ET 10-0026, License Amendment Request for Deviation from Fire Protection Requirements - Reactor Coolant System Subcooling During Alternative Shutdown

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License Amendment Request for Deviation from Fire Protection Requirements - Reactor Coolant System Subcooling During Alternative Shutdown
ML102720417
Person / Time
Site: Wolf Creek Wolf Creek Nuclear Operating Corporation icon.png
Issue date: 09/22/2010
From: Garrett T
Wolf Creek
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
ET 10-0026
Download: ML102720417 (26)
v • d • e


Text

W@LF CREEK'NUCLEAR OPERATING CORPORATION Terry J. Garrett September 22, 2010 Vice President, Engineering ET 10-0026 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: License Amendment Request (LAR) for Deviation from Fire Protection Requirements - Reactor Coolant System Subcooling During Alternative Shutdown Gentlemen:

Pursuant to 10 CFR 50.90, Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to the Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station (WCGS). This license amendment request is seeking approval by the Commission, pursuant to License Condition 2.C.(5), to make changes to the approved fire protection program as described in the Updated Safety Analysis Report (USAR). A deviation from a commitment to certain technical requirements to 10 CFR 50, Appendix R, Section III.L.1, as described in Appendix 9.5E of the WCGS Updated Safety Analysis Report (USAR) is requested. WCNOC is proposing to revise USAR Table 9.5E-1 to include information on Reactor Coolant System process variables not maintained within those predicted for a loss of normal ac power as evaluated in Evaluation SA-08-006 Rev.1, "RETRAN-3D Post-Fire Safe Shutdown (PFSSD) Consequence Evaluation for a Postulated Control Room Fire."

Attachment I provides the evaluation and justification for the proposed license amendment.

Attachment II provides markups of the USAR including Appendix 9.5E. USAR Appendix 9.5E provides a design comparison to 10 CFR 50 Appendix R.

It has been determined that this amendment application does not involve a significant hazard consideration as determined per 10 CFR 50.92. The amendment application was reviewed by the WCNOC Plant Safety Review Committee. In accordance with 10 CFR 50.91, a copy of this application is being provided to the designated Kansas State official.

P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HC/VET ' -'

ET 10-0026 Page 2 of 3 WCNOC requests approval of this proposed amendment by September 29, 2011. Once approved, the amendment will be implemented within 180 days of receipt.

There are no commitments associated with this submittal. If you have any questions concerning this matter, please contact me at (620) 364-4084, or Mr. Richard D. Flannigan at (620) 364-4117.

Terry J. Garrett TJG/rlt Attachments: I Evaluation of Proposed Change II Markup of USAR Pages cc: E. E. Collins (NRC), w/a T. A. Conley (KDHE), w/a G. B. Miller (NRC), w/a B. K. Singal (NRC), w/a Senior Resident Inspector (NRC), w/a

ET 10-0026 Page 3 of 3 STATE OF KANSAS )

SS COUNTY OF COFFEY )

Terry J. Garrett, of lawful age, being first duly sworn upon oath says that he is Vice President Engineering of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

By Terriy).arrett Vice Kesident Engineering SUBSCRIBED and sworn to before me this 4_.214day of oefrnvl be (6 , 2010.

GAYLE Notary SHEPHEARD Public - Statp of Kansas*

Notary F3blic My Appt. Expires 7 0.1,/ I Expiration Date N, /Mpo

Attachment I to ET 10-0026 Page 1 of 20 EVALUATION OF PROPOSED CHANGE

Subject:

License Amendment Request (LAR) for Deviation from Fire Protection Requirements - Reactor Coolant System Subcooling During Alternative Shutdown

1.

SUMMARY

DESCRIPTION

2. DETAILED DESCRIPTION
3. TECHNICAL EVALUATION
4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 4.2 Significant Hazards Consideration 4.3 Conclusions
5. ENVIRONMENTAL CONSIDERATION
6. REFERENCES

Attachment I to ET 10-0026 Page 2 of 20 EVALUATION

1.

SUMMARY

DESCRIPTION This evaluation supports a request to amend Renewed Facility Operating License NPF-42 for the Wolf Creek Generating Station (WCGS). The proposed amendment would revise the Renewed Facility Operating License to deviate from a commitment to certain WCGS Fire Protection Program requirements. A deviation from certain technical requirements to 10 CFR 50, Appendix R, Section II.L.1, as described in Appendix 9.5E of the WCGS Updated Safety Analysis Report (USAR) is requested. WCNOC is proposing to revise USAR Table 9.5E-1 to include information on Reactor Coolant System (RCS) process variables not maintained within those predicted for a loss of normal ac power as evaluated in Evaluation SA-08-006 Rev.1 (Reference 1), "RETRAN-3D Post-Fire Safe Shutdown (PFSSD) Consequence Evaluation for a Postulated Control Room Fire." NRC approval is requested pursuant to License Condition 2.C.(5) as the proposed change would adversely affect the ability to achieve and maintainsafe shutdown in the event of a fire.

