit a report summarizing the results of the cable and conduit racesmys assessment to verify its adequacy. The report should also detail the criteria and methodology mentioned in Reference 7.

j- FPC Response:

The evaluation of Cable and Conduit raceways was discussed during the NRC Audit.

The walkdowns by the Staff indicated agreement that the CR-3 raceways are a rugged design. In addition, sample design basis calculations were provided to the NRC. FPC has received feedback from the NRC that indicates there are no op:n issues with FPC raceways.

I Ilowever, FPC said it would provide additional justification. FPC management voluntarily initiated a structural extent-of-condition study, based on the IPEEE seismic i

margins program to verify adequacy of plant structures. To provide this justification, the seismic margin analysis included an expansion of scope to address electrical raceways. This involved a more detailed review of electrical raceways than is typically l performed in a seismic margins program. FPC has performed two Limited Analytical Reviews (LAR) for raceways using the GIP method. One LAR on a cable tray was performed to address the Thermo-Lag issue. These calculations showed positive results y-. __-- . .s -- . _ . - . s._.., ,,-,

U.S. Nuclear Regulatory Commission Enclosure 2 3F1297-24 Page 7 of 26 and further evaluated the acceptability of FPC's raceways. These LARs are included as part of FPC Calculation S97-0542. This calculation is incleded as Attachment D of this submittal.

Oversight of this effort was petformed by Dr. Robert Kennedy. Dr. Kennedy concluded that CR-3 electrical raceways should not require further review, especially if some LARs (mentioned above) are performed as a part of a seismic margin study.

SER Concei 3:

The proposed guidelines for emluating the seismic adequacy of tanks and heat exchangers are acceptable uhen proper documentation is provided. 1he licensee should include the evaluation of seismic adequacy of coreflood tanks in the A-46 program activitiesfor resolving USI A-40.

FPC Response:

FPC has completed analysis of the SSEL tanks and heat exchangers identined on the SSEL. There are eighty-seven (87) tanks and heat exchangers (equipment class 21) currently listed in the SSEL. The following list of calculations contain qua'ification of the listed tanks and heat exchangers. There is a subset of tanks and heat exchangers that are quali0ed by similarity and not specifically referenced in a calculation. That is, the item is qualified by reference to an item that is qualified by a calculation listed below. These calculations are included in the following FPC calculations:

1. S94-00ll, Revision 0, submitted previously as Attachment C, in Reference Letter 5. This calculation includes calculations for five tanks and heat exchangers.

2. S96-0013, Revision 0, submitted previously as Attachment 5, in Reference Letter 4. This calculation includes calculations for forty-two tanks and heet exchangers.

3. S96-0013. Revision 1, included with this submittal in Attachment C. This revision adds calculations for RC Bleed tanks (tag numbers WDT-3A,3B, and 3C), RC Drain tank (tag number WDT-5), and the Main Condensers (tag numbers CDilE-4A and 4B).

4. S97-0316, Revision 0, included with this submittal in Attachment B. This calculation includes the new calculation to resolve the outlier identified for the Emergency Feedwater Tank (tag number EFT-2).

The Condensate Storage Tank (tag number CDT-1) has been recently identified as an outlier. An error in the existing calculation (S94-00ll) was found during a review by

U.S. Nuc; car Regulatory Commission Enclosure 2 3F1297-24 Page 8 of 26 1

Dr. Robert Kennedy. Ilis review was part of a third party assessment of FPC's l structural extent of condition review. A Precursor Card (PC 97-7423) was generated to document this error. This issue was discussed with the NRC during the audit.

CDT-1 is included in the group of post-restart outliers. For an operability assessment, FPC has performed a seismic margins calculation. This calculation is contained in Attachment D of this submittal. FPC has made an attempt to locate original design basis calculations for this tank and has been unsuccessful to date. FPC has initiated another Precursor Card (97-8523) to document our inability to locate this calculation in a timely manner.

The Core Flood trsnks were also discussed during the audit. These tanks are part of the Nuclear Steam Supply System (NSSS). The NSd. has been specifically excluded from review by the A-46 program. Furthermore, drawing were produced and shown to the Staff during the audit that showed the core flood tanks clearly do not fall within the scope of USI A-40. USI A-40 was created to verify the acceptability of large flat bottomed tanks. The drawings showed the core flood tanks to be elevated tanks, supported by legs. Therefore, FPC concludes that no further work is required for these tanks.

SER Concern 6:

The ground response spectra and the approach used in developing the in-structure response spectra are acceptable provided that the licensee's verzfication of the seismic adequacy of equipment and anchorages is in accordance with the staf position delineated in Sections 2.3 and 2.7 of this evaluation and in Reference 8.

I'PC Response:

In Section .2.3 of the SER, the NRC took exceptior. to FPC's anchorage pasition.

Since the issuance of the SER, FPC has undertaken additional steps to provide assurance that the equipment identified on the SSEL is adequately anchored. These additional steps have been discussed with the Staff during the Audit and are again extensively addressed in this submittal. Specifically, anchorage is addressed in FPC's responses to SER Concern 3, RAI Question 12, and Attachment A of this submittal, in Section 2.7 of the SER, the NRC took exception to FPC's exclusion of certain GIP caveats. The information provided during the audit and with this submittal addresses the Staff's comments. FPC has reviewed the differences between the PSP and the GIP caveats and found that they are acceptable and do not adversely affect any of the conclusions of our program. This issue is discussed further in FPC's response to RAI Questien I'l and in Attachment A of this submittal.

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U.S. Nuclear RegulCtory Commission Enclosure 2 : -l 3F1297-24 Page 9 of 26  :

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- SER Concern 7:- i lhe imposed criterl.1 and procedures for addressing the genericcaveatsfor equipment classes are not completely acceptable. ,Ihe staff uill tese GIP generic caveats to evaluate the:.  ;

licensee's assessn.ent of equipment seismic adequacy in its A-46 implementation summary  :

report.

4 i FPC Response:

J.a stated above, information provided during the audit and with this submittal are intended to address all the Staff's concerns on this issue. This issue-is discussed' further in FPC's response to Question 11 and in Attachment A of this submittal.

