ML100541604

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Request for Exemption from 10 CFR 50 Appendix R Section III.L.2.b, Performance Goal for Satisfying the Reactor Coolant Make-Up Function
ML100541604
Person / Time
Site: Pilgrim
Issue date: 02/03/2010
From: Jeffery Lynch
Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
2.10.011
Download: ML100541604 (48)
v • d • e


Text

v=Entergy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 February 3, 2010 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No.: 50-293 License No.: DPR-35

Reference:

Request for Exemption from 10 CFR 50 Appendix R Section I1l.L.2.b, Performance Goal for Satisfying the Reactor Coolant Make-up Function (1) NRC Regulatory Issue Summary (RIS) 2006-10, "Regulatory Expectations with Appendix R Paragraph III.G.2 Operator Manual Actions," June 30, 2006 (2) Entergy Letter, "Request for Exemption from 10 CFR 50 Appendix R Section III.G.2 to allow Hot Shutdown Manual Operator Actions,"

dated March 6, 2009.

(3) Entergy Letter, "Response To NRC RAI For Exemption To Allow Hot Shutdown Manual Actions At Pilgrim Nuclear Power Station (TAC NO. ME831)," dated December 8, 2009.

LETTER NUMBER: 2.10.011

Dear Sir or Madam:

In accordance with the provisions of 10 CFR Part 50.12, Pilgrim Nuclear Power Station (Entergy) hereby requests an exemption from certain provisions of 10 CFR Part 50.48, Appendix R to Part 50.48,Section III.L "Alternative and dedicated shutdown capability."

Specifically, an exemption from paragraph III.L.2.b is requested to permit the use of our Automatic Depressurization System (ADS) safety relief valves (SRVs) in conjunction with either the Core Spray (CS) System or Residual Heat Removal (RHR) System in the Low Pressure Coolant Injection (LPCI) mode to achieve and maintain safe shutdown for fires in certain fire areas and fire zones. Reliance on the manual operation of ADS and low pressure systems may result in short term core uncovery which does not meet the IIl.L.2.b requirement for ensuring that the reactor coolant makeup function shall be capable of maintaining the reactor coolant level above the top of the core for Boiling Water Reactors (BWRs).

This exemption is necessary based on a re-examination of Appendix R compliance strategies resulting from NRC guidance identified in Regulatory Information Summary (RIS) 2006-10 (Reference 1), Regulatory Guide (RG) 1.189 R2, and our pending exemption request documented in Entergy Letter dated March 6, 2009, (Reference 2),

as revised by Entergy Letter dated December 8, 2009 (Reference 3).

&W4~

Entergy Nuclear Operations Inc.

Letter Number: 2.10.011 Pilgrim Nuclear Power Station Page 2 This letter submits Entergy's request for exemption in accordance with 50.12(a)(2)(ii) which states, "application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule."

This letter contains no new commitments.

Please do not hesitate to contact Mr. Joseph R. Lynch, (508) 830-8403, if there are any questions regarding this submittal.

Sincerely, Joseph R Lynh Licensing Manager FXM Enclosure cc:

Mr. James S. Kim, Project Manager Plant Licensing Branch I-1 Division of Operator Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission One White Flint North O-8C2 11555 Rockville Pike Rockville, MD 20852 Regional Administrator, Region 1 U.S. Nuclear Regulator Commission 475 Allendale Road King of Prussia, PA 19406-1415 Office of the Resident Inspector U.S. Nuclear Regulatory Commission Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 to Entergy Pilgrim Letter 02.10.011 Request for Exemption from 10 CFR 50, Appendix R, Section II.L.2.b for allowing reactor water level to drop below the Top of Active Fuel (TAF) 20 Total pages

Entergy Pilgrim Letter 2.10.011 Request for Exemption from 10 CFR 50 Appendix R Section III.L.2.b for allowing reactor water level to drop below the Top of Active Fuel (TAF) 1.0 EXEMPTION REQUESTED In accordance with the provisions of 10 CFR Part 50.12, Pilgrim Nuclear Power Station (PNPS) hereby requests an exemption from certain provisions of 10 CFR Part 50.48, Appendix R to Part 50.48,Section III.L "Alternative and dedicated shutdown capability."

Specifically an exemption from paragraph III.L.2.b is requested to permit the use of our Automatic Depressurization System (ADS) safety relief valves (SRVs) in conjunction with either the Core Spray (CS) System or Residual Heat Removal (RHR) System in the Low Pressure Coolant Injection (LPCI) mode to achieve and maintain safe shutdown for fires in certain fire areas and fire zones. Reliance on the manual operation of ADS and low pressure injection systems may result in short term core uncovery which does not meet the III.L.2.b requirement for ensuring that the reactor coolant makeup function shall be capable of maintaining the reactor coolant level above the top of the core for Boiling Water Reactors (BWRs).

2.0 BACKGROUND

The initial PNPS response to 10 CFR 50, Appendix R was submitted for NRC review in 1982 (Reference 1). This submittal enclosed our initial Appendix R analysis report which identified the systems and support systems that would be utilized to accomplish safe shutdown in the event of fire in any plant area. This report identified methods and assumptions used in the safe shutdown analysis as well as a detailed fire zone summary for each fire zone. The fire zone summary identified fire hazards and fire protection features for each fire zone.

The initial Appendix R analysis report also identified the primary systems relied on for safe shutdown. These systems included the Automatic Depressurization (ADS) System, the High Pressure Coolant Injection (HPCI) System, the Reactor Core Isolation Cooling (RCIC) System, and the Residual Heat Removal (RHR) System. HPCI and RCIC were identified as systems which would be relied on for reactor coolant inventory control. As an alternative to the use of the high pressure HPCI and RCIC Systems, reactor depressurization via manual operation of the ADS valves and use of the low pressure injection systems (Core Spray or RHR-LPCI) was identified.

The initial Appendix R analysis report identified seventeen (17) of sixty-nine (69) fire zones did not comply with Section III.G.2 compliance. The report proposed specific modifications and/or procedure actions (i.e, operator manual actions outside the Control Room) to resolve the identified non-compliance issues. These fire zones were identified based on lack of cable separation between redundant safe shutdown systems. The availability or unavailability of high pressure systems used for reactor coolant injection was not used as a basis to identify III.G.2 compliance or non-compliance.

NRC review of the initial Appendix R analysis report resulted in issuance of a safety evaluation in 1983 (Reference 2). This safety evaluation approved the conceptual design for the proposed modifications and/or procedure actions applicable to the 17 fire zones identified as not meeting III.G.2 requirements. This safety evaluation also identified that the use of low pressure systems in conjunction with ADS for reactor inventory control was acceptable for use to demonstrate safe shutdown capability.

Page 1 K

Entergy Pilgrim Letter 2.10.011 Due to subsequent industry concerns and the fact that the safety evaluation was issued for approving the proposed design for III.G.3 and Ill.L, it is not clear if the NRC intended to approve the use of procedure actions or the use ADS and low pressure injection systems for III.G.2 compliance.

With respect to procedure actions (i.e., operator manual actions (OMAs) outside the Control Room to support Hot shutdown), the NRC issued RIS 2006-10 (Reference 3) which clearly indicated that OMAs are not allowed to demonstrate III.G.2 compliance, unless an exemption is requested and approved. In response to RIS 2006-10 instructions, PNPS submitted an exemption (Reference 4) from III.G.2 requirements to address the previously approved OMAs for specific fire zones. PNPS has subsequently revised the exemption (Reference 5) based on NRC information requests and recently issued NRC guidance contaned in RG 1.189, R2 (Reference 6). This exemption is still under review.

With respect to III.G.2 compliance and use of ADS and low pressure injection systems for reactor inventory control, NRC and industry concerns were reviewed subsequent to the 1983 safety evaluation. In 1999, the BWR Owners Group submitted information to the NRC which demonstrated that one train of ADS and low pressure injection systems is adequate to demonstrate redundant safe shutdown capability (i.e., acceptable for III.G.2 compliance). NRC review of these submittals is documented in NRC Letter dated December 12, 2000 (Reference 7).