2. DETAILED DESCRIPTION Table 9.5E-1 of USAR Appendix 9.5E provides a design comparison to 10 CFR 50, Appendix R and how WCNOC compares with the specific requirements of 10 CFR 50 Appendix R.

WCNOC is proposing to revise Table 9.5E-1 to include the below information related to the comparison to Appendix R,Section III.L.1 requirements:

Analysis demonstrates that the performance goals of III.L.2 are satisfied. The performance criteria of 1II.L.1 are also satisfied, with the exception of maintaining reactor process variables within those predicted for a loss of normal ac power. This is acceptable, as long as a control room fire will not result in the plant reaching an unrestorable condition, which could lead to core damage. The criteria for "not reaching an unrestorable condition" are; natural circulation is maintained and adequate core cooling is maintained (core exit temperature is less than 712°F).

This change is being made based on the results of Evaluation SA-08-006 Rev.1, "RETRAN-3D Post-Fire Safe Shutdown (PFSSD) Consequence Evaluation for a.Postulated Control Room Fire."

Evaluation SA-08-006 Rev.1 shows that during all postulated control room fire transients, the process variables predicted for loss of normal ac power are maintained and the fission product boundary was not affected. However, two of the scenarios where one pressurizer PORV spuriously opens resulted in void formation in the upper core. The void formation was of limited size and duration and would not prevent natural circulation assuming the PORV remains open for 3 minutes. Void formation in the upper core is not predicted during the loss of normal ac power evaluation because loss of normal ac power is a non-LOCA event. Therefore, since void formation is not predicted in the loss of normal ac power event, the process -variables predicted for a control room fire are not maintained within those predicted for a loss of normal ac power.

Attachment I to ET 10-0026 Page 3 of 20

3. TECHNICAL EVALUATION

Background

The WCGS USAR, Appendix 9.5E, provides a comparison of the WCGS Fire Protection Program against the requirements of Section III of Appendix R to 10 CFR 50. Although WCGS obtained its operating license after January 1, 1979, the NRC stated, in NUREG-0881 (Reference 3),

"Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1,"

hereafter referred to as SER, that they will condition the WCGS operating license to require WCGS to meet the technical requirements of Appendix R to 10 CFR 50, or provide equivalent protection. However, the Condition never appeared in the WCGS full power operating license when it was issued on June 4, 1985. Therefore, although Appendix R does not apply, WCGS's commitment to Appendix R is established in Appendix 9.5E as part of the approved Fire Protection Program documented in the USAR.

During an NRC audit of fire protection during the week of July 30, 1984 assumptions applied in the control room fire hazards analysis were questioned. The concern resulted in meetings involving SNUPPS, WCGS, Callaway, and the NRC in August 1984 to resolve the issues.

Meetings were on August 10, 14, 15, and 22. As a result of the meetings, SNUPPS developed letter SLNRC 84-0109 (Reference 5) which documented the SNUPPS "Response Plan for Immediate Evacuation of the Control Room Due to Fire," addressed spurious actuations, and established Phases A through F and their actions to be taken to mitigate the consequences of a fire in the control room.

One of the Phase A (0 to 5 minutes) actions in SLNRC 84-0109 is to remove control power from pressurizer PORVs BBPCV0455A and BBPCV0456A to fail the PORVs closed.

NUREG-0881, Supplement No. 5 (Reference 6) Section 9.5.1.5, paragraph 5 and USAR Appendix 9.5B reference the August 23, 1984 letter (SLNRC 84-0109). The NRC concluded in Reference 6 that "On the basis of its review of the phased procedural approach outlined in the applicant's letter dated August 23, 1984, ... the SNUPPS alternative shutdown capability is acceptable... after procedure changes and modifications are complete." None of the procedure changes or modifications involved revising the response to failed open pressurizer PORVs.

Therefore, the NRC approved, via NUREG-0881, Supplement 5, the 5 minute time to close a failed open PORV. The current time to close a failed open PORV is 3 minutes.

NRC Inspection Report 05000482/2005008 (Reference 7) identified apparent violation 05000482/2005008-02 concerning the failure to ensure that the RCS would not lose subcooling during an alternative shutdown scenario if a fire caused both pressurizer power operated relief valves to spuriously open. Subsequent to this inspection (2005 triennial fire protection inspection), a number of changes were made to procedure OFN RP-017 to improve the timing of actions. A thermal hydraulic analysis was performed in Evaluation SA-08-006 Rev.1 to assess RCS performance under various control room fire scenarios. This evaluation determined that the spurious opening of a pressuizer power operated relief valve (PORV) results in depressurization of the RCS and short term voiding within the reactor vessel until the loss of coolant event is terminated. In response to this evaluation, USAR Table 9.5E-1 response to Section III.L.1 was revised to provide a deviation to the requirement for maintaining RCS process variables within those predicted for a loss of normal ac power.