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U.S. Nuclear Regulatory Commission Enclosure 2

- 3F1297 24 Page 10 of 26 l

NRC RAI REQUEST NUMBER 2 On Page 14, third paragraph, the walkdown summary report (Reference 4) states that "the i methodology used to identify the safe shutdown paths and components is in accordance with the Plant Specific Procedure (PSP) except as noted herein." llowever, the exceptions are not found in the report. Please identify the exceptions clearly so that the staff can evaluate their )

impact.

FPC SUPPLEMENTAL RESPONSE:

In a submittal dated August 15,1994 (3F0894-02), FPC provided the discussion of methodology for achieving and maintaining liOT STANDBY following a design basis seismic event at CR-3. The FPC A-46 program scope includes the systems and corresponding equipment necessary to ensure that ilOT STANDBY can be achieved and maintained for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following an SSE. For CR-3, given a loss of offsite power, it is not possible to cooldown to a llOT SilUTDOWN condition within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. FPC stated that for this condition, an additional 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> are required to cooldown to the decay heat system entry point. This is due to the limited capability to relieve steam through the atmospheric dump valves.

Recently, FPC has performed additional calculations of the plant response to a natural circulation cooldown using a more sophisticated model and more realistic assumptions. This analysis was performed to demonstrate compliance with 10 CFR 50, Appendix R, Section Ill.L.l(d), which is more restrictive than the PSP in that it requires it be demonstrated that the plant can be cooled to cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. This model used to demonstrate compliance with Appendix R includes a more realistic decay heat medel and takes cr:dit for the heat removed through operation of the turbine driven emergency feedwater pump. The results of this calculation indicate that it is possible for CR-3 to cool down to the point at which the decay heat removal equipment could be used in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a loss of off-site power as stated in the PSP. The result of this analysis is presented in more detail in letters to the NRC dated August 29,1997 (3F0897-48), and December 10,1997 (3F1297-04).

This revised analysis allows FPC to conclude that the methodology used to identify the safe shutdown paths and components is in accordance with the Plant-Specific Procedure PSP.

Therefore, based on the above resolution, FPC has no identified exceptions.

U.S. Nuclear Regulatory Commission Enclosure 2 3F1297-24 Page 11 of 26 NRC RAI REQUEST NUMBER 4 in item No.10 on Page 16 of the report, the equipment types that were not included for seismic evaluation include " equipment...which, upon loss of power, will fail in the desired position or state...." Please verify that, under all concerned plant conditions, the control devices of such equipment that may cause a failure of the equipment in an undesirable state have been included in the safe shutdown equipment list.

FPC SUPPLEMENTAL RESPONSE:

When FPC expanded the scope of the SSEL in response to Revision 1 of the PSP, all electrical enclosures (e.g., cabinets, panels, switchgear, MCCs) were evaluated. The evaluation excluded only enclosures that are clearly not important to safe shutdown of the plant. For example, these systems included fire service, and security. The remaining electrical enclosures that might contain contro) devices were included in the SSEL. The intent was to include any enclosure that might contain relays or other control devices associated with equipment on the SSEL. 'n the process, FPC conservatively included enclosures containing control devices for equipment not required for safe shutdown. This included equipment not identified during the original review. Therefore, there is an expectation that the control devices for couipment not on the SSEL have been evaluated in the same manner as the control devices for equipment that are on the SSEL. Thus, there is a high degree of confidence that these control devices will not cause a failure of equipment (which is not on the SSEL) in an undesirab;c state.

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-_U.S. Nucle:r Regulatory Commission Enclosure 2

- 3F1297 24 l Page 12 of 26: -

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NRC RAI REQUEST NUMBER 10

, The report on Page 36 Section 5.1.3 states that "all [ underline added] reinforced concrete pads are integrally attached to the concrete floors by dowels." Please explain how this was-verified. -

FPC SUPPLEMENTAL RESPONSE:

FPC acknowledges the Staff's concern about anchoring equipinent to raised pads that may not be reinforced, or attached to the base slab. As part of the A-46 review FPC performed walkdowns and reviewed applicable plant drawings to show that equipment pads are reinforced. Where applicable this information is documented on the SEWS. If this information could not be readily determined by a review of the plant drawings, then FPC '

conservatively declared the component an outlier. For example, the Control Complex HVAC Al.r Compressors (tag numbers AHP-01 A, OlB, OlC, and OlD) are included as outliers because pad reinforcement could not be determined. The resolution of these outliers required

- an ultrasonic test be done on the anchor bolts. The UT exam showed the anchor bolts themselves were doweled through the raised pad into the floor. Therefore, the anchorage for these items is considered adequate.

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1 U.S. Nuclear Regulatory Commission Enclosure 2 3F1297 Page 13 of 26 l

NRC RAI REQUEST NUMBER 11  :

In Reference.1, the staff has stated that meeting the caveats is an essential element of the experience-based approach documented in the Generic Implementation Procedure (GlP) and that it would use the GIP caveats to evaluate the licensecs' USI A-46 resolution program.

There are several caveats that are listed in the GIP but not in the PSP (Reference 2). Itis acknowledged that some justifications are provided in the Technical Basis document (Reference 3) to show that_the missing caveats are not of concern for Crystal River, mostly because of low seismicity. But, as the staff had 'already pointed out, meeting the caveats is a prerequisite for application of the experience-based-approach. Caveats were prepared by experts considering potential vulnerabilities of equipment. - The purpose was that an experienced engineer would go over the entire checklist of caveats to verify that there were no concerns for the identified vulnerabilities. For example, consider Caveats 4 and 7 of Equipment Class 1. One may make a plant-specific case for exceeding caveat limits on

-attached weights and cutouts but there should be some limits even for a low-seismicity site.

Elimination of the caveats from the list makes the engineer systematically verify site-specific conditions and judge whether such conditions are acceptable given the identified generic vulnerability concerns. Therefore, the staff does not consider the justifications provided in Reference 3 to be adequate and please demonstrate how the missing caveats (a potential list is provided below) were satisfied for Crystal River 3.