Based on NRC review, use of ADS and low pressure systems is acceptable as "redundant" safe shutdown systems under Appendix R and therefore, may be used to demonstrate III.G.2 compliance. This information is also reflected in RG 1.189 R2 which was issued in October 2009. As a result, Appendix R 1II.L requirements do not apply to areas that comply with III.G.2 and no IIl.L.2.b exemption is needed for III.G.2 areas that rely on ADS and low pressure injection systems.

The following exemption request applies to the specific the fire zones that credit Alternative Shutdown (i.e., III.G.3 and III.L compliance) and also rely on ADS and low pressure injection systems for reactor inventory control. The focus of this exemption is to obtain an exemption from 10 CFR 50, Appendix R, Section IIl.L.2,.b for maintaining reactor coolant above the top of active fuel (TAF).

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Entergy Pilgrim Letter 2.10.011 3.0 EXEMPTION BASES 3.1 Plant Specific Analysis of using ADS and Low Pressure Injection Systems Prior to the NRC's review and acceptance of using ADS and low pressure injection systems for post fire safe shutdown in 2000 (Reference 7), PNPS contracted General Electric to perform a plant specific analysis of a post fire safe shutdown event where ADS and low pressure injection systems are relied on to demonstrate safe shutdown capability. This analysis (Reference 8) modeled a postulated Appendix R event where offsite power was lost; HPCI and ROIC systems were not available; two SRVs are manually operated to depressurize the reactor; and one train of low pressure systems (either CS or RHR) used for reactor injection. After reactor inventory is recovered one train of RHR (one pump) is operated in the suppression pool cooling mode.

Key assumptions and initial conditions included the following:

1. The reactor is operating at full operating power at normal water level at the start of the event.
2. The reactor is scrammed either from a loss of off-site power (LOOP) or from manual initiation from the Control Room.
3. MSIVs begin to close at event initiation either as a result of LOOP or due to manual closure.
4. Feedwater flow will ramp to zero within 5 seconds after event initiation.
5. The 1979 American Nuclear Society (ANS) decay heat correlation is used to realistically model decay heat.
6. The initial temperature of the suppression pool is 80 degrees F and containment pressure is 14.7 psia.
7. Water level in the suppression pool is the lowest level allowed by Technical Specifications.
8. The suppression pool air space is pressure is in equilibrium with the drywell pressure.
9. The suppression pool and the drywell are conservatively assumed to be insulated volumes.
10. No HPCI or RCIC is available during the event.
11. The Core Spray System for core cooling is available after reactor depressurization.
12. The RHR System for suppression pool cooling is assumed available at one hour after event initiation in the initial analysis. Suppression pool analysis is also performed for an RHR pump initiation time of two hours.

The acceptance criteria used in evaluating this fire event include criteria for fuel cladding integrity, and for suppression pool integrity.

The reactor level response provides a good indication whether fuel damage (fuel cladding perforation) is expected. If the core remains covered, no fuel cladding damage would occur. If the top portion of the core is uncovered for a brief period, the combination of low power level for this portion of the fuel and the steam updraft cooling will prevent any significant heat up or fuel cladding damage. To ensure the fuel cladding integrity, the calculated peak cladding temperature (PCT) should be less than the temperature (approximately 1500 degrees F) at which cladding damage can occur.

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Entergy Pilgrim Letter 2.10.011 To judge the acceptability of the suppression pool response three criteria were used:

1. Temperature and pressure in the pool should remain below the design conditions of 281 F and 56 psig.
2. The suppression pool temperature should remain below the Heat Capacity Temperature Limit (HCTL) for the pool while the reactor vessel is pressurized.
3. Adequate NPSH to the low pressure pump should be available in order to prevent possible pump cavitation. A suppression pool temperature of approximately 190 F for the RHR pump and a temperature of approximately 185 F at atmospheric pressure will provide adequate NPSH for the CS pump.

Analysis results identified the following:

The indicated water level remained above the top of the core for approximately 24 minutes at which time the reactor vessel is depressurized (the actual water level remained above the TAF until 27 minutes). When the reactor is depressurized to the operating pressure of the core spray pump, the pump operated to maintain reactor inventory. When reactor inventory was sufficiently recovered, the operator started the RHR System to provide suppression pool cooling. During the depressurization, some core uncovery was expected. This core uncovery resulted in a PCT of approximately 1320 F which is below the temperature at which cladding perforations can occur.

Therefore, the postulated fire event will not cause fuel cladding damage.

The suppression pool response to the event was also evaluated. The pool temperature before reactor depressurization (after 24 minutes) was about 106 F and the pool pressure was about 2 psig. The pool temperature was well below the HCTL for the corresponding reactor pressure. Utilizing the RHR system to provide suppression pool cooling after reactor depressurization, the maximum pool temperature and pressure are expected to remain below 180 F and 11 psig, respectively. Further, this combination of pool temperature and pressure is below the design limits, and will provide adequate NPSH to assure the operation of the RHR and Core Spray pumps for coolant injection.

Therefore, the requirements identified for the suppression pool are satisfied.

The analysis concluded that the PCT is low enough-to ensure fuel cladding integrity; that adequate margin to assure containment and suppression pool integrity for initial pool temperatures up to 100 F and the initiation of the RHR pump at up to two hours; and that available NPSH will maintain an adequate margin above required NPSH for the core spray and RHR pumps.

This analysis is the analysis of record and defines the acceptance criteria for the Appendix R fire event. This analysis is included as Attachment 2 to Enclosure 1.

Subsequent plant changes involving an increased ultimate heat sink temperatures were evaluated (Reference 9) to ensure Appendix R acceptance criteria were satisfied.

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Entergy Pilgrim Letter 2.10.011 3.2 Alternative Shutdown Zones that Credit ADS and Low Pressure Iniection Systems:

The following fire zones require an exemption from 10 CFR 50, Appendix R Section IIl.L.2.b.

Fire Area 1.9 Reactor Building Zones Zone 1.1 Reactor Building, El. -17'6", RHR and Core Spray Pumps Room "A" Zone 1.6 Reactor Building, El. -17'6", CRD Pump Room Zone 1.8 Reactor Building, El. 2'9", CRD Pump Room Mezzanine Zone 1.9* Reactor Building, El. 230", CRD Modules Area - east Zone 1.11

  • Reactor Building, El. 51'0", Open Area - east half Zone 1.14 Reactor Building, El. 74'3", Open Area - north half and south half Zone 1.13 all zones above Zone 1.32 Reactor Building, El. 23'0.", Steam Tunnel Fire Area 1.9 Turbine Building Zones - Stairwell Area 2.16 Zone 2.16* Radwaste and Control Building, Stairway #8, El. -1'0" to 37'0", Stairwell Fire Area 1.9 Turbine Building Zones - Vital MG Set Area 3.5 Zone 3.5* Radwaste and Control Building, El. 23'0", Vital Motor Generator Set Room Fire Area 1.10 Reactor Building Zones Zone 1.2 Reactor Building, El. -17'6", RHR and Core Spray Pumps Room "B" Zone 1.3 Reactor Building, El. -17'6", HPCI Pump Room Zone 1.4 Reactor Building, El. -17'6", HPCI Pump Panel and Valve Room

,Zone 1.5 Reactor Building, El. -17'6", RCIC Pump Room Zone 1.7 Reactor Building, El. 2'9", RCIC Pump Room Mezzanine Zone 1.10* Reactor Building, El. 23'0", CRD Modules Area - west Zone 1.12* Reactor Building, El. 51'0", Open Area - west half Zone 1.30A* Reactor Building, El. -17'6", Torus Compartment The plant fire zones listed above are zones that either directly rely on OMAs (annotated with (*) or are located in Reactor Building locations in close proximity to the zones that rely on OMAs. With respect to the fire zones in the Reactor Building that are adjacent to zones the, require OMAs, these zones utilize the same shutdown systems and follow the same safe shutdown procedures (References 10 and 11). However, theses zones are separate from the alternative shutdown room or zone under consideration per Generic Letter 86-10 (Reference 12) Enclosure 2, Question 3.1.5 and do not require the same level of fire protection as that defined in III.G.3.