Attachment I to ET 10-0026 Page 4 of 20 The NRC Triennial Fire Protection Inspection Report (Reference 8) identified an unresolved item for a concern that a change was made to the approved fire protection program that could adversely affect the ability to achieve and maintain safe shutdown in the event of a fire without prior NRC approval. Subsequently this was identified as a Severity Level IV noncited violation in Reference 2. The NRC concluded that the change to USAR Table 9.5E-1 would adversely affect the ability to achieve and maintain safe shutdown in the event of a fire since it allowed WCNOC to violate a requirement (Appendix R,Section III.L.1) without an NRC approved deviation.

Evaluation Evaluation SA-08-006 Rev.1, "RETRAN-3D Post-Fire Safe Shutdown (PFSSD) Consequence Evaluation for a Postulated Control Room Fire," documents the basis of the WCGS RCS thermal hydraulic response for different scenarios potentially caused by a fire in the control room. Sixteen control room fire scenarios were developed to bound the spurious actuations that could potentially result in the worst case mass loss from the primary system and the worst case pressure reduction of the primary system. Mass loss and uncontrolled cool down of the primary system can cause the pressurizer water level to fall off scale. Pressure loss can cause voiding outside the pressurizer which could prevent natural circulation if left unmitigated. The scenarios were developed to model the effects of potential fire-induced spurious operation of equipment due to a fire in the control room. The scenarios assume a single spurious operation occurs in conjunction with and without a loss of off-site power and with and without an auxiliary feedwater actuation signal (AFAS). The AFAS was modeled since it could adversely impact post-fire safe shutdown (PFSSD) due to the potential for uncontrolled cooldown.

The sixteen scenarios are grouped into four categories and are summarized as follows:

Spurious Behavior of PORV Failed Open Scenario 1 (SN1-40) Loss of off-site power, PORV open Scenario 1A (SN1A-40) Loss of off-site power, PORV open, AFW pumps auto start for unplanned cooldown Scenario 2 (SN2-40) No loss of off-site power, PORV open Scenario 2A (SN2A-40) No loss of off-site power, PORV open, AFW pumps auto start for unplanned cooldown Spurious Behavior of SG ARVs Failed Open Scenario 3 (SN3-40) - Loss of off-site power, ARV open Scenario 3A (SN3A-40) - Loss of off-site power, ARV open, AFW pumps auto start for unplanned cooldown Scenario 4 (SN4-40) - No loss of off-site power, ARV open Scenario 4A (SN4A-40) - No loss of off-site power, ARV open, AFW pumps auto start for unplanned cooldown Spurious Uncontrolled Letdown Scenario 5 (SN5-40) - Loss of off-site power, letdown open Scenario 6 (SN6-40) - No loss of off-site power, letdown open Scenario 6A (SN6A-40) - No loss of off-site power, letdown open, no Pressurizer heaters Scenario 6B (SN6B-40) - No loss of off-site power, letdown open, no Pressurizer heaters, Auxiliary Spray on

Attachment I to ET 10-0026 Page 5 of 20 Spurious Behavior of MSIV Failed Open Scenario 7 (SN7-40) - Loss of off-site power, MSIVs open Scenario 7A (SN7A-40) - Loss of off-site power, MSIVs open, SG blowdown open Scenario 8 (SN8-40) - No loss of off-site power, MSIVs open Scenario 8A (SN8A-40) - No loss of off-site power, MSIVs open, SG blowdown open The loss of normal ac power event is a non-LOCA transient, while four of the PFSSD scenarios are small LOCA events due to an open PORV providing water relief for three minutes, with two of the scenarios (scenario 1 and 1A) indicating the presence of an upper core void fraction of a limited duration. The upper core void fractions are approximately 5-6% and are present 10-12 minutes.

10 CFR 50, Appendix R Section III.L.1., states, in part: "During the postfire shutdown, the reactor coolant system process variables shall be maintained within those predicted for a loss of normal ac power, and the fission product boundary integrity shall not be affected; i.e., there shall be no fuel clad damage, rupture of any primary coolant boundary, of rupture of the containment boundary."

USAR Section 15.2.6, "Loss of Non-Emergency AC Power to the Station Auxiliaries (Blackout),"

describes the loss of normal ac power event. Note that calculation AN-09-012 (Reference 9),

"RETRAN-3D Analysis of the Loss of Non-Emergency AC Power for the Feedwater Check Valve Relocation Project (DCP #12792)," is the analysis of record (AOR) but is not currently reflected in USAR Section 15.2.6. Calculation AN-09-012 does not provide bounding thermal hydraulic parameters for a control room fire, as it does not postulate additional spurious equipment actuation concurrent with the loss of power event. A fire in the control room requires the assumption of spurious equipment actuation concurrent with the loss of off-site power.

A loss of normal ac power event is classified as an ANS Condition II event. It may be caused by a complete loss of the offsite transmission network accompanied by a turbine generator trip at the plant, or by a loss of the onsite ac distribution system. The reactor will trip on either 1) turbine trip, or 2) upon reaching one of the trip setpoints in the primary or secondary system as a result of the flow coastdown and decrease in secondary heat removal, or 3) loss of power to the control rod drive mechanisms as a result of the loss of power to the plant. The loss of normal ac power event begins with a loss of normal feedwater flow. The loss of normal ac power event results in a primary system heatup with parameters selected to maximize the pressurizer fill. The single failure assumed is the loss of the turbine driven auxiliary feedwater pump. Note: The loss of normal ac power event is bounded by other events for minimum departure from nuclear boiling ratio (DNBR) concerns.