Class 1 Caveat 4 - Attached weight of 100 pounds or less 7 - Cutouts not iarge 8 - Door /t, rackets secured 9 - Natural frequency relative to 8 Hz limit considered Class 2 Caveat 3 - Side-to-side restraint of breaker 5 - Attached weight of 100 pounds or less 8 - Cutouts not large 9 - Door secured Class 3 Caveat 5 - Attached weight of 100 pounds or less 8 - Cutouts not large 9 - Doors secured Class 4 Caveat 8 - Weak-way bending 10 - Doors secured Class 5 Caveat 4 - Check of long unsupported piping 8 - Relays (if any)

Class 6 - Caveat 3 - Check of long unsupported piping 6 -' Relays

U.S. Nuclear Regulatory Commission Enclosure 2 3F1297-24 Page 14 of 26 Chss 7 Caveat 2 - Valve body not of cast iron 3 - Valve yoke not of cast iron for piston-operated valves and spring-operated pressure relief valves 4 - Mounted on one-inch diameter pipe line or greater 5 - Valve operator cantilever length for air-operated diaphragm valves, spring-e erated pressure relief valves, and light-weight piston-operated valves 6 - Valve operator cantilever length for substantial piston-operated valves 7 - Actuator and yoke not independently braced Class 8A Caveat 2 - Valve body not of cast iron 3 - Valve ske not of cast iron 4 - Mounted on one-inch diameter pipe line or greater 5 - Valve operator cantilever length for motor-operated valves 6 - Actuator and yoke not independently braced Class 8B Caveat 2 - Valve body not of cast iron 3 - Valve yoke not of cast iron 4 - Valve operator cantilever length 5 - Actuator and yoke not independently braced Class 9 Caveat 4 - No possibility of excessive duct distortion causing binding or misalignment of fan Class 10 Caveat 3 - Doors secured 4 - No possibility of excessive duct distortion causing binding or misalignment of internal fan 8 - Relays Class 11 Caveat 2 - No reliance on weak-way bending of steel plate or structural steel shapes 5 - Relays Class 12 Caveat 5 - Relays Class 13 Caveat 6 - Relays Class 14 Caveat 2 - Contains only circuit breakers and switches 3 - Doors secured Class 16 Caveat 4 - No reliance on weak-way bending of steel plate or structural steel shapes 6 - Doors secured

U.S. Nuclear Regulatory Commission Enclosure 2 3F1297-24 Page 15 of' 6 Class 17 Caveat 6 - Relays Class 18 Caveat 2 - Evaluate computers and programmable controllers separately 5 - Natural frequency relative to 8 Hz limit considered Class 20 Caveat 2 - Eialuate computers and programmable controllers separately 3 - Evalaate strip chart recorders separately 7 - Doors secured FPC SUPPLEMENTAL RESPONSE:

FPC states that all GIP caveats weie considered by the walkdown team. This is demonstrated by the term having identified outliers based on prescreened GIP caveats that were not included with the PSP SEWS. To further justify this position, FPC initiated a confirmatory walb Jown.

This walkdown was done by a SRT that included members from the original walkdown team and an independent team member. Mr. Walter Djordjevic of Stevenson & Associates was contracted as an independent review member to walk down a sample list of equipment. The walkdown consisted of a sample (approximately 10%) of equipment on the SSEL. The sample was based on a selection of at least one sample from all the twenty (20) equipment classes, except the tanks. The tanks were excluded from the sampling as the GIP and PSP SEWS are the same. This sample walkdown was done using the GIP SEWS that contain all the GIP caveats. This effort was designed to be independent of any existing SEWS based on the PSP.

After the walkdown, the SRT compared the conclusions of the FPC PSP SEWS and the GIP SEWS. The specific results of this review are shown in Table 1 of the report from Stevenson and Associates. This letter is included in At:achment A of this submittal. FPC has shown that

'he differences between the PSP and GIP caveats are acceptable and & not adversely affect i any of the conclusion of our program, i

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1 U.S. Nuclear Regulatory Commission Enclosure 2 31 1297-24 Page 16 cf 26 NRC RAI REQUEST NUMBER 12 in Reference 5, the Staff identified the need for adherence to the GlP for anchorage evaluation which is a critical item in equipment seismic adequacy verifkation. Based on information provided in Section 5.1.3 on Page 36, it is not clear whether anchcrage verification was adequately performed. Statements such as "where practical, anchor bolts were tightness tested by hand to assure that they did not freely spin in place" do not provide an assurance of

" wrench tightness" discussed in the GIP and endorsed by the PSP (Reference 2). Please provide documentation to demonstrate that equipment anchorage was evaluated per Section 4.4, Appendix C and GIP's equipment specific anchorage caveats.

FPC SUPPLEMENTAL RESPONSE:

FPC acknowledges there are three important attributes to verify adequate anchorage:

verifying actual anchorage is installed, verifying proper installation of the anchorage, and verifying there is adequate capacity of the anchorage.

Verifying Anchorage Installation Exists:

The primary action to verify the existence of anchorage has been satisfied. FPC performed a 100% inspection of accessible anchorage and hands-on check of all non-energized equipment and energized equipment with external anchorage. There is a specific caveat on the PSP SEWS to verify the existence of anchorage. This has been documented on all applicable SEWS (the exception is line mounted equipment). Where the anchorage was missing, or considered poor, then that equipment was declared an outlier. Cut of the one-hundred-thirteen (113) outliers, approximately twenty-seven (27) are because of missing or poor anchorage. At the conclusion of the outlier resolution effort, there will be no unanchoied SSEL equipmem. The PSP clso required the Seismic Review Team (SRT) to verify anchorage load path, in addition to the USI A-46 effort, there has been many inspections and other programs at FPC that help verify ths existence, or other concerns wah anchorage. The Structural Maintenance Rule Inspections made hands on inspection, of virtually every accessible anchorage in the plant. The System Readiness Review Phn performed walkdowns of plant equipment.

Verifying Proper Anchorage Installations:

The second area regarding proper installation has also been satisfied. All anchor bolts were hand checked and visually inspected where accessible during the A-46 walkdowns. Again, any concerns noted were required to be documented on the SEWS.

In addition to the A-46 walkdowns, the Structural Maintenance Rule Inspection performed a visual inspection of anchorage in the plant (including non-SSEL and SSEL

U.S. Nuclear Regul: tory Commission Enclosure 2 3F1297-24 Page 17 of 26 equipment). The Maintenance Rule walkdowns also included a wrench tightness test that _ was performed for one-hundred-ninety three (193) expansion anchors using methods based on the GIP. The one-hundred-ninety three (193) expansion anchors tested were a subset of the three-hundred-fifty four (354) expansion bolts included in the scope of the Maintenance Rule. A foilow up analysis concluded the loose bolts did not adversely affect the adequacy of the equipment. This information is contained in the Structural Maintenance Rule inspection Report.