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Entergy Pilgrim Letter 2.10.011 3.3 Previously Approved Appendix R Exemptions in Applicable Fire Zones Due to the fact that the above referenced plant fire zones may be redefined as part of this exemption request to be Alternate Shutdown (III.G.3 and IIl.L) zones, Appendix R Section III.G.3 requirements for providing fire detection and fixed suppression must be satisfied or an exemption needs to be requested to justify the adequacy of fire protection features provided.

Table 1 provides a compliance summary for each fire zone that requires the IIl.L.2.b exemption. References to the approved / pending fire protection exemptions are identified. It must be noted that some Reactor Building fire zones are located in the same fire area as the fire zones that require OMAs and can not be evaluated to be separated by fire hazards analysis. These zones utilize the same train of safe shutdown systems and rely on the same safe shutdown procedures to accomplish shutdown and will therefore also require exemption from IIl.L.2.b. However, these fire zones do not contain redundant safe shutdown cables, do not require OMAs, and may only be of consequence if fire starts and spreads to the adjacent zone where OMAs are required.

Therefore, consistent with GL 86-10, Enclosure 2, Question 3.1.5, alternate shutdown is provided on the basis of rooms or zones and the provision of fire detection and fixed suppression is only required in the room or zone under consideration (i.e., the zone requiring the OMAs).

The approved Appendix R exemptions applicable to the zones identified in Section 3.2 above, are identified and summarized below. The exemption summaries identify the purpose of the exemption and provide reference to submittal and approval letters.

Attached figures 1, 2, 3, 4, and 5 depict the Reactor Building floor elevations and applicable fire zone layouts.

These exemptions and the 1983 safety evaluation were utilized as the basis to implement the existing plant safe shutdown program. As such, the current safe shutdown program (shutdown procedures and fire protection features) are ensured to be consistent with the bases and analysis identified the approved exemptions. On-going reviews are performed to ensure fire protection program features are not adversely affected by the proposed plant changes.

In most instances, the approved exemptions address the adequacy of existing fire protection features (i.e., detection, suppression, and barriers) to satisfy Appendix R requirements. These exemptions do not justify the adequacy of primary safe shutdown systems (e.g., HPCI, RCIC ADS, CS, and RHR) to meet system performance goals applicable to Appendix R Section III.L (i.e, maintain reactor level above the top of the core when ADS and low pressure injection systems are used). Therefore, the previously approved exemptions are not affected by this new IIl.L.2.b exemption request.

Exemption #1B - Control Room Suppression and Detection PNPS requested an exemption from providing full area suppression in the Control Room. The area does not meet Appendix R requirements because it is an Alternate Shutdown area and Appendix R Section III.G.3 requires fire detection and fixed fire suppression to be installed in the area, room, or zone under consideration. Fire detection is provided in specific panels.

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Entergy Pilgrim Letter 2.10.011 Submittal/Approval Documents:

PNPS Letter 2.81.059, dated 3/18/81 NRC SER (Letter 1.81.336) dated 11/10/81 Exemption #5 - Torus Area Summary PNPS requested an exemption from providing an automatic suppression system and a fire detection system in the Torus Area.,- The area does not meet Appendix R requirements because automatic suppression and detection systems are not provided and redundant trains of torus temperature and level indication are in the area.

Submittal/Approval Documents:

PNPS Letter 2.83.130, dated 5/17/83 PNPS Letter 2.84.049, dated 4/02/84 NRC SER (Letter 1.85.191) dated 6/10/85 Exemption #7 and #8 - Reactor Building 23' and 51' Water Curtains PNPS requested an exemption from the requirement of providing full area suppression on the 23' and 51' elevations of the Reactor Building and from the requirement of having 20 feet of separation with no intervening combustibles or fire hazards. A water curtain was installed in the open area separating Fire Zones 1.9 from 1.10 and 1.11 from 1.12 to establish an east-west separation zone between redundant safe shutdown systems.

Submittal / Approval Documents:

PNPS Letter 2.83.130, dated 5/17/83 PNPS Letter 2.84.049, dated 4/2/84 PNPS Letter 2.86.084, dated 6/17/86 PNPS Letter 2.86-009, dated 2/3/86 NRC SER (Letter 1.84.379) dated 12/18/84 NRC SER (Letter 1.85.191) dated 6/10/85 NRC letter (1.86.282) dated 8/19/86 Exemption #9 - Lack of Suppression in Zones 1.9. 1.10. 1.111, 1.12. and 3.5 PNPS requested an exemption from the Appendix R Section III.G requirement to have a fixed suppression system in areas which credit or rely on Alternate Shutdown capability.

The following fire zones were specifically identified:

Fire Zone 1.9 - East Side Reactor Building, EL 23 ft.

Fire Zone 1.10 - West Side Reactor Building, EL 23 ft.

Fire Zone 1.11 - East Side Reactor Building, EL 51 ft.

Fire Zone 1.12 - West Side Reactor Building, EL 51 ft.

Fire Zone 3.5 - Vital M.G. Set Room, EL 23 ft.

Submittal / Approval Documents:

PNPS Letter 2.83.130, dated 5/17/83 PNPS Letter 2.85.049, dated 3/20/85 NRC SER (Letter 1.85.191) dated 6/10/85 Page 7

Entergy Pilgrim Letter 2.10.011 Exemption #11-Torus to NE Quad Fire Barrier PNPS requested an exemption from the requirement to provide a three-hour rated fire barrier between the Torus Area (FZ 1.30A) and the Northeast Quadrant of the Reactor Building below the 23-ft. floor elevation (FZ 1.6 and 1.8). Fire Zone 1.30A is associated with Appendix R Fire Area 1.10. Fire Zones 1.6 and 1.8 are associated with Appendix R Fire Area 1.9.

Submittal / Approval Documents:

PNPS Letter 2.83.281, dated 11/16/83 PNPS Letter 2.86.110, dated 7/28/86 PNPS Letter 2.86.176, dated 11/14/86 NRC SER (Letter 1.88.254) dated 7/14/88 Exemption #12 - Torus to SE Quad Fire Barrier PNPS requested an exemption from the requirement to provide a three-hour rated fire barrier between the Torus Area (Fire Zone 1.30A) and the Southeast Quadrant of the Reactor Building below the 23-ft. Floor elevation (Fire Zone 1.1). Fire Zone 1.30A is associated with Appendix R Fire Area 1.10. Fire Zone 1.1 is associated with Appendix R Fire Area 1.9.

Submittal / Approval Documents:

PNPS Letter 2.83.281, dated 11/16/83 PNPS Letter 2.86.110, dated 7/28/86 NRC SER (Letter 1.88.254) dated 7/14/88

  1. 13 - Structural Steel in Torus Area PNPS requested an exemption from the requirement to provide fire protection to structural steel members supporting the fire barrier that separates the Torus Area (Fire Zone 1.30A) from Reactor Building locations on the 23' floor elevation (Fire Zones 1.9, 1.9A, 1.10, 1.10A, 1.10B, 1.25, and 1.32). Fire Zones 1.30A, 1.10, 1.10A and 1.10B are associated with Appendix R Fire Area 1.10. Fire Zones 1.9, 1.9A, 1.25, and 1.32 are associated with Appendix R Fire Area 1.9.