The basic timeline description for the loss of normal ac power event is as follows:

- Main feedwater flow stops

- Reactor trip on steam generator water level low low

- Control rods begin to insert

- Reactor coolant pumps (RCPs) coast down

- Peak water level in the pressurizer occurs

- Auxiliary feedwater (AFW) pumps start

- AFW lines are purged and the steam generators begin to receive cold AFW

- Core decay heat decreases to the AFW heat removal capacity

Attachment I to ET 10-0026 Page 6 of 20 USAR Figures 15.2-9 and 15.2-10 show the process variables in relation to time for those variables included in the analysis of the loss of normal ac power event. Those process variables include Core Power, Core Mass Flowrate, Steam Generator Pressure, RCS Temperature, RCS Pressure, and Pressurizer Level. These same process variables were utilized in calculation AN-09-012.

Demonstration that Process Variables are within that Predicted for Loss of Normal AC Power Each of the 16 control room fire scenarios included in Evaluation SA-08-006 Rev.1 was modeled using RETRAN-3D. For each scenario, six (6) graphs were developed to allow comparison of the process variables with those predicted for a loss of normal ac power as described above.

A comparison of the process variables determined from Evaluation SA-08-006 Rev.1 with the graphs from the AOR (calculation AN-09-012 (Reference 9)) is presented in the following paragraphs. For the graphs presented below, the sixteen scenarios are compared with the process variables from the AOR for the pressurizer overpressure case and the pressurizer overfill case.

Core Power (MWth)

The conservative scram function and decay heat model used calculation AN-09-012 (Reference 9), "RETRAN-3D Analysis of the Loss of Non-Emergency AC Power for the Feedwater Check Valve Relocation Project (DCP #12792),"are used in the analysis of the PFSSD scenarios. As shown in Figure 1, for all scenarios in Evaluation SA-08-006 Rev.1 and the loss of normal ac power presented in calculation AN-09-012, the calculated end results are essentially identical, with core power reducing to a value of approximately 100 MWth after 280 seconds from scram initiation. This process variable for all control room fire scenarios is entirely bounded by that predicted for the loss of normal ac power.

4500 4000 3500

" SN1.40

" SNIA-40 SN2-40 3000 SN2A40

  • SN3-40
  • SN3A,4(0 2500 *SNS-4fl 2000 I

SW.A-40 SNZ-40 1500 SN7A,40 SNBA-40l 1000 -Safety Analysi 500 0

0 30 60 90 120 150 180 210 240 270 300 Time (seconds)

Figure 1A - Overfill Case of AOR

Attachment I to ET 10-0026 Page 7 of 20 4000 3500

  • SNI-40 3000
  • SNIAADf SN2-40 SN2X40 2500 I SN3-A40
  • SN4-40
  • SN4A-40 2000 I SN5~40 0

1500 SN7.40 SN7A-Afl SNBA-40l 1000 500 0 30 60 90 120 150 180 210 240 270 300 Time (second*)

Figure 1B - Overpressure Case of AOR Figure 1 Core Power vs. Elapsed Time Core Mass Flowrate (Fraction of Nominal)

The conservative RCP coastdown model is used in calculation AN-09-012 (Reference 9) and the analysis of the PFSSD scenarios, with the exception of a minor change in the pump inertia. As shown in the figures for all PFSSD scenarios and the loss of normal ac power figure presented in calculation AN-09-012, the calculated results are similar after RCS pump trip. The mass flowrate is slightly different due to model uncertainties and biases.

Attachment I to ET 10-0026 Page 8 of 20 1.2 1.0 S NI-40 SNIA-40 SN2-48 08 SN2A-40

  • SN3-40
  • SN3A.40 z

16 *SN4-40

  • SN4A-4D 1 06 *SN8.40 SNSB-40J o 0.4 SN7A,40 SN7A.40 SS5A,40 02 0.0 600 1200 1800 2400 3000 3600 Time (seconds)

Figure 2A - Overfill Case of AOR 1,2 S40A SM

  • SNIA-AD SN2-40 SN2A,40 SN3-11J E
  • SN3A,40 z SN1A-40

-SN4AA-4

  • S6-.40 SW-40 S1.A-40 SNE8.40 SU7.40 SN7A,40
  • SNO-40 SN0A4O

-Safety Analysis 0 600 1200 1800 2400 3000 3600 TIme (seconds)

Figure 2B - Overpressure Case of AOR Figure 2 Core Flow Rate (fraction of Nominal) vs. Elapsed Time

Attachment I to ET 10-0026 Page 9 of 20 Steam Generator Pressure (psia)

The steam generator pressure peaks in the loss of normal ac calculation at approximately 1250 psia and the PFSSD scenarios peak at less than 1200 psia. This process variable is bounded.