Verifying Anchorage Capacity:

The third area deals with the capacity of the anchorage. The program at FPC relied on the judgment and experience of well trained Seismic Capability Engineers to make the determination if the capacit;. of the anchorage is adequate. FPC understands the Staff's concern about making an engineering judgment about the adequacy of an anchcrage capacity. To address this issue, FPC has perfo. acd a bounding anchorage calculation based on the requirements of the GIP.

This bounding calculation took a subset of the eiectrical equipment contained in the SSEl, and performs a bound!ng calculation. The intent of this bounding calculation is to envelope the equipment at CR-3 and to ensure equipment anchorage adequacy.

These calculations are done to the requirements of the GIP. FPC maintains that mechanical equipment is inherently more rugged and generically better anchored.

Therefore, no further review of mechaaical equipment is warranted. This study included the following equipment classes:

Equipment Class 1 Motor Control Centers Equipment Class 2 Low Voltage Switchgear Equipment Class 3 Medium Voltage Swi tchgear Equipment Class 4 Transformers Equipment Class 14 Distribution Panels Equipment Class 16 Battery Chargers and Inverters Equipment Class 20 Instrument and Control panels and cabinets A summary of the results of the calculation is included in Attachment A of this submittal. Backup calculations to this study are also includea in Attachment A (FPC Calculation S97-0541),

i The NRC has noted several items of equipment that they felt needed further review.

These items are "noted" in Attachment A. The specific items of equipment and the l findings of the FPC review are documented in Attachment A of this enclosure. Other l than the item discussed next, all equipment anchorage were found to be acceptable.

In addition, the NRC has identified an !ssue with an electrical enclosure that was anchored throegh the base sheet m;tal. Specifically, this is the SCR CaNnet for l

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U.S. Nuclear Regulatory Commission Enclosure 2 3F1297-24 Page 18 of 26 AllllE-4A and AilllE-411 (tag number AllCP-4). To address this pecific concern, FPC has generated a Precursor Card to document the new anchorage outlier. FPC also has generated an "extert-of-con <lition" review to verify there are no other cabinets with this concern. The " extent-of condition" review found ne additional items (see Attachment A of this enclosure for more information).

As part of our confirmatory self assessment, FPC will also perform additional anchorage calculations for approximately one hundred (100) electrical components.

This is approximately 50% of the total scope of SSEL clectrical components. Based on the results of the inspection, FPC may expand the r, cope of the assessment utilizing statistical methods.

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U.S. Nuclear Regulatory Commission Enclosure 2 3F1297 24 Page 19 of 26 NRC RAI REQUEST NUMilER 14b The following requests pertain to the equipment outlier list provided in Table 5-4 of the report:

b) For unanchored cabinets (e.g., SEQ 652 659), the resolution plan was to address

" overturning / sliding potential." The potential for rattling is not necessarily climinated Sy addiessing the " overturning / sliding potential." Please provide information to demonst, ate how the equipment performance will be assured without climinating the potential for cabinet rattling.

FPC SUPPLEMENTAL RESPONSE:

FPC agrees that equipment performance cannot be assured without the climination of cabinet rattling. FPC has declared several pieces of SSEL equipment as outilers because of this issue.

This information is documented on the SEWS. Out of the one hundred thirteen outliers, epproximately twenty seven are due to anchorage.

The outliers have been classified as either restart outliers, or post restart outliers. FPC will have completed the resolution of all restart outliers prior to restart from the current outage.

The remaining outliers dealing with this issue will be resolved prior to restart from Refuel 11.

In the interim, FPC will perform an operability evaluation to confirm that CR-3 can be safely shutdown following a seismic event using only seismically verified equipment to show the post-restart outliers are acceptable in the interim (Enclosure 5 3F1297-24-03),

in addition, upon completion of ti,e outlier resolution, there will be no adjacent cabinets that are not bolted, or otherwise attact'ed together, and there will be no unanchored SSEL equipment.

U.S. Nuclect Regulatory Commission Enclosure 2 3F1297 24 Page 20 of 26 NRC RAI REQUEST NUMBER 14c The following requests pertain to the equipment outlier list provided in Table S-4 of the report:

c) No resolution plan was provided for poor rack construction (SEQ 198 and 202). Please describe how this issue was resolved to assure equipment functionality.

FPC SUPPL.EMENTAL RESPONSE:

The above two Sequence Numbers (198 and 202) refer to the two safety related station batteries DPBA 1 A and DPBA 1B, respectively. These are the battery racks in the Control Complex. These racks have been declaied outlicts due to their poor construction. To address this, FPC has initiated a modification (MAR 97-0810 01) to modify the racks.

The design work for this modification is complete. The Ocid modification is scheduled for completion by end of the current outage. Once the modification is complete, the outliers associated with these two tags will be resolved.

U.S. Nuclear Regulatory Commission Enclosure 2 t 3F1297 24 Page 21 of 26 ,

NRC RAI REQUEST NUMBER 16  !

Regarding cable and conduit raceways, the staff had previously rejected your reasons for not  ;

adhering to the GIP on the basis that they are qualitative (References 1 and 5). Therefore, the staff is requesting additional information that (1) identifies the cable and conduit raceways examined by the seismic capability engineers (SCEs) during its plant specific walkdown, and (2) summarizes the results of the assessment and the basis for the conclusions reached by the -

SCEs in verifying cable and conduit raceway seismic adequacy.

. The requested information should also detail the criteria and methodology mentioned in the letter from P. Beard (FPC) to NRC Document Control Desk (on Generic Letter 97 02), dated August 27,1993. ,

a The need for the walkdown review of the cable and conduit raceway systems is evidenced by the identification of potential weak links by the Third Party Review. For example, the beam ,

clamps identified in the Third Party Review are the types of plant specific details that need to be verified. This reinforces the need for an A 46 review of the scismic adequacy of the cable +

- and conduit systems by the SCEs. For the beam clamps, please provide documentation (loading, capacity, etc.) to demonstrate that they pass the GIP criteria for supports.