Submittal / Approval Documents:

PNPS Letter 2.83.281, dated 11/16/83 PNPS Letter 2.86.110, dated 7/28/86 PNPS Letter 2.86.176, dated 11/14/86 PNPS Letter 2.87.062, dated 4/21/87 PNPS Letter 2.87.132, dated 8/4/87 PNPS Letter 2.88.010, dated 1/19/88 NRC SER (Letter 1.88.254) dated 7/14/88 Exemption #14 - Structural Steel in Main Steam Tunnel PNPS requested an exemption from the requirement to provide fire protection to structural steel members supporting the ceiling fire barrier that separates the Main Steam Tunnel (Fire Zone 1.32) from Reactor Building and Turbine Building locations on Page 8

Entergy Pilgrim Letter 2.10.011 Enclosure.1 the 51' floor elevation (Fire Zones 1.11, 1.12, 1.23A and 1.23B). Fire Zone 1.32 and 1.11 are associated with Appendix R Fire Area 1.9. Fire Zones 1.12, 1.23A and 1.23B are associated with Appendix R Fire Area 1.10.

Submittal / Approval Documents:

PNPS Letter 2.83.281, dated 11/16/83 PNPS Letter 2.86.110, dated 7/28/86 PNPS Letter 2.88.010, dated 1/19/88 NRC SER (Letter 1.88.254) dated 7/14/88 Exemption #15 - Radwaste/ Control Building Corridor at the (-) 1' Elev PNPS requested an exemption from the requirement to install full area fire detection and automatic suppression in the Radwaste and Control Building between Corridor #137 on elevation 23 feet and Corridor #49 on elevation (-) 1 foot. These plant areas contain redundant Division A and B safe shutdown power cables. Corridor #137 is located in Fire Zone 3.9A and Fire Zone 3.9A is associated with Fire Area 3.3. Corridor #49 is located in Fire Zone 3.7 and Fire Zone 3.7 is associated with Fire Area 1.10..

Submittal / Approval Documents:

PNPS Letter 2.87.135 dated 8/10/87 NRC SER (Letter 1.88.261) dated 7/20/88 Exemption #18 - Reactor Building Annex PNPS requested an-exemption from the requirement that no intervening combustibles be present between redundant safe shutdown systems located in the Reactor Building Annex area. Train A safe shutdown cables associated with the RBCCW, SSW, and EDG Fuel Oil Systems are routed through the Acid Neutralizing Sump (1.21 A) which is located at the 13 foot elevation. Redundant Train B safe shutdown equipment and cables are located in the "B" RBCCW Room (FZ 1.22 and 1.3A) which is on the 3 foot elevation. Although the Acid Neutralizing Sump and the RBCCW Pump Room are separate fire areas, non-rated floor/ceiling barrier penetrations allow both of these areas to communicate with the Water Treatment Area (FZ 1.29), located above on the 23 foot elevation. Fire Zone 1.21 A is associated with Fire Area 1.21. Fire Zones 1.3A, 1.22, and 1.29 are associated with Fire Area 1'.10. The exemption identifies that adequate fire protection is provided to ensure that fire will not spread to and impact redundant safe shutdown systems or cables.

Submittal / Approval Documents:

PNPS Letter 2.87.135 dated 8/10/87 NRC SER (Letter 1.88.261) dated 7/20/88 Exemption #21 - Hot Shutdown Repairs:

PNPS requested and exemption from the requirements of Appendix R,Section III.G.1.a for performing a series of pre-planned procedurally-controlled operator actions to transfer control, then replace control circuit fuses or install jumper wire if necessary, on specific equipment required to achieve and maintain hot shutdown. The exemption applies to Fire Zones 3.1, 3.2 and 1.9.

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Entergy Pilgrim Letter 2.10.011 Submittal / Approval Documents:

PNPS Letter 2.87.160, dated 10/2/87 NRC SER (Letter 1.88.120) dated 04/14/88 Exemption #22 - Torus Level Instrumentation Separation in Zones 1.23, 3.4, and 3.11 PNPS requested an exemption from the requirement for providing fire detection and automatic suppression for torus water level indication cable located in specific Reactor Building and Radwaste/Control Building locations. The exemption applies to the Radwaste/Control Building areas (FZ 1.23B, 3.4, and 3.11) that are associated with Fire Area 1.10, and credit the availability of the alternate shutdown torus level indication provided on Panel C165 (FZ 1.10) which is in Fire Area 1.10. Control Room indication and ASD Panel indication will not be impacted by common fire.

Submittal / Approval Documents:

PNPS Letter 2.87.135 dated 8/10/87 NRC SER (Letter 1.88.261) dated 7/20/88 4.0 SPECIAL CIRCUMSTANCES:

Achieve and Maintain Hot Shutdown During a Fire Event The underlying purpose of 10 CFR 50 Appendix R, Section III.G.1 is to provide

-/

reasonable assurance that at least one train of systems necessary to achieve and maintain safe shutdown conditions from either the control room or emergency control stations(s) is free from fire damage. Pilgrim station satisfies the underlying purpose of 10 CFR 50 Appendix Section III.G.1 for all plant fire areas.

However, as stated in our initial 1982 response to Appendix R (Reference 1), Pilgrim Station did not meet the separation requirements of Appendix R Section III.G.2 for seventeen (17) of the sixty-nine fire zones evaluated. Consistent with compliance strategies described in the initial Appendix R submittal for these seventeen (17) fire zones, and the NRC SER (Reference 2) which approved the conceptual design for alternate safe shutdown, Pilgrim Station implemented modifications and developed safe shutdown procedures which relied on operator manual actions to ensure safe shutdown capability in 1987.

Fire Zones 1.9, 1.10, 1.11, 1.12, 1.30A, 2.16, and 3.5, were included with the list of seventeen (17) zones that did not meet Appendix R,Section III.G.2 requirements.

Modifications were not implemented to resolve compliance concerns (ie, OMAs) in these zones and procedure actions are relied on to ensure safe shutdown capability. The previously approved procedure actions are now considered acceptable for use in Alternative Shutdown (meets II.G.3 compliance) zones. Additionally, these fire zones rely on the use of low pressure systems used in conjunction with ADS to restore reactor inventory.

Reliance on these systems for reactor inventory control inherently results in short duration uncovery of the reactor core which does not satisfy the Appendix R, Section IIl.L.2.b requirement for maintaining reactor coolant level. As a result an exemption from Page 10

Entergy Pilgrim Letter 2.10.011 the requirements of Appendix R, Section IIl.L.2.b (to maintain the reactor coolant level above the top of the core) is required.

Reliance on ADS and low pressure injection system for reactor inventory control has been evaluated by the NRC to be an acceptable means to accomplish safe shutdown for both 114.G.2 and for III.G.3 compliance. NRC review and approval of this safe shutdown strategy is documented in the PNPS safety evaluation (Reference 2) and in generic industry reviews (Reference 7). As such, reliance on these systems is acceptable and the underlying intent of the 10 CFR 50, Appendix R Section III.G.1 regulation to ensure safe shutdown capability is satisfied.

Special Circumstance A review of NRC Regulatory Guide (RG) 1.189, Revision 2, dated October 2009 identifies that special circumstances exist when an approved SER is issued and no corresponding exemption has been approved. The RG identifies that pre-1979 licensees need an exemption, even if a staff decision in an SER approves an aspect of the Fire Protection Program (FPP) that does not comply with regulatory requirements.

The RG specifically identifies (on page 13 under SERs) that pre-1979 licensees that have SERs, but not a corresponding exemption from the regulatory requirements must request an exemption under 10 CFR 50.12 by (1) highlighting the special circumstances of 10 CFR 50.12(a)(2), (2) citing the SER as the safety basis, and (3) confirming that the safety basis established in the SER remains valid.

In accordance with the instruction provided in RG 1.189, the special circumstance is that the provision of 10 CFR 50.12(a)(2) applies; the previously approved SER provides the safety basis; and the safety basis established in the SER remains valid.

5.0 CONCLUSION

Based on the technical justification and the special circumstances detailed above, Pilgrim Station requests an exemption from the requirements of 10 CFR Part 50, Appendix R, Section I1l.L.2.b for Alternative Shutdown fire zones that credit the use of ADS and low pressure systems for reactor inventory make-up. Reliance on these previously approved systems may result in short term core uncovery but will not jeopardize capability to achieve and maintain safe shutdown.

6.0 REFERENCES

1.