Results from these calculations demonstrate that the acceptance criteria with respect to overpressure concern are met. That is, the maximum secondary pressure attained remains below 100% of design pressure (1318.5 psia).

1400 1300

  • SN1-40
  • SN1A-40 1200 - S2-40

""*'*J;*,, *SN3A-40

  • 1100 - SN4-40 SW,4 SN4B-40 SN7-4O SIB-A-40 900.- SNB-40

"*'*'*"*-Safety Analysis 600 1200 1800 2400 3000 3600 Time (seconds)

Figure 3A - Overfill Case of AOR

Attachment I to ET 10-0026 Page 10 of 20

  • SN1-4L SNIA-40 SN2A,40 SN3-AIJ
  • SN3A,40 0

SNA440 0

U U

0 a.

SNAA-40 Co SNB-40l SNBA40

-Safety Analysis 0 600 1200 1800 2400 3000 3600 Time (seconds)

Figure 3B - Overpressure Case of AOR Figure 3 Steam Generator Pressure vs. Elapsed Time

Attachment I to ET 10-0026 Page 11 of 20 RCS Coolant Average Temperature (F)

The RCS coolant temperature includes the core exit (hot leg) temperature, core inlet (cold leg) temperature, and vessel average temperature. The vessel average temperatures remain below approximately 595 OF for the loss of normal ac power calculation and below 590 IF for all PFSSD scenarios. This process variable is bounded.

620 610 600 *SNI-40

  • SNIA,40 Stu-40 590 SN2A-AD SN3-40 580 SNIA-lO

-SN4AA-A 570 -SN5-A0 S &40 560 SNGB-40 SN7-40 550 SN7AADf SW-40 540 530 520 600 1200 1800 2400 3000 3600 Time (seconds)

Figure 4A - Overfill Case of AOR 600 590 580 SN1-40

  • SNIA-IS SW2-40 U" 570 SN3-40
  • SN2A,40 SN4-40
  • 560 SN4A-A0 E 0yNA-40 550 SW,440 067-40 SW7A-40 540

-Safety Amalyve 530 520 600 1200 1800 2400 3000 3600 Time (seconds)

Figure 4B - Overpressure Case of AOR Figure 4 RCS Coolant Average Temperature vs. Elapsed Time

Attachment I to ET 10-0026 Page 12 of 20 RCS Pressure (psia)

The pressurizer pressure peaks at approximately 2450 psia (Overfill case) and over 2600 Psia (Overpressure case) for the loss of normal ac power calculation and less than 2400 psia for all PFSSD scenarios. This process variable is bounded. Results from these calculations demonstrate that the acceptance criteria with respect to overpressure concern are met. That is, the maximum RCS pressure attained remains below 110% of design values (2748.5 psia).

Note: The cold leg pressure is slightly higher than pressurizer pressure.

2600 -

2400 2200 SN1-40

  • SN1A-4O SN240 SN2A40 2000 - SN3-ro Ta Si'3A-48
  • ,SN4.40 "a * -o~SN4~A.O 1800 m ~S%640
  • . ~SNA40

-- SNBB40

-Pressurnzer

-Cold Leg 1200 1000 0 600 1200 1800 2400 3000 3600 rime (seconds)

Figure 5A - Overfill Case of AOR

Attachment I to ET 10-0026 Page 13 of 20 2800 2600 2400

  • SN1-40 SNIA-40 SWi-40 2200 SN2A,40 SN3-40
  • SN3A-40 2000 SNA-40

-SN4A,40

  • SN-40 SW6-41J
a. 1800 SS~A-4 ShaB-40 SN7-40 1600 SN7A,40 SNB-40 SNBA.4

-Pressurize 1400

-Cold Leg 1200 1000 0 600 1200 1800 2400 3000 3600 Time (seconds)

Figure 5B - Overpressure Case of AOR Figure 5 RCS Pressure vs. Elapsed Time Pressurizer Level (ft)

The pressurizer level is 40 ft. [-90%] in the loss of normal ac power calculation (the analysis being designed to maximize the potential for pressurizer overfill), compared to a pressurizer level at less than 30 ft (or 70%) for the PFSSD scenarios. Pressurizer level, both loss of normal ac power event (Overfill and Overpressure case) and all PFSSD scenarios, is on scale. Results from these calculations demonstrate that the pressurizer will not become water solid following these postulated events. As such, the potential for filling the pressurizer relief tank and spilling radioactive coolant into the containment building does not exist. Thus, the postulated scenarios would not propagate to a more severe event (i.e., Condition III or IV event).