FPC SUPPLEMENTAL RESPONSE:

The evaluation of Cable and Conduit raceways was discussed during the NRC Audit. The walkdowns by the Staff indicated agreement that the CR-3 raceways are a rugged design, in -

addition, sample design basis calculations were provided to the NRC. FPC has received feedback from the NRC that indicates there are no open issues with FPC raceways.

Ilowever, FPC can provide additional justification. FPC management voluntarily initiated a structural extent of-condition stud,, based on the IPEEE scismic margins program, to verify adequacy of plant structures. This seismic margin type study included an expansion of scope to address electrical raceways. This involved a more detailed review of electrical raceways than is typically performed in a seismic margins program. FPC has performed two Limited

. Analytical Reviews (LAR) for raceways using the GlP method. One LAR on a cable tray was performed to address the Thermo42g issue. -These calculations showed positive results and further evaluated the acceptability of FPC's raceways. These LARs are included as part of FPC Calculation S97 0542. This calculation is included as Attachment D of this submittal.

Oversight of this effort was performed by Dr. Robert Kennedy. Dr. Kennedy concluded that CR-3 electrical raceways should not require further ' review, especially if some LARs (mentioned above) are oerformed as a part of a seismic margin study.

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U.S. Nuclear Regulatory Commission Enclosure 2 3F1297 24 Page 22 of 26 1

l The question above also mentions the Staff's concern on beam elamps, ibis item was I reviewed by the Staff during the Audit. FPC understands this issue was found to be  !

acceptable with no further information required. {

FPC has confirmed to the Staff during the Audit and associated walkdowns that CR-3 i raceways are of rugged construction. This was verified by the various walkdowns and observations.

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L X 3 Ne! ear Regulatory Commission Enclosure 2 3FI2V, ' t Page 23 of 26 l ,

, N'RC RAI REQUEST NUMBER 18g The following requests pertain to the screening verification data sheets included in Appendix C to the report:

g) For cabinets RCPM 3A and 3B, it was stated on Page 43 in Appendix C that their anchorage and interaction verifications are not applicable, it is understood that the inspection of these cabinets have been deferred (see page 55 of the report, Table 5-5 SEQ Nos. 533 and $34) and it is expected that the anchorage and interaction verifications will be donc at a future outage. Therefore, please justify why the table in Appendix C shows that the anchorage and interaction verifications of these Class 20 equipment iteins are not applicable even though the PSP requires such verifications.

FPC SUPPI.EMENTAI, RESPONSB a

When the walkdown report was issued in December 1995 FPC had not completed the walkdowns and evaluations of these thirty-five (35) inaccessible components identified in ,

Table 5 5 of Appendix C (Reference 1). Since then, FPC has completed its walkdowns including RCPM-3A and RCPM 3B, The results of the review inificated that the above components have acceptable anchorage and interaction reviews.

A copy of the PSP SEWS for these two components is included in Attachment E.

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U.S. Nuclear Reguttory Commission Enclosure 3 3F1297-24 Page 24 of 26 NRC RAI REQUEST NUMBER 20b The following requests pertain to Section 2.2, in-Structure Response Spectrum (Page 7):

it is not clear how the FRS presented in the seismic evaluation report were developed. Please provide a discussion which includes deviations, if any, from the staff safety evaluation on the subject, dated December 16, 1993. Please provide detailed information of the spectra including damping values, the input ground motion used and the structural model as well as the Hnal results that are used for the plant, in particular, please provide a detailed description of the development of the FRS for the interior of the Reactor Building at the 160-foot elevation which is shown in the Figure 2 3, page 11.

FPC SUPPLEMENTAL RESPONSE:

FPC's understanding is that our previous response concerning floor response spectra (Reference 5) is acceptable.

ilowever, the previous response to this RAI and other submittal on this subject, failed to document an exception to the llousner spectra. The information that follows was shared with the Ftaff during the Audit.

FPC's response to Generic Letter 87-02, Supplement 1, provided the requested information for the bor response spectra (F".S). The required information was sent to the NRC via two submittals. The first submittal was FPC letter 3F0493-09, dated April 16, 1993. The second submittal was FPC letter 3F1093-04, dated October 6,1993. These two letters contained floor response spectra for the Auxiliary Building, Control Building, and Reactor Building.

This spectra was based on the site specific ground response spectra.

During a recent review, FPC discovered an oversight on the part of FPC Engineering to not verify the licensing basis for EFT-2. In resolving an A-46 outlier associated with Emergency Feedwater Tank (EFT-2) and performing a review of the A-46 documentation in preparation for the NRC audit, an omission was found in FPC submittal to the NRC. In various submittal to the NRC concerning USl A-46, FPC stated that the site specific ground response spectra was to be used for the CR-3 A-46 program, it was later discovered that the emergency feedwater tank building was analyzed to Regulatory Guide 1.60, " Design Response Spectra for Seismic Design of Nuclear Power Plants," ground response spectra. This is the current CR 3 FSAR commitment for this building. This tank was built later in the life of CR-3 and used the Regulatory Guide 1.60 design response spectra. This spectra is more conservative than the FPC site specific ground response spectra.

A review of the CR 3 FSAR shows that only the Emergency Feedwater Tank Building (and building contents - including the tank) is designed to Regulatory Guide 1.60. All other structures and components in the plant are designed to the site specific ground response

U.S. Nuclear Regulatory Commission Enclosure 2 3F1297 24 page 25 of 26 spectra. The Emergency Feedwater Tank (EIT-2) is the only item of equipment in the SSEL .

that was designed to Reg. Guide 1.60. Therefore, FPC has concluded that there is no extent of condition concern with this item.

This exception does not invalidate any work done to date for the CR 3 A-46 program. The scismic adequacy of the tank has bcen verified using Reg. Guide 1.60 design response spectra.

Attachment B of this submittal includes the latest calculation for seismic adequacy of the tank.

U.S. Nuclear Regulatory Commission Enclosure 2 3F1297 24 Page 26 of 26 ,

NRC RAI REQUE T NUMBER 23 The report states that no significant or programmatic deviations from the PSP were made (Page 64). Please provide a clear explanation of what "no significant deviation" means.