Pilgrim Letter 2.82.180, "Boston Edison Response to Appendix R", dated June 25, 1982.

2.

NRC Letter dated November 2, 1983, "Safety Evaluation for Appendix R to 10 CFR Part 50 Items G3 and III.L", (PNP Letter No. 1.83.267).

3.

NRC Regulatory Issue Summary 2006-10, "Regulatory Expectations with Appendix R Paragraph III.G.2 Operator Manual Actions"' June 30, 2006.

4.

Entergy Letter, "Request for Exemption from 10 CFR 50 Appendix R Section III.G.2 to allow Hot Shutdown Manual Operator Actions," dated March 6, 2009.

Page 11

Entergy Pilgrim Letter 2.10.011

5.

Entergy Letter, "Response To NRC RAI For Exemption To Allow Hot Shutdown Manual Actions At Pilgrim Nuclear Power Station (TAC NO. ME831 )," dated December 8, 2009

6.

NRC Regulatory Guide 1.189, R2, dated October 2009, "Fire Protection For Nuclear Power Plants"

7.

NRC Letter dated December 12, 2000, Subject "BWR Owners Group Appendix R Fire Protection Committee Position on SRVs + Low Pressure Systems used as "Redundant' Shutdown Systems Under Appendix R (Topical Report GE-NE-T43-0002-00-03-ROl (TAC NO. MA8545)"

8.

General Electric Company Report, EAS 82-0787, July 1987 (PNPS SUDDS/RF 87-889), Safe Shutdown Analysis for the Pilgrim Nuclear Power Plant.

9.

General Electric Company Report, GE-NE-T23-00749-01, December 1997 (PNPS SUDDS/RF 97-96), "Pilgrim Nuclear Power Station Containment Heatup Analysis with ANS 5.1 + 2a Decay Heat."

10.

Pilgrim Procedure 2.4.143.1, "Shutdown with fire in the Reactor Building East (Fire Area 1.9)", R15

11.

Pilgrim Procedure 2.4.143.2, "Shutdown with fire in the Reactor Building West (Fire Area 1.10)", Rev. 15.

12.

NRC Generic Letter 86-10, "Implementation of Fire Protection Requirements", April 24,1986.

13.

Pilgrim Station Power Systems Calculation No. PS-32, "Appendix R Safe Shutdown Analysis Report", Revision 5

14.

Pilgrim Station Updated Fire Hazards Analysis, Report 89XM-1 -ER-Q, Revision E5.

Page 12

Enclosure I to Entergy Pilgrim Letter 2. 10.011 Table 1 Fire Zones Compliance Summary Fire Area/

Zone III.G.2 Alternative High ADS / Low Hot Approved III.G.3 and Fire Zone Description Compliance Shutdown Pressure Pressure Shutdown Appendix R III.L.2.b under Problem (ASD) Zone System System Operator Exemption Compliance consideration (Initial Available Available Manual (See Note 1)

Submittal/

(Current)

(Initial Actions SER and Submittal (OMAs) proposed and resolution)

SER)

Area 1.9 Zone 1.1 RHR A Quad No No No Yes No

  1. 12 - Torus No OMAs; to RHR A Exemption Quad Barrier requested from IIl.L.2.b.

Zone 1.6/

CRD Quad Yes (Zn 1.8 No No Yes No

  1. 11 - Torus No OMAs; 1.8*

only-To CRD Exemption proposed Quad Fire requested mod)

Barrier from IIl.L.2.b.

Zone 1.9*

RB East 23'

-Yes Yes No Yes Yes

  1. 7 - RB wtr Existing (proposed Curtain; #9-III.G.3 mods and Full area FP exemption; OMAs)

-Sup and Exemption detect in Alt requested Shutdown from IIl.L.2.b zone; #21 -

Hot Shutdown Repairs; #22 Torus Instruments Page 1 to Entergy Pilgrim Letter 2. 10.011 Table 1 Compliance Summary Fire Zones Fire Area/

Zone III.G.2 Alternative High ADS / Low Hot Approved III.G.3 and Fire Zone Description Compliance Shutdown Pressure' Pressure Shutdown Appendix R III.L.2.b under Problem (ASD) Zone System System Operator Exemption Compliance consideration (Initial Available Available Manual (See Note 1)

Submittal/

(Current)

(Initial Actions SER and Submittal (OMAs) proposed and resolution)

SER)

Zone 1.11*

RB East 51' Yes Yes No Yes Yes

  1. 8 - RB wtr Existing (proposed Curtain; #9 -

III.G.3 mods and Full area FP exemption; OMAs)

Sup and Exemption detect in Alt requested Shutdown from IIl.L.2.b zone; #22 Torus Instruments Zone 1.14*;

RB 74' Yes (Zone No HPCI Yes No None No OMAs; (and all Rx 1.14 only-Exemption Bldg zones at proposed requeste~d or above 74')

cold from IIl.L.2.b.

shutdown op action)

Zone 1.32*

RB Steam Yes No No Yes No

  1. 14-No OMAs; Tunnel (proposed Structural Exemption cold Steel in Fire requested shutdown Barrier from III.L.2.b.

op action)

I Page 2 to Entergy Pilgrim Letter 2. 10.011 Table 1 Fire Zones Compliance Summary Fire Area/

Zone III.G.2 Alternative High ADS / Low Hot Approved III.G.3 and Fire Zone Description Compliance Shutdown Pressure Pressure Shutdown Appendix R IIl.L.2.b under Problem (ASD) Zone System System Operator Exemption Compliance consideration (Initial Available Available Manual (See Note 1)

Submittal/

(Current)

(Initial Actions SER and Submittal (OMAs) proposed and resolution)

SER)

Zone 2.2*

Swgr Rm "A" Yes No No Yes No None Op actions and 2.3 and Battery (proposed important to Rm OMAs) safe shutdown; III.G.2 Compliance based on fire barriers Zone 2.16*

Stairway # 8 Yes Yes No Yes Yes Pending -

Exemption (proposed Lack of fixed pending; OMAs) suppression Exemption in ASD area requested from IIl.L.2.b Zone 3.5*

Vital MG Set Yes Yes No Yes Yes

  1. 9-Lack of Exemption Rm (proposed Fixed Supp approved; OMAs) in ASD Area Exemption requested from IIl.L.2.b.

Fire Area 10 Zone 1.2 RHR B Quad No No No OMAs; Exemption requested from IIl.L.2.b.

Page 3 to Entergy Pilgrim Letter 2. 10.011 Table 1 Fire Zones Compliance Summary Fire Area/

Zone III.G.2 Alternative High ADS / Low Hot Approved III.G.3 and Fire Zone Description Compliance Shutdown Pressure Pressure Shutdown Appendix R IIl.L.2.b under Problem (ASD) Zone System System Operator Exemption Compliance consideration (Initial Available Available Manual (See Note 1)

Submittal/

(Current)

(Initial Actions SER and Submittal (OMAs) proposed and resolution)

SER)

Zone 1.3/1.4 HPCI Room No No No Yes No None No OMAs; and Exemption requested from IIl.L.2.b.

Zone 1.5/1.7 RCIC Quad No No' No Yes No No OMAs; Exemption requested from III.L.2.b.

Zone 1.10*

RB West 23' Yes Yes No Yes Yes

  1. 7 - RB wtr Existing (proposed Curtain; #9-III.G.3 mods and Full area FP exemption; OMAs)

Sup and Exemption detect in Alt requested Shutdown from IIl.L.2.b zone; #22 Torus

_Instruments Zone 1.12*

RB West 5i' Yes Yes No Yes Yes

  1. 8-RB wtr Existing (proposed Curtain; #9-III.G.3 mods and Full area FP exemption; OMAs)

Sup and Exemption detect in Alt requested Shutdown from IIl.L.2.b zone Page 4 to Entergy Pilgrim Letter 2.10.011 Table 1 Fire Zones Compliance Summary Fire Area/

Zone III.G.2 Alternative High ADS / Low Hot Approved III.G.3 and Fire Zone Description Compliance Shutdown Pressure Pressure Shutdown Appendix R IIl.L.2.b under Problem (ASD) Zone System System Operator Exemption Compliance consideration (Initial Available Available Manual (See Note 1)

Submittal/

(Current)

(Initial Actions SER and Submittal (OMAs) proposed and resolution)

SER)

Zone 1.22 RBCCW "B" Yes No No Yes No

  1. 18 -

No OMAs; Pp Rm (proposed Redundant III.G.2 mods)

Train demonstrated Separation with fire barriers.