Attachment I to ET 10-0026 Page 14 of 20 45 40 SNIA-40 SN2.40J 35 SN2A,40 SN3-40 30 SN4-40

  • SN5-A,4 "i 25 S*8-40 SN740 20 SN7A,40

-Safety Analysis 15 10 600 1200 1800 2400 3000 3600 Time (e)

Figure 6A - Overfill Case of AOR 35 30

  • SN1-40 SNlA,40 SN2-40 SN2A,40 SN3-AO g SN4.40
  • SN4A,4fl SN6-40 20 SN7-4O SN7A,40 SN640A SNB8A.40 15 10 1800 3600 Time (e)

Figure 6B - Overpressure Case of AOR Figure 6 Pressurizer Level vs. Elapsed Time

Attachment I to ET 10-0026 Page 15 of 20 Demonstration of Fission Product Boundary Integrity Fission product boundary integrity is demonstrated by meeting the following three criteria:

No fuel clad damage Fuel cladding integrity is usually demonstrated by ensuring the calculated minimum DNBR remains above the 95/95 DNBR limit. To demonstrate cladding integrity, an examination of an analogous calculation, AN-96-026 (Reference 10), "DNB Analysis of Complete Loss of Flow Event with Reduced Thermal Design Flow with Tavg = 588.4 OF for Cycle 9," pertaining to the complete loss of flow was examined as it also featured a coast down of the reactor coolant pumps with conditions that minimize the calculated departure from nucleate boiling (DNB) ratio. It is observed that the limiting DNB occurs within a few seconds of the initiation of the event. For the time period in which DNB is a concern, the discussion above demonstrates that the process variables relevant to DNB for the PFSSD cases are bounded by the USAR Chapter 15 analysis of record for this event class. Although a small degree of upper core voiding (less than 5%) occurs in failed open PORV scenarios 1 and 1A, the voiding occurs later in the scenarios. The effect of some core voiding later in the transient does not present a challenge to fuel integrity as long as natural circulation is maintained and core exit temperatures remain below 712'F. Using the methodology of WCAP-14696-A (Reference 12), "Westinghouse Owners Group Core Damage Assessment Guidance," with indicated core exit temperature less than 7120 F, no core damage is indicated. For these scenarios, continuous positive core mass flow rates demonstrate that natural circulation was maintained and core exit temperatures less than 600OF confirm that cladding integrity was not challenged.

No rupture of the primary coolant boundary The criterion for the design basis accidents is to maintain pressures below 110% of the design values (1318.5 psia for the steam generators and 2748.5 psia for the RCS). An examination of the PFSSD process variable Figures 1 through 6 and the discussion above of the process variable comparisons with the loss of normal ac power event, leads to the conclusion that the pressure values (less than 1200 psia for the steam generators and less than 2400 psia for the RCS) demonstrate that the pressure criteria is not exceeded.

No rupture of the containment boundary The PORV is only open for three minutes for four PFSSD scenarios (1, 1A, 2, and 2A) and there is no further relief needed due to overfilling of the pressurizer in any of the PFSSD scenarios. It is not reasonable to assume the containment boundary is challenged by water relief from a PORV open for three minutes, as compared with the significantly larger mass and energy releases of the main steamline break (MSLB) inside containment and the large break loss of coolant accident (LOCA) analyses of record.

Maintenance of inventory is demonstrated by the pressurizer remaining on scale. Based upon an examination of the RETRAN analyses of all scenarios, the conclusion that subcritical reactor conditions are maintained can be obtained. In addition, hot standby conditions will be achieved and maintained.

Attachment I to ET 10-0026 Page 16 of 20 The results of Evaluation SA-08-006 Rev.1 shows that WCGS meets the guidelines of Generic Letter 86-10 (Reference 11), Enclosure 2, "Appendix R Questions and Answers," Section 5.3.10.

Section 5.3.10 of Generic Letter 86-10 identifies, for a control room fire scenario, that loss of offsite power is required to be assumed concurrent with the following additional spurious equipment actuation assumptions:

(a) The safe shutdown capability should not be adversely affected by any one spurious actuation or signal resulting from a fire in any plant area; and (b) The safe shutdown capability should not be adversely affected by a fire in any plant area which results in the loss of all automatic function (signals, logic) from the circuits located in the area in conjunction with one worst case spurious actuation or signal resulting form the fire; and c) The safe shutdown capability should not be adversely affected by a fire in any plant area which results in spurious actuation of the redundant valves in any one high-low pressure interface line.

Safe shutdown is not adversely affected by the formation of voids in the reactor head for a short time following a fire in the control room and spurious temporary opening of the PORV.

Conclusions The thermal-hydraulic analysis results demonstrate that the control room shutdown capability meets the Appendix R lII.L criteria with one exception. The RCS process variables are not maintained within those predicted for a loss of normal ac power. Specifically, voiding is predicted outside the pressurizer in the case of a spuriously open PORV whereas no voiding is predicted outside the pressurizer in the case of a loss of normal ac power. The voiding only lasts for a limited amount of time, about 10 to 12 minutes.

The results shown in Evaluation SA-08-006 Rev.1 demonstrate that this voiding outside the pressurizer does not cause any damage to a fission product boundary, nor does it cause a loss of natural circulation. The core mass flow rate calculated for the worst case scenario, one PORV spuriously open for 3 minutes, demonstrates that natural circulation is maintained. The average core exit temperature for the same scenario is calculated to be less than 6000 F. Based on the methodology of WCAP-14696-A (Reference 12), "Westinghouse Owners Group Core Damage Assessment Guidance," if the indicated core exit temperature is less than 712' F, no core damage is indicated. Therefore, since natural circulation is maintained and no core damage is indicated, the capability to achieve and maintain safe shutdown is assured.