Please itemite those evaluations / methodologies in PSP which you did not follow or from which you deviated. You should discuss what the deviations are and why they are justined. A  ;

definition including the use of examples as to what is considered significant should be provided.

- FPC SUPPI EMENTAI, RESPONSEt The original response to this RAI stated that minor programmatic deviations were discovered in that a few anchorage calculations were not performed for tanks and heat exchangers, although their anchorage had been judged adequate. FPC committed to perform these anchorage calculations for tanks and heat exchangers to confirm anchorage adequacy. The calculation for these tanks and heat exchangers is complete with the exception of CDT-1. The calculations are included as Revision I to Calculation S96-0013. This calculation is included in Attachment C of this submittal.

The exception to this comment is the Condensate Storage Tank (CDT-1). This tank has recently been identined as an outlier. This issue was discussed with the NRC during the audit.

CDT-1 is included in the group of post restart outliers. For an operability assessment, FPC has performed a seismic margins calculation. This calculation is contained in Attachment D of this submittal.

}

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l ENCLOSURE 3 1

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 RESTART OUTLIER RESOLUTION SCHEDULE AND STATUS l

U.S. Nuclear Regulatory Commission Enclosure 3 3F1297 24 CRYSTAL RIVER UNIT 3 USI A-46 RESTART OUTLIER RESOLUTION SCllEDULE AND STATUS UNRESOINED SAFETY ISSUE (USI) A-46, GENERIC LETTER 87-02 By letter dated August 1,1997 (3F0897-01), FPC committed to resolve seventy (70) outliers which were part of our level I and level 11 systems (safety related), as dermed by the System Readiness Review Plan, prior to startup from the current outage.

This enclosure provides the restart outlier resolution schedule and status. This schedule indicates that, to date, sixty (60) restart outliers are resolved, leaving a balance of ten (10) to be resolved prior to restart.

L l

l l

U.S. Nuclear Regulatory Commission Restart Outlier Resolution Status Enclosure 3 3F1297-24 Page 1 of 4 Tag Number Description Disposition Status Disposition Document Engineering - 100%, Field Work AHF-17A CONTROL COMPLEX NORMAL SUPPLY FAN A PEERE 1512 (NUO345652)

Pending Engineering - 100%, Field work AHF-17B COMTROL COMPLEX NORMAL SUFFLY FAN B PEERE 1512 (NUO345652)

Pending "9 "" "9 ~ ' ' ~ ~

DPBA-1A 250/125V BATTERY A Pending Calc. 597-0330 "9 "" "9 ~ * * - ~ ~

D84A-1B 250/125V BATTERY B Pending Calc, 597-0330 Engineering - 100%, Field Work MAR 97-08-10 ,,1 (NUO349261) and MTSW-2C 4160V ES 3A (NORTH) Pending Calc. 597-0330 Engineering - 100%, Field work MTSW-2F 4160V ES 3B (SOUTH) NUO344313 (inspect) NUO349156 (Mod)

Pendi N VBX5-1B ngineering - 100%, Field Work VITAL BUS TRANSFER SWITCH B NUO344786 (to be scheduled)

Pending VBX5-ID VITAL BUS TRANSFER SWITCH D "9 ** "9 ~ '

• Pending NUO344788 (to be scheduled)

VBXS-3B EFIC VITAL BUS TMNSFER SWITCH B "9 "" "9 ~ ' '

Pending NUO344789 (to be scheduled) i VBXS-3D EFIC VITAL BUS TRANSFER SWITCH D 9 "" 9 *

• p NUO344790 (to be scheduled) t ACDP-68-T "#C " """

ES DISTRIBUTION PANEL 3A8 TRANSFORMER _ g fMii I

• g AH-196-PO51 AHD-1 CONTROL -

7 (Mi E Wi MAR 97-08-05-02 removes equip.

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AH-196-POS2 AHD-2 CONTROL , y, [M[ $. a MAR 97-08-05-02 removes equip.

AH-196-PO53 AHD-3 CONTROL ~ UMY '

TT NU 0349305 AH-967-SV AHD-1 & AHD-ID CONTROL - **

Y ~ [M[ -

Engineering Evaluation AHD-01D CONTROL COMPLEX MAKE-UP AIR * - . . .?M? , ... Engineering Evaluation ey Y:':.

[ Mpg !'

J:

ATCP-1 ANTICIPATED TRANSIENT WITHOUT SCRAM LOGIC CABINET s Ng TT NUO344792 (completed)

CAP-1B BORIC ACID PUMP B ~ %; _ [M[ ' - TT NUO344793 (ccepleted)

CEILING CONfROL ROOM CEILING #_ ,

v.-[M[ [ MAR 97-03-04-01 DCP-1B DECAY HEAT CLOSED CYCLE COOLING PUMP B [Mj -

PC 97-0048 RESTART.XLS Page 1 12/16/97

i U.S. Nuclear Regulatory Corrmssion Restart Outlier Resolution Status Enclosure 3 3F1297-24 Page 2 of 4 Tag hundr-r Description Disposition Status Disposition Docisment

~ ~ '

DFT-3A DIESEL GENERATOR FUEL OIL DAY TANK A I  : M y ac

. 597 330

'~

DFT-35 DIESEL GENERATOR FUEL OIL DAY TA4K B } f \ [ ,

C 59 33 DHT-1 SORATED WATER STORAGE TANK ,

s.y Revised Calc. 594-0011

- DRRD-2-1 CRD DC BREAKER CABINET UNIT 1 & 2 . ^

. 2. 7 ggn-a 6 OnsMW, W mM ng Evaluation to close outlier 6 Onsped. W med ng DRRD-2-2 CRD DC BREAKER CABINET UNIT 3 & 4 3$-;l_=m.MMm< f

_ m '-

Evaluation to close outlier 1 4_ N 344316 O mpect). E@ mering OkRD-2-3 CRD DC BREAKER CABINET TRIP RE5rT _

= C M~= bn ' p g Evaluation to close outlier EFT-2 EMERGENCY FEEDWATER TANK .- e, .