Zone 1.30A*

Torus Yes Yes No Yes Yes

  1. 5 - Torus Existing Compartment (proposed Instrument III.G.3 mods and and lack of exemption; OMAs) suppression Exemption and requested detection; from III.L.2.b
  1. 11, #12, and #13 -

Fire Barrier and Structural Steel to adj Quad zones.

Page 5 to Entergy Pilgrim Letter 2. 10.011 Table 1 Fire Zones Compliance Summary Fire Area/

Zone III.G.2 Alternative High ADS / Low Hot Approved III.G.3 and Fire Zone Description Compliance Shutdown Pressure Pressure Shutdown Appendix R IIl.L.2.b under Problem (ASD) Zone System System Operator Exemption Compliance consideration (Initial Available Available Manual (See Note 1)

Submittal/

(Current)

(Initial Actions SER and Submittal (OMAs) proposed and resolution)

SER)

Zone 2.1*

Swgr Rm "B" Yes No No Yes No None Op actions and 2.4 and Battery (proposed important to Rm OMAs) safe shutdown; III.G.2 Compliance based on fire barriers Zone 2.10*

Condenser Yes (Zone No No Yes No None Op actions (and other Bay 2.10 only -

important to Turbine Bldg proposed safe zones 2.5 OMAs) shutdown; thru 2.13)

III.G.2 Compliance based on fire barriers Zone 1.23 SBGT 51' No Yes Yes Yes Yes

  1. 22 - Torus Exemption Instruments approved for and Lack of III.G.3:

FP Sup and IIl.L.2.b detection in exemption not Alt required.

Shutdown Zone Page 6 to Entergy Pilgrim Letter 2. 10.011 Table 1 Fire Zones Compliance Summary Fire Area/

Zone III.G.2 Alternative High ADS / Low Hot Approved III.G.3 and Fire Zone Description Compliance Shutdown Pressure Pressure Shutdown Appendix R IIl.L.2.b under Problem (ASD) Zone System System Operator Exemption Compliance consideration (Initial Available Available Manual (See Note 1)

Submittal/

(Current)

(Initial Actions SER and Submittal (OMAs) proposed and resolution)

SER)

Zone 3.4 CR HVAC No Yes Yes Yes Yes

  1. 22 - Torus Exemption Fan Rm 51' Instruments approved for and Lack of III.G.3:

FP Sup and III.L.2.b detect in Alt exemption not Shutdown required.

Zone Zone 3.11 Control Rm No Yes Yes Yes Yes

  1. 22 - Torus Exemption Annex 37' Instruments approved and Lack of from III.G.3:

FP Sup and IIl.L.2.b detect in Alt exemption not Shutdown required.

Zone Fire Area 1.21 Zone 1.21

  • RBCCW "A" Yes No No Yes No
  1. 18-No OMAs; Pp Rm (proposed Redundant III.G.2 mods)

Train demonstrated Separation with fire barriers.

Page 7 to Entergy Pilgrim Letter 2.10.011 Table 1 Fire Zones Compliance Summary Fire Area/

Zone III.G.2 Alternative High ADS / Low Hot Approved III.G.3 and Fire Zone Description Compliance Shutdown Pressure Pressure Shutdown Appendix R IIl.L.2.b under Problem (ASD) Zone System System Operator Exemption Compliance consideration (Initial Available Available Manual (See Note 1)

Submittal/

(Current)

(Initial Actions SER and Submittal (OMAs) proposed and resolution)

SER)

Fire Area 3.1 Zone 3.1*

Control Rm Yes Yes Yes Not currently Yes

  1. 1B - Lack Existing (proposed credited of Fixed III.G.3 mods and Suppression; exemption; OMAs)
  1. 21 - Hot HP Systems Shutdown available, repairs III.L.2.b exemption not required.

Fire Area 3.2 Zone 3.2*

Cable Yes Yes Yes Not currently Yes

  1. 21 - Hot In compliance Spread Rm (proposed credited Shutdown with III.G.3; mods and repairs HP Systems OMAs) available, IIl.L.2.b exemption not required.

Note 1-The fire zones annotated with an asterisk (*) were identified in the original Appendix R Submittal as a zone where III.G.2 compliance was not demonstrated.

Page 8 to Enclosure 1 Entergy Pilgrim Letter 2.10.011 Pilgrim Station Exemption request from 10 CFR 50 Appendix R,Section III.L.2.b Reactor Building Fire Zones Figures 1, 2, 3, 4, and 5 5 Total pages Page 1 to Entergy Pilgrim Letter 2.10.011 Reactor Building Figures Reactor Building (-) 17' elevation Figure 1 Page 2 to Entergy Pilgrim Letter 2. 10.011 Reactor Building Figures Reactor Building 23' elevation Figure 2 Page 3 to Entergy Pilgrim Letter 2.10.011 Reactor Building Figures Reactor Building 37' elevation Figure 3 Page 4 to Entergy Pilgrim Letter 2. 10.011 Reactor Building Figures a AAU 4

  • f Reactor Building 51' elevation Figure 4 Page 5 to Entergy Pilgrim Letter 2.10.011 Reactor Building Figures SEEt B

.6 a

ft A 111 Reactor Building 74' elevation Figure 5 Page 6

f to Enclosure I to Entergy Pilgrim Letter 2.10.011 Pilgrim Station Exemption request from 10 CFR 50 Appendix R, Section III.L.2.b General Electric Analysis of Pilgrim Station Appendix R Fire Event EAS82-078 DRF AOO-03144 Rev. 0 July 1987 17 Total Pages Page 1

AM 802-0787 y 18i" A00,0S14

.Judl 187 9883002-73 SAFE SHUTDOWN APErDIX R A.ALYSE FOR THE PILGRxI NUC.AR powuE PLANT Prepared b' ble.

P ARM"-

C3' NiI

.. 0.

Reviewed by Approved by K.-.L-. U~dfq&d -In ginee.:

Applicatio6 Snino.*r*ng Services L. L. tbi, beni*or Engineer Application Engineering Services Application BnigLeoring Services NUCLEA INERGY BMNI OPIATIONS 0 GUNERmN.

KIC1 COUPNm OCORM4A 86 SEN hAL

  • ELICTIgS

- a

  • IJIORTANT NOTICE R]MIADI* G cONTMs OF THIS. iZCOT Please Read, CArefuly The 6nly n... g.

G.,enra i

ls t Tuep

r. *9i.tug-

+.4 1W It contained in theW ;ýntraot tweJ~the

  • customer and.?-*on Eectric*.

Com,..pa@y oo Identif tled in otbpeu-chas tconstrued.

as c*hngingthe contract.- The use of ht,*a Lfoa mmtl.11*

a*4ori n

t)'WI customer..or, for any purpos**- other than "whih $t is

itended, is not aehorized; and with respectlto any unauti*btZ'd use Ganeral Electric Company makes n& r0presen8tainor varranty, and assumes no liability as to the completeness, accurayo.

0"r' useiful6"e

ýof the information contained in this 6ocument.

N :.

I :

1. INTRODUCTION

1.1 BACKGROUND

In accordance with 10CFR50.48 and 10CFR50 Appendix R, adequate protection of plant equipment is required to ensure safe shutdown of a nuclear power plant in the unlikely event of a fire in any,,plant lor~ation.

The. fire event is'assumed to have concurrent loss of off-site power (LO0S?).

loss of all automatic function affected by the fire, spurious operation of plant equipment,

and, if necessary, control room.evacuation.