4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria 10 CFR 50, Section 48, Fire Protection, paragraph (a)(1) states, in part: "Each holder of an operating license issued under this part or a combined license issued under part 52 of this chapter must have a fire protection plan that satisfies Criterion 3 of appendix A to this part."

Paragraph (b) states, in part: "Appendix R to this part establishes fire protection features required

Attachment I to ET 10-0026 Page 17 of 20 to satisfy Criterion 3 of appendix A to this part with respect to certain generic issues for nuclear power plants licensed to operate before January 1, 1979."

10 CFR 50, Appendix R Section III.L. Alternative and dedicated shutdown capability, states in part:

1. Alternative or dedicated shutdown capability provided for a specific fire area shall be able to (a) achieve and maintain subcritical reactivity conditions in the reactor; (b) maintain reactor coolant inventory; (c) achieve and maintain hot standby 2 conditions for a PWR (hot shutdown 2 for a BWR); (d) achieve cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and (e) maintain cold shutdown conditions thereafter. During the postfire shutdown, the reactor coolant system process variables shall be maintained within those predicted for a loss of normal a.c. power, and the fission product boundary integrity shall not be affected; i.e.,

there shall be no fuel clad damage, rupture of any primary coolant boundary, of rupture of the containment boundary.

2 As defined in the Standard Technical Specifications.

10 CFR 50, Appendix A, Criterion 3-Fire protection. Structures, systems, and components important to safety shall be designed and located to minimize, consistent with other safety requirements, the probability and effect of fires and explosions. Noncombustible and heat resistant materials shall be used wherever practical throughout the unit, particularly in locations such as the containment and control room. Fire detection and fighting systems of appropriate capacity and capability shall be provided and designed to minimize the adverse effects of fires on structures, systems, and components important to safety. Firefighting systems shall be designed to assure that their rupture or inadvertent operation does not significantly impair the safety capability of these structures, systems, and components.

The WCGS USAR, Appendix 9.5E, provides a comparison of the WCGS Fire Protection Program against the requirements of Section III of Appendix R to 10 CFR 50. Although WCGS obtained its operating license after January 1, 1979, the NRC stated, in the WCGS SER (NUREG-0881) dated April, 1982, that they will condition the WCGS operating license to require WCGS to meet the technical requirements of Appendix R to 10 CFR 50, or provide equivalent protection.

However, the Condition never appeared in the WCGS full power operating license when it was issued on June 4, 1985. Therefore, although Appendix R does not apply, WCGS's commitment to Appendix R is established in Appendix 9.5E as part of the approved Fire Protection Program documented in the USAR.

Table 9.5E-1 in Appendix 9.5E of the WCGS, USAR in response to the provisions in 10 CFR 50 Appendix R Section III.L.1, does not specify a deviation to the requirement for maintaining RCS process variables within those predicted for a loss of normal ac power.

4.2 Significant Hazards Consideration The proposed amendment would revise the Renewed Facility Operating License to deviate from certain Wolf Creek Generating Station (WCGS) Fire Protection Program requirements.

Specifically, a deviation from certain technical requirements to 10 CFR 50, Appendix R, Section III.L.1, as described in Appendix 9.5E of the WCGS Updated Safety Analysis Report (USAR), is requested for Reactor Coolant System (RCS) process variables not maintained within those

Attachment I to ET 10-0026 Page 18 of 20 predicted for a loss of normal ac power as evaluated in Evaluation SA-08-006 Rev.1 (Reference 1), "RETRAN-3D Post-Fire Safe Shutdown (PFSSD) Consequence Evaluation for a Postulated Control Room Fire."

WCNOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The design function of structures, systems and components (SSCs) are not impacted by the proposed change. Evaluation SA-08-006 Rev. 1 has demonstrated that the formation of voids in the reactor head for a short time following a fire in the control room and spurious temporary opening of the pressuizer power operated relief valve (PORV) does not result in damage to a fission product barrier and does not result in a loss of natural circulation cooldown. The proposed change does not alter or prevent the ability of SSCs from performing their intended function to mitigate the consequences of an initiating event within the assumed acceptance limits. Therefore, the probability of any accident previously evaluated is not increased.

Equipment required to mitigate an accident remains capable of performing the assumed function.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change Will not alter the requirements or function for systems required during accident conditions. The design function of structures, systems and components are not impacted by the proposed change. The thermal hydraulic analysis of the reactor coolant system identified that the process variables are not maintained within those predicted for a loss of normal ac power, however, the fission product boundary integrity is not affected.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on departure from nuclear boiling ratio (DNBR) limits, heat flux hot channel factor (FQ(Z)) limits, nuclear enthalpy rise hot

Attachment I to ET 10-0026 Page 19 of 20 channel factor (FAH) limits, peak centerline temperature (PCT) limits, peak local power density or any other margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, WCNOC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.3 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5. ENVIRONMENTAL CONSIDERATION WCNOC has evaluated the proposed changes and determined that the changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
6. REFERENCES
1. WCNOC Evaluation SA-08-006, "RETRAN-3D Post-Fire Safe Shutdown (PFSSD)

Consequence Evaluation for a Postulated Control Room Fire," Revision 1, February 1, 2010.