N(, zj Calculation 597-0316 p

EMERGENCY DIESEL GEN A ELECTRICAL EQUIP *ENT CABINET g~~_ ,

g' 9# - " -

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.; MAR 97-08-10-01 (NUO349322) and Calc. 597-0330 ECCP-2B CABINET ZX

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MAR 97-08-10-01 (4U0349519) and Calc. 597-0330 ESCP-4A ENGINEERED SAFEGUARDS ACTUATION RELAY CABINET 4A

N% ' '

Verbal agreement with Ops. ,

ESCP-4B ENGINEERED SAFEGUARDS ACTUATION RELAY CABINET 4B , [Mll~ ~

Verbal agreement with ons.

M

~

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ESCP-4D ENGINEERED SAFEGUARDS ACTUATION RELAY CABINET 4D -

3M5 ^ Verbal agreement with Ops.

a ^, ,,- -

ESCP-5A ENGINEERED SAFEGUARDS ACTUATION RELAY CABINET SA _

7 -5MWgx , NU 0342044 (completed)

[N IAP-1A INSTRUMENT AIR COMPRESSOR A g 9

? Engineering Evaluation (T 27n :L.

g - . - .

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RESTAItT.XtS Page 2 12M 637 i

U.S. Nuclear Regulatory Commrssion Restart Outlier Resolution Status Enclosure 3 3F1297-24 Page 3 of 4 Tag Mun6er Description Disposition Status Disposition Document s%R 97-08-10-01 (NUO349328) and

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MUV-051 LET-DOWN FLOW CONTROL VALVE s 7_~ M .

Further Engineering Evaluation  !

u- s .. - - m :.e MAR 97-08-10-01 (NUO349520) and h5 . M.

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l MUV-200 LETDOWN ISOLATION VALVE TO DEMINERALIZER MULM-1A +As

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'b fMhh N NUO344312 NI&P-D2 NI&P SYSTEM SUBA55EMBLY D CABINET 2 - sM_. _

Verbal agreement with Ops.

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l NI-1-A3 PROPORTIONAL COUNTER ASSE"BLY

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RSA-1 REMOTE SHUTDOWN RELAY CABINET A-1 i ' ' ~

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U.S. Nuclear Regulatory Commission Restart Outlier Resolutxm Status Enckwe 3 3F1297-24 page 4 of 4 Tag Number Description Dispositie Document D. fosition Status h

SWP-1A EMERGENCY NUCLEAR SERVICE SEA WATER PIMP 3A  :  : NUO344310 (inspect - cogleted) 3M 64 ~P NUO344796 (inspect) & NUO149157 SWP-18 EMERGENCY NUCLEAR SERVICE CCC PUMP 38 m me:pS .- N

, m ., e e n ~ r ~@ . 4j 0 (co mleted ned.)

SWP-1C NORVAL NUCLEAR SERVICE CLOSED CYCLE COOLING PUwP 2 m e -w+- h N

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pg;pfgg g Ngo344787 (completed) en._d pp %1. *L Closed based on inspection of WDT-1A WASTE CAS DECAY TANK 1A hyT._M_ Cr

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. . .J. M i Mmwp +l Closed based on inspection of obotoneachs RESTART.XLS Page 4 12/16S7

1

)

i ENCLOSURE 4 FLORIDA POWER CORPORATION  !

, CRYSTAL RIVER UNIT 3

DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 i

POST-RESTART OUTLIER RESOLUTION SCHEDULE AND STATUS 1

U.S. Nuclear Regulatory Commission Enclosure 4  !

3F1297 24 1

CRYSTAL RIVER UNIT 3 ,

USl A-46 POST RESTART OUT1,lER RESOLUTION SCllEDULE AND STATUS I

UNRESOINED SAFETY ISSUE (USI) A 46, GENERIC LEITER 87 02

!!y letter dated August 1,1997 (3F0897 01), FPC committed to develop a work-off curve for completion of post restart outliers prior to restart from the current extended outage. Forty-three (43) outliers in level 11 and level 111 systems (non safety related) are being resolved prior to startup from Refueling Outage 11.

This enclosure provides the post restart outlier resolution schedule and status. To date, seventeen (17) of forty-three (43) post restart outliers are resolved, leaving a balance of twenty six (26) to be resolved. This schedule indicates resolution of the last outlier to be by the end of January 1999.

These post restart ouillers are receiving an operability review prior to restart from the current outage to confirm that safe shutdown is still achievable using oth:r safety related equipment.

U.S. Nuclear Regulatory Commission Restart Outlier Resolu+.icn Status Enclosure 4 3F1297-24 Page 1 cf 3 5 5 "

Tag Number Description 3

Disposition Document AHCP-4 SCR CABINET FOR AHHE-4A/4B 6/15/98 0% CCaR to be developed DPBA-1C 250/125V BATTERY C 8/15/98 0% CCaR to be developed ER1 EVENTS RECORDER CABINET 1 9/15/98 0% CCaR to be developed ER2 EVENTS RECORDER CABINET 2 9/15/98 f 0% CCaR to be developed ER3 EVENTS RECORDER CABINET 3 9/15/98 0% CCaR to be developed ER4 EVENTS RECORDER CABINET 4 9/15/98 0% CCaR to be developed ER5 EVENTS RECORDER CABINET 5 9/15/98 0% CCaR to be developed ER6 EVENTS RECORDER CABINET 6 9/15/98 0% CGR to be developed ER7 EVENTS RECORDER CABINET 7 9/15/98 0% CCaR to be developed ERG EVENTS RECORDER CABINET 8 9/15/98 0% CCaR to be developed ICS-5 A OL W 8 GB N 10/1/98 0% CCaR to be developed 5

g_g ADV BACKUP NITROGEN SUPPLY TANKS 10/1/98 0% CCaR to be developed (10)

NNI-5 AUXILIARY CONTROL SYSTEM CABINET 5 10/15/98 0% CCaR to be developed NNI-6 AUXILIARY CONTROL SYSTEM CABINET 6 10/15/98 0% CCaR to be developed MTMC-09 480V PRESSURIZER HEATER MCC 38 11/15/98 0% CCWR to be developed MTSW-3D 480V REACTOR AUXILIARY BUS B 11/15/98 0% CCWR to be developed TRANSMITTER POWER SUPPLY CABINETS TPC A&B 11/15/98 0% CCaR to be developed MTMC-12 480V TURBINE MCC 3A 12/15/98 0% CCaR to be developed PORESTAR.XLS Page 1 12/16197