For* fire events V1 which, may, result in: control room evacuation, Boston Edison Companyr (bMco) o has elected to ImplIment a remote shutdown system (1S5) for the Pilgrim Nuclear Power Station (IMPS).

Part of the RSS includes two panels with manual control for pressure relief.

These panels include all four safety relief valves (SRVs); two SRVs at each panel.

A fire in the vicinity of one of these shutdown panels may result In a loss of the two SRVs controlled from the panel.

Also, the high pressure make up systems may not be available for safe shutdown since the panel area has cables associated with HPCI and RCIC equipment.

A fire event may also render some of the pumps for the Residual Heat Removal (RHR) system and Low Pressure Core Spray (LPCS) system unavailable for core tooling, This Implies that in the event of a fire at one of the panels, only two RVs end one LPCS pump will be available for core cooling, and one RHR pump will be available for decay heat removal for safe shutdown.

In order to demonstrate that safe shutdown can be achieved with the equipment specified above, analyses were performed to determine if there is adequate capability to depressurize the reactor to ensure no buel daa, and no failure of the primary pressure vessel or containaeut.

Ptevious

I analyses (Reference 1) have been performed using the Low Pressure Coolant injection (LPCI) mode and normal shutdown cooling mode of the RUR sys~ou.

Hovever, due to procedural considerations, DECo has decided to operate the MH&

pumps in the suppression pool cooling node for decay host removal.

The RHR pump will be used to remove the decay beat from the suppression pool after the LYCS pump has been placed In service for core cooling.

Further, due to potential loss of control circuits, the RHR pump may require a longer time to become available than previously assumed.,

0-.

1.2 SCOPE AN~D ODJECTIV5 The, spcfcobetv f the analysesis to demonstrate that the plant can achieve safe shut down and that there is no fuel "da'se, and there Is mmre than adequate margin to assure containment and suppression pool integrity.

The system performance requirements considered In those analyses were consistent with those specified in Section I1.L of 10CFRSO Appendix R and NRC guidelines (References 1 and 2).

The analyses. and their results are described and awmarizod in the following sections.

CD

/

~.J 101

k0.

At the start of the. hypothetical fire event at nse of the two shutdown panels, the reactor it assumed to be operating at full power, normal water

level, and steady state conditions.

The evaluation of the fire event assumes :the concurrent LOOSP.

During a LOOSP event, the loss of power to the Reactor Protection System. (RPS) and the fail-safe C46sn of the isolatLon eystem ill result in an automatic scram and Isolation.

Inmdistely after scram and isolation, the reactor pressure increase is limited by the SRrs operating*in the pressurc actuation mods.

Tbh*function

.of the SRVs 'in thisý modes is not affected by. th fire, since the SRVP-. are located in the Snerted gontainsent and they function i.n mechanicAlode which does. not rely on external power.

There Is no potential: for overpressuritation because the SOVo are designedsand sized to accemeodate this type of Isolation event.

There is no potential for fuel dauage because this event is similar to many transient events, involving LOOSP end closure of all main Steam Isolation Valves (KSIVs),

previously analyzed as part of the plant's design bases safety evaluation.

The LOOSP also results in a loss of foedwater flow to the reactor.

Normally, the high pressure make up systems (HPCI and RCIC) will operate to maintain reactor water inventory when substantial reactor inventory is lost due to the SaVs actuations.

As discussed earlier, for this postulated fire at one of the two shutdown panels, both FIPI and RCIC are assumed to be unavailable. Without the high pressure make up systems, the Skye actuations result in a gradual lose of reactor water inventory.

This "boll off continues with the reactor maintained at high pressure (around 1100. pat) until the operator manually actuates the two SilV to depreesurlse the reactor.

The operator would monitor reactor pressure, reactor weter Level, suppression pool water temperature and level.

Based on this informatlm, the operator would take appropriate actions to achieve safe shutdown.

A.

14,

In accordance With

.te plant

.mergnhy c u

hen the Indlcated eter level rehim the: Op: 0o'If,; active fuel. the operator will initiate aualdepre.surizatiowit t

ie Rova'labl, and ali, the low

..pressure systems for recovery of reactor Luventory-nd removal of the decay heat-.-.

When-the r.eactor inventory i%: recovered and when the reactor

.preaure ii 2ov enoug.,-the operator will begla operating the Residual Heat R.moval (RHR) system I. SuPPression.

Pool*.Cooli.

(M) m.ode.

W.

0

'a,

3. WENTl ANALYSIS For the postulated fire event. the evaluation of the reactor response was performed, utilizing, the 02 blowdown model (SAVE).

The SAF Code dotermines the reactor coolant inventory (water level) response.

If the resuts show core uncovery occurs, the fuel cladding integrity r. lluation, is-performad by determining the resultn.g pea* cladding temperature (P?') for the duration of the core.ncovery.

The PCT calculetions were; perormed by incorporating the WAE output In the corc eau a

,ayi code(cM?)

The integrity of the suppression, pool ye evalted by m

ass

+.+and:

energy blanceia on4 the suppression pool accountnlg for the uses and

!!}?iii+:i)ien-ergy iDadded by t~he SRVs.

++++++++ The basic +anaysis assumpti+os are consistent-wit Ifl OCQ+++S, loGPJ.5 Appendix R and the plant emergency procedure guidelines (EPGB)..

C4

3.

.ASSUkPTIONS ANID INITIAL CONDITIONIS The key model or Input assumptions and initial conditions used In the analyses are summarize* below;

a.

The reactor is assumed to be operating at full power. and at normal water level at the time of event Initiation.

b.

The event initiation occurs concurrently with OOSP.

c.

The reactor scrams at event initiation either as a result of the concurrent LOOSP or by manual action from the control room.

d.

The Main Steam Isolation Valves (SIVs) begiLn to close at event initiation either as a result of LOOSP ot due to manual closure.

es Feedwater flow is assumed to camp to zero in five seconds after eVeu initiation.

+ + ++,*

+++

++++ " + + : + ++++++* +++ ++ ++++++3.1+

44,L.

.4 4

-4,.-

+

,.*+.

+ * +

"*+

-*;:+;+* *:

  • *1+*+ +--

+ *.+: m

+

+: 'f :,7

f.

The 1979 AiS decay hoat correlation Is used to realistically medel theh reacotr decay beat..

.0.i i *!*i The.

r initial temperature of the suppression pool IS 800, ald the containment pressure is 14.7 pala.

h.

The water level in xthe suppression pool is at the lowest level allowed

ýýby the plant Technical Specifications.

I. The Suppression pool ýairspace pressure is in equilibrium with the

drywall Pressure.
j.

The suppreSsion pool and.the dryi-yll are conservatively assumed-tcbe Insulated volumes..

k.

No HPCI or RCIC systems ave available during fire event.

1.

LPCI systems are not used for core cooling in this analysis.

M.

One cof. spray system for core cooling is available after reactor depressurization.

n.

One RHR pump for suppression pool cooling is assumed available at one hour after event initiation in initial analyses. Suppression pool analyses are also performed for an RHR pump initiation time of two hours.

3.2 ACCEPTANCE CRITERIA The acceptance criteria used in evaluating this fire event include criteria for fuel cladding integrity and for suppression pool Integrity.

The reactor level response provides a good indioatton whether MtL deGege (fuel cladding perforation) to expected, It the core remains awenrd

(:

Y

1.'

throughout the analyzed event, then no fuel cladding, damage would occur.

If. the top portion of tho core it uncovered for a bris, period, the comblnation of low power level for this portloa of the fuel and the steam updraft cooling 11l prevent any significant heatup or fuel clawding damage.

If the core is uncovered to a mipgnificant depth and duration, then the :resultant fuel heat up must be evaluateG.

To ensure the fuel cladding interity. _the calculated PC?* should be less than th t

ra e

(approximately 1500 F) at which cladding. d*mae may occur. (Re ne

).