2. Wolf Creek Generating Station - NRC Integrated Inspection Report 05000482/2009004, November 10, 2009.ý
3. NUREG-0881, "Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1," April 1982.
4. NUREG-0881, "Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1," Supplement No. 3, August 1983.
5. Letter SLNRC 84-0109, "Fire Protection Review," from N. A. Petrick, SNUPPS, to H. R.

Denton, USNRC, August 23, 1984.

6. NUREG-0881, "Safety Evaluation Report related to the operation of Wolf Creek Generating Station, Unit No. 1," Supplement No. 5, March 1985.

Attachment I to ET 10-0026 Page 20 of 20

7. Wolf Creek Generating Station - NRC Inspection Report 05000482/20059008, February 1, 2006.
8. Wolf Creek Generating Station - NRC Triennial Fire Protection Inspection Report 05000482/2008010, January 2, 2009.
9. WCNOC Calculation AN-09-012, "RETRAN-3D Analysis of the Loss of Non-Emergency AC Power for the Feedwater Check Valve Relocation Project (DCP #12792)," Revision 0, July 31, 2009.
10. WCNOC Calculation AN-96-026, "DNB Analysis of Complete Loss of Flow Event with Reduced Thermal Design Flow with Tavg = 588.4 OF for Cycle 9," Revision 0, March 8, 1996.
11. Generic Letter 86-10, "Implementation of Fire Protection Requirements (Generic Letter 86-10), April 24, 1986.
12. WCAP-14696-A, "Westinghouse Owner's Group Core Damage Assessment Guidance,"

July 1996.

Attachment II to ET 10-0026 Page 1 of 3 Markup of USAR Pages

Attachment II to ET 10-0026 Page 2 of 3 WOLF CREEK TABLE 9.5E-1 (Sheet 25) 10CFR50 Appendix R WCGS covered by any complete shift personnel complement. These duties include command control of the brigade, transporting fire suppression and support equipment to the fire scenes, applying the extinguishant to the fire, communication with the control room, and coordin-ation with outside fire departments.

g. Potential radiological and toxic hazards in fire zones.
h. Ventilation system operation that ensures desired-plant air distribution when the ventilation flow is modified for fire con-tainment or smoke clearing operations.
i. Operations requiring con-trol room and shift engineer coordination or authorization.
j. Instructions for plant operators and general plant personnel during fire.

III. L. Alternative and Dedicated Shutdown Capability

1. Alternative or dedicated An auxiliary shutdown shutdown capability provided panel, described in for a specific fire area shall Section 7.4, in conjunction be able to (a) achieve and with certain local maintain subcritical reac- controls, provides a means tivity conditions in the of achieving and maintaining reactor, (b) maintain hot standby in the event reactor coolant inventory that the main control room (c) achieve and maintain is uninhabitable.

hot standby'7' conditions Rev. 23 Analysis demonstrates that the performance goals'of III.L.2 are satisfied. The performance criteria of III.L.1 are also satisfied, with the exception of maintaining reactor process variables within those predicted for a loss of normal ac power. This is acceptable, as long as a control room fire will not result in the plant reaching an unrestorable condition, which could lead to core damage. The criteria for "not reaching an unrestorable condition" are; natural circulation is maintained and adequate core cooling is maintained (core exit temperature is less than 712 0 F).

Attachment II to ET 10-0026 NO CHANGES THIS PAGE - INCLUDED FOR INFORMATION ONLY Page 3 of 3 WOLF CREEK TABLE 9.5E-1 (Sheet 26) 10CFR50 Appendix R WCGS for PWR (hot shutdown17 1 for The auxiliary shutdown a BWR); (d) achieve cold panel contains the con-shutdown conditions within trols and indication 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and (e) maintain necessary to maintain cold shutdown conditions reactor coolant system thereafter. During the inventory, remove decay postfire shutdown, the heat, and provide the reactor coolant system required boration for process variables shall hot standby. Adequate be maintained within operations shift staffing those predicted for loss is provided to achieve and of normal ac power and maintain post-fire safe shut-the fission product down "Hot Standby Conditions" boundary integrity shall in the event of a fire.

not be affected i.e., Cold shutdown can be there shall be no fuel achieved and maintained clad damage, rupture of from outside the control any primary coolant room by additional manual boundary, or rupture operator action at local of the containment control sites.

boundary.

The auxiliary shutdown

2. The performance goals for panel is included in the shutdown functions shall the fire hazards anal-be: ysis, Appendix 9.5B.
a. The reactivity control function shall be capable of achieving and maintaining cold shutdown reactivity conditions.

7' - As defined in the Standard Technical Specifications.

Rev. 23

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