U.S. Nuclear Regulatory Commission Restart Outfier Resolution Status Enclosure 4 3F1297-24 Page 2 of 3 >

s D on Tao Nu4er Description Disposition Document MTSW-3A 480V TUP.BINE AUXILIARY BUS A 12/15/98 0% CCnR to be developed MTSW-3C 480V REACTOR AUXILIARY BUS A 12/15/98 0% CCaR to be developed MTSW-33 480V PLANT AUXILIARY BUS 12/15/98 0% CGWR to be developed MTSW-33-T 4160/48W MW AMIAW BUS CCaR to .se developed TRANSFORER 1/15/99 0%

PORV & TEMPERATURE SATURATION PORV/TEMF 1/15/99 0% CGWR to be developed CABINET 0W RFL MPLXR 1/15/99 0% CGWR to be developed CDT-1 CONDENSATE STORACE TANK n/a 0% Calculation to be de< eloped SPENT C NT FLOW SF-9-FIT n/a 0% Engineering Evaluation OL @N WE AIR a h rdnspecdon of plant M

~

AHP-01A n/a COMPRESSOR A drawings i

1. ^ a by m hspection of plant AHP-018 COMPRESSOR B n/a M drawinos  !

L CmW WAC AIR AHP-01C COMPRESSOR C n/a i [ ~ N .. .

~

=

Qualified by reinspection of plant drawings

^ ^ Qualified by reinspection of plant AtlP-01D COMPRESSOR D n/a Mm -~-

4 drawings CDHE-4A MAIN CONDENSER A n/a %M ] Q Calculation 596-0013. Rev. 1 CDHE-4B MAIN CONDENSER B n/a O M Calculation 596-0013 Rev. 1 Qolified by reinspection of plant DPBC-1G BATTERY CHARGER G n/a -

- M ~

dravrings and Engineering Evaluation

'}uali*ied by reinspection of plant CDBC-1H BATTERY CHARGER H n/a c.

+ M I. ;is- .

drawings and Engineering Evaluation DPBC-II BATTERY CHARGER I n/a , ;M

• D " "5E*C

" PI#".

drawings and Engineering Evaluation s DPDP-1C 250/125V DC MAIN PANEL 3C n/a ^1 9M y a by reinspection of plant dras<ings and Engineering Evaluation i

PORESTAR.XLS Page 2 12/16!97

4 U.S. Nuclear RegulMory Commission Restart Ou*?er Resolution Status Ent.losure 4 3F1297-24 Page 3 of 3 Tag Murdper thsposition Det

~

Description l

DPDS-1C BATTERY 3C DISCONNECT SWITCH n/a --~

I

~"."*"C.

drawings and Engine <.fino Evaluation

" E *".

DPXS AC DPBC-1 INPUT POWER TRANSFER SWITCH n/a bfMI-:'y ' F' W" "*"C " E ""

_=~ --

dr d nas and Engincer,.no Evaluation R'#-1 NORMAL NUCLEAR SERVICES SEA WATER ,, j, _ _ _ _

"j Calculation 570-0001, Rev. O PUMP MOTOR COOLER WDT-3A RC BLEED TANK 3A n/a Calculation S96-0013, Rev. 1 WDT-38 RC BLEED TANK 3B n/a Calcu?ation 596-0013 Rev. 1 WDT-3C PC BLEED TANK 3C n/a Calculation 596-0013. Rev. 1 WDT- 5 REACTOR COOLANT DRAIN TANK n/a Calculation S96-0013 Rev.1 t

l i

f a

i PORESTAR.XLS Paga 3 12/16/97

l l

ENCLOSURE 5 FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 LIST OF REGULATORY COMMITMENTS

- - - - - - - - _ , , w,------, , .~ --- ,_n ,. . . - ,, , , , --,

U.S. Nuclear Regulatory Commission Enclosure 5 3F1297 24 Page1ofI List of Regtilatory Commitments The following table identifies those actions committed to by Florida Power Corporation in this document. Any other actions discussed in the submittal represent intended or planned actions by Florida Power Corporation. 'lhey are described for the NRC's information and are not regulatory commitments. Please notify the Manager, Nuclear Licensing, of any questions regarding this document or any associated regulatory commitments.

ID NUMilER COMMITMENT DATE DUE 3F1297 2441 As pah of closure of our program, a connnnatory self- Prior to startup assessment will be conducted. Fit will inform the NRC about from Refueling the results of the audit conecrning all open issues regarding USI Outage 11 A-46 for CR-3. Review to include: audit of SSEL, resolution of all outliers, and review of anchorage calculation.

3F1297 24-02 FPC has improved upon the schedule and will be completing Prior to stanup (Ref. 3F1297 33 2) resolution of all outliers prior to startup from Refueling from Refueling Outage 11. FPC will provide conHrmation of mmpletion of Outage 1i resolution of all outliers.

3F1297-24-03 The remaining twenty-six (26) post restart outliers will Prior to restart receive an operability assessment prior to restart from the from current current outage to confirm that the failure of any of the outage outstanding non safety related outliers will not affect the ability to achieve safe shutdown of the plant. The assessment will also include a review of any seismic interaction a mcerns.

3F1297-2404 Revise Abnormal Procedure AP-%1, Eanhquake, to include fly Mode 2 identincation of the had actor relays and guidance for operators to a)pe with relay chatter subsequent to a seismic event.

3F1297 2445 FPC will perfonn GIP anchorage calculations for approximately Prior to stanup 50% (100) of electrical components currently identi6ed on the from Refueling SSEL. FPC may expand the scope of the almlations utilizing Outage 11 statistical methods.

3F1297 2446 FPC will revise the PSP to incorporate mmments on relay Prior to startup review. from Refueling Outage 11 3F1297-2447 At the mnclusion of the outlier resolution etfort, there will be no Prior to startup unanchored SSEL equipment, from Refueling Ouage 11 3F1297-2448 FPC will have mmpleted the resolution of all restart outliers Ily Mode 2-prior to restart from the current outage.

3F1297-2449 At the mmpletion of outlier resolution, there will be no SSEL Prior to startup related aQacent cabinets that are t)ot bolted or otherwise attached from Refueling together. Outage 11 3F1297 24-10 Submit AP-961 to NRC for information upon cumpletion of January 31, revision (upon mmpletion of Commitment 3F1297-2444) 1998 j