'The use-of a peak cladding. temperature a,4 des0 requirement rather than no.. core un-overY is consistorit with MIR guidellnes (Reference 5) which Is.

esablishes that.co U.novory of the U poprton of the Coro d=lus epesutization prior to reflo~oding Is acceptable for a urn.

oaTo uet acceptabiltity. of the suppression pool Ve..oote e cteria S,

are usead.

Tihs critriia "a.

a. The temperature and pressure in the pool should remain below the design conditions of 2810 F and 56 posi.
b.

The pool temperature should remain below the Heat Capacity Temperature Limit (H1CT)' for the pool *w*le'the reactor vessel io pressurized.

1, This criterion, established under ZVQs, is imposed to ensure that the operator will have sufficient time to transfer controls to tbe shutdown panel without the need to depressuri*e the reactor.

0 C.

Adequate NPSH to the low pressure pump should be availablo in order to prevent possible pump cavitation.

The NPSH available to the pmp is a function of both the pressure looses in the suction piping ead the pressure and temperature in the suppression pool., A pool temperature of approximately 1900 F at atmospheric pressure will provide adequate PS l for the W

pump and a temperature of approximately 11% F at atmospheric pressure will provide adequate NPSK for the.LMC p (Reference 6).

The results of the evaluations for the Identified tir evnt"at' in Seottoa 4.

-t I

-r;

4 RESULTS For the postulated fire event, the reactor pressure and reactor wast.r level are shown in Figures 4-1 and 4-2.

During the early period of the translent the reactor water level is slowly depleted by each 8RV actuation.

The indicated water level remains above the TAP until about 24 minutes at which the reactor vessel is depressurized. (The actual water level remains -above the TAF until about 27 minutes.)Therefore, the operator has sufficient time to take necessary actions to start depressurzinfg the reactor with the tvo SRV*. available..

Vhen, the reactor Is depressurized to the operating pressure.

of-:th core spray pi,,

this pump will be operated to Maintain reactor inventory.

Vhen the. reactor ilnventory is sufficiently recovered, the oprtr will'start the RHR-system. to provide suppression pool cooliing.

During, the depressurizationt, some-core uncovery is expected.

Thi. core un.overy, results in a PCT of approximately 13200 F (Figure 4.3) which is well below the temperature oa.t which cladding perforations can occur (approximately 15000 F).

Therefore, the postulated fire event will not cause fuel cladding damage.

To ensure that the identified limits for the suppression pool are not exceeded for the duration of hot shutdown, the suppression pool tevmperature response for this event was evaluated.

The pool temperature before reactor depressuritation was about 1060 F (Figure 4-4),

und the pool pressure was about 2 psig. The pool temperature is well below the HCTL for the corres-o ponding reactIr pressure. Utilizing the RHR system to provide suppression pool cooling after reactor depressurization, the maximum pool temperature and pressure are expected to -.dmax' below about 1800F Ond 11 pats, respectively.

Further, this combinatlor, of pool temperature and pressure is below the design limits, and will pro,'de adequate NPSH to assure the operation of the lHR and core spray V.umps for cool.rn.

injecttio.

Therefore, the requirements Identified for supiression *ool.ar satisietd.

The above pool temperature response, which is based on an Initial peel temperature of 800 F and MR pump initiation at one hour, shows a Ian" L

I margin to the Identified design limits for the suppreshion pool.

The margins, to the l1imits are large enough to satisfy the limits for the suppression.pool. and the NPSH requiroments even if the initial pool temperature:.. Isat 1

P oand

.r R pump Initiation for suppression pool cooling.does not begin until two houre after shutdown begins (Figure 4-5).

Therefore ýthe. sal als&s demonstrate that the remote shutdown system at PMP meets the requirements. specified in Section 111.L of 100FRO Appendix Rt for this, postulated 'event of fire at one of: the two shutdown PanelS.

La 0~'.-

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0 3 7 3 I

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  • i!:!6 *r!*ii~i!**~i,*,ii i~i *p*~~*!!!

cc LL) aci npare 4.1:

RoCCor Presure Response For fire At one Shutdmm hual

0 3 731.

20P I ;

'" -c Nt UL LUI LUI

-J wj T IME (SEC)

Flpwo 4-2: t.ctor water Level Response For nir. At One shutdon Panel

201 3 1 Pilgrim Power Plant Peak Cladding Tepern' r vS.o*

Trn

- I 1.4 1.3 1.2 1.1

.1 0.9 0.8 0.7 0.6 0.5 0,4 0

0.4 0.8 1.2 (Thousands) rime (Iusc) 1.6 2

fi~gmn 4-3:

tipre4-3:Peak Cla~ddiU Temperature Tror 71re At one Shuatdow Pm lns

0373 2ol9 PILGRIM NUCLEAR POWE.R STATION 50WPOO 1.VS 1

!,~~~~u MO-pm'"

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wStan*,

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.OA UK 00103 -memo v*

VS ~ ~

~

~

~

~

~

~~N 8r 4-:kpuuimF.

?qwt, bpm.?r Fire Ives At SUWU. P506

7L,

!! i'** ii*!~i~* **!.!'*

V i;7v.

I;

2 0 2 0 "!

":i~! * **<%:'

03731 Pilgrim Suppmssion Power-.- 'Ploant..

Pool r.~. Response 190 iao 170 160 150 140 130 120 110

  • IPs-ofW

.4. IU 100 gO II 80 0

4 8

12 16 (Thousands)

Tine ("0c) 20 24 28 32 FipVA 4-5:

SuPpresslon Pool TWq~ratuze for 60 F. 100 F Initial Fool

?.q~.rature and I Hour, 2 Hour RH nit Ltioncl Uus"

The analysis results show that safe shutdown can be achieved for a fire

'event at one of the two remote shutdo panels.

The event is postulated tod result in the -loss. 'of control of two SRVs, the HPCI, RCIC, and LPCI systems; however, two:: SMVe are available for reactor pressure control, one LPCS-pump is-assumed available for reactor cooling and one RUH pump

.is aasumedavailable.:for suppressib.

pool cooling, The" operator will have-sBuffi ient'. t to..have thet core spray system available to provide "ore makeup.

The-peak cladding temperature (FCT) Is low enough to ensure Fuel: clading integrity.

There Is also more than adequate margin to assure ::coutaiument: 'and-i suppression pool integrity for initial pool

.,,temperature of up. to 100F and the..initiation of the ýRHR pump at up to two hours".The"analysis also..shows that the available..NFSH will maintain

a.

adequate margin above the required NPSH fo the core spray and RHR 0 *p.mps.

0 5'4

6.

RIttOEMCES

/.

"U'.

..: at; 1,++

C. Saxenas, 'Safe Shutdown Appendix R Analyses for Fire96 vent One Shutdown Panel for PLlgria Nuclear Power PlantO, Revision draft June 1986.

!ii (4

2.

Letter, D. a.' FBisenhut (NRC)ý to: All PowrRatrLcne. arA A.Applicarito Fir: Power,Reactor Licence, Fre-: rotect an PoiySteering Committee Report (Genarid ILett*

5O) zt January 9,. 1983.

I 3,

Kaema, 11. J. Dix.Akw (WRC) to The NRC Commissionkers, uStaUf

'Recosmmovdations Rego rding the Implen entation of Appendix R, to 10 C.. :CIP 50 N, uclear Regulat ry Co.mis.on, Rule Xaking Neu,.*

(Notation Vote), SRCY-85-306: September 17, 1985.

4.

R. VanHouten, "Fuel Rod Failure as a Conaequev'ee of Departure from nNucleate Boiling or bryout",

U.S.

,e,

-- gulatory Commission, NURIG - 0562. June 1979.

5.

Memo. W. J. Dircks (NRC) to the NRC Commissiotners, 'Fire Protection Rule for Future Plants (SECY 82 - 267)", Rule Making Issue (Information) SECY 269, July 5, 1985.

I~.

0

6.

Letter, R. N. Swanson to R. G.. Furgeson, "N:

ECCS Systems". July 9, 1987 (ELGE87-07?

i~eme.~For 84

-